ML20117M999

RTM-90 Response Technical Manual Workbook

ML20117M999
Person / Time
Issue date:	06/30/1990
From:	Brown G, Jackson K, Mckenna T NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
To:
References
NUREG-BR-0150, NUREG-BR-0150-V02, NUREG-BR-150, NUREG-BR-150-V2, NUDOCS 9609180282
Download: ML20117M999 (132)

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~ s p* "c, ui h s,,.....f NUREG/BR-0150, Vol. 2 U.S. Nuclear Regulatory Commission RTM 90 Response Technical Manual Workbook i June 1990 T McKenna, K. Jackson, G. Brown, W. Reckley, R. Hogan, J. Jolicoeur Division of Operational Data Office for Analysis and Evaluation of Operational Data

i PREFACE This workbook is to be used by students of the Response Technical Manual (RTM-90), NUREG/BR-0150, Volume 1. Not all of the steps, or the sequence of steps in the manual may be self-explanatory. Study and use of the manual should be supplemented by class room instruction or attention to the video tapes. The student should also read NUREG-1210, Volumes 1-5 as part of the response training prerequisites before attending the training course or viewing the tapes. Please direct any questions concerning this workbook to the Incident Response

Branch, Office for Analysis and Evaluation af Operational Data, Division of Operational Assessment.

i NOTE: This workbook replaces Volume 6, Rev. 1, NUREG/BR-0132, dated August 1989. i l i i

l ~ TABLE OF CONTENTS PAGE l QUESTIONS: Overview. 1 Section A - Classification. 3 Section B - Core & Containment 14 Section C - Rx Accident Consequence 25 Section D - RASCAL. 34 Section E - IRDAM 38 Section F - 100 Mile or Rain RX Dose. 39 Section G - ARAC 44 Section H - Dose Hand Calculations. 53 Section I - Plume PAR 59 Section J - Intermediate PAR. 79 Section K - Ingestion PAR 83 Section L - Emergency Work PAR. None Section M - KI and Thyroid Monitoring 86 Section N - NRC TLD System. None Section 0 - Medical Assistance 90 Section P - Rad. Field Monitoring 91 Section Q - SI Units. 92 Section R - Radiation Chart None Section S - Half Life None ANSWERS: Overview. 94 Section A - Classification. 95 Section B - Core & Containment. 102 Section C - Rx Accident Consequence 104 Section D - RASCAL. 109 Section E - IRDAM 110 Section F - 100 Mile or Rain Rx Dose 111 Section G - ARAC 113 Section H - Dose Hand Calculations. 114 Section I - Plume PAR 117 Section J - Intermediate PAR. 121 Section K - Ingestion PAR 123 Section L - Emergency Work PAR. None Section M - KI and Thyroid Monitoring 125 Section N - NRC TLD System. None Section 0 - Medical Assistance 126 Section P - Rad. Field Monitoring 127 Section Q - SI Units. 128 Section R - Radiation Chart None Section S - Half Life None i iii

I 1 l i _ QUESTIONS 6/90 1 l l Overview: 1. Q. Find the following in the " Response Technical Manual." Answers should state the page number. a. EPA recommended protective action to reduce whole body l and thyroid dose from exposure to a gaseous plume l l b. Guidance on the NRC assessment of severe reactor accident public protective actions c. Water that must be boiled to cool a core d. Dose factor ground shine e. Reactor accident dose projections at long distances or with rain f. Relationship between plant accident conditions and possible offsite doses and consequence i g. FDA protective action guidance (PAG) concerning possible l ingestion of contaminating food h. EPA recovery / reentry dose conversion factors for 1 year of gamma exposure 1. A quick reference for use in assessment of an accident classification l j. Population around the site k. A chart that can be used to explain the danger from various levels of radiation to the public 1. State and local decision makers m. The effectiveness of KI for thyroid blocking n. Recommendation on contamination screening levels o. Derived preventive response levels for drinking water (5 day) p. Mark III containment monitor response to a severe accident q. Immersion dose at 1 mile from release of 1 Ci of Cs-134 l l

2 QUESTIONS 6/90 1. continued: r. Fuel cycle / material accident classification s. Coolant concentration for various levels of core damage

QUESTIONS 6/90 3 Section A 1. Q. Why is assessment of the event classification most important to assuring the appropriate protective actions will be taken? A. 1 ^l 1 4 1 1 4 I N

-..m. an 4 QUESTIONS 6/90 l Section A i 2. Q. Licensee reports Steam Generator Tube Rupture (SGTR) of a few .hundred gpm primary to secondary. What level of emergency should be declared on this information alone? A. i t I j l l L f L t L ? i f f ( r i

_ QUESTIONS 6/90 5 Section A 3. Q. At Crystal River Unit 3 on February 26, 1980, while the reactor was operating at full power, a transient resulted from a trip and one of two non-nuclear instrumentation (NNI) dc power supplies caused by a short circuit in an NNI component. The NNI power supply trip resulted in erroneous readings on control board indicators served by the power supply. The trip concurrently caused command signals opening the pressurizer power-operated relief valve (PORV) and the pressurizer spray valve. Erroneous signals were also sent to the integrated control system and resulted in control rod withdrawal (which in turn caused increased reactor power production) and simultaneous reduction in feedwater flow to the steam generators, decreasing the reactor heat removal capacity. This caused reactor and turbine trips, automatic initiation of high-pressure injection (HPI), rupture of the reactor coolant tank rupture

disk, and pressurization of the containment building to approximately 4 psig.

Peak radiation levels of approximately to 80 R/hr (see question B-5) occurred in the containment building. Approximately 4100 gal of primary reactor coolant water were discharged through the pressurizer and code safety valves into the containment building sump before natural circulation cool-down was established. Classify the various phases of this event. A.

6 QUESTIONS 6/90 Section A 4. Q. An earthquake of unknown magnitude occurred in the vicinity of an operating commercial nuclear power plant in the southeastern part of the United States. Earthquake tremors were felt in the control room and throughout the plant. Control room operators manually tripped the reactor as a precaution. Licensee is presently inspecting the plant for damage. What emergency classification should the licensee declare for this event? A.

1 QUESTIONS 6/90 7 Section A 5. Q. A plant is in an ALERT emergency class following a reactor shutdown that occurred an 1-1/2 hours ago. The licensee is j asked by the NRC over the ENS for an update of their ability to maintain reactor coolant inventory and ability to remove decay heat. They report: 1. They are cooling the core by feed and bleed through a reactor coolant system leak. 2. They have been unable to maintain more than 100 gpm makeup for the past 1/2 hour. 3. There has been a rapid increase of the incore thermocouple temperature in the past 15 minutes to in some cases off scale (1000* F). When asked if they anticipate a change in classification, the licensee stated that they are reviewing the classification in accordance with their classification procedure on " Core Fuel Damage Classification" (attached) and cannot change the classification u..til they obtain a coolant sample and analysis within the next hour. What is your assessment of this response? What is the proper classification for this event? A.

8 QUESTIONS 6/90 CHART FOR SECTION A, QUESTION #5 CORE FUEL DAMAGE CLASSIFICATION EMERGENCY CONDITION INDICATIONS CLASS. Core damage with in-1. Confirmed primary coolant Site Area adequate core cooling sample results indicate: Emergency determined A. DOSE EQUIVALENT I-131 > T.S. OR B. >100%/E yCi/ gram specific activity AND 2. The incore thermocouple indicate superheated con-ditions in the core Core damage with 1. Confirmed primary coolant General other plant conditions sample results indicate Emergency ) makine a release of >300 pCi/ gram DOSE large amounts of EQUIVALENT I-131 radioactivity AND possible 2. Incore thermocouple temperatures >700* F AND l 3. Containment radiation level is > 10' R/hr i

4 QUESTIONS 6/90 9 ^ Section A 6. Q. At 4 :30 p.m. on Friday, A BWR-4, MK-I licensee reported to the NRC Headquarters Operations Center that they had been in full power operation for 11 months when at 4:15 p.m., a loss of condenser vacuum resulted in a reactor scram and a turbine trip. Safety / relief valves (SRVs) had opened to relieve reactor pressure after the main turbine trip. The MSIVs closed and the HPCI system started. SRVs are now recycling to remove decay heat. The Shift Supervisor had immediately informed the Plant Manger of the scram who said to go to cold shutdown condition for refueling since this had been planned for Monday anyway. A control room operator had started the RHR system to cool the suppression pocl but the RHR pump motors kept tripping off. What is the proper classification for this event? A.

10 QUESTIONS 6/90 section A ~ 7. Q. This is a continuation of Section A, Question #6. At 9:30 p.m. the same evening, the licensee reported to the NRC that they still did not have the RHR pumps. running, the temperature of the pressure suppression pool was continuing to rise, and they had no means available to cool the suppression pool water. What is your evaluation of this situation? What is the proper classification for this event? A.

~ QUESTIONS 6/90 11 Section A 8. Q. A BWR-4 operating at 100% power for six weeks with slightly leaking fuel, incurred a rupture of the 26 inch diameter "B" main steam line outside primary containment between the outboard main steam isolation valve (MSIV) and the containment. (See Attached Figure). All MSIVs auto closed due to high main steam line flow except the "B" line in board MSIV which failed full open. The reactor auto scrammed due to the closure of the MSIVs. What is your evaluation of this situation? What is the proper classification for this event? A. m

U SAFETY / RELIEF ^ INBOARD VALVE ^' ^ MAIN STEAM A ^' ISOLATION N STEAM VALVE y s i, FLOW c RESTRICTOR [ h M v 4 t 3 1 . --1 m l f i 6 / w 26 INCH DlAMETER M r MAIN STEAM LINE O I s t VESSEL ~ REACTOR o TO y PRESSURE SUPPRESSION o CONTA N NT (DRYWELL) d O E MAIN STEAM LINE BREAK OUTSIDE CONTAINMENT o t 3 i a i' 5 R 8 '. s

... _ _ _. _ _. _. _. _ _. _. _ -.. _ - _ _ _ _.. _ _ _... _. ~. - _ _.. - _ _ QUESTIONS 6/90 13 J Section A i 9. Q. The NRC Operations Center receives a call at 0200 in regard l-to a fire at a UF conversion facility. The fire is'in the 6 vicinity of UF cylinders and has been classified as an Alert 6 by the licensee. Licensee personnel. int he area of the fire. have been evacuated and the fire brigade is responding. By 0215, communications have been established with the response personnel from the HRC and state agencies. The licensee reports that the fire is heating a cylinder and slight smell i of chlorine indicates that a small leak may have developed. Monitors near the site boundary do not indicate a significant offsite release is occurring. After an additional 30 minutes, the fire has been extinguished, the cylinder is being cooled by the fire hoses, and preparations are being made for i transfer of the damage cylinder contents to another cylinder. During the heating up of the cylinder, should this event be classified as a Site Area Emergency? A. 1

14 QUESTIONS 6/90 Section B ~ 1. Q. The core of a PWR reactor has been uncovered with no flow for 1/2 hour. What is your assessment of the condition of the core? A. i l

QUESTIONS 6/90 15 Section B 2. Q. A PWR reactor core has been recovered with water after being uncovered for over an hour. There have been several indications (e.g., very high radiation in the containment) that confirm that the core was severely damaged. Is threat of vessel melt through over? A.

~ ~ l' QUESTIONS 6/90 Section B 3. Q. A PWR 1000 MW(e) has been shutdown for 2 hours. They are cooling the reactor core by feeding (injecting) 100 gpm into the primary system and bleeding through a stuck open PORV into the containment. Is this sufficient to protect the core for an extended period? A. t + b 4 i i

QUESTIONS 6/90 17 Section B 4. Q. The maximum radiation level in the TMI contamination resulting from the accident was between 7000 R/hr and 60,000 R/hr. The containment sprays were operational. What level of core damage does this indicate? Is this a good indicator of the maximum level of core damage? Can you explain the wide range in the estimated maximum reading? A.

18 QUESTIONS 6/90 _' Section B 5. Q. This figure charts the reactor building radiation monitor response for the first thirty minutes following the Crystal River Unit 3 event. Crystal River is a PWR. There were no sprays. All of the monitors are located in the containment. 105 her<em tur duiiding '1 Dome Monitor 104 J 103 i E m 5 Fuel Handling Bridge [ nc E 102 Reactor Building Incore Instrumentation Area Personnel Hatch 1 O 30 Minutes Explain the shape of the response curves for the 4 instruments and what level of core damage (if any) is indicated. ( A. j i ( (

QUESTIONS 6/90 19 Section B 6. Q. The PWR core thermocouple are reading in the range of 630 degree F and the primary system is at 1500 psi. What does this indicate? A. C

l 20 QUESTIONS 6/90 Section B 7. Q. PWR coolant concentration has been analyzed to be about 10,000

1. Ci/g I-131.

What is the status of the core? A. i /

QUESTIONS 6/90 21 Section B 8. Q. There is 20% H concentration in a Mark I, BWR containment. 2 What does this indicate? A. I

22 QUESTIONS 6/90 Section B 9. Q. The TMI core was about 1/3 melted and another 1/3 lost all cladding and was severely heated. These conditions should drive out most of the more volatile fission products (I, Cs). A coolant sample was taken about 48 hours after core damage. The results are shown on page B-14, "PWR Baseline Coolant Concentration." Can you explain the results relative to baseline concentrations shown in the case. Why are there inconsistencies? A. i l 1

QUESTIONS 6/90 23 ~ Section B 10. Q. A PWR reactor has been shutdown for an hour. The pressure is at about 1000 psig and the temperature is 503*F. They have lost feedwater and are injecting about 250 gpm and bleeding through the PORV. Is the core going to be protected under this condition? A.

24 QUESTIO: '90 Section B 11. Q. This figure charts the reactor building radiation monitor response for the first five hours following the Crystal River Unit 3 event. No sprays were on and all of the monitors are inside this PWR containment. 105 Reactor building dome monitor 104 d Reactor building incore instrure^ation area 103 m Ea 102 f 2 ( Personnel hatch t Fuel handling bridge to 1 O 1 2 3 4 5 Time (hours) l Explain the decrease in dome monitor after half an hour and in incore instrumentation area after an hour. A.

QUESTIONS 6/90 25 Section C 1. Q. For the following accident conditions estimate the whole body and thyroid dose at 1 mile. a. PWR Large Dry Containment Core Melt release to the atmosphere by containment leakage at design rate sprays on release starts 2 hours after release of the fission products from the core into the containment from reactor coolant system b. PWR - Stem Generator Tube Rupture once through Steam Generator - OTSG (B&W) 1 normal coolant concentration (confirmed by a sample the day before the accident) rupture of 1 tube at full pressure c. BWR - Leakage out of Wet Well (through suppression pool) gap release to containment { subcooled suppression pool holdup of 2 hours in containment before release catastrophic containment failure d. A by-pass of the containment by failure of the low /high pressure interface the core has been uncovered for about 40 minutes not a filtered release release rate 100% per day A.

26 QUESTIONS 6/90 Section C 2. Q. A BWR has experienced core damage and the containment monitor has increased to 10,000 R/hr in the past 1/2 hour. The containment sprays are not operational. The containment pressure is above design and increasing. The TSC director is considering venting and has ask the NRC for a fast comparison of the possible consequence if the containment is vented through the subcooled suppression pool now vs allowing containment failure in a few hours. Provide your analysis. A. i \\ i 1 J

QUESTIONS 6/90 27 Section C 3. Q. Describe what is a " gap" release and why it important? A.

QUESTIONS 6/90 28 Section C 4. Q. What is a typical I-131 core inventory for a 1000 MW(e) LWR (PWR or BWR) at shutdown? A.

QUESTIONS 6/90 29 Section C 5. Q. Estimate the dose at 1 mile from release of 100% of the core " gap" 6 hours following shutdown. A. j 4 4 5 1 1 1 4 b

~ QUESTIONS 6/90 30 Section c 6. Q. For the following conditions estimate the whole body dose at 1 mile: PWR Large Dry Containment Reactor shutdown for 1 month Core melt Release to atmosphere at design leakage rates No spray Release starts 2 hours after release of the fission products from the core into the containment from the primary system. A. l 1

\\ l 4 QUESTIONS 6/90 31 Section C 7. Q. A 1000 MW(e) PWR, operating for a year, is located 1-1/2 miles from a beach in Northeast United States. It is early Sunday-morning in July - a beautiful day. The only other information you have been given is that core is melting and early containment failure predicted. You are to estimate offsite consequences on the beach. A.

32 QUESTIONS 6/90 Section C 8. Q. Fill out the table below using Section C, " Reactor Accident Consequence Assessment From Plant Condition" for a PWR Large Dry Containment Leakage case" (Red Tab). Whole Body Dose at 1 mile 100%/hr release rate Core Condition .5 hr hold-up .5 hr hold-up 2/12 hr hold-up Spray-on Spray-off Spray-off Gap Grain Boundary Melt Examine the results and discuss what accident conditions are required to produce early deaths offsite. A.

~. QUESTIONC 6/90 33 Section C 9. Q. Assume a loss of water from a spent fuel pool at a PWR. The PWP fuel pool contains 5 batches (5 reloads). The last reload was several months ago. What is your estimate of the possible consequences if the pool is drained? A. I

34 QUESTIONS 6/90 Section D 1. Q. A RASCAL dose projection indicates a dose of about 10 rem at 2 miles from the plant while the licensee model projects 40 rem. What is the significance of this difference? A. 1 l l

QUESTIONS 6/90 35 Section D 2. Q. A dose model projects a dose of about i rem at the site boundary NW of the plant. A team is sent to that location and reports a single reading of 15 rem. What is the significance of this reading? Is the model wrong? A. 1

= - _ -...... 0 36 QUESTIONS 6/90, Section D ~ 3. Q. What are the sources of uncertainty in dose projections and what are the possible ranges? What are the uncertainties if the release rate is known? A. ) I l I l I i )

s QUESTIONS 6/90 37 Section D 4. Q. What RASCAL dose calculations are used to gain insights into early health effects and for comparison with EPA PAGS. A.

s s 38 QUESTIONS 6/90 Section E 1. Q. State why IRDAM should not be used to assess severe accident consequences. A. l

\\ l _ QUESTIONS 6/90 39 4 Section F j 1. Q. A PWR has suffered a core melt, early major containment leakage with the sprays operational. Quickly estimate the 24 hour dose for: Total whole body j Whole body from ground shine Whole body from cloud shine 1 5 at 50 miles from the site. The winds are in a constant I direction, it is not raining. 3 ,i A. 1 i i l 3 ?i I i

40 QUESTIONS 6/90 Section F 2. Q. What pathway appears to be most important source of dose from the plume resulting from a severe reactor accident. Use PWR

1. 1A (ground level release) and PWR 1B (highly elevated), no rain (pages F-4 and F-5) to illustrate your conclusion.

A. i i

QUESTIONS 6/90 41 Section F 3. Q. Explain the difference between the PWR #1A (page F-4) and the

1. 1B (page F-5), Section F, "Long Distance Reactor Accident Projection (WASH-1400)".

A.

42 QUESTIONS 6/90 Section F 4. Q. What insight does the answer to question F-3 provide about the peeformance of dose / consequence assessments for severe reactor accidents? A.

l QUESTIONS 6/90 43 Section F 5. Q. Compare the consequences of a major release with and without rain. A core melt accident with total early containment failura could (under worst conditions) be similar to a PWR 1A (ground level release) or - a PWR 1B (elevated release). Complete the Unarts below using the Figures on pages F-4, F-14, F-5 and F-15. Discuss the impact of rain on possible offsite doses. A. 24 Hour Whole Body Dose No Rain Rain PWR 1A (F-4) PWR 1B (F-5) PWR 1A (F-14) PWR 1B (F 15) f (Ground) (Elevated) (Ground) (Elevated) 1 Mile 2 Miles 5 Miles 10 Miles 20 Miles

44 QUESTIONS 6/90-, Section G

1..

Q. At 12:00 a.m. EST (9:00 a.m. on the West Coast) you contact LLNL to have them provide an estimate of possible off-site doses using ARAC. You provide all the pertinent information that is presently available. When can you reasonably expect to get a response? A. f f i t 2 A

~ QUESTIONS 6/90 45 Section G 2. Q. What are the limitations of the dose projections made by ARAC? A. i h

46 QUESTIONS 6/90 Section G 3. Q. What is the maximum dose projected for the infant thyroid on the ARAC result shown in Figure for Section G, Question #3. A.

O c N

3 o

Z tn V V V V V V V) V V Refined ARAC CoIculotion (H+18) Generated: 153UN89 1930 2 N l CAYUTA CREEK < Remarks: 1-131. 1-133 & I-135 o i Inhalation Child Thyroid Dose Ac tuo i 5 Hour Release Amounis <. Rainout included g '\\ ( Integrated: c 14JUN89 2200 2 to 15JUN89 1000 2 5 Hateriol: IODINES at 1.5 m y ? x Exposure Action Leveis: Q 0 jf \\g} Con tour Unlues (in Rem ) s g s cn 4 [' {>CxC>l > 2. 5E+ 01 to ) --h 17C area Covers 58.49 sq km O i> j 5.0E+00 o Z 17 Area Covers 114.11 sq km NY CHENUNG RI I'//////)) 1.0E+00 ,0 l 5 j(Area Covers 146.13 sq km g / k% % %]> 1.OE-01 g p Area Covers 171.14 sq km g g 1 <l l l l l l l ll) 1.OE-03 s s g 1) Area Covers 195.12 sq km O = V b ENS Source Lo c o t, i o n,,: j( V V Lot: 41*57 00 N I 10000.O Ft Lon: 76*3l'00" u ij 370.0 ~ ~~ ~~ O

i \\ 48 QUESTIONS 6/90 Section G 4. Q. The attached ARAC plots, see Figure on next page, show " pockets" of dose while other plots show similar " pockets" of deposition. How would you explain these (assuming that you believe them)? A. i i i l I i T i b t I I l r i r I i f b I r -.,---,---.~.-..--.,.~~.-,..n.-

g C$, d 0 - -- y y -- gr V V V-Vi' V V InitioI ARAC CoIcuIo1 ion (H+6) CAYUTA CREEK Effective Hhole Body 1.0 Ci/s for 10 Hours 1.0 E-04 [ < Rainout Coeffitient = Integroted: Q .N I 4 14JUN89 2200 2 to 15JUN89 0800 2 N %h\\ \\ rio t e r i a l : KR-88 of 1.5 m g 'hN Contour Values ( in Rem ) 'f Exposure Actson Leueis: O g ) \\ .I ' '\\ 17C ')]> 2.7E-02 \\ Area Covers 3.23 sq km Q \\s ~ 2/ ' \\ j _ _j ) 2.7E-03 g \\ s \\ 17 Area Covers 17.10 sq km z p NY CHEMUNG R1 V/////A > 2. 7E-04 o \\y Area Covers 49.65 sq km i / h%%\\l) 2.7E-05 s i Area Covers 109.40 sq km y pg __) <l l l l l l l ll) 2.7E-06 g Area Covers 156.75 sq km g T ENS Source Lo c o t, i o n,: V V Lot: 41* 57 00 N f 10000.0 ft Lon: 76"31'00" H ... 3gg,g _ 3jg, g. __L [ ,l i l

) QUESTIONS 6/90, Section G 5 Q. What is the basic difference between the ARAC results shown in Figure Sa and Figure 5b attached? A.

o C r Mme H Oz C's g v v V-v= v v v v v g qg g O Ge n e r a t e d;_14 J Lit 4 8 2dQ3 0 Z h < Remarks: Al1 N uc l ide s Gr oun d Exp. Tot. Hody Dose Rate sO 3 Ac t ua 1 5 flo u r Heiease Amou nt s g ' Ra i no u t ineluded N-g- To t al Deposition: 4 N 's . 14 4 M!!8 9_2 20 0 Z.t g.15 Jy H 89.1 Q 0 0, z j 6 .Y Y. N Material: Alete at 0al y 6 / Ex p os u re Action Levels: O i o \\ \\ Contour values (in u Rem /h r ) M M > 1.0Et06 M 17 0 Area Cove rs 3.43 sq km O8 i l 1.0E+05 s 0 \\ 17 Ar e a Covers 28.29 sq km 7 ny gg . j [//////M >

1. OE + 04 3 ( Area Cove rs 128.62 sq km 4

</ / \\ Y 6 - ^ b k M > 1.OE*O3 g c D / b ) Area Covers 179.38 sq km C [{ ] > 1.0E*02 PA 0 d ~ Area Covers 219.13 sq km g O 0 2: =m AT IENS Source Lo ca t i o n : 0 V \\/ Lat : 41 57'00' N 10000.0 ft Lon: 76 31'00" W (. l 360.0 370.0 m H ./

m M V ~ ~~- V V" V V ~V-v v Re [ jpe d _b R AC C a 1 e u l_ a t_ i o n ( H _+ 1 8 ) _ Gen er_a t e si: 14 J UN BL1010 - Z \\ j Re m ar k s : Al1 N uc l id e s Ground Exp. Tot. Body Dose Rate H h fl Act ua 1 5 ilo u r Re 1 ea se Amounts Ra i no u t ineluded Tot al Depos it i on : 4 l[ 14 J UN 8 9 2200_Z_t o _1 S Jy N 8 2__0 4 0_.0 _ Z y 6;, Mat er i al _ A LL_ _at o o_r y m e, [ o g Exposure Action Levels: O \\ Co n t o u r Values (in uRem /h r ) 17 0 Area Covers 3.43 sq km o $.3 ,L_-. 1.0E+05 Area cove rs 23.51 sq km A 17 } l-f////// >

1. 0E + 04 t4 Y

\\m l< Area covers 84.19 sq km 4 _ M > 1.OE+03 g 6 - - - - - ~ ~ ~ ~ ~ Area Covers 152.38 sq km C 5 v ))> 1.OE+02 k PA 0 j Ar e a Covers 185.95 sq km o 0 ENS m T I Source I,o ca t lo n : V V Lat: 41 57'00" N O h 10000.0 ft Lon: 76 31'00" W j __ L a H 360.0 370.0 O Z CD m \\ s

• e

,O i-m m a. - ---+= ems a%9 rmw e.m*w-w w ev=F-e'=F*

l 4 QUESTIONS 6/90 55 i Section H l 3. Q. You need a quick estimate of the effects of a 50-Ci release I of Cs'37? Which organ is affected the most? What are the l acute health -ffects? A. \\ l i l I l L r I l l t (

~ a QUESTIONS 6/90 56 Section H 4. Q. You have nagging doubts about the validity of a surface 2 contamination sample report of 3000 pCi/100 cm taken one hour af ter a 1-Ci release of Co". The sample was taken 1 mile from the reactor. While you await further field measurements, can you use your assessment tools to estimate whether the sample results are believable? A. t l .i

l l QUESTIONS 6/90 53 Section H 1. Q. A field measurement for I-131 of 1 x 10 5 Ci/m was taken at 1 3 mile from the reactor. The meteorology includes a heavy cloud cover with no precipitation. The licensee has made no recommendation yet, pending results of a computer calculation. Based on this information and assuming a 2 hour exposure time, what is your recommendation for residents within 2 miles of the reactor? Note: EPA PAGs: for protective actions: 1-5 Rem Whole Body (red bone marrow) 5-25 Thyroid A. l l

54 QUESTIONS 6/90 Section H ~ 2. Q. A worker's dosimeter indicated that he received 150 R while taking part in a search and rescue operation over a 2-hour period. Preliminary results from a whole body count indicated minimal internal activity. His supervisor told him to take the day off and rest but, he is extremely concerned about his immediate health and has turned to the NRC for guidance. What is your assessment? A. \\

) QUESTIONS 6/90 57 Section H 5. Q. The release of Co" has just started at the rate of 50 ci/sec. l It is not known when the release can be stopped. Wind speed is swirling at about 2 m/sec, and there is a heavy overcast. It appears that a downpour could come at any minute. The nearest houses are 1 mile from the site. There is only one route to evacuate for the affected sector which is directly in the path of the plume. The current recommendation is for administering KI and sheltering the population within a 2-mile radius and evacuating from there out to 5 miles. Do you concur? A. l \\

QUESTIONS 6/90 58 Section H ~ 6. Q. You have responded to an accident to find that a release to 2a the atmosphere of Am has been underway for the past hour at the rate of 1.4 x 10~3 Ci/sec It appears that the release can be stopped in a few minutes. It is a warm, sunny Spring day without a cloud in the sky. The wind is a little breezy, you can feel the wind in your face and see leaves rustling on the trees. You are advised of an elderly couple who live in an old one-story frame house about 2 miles from the site. They i were advised by the police to evacuate, but the husband i informed them that his wife was ill and it would be very detrimental to move her. The police have asked for your recommendation. A. j /

~. QUESTIONS 6/90 59 Section I 1. Q. What are the objectives, in order of priority, for taking protective actions in response to a severe accident? A. J d s l 'N

60 QUESTIONS 6/90 .~ l Section I 2. Q. What reactor accident conditions would be required for a release to result in early deaths or health effects offsite? ? A. I I 1 i i

... _ -........ - -... -... - ~ 4 QUESTIONS 6/90 61 f Section I 3. Q. When must protective actions be taken in the event of a severe reactor accident to reduce risk substantially? A. i i

~ QUESTIONS 6/90 62 Section I 4. Q. List the major containment failure mechanisms and indicate those that can be forecasted with accuracy. A.

i QUESTIONS 6/90 63 i Section I 5. Q. Can shelter in a frame house prevent early health effects close to a severe reactor accident? l l A. l I l l \\ I

64 QUESTIONS 6/90 ~ section I 6. Q. State some of the reasons why, for a severe reactor accident, public protective action should be based on core conditions and not projected doses? A. 1 i t 1 b k )

QUESTIONS 6/90 65 Section I 7. Q. To what distance around the reactor is the public at greatest risk in the event of a core damage accident? A.

h ~ 66 QUESTIONS 6/90 section I 8. Q. Under what conditions should protective actions only be taken in the projected downwind direction given a core damage accident? A. i I l l l i

~ ', QUESTIONS 6/90 67 Section I 9. Q. If there is no core damage or the facility is not a reactor, what is the basis for public protective actions? A. ) i ) i l 1 l 1

68 QUESTIONS 6/90,' Section I 10. Q. There are direct indicators of core damage (thermocouple and containment monitor reading). The plant has declared a j General Emergency. The containment pressure is increasing and 1 if this trend continues the containment pressure is expected to reach design pressure in about 2 hours. Based on this I information the licensee decides that containment failure is imminent and recommends shelter out to 5 miles in all directions and 10 miles downwind. What is your assessment? (Cite the procedure used in your assessment.) A. i J f

i ', QUESTIONS 6/90 69 j section I 11. Q. What special precautions are recommended for school children in EPA guidance in the event of a severe accident? What does NRC guidance indicate is appropriate? A. 1 1 4

~ i 70 QUESTIONS 6/90,' Section I l 12. Q. An accident is declared a General Emergency and the licensee recommends evacuation of the preplanned area near the site (2-3 mile) in. all directions. The offsite officials are preparing'to make prearranged radio announcements in a few minutes to start the evacuation. Your analysis using the i Reactor Accident Assessment Method, Section C, indicate that doses in excess of EPA PAG levels, are projected beyond 3 miles if the accident continues to progress. This indicates that evacuation beyond 3 miles is appropriate. What recommendations should be made to offsite officials at this time? A.

l

1 i l l l ~ ', QUESTIONS 6/90 71 Section I i P 13. Q. Why might a minor suggestion from the NRC concerning [ protective action (e.g., evacuate 3 miles vs 2 miles) delay i their implementation? P A. l 4 i k i ) l s 4 4 2l-1 1 !t l 1 i ? i I i i i 5 J 4 1

1 I 72 QUESTIONS 6/90 ~ i Section I 14. Q. What organization is the most likely offsite protective action decision makers near a nuclear power plant i State office of rad-protection 11 NRC iii Local government iv State. office of Civil Safety v State Governor A. t l l l

~ i QUESTIONS-6/90 73 i Section I i 1 15. Q. It is 2:00 a.m. on a clear autumn night and you'haveljust j arrived at your Headquarters (or Region) response center. You have gotten a call at 1:00 a.m. that a plant with a PWR dry containment had an accident. The initial-information was: Scram at low pressurizer pressure Safety Injection did not come.on automatically but was started by the operators manually Containment pressure 12 psig - (operators did not start containment spray because RWST was decreasing fast) Some core exit TC's off scale (1000 *F+) - most at 700 Containment high rad monitor above 10,000 R/hr. (No other high readings outside containment.) Licensee had called State and local authorities to announce General Emergency but had made no protective action recommendations because there are no monitored release rates or field measurements indicate projected doses above EPA PAG offsite. Licensee is still on phone. Your boss arrives, asking "What is going on? Is there an offsite release? What is going on with the core? What is the projected dose and dose rate and, what is the magnitude of the source for these numbers - at two little villages 1-1/2 miles northwest, and 2 miles southeast. What is your response? A. i i 3 i -l

74 QUESTIONS 6/90 section I ~ 16. Q. It is now 5 minutes later - 2:08 a.m. - in the same situation as I-15 above. You are hard at work when you are instructed to use IRDAM to get a quick, offsite consequence analysis. No new information has been received from the plant. Give your response. A. I l I i

l i ', QUESTIONS 6/90 75 Section I 17. Q. Assume that a person is traveling away from the source of a release, under the plume. The plume follows the evacuee forever. If we also assume that the plume travels at the same speed as the evacuee we can use the graph on the next page to 1 provide an upper bound of the dose received by the evacuee. This graph is a reasonable method if it is a ground level release and the dose falls off at I \\ '*5 D =D ro 3 o r3, where dose down wind at point 1 D = i dose at starting point D = o distance from source of starting point r = o distance from source of evacuee r = 3 NOTE: - Later you could compare the whole body dose shown on page F-8 with this assumption and see if it is reasonable. - Based on Health Physics Journal, Vol. 32, April 77, pp. 305-307. The graph shows as a function of starting distance (r ) the o evacuation speed in mph (w) that must be exceeded so that the dose received while evacuating (D,) is less than the dose while sheltered (D,) for various exposure times (T). It is also assumed that the sheltering is a frame house providing a 1/2 reduction in total dose. For example, starting at 5 miles an evacuee must travel >10 mph to receive a dose less than a person exposed in a frame house at 5 miles for 2 hours. Assume a major release. a. Under what conditions is it clearly better to shelter? b. Under what conditions is it clearly better to evacuate? c. List why the assumptions used would not be appropriate. d. How will rain affect the result? e. For a short puff release (<.5 hr) what exposure time (T) should you assume?

a 76 QUESTIONS 6/90, Section I FIGURE j J0W1 W::l] JOSI C0:':?A:1::501 S:?I3] 70:1 IVAC JOS3 < S' L"3:1 HINIMUM EVACUATION SPEED (MPH) (V) / / 30 Puff-N0 GROUND SHINE N/ 20 / 10 ( l 1 On-r 0 1 2 3 4 5 STARTING POINT (r) HILES EXPOSURE T =.5 h l EXPOSURE T = 1h + EXPOSURE T = 2h O EXPOSURE T = 4H ' = SELTER EXPOSURE TIE i i

', QUESTIONS 6/90 77 Section I 18. Q. Examine Figure for Question #3, Section J (page 82) and Figure for Question #18, Section I (next page), Do you think a computer dose model could project these a. patterns? b. Can you explain the large area of contamination about 100 miles from Chernobyl? c. Do you see any patterns for these two actual events? A.

~ 78 QUESTIONS 6/90,' Figure for Section I Question #18 Windscale Contamination NRITH W O R KIN G T O N DERWENT I WATER LAwATER CRUMMOCK WATfR WHIT EH AV E N ENNERDA F HAWE5 WATER wA R WIND 5C A LE L AK E WINDERMERE SE A5C ALE LAKE ONI a eNon, f IRISH SEA BARROW e LANC AST ER ? y 27 Kn....... g ) 0.15 mR l Q >0.05 mR E >0.02 mR i I

QUESTIONS 6/90 79 Section J 1. Q. A reactor has had a severe accident resulting in considerable core damage. There is a large inventory of fission products in the containment and pressure is twice design. Does the " Intermediate Phase General Population Protective Action Assessment" procedure apply? A. i i i l 1 i i l l

80 QUESTIONS 6/90, Section J 2. Q. There is a large release from an accident. The release has stopped and further releases are not possible. The local officials have asked the NRC advice on allowing evacuated people to return to their homes. What procedure do you use and what are your first actions? A.

f l ', QUESTIONS 6/90 81 Section J 3. Q. Based on Figure for Section J, Question #3: a. Estimate the 1st year dose for the closed zone. b. Compare this with the Intermediate Phase PAGs to determine if people would be allowed to return. c. What area do you estimate people would not be allowed to live in based on EPA PAGs. d. Do you know the current USSR Intermediate PAG 7 NOTE: Table 1, page K-4, melt release fractions are in reasonable agreement with the observed results of the release from Chernobyl. A. i l \\

=. _. 4 82 QUESTIONS 6/90 Figure for Section J, Question #3 Radiation Hotspots Resulting from Chernobyl Nuclear Power Plant Accident h # %. + . s q ~r a\\_ h e s u r p .a. -/ ( V w./ y q \\. (h Belorussian + S. S. R. GOMEL' t \\ M ( - / I ey* y ,J \\,.w- ~

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QUESTIONS 6/90 83 ~ Section K 1. Q. Several hours after a release and activation of the NRC response you are requested to evaluate possible milk pathway doses based on some field monitoring results which show I-2 131 concentrations as high as 1.5 microcuries/m. What should be your first action? A. l l t l l l l \\ \\ a l I I

84 QUESTIONS 6/90 Section K 2. Q. Again assume that a release has occurred and the plant is now stable. The below table gives the results for the maximum levels observed at or near the projected plume centerline at various distances for samples of leafy vegetables being grown in the area. Harvesting is planned in 2-3 days for all of the leafy vegetables. DISTANCE IN MILES (pci/Kg) 5 10 20 30 6 6 5 5 I-131 3.7x10 1.1x10 3.5x10 2.1x10 Cs-137 4.8x10 1.4x10 4.5x10' 2.7x10' 5 5 Another agency has calculated that, for the nuclides, I-131 and Cs-137, for short term ingestion of the harvested vegetables, the Preventive PAG level is exceeded at about 5 miles. You are asked to check that these distances are proper. Perform a calculation for these isotopes only. A.

l QUESTIONS 6/90 85 Section K 3. Q. General Emergency has been declared. Assume that a release of noble gas, iodine and particulate has occurred over a 4 hour period (1100-1300) and the plant is now stable with no further releases expected. The following data was obtained from laboratory analysis of milk samples taken the next morning from cows. The milk sample was taken at 0500 milking Day 2 (12 hours after the plume past the dairy location). Cous were left on pasture during and after the plume passage due to not having stored feed available. The sample showed: pCi/ liter 0 0500 Day 2 I-131 8600 t 430 Cs-134 850 60 What Protective Actions should be recommended? ^* U

_ _ _. ~ _... QUESTIONS 6/90 86 Section M l l 1. Q. In the aftermath of a severe reactor accident that occurred about noon on Wednesday, a local law enforcement officer worked directing traffic until about 8:00 p.m. Preliminary monitoring and analysis indicated that he may have inhaled I-pills. He i 131. A State HP' gave him three 130 mg KI i started taking the pills with his morning coffee. What percentage of I-131 which he may have absorbed during traffic direction was blocked by the pills? A. i l l l

_. =.... , QUESTIONS 6/90 87 Section M' 2. Q. In a convention hall used as a relocation center, a rumor has spread through some 200 evacuees - that they may have been overexposed by inhalation of I-131. Describe a simple screening process to help allay their fears. 'A. 1 i

QUESTIONS 6/90 88 Section M 3. Q. A standard. Civil Defense GM instrument, CDV-700 with a Victoreen 6306 tube, registers about 2000 cpm in the procedure described in question M-2. What is the dose to the thyroid? A. I l l l 1 t l l J l 1 r l

~ QUESTIONS 6/90 89 Section M 4. Q. How long should someone take KI? A.

i QUESTIONS 6/90 90 Section o 1. Q. You receive a call at 10:00 am EST from an emergency room nursing supervisor at a hospital. She tells you there has been a traffic accident involving radioactive material. They are treating the driver and want to know if there are any special precautions that should be taken? A. t \\ P T i --w.

QUESTIONS 6/90 91 Section P 1. Q. There has been a fire at a materials facility in Ohio. There are indications that there has been a considerable (>300 C1) release of I-131. The state is reponding; but, NRC management directs you to have NRC conduct monitoring to assure public health and safety. What do you do? A.

.. -. - ~ -.. - - f t i 1 92 QUESTIONS 6/90 t 1 i Section Q 1 i 1. Q. Convert the following from SI to the units used in the U.S.: 50 Bq = Ci 3.5 Sv = rem i 2200*C = F l l A. J i 4 2 n L 3 12' 2 4 i t l t i s i 1 l 1

a ANSWERS 6/90 93 I l l I l I l ANSWERS AND COMMENTS \\ f 1

94 ANSWERS 6/90 Overview 1. a. Page I-4 WHOLE BODY THYROID ACTION < 1 REM < 5 REM No action 1 to < 5 5 to < 25 Evacuation 5 and above 25 and above Evacuation b. Page I-3, " Initial General Emergency Public Protective Actions" c. Page B-4, " Water Required to Cool Core by Boiling. d. Page H-19, Ground shine dose factors. e. Section F. f. Section C, C-2, F3. g. Pages K-6 and 7, Table 3, " Ingestion Actions Preventive / Emergency PAG." h. Page J-7, " Gamma Exposure Rate and Ef fective Dose Equivalent. " 1 1. Page A-3, Reactor Quick Assessment. j. Page I-5, " Summary of Permanent and Transient Population at Nuclear Power Reactors." k. Page R-1, " Radiation Explanation Chart for Public" 1. Page I-7, the controlled copies in HQ or at the regions would contain the local decision makers. m. Page M-3, " Percent of Thyroid Blocking Afforded by 130 mg of KI." n. Pages J-12 and J-13, " Recommended Surface Contamination Screening Levels...". o. Page K-10, Table 6. p. Page B-12, "BWR Mark III Containment Response." q. Section H-9. r. Page A-30, " Material Incident Emergency Classes." s. Page B-14, " Coolant Fission Product Concentrations."

ANSWERS 6/90 95 Section A 1. Correct classification is all important because all response actions to include protective actions have been preplanned based on classification. If a general emergency is not declared when appropriate, then the appropriate protective actions will not be taken. 2.

Alert, See Classification Assessment procedure, Reactor Quick Assessment, page A-6, Primary System Breaks & Leaks, "PWR rapid steam generator tube ruptures and -

leak > 300 gpm." OR Use the fission product barrier (challenge) indication matrix (A-27). This would involve the challenge to the RCS and possible challenge to containment. If control of release via secondary system is not achieved, containment barrier would be assumed to be lost and event would escalate to a Site Area Emergency. 3. Section A, procedure " Classification Assessment," Reactor Quick Assessment indicates: Inventory Control Unusual Event - ECCS start (page A-4) Primary System Leaks / Breaks Alert - primary system leak (page A-6) (through PORV) > 50 gpm j i Radiological Release Alert ' Inplant radiation levels (page A-7) 1000x normal Activation of Centera (?) Alert / Site Area Emergency Plant (page A-8) conditions that warrant activation of emergency centers i OR Use of the Barrier Challenge or Loss Indicator Matrix (page A-27) would give multiple indications of a challenge to the RCS barrier. Loss / challenge to RCS would lead to an Alert classification. 4. Since the magnitude of the earthquake tremors did not meet the licensee's emergency plan criteria for declaring an emergency, the licensee did not declare an emergency event and did not notify state and local officials of the earthquake. One' point to emphasize here is that the emergency plan criteria for declaring an emergency varies among licensees and may be different from the criteria specified in NUREG-0654, especially for seismic events (earthquakes). 1

i i d 96 ANSWERS 6/90 _' l Section A - continued l One reason for the differences is the geological characteristics of the particular plant site. For example, the Millstone plants are built on Connecticut Bedrock whereas the Palo Verde plants are located in the Arizona desert. } l S. As can be seen from " Water Required To Cool Core by Boiling," page i B-4, it will take more than 200 gpm to cool a core boiling 1-1/2 i hours after the reactor is shutdown. Thus, this situation clearly indicates that the plant has lost the { ability to keep the core covered and cool and core damage has j started. 1 Due to hich core temperatures the Classification Assessment procedure, Inventory Control (page A-4), Primary System Breaks & Leaks (page A-6) and Fuel Cladding Loss (page A-6), indicate that these conditions should be classified as a GENERAL EMERGENCY. It j is ngt appropriate to wait for coolant concentration analysis to confirm core damage because it is too slow and may not be correct, (there may be no flow through the sample line or coolant to l sample). i OR Use of the Barrier Challenge or Loss Indicator Matrix would classify the initial event at least at the Alert level. Per the table

notes, trends approaching criteria without adequate mitigation systems available should be classified assuming trends continue and barriers are lost.

Inadequate core cooling will result in increasing thermocouple temperatures and decreasing reactor levels. Either parameter is indicative of core melt sequences and declaration of a General Emergency. 6. Procedure, Classification Assessment, Reactor Quick Assessment Heat Removal Loss (page A-4) indicates that an ALERT emergency class should be declared because of the loss of a function needed for plant cold shutdown condition. 7. Steam produced by reactor decay heat is relieved from the reactor vessel by the SRVs and is condensed in the pressure suppression pool. The temperature of the suppression pool increases which causes the pressure in the primary containment (drywell) to increase. l

. -. - -. -. ~.. - 1 ANSWERS 6/90 87 Section A - continued This is a difficult situation to evaluate since there may be no immediate threat to the public. With no means available to cool the suppression pool water, the steam produced by the reactor decay heat will continue to heat the suppression pool water. The water in the pool will begin to boil in approximately 7.5 hours. (See j Figure 7a attached). 3 j As the suppression pool water continues to boil, the primary containment (drywell) design pressure will be reached in i approximately 21.5 hours and the primary containment (drywell) j failure pressure'will be reached in approximately 35 hours. (See Figure 7b attached). Procedure, Classification Assessment, Reactor Quick Assessment Heat Removal Loss (page A-4) indicate that a GENERAL EMERGENCY should be declared if the reactor shuts down but there is an extended failure of decay heat removal systems. The issue here is at what point should the licensee declare a GENERAL EMERGENCY in this situation? What questions do you think your managers would want j answered in a situation like this? r i Assuming normal or emergency systems could not be restored, parameter trends and evaluations would determine that various parameters would exceed criteria and also warrant classification of a General Emergency using the Barrier Challenge or Loss Indicator Matrix (page A-27).

gg ANSWERS 6/90 Figure 7a UNMITIGATED LOSS OF DHR - SUPPRESSION POOL TEMPERATURE ORNL-0WG 83-4227 ETD 8-NO POOL HEATUP DURING C,J 3' REPRESSURIZATION

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l l ', ANSWERS 6/90 99 Figure 7b UNMITIGATED LOSS OF DHR - DRYWELL PRESSURE (WITH EARLY TRIP OF DRYWELL COOLERS) ORNL,-DWG13./ 229 ETO S O-1 2-DRYWELL F AILURE PRESSURE 1 R. .4 RATE OF PRESSURE INCREASE SLOWED DURING REACTOR / VESSEL REPRESSURIZATION G-

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100 ANSWERS 6/90 Section A - continued 8. The integrity of the second fission product barrier is lost due to the break in the primary system. The integrity of the third fission product barrier (containment) is lost since the steam line break is outside containment. Some radioactive material will probably be released through the break into the outside atmosphere since there is a breach in some of the fuel cladding, the first fission product barrier. However, the magnitude of the fuel leakage is presumed to be below technical specification limits since the plant has been operating with "slightly leaking fuel" for six months. What plant equipment is designed to limit the consequences of the accident? The steam flow restrictor is designed to limit steam line flow to approximately 200% of rated flow for that steam line. By restricting the rate of steam flow, the loss of coolant from the reactor vessel is limited, the differential pressure across the reactor vessel internals is limited, and the rate of release of radioactive material is limited. It this primary system leak a loss of coolant accident? A loss of coolant accident is a primary system leak at a rate in excess of the capability of the plant's reactor normal coolant makeup systems to furnish sufficient water to keep the core covered and cooled. In this scenario, primary pressure blows down to atmospheric in a very short period of time and expels primary coolant out the break. The core at least partially uncovers. There has been a loss of the coolant necessary to keep the core covered and cooled. What other action (s) could the control room operators take to improve this situation? Control room operators would insure that the low pressure emergency core cooling system comes on to reflood the core when the low level in reactor vessel is reached, and would continue to pay particular attention to reactor water level indications, and indications of water and fuel temperatures. Should this situation improve or deteriorate with time? If present plant conditions remain unchanged, this situation should improve since the amount of decay heat being generated will decrease with time. So what is the proper classification of this event? i

ANSWERS 6/90 101 Section A - continued Classification Assessment, Reactor Quick Assessment " Primary System Breaks and Leaks" (page A-6) indicates that a SITE AREA EMERGENCY should be declared due to a BWR steam line break outside containment without MSIV closure. However, it could also be a GENERAL EMERGENCY due to a loss of any 2 of 3 fission product barriers and a potential loss of a third barrier. This accident should probably be classified as a SITE AREA EMERGENCY. Offsite radiation dose rates should be monitored and as specified in Initiating Plant Condition 3, a GENERAL EMERGENCY should be declared if radiation measurements at or beyond the site boundary will exceed the levels established by the environmental Protection Agency Protective Action Guides. OR Use of the Barrier Challenge or Loss Indicator Matrix involves indications of degraded fuel cladding, cha:1.lenge to or loss of RCS (reactor pressure, reactor vessel level, ECCS injection) and loss of containment (MSIV open, no containment pressure increase). Challenge / loss of RCS and loss of containment would be classified as a Site Area Emergency. If event challenged fuel cladding, escalation to General Emergency would be warranted. 9. Upon establishment of communications, the first task is to evaluate the event and the licensee's classification. The Alert classification based upon the fire in the area of the UF cylinders 6 seems appropriate. Classification as a Site Area Emergency involves response by state and local organizations rather than only notification. The specifics of the event, including consideration of whether the situation is under control or is degrading, should be factored into the judgment to request offsite response. The Materials Incident Emergency Classes are described on pages A-30 and A-31.

102 ANSWERS 6/90 Section B 1. Page B-2, " Core Heatup Once Uncovered" shows that the core temperature will be about 2400

• F (600
• F + 1*/sec x 1800 sec).

Page B-3 " Core Damage Progression Once Uncovered" shows that at this temperature there will be fuel cladding failure, H, release and rapid release of iodine, cesium and noble gas from the core. 2. No. On Page B-2, " Core Damage Progression Once Uncovered" the note at the top and the chart show that once the core is severely damaged it may become uncoolable. 3. No. Page B-4, " Water Required to Cool Core by Boiling" shows that more than 200 gpm must be boiled to remove the decay heat. Since coolant (water) would also be lost through the PORV considerably more than 200 gpm may be required to assure the core is protected. 4. Page B-10, "PWR Containment Monitor Response" bottom chart (spray on), indicated damage between 100% gap release (cladding failure) and about 20% melt. (In fact about 1/3 of the TMI core melted and another 1/3 of the core lost cladding.) This should be considered the minimum level of damage for the reasons stated in the " CAUTION" on page B-6. The uncertainty results from the difficulty in accounting for monitor shielding since the mix in the containment atmosphere was unknown (see first NOTE on B-6). 5. The top figure on Page B-10, "PWR Containment Monitor Response" indicates between normal 100% coolant and 1% gap release. As indicated by NOTE 2 on page B-6, " Containment Radiation Levels" difference in monitor readings can result from poor mixing. The elevated indications of the reactor building incore instrumentation l area and, especially, the reactor building dome monitor, apparently result from the hot steam discharging from the reactor coolant tank rupture disc and carrying the activity to the top of containment. Even though the bridge monitor and the dome monitor are both i exposed to large gas volumes in containment, there is not enough mixing to prevent radiation dose rates several orders of magnitude different. The initial dip in the fuel handling bridge and the reactor building incore instrumentation area monitors results from the decay of N-16 immediately following reactor trip. 6. Page B-13, " Saturation Table (P/T)" shows that the water in the vessel is boiling. This would be an indication of imminent core damage. 7. From the table on page B-14, "PWR Baseline Coolant Concentrations," it can be seen that probably more than gap concentration is in coolant. More than cladding has been damaged and some UO2 may be severely heated or melted. These I-131 concentrations are similar to TMI concentrations.

) ANSWERS 6/90 103 ) Section B - continued 8. Page B-16, " Hydrogen Flammability" indicates that if the containment is not inerted that a H detonation is possible. It 2 is uncertain if this would result in containment failure. Page B-14, " Containment Hydrogen vs Core Damage" indicates that there may have been metal water reaction with a large fraction of the available metal in the core. 9. The I-131 concentration (1.30 x 10' gCi/g) falls between the gap and melt baseline concentrations as expected. The I-133 concentration is below that for gap and this may be due to decay of I-133 (half life 20 hours). The Cs levels are also below the gap baseline. The baselines were for an old core. The TMI core was new and the Cs had not had an opportunity to build up (see NUREG/CR-3108). This may explain the difference. 10. From page B-12 " Saturation Table" we see that the primary system is subcooled (not boiling). Therefore, the 250 gpm injection will be insufficient to remove all the decay heat (see CAUTION page B-4). Therefore the core will start to heat up. The reason that the core will heat up is that the amount of heat (BTUs) necessary to raise a pound of water to the temperature of the system is considerably less (may be a factor of 5, depending on pressure) than the amount of heat necessary to transform that pound of water into steam by boiling; consequently, boiling water removes much more heat from the source (the reactor core) than simply conducting heat away by transferring it to water. 11. The decrease in both the dome monitor and the reactor building incore instrumentation area reading corresponds to the mixing provided by the reactor building fan coolers.

104 ANSWERS 6/90 Section C 1. Section C " Reactor Consequence Assessment Based from Plant Conditions" PMT-402. a. Tab Red - 3, Figure 4 (Page C-11), 4th branch from the bottom i Whole Body 4.8 E-3 rem Thyroid 4.0 E-1 rem As noted in Step 6 of the procedure, the specific dose projections should not be used in presenting your assessment. The possible consequences should be presented such as: "These conditions should result in doses offsite that are only a small fraction of the EPA PAGs." b. If you assume normal coolant, use Tab Blue - 1, Figure 9 (Page C-29), bottom tree, 2nd branch from the top. Whole Body 1.75 E-3 rem Thyroid 1.75 E-2 rem Assessment: "These conditirsns should result in doses offsite that arte only a small fraction of the EPA PAGS."

However, there could have been a " spike" when the tube ruptured and reactor shutdown. Therefore, to provide an upper bound you could also use the 100X normal tree (top tree) on page C-29, 2nd branch from the top.

Whole Body 1.75 E-1 Thyroid 1.75 E+0 c. Tab Yellow - 1, Figure 19 (Page C-55), 6th branch from the bottom (catastrophic = 100%hr) Whole Body 7.0 E-1 Thyroid 1.0 E+1 Assessment: "These conditions could result in doses of fsite just below the EPA PAGS." i

l i ', ANSWERS 6/90 105 Section C - continued d. Based on " Core Heatup Once Uncovered" the core temperature can be estimated as about (3000* = 600* + 2400* (l'F/s x 2400 sec)). This corresponds to a grain boundary release from the core (page C-61). Tab Green - 2, Figure 23 (Page C-71), 2nd branch from the top. Whole Body 3.2 E+1 rem Thyroid 3.2 E+3 rem Since there are doses in excess of PAGs beyond 1 mile, you must determine the distance to which consequences are possible. From Step 5 we see: Whole body at 1 mile = 3.2 E+1 rem at 2 miles = 1.2 E+1 rem at 5 miles =.28 E+1 rem Thyroid at 1 mile = 3.2 E+3 rem at 2 miles = 1.2 E+3 rem at 5 miles = 2.8 E+2 rem at 10 miles = 9.6 E+1 rem Assessment: These conditions could result in doses in excess of the EPA PAG several mile from the plant and doses in excess of the EPA Thyroid PAG more than 10 miles from the site. However, the whole body doses should be below those required to produce early health effects offsite. 2. Using procedure " Reactor Accident Consequence Assessment Based from Plant Conditions" Section C, this problem can be bounded by comparison of a wet well release (venting through suppression pool) and dry well release (containment failure). Tab Orange - 3 = "BWR containment Dry Well Leakage / Failure" for a core melt o Figure 18 (Page C-47) for sprays off, 2-12 hours hold up, 100%hr (catastrophic) release (2nd branch from the top) can be used to bound the near term major containment failure. This type of failure could result in the following doses at one mile Whole Body 220 rem Thyroid 20,000 rem

106 ANSWERS 6/90 _' Section C - continued Tab Yellow - 3 = "BWR Containment Wet Well Leakage / Failure" for a core melt o Figure 21 (Page C-59) subcooled pool, 2-12 hour hold up time, 100% day release (5th branch from bottom) can be used to bound a vent through the pool. Note 100%/ day release rate is used because this is a controlled vent not a failure case. A one hour release at this following dose at one mile. FILTERED NOT-FILTERED Whole Body .8 rem 1 rem Thyroid .2 rem 20 rem Clearly several hours of venting through the pool results in a major reduction in the offsite consequence but there are other questions before venting can be considered. Can venting be stopped? Are we sure the containment will fail? What will be impact on the venting system duct work? What will be the impact on the accessibility of various areas of the plant? 3. As discussed in NUREG-1210, Vol. 2 and NUREG-1228. The reactor fuel consists of uranium dioxide (UO ) pellets about 1 cm in 2 diameter by 1.5 cm tall. The pellets are stacked into a fuel can, called cladding, made of zircaloy about 11 feet long which is sealed. The fuel cans are also called fuel pins. The clearance space between the pellets and the inside of the cladding is called the gap. As the reactor operates, gaseous fission products are released from the fuel pellets and are trapped in the gap. The highly volatile gaseous fission product (i.e., 3% of Xe, Kr, 2% I, 5% Cs) constitute the gap gasses (See Section C, " Core Release Assumptions, page C-84). They are of special interest because if the fuel pins (cladding) are damaged, the gap gasses will be released very quickly at about 7.200*-1400' (see page B-3) and may provide the fir.st indication of core damage. This is called a gap release. 4. Section C, " Core Inventories" (page C-83) shows the inventories foranoJdcore. Iodine 131 inventory for a 1000 MWe reactor is 8.5 x 10 Ci (85,000 C1/MW(e) x 1000MW(e)). 1 a

i ANSWERS 6/90 107 Section C - continued 5. Section C, page C-89, "One Mile Dose Used in the Event Trees" indicates that the dose from the cloud shine, acute inhalation and 3 hours of ground shine (see NOTES on table) would be: Noble gas .6 rem = Particulate 100.0 rem = TOTAL Dose 100.6 rem = j 6. Section C, page C-11, forth branch from the top estimates that at 4 j shutdown the whole body dose would be 8.8 x 10 rem. Adjusting this dose based on " Shutdown Reactor Correction Factors" page a C-91 the dose becomes: 4 4 8.8 x 10 x.1 = 8.5 x 10 rem l 7. With this limited information the best tool to use is the " Reactor Accident Consequence Overview" page C-2. From this tree it is clear that this is most like one of the top two branches of the tree. In this case early health effects are possible offsite. 8. From Saction C I Whole Body Dose at 1 mile i 100%/hr release rate (Page # in ( }) 4 i I Core Condition .5 hr hold-up .5 hr hold-up 2/12 hr hold-up Spray-on Spray-off Spray-off Gap (Page C-7) 5 rem 42 rem 4 rem Grain Boundary (Page C-9) 100 rem 840 rem 100 rem Melt (Page C-11) 230 rem 2080 rem 220 rem Deaths might occur (very unlikely) at a bone marrow (whole body) dose above 220 rem at dose rate of > 5 rem /hr (see Section H, Table 10, " Health Effects," page H-24) with minimal treatment and at 330 rem with supportive treatment. Therefore, at least a grain boundary release from all of the core (major core damage) combined with early failure of the containment and the failure of the sprays is required to result in deaths offsite.

f los ANSWERS 6/90 Section C - continued 9. Section C, page C-80. The top branch of the trees indicates that the release of gap (reasonable worst case) from 15 batches would result in about a 10 rem dose at 1 mile. Since this pool only has 5 batches, the estimate would be adjusted to 3 rem. Here again, you should not provide numerical estimates. The assessment should indicate that doses in excess of EPA PAGs are possible offsite but early health effects are not possible, i i i ) r

ANSWERS 6/90 109 Section D 1. None. Given the great uncertainties, these two projection are very close (See NUREG-1210 Vol 2). 2. In view of the uncertainties, this measurement appears to confirm l the projection. But never assume anything based on a single, or only a few, environmental measurements. 3. As shown in NUREG-1210 Vol. 2, the source term could be off by a factor of 1,000,000 or more for a major unmonitored pathway. There would be that much uncertainty, for example, in whether containment failed or not. Even if the release rate is known (unlikely), the transport code estimates of concentration could be off by a factor j of 10 or more Finally plume movement (direction) is very uncertain. acute dose should always be used for 4. Whole body total dose insight for early health effects because early health effects are only seen at high dose rates required over relatively short periods. For reactor accidents bone marrow dose is considered to dominate early health effects. For some non-reactor accident, acute lung dose may dominate early health effect or deaths. For comparison with the Whole Body EPA PAG use the cloud shine dose and for the thyroid EPA PAG use the thyroid inhalation dose. See Section D "Use of RASCAL." l l 4 f r l l

110 ANSWERS 6/90 Section E 1. Section E, Use of IRPAM, page E-1 states in the " CAUTION" why IRDAM should not be used. It does not consider ground shine or inhalation for Whole Body dose. It only considers noble gases for Whole Body. i i I

ANSWERS 6/90 ill Section F 1. Since you need a quick dose estimate at 50 miles, use Section F, "Long Distance and Rain Reactor Projections." Figure 1, page F-3 indicates that the accident conditions are consistent with a PWR-4. The graphs on page F-8 "PWR #4 No Rain" show the following l doses for 24 hours at 50 miles: Total Whole Body .8 rem (top chart) Whole Body from Ground Shine .7 rem (middle chart) Whole Body from Cloud Shine .1 rem (middle chart) 2. As can be seen from the 24 hour components for the ground level release (PWR #1A page F-4) for Whole Body dose the ground shine is about 10 x the cloud shine and could result in doses > 1000 rem. Ground shine then is the source of the doses that could result in early deaths. For the highly elevated release (PWR #1B, page F-5), the ground shine does not contribute until the plume touches down. At 10 miles the ground shine still results in the highest i source of dose over a 24 hour period. 3. The deses from the IB case (page F-4) are much lower than the 1A case (page F-5) because the 1B has 20 times the energy of the 1A case and therefore results in a very elevated plume (See F-25, 26). The elevation of plume is what greatly reduced the local effects of the Chernobyl accident. In fact, these two cases were the basis for the early NRC and USA assessment of the Chernobyl accident. 4. The effective height of the plume will not be known, even after a release. Therefore, even if the composition of the source term were known (which it will not), it is unlikely that an accurate l (within a factor of 10 to 100) dose can be projected. Research on l U.S. reactors indicates that many severe reactor accidents will result in ground level releases. l l l

___ - -.. - ~.. i 112 ANSWERS 6/90 Section F - continued 5. 24 Hour Whole Body Dose No Rain (F-4 & 5) Rain (F-14 & 15) PWR 1A PWR 1B PWR 1A PWR 1B 1 mile 1,700 15 30,000 30,000 2 miles 1,700 2.5 10,000 13,000 5 miles 1,000 .4 2,000 4,000 10 miles 300 8 100 200 20 miles 100 20 4 4 Rain greatly increases the dose close to the plant. For an elevated release rain will result in doses close to a site similar to a ground level release during the rain by bringing the non-nobles down to ground level. Rain also increases the fall-off of dose with distance (by wash out). As a result, rain greatly increases the risk close to the site (or wherever it meets the plume); but, may reduce the distance from the site that requires action. s l l l l

I i ANSWERS 6/90 113 Section G 1. As stated in the first note on page G-1 "Use of ARAC" it is prudent not to expect a response from LLNL using ARAC in less than one hour. The one hour delay is after all information to include source term estimates has been provided. 1 2. ARAC has the same limitations as all other dose projection models, and in addition, ARAC is not presently capable of projecting early health effects, i.e., acute doses. 3. As discussed in Attachment 3 (page G-7) of "Use of ARAC", Section l G one cannot derive an absolute maximum dose projected from Figure j G-3. The only answer that can be given is that the maximum dose projected for the infant thyroid in Figure G-3 is greater than 25 rem. i l 4. Some of the " pockets" referenced may be only artificialities due ] to the grid-square averaging that takes place within the computer code. Others may and probably are caused by differences in terrain elevation. When dealing with computer model graphical output one l must keep in mind that we are dealing with a mathematical construct and not " ground truth." The object of the modeling effort is to produce a product that will mimic " ground truth." Modeling l assumptions, sophistication of the treatment of Atmospheric physics within the model and the quality and quantity of meteorological data that the model can utilize all affect the quality and accuracy of the modeling effort. S. Figure Sa is a Total Deposition Plot for all nuclides from 2200Z l on June 14, 1989 to 1000Z June 15, 1989. While Figure 5b is also Total Deposition for all nuclides; however, the time period is 2200Z on June 14, 1989 to 04002 on June 15, 1989. There is a six hour difference in the time over which the deposition occurs. The clue to the difference is the difference in the areal coverage (in l sq km) which is given along with the contour values in the accompanying notes next to the graphical plot display. l l l

114 ANSWERS 6/90 Section H ~ 1. Two equations would be needed to calculate the dose at two miles. The " Projections Downwind From Field Measurements" in Worksheet 3, and the inhalation equation in Worksheet 2. Use Table 1 criteria to estimate stability class based on the heavy cloud cover observation given in the problem. Table 4 contains the dispersion (xu/Q) factors needed for 1 and 2 miles. Since this involves I131, use Iodine dose factors provided in Table 11. WORKSHEET 3 xu/Q (2 miles) = 2. 0E-05 m 2 (Table 4)

5. 4E-05 m 2 (Table 4]I xu/Q (1 mile)

= 3 l 1 mile measurement = 1.0E-05 Ci/m I (given) (2. 0E-05 m-2) x (1.OE-05) Ci/m ) / (5.4E-05 m-2) 3 3 3.7E-06 Ci/m WORKSHEET 2 3 Air Concentration = (3.7E-06 Ci/m ) (Worksheet 3) 3 Dose Factor = 1.1E+06 (rem) (m )/ (Ci) (hr) (Table 11) Exposure Time = 2 hour (given) 3 3 (3.7E-06 Ci/m ) x (1.1E+06 rem-m /Ci-hr) x (2 hr) = 8.1 rem CONCLUSION The inhalation dose to the thyroid by I'3' is the greatest concern. Based on this estimate, any PAR should include dispensing of potassium iodide and, since EPA PAGs are exceeded, evacuation should be considered. 2. The problem implies that the 150 R was received over a 2-hour period during which the worker was involved in search and rescue operations. This would indicate that he received the exposure at a rate of 75 R/hr, if he were in a uniform field for the entire period.

However, this is unlikely in a search and rescue operation.

He could have received the entire dose over a period i ranging from the entire two hours, to less than a minute. If he received the entire dose within 9 minutes or less, the dose rate would exceed 1000 R/hr. Table 10, "Early Health Effects of Exposure to Radiation", show that a dose of 150 R at 1,000 R/hr or greater might be fatal in absence of supportive treatment. Medical attention should be provided for the worker.

ANSWERS 6/90 115 Section H - continued 3. Tables 2 and 3 provide dose projections for 1 Ci at distances of 1/4 mile and 1 mile, respectively. The tables project a dose of 8.5E-04 R to the bone marrow, and 1.3E-03 to the lungs for one hour at 1/4 mile for a 1 Ci Cs"7 release. Multiplying by 50 Ci gives doses of 42.5 mrem whole body and 65 mrem to the lungs at 1/4 mile. These doses are well below any early health effects projected in Table 10, and are below EPA PAGs as well. 2 2 4. pCi/100 m is converted to Ci/m by multiplying by 1.0E-10. 3000 2 pCi/100 cm converts to 3.0E-07 Ci/m. Projections in Table 6 for ground concentrations show 2.9E-07 for a 1 Ci release at 1 mile. This compares favorably with the field measurement. 5. Since it is not known when the release can be stopped, a prolonged release should be anticipated. Table 3 projects that doses to the residents at 1 mile would exceed EPA PAGs even if the release only lasted one hour. 50 Ci/sec x 3600 sec/hr = 1.8E+05 Ci in one hour Table 3 projects an immersion dose of 1.0E-05 rem for a 1 Ci release at 1 mile. The total Curies released multiplied by this 1-Curie dose gives: 1.8E+05 x 1.0E-05 = 1.8 rem Even if the sheltering factors indicated in Table 8 were included, those in woed-frame houses would still exceed the PAG: Table 8 sheltering factor for wood frame house =.9 1.8 rem x.9 = 1.6 rem Considering the dispersion factors in Table 4 it appears that their dose could be cut in half by just moving them 1 mile further downwind, even if they travel all the way directly in the plume centerline. The dose could be reduced by an order of magnitude if they could be moved 5 miles: xu/Q at 1 mile = 5.4E-05 xu/Q at 2 miles = 2.0E-05 xu/Q at 5 miles = 5.0E-06 Not to be ignored is the threat of a downpour. If it occurred, the ground shine dose would become a

\\ l 116 ANSWERS 6/90 Section H - continued ~ i significant factor in the 1-2 mile radius, while all doses would be proportionately reduced further away, thus, further favoring evacuation over sheltering. i 6. Let the old folks stay where they are. A warm, sunny Spring day with leaves rustling equates to B stability @ 2 m/s. See Table i 1 on page H-3. We can make a quick check by using Table 3, which shows 1 mile downwind dose for D stability, 2 m/s, a 1 curie release of Am-241, the following, in Rem / curies released: Inhalation Immersion Ground Bone Lung W. Body W. Body 4.3E-04 9.9E-03 3.1E-08 1.8E-08 From Table 4, the X/Q for D at 1 mile is 5.4E-5; the X/Q for B at 2 miles is 1.5E-6. Thus the doses above should be multiplied by 5 (for 5 curies) and by 1.5E-6 (for dilution further downwind) 5.4E-5 We can ignore shielding of an old frame house and we assume on a nice day the windows were open. Consequently the doses in Rem for 1.4 x 10'3 Ci/sec x 3600 sec, i.e., 5 Ci, are: Inhalation Immersion Ground Bone Lung W. Body W. Body 5.9E-05 1.35E-03 4.2E-09 2.4E-09

ANSWERS 6/90 117 Section I 1. 1) Prevent early deaths l 2) Prevent early health effects 3) Reduce doses above EPA PAGs 4) Control long term doce (reference: NUREG-1210, Vol. 4, page 4) 2. Severe core damage combined with early containment failure and failure of the systems used to mitigate a release (e.g., sprays). (reference: NUREG-1210, Vol. 2, page 21) 3. Before or shortly after the start of the release. (reference: NUREG-1210, Vol. 4, page 12) 4. Bypass (e.g., steam generator tube rupture) Isolation Failure Overpressurization Direct Containment Heating (high pressure vessel melt-through) H Burn Melt Through Intentional Venting only Intentional Venting is predictable. (reference: NUREG-1210, Vol. 2, pages 32- ) 5. For the worst accidents projected, no. (reference: NUREG-1210, Vol. 4, page 17) 6. As discussed in NUREG-1210 Vol. 2 Greatly increased risk of major release for core melt accident It is easy to detect core damage Core damage indicates that releases that could result in health effects are possible It is difficult to project containment response Protective action should be taken before or shortly after a release There are large uncertainties in doing dose projections 7. 2-3 miles. (reference: NUREG-1210, Vol. 4, page 12) 8. Protective actions should be taken in all directions for the area of greatest risk (2-3 miles). (reference: NUREG-1210, Vol. 4, page 25) ( ,-.,m

118 ANSWERS 6/90 Section I - continued 9. Dose projection or monitoring results compared to EPA PAGs. 10. Section I " Plume Phase General Population Protective Action Assessment" indicates the basic recommendation should be close in evacuation and not shelter. Shelter is not appropriate because the increasing containment pressure does not allow prediction of a short duration release before evacuation. In fact, only venting of the containment, before an evacuation is possible, or a major weather problem are judged to provide a basis of sheltering vs evacuation close to the plant in the event of a General Emergency. In addition, containments are estimated to fail at 2 to 3 times design pressure, not at design pressure. See NUREG-1210 Vol. 2, for a detailed discussion of severe reactor accident decision making. Since shelter is not the appropriate protective action, an attempt should be made to discuss your assessment with the licensee. Following this discussion, the ET or DSO should be informed immediately and provided with a description of the basis for your assessment and the position of the licensee. 11. Section I, Attachment 2, page I-4. EPA PAGs specify a range of 1-5 rem Whole Body and 5-25 rem Thyroid with the stipulation that children and women of childbearing age should be protected at the lower level of the range. However, for severe reactor accidents (General Emergencies) the Attachment 1 (page I-3) guidance indicates the entire population should take the same early predetermined protective action. These actions would be determined based on plant conditions. 12. As indicated by the " Plume Phase General Population Protective Action Assessment, Section I, Step 4, CAUTION (page I-1). "Do not interfere with or question offsite action if an announcement of protective action recommendations for the nearby population is imminent or has begun." In this case, unless there is an indication that offsite officials are not going to act on the recommendation, you should not take any action until the close-in population has been requested to evacuate. Once the preplanned evacuation has started, it may be appropriate to contact the licensee or offsite officials to discuss the need for further action. However, it is doubtful if these early assessments could provide an adequate basis for such a discussion. l

ANSWERS 6/90 119 Section I - continued 13. Implementation of protective action has been predetermined. The areas to be evacuated, the message to be used, the routes to be followed, are in place for fast implementation. Experience during exercises shows that offsite decisionmakers often delay taking i action in order to discuss any suggestions, even minor changes to a recommendation. 14. Local government early in the event (1st few hours). 15. You must act on what little information you have. Your first l action should be to inform NRC management of the facts and that a protective action recommendation should be made to evacuate people in a 2 mile radius around the plant based on Section I, procedure " Plume Phase General Population Protective Action Assessment." You should make it clear that at this time this is the most important issue and should be discussed with licensee management immediately to determine if there are reasons (e.g., severe weather) that make early evacuation near the plant impossible. 1 i 16. There is insufficient information to warrant running any of the models. The Reactor Accident Assessment Based on Plant Conditions a methods, Section C, would provide a reasonable estimate of offsite i consequences. In addition, IRDAM would not be appropriate in any case. RASCAL should be used if a model run is required. RASCAL i would allow consideration of plant conditions in estimation of the source term. 17. a. For very short puff release that do not result in much ground contamination or for very slow evacuation speeds. l b. For most reasonable evacuation speeds if you start close to the site. c. It may be elevated release. This could result in higher doses for the evacuees but as we showed earlier (Question F-3), an elevated release is less of a threat. d. If the release contained considerable non-noble fission products, rain would greatly increase the ground contamination and therefore the dose close to the site. The dose will also fall off very fast with distance. The result is that shelter will be less effective (see Problem F-4). It would depend on the ground contamination. If there were e. no ground contamination you would use .5 hr. But, if there were major ground contamination, you would use the time sheltered since they will continue to be exposed after the puff has passed, i h

[. 1 l 120 ANSWERS 6/90 4 j Section I - continued i 18. a. We don't think so. j .i b. We understand it resulted from rain. i i c. In the case of Chernobyl, the contamination puddled around the ~ i site. In neither case do the patterns look like those I i projected by simple plume models. 4 4 ) i 4 i i I I E 1 i i Y h l x

k l ANSWERS 6/90 121 Section J 3 1. No. As indicated by Step 1 on page J-1 the source of the threat is not under control. There is a radioactive inventory capable of release and a threatened barrier (high containment pressure). 2. PMT-409 " Intermediate Phase General Population Protective Action Assessment" page J-1. Actions 1 assure threat is under control (Step 1 page J-1) 2 get together with EPA, USDA and FDA, etc.,(Step 2 page J-1) and work as a group to address the issue. 3. Based on Section J " Intermediate Phase General Population Protective Action Assessment," Step 3, you are directed to page J-5, " Intermediate Phase Dose Projection Considerations." a. Since only Cs-137 levels are shown, note the second note on page J-5 indicates that the contribution of other fission products should be considered. The melt fission product fractions for a melt release from Table 1, page K-3, were used to estimate the other fission products. (Note the mix of fission products released from Chernobyl are in reasonable agreement with the melt fractions.) Once the ground contamination level for all the fission products is estimated, Table 2 (page J-7) is used to calculate the 1st year dose. The results are shown in the table on next page. The total { is 12.6 rem the 1st year which is 6 times the Intermediate Phase PAG of 2 rem. b. No. c. Assuming our calculation is correct, the entire Closed Zone (about 45 x 45 miles) of 2045 square miles would at least be evacuated. d. We understand it is about 30 rem.

i M M CHERNOBYL CIO8ED AREA (>40 Ci Cs-137/Km) Assumed ISQg,3 40 Ci Ground Contami-p a rem /1st vr 6tb> Cs-137 x (Cs-137) nation Ci/Km x 10 = m x pCi/m"'# m rem /1st year = 6.8 x 10'

6. 8x10' x 7.1x10

= 4 8. 2 x10' = 4. 8 x10' M Ru-103 .17 x 40 = = r 2 x 10' 2x10' x 1.2x10 = 2.4x10' 2.4x10' Ru-106 .05 x 40 -5 = = = M Te-132 7.5 x 40 300 x 10' e 2 2 3x10 x 3.2x10 = 9.6x10 = 9.6x10 = = t-s r 700 x 10' I-131 17.5 x 40 s 2 2 7x10 x 1.3x10 = 9.1x10 9.1x10 j = = = I-133 16.5 x 40 660 x 10' s 6.6x10 x 2.1x10'# g 13.8x10' = 1. 38x102 = = = O I-135 2.0 x 40 80 x 10' 8x10 x 1~.6x10'I

12. 8x10' = 1. 2 8 x10'

[ j = = = O 60 x 10' Cs-134 1.5 x 40 3 3 3 i 6x10 x 1. OxlO = 6.Ox10 = = 6.0x10 = C4 Ba-140 6.5 x 40 260 x 10' 8 3 3 2.6x10 x 1.1x10-5 = 2.8x10 = = 2.8x10 = l Cs-137 1 x 40 40 x 10' 4 2 3 = 4x10 x 4.5x10 = 18x10 = 1.8x10 g = C 4 1.26xlO (D M mrem ist year

== u ('8 Based on " Typical 2nd Day Mix of Ground Contamination Resulting From a Major Reactor Accident" (page K-3) (only those isotopes listed in both Table 2, J-7 and Table 1, K-3 are considered). (b) 6 3 10 pCi/m y 2: ") Table 2, J-7 lE: tn lC 03 Os No O i l r i

', ANSWERS 6/90 123 Section K 1. In accordance with " Ingestion Pathway Protective Action Assessment," Section K, page K-1, Step 1, you should contact the HHS representative. Their representatives should be located at the HQ Operations Center within a few hours of the start of the event. 2. As noted by the " Caution" on page K-1, other fission products should also be expected if there is I-131 and Cs-137 ground contamination. Table 1 shows a starting list to look for. Table 8 (page K-14) the " Ingestion Pathway Protective Action Assessment" procedure shows that the critical receptor for leafy vegetables is the child for 30 day ingestion (lowest DRLs). Step 3 (page K-1) directs you to take the individual DRLs from Table 8 and divide the DRLs into the measured values and sums /the individual ratios for each distance. This problem considered only 2 isotopes, I-131 and Cs-137. However, if all the other isotopes in Table 1 were considered in the calculation, the result will show that the preventive PAG is exceeded at between 20 and 25 miles. 6 The table below shows these calculations (note 10 pCi/kg = 4Ci/kg). Leafy Vegetables Ratio to the 30 Day Preventive Derived Response Level (K-14) Distance in Miles Isotope 5 10 20 30 I-131 (gCi/kg) 3.7/.67=5.5 1.1/.67=1.6 .35/.67=.52.21/.67=.31 Cs-137(gCi/kg) .48/1.1=.43.14/1.1=.12 .045/1.1=.04.027/1.1=.02 Total 5.98 1.72 .56 .33 Since the totals are >l out to 10 miles, the preventive PAG will be exceeded beyond 10 miles even without consideration other isotopes. 3. Follow the guidance in Section K " Ingestion Pathway Protective Action Assessment" l 1st Arrange for uncontaminated stored feed and water (if needed) { to assure that the cows do not continue the intake of contamination (Table 3, " Ingestion Protective Actions," page K-6). I 1

-=- 124 ANSWERS 6/90, Section K - continued 2nd Convert pCi/l to gCi/1 (see section Q-1) I-131 pCi/l 8.6x10 x 10 = 8.6 x 10'3 pCi/l 3 Cs-134 pCi/l 8.5x10 x 10 = 8.5 x 104 2 gCi/l 3rd For the milk which is in the tank: Calculate the ratio of the derived response level to the milk concentration level. For I-131 From the FDA Table .015 gCi/ liter Preventive Derived Response Level = (infant) (page K-8) Ratio =.0086/.015 =.57 or 57% of infant DRL (PAG) Similarly for Cs-134 Preventive Derived Response Level =.15 pCi/ liter Ratio =.00085/.15 =.0057 or 0.6% of the infant DRL If the decision on Protective Actions is made based on the above analysis, one would conclude that this milk is below the j derived Preventive PAG, and would be wrona. Using Figure 1, page K-22, for the buildup of the various nuclides in milk, one can calculate the fraction of the maximum concentration which would be reached to that at 12 hours. Measured at Fraction of Maximum Ratio of maximum 12 Hours Maximum (Unit Concentration concentration gCi/ liter Concentration gCi/ liter to DRL at 12 hours (12 hour Divided by concentration unit maximum divided by on Figure 1) fraction of maximum) 1-131 .0086 0.44 .0195 . 0195/. 015 = 1. 3 Cs-134 .00085 0.30 .0028 . 0028/.15 =. 019 The ratio ef the I-131 concentration to the DRL is greater than 1. Therefore, this milk sample is greater than the Preventive DRL (PAG) and some action should be taken. From a purely technical basis, diversion to cheese, butter, or dry milk solid would be acceptable as discussed in Table 3, page K-6.

ANSWERS 6/90 125 Section M 1. Practically zero. As seen from "Use of KI and Thyroid Monitoring" Figure 2, page M-3 to be effective as a blocking agent, KI should be taken within 10 hours before or 2 hours after exposure. 2. Procedure "Use of KI and Thyroid Monitoring,", Section M, page M-2, item 4, describes surveying the thyroid gland by placing a Geiger Muller tube directly under one's Adam's apple. An evacuee, j who has not been exposed, may feel reassured as he and the surveyor both observe the response on the radiation meter. 3. Procedure "Use of KI and Thyroid Monitoring," Section M indicates (Page M-2) that 500 cpm = 1 pCi I-131 therefore 2000 com = 4 pCi x 6.5 BAD = 26 RADS 500 cpm /yCi 4Ci This dose is at the threshold for administering KI. 4. Step 2, "Use of Potassium Iodide (KI) and Thyroid Monitoring" page M-1, indicates 3 days following exposure.

-126 ANSWERS 6/90 Section 0 1. As indicated by procedure 0-1, you should tell her or transfer her to REAC on 615-576-31311 { i l l l l [

=. , ANSWERS 6/90 127 Section P 1. Following the guidance in " Radiological Field Monitoring Assistance" page P-1, you should request DOE assistance. DOE i should be requested to send a RAP team and also to establish a FRMAC to coordinate the effort. The RAP team can be called out directly by calling the DOE Regional Coordinating Office for Ohio, i.e., the Chicago operations Office, as indicated on Attachment 1 on page P-4. FRMAC assistance should be coordinated through the HQ Operations Center. i I i I m m V

.o 128 ANSWERS 6/90 j Section Q 1. Using the SI conversion on page P-1 1.3 nCi 50 Bq = 350 rem 3.5 Sv = 2200*C 3992*F i i

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