ML19290E042
| ML19290E042 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 02/15/1980 |
| From: | Moon C Office of Nuclear Reactor Regulation |
| To: | Doherty J DOHERTY, J.F. |
| References | |
| NUDOCS 8003040060 | |
| Download: ML19290E042 (21) | |
Text
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UtilTED STATES OF AMERICA 2/15/80 fiUCLEAR REGULATORY C0fC11SS10N BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
1:0VST0Tl LIGHTIfiG & POWER COMPANY
)
Docket flo. 50-4C6
)
(Allens Creek fluclear Generating
)
Station, Unit 1)
)
NRC STAFF RESP 0f1SE TO JOHN F. DOHERTY'S ELEVENT;! SET OF INTERROGATORIES The NRC Staff responds as follows to the eleventh set of interrogatories pro-pounded by John F. Joherty in this proceeding.
11-1.
Does Staff have any data on gap conductance uncertainty where linear power of the fuel rod is constant and the uncertainty is a function of fuel burn-up?
a/
If so, please cite such data.
P,esponse (A) We do not have any experimental data on gap conductance uncertainty where the linear power level of the rod is held constant and the uncertainty is a function of burn-up.
Tilis type of information Id be very desirable. However, the difficulties associated with maintain.-
constant local power level in a fuel rod over a a
long period of time have precluded experimental results in this area.
As an 8003040
example, the fiRC/PNL Halden Test Assembly IFA-431 was designed to assess the uncertainties in gap conductance and other fuel rod parameters.
Although this assembly had a design power of 328 W/cm (10 kW/ft), the data reportd shows large variation in the actual local linear heat ratings due to reactor scrams and other changing conditions.
Tne scram frequency of a commercial power plant is typically less than that indicated for the Halden reactor, which is a test facility. However, commercial reactor fuel is also shuffled during its exposure lifetime which leads to additional variations in linear power level.
Because of these realistic variations in the power level of a fuel rod, the gap conductance models used in plant safety analysis usually have the ability to follow a changing power history. Once these codes have been verified with experimental data from sources, such as IFA-431 mentioned above, they can be used to predict gap conductance for a hypothetical case involving no power changes.
Under these conditions, the relative uncertainty in gap conductance increases with time (burn-up). However, "the output uncertainties (fuel temperature, stored energy) decrease dramatically when the change is made from open gap to assumed fuel cladding contact."l/ Fuel cladding contact usually occurs in LWR S. R. Hann, E. R. dradley,ft. E. Cunningham, D. D. Lanning, R. K. fiarshall C
and R. E. Williford, " Data Report for the NRC/PfiL Halden Assembly IFA-431,"
Battelle Pacific Northwest Laboratories Report PNL-2494, April 1978.
S. E. Cunningham, d. D. Lanning, A. R. Olsen, R. E. Williford and C. R. Hann, M
" Stored Energy Calculation:
The State of the Art," Battelle Pacific Northwest Laboratories Report PNL-2581, May 1978.
fuels as elevated burn-up, so the relative uncertainty increases with time but so does the overall value of gap conductance.
(B) The documents are listed as referenced in 11-1 (A).
(C) No other references were reviewed for the answer.
(D)
J. C. Voglewede (USNRC).
(E) The Fuel Behavior Research Branch is supporting some effort in this area as part of the fuel performance code (FRAP) development at EG&G-Idaho. Other studies are being performed by Battelle-Northwest as part of the Halden program.
(F)
J. C. Voglewede of the Core Performance Branch will be available to testify on this matter.
11-2.
Does Staff have any data on BWR fuel rods that show the effect of burn-up between 15,000 mwd /t and 30,000 mwd /t on fuel rod integrity with energy deposition as the other variable, such as shown on page 24 of NUREG-0581?
a/ If so, please cite such data.
Response
(A) This interrogatory has been answered in response to Interrogatory 8-10.
Three references and a copy of a figure showing fuel ro? integrity as a function of energy insertion vs. burn-up were enclosed.
. 11-3.
Does Staff have any summary or cther report of a meeting between NRC Staff and General Electric personnel as part of a presentation on rod bowing, mentioned on page 4.2-14f(l) of the PSAR?
a/ If so, please cite such summaries or reports so that an F0IA request may be made, and please give its length, too.
Response
(A) We have enclosed a copy of the meeting summary of the General Electric Fuel Rod Bowing Meeting of November 13, 1974.
11-4.
Does Staff have any report or summary of an accident involving a foreign BWR where prolonged blowdown led to loss of integrity of the suppression pool by unstable steam condensation? (Note:
See Page 2 of Attachment to ACRS memorandum from S. H. Hannauer (NRC) to M. Carbon, (Chairperson of ACRS), July 6, 1979, NRC Accession number 7907200512.
a/ If so, please cite such report or sumary sufficiently that it may be obtained through an FOIA request and give its length if possible.
Response
See the response to 9-2.
11-5.
dVREG-0006, " Water Reactor Safety Research Program," on pg.123, states:
Present procedures for determining susceptibility to intergranular attack are given by ASTM standard A 262-70.
The tests described in th.is standard detect suscepti-bility in austenitic stainless steels to intergranular attack by highly corrosive chemical solutions in the absences of stress and have been found unreliable for application to BWRs. The required improvement is a test that will detect susceptibility to IGSCC in the half heat-affected zones of austenitic stainless steel welds for BWR service.
a/ Does Staff maintain the " required improvement" has been achieved?
b/
If not, what progress has been made in detecting susceptibility?
c/
Is Applicant currently required to meet only the ASTM standard stated here?
d/ Will there be any half heat-affected zones of austenitic stain-less steel welds in Applicant's recirculation risers or other piping? (D-44)
Response _
(A) a/ No.
b/ An acceptable in situ test has not yet been developed.
c/ Yes.
d/ Yes.
(B) " Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," !!UREG-0313, Rev.1, October 1979, page 9.
(C) None.
(D) Not applicable.
Answer provided by NUREG-0313, Rev.1.
(E) None except as described in NUREG-0006.
(F) Since the in situ test method under development is not required no testimony is planned.
. 11-6.
Is the combination of oxygen dissolved in coolant plus high stress plus sensitization of stainless steel that is most likely to produce IGSCC known at this time? (D-44)
Response
(A) Stress levels approaching yield and oxygen levels of about eight parts per million and greater are known to be necessary for IGSCC to occur in highly sensitit:j austenitic stainless steel.
The research referred to in response to 11-5 is aimed at better relating the degree of sensitization to the sus-ceptibility to IGSCC.
(B) " Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping of Boiling Water Reactor Plants," PB-246 645 (NUREG-75/067) October, 1975.
" Investigation and Evaluation of Stress-Corrosion Cracking in Piping of Light Water Reactor Plants," ilVREG-0531, February 1979.
"Investiga; on of Cause of Cracking in Austenitic Stainless Steel Pipes,"
NED0-21000, General Electric Company, July,1975.
" Prevention of Stress Corrosion by Limitation of Applied Static Load in BWR Piping and Components," ilED0-23684, General Electric Company, September 1978.
(C) None.
(D) Response is based on published reports.
(E) See response to 11-5 and documents listed under (B).
(F) None planned.
11-7.
Has General Electric submitted to the NRC any experimental data on Fission Gas Release (FGR) like that shown for the German PWR in Fig. 6 (page 20) of NUREG-0418, " Fission Gas Release from Fuel at High Burnup"?
(D-20)
Response
(A) Development and qualification of the current General Electric fission gas release modelS is described in a report by Hoffman and Coplin. 4/ The data presented were limited to exposures of less than 11,000 mwd /tU.
This burn-up range is significantly less than that given for the KWU Obrighein data in Figure 6 of NUREG-0418.
A revised fission gas release model__/ has been submitted by General Electric 5
for Staff review.
This model is based on data governing both the release fraction
_3_/"GEGAP-III: A Model for the Prediction of Pellet-Cladding Thermal Conductance in BWR Fuel Rods," General Electric Company Report NED0-20181, iiovember 1973.
d/. P. Hoffman and D. H. Coplin, "The Release of Fission Gases from Uranium J
Dioxide Fuel Operated at High Temperatures," General Electric Company Report GEAP-4596, September 1964.
5/. B. Johansson, G. A. Potts and R. A. Rand, "GESTR: A Model for the E
Prediction of GE BWR Fuel Rod Thermal / Mechanical Perfonnance," General Electric Company Report NED0-23785, March 1978.
and burn-up range of the Obrighein data described in NUREG-0418.
Review of the revised model is not yet completed and no judgment has been made on the applicability of the data presented.
(B) The documents are listed as referenced in 11-7 (A).
(C) No other references were reviewed for the answer.
(D)
J. C. Voglewede (USNRC).
(E) Other fission gas release programs and references were discussed in our reply to Interrogatory 10-5.
(F)
J. C. Voglewede of the Core Performance Branch will be available to testify on this matter.
11-8.
Will the five criteria listed in P.6 of NUREG/CR-0727* to give reasonable assurance that future off-gas detonations will be minimized be met by Applicant?
- Titled:
"BWR Off-gas System Evaluation" (T-33).
Response
(A) The capability of the off-gas system at Allens Creek Nuclear Generating Station to withstand and to mitigate the effects of a hydrogen explosion were evaluated in the Safety Evaluation Report (SER) for Allens Creek, November
\\
9_
1974, and in SER Supplement No. 2, March 1979. These documents indicated that the off-gas system has redundant hydrogen analyzer to indicate and alarm the presence of excess hydrogen levels and that it is designed to withstand a hydrogen explosion.
These provisions have the effect of mitigating the effects of an explosion internal to the system.
The criteria set forth in NUREG/CR-0727, "BWR Off-Gas Systems Evaluation,"
June 1979, include five criteria for preventing explosions external to the systems as a result of system venting or leakage of potentially explosive mixtures.
These five criteria provide specific detail as to the means for evaluating Acceptance Criteria II.3 of NRC Standard Review Plan 11.3 and therefore the Applicant's off-gas system will be required to meet them.
Three of the five criteria invol"e operatienal procedures for the off-gas system eind have not yet been reviewed; however, they will be reviewed at the time of the submittal for an operating license.
Two of the five criteria involve system design to prevent leakage paths from the system, to adequately ventilate surrounding areas, and to monitor off-normal conditions in the system. A review of the PSAR indicatas that liquid seals in the system are vented to the main condenser where the addition of a stream of gas would not substantially increase the probability of an internal explosion or burn.
In addition, there are monitors and alarms indicating low dilution steam flow, high hydrogen analyzer level, and high and low temperature in the recombincr.
Examination of the preliminary design of the ventilation system (Table 9.4-11 and Figure 9-4.4 of the PSAR) indicates no barrier to achieving compliance with the ventilation requirements of the guide in the final design.
Thus, the system can be designed to meet the design criteria of NUREG/CR-0727. A more detailed review of all five criteria will be completed at the time of the submittal for an operating license.
(C) See (A).
(C) None.
(D)
Frank Cardile of the Effluent Treatment Systems Branch.
(E) None planned.
(F) None planned at this time.
11-9.
NUREG-0006, " Water Reactor Safety Research Program" 2/79, on Page 60 states:
Most of the reliable experiments (on measuring gap conduct-ance)... utilized small (6200g) diametrical gaps.
There is little well-characterized data for thermal reactor fuel... in the 33-to 50-kW/m (10- to 15-kW/ft) operating power range.
a/ Is there any data you can cite which is an improvement on the current state of research on this topic for the operating power range mentioned?
b/ If so, summarize it very briefly please.
Response
(A)
Yes, there are data which result in an improvement in our understanding of gap conductance for the conditions mentioned. As an example, a six-rod test assembly designated IFA-432 is currently being irradiated in the Halden Boiling Water Reactor. This assembly is well-characterized and heavily instrumented.
It was built and is being irradiated by Pacific Northwest Laboratories under the sponsorship of the NRC. The assembly was designed to operate in the linear power range of 35-50 kW/m (10.7-15.2 kW/ft) ar.d has initial diametral gap of 75 to 380j4m. The diaretral gap values may be compared to that of a typical BWR fuel pin, which is approximately 230 m.
This test assembly is still being irradiated and is not due to be discharged until early in 1981. However, data taken from the experiment to date are available.S The data continue to support the concept of gap conductance as a complex function of pellet cracking and relocation, pellet swelling, thermal expansion, and cladding creepdown.
The change in gap conductance due to such operating variables as rate of power increase, number of power cycles, and power level is described by Lanning, Barnes and Williford.-- /
S. R. Hann, E. R. Bradley, M. E. Cunningham, D. D. Lanning, R. K. Marshall, C
and R. E. Williford, " Data Report for the NRC/PNL Halden Assembly IFA-432,"
U.S. Nuclear Regulatory Commission Report NUREG/CR-0560 (PNL-2073), August 1978.
S. D. Lanning, B. O. Barnes, and R. E. Williford, " Manifestations of Non-D linearity in Fuel Center Thermocouple Steady-State and Transient Data:
Implications for Data Analysis," U.S. Nuclear Regulatory Commission Report NUREG/CR-0220 (PNL-2692), January 1979.
(B) The documents are listed as referenced in 11-9 (A).
(C) Other gap conductance references were provided in our response to Interrogatory 8-12.
(D)
J. C. Voglewede (USNRC).
(E) The response to ll-9(A) describes the future work in this area.
(F)
J. C. Voglewede of the Core Performance Branch will be available to testify on this matter.
11-10.
Does the change in Applicant's steel shell containment to a hemispherical from semi-elipsoidal shape reflect any staff position, requirement or other change?
Response
(A) The change does not reflect any Staff position or other change. The choice of configuration is the designer's prerogative as long as the design criteria are satisfied.
It is the Staff's expectation that this change will enhance the stability at the dome-cylinder junction.
(B) 10 C.F.R. Part 50 does not specify containment dome configuration.
(C) None.
. (D) As indicated under (A) and (B), the response is based on the content of the Comission's regulations.
(E) None directed at the question of dome configuration.
The Staff is updating its review of design criteria to assure that no changes of criteria were effected by Amendment No. 54, or that any changes made are acceptable.
(F) The fact that the configuration changed does not require testimony.
QUESTIONS REMAINING FROM SEVENTH SET 7-6.
Relative to Doherty Contention #3, on p.ge 37 of the Enclosure of the 2/15/79 letter from R. Mattson (NRC) to G. G. Sherwood (G.E.), it says, "PCI failures are considered to be more likely to occur during power-increasing than reduction-in-flow events because during the former the fuel pellets heat up and expand more rapidly than the cladding whereas in the latter type of event the opposite thermal expansion effect occurs."
Would this analysis only he'd true for rods which essentially not operating?
(D-3)
Response
Tne interrogatory as stated is hard to understand.
What is meant by a " rod essentially not operating"? Perhaps the discussions to follow for 7-7, 7-8 and 7-9 will sufficiently cover the intended concern.
If not, please provide further clarification of the question.
7-7.
Why (in the above) would the cladding expand more in all rods in a reduction-in-flow event, thus reducing PCI effects?
(D-3)
Response
Tne NRC statement concerning PCI failures during a reduction-in-flow event might better have been worded to reflect the absence of PCI during these events.
In the Allens Creek PSAR, four reduction-in-flow events are analyzed:
(1) trip of one recirculation pump (Fig. 15.1.21-1),(2) trip of two recirculation pump notors (Fig. 15.1.22-1),(3) seizure of one recirculation pump (15.1.23-1) and (4) flow controller failure with decreasing flow (Fig. 'c 1.24-1).
In all of these analyses, the BWF core parameters respond in t.m same manner:
( l '.:.e neutron flux drops rapi.!1y in the first few seconds and except for the one pump trip event, is reduced by 95% within 10 seconds,(2) the fuel center temperature follows the general trend of the neutron flux and with the noted exception is diminished by approximately 90% within 30 seconds, and (3) the average surface heat flux follows the same relationship as the fuel temperature.
In the one exception (trip of one pump), all parameters are reduced within 10 seconds but then stabilize at about the 60% level.
There are no rapid power or temperature increases in any of these events.
Under such conditions, PCI would not be expected.
Cladding temperature is a function of the average heat flux and the cooling capability of the coolant. The above analyses show the average heat flux is reduced rapidly.
The cladding temperature would also oe expected to diminish in the same time frame.
Unless the coolant conditions lead to departure from
. nucleate boiling (DNB), the cladding temperature would be expected to remain within some increment (50 F) above the coolant temperature.
If DNB occurs, the cladding temperature will increase.
For failure calculations, a fuel rod in DNB is assumed to fail although research results obtained at PBF-Idaho show that this is not always the case.
As discussed above, the components of the fuel rod will be cooling and contracting during a reduction-in-flow event.
In relative terms, the fuel will cool more than the cladding.
Thus, the fuel is moving away from the cladding.
If DNB occurs, the cladding temperature will increase and would expand away from the fuel.
In either case, PCI would not be expected.
7-8.
Referring to 7-6 isn't there really only a less severe PCI in the reduction-in-flow event? That is isn't the fuel going to swell faster than the clad in either type accident?
(D-3) (0-20)
Response
No. For the reasons given in answer to Interrogatory 7-7 above, PCI is not only less severe in the reduction-in-flow event but it is not expected to occur at all.
7-9.
What data support the idea the reduction-in-flow accident will lead to clad getting farther from the pellets? (D-3) (D-20)
Response
The data that support the concept of fuel behavior during a reduction-in-flow event are presented in the PSAR and copies of the analyses for these events are enclosed.
All PCI studies reported to date are based on power increases.
. 7-10.
Referring to Page 43 of the Enclosure mentioned in Int. 7-6, (supra.)
what is meant by, "For all BWR plants, including... EMG, the suppression pool temperature limit remains an open item for ATWS?
(D-5,D-8) a.
Has applicant been given any guidance on how to deal with this "open" item?
b.
Is applicant's design required to keep this limit below 200 F?
Is Staff considering increasing the number of SRVs to more evenly c.
distribute the steam to the suppression pool during an ATWS?
(D-5) d.
May the local temperature exceed 200 F?
e.
If the answer to "d" is "yes," by how many degrees?
f.
If the answer to "d" is "yes," for how large an area may the pool temperature exceed 200 F?
Response
See the response to 9-1.
7-11.
Is the NRC or a contractor currently studying the problem of adequacy of relief valves when steam-water in supercritical state flow through said valves? (D-17)
Response
By letter of October 10, 1979, the Staff stated its requirement that Houston Lighting & Power Company implement a relief and safety valve testing program.
In its response by letter dated November 14, 1979, the Applicant committed "to incorporating in the ACNGS design valves verified by the resolution arrived at by ongoing industry /NRC activities."
By letter of November 9,1979, the Staff provided Houston Lighting & Power Company with the following clarification of Item 2.1.2 in its Oc:ober 10, 1979 letter:
PERFORMANCE _ TESTING FOR BWR AND PWR RELIEF AND SAFETY VALVES (l[~1.2)
POSITION Pressurized Water Reactor and Boiling Water Reactor licensees and applicants shall conduct testing to qualify the reactor coolant system relief and safety valves under expected operating conditions for design basis transients and accidents.
CLARIFICATION 1.
Expected operating conditions can be determined through the use of analysis of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70.
2.
This testing is intended to demonstrate valve operability under various flow conditions, that is, the ability of the valve to open and shut under the various flow conditions should be demonstrated.
3.
Not all valves on all plants are required to be tested.
The valve testing may be conducted on a prototypical basis.
4.
The effect of piping on valve operability should be included in the test conditions. Not every piping configuration is required to be tested, but the configurations that are tested should produce the appropriate feedback effects as seen by the relief or safety valve.
5.
Test data should include data that would permit an evaluation of discharge piping and supports if those components are not tested directly.
6.
A description of the test program and the schedule for testing should be submitted by January 1,1980.
7.
Testing shall be complete by July 1, 1981.
The results of the Staff's review of the Applicant's letter of November 14, 1979, and of the test program description, if available, will be included in a supplement to the Safety Evaluation Report. As noted in the clarification (Item 1),the expected operating conditions are to be determined for incorporation in the test program.
In view of the Applicant's commitment it is not anticipated that there will be a need for testimony in addition to the Safety Evaluation Report.
Dated at Bethesda, Maryland, this 15th day of February, 1980.
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)
)
AFFIDAVIT OF CALVIN W. MOON I hereby depose and say under oath that the foregoing NRC Staff responses to interrogatories propounded by John F. Doherty 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.
N Acjw Calvin W. Moon Subscribed and sworn to before me this/ erd day of A / : " < j 1980.
/
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Notary Public My Commission expires:
be / 4 6 Nf 2 u
v
UNITED STATES OF AMERICA NUCLEAR REGULATURY 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 RESPONSE TO JOHN F. DOHERTY'S ELEVENTH SET OF INTERR0GATORIES" and " AFFIDAVIT OF CALVIN W. M0ON" in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class, or, as indicated by an asterisk by deposit in the Nuclaar Regulatory Commission internal mail system, this 15th day of Febru6ry 1960:
g 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 R. Gordon Gooch, Esq.
Baker & Botts Mr. John F. Doherty 1701 Pennsylvania Avenue, N.W.
4327 Alconbury Street Washington, DC 20006 Houston, Texas 77021 J. Gregory Copeland, Esq.
Mr. and Mrs. Robert S. Framson Baker & Botts 4822 Waynesboro Drive One Shell Plaza Houston, Texas 77035 Houston, Texas 77002 Mr. F. H. Potthoff, III Jack Newman, Esq.
1814 Pine Village Lowenstein, Reis, Newman & Axelrad Houston, Texas 77080 1025 Connecticut Avenue, N.W.
Washington, DC 20037 D. Marrack 420 Mulberry Lane Carro Hinderstein Bellaire, Texas 77401 8739 Link Terrace Houston, Texas 77025
Texas Public Interest Margaret Bishop Research Group, Inc.
11418 Oak Spring c/o James Scott, Jr., Esq.
Houston, Texas 77043 8302 Albacore Houston, Texas 77074 Glen Van Slyke 1739 Marshall Brenda A. McCorkle Houston, Texas 77098 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.
Polian, Nicholson & Doggett Rosemary N. Lemmer P.O. Box 592 11423 Oak Spring Rosenberg, Texas 77471 Houston, Texas 77043 Bryan L. Baker Charles Andrew Perez
,1118 Montrose 1014 Montrose Blvd.
Houston, Texas 77019 Houston, Texas 77019 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 Mrs. Connie Wilson 11427 Oak Spring Atomic Safety and Licensing
- Houston, Texas 77043 Board Panel U.S. Nuclear Regulatory Commission Washington, DC 20555 Docketing and Service Section
- Office of the Secretary Carolina Conn U.S. Nuclear Regulatory Commission 1414 Scenic Ridge Washington, DC 20555 Houston, Texas 77043 Mr. William J. Schuessler Mr. Robert Alexander 5810 Darnell 10925 Briar Forest #1056 Houston, Texas 77074 Houston, TX 77042 m
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~Stbphef M. Schinki
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Counsd for NRC Staff