ML19330B293
| ML19330B293 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 07/29/1980 |
| From: | Moon C Office of Nuclear Reactor Regulation |
| To: | TEXAS PUBLIC INTEREST RESEARCH GROUP |
| References | |
| ISSUANCES-CP, NUDOCS 8007310312 | |
| Download: ML19330B293 (27) | |
Text
.
July 29, U80 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
HOUSTON LIGHTING & POWER COMPANY Docket No. 50-466 (Ali vs Creek Nuclear Generating
)
Station, Unit 1)
)
HRC STAFF'S PARTIAL RESPONSE TO TEXPIRG INTERR0riATC'IES AND PRODUCTION OF DOCUMENTS The NRC Staff responds, in part, as follows to the interrogatories propounded by TEXPIRG in the captioned proceeding. The remaining responses will be filed as soon as the necessary Staff reviewers complete current review assignments, but in any event, no later than two weeks.
No. 8 Natural Gas Alternative _
1.
Provide the factual basis and a summary of the testimony each witness is expected to provide to the following issues or questions.
(a) What are the reasons that natural gas cannot be used as a fuel to replace the need for ACNGS?
(b) What sections of what laws prevent the use of natural gas for utility fuel after 19907 (c) Is natural gas an environmentally safer fuel than nuclear power? (d) Will the expected supply of natural gas be sufficient to supply the Applicant with enough energy to replace ACNGS?
(e) How do the future prices of nawral gas compare to that of nuclear and coal? (f) Have any new laws either been passed or proposed in the U.S. Congress that would amend the Industrial Fuel Use Act of 1976 to allow utilities to continue to burn natural gas past 19907 (g) Do the present laws allow utilities in the Non-attainment areas for oxidants and sulfur dioxide to be exempted from the required conversions away from the use of natural gas?
(h) How do the present and future costs of electricity compare when generated by natural gas versus nuclear?
3007310 3) A
Q 4 Response 1(a). The reasons stated on page S.9-2 and S.9-3 of the Final Environmental Statement (FES),NUREG-0470, August 1978.
1(b).
Sections 2(a), (b), and (c) of the Energy Supply and Environmental Coordination Act of 1974 (Public Law 93-319 ESECA) and the Industrial Fuel Use Act of 1976.
1(c). As stated on page S.9-3 of the FES, the Staff does not find natural gas to be a viable fuel for a 1200- W e base-load power station.
Hence it did not compare the environmental impacts for gas versus nuclear.
1(d). The Staff has not attempted to make a determination of the expected supply of natural gas.
1(e). As noted on page S.9-2 of the FES, deregulated costs of natural gas would be equivalent to the current cost of fual oil and, therefore, too costly for boiler fuel.
1(f). None have been passed.
Proposals would not be applicable at this time.
1(g). The Staff does not know.
J
l l 1 1
1(h). Tha Staff aid not find natural gas to be a viable fuel for a 1200-MWe l
I base-load power station and, therefore, did not compare costs of 1
generating electricity when generated by natural gas versus nuclear.
2.
Identify any studies or documents which show that a new natural gas plant will not produce electricity more economically than a nuclear plant.
Response
None known to the Staff.
No. 12 (Conservatism and Interconnect) 1.
What conclusions reached by Taylor in his book, Energy:
the Easy Path, do you disagree with?
Response
The Staff has not reviewed the reference for the purpose of determining agreement or disagreement with conclusions of the author.
2.
What steps have you taken to promote conservation within the last 10 years, and what effect have they had to date?
Response
None 1C has no charter or authority to promote energy conservation.
i 1
4 3.
If the Applicant undertook a major effort (spending at least 2 billion dollars within next 10 years) to promote the conservation of energy of its customers, what % of the energy now wasted could be saved?
Response
The Staff has not performed studies of the type suggested specifically for the Applicant's service area. The Staff's analysis of potential impacts of con-servation are discussed in Section S.8.2.3 of the Supplement to the FES.
4.
If you doubt that more than 30% of energy could be conserved then give the basis of ycur conclusion.
Please cite all studies or reports that you rely on for this conclusion.
Response
See the response to 3.
5.
How much more generating capacity than necessary to supply peak demand do each of the utilities within 500 miles of Houston, Texas have at the present time? (Answerat least as to those utilities taking part in the ongoing
" Federal Interconnect" proceedings.)
Respone.e As stated in Section S.9.1, the Staff concluded that the purchase er exchange of power was not a viable alternative.
C
4
. 6.
What effect on necessary reserve margins (now about 15%),
wi11 the proposed "DC Interconnect" proposed by HL&P and CSW have on the Applicant?
Response
The Staff does not have any firm indication of effects on reserve margin by the proposed DC Interconnect.
In general, if there is no purchase, reduction of more than a few percent would not be expected.
7.
What studies or reports do you have to indicate that the demand for electricity within the Applicants service area will not be less than that projected in the ER and FES because of Harris County and Galveston County being des-ignated as "non-attainment" areas by EPA and thus restricting the growth of pollution causing (and electricity using) industries.
Sumnarize your position on this issue.
Response
None.
8.
Identify and summarize the results of any studies which show that it would not be economically feasible for Applicant to reduce its reserve margin through " interconnection of grids."
Response
The Staff knows of none.
(See Section S.9.1.1 of the supplement to the FES.)
. No. 26 (Error in Computer Program) 1.
What assurance do you have that the computer program (s)
~
used by the Applicant to calculate forces and stresses on the rcuctor, associated piping, and containment do not hr.ve errors involving the subtraction of forces that should be added?
Response
The computer codes used have been benchmarked and produce correct results.
The errors postulated in the interrogatory are not present.
In addition, the results obtained for ACNGS will be audited and reviewed by the Staff.
No. 31 (Technical Qualifications) 1.
Specify in what ways the quality assurance program of the Applicant for ACNGS differs from that proposed by the applicant for the South Texas Project.
Response
The quality assurance program of the Applicant for ACNGS is described in Chapter 17 of the ACNGS PSAR.
Likewise, the quality assurance program of the Applicant for the South Texas Project is described in Chapter 17 of the South Texas Project PSAR.
The descriptions of the Applicant's quality assurance progrea are virtually identical in these two PSARs except as noted below.
1
. a.
Job Titles and Reporting Relationships of QA Management South Texas Project ACNGS Executive Vice President Vice President Power Plant Construction
& Technical Services I
Manager Manager Quality Assurance Quality Assurance i
i Supervising Engineer Projects South Texas Project QA Manager Project QA Supervisor Home Office Site Supervising Engineer Site QA QA Supervisor Site QA Site QA Group Group b.
The Quality Assurance Program Evaluation Committee is chaired by the Manager -
Quality Assurance per the South Texas Project PSAR and is chaired by the Vice President - Power Plant Construction and Technical Services per the ACNGS PSAR.
c.
The South Texas Project PSAR shows a QA Plan divided into four major areas:
1.
Organization 2.
Project Engineering Administration 3.
Project Construction Quality Surveillance 4.
Auditing o
l l _.,
. The ACNGS PSAR shows a QA Plan divided into eight najor sections:
1.
Introduction 2.
HL&P Organization 3.
Project Administration 4.
mngineering 5.
Procurement 6.
Construction and Fabrication 7.
Records 8.
Auditing d.
The ACNGS PSAR includes (in part 17.1.11A) a paragraph on procedures for system turnover from the constructor to the applicant which is not included in the PSAR for the South Texas Project.
e.
The ACNGS PSAR incorporates the quality assurance programs of General Liectric and Ebasco instead of Westinghouse and Brown and Root.
2.
Do you maintain that Ebasco would carry out a better quality assurance program than Brown and Root? Why or why not.
Response
We do not maintain that Ebasco would carry out a better (or worse) quality assurance program than Brown and Root.
Either organization would be required to meet the l
requirements of Appendix B to 10 C.F.R. Part 50.
I
. 3.
Do you deny that the Applicant underestimated the amount of c)nstruction materials to be used in the S. Texas project in' the amounts specified in (b) of the contention? Why?
Response
The Staff does not deny or affirm the underestimates cited because it does review materials estimates.
4.
Is the design of the ACNGS plant further along than the S. Texas plants were at the time of their Construction permit hearings? What percent complete were the design plans of S. Texas at the time of its CP hearings, and what
% complete are the design plans of ACNGS at the present time?
Response
The Staff does not monitor the pre-construction permit status of design plans.
Because of the delays on Allens Creek it is reasonable to anticipate that the design of ACNGS may be further along than was the design of South Texas at the time of its CP hearings.
5.
What amounts of steel, concrete, rebar, piping, wire and cable, terminations, cable trays, and conduit are now expected to be used in ACNGS? What is the total projected cost of each of these items such as concrete?
Response
The Staff does not monitor amounts or costs of various construction materials.
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7
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. 6.
How many of the applicants direct employees have a Ph. D.
degree in either a science or engineering field? For each give their name, position with applicant, years experience since receiving Ph.
D., sumary of present duties, university degree (Ph. D) received from, and undergraduate grade point average. Which have degrees in nuclear physics or nuclear engineering?
Response
The Staff does not require that the Applicant provide such infonnation.
Experience and education is provided for managerial personnel in Appendix 13.1A to Section 13.1 of the Preliminary Safety Analysis Report.
7.
Do you feel that the NRC was justified in fining the applicant
$100,000 dollars so far in 1980 for violations of the NRC regulations? Why or why not?
Response
The Office of Nuclear Reactor Regulation has no reason to question justification for the actions of the Office of Inspection and Enforcement.
8.
List the violations charged against the applicant that resulted in the 100K fine over problems at the S. Texas project.
Response
See Appendix A, " Notice of Violation" enclosed with letter of April 30, 1980, from NRC Office of Inspection and Enforcement to HL&P.
. 9.
What effect in the way of increased cost of construction at the S. Texas plant has the quality assurance program there had? What % increase has that been?
Response
The Staff does not have such information.
- 10. What % of the increased cost of the S. Texas project, do you think was caused by a) intervenors, b) increased costs of NRC regulation changes, c) miscalculation of original estimates, d) technical incompetence of applicant, and other causes? Please detail the basis of each part of the answer.
For example in b), list each of the regulation changes, the date of change, and increased cost to S. Texas from such change.
Response
The Staff does not have such information.
- 11. What specific changes has the applicant made to increase its ter".nical competence as a result of a) Three Mile Island lessons learned, b) new NRC regulations, c) NRC report and fine concerning quality control at S. Texas?
Response
None has been reported in the Allens Creek PSAR.
- 12. What % of total cost of S. Texas and ACNGS has and is expected to result from a) technical costs such as design and technical employees, b) cost of materials such as concrete and steel, c) interest,and d) overhead such as i
attorneys and bookkeepers, and office space.
l l-,.
7
n..
. Response As noted in Section 20.2 of Supplement No. 2 to the Safety Evaluation Report, the Staff only makes a rough check of cost estimates made by the Applicant and does not attempt to make detailed engineering cost estimates.
No. 21 (Radiation Exposure of Workers) 1.
How many employees are planned to be on each crew (shift) for the operation of ACNGS? How many man-rems of exposure are expected each year of operation (1st through 40th)?
Which positions such as operator or electrical repair, etc.
are expected to average the highest exposure / year? How much?
Re;ponse We cannot meaningfully estimate the average number of workers per shift, or the worker breakdown by job type per shift for ACNGS or any other proposed plant.
This number will vary widely during any single year depending on the operational status of the plant. The number of workers per shift and worker type during a refueling / maintenance outage will differ greatly from the worker number and type onsite per shift during normal plant operation.
Figures for the total number of persons receiving measurable exposures per year by plant is given for each operating plant in the report, " Occupational Radiation Exposure at Conmercial Nuclear Power Reactors 1978." This report also contains a breakdown of annual collective dose by work function for all operating reactors. This data was compiled from the individual plant dose breakdowns per work function presented in Appendix C of this report. We have not estimated the man-rems per year for l
i L
u-
. each year of operation or by worker category.
For our comments on dose estimates, see our response to Interrogatory 21(4).
2.
What has been the total occupational exposure from each of the BWR's that started operation during the last 10 years for each of those years in operation? Restrict to those in US.
For example, Browns Ferry #1 was 267 man-rem in 1978 and 345 man-rem in 1979.
Response
See response to Interrogatory 21(6).
3.
What changes have been made in ACNGS as compared to older BWR's that would reduce the expected occupational exposure of ACNGS? Detail expected reductions to be expected from each change whether design (thicker concrete) or operational (new rules).
i
. Response In the course of our review of the Allens Creek application for a construction permit, we considered actions taken by the Applicant to reduce occupational radiation exposures, against the provisions of Regulatory Guide 8.8, "Information Relevant to Ensuring that Occupational Radiation Exposure at Nuclear Power Plants will be as Low as is Reasonably Achievable" (Revision 2). On the basis of the information submitted, we have concluded that the Applicant's radiation protection design is acceptable. There are alternate acceptable ways to meet the various provisions of Regulatory Guide 8.8.
The acceptability of the complex array of alternate actions selected in the design of a particular plant cannot be compared directly on the basis of comparing specific selected design features.
4.
Detail any reasons that you believe that it is not " reasonably achievable" for Allens Creek to keep its total occupational exposure equal to or less than that of the average exposure of the exposure experienced by the BWR in 2 above that has kept its exposure the lowest.
Response
Past experience from operating nuclear power stations has been used to provide a widely applicable estimate of occupational exposure to be used for all light water reactor pmver plants of similar type and size. This experience indicates an riarage value of 500 man-rems per year per unit.
O
. On this basis our FES estimated an average occupational dose of 500 man-rems for determination of impact.
It is not possible to project meaningfully the occupational radiation exposures that will occur at any given plant such as Allens Creek, either on the average or in any given year.
Reactor exposure experience overall varies widely from year to year and plant to plant for reasons that cannot be predicted in advance.
For example, at one plant a highly radioactive component may fail unexpsctedly and need to be repaired or replaced with a large resulting dose.
Another similar plant may not experience the same difficulty. Both plants could be maintaining doses ALARA.
5.
What has the mean, median, and range of exposures to workers been at the plants listed in 2 above, i.e. that to individual workers such as in 1977, Browns Ferry mean was O.89 rem, median was 1.0 rem, min was 0.07 rem, and max was 15.9 rem.
Response
See response to Interrogatory 21(6).
c-~
. 6.
For each of the plants in 2 above, and for each of the years since 1970, show the number of workers that received more radiation than that now allowed by the NRC regulations.
Response
Infomation required of NRC licensees about persons exposed, annual man-rem doses, and distribution of doses among different personnel are published annually. The most recent report, " Occupational Radiation Exposure at Commercial Nuclear Power Reactors 1978," published November 1979, is enclosed.
Infomation on the number of workers exposed to radiation in excess of our limits is contained in the public document room in each plant's docket file.
No. 28 (Control Room Design) 1.
Detail the major ways that the control room layout and design differs from that of TMI.
Explain why these differences make it easier for the operators to control the plant under all accident conditions.
2.
What is the maximum distance between instruments that might need to be read by one operator? What are those instruments and where are they placed in the control room.
3.
What are the maximum number of a) visual and b) noise alams that might be activated at the same time (not start at same time,.but still be in the alam state while other alams are active).
4.
During the worst accident conditions planned for at ACNGS, how many people would be allowed in the control room, and what are each of their positions (job not physical),
and duties.
5.
Detail the features of the ACNGS control room design that make it impossible for an operator to make an error that would make an accident happen or make an accident worse that was in progress.
. Response The Staff has not completed its revised requirements for control room design that reflect lessons learned from TMI, including application of human factors.
When any new or revised requirements are forwarded to the Applicant for HL&P, all parties will receive copies.
In addition, in accordance with 10 C.F.R. 550.34, an applicant for a construction permit is required to submit in its PSAR only the preliminary design of the facility. Therefore, detcils as to the specifics of control room layout and design, instrumentation, alarms, and staffing are not available for Staff review at this stage of licensing.
No. 39 (Reactor Vessel) 1.
What is your understanding of the purpose and status of Task A-ll?
2.
Why were PWR's allowed to use less safe reactor vessel materials than BWR's? Explain.
3.
What is the basis for the concern expressed in Task A-ll?
4.
What is the measure of reactor fracture toughness, i.e.
/l pounds / square inch. How is it measured with small samples and in reactors?
5.
How does reactor fracture toughness vary with a) the material used, b) the temperature of the material, c) d past radiation exposure of the material, d) the thic.; ness of the material, e) the shape of the material, and f) its past history of temperature cycles, extremes, radiation, welding, atmosphere of ambient air, etc.
6.
What is the accuracy af these measurements when applied to large operating reactor vessels, i.e. % error? What has been the range of values measured for past reactor vessels?
What is the " safety factor" required for ACNGS? Are there
bs.
. requirements or plans to test the pressure capability of the ACNGS by actually applying overpression to the actual reactor vessel at various times during its operating 1.ife?
Response
Task A-11 was initiated when the Staff observed from the results of surveillanc test data that the reactor vessel in a small number of the older operating reactors contained beltline region material with potentially reduced toughness properties. The purpose of Task A-11 is to develcp licensing criteria to ensure that an adequate margin of safety is maintained during nonnal operating and The criteria are postulated accident conditions for those reactor vessels.
being developed and the objective of Task A-11 is being met.
l, f
s f
However, for plants now in the licensing stage, including Allens Creek Nuclear of 10 C.F.R.
Generating Station, Unit 1, current criteria set forth in Appendix G Part 50, as well as Appendix G of Section III of the ASME Code, require that the 4
materials used for fabricating the reactor coolant pressure boundary are adequate l
to ensure, throughout the design life, suitable fracture toughness properties and adequate margins of safety.
Current criteria equally apply to materials in all reactors--both PWR's or BWR's.
Reactor materials are required to meet the fracture toughness properties of Appendix G of 10 C.F.R. Part 50. The acceptance criteria are stated in Standard
. Review Plan 5.2.3 and 5.3.1.
The stress intensity value required to cause The stress intensity fracture is a measure of a materials fracture toughness.
is expressed in psb, and a referenced value is obtained for each reactor v Irradiation results in a decrease of fracture toughness, and the irradiation eff The nil-ductility temperature of a material are stated in Regulatory Guide 1.99.
is measured by measuring the Charpy impact resistance as a function of
{
temperature.
1 The Allens Creek facility meet the regulatory requirements for fracture toughness.
I No. 41 (Over-pressure in Vessel)
What is the ASME Boiler and Pressure Vessel Code upper 1.
What values pressure limit for the ACNGS reactor vessel?
are the applicants relief valves set at for the reactor vessel?
Response _
As shown in PSAR Table 5.2-1, the reactor vessel design pressure is 1250 ps f
As shown in PSAR Table 5.2-6, the spring set pressures of the main steam
~
See relief valves varies between 1165 and 1205 psig, depending on the valve.
s response No. 7 below for a discussion of ASME Code pressure limits.
t-
i l
. 7.
What is the basis for allowing the ACNGS overpressure capacity to be greater than the reactor coolant pressure boundary design pressure? Where is the large safety factor here?
Response
The ACNGS main steam safety / relief valves will be sizW so that under upset plant conditions, the maximum pressures in the reactor coolant pressure boundary will respectively be no greater than 1.1 times the design pressure.
Under emergency plant conditions, the pressures will be limited so that Service Limit C of the ASME Code will not be exceeded. This is in accordance with the ASME Boiler and Pressure Vessel Code and is based on excepted practice and i
g experience in all types of preswee usages for a period of time comparable to I
i ACNGS plant and is therefore acceptable. We are satisfied that compliance with the ASME Code provide a safe design.
I I
f Nos. 40 and 53 (Hydrogen Explosion)
State all the ways that the ACNGS hy(drogen gas detection i
1.
systems differ from those of THI?
both for containment and vessel).
i
Response
Reactor vessels for nuclear power plants do not have hydrogen gas detection j
detection system inside the con-t systems in them. TMI-2 did not have an H2 concentration at TMI-2, samples of the containment i
tainment. To obtain the H2 ACNGS atmosphere had to be taken through a penetration and then to a gas analyzer.
L..
~
. will have an H detection system that can take samples of the containment atmosphere 2
automatically and provide a continuous readout in the control room of the H2 concentration at eight preselected locations inside the containment.
2.
Do vou believe that it would be impossible for hydrogen explosion to take place in the vessel or containment of ACNGS? Why?
Response
could exist inside the The TMI accident showed that an excessive amount of H2 reactor coolant system (RCS) without there being any chance of an explosion.
This is because the free oxygen concentration inside the vessel could not reach the flammability limit due to the recombination of free oxygen that was needed to produce an continuously taking place. The concentration of H2 I
explosion in the containment is 18%.
Using the most current analytical methods, and assuming the worst single failure with no credit for using non-safety grade It.
I concentration inside the ACNGS will be less than 4%.
i systems, the maximum H2 t
3.
What is the minimum explosive force or pressure within a) vessel and b) containment that it would take to cause i
a crack? Explain? What pressure would it take to also cause the concrete shell to shatter?
1
Response
f As described in the PSAR, the Applicant will design the reactor vessel in I
3(a).
accordance with the ASME Boiler and Pressure Vessel Code for specific Built into the ASME Code design equations are design and peak pressures.
A.
. substantial factors of safety such that the design and peak pressures would have to be greatly exceeded before actual failure might be expected.
In our review we are not required to detennine any such faiNre We would add that the ASME Code considers constant, or static, pressure.
pressures to be more severe than a temporary pressure spike due to an explosion.
Consequently, there is an even larger, though unquantified, factor of safety against failure due to explosive pressure inherent in the ASME Code design.
3(b). The response to (a) is generally applicable. The containment is an ASME Code vessel. The pressure to cause failure of the concrete shield building has not been calculated.
5.
What do you think was the results of the hydrogen gas explosion (pressure transit) at TMI? What was the maximum pressure measured during the transit? What was the con-centration of hydrogen gas just before the explosion?
t just after?
Ii i
Response
The pressure spike that occurred during the hydrogen burn (approximately 9-1/2 hours after the accident started) reached 28 psig. The containment air samples taken concentration of about 2%. Since the several days after the accident showed a H2 gas monitor, the con-TMI-2 containment was not equipped with a continuous H2 centration of H inside the containment that existed before the pressure spike 2
)
, into the containment, Because of the release point of the H2 is not known exactly.
it is possible that the majority of the gas releasad was confined to a small considerably portion of the containment, making the possible concentration of H 2
was uniformly distributed throughout the containment.
higher than if the H2 At this point in time, with no actual inspection of the equipment inside the No containment, there are no known adverse results from the pressure spike.
change in the containment leakage rate was noticed, and no safety related equipment failed; even the non-safety related equipment (fan coolers, va instruments, etc.) was not affected.
I I
If a condition existed for a hydrogen explosion of sufficient i
force to crack the vessel, containment and concrete wall, 6.
would you recommend that the operating crew stay in the control room or leave fast?
i t
I
_ Response The control room is designed for occupancy during all design basis events.
The scenario postulated is not a design basis event.
g.
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a g
D
. No. 50 (Latching)
Do you deny that the radiation from a nuclear plant can 1.
cause the phenomenon of " latching"? Do you agree that
" latching" can cause an electronic manfunction?
Response
We do not agree that the radiation intensity fonn a nuclear power plant is To the best of our sufficient to cause the phenomenon of " latching."
knowledge the tenn " latch-up," as applied to electronic systems when exposed to radiation environments, refers to the random change of state of electronic i
5 3
f circuiting when exposed to very high levels of pulsed radiation, well above As such, its occurrence represents an electronic malfunction.
4 10 R/Sec.
i Provide your basis, if any, for taking the position that the f
2.
radiation from ACNGS could not increase the probability that aircraft near it would be more likely to fall into it than if no radiation was emitted into the air.
j
Response
The basis for the Staff position rests on the following:
The radiation levels around the ACNGS will be well below the radiation 1.
levels that an aircraft would encounter in normal flight operations.
l i
I The phenomenon of " latch-up" cannot be caused by the ACNGS, neither during 2.
normal nor accident conditions, but can be produced only during conditions of intense radiation fields, many orders of magnitude greater than that fourd near a nuclear power plant, such as during nuclear weapons bursts.
UNITED STATES OF AMERICA NUCLEAR REGULATORY C0:@11SSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of HOUSTON LIGHTING & POWER COMPANY Docket No. 50-466 (Allens Creek Nuclear Generating
)
Station, Unit 1)
)
AFFIDAVIT OF CALVIN W. MOON I hereby depose and say under oath that the foregoing NRC Staff re.sponses to interrogatories propounded by TEXPIRG were prepared by me or under my supervision.
I certify that the answers given are true and correct to the best of my knowledge, information and belief.
h
'% N~
~
Calvin W.~ Roon Subscribed and s rn to before me thiQfiay of 1980.
%9 hv Ohb Notdry Pub 1fc My Comission expires:
/, / 98 y
o k
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of HOUSTON LIGHTING & POWER COMPANY Docket No. 50-466 (Allens Creek Nuclear Generating
)
Station, Unit 1)
)
CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF'S PARTIAL RESPONSE TO TEXPIRG INTER-ROGATORIES AND PRODUCTIO" "~ DOCUMENTS" and " AFFIDAVIT OF CALVIN W. M00N" in the above-captioned proceeding.iave been served on the following by deposit in the United States mail, first class or, as indicated by an asterisk by deposit in the Nuclear Regulatory Commission internal mail system, this 29th day of July, 1980.:
Sheldon J. Wolfe, Esq., Chairman
- Richard Lowerre, Esq.
Atomic Safety and Licensing Board Panel Asst. Attorney General for the U.S. Nuclear Regulatory Commission State of Texas Washington, DC 20555 P.O. Box 12548 Capitol Station Dr. E. Leonard Cheatum Austin, Texas 78711 Route 3, Box 350A Watkinsville, Georgia 30677 Hon. Jerry Sliva, Mayor City of Wallis, Texas 77485 Mr. Gustave A. Linenberger
- Atomic Safety and Licensing Board Panel Hon. John R. Mikeska U.S. Nuclear Regulatory Commission Austin County Judge Washington, DC 20S55 P.O. Box 310 Bellville, Texas 77418 Mr. John F. Doherty
~~
4327 Alconbury Street Houston, Texas 77021 5
J. Gregory Copeland, Esq.
Mr. and Mrs. Robert S. Framson i
Baker & Botts 4822 Waynesboro Drive One Shell Plaza Houston, Texas 77035 Houston, Texas 77002 Mr. F. H. Potthoff, III Jack Newman, Esq.
7200 Shady Villa #110 Lowenstein, Reis, Newnan & Axelrad Houston, Texas 77055 1025 Connecticut Avenue, N.W.
Washington, DC 20037 D. Marrack 420 Mulberry Lane Carro Hinderstein Bellaire, Texas 77401
..e 8739 Link Terrace Houston, Texas 77025 e
s __
s'
.' Texas Public Interest Margaret Bishop Research Group, Inc.
11418 Dak Spring c/o James Scott, Jr., Esq.
Houston, Texas 77043 13935 Ivymount Sugarland, Texas 77478 Brenda A. McCorkle 6140 Darnell Houston, Texas 770/4 J. Morgan Bishop 11418 Oak Spring Mr. Wayne Rentfro Houston, Texas 77043 P.O. Box 1335 Rosenberg, Texas 77471 Stephen A. Doggett, Esq.
Pollan, Nicholson & Doggett Rosemary N. Lemmer P.O. Box 592 11423 Oak Spring Rosenberg, Texas 77471 Houston, Texas 77043 Bryan L. Baker
,1923 Hawthorne Houston, Texas 77098 Robin Griffith Leotis Johnston 1034 Sally Ann 1407 Scenic Ridge Rosenberg, Texas 77471 Houston, Texas 77043 Elinore P. Cummings Atomic Safety and Licensing
- 926 Horace Mann Appeal Board Rosenberg, Texas 77471 U.S. Nuclear Regulatory Commission Washington, DC 20555 Atomic Safety and Licensing
- Board Panel U.S. Nuclear Regulatory Commission Mr. William Perrenod Washington, DC 20555 4070 Merrick Houston, TX 77025 Docketing and Service Section
- Office of the Secretary Carolina Conn U.S. Nuclear Regulatory Commission 1.414 Scenic Ridge Washington, DC 20555 Houston, Texas 77043 Mr. William J. Schuessler U.S. Nuclear Regulatory Commission 5810 Darnell Region IV Houston, Texas 77074 Office of Inspection and Enforcement 611 Ryan Plaza Drive The Honorable Ron Waters Suite 1000 State Representativd, District 79 Arlington, Texas 76011 3620 Washington' Avenue, No. 362 Houston, TX 77007 d.
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Richard L. B44ck 1
Counsel for NRC Staff l
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