ML19260C661
| ML19260C661 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/21/1979 |
| From: | Weiss E SHELDON, HARMON & WEISS, UNION OF CONCERNED SCIENTISTS |
| To: | NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| References | |
| NUDOCS 8001080259 | |
| Download: ML19260C661 (28) | |
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y
)
In the Matter of
)
)
METROPOLITAN EDISON COMPANY
)
Docket No. 50-289
)
(T hree Mile Island Nuclear Station
)
Uni t 1)
)
)
UNION OF CONCERNED SCIENTISTS INTERROGATORIES TO THE NUCLEAR REGULATORY COMMISSION STAFF Pursuant to 10 CFR 52.740(b) and 2. '/2 0 ( h ) ( 2 )( ii ), the Intervenor Union of Concerned Sciantists ("UCS") requests that the attached Interrogatories be answered fully, in writing, and under oath by any members of the Staf f who have personal know-ledge thereof.
The answer to each interrogatory should contain the name(s) and identification of the person (s) supplying the answer whether or not he or she has verified the answer.
Each question is instructed to be answered in 5 parts as follows:
A.
Provide the direct answer to the question.
B.
Identify all do~1ments and studies, and the particular parts thereof, relied upon by the Staff, now or in the past, which serve as the basis for the answer.
In lieu thereof, at Staff's option, a copy of such document and study may be attached to the answer.
C.
Identify all documents and studies, and the particular parts thereof, examined but not 1702 225 relied upon by the Staff, which pertain t 8 0 010 8 0 2LS 7
the subject matter questioned.
In lieu thereof, at Staff's option, a copy of each such document and study may be attached to the answer.
D.
Explain whether the Staff and/or any indepen-dont contractor are presently engaged in or intend to engage in further research or work which may bear on the issues covered in the interrogatory.
If so, please identify such research or work and the person (s) responsible therefor.
E.
1)
Identify the exoert(s), if any, whom the Staff intends to have testify on the subject matter covered in the interrogatory.
State the qualifications of each such expert.
2)
Present a summary of each expert's orocosed testimony on each UCS Contention.
3)
Identify all cases in which any such expert has previously testified and state the subject matter of such tes timo ny.
Answer each of the following five preliminary questions for everv Contention:
1.
Exclain the present Staff position on UCS Contention N (N=1-20).
2.
Does the current oosition differ from the position of the Staff in any prior cases ?
If so, identify the case ( s ), explain the crior position, and 1702 226
- explain the basis for the change in position.
3.
Identify any members of the Staff who dissent from the present Staff position on UCS Contention N.
Explain the reasons for which any Staff members dissented from the present Staff oosition on UCS Contention N.
4.
Identify the specific sections and page numbers of the SER and/or FSAR for TMI, Uni t 1, which are relied upon in formulating the Staff position on UCS Contention N.
5.
Identify all sections and page numbers of the SER and/or FSAR which contain subject matter pertaining to UCS Contention N.
CONTENTION 1 1-5. Answer each of the five preliminary questions with respect to Contention 1 and number the answers 1-5.
6.
Explain whether or not natural circulation is an adequate means for removing decay heat from the reactor core in the event of a small loss-of-coolant-accident ("LOCA").
7.
Explain in detail which of the short or long term measures recommended by the Director of Nuclear Reactor Regulation will prevent the formation of voids in the reactor cooling system as occurred at TMI-2.
8.
Doe s the Staff take the posicion that implemen-tation of the short term and/or long term 1702 227
_4_
measures identified by the Staff will consti-ture conformance with the requirements of GDC 34?
If yes, identify the specific measures and explain how their implementation will meet the requirements of GDC 34.
9.
If the answer to question 28 above is "no,"
what additional measures will be required by the Staff to achieve conformance with GDC 34?
10.
Wha t is the schedule for implementation of the additional requirements, if any, described in question #9 above?
11.
Doe s the Staff take the position that the short term and/or long term measures identifed by the Staff will constitute conformance with the requirements of GDC 35?
If yes, identify the specific measures and explain how their imple-mentation will meet the requirements of GDC 35.
12.
If the answer to question #11 above is "no,"
what additional measures will be required by the Sta ff to achieve conformance with GDC 35 ?
13.
Wha t is the schedule for the implementation of the additional requirements, if any, described in question #12 above?
CONTENTION 2 14-18. Answer each of the five preliminary questions with respect to contention 2 and number the answers 14-18.
19.
Explain how the Licensee will provide forced cooling flow to the reactor during small LOCA'1702 228 c
. 20.
Explain how any and all proposed methods for providing " enhanced" cooling to the core after a small LOCA will comply with all applicable NRC safety regulations and are sufficiently reliable to safeguard the public health and safety.
21.
If the licensee plans to rely on the reactor coolant pumps, explain how the formation of voids will be prevented.
22.
Explain how the reactor coolant pumos will comply with GDC 17 (on-site power supply).
23.
Explain how the reactor coolant pumps will comply with IEEE 279 (10 CPR 50.55a(h)-controls),
and 24.
Explain how the reactor coolant pumps will comply with GDC 's 2 and 4 (seismic and environmental qualifications ).
25.
If the licensee olans to rely on the emergency core cooling system ( ECCS ) in a " bleed and feed " mode, explain how there will be suffi-cient capacity and radiation shielding for the storage of the radioactive water bled from the primary coolant systen.
CONTENTION 3 26 -3 0. Answer each of the five preliminary questions with respect to Contention 3 and number the answers 26-30.
1702 229
. 31.
The staff has recognized that the " ma in te nance of natural circulation capability is imoortant to safety ( and ) de pends on the maintenance of oressure co ntrol (which) is normally achieved through the use of pressurizer he a te rs. "
p.
A-2.
Explain then, why pressurizer heaters and their associated controls are not classified as
" co mpone n ts important to safety" as discussed in GDC 17 and the Introduction to Appendix A to CFR P ar t 50.
32.
Explain in detail whether and in what matter the following design criteria would be met with respect to the pressurizer heater and its associated controls,
- a. GDC 22 ( diversi ty )
b.
GDC 2 and 4 (seismic and environ-mental qualifications )
c.
GDC 20 (automatic initiation) d.
GDC 3 and 22 (separation and independence )
e.
GDC 1 (quality assurance )
f.
GDC 17 (adequate, reliable on-site oower supply) and g.
the single failure criterion.
1702 230
CONTENTION 4 33-37. Answer each of the five preliminary questions with respect to Contention 4 and number the answers 33-37.
38.
Explain why the addition of the pressurizer heater to the on-site emergency power supplies will not degrade the capacity, caca-bility, and realibility of the on-site emergency power source in violation of GDC 17.
39.
Explain why installation of an independent and redundant on-site emergency oower suonly for the pressurizer heater would not orovide greater reliability of power supply to pressurrizer heaters.
CONTENTION 5 40-44. Answer each of the five preliminary questions with respect to Contention 5 and number the answers 40-44.
45.
Does the Staff agree that proper operation of PORV's, associated block valves and the instruments and controls for these valves is essential to mitigate the consegunces of accidents ?
Explain your response fully.
46.
Does the staff agree that failures of these valves, ins trumen ts and controls can cause or aggravate a LOCA?
Explain your response fully.
47.
Provide the justification for the failure to classify power operated relief valves (PORV's) and associated block valves and their respec-tive instruments and controls as "comoonents 1702 231 important to safety," requiring compliance with safety-grade design criteria.
48.
Explain how the motive and control components of the PORV's and their associated block valves and the vital instruments shall be supplied by the on-site emergency power source when offsite power is not available without degrading the capacity, capability and reliability of emergency power in violation of GDC 17.
49.
How have the devices through which motive and control power components for the PORVs and their associated block valves are connected to emergency buses been qualified in accordance with safety-grade requirements?
CONTENTION 6 50-54. Answer each of the five preliminary questions with respect to Contention 6 and number the answers 50-54.
55.
Describe in detail the methods by which the reactor coolant system relief and safety valves have been qualifed to verify the capability of these valves to function during normal, transient and accident conditions.
This description should include specification of the environmental conditions assumed during normal, transient and accident situations and the means by which these environmental conditions were derived.
1702 232 Provide ref erences to the Regulatory Guides applied in this analysis.
56.
Did the Staff fully apply the analysis of accidents and anticipated operational occur-rences referenced in Regulatory Guide 1.70.
Revision 2,
to determine the expected valve operating conditions?
If not, provide the justification for f ailing to do so.
57.
Explain how the licensee chose the single failures applied to these analyses so as to maximize the dynamic forces on the safety and relief valves.
58.
Explain how the test pressures utilized in these analyses were determined to be the highes t pressures predicted by conventional safety analysis procedures.
59.
How did the licensee determine the test conditions for qualification of the control circuitry, piping and supports associated with the reactor coolant system relief and safety valves ?
60.
Explain how the qualification testing of the reactor coolant system relief and safety valves and associated control circuitry, piping and supports complies with GDC 1, 14, 15 and 30.
1702 233 CONTENTION 7 61-65. Answer each of the five preliminary questions with respect to Contention 7 and number the answers 61-65.
66.
Would any of the short and/or long term measures recommended by the Staf f provide instrumentation to directly measure the water level in the fuel assemblies ?
Explain your answer fully.
67.
If the answer to question #66 is "no,"
explain the basis for concluding that the provisions of IEEE 279, 54.8, as incorporated in 10 CFR 50.55a(h) have been complied with.
68.
Would a direct measurement of the reactor coolant level be of assistance to the reactor operator in determining the most apprcoriate remedial actions during a small break LOCA?
69.
Explain how present procedures and instrumentation permit prompt recognition of low reactor coolant level and inadequate core cooling.
70.
What chort term modifications of existing oroce-dures and/or instruments nave been made at TMI-l for monitoring water level in the fuel assemblies ?
71.
How do any such new procedures and/or instruments differ from those in places prior to the accident at TMI-2 ?
72.
Describe the training program to inform reactor operators of new procedures.
Document the number of hours of instruction on these new procedures.
1702 234
. 73.
Explain how any modifications proposed by the Staff will provide more direct measurement of reactor cooling level and inadequate core cooling.
Wha t is the implementation schedule for any proposed modification?
74.
Discuss how the reliability of information from any proposed instrumentation compares with the reliability of the direct measurement of the reactor coolant level.
CONTENTION 8 75-79.
Answer each of the five oreliminary questions with respect to Contention 8 and number the answers 75-79.
80.
Discuss in detail what analysis has been performed for a spectrum of small break locations which demonstrates that the specific parameters of 10 CFR 50.46 will not be exceeded, with particular attention to peak cladding temperature (50.46(b)(1))
and hydrogen formation (50.4#-(h)(3)).
81.
Discuss in detail how the analysis described in question #80 above differs from the analysis performed in the FSAR and SER for TMI-1.
82.
Wha t steps have the licensee and the Staff taken to prevent the fuel cladding temperature from exceeding 2200'F and to prevent the reac-tor of more than 1% of the cladding with water or steam to produce hydrogen in the event of a
small LOCA.
1702 235 CONTENTION 9 83-87.
Answer each of the five preliminary questions with respect to Contention 9 and number the answers 83-87.
88.
When did the Staff make the formal decision against "backfitting" of Regulatory Guide 1.47?
89.
Provide all contemporaneous documentation supporting the Staf f's decision not to backfit Regulatory Guide 1.47.
90.
Does TMI-l comply with the orovisions of Regulatory Guide 1.47?
If not, explain how the design of TMI-l differs from the require-ments of the Regulatory Guide.
91.
Is it the position of the Staff that TMI-l can be operated with adequate protection for the public health and safety without a determination by the Staff that TMI-l is now, finally, in full comoliance with Req.
Guide 1.47?
Explain your answer fully.
CONTENTION 10 92-96. Answer each of the five preliminary questions with respect to Contention 10 and number tha answers 92-96.
97.
Is it the position of the Staff that " protection system" as referred to in 64.16 of IEEE 279 does not refer to high-oressure ECCS, low-pressure ECCS, containment isolation, emergency power or other prescribed safety functions ?
Explain your answer fully.
1702 236 98.
Does the Staff take the position that the design of TMI-1 complies with S4.16 of IEEE 279 as incorporated in 10 CPR 50.55 ( a )( h ) in light of the TMI-2 7.ccident ?
Explain your answer fully, particularly with reference to the operator's shut-off of the ECCS.
99.
Does the Staff take the position that the TMI-2 operator prevented a protection system action from going to completion?
100.
If the answer to question 399 above is "yes,"
what specific design features have been recom-mended to prevent this from recurring?
CONTENTION ll 101-105. Answer each of the five oreliminary questions with respect to Contention 11 and number the answers 101-105.
106.
In the opinion of the Staff, what should be the hydrogen capacity for which the TMI-l hydrogen recombiner should be designed and installed?
How does this capacity compare with the total amount of hydrogen which could be produced theoretically in the event that 100% of the fuel cladding reacted with water or steam?
107.
Provide the detailed technical justification for the Staff conclusion that the croposed design capacity for the TMI-l hydrogen combiner is suitably conservative.
i702 237 CONTENTION 12 108-112. Answer each of the five oreliminary questions with respect to contention 12 and number the answers 108-112.
113.
With respect to TMI-l list the " structures, W
sys tems, and components imoortant.to saf,ety" within the containment and auxiliary buildings to which GDC 4 presently applies.
114.
What were and are the maximum environmental parameters which each such GDC 4 structure, sys tem, and component is qualified to withstand?
115.
To what extent did *he actual accident conditions at TMI-2 exceed the past and present maximum environmental parame ters for each such structure, system, and component discussed in question #113?
116.
What was and is the length of time in an accident environment for which each such structure, system, and component is qualified to remain operable?
117.
To what extent did the actual accident conditions at TMI-2 exceed the past and oresent time periods for which each such structure, system, and compo-nent is qualified to remain operable under accident conditio ns ?
118.
Describe in detail the method used to qualify each such safety structure, system and component as me eting GDC 4.
Provide the relevant documentation.
119.
Is each such structure, system and component qualified according to the criteria of IEEE-3 2 3-19 74, 1702 238 as modified by Reg. Gui de 1. 8 9 ?
120. For each system, structure or component not qualified according to the criteria of IEEE-323-1974 as modified by Regulatory Guide 1.89, provide the criteria by which it was qualified.
121. List all equioment within the containment and auxiliary buildings previously deemed to be qualified which f ailed, either wholly or partially, during or after the accident at TMI-2.
Describe the way in which each such piece of equipment failed and the reason (s) fo r the failure.
122. When was the decision made not to "backfit" Regulatory Guide 1.89, incorporating IEEE-323-1974?
.23.
Provide all contemporaneous documentation sucporting the decision against backfitting Regulatory Guide 1.89, incorporating IEEE-3 23-19 74.
124. What is the Staff's present rationale for f ailing to now require compliance with Regulatory Guide 1.89, incorporating IEEE-3 2 3 -19 74, for TMI-l?
125. Is it the position of the Sta f f tha t the methods and assumptions used during the licensing and review of TMI-l to environmentally quality safety-related systems, structures and components (including environmental parameters, testina and analysis methods, length of time for whjah equip-ment must remain functional, and identification of " safety-related" equipment ) are adequate to 1702 239 fully comply with a ) GDC 4 and b) Regulatory Guide 1.89?
126.
If the answer to ques tion #125 is other than an unqualified, "yes, " state in what manner those methods and assumptions were deficient in any respect.
127.
If the answer to question #125 is other than an unqualified "yes," state which of the short and/or long term measures recommended by the Staff will correct those deficiences.
Explain your answer fully.
CONTENTION 13 128-132. Answer each of the five preliminary questions with respect to Contention 13 and number the answers. 128-132.
133.
Does the Staf f take the position that all credible accidents have been included within the design basis for TMI-l?
Provide all relevant documentation supporting your conclu-sion.
134.
What is the probability that an accident beyond the design basis for TMI-l will occur?
Provide all relevant documentation supporting your conclusion.
135.
What is the probability that the following ac cident scenarios described in NASH-1400 will occur at TMI-l?
a.
PWR 5 b.
PWR 4 1702 240 c.
PWR 2 Provide all relevant documentation st'pporting your conclusion.
136.
State what the health rafety and environmental consequences would be (short and long term) of each of the accident scenarios identified in question al35 above, including procerty damage.
137.
Define " Class 9" accidents.
138.
Prior to the accident at TMI-2, did the Staff have an opinion as to the probability of the TMI-2 accident ?
Wha t was that opinion?
139.
Identify the members of the Staff who worked on Section 3.3 (p. 3 3-6) and Appendix A#10 (p. A A-15) of NUREG-0 5 85, TMI Lessons Learned Task Force Final Report.
140.
Identify the members of the Staff who have been assigned or will be assigned to work on the proposed rulemaking relating to the conse-deration of design features to mitigate accidents that would result in core melt and severe core damage.
(See NUREG-0585, TMI-2 Lessons Learned Task Force Final Report o.
A A-15.)
141.
Provide all draf t and final analyses, memoranda, re po rts, recommendations or other documents produced by, relied upon or consulted by the Staff relating to the probability or consequences of accidents beyond the current design basis and/
1702 241
. or to measures designed to mitigate the conse-quences of such accidents.
142.
Will the Staff take the position in the proposed rulemaking described in question #140 above that accidents beyond the present design basis should be considered in the safety review for nuclear power plants ?
" xplain the Staff's position fully.
143.
Does the Staff take the position that the accident at Three Mile Island exceeded many of the pres en t design bases by a wide margin and was evidently a significant orecursor of a core-melt accident.
' 'TJREG-0 5 8 5,
p.
3-5 ).
144.
Does the Staff take the position that "[b]ecause the accident at Three Mile Island exceeded many of the present design bases by a vide margin and was evidently a significant precursor of a core-melt accident,
the NRC should begin to formulate requirements for design features that could mitigate the consequences of core-melt ac cide nts. " (NUREG-0585, p.
3-5) 145.
Given that TMI-2 has been identified by the Staff as a Class 9 accident and Class 9 accidents pose the gravest threat to the public safety of all possible nuclear reactor accidents, explain how the Staff can adequately orotect the public without consideration of Class 9 accidents and consequences.
1702 242 146.
is it the position of the Staff that their present method of reviewing the safety of nuclear power plants which excludes consideration of accidents beyond the design-basis, is sufficient to ade-quately protect the public health and safety?
147.
In Offshore Power Systems (Floating Nuclear Power Plants)
CLI-79, Sept. 1979), the Commission asked the Staff, pending generic rulemaking on the consideration of class 9 accidents at land based reactors, to bring to the attention of the Commission "any individual cases in which (the Staff ) believes the environmental conse-quences of Class 9 accidents should be considered."
Sl.op. at 9-10.
Explain why the Staff does not believe the consequences or likelihood of another Class 9 accident should be considered during the present review of TMI-1.
148.
Provide the criteria used by the Staff pursuant to the Commission's directive in the Offshore Power Sys tems case, supra, to determine which individual cases should include a consideration of the environmental consequences of Class 9 ac cide nts.
149.
In its Memorandum and Order on the Offshore Power Sys tems case, suora, the Commission directed the Staff to
"[ p3 rovide us with rec-ommendations on how the interim guidance of the Annex might be modified, on an interim basis 1702 243
. to reflect developments since 1971 and to accord more fully with current Staff oolicy in th s area.
(Sl.op. at 9).
a.
Have any such recommendations been provided to the Commission?
If so, supply them.
b.
If no such recommendations have yet been provided to the Commission, why not?
When will they be provided?
c.
Provide any draft memoranda or recommenda-tions which have been orepared in response to the Commission's directive d.
Identify the Staff members who have been assigned to work on a response to the Commission's directive.
e.
What are the " developments since 1971?"
f.
What is " current Staff oolicy in this area?"
CONTENTION 14 150-154. Answer each of the five preliminary questions with respect to Contention 14 and number the answers 150-154.
155.
Identify all systems and components cresently classified as non-safety-related which contri-bu ted to the cause of the TMI-2 accident, aggra-Vated the accident or were called uoon to attempt to mitigate the accident.
Discuss their role 1 7 qL p_
!7d4 in the accident sequence.
i
156.
Does the Staff agree that some systems and components presently classified as non-safety-related can have an adverse effect on the core because they can directly or indirectly affect temperature, pressure, flow and/or reactivity?
Identify all such systems and components related to the core cooling systems.
157.
Which of the short and/or long term measures recommended by the Staff are directed toward preventing adverse effects on the integrity of the core caused by non-safety-related systems and components?
How will these measures correct the de ciciencies identified in NUREG-0 5 78, Section 3.2?
Wha t is their schedule for implementation?
158.
Does the Staff propose to classify the reactor coolant pumps as safety-related?
If not, explain your answer fully.
159.
Does the Staff propose to classify the steam generators as safety-related?
If not, explain your answer fully.
160.
Which of the General Design Criteria applying to safety-related systems, structures and components does the Staff cropose to apoly to the reactor coolant pumps ?
If the Staff proposes to apply less than all of them, explain why those not applied have been excluded.
1702 243 161. Answer the same question as #160 above with regard to the steam generators.
162. Explain how the Sta ff can assure the adequv e protection of the public health and safety when systems and components, which are classi-fied as non-safety-related but already have been demonstrated to contribute to the aggrava-tion of mitigation of the TMI-2 accident have not been classified as components important to safety and required to meet all safety grade design criteria.
CONTE NTION 16 163-167.
Answer each of the five oreliminary questions with respect to Contention 16 and number the answers 163-167.
168. What basic assumptions and methodology were used to define the 10 mile emergency planning zone for the plume exposure cathway?
What were the parameters of the accident assumed to occur?
Wha t assumptions were made about meteorology?
169. Pro vide the NRC and EPA Sta ff input into NUREG-0396, EPA 520/1-78-016.
" Planning Basis for the Development of State and Local Gover nment Radiological Emergency Response Plans.
This should include draf t and final memoranda, reports and other documents.
1702
?45
176.
What is the effectiveness of potassium iodide as a thyroid blocking agent?
171.
Potassium iodate is already stockpiled in England and has a longer shelf-life than potassium iodide.
Has the Staff analyzed the use of potassium iodate as a thyroid blocking agent and what are its advantages and disadvan-tages compared to potassium iodide?
172.
Does the Staff propose to require the distribu-tion of potassium iodide ?
Explain your answer fully.
173.
Explain in full the f actual bases for the 10 mile emergency planning zone for the plume exposure pathway.
174.
Explain how and in what sections the TMI-l emergency response olan meets the "Near Term Requirements for Improving Emergency Preparedness" issued by the NRC Division of Opera ting Reactors on Sept. 13, 1979.
175.
Explain how and in what sections the TMI emergency response plan meets the " Draft Emergency Action Level Guidelines for Nuclear Power Plants" NUREG-0610 (the successor to Reg. Guide 1.101) 176.
By what method will the public be notified of recommended protective sections and how 1702 247 long will it take to notify all persons within 10 miles of the reactor?
177.
To what extent did people outside the recommended class of evacuees and the 5 mile radius from TMI-2 evacuate of their own volition?
178.
Explain how the emergency response olan for TMI-l takes into consideration the likelihood of " spontaneous" evacuation outside the 10 mile EPZ wnich may interfere with evacuation efforts within the EPZ.
179.
Explain how spontaneous evacuation by the public will be discouraged.
180.
To what distance would the entire 360 degree circumference of the reactor be evacuated, regardless of wind direction during a major atmospheric release?
181.
Explain how the angle size and length of evacuation sectors will be determined.
182.
How much time will be necessary to evacuate each 45, 67.5, and 90 sector around TMI out to 10 miles ?
Include the estimated time necessary for notification, for mobiliza-tion or preparation, for evacuation, and for actual transit time.
1702 '48 183.
What is the spectrum of radioactive alume speeds factored into the emergency response plan ?
184.
Explain how the decision will be made whether to order sheltering or evacuation during an atmospheric release of radioactivity.
185.
Who will order the appropriate protection action for the public and on the basis of what information?
186.
In endorsing the concept of EPZ planning guidance, the Commission stated "it is appropriate and prudent for emergency planning guidance to take into considera-tion the principal characteristics of a spectrum of design basis and core melt accidents. "
44 FR 61123 (Oct. 23, 1978)
Explain how the TMI-l emergency response plan considers the principal characteristics of a spectrum of core melt accidents.
Dr. Jan Beyea has written a draf t report to the Presidents Council on Environmental Quality entitled "Some Long Term Conse-quences of Hypothetical Major Releases of Radioactivity to the Atmosphere from TMI. "
T able I on page 5 of the recort summarizes the health consequences to the population over 50 miles from TMI as the result of 6 hypothetical radioactive atmospheric releases.
Beyea calculates that a PWR 2 core melt with breach of containment could result in up to 60,000 long term cancer deaths, 60,000 genetic defects, and 450,000 thyroid nodules.
The technica 4g calculations which provide the basis for these projected figures are provided in Appendix E of Beyea's report.
( co oy attached )
187.
Does the Staff agree with Beyea's dose calcula-ti ons, dose / effects coefficients, population calculations, and interdiction criteria for agricultural and human use ?
Dis cuss fully any Staff disagreement with any of these calcula-tions.
188.
Does the Staff agree with Beyea's conclusions regarding projected health consequences as the result of radioactive atmospheric releases?
Discuss the technical basis for any disagreement.
189.
Explain how the Staff is taking into account the ongoing expedited rulemaking on emergency planning in their review of the adequacy of the TMI-1 emergency response plan.
The Union of Concerned Scientists By:
Ellyn R.
Weiss SHELDON, HARMON & WEISS 1725 I Street, N.W.
Suite 506 Washington, D.C.
20006 (202) 833-9070 Date:
December 21, 1979 1702 250
e Appendix E TECHNICAL DETAILS OF CALCULATIDNS This appendix has been written for readers who are familiar with accident consequence calculations.
Background information, for those who are not, can be found in Appendix VI of WASH-1400 (the Reactor Safety Studv) or in Ref. E-1.
IT Dose Calculations A.
Meteorological Model Calculations were made for typical weather conditions:
5 m/see wind speed; Pasquill stability Class, D;.01 n/sec deposition velocity. A time-independent Gaussian plume model was used with " top hat" approximation.
Dispersion parameters were taken identical to those used in WASH-1400 (for a 30 minute release duration).
Although experimental data used to determine the dispersion parameters are scarce beyond 20 miles, the model is satisfactory for calculating health effects when a linear relationsh%p is used between dose and response.
In such a case, the total number of health effects depends only upon the total population dose, which in turn is rather insensitive to the dispersion parameters and otaer modelling details -- if the population dis tribu tion is uniform.
Variation in the population density with distance from the reactor can introduce a model dependence Lnto the results, but a large population effect, which dwarfs the radial variations, has already been included by calculating health effects for different wind directions.
In a uniform population distribution model, the inhalation dose cocponent of the population dose tends to vary inversely with the deposition velocity.
- See Appendix VI of WASH-1400. Note that full Gaussian calculations were made when calculating contaminated areas.
1702 251
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UNITED STATES OF A'iERICA NUCLEAR REGULATORY COMMISSION s
a
)
4 m
In the Matter of
)
)
METROPOLITAN EDISON COMPANY
)
Docket No. 50-289
)
(Three Mile Island Nuclear Station
)
Unit 1)
)
)
CERTIFICATE OF SERVICE I hereby certify that a copy of " Union of Concerned Scie nti s ts Interrogatories to the Nuclear Regulatory Commission S ta ff" were mailed postage prepaid this 21st day of December to:
Secretarv of the Commission ATTN:
Chief, Docketing and Service Section U.S.
Nuclear Regulatory Commission Washington, D.C.
20555 Ivan W.
Smith, Esquire Atomic Safety & Licensing Board Panel U.S.
Nuclear Regulatory Commission Washington, D.C.
20555 Dr. Walter H.
Jordan 881 W.
Outer Drive Oak Drive, Tennessee 37830 Dr. Linda W.
Little 5000 Hermitage Drive Raleigh, North Carolina 27612 George F.
Trowbridge, Esquire Shaw, Pittman, Potts & Trowbridge 1800 M Street, N.W.
Washington, D.C.
20006 James Tourtellotte, Esquire Offios of the Executive Legal Director U.S.
Nuclear Regulatory Commission Washington, D.C.
20555 Ellyn R.
Weiss 1702 252