ML20027C306
| ML20027C306 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 10/12/1982 |
| From: | Blauer H, Cordaro M, Schmitt J LONG ISLAND LIGHTING CO. |
| To: | |
| Shared Package | |
| ML20027C225 | List: |
| References | |
| ISSUANCES-OL, NUDOCS 8210150257 | |
| Download: ML20027C306 (289) | |
Text
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LILCO, Octobor 12, 1982 UNITED STATES OF AMERICA !
NUCLEAR REGULATORY COMMISSION l Before the Atomic Safety and Licensing Board In the Matter of )
)
LONG ISLAND LIGHTING COMPANY ) Docket No. 50-322 (OL)
) (Emergency Planning--
(Shoreham Nuclear Power Station, ) Phase I)
Unit 1) )
TESTIMONY OF H. MARK BLAUER, MATTHEW C. CORDARO, AND JOHN F. SCHMITT FOR THE LONG ISLAND LIGHTING' COMPANY ON PHASE I EMERGENCY PLANNING CONTENTION 10(B) --
REAL TIME MONITORS
. PURPOSE The purpose of this testimony is to establish that the equipment LILCO intends to use to monitor releases provides information regarding radioiodine, particulate and noble gas releases to the environment. Downwind survey teams can scan for the plume centerline and the plume boundary and determine atmospheric radioiodine concentrations. The REMP can collect specific media in each portion of.the aquatic, terrestrial and atmospheric environment including direct accumulated radiation dosimetry using TLD's. Together, these systems, equipment, and methods provide timely, accurate, and detailed information for predicting and-calculating dose assessment. Thus, LILCO has i
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8210150257 821012 PDR ADOCK 05000322 f
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l the ability to assess accidents and monitor radiological releases from the Shoreham facility in the event of a radiolo-gical emergency. Real time monitors are'not required by NRC regulations and guidelines, and are unnecesary for timely moni-toring of the plume.
Attachments to this Testimony:
Attachment 10(B)-1 Professional Qualifications of H. Mark Blauer Attachment 10(B)-2 Professional Qualifications of Matthew C. Cordaro Attachment 10(B)-3 Professional Qualifications of John F. Schmitt Attachment 10(B)-4 SECY 82-111, "NRC Staff Recommendations on the Requirements for Emergency Response Capability" Attachment 10(B)-5 NUREG/CR-2644, "An Assessment of Offsite Real Time Dose Measurement Systems for Emergency Situations" Attachment 10(B)-6 SP 69.026.01, " Protective Action Recommendations" Attachment 10(B)-7 AIF/NESP-023, " Evaluation of an Environs Exposure Rate Monitoring System for Post-Accident Assessment"
LILCO, Octobar 12, 1982 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION i
Before the Atomic Safety and Licensing Board In the Matter of )
)
LONG ISLAND LIGHTING COMPANY ) Docket No. 50-322 (OL)
) (Emergency Planning--
(Shoreham Nuclear Power Station, ) Phase I)
Unit 1) )
TESTIMONY OF H. MARK BLAUER, MATTHEW C. CORDARO, AND JOHN F. SCHMITT FOR THE LONG ISLAND LIGHTING COMPANY ON PHASE I EMERGENCY PLANNING CONTENTION lO(B) --
REAL TIME MONITORS _
Ql. Please state your names and business addresses.
A2. [Blauer] My name is H. Mark Blauer; my address is 175 Old Country Road, Hicksville, New York, 11801.
[Cordaro] My name is Matthew C. Cordaro; my address is 175 Old Country Road, Hicksville, New York, 11801.
[Schmitt] My name is John F. Schmitt; my address is Shoreham Nuclear Power Station, P.'O. Box 628, Wading River, New York, 11792.
Q2. What are-your positions with LILCO? .
RA2. [Blauer] I am Emergency Planning Coordinator and
. Chairman of the Emergency Planning Task Force.
i.. .
[Cordaro} I am Vice President, Engineering for LILCO.
[Schmitt] I am the Radiochemistry Engineer at Shoreham.
Q3. Please state your professional qualifications.
A3. {Blauer] My professional qualifications are attached to this testimony (Attachment 10(B)-1). My familiarity with the issues surrounding the use of real time moni-tors stems from research in conjunction with my teach-ing position, and my duties involving specification and installation of real time monitors during the course of employment in various positions.
[Cordaro] My professional qualifications are attached to this testimony (Attachment 10(B)-2). I am sitting with this panel to provide the LILCO management per-spective regarding Emergency Planning, and to answer any questions pertinent to management. My role in Emergency Planning is to ensure that the needs and re-
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quirements of Emergency Planning are being_provided,
- and that the technical direction and content of Emergency Planning are being conveyed to corporate man-agement.
[Schmitt] My professional qualifications are attached to this testimony (Attachment 10(B)-3). My familiarity o
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.with'the issues surrounding the~use of real time monitors stems from my familiarity with the Radiation Monitoring: System, and the effluent qualification and dose calculation methods we employ.
-Q4. Are you-familiar with Suffolk County Contention EP 10(B)?
A4. [Blauer, Cordaro, Schmitt) Yes.
QS. What does that contention say?
4 AS. [Blauer, Cordaro Schmitt) EP 10(B) provides:
1 3.
Suffolk County contends'that the LILCO plan (See Chapter 6) is inadequate-with respect to e
f its ability to assess and mitigate accidents and monitor radiological releases from the Shoreham facility in the event of a radiolo-gical emergency. Thus, LILCO has failed to comply with 10 C.F.R. 95 50.47(b)(2), (4),
(8), (9), and (10), 10 C.F.R. Part 50,
- Appendix E and NUREG 0654, Items II.B, D, H, I, and J in the following respects
(B) LILCO does not intend to use real time monitors at fixed locations that can be remotely interrogated.
Q6. Are you familiar with the regulatory requirements and guidelines cited in the preamble to EP 10(B)?
v A6. [Blauer, Schmitt) Yes.
1 Q7. Are alloof the regulations cited by the County appli-cable to the issue of real time monitors?
i-( .,. :g
A7. [Blauer, Schmitt] Arguably, 10 C.F.R. 5 50.47(b)(9),
10 C.F.R. 9 50, Appendix E, Item IV.E, and NUREG-0654, Items II.H(6)(b) and I(8) are related to the issues surrounding real time monitors.
Q8. What do those regulations and guidelines say?
A8. [Blauer, Schmitt] 10 C.F.R. 5 50.47(b)(9) provides:
Adequate methods, systems and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emer-gency condition are in use.
10 C.F.R. Part 50 Appendix E, provides:
Item IV.E(2)
Adequate provisions shall be made and de-scribed for emergency facilities and equip-ment including: . . . (2) equipment for de-termining the magnitude of and for continu-ously assessing the impact of the release of radioactive materials to the environment.
NUREG-0654, Item II.H(6)(b) provides:
Each licensee shall make provisions to acquire data from or for emergency access to offsite monitoring and analysis equipment in-cluding: . . . (b) radiological monitors in-cluding rate meters and sampling devices.
Dosimetry shall be provided and shall meet at a minimum the radiological assessment branch technical position for the environmental radiological monitoring program.
NUREG-0654, Item II.I(8) provides:
Each organization [ licensee, state and local], where appropriate, shall provide methods, equipment and expertise to make rapid assessment of the actual or potentie.1 magnitude and locations of any radiological hazards through iiquid or gaseous release pathways. This shall include activation, o
. notification means, field team composition, transportation, communication, monitoring equipment and estimated deployment times.
Q9. Do the regulatory requirements and guidelines cited by the County, or any other requirements or guidelines, require LILCO to install real time monitors? i A9. [Blauer] No. In SECY 82-111, "NRC Staff ,
Recommendations on the Requirements for Emerger._y l Response espability" (Attachment 10(B)-4 to this testi- i mony), the NRC Staff states at part 2. A on page 13 that i i
" continuous offsite dose monitors are not required pending their further development and consideration as requirements." In addition, NUREG/CR-2644, "An Assessment of Offsite Real Time Dose Measurement- l Systems for Emergency Situations" (Attachment 10(B)-5 to this testimony) summarizes that "in general'it is highly questionable that a 16-32 unit emergency moni- I toring system can provide sufficiently reliable techni-cal information to be of use in a decision making pro-cess in the event of an emergency situation." Real i
time monitoring capability is not recommended by l l
NUREG-0654, Rev. 1, " Criteria for Preparation and
{
Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants." The 4
schedule for implementing the use of 16-20 environs i
x a
radioactivity monitors in the proposed revision to Reg.
Guide 1.97, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident", dated
' December 19, 1979, has been delayed in Revision 2 to Reg. Guide 1.97, dated December 1980. (errata, July 1981).
Q10. What are real time monitors?
A10. [Blauer) An offsite dose measurement system (real time monitor system) consists of fixed field stations com-prised of (1) a detection device (the real time moni-tor) and (2) a microprocessor with data transmission capability. The detection device, normally a pressuri-zed ionization chamber, and. associated electronics, such as preamplifiers, amplifiers, high voltage sup-plies and other equipment, measure real time dose rates. The microprocessor converts the field detection data to output data for transmission by either radio telemetry, dedicated phone line, or hard wire cennec-tion to a remotely located, central processing unit (CPU). The CPU acquires, reduces, and stores data de-scribing radiation dose rate conditions existing at each field station.
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-Q11. Does LILCO plan to install real time monitors?
All. [Blauer, Schmitt] No.
Q12. Why not?
A12. [Blauer, Schmitt] LILCO does not intend to use real time monitors because (1) LILCO has in-plant monitors that monitor radiation releases, and field teams that monitor the plume, making real time monitors unneces-sary, and (2) there are certain functions that cannot be performed by a real time monitor that can be per-formed by LILCO field monitoring teams.
Q13. How does LILCO monitor releases and assess doses?
A 2. [Blauer, Schmitt] LILCO uses three methods to monitor releases and assess doses:
(1) Radiation Monitoring System; (2) field monitoring teams; and (3) Radiological Environmental Monitoring Program.
Q14. What is the Radiation Monitoring System?
.A14. [Blauer, Schmitt) The Radiation Monitoring System-(RMS) is a computer-based system that provides the status of radiation levels throughout the plant to the control room operators. The data from radiation
monitors are inputted to the computer to provide continuous real time data on plant conditions. These data are stored as an historical file for report gener-ation. Set points are provided for each radiation mon-itor and automatically trip to alert plant personnel of unusual conditions.
In addition to keeping the staff aware of conditions in the plant, the RMS performs calculations to determine the effect of effluents on the population. For gaseous releases, the RMS uses radiation detector readings at the point of discharge, with current meteorological data from the offsite meteorological tower instruments.
These data are used in the accident situation in for-mulas concerning dispersion and biological effect of each radionuclide in the release. Using the data, the RMS calculates radiation doses to the thyroid and whole body to individuals at various distances in the down-wind direction from the plant.
Q15. What is the Radiological Environmental Monitoring l Program?
A15. [Blauer] The Radiological Environmental Monitoring 1
Program (REMP) identifies and determines the background radioactivity present in various biological, hydrologi-cal, lithological, and atmospheric media.found in the
_9 Shoreham environs. Samples of these media are chosen i'
for monitoring to best determine the radioactivity pre-sent in pathways to humans. The samples are sensitive indicators of changes in radioactive levels. Results j l
from analyses of these samples will indicate any envir- !
onmental impact Shoreham may cause. <
d Media samplad within the aquatic environment include surface water, aquatic plants, fish, invertebrates and sediment. From the atmospheric environment, airborne particulates and iodine are collected. From the ter-restrial environment, milk, potable water, food prod-ucts, game and soil are obtained. In addition, direct accumulated radiation doses are measured using thermo-1 luminescent dosimeters (TLD). During an emergency, the REMP can be increased in size, scope and frequency of sample collection.
Emergency response capabilities to analyze these sam- !
ples within approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> have been establi-shed. Thus, the REMP can provide a historical. record
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l to monitor releases and assess doses.
Q16. What is the composition of the field monitoring teams?
i A16. [Blauer) Should an offsite radiological release. occur, LILCO has the capability to dispatch three field.
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monitoring teams. Each team would consist of a health physics / engineering technician and a health,phy-sics / engineering aide. Depending upon the' level of emergency declaration and the emergency response facil-ity that has been activated, the team will be dis-patched either from the Technical Support Center (TSC) or the Emergency Operations Facility (EOF).
Field teams are dispatched to measure offsite radiation I levels. Each team is issued a downwind survey kit, a ,
i portable two-way radio, and a field monitoring vehicle. ;
Included in each kit are: ]
(1) a portable dose rate survey meter; I
(2) a portable high volume air sampler; (3) a portable count rate meter; >
(4) personal dosimeters; 4 (5) respiration equipment; (6) protective clothing; and :
(7) potassium iodide pills.
P Q17. What is a portable dose rate survey meter? [
i A17. [Blauer] A portable dose rate survey meter measures-radiation levels. It is a hand-held box, about the ,
size of a small shoe box, with a probe attached by coaxial cable. The portable dose' rate survey meter can s
be used to determine the location of the plume by comparing (1) the dose rate shown when the probe is held four feet above the ground and pointed to the sky to (2) the dose rate shown when the probe is held three inches off the ground and pointed toward the ground.
The field monitoring teams can compare the two dose rates to determine whether the plume is above them, has settled on the ground, or is around them. In addition, by scanning with the portable dose rate survey meter as the field teams are traveling through an area, the field teams can locate the plume boundaries and center-line.
Q18. What is the portable high volume air sampler?
A18. [Blauer, Schmitt) A portable high volume air sampler is a large box that has the ability to suck air through a filter-cannister assembly, and exhaust the air through to the other side. Particulate radioactive ma-terial that may be in the atmosphere is deposited on the filter, and radioiodine is collected on silver-i impregnated resin located behind the filter. The car-tridge containing the filter and the resin can then be removed from the air sampler and taken to the field monitoring vehicle, where instruments are available to analyze the sample deposited in the cartridge and on I
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l the. filter. The results of these analyses are reported by the field monitoring team to the dose assessment staff, using two-way radios. The dose assessment staff uses that information in their dose assessment calcula-tion, using the dose assessment model described in the testimony filed for emergency planning contention 14.
Q19. What is a portable count rate meter?
A19. [Blauer] The count rate meter is similar to the dose rate survey meter, but measures the quantity of radioi-sotopes in the environment. It uses a compensated GM detector. The count rate meter is used to determine the amount of radioactive material collected in an air sample. The portable count rate meter enables the field monitoring teams to determine the particulate and radiciodine concentrations in the plume. The portable count rate meter and GM detector, when used with the TCS high volume air sampler, is capable of detecting as little as 10 -7 uCi/cc I-131.
Q20. How are personal dosimeters, respiration equipment, and protective clothing used?
A20. [Blauer] These items allow the field monitoring teams to travel throughout the ten-mile EPZ to monitor the
, plume, while maintaining doses as low as reasonably achievable (ALARA).
Q21. How is - the information obtained Irom the Radiation Monitoring System, the field monitoring teams, and the Radiological Environmental Monitoring Program used by LILCO in radiological assessment?
A21. [Blauer, Schmitt] The Radiation Monitoring System pro-vides release rates, source terms, and concentrations of effluent releases. During accident situations, real time site meteorology is used with emergency dose soft-ware to calculate offsite doses. These calculated pro-jected doses are used in accordance with SP 69.026.01,
" Protective Action Recommendations" (Attachment 10(B)-6 to this testimony), to determine the appropriate recom-mendation.
Downwind survey teams provide actual radiation measure-ments of dose rate, particulate, and radiciodine air concentrations. These measurements are transmitted by radio to the dose assessment staff for determination of measured projected doses. The survey teams provide !
more accurate and reliable measurements than the Radiation Monitoring System for determining protective action recommendations, and can verify projected doses from the Radiation Monitoring System.
The Radiological Environmental Monitoring Program allows for the collection and analyses of environmental
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samples. These'results provide an historical record to be used in assessing the radiological environmental impact of an accident.
-Q22. What is the capability of a real time monitoring sys-tem?
A22. (Blauer] A rer.1 time monitoring system features con-tinuous monitoring of ambient radiation levels, contin-uous 24-hour opertion, and remote interrogation from
, any number of locations.
Q23. Did LILCO consider using a real time monitoring system as part of its emergency plan?
A23. [ Blauer,] Yes. The Emergency Planning Task Force investigated the merits of a real time system to deter-mine offsite doses by field monitoring and telemetering field readouts to the plant control room and the emer-gency response facilities. The Emergency Planning Task Force concluded that such a system would not sufficien-tly enhance LILCO's present monitoring capabilities to an extent that would justify the expense, because of the following:
(1) The present Radiation Monitoring System already calculates dose and dose rate values using real time meterclogical information for the downwind
ib sector, from the exclusion area out to a distance of 100 kilometers from the plant, based on current accident or sampled isotopic releases.
(2) The RMS can also display on a Long Island F.ap all areas having the potential for doso or dose rate values equal to or greater than protective action guidelines.
(3) Remote real time systems cannot reasonably deter-mine isotopic concentrations and may not, in most
- cases, be in the plume path. Additionally, remote real time systems are not sufficiently sensitive to provide timely indications of actual public dose.
(4) Downwind r,urvey teams can more accurately define the plume and sample the environmental radioacti- ,
vity. Actual samples can be taken and analyzed.
If one of the existing environmental stations is located in the plume pathway, its monitors and sam-ples can also be used.
I Q24. Are there any other limits to the real time monitoring system.that the County proposes LILCO use?
l A24. [Blauer] Yes. Calibration of the overall real time monitoring systera is extensive. The detectors and l
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. electronics, as an integrated system for each field station, must be calibrated to obtain conversion fac-tors relating detector current to dose rate (R/hr).
Report AIF/NESP-023, " Evaluation of an Environs Exposure Rate Monitoring System for Post Accident Assessment" (Attachment 10(B)-7 to this testimony),
states that the initial' calibration energy range for detectors should be from 60 to 3,000 kev. Also, the initial detector calibration from the system vendor should be done at the National Bureau of Standards (NBS) or with an air cavity ionization chamber calibra-ted by the NBS. Less extensive field calibration, that is, use of fewer energies, should be conducted semi-annually.
In addition, a real time monitoring system would give no indication of an impending release, and hence.would not provide warning time to initiate protective action.
The monitors must be located close to the plant, pref-erably within one mile, to provide the sensitivity and response time necessary for predicting and calculating doses. Under certain meteorological-conditions during a nuclear incident, the plume may rise and skip"over the. detectors. Therefore, the real time monitors would l l
not enhance LILCO's' capabilities to determine onsite and offsite doses.
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Furthermore, s'ixteen_ locations (one for each' sector for a 360 degree planning area around the plant) do not provide any precise information regarding the plume's centerline or plume boundary. As I previously men-tiened, downwind survey teams are able to scan forithe plume centerline and the plume boundary,_providing more accurate and detailed information for predicting and calculating doses. The equipment LILCO will use to monitor the plant effluent releases provides timely and accurate information regarding actual quantities of ra-dioactive iodine, particulate, and noble gas releases to the env'.ronment.
The real time monitoring system's potential deficien-cies are summarized aptly in NUREG-CR-2644 as follows:
(1) While a ring of detectors around a nuclear power station can provide the means for monitoring releases, the number of stations required for two detectors to provide information within a factor of five of each other can be as large as 50 or more for one installation.
(2) The use of short time (15 minutes) data from a fixed offsite monitoring system to project downwind dose rates is a complex and highly uncertain process. Based on our study, the uncertainty associated with a projected value is at least a fac-tor of 10 or more.
(3) The use of a= fixed offsite monitoring system to determine the magnitude.of an unmonitored release in the presence of-a monitored release is highly questionable.
Depending on the ratio of an unmonitored
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. release- to .the monitored release, uncertainties of factors of 25 and 50. are
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(4) Several vendors of monitoring equipment were' contacted relative to cost and per '
formance-characteristics of the available-instrumentation. In addition, w 'e con -
tacted several power stations and_ state-agencies involved in the installation of~
fixed real time environmental monitoring systems. While the cost factors for the instrumentation were relatively fixed, the installation costs were highly varia- :
ble. Based on this study, the cost per- -
monitoring statica ranges from $25,000 to
$65,000. Depending upon the specific site characteristics, the. cost for a 32
-station system could easily exceed l
$1,000,000 while only providing data-with
-uncertainties in the range of factors 10 to 50.
(5) The placement of the simple limited
($500,000) detector system in proximity (0.5 miles) to a reactor may not provide reliable information in~the case of an emergency for several reasons. Of prime importance is the limited number of sta- 4 tions (8-16) that could be installed and the consequence that a plume might go undetected.
- For all of these reasons, LILCO has determined that real time monitors are unnecessary for the Shoreham- '
plant.
Q25. Please summarize your testimony.. ,
A25. (Blauer, C o r d e.r o , Schmitt] In summary,'LILCO-does not ~
intend to use real' time monitors at fixed-locations o
that can be remotely interrogated. .The real time
- monitoringLsystem limitations include:
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1
- 1. no indication of an impending release,
- 2. little warning time to initiate protective actions,
- 3. uncertainty that the plumes may rise above or pass to the side of a monitoring location and thus miss the fixed monitor completely,
- 4. inability to measure radioiodine air concentra-tions,
- 5. uncertainty in determining plume center line and plume boundaries,
- 6. low sensitivity to determine the magnitude of an unmonitored release in the presence of a monitored release, and
- 7. uncertainty prcjecting downwind dose rates. I The equipment LILCO intends to use to monitor releases provides information regarding radiciodine, particulate and noble gas releases to the environment. Downwind survey teams can scan for the plume centerline and the plume boundary and determine atmospheric radioiodine concentrations. The REMP can collect specific media in-each portion of the aquatic, terrestrial and atmos-pheric environment including direct accumulated radia-tion dosimetry using TLD's. Together, these systems, equipment and methods provide timely, accurate and de-tailed information for predicting and calculating dose assessment. Thus, LILCO has the ability to assess
accidents and monitor radiological releases from the Shoreham facility in the event of a radiological emer-gency.
t F-
4-p Attachront 10 (B) . C , , ,
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PROFESSIONAL QUALIFICATIONS H. MARK BLAUER Chairman, Emergency Planning Task Force Emergency Planning Coordinator ,
LONG ISLAND LIGHTING COMPANY My name is H. Mark Blauer. My business address is Lona Island Lighting Company, 175 East Old Country Road, Hicksville, New York 11801. I am Chairman of the Emergency Planning Task -
i Force and Emergency Planning Coordinator. In this capacity, I
- . report to the Vice Presdient, Nuclear, and the Vice President, Engineering. I also report to the Manager, Nuclear Engineering Department. My duties inc.?.ude overall technical and cdminis-tration responsibility for the Emergency Planning Task Force.
The Task Force is' responsible for developing and maintaining the-Shoreham Nuclear Power Station Emergency Plan; Emergency Training curriculum, manuals, and lesson plans;' qualification and selection of emergency response personnel; Emergency Plan-procedures; onsite and offsite emergency support facilities;
- the Prompt Notification System; the interfacing with Federal l
(NRC,-DOE, FEMA, Coast Guard), State (Department of Health, -
Disaster Preparedness Commission).and Local (Suffolk County, E
d e- ,. ~ ,e v ~w6 -v -
e .o e,--, em.
c ,
t hospitals and fire departments) authorities as well as other
- nuclear industry support groups (INPG). .
I received my Bachelor and Master of Science degrees from the State University of New York at Stony Brook in 1968 and 1971, respectively. I received my Doctorate in Nuclear Chemistry from the University of Glasgow, Scotland in 1977.
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From 1971 to 1975 I was a Research Assistant (U.S.
equivalent: Assistant Professor) at the University of Glasgow teaching nuclear chemistry and researching low-level tritium techniques. I was a E.. search Assistant (U.S. equivalent:
Assistant Professor) at University College, London from 1975 to 1977 teaching isotopo geology, researching major and trace ele- -
ment techniques and acting as consultant to several water auth-orities. During this period the following were~ published:
Anderson, A., Blauer, H. M. and Baxter, M.S.
' (1977). A controlled power supply for the electrolytic enrichment of tritium, J. Physics, V10, pp. 1286-1294.
Beckinsale, R.D., Bowles, J.F.W., Pankhurst, R.J.,
Wells, M.K. and Blauer, H.M. (1977).
Rubidium-strontium age studies and geochemistry of acid veins in the Freetown complex, Sierra Leone, Mineralogical Magazine, V41, pp. 501-511.
Blauer, H.M., Baxter, M.S. and Anderson, A.
(1978). An improved technique for the electrolytic enrichment of tritium, Analyst, V103, pp. 823-829.
Hope, C.A., Blauer, H.M. and Reiderer, J. (1980).
Recent analysis of 18th dynasty pottery in "Studien zur
" Altagyptischen Keramic," edited Dorothea Arnold, Philip von Zabern, Mainz.
In 1977 I returned to the United States and assumed the
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4 position of Assistant Professor at the University of Pittsburgh, Department of Radiation Health from 1977 to 1980.
I taught radiation health, radiation chemistry and nuclear chemical separation techniques; researched bioassay techniques and low-level environmental measurement techniques; directed an EPA certified radio-chemical laboratory; and consulted with several major uranium producers. During this period the fol-lowing were published:
Dennis, Nancy A., Blauer, H. Mark, and Kent, Jacquelir.e E. (1981). Dissolution fractions and half-times of single source yellowcake in simulated lung fluids, Health Physics, V41.
Culp, P..and Blauer, H.M. (1979). Dissolution rates of radionucleides from coal and coal ashes, Twenty-fourth Annual Meeting of the Health Physics Society, Philadelphia, PA.
Dennis, N.A. and Blauer, H.M. (1979). Dissolution rates of uranium in yellowcake in simulated lung fluids, Twenty-fourth Annual Meeting of the Health
- Physics Society, Philadelphia, PA.
Padezanin,.T. and Blauer, H.M. (1979). Comparison of
~~ uranium urinalysis methods, Twenty-fourth Annual Meeting of the Health Physics Society, Philadelphia, PA.
Blauer, H.M. and Dennis, N.A. (1979).' Dissolution rates of uranium from single source yellowcake in both simulated interstitial and surfactant lung fluids, Twenty-fifth Annual Conference on Bioassay, Enviromental and Analystical Chemistry, Las Vegas, N.Y.
Maitz, A.H. and Blauer, H.M. (1980). Pure uranium' oxides: their dissolution rates plus relationship to yellowcake dissolution characteristics,-Twenty-fifth Annual Meeting _of the Health Physics Society, Seattle, WA.
Blauer, H.M. and Brown, S.H.~(1980). Physical and chemical parameters affecting dissolution-h
%r
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4 characteristics of yellowcake in simulated lung fluids, Twenty-fifth Annual Meeting of the Health Physics Society, Seattle, WA. .
Brown, S.H. and Blauer, H.M. (1980). Characterization of_ yellow-cake (U308) from multiple sources and some implications regarding-uranium mill bioassay, Twenty-fifth Annual Meeting of the Health Physics Society, Seattle, WA.
From 1980 to 1981 I was Environmental Scientist at
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Three Mile Island Nuclear Generating Station responsible for audits, the Radiological Environmental Monitoring Program, offsite dose calculations and health effects studies. During this period the following positions and procedures were writ-ten:
Blauer, H. Mark (1981). Three Mile island Nuclear Station, Comments on the Articles "The First Catualty at TMI" and "The Lethal Path of TMI Fallout" by Ernest J. Sternglass.
Blauer, H. Mark (1981) TMI Enviromental Controls REMP Procedure, Determination of REMP investigation levels and subsequent actions, ECP 1507, Rev. 1.
Blauer, H. Mark (1981) TMI Environmental Controls Emergency REMP Procedure, operating procedure for the CRT, ECP 1601, Rev. O. j Blauer, H. Mark (1981) TMI Environmental Controls f Emergency REMP Procedure Determination of Off-Site ,
Dose, ECP 1602, Rev. 1.
Blauer, H. Mark (1981) TMI Environmental Controls Procedure Ge(li) detector system using series 80, ECP .
1719, Rev. O.
I joined LILCO in 1981 as Senior Scientist, Nuclear (
Licensing Division. My responsibilities include providing sup-port to corporate and plant staff in the areas of Radiation i
-S-Protection, Health Physics, ALARA, Emergency Planning and REMP.
In 1982 I became Chairman of the Emergency Planning Task Force responsible for all technical and administrative functions.
During this period, the following courses and procedures were prepa12d:
General Physics - BWR Familiarization Course (1 week)
LILCO - BWR Familiarization Course (2 weeks)
Blauer, H.' Mark (1981) REMP data receipt and running tables, RP 4.2, Rev. O Blauer, H. Mark (1981) Anomalous data results - LLD and positive value exceptions, RP 4.3, Rev. O Blauer, H. Mark (1982) Acceptance Criteria, RP 4.4, Rev. O Blauer, H. Mark (1982) Determination of REMP investiga-tion levels and subsequent actions, RP 4.5, Rev. O.
I am certified by the American Chemical Society and a member of the American Geophysical Union and Health Physics Society.
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, Attachment 10(B) -2 7
?ROFESSIONAL QUALIFICATIONS MATTHEW C. CORDARO Vice President of Engineering LONG ISLAND LIGHTING COMPANY My name is Matthew C. Cordaro. My business address is Long Island Lighting Company, 175 East Old Country Road,
'Hicksville, New York 11801. I am currently Vice President of Engineering and have held this position since the spring of 1978. As Vice President of Engineering, I am responsible for all of LILCO's engineering activities. This includes responsi-bility in the areas of facility planning and engineering for nuclear and fossil electric generating plants, as well as elec-tric and gas transmission and distribution systems. In addi-tion, I am responsible for assessing the environmental impacts of all LILCO operations.
I received my Bachelor of Science degree in Engineering Science from C. W. Post College in 1965. I received my Master of Science degree in Nuclear Engineering from New York University in 1967. I received my Doctorate in Applied Nuclear Physics from the Cooper Union School of Engineering and Science in 1970. I was awarded the Atomic Energy Commission Special Fellowship in Nuclear Science and Engineering
% %, 6 My past professional affiliations include a position as-Guest Research Associate at Brookhaven National Laboratory, Adjunct Associate Professor of Nuclear Engineering at
. Polytechnic Institute of New York and Adjunct Assistant Professor at C. W. Post College.
I joined LILCO in 1966 and from 1966 to 1970 I held the positions of Assistant Engineer (1966), Associate Engineer-(1967), Nuclear Physicist (1968) and Senior Environmental
, Engineer (1970). In these earliest positions with LILCO I was involved as a principal in all phases of nuclear power plant design, licensing and fuel management. I was also a lead wit-ness for the Company in Federal and State licensing proceedings for the Shoreham and Jamesport Nuclear Power Stations.
In 1972 I assumed the position of Manager of Environmental Engineering. In this capacity I was responsible for the environmental impact of all LILCO operations. This position involved the supervision, administration and direction of all environmental programs aimed at demonstrating compliance with applicable standards.
I am a member of a number of related professional-organizations including: the Board of Directors, Adelphi University's Center on Energy Studies; and the Council'of >
+
- Overseers, C. W. Post College. Other related professional
-2
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affiliations are: the Technical Resources Advisory Council to the New York State Department of Environmental Conservation; 4
the New York Power Pool Environmental Committee; Advisory Task
. Forces and Committees of the Atomic Industrial Forum; the Long Island Association of Commerce and Industry Environmental Committee; the Advisory Board to Environmental Techno5 y a
Seninar; the Environment and Energy Committee of the Edison Electric Institute; and the HSA Environmental Task Force. I have also been a member of the Research Planning Advisory Committee for the New England River Basins Commission Study of Long Island Sound, the Marine Advisory Council to the New York State Sea Grants Seminar, and the Nassau-Suffolk Health Systems Agency (HSA), Suffolk County Council.
In addition, I am a member of the American Nuclear 1
Society, the Health Physics Society and the Environmental Technology Seminar.
My most recent publications include a paper on methodology for power plant site selection, papers presented at the World Energy Conference on space heating alternatives and power plant cooling systems, a paper related to power plant waste heat utilization, and a paper on the transportation of nuclear wastes. I have also published journal articles in-the fields of environmental science and nuclear science, as well as i .. -. . - . . . . . . . . . . . ~ .
numerous studies and reports related to the environmental effects of energy production.
I recently testified before Congressional Committees on Nuclear Waste Transport and the Economics and Environmental Impacts of Coal Utilization.
1 i
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, Attachm:nt 10(B) -3 l
1
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l PROFESSIONAL QUALIFICATIONS JOHN F. SCHMITT Radiochemistry Engineer LONG ISLAND LIGHTING COMPANY My name is John F. Schmitt. I am the Radiochemistry Engineer of the Shoreham Nuclear Power Station, a position I I
have held since January 1975. As such, I am responsible for developing and implementing the chemistry, radiochemistry and effluent monitoring program for Shoreham. This includes, among other things, directing all work related to conducting the chemical and radiochemical analyses and treatments of plant '
process systems; detecting and controlling environmental re-leases; implementing the ALARA policy for these releases; and preparing records and reports of chemical surveys.
I graduated from Manhattan College in 1966 with a ;
Bachelor of Science degree in chemistry and received a Master of Science degree in Environmental Health Science, specializing '
in Radiological Health (Health Physics), from the University of Michigan in 1974 and became a Certified Health Physicist in !
l 1982. I completed the General Electric Boiling Water Reactor j
Chemistry Course in November 1975. I have also completed many industry seminars and training programs, including:
4,
- a. Radiation Protection - LILCO Evening Institute
- b. Radiation Protection Workshops - General Electric Company
- c. BWR Chemistry Training - General Electric Company
- d. Health Physics Review - Rockwell International
- e. Accelerated Health Physics Instruction - NUS
- f. Accelerated Nuclear Plant Chemistry Instruction -
- g. Health Physics Review - Brookhaven National Labs
- h. Environmental Radiation Surveillance - Harvard i School of Public Health
- i. Radioactive Waste Management for Nuclear Power Reactors - ASME/ University of Virginia
- j. Post Accident Sampling Workshops - Sentry Equipment, EPRI
- k. Control of Plant Radiation Fields - EPRI, General Electric Company
- 1. Atomic Absorption / Atomic Emission Spectrometry -
Instrumentation Labs
- m. Gamma Spectrometer Operation - Canberra Industries I started work for the Long Island Lighting Company in 1966 as an Assistant Engineer at the Far Rockaway Power Station. I took a military leave of absence from 1967-1972 to serve as an officer in the U.S. Air Force. Returning to LILCO in 1972, I was an Associate Engineer at the Glenwood Power Station. From 1973 until assuming my present position in 1975, I was assigned to the staff of the Shoreham Nuclear Power Station as an Associate Engineer and Plant Engineer. During this time, I studied health physics at the University of
Michigan and received training at the AEC's Savannah River Plant and Commonwealth Edison's Dresden Nuclear Power Station.
I am a member of the Health Physics Society, New York Chapter of the Health Physics Society, Power Reactor Health Physicists, and the Long Island Chapter of the American Nuclear Society.
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y-5 Attachment 10(B) 4, ;. s <- -
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1 NRC STAFF RECOMMENDATIONS ON THE REQUIREMENTS FOR l
EMERGENCY RESPONSE CAPABILITY i l
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. . CONTENTS Pace ,
. 1. INTRODUCTION.................................................. I
- 2. USE OF EXISTING DOCUMENTATION................................. 3
- 3. COORDINATION AND INTEGRATION OF INITIATIVES................... 4
- 4. SAFETY PARAMETER DISPLAY SYSTEM (5PD5)........................ 7
- Current Regulatory Requirements Functional Statement -
Recommended Requirements Integration
- Reference Documents
- 5. OETAILED CONTROL ROOM DESIGN REVIEW........................... 10 Current Regulatory Requirements Functional Statement Recom= ended Requirements
. 'k - Documentation and NRC Review
- Integration
- Reference Documents .
- 6. REGULATORY GUIDE 1.97 - APPLICATION TO EMERGENCY RESPCNSE
- FACILITIES.................................................... 13
' Current Regulatory Requirements Functional Statement Recommended Requirements .
Documentation and NRC Review
- 7. UPGRADE EMERGENCY OPERATING PROCEDURES (E0Ps)................. 15 Current Regulatory Requirements .
Functional Statement Recommended Requirements l
- Documentation and NRC Review -
Reference Documents B. EMERGENCY RESPONSE FACILITIES................................. 17 )
.- 1 Current Regulatory Requirements .
l, Technical Support Center.................................... 19 Functional Statement Recommended Requirements
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CONTENTS (Continued)
Operational Support Center.................................. . 21 Functional Statement Reco= ended Requirements
- Emergency Operations Facility.............................. 22 Functional Statement Recommended Requirements Documentation and NRC Review
- Reference Documents Table 1 - Emergency Operations Facility Locction Options.......... 25 Table 2 - Minimum Staffing Requirements for NRC Licensees for Nuclear Power Plant Emergencies......................... 26 o
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EMERGENCY RESPONSE CAPABILITY .
. 1
- 1. INTRODUCTION .
I This report was prepared as a result of a review by the Committee to l Review Generic Requirements (CRGR). The recommendations herein have been developed by the program offices and are supported by CRGR. The report represents the staff's attempt to distill the fundamental requirements for nuclear plant Emergency Response Capability from the wide range of guidance documents that NRC has issued. It is not intended that these guidance documents (NUREG reports and Regulatory Guides) be ignored; they are still useful sources of guidance for licensees and NRC staff
. regarding acceptable means for meeting the fundamental requirements contained in this document.
These fundamental requirements are further specification of the general guidance specified previously by the Commission in its regulations, orders and policy statements on emergency plannin5 and TMI issues. It is intended that these fundamental requirements would be applicable to licensees of operating nuclear power plants and holders of construction 1
.( permits for nuclear power plants. For applicants for a construction Permit ICP) or manufacturing license (ML), the requirements described in th.is document must be supplemented with the specific provisions in the rule specifying licensing requirements for pending C F and ML applications.
In this regard, it is expected that the staff would review CP and ML applications against the guidance in the current Stendard. Review Plan, -
and this might lead to more detailed requirements than prescribed in this -document.
Based on discussions with licensees, the staff has learned that many of
- the Commission approved schedules for emergency response facilities probably will not be met. In recognition of this fact and the difficulty
! of implementing generic deadlines, the staff proposes that plant-specific l schedules be established which take into account the unique status of each plant. The following sequence for developing implementation schedules
- is proposed.
When the basic requirements for emergency response capabilities and j facilities are finalized, they should be transmitted to licensees by a
- generic letter from NRR, promulgated to NRC staff, and incorporated as .
regulatory requirements (e.g., in the Standard Review Plan or by regulation or Order, as appropriate). The 'etter to licensees should request that licensees submit a proposed schedule for completing actions to comply with the basic requirements. Each licensee's preposed schedules would then be reviewed by the assigned NRC Project Manager, who would discuss
( . the subject *With the licensee and mutually agree on schedules and completion i
dates. The implementation dates would then be formalized into an enforceable document. ;
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- The basic requirements in this document do not a?ter previously issued guidance, which remains in effect. This document does attempt to place that guidance in perspective by identifying the elements that the NRC staff believes to be essential to upgraded emergency response capabilities.
The proposal to formalize implementation dates in an enforceable document
- reflects the level of importance which the NRC staff attributes to these basic requirements. The NRC staff does not recommend that existing guidance be imposed in this manner, but rather that it be used as guidance to be considered in upgrading emergency response capabilities. This indicates the distinction which the staff believes should be made between the basic requirements and guidance.
The following sections describe NRC staff recommendations on basic re-quirements, their interrelationships, and NRC actions to improve manage-ment of emergency response regulation. Reference documents are cited with a description of content as it relates to specific initiatives.
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- 2. USE OF EXISTING DOCUMENTATION The NRC~ staff recommends that the following NUREG documents are intended to be .
used as sources of gi'idance and information, and the Regulatory Guides are to be considered as guidance or as an acceptable approach to meeting formal requirements. The items by ' virtue of their inclusion in these documents shall not bs misconstrued as requirements to be levied on licensees or as inflexible criteria to be used by NRC staff reviewers.
NUREG Report Titles 0696 - Functional Criteria for Emergency Response Facilities 0700 - Guidelines for Control Room Design Reviews 0799 - Draft. Criteria for Preparation of Emergency Operating Procedures .
0801 - Evaluation Criteria for Control Room Design Reviews 0814 - Methodology for Evaluation of Emergency Response Facilities ,
0818 -
Emergency Action Levels for Light Water Reactors .
(
L . - 0835 , . Human Factors Acceptance Criteria for SPD5 Regulatory Guides . .
1.23 (Rev. 1) - Meteorological Measurement Program for Nuclear Power Plants 1.97 (Rev. 2) - Instrumentation for Light-Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following .
an Accident 1.101 (Rev. 2) -
Emergency Planning for Nuclear Power Plants 1.47 - Bypassed and Inoperable Status Indication for Nuclear Power.
Plant Safety Systems 9
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- 3. C'00RDINATION AND INTEGRATION OF INITIATIVES
- 1. The design of the Safety Parameter Display System (SPDS), design of instrument displays based on Regulatory Guide 1.97 guidance, control room design review, development of symptom oriented emergency operating proce-dures, and operating staff training should be integrated with respect to ,
the overall enhancement of operator ability to comprehend plant conditions l and cope with emergencies. Assessment of information needs and display
- formats and 1ccations should be performed by individual licensees. The SPDS could affect other control room improvements that licensees may consider. In some cases, a good SPDS may obviate the need for large-scale control room modifications. 14owever, installation of the SPDS should not be delayed by slower progress on other initiatives. The SPDS should not be contingent on completion of the control room design review. NRC does not plan to impose additional requirements on licensees regarding SPDS.
- 2. Implementation of part or all of Regulatory Guide 1.97 (Rev. 2) represents a control room improvement. The implementation of control room improve-ments is not contingent on implementing Technical Support Center (TSC) and Emergency Operations Facility (EOF) requirements.
- 3. The Technical Support Center (TSC) and Emergency Operations facility (EOF)
\ are dependent on control room improvements in terms of communication and
( instrumentation needs among the TSC, EOF, and control room. TSC and EOF
. facilities are not necessarily dependent on each other. The Operational-Sbpport Center (OSC) is independent of TSC and EOF.
- 4. The three groups of initiatives--SPDS, control room improvements, and ,
emergency response facilities (TSC, EOF, OSC)--should have the following interrelationships:
. a. The SPDS is an improvement in the control room because it enhances operator ability to comprehend plant conditions and interact in situations that require human intervention. The SPDS could affect -
other control room improvements that licensees may consider. In some cases, a good SPDS could obviate the need for extensive modifications t
. to control rooms.
- b. New instrumentation that may be added to the control room should be considered a requirement for inclusion in the design of the TSC and EOF only to the extent that such instrumentation is essential to the t
performance of TSC and EOF functions. _
- c. The SPDS and control room improvements are essential elt.ments in operator training programs and the upgraded plant-specific emergency j .
operating procedures. ,
l i (( d. Acquisition, processing, and. management of data for SPDS, control room improvements, and emergency response facilities should be coordinated but need not be centralized. -
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. 5. Specific implementation plans and reasonable, achievable schedules she'uld be established by' agreement between the NRC Project Manager and each individual licensee. The NRC office responsible for implementing each .
requirement should develop procedures identifying the foliosing:
- a. The respective rol'es of NRR, IE, and Regional Offices in managing implementation, checking licensee rate of progress, and verifying compliance, including the extent to which NRC review and inspection is necessary during impi oentation.
- b. Procedural methods and enforcement measures that could be used to
. ensure NRC staff and licensee attention to meeting mutually agreed upon schedules without significant delays and extensions.
- 6. The NRC Project Manager for each nuclear power plant is assigned program management responsi 9ity for NRC staff actions associated with imple-
-menting emergency ; Jense initiatives. The NRC Project Manager is the principal contact for the licensee regarding these initiatives.
- 7. NRC will make allowances for work already done by licensees in a good-
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faitn effort Ic meet requirements as they unoerstanc tnem. For each case in which a licensee would have to remove or rip out emergency response facilities or equipment that was installed in good faith to meet previous
.{ guidance in order to meet the basic requirements described in this docu-ment, the Director of the Office of Nuclear Reactor Regulation or Inspe2- ,
. tion and Enforcement will review the circumstances and determine whether removal is recnsary or existing facilities or equipment represent an acceptable alternative. Any regulatory position that would require the
- removal or major modification of existing emergency response facilities or equipment requires the specific approval of the Office Director.
I
- 8. NRC recognizes that acceptable alternative methods of phasing and inte-grating emergency response activities may be developed. Each licensee .
needs flexibility in integrating these activities, taking into account the varying degree to which the licensee has implemented past requirements and guidance. An example of a way in which these activities could be inte-grated is discussed below. Other methods of integration proposed by licensees would be reviewed considering licensees' progress on each .
4 initiative. .
< a. SPDS l
(1) Review the functions of the nuclear power plant operating staff that are necessary to recognize and cope with rare events that
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(a) pose significant contributions to risk, (b) could cause j
operators to make cognitive errors.in diagnosing them, and (c) are not included in routine operator training programs. .
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i (2) Combine the results of this review with accepted human factors prihciples to select parameters, data display, and functions to be incorporated in the SPDS.
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(3) . Design, build, and install the SPDS in the control room and
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train its users.
b.. Tc be done parallel without delaying SPDS, complete emergency opera-ting procedure technical guidelines that will be used to develop plant-specific emergency operating procedures.
- c. Using these_E0P technical guidelines, the SPDS design, and accepted '
human factors principles, conduct a review of the control room design. Apply the results of this review to:
(1) Verify SPDS paramete'r selection, data display, and functions.
(2) Develop plant-specific E0Ps.
(3) Design control room modifications that correct conditions ;
adverse to safety (reduce significant contributions to risk),
and add additional instrumentation that may be nectssary to ,
implement Regulatory Guide 1.97.
(4) Train and qualify plant operating staff regarding EDPs and modifications.
- d. Verify, prio- te finali ation of designs for modifications and of procedures and training, that the functions of control roon operators 4
in emergencies can be accomplished (i.e., that the individual initia-
-( tives have been integrated sufficiently to meet the needs of control
. room operators and provide adequate emergency response capabilities).
- e. Implement E0Ps and install centrol room modifications coincident with scheduled outages as necessary, and train operators in advance of ,
these changes as they are phased into operation.
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- 4. SAFETY PARAF.ETER DISPLAY SYSTEM (SPDS)
Current Reculatorv Recuirement_s .
4
. No licensee action is required.
Functional Statement The SPDS should provide a concise display of critical plant variables to the control room operators to aid them in rapidly and reliably determining the safety status of the plant. Although the SPDS will be operated during normal
. operations as well as during abnor:ial conditions, the principal purpose and function of the SPDS is to aid the control room personnel during abnormal and emergency conditions in determining the safety status of the plant and in assessing whether abnormal conditions warrant corrective action by operators to avoid a degraded core. This can be particularly important during , anticipated transients and the initial phase of an accident. .
Recommended Recuirements
- 1. Each operating reactor shall be provided with a Safety Parameter Display System that is located convenient to the control room operators. This
- system will continuously display information from which the plant safety
^
[ status can be readily and reliably assessed by control room personnel who a,re. responsible for the avoidance of degraded and damaged core events.
- 2. The control room instrumentation required (see General Design Criteria 13 and 19 of Appendix A to 10 CFR 50) forms the basic safety components '
required for safe reactor operation under normal, transient, and accident conditions. The SPDS is used in addition to the basic components and serves to aid and augment these components. Thus, requirements applicable to control room instrumentation are not needed for this augmentation (e.g. , GDC 2, 3, 4 in Appendix A; 10 CFR Part 100; single-f ailure require- -
ments). The SPDS need not meet requirements of the single-failure criteria and it need not be qualified to meet Class 1E requirements. The SPDS shall be suitably isolated from electrical or electronic interference with equipment and sensors that are in use for safety systems. The SPDS need not he seismically qualified, and additional seismically qualified .
indication is not required for the sole purpose of being a backup for SPDS. ' After the SPDS has been installed, operating procedures should be available that will allow timely and correct safety status assessment when the SPDS is not available. .
- 3. There is a wide range of useful information that can be provided by varfous systems. This information is reflected in such staff documents as NUREG-0596, NUREG-0835, and Regulatory Guide,1.S7.
Prompt. implementation of an SPDS can p'rovide an important contrib' u tion to i ,
plant safety. The selection of specific information that should be provided for k particular plant shall be based on engineering judgment of individual plant-licensees, taking into account the importance of prompt
- implementation. ,
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. The SPDS display shall be designed to incorporate accepted human factors principles comprehended sobythat SPDS theusers.
displayed information can be readily perceived and
- 5. '
Minimum information tion to plant operatorsto be provided shall be sufficient to provide informa-about:
- a. Reactivity control b.
- c. Reactor core cooling and heat removal from the primary system d.
Reactor coolant system integrity Radioactivity control -
- e. Containment conditions'
- The specific parameters to be displayed shall be determined by the licansee.
5.
The licensee shall prepare a written safety analysis describing the basis on which the selected parameters are sufficient to assess the safety status include of each identified symptoms of severe function for a wide range of events, which accidents.
f Such analysis, along with the
( specific implementation plan for SPDS shall be reviewed as described below.
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- 7. '
.in accordance with the licensee's technical specificat "whether the changes involve an unreviewed safety question or change of technical specifications.
normal fashion with prior NRC review.If they do, they shall be processed in the If the changes do not involve an the licensee may implement such changes without pr However, the licensee's analysis shall be submitted to NRC promptly on completion of review by the licensee's offsite committee.
Based on the '
results of NRC review, the Director of IE or the Director of NRR may question is posed.by the licensee's proposed system, o analysis is seriously inadequate.
Inteoration .
Prompt implementation of an SPDS is as design goal and of primary importance.
- The. schedule for implementing SPDS should not be impacted by schedules for the control room design review and development of symptom-oriented -
emergency operating procedures. For this reason, licensees should develop and propose an integrated schedule for implementation in which the SPDS design is an input to the other initiatives. If reasonable, f-
. this schedule should be accepted by NRC. -
( Reference Dotuments -
-- Need for SPDS identified .
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March 11,1982 i TV s' .R 1 g ($-) j SECY-82-111
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POLICY ISSUE For: (Notation Vote). LICENSING DMSl0N .
.The Commissioners bIdr,iiiRY
, From: William J. Dircks Executive Director for Operations Sub.iect : REQUIREMEhTS FDR EMERGENCY RESPONSE CAPABILITY Puroese: To request Commission approval of a set of basic re-quirmnents for energency response capability and approval for the staff to work with licensees to develop plant-specific impimentation schedules.
Discussion: One of the first issues reviewed by the Concittee to Review Gene-ie Recuirenen*.s (CRGR) was the broad arst cf ame gency resoonst facilities anc cacaoflities at nucisar plants. The Committee found tnat imple-
-(
. mentation schedules were not being coordinated vithin the NRC. In additior., existing NRC documents published as guidance to licensees were sometimes beine used as firT recuirements. Discussions with incust y represencatives and the staff indicated that some licenstees had slowed down on work in this area pending NRC clarification of its requirements.
.Some utilities have virtually stopped work on some of the itens, while others have proceeded and, in some cases, completed some of the items. The Committee reconnended *Jmt steps be taken by the Office Directors involved to clarify the requirements and implementation schedules for the Safety Parameter Display System (SPDS), Control Room Design Review, upgraded Eme gency Operating Procedures, Regulatory Guide 1.57, Technical Support Center (TSC), Operational Support Center (OSC), and Emergency Operations Facility (EOF). In my memo to the Commission dated Deceber 31, 1981, I noted that the DEDROGR staff would work with the program offices to clarify the basic requirements in this area and esablish a revised implementation plan.
V l
Enclosed are the staff's recommendations for the requirements in the broad area of emergency response facilities and capabilities outlined above. The
( requirements were developed by the program offices ,
Contact:
V. Stallo, Jr., DEDROGR 49-29704
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4 The Comnissioners and are supportec ey CRGR. Tne enclosure represents a distillation of fundamental requirements from the -
broad range of guidance documents that NRO has issued (principally NUREG reports and Regulatory Guides) . The staff intends that the guidance documents referred to in the enclosure not be used to impose requirements on licensees, but rather that they be l used as sources of guidance for NRC reviewers and '
licensees regarding acceptable means for meeting the :
fundamental requirements proposed. )
i In discussions with owners' groups and individual licensees, the staff has learned that the Commission
, approved schedule of October 1,1982, for implementation of the TSC and EOF probably cannot be met. In <
recognition of this fact and the difficulty of implementing generic deadlines, the staff is proposing that plant-soecific schedules be established which
~. ant in= accoun*, irs unious s. acus oi nact, pian ~..
Eacn licensee would be requested te submit a p oposed schedule for completing the actions to comply with
( the fundamental reouirements. The NRC Prejact Manage- for each plant should be knowledgeable of ,
the ove all work e* fort going on at a plant and, based on guidance received from HR0 management, could reach agreement with licensees on schedules which optimize use of utility and NRO resources. The agreed 4rpon comoletion dates would be formalized in .
an order. By this approach, future staff coordination problems regarding implementation schedules will be avoided.
Resource The costs to licensees to implement the requirements Estimates: proposed in the enclosure were included in the estimates set out in NUREG-0660.
Recommendation: That the Commission:
I
- 1. Anorove the fundamental requirements described in the enclosure.
- 2. Anorove the issuance of the requirements in the enclosure by 50.4f letters as a revision to NUREG 0737.
- 3. Accreve the method for establishing plant-( specific implementation schedules described in the enclosure.
( '
The Commissioners 4 Acoreve tne implemenca-ior. of these requirements tnrougn plar.t-specific orders.
- 5. Note that the staff intends to use the previously
- issued NUREG reports and Regulatory Guides as guidance documents only.
Scheduline- Licensees are currently required to establish a TSC and EOF by October 1. Prompt action on this paper is required in order to provide guidance to licensees.
William J. Dircks Exec::tive Director for Operations
Enclosure:
NR; 3 aff Recomenaatier.
on tne kequiramen s for Emergency Response Capability z
l 4
Cosm:issioners' comen.s should be providec directly to the Of* ice of the Secretary by c.c.b. Mondav, March 29, 1982.
Cons:ission Staff Office comunents, if any, should be submittad to the-Casurissioners NLT Monday, March 22, 1982, with an information copy to the Office of the Secretary. If *.he paper is of such a nature that it requires additional time for analytical review and comment, the Consissioners and the Secretariat should be apprised of when coments may be expected.
DISTRIBUTION counrissioners Comarission Staff Offices Exec Dir for Operations l Exec Legai Director
Secretariat S
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(: Specified SPDS NURdG-0737 .
NUREG-0596 --
Functional criteria for SPDS ,
NUREG-OS35 --
, Specific a::eptan:e criteria keyed te 0596
< Reg. Guide 1.97 (Rev. 2) --
Instrumentation for Light-Water Cocied Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident e
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- 5. DETAILED CONTROL ROOM DESIGN REVIEW Current Reculatorv Reouirements As specified in Item I.D.1 in NUREG-0737, the implementation schedule is still j.o be developed.
Functional Statement The objective of the control room design review is to " improve the ability of nuclear power plant control room operators to prevent accidents or cope with accidents if they occur by improvirig the information provided to them" (from NUREG-066n, Item I.D.1). As a complement to improvements of plant operating staff capabilities in response to transients and other abnormal conditions that will result from implementation of the SPDS and from upgraded emergency opera-ting procedures, this design review will identify any modifications of control room configurations that would contribute to a significant reduction of risk i
and enhancement in the safety of operation. Decisions to modify the control room would include consideration of long-term risk reduction and any potential temcorary decline in safety after modifications resulting from the need to relearn ma'intenance anc opera-ing pro:edures. This should be carefully
, reviewed by persons competent in human factors engineering and risk analysis.
I
( Recommended Reovirements
- 1. fonduct a control room design review to identify human engineering dis-crepancies. The review shall consist of:
- a. The establishment of a qualified multidisciplinary review team and a review program incorporating accepted human engineering principles.
- b. The use of function and task arialysis (that had been used as the basis for developing emergency operating procedure Technical Guide- .-
lines) to identify control room operator tasks and information and control requirements during emergency operations. This analysis has multiple purposes and should also serve as the basis for developing training and staffing needs and verifying SPDS parameters.
- c. A comparison of the display and control requirements with a control room inventory to identify missing and surplus (distracting) displays and controls. -
- d. A control room survey to identify deviations from accepted human factors principles. This survey will include, among other things, assessment of control room layout, the usefulness of audible and visual alarm systems, information recording and recall capability, i and control room environment.
1 . .
- 2. Assess which human engineering dis'crepancies are significant and should be
. corrected. Select design improvements that will correct those discrep-ancies. Improvements that can be accomplished with an enhancement program
- (paint-tape-label) should be done promptly.
l l
N. ,
i 11 . i
( ,
. 3. ' Verify that each selected design improvement will provide the necessary ;
correction, and can be introduced in the control room without creating any unacceptable human engineering discrepancies because of significant ,
contribution to increased risk, unreviewed safety questions, or situations in which a temporary reduction in safety could occur. Improvements that are introduced should b'e coordinated with changes resulting from other improvement programs such as SPDS, operator training, new instrumentation i (Reg. Guide 1.97, Rev. 2), and upgraded emergency operating procedures, i l
Documentation and NRC Review
. 1. All licensees shall submit a program plan within two months of the start of the control room review that describes how items 1, 2 and 3 above will be accomplished. NRC approval is not required before licensees conduct their reviews. !
- 2. Selected licensees will undergo an in progress audit by the NRR human factors staff based on the program plans and advice from resident t inspectors. ;
j 3. All licensees shall submit a summary report outlining proposed control ;
room changes. The report will also provide a summary justification for !
. human engineering discrepancies with safety significance to de left !
{ uncorrected or partially corrected.
- 4. , Mthin two weeks after receipt of the licensee's summary report, the NRC will inform the licensee whether it will conduct a' pre-implementation onsite audit. The decision will be based on the content of the program plan, summary report, and results of NRR in progress audits if any. The licensee selection for pre-implementation audit may or ma; at include licensees selected for in progress audits under paragraph 2. ,
i
- 5. For control rooms selected for pre-implementation onsite audit, within one j month after receipt of the summary report, the NRC will conduct: -
j
- a. A pre-implementation audit of proposed modifications (e.g. , equipment additions, deletions and relocations, and proposed modifications). :
7
- b. An audit of the justification for those human engineering discrep- [
i ancies of safety significance to be left uncorrected or only l partially corrected.
The audit will consist of a review of licensee's record of the control !
room reviews, discussions with the licensee review team, and usually a '
control room visit. Within a month after this onsite audit, NRC will i issue its safety evaluation report (SER). ,
.( 6. For control rooms for which NRC does not perform a pre-implementation !
onsite audit, NRC will conduct a review and issue its SER within two l 1
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' months after receipt of the lic'ensee's summary report. The review shall be simila'r to that conducted for pre-implementation plants under para-graph 5 above, except that it may or may not include a specific audit.
The SER shall indicate whether, based on the review carried out, changes in the licensee's modification plan are needed to assure operational safety. Flexibility is considered in the control room review, because certain control board discrepancies can be overcome by techniques not involving control board changes. These techniques could include' improved procedures, improved training, or the SPDS. .
- 7. The following approach will be used f6r OL review. For OL applications with SSER dates prior to June 1983, licensing may be based on either a Preliminary Design Assessment or a Control Room Design Review (CRDR) at
- the applicant's option. However, applicants who choose the Preliminary Design Assessment option are required to perform a CRDR after licensing.
For applications with SSER dates after June 1983, Control Room Design -
Review will be required prior to licensing.
l Intecration
. Promot im'iemer.tatier.
o of ar. SpDS is a desier. cotl and of prima y imaortan:e.
~
l The schedule for. implementing SpDS should not be impacted by schedules "
for the control room design review and development of symptom' or.ented - .
( emeroency operating procedures. For this reason, licensees should .
4
- devef oo and propose an integrated schedule for implementation in which the SPDS design is an input to the other initiatives. . If reasonable, .
this schedule should be accepted by NRC. ,
Reference Documents ,
NUREG-0585 --
States that licensees should conduct review.
NUREG-0660, Rev. 1 -- States that NRR will require reviews for operating '
reactors and operating licensee applicants.
NUREG-0700 --
Final guidelines for CRDR.
-- States that requirement was issued June, 1980, final'.
guidance not yet issued.
. NUREG-0801 -- October 1981 draft for comment; staff evaluation .
criteria. .
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REGULATORY GUIDE 1.97
- 6. APPLICATION TO EMERGENCY RESPONSE FACILITICS
. Currant Reculatorv Reevirements No licensee action is required.
Functional Statement Regulatory Guide 1.97 provides data to assist control room operators in pre-4 - venting and mitigating the consequences of reactor accidents.
Recommended Recuirements
- 1. Control Room ,
Provide measurements and indication of Type A, B, C, D, E variables listed -
in Regulatory Guide 1.97 (Rev. 2). Individual licensees may take excep-tions based on plant-specific design features. BWR incore thermocouples
- and continuous offsite dose monitors are not required pending their further development and consideration as requirements. It is acceptable g
to rely on currently installed equipment if it will measure over the range ~
- indicated in Regulatory Guide 1.97 (Rev. 2), even if the equipment is p_resently not environmentally qualified. Eventually, all the equipment required to monitor the course of an accident would be environmentally qualified in accordance with the pending Commission rule on environmental qualification. .
Provide reliable indication of the meteorological variables (wind direc-tion, wind speed, and atmospheric stability) specified in Regulatory Guide 1.97 (Rev. 2) for site meteorology. No changes in existing meteoro-logical monitoring systems are necessary if they have historically provided reliable indication of these variables that are representative of .
meteorological conditions in the vicinity of the plant site. Information on meteorological conditions for the region in which the site is located shall be available via communication with the National Weather Service.
- 2. TechnicalSucoortCenter(TSCI The Type A, B, C, D, E variables that are essential for performance of TSC functions shall be indicated in the TSC.
- a. BWR incore thermocouples and continuous offsite dose monitors are not required pending their further developmsnt and consideration as ;
requirements.
(4 b. The indicators and associated circuitry shall be of reliable design but need not meet Class 1E, single-failure or seismic qualification
. requirements.
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- 3. Emergency Operations Facility (EOF)
- a. Those primary indicators needed to monitor containment conditions and releases of radioactivity from the plant shall be provided in the EOF.
- b. The EOF data indications and associated circuitry shall be of reliable design but need not meet Class 1E, single-failure or seismic qualification requirements.
Documentation and NRC Review ,
NRC review is not a prerequisite for implementation. Staff review will be in the form of an audit that udll include a review of the licensee's method of implementing Regulatory Guide 1.97 (Rev. 2) guidance and the licensee's sup-porting technical justification of any proposed alternatives.
t
- The licensee sh'all submit a report describing how it meets these requirements.
The submittal should inclade documentation which may be in the form of a table that includes the following information for each Type A, B, C, D, E variable
- shown in Regulatory Guioe '_.S7 (Rev. 2):
} (a) instrument range -
(b) environmental qualification (as stipulated in, guide or state criteria)
(c) seismic qualification (as stipulated in guide or state criteria) -
(d) q'uality assurance (as stipulated in guide or state criteria) '
(e) redundancy and sensor (s) location (s)
(f) power supply (e.g. , Class IE, non-Class 1E, battery backed) '
(g) location of display (e.g., control room board, SPDS, chemical laboratory)
(h) schedule (for installation or upgrade) i Deviations from the guidance in Regulatory Guide 1.97 (Rev. 2) should be explicitly shown, and supporting justification or alternatives should be ,
presented.
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- 7. UPGRADE EMERGENCY OPERATING PROCEDURES (EOPs)
Current Reculatory Recuirements
.' NUREG-0737. Item I.C.1, which has been approved by the Commission for imple-mentation.
Functionel Statement Symptom-based emergency cperating procedures will improve human reliability and the ability to mitigate the consequences of a broad range of initiating events
- . and subsequent multiple failures or operator errors.
Recommended Recuirements
- 1. In accordance with NUREG-0737, Item I.C.1, reanalyze transients and accidents and prepare Technical Guidelines. These analyses will identify operator tasks, and information and control needs. The analyses also -
serve as the basis for integrating upgraded emergency operating procedures and the control room design review and verifying the SPDS design.
/~ 2. Upgrade E0Ps to be consistent witn Tecnnical Guidelines and an approp-iate
- procedure Writer's Guide. ,
(
- 3. Provide appropriate training of operating personnel on the use of upgraded
. E0Ps prior to implementation of the E0Ps.
- 4. Implement upgraded E0Ps. .
Documentation and NRC Review
- 1. Submit Technical Guidelines to NRC for review. NRC will perform a pre-
- implementation review of the Technical Guidelines and the Writer's Guide.
Within two months of receipt of the Technical Guidelines and Writer's .
Guide, NRC will advise the licensees of their acceptability.
- 2. Each licensee shall submit to NRC a procedures generation package at least three months prior to the date it plans to begin formal operator training .
on the upgraded procedures. NRC approval of the submittal is not necessary prior to upgrading and implementing the E0Ps. The procedures generation package shall include:
- a. Plant-Specific Technical Guidelines - plant-specific guidelines for
. plants not using generic technical guidelines. For plants using
. generic technical guidelines, a description of the planned method for
- developing plant specific E0Ps from the generic guidelines, including
- plant specific information. .
( b.
A Writer's Guide that details the specific methods to be used by the licensee in preparing E0Ps based on the Technical Guidelines.
4
.g y -- --- --- --..- - , -- - --,-
16
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- c. A description of the program for validation of the E0Ps.
- d. A brief description of the training program for the upgraded E0Ps.
- 3. All procedures generation packages will be reviewed. On an audit basis for selected facilities, upgraded E0Ps will be reviewed. The details and extent of this review will be based on the quality of the procedures generation packages submitted to NRC. A sampling of ugpraded E0Ps will be reviewed for technical adequacy in conjunction with the NRC Reactor Inspection Program.
Reference Documents NUREG-0660, Item I.C.1, I.C.8, I.C.9 NUREG-0799
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- 8. ' EMERGENCY RESPONSE FACILITIES Current Reculatorv Reouirements* ,
10 CFR 50.47(b)(6) (for Operating License applicants) -- Requirement for prompt
~
l communications among principal response organizations and to emergency personnel and to the public.
I 10 CFR 50.47(b)(8) -- Requirement for emergency facilities and equipment to support emergency response.
. 10 CFR 50.47(b)(9) -- Requirement that adequate methods, systems and equipment i for assessing and monitoring actual or potential offsite consequences of a I
radiological emergency condition are in use.
10 CFR 50.54(q-) (for Operating Reactors) -- Same requirement as 10 CFR 50.47(b)
, plus 10 CFR 50, Appendix E.
10 CFR 56, Appendix E, Paragraph IV.E Requirement for:
"1. Equipment at the site for personnel monitoring;
-( "2. Equipment for determining the magnitude of and for continuous 1'y
. assessing the impact of the release of radioactive materials to the environment; .
"3. Facilities and supplies at the site for decontamination of
- onsite individuals; "4. Facilities and medical supplies at the site for appropriate emergency first aid treatment; "S. Arrangements for the services of physicians and other medical personnel qualified to handle radiation emergencies on site; "6. Arrangements for transportation of contaminated injured individ-uals from the site to specifically identified treatment facili .
ties outside the site boundary; "7. . Arrangements for treatment of individuals injured in support.of licensed activities on the site at treatmert facilities outside the site boundary; -
- " 8. A licensee onsite technical support center and a licensee
- near-site emergency operations fac.ility from which effective j
direction can be given and effective control can be exercised -
w . during an emergency; . !
. "S. At ieast one onsite and one offsite communications system; each system shall have a backup power source.
i 18 i . ,
I ~
.All communication plans shall have arrangements for emergencies, including titles and alternates for those in charge ,at both ends of the communication links and the primary and backup means of communication. Where consistent with the function of the governmental agency, these arrangements will include:
" a. Provision for communications with contiguous State / local governments within the _ plume exposure pathway (emergency planning zone) EPZ. Such communications shall be tested monthly.
"b. Provision for c'ommunications with Federal emergency I response organizations. Such communications systems shall !
be tested annually.
"c. Provision for communications among the nuclear power ,
reactor control room, the onsite technical support center, and the near-site emergency operations f acility; and among the nuclear facility, the principal State and local emer-gency operations centers, and the field assessment teams.
- Such communications systems shall be tested annually.
" d. Provision for communications by the licensee with HRC .
( Headquarters and the appropriate NRC Regional Office Operations Center from the nuclear power reactor control t room, the onsite technical support center, and the near- ,
site emergency operations facility.' Such communications shall be tested monthly." ,
Within this section on emergency response facilities, the Technical Support Center (TSC), Operational Support Center (OSC) and Emergency Operations
. Facility (EOF) are addressed separately in terms of their functional statements -
and recommended requirements. The subsecticns on Documentation and NRC Review and Reference Documents that follow the EOF discussion apply to this entire .
section on ~ emergency response facilities.
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- Technical Support Center (T5C)
Functional Statement
.' The TSC is the onsite technical support center for emergency response.
activated, the TSC is staffed by p* redesignated technical, engineering, senior When
' management, and other licensee personnel, and five predesignated NRC personnel.
During periods of activation, the TSC will operate uninterrupted to provide l
plant management and technical support to plant operations personnel, and to l relieve the reactor operators of peripheral duties and communications not
! directly related to reactor system manipulations. The TSC will perform EOF
- functions for the Alert Emergency class and for the Site Area Emergency class and General Emergency class until the EOF is functional.
Recommended Recuirements The TSC will be: -
l
- 1. Located within the site protected area so as to facilitate necessary interaction with control room, OSC, EOF and other personnel involved with the emergency.
- 2. Sufficient to accommodate and support NRC and licensee predesignated
-( personnel, equipment and documentation in the center.
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- 3. SItructurally built in accordunce with the National Uniform Buf1 ding Code.
- 4. Environmentally controlled to provide room air temperature, humidity and -
cleanliness appropriate for personnel and equipment.
- 5. Provided with radiological protection and monitoring equipment necessary to assure that radiation exposure to any person working in the TSC would not exceed 5 rem whole body, or its equivalent to any part of the body, -
for the duration of the accident. .
- 6. Provided with reliable voice and data communications with the control room and EOF and reliable voice communciations with the OSC, NRC Operations Centers and state and local operations centers. .
- 7. Capable of reliable data collection, storage, analysis, display and communication sufficient to determine site and regional status, determine changes in status, forecast status and take appropriate actions. The ^ -
following variables shall be available in the TSC:
(a)
- the variables in the appropriate Table 1 or 2 of Regulatory
- Guide 1.97 (Rev. 2) that are essential for performance of TSC functions; and (4 ~
(b) the meteorological variables in Regulatory Guide 1.97 (Rev. 2) for
. site vicinity and National Weather Service data available by voice communication for the region in which the plant is located.
20
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Principally those data must be available that would enable evaluating incident sequence, determining mitigating actions, evaluating damages and cetermining plant status during recovery operations.
- 8. Provided with accurate, complete and current plant records (drawings, schematic diagrams, etc.) essential for evaluation of the plant under accident conditions.
- 9. Staffed by sufficient technical, engineering, and senior designated licensee officials to provide needed support, and be fully operational
- within approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> af,ter activation.
- 10. Designed taking into account good human factors engineering principles.
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Operational Support Center (OSC)
Functional Statement When activated, the OSC will"be the onsite area separate from the control room I where predesignated operations support personnel will assemble. A predesignated licensee official shall be respensible for coordinating and assigning the personnel to tasks designated by control room, TSC or EOF personnel.
. Recommended Recuirements The OSC will be:
- 1. Located onsite to serve as an assembly point for support personnel and to facilitate . performance of support functions and tasks.
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Emergency Operations Facility (EOF)
Functional Statement .
The EOF is a licensee controlled and operated facility. The EOF provides for management of overall licensee emergency response, coordination of radiological, and environmental assessment, determination of recommended public protective actions, and coordination of emergency response activities with Federal, State, -
and local agencies.
When the EOF is activated, it will be staffed by predesignated emergency personnel identified in the emergency plan. A designated senior licensee
~
official will manage licensee activities in the E0F.
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 radioactive releases from the plant and to determine dose projections.
< Recomended Recuirements
( The EDF will be:' ,
~
- 1. l.ocated and provided with radiation protection features as described in -
Table 1 (previous guidance approved by the Commission) and with appropriate radiological monitoring systems.
- 2. Sufficient to accommodate and support Federal, State, local and licensee predesignated personnel, equipment and documentation in the EOF.
. 3. Structurally built in accordance with the National Uniform Building Code. .
~
- 4. Environmentally controlled to provide room air temperature, humidity and cleanliness appropriate for personnel and equipment. ;
4
- 5. Provided with reliable voice and data communications facilities to the TSC and control room, and reliable voice communication facilities to OSC and ,
to NRC, State and local emergency operations centers.
- 6. Capable of reliable collection, storage, analysis, displays and communica-tion of information on containment conditions, radiological releases and meteorology sufficient to determine site and regional status, determine changes in status, forecast status and take appropriate actions.
Variables from the following categories that are essential to EOF
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functions shall be available in the EOF: .
(4 .
variables from the appropriate Table 1 or 2 Regulatory Guide 1.97 4
(a)
(Rev. 2), and O
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. - - - - , . , , n - - - , - , , . - , - - , . - , . - . , , - - , - - , ..- ----y -- -
23 .
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t (b) the meteorological variable's in Regulatory Guide 1.97 (Rev. 2) for i site vicinity and regional data available via communication from the National Weather Service. . ;
.' 7. Provided with up to date plant records (drawings, schematic diagrams, etc.), procedures, emergency plans and environmental information (such as i geophysical data) needed to perform EOF functions. j t
- 8. Staffed in accordance with Table 2 (previous guidance approved by the l Commission). Reasonable exceptions to the 30-minute and 1-hour time limits for staffing should be justified and will be considered by NRC l
staff. ;
I
- 9. Provided with industrial security when it is activated to exclude ,
unauthorized personnel and when it is idle to maintain its readiness. [
- 10. Designed taking into account good human factors engineering principles. -
l
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Documentation and NRC Review
- The conceptual design for emergency response facilities (TSC, OSC, and EOF) l have been submitted to NRC for review. In many cases, the lack of detail in
., i these submittals has precluded an NRC decision of acceptability. Some designs !
\ have been disapproved because they clearly did not meet the intent of the >
applicable regulations. NRC does not intend to approve each design prior to !
implecientation, but rather has provided in this document those " recommended i i
requirements" which should be satisfied. These recommended requirements C
, provided a degree of flexibility within which licensees can exercise management I prerogatives in designing and building emergency response facilities (ERF)'that satisfy specific needs of each licensee. The foremost consideration regarding f I
ERFs is that they provide 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. . j Licensees should assure that the design of ERFs satisfies these basic.
- j requirements. Exemptions from or alternative methods 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 i according to schedules that will 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 capable of performing their intended functions. Licensees need not document their actions on each specific l item contained in NUREG-0696 or 0814.
- t
- Reference Documents (Emergency Response Facilities) [
[( 10 CFR 50.47,(b) -- Requirements for emergency facilities and equipment *for OLs. f;
[
10 CFR 50.54(q) an( Appendix E, Paragraph IV.E -- Requirements for emergency !
facilities and equipment for ors. .
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NUREG-0550 -- Description of and imp 1'ementation schedule for TSC, OSC and EOF.
Eisenhut letter to power reactor licensees 9/13/79 -- Request for com=itment to meet requirements.
Denton letter to power reactor licensees 10/30/79 -- Clarification of requirements and implementation schedule.
Eisenhut letter to power reactor licensees 4/25/B0 -- Clarification of requirements. ,
NUREG-0554 -- Radiological Emergency Response Plans NUREG-0596 -- Functional criteria for emergency response facilities.
NUREG-0737 -- Guidance on meteorological monitoring and dose assessment. .
Eisenhut letter to power reactor license 2/18/81 -- Commission approved guidance on location, habitability and staff for emergency facilities.
Request and
' deadline for submittal of conceptual design of facilities.
( NUREG-0814 (Draft Report for Comment) -- Methodolo h for evaluation of .
. emerge,ncy response facilities. -
NUREG-0818 (Draft Report for Comment) -- Emergency Action Levels .
Reg. Guide 1.97 (Rev. 2) -- Guidance for variables to be used in selected emergency response facilities.
COPJA-80-37, January 21, 1981 -- Commission approval guidance on EOF loc.ation and habitability. ,
Secretary memorandum 58l-19, February 19, 1981 -- Commission approval of NUREG-0696 as general guidance only.
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TABLE 1 * ,
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_ EMERGENCY OPERATIONS' FACILITY
- Option 1 Option 2 .
- One Facility Two Facilities Reduce liabitability" o At or Deyond 10 miles.
\. Close-in, Primary:
o No special protection factor. ,
o within 10 miles o If beyond 20 miles, specific o protection factor = 5 approval required by the o ventilation isolation Commission, and some provi- -
with IIEPA (no charcoal) sion for NRC. site team closer*
to site. ,
o Strongly recommended location be coordinated with offsite authorities.
B. Backup EOF -
l , m o between 10-20 miles
- o no separate, dedicated facility '.
o arrangements for portable backup equipment ,
a strongly recommended location be coordinated with offsite authorities o continuity of dose projection and decision making capability For both Options:
- located outside security boundary
- - space for about 10 HRC employees
- none designated for severe phenomena, e.g. , earthquakes "llabitability requirements are only for the part of the EOF in which dose assessments communications and decision making take place.
If a utility has begun construction of a new building for an EOF that is located with 5 miles, that new ftcility is acceptable (with less than protection factor of 5 and ventilation isolation and ilEPA) provided .
thst a backup EOF siellar'to "B" in option 1 is provided. . ,
e
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TABLE.
MlHIMUN STAFFING REQUIREMENTS FOR NRC LICENSEES
. . . FOR N0 CLEAR POWER PLANT EMERGENCIES Capability for Additions
- Position Title On Major Functio'nal Area Major Tasks or Expertise Shift
- 30 min. 60 min.
Plant Operations and Shift suporvisor (SRO) 1 -- --
Shift foreman (SRO) 1 Assesshent of Operational Aspects Control room operators 2 -- -- -
- Auxiliary operators 2 -
Emergency Direction and Shift technical advisor, 1** -- --
shift supervisor, or Control (Emergency designated facility Coordinator)*
- manager Motification/ Nority licensee, state 1 1 ~2 no I Communication **** local, and federal personnel & maintain communication ,
- Radiological Accident Emergency operations Senior manager -- --
1 ~
- Assessment and Support facility (EOF) director of Operational Accident Offsite dose Senior health physics --
1 --
Assessment assessment (llP) expertise i -
- Offsite surveys --
2 2 -
Onsite (out-of plant)- --
1 1 Inplant surveys llP technicians 1 1 1 i Chemis try/ radio- Rad / chem technicians 1 --
1
- - chemistry N0lE
- Source of this table is NUREG-0654, " Functional Criteria for Emergency Response Facil.iLies."
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TABLE 2 (Cir. ) -
- - Capability for Additions .
. Posillon Title on Major Functional Area Major Tasks or Expertise Shift" 30 min. 60 min.
Plant System Technical support Shift technical advisory 1 -- --
Engineefing, Repair Core / thermal hydraulics --
1 --
Electrical -- --
1 -
. and Corrective Actions --
1 Hechanical --
Repair and corrective Mechanical maintenance / 1** --
1 actions Radwaste operator 1
. Electrical maintenance / 1** 1 1 - .
Instrument and control 1
~
(I&C) technician 1 --
Protective Actions Radiation protection: ilP technicians 2** 2 2 (In-Plant)' tj
- a. Access control
- b. IIP Coverage for repair, correc-2 - tive actions, search and rescue first-aid, & ,
- firefighting l c. Personnel monitor-ing
- d. Dosimetry , ,
Firefighting -- --
Fire local brigade support per techni-cal specifi-I cation i '
Rescue Operations .
2** Local .
and First-Ald -
support .
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Attaciiment 10(B) -5 NUREG/CR-2644 O ENICO-1110 :
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> 4 An' Assessment o" O " site, Rea-Time .
Jose Measuremen: Systems for ~
l Emergency Situations .
i 1
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l Prepared by W. J. Maeck, L G. Hoffman, B. A. Staples, J.11. Keller i Exxon Nuclear idaho Co., Inc. !
i Prepared for !
U.S. Nuclear Regulatory l
Commission .
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' This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the
. United States Government nor any agency thereof, or any of : ,
their employees, makes any warranty, expressed or implied, or I assumes any legal liability or resp onsibility for any third party's 5 use, or the results of such use, of any information, apparatus l
product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned .
l j rights.
l d
Available from GPO Sales Program Division of Technical Information and Documer. Control U. S. Nuclear Regulatory Commission j k'ashington, D. C. 20555 Printed copy price: $4.75 .
and i
-[- National Te::hnical Information Service Springfield, Virginia 22161 O
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NUREG/CR-2644 Q ENICO-1110 y-An Assessment of Offsite, Real-Ti~me Dose Measurement Systems for <
Emergency Situations
=
M:nuscript Completed: March 1982 D:te Published: April 1982 Prepared by W. J. Maeck, L G. Hoffman. B. A. Staples, J. H. Keller f~ Exxon Nuclear idaho Co., Inc.
f.O. Box 2800 f ttho Falls, ID 83401
' Prepared for Division of Systems integration Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 NRC FIN A6461 O
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$ U NOTICE r 1 Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the folio ing sources:
t 1. The NRC Public Document Room,1717 H Street, N.W. ,
g Washington, DC 20555
- 2. The NRCIGPO Sales Program. U.S. Nuclear Regulatory Commission.
Washington, DC 20555
- 3. The National Technical Info'rmation Service. Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive.
Refere'nced documents available for inspection and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers;and applicant and licensee documents and correspondence.
The following documents in the NUREG series are available for purchase from the NRCIGPO Sales Program: formal NRC staff and contractor reports, NRC sponsored conference proceedings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of Federal Regulations, and Nuclear Regulatory Commission issuances.
(j Documents available from the National Technical Information Service include NUREG series reports and techrical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.
Documents available from public and special technical libraries include all open literature items, such as books, journal and periodical articles, and transactions. Federal Register notices, federal and state legislation, and congressional reports can usually be obtained from these libraries.
Documents such as theses, dissertations, foreign reports and translations,and non NRC conference proceedings are available for purchase from the organization sponsoring the publication cited.
Single copies of NRC draft reports are available free upon written request to the Division of Tech-nical Information and Document Control, U.S. Nuclear Regulatory Commission, Washington, DC 20555.
f ' Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library,7920 Norfolk Avenue, Bethesda, Maryland, and are available there for reference use by the public. Codes and standards are usually copyrighted and may be
- purchased from the originating organization or, if they are American National Standards, from the American National Standards institute,1430 Brondway, New York, NY 10018.
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A8STRACT
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l e An evaluation is made of the effectiveness of fixed, real-time mon-itoring systems around nuclear power stations in determining the magni-tude of unmonitored releases. The effects of meteorological conditions hn the accuracy with which the magnitude of unmonitored releasks is de-termined and the uncertainties inherent in defining these meteorological' conditions are discussed. The number and placement of fixed field de- ,
tectors in a sys' tem is discussed, and the data processing equipment re-Quired to convert field detector output data into release rate informa-tion ,is det"Ibed. Cost data relative to the purchase and installation of specific systems are given, as 'well as the characteristics and in-formation return for a system purchased at an arbitary cost.
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SUMMARY
(( .
, The Nuclear Regulatory Commission has been considering a recuirement
. ,\
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that each operating commercial nuclea'r power station be _ fitted with an y effsite real-time emergency monitoring system. Currently, seve'ral power ,
stations have installed, or are-in the process of installing, mhnitoring irrystems of-varying degrees of complexity and sophistication. I Prior to deciding whether to reoutre all stations to install an 4
offsite real-time emergency monitoring system, the NRC recuested an independent evaluation of the usefullness of such a system and an assessment of the validity of the information obtained from the system.
The .information provided by this stuoy will be used to aid the NRC in their determination of whether or not to recuire that fixed offsite real-time. emergency monitoring systems be installed at all operating and planned e.cmcercial nuclear power stations.
1 This study addresses several aspects of the offsite real-time emer-gency monitoring system concept. The primary items receiving attention e in this study are:
- 1. The ability of a fixed real-time monitoring system to detect and cuantify monitored and unmonitored releases.
- 2. The ability of the system to detect and cuantify an unmonitored release in the presence of a known release.
- 3. An assessment of the uncertainties associated with estimating l the magnitude of an unmanitored release.
- 4. The number of stations reouired to detect a release and the h uncertainty associated with the detected value.
. t
- 5. The availability, cost, and the instrumentation reduirements for a system.
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An augmented effort of the study was to determine the characteris-n tics and information return that might be obtained from a close-in (0.5 g
(I mile) system with capital costs limited to $500,000.
/
t A matrix approach was used in this evaluation in which. the three
{ major parameters were,1) the measurement range of the oetect}r, 2) the
'. accuracy of the final results, and 3) the costs.
e The general conclusions from this study are presented below. The uncertainty estimates are based on the use of simple error analyses of the meteorological expressions reouired to describe plume shapes and -
atmospheric transport. , ,
- 1. While a ring of detectors around a nuclear power station can provide the means for monitoring releases; the number of sta-tions reouired for two detectors to provide information within
~
a factor of 5 of each other can be as large as 50 or more for one installation.
C
,h 2. The use of short-time (13 min) data from a fixed offsite moni-toring system to project downwind dose rates is a complex and highly uncertain process. Based on our study the uncertainty associated with a projected value is at least a factor of 10 or more.
- 3. The use of a fixed offsite monitoring system to determine the magnitude of an unmonitored release in the presence of a moni-tored release., is highly ouestionable. Depending on the ratio of the unmonitored release to the monitored release, uncertain-ties of factors of 25 and 50 are common.
f i 4. Several vendors of monitoring ecuipment were contacted relative to cost and performance characteristics of the available in-strumentation. In addition, we contacted several power stations and state agencies involved in the installation of fiixed real-(
time environmental monitoring systems. While the cost factors L) vi l -
l t
h
-;k for the instrumentation were relatively fixed, the installation costs were highly variable. Based on this study the cost per
/ monitoring station ranges from 525,000 to 565,000. De,pending upon the specific site characteristics the cost for.a 32 station '
p system could easily exceed 51,000,000 while only provid g data with uncertainties in the range of factors of 10 to 50
- 5. The placement of a simple limited (5500,000) detector system in .
proximity (0.5 mi) to a reactor may not provide reliable in-
-formation in the case of an emergency for several reasons. "Of prime importance is the limited number of stations (8-16) that could be installed and the consecuence that' a plume' might go undetected. A second serious oroblem, especially in the case of a BWR, is the building shine factor which could give a sufficiently high background signal to negate detection of the d
plume radiation.
h In general, it is highly ouestionable that a fixed station (16-32 i i units) emergency monitoring system can provide sufficiently reliable technical information to be of use in a decision-making process in the event of an emergency situation.
This conclusion should not preclude consideration of the installa-tion of such a system. A monitoring system could be used to develop site specific meterological information and could develop improved public relations with the populace. It should be emphasized, however, that the stations should 'be judiciously placed so as not to convey false-t information.
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l TABLE OF CONTENTS Qi ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . iii
SUMMARY
.......................... v
.1. 0 INTRODUCTION , . . . . . . . . . . . . . . . . . . . . . . ' . 1
! 1.1 Background ....................1 1
. 1.2 Objective ...................... I 1.3 Evaluation Criteria . . . . . . . . . . . . . . . . . . 2
. (
2.0 QUANTIFICATION AND ASSESSMENT OF THE UNCERTAINTIES ASSOCIATED WITH THE MEASUREMENT OF AN UNMONITORED RELEASE .......................... 5 2.1 Prediction of Downwind Atmospheric Concentration .
Values .................'....... 6 2.2 Prediction of ~ Downwind Atmos'oheric Dose Values . . . . 15 4 2.3 Uncertainties Associated with the Quantification of an Unmonitored Release . . . . . . . . . . . . . . 21 1 ;
3.0 DETECTOR PLACEMENT AND REQUIREMENTS . . . . . . . . . . . . 25 3.1 Detector Placement and Response Functions . . . . . . 25 3.2 Building Shine and Background . . . . . . . . . . . . 30 O 4.0 INSTRUMENTATION REQUIREMENTS, AVAILABILITY AND SYSTEM f(/ COSTS ........................... 34 t
4.1 Instrument Description and Recuirements . . . . . . . 34 I A.2 Instrument Availability . . . . . . . . . . . . . . . . 38 A.3 System Costs ..................... 40
(
5.0 MATRIX EVALUATION ..................... 48 6.0 MINIMUM COST EMERGENCY SYSTEM . . . . . . . . . . . . . . . 52 .
4 l
7.0 REFERENCES
. . . . . . . . . . . . . . . . . . . . . . . . . 56 '
APPENDIX A. BRIEF SUM 4ARY OF EXPERIMENTAL RESULTS -
TO COMPARE MEASURED AND ~ PREDICTED GROUND LEVEL I CONCENTRATION VALUES A.1 85Kr Experiment at Savannah River Plant . . . . . . A-1 g A.2 ORNL Assessment of Hanford Experiment . . . . . . . . A-1
. .- A.3 Excerpts from a Workshop on the Evaluation of Models used for Environmental Assessment of Radionuclide -
Releases ......................j. .
A-3
) iX
A.4 Results of a Survey of Programs for Radiological Dose Computations ..................... A-5 APPENDIX B. VALUES FOR oy AND cz USED IN 0
I-( - CALCULATION
/
FIGURES N
. Q
, i 1. Matrix Parameters ...................... 4
~L
- 2. Model to Evaluate the Esticate of an Unmonitored Release in the Presence of a Known Release ............. 6
- 3. Comparison of Short-Term Diffusion Factors (Stability Class A)
Depicted for 6 Different Diffusion Parameter Systems . . . . . 13 4 Comparison of Short-Term Diffusion Factors (Stability Class D)
Depicted for 6 Different Diffusion Parameter S'ystems . . . . . 13
- 5. Comparison of Short-Term Diffusion Factors (Stability Class F)
Depicted for 6 Different Diffusion Parameter Systems . . . . . 14
- 6. Projected Center Line Dose as a Function of Stability Class and Distance For a Ground Level Release ........... 19
- 7. Effect of Release Height on Dose Rate as a Function of Distance for Stability Class B, D, and F. h
.......... 20 J
- 8. Uncertainty in Calculated Values of an Unmonitored Release l in the Presence of a Monitored Release ........... 23
- 9. Plume Shape Analysis for Determining Detector Recuirements ........................ 26
- 10. Number of Detectors Reouired at 1600 m to give Response within 200, 300, and 500% . ................. 27
' 11. Detector Reauirements as a Function of Cloud Dose Gansna Ray Energy .................... 29
- 12. Effect of Building Shine on Detector System Response .... 32
- 13. Schematic of Offsite Monitoring System Basic Components . . . 35 14 Matrix Parameters ...................... 51 A-1 Measured to Predicted 85Kr Concentrations ......... A-2
, A-2 Comparison of Different Dose Calculation Models, Class F .. A-6 A-3 Comparison of Different Dose Calculation Models, Class C .. A-7 Q
) x
- a , -,~
TABLES
(( < -
i I. Errors in x (Ground-Level Average Concentration)
- _, for a One Unit Assignment Error in Stability' Class . . . . .
9 l
t II. Errors in X (Ground-Level Average Concentration)
- -for a One Unit Assignment Error in Stability Class . . i. . . .. 10
-} III. Variability in Stability Class Assignment Based on Two Different Measurement Methods ............ 11
. IV. 9ange of, Uncertainties Which can be Associated with
- an Unmonitored Release Having a True Value Of I . . . . . . 24 V. Detector Reouirements ................... 29 VI'. Vendor Data for Real-Time Monitoring Systems . . . . . . . . 43 VII. Climatronics Meteorological Accessory Package . . . . . . . . 46 VIII. Installed Real-Time Monitoring System . . . . . . . . . . . 47 A-1. Evaluation of Handford Experiment by ORNL . . . . . . . . . . A-4 B-I. Values for ey and az Used in Dose Calculations . . . . . . . B-1 4
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1.0 INTRODUCTION
1.1 Backaround I
It has been recomended that systems of offsite, real-time envi-
-2 ronmental monitors be installed around nuclear power statbns. The
~
Apremise is that the data obtained from such a system could, when coupled with meteorological data, provide information relative to unmonitored, as well as monitored radioactive effluent releases, and provide the basis for making downwind dose rate projections during an emergency accident situation.
1.2 Objective The purpose of 'this study is to evaluate this proposal and to pro-vide information to aid the NRC in determining whether or not to require that a fixed offsite monitoring system be installed at all nuclear power stations.
The primary items considered in this study are:
- 1) The ability and related accuracy of a fixed real-time monitoring system to detect monitored and unmonitored releases.
[) The ability of a fixed real-time monitoring system and associ-i ..
ated calculational methods to detect and quantify the magnitude of an unmonitored release in the presense of a known release.
- 3) To provide an estimate of the credibility (uncertainty) of the information associated with the estimated value of an unmoni-tored release.
.e
- 4) To determine, using calculational methods, the number of fixed i
stations required to detect a release and to provide an estimate of the uncertainty in the measured dose as a function of the number of stations.
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- 5) To provide cost data relative to the installation, operation, g and maintenance of a fixed r,eal-time monitoring system.
- 6) To determine the characteristics and information retafn for an
, 800 m (0.5 mile) (probably onsite) emergency sy_ stem with
. capital cost limited to $500,000. 3 1.3 Evaluation Crit'eria .
The variables to be considered in this evaluation are listed below and shown in a matrix array'in Figure 1.
Range of Detector (Assume a. (0.1 x background) to 10 R/hr Background of 10 pR/hr) b. (1.0xbackground)to10R/hr
- c. (10 x background) to 10 R/hr
- d. (100 x background) to 10 R/hr Accuracy of Dose to: a. factor of 2
- b. factor of 5 g
,, c. factor of 10
- d. factor of 50
- e. factor of 250 Order of Magnitude Costs a. $ 250,000 for Installed System (Exclud- .b. 3 750,000 ing Costs for Detectors) c. $2,000,000 The following assumptions are used throughout the evaluation:
- 1. The detectors will be available as "off the shelf" items and will have the sensitivity to make the required measurements.
{ Calibration procedures will be available to assure a detector
, response accurate to 25f. .
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- 2. The monitoring stations will be located within 3200 m (2, miles) of the plant and the measurements will be averaged on a
{f 15-minute time scale. The costs of the detectors will not be considered; but costs' for signal averaging, transmissdon, and correction for background will be included.
?
. b
- _ 3. Meteorological information reouirements will be those reouired -
. to satisfy NUREG-0654, Regulatory Guide 1.97 and the Proposed Revision to the Regulatory Guide 1.23.
4 Computerized analysis of the dectector and meteorological input will use in-house or "off the shelf" hardware and software to provide accurate and intelligible output for use in control roca decisions. For offsite, real-time monitoring system output to be intelligible, the information presented to the operator in the control room must describe in real time the significant fenures of the release, such as dose distribution and contours within two miles and characterization of the source. In addi-(. tion, the computer analysis must provide for downwind dose pre-
/ diction capability beyond two miles.
- 5. The source term to be evaluated will be limited to mixtures of radionuclides which are nondepositing, i.e., only the noble gases without radioactive daughters.
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2.0 QUANTIFICATION AND ASSESSMENT 0F THE UNCERTAINTIES ASSOCIATED WITH THE MEASUREMENT OF AN UNMONITORED RELEASE
/ To provide an evaluation of the accuracy which might be obtained from a fixed offsite real-time monitoring system we used simp 15 statis-tjcal methods of error analysis. Of particular concern was the- quality
, and credibility of the values obtained for an unmonitored release in the presence of a known release.
The model used for this evaluation is shown in Figure 2, in which Dg is dose related to background, Ry is the known or monitored release, R2 is the u_nknown release, D is the dose related to R g, i
0 is the dose related to R , and 2 2 DT is the total dose measured by the receptor.
Thus, the total dose, DT, is the su i of D , y0 2 and Dg which are in some
( form proportional to Rg and R
- 2 1
DT=Dg+D2*UB Il) where Dy=Ry and D2*N2 To obtain a value for the unmonitored release in the presence of a known release, the following procedure is used. First, the measured value for R y is converted to a dose, 0, using the equations given 1
in Section 2.1. Second, the calculated value D i
is subtracted from the measured value D T to give a value for D 2. Third, the value D 2 is then converted to a value for R 2, using the same equations to obtain
, T.
1 It is assumed that 0 3 is sman in comparison to Dg and D2 and can therefore be ignored. ,
. i
- The following is a discussion of the errors associated w'ith each step in the calculational procedure and an assessment of the uncertainty
} in the value of R2*
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p( Model to Evaluate the Estimate of an Unmonitored Release s in the Presence of a Known Release 2.1 Prediction of Downwind Atmospheric Concentration Values i
The first calculational step involved in the model given in Figure 2 is conversion of the measured release R g to a dose D y. The most ,
commonly used meth,od for calculating the exposure to a receptor involves converting the known release value to an atmospheric concentration value at some downwind distance and then integrating the concentration over c i the volume of the plume. The exposure is then proportional to the pro- c duct of the integrated concentration and the decay energy of the radio- e v nuclides present in the plume, expressed as an exposure rate per unit a<
i { release (R/hr)/(C1/s) at 1 m/s wind speed. The detector response cal- e
( . culated in this study is in exposure rate. However, in the remainder of t
' this report the authors ecuate exposure rate and " dose rate", as is com- g d mon practice. .t f
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. The atmospheric concentration value at some downwind distance is I usually calculated using the Gaussian plume ecuation. This is an empiri-
/
cal diffusion formula which assumes constant wind speed, no wind shear, and flat topography. The ecuation for a continuous point source release is:
}
x(x,y,z) = #1 exp ( d (y/c y)2) (2) 2iroyoz u where:
x= atmospheric concentratio'i at a calculated point (x,y,z) for a release point h meters above the ground, Ci/m Q= source term (release rate), Ci/ seconds G(z) = exp - ((z-h)/o )2 + ero -S((z + h)/o )2 g c, = horizontal atmospheric c1fiusion parameter, m og = vertical atmospheric diffusion parameter, m
' /.(,
I = average wind speed, m/se; y = cross wind distance, m h = release height, m x,y,z =
coordinates of the Doint where the concentration is -
calculated
[
l In this relationship the most critical terms are the values for and '
a, o,. Both of these terms carry a different value for each class of atmospheric stability and downwind distance. Unfortunately, the values for c, and c, are not explicitly mathematically defined and as such must be determined empirically. A number of different field eiperiments have been conducted to determine a and c, as func- '
tions of atmospheric stability concitions (weather class) and downwind di' stances.
9
)
4 1
Currently, the most widely used data sets for o, and o g are g i I those based on the Pasquill-Gifford model for atmospheric diffusion.
- Several methods have been used to establish the atmospheric stability class which must be determined prior to obtaining the valhs for o y l One general classifying scheme is based on isolagion, cloud f and a,.
. cover, and wind speed. The standard deviation of the hcrizbntal wind
' direction is also used to establish the stability class. Another method, recommended by the NRC2 (Reg. Guide 1.23) uses the temperature gradient between -10 and 60.m.(or the release height) above the ground to determine the stability classification. None of these methods are without uncer-tainties, and in many cases the selection of the proper atmospheric stability class ma"y be in error bi one or more classes.
Assuming an error of one stability class in the assignment process (i.e. - assigning class D for a real class E conditions), we determined 4 -
the error which would be introduced in the value for x based on the Pasouill-Gifford curves 3 for adjacent atmospheric stability classes.
- g. The effect on the value for x at distances of 1000 m and 3000 m for g release heights of 10 m and 100 m is given in Tables I and II, respec-J tively. For a near ground-level release, the error in the predicted groundlevel average concentration could range from a f actor of 2 to 10 i for a one unit misassignment of the stability class. For a 100 m release the errors can be much larger.
To establish the frecuency with which the stability class may be in-Question, four months of meteorologics.1 data for an inland nuclear power station were evaluated. For this station, both the standard deviation of the horizontal wind directinn and the temperature gradient data were
- available on an hourly basis. An analysis of these data indicates that the assigned stability class based on these two methMs differed by one
- class- d3% of the time, and by two classes, up to 25% of the time.
- The results shown in Table III indicate that the stability class assign-I .
s ment based on the two methods differed about 60% of the time. Thus, the I~ ;
downwind ground-concentration value could be in error by al factor of 5
! - *about half of the time just from this source.
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Table I. ERRORS IN X (GROUNO-LEVEL AVERAGE CONCENTRATION) , , , , , ,
. FOR A ONE UNIT ASSIGNMENT ERROR IN STABILITY C!. ASS
Release Height 10 m Error Factors True Assigned D = 1000 m 0 = 3000 m Class Class Over-predict Under-predict Over-predict Under-predict A 8 5 10 B A 5 t 10 B C 3 4 C B 3 4 j C D 3 4 D C 3 4 h 0 E 2 2 E D 2 2 E F 2 2 F E 2 2 .
Example: If the true stability class is C and the assigned class is D, the Gaussian plume model using Pasquill-Gifford diffusion values for class D at 3000 m over-predicts the ground-level
[ average concentration by a factor of 4.
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- - m - - - -
p .
p
TableII.'ERRORSINX(GROUND-LEVELAVERAGECONCENTRATION)
FOR A ONE UNIT ASSIGNMENT ERROR IN STABILITY CLASS "8" Release Height, 100 m .
Error Facters True Assigned D = 1000 m D = 3000 m Class Class Over-predict Under-predict Over-predict Under-predict A B 5 30 8 A 5 30 B C 1 6 C 8 1 6 C D 12 1 0 C 12 1
' O E 15 2
- ? E D 15 2 5 E F 800 33 F E 800 33 i
e 39 * . , , g.
9
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1 q- ' Table III. VARIABILITY IN STABILITY CLASS ASSIGNMENT
' BASED ON TWO OlFFERENT MEASUREMENT METH005
{ .
1 Date No. Observation One Class Difference Two Class Offference June-1974 640,a 274 (43%)D 39(61)
JQly-1974 430 186 (43%) 113 (26%) -
Aug.-1974 613 262 (43%) 76 (12%)
Sept.-1974 661 281 (43%)
152 (23%)
a) Number of hourly observations for which both wind variability and temperature differential data were available. .
b) Percentage of the time that the stability class assignments were different.
At this point it might be well to recognize that the Gaussian plume ecuation only provides concentration estimates and not dose estimates.
In general, the uncertainties in the dose values are not as variable as the ground-level concentration values, because the cloud gamma dose is an integrated value as opposed to a point concentration value. This fact, however, should not preclude consideration of the uncertainties in concentration valces predicted by the Gaussian plume ecuation because the ground-level concentration values are more important with respect to the beta dose factor, the inhalation dose factor, and the ground-level concentration valua for radiciodine, which may be the dominant factor in an accident case. The un. certainties associated only with the dose values will be treated in detail later in this Section.
Another item which must be considered regarding the uncertainties asisociated with the Gaussian plume ecuation and ground-level concentra-
}- tion values, is the validity of the primary diffusion data based on the Pasouill scheme. The basic Pasouill diffusion data were derived from
~
tracer experiments which inve,1ved a ground-level release over. vdry flat f4 terrain with sampling periods of a few minutes at distances of up to about 1 km. Unfortunately through time and widespread usage, the
) , .
original . nature of the e' periment x seems to have been forgotten by many 1 users of the data, and the original results have been extrapolated to j include elevated release points (up to 100 m) and to distances of up to 100 km. Pascutil diffusion parameters are primarily applicable to short g
term releases at or near ground-level over relatively short distances (1
'km) and cuite flat terrain.
]
. 1 Because of the' restrictive nature of the Pasauill scheme, mora re- !
cent experiments have been conducted to attempt to better cuantify the diffusion parameters for the more realistic cases (i.e. hills, rough 3
~
terrain, forests, metropol1 tan areas, and elevated releases). Some ex-7 8 amples are given in References 4,, 5, and 6. Vogt and Brenk have [
reviewed these experiments in some detail and compared the diffusion [
parameters derived from these experiments to each other and to Pascuill. !
In some cases the downwind concentration values may differ by factors of 10 to 1000, depending upon the stability class involved. t E
Figures 3, 4 and 5 taken from Brenk give comoarisons of the !
{
short-term diffusion factors for the various experimental results for h!
y stability classes A, D, and F as a function of distance, for a release ,
height of 100 m. For class A, unstable diffusion, the data,are in good j agreement. However, with increasing atmospheric stability, significant j differences are evident (Figures 4, 5). For class O stability at a dis- ,
I tance of 1000 m, the difference between the Pasautil diffusion factor ;
and the majority of the other systems is a factor of 10 to 15. At 3000 [
m the difference is about a factor of 5. ;
e For class F stability there is little agreement in the diffusion f factors for the various systems and differences of a factor of 100 to ;
1000 are common. l
- i These data are presented not to dwell on the large differences be- ;
i ' tween the various systems, but rather to emphasize the need for selecting ;
' the most applicable system for a given site. Ideally, the preferred
- situation is to develop site specific data. Unfortunately, (experiments i
{
Ef this type are difficult and expensive to conduct. Brief descriptions
(
e- j l .
O l f(' 10-4 '
- Diffusion Category D 5 -
2- ,gn. Release Height 100 m -
t 10-5
-' ,/ -
- m_ w
'^ '
4 I / %
g 5- p/'j/ / ,./W-Q ~
~ ~ r -
/ .-
- I
/[ / ,/ / g,%%g wh
- / / ? '
/
I
- 10-6, / l s 2
l ! -
l ' ~:% .
. e../
l e p / '
i I =
n 2'/
/
l l
/ ,/ ./
,I C 3y7 I i , l i
- : 1 I
E 5[ t I I
l
'5 l j j , Brookhaven o 2 "
j s j, f --
- St. Louis N 10'8 [ l f ---- Pasquill 5
j
! ' ~
-l Julich 50
- !' l J I i .I I Julich 100 2 .
1 10-9 102 2 5 103 2 5 10 4 Source Distance (m) .c,..s.m.
h Figure 3. Comparison of Short-Term Diffusion Factors (Stability Class A)
Depicted for 6 Different Diffusion Parameter Systems (From Brenk, ref. 8)
{
[ ]
j 10~'
5 '
Diffusion Category A 2 -
== Release Height 100 m
~
l '" S .
/// / '%
c o
10 5 L f
lI. ,I ff f f kgN w i l %
=
g g II t ol ll 1 i
%.%. % ' s.. % ..%
^ '
5 7
, f/ Ii N. N . .N g 11 !
I j/ \ NN, 5 ' g'/ b 5 2 . l/ fI --~~-St.Brookhaven Louis g'g m f,' I I ----Pasquill y a 10 3 N
' /f i ! Klug \
. 5 hl j. f ----Julich 50 1
s If jf . ----Julich 100 \. !
10 2 2 5 10 3 2 5 10 4 Source Distance (m) _ _ , ,
Figure 4. Comparison of Short-Term Diffusion Factors (Stability Class D)
)- Depicted for 6 Different Diffusion Parameter Systems (From Brenk, ref. 8) 4 i
_ . , , , _ _ . _ . . . _ ~ , , _ . _ . . . _ - , . .
m . .r -3, i
. .,4..
10-4 5
- Brookhaven 2 ----- Sl. Louis Release Height 100 m '
^ 10-5 - - - - Pasquill , _ _ _
4 __..- Klug -
~~~
E 5 '
j -- - - Julich 50
/ 's
.B. 2 Julich 100/ / - - - - . .
u s2 10-6
, ,,.,.. ./
?
8 5 .
I s'
../ .
l j
? E 2 -
/ . / .. .,t - _ , '
, O 10-7 ,! ,
- ./ ,
.E 5 /
/
/ -l l H I : / -/
+ 2
,/ / j !
i o / .
/
i '
N 30-8 ; l/ ! !
- / / /
[/
i 5
/
/ /
j' 3
.- i ...
j j . s-i f / /
i 10-9 - -
' 2 2 5 10 10 3 2 5 10 4
- Source Distance (m) icer s rsis Figure 5. Comparison of Shori-Term Diffusion Factors (Stability Class F)
Depicted for 6 Dlfierent Diffusion Para er Systems (From Brenk, ref. 8)
--_--a. -m--ae+m__-- - - - e-.----_.v----.m -wwase-, e,-----s .we-,, *r w ,yr e ,
W e
{O(
g and the result and data obtained from some recent experimental programs are given in Appendix A. Included are reviews of the Savannah River i 85 Kr experiment,9 the ORNI. assessment of the Hanford experiment 10 ,
- excerpts from a Workshop on the Evaluation of Models Used for the En-hironmental Assessment of Radionuclide Releases,II and resthts of a (survey of programs used for radiolcgical dose computations.I2 Presently, .it is virtually impossible to give a definitive estimate -
of the -overall uncertainty to be associated with the prediction of down-wind concentration values, especially for data related to short time pe-
- riods, however, based on our study and those of, others9 ,10,11,12 ,,
believe that a predicted value which may vary by a factor of 10 to 25 i
from the true downwind concentration is not unreasonable. Even this estimate may be low if site specific diffusion parameters are not i available.
2.2 Prediction of Down.ind Atmosaheric Dose Values 4
( The calculation of the cloud gamma exposure from a plume is a two-s-
step process. First, the radionuclide concentration of the plume is calculated using the Gaussian plume dispersion ecuation given in Section j 2.1 (Eo. 2). Second, tne total cloud gansna exposure rate at the detec-tor is calculated by using a point source approximation and integrating over the source distribution (i.e. the volume of the plume). Both com-ponents of the exposure rate calculation have been incorporated into a code developed by Science Application, Inc.,13 which was used in this ;
study to establish detector response values.
The assumptions and parameters used to calculate the cloud gamma values presented in this report are given below.
- 1. The Gaussian plume ecuation given in Section 2.1 (Eo. 2) was a
used to establish the plume dispersion and downwind concentra-tion values. The values used for o and o are given in Appendix-II.
1 (S
}
' ~
i
___ . _ . . _ _ . _ _ ~ , _ _ _ . _ . . . _ _ . _ . _ , _ . , , _ _ _ _ . . . . _ _,. _ . _ . .
- 2. The cloud gama exposure rate at a receptor was obtained by using a point source approximation and integrating over the Q
( volume of the plume. This involved an extensive numerica'l sum-I mation of small volume elements. Although this Is a lengthy process, we believe the results are more representative than
! 7 those obtained from the use of infinite or semi-infinite cloud
.- approximations. The following is the methodology used to cal-culate the cloud gamma exposure rate to a receptor.
exposure rate D(f)=Cpa-EB(uR)T (3) where R-C = 6.87 x 10-5 O
r( .
a p = mass absorption coefficient for air at energy E (m2fg)
J E = energy per photon MeV/ photon B(uR) = buildup factor T = photon flux (Photons) mZ-s photon flux T (photons) ,_s e-pr m2-s (4) 4wr2
~
where s = photon emission rate (photons /s) j r
= distance from source (m)
. p
= total linear attenuation coefficient for air (m"1) s i
1 ) .
- . , . . - . _, . - _ _ ,--,,,..,_,,e,
The photon emission rate, s, was determined by assuming a small volume, I dV, at concentration x as follows:
')
?
5 (photons /s) = 3.7 x 10 10 X I dV
- . g . (5) ,
n >
where -
3.7~x 1010=. the the number of disintegrations per second per curie
- X= radionuclide concentration in the small volume element I 3
dV(C1/m) j 1,c = number of photons of energy E per. disintegration f
' dV = volume element considered (m ) [
i Combining equations 4 and 5 t' l
x 06 b(h)= { EIe Xe-ur B(ur) dV (6) ;
} $
' 'quation 6 is the contribution to the exposure rate at the detector due t
+
to the small volume element dV. The total exposure rate was obtained by ;
integration over the volume of the plume. When using the code, XE/Q was f used in equation 6 instead of X to subsequently give results in terms ;
of D E/Q or exposure rate per unit release rate (R/h)/(Ci/s) at I meter !
per second wind speed. i i
l 5
t i
I !
.s.
( ,
i
-17 ~ l
V Several calculations were made to evaluate the doss rate to a recep-tor as a function of stability class, distance, and release height. -The h
j dose rate as a function of distance for several stability classes for a ground level release is shown in Figure 6. /.t a distance of 3200 m (2 miles), the centerline dose can vary by at least four orders yf magnitude 7 ~
over the extreme stability class range of A to F. The uncertainty in I,
the dose as a function of adjacent stability classes can also be esti- C mated from Figure'6. For example, at a distance of two miles the dif-ference in the maximum centerline dose between stability class B and C is approximately 8, and between stability class C and D, approximately
- 3. These values are for an average gama ray energy of <80 kev (133Xe). Tht differences are only slightly less for an average energy of 250 kev.
The effect of the release height on the dose rate as a function of distance for three different stability classes is shown in Figure 7.
For the worst case, class F, the dose rate at short distances (500-1000 m) can vary a factor of 6-12 between a release height of 0 to 100 m.
This difference decreases as a function of distance. At 3200 m the dif-g
- ,f ference is approximately 2.5.
In the discussions presented up to this point, we have assumed that the centerline of the plume has passed directly over the receptor.
thereby giving the maximum dose value. The probability of this happening is ouite remote. The number of detectors and their placement reauired to give accurate dose readings will be discussed in detail in Section 3.
Based on the calculated data given in Figures 6 and 7 and the pro-blems presented with respect to an accurate assessment of. the prevailing weather. class and to a knowledge of the location of the source term, it is our opinion that the calculated downwind dose value must carry an
. .-" associated uncertainty of at least a factor of 10 or more.
?.
~
6
) .
S -
I q
O -
i 1 , ;
10 2 F. :. j D\
C F .
- - F. ,, !
B l
C o\ D
. A FN -
I
- B i l C i
%C F -
f
- /
F [
F A B l 10 -
kB C -
D'D A !
- B C
)
7 -
A f 5 t
~
U >
b 1 10'4 -- -
1 5 -
/ E l
_ 8,8 k i
I A
?
"- i
- Stability Class !
t 10 -
A
/ l r
\
- Case: h=o !
I l
?
- {
10 6 ' ' ' ' i 0 500 1000 2000 ! 3000 1
. l ICPP 5 7910 Distance (m) t Figure 6. Projected Center Line Dose as a Function of Stability Class and Distance For a Ground Level Release !
- i i
i
- - - --,,- .,--.- ---- ,. ,.,--n. , , , , . . , . . . , , - - - - , - . - , , - , ,e., ,
0 100 ,
, i t
2 .
i; r
10.0 -
E E E
en e x z i.
I o *
{
E
. Class - F x
g
..f R e Class - D C
2
=
5= 1.0 -
m Class - B L
- C.1
1000 2000 3000 Distance (m) -
' cap s.m2 Figure 7. Effect of Release Height on Dose Rate as a Function of
~ ,
Distance for Stability Class B, D, and F g
} .
d
- , - ,,,,rr,,-- , , - -
g--,--r- - - - - g , --,yn ---w--
t 2.3 Uncertainties Associated with the Qudntification of an Unmonitored Release -
t The uncertainties and range of values associated.with quantifying thehagnitude of an unmonitored release (R ) in the presence of a.known 2
relerase (Ry ) were calculated based on the model given in Figure 2 and the' relationship. -
DT=Di+D2 i -
$ where D i = R 1, and D2=R' 2 The calculation of the expected error in R assumed the following 2
conditions:
O 1. R(coastant) 2 1 1 1
.k Ry (variable) 10 1 0.1 3
- 2. The uncertainties assigned to 0 were:
1 factor of 2 (200%)
, factor of 5 (500%)
f factor of 10 (1000%)
factor of 25 (2500%)
I.
j 3. The same uncertainties were assigned to D ; however, in many 2
cases the uncertainty associated. with D may be larger Oan 2
} D1 because the height of the release is probably unknown.
i :
- 4. No significant error was assumed in the measured dose, D
- j ' T j 5. The ' background contribution is small. If the background is significant with respect to the measured D value the resul-T tant error will increase.
?
) "
'e
o The results of the error analysis are given graphically in Figure 8 g
.and listed in Table IV. In Figure 8, the range in the values for R
,.( 2 as a function of the ratio R g/R2 are given for.a family of uncertainty assignments for D g and 0 . From this simple error analysis it is 2
( , concluded that uncertainties of factors of 10 to 25 are possible for -the i'
calculated value for the unmonitored release, especially ahen the magni-
} tude of the unmonitored release is eoual to or smaller than the known release. For the case where the unmonitored release is large with re-Spect to the known release the uncertainty in the unmonitored release will approach the error associated with the values for 01 and D 2*
For example, in the case where the known. release and the unmonitored release, R g and R respectively, are of coual' magnitude (in this 2
case,1) and the assumed uncertainty in the calculated values for D g and 2D is a factor of 10, the value for R can have a range of 0 to 2
19 for a true value of 1. For the case where the unmonitored release is 10 times larger than the known release and the uncertainty in D. and D is a factor :f 10, the value of R C#" h***
2 2 * #8"9' 0t ll for a true value of 1. For the case where R is nly one-tenth of 2
-Q R
g the uncertainty in the value 1 for R increases dramat ic ally, 2
having a range of 0 to 100 for an uncertainty of a factor of 10 in D 1
and D '
2 This error analysis only presents the range of relative values to be" associated with an unmonitored release having an assigned value of
- 1. It does not provide an estimate of the true value of the unmonitored value. The accuracy of the true value, for the unmonttored release de-pends on the location of the plume relative to the detector. If the l
plume centerline is several degrees removed from the cetector, the mea-sured value for DT could be low by a factor of 2 to 10 depending on the proximity of the plume to the detector. This effect is discussed in detail in Section 3.
l
(
l l .
r
(
G>
) .
O Db l
? (
/
. t
. 10.0 -, x -
4 -
. u 2
. . or 5 o 10 i
t
~
i o 25 o x
- 1. 0 . .
[
E -
E .
a u
t t h4 E o
S o
( E c -
i .5 C O b b
- =
c N C
True Value of R2'1
,o
~
0.1g" x o a c 4 m .
s 0 '
10 20 30 40 50 60 70 80 90 100 120 Possible Value For Unmonitored Release, R 2 'C'"' '
r
(-
) Figure 8. Uncertainty in Calculated Values of an Unmonitc,,uu nelease in the Presence of a Monitored Release
. Table IV RANGE OF UNCERTAINTIES WHICH CAN BE ASSOCIATED WITH g, AN UNMONITORED RELEASE HAVING A TRUE VALUE = 1.
\(
) .
UNCERTAINTY RANGE IN CALCULATED i CASE: 1 D,D2 1 VALUE05R2
?
R1 = 10 200% 12 to -4.5 R 1 2= 500% 45 to -7.8 RT = 11 1000% 100 to -8.9 2500% 265 to -9.6 CASE: 2 R1=1 200% 3 to O R2=1 500% 9 to -0.6 RT=2 1000% 19 to -0.9 2500%
a 49 to -0.9 h CASE: 3 Rg = 0.1 200% 2.1 to 0.5 R2=1 500% 5.4 to 0.1 RT = 1.1 1000% 10.9 to 0.01 2500% 27.4 to--0.1 j
~.
4 I
(
) .
I
\
l l g 3.0 DETECTOR PLACEMENT AND REQUIREMENTS
- Q.)
{
~
3.1 Detector Placement and Resoonse Functions ..
3 The response functions and reouirements for a ring of; detectors Iwere determined by calculating the dose rate from a plume at various
~
-distances from the plume centerline. Figure 9 gives the dose rates at 1600 m for three different stability classes (A, C, and F) as a function of distance from the plume centerline for a ground level release of 1 Ci/s. The curves given in Figure 9 describe one-half of the plume shape; from the centerline to one edge. The plume shapes and dose rates . vere calculated for 80 k'eV gansna rays (133 Xe) using th'e ecuation and input factors gisen in Section 2.2.
The numoer of detectors reouired for two adjacent detectors to give responses within factors of 2, 3, 5, anc 10 of each other was determined caseo on the plume shape (i.e., tne wic n of the plume). For the plume b shape corresponding to stability class C (Fig. 9), the lateral distance
( from the plume centerline which gives a signal eoual to one-half of the maximum was determined to be e7.8 ce;rees. Dividing a 360 degree circle by this value gives a value of 46, which is the number of detec-tors recuired for two adjacent cetectc-1 to give a response within a factor of two of each other. The same process was used to establish the number of detectors reouired to give readings within factors of 3, 5, and 10 of each other for each stability class. In all cases - it was assumed that the plume centerline was directly over one detector. This is the worst case situation.
Figure 10 shows the number of detectors at 1600 m reouired to give responses agreeing within 200%, 300%, and 500% as a function of stability glass. These results are for straight line meterology, a release height of 100 m, and an average gamma ray energy of 80 kev. For class F weather (the worst case) about 85 detectors are recoired for two adjacent detec-tors to give signals within a factor of two of each other. For,a ground-
@ level release, approximately 100 detectors would be reouired for a factor i
of two agreement. Even for class B weather and a release height of
.) 100 m, acout 36 detectors would be required f or agreement within a f actor of two. -
G,
(<'
f 10 2 _
Key: -
3 1/3 ( ; ) l f ,
h #
/ \ .
\\F 1/2 (105) Fraction of / Number of Maximum Value Detectors '
r i
Class F !
10-3._.-
Case:
. Class C D = 1600 m
, 1/2 (46) ,
c1/3 (36) h =0m !
p r E = 80 kev ,
_ e 1/5 (30) i \ !
! 8 30 4 _ 1<'10 (24) s - f i
~
.c - ,
( E J -
F
~ %x C !
Class A 3
, 1/2 (24) 10-5 -~
\ ,
l A i
i i
- ~
10-6 ' ' ' ' ' i i e i , ,
0 2 4 6 8 10 12 14 16 18 20 22 24 Degrees From Centerline ice..s.ms Figure 9.
Plume Shape Analysis for Determining Detector Requirements
( :
i i
O
/
/
100 i i e i -
a
/
Distance: 1600 m (1 mile)
, - / 200'/o
- y 90 -
Release Ht: 100 m -
~ /
' /
O m *
/
/ 200%
80 - Energy: 80 kev
/
/ -
/
/
./ 1 70 -
/
/
[
/x O 300to 5 60 -
E / -
y /
/
z !
50 -
(O(
\
f o 500'c j x
' 5 o f Agreement f 40 -
,/ Between 2 -
/
1 Detectors 30 -
2 X
20 -
O 10- -
l 0- 1 9 f f I f A B C D E F Stability Class 'C "" "
(4 Figure 10. Number of Detectors Required at 1600 m to give Response within 200,300, and 500%
,27-
-G 10-2 , q
)
-x--
, N 1500 kev x 1/2(40) jj.
x [J 0
1/5(25) ;
o 250 kev i 10-3.- 1/2(42) ;
y t 80 kev l 1/2 (44) 1/5 (27) x N 1500 kev f i
40 1/5 (30 :
U '
t 10 !
5.
- (
) -
l l
_ Case: 250 kev Distance - 1600m [
Release HT. - 100m i
.3 0-5._ Stabil:ty Class - C -
Key: [t g- 1/5 (
) l N Number of !
7 Detectors !
Fraction of 80 kev :
Maximum Value !
l 10-6 ' ' ' ' ' ' ' ' ' ' ' '
.~ 2 4 6 8 10 12 14 16 18 20 22 24 ' " ' ' " *
[
Degrees From Centerline Figure 11. Detector Requirements as a Function i of Cloud Dose Gamma Ray Energy
)- . . s
[
~ , . ., _ , , , .,wyym . , .g.. -r *
- . . - - - - - _ . _ _ _ - -v. .,,.i,.,; Table v. DElECTOR REQUIREMENTS y
/
WEATHER CLASS { DISTANCE, RELEASE AGREEMENT BETWEEN A, B, C O' E, F, m HEIGHT, m TWO DETECTORS NUMBER OF DETECTORS x2 23 30 40 38 69 90 800 0 x5 15 19 25 36 42 51 x 10 -- 16 20 2E 33 3E x2 ~d5 30 39 43 46 SE 800 100 x5 16 19 24 26 29 30 x 10 -. 16 19 21 31 24
@ 1000 0 x4 2* 32 42 65 --
100
/
x2 23 34 45 62 72 er 1600 100 x3 lE 28 3o 49 57 c-x5 15 23 30 40 46 50 x2 -- 44 55 69 103 135 3200 0 . x5 34 40 46 65 80 x 10 -- 32 36 40 55 66
Table V gives aoditional data for distances ranging from 800 m to 3200 m. As expected, the number of required stations increases with distance. O
) These data also show the dependence on release height, with the worst case being a grouno level release. -
[ The effect of more energetic gama rays from shorter-lived noble gas nuclices on the number of detector requirements has been evaluated. Plume shape and detector requirement calculations similar to those shown in Figure 9, page 26, were made for tiiree different gama ray energies: 80 kev, 260 kev, and 1500 kev. The conditions assumed were class C weather staoility, and a release height of lubm. The results given in Figure 11 show little change in the overall plume shape with respect to gamma ray energy and hence, little significant difference in the number of detectors required to give responses within factors of two or five of each other. Based on the results of these calculations, it is quite evident that offsite real-time monitorinc systems consisting of 16 or even 32
- 4. units may not p-ovide information on centerline dose values and plume
) -
location because of the limited .;mber of detectors. In some cases, especially for extremely narrow plumes (stability classes E and F), the plume might, pass between two detectors and go undetected, or if detected, tne magnitude of the dose associatec with the plume coulo be greatly underestimated unless it passed directly over one of the sparsely placed detectors. Conversely, in our opinion, the installation of a 100 unit cetector system is not practical, feasible or cost effective. 3.2 Building Shine and Background Some consideration has been given to the installation of real-time monitoring systems within the confines of the site boundary; distances
. of 500-800 m are typical. In tne event of an accident, it is quite pro-
' bable that the background resulting from building shine could result in a significant signal to near-by detectors. To evaluate the magnitude of this component we calculated the dose to a receptor as a function of
' distance for the following condition:
(- h .) .
f n 1. 100% of the Krypton and Xenon isotopes and 50% of the iodine isotopes were released from the core.
.g
- 2. Of these amounts 1% of each leaked to the reactor building.
I 3. The following reactor building contents (based nn WASH 11400II#I for a 12 he decay period). 87 I Kr 2.4 Ci I 1.2 x 105 Ci
- 88 #
Kr 1 x 10 Ci 133 I 4.8 x 103C1 Xe 4.8 x 105 Ci I 1.7 x 10$C1 135 # 134 Xe 4.1 x 10 C1 I 22 C1 135 # I 6.5 x 10 C1 4 No significant shielding (SWR).
- 5. Building volume = 5 x 10 # n. 3 Using the ouilding contents given abcve, the dose rate from this h source was calculated for various distances from the building using the code IS05K.0-II(15) .
The results for the rare gas and iodine compo-nents are given seoarately in Figure 12. Tnese data indicate a signifi-cant increase in the normal background level (0.01 mR/hr) due to the contents o# the bailding, especially at distances of less than 1000 m. The cuestion of shielding the detectors from this source has not, in our opinion been adeouately resolved. Complete shielding of the de-tector from this source would only negate the signal from a plume. The value of partial shielding in the direction of the building shine is cuestionable considering the scattered radiation from the builtiing. To evaluate the impact of the building shine on a signal from a passing plume we have included in Figure 12 the contribution from a plume I of Xe based on the building contents given above and a leak rate of 1%/ day, giving a source term of 0.055 Ci/s. Also assumed was . class E weather, a. wind speed of 1 m/s, a ~ release height of 100m, and that the f bulk of the iodine was retained in the reactor building or trapped by I the ' filter system and therefore' had no significant contribution to the plume dose.
~31-
. . . . i
-l- _. -- - - - - " - - " - - - - " - - - - - - - -
O 4
) 103 ,
t -
\ -
102 -
- Noble Gases
~
' Radiciodines I 10 - \
iii :
\
.0 .i LS -
a 2 +a e _
'd, 5
) 3
, m 1_
2 Release. HT. - 100m
~
Stability Class - D
- _ Leak Rate - 1'/o/ Day of Building inventory 10 - X Noble Gas Plume Dose
-x-x_
10-2 ' ' ' ' ' 200 400 600 800 1000 'C" 5-'"
- Distance (m)
Figure 12. Effect of Building Shine on Detector Response
]
{ ..
1 ge The results of this calculation clearly show the significance of
, the building ~ shine factor relative to the plume dose. For the accident ,
case where significant ouantities of the volatile radioactive products I are in the reactor building, little or no information regarding the plume ! dose could be obtained from detectors located close to i:he reactor ' i building.
- . i P
i
.I P - 4 6
. i
.I ,
e es b
- ,-3 - \
y-
4.0 INSTRUMENTATION REQUIREMENTS, AVAILABILITY AND SYSTEM COSTS J. The basic components of an offsite real-time monitioring system are
; shown in Figure 13. Aiso identified, are the major cost areas to be
, $, considered in establishing an offsite, real-time monitoring system.
, Currently . virtually all of the existing real-time monitoring systems which have been installed are for the purpose of monitoring routine re- !
leases rather t'han for use in emergency tituations. Although the imme-diate use is different, the ecutoment and costs should be similar. Be-I' cause many of these systems have only recently been installed or are in
- the installation stage, little information or cost figures are available in the open literature.
To obtain the information necessary to establish an estimate of the costs involved in the installation of a system for use in an emergency situation, several utility scations and state agencies were contacted. 3( These conversations ranged f-om rather open discussions to cuite guardec comments, and in some cases, a reluctance to cuote cost values. Als 0-
) vendors of potentially useful instrumentation were contacted.
A review of the instrumentation recuirements and availability c' real-time monitoring systems is given in Section 4.1. Section 4.2 gives a review of total system cost and an estimate of the installation costs based on information gathered for existing or planned ' systems. 4.1 Instrument Description and Reouirements The basic reouirements for an offsite, real-time monitoring systera are listed below and shown diagramatically in Figure 13. 4.1.1 Field Stations. Field Stations will consist of radiation . detecting devices and associated electronics. The stations would pre-ferrably have the capability of signal averaging and onsite readout. The radiation detection system should be capable of measuring dose rates Q'
)
.- <-v.r,-r* r-% . , e - _ - . . --y y....w ; -m ,. , - - .- , -- 1 ny, ,--.w,- -- ----,,,w-
b i / OFFSITE, REAL-TIME MONITORING SYSTEM MAJOR COST AREAS MAJOR COMPONENTS i Power v v Sen rs & Detector + lElectrometer
'h Field Micro Unit j Trans.
~
> Processor c lator MODEM
=
E 3 a. Telemetry y Data y b. Dedicated Transmission g
.c
- c Phone s
\ -
( 3. c. Hardwire \ j - - - - - - - _\----
\
_s ' \ c
- -! \
Receiver I Central y Computer
) Control Processor Station 5 /
v . V p. Storage Printer I s Maintenance ICPP.S 7838 . Calibration j Operation ( l Figure 13. Schematic of Offsite Monitoring System Basic Components , 1
-n.
l
~
from ipr /hr to 10 R/hr with reasonable accuracy (+ 10%) and respond in ( a relatively flat manner to photons of 50 to 3000 kev. The detector (s) should be weather proof and the associated electronics enclosure main-h j titined at suitable operating conditions. This may reautre heating or cooling depending on site conditions. For the winter of 1991-2, the heating recuirement could be significant'. A provision for backup power
should be made.
Additional instrumentation, such as meteorological sensors and io-dine sampling devices may be added to the field stations. This addi-tional instrumentation may provide useful data but the ecs: per field station will be increased. 4.1.2 Data Transmission. Three practical methoos exist for transmitting data from the field stations to the central processing unit and comands from the central processing unit to the field stations. These include direct hard wire connections, dedicated telephone lines and radictelemetry. The choice for specific site will depend on economic and envircr. mental f a: tors. Tne selected system must ce capable of: i h J 1. Bidirectional operation,
- 2. Error detection and correction, '
- 3. A useful transmission rate, and 4
A transmission structure compatible with the accumulated data. Direct wire connections often provide the most reliable connections. However it may be impractical to use ha: dwire connections over water or at distances greater than one mile. ( G
}
g V Telechone systems using voice grade lines for data and comand transmission can be installed by and then leased from telephone compa-nies. Bidirectional transmission is preferred, although half-dupler 'is
/
adecuate.
. Several commercial vendors including those of real-time. environ-mental monitoring systems supply compatible telemetry systems. Line of sight transmission of up to one mile can be performed using an FM system.
As with other field stations electronics, transmitters and encodr.rs must be protected from the environment and, at colder sites, heated. 4.1.3 Central Processing Unit (CPU). The CPU performs the i.ccui-sition, reduction and storage of data describing radiation dose rate conditions existing at each field station. The CPU also performs the following functions: 1) diagnostic testing of these data to provide dose rate and meteorological condition time average values for each
,., station, 2) the com::a*ison of radiation data to alarm points, 3) compi-lation of historical data files, and 4) polls the field stations for radiation dose rate 'ivels at recuested intervals. The CPU also shoulo have an interface for transferring the accuired data to an external computer for plume analysis, characterizations, and the ultimate predit-tion of downwinc dose values. Hardware recuired for these as;.s inclu e:
- 1. Data receiver and decoder
- 2. Microprocessor,
- 3. Data storage device, 4 Printer,
- 5. Command entry device, and
- 6. Back-up power supply, b
4.2 Instrument Availability The capital costs of obtaining and installing an emergency monitor-C j ing system were estimated from costs of existing routine monitoring sys-tems. To prepare capital cost estimates for an emergency monitoring system, three vendors of routine monitoring systems (GCA Corporation. Harshaw Chemical Company and Reuter-Stokes) were contacted. None of these vendors offer systems which are capable of simultaneously monitor-ing routine radiation releases and meteorological conditions, transmit-ting this information to a CPU which subsequently models the release and provides dose rate characteristics. The existing systems provide real time remote location dose rete data which is transmitted to a CPU and converted to information such as count or c:se rate averages, anomalies and alarm points. All three venoors offer a CPU which can be interfaced with an external computer for characterizing and predicting dose rates. It is interesting to note that the only external computer that each of tne tnree venocrs recommends interf acing to their CPU is the Digital E ;uipment Cor;:cra-icns P c-11/30 Oi ailt of the features, capabilities and price fer each of the thres systams are discussed below and summari-zed in Table XI. Q.
.)
GCA Corocration The GCA Corporaticr. has installet severd of their " Guardian" sys-tems at power stations in the United Kingdom for the purpose of provid-ing routine real-time environmental monitoring. The " Guardian" system employes two GM detectors (low range, 10-6 - 10-2 R/hr, and high range 10 10 R/hr) at each field station for radiation detection. In addition, a "Maypac" particulate and iodine filter system with con-l stant air pump can also be placed at each field station. Data transmis-sior. from the field stations to tne central processing unit is ucually l performed by VhF radiotelemetry, but other methods are possible. The
" Guardian" CPU provides imediate hard copy and visual display of current field station readings, system diacnostics, and data logging. Although i
the " Guardian" system has beer, marketed in the USA since 1981 it has not been installed and operated at any power station in this country. In-h* {' stallation costs and operational characteristics of this system can be ootnined by contacting power station personnel in the U. K.
,, T_he Harshaw Chemical Comoany
( ,'I
~
'{ The Harshaw TASC-a systems may be used for routine real time envi-i ronmental radiation monitoring. This system uses two scintillation de-tectors per field station to give monitoring capabilities over seven decades of signal. Data transmission from the field stations to the CPU
; is by dedicated hard wire systems because generally the units are used
, inside of buildings where the distances are short. An advantage of the Harshaw TASC 4 system is that all field station electronics, except the creamplifier, can be placed in the CPU thus minimizing the effects of weather on the system, lowering the potential for tampering at the field stations, and centralizing much of the maintenance. Components of the CPUs of these systems also include counter-timers, printer, and computer interface modules. To date none of these systems have been installed to function as routine real time m:nitoring devices at distances being con-sidered in this study.
Esuter-S okes 8 Tre Reuter-Stokes Sentri-1011 system, designed specifically for real time routine radiation conitoring, has been installed at several nucles- cower stations in the USA. The field stations of the Sentri 1011 systems are eouipped with high range (10~ - 10 h/hr) and low rance (10-D - 10-2 R/hr) pressurized ion chamber detectors and as-sociated instrumentation. Reuter-Stokes is presently ceveloping a single detector to provide accurate monitoring over seven decades of signal which should result in a reduction of a capital cost and installation. Data transmission from the field station to the CPU of the Sentri-lOli
~
system can be accomplished by radiotelemetry, dedicated telephone lines, or hard wire. The Sentri-1011 CPU performs field station data reduc-tion, system diagnostics, and data logging. Historical informat' ion can be obtained in hard copy and the unit contains an interface port for an external computer. t
-39
, -, e
Reuter-Stokes is the only vendor which offers a compatible meteorologi-cal accessory cackage (cr their field stations. This package, the "3-D (j Wind System," is marketed by Climatronics and is described in Table '/II. J 4.3 System Costs Several commercial power reactor stations and state radiological
. monitoring agencies were contacted relative to obtaining cost information regarding the purchase and installation of offsite, real-time monitoring systems. Although the systems which have or are being installed are for the purpose of monitoring routine releases, the basic instrumentation and cost data should be similar for an emergency montoring system. In some cases, detailed cost information was not available because the sys-tems were still being installed or existing systems were being modified or expanded. Information regarding date of installation, number of fixed stations, distance from the source, and type of data transmission is given in tam e VIII. All of these systems are using the Reuter Stokes Sentry-101: T.:nitoring system.
k The cM- f actors for these systems are ouite variable because of
,j varying de;ress of instrumentation complexity and whet'.er a subcontractor was involved in tne design, purchase, and installa-ion of the system.
The range i . ne c:sts, per monitoring unit is apprer'iately from $20,000 to 540,000/ unit. In general, the higher priced systems included a mete-orolo31 cal sensing component and/or additional subcontractor costs. For purposes of this survey an average cost of about 530,000 per unit appears f reasonable. This value includes the -costs of all monitoring and data transmission instrumentation. The cost of a central data processing ! unit and a computer for extended data handling and reducing capibilities is variable depending on whether dedicated or existing hardware is used for this purpose. A much more ambiguous cost is that regarding the installation of the field units. If the monitoring unit is installed on existing sup-ports (power transmission poles) and the power source is readily avail-able, the installation costs may only amount to a few thousand dollars (53,000 - 5,000) per unit. Conversely, if special suppo-ts are reouired g
-4 0-i
.~, ,,p=-- + < - ,
or if the units are installed over water, the average station costs could l , increase five fold (525,000/ unit). Additional cost would also be incur-
-red if special power lines and installation are reouired. For ex'mple, a l
i use of uninterrupted power free the Auxiliary Building could add several t hundred thousand dollars to the overall costs. Other costs which must be considered but are difficult to cuantify include design and engineering, purchase of land if necessary, deprecia-tion, routine maintenance, dedicated telephone line leasing fees, and operating cost. The last item could be significant if a group of dedi-cated operators (meteorologists) were assigned to oDerate the system and
- evaluate the data.
Based on the data currently available, the following range of cost
- figures are given for a 16 unit station at a distance of 2 miles.
Range e' Cests '? "i k, 1. Instrumentation 520-O:/ unit 400 - 40
- 2. Data Collection and orocessing eoui:;. 40 -
11?
, 3. Installation SS-25K/ unit 80 -
400 4 Design and Engineering 50 - 200
- 5. Contingency 100 - 200 5670 - $1.610 e
4
+
, , . g -
g--,, ,r - - , -
Th'e lower cost figure does not insure uniform placement of the mon-( itoring units because existing support Doles are considered for use. O Thus'. it is ouite probable that a release could go undetected if it con-sisted of a compact plume (stability class E or F). Considering that we are only referencin'g a 16 unit system this same cornment could apply even 1
, f the monitoring units were uniformly spaced on a ring. Therefore, one
~.could raise a cuestion regarding the technological validity of the entire concept.
To increase the number of stations to give highly reliable measurements would result in increasing the overall cost of a system by several milliond'ollars.
,(.
G
/
~
( O
)
- 42. .
-i _ _ _ _ _ _ _ _ _ _ _ _
- q. A ., %
l IABLE VI VENDOR DATA FOR REAL-11t4L NON110 RING SYSTEMS Field Station Data Central Processing Detector Electronics Transmission Unit (CPU) liarshaw Chemical , Co. TASC-Il Features Two weatherproofed liard wire. All field station electronics CaF2 (Eu) Scintillation except preanplifier are at CPU. detectors. Wireal power backup, b* Capabilities Two detectors span seven liard wire use demonstrated Field station electronics at CPU
' decades of signal up to one mile. record and print out monitoring (lpR/hr - 10 R/hr). data. External computer required and withstand tempera- for other data reductions and dis-ture variations of p lays.
1500/hr.
.1981 Prices $7K per station including included in field station 14.2K for CPU consisting of com-electronics. price. puter interface, data recorders, counter and timer.
9 L. - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ . _ _ _ _ __
x TABLE VI (Cont'd) VENDOR DATA FOR REAL-TIME MONITORING SYSTEMS Fleid Station Data Central Processing Detector Electronics Transmission Unit (CPU) GCA Corporation
" Guardian"
. Features Two weatherproofed RF telemetry, GM detectors, digital CPU has teletype printer, visual dedicated telephone, data display, data storage, alarm display, detector range hard wire. and status system, and external
, changer, solar powered g battery backup and conputer interface.
"Maypac" particulaLe anil iodine filters. Auto-matic detector range changer.
Capabi li t ies Two detectors span seven liard wire or dedicated decades of signal Processes data from up to 31 field phone line possible. stations. Field stations scanned (lpR/hr - 10 R/hr) with at 5 second intervals in system
+ 5% accuracy at 10pR/hr. alarm mode, 5 salnute intervals in Ynergy response between in non-alarm mode.
60 kev and 3 MeV is + 20% Tenperature operating ; range -1000 to +60 0C.
-o 1981 $15K per station with included in station $50K Prices telemetry cost
$10K per station with
- hard wire or telephone.
i #
@ , @ O
b,
- A- .
M hl 1 Alli C VI (limt 'd) . VENDOR 11A1 A TOR RI.AL-LIME MONITORING SYSTEMS
- ' ~
field Station Data Central Processing Detector Electronics Transmission Unit (CPU) Reuter - Stokes Sentry 1011 . Features Two weatherproofed pres- Hard wire, dedicated tele- Basic CPU processes input from 16 surized ion chamber phone, or RF telemetry, field stations, reduces and stores
, detectors. Digital dis- exposure rate data. CPU also has play, strip chart re- alarm and system diagnostics, and corder, 8-hour battery external conputer interface.
i power backup, serial i' . readout and automatic detector range charger. - Capabilities Two detectors span seven Basic CPU can be upgraded to pro-decades of signal cess input from up to 48 field (lpR/hr - 10 R/hr) with stations. . Scan of field stations 8 5% accuracy at low can be done at 5 second to 5 min-Tevels. Temperature ute intervals. operating range is -2500 to + 550C. 1981 Prices $11.3K per station $3.5K per station for tele- $39K to process input from 16 (detectors cost $3K metry. liard wire about $2K stations. . a piece) per station at 0.5 miles. 170K to process ir.put from 48 stations. O w -,.... , . , , - - 3 . e_ - .+ . . - - - - , , , _- = .. , w-- e- -y-, , - -- - ,- -e- -. --m-. - - -
.- e m .r,s -_
.,c-.
, n. ,
e TAlllE Vil Climatronics "3-D" Metorological Montoring System Field Station Data Electronics Central Processing
, Transmission, Unit (CPU)
Features Sensor package inputs to Dedicated 2-way telemetry field computer to de- Field computer catput transmitted system recommeruled for real- directly to modeling system to ' termine e and
- time monitoring but tele-Can use dedicated tele- phone and hard wire is determine stability class.
' metry for radiation de- possible. tection electronics for e transmission. Ten meter
$8 tower, sensor heater and electronics hous ing re-quired.
Capabilities Solarpoint dew power, tengierature sensors also Two-way RF can relay commands Station sensors used with doppler available. to station such as time aver- monitor for forecasting would give ages anil scan times. mure information for modeling. Costa $10K/ station for the 13.5K/ Station for P-way RF $3K for RF central processing unit above features and capa- telemetry. bilities. o a1981 purchase prices g .h ~
.a' l mg s"c d' o TABLE Vill REAL-TIME INSTALLED MONITORING SYS1 EMS Number of Distance Faciltity '
9ta11ation Monitoring from Sources Data
._ (0wner) J ate Units (miles) ,
Transmission Diablo Canyona 191 12 5-10 Phone (PG & E) La Salle 1980 8b $ Phone (Comm. Ed) A 8erwick 1980 2c Y 15-26 None (Penn. P & L) Virgli C. Suumera 1981 8 0.5 liardwire (SC G &-i) Three-Mile Is. 1980 12 0.1-4 Phone (Metro. Ed) Indian Pt. -2a 1980 16 O.5-2.5 Telemetry (Con. Ed.) San Onofre 1981 9d 0.6 Phone (50. Cal.Ed.) a) Field stations include meteorological accessories b) Wien completed system will have 16 fie,d units c) Complete system will have 7 field units d) Conclete system will have two rings of 9 stations each. One at 1000m and the other at 2000m.
5.0 MATRIX EVALUATION h ( One .of the primary objectives of this program is to -evaluate the
, concept and usefulness of 'an offsite real-time monitoring system in the light of a matrix array and associated parameters. The matrix and its
,three major conponents, accuracy, cost, and detector sensitivity were' presented in Section 1.
Based on the review and studies conducted in the prior sections of this report our evaluation of the matrix is as follows. For ease of reference tne matrix array is reproduced on page 51 as Figure 14. Accuracy - The accuracy level was evalu'ated based on :ne uncertainty associated with the quantification of an unmonitored release in the presance of a moni-tored release. Based on our study we propere el: 'nating all ccnditions associa e: with at:.- racy values for factors of 2, 5 an; 10. In s: e cases, especially for the case where the unm:r - g (. tored release is small compared tc the monitcret release, even the accuracy factor of 50 for the unmonitorec release may be a question. Detector Range - The recuirement for detectors sensitive to the measurerent of a quantity of radiation equivalent to 0.1 background (1 pR) cannot be justified, especially when the background can fluctuate more than this amount. A similar argument can be made for detector systems having a lower range equiva-lent to background (10 pR/hr) because the un-certainty in the signal would be large and a reading equivalent to background in an emergency
, situation wccis not be significant relative to the initiating protective action in the surround- !
ing areas. One might make a case for the use of G 1(- detectors having a lower range of 10 pR/hr 'in
-} , establishing site specific diffussion models based
-4 8-
q.g on the monitoring of normal releases. However, , N'# if the detectors are placed at a two-mile dis-t( tance from the plant, the dose rate from normal I releases would be so small as to be garbled in the normal background fluctuations. In our opinion, detectors with a lower range of 10 i times the background level or 0.1 mR/hr should be adecuate for an offsite real-time monitoring 4 4 system, because readings of less than 0.1 mR/hr
- are of little significance from a hazard stand-point. This is a point which should be pre-sented as part of the public relations effo*t of the etility. This conclusicn eliminates the two lower levels of the matrix.
Cost Facters - Tnis item is more difficult to assess cecause of the wide range of values associate: :ith the it:alla:4cn ecs:s. We cat. no.ever. 3As sore
<l p generai comments. If a low cost sys et is in-1 s alle: witn a minimum ;9-12) r..cte- :f sta-l tions, .here is a high orocability Of :issing a plume, in which case the system has little te:nneinci:al value. To install a minimum sy -
tem witn detectors only near population centers may have acceal from a Duolic relations stano-point but it does not provide the technical data whicn is necessary to assess the impact of a plume to the rural areas which could be popu- ! lated by grazing milk cows. Also, if the detectors were not uniformaly spaced near the Doculation centers, false information relative
. to the intensity of the plume dose could result.
O 1,
-4 9. -
._ ~.
We do not support the installation of a minimum i.. system, which we are associating with a 3250,000 k cost- value, because the technical information obtained from such a system would be of ques-tionable use in a decision making process. i
. Similar arguments can - be made for a 5750,000 system; however, at this level each installation would have to be evaluated on an individual basis because of site specific characteristics.
Obviously the requirements for a monitoring system in a flat terrain situation is different from one involving water, off-shore and on-shore-breezes etc. In some cases, a system con-structed for a cost of 5750,000 might provide reasonable technical information. Thus, we
- r. ave decided to leave this area of ths matrix nen but emphasize the site specific character-stic of the case.
h [(.i For 52,000,000 one might construct a reasonable system, but in no case would information ac:ar-ate to a factor of 5 or 10 be obtained. In fact, almost no sum of money would insure ob-i taining dose values to this level of accuracy. (- Based on the above discussion and evaluation, the bulk of the matrix has been eliminated. The remaining areas which we feel identify the potential benefits and aasociated uncertainties from the installation of a fixed off-site real-time monitoring system are shown in gray in Figure
- 14.
While it is acknowledged that our conclusions are argumentative, we ! believe they are representative of the current state the of art. l - \ 7 O J _,
v.- )
'~
s^ h u h s ,a - s'
(5 ~/ / / P'9c%?'
/
~/ / / / // / / /
1000 g R/hr [
'i' . 100# RIhr /
. cn E . . . _.
EC h 10 g R/hr /
^
1 p R/hr
/
ICPP5 Feet X2 XS X10 X50 X250 Accuracy Factor ' ' ' ' Figure 14. Matrix Parameters i
l 6.0 MINIMUM-COST EMERGENCY SYSTEM _{ O An augmented effort to the general program involved the characteri-zation and evaluation of a specific, minimum cost emergency ' system with
, close proximity to the plant. The constraints to be applied to the eval-s uation of such a system are as follows:
- 1. Total system cost - not to exceed 5500,000,
- 2. Detector assembly cost - not to exceed 57,000/ unit
- 3. Detector distance - no further than 800 m (0.5 mi),
4 Detector sensitivity - 0.1 mR/hr to 10 R/hr, and
- 5. Accuracy - within a factor of 10.
Using the cost data presented in Section 4, the following values '
- k. were used to establish the magnitude of the systert wnicn could be ins al-led within the $500,000 constraint.
Fixed Costs Central Processor (with modeling, 5110,00C 48 station capacity) Design and Engineering 40,000 S150,000 This leaves a balance of $350,000 which can be allocated to the cost of the detector assembly, data transmission, and installation. The cost per station is estimated as follows:
)
-52' -
9
- --- - a
Dectector Assembly 5 7,000 Data Transmission / Unit 8,000 a Installation / Unit 3-15,000 D I Total / Unit $18-30,000 a Includes capital cost and instal ution b highly variable desending on so'cific e location For this exercise it was assumed that the data transmission would involve a telemetry system because the cost of installing hardwired or cedicated phone systers is nignly variable. For example, climatic fa:- toes may di:: ate the :arial Ir.:/cr the use of special material- in et:n h of these data transmission sys e s. The installation costs are basec on ( , simple units all installed c #iat solid terrain. If uniform place snt of the detector assemolies reccirec installation in cooling ponds, r'v-ers, or other bodies of a ater, tne installation costs would increase significantly, perhaDs ey is eu r as a factor of five for those uniti in such a location. Another significant expense item is the power source. If an uninterrusted power supply from the Auxiliary Building is used the cost per station would be significantly more, especially if underground or underwater lines were used. Based on an after fixed-cost balance of $350,000 which can be allo-cated to the detector units, and a range of average station costs of 513,000 - 530,000, from 12 to 20 detector units could be installed depending on the actual placement of the units. The 12 unit system aould insure e ual placement of the units regard-less of the location. The 20 unit system could have voids in the moni-A toring grid and be coerated with normal power sources. Il
.) i I d I
The. estimate appears reasonable based on information obtained from
,( two utilities which provided cost information for a comparable system. O
- J One station which recently comoleted . installation of an 8 unit system at a distance of 800 m (0.5 mi) ouoted a cost of aboui. 5435,00,0 for the purchase and installation of the package. In this case, each, unit also
{ included a meteorology station and the output from the unit was hard-wired to the control station and coupled to an existing HP-1000 data processer. Thus, the total cost per unit is approximately 554,000. The overall cost would increase if a dedicated CPU were used and probably decrease if some other form of data transmission were used. The ' cost per unit would also decreasing if meteorology sensors were not installed with each unit, but the validity of any down-wind projection would also decrease. A second utility while not providing complete cacital ccst data did provide sufficient informatior. to estimate the cost for a ring of nine units at a distance of 1000 m. The central processing unit for tnis ( system has not been purchased. However, the purchase cost of the nine h
')
field stations was about $135,000, or about 515,000 ce- sta-ion. For the nine unit system using a dedicated phone system for data trans:-is - sion, the installation cost per unit was ouoted at about 523,000 ce* unit or about 5200,000 for the system. Tr's is somewhat nigr.s- than ou-estimate but gives some idea of the costs involved just for installation. Assuming fixed costs of 5150,00'. #oe design, engineering, eno a central data processer, about 5135,003 for instrumentation and S200,000 for in-
, stallation gives a sum within the 5500,000 constraint. The unit cost for the nine detector system is about $55,000, which is similar to the first system discussed.
i To estimate the credibility of the data which could be expected l
. from an 8 to 20 unit system we first considered the data given in Table V. Based on these data, a minimum of #90 ecually soaced sta-
. tions would be reouired fo.- two adja:en units to give a reading within a factor of 2 of each other when tne release was at ground level
( 0
)
-5L-
I and the stability class was F. For two units to agree within a factor h of 10 would re:uire a system of 35 40 units. About 30 detectors would be recuired for more common class D weather. These numbers are based or the assumption that the centerline ~of the plume passes directly over oi of the stations. This is a hignly unprobable event. The passage of a plume between two detectors would give a response which underestimates the true magnitude of the release. A second factor wnich must be considered for a 500-800 e system, is the effect of the building shine factor, especially for a BWR. For the case given in Figure 12, the plume dose at 800m for a leak rate of 1% per day of the building noble gas inventory is considerably less than the building shine background. The effect of building shine will be much less for a PWR. 1 The effect cf building wake and dispersion of the flow regime by other ouilcings f ctner tna'i tne reactor building) is a third factor which shof.: ce :,nsi:E sc. ~nis ef's:: could significantly a'ter the mea-h surecent of the tr e ocse from tne plume. k While a ci:se-in detector system might in some ins ,ances provide some informaticr. in an emergency situation, the ability t0 extracolate and creject tne 'r#r:.atior to gi e vconcentration or dose u 'ues at some extended downxind sistante 15-10 mi) is highly ouestionable. This could only be. done with a reas:nable degree of confidence if site specific moce'.ing and adcitional downxind meteorological data were available. 9 9 s
)
. -ss- -
h '" '-
7.0' REFERENCES ,
- 1. F. - Pasouill, "The . Estimation of the Dispersion of Windborne O ';
'( Material," Meteoro1. Mag. , 90:33-49 (1961). -
- 2. U. S. Nuclear Regulatory Cornission, " Proposed Revision 1 to f Regulatory Guide 1.23, Meteorological Programs in Support of i Nuclear Power Plants" (Sept.1980).
. I 3. D. H. Slade (Editor) "Meterorology and Atomic Energy-1968," USAEC
~
Rept., TID-24190, Environmental Science Services Administration ! (1968). i 4 I. A. Siriger, M. E. Smith, " Atmospheric Dispersion at Brookhave*. I National Laboratory," Air and Water Poll. Int. J., Vol. 10, oc ! 125-135.(1966). ; i S. J. L. McElroy, F. Pooler, "St. Louis Dispersion Study Volume II-Analysis " U. S. Dept. of Health, Education, and Welfa-e, Arlington, Virginia (1968). '
- 6. K. ' J. Vogt, H. Geiss, " Tracer Experiments on the Dispersion of plumes over Terrain of Major Surface Roughness," JUL-1131-ST ,
Julich (1974) see also Ref. 7.
- 7. K. J. Vogt, "E .sirical Investigation of the D'ffusion of *.{tste Air f Plumes in tne Atmosoners," Nucl. Te:nnoi. 34:4':-57 (1977).
- 8. H. D. Brenk, "Atmossheric Discersal," Lecture presarey for the
. Third Annual Healtn Pnysics Society So.w er Scnool at tne - ce-si v i cf Washington, Seattle, WA (1980).
- 9. C. V. Googlak, H. L. Seck, M. M. Pendercast, " Calculated and f Observed 8-Kr Concentrations within 10 km 'o the Savannan Rive
- i Plant Chemical Sesarations Facilities," Atr.csoneric Enn onment, i Vol. 15, pp 497-507 (1981). ;
- 10. R. O. Chester, C. W. Milier, .ORNL Unpublished Data, Breeder Prog-ar l Technical Status Rept. (August 1981).
- 11. F. Owen Hoffrran, Gen. Chairman, " Proceedings of a Worrshop on The j Evaluation of Models used for the Environmental Assessment of !
Radionuclide Release," CONF-770901, Gatlingurg, Tennessee t
. (September 1977). ,
i
- 12. Isaac Van der Hoven, W. P. Gamoill, "A survey of Programs fo*
Radiological-Dose cocou.ation," Nuclear Safety, Vol. 10, No. E ; (1969). i
- 13. C. D. Thomas, Jr. , J. E. Cline, P. G. Voilleoue, Evaluation Of Ar_ l Environs Exposura Rhte Monitorina System for Post-Accioenc j Assessment, Draf t Report to AIF Science Applications inc., Nuclear l Environmental Services, (Sept. 1981). '
( Ol 1 !
', -sr6- !
l 1
c)
'O REFERE?CES (cont'd) g( 14 U. S. Nuclear Regulatory Comission, NUREG-74/014, Reactor Safety Study - An Assessment of Accident Risks in U.. S. Commercial Nuclear '
Power Plants, (Reprint of WA5n-1400), (1975).
- 15. R. l.. Engel, J. Greenborg, M. M. Hendrickson, ISOSHLO - Comouter
- . Code for General Purcose Isotope Shielding Analysis. BNWL -
236 7 (1966). 0 (
)
O t T J
m m i{ J s 3 APPENDIX - A O BRIEF SU!tv.ARY OF EXPERIMENTAL RESU : 70 f(- j C0!i: .F.E MEASURED AND PREDICTED GROUND LEVEL CO,TE*;TMTION VALUES
,a"~%
r --V
.}
e
o l h A.1 85Kr Expe. iment-at Savannah River Plant 9 {(.r In this experiment, the release of 85 Kr from the Savannah River Plant chemical separations f acility was monitored for over a year at six 3 sites within 10 km of the release points. Using the Gaussian p'lume model [foracontinuoussource',theratioofthepredictedconcentrationtothe measured concentration was determined. The dispersion parameters were those based on the ideal case of flat terrain, short distance, and steady meteorological. conditions. The general results showed that the annual averace concent ations were over-predicted by a factor of 2 to 4 compared to the measured values. For the short-term (10 hours), the predicted values were within about a factor of 10, and in many cases, particularly in calm or stable conditions, reasurable concentrations were predicted when none were ob-served. The results of the short-tert data are shown in T ;ure A-1 from I g Reference 9. A.2 ORNL Assessment of Har. ford Exceriment10 As part of a DOE sconsorec prograT. associatec .itn ne Breeder Reactor Program, ORM. is evaluating experimental data cotained from an experiment conducted at Hanford in which zinc sulfide fluorescent oarti-cles were released from a height of 111 m over relatively smooth terrain.
'Crosswind-integrated ground-level air concentration measurements were compared with predicted values using a Gaussian plume atmospheric dis-persion model. Of interest was the use of three different sets of cea-surements to calculate the atmospheric stability class.
- a. The vertichl temperature difference between 10 and 122 m above ground-level,
- b. The standard deviation of the wind direction measure at a neight of 122 m, and A-1 as
- am
( 0 ( Composite Of All Locations : 10 4
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/
/
/
/
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/
/
3 ' 10 -
/
/
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t .
. /
^ / . /
NN
*/ - i
/ /
5
* */
/
o ' / r/. . *
- c. * * /
2 , f
~ 10 , _
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- s 3 ,
5 /. .
. .- . /
( = i .. ...- i ac /* i .
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- l
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0 < . . 10 0 3b1 3b2 gb3 10 4 Measured (pCi-h//) J l 1 ) l ICPp S 7924 l Ficure A-1. Measured to Predicted 85 Kr Concentrations (ref. 9) A-2
l
-j -f h c. -A combination of a and b.
! -k For the Hanford data, methocs a and b, with one exception, indicate Pasquill stability classes E or F, while method c always indicates class D. In this study ORNL compared the results obtained as a function of' "the di spersion factor, c, 7 based on five different sets of diffusion model s. Basically, these include those data sets previously discussed and reviewec by Brenk8 (Pasouill, St. Louis, Briggs' Rural, Brockhaven, and Julich-100 m). Separate comparisions were made betaeen measured and predicted concentration values using each of the fwe sets of c,- values and three stability class determinations. A s u=.ar y of the ob-served and predicted concentrations values is given in Tanle A-1. Tnese cata (Taoie A-1) snow that the precic ed values cif fer from tne msasurec values Oy a fatter c# 5 I: 10 r:re tr.n 50t of tne time anc g that the precictec value may oe .To e or less 1 ar. tne measurec value (. depending on the disDersion sys:em used anc ~t associated discersion ( factors. About 40% of the time tne :1fference :st..een tne credicted and observed values can be a facter of 10 or greater; again in either 4 direction. 1 Tnese data tend to support our initial comments regarcing uncertain-ties n~,::tatec witn the use of tne standars Pasauill f actors and the
- neec to oeveloo site specific data.
A.3 Exterots from a Workshes on the Evaluation of Models Used for Enviccnmental Assessment of Racicnuclide Releases l1 The crdng group suggested some tentative acct . acy statements on the estimation of airborne concentrations. Tnese statements are largely based on s:Mcific jucgement; there are no enough data upon which to l hase a rellaMe statistical estimate. For the ideal situatson of a high-ly instrumentec flat-field site from which previous data on meteorology i A-3 we - e .
. , - - . - - - + , , - - , ~ ., , - , . - - - -, , - - - - - , , - -
A. & V i .
% . m
. e i<
Table A-1. EVAlllATION OF liANFORO EXPERIMENT BY ORNL.
% of Observations Exceeding Limits' Factor of 5 or Greater Factor or 10 or Greater Stability Assigninent I4ethod* M (b) (d (b) 43% UP** 38% UP Pasquill-Gifford 62% 52%
57% 38% UP 38% UP St. Louis (Smith) 62%
?4% UP 33% UP 48%
Briggs' rural. 43% 52% OP 52% OP 62% 52% Brookhaven 62% 43% OP 38% OP Julich (100 m) 62%
.y
* (a) Stability class based on vertical temperature riif ference between 10 and 122 m above ground level.
(b) The standard rieviation of the winal siirection rmta: ureal at a height of 122 m 3;
** UP - Model un<lerpredt ts grounti level concentration relative to observed values (i.e _obs. pred.
majorfly of time). obs. <g OP - Model overpreilicts gronnel level concentrallon relative to observed values (i.e pred.. maj ority ol' time), l l 0 C o
m l h and airborne concentrations were available, it should be possible to T estimate to within 120% the ground-level centerline concentrations from a continuous point source at cownwind distances of less than 10 km. For a specific hour and downwind receptor point, the ac' curacy is
?very dependent on the calculation of the exact plume trajectory during a
-short period. For flat terrain and relatively steady meteorological .
4 conditions and distances of 10 km or less, the airborne concentrations , for an indivioual case shoJld be estimated to within about a factor of 110. For annual average concentrations values, the accuracy estimate is about a factor of 2. For a complex terrain er meterological situations (e.g., sea breeze regimes) a few experiments have indicated departures from estimates from the Pascuill-Giffore curves of mere than a factor of 10. h t <.e ve r , there are insufficient cata unor. wnich to base even a " scientific .iuccement" estimate Of accuracv. h
-( A.a Results c# a Survey o# Procrams for Radiological C se Comou-tations U A standarc acciden: release problem was presented to sets 31 nuclear facill:1es with tne recuest tnat the cloud gamma dose be calcu~.ated as a function of distance. The same ircut data were used by all articipants.
The results of tne various calculations using identical input are shown in Figures A-2 and A-3. The range in the calculated values is a factor of .r10 at the 1000-3000 m cittance. Considering that there are no absolute standards by which to judge the accuracy of the dose calcula-tions, one might ouestion, "now close is the range of values presented by these calculations to the true absolute value?" \ . D t
)
A-5
u l l 1 0; i(.
/
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. I \\
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\. B. G, J
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105 4 . s s '\
- 4. 's\s \
' g,s s
\
s N - g
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~
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N
\ ~
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\
.(.-
m :b \\s '\
$ %' I.p.\
s
\N F l
10 3 _ ., s, l
., g -
\
bB G.J
'- , ., D
's Cloud Gamma Dose A 10 2 sK C
Type F 1 m/ roc 1-hr Reisase. Ground Source inversion Conditions l i 10 - 1 -i = J t i I I 5000 10.000 100 200 500 1000 2000 Distance (m) Fl icep.s.792s O
) Figure A-2. Comparison of Different Dose Calculation Models, Class F (ref. .2)
A-6
- -r .-.v. , - - , _ -_ ..r-,,_._g.,,
O \ !(', 103 ' l I l l ] 1 - I Cloud Gamma Dose
\ Type C 4 m/sec s
102
\
g 24-hr Release. Ground Source _
\
\
\ -
s
\ N. Neutral Conditions
\
\
\
10 - 7 N \ E 4 \ f:
'(
o s,N. gs. t
\
y a hs, x\ s \ 1 - .k. ; - x N 10-1 _ w%*k g\ \ , x 4 A B. I
.E, K D l J
10-2 1 ! I I I H 100 200 500 1000 2000 5000 10,000 Distance (m) ICPP.S.7917 l Figure A-3. Comparison of Different Dose Calculation Models, Class C (ref.12) A-7
. . . . . _- ----------------__-_--__---.---_-.-----------------------J
9 k I t,'y 1
%; 1
( e O e l l 4 {-Ji VALUES FOR _y c AND o gUSEE :: DOSE CALCULATIONS
=e
r TABLE B-l VALUES FOR oy AND or llSED IN DOSE CALCULATIONSa , . ,, Distance 500 m 000 m 1600 m 3200 m Stability Class oy of ny .if oy or Oy Oz A 106 128 .164 326 312 1530b 586 8294b 8 80 53 125 98 239 278 449 920 C 57 35 88 54 167 99 313 176 0 37 19 57 27 108 45 204 71
- Y' E 28 14 43 20 81 31 151 47 F 19 9 30 12 56 19 105 27 6
a , Listed data are calculated values based on an eauation developed (13) to fit the data given in Reference 3. b Values greater than 1000 are not realist ic becaisse the maring layer depth (#1000m) can restrict verticle planne growth. 9 o
SE A 88 Pe##r DOC / U'S NUCLE AD CECULATOAV COMMIS$10N 644 BIBt.IOGRAPHIC D ATA SHEET f
,0 A
' " ' ~ " ' " ' ' ' " " " ~ ' ' ' ' ' ' ' ' ' '
'n Assessment of Offsite, Real-time Dose Measurement Systems for Emergency Situations 2. RECiPiENr$ 1.CcEss ON NO
- 7. AUTMORis!
- 5. r ATE REPORT COMPLE.TED h Tre W. J. Maeck, L. G. Hoffman, B. A. Staples, J. H. Keller March (VEAm 1982
- 9. PERFORMING ORGANi2ATsON NAME AND M AILING ADoRESS traervar I,a Come/
,t, ATE REPORT ISSUED Exxon Fuclear Idaho Co., Inc. "o*'" ("^^
P. O. Box 2800 Aoril m2 Idaho Falls, Idaho 83401 5''*'"
- 8. (Leave etonal 12 SPONSORING ORGANIZ ATIC'4 NAME AN" M AILING ADDRESS Isac,vae I,o Comes p
Division of Systems Integration Office of Nuclear Reactor Regulation ii. CONTRACT NO. U.S., Nuclear Regulatory Cocunission Washington, DC 205a: NRC FIN A6461
- 13. TYPE OF REPORT et nios cove mt o (mv.,s..e maarsJ Technical September 198' - March 1982
- 15. r,.P*LEMENTARY NoTsS 14 .;esve s e5&J 1s assiRAc7 2:: ~ ores a esu h .n evaluation is made of tne effectiveness of fixed, real-time mor.itoring systems around nuclear power stations in determining the magnitude of ung-itored releases.
(' The effects cf meteorological conditions on the accuracy with whi:r Ine magnitude of lunmenitored releases is determined and the uncertainties inherent in defining these. meteorological conditions are discussed. The number and placement of fixed field detectors in a system is discussed, and the data processing equipne. required to convert fiel:: detecter Outet: data into release rate information i cescribed. Cos: cata relative to the purcr.ase anc installation of specific systemt are given, as well as the characteristics and information return for a system purchased at an arbitary ces . 17 KEY WORDS AND DOCUMENT ANALYSIS 17e DEscRIPTORS Offsite, Real-Time Dose Measurement, Emergency 4 17e :DENTIFif RS OPEN ENCE3 TER A*S
,P9 is AV AILASIUTY STATEMTNT
' 19. SE CURITY CLASS (Tees recorrt 21. NO OF P ACES
~\
Unlimited Unclassified 2o SgTggg ,.. ,r> : price evmCsonu 333 1777, a j
- _ __ _ _ _ _ _ ___ _ _ . . - - . - - ~
(. Attachmant 10(B) -6 Submitted: 1 4dN f,$ ( - 1 Reviewed /OQA ngr .) w Approved / Plant Mgr.: l' SP Number 69 026.01 Revision: 0 Date Eff.: 7/09/82 TPC, TPC TPC PROTECTIVE ACTION RECOMMENDATIONS ( 1.0 PURPOSE This procedure provides guidelines for determining protective action recommendations to be given to offsite authorities. 2.0 PISPONSIBILITY The Radiological Control h 7ager, Radiation Protection Manager or In plant Radiation procedure. Monitoring Technic.ian is responsible for ensuring compliance with this PPF 1021.600-6.421 (
- - _ - - - - - - _)
I 3.0 DISCUSSION ( 3.1 The decision process used in determining a recommended protective action is based upon a number of factors. These factors are: release duration; magnitude of release; plume travel time; evacuation time estimates; dilution factors; shelter factors; dose limits; and dose savings. 3.2 Af ter determining a protective action, the Response Manager / Emergency Director will give approval to such an action before it is recommended. to offsite authorities. 3.3 Radiological Control Manager / Radiation Protection Manager or his staff may be available to perform this procedure for different distances used in SP69 022.01 Determination of Offsite Doses. This procedure explains the methodology for determining protective actions for points sequentially. 3.4 Topics covered in this procedure include: Page 8.1 Waterborne Protective Actions 3 8.2 Airborne Protective Actions 4 Appendix 12.1 - Waterborne Protective Action Guidance Chart Appendix 12.2 - Airborne Protective Action Guide Worksheet, SPF69.026.01-1 Appendix 12.3 - Evacuation Times by Wind Direction Appendix 12.4 - Shielding Factors from a Camma Cloud Source ( Appendix 12.5 - Thyroid-and Whole Body Guidance Charts Appendix 12.6 - Protective Action Map, SPF69.026.01-2 4.0 PRECAUTIONS Because protective action recommendations could be influenced by factors not considered here, use this procedure with common sense and judgement. 5.0 PREREQUISITES , SP69.022 01, Determination cf Offsite Doses and/or SP69 024.01, Waterborne Release Dose Projection have been initiated. 6.0 LIMITATIONS AND ACTIONS The Emergency Director / Response Manager *ust approve of a protective action before it is given to offsite authorities. 7.0 MATERIALS AND EQUIPMENT
)
N/A l l k SP 69.026.01 Rev. 0 7/09/82 Page 2 -
8.0 PROCEDURE 8.1 Waterborne Protective Actions 8.1.1 Dose Assessment Staff Member or In plant Radiation Monitoring Technician, perform the following: 7 8.1.1.1 Compare projected swimming (whole body and skin) and boating doses obtained from SP 69 924.91, Waterborne Ralease Dose Projection with the Waterborne Protective Action Guidance . Chart (Appendix 12.1). 4' 8.1. J . 2 Af ter consultation with the Radiation Protection Manager
, / Radiological Control Manager (if available), give the filled out worksheets (Appendix 12.1 and SPF 69.924.91-1) to the Emergency Director / Response Manager for subsequent approval and transmission to offsite authorities in accordance with SP69 999.91, ' Notifications.
1 8.2 Airborne Protective Actions 8.2.1 Dose Assessment Staff Member or In plant Radiation Monitoring Technician, complete the Airborne Protective Action Guide Worksheet (Appendix 12.2) as follows: 8.2.1.1 Obtain distance (item la), direction and affected downwind sector (item Ib), expected release duration (item 2), ground or elevated windspeed (item 3) and projected doses (item 10) from SP69.922 91, Determination 2 .( of Offsite Doses. 8.2.1.2 Using the distance (item la), the affected downwind sector (item lb) along with the Protective Action Map (Appendix 12.5) determine the zone (item Ic). 8.2.1.3 Calculate the plume travel time (item 4). 8.2.1.4 Contact the Control Room and determine the time the release started (item Sb) or the time the release is expected to start (item Sg). 8.2.1.5 Enter the current time (item Se or item Sh) and cocplete the remaining items (either items 5d-e or 51-j) for the 4 appropriate situation. 8.2.1.6 Determine the prevailing weather conditions and circle this in item 6. Adverse weather consists of conditions which will significantly reduce traffic speeds, such as rain or light snow. If severe weather (e.g. , flooding or blizzard) conditions exist a separate evacuation time will have to be estimated. i l I k l SP 69.926.01 Rev. 9 7/09/82 Page 3 i l
=- -, - -- -
i
- 8. 2.1. 7 - Determine the evacuation time and record this in item 7.
The evacuation time is found by turning to the correct
. table of Appendix 12.3 for the prevailing weather
_(. conditions (items 6a and 6b). Then with the affected
-downwind sector-(item lb), find the left most value that
,contains the zone (item ic) and pick the evacuation time for either day or night. (item 6c) conditions.
, 8.2.1.8 Complete items 8 and 9,to determine the time a person
- evacuating will be exposed to the plume (evacuation exposure period).
, 8.2.1.9 If field monitoring teams are deployed near the area of. concern, record the whole body dose rate and thyroid dose j commitment values in item 11. NOTE: Thyroid dose consitment is obtained by converting air sampler cpn readings by use of SP69 923.91, Thyroid Dose Commitment using the TCS Sampler. 8.2.1.10 Determine most reliable projected dose (item 12) based upon reliability of field taas measurements (if available). Record the projected. dose (item 12) on SPF69.922.01-2 Tabulated Dose and Protective Action l Worksheet. l ! 8.2.1.11 Complete items 13-15, and circle the higher recanmended protectire action for ite: 16.- Record these on the Tabulated Dose' and Protective Action Worksheet ( .._.... _ __ SPF69. 922 91-2. i 8.2.1.12 Record the protective action on the Protective Action Map (Appendix 12.5). 8.2.1.13 Repeat this procedure for other distances used in l_ SP69 922.91 Determination of Offsite Doses
.1 Consider recommending the same pr,otective action for adjacent zones.
.2 Consider recommending the same protective action for adjacent zones as distance from the plant increases.
8.2.1.14 After consultation with the Radiation Protection Manager / Radiological Control Manager (if available) give the completed worksheets (Appendix 12.5 and SPF 69 922 91-2) to the E=ergency Director / Response Manager for subsequent approval and transmission to offsite authorities in accordance with SP69.909.01 Notifications. l( SP 69.926 01 Rev. 9 7/09/82 Page 4
9.0 ACCEPTANCE CRITERIA ( N/A 10.0 FINAL CONDITIONS A protective action recommendation has been determined and approved by the Emergency Director / Response Manager.
11.0 REFERENCES
11.1 Shoreham Nuclear Power Station Emergency Plan
'11.2 SP69.922.91, Determination of Offsite Doses 11.3 SP69.924.91, Waterborne Release Dose Projection 11.4 SP69 923.91, Thyroid Dose Commiteent using the TCS Air Sampler 11.5 SP69.999 91, Notifications 12.0 APPENDICES 12.1 Waterborne Protective Action Guidance Chart 12.2 Airborne Protective Action Guide Worksheet, SPF69 926.91-1 12.3 Evacuation Times by Wind Direction 12.4 Shielding Factors from a Gamma Cloud Source 12.5 Thyroid and Whole Body Guidance Charts 12.6 Protective Action !!ap, SPF69.926.01-2 i
t-l ( SP 69 926.91 Rev. 9 7/09/82 Page 5 __- - - --__ ea---.
APPENDIX 12.1 l Page 1 of 1 l l
- ( l l WATERBORNE PROTECTIVE ACTION GUIDANCE CHART l
\ l IF THEN Projected whole body or skin Instruct the U.S. Coast Guard dose due to swimming is equal to remove all swimmers within to or greater than I rem. a 1 mile distance of the plant Projected whole body dose due Instruct the U.S. Coast Guard to boating is equal to or to evacuate all boats and greater than I rem. vessels within a 1 mile distance of the plant ( . . _ . . . _ . . . . . ( SP 69 026 01 Rev. 0 7/09/82 Page 6
Appendix 12.2
'Page 1 of 4
( AIRBORNE PROTECTION ACTION CUIDE WORKSHEET
- 1. Area of Concern
{. - a.' Distance miles (from SP69.022 91)
- b. Direction degrees, Affected Downwind Sector (from SP69 922 91)
- c. Zone (A-S, from Appendix 12.6)
- 2. Expected release duration _ hrs (from SP69.922 91) l
- 3. Windspeed miles /hr (from SP69 922.01)
NOTE: For ground releases use 33 ft. windspeed; for elevated releases use 150 f t. windspeed.
- 4. Plume travel time = item la/ item 3
= / = hours
- 5. Time until exposure begins (choose a or f)
- a. If release has begun:
> b. Time release has started (use 24 hour clock) ( c. Time of calculation (use 24 hours clock)
- d. Time difference = item Sc - ites 5b = hrs.
- e. Time = item 4 - item 5d = hrs.
1 i NOTE: If item Se is a negative number, enter zero hours.
- f. If release will begin later:
- g. Time release is expected to start (use 24 hour clock)
- h. Time of calculation (use 24 hr clock)
- 1. Time difference = item 5g - ite: Sh = hrs.
- j. Time = item 4 + item 51 = hrs.
SPF 69.926 91-1 Rev. 9 ( SF 69 026.01 Rev. 9 - 7/09/82 Page 7 3
Appendix 12.2
,Page 2 of 4
( 6. Weather condition and season (circle one for a,b and c):
- a. Ideal Adverse Severe
- b. Seasonal Non-Seasonal
- c. Day- Night
- 7. Evacuation time:
Use Appendix 12.3 along with information recorded in items 1 and 6 to determine the evacuation time. See procedure Step 8.2.1.7. ' Evacuation time hrs.
- 8. Exposure time =
item 7 - (item Sa or 5f)
= - = hrs.
NOTE: If item 8 is negative, enter zero hours.
- 9. Evacuation Exposure Period:
Smaller of item 8 or ites 2! hrs.
'DIYROID WOI.E BODY
( 10. Projected Dose (from_SP69.922 91) . 10. rem 10. rem
- 11. Measured dose from field monitoring teams (if applicable):
Monitoring Team Dose Rate X item 2 res/hr WB x hrs. 11. rem Thyroid Dose Commitment from TCS Air Sampler (from SP69 923.91) 11. res SPF 69.026.91-1 Rev. 0 ( SP 69 926.01 Rev. 9 7/09/82 Page 8 i
Appendix 12.2 Page 3 of 4 ( THYROID WHOLE BODY
*12. Most reliable projected dose 12. rem 12. rem (item 10 jg; 11)
*13. Evacuation Dose item 9 x item 12/ item 2 x / (Thy.) 13. rem x / (WB) 13. rem
*14. Shelter Dose Thyroid (a jy; b)
- a. For item 2 less than or equal to 2 hours item 12 x 0.33 = x 0.33
- b. For item 2 greater than 2 hours item 12 x (1 -
1.34) item 2 , ( x (1 - 1.34)
- 14. ren (a cy; b)
Whole Body Item 12 x Structural Shielding Factor (Appendix 12.4) x 14. rem
- Record these valves on SPF 69 922 91-2 SPF 69.926 91-1 Rev. 9 l
( SP 69.926 01 Rev. 9 7/09/82 Page 9
Appendix 12.2 Page 4 of 4 ( THYROID WHOLE BODY
*15. Refer to the Thyroid and Whole Body No Action No Action Guidance Charts (Appendix 12.5) and ,
Circle the appropriate action for each Shelter Shelte r Evacuate Evacuate I i 1 1
- 16. Protective Action Recommendation (Circle One)
No Action Shelter Evacuate I
- 17. Indicate item 16 on the Protective Action Map I (Appendix 12.6) for the affected zone i
( . . - - . . . . . . .
- Record these valves on SPF 69.922.91-2 l
SPF 69.926.91-1 Rev. 9 l l i l I l l l k SP 69.026.91 Rev. 9 7/09/82 Page 10
" ^
. m .'
e e EVACUATION Tlhied BY WIND DIRECTION , SEASON /AL (IDE?tt a.*Moono ts) .. ,. 0-2 MILES 7 0-5 MILES O 10 MILES WIND DIRECTION ZONE (8) WECK WEGK WEEK WEEK WEEK WEEK pn y)_o p y ZONES D8Y HIGHT
- ZONES DAY NIGHT (toward)
W by WNW A _ 2. 2S fer Af f. 08 . $0F AFKG sC8 .r. Cf - W A 2.25 2.25 A F G' Col f.e f AFGKO S~T8 4~.ZY , W by WSW As 2.50 2.56 ADFG .f. 0f f. d f ABFOKO- s r3 g r8'
.W S W Aa .t.so 2.so A a r.'O S.of C. o f AaFGnot s~rs s.ra WSW by SW As J ro 2.sb A D F'O 48f . S 08
- ABFOKRL' S.17 .r/ 7 SW Ao .7.ro 25D A B G' J.a* 3.re ABOKRLu ./ E /.J3 SW by SSW ABC 2.se 2.sb A8CGH J.93 ' 4.13 AaCGHKRLu 4.47 J.L 7 SSW ABC 2 58 2.56 ABCGH J.#7 f.fJ ABCQHRLuN S~#D  SSW by S DC 2 4z BCGH 3,I7 aCGnLun 3.s7 242 3.I7 2.17 8 BC 2.42 2.42 DCGHI. 3 17 3.o7 aCGHsuH0 3 67 3.s 7 S by SSE __
CD 2.17 2.s? CDnl 3.17 2 57 CDnluno 3.s7 2 7f ' u u, SSE CD 2.o7 2.t7 CDnI ' 3. s y 2 47 CDnIno 3.L 7 z.7/
** SSE by SE CD CDnI 2 17 2 17 3.17 247 CDunno 3.67 1 7s'
- c. .
y SE CDE 2.25 2 2r CDEIJ 4.67 4.or CDEIJ08 4 47 J.12 W SE by ESE CDE 2.25 2. 2 f" CDEla 4.67 d. or. CDEsJop8 4.L7 3.13'
? E S_ E CDE 2.2f 2 2f' CDElJ' d. 6.7 d.ef CDGlJOPS 4.47 3.13 ,
l yy ESE by E DE 2./7 2.s7 DElJ 4.47 4.df D EliO P S g.47 J.13 E? E E
- 2. co 1.au EJ 4.95 3.53 EJoP8 4.o1 J.3a .
C* E by ENE E _ f.do 2.ao EJ M.#7 J. 33 EJP /. ## ' 3 33
,4
. EM NO WIND ADCDE 2.30 2,sa N/A
- NIA p
.P,. E TIMES ARE EXPRESSED IN HOUnB AND INCLUDE 20 MlH. FOR MODILIZ ATION *
~
? e e EVACUATION Tilwdo BY WIND DIRECTION ', .
'~
neAl . samoa t. (swm. cwomaa0 ..' . 0-2 MILES 0-5 MILES 0-10 MILES WEEK WEEK WEEK WEEK WEEK WEEK WIND DIFIECTION ZONE (S) .__ DAY _ ZONES 'ZCNES DAY NIGHT N8DlLT_ DAY, N I RH.T_ (toward) ~~ W by WNW A l. r3 fo ra AF 4.77 1 4.xf AFKQ CSS f.39 W A ( .. (3 /.13 A F G* 4 7t* 4.7f AFGKa cas s.ss ADFG 4.7f 4.I ABFOKO F 33 s'. 3 9
- W by WSW 1; .z. 50 J.J6 S3S
- 3".89 WSW Aa J..rs .2.fD ADFO 4.7t~ 4.7f ABFOKQL J,JD ABF'O /.7f 4.7f
- ABFGKRL' 4. 9 2. ~ g. f a.
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] - - .
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.sensaan. coueur esaarrros) _
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APPENDIX 12.4 .g. . Page 1 of 1 RZPRESENTATIVE SHIELDING FACTORS FROM GAMMA CLOUD SOURCE (1) SHIELDING ' STRUCTURE OR LOCATION FACTOR (a) REPRESENTATIVE RANGE Outside 1.0 ___ Vehicles 1.0 --- Wood-frame house (b) 0.9 --- (no basement) Basement of wood house 0.6 0.1 to 0.7 (c) { Masonry house (no basement) 0.6 0.4 to 0.7 (c) Basement of masonry house 0.4 0.1 to 0.5 (c) Large office or industrial 0.2 0.1 to 0.3 (c), (d). building
- 1. --
(a) The ratio of the dose received inside the stre * - .9 the dose that would be
-received outside the structure.
(b) A wood frame house with brick or tone veneer is approximately equivalent to a masonry house for shielding purposes. (c) This range is mainly due to different wall materials and different geometries. (d) The shielding factor depends on where personnel are located within the building
.(e.g. the basement or an inside room).
(1) Ref: Sand 77-1725 (Unlimited Release) I s SP 69.026.01 Rev. Q 7/09/82 Page 15
APPENDIX 12.5 Page 1 of 1 ( THYROID CUIDANCE CHART IF THEN Projected dose (Item 12) is less than 5 res No action Shelter dose (Item 14) is less than 25 rem Shelter
- for children and women of childbearing age.
Shelter dose (Item 14) equal to or greater than 25 rem and evacuation dose (Item 13) Shelter
- equal to or greater than shelter dose.
Sheltr dose '(Item 14) equal to or greater than 25 rem and evacuation dosa (Item 13) Evacuate less than shelter dose. Shelter is to be with ventilation control. Ventilation control means turning off air conditioners or fans, closing doors and windows thus preventing (- access of outside air. . WHOLE BODY GUIDANCE CHART IF THEN Projected dose (Item 12) less than 1 rea No Action Shelter dose (Item 14) less than 5 rem Shelter
- Shelter dose (Item 14) equal to or greater Shelter
- than 5 rem and evacuation dose (ltem 13) equal to or greater than shelter dose.
Shelter dose (Item 14) equal to or greater Evacuate than 5 rem aad evacuation dose (Item 13) less than shelter dose.
- Shelter is to be with ventilation control. Ventilation control means turning off air conditioners or fans, closing doors and windows thus preventing access of outside air.
k SP 69.926.91 Rev. 9 7/09/82- Page 16
.'I PROTECTIVE ACJl^N MAP
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1 l l AIF/NESP TASK FORCE i( ,
' Gerald R. Davidson Kevin Rooney Sargent & Lundy Sargent & Lundy Task Force Co-Chairman Task Force Co-Chairman s
Task Force Members Mary L Birch Alan D. Miller Duke Power Company EDS Nuclear. Inc. Frank Congel Nicholas Panzarino U.S. Nuclear Regulatory Commission Yankee Atomic Electric Company Liaison - Task Force Sidney Porter Dean L. Erickson Porter Consultants, Inc. Cincinnati Gas & Electric Company David Sommers
,( John C. Golden Consumers Power Company Commonwealth Edison Company Edward A.Warman A. Scott Leiper Stone & Webster Engineering Corporation Atomic Industrial Forum. Inc.
Task Force Secretary ((
) .
M = " 'D C~FM *ST* 2 . ,n-*v t.wr;- ----.r--,
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AIF/NESP-023 I i . EVALUATION OF AN ENVIRONS EXPOSURE RATE MONITORING SYSTEM FOR POST-ACCIDENT j ASSESSMENT Prepared for the National Environmental Studies Project ofthe Atomic Industrial Forum, Inc. by I. , Charles C. Thomas, Jr. James E. Cline Paul G. Voillequ6 December 1981
')
- SCIENCE APPLICATO','S, INC.
3 Choke Chern Roa::. Rockvigt. Y..r,.4.nd 20850 Lag '"*--~' 1 r re ;- - - ' 7 T ;" " ~7 '*" 3'"" ' '~ ~ ' ' ~ ' ~ ~ ' ' " '
NOTICE I ( This report was prepared as an account of work sponsored by the Atomic industrial Forurn. Neither the Atomic Industrial Forum, nor any of its employees, makes any warranty, expressed or implied. or assumes legal liability or responsibility for the accuracy, completeness or usefulness of any inf ormation, cpparatus, pro-duct or process disclosed, or repres?nts that its use would not infringe pr;vately-owned rights. The opinions, conclusions, and recommendations set forth in this report are those of the authors and do not necessarily represent the views of the Atomic Industrial Forum,Inc its members, or its consultants. Because NESP is supported in part by Federal funds, the following notice is required by Federai regulations: The Atomic Industrial Forum's NESP activities are subject to Title VI of the Civil Rights Act of 1964, which prohibits discrimination based on race, color, or national origin. Written complaints of exclusion denial of benefits, or other discrimination on those bases under this program may be filed with (among others) the Ten-nessee Valley Authonty (TVA) Office of EEO,400 Commerce Avenue EPB14. Knoxville. TN 37902, and must be filed not later than 90 days from the date of the alleged discrimination. Applicable TVA regulations appear in part 302 of Title 18, Code of Federal Regulations. Copies of the regulations, or further information, may be obtained f rom the above address on request. (# Copyright @ 1981 by {( Atc mic Industrial Forum, Inc.
) 7101 Wisconsin Avenue Wt. thington. D.C. 20014 S2O oo Sponsors's50.OO Non sponsors , All rights reserved.
.ae . % h M % .s ggg , ; g- - y-e- ,-- y e % - . s C u . ,,
t_-
PREFACE ; s t The accident at Three Mile Island prompted the Neefear Regulatory Commission to propose substantial Il changes to Revision 2 of Regulatory Guide 1.97, and a draf t including these changes was published in Decem-ber,1979. Among other things this draf t document called for the presence of a ring of permanently installed, automatic radiation rnonitors around nuclear power plants. These monitors were to be sensitive over a very wide range (1pR/hr to 10 R/hr). and capable of providing real time information on accidental releases from both monitored and unmonite ed points in the plant. The purpose of this study, begun in the summer of 1980, was to examine the ramifications of the proposed. new requirements in the light of both costs and the technical limitations of the equipment available **off-the shelf ** at that time. While this NESP study was being developed, the proposed draft to Revision 2 of Regulatory Guide 1.97 was considerably revised as a result of the review process, and reissued in December. ' 1980. This latest version left wide latitude to utilities with regard to the approach used in obtaining post- ' accident monitoring information. Specifically a fixed a ray of exposure rate monitors was no longer required. This NESP study concludes that no system of instrumentation commercially available in August.1980, could have met the requirements of the December,1979 draft of Revision 2 of Regulatory Guide 1.57. The study [ further identifies the nature of the shortcomings of the systems evaluated. This is not to say that fixed moni- - tors cannot provide valuable information, esoecially in the early stages of a radioiopical emergency prior to the arrival of mobile survey teams. Furthermore, several vendors have recently devoted considerabie effort to de-veloping more effective equipment. Where the difficulties inherent in fixed monitors can be overcome they may, in some instances, prove to be effective in meeting the intent of the recuirements in the current version of Regulatory Guide 1.97. r I i This study was des;gned and guided by tne NESP Tast. Force listed on the inside frora cover. Soecial thanks b-y are due the Task Force's co chairmen- Geraic R Cavioson, who coordinated the ima; reviews, and Kevin
- z Roone), who guided the study in the beginning and f as continued to contribute signiMetty to it Also deserv.
( . ing credit for providing comments and reccmmendatens are the Al? Subcommittet t.r Emergency Prepared-ness and Siting Policy, chaired by Stever, J Miliots of Amencan Electric Powe* SerWet Cc pora;.on. and Paul L J. Pettit, now of Ha!!iburton Services, whc preccoe: r e as NESP Project Manager and Secretary far the Task Force. Finaty. the information and assistance prowded by the vendors listed ir. Accencir A. eroecially the { 1, Harshaw Chemical Company and Reuter-Stokes Inc . are most gratefuliv ackno..:enge: 5 li l A. Scott 1.eiper Project Manager, NESP h, Atomic industrial Forum. inc. i t k t i F il i t. 3 5 h t (' 5 t 4 4 W"M*.CC.C.'5"ril' . .l.' 2 ~NZ,r.& .W_ = - w% TEG*'M;.ng.W.~.MwWW.T?. $
~ . _ _ ... _ . _.
1
. l l
! l 1
i l
.?
u rf TABLE OF CC.NTENTS . [! .
/ '
. ; Pace J.O IDENTIFICATION OF COMMERCIALLY AVAILABLE INSTRUMENTATION........ 1 1.1 Approach................................................... 1-1 I~ 1.2 Findings................................................... 1-1 [; 1.3 Systems Offered By the Two Vendors Visited................. 1-2
; 1.3.1 Harshaw.............................................. 1-2 1.3.2 Reuter Stokes....................................... 1-3 1.4 Systems Offered by Other Vendors........................... 1-4 1.5 Calibration................................................ 1-4 1.6 Environmental Qualification................................. 1-4 i
2.0 CRITERIA FOR LOCATION OF FIMOTE STATIONS........................ 2-1 2.1 Description of Model E 71oyed.............................. 2-1 ' 2.2 Parametric Studies I e rformed. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-4 2.3 Re sults of Pa rame tr r Stu die s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-4
/
i 3.0 ESTIMATED DOSE RATES FOR ACCIDENT CONDITIONS. . . . . . . . . . . . . . . . . . . . 3-1 f 3.1 Design Basis Accider......................................... 3-1 + 3.1.1 Radionuclide Source Term and Estimated Plume I 4 Doses............................................... 3-1 ' 3.1.2 Building Shine Dose Rates For a Boiling Water Reactor (BWR)........................................ 3-7 3.1.3 Minimum Detectable Release Rates..................'.. 2-10 4 3.2 Analysis .of a..tiypothetical Environs Monitoring System at Three Mile Island....................................... 3-15 ' 3.2.1 Radionuclide Release Rates.......................... 3-15 3.2.2 Meteorology......................................... 3-17 3.2.3 Calculated Dose Rates............................... 3-17 3.2.3.1 Period of First Releases................... 3-17 l 3.2.3.2 Periods of Maximum Releases................ 3-2"i 3.2.4 Conclusions......................................... 3-23 l i- k.-_ l . ' ~r',w w:=t--- -n -...r~~=.,m w **. . s , w . - - - m------ -w.~- c- * ' * ~ '
TABLE OF CO!.~fENTS (continued) as (' , Page 3.3 Monitoring System Response to Hypothetical Accident at a Boiling Water Reactor.............................. 3-28 [L 3.3.1 Radionuclide Release Rates. . . . . . . . . . . . . . . . . . . . . . . 3-28 3.3.2 Meteorol ogical Da ta . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-28 3.3.3 Calculated Dese Rates . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-29 . 3.4 Use of the Environs Monitor to Project Doses For Other Loca t i on s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 -3 5 4.0 s'RITERIA FOR INSTRUMENT
- TION ANO ANALYSES.................... 4-1 4.1 Remot e S tation s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 -1 4.1.1 Detectors........................................ 4-1 4.1.1.1 Ene r gy Re spon se . . . . . . . . . . . . . . . . . . . . . . . . . 4 -1 4.1.1.2 EX pO sur e Ra t e . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 - 3 4.1.1.3 Environ = ental Requirements.............. 4-3 i'
4.1.2 Electronics...................................... 4-5 4.1.2.1 Ranga, Accuracy and Computational (, Re quir ements . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 - 5 5 4.1.2.2 Environmental Requirements.............. 4-5
, 4.1.3 Power Require..ents............................... 4-6 4.1.4 Transmitters..................................... 4-6 4.2 Central Station......................................... 4-7 4.2.1 Receiver an d ransmi tter. . . . . . . . . . . . . . . . . . . . . . . . . 4-7 4.2.2 Central Prc:essor and Peripheral Hardware........ 4-7 4.2.3 Data Analysis and Interpretation . . . . . . . . . . . . . . . . . 4-8 4.3 Environmental Qualifica tion s. . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-10 4.4 Quality As surance Levels . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-10 4.5 Sy stem Ca libration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 -10 4.5.1 Primary Calibration.............................. 4-11 4.5.2 Fiel d Cal ibra tion s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 -11 e
5.0 COSTS 5.1- Detectors and Electrcnics at the Remote Stations......... 5-1 . l Data, Transmission Fre the Remote Sta tion. . . . . . . . . . . . . . . 5.2 5-1 l
.k
; 5.3 Central Receiving and Processing System.................. 5-1 f.4 Installation............................................ 5-2.
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i I i l TABLE OF CONTENTS (continued) T g Page 5.5 Estimated Costs For a 16 Station Ring................... 5-2 5.6 Estimated Costs For Greater Than 16 Stations. . . . . . . . . . . . 5-2 REFERENCES......................................................... R-1 APPENDIX A Survey Information..................................... A-1 APPENDIX B Detailed Results.of Parametric Study................... B-1 APPENDIX C Listing of Program Plume - Routine to Calculate t.ne Response of Detectors to the Activity in a Plume Released During an Accident............................ C-1 APPENDIX D Estimate Noble Gas Release Rates of TMI-2 From TDR-TMI-116................................................ D-1 i I. .(, 4 I I t e i . t. p i i t
)
i m6 L*m W! -n..,- .. .63 m i ._ secV !dm.my.. , , w, . m Jr.- Y w ;- mc a ,. w 2 y * + - ,, = . ,., " - **. .
- t .-* -
l ( o . ILLUSTRATIONS Figure Page 2.1 Dose Rate Variation With Radius For Stability Class A....... 2-6 2.2 Dose Rate' Variation With Radius For Stability Class F....... 2-7 2.3 Dose Rate variation With Stability Class For Release Height of O Meters.......................................... 2-8
- 2. 4' Dose Rate Variation With Stability Class For, Release Height of 100 Meters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-9 2.5 Dose Rate Variation With Release Height..................... 2-10 3.1 Typical Noble Gas Core. Inventory After Shutdown............. 3-2 3.2 Relative Energy Release Rate For a Typical Core Inventory... 3-3 3.3 Dose Rate at 500-m From A SWR LOCA Source Term in the Reactor Building............................................ 3-8 3.4 Dose Rate From A BWR Reactor Building Eight Hours After A LOCA........................................................ 3-9
.{
D 3.5 Hypothetical Location of Monitoring Stations For TMI... ..... 3-16 3.6 Dose Rate at Maximum Station and Directly Under the Plume For the Period 0700-1300 28 March 1979...................... 3-25 3.7 Dose Rate at Maximum Station and Directly Under the Plume For the' Period 2000 28 March 1979 to 0200 29 March 1979..... 3-27 3.8 Oyster Creek Meteorological Parameters...................... 3-30 3.9 Exposure Rate Resulting From a Hypothetical Accident at a Boiling Water Reactor. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-31 3.10 L:pcsure Rate Resulting From a Hypothetical Accident at a Boiling Water Reactor....................................... 3-32 3.11 Exposure Rate Resulting From a Hypothetical Accident at a Boiling Water Reactor. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-3 3 3.12 Exposure Rate Resulting From a Hypothetical Accident at a Boiling Water Reactor....................................... 3-34 . 4.1 Block Diagram For Environs Exposure Rate Monitoring System.. 4-2 4.2, Distribution of Scattered Gamma-Ray Energies. . . . . . . . . . . . . . . . 4-4
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1 1 TABLF.S E ! Table Page 2.1 Summary of Parametric Studies Performed...................... 2-5 3.1 Release Rates For Major Dose Contributin Noble. Gas Radio-nuclides................................g..................... 3-5 3.2 centerline Dose Rate (mR/hr) For 1% Per Da Leak Rate and Wind Speed of 3 m/s.......................y................... 3-6 3.3 Minimus Detectable Release Rates With No Other Identified Releases, Wind Speed of 3 m/s (mci /sec) . . . . . . . . . . . . . . . . . . . . . . 3-11 3.4 Minimum Detectable Ground Level Release Concurrent With Design Basis Accident Release, Centerline Station (mci /sec).. 3-13 3.5 Minimum Detectable Ground Level Release With The Presence of BWR Shine (mC1/s)............................................. 3-14 3.6 Adjusted Noble Gas Release Rates at TMI-2 From 0700 Until 1300 March 28, 1979 { (Ci/sec).................................. 3-18 3.7 Adjusted Noble Gas Release Rates at TMI-2 From 200 28 March 1979 Until 0200 29 March 1979 (Ci/sec)....................... 3-19 3.8 Meteorological Parameters For TMI-2 0700 to 1400 28 March 1979..... ................................................... " 3-20 3.9 Meteorological Parameters For TMI-2 2000 28 March 1979 To 0 2 00 29 Ma rch 19 7 9 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 21 3.10 Dose Rates Above Natural Background At Hypothetical Monitor-ing Stations Around THI-2 For The Period 0700 to 1300 March 19 79 ( pR/hr) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ......... 28 3-24 3.11 Dose Rates Above Natural Background at Hypothetical Monitor-ing Stations Around TMI-2 For The Period 2000 28 March 1979 to 0200 29 March 1979 (pR/hr) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-26
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SUMMARY
Introduction The Three Mile Island accident initiated an extensive review of post-accident systems required for nuclear power plant operation. Following this. review, the Nuclear Regulatory Commission proposed revisions to Regulatory Guide 1.97, " Instrumentation For Light Water Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Follow-ing An Accident" Q). Among the proposed revisions was the requirement for an environs nonitoring system. Although the environs monitoring system was proposed for post-accident monitoring, a lower detection linit of 1 1,R/nr was specified. This lower ( limit was presumably chosen on the assumption that a single detector would r be used and that such a lower limit would insure a positive detector read-ing at all times. Study Goals l This iiiudy examined an environs monitoring syster only as it relates to post accident monitoring. It did not consider the merit of such a system I for other purposes such as environmental monitoring. The study addressed the following. six aspects: '
- 1. Availability of Instrumentation An evaluation was made in mid-1980 to determine the availability of environmental radiation monitoring systems to meet the appli-cable requirements being considered in 1980 for Revision 2 to Regulatory Guide 1.97 and to determine if the range 1 UR/hr to a 10 R/hr could be met with a single detector. ,
. S-1 b4 e = n U M . 4 m M Sp e b" .8 "*W"9. NI ** ' ^- ^ ^ '*
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- 2. Dose Rate Projectio_n, The study investigated the use of environs monitoring system data to project dose rates at other locations. l l
- 3. Response of a System to Accident Conditions F
. ; The response of an environs moni*oring system to an accident condition was evaluated i<or a design basis accident. Also in-
! vestigated were the responses under actual meteorology to the accident at Three Mile Island and to a hypothetical accident at a boiling water reactor (BWR). ,
- 4. Rf, lease Rates From Unmonitored Pathways .
h e study addressed the feasibility of using an environs monitor- , ing system for the purpose stated in an early draft of Revision 2 to Regulatory Guide 1.97:
"For estimating release rates of radioactive materials released during an accident from unidentified release paths (not covered by effluent monitors) .. .." , 5. Number of Stations Required to Determine Maximum Plume Exposure
{ o Rate t An investigation was made of the number of stations required to enable determining the maximum plume exposure rate (i.e., the rate directly under the plume centerline) within a factor of two. 1 6. Costs i The final objective of this study was to estimate the costs of an environs monitoring system. 4 Conclusions i n e following summarizes the results of this study:
- 1. Availability of Instrumentation As of August 1980, no system was available "off the shelf" that could measure over the range 1 UR/hr to 10 R/hr with a single de-tector. Nor was there available any system of multiple detectors g
covering this range with each detector.providing a positive read-f J ing at all times. S-2 l l
~ ~ ~ " ~ -
- m.:r.tsemm.ar .
l
- 2. Dose Rate Projection l
Using data from an environs monitoring system to project dose ( rates at other locations is a two-step process. 'Ihe first step is to deduce a source term from the environs data. The second is to make projections using this source term. Making accurate J- projections would be extremely difficult and in some cases im-possible because it would require accurately knowing:
- a. either plume centerline dose rate or location of the plume c.enterline relative to the dete cors;
- b. effective heights of all releases (monitored and unmonitor-ed);.
- c. energy compositions of releases; ,
, d. shine contributions to detector dose rates due to contained sources (e.g., airborne activity on BWR refueling floor);
- e. meteorological stability class;
- f. local meteorological phenomena (e.g. , looping, fumigation) .
l g 3. Response of a Syster. to Accident Conditions
- a. Resoonse for Actual Meteorology
'( .c ,
For actual meteorology at Three Mile Island and at a typical BWR, we found large fluctuations in dose rates that'would be measured by fixed environs detectors. Peak readings of a 16-
, detector system would change by a factor of 10 or more without any significant change in release rate. This would complicate the problems of following accident trends and of detecting re- $ leases by unmonitored paths. ! b. BWR Shine i
f For a Ma'rk I or Mark II BWR, shine from the refueling floor i could produce dose rates comparable to plume centerline dose { rates for close-in detectors. This would complicate data interpretation and make more difficult the detection of re-I leases by unmonitored paths.
- c. Design Basis Accident We examined a postulated design basis accident with a direct It/ day leak rate from primary containment to the environs.
At 250 m the peak dose rate was approximately one-tenth the , 10 R/hr upper limit spe-ified by the NRC. This was for the
, S-3 9.W =~~ m . - - ,.. _ .. - n _ __ . _ . __ - - - - . - - .
- worst case considered (Class 7 meteorology, ground level I
# release, 3 m/sec wind speed) . Therefore, 10 R/hr appears to be a reasonab'le upper limit at this radius for monitor-
'i ng a design basis accident of this type.
- 4. Release Rates From Unmonitored Pathways _
- a. In the Absence of Monitored Releases 5f there were no monitored releases and no shine from con-tained sources, an environs monitoring system would be 10 to 50,000 times less ' sensitive than a typical stack monitor.
Furthermore, unless one knew the release height and the plume centerline location relative to the detectors, there could be considerable uncertainty in the release rate de-duced from environs monitoring system measurements.
- b. In the Presence of Monitored Releases
.I
\
Honitored releases from our postulated design basis accident a- would decrease the sensitivity for detection of additional unmonitored releases by several orders of magnitude. To determine the size of an unmonitored release, effects of known releases would have to be accurately calculated and subtracted. In most practical circumstances, it would be nearly impossible to detect an unmonitored release in the
. presence of monitored releases.
- 5. Number of Stations Required to Latermine the Maximum Plume Exposure
~
Rate We analyzed the response of a ring of 16 detectors to a Class F 80-kev ground level relhase passing halfway between two detectors. For rings of radii 250 to 1000 m, none of the detectors would read within a factor of two of the dose rate below the plume centerline. We were unsuccessful in attempts to find a mathe-matical fitting function that would enable determination of the _ centerline dose rate within a factor of 2 for all stability classes.
. In our 1980 market survey, we found no vendor who had available a b}
' data analysis program to accouplish this.
l S-4 f
- - - - , ,. . ~ . , . . . , - , -
- 6. Costs For a ring of 16 detectors and associated central processing
[( ~, equipment, we estimate the following costs in 1980 dollars: Equipment $226,000 to $364,000 Installation: $40,000 to $2,000,000
~.
In addition, the cost for annual operation and maintenance should be included. The wide variance in the installation costs primarily results from individual site requirements which can greatly affect the cost, particularly for hard wired systems, t I i
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l Section 1 7-1.0 IDENTIFICATION OF COMMERCIALLY AVAILABLE INSTRUMENTATION 1.1 APPROACH To determine the availability of commercial equipment to meet the requirements of an early draft of Revision 2 to Regulatory Guide >1.97 Q), a very brief questionnaire was sent to forty-nine possible vendors. Thirteen of these were contacted by telephone as well as by letter. The questionnaire, the mailing list and a list of those contacted by telephone are included in Appendix A. Each vendor was also requested to provide literature describing an appropriate system. All of the vendors contacted responded. i As a result of the survey two vendors were visited to examine their facilities e and further discuss their capabilities. On August 6, 1980 we met with person-nel from Harshaw Chemical Company, a supplier of a system using NaI(Tl) de-tectors but who propose to use CaF 2 scintillati n detectors. The following day we met with Reuter-Stokes who provide a system using two ion chambers. 1.2 FINDINGS 2he results of our study indicate that as of August,1980: e l No system was, available "off the shelf" that could meet the pro-jected requirements. e Two suppliers, Hardaw and Reuter-Stokes, had manufactured and de-livered systems having similar but not identical specifications. Their emphasis had been on routine low level environmental monitor-ing rather than on accident monitoring. One supplier, PAR Systems Corporation, had a system available using two energy compensated Geiger Counters. e Many vendors were considering adding such a system to their product lines when the requirements and specifications became firm. 1-1 . o
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- Some vendors stated that they would accept orders to build such systems as special orders but only according to detailed specifi-
\ cations of customers. .
9 No vendor had addressed the problem of data reduction, presentation or interpretation. They provided chart recordings or data listings i. from each detector station. Any additional data reduction would be
- up to the user.
e Both vendors visited stated that they intended for the future' to make ,on-site maintenance available to customers but neither current-ly provided a "fsst action" repair service. 1.'3 SYSTEMS OFFERED BY THE 'IWO VENDORS VISITED 1.3.1 Harshaw
~
Harshaw felt they could meet the desired range from 10 R/hr to 10 R/hr with a single CaF 2 scintillator detector operated in the current mode ratJter than l pulse mode. Although they had not yet produced such a system, they appeared to have developed most of the required components. A perimeter monitoring system was part of their marketing plan. Their system as then envisioned consisted of a remote station with the Cary detector, a battery backed up power supply and a local microprocessor hard wired to a central station. The central station would contain an adjust-able high voltage power supply for each remote station. Plans called for the remote stations to be monitored in sequence, but the system could be altered to provide simultaneous individual interrogation. The use of the single de-tector would satisfy the NRC desire to have a system that, when functioning, always gives a positive reading. The remote unit was small: the detector was about 3" in diameter and 7" long and the electronics were in an 8" x 8"
- x 12" box.
I Harshaw was marketing an environmental system, TASC-4, capable of remotely i monitoring fields from 1 UR/hr to 10 mR/hr. The detectors had alpha stabilized gain (implanted Am source) and were operated as pulse counters. The systems
- already delivered supplied an analog signal to a panel meter or strip chart C recorder at a central location and had limited data reduction capability.
1-2 s W
, TASC-4 systems had been supplied to two utilities in the U.S. and a number of overseas companies.
/
1.3.2 Reuter Stokes
. Reuter Stokes was marketing a system, Sentri 1011, which used two detectors to monitor the range from 10 to 10 R/hr. A high-pressure ionization cham-ber (HPIC) was used to measure fields from 1 UR/hr to 100 mR/hr. This de-tector had two ranges with automatic switching in the electrometer. Radia-tion fields from 10 mR/hr to 10 R/hr were measured with a smaller ionization chamber. A remote station consisted of the two detectors, a power supply with a back-up battery, the range changer, a strip chart recorder, a digital display and data transmitter. Data transmittal could be by hard wire, tele-phone or telemetry. The central processor consisted of a computer with a storage device and printer. Reports could be generated on a s'chedule as well as on operator command. The system handled up to sixteen stations and inter-rogation of the remotes was through a direct link to each station. The system provided a continuous positive active response from only the low-range de-tector. The high range detector could be provided with an internal beta-emitting source to provide positive indication that the high range chamber was functioning properly. A practical way to provide continuous assurance '
that the electrometer switching device was functfonal had not been conceived. I Two systems were to be delivered soon to eastern utilities. Both systems were to have a central processor with telephone links to sixteen stations at one , and to ten stations at the other. Another system of ten stations was soon to be delivered without a central. processor. Processing for this system was to be by , a plant computer. In addition to these they planned to deliver two more systems of eight stations in the fall of 1980. Both would link to the plant computer. None of these systems was a telemetered system but Reuter Stokes had done some ; work with telemetry and was setting up a telemetering link between two of their buildings in Solon, Ohio. They felt that two miles would be a comfortable dis-tance for such a system. One problem they had encountered in their telephone transmittal systems was the inability of the telephone company to provide con-nections for an outdoor environment.
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1.4 SYSTEMS OFFERED BY OTHER VENDORS PAR Systems offered a complete syrrem that was manufactured in Britain and was not visited. Although a nurber of other vendors supplied information
, Other than on available detectors, none offered a complete system package.
those previously mentioned, five vendors stated that they could provide two Five respondents
- detectors that would meet the range 1 UR/hr to 10 R/hr. ,
L
~
indicated that they had a detector capable of measuring 10 R/hr with a lower limit greater than 1 UR/hr. Of these, three had a lower limit of 1000 pR/hr, one was 100 UR/hr and the fifth was 10 VR/hr. 1.5 CALIBRATION Both Harshaw, using a CaF7 current output system,, and Reuter Stokes, using a pair of ion chambers, had detectors whose responses were approximately energy independent. Calibration could be accomplished by cross calibration with any chamber calibrated to yield accurate exposure rates or using a gamma ray standard. Harshaw had calibration facilitier that they used to calibrate and control Co source that they had their TLD production and service work. They had a
/
calibrated to be a secondary star.dard through the use of air cavity ioniza-tion chambers that were cross-calibrated at the National Bureau of Standards. They seemed to have a fair grasp on the necessity and requirements of the calibration process. They believed that systems supplied by them should be field calibrated semiannually. a Beuter Stokes used procedures developed by DOE (HASL) for the calibration of the HPIC. They also depended on Victoreen Instrument Company for calibration services. Reuter ' stokes had a HPIC calibrated by Victoreen that was used as a secondary standard. All chambers marketed by Reuter Stokes were being cali-bratt.d bv referencing to this standard chamber. 1.6 ENVIRONMENTAL QUALIFICATION
'Ihe two vendors felt that a system could be qualified to IEEE Standard 323-1974. However, neither was willing to state that its instrumentation was qualified to this standard. Of major importance in this reluctance was the requirement for aging, a generic problem.
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t.
IEEE Standard 323-1974 provides guidance for demonstrating qualification
- I of Class 1E equipment. For equipment not located in containment and not l( /
subject to the extreme post LOCA environment, it can be interpreted to say that it should be shown that "the equipment can operate reliably under con-ditions that it is likely to encounter". These conditions are not severe gfor this type of equipment and both Reuter Stokes and Harshaw have had ex-
-perience with systems in the field. They both felt that they could qualify their systems to meet the conditions that would be found in field operations.
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Section 2 2.0 CRITERIA FOR LOCATION. OF REMOTE STATIONS A parametric study to evaluate the detector response under different assumed accident conditions attempted to establis?_ criteria for location of the re-mote monitoring stations. The location criteria include the requirement that the system be capable of determining, within a factor of two, the maximum dose rate on the ground directly under the plume at the radius of the detector ring.
2.1 DESCRIPTION
OF THE MODEL EMPLOYED t
; A finite plume model was developed to do this study that was based on the concentrations cbtained with the Pasquill-Gifford (2) model for atmospheric diffusion. This model calculates the exposure rate a.t any detector location I
for any given release rate and ga:=a-ray energy. The commonly accepted ex-pression " dose rate" is subsequently used for exposure rate in this report. f The concentration, x jC 3
, is dependent on atmospheric conditions (wind speed
; and stability class),(mre ease height and release rate (Ci/s). The concentra-
! tion and, therefore; the dose rate is directly proportional to the release I; rate and inversely proportional to the wind speed. The model was used to
} compute the dose rate per unit release rate (R/hr)/(Ci/s) at a unit wind speed
[ (1 m/s). The dose rate under specific conditions is then determined from the j calculated value by multiplying by the actual release rate (in Ci/s) and divid-ing by the wind speed (in m/s). i The plume concentrations were computed using Equation ~ (1) from Reference 2. i I 2-1 e e v-==a wwwM ----m,-w' r-
--n
- - - - w w.. . .4MPym,-_ _ __ __ g e w W, M***g #
_m___ _ _ _ _ _ _ _ _ _ _ - _ . _ _
(1) y(x,y,z) = exp (-h(y/cy) ) YZ ((I-where (2) G(z) = exp (- z-H) /0,) + exp(- ((z+H)/cg ) ) x = nuclide concentration (Ci/m ), Q = nuclide release rate (Ci/3), G = average wind speed (m/s), e = horizontal atmospheric diffusion parameter (m), y
= vertical atmospheric diffusion parameter (m),
c, - H = height of release (m), and x,y,z
= coordinates of point at which the concentration is computed.
The dose rate at the detector can then L; obtained by using a point source approximation and integrating over the source distribution.
- The exposure rate is given by:
(3)
= C E B(UR) T ,
Dl p o t hr , where C = 6.87 x 10" g (m /g),
= mass absorption coefficient for air at energy E En = ~ energy per photon MeV/ photon, B(UR) = buildup factor, and "*
T = uncollided photon flux The uncollided photon flux T is:
-U (4)
S e m 2-s I(ohotonsi 4wRZ
)
. 2-2 k -- - . _ . . . ** -- . . . . .
~ , ,
wh ra S = photon emission rate (photons /s), [(' i ' R = distance from source (=) , and u = total linear attenuation coefficient for air (m-1). The photon emission rate, S, can be found by assuming a small volume, dV, at concentration x as follows: S (photons /s)' = 3.7 x 10 xI k dV, (5) where 0 3.7 x 10 = the number of disintegrations'per second per C1, x = radionuclide concentration in the small volume element ; dV(C1/m ) , I k= rr.mber of photcas of energy En per disintegration, [
! and l dy = volume element considered (m ). ,
l
?f( Combining equations 4 and 5 with equation 3 gives: 1 J
Eo IkXe" B(UR) dV,
~
D = (6) R2 I l which is the contribution to the exposure rate at the detector due to the i small volume element dV. The total exposure rate is obtained by integration over volume of the plume. Calculational convergence criteria reduced com-putational time. In the parametric study Xu/Q was used in Equation (6) rather than X. This Produced an answer in terms of Du/Q or dose rate per unit release rate (R/hr)/(Ci/s) at 1 meter per second wind speed i
)
! 2-3 .
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2.2 PARAMETRIC S1UDIES PERFORMED
- i d The model can calculate 'the dose at any distance and directionThe for any des r
' combination of gasena-ray energy, release height and stability class.
dependence of the exposure rate on these variables was examined i ble. in a' p metric study that considered a reasonable range of values for each var a p a low-energy gamma-ray,
- Three energies were chosen for the present study:
133 Xe, a medium energy, 250 kev, representative 80 kev, representative of 135 Xe, and a high energy,1.5 MeV, representative of nuclides such as of 88 Xr. Three detector circle radii were chosen for examination, 250 m, 500 m, 1000 m. These were considered to be reasonable locations for on-site, site boundary and off-site locations for the detector ring. ii Pasquill-Gifford Class A, Class C and Class F atmospheric diffusion cond t Three release heights were chosen: a ground level, 50 m, were considered. and 100 m. These were chosen Seven locations were considered for each detector ring. O', 11.25', 22.5', with respect to the plume centerline and were at angles This choice was made so the study could evaluate 45', 90', 135' and 180'.
' the number of detectors required at a given radius and allowed examination In addition of four, eight, sixteen and thirty-two locations on each ring.
the choice 11.25' allows examination of the plume centerline half way between two detectors if the ring censists of 16 stations.
- A summary description of uhe parametric calculations performed is given in Table 2.1. Calculations were done for each parameter This variation for each of 1
required a total the seven detector angular locations discussed above. of over 500 calculations. 2.3 RESULTS OF THE PARAMETRIC STUDIES Plots Representative results of the calculations are given in Appendix B. of the computed exposure rates for a source of 80-kev gansna rays are shown in Figures 2.1 to 2.5. As can be seen from those figures, the shapes of the At a detector radius of 500 m with a ground response curves vary greatly.
- 1evel release and class A meteorological conditions (Figure 2.1), the dis-However, as can be seen in tribution with angle appears nearly Gaussian.
) Figure 2.2 for Class F meteorological conditions, the curve is ' cusp shaped and the dose rate falls off very , rapidly with angle.
2-4
- -- , ,_ _ , _* * * ' * ' - --~- , . _ _ . _ , . , , _ _ _ , _
l l I (. gI' .
/ Table 2.1 .
SUMMARY
OF PARAMETRIC STUDIES PERFORMED
' i RADIUS OF DETECTOR ATMOSPHERIC RELEASE HEIGHT ENERGY PING (METERS) STABILITY CIASS (METERS) (kev) 250 A 0 80, 250 & 1500 j SO 80, 250 & 1500
( 100 - 80, 250 & 1500 C 0 80, 250 & 1500 : 50 80, 250 & 1500 6 t 100 80,, 250 & 1500 ! F 0 80, 250 & 1500 ' 50 80, 250 & 1500 100 80, 250 & 1500 [ [( 500 A 0 80, 250 & 1500 :
/ -
50 80,.250 & 1500 100 80, 250 & 1500 C 0 80, 250 & 1500 50 80, 250 & 1500 l 100 80, 250 & 1500 F 0 80, 250 & 1500 50' 80, 250 & 1500 100 80, 250 & 1500 1000 - A 0 80, 250 & 1500 100 80, 250 & 1500 l C 50 80, 250 & 1500 i F 0 80, 250 & 1500 50 80, 250 & 1500
, 100 80, 250 & 1500 t
! I i i - i(-
- ) -
. 2-5 -
k' I
. , . - . . - . _ . , . - - . . . ~ -- - - - .-~ ~ . ---
- a m-v
Q ' Dose Rate Variation with Radius for Stability Class A 10-2
. Energy 80 kev
[ Height 0 m
~
Radius e 250 m a 500 m _ o 1000 m
. 10-3 _
E ~~ 3 3 8.s v> = r(_ ' 3 S 10~4 - EE ~ 55 _ 2a cc _
$ 10-5 _
i l i I e 10-6 0 22.5 45 315 337.5
' Location of Detector Relative to Plume (degrees)
{ Figure 2.1
")
. 2-6 l
l Dose Variation with Radius for Stability Class F (,
~
Energy 80 kev
~
]
Height 0 m 10-2 Radius e 250 m O 500 m
, O 1000 m E
" 10-3 __
o -
= -
G 8.2 e :: D U c~ -
*** C
- c. iit .c I'
Ei.-E 10-4 e_ E _ G 7 O o _ 10-5 N ee WEmuuD 10-6 i i i i 292.5 315 337.5 0 22.5 45 90
~ .
Location of Detector Relative to Plume (degrees) Figure 2.2 2-7'
...:. .- -_. .:,=---. - _. . . ~ _ , . - , . . _ . .
~ 9~ = _ _ . . . . -- --
Dose Rate Variation with Stability Class f for Release Height of 0 Meters .
~
Energy 80 kev
. 10-2
- Radius 500 m Stability
- Class A a Class C O Class F J
.E 10-3 E _
o - u
$~
&s -
(./ ue
.E E 10-4 is.c _
5E - 2 - a 3 0 8 10-5 - I I I I I 10-6 22.5 45 315 337.5 0 Location of Detector Relative to Plume (degrees) [ Figure 2.3
- 2-8 l
Io ~
i
/
1 Dose Rate Variation with Stability Class ! for Release Height of 100 Meters 10-3 Energy 80 kev j Radius 500 m l Stability e Class A
. a Class C 3 -
o Class F l t 10-4 i e E. E m. 2 (..,t2 -
'E3 .
3 .c '
.-82
$ 10-5 -
cc :- 3 g - . l 10-6 i l i i 315 337.5 0 22.5 45 Location of Detector Relative to Plume (degrees) Figure 2.4 2-9
)
- e. - . , . m .. - . -. ..--..,m.....
. _ .. ., . _ ..-. . - w,r-
Dose Rate Variation with Height {t. Energy 80 k'eV
/
Radius 500 m 10-2 Stability Class F e0m
- 0 50 m 0100 m y 10-3
- \
E ~
. .T" o -
a e es m~
".5 10-4
{IJ 3E u ~~~ S6 ~ e CC ~ e m O O 10-5 - l i l i i i 315 337.5 0 22.5 45 Location of Detector Relative to Plume (degrees) t. Figure 2.5 2-10 'E...p._..... . ..---...-....m... _ ,, ,,,,,,.., , . :wsuppr=_T;
A limited attempt was made to find a single function that could fit all the
"~ .
distributions resulting from the calculations. Both linear and logarithmic
, parabolic fits failed to calculate the maximum value within a factor of two accuracy. Two other functional forms were tried. These are i
Eh! - A
- B(Sin 2
(0+0O)
, (7)
Q Sin (0+0o) and 2 EE = Exp (Ae
~
).. (8)
Q The first of these is an approximation to the gamma-ray attenuation and the 1/r form of dose that describes the Class F conditions reasonably well. The other is a Gaussian in the log of the dose rate that fits Class A releases reasonably well. Neither of these functions provided a satisfactory fit to all of the calculational results. j
- r. It is clear that different functional forms are required to obtain the desired it
\ interpolation accuracy if the plume does not pass directly over one of the
/
detectors. The hope in finding such a function is that from the response of detectors within about 90 of the plume one might determine the peak value to be exoected for the clume centerline at any position between two de ectors. Nonetheless, we can examine the data as plotted to draw conclusions about l detector locations. The meteorological stability class has a pronounced effect on the dose rate. This can be clearly seen in Figure 2.3. For a given release rate and wind speed, the centerline dose rate at 500 m may differ by a f. etor of 10 from Class A to Class F. If the detector ring were at 1000 m this difference would be almost a factor of 50. At 250 m the difference is only a factor of 5. The effect of the radius of the detector ring can be seen in Ficure 2.1 and 2.2. For a Class F condition the dose rate at 250 m would only be about 2-2 times what would be dete ted at 1000 m. For Class A it would be a factor of abot 20.
)
2-11
- rr*'5iB!L_,_c.2. m m .m. . : - .. m- : + . . mm s v ;. s + ~. m, . .,. -
. . m.~=-. ==v * -.-L-
a
's For detectors 22.5" The effect of release height can be seen in Figure 2.5. .The degrees or greater off centerline the dose rate is almost the same.
centerline dose rates, however, differ by a factor of about 10. If the plume centerline passed halfway between two detectors on a ring that-had sixteen stations, the dose rate information available would be from loca-tions i 11.25* , 1 33.75', i 56.25' , etc. Figure 2.2 shows the centerline dose Thus even rate much greater than a factor of 2 times any measured dose rate. t with sixteen detectors it will be necessary to fit the data from the differen stations to obtain the centerline dose rate. Based on the results of these parametric studies, it appears that with no additional information, data from a ring of sixteen stations would not be adequate to meet the criterion of projecting the true dose rate within a factor of two. Furthermore,'because of the strong dependence on stability class and release height, further effort would be required to establish alge-rithms to estimate dose rates within a factor of two for plumes passing be-The simultaneous availak,ility of station meteorological
# tween two stations.
data would be essential to accomplish this cornplex task. G ) . 2-12
p .
. Section 3 3.0 ESTIMATED DOSE RATES FOR ACCIDENT CONDITIONS 3.1 DESIGN BASIS ACCIDENT.
The response of an environs exposure rate monitoring system has been evaluated for a postulated design basis accident. The analysis assumed a direct leak rate from primary containment to the environment of 1% per day. This rate was chosen to simplify scaling to other release rates. In a departure from normal accident assumptions, only the release of noble gases was considered. l 3.1.1
,k Radionuclide Source Tez1n r_nd Estimated Plume Doses It is necessary to estimate the radionuclide composition and the release rate from the source to estimate the dose rate that will be measured at any detector location. For purpcses of this study, only releases of the noble gases have been considered. Although the iodines are a potentially important contribution to the dose rate, experience at TMI indicates a relatively small release of iodines.
To obtain a representative radionuclide release for the present study, we chose the TMI-2 core inventory Q) as being typical. Figure 3.1 shows the total core inventory (Ci) for the major dose contributing nuclides for the first seventy hours following reactor shutdown. Using the average gamma ray . energy for each of these nuclides one obtains the relative dose rate con-tribution (MeV/s) from each nuclide as a function of time. Figure 3.2 shows > average gamma ray energies and the energy release rates for a full core in-I ventory. It can reasonably be assumed that the noble gases would all be released at the same rate so the relative contribution to gamma ray source term at any time can be deduced from Figure 3.2. 3-1 m_ --
~
1 l ( : : : 133 xe . ! 30 s ~ 1357 , 133m Xe 10 7
.= _
o 5 -
~
88 Kr 2 g _ E (.. e 10 6
~
1 o
~
87 . o _ Kr m _ E
>=
Typical Noble Gas Core 105 _ Inventory After Shutdown ( i I I I I I I i i i I 10 4 70 O 10 20 30 40 50 60 f Time After Shutdown (hours)
) Figure 3.1 3-2
l 10 18 f ,
' 88 Kr (1.9 MeV)
Relative Energy Release Rate for a Typical Core inventory 18 87 10 , Kr (0.79MeV) i 135 .Xe - (0.25MeV) 8 i A i o - 1 - 133 2 - e Xe (0.05MeV) ..
\
/
10 17 1 e ( -
, 133M 04Mev) l l I 1016 1 .
10 20 30 40 50 60 70 Time (hours)
) Figure 3.2 3-3
- - s v . m -_ . - ,1...:.m s - - - - = _ m;, a=v. m m &i -,
1 1 I
\ d are available
.It-is assumed that 100% of the noble gases escape the core an Table 3.1 shows
/ for release to the environment at a rate of 14 per day.
the resulting release rate for the important dose contributing nuclides at these values are directly
, I hour,12 hours and 24 hours after sliutdown,
- proportional to the leak rate and can be adjusted by the ratio of other desired rates to the 1%.
Using this source term, and a typical wind speed of 3 m/s, the dose rate Table 3.2 expected for a detector directly under the plume was calculated. The table gives the results for both ground level and' shows these results. As can be 100-meter release points for Class A and Class F stabilities. seen from Table 3.2 at any location and time, the observed dose rate can vary more than an order of mac,nitude depending on the release point and meteorological condition. This makes interpretation of the data difficult as will be discussed in Section 3.4. The dose rates calculated for the postulated accident are within the range of the detector response (1 UR/hr
' to 10 R/hr) for detector rings from 250 m to 1000 m.
}
I
~
l l l l l 3-4
* ~ ~ ~ ~ - - ..._;_ _, - - - . - . . - _ _ . . . . , .. ~ _ xxt _ w ;
m...
l l [(' i D l Table 3.1 RELEASE RATES FOR MAJOR DOSE CONTRIBUTING NOBLE GAS RADIONUCLIDES Computed Release Rate (Ci/s) # Nuclide T = 1 hr T = 12 hrs T = 24 hrs 131. M 17 17 16 135 , Xe 5 7 4 I 133 m I Xe 2 2 2 87 l Kr 3 0.01 - I 88 Kr 6 0.02
- f. '
0.4 33.00 26.41 22.02 (a) Assumes 100% noble gas release to containment and leakage to the environment at 1% per day. l
~
Wl 1
)
- 3-5 --
f
^^~
y . M e .g ',a _5 A* N *is e m'^- ge ;,- - + * -
l g-J TABLE 3.2 CENTERLINE DOSE RATE (mR/hr) FOR 1% PER DAY
~ LEAK RATE AND WIND-SPEED OF 3 m/s Class F Class A Ground Elevated Ground - Elevated Release Release Level "
Detector Imvel Release _ H = 100 m_ Release _ R= 100 m Distance 250 meters 1100 52 180 52 I hour 13 13 250 12 hours 42 140 7.2 24 hours 24 7.2 500 meters y 46 550 53 1 hour 61
> 130 13 13 11 12 hours 7,1 6.1 70 24 hours 7.5 1000 meters 50 12 330 I hour 11 78 12 12 hours 2.4 2.4 '
44 6.7 24 hours 1.3 1.3 l l l i 3-6
.=
3.1.2 _ Building Shine Dose Rates For a Boiling Water Reactor (BWR) ne radiation field from a Mark I or Mark II BWR reactor building following
/ an accident could be substantial.* A representative dose rate for a design {
3 basis accident is shown in Figure 3.3. It was assumed that 100% of the noble gases and 50% of the halogens were released immediately to the containment and subsequently leaked to the reactor building at it per day. The noble. gases mixed uniformly in the reactor building and were exhausted at a rate of two air changes per day by the standby gas treatment system. The aret below the refueling floor was well shielded so that only about 30% of the reactor building activity contributed to the dose rate. This data is pre-sented.only to provide an approximation to the building shine contribution. The time of release into the containment and the leak rate into the reactor building can substantially alter the amounts of high energy ganca-rays from short-lived krypton isotopes and change the dose rates appreciably. Figure 3.3 shows the postulated dose rate at 500 meters from the reactor building. The maximum dose rate of 17 mR/hr occurs about eight hours after shutdown. This has been calculated for other distances frore the building and is shown r, in Figure 3.4.
/
The building shine dose rates can be compared to the calculated plume dose rates shown in Table 3.2. At 500 m and below, dose rates from the unshielded BWR reactor building can approach or exceed centerline dose rates due to a Plume. Correction for this shine dose rate may be difficult. The shine may be anisotropic due to building shape and shielding irregularities. Shielding the detectors against reactor building shine would also be difficult. It can be shown that a substantial fraction'of the building shine dose rate is cue to radiation scattered from the sky above and around the reactor building. Reactor building shine calculations of this type would be required on a site specific basis for establishing a suitable detecter ring radiuc. r
- The secondary containment bf a* Mark III BWR would not produce a significant shine dose rate following a design basis accident.
) 3-7 -
i
,_ w--ww - ' '
l l5 l sm- ..m. A 1 i!' Dose Rate at 500 m from a BWR LOCA
' Source Term in the Reactor Building i 21 1
' 19 -
- 17 -
. c 15
.c j j j i3
~
i v. :: li
- m 3a 11 -
!! L CC i :
- o 9
- u> 0 o 7 - 5 - l I 3 - i
! j' 100 1 10
! fla . Time After Reactor Shutdown (hours) i,y2: I II
.t ' .i ; - . - - - . . . . _ _ - , . . _
',1 l
1 Dose Rate From a BWR Reactor
- Building Eight Hours After a LOCA !!
tr 1000 _ t l 100 l
.c t !
l - E - 1 (' 10
, c$
c - e __ i m O _ Q 1 - !
~
i i i i i I i Ill 1 i i i i 10 100 1000
- r. Distance from Reactor Building (meters) l 1
Figure 3.4 3-9 . l
'-'sCa - - - 1iI *wi""..
L E--Ap%* t._ dh h - " - " ""
. .. s l
, - - . .= . .- .- - . _
i s 1 3.1.3 Minimum Detectable Release Rate _s d level releases The response of an environs imonitoring system to small groundet second has been examined to determine the min mumFirst with no other rele
' three different conditions. h a BWR " shine" com-with a design basis accident release and finally wit ponent present.,
established that For the first case, with no other release, a criterion wasincrease its re a plume passing directly over a monitor should increase in both monitors For a plume passing directly between two monitors,anThis criterion of 2.5 VR/hr should be detectable. The dose rate was determined such increases should be readily observable. t 1 hour,12 hours, for a total release of the entire noble gas core inventory aT and 24 hours following shutdown. i d to produce the obtained by taking the fraction of this inventory requ re2.5 UR/hr for a plume
.I exposure rate of 5 WR/hr for a centerline monitor orTne resulti passing between two monitors. d of 3 m/s; these values can j are shown in Table 3.3 for an assumed wind spee 7
i f the desired be adjusted to other wind speeds by multiplying by the rat o o wind speed to 3 m/s. tes that produce i Table 3.3 shows that there is a wide variation i in release This is becausera the minimum detectable dose rate at the environs mon ii tors.to release height the dose rate at a particular point can be very sens t ve line. and to location of the point relative to the plume center l
/
\ O i (
) '
3-10 I i 0. ..
- - - - - - - . . _ . _ _ _ , , , , - - - - . - - . . . . . ~------.= __ mm.
1 , Tabli 3.3 l n2m nun oc7tenstz met:Ase mans win no omen romT2rzro mm.zasts, w!ND SPEED OF 3 m/s teci/s) i Class A Class r
.Cround Inval pelease Elevated mejses,n . 100m Ground I4 vel pelease (1evated pelease H = 100 m Detector centerline Off Center Centerline Off center Centerline off Center cent'erline off center Detector _ Detector' Detector Detector Detector Detector Detector Detector un2 i 1 hour 1 1 3 2 0.2 1 3 2 12 hours 3 . . 3 10 3 0.5 5 10 8 23 hours 5 5 15 13 0.s 7 15 14 5_00_m a bout 3 1 4 3 0.3 3 3 6 12 hours 10 12 12 15 15 18 1 10 24 nours 15 21 13 24 2 30 15 33 3000 a 1 hour 16 19 14 20 1 22 3 24 g
7 hours 53 87 53 94 2 127 11 144 hours 85 264 85 163 2 333 16 354 O '
)
3-1.1
% W - ---
bo $$ g g.7db~Q ((M 9 Ief f- "'-^2%..%,-- ,'m- _ w w g g ,- y- --79,6
t (E The sensitivities in Table 3.3 may be compared with the sensitivity of I Typical plant a typical noble gas radioactivity stack effluent monitor. Monitors with vent stacks have' exhaust rates of 100,000 to 500,000 cfa. For a 500,000 a sensitivity of 10 pCi/cc are commercially available.
. /
cfm exhaust rate, such a monitor could detect a release rate of .0.02 mci cc.
~
Under the most-favorable circumstances (Class F ground level release with , Pl ume centerline directly over a detector), the environs monitoring system In some cir-is at best 10 times less sensitive than the stack . monitor. f cumstances (Class F off-center releases), the environs monitoring system For i can be more than 10,000 times less sensitive than the stack monitor. a 100,000 cfm stack' exhaust rate, the environs monitoring system would be 50 to 50,000 times less sensitive than the stack mon.itor for the conditions i of Table 3.3. . The exposure rates resulting from a design basis accident release are given l in Table 3.2. The criterion chosen for an additional ground level release fnus,- , to be detectable was that the monitor. reading increase by one third.*
~
for Class A meteorology, at I hour a monitor at 250 m would have to increase , from 180 mR/hr to 239 mR/hr. Based on this criterion the minimum detectable additional release rates for a centerline station are given in Table 3.4. l s i i Th,e same criterion was used for the minimum detectable limit in the presenc > of a BWR shine component, i.e., the monitor should increase by a factor of :
- j one third. The dose rate contribution from shine and the minimu ground level release rate are given in Table 3.5 l i
I l L-i l 1 suggest that this criterieiI
- Results presented later (in Sections 3.2 and 3.3) k may be somewhat optimistic due to the effects of meteorological fluctuat ,
,, Such fluctuations can cause a monitor reading to change by a factor of 10 ,
without a significant change in release rate. _ 3-12
. l j
* ~.
. . . . . . . . , , _ , _ , =
. . . . ..... , , , ~~~cm. . , . . . ;
l.' - . . _ . _ - . . _ . . _ - . . .., ,_ _ . - .- --,,
1 ((ao Table 3.4 l 1 MINIMUM DETECTABLE GROUND LEVEL RELEASE CCNCURRENT* WITH DESIGN BASIS ACCIDENT RELEASE, CENTERLINE STATION I (mci /s) l Class A Class F Normal Normal Normal Normal Ground Elevated Ground Elevated Detector Level Release Level Release Distance Release H = 100 m Release H = 100 m 250 meters I hour 11000 3200 11000 520 12 hours 8700 2700 8700 450 24 hours 7300 2200 7300 370
,0 . 500 meters
~( 1 hour 11000 8200 11000 1100 12 hours 8700 7400 8700 870 24 hours 7300 5900 7300 740
*s.,
1000 meters ' s 1 hour 11000 12000 11000 1700 12 hours 8700 8700 8700 1300 24 hours 7300 7300 7300 1100
- These results are based on the "1/3 increase" criterion, which may be optimistic. See footnote on page 3-12.
n ~:, - I \ 3-13 4
-A t s, m a rcs 4-m w.sc a 4._:.:.rs.w,.m u.x,c-m..- d,== m u ; = = . aw- a si. na c. .
m I Table 3.5 MINIMUM DETECTABLE GROUND LEVEL RELEASE
- WITH THE PRESENCE OF BWR SHINE (mC1/s) j Minimum Detactable BWR Shine Minissa Detectable Limit Limit.
Dose mate Class F (mR/hr) class A , (aci/s) Detector (aci/s) _. Distance. 250 meters 1800 11000 l1 1 hour 180 6400 :l 184 38000 ' 12 hours 3500 67 20000 24 hours 500 meters 260 [ 1 hour 13 2300 640 9.6 6400 12 hours 150 1.4 1400 24 hours 1000 meters 10 220 I hour 0.22 20 0.15 540 12 hours 3 0 02 110 24 hours i l O
- These results are based on the "1/3 increase" criterion, which may be optimistic. See footnote on page 3-12.
(
) '
3-14
- - .. . . _ . . . ~~-- ---. ... _ _ ,, -
-- . ..aas.-- ..-. .--. :. .y :
3.2 ANALYSIS OF A HYPOTHETICAL' ENVIRONS HONI70 RING SYSTEM AT THREE MILE ISLAND 8, The meteorology and estimated radionuclide release rates for the Tr(I-2, ,
, accident were used to estimate the performance of an environs monitoring system had one been installed. The hypothetical system consisted of a ring of sixteen equally spaced stations at a radius of 1000 meters. The assumed ring is shown in Figure 3.5. The stations are numbered one through sixteen, clockwise, with station #1 located +11.5 from north. This parti-cular arrangement,was chosen because only one station (#10) would be located on the river.
The response of the monitoring system was evaluated for two six hour time periods. The first period was from 0700 until 1300 an March 28, 1979, when the first releases occurred from the plant. The second period was from 2000 on March 28 until 0200 March 29 and included the period of maxi-mum releases. It was assumed that all releases occurred at an elevation of 25 m.
- 3. 2.1 Radionuclide Pelease Rates I'
s The release rates were obtained fro- the Assessment of Offsite Doses From The Three Mile Island Unit 2 Accident, TDR-TMI-ll6 (4_) . This data is repro-g duced in Appendix D. There were sor.e adjustments made to the data as pre-sented. Due to the low contribution to the dose rate from *Xe, it was not included in the evaluation. However, Kr was included because of its relatively high energy gamma radiation. The data from Anpendix D were used to estimate average release rates for 15-minute pc 'n order to correspond with the reported meteorological data. ( Because of the short half lives of Kr, Kr and "Xe, the data from Appendix D was corrected for decay for the early time period. The average 88 release rate of Kr from 0700 until 0900 is reported as 5.5x10 pCi/sec. The release rate used was 7.8x10 pCi/see at 0700 and decreased to 3.9x10 3 i . pCi/see at 1000. This produced an average release rate of 5.5x10 pCi/see 135m f for the three hour period. Xe was treated in a similar manner. The
! ratio of Kr to Kr in the TMI-2 core inventory at 0700 was 0.29 (4) .
The release rate for Kr was assu.ed to r,e 0.29 x 7.8x10 pCi/see a 0700 i
) -
3-15
** " wak ' N "
g , k ' #
l l rigure 3.5 l y e HYPOTHETICAL LOCATION OF MONITORING STATIONS FOR TMI ,
~
\ HILL l f p g ISLAND Igr S N
\ N~k 1 $ , ('
\ , ..
\$ \ l I
\c# I 1 SHELLEY ,
)
d ISLAND' i l 3 ) (s 7 s 3
/ ,
s \, J ! /
\_. - t E s' l
\ l\ l i
i / , 1( No .,5 i
~ t k \ _
n (
~"
)
. = . . _ -
_ __ a
l l and it was also decayed. The release rates used for this study are shown in Tabl.es 3.6 and 3.7.
/ 3. 2. 2 Meteorolo g The 15-minuts average raw meteorological data was obtained frort Pickard, Lr.e
.and Garrick for the first day of the accident at TMI-2.
Tables 3.8 and 3. 9 give the meteorological parameters used for the two
~ different time periods. Also listed in these tables are the hypothetical station identifications for the monitor nearest the plume centerline and the deviation of plume centerline from the nearest station. For a zero I
degree deviation the plume would pass directly over a monitor. 8 During the first period of releases, meteorological conditions were mostly neutral (Class D) or extremely unstable (Class A) with one 15-minute period of slightly unstable (Class C). During the six hour period the win? direction was quite variable and blev in the direction of almost every station for some period. (
/
During the second period when the highest releases occurred, conditions
' were slightly stable (Class E) for the entire period. The wind speed did not change greatly and was blowing in the direction of Station 15 (%NW) most of the time.
- 3. 2. 3 Calculated Dose Rates Using the meteorological parameters and the release rates, the dose rates
~
were calculated for each fifteen minute period from O '00-1300 and f.om 2000' 0200. Since the meteorological data was available on ty for fifteen minute # *'" vals, the dose rates were considered to be constant during each ir.terval. I ""*
' meander and changing wind speeds during each interval could not be consider ud.
i
- 3. 2. ,.1 Period of First Releases The dose rates that would be measured at each hypothetical station during this period are shown in Table 3.10. A background reading of 7 UR/hr was i i T
/
t
- 3-17 '
a ' l
-t d .ed.AL-
,,1 - . ng&A T .g
- 8'
f Table 3.6 (t ADJUs1ED NOBIE, GAS RELEASE RATES AT TMI-2 FROM 0700 UNTIL 1300 MARCH 28,1979 ~ (Ci/sec) 135x , 135m xe Total 87 -88 133,j 9g j 4.7-03 0.123 9.8-02 1.0-02 ; 0700 2.2-03* 7.8-03 0.122 1.0-02 4.6-03 2;0-03 7.3-03 9.8-02 0715 4.5-03 0.121 9.8-02 1.0-02 0730 1.7-03 6.9-03 0.120 1.0-02 4.3-03 l 6.5-03 9 8-02 0745 1.5-03 4.2-03 0.120 9.8-02 1.0-02 0800 1.3-03 6.1-03 0.119 1.0-02 4.1-03 5.7-03 9.8-02 0815 1.1-03 4.0-03 0.118 9.8-02 1.0-02 9.9-04 5.~4-03 0.118 0830 3.9-03 l 9.8-02 1.0-02 0845 8.7-04 5.1-03 0.117 1.0-02 3.8-03 7.6-04 4.7-03 9.8-02 0900 3.7-03 0.117 4.5-03 9.E-02 1.0-02 0915 6.6-04 3.6-03 0.116 4.2-03 9.8-02 1.0-02 0930 5.8-04 3.5-03 0.116 {. g 3.9-03 9.8-02 1.0-02 0945 5.0-04 2.9-01 7.4 ( 3.3-01 5.8 0.9
~
1000 3.9-02 -2.8-01 7.3 5.8 0.9 1015 3.4-02 3.1-01 73 0.9 2.7-01 3.0-02 2.9-01 5.8 1030 2.7-01 7.28 2.8-01 5.6 0.9 1045 2.6-02 2.6-01 7.24 5.8 0.9 2.3-02 2.6-01 7.22 J200 0.9 2.5-01 2.0-01 2.5-01 5.8 1115 2.5-01 7.20 5.8 0.9 1130 1.7-02 2.3-01 7.2 5.8 0.9 2.4-01 1145 1.5-02 2.2-01 9.2 7.4 1.2 2.9-01 1200 1.7-02 2.6-01 9.15 7.4 1.2 2.8-01 1215 1.5-02 2.5-01 9.11 7.4 1.2 2.7-01 1230 1.3-02 2.3-01 9.10 7.4 1.2 2.7-01 1245 1.1-02 2.2-01 5
~3
- 2.2-03 means 2.2x10 .
JI
. 3-18
, Table 3.7 ADJUSTED NOBLE GAS RELEASE RATES AT TMI-2 FROM 2000 28 MARCH 1979 UNTIL 0200 29 MARCH 1979
( . (C1/sec) Time 87 E0 133 ' Kr h xe xe , Total 2000 6.4-04* O.15 16 5.4 0.49 22 2015 6.4-04, 0.15 16 5.4 0.49 22 2030 6.4-04 0.15 16 5.4 0.49 22 2045 6.4-04 0.15 16 5.4 0.49 22 2100 1.6-03 0.38 52 17 , 1.4 71 2115 1.6-03 0.38 52 17 1.4 71 2130 1.6-03 0.38 52 17 1.4 71 2145 1.6-03 0.38 52 17 1.4 71 2200 2.7-03 0.63 110 34 2.7 147 2215 2.7-03 0.63 110 34 2.7 147 2230 2.7-03 0.63 110 34 2.7 147 2245 2.7-03 0.63 110 34 2.7 147 {~. 2300 1.4-03 0.32 180 25 1.8 207 (" 2215 1.4-03 0.32 180 25 1.8 207 2230 1.4-03 0.32 180 25 1.8 207 ,
, 2245 1.4-03 0.32 180 25 1.8 207 2400 1.4-03 0.32 180 25 1.8 207 0015 1.4-03 0.32 180 25 1.8 207 0030 1.4-03 0.32 180 25 1.8 207 0045 1.4-03 0.32 180 25 1.8 207 0100 1.5-04 3.5-02 32 43 0.25 37 0115 1.5-04 3.5-02 32 43 0.25 37 0130 1.5-04 3.5-02 32 43 0.25 37 0145 1.5-04 3.5-02 32 43 0.25 37
- 6.4-04 means 6.4x10 .
p,9 ( 3-19 4 e S
- h _e d -
#7'-- "E=W i ~ .e *. e .g 9 e
- Tablo 3.8 METEOROIDGICAL PARAMETERS FOR TMI-2
) 0700 To 1400 28 MARCH 1979
( Centerline Deviation From Pasquill-Gifford Wind Speed Direction Centerline (degrees) (m/s) (degrees)_ Station Time Stability class
'2 81 12 0700 D 1.3 12 0 D 0.8 101 0715 .
11 2.0 90 12-13 0730 D 12 -10 2.6 69 0745 D 11 8 2.8 64 0800 D 12 0 3.1 79 0815 D 11 9 1.9 65 0830 D 12-13 11 D 2.1 90 0845 2 1.3 171 16 0900 C 9 -8 1.2 3 0315 A 16 -1 D 1.3 168 0930 11 1.8 225 2-3 D ( 0945 2.1 237 3 0 (( 1000 A D 1.3 203 1-2 11 1015 7 -9 D 1.4 317 1030 8 289 5 1045 A 1.6 5 -9 A 1.8 272 1100 0 1.7 259 4 1115 A 3-4 11 A 1.9 247 1130 16 -10 D 2.3 159 1145 16 9 A
~
2.1 178 1200 16 -7 D 2.1 162 1215, 16 -5 D 2.6 164 1230 2 -9 A 2.1 205 1245 (1[Windfrom...dagrees. t
) 3-20
.=
' ~ ~ --- .. . . , _ _ __
[ Y.Q.
l Table 3.9 I METEOR 0IDGICAL PARAMETERS FOR TMI-2 -
, 2000 28 MARCH 1979 TO 0200 29 MARCH 1979 Deviation from Pasquill-Gifford Wind Speed Direction ( Centerline Centerline Station Tium Stability Class (m/s) (degrees) Station (degrees) 2000 E 3.4 138 15 -8 2015 E 3.4 137 15 -9 2030 E 3.1 143 15 -3 2045 ,
E 3.3 145 15 0 2100 E 3.2 - 149 15 3 2115 E 3.6 152 15 5 t30 E 4.0 150 15 3 2145 E 3.6 150 15 3
, 2200 E 3.8 157 15 11 6
; 2215 E 4.2 147 15 0.5 I g.
2230 E 3.8 146 15 -0.5 hI I 2245 E 3.5 152 15 5
- l. ',, 2300 E 4.0 155 15 8 2315 E 3.8 146 15 -0.5 2330 E 4.0 146 15 -0.5 2345 E 3.4 138 15 -E 2400 E 3.5 157 15 10 0015 E 2.6 176 16 7 0030 E 3.1 191 1 0 0045 E 2.5 173 16 4
~
0100 E 2.4 169 16 0 0115 E 3.2 155 15 8 0130 E 3.0 124 14 0 0145 E 1.9 113 13-14 11 (1) Wind from.. ... degrees.
;-y I
3 - 2,1-i Y =:seRW sw d* ~ > L aCM' LY.TM
I l l asstamed (4) for each station and only readings above this estimated back-
' ground are given in Table 3.10. Also shown in this table are the maxi- '
usan dose rates that would occur directly under the plume centerline and ' the ratio of the dose rate at the maxistas station to the dose rate di :
. ly under the plume centerline.
During the period from 0700 until 1000 hours, the release rate was quite low and only two or roe stations would have registered above background. The effects of wind speed and direction are quite apparent if the dose rate Although the release rate is at 0715 is comparM to the reading at 0730. essentially the same, the wind was blowing directly at station 12 at 0715 The esulting dose rate at 0715 was 660 and in between 12 and 13 at 0730. WR/hr and at 0730, 28 VR/hr a' factor greater than 20. The release rate in-creased at 1000 hours and is reflected in the number of stations reading above background as well as the general increase in the maximum station reading. The maximum station reading along with the dose directly under the plume is plotted in Figure 3. 6. (
- ' 3. 2. 3. 2 Periods of Maximum Releases Table 3.11 shows the dose rates that would have occurred during the period Only stations 1,2,3,12,13,14,15 and 16 would of maximum releases at TMI-2.
The wind direction have shown readings above background during this period. anil speed were more nearly constant during this period and directed toward station Number 15 most of the time. The Pasquill-Gifford stability was As would be Class E for the entire period resulting in a narrow plume. expected for the Class E condition, the deviation of the maximum r.tation This can be clearly reading from the piume centerline was more pronounced. seen in Figure 3.7. It should be noted that the large differences between the dose rate directly under the plume and the dose rate at the maximum station that occur at 2200, 2400 and 0145 were when the plume passed half-
- way between stations. If the number of stations had been doubled so that there were 32 stations on the ring, one of these stations would have been
~
directly under the plume. (-
)
3-22 O' i -- - - - . ---_ .
3.2.4 Conclusions j This analysis of a hypothatical monitoring system applied to the TMI-2, ( accident conditions generally supports and reinforces the conclusions ! of the ideal model studies performed in Section 2. Even for periods of constant release rate and relatively unchanging meteorological conditions, the dose rate at the maximum station can vary as much as a factor of 10. The results presented in Figures 3.6 and 3.7 clearly demonstrate the need for an algorithm to estimate dose rates from plumes pas ing between two detectors. a
- i. (
eo e it i
) 3-23 1 .
E
"^ #5 *q' kMa4 ed'h , .ee Ti' --- k l U- * "
g" g' M*h
. 74 l
' N
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-nv .
1 . l I i
. e l
f Table 3.10 S AROUND 'Dil-2 DOSE RATES FOR ABOVE NATURAL BACKGROUND AT HYPOTHETICALnatioMONITORI THE PERIOD 0700'r0 1300 28 MARCH 1979 (UR/hr) . Itemtensa Flame sInsts am statten _ statlan member 15 16 _
- Centerline Flasne centertlne_
12 13 le
- a 9 to il .
6 7 423 .9 3 4 . 5 3 Tiene 1 2 _ 9.2 390. 7.5
! 1.0 660 0700 9.3 660. 9.3 253 0.1 0715 20, 28.
107 0.1 0730 16. 27.
,' 167 0.1 0745 42. s.4
) -
150 *0 0800 1.4 150. 1.9 f 229 .1
's 0815 46. 12.0 201 .1 0830 21. 21.
'f} '
0845 3.0 170. 171 .9 w 12 .8 0.0 h 0900 6.7 9.8 2.1 2.5 200. 294 1.0 0915 205 .1 0930 4.0 3.6 490 1.0 0945 21. al. 150. 20. 3.6 58. 20600 0.1 1000 20. 150. 490. . 5.2 { 8.2 18500 0.2 1015 2200. 2200. 54. 34, 1300. 370n. 75. 5.6 559 .9 4.8 1030 44. 1. .8 500. 400. 414 2.3 66. 1045 2. 1.0 320. 390. 59, 500 6.1 36. 1100 e.4 1 500. 150. 19. 418 1115 * ' 3.4 19. 150.
- 39. 6.2 9940 .1
- 39. 300. 300. 28. 1400.
1130 6.2 .9
- 50. 300. 41%
640. 150. 2.5 1.% 1145 4 4.A 52. 5000. 13300 1200 290. 33.
- 7 so. 6900. 10400 1215 470. 16. 4.t fl 4.2 23. 412 .8 230. 9.n 1230 350. 48, 6.3 1245 270.
i
a. t Figure 3.6
- DOSE RATE AT MAXIMUM STATION
- AND DIRECTLY UNDER THE PLUME
_ FOR THE PERIOD 0700-1300
, 28 MARCH 1979 Ps
!I I I F-1@ -
g I g i1
~ '
s t it I l il \ l ' It i
- l-I IJ l I1i ;
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o 102 - i g t i I
- 1 ll gg ___ DIRECTLY UNDER PLUME 101 - -
Y MAXIMUM STATION 6 6 9
'O 1 I i 1 I I I
~
0700 0800 Ca00 1000 1100 1200 1300 TIME f (
.) '
3-25
=
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1 1 I i 7A _- -J_m --- - - - = greg 1.., , - _ - , , , _ . = 4 g grg w1_ _ _ _. ., _ _ . ,g l
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( F Table 3.11 t!MG STAT 30NS AROUSE TM1-2 FOR litZ DOSE RATES ABOYE 1Eh1(WR/hr) URAL BAOCMERS AT EYPCmtETICAL Ratio M _ aximum Station _ M
. azisman Plume centerline 5tation M M r 2 3 _ Pi m centerline _
16 1 0.1 g 15 21000 12 13 TA 490 2500 18 21000 0.1 2000 11 0.8 2500 18 23000 11 490 2015 47 1.0 130 10000 22000 2030 6.6
- 4.8 ,
0.8 22000 74 70000 r 4.8 74 2045 400 16 0.4 8.5 120 55000 63000 2100 620 18 0.8 6.4 70 25000 56000 ! 2115 320 13 0.3 6.5 96 44000 63000 2130 360 17 0.04 7.6 110 49000 120000 2145 47 5.9 1.0 5600 5600 110000 5.9 47 2200, 300 16, 1.0 , 16 300 110000 120000 2215 330 18 0.3 330 120000 130000 2230 18 32 4.5 0.1 33000 1200 4.8 38 120 130000 2245 1900 21 0.9 h 37 12000 140000 2300 210 9.4 0.9 9.1 210 130000 130000 s 2315 200 e.9 0.1 s.9 200 120000 150000 2J30 43 0 04 25 2200 14000 150000 2345 5700 28 0.3 28 5700 3.3 200000 2400 24 73 62000 1500 1.0 . 7.7 12 160000 ' 2415 260 0.6 16 260 160000 210000 2430 760 19 1.0 120 120000 9.6 37000 2445 49 0.3 49 37000 28000
- 0100 370 2.9 1.0 56 2600 30000 0115 30000 39 0.04 39 47000 0130 5.9 0345 5.9 1a00 1500
/
e* I I i l ( n -
)
3-26 W ~- * ~ .~~ -- . - - - . . ..
Figure 3.7
. DOSE RATE AT MAXIMUM STATION AND DIRECTLY UNDER THE PLUME
- FOR THE PERIOD 2000 28 MARCH 1979 T0 0200 29 MARCH 1979 A
j ft%M./ I .
- e i d
105 _ s"# \ I , I h ' a%v ) - t m < [ t f a J l 1 w g - 4, p .1 !'i ti ! _ u
;~'
e o 1 1M - , I s . _ ~ DIRECTLY UNDER PLUME
- MAXIMUM STATION
. 1
- 103 ' t , , ,
7 M 2100 2200 2300 2400 0100
~
TIME . O . d '
)
' 3-27 O
~$_' ==w1_-_2 ,
, = " e -u i, ~ - _. y,
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NG 3.' 3 MONI*IORING SYSTEM RESPONSE TO HYPorHETICAI, ACCIDENT AT A B"~. I
- WATER REAC1OR Calculations of the response of an environs monitoring system te 2 loss of coolant accident at a boiling water reactor were also made. m Oyster Creek site was selected for the calculations because of the aval:2bility
- of site meteorological data. The monitoring system was assumed r: consist of 16 stations located at a distance of 1000 m from the reactor 1;:llding. ;
The average response of each detector was calculated during a 16-Nur period beginning one hour after the presumed reactor scram. 3.3.1 Radionuclide Release Rates _ The activity available for leakage from the drywell was taken te M the entire noble gas inventory one hour after the reactor scram. A noble gas activity leak rate of 0.lt/ day direct to the enviror..ent was assumed (via the standby gas treatment system and the %100-m chi-..ey, with-out dilution in the reactor building) . An additional 0.lt/ day 2cak was The assumed to be leaked into and mixed ur.iformly in the reactor bui;. ling. radiation field due to the airborne activity in the building wa, .-amputed and added to that due to the postulated release from the chimney. The noble 5( gas release was assumed to start at 0700 on 28 March 1979. 3.3.2 Meteorological Data The actual 15-minute average meteorological data collected at the oyster Creek site by Pickard, Lowe and Garrick were used to compute the dose from the noble gas plume. Figure 3.8 shows the average wind directions measured at the 116-m level.during the 64 15-minute periods between 0700 and 2300 on 28 March 1979. A weather system was passing through the area during the I period and a dramatic change in wind direction occurred. Wind speeds gen-l erally exceeded 4 m/s; hourly average values are shown in Figure 3.8. Atmosphere stability (based on the vertical temperature gradient) decreased from Class E to Class B and then returned to Class E. The stability classes
~
for each hour are indicated by the letters on the right side of rigure 3.8. l . l 1 . J
,3-28 e
b ., . . . . n ._ . . ,-
1 i 3.3.3 Calculated Dese Rates I
, } Figures 3.9 through 3.12 show the ec=puted responses of the 16 detectors
( to the' total radiation field from the plume and the reactor building. The Plotted values include the natural background exposure rate of 7 pR/hr. The field from the reactor building (0-2 pR/hr) was the principal source of radiation at detector locations 2, 3 and 4 during the 16-hour period. The plume component of the field was dominant at the other stations when the wind carried the plume toward them. The highest dose rate calculated was about 2.4 mR/hr during the period of lower wind speeds. Predicted doses for later times were reduced by higher wind speeds, radioactive decay, and during some periods, better dispersion conditions. (\
,i t .O
( 3-29
" M "A =
nC M"
'h4 K MM K = - M 1 ", f m*, m * '
P. r O] -- Q - s 9.s !~ . .. .
.N . .
AV(R AGE WIND DIRECTION WSW SW SSW S SSE SE ESE E ENE NE NNE N NNW NW WNW W gy, ,f, 7 I I I I I I I I I I fl i I I E.D 4.8 I I g - , Q D 2.8 i g , I ~ H D 2.7 10 - D.C 3.0 11 ~ . 8.C 4.3 - 12 - - 5 8.C 4.7 E 13 ~ 8 8.0
- C 8.7
, $ 14 -
* '8 o D.C 7.0 Y. g z 15 -
o D.C 7.8 { 16 -
- D 7.1
' 17 18 ~ D.E 8.9 . 18' ~ E 8.8 20 - E 8.4 E 8.2 , j .. 21 -
) i l i I I I I I I I I
- 1. I I 5 1 4
< 22 10 9 8 7 8 5 15 14 13 12 11 3 2 1 16
{ DETECTOR TOWARD WHICH THE WIND IS BLOWING , f, il
.it -
c _ _ - . _. -,_____ . . . _ , _ . _ _ _ _ _ . . . _ . _ _ _ _ _ _ . ~ _ _ _ - - _ - _ _ - . . _ . - . . _ . -
I' - 5 i
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v7 - O. ,9 - f .' b, Figtire 3.9 h
- l. -
EXPOSURE RATE RESULTING FROM A IIYPOTIIETICAL ACCIDENT AT A BOILING WATER REAC'IOR I'
*\ '
e'
.)
I]' AVERAGE EXPOSURE RATE ( R/hri it, [
.s8 asMEiEis
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. DETECTOR S
= .
as 3 I ; 10 3 g - 0 2 M a.3 5 a M4 - , 4 0000 -
"4 j 2000 - -
a " " t 1000 - (4 w a M H g - t 8< a: g 200 - a. N
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$ 20 - DETECTOR 10 10 - A
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= 2000 -
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t t t t I I I I 1 1 1 1 1 3
) 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 ,
HOUR 3-32 k--- .-
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1000 500 - DETECTOR 12 1 C. s 100 - ( 50 - l ) 20 -
" 10 -
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N oc i 1000 - 3 500 . L 0 DETECTOR 13 2 3 m - - o m 100 -
< ; 50 -
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et m 1 1 o M e M g 1000 w iy y w 500 - tn g $ DETECTOR 14 m e 200 - o. c
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< 50 -
I E - o m 20 - N O 10 - 2 -- U 5 - m N 1000 - g 500 - DETECTOR 15 fg M 100 - 5 I 3 50 - 20 - 10 -- nll I 5 - D I t i I i , e i I e f I
, 7 8 9 10 11 12 13 14 15 16 17 18 13 20 21 22 HOUR
[ 3-33
$ ,M E 'MM .^ ' -l, 'MWi'_* T M M C M 'sbE2-+**N * .
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$ 5 100 - -
)
M w , 8 F 50 - i
< 4 m:
f
- 20 -
'h gm e 10 - _
( 5 - DETECTOR 1
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i e= 0 t i e g 10 -_ _ m < 5 - DETECTOR 2 . ! E D
.2 3
m 10 - E 5 - g ~ DETECTOR 3 E DETECTOR 4 I l .g 10 - g ' ' ' ' ' ' ' 5 ' ' ' ' ' ' ' 5 7 3 's 10 11 12 13 14 15 16 17 18 19 20 21 22 , HOUR I Figure 3.12 g t.
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3-34 I e e 9 96% 9 GD* M g - ,, _
3.4 USE OF THE ENVIRONS MONITOR TO PROJECT DOSES FOR OTHER LOCATIONS Current computational techniques use radiation release source tems and (_' meteorological models to project dose rates. Derivation of the source ( term from environs monitoring data requires detailed knowledge or assump-l tions about the effective release height and energy spectrum of the source, and the meteorological conditions between the source and the monitors. Uncertainties in such a derivation could result in severe uncertainties in the estimated release rates. These would be compounded when combined l with the uncertainties in the meteorological models used to compute dose rates at locations other than the monitor location. l The uncertainties in source term resulting from uncertainties in the release height or in the energy spectrum can be seen from the r'esults in Section 2. Further, the vendors contacted did not have software that co.ald estimate l accurately the maximum dose along the plume centerline when the axis lies between two detectors. As can also be seen in the results of Section 2, this alone can severely affect the uncertainty in dose projection. The uncertainties in dose rate estimates associated with meteorologic redels
} '
will depend upon a number of factors. Principal among these are knowledge of the release location, the current meteorological conditions, and the dif-fusion climatology of the site. Even for relative:1y constant conditions, g the respcase of a monitor in a fixed location will exhibit variability associ-ated with minor changes in wind speed and direction that always occur. When meteorological conditions are changing, two types of uncertainty arise: (1) uncertainty in the location of the highest exposure rate at a particular distance due to changes in wind direction (both at the source and with dis-tance from the source), and (2) uncertainties in the magnitude of that ex-posure rate due to changes in wind speed and stability. Stability changes can depend on season, time of day, and the existing regional weather pattern. Short tem changes in wind speed and direction cannot be predicted reliably. l Predictions of exposure rates at specific points are highly dependent on wind direction at the release point and the influences of terrain, buildings, wind shifts with distance and other factors. If either the site terrain or weather l pattern is complex, then predictions of plume travel path will be highly un-certain and the dose rates predicted for specific points downwind may be grossly
\ in error.
' ) 3-35
~
f O
.._ . _ . . . . . . . . _ _ )
l-((s In sunnary, the accuracy that can be achieved when environs monitoring ii system data are used to predict dose rates at other locations depends *l l upon the ability to calculate the source term and the accuracy of the *; meteorological models to project downwind doses. Considerable effort must be expended to achieve the first of these requirements and sor.e I of the computational algorithms were not available ccamercially at the ! time of our market survey. ! I
?
e 6 l L I s h f 6 h t
) t 3-36 :
- -~.----.. ...... .-
- ~~
~ ~.-. -
I
) SECTION 4 4.0 CRITERIA FOR INSTRUMENTATION AND ANALYSES Instrumentation is required at both the remote stations and a central control location. Some data reduction capacity is needed at both loca-tions. A functional block diagram of the instrumentation for a perimeter monitoring system is shown in Figure 4.1.
4.1 REMOTE STATIONS Each remote station will likely consist of a detector (or set of detectors), l prear plifiers and current monitors (electrometers) , a data storage and con-l l trol microprocessor, and a data trar.sr.itter. Required ancillary equipment consists of power supplies, enclosures and environmental conditioners.
!(
) 4'.1.1 Detectors The detectors will probably be ionization chambers or scintillation detectors.
Both of these detector types can be used to measure dose rate. 4.1.1.1 Enercy Response Detectors used in the remote stations should be capable of measuring dose rate with reasonable accuracy (110%) and are required (1) to respond to photons of energy 60 to 3000 kev. Releases of gaseous activity soon after an accident will contain a mixture of several noble gases that provides a complicated energy spectrum over the range 80-3000 kev. Because of this, it is highly de-sirable that the detector energy response be flat over the range from 60 to 3000 kev. Detector response should approximate that of an air ionization cavity I as much as possibic. .This requires that the material of the detectors housing shoul:1 be a low atomic number material so that the forward scattered Compton ( distribution from it is similar to that fcr air, and the scattered electronic equilibrium should also be maintained in the housing.
..) 4-1 , - % m w . _ .. m w - n. = -~ ~~~ - - - -
~
High Voltage
-il S@
g/ g Supply
'l \ Power Supply elector (s}
- s. - Current Power Lines preamplifier l
, , ,n o -
l
"",5 "',i (Electrometer) n Typical Remote Station ;
Micro Processor
- u Transmitter
\U /* .
Receiver _ n i Link to Any Central Other Computer i i Processor Or Transmitter I
). !
1' y i Bulk Storage Printer ' t t I Central Station i l Block Diagram for Environs Expos!.re Rate , Monitoring System l 1 r Tigare 4 1 i L l
) ,. .
- 1
_ -- ,..,.. .u.= w m,e .
l l The required detector response at 60-kev demands relatively thin detector )
< wall and environmental housing thicknesse!$. Another consideration that l
( favors thin walls is that of dose build-up factors. The build-up
) results' from Coi.pton scattering of primary and secondar'y photons in the air surrounding the plume and the detector. The build-up factors are a function of both primary photon energy and distanc6 Spectra of scattered photons, also a function of primary energy, have been calculated (5_,6_) .
Figure 4.2 shows probability distributions for the scattered photon energies as a function of incident photon energy. As can be seen from the data in the figure, most of the scattered gama-rays have energies above 50 kev. For primary photons of 80 kev ( 33xe), 70s of the scattered radiation has an energy above 50 kev. Therefore, to accurately record dose from the scattered photons, a detector casing must be thin enough to pass 50-key ga:= .a-ra ys. A relative detector response of >70% at 50 kev is prcbably adequate. 4.1.1.2 Ext.osure Ratt
, The exposure rate rar.gs requested (l_) for the perimeter monitors is from 1 UR/hr to 10 p R/nr. This range of seven decades is difficult to achieve with
} a single detector. If it cannot be dor.e aith a single unit, multiple detectors must be used. The use of multiple detect *.rs complicates the system, however, because of duplicatien of components and . ,ci: elements at the rerete s:a-ions.
It is also difficult with multiple detect.rs to satisfy the desire of the NRC staff tcr have the detectors actively resp. .d to ambient radiation levels at all times. This point is discussed further if Section 4.5. 4.1.1. 3 Environmental Recuirements The remote stations will usually be located outside of any main buildings and either inside or outside the plant compou d. Hence, the detectors.will be sub-ject to the weather conditions of the sit'., ar f they must be capable of operat-ing in a temperature environment of -40 t, +15 C (-40 to +131 F) and in rain, snow and wind. To be er.vironmentally qua:ifd+f, the detectors must probably be placed inside small weatherproof enc 1:wru that may contain considerable ins,ulation to temper the rate of temperat re :hanges. Heating of the enclosure b(, in winter may also be required at the ec'ser sites.
) .
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eQ n .~% . m" I I Percent ol Scattered Gamma nadiation' Spectrum Havino
+
Energy Values Less than the Ordinate Value 99.9 99.8 99 98 95 90 80 70 60 50 40 30 20 10 5 2 10.50.2 0.05 0.01 99.99
'O' I I I i I I I I I II ,
II I I I I {l I 800 - Primary Energy N Diutribution
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.]
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4.1.2 Electronics
, The electronic instrumentation for the remote station typically consists of a detector high-voltage supply, preamp 11fier, current-to-voltage con-verter (electro =eter), analog-to-digital converter (ADC) and a micropro-Cessor. , 4.1.2.1 Range, Accuracy and Computational Requirements The range of the system as discussed above is from 1 UR/hr to 10 R/hr. If more than one detector is used to obtain this range, more than one ses of electronics is probably required. Furthermore, a decade of operational range is necessary as an overlap between detectors to assure an adequate cross calibration. The microprocessor must either ascertain the proper detectoz data to transmit (and must transmit the data from both detectors during the overlap period) or must always transmit data from all detectors.
Minimun a: aracy of data should be about 10% of the reading to achieve the desired Q) overall accuracy of half a decade. Tnis degree of accuracy can
'( be ach.sved with a low-resolution ADC (64-256 channel) . The use of loga-rithmic current decoders is also desirable. Minimum requirer.ents on the microprocessor are to (1) receive the data from the ADC(s), (2) periodi-cally store these data (typically every 10 seconds), (3) calculate and store time averages (nominally 1-minute and 10-minute averages) and (4) on demand, transmit these data through the transmitter to the central controller. It is suggested that at least three hours of these averages be retained in the memory of the processor. A " push-down" stack remory is a pessible technique.
4.1.2.2 Environmental Requirements The electronic circuitry, as the detectors, must be able te. withstand the extremes of the weather that include a temperature range of -40 to +55 C and heavy rains and winds. Unlike the detectors, the electronics may be Pl aced in heavy duty enclosures either above or below the ground. place-ment below the ground could offer additional advantages in physical security. Heating or cooling or both may be required to satisfy the temperature require-ments. { l
) 4-5 .mnh e-.-x _ . 5 ,,-,.ew .m : , , , ,, , .. nm_, , s - - - . - m , .,,_ _ . , . .,._ y ,
e 4.1.3 Power Requirements The power requirements (1) for the perimeter monitor are that it be "high (( reliability non-class lE power, battery backed when momentary interruption is not tolerable". A battery backed system is usually more reliable because of the susceptibility'of supply power lines to environmenta) damage.. A battery opera'.;ed system under " trickle charge" from the AC power is a very feasible method, but the capacity of a fully charged battery should be at least ten hours. 4.1.4 Transmitters Three principal means exist to transmit data from the remote sites to the central station. These are:
- 1. direct wire connection,
- 2. dedicated telephone line, and
- 3. telemetry.
s Direct connections often provide the most reliable method of data transmission. l For many plants Ic,ented on large rivers, this method may not he practical for stations across the river from the plant. tocations for run..ing the cables
. . via bridgas, underwater, or overhead may not exist at the plant.
)
Practical telephone syster.s can be supplied by either AT&T or nost independent telephone corpanies. An adequate telephone modem is produ:cf 'yr Western Elec-tric Company. Data transmission can be over dedicated voice grade phone lines at up to 1600 baud rate. Half-duplex transmission (bidirectional, but one direction at a time) is adequate. The lines are usually dedicated and rented, from a telephone company. Several commercial vendors can supply telemetry systems. When the sender and receiver are on a line of sight and the distance between them is small (i.e. , less than a mile or so), a low power frequency modulated (FM) . system is adequate and does not require an FCC license. The central antenna for such a system can be elavated (on the meteorological tower, stack, etc.) to improve communication between the central and remote stations. h A typical system will be tailored to a specific plant site and roay incorporate j a combination of a.$.1 of these means of data transmission. Data transmission k 4-6, 1 ;
... ~ . . . . . . , . .
l 1
'I i
by direct wire and telephone is usually by digital pulses and that by telemetry l l is ordinarily by modulation of the FM carrier frequency. In all cases, the 1( encoders and transmitters must be conditioned to withstand the extremes of the environment. They also will likely require an insulated weatherproof enclosure that may contain heating as well. 4.2 CENTRAL STATION The central controlling station will be located within a plant building in an area such as the Technical Support Center or Control Center. It would con-sist typically of a data receiver and transnitter to communicate with the re-mote stations, a processor to analyze and store,the data, a printer-keyboard and an interface to another device such as the plant computer or a data net. Communication will normally be through the keyboard. 4.2.1 Receiver and Transmitter The receiver and transmitter unit must be capable of two-way c mmunication with the satellite remote stations. The units are ma:es to the field ur.i s and the oparating characteristics of the receiver and decoder must natch. It is likely that the stations may be interrogated sequentially rather than sir.ul-taneously and that a " daisy chain" interrogation is acceptable. 4.2.2 Central 7rocessor and Peripheral Har: ware The requirements of the controlling and data receiving processor are that it periodically (1) receive the sets of time averages (e.g. , one-minute and ten-minute) from each remote station, (2) reduce further the data to obtain and store in permanent files time averages from each station, and (3) provide an interpretation of the data to give information on the magnitude of the acti-vity in the plume and the directica in which it is traveling. Hardware re-quirements for this task are:
- 1. Automatically sequenced data receiver and decoder (or set of these devices),
- 2. Mini- or micro-processor with a probable minimum storage capacity of 4096 words, j
- 3. Bulk data storage device such as magnetic tape, hard disk or floppy 1
disk, l (1
. I y 4. Printer and concand entry keybcard, and )
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- 5. Associated interfaces including an interface (e.g., RS-232C) i to another computer or phone link.
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The processor should have automatic power-fail recc.ery so that data are not lost by short power interrupts such as might occur in a plant shutdown.
. 4.2.3 Data Analysis and Interpretation In a typical application each station may transmit averages every five minutes.
This results in 192 individual data points each hour for a typical system with sixteen remote stations. A simple listing of these data would not be parti-cularly useful. There normally would be too much data to compile for an environ-mental or accident report. In an accident, the desired use of the system is to provide timely and as-accurate-as-possible data to aid in environmental assessment. Neither time nor personnel are generally available to perform lengthy analyses. Hence, the data analysis system should be capable of some P minimum data interpretation as well as data reduction. ' The data received by the central station mus- be identified by time and date and stored in a retrievable form in bulk storage. Additional desired data re-k duction would consist of calculating longer time averages (e.g., hourly and daily), storing these averager, writing raitp:,rts and transmitting data to another system or := a data net either c.n demand or r:.utinely. The system must aise ; be able te write reports on demand. These re;.:u enents are not partirularly demanding and require relatively simple programming. It is highly desirable that the prr; ramming be done using a universally understood source language such as FORT?.AN IV or BASIC. The analysis program should provide analysis and reporting options so different analyses can be performed and reports print- l ed as required. The principal purpose for this system stated by 'he NRC in early drafts of Reference 1 was analysis of otherwise unmonitored releases from a plant. Ideal data interpretation from the Ring-Around-7ne-Station (RATS) would give the release rate (in Ci/s) from the unmonitored release. In practice this is nearly ir.p:ssible. One must account for the contributions to the plumes from all of the monitored releases and one must know the size, location, altitude and gamma-ray energy spectrum of the plume activity to begin to ralculate ::.e unmonitored release rate. Most of these data are not veil l
-) i 48
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known. In reality, the best practical analysis uses the data from the p perimeter monitors to estimate the dose rate on the ground directly
} under the plume and verify the direction of transport of the plume. I Data interpretation at boiling water reactors (BWR) presents additional complications. An accident at a BhR can release a substantial fraction of the core inventory of fission gasses into the secondary containment of the reactor building. In such a case the direct radiation (shine) from the nearby unshielded reactor building can contribute a substantial dose rate to the perimeter stations (as much as or more than the plume itself).
This contribution must be subtracted from the individual station readings before the dese from the plume can be computed. Some of the remote stations are partially shielded from the reactor building by various facilities, such as the turbine building. Hence, the response of each station to reactor building shine is both site and location dependent. It must be established at the time of system installatio. and included in the interpretative soft-ware. Data interpretation for normal cor.ditions at a plant wil". is dependent princi-pally upon the desires of the cust:mers. At a minimum ti:e-; should consider radiation fields from radioactive waste handling and s:crage areas, liquid storage tanks, and the turbine building at c. BW.:.. There r.ay be other contri-butions to remote station readings that are site spe:ific. If the detectors are to Operate and the reporting is to be at environmental levels, the analyses might include _.rl effects of rain-out of raden daughter activities, diurnal radon concentration variations and the effects of plu:nes from nearby industrial releases. Good environmental level analyser and preter.tation coupled with a Scod public relations program could be highly valuable te a utility. Graphics i could be used advantagecusly. 4-9
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- l 4.3 IM'IRONMENTAL QUALIFICATIONS The NRC has not required that the perimeter monitor including the installation l :
# be seismically qualified; but it should conform to the reqairc aents of NUREG-0588. This has been further interpreted to state Q), " Instr.: mentation com-ponents should be installed consistent with the criteria of the Uniform Build-f f
ing Code applicable to the building or structure in which the instrumentation components are mounted or housed", and " Assurance must be obtained that equip-ment has been designed and qualified to meet its performance and service re-quirements". Our interpretation is that it must be demonstrated by design and construction review that system will operate under the nermal and accident con-ditions to which it will be subjected. 4.4 QUALITY ASSURANCE LEVELS Requirements for a vendor's quality assurance program are highly subjective. In most cases they will be established by the purchaser rather than by the NRC. f At a minimum the vendor must have a quality assurance and quality control pro- I s gram and must provide complete engineering drawings, manuals and dr.:u tentatier. of calibrations. The calibratiens must be traceable to the National Buresu of ([ f Standards (NBS) . ! I 4.5 SYSTEM CALIBRATION : The detectors and electronics for each remote statier. must b - cali:r:-ted tc giW the conversion factors relating detector current output to dose rate (R/hr). l
'this in turn must be related to give the signal at the output of the needer at the central station in terms of dose rate at each detector. This r.:st be f done as a function of gamma-ray energy over the range from 60 to 3000 kev. It k is probably sufficient to do this r.t only a few energies (e.g., 60, 355, 663 and 1275 kev). Because of the nature of the detectors that will prohably be l used, only periodic verificatior. calibrations need be performed after the {
initial extensive calibration is completed. Members of the NRC staff have e>: pressed a desire that all detectors in a t*:r'er be actively reading at all times. This would include all detectors used t:
.- achieve the seven-decade range requested (l_) . It is possible to achieve this T(
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4-10 ; I
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2 feature in the higher-range detectors by depcsiting a beta emitting source , on the inside of the detector chad er. Such a source could also serve as l n [ a' continuous calibration verification for these detectors. At the sacrifice
/ of very low-level environmental monitoring, such a source could also be in- -
i cluded with the low-level sensor. Alternatively, the time variations in the a background exposure rate could be used. t i 4.5.1 Primary Calibration ( Se system vendor will likely do the initial or primary detector calibration, j Sources, denoted here as primary sources, should be used. The primary sources ! are calibrated with an air cavity ionization charber. This calibration should ; either be done at the NBS or with a cha2er calibrated by the NBS. There should be a collection of these sources that covers the energy range 60 to ; 3000 kev. Typical sources could be 241-Am, 133Ba, 137Cs and 60Co. t System calibrations should yield the conversion constant for the output of the de-f coder at the central station. This should be determined for each energy over ; the range of the entire detec c: syste:n C 2R/hr to 10 R/hr) with measurements , at values such as 10 uR/hr,100 uR/hr and 1 mR/hr,100 mR/hr and 1 R/hr. Docu- t mentation of these measurements must be supplied to the purchaser. J l Calibrations must effectively treat the effect of dose buildup from scattered { , radiation. The normal and acceptable tech..ieue is to use a shadow shield between the source and detector :c r.easure the buildup dose. Subtraction of
- this reading from the dose measured with no shield yields the dose from the unscattered radiation. De primary sources must have been calibrated in the same manner. i 4.5.2 Field Calibrations Periodic field calibration verifications are necest.ary. These can likely be.
done at only a few energies such as 60 and 663 kev and at only one or two
- dose rates in each detector range. The detectors may be supplied with the
~
beta source on the inner surface of the chamber se as to have each detector be actively reading at all times. If this were de.e, some calibration veri-fication is done continuously and additional calibration requirements can be less extensive. Field calibratioh verifications should be done at least every six. months.
) 4-11 l-
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- If the verification measurements show that a detector has changed its efficiency by more than an established a2nount (e. g. , 25%) , it should be replaced. The replacement detector should receive a pris.ary calibration to relate output current to dose rate and the output current must be re-
' lated to the output of the central station decoder.
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I i f-1 J Section 5 5.0 COSTS In determining the initial cost for a perimeter monitoring system the following items must be considered:
- 1. Detectors and associated electronics at the remote stations.
- 2. Eata transmission fror. the remote station..
- 3. Central receiving and processing system.
- 4. Installation.
The cost estimates for detectors, data transmission and the central system I were chtsi ed fr= ver." rs of su:h epi, ment while the ecsts ' ass::leted with installatior, cf t'.e spi; .en were supplied by various util:::es. . p k 5.1 DETECTORS AND E:.ECT:.ZICS AT THE REMOTE STATIONS Each remote station will include a detector with its associated electronics, including a microprocesser data handling device. This equi,=.er.. is estimated to cost S10,000 to S15,000 for each remote station. 5.2 DATA TEA'SMISSION FROM THE REMOTE STATION Three methods of data transmission were considered: telephone, telemetry via FM signal and a hard-wired cable transmission. A signal processor for telephone transmission will cos: 31,000 to S1,500 for each station. Telemetry g via FM transmission would cost about $4,000 per station. The costs for a I I hard-wire system are included i the installation costs given below. i 5.3 CENTRAI. RECEIVING A'C PRO II, SING SYSTEM The central data re:eiving and .trocessing system will include a data receiver
'} and transmitter, a central processing unit and a printer / keyboard I/O device.
The cost for this is estimated :: be $50,000 to S60,000.
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1 i j 1 5.4 INSTA*.LATION I For a telephone communication cr a telemetry system about $500 per station ,j i should be sufficient for the actual installation. -In a'ddition about S2,000 l' to $3,000 per station would be required for design, engineering and fabrica-tion. The costs for installation of each station could be increased another
$1,000 if it is necessary to . include enclosures and temperature control de-vices to heat and/or cool the electronics.
Installation costs for a hard-wired system could be quite large. One utility , that has installed a hard-wired system of six stations estimates the present cost of installation would be about 567,000 per station. Another utility has made a recent cost estimate for a 16-station hard-wired system to be locatec i 500 m from the center of the Auxiliary Building of a 2-unit nuclear generating . station. Seven of ti.e locations would be in the cooling pond at that distance, - a factor that has a significant bearing on the costs. The total costs for four l alternatives vary from S1.1-1.2 million to Sl.9-2.0 milli on. The lower range applies to options that sacrifice unifor::.ity of radial d: stance :: cbtain sites .
/ on solid ground. That is, the cost of installing 'e systems in a cooling pond 8
is %SO.8 million more than installing them on dry land. The vari..nce ' within j the two ranges given is $60-70 thousand which represents the cost differential l between providing an uninterrupted power supply (UPS) from the Auxiliary Build-ing and providing power from the nearest avai' able sour:es. 4 5.5 ESTIMATED COSTS FOR A 16 STATION RING Based on the information above the range of cost for a 16 station environs monitoring system would bes
.\
' Detectors and' Associated Electronics $160,000 to $240,000 l
Data Transmission 16,000 64,000 l Central Processor 50,000 60,000 t Installation 40,000 2,000,030 l Total System Cost
- S266,000 52,364,000 S.6 ESTIMATED COSTS FOR GREATER THAN 16 SIATIONS The recent cost data employed in making the estimates given above car. be used to evaluate the costs of larger hard-wired systems at the sa a M :e.
5-2
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- Doubling tha~ number of stations to 32, all on or near the 500 m arc, would
~
approximately double the cost estimates given above for the various options considered. Thus a totally land based system containing 32 stations at 4500 m would be expected to cost $2.2-2.4 million. The cost of a double ring of stations can also be estimated. For 16 stations on both the 500 m and 1000 m arcs, the total costs estimated on the same basis would be S3.4-3.6 million. l ( . I e. 1 ( g 5-3
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i I. o REFERENCES
- 1. U.S. ' Nuclear Regulatory Commission, Regulatory Guide 1.97 Revision 2,
" Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident" (Decem-ber 1980).
The present study was begun shortly after issuance of Draft 2 of Fro-posed Revision 2 of Regulatory Guide 1.97 (June 4,1980) . This study addresses criteria in that draft. These included the following: Purpose "For estimating release rates of radioactive materials rt-leased during an accident from unidentified release paths (not covered by effluent monitors) - continuous readout capa-bility. (Approximately 16 to 20 locations - site dependent) ." Requirements
- 1. The ability to censure dose rates in the range 10 -6 F ?.: :c 10 R/hr.
- 2. Environmental Qualification per Regulatory Guide 1.2i UREJ-05SE ) .
, 3. A high reliability power source, non-Class 1-E, satter*; :sched when momentary interruption is not tolerable.
Periodic testing per Regulatory Guide 1.116.
~
4.
- 5. Detecters should respond te gamma radiation phot:nr >- f: n an. s.ergy range from 60 kev te 3 kev with an accuracy of 1 204 at any specific photon energy from 0.1 MeV to 3 MeV. Overall system accuracy should be within 1/2 decade over the entire range.
- 2. D. H. Slade (Ed.), Meteorology and Atomic Energy, USAEC Division of Technical Information, TIO-24190 (1961).
- 3. R. G. Canada, NSAC EPRI ORIGEN CODE Calculatien of the TMI-2 Fission Product Inventory, Technology for Energy Corpcration Report R-80-012 (May 1980).
- 4. Assessment of Offsite Ocses From the Three Mile Island Unit 2 Accident, TDR-TMI-ll6. e
- 5. H. Goldstein and J. E. Wilkins, Jr. , Calculati:ns of the Fenetratien of GLmma Rays, NYC-3075 f. une 1954) .
- 6. L. D. Gates, Jr. and C. Eisenhauer, Technical Analysis Report, AFSW7
-502A (January 1954).
R-1
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J APPDiDIX A Questionnaire Vendors Centacted By Mail Venders Contacted By Telephone (
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' QUESTIONNAIRE
([. , J INSTRUMENTATION AVAILABLE FOR APPLICATION
'ID A REMOTE MONITORING STATION SYSTEM I
- 1. Do you manufacture an instrument with a single detector capable of measuring the range from 10-6 R/hr to 10 R/nr? Yes No l' If the answer to question 1 is "No": Do you provide a' system i-2.
using multiple detectors capable of. measuring the range from 10-6 R/hr to 10 R/hr? Yes No
- 3. If the answer to question 1 is "No": Do you manufacture an instrument containing a single detector capable of measuring an upper limit of 10 R/hr? Yes No k*r.at is the icwer sensitivity of this device?
, j(,
- 4. Is this instrument presently commercially available?
Yes No J
- 5. Can this instrument be environmentally qualified (i.e. , non-s eisr.i c , IEEE 323,1974)1 Yer Nr l
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VENDORS CONTACTED BY MAIL I 1. Applied Physical Technology, In c . D. M. Walker
' ~
- 2. Atomic Products Corporation James Reiss
- 3. Can'berra, Industries, Inc. G. Leskowski
- 4. Catalytic Products International, Inc. E. Betz
- 5. Digital Data Dosimetry Inc. L. Lay
- 6. Dosimeter Corporation of America M. Srybnik
- 7. Eberline Instrument Corporation Eri c L. Geiger
- 8. EG&G Ortec Inc. John Haynes
- 9. Electrometer Corporation Al Zirkes i
- 10. Fluid Components, In c. M. M. McQueen
- 11. Foxboro Analytical Carol Tunick
- 12. General Atcmic Co. R. J. Bobby
- 13. The Marshrv Chemical Company Harry Cobble
- 14. Heal.th Physics Instruments, Inc. J. Handloser
- 15. Migh Voltage Engineering Corp. R. L. Malnati
- 16. Johnston Laboratories, Inc. L. Gevins
- 17. YJmmn Sciences Corporation Jim Lee
- 18. Kimmel, Inc. William L. Weiss
- 19. Kurz Instrunents, In c . Bob Steinberg
- 20. Leeds and Northurp H. A. Selko
- 21. Metecrology Research, In c . A. L. Severson
- 22. Murray and Trettel, In:. Jack Cchlen:
- 23. Naticnal Nuclear C r; oration H. Miller
- 24. Nuclear Equipment T:.cmical Corporation B. Backstroz
. fj 25. Nuclear Instrument Ccapany K. Gerrish
\ 26. Nuclear Measureme..ts Corporation Mr. Hil dentrand
/ 27. Nuclear Power Outfitters Earl E. Jacobson
- 28. NUS Corporation L. R. Love
- 29. Huelear Fesearch Corporation D. T. McIntyre
- 30. Far Systems Mike Ryder
- 31. Princeton Gamma-Tech, In c. R. J. Willi ams
- 32. Radiation Management Coqporation Thomas C'Malley
- 33. Radiation Monitoring Devices. In c. Elisa Redler
- 34. Re ute r-Stoke s , Inc. Annette Leckhout
- 35. SAI Technology Trancis Smith
- 36. Schmidt Instrument Campany A. C. Sch-idt
- 37. Sierra /Misco, In c. J. R. Ar.dre
- 38. . Technical Associates- J. R. Starr
- 39. Technology For Energy Corporation Bill Hartman
- 40. Air Monitor Corporation Ted Andre s
- 41. Teledyne Analytical Instruments Victor Black
; 42. TIRA Corporation Robert W. Felter.
, 43. Terradex Corporation J. Gingrich
- 44. Texas Instruments, Inc. J. T. Major
- 45. Victoreen Instrument, In c . Jeanee C. Moriarty
- 46. Westinghouse Instr ument Di vi sica R. Higgenbothan
- 47. 'Weston Co=ponents and Controls Division J. L. Walsh
- 48. Xetex. Inc. K. F. Sinclair
[ 49. General Electric
\
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VENDORS CONTACTED BY TELEPHONE ( - 1
- 1. Applied Physical Technology, Inc. D. M. Walker
- 2. Eberline Instrument Corporation Eric L. Geiger
- 3. EGGG Ortec Inc. John Haynes
- 4. General Atomic Company R. J. Robby
- 5. The Harshaw Chemical Company Harry Koppel
- 6. Kaman Sciences Corporation Jim lee
- 7. Kicmel, Inc. William L. Weiss
- 8. Nuclear Heasurements Corporation Mr. Hildenbrand
- 9. Par Systems Mike Ryder
- 10. Radiation Management Corporation Thomas Q'Malley
- 11. Reuter-Stokes, Inc. Annette Eeckhout
- 12. Technology For Energy Corporation Bill Hartman
- 13. Westinghouse Instrument Division R. Higgenbotham i.
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APPENDIX B Results Frc= Paran:etric Study t I f f I ie I
,e
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f . APPENDIX C LISTING OF PROGRAM PLtDIE A routine to calculate the response of detectors to , the activity in a plume released during an accident. J e f ( . t I ( i
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FORTRAN IV VO2.2-1 TUE OS-DCT -G1 15 17:20 PAGE 002
, LP: rDL1 : PLUME C PROGRAM PLUME C
,C ROUTINE TO CALCULATE THE RESPONSE OF DETECTORS PLACED IN A f C RING ABOUT A NUCLEAR POWER PLANT TO THE ACTIVITY IN A PLUME
\ C RELEASED DURING AN ACCIDENT.
C C THE CALCULATION MODEL ASSUMES THAT N DETECTORS ARE PLACED C AROUND THE PLANT ON A CIRCLE OF RADIUS, RAD, METERS. THE C RELEASE POINT IS AT THE CEN1ER OF THE CIRCLE AT A HEIGHT, C H, METERS. C C THE PLUME MOVES ALONG THE X AXIS OF THE COORDINATE SYSTFM C AWAY FROM THE PLANT. THIS AXIS IS AT AN ANGLE, PHI, TO C ANY DETECTOR. THE PLUME DIVERGES ACCORDING TO THE STABILITY C CLASS (NCLASS = A TO C) C THE CALCULATIONS ASSUME THAT THE WIND SPEED AND THE RELEASE C RATE ARE 1 UCI PER SECOND AND 1 METER PER SECOND,RESPECTIVEl.Y. C C THE CONCENTRATION AT ANY POINT WITHIN THE PLUME IS COMPUTED C THROUGH AN ALGORITHM CIVEN BY J.N.llAMAWI, OCTOBER 1971, C ..PLUMDOS., YANKEE ATOMIC ELECTRIC CO. C C CAMMA-RAY ATTENUATION COEFFICIENTS HAVE BEEN FIT TO A GUADRATIC C IN THE t.OG OF THE ENERGY AS: C .MU=EXP(-3 157'/-C 0265 LN(E) -0.03352 (LN(E))**2). TOTAL t' ANls .. .... .. .. .. - C XMU=EXP(-1 877C-1 6300 LN(E) 40.16800 (LN(E))**2) (400 KEV { C .
=EXP(-9 0220+1.1146 LN(E) -0.09175 (LN(E))**2) >400 KEV C ABSOHPTION COEFFICIENT......
( i ( BUILD-UP FACTORS HAVE DEEN 'lAKEN FROM: NUCLEAR SCIENCE AND ENGINEERING V.73 1980 97-107 CHILDEN,EISENHAVER, & SIMMONS THF V HAVE LEEN FI~ TO THE EGUATION-
~
EUF=1 +EXP ( 41* A2 LN(MFF ) +A3 (IN(MPP))**2) WHERE
=4 Als3 36411-0.24236 LN(E) -0.02765 (LN(E))**2 <POO KEV 4
2723 -0 907E4 LNIE ) +0 ODOO71(1.N(E))**2 >200 l'.EV A2=-2 47496+1.o1249 LNIE) -0.16468 (LN(E))**2 <200 MEV i = 0.99512+0 23052 LNfE) -0.'027981(LN(E))**2 >200 KEV
~
A3=-3.35150+1 45130 LN(E) -0.14678 (LN(E))**2 <200 KEV c
= 0.93913-0.46538 LN(E) +0.02786 (LN(E))**2 >200 KEV C INTEGRATION IS OVER Y , 7. Y, RESPECTIVE!Y, WITH TERMINATION C WHENEVER A PARTICULAR VALUE IS L ESS TI(AN MAX /2OOO.
C THE DIN S17ES FOR THE NUMERICAL INTEGRATION IN Y AND Z COORD. C ARE SET EGUAL TO SIGMA Y'/5 AND SIGMA Z /5,REOPECTIVELY. C THAT FOR X IS AN INPUT PARAMETER. C 0001 DIMENSION PYk i 4, - ) . P 7K ( 4,7 ), E ( 3) A11 (6), A22(6 ) A33(6), U( 3 )
- 1. RES f 30.
- 1. UA ( 6 } . FHI ( 9 ), CUE ! 20, 2 ) I CUF ( L'O, d* )
C COEFFICIENTS FOF. PL.UME DIFFUSION. 0007 DATA X1,Y2.V3/4 6052,6.9075,9.2103/ { C Y DIFFUSION COEST IC IEN15.
'3 DAT A P YF / ~2 135,0 4439.-O.01012,-0.OOJ682,D.7?7,0.9729,-O 01535
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- ORTRAN IV VO2.2-1 TUE 06-OCT-81 15:17:20 PACE 002
.LP:=DL1: PLUME 1,-0.000177,2.518,0.9402,-0.007994,-0.0004241,2.104,0.9334 2,-0.006178,-O.001078,1.609,0.9367,-0.00?51,-0.000731,1.386 ;
; 3,0.9629.-O.01615,0.001037,0.9933,3.9306,-0.003Y81,-0.001395/
C Z DIFFUSION COEFFICIENTS.
){ , DATA PZK/2.741,1.53,0.2OO9,-0.037,2.416,1.081,0.1612.-0.003L59 1,2.016,0.9428,-0.02774,-0.0002219,1.571,0.8259.-O.04077,0.00116 2,1.261,0.8142,-0.05579,0.0003467,0.8502,0.7709,-O 05184,0,00035 3 33,0.4383,0.7543,-0.0581,0.0006759/ -
C CAMMA-RAY ENERGIES. 2005 DATA E/80. 250.,1500./ ' C BUILD-UP FACTOR COEFFICIENTS.
>O06 DATA A11/3.16911,-0.29236,-0.02765,4.2723,-0.80324,0.028071/
2007 DATA A22/-2.47496,1.61249,-0.16469,0.99512,0.23052 -0.027981/ 300]. DATA A33/-3.3515,1.4513,-0.14678,1.93913,-0.46538,0.02786/ C CAMMA-RAY ATTENUATION COEFFICIENTS.
>009 DATA U/-3.1577,-0.0265,-0.03352/
C CAMMA-RAY ABSORPTION COEFFICIENTS. 1010 DATA UA/-1.877,-1.63,0.168,-9.022,1.1246.-0.091753/ C ANGLES TO BE CALCULATED. 1011 DATA PHI /O., 0.0491,0.0982,0.1963,0.3927,0.7854,1. 5708,2.3562,3.1426/ C C ............. INPUT DATA. . . ...... . .......... 5012 TYPE 2 6033 ACCEPT 20,INC
*014 DO 500 IC=1,INC '015 TYPE 5 !n%DIUS 0.: DETECTOR CIRCLE? '016 ACCEPT 10, CUE (IC,1)
Of TYPE 30 ! SLACK HEIGitT7
#(
s . ACCEPT 10 CUE (IC,2) TYPE 35 ! STADIL ITY CL ASS?
>020 ACCEPT 15,ACLASS 6021 IF( ACLASS. EG. ' A ') ICUE ( IC,1 ) =1 1023 IF(ACLASS.EG.'D')ICUE(IC.1)=3 025 IF( ACLASS. EO. 'C ' )ICUE ( IC,1 )= - >O27 IF ( ACL ASS. EG. 'D ') ICUE ( IC.1 )=4 >029 IF(ACLASS.EG.'E')ICUE(IC,1)=5 , '031 IF(ACLASS.EG.*F')ICUE(IC,1)=6 033 IF( ACLASS. EG. 'G ' )ICUE(IC,1 )=7 035 TYPE 40 ! XDIN LENGTH 036 ACCEPT 20,ICUE(IC.2) 037 500 CONTINUE C i 039 CALL ASSICN(6,'LP: ')
C C CALCULATION LOOP 039 DO 3000 IC=1,1NC 050 RAD = CUE (IC,1) 041 HAnCUE(IC.2) 042 NCLASS=ICUE(IC,1) ; 4043 NXBIN=ICUE(IC,2) C ...... . . . . . RELEASE HEIGHT L OOP. . . ... ........ 1044 IF(ICUE(IC,1).EG.1)ACLASS='A' IO( IF(ICUS(IC,1).EO DiACLASSr'B' 6D( IF(ICUEi1C,1).EG 3/ACLASS="C' C-3 f..u, w., :'.. ~f~ P2"."i.' 3I3"l2r?. r.%-.O.-'TJ'EU.T i.T ';'3I M *J.5 , o Q '~~ N .' '
T I
FORTRAN IV VO2 2-1 TUE 06-OCT -88 15: 17: 20 PAGE 003 o LP: =DL3 : PLUME , 0050 IF(ICUF(JC,1).EO.4)ACLASSa'D' 0052 IF(I'UttIC,1).EG.5)ACLASS='E' C i '54 IF(ICUE.(.C,1).EG.6)ACLASS='F' , IF(ICUE(IC,1).EO.7)ACLASS='G' [ 56
, $8 DO 2000 IHal,1 l
l 0059 H=HA . 0060 IF(IH.EG.2)H=O 0062 IF(IH.EG.3)H=100. C .............. DETECTOR LOOP................... 0064 DO 1900 IN=1,9
@O65 TAU = PHI (IN) 0066 RES(IN,1)-TAU C
C .............. ENERGY LOOP................. .... 0067 DO 1800 IEN=1,3 0068 RSPMAX=1E-15 ! MAXIMUM RESPONSE 0069 DRESP=O ! INTEGRAL VAL UE 0070 ,.XLNE=ALOG(E(IEN)) 0071 XMU=EXP(U(1)+XLNZ*(U(2)+XLNE*U(3))) ! ATTENUATION COEFFIC. 0072 I=1 ! LOWER ENERGY FIT 0073 IF(E(IEN).GT.2OO)I=4 !USE HIGHER ENERGY FIT 0075 Al=A11(I)+XLNE*(A11(1+1)+XLNE*A13(I+2)) ! BUILD-UP 0076 A2=A22(2)+XLNE*(A22(I+1)+XLNE*A22(I+2)) ! FACTOR 007' A3.-A33 ;)+XLNE*(A33(I+1)+XLNE*A33(1+2)) ! COEFFICIENTS. 007G !=1 0079 icC_ ?EN).OT 400)l=4 0081 X MU/ = = XP ( U A ( I I + X LNE * ( UA ( I +1 ) +XLNFe UA ( 1 +2 ) ) ) ! ABSORB. COEFF. I(,.02 i . D3 l'OO IX = 15,
... X L OOP. . .. . ....
15000+NXDIN/2, NX181N 0063 R=In O?O4 T. 5-sal *COSITAV) 006" .f- 0) . = O
;0i. - 1 ) =AL3Z. ( R .
0087 ? Y='-YK ' 1, NCLASM ) + ( X X-X 1 ) *Pv" (
- d, NCL. ASS ) * ( X X-X 1 ) * ( X X-X2 )
- PYM t 1 3, NCL. ASS ) + ( X X-11 ) * ( XX-X2) * ( X X-X3) t PYK (4. NCI ASS 5 0055 G Z=FIF ( 1, NCLASS )+( XX-X 1 ) *P ZK(2, NCLASti) + ( XX-X1 ) + ( AX-XP) rF7K -
) 3,NZLASS)+(XX-X1)*(XX-X2)*(XX-X3)*PZK(4,NCLASS) 0009 SIOM AY=E XP ( P r')
0090 SIGMAZ=EXP(PZ) 0091 NDELY=SIGMAN/.5 0092 NDELZ=SIGMAZ/5 0093 IF(NDELY.LT.5)NDELYm5 0095 IF(NDELZ.LT.2)NDELZ=2 C S .. ... . . . . . Z LOOP. ..... ....... . 0097 UO 1600 IZ=NDELI/2, 4GO+NDEl.Z /;!, NDELZ 005S Z=12 0099 D3 ISSO JJ=1,2 0100 IF(NEGZFL :45 0. AND. JJ. EO. 2)GO 10 1600 !DELOW CROUND 010 ~. IFiJJ EG.2.t.ND Z.GT.H)GO TO 1000 !GET NEC4FL 0104 . IF(JJ EG.2)2e-Z ( 106 NEGYFL=0 !RESE1 NdGATIVE Y FIAC 07 GZ =EXP (-( Z *I) / ( 2* ( SIGi1AZ* k2) ? ) +L- XP (-( ( 7 +2*H) * *2) / ( P* (
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( (SIGMAZ**27)) C f ')B Y t ilOP. . . . DO 1000 1 Y-rNDEl.Y /2, 15000*NDIlY/2,. N1ELY . 19 ( ~ YelY
,20 1 CONCe(1/(2*3.1416*SIGMAYeSICMA7))*CZ*
0111- EXP((-(Ye*2)/(2*(51GMAY**2))))
*0112 DO 1490 13=2.2 0134 3F(NECYFL IF (11. EG, 2)NF.
Yu- O. Y AND. II. EG. 2)CO 10 1500 0116
~1 DISTSG=(ABS (Z+H)*(Z+H) (PAD
- SIN (TAU))-Y)*(ADS (RAD
- SIN (TAU))-Y)+X*X+
0117 0119 DIST=SGRT(DISTSG) 0119 XMPFrDIST*XMU 0120 BUF=1+A1+ALOG(XMPF)*(A2+ALOC(XMPF)*A3) 10122 IF(BUF LT.1)DUFr1.0 0123 EYPO=EXP(-XMPF) D RESP =(CONC.*DUF* EXPO)/(12.566*DISTSG) TYPE SOOO.IX,1Y,1Z.IEN,IN. RESP,RSUMAX D5000 FORMAT (' , 5( 2X.14 5. 2E 12. 5) 0124 0 DRESP=DRESP+ RESP *NXDIN*NDELY*NDE!Z 012S CH'J CK SI ZL. . . . ...... 9127 IFinESP.G7 REPMAX)GO TO 1400 !!4FT NEW RSPMAX 0129 IF(PESP EO 01G0 TO 1440 IF(RESP !WE 'RE OU1 OF PLUME 0133 CE PSPMAr/2OOO.300 TO 1490 'OK..CO ON 0133 IF ': 17 EO. NDI I/2 A'C. IY FO.NDhtY/2)CD TO 1770 !DCNE 0135 IFtv LT OSOC TO 1;.c0 'ND MORF NFCA11VE Y Al THIS Z ft '7 IF i lY EG T4DE L f /2 iOC TO 1700 !Nr.XT X VA!UE CO TO 1590
'( 1940 !NEX1 2 YAt UE
,. v IF(!Y.EQ NDELY/21C- 10 1090 !NEXT 7 VALUE 141 CD TO 1700 J ' Af) NEGYFL*1 !NEXT X VALUE
'14D
!GET NO -Y VALUES FLAC CD TO 1500 d a~ 10F0 RSPMAX= RESP
! FINISH Y 10DP 1; 1990 CONTINUE !NEW MA):IMUM VALUE
- 1 4 '. 1 * > Ori CONTINUE !Y INTERNAL LOOP
.*iah !Y LOOP **C' CO TO 1590 l !'O NECZFL=1 0146 GO TO 1600 ! SET FOR NO Z VALUES DELOW GROUNO 'la- l 'WO CON 71NUE !NEXT 2 VALUE 0150 1600 CONTINUE !? INTERNAL LOOP 0101 1700 CONTINVE !? LOOP 0152 1790 !X tOUP 0353 10hO RES(IN.IEN+1)=DRESP*Fs1EN)rXMUA*J V/
CONTINUE 0154 1900 CONTINUE ! ENERGY LODP C !DETEClOR LOOP C . . 3195 WRITF(6 101) RAD,H PRIN7 HESUL10.. . .. . ...... .... 3156 WPITF(6,99)1C }157 0159 WRITEf6 105)ACLASS WRITE (6,98)NXBIN 0159 WRITE (6,104) 0160 WHITE (6 1021 Os DO 100,I=:.9 t( -
) :-
a ,w.m,a m -,g-~n.pe w ,4 & :w&-- - . ~ = ." 9-n:~ ~ ~ -, ~gy- ': . -*- n p.sg~' 7' '. 7 .sp37-. ;. .r'a- ~~
FORTRAN IV VO2.2-1 TUE 06-0CT -81 15:17:20 PAGE 005
. LP : =DL.I : PLUME 0162 C= (-1. 7918+ ( RES ( I,2 )-RES t I,3 ) ) +1.1394 * ( RES( I,3)-RES ( I, 4 ) ) ) /
1 (-5.9831) I 63 B=(RES(I,2)-RES(I,3)+11.2844*C)/(-1.1374) ( 24 A=RES(1,2)-4.382*D-19.2022*C d165 100 WRITE (6,103)I,(RES(I,J).J=1,4),A,D.C 0166 2000 CONTINUE ! RELEASE HEICHT IDOP CLOSE 0167 3000 CONTINUE ? CALCULATION LOOP CLOSE C 0168 CLOSE(UNIT =6) C
'0169 2 FORMAT ('5 NUMBER OF SAMPLES ? ')
0170 5 FORMAT ('fRADIUS OF DETECTOR CIRCLF (METERS)? ') 0171 10 FORMAT (F10.0) 0172 1D FORMAT (A4) 0173 20 FORMAT (12) 0174 30 FORMAT ('5 STACK OR RELEASE POINT HEIGHT (NETERS) ? ') 0175 35 FORMAT ('5 STABILITY CLASS CA-G3? ') ' 0176 40 FORMAT ('SX BIN LENGTH ') 0177 98 FORMAT (' ' 5X,'X BIN LENGTH e ',13) 0178 99 FORMAT ( 'O ',6X, 'CALCULAT7 DN NUMBER ' ,1P) 0179 101 FORMAT ('1',///,' '
,20X,'S U M M A R Y OF DETECTOR
' ,5X,' RADIUS OF 1R E S P O N S E T0 P L U M E',///,'
2 DETECTOR CIRCLE = ', F8. 2, ' METERS ', / , ' ' ,5X,'HEICHT OF RELEASE 3 POINT = ',F7.3,' METERS') 01FO 10? FOR M AT ( ' O * . .', ' ',6X.* DETECTOR',7X.'ANCLE',jeX,' ENERGY (KEV)'
- 1. /, ' t * ,19 F . ' ( R ADI ANS 1 ' , 6X , '80 ', I L.X ' 21 O * ,16X, ' 1500 ' ,10X, ' A ',
PSX,'B',8X,'C',/,* '
,80(' '),///)
81 103 FORMAT (' ' . EX,12 T21. F7. 3, T30. 3( I PL- 12. 5, 7X ) ,3 ( 2X,1PE 12. 5 ) ) k 22 104 FORMA 1 ( 'O ',20X, ' DETECTOR 'itESPONSES TO PLUME (R/HR)/(CI/SEC)',//) T,163 IOS FORMATi *
,5X, '51 ADILITY CLASS 6 ',A4) t D j A FIJII
\
l 1 i I
.]
C-6 , UE-Y'T. i.C*2CC'.?:T:1M e c t y -a ;r,u :, .;., - n ,
. 7."C a 3,3. r 3- - -..-
_3
I i t f
~l
.j J
0 y e ! e APPENDIX D i { [ . ESTIMATED NOBLE GAS RELEASE RATES OF TMI-2 FROM TDR-TMI-116 l-s
- ?
b f , i. I-1 '
&'i.
1 , 1 1 t i. h, i { a e
- v. g 4
9 0 ( : j.
-I a
e it e 0 f e a t i j l e , f-n _) - D-1 M *l F 7 *, ** y M
- T O3.@- et h 4 -% %
. ..
- s . .*J'i- 7'?%Rou %-
O b N 't n ri - 4 ' *' s'W 4 8
* ^ d '4
. -.. ~. .
.% 5'e . . , .- . . . . -- -
- ., -..---.-..;------ - -~ -- - - '
ESTIMATED NOBIE GAS RELEASE RATES OF TMI-2 FROM TDR-TMI-116 Average Release Rate (UCi/s) ** 88 Time 1333 , 133my , 135x , 135mx , Kr 400 500 600 0 0 0 0 0 700 800 900 0.98+05* 0.94+03 0.10+05 0.40+04 0.55+04 1000 1100 0.58+07 0.75+05 0.90+06 0.26+06 0.27+06 1200 1300 0.74+07 0.94+05 0.12+07 0.26+06 0.21+06 1 l 1400 ' I 1500 0.79+08 0.95+06 0.14+08 0.22+07 0.13+ 07 1600 1700 1800 0.93+08 0.7f+06 0.12+08 0.13+0' C. F +0f 1900 0.47+08 0.11+07 0.16+06 0.16+07 0. 5 7+ 06 2000 0.16+08 0. 3S+06 0.54+07 0.49+06 0.15-O' l 2100 0.52+08 0.12+07 0.17+08 0.14+07 C . 2':- 26 2200 0.11+09 0.26+07 0.34+08 0.27+07 0.E:-06 l 2300 1 2400 0.18+09 0.25+08 0.25+08 0.18+07 0.32+06 0100 0200 0.32+08 0.38+C6 0.43+07 0.25+06 0.32+0E
- 0.98+05 means 0.98x10
** Release rates are averages for the corresponding hour and previous hours of assumed constant R(t) .
t
/ D-2 4
L2?:Y C ':' n w . a x + ~ W a:c.wa..:;;;.Ww. .a;;7 ~r . .:n,; , ~ z -~ ;- ~
t
~
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