ML19261D706
| ML19261D706 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 05/07/1979 |
| From: | Gay G WEST TEXAS LEGAL SERVICES |
| To: | |
| References | |
| NUDOCS 7906260084 | |
| Download: ML19261D706 (36) | |
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UNITED STATES OF AMERICA ome
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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD IN THE MATTER OF S
I TEXAS UTILITIES GENERATING COMPANY, S
DOCKET NOS. 50-445 l
(Comanche Peak Steam Electric
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t Station, Units 1 and 2)
S SUPPLEMENTAL PETITION AND CONTENTIONS OF INTERVENORS, ACORN, MARY AND CLYDE BISHOP AND ODA AND WILLIAM WOOD i
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NUCLEAR REGULATORY COMMISSION side
- BEFORE THE ATOMIC SAFETY AND LICENSING BOARD Ng IN THE MATTER OF S
TEXAS UTILITIES GENERATING COMPANY, S
DOCKET NOS. 50-445 ET AL 50-446 S
(Comanche Peak Steam Electric Station, Units 1 and 2)
S SUPPLEMENTAL PETITION AND CONTENTIONS OF INTERVENORS, ACORN, MARY AND CLYDE BISHOP AND ODA AND WILLIAM WOOD Intervenors, ACORN, MARY and CLYDE BISHOP, and ODA and WILLIAM WOOD, by and through their attorneys of WEST TEXAS LEGAL SERVICES, pursuant to 10 C.F.R.
- 52. 714 (b), respectfully submit this supplement to their amended petitions to list and provide the bases for contentions which they seek to have litigated in this matter.
Nothing contained herein is to be deemed a waiver of Intervencrs Motion for Continuance of this matter, and irrespective of the final ruling on that Motion, Intervenors request the right to amend or supplement this filing with contentions that arise out of investigations of the Three Mile Island accident or the nuclear industry and the regulatory process.
Intervenors may, after conferring with staff, request leave to supplement or amend this filing to refine wording or add greater specificity.
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e l.
SAFETY CONTENTIONS As a basis for issuing an operating license for the Comanche Peak plants, NRC regulations require that the following findings be made:
The facility will operate in conformity with the Application as amended, the provisions of the Act, and the rules and regulations of the Commission.
There is reasonable assurance (i) that the activities authorized by the operating license can be conducted with-out endangering the health and safety of the public and (ii) that such activities will be conducted in compliance with the regulations in thi.s chapter.
The issuance of the license will not be inimical to the common defense and security and to the health and safety of the public.
Because of the following safety problems, these findings cannot be made in this case, and therefore, Comanche Peak should not be licensed for operation.
2.
SAFETY CONTENTION ONE The Comanche Peak design fails to adequately account for the ef fect of asymmetric loading resulting from a pipe break in the area between the reactor vessel and the shield wall.
3.
EXPLANATION A pipe break in certain locations between the vessel and the shield wall would cause instantaneous extreme pressure differentials, causing forces which could tip the vessel, 2312 342
shearing the pipes and preventing cooling.
In addition, these forces could damage the fuel spacer grids and distort the fuel geometry.
The failure of present PWR designs to adequately consider the effects of asymmetric loading was not identified in the Westinghouse design.
A detailed description of the problem can be found in Appendix A, Task A-2 of the staff testimony in the Black Fox proceeding entitled " Testimony of Michael B. Aycock, Laurence P.
Crocker and Cecil O.
Thomas, Jr.
Relating to the Status of NRC Staff Activities Regarding Generic Safety Issues."1/
It is acknowledged by the staff in both NUREG-0410 and NUREG-0510 to fall within " Category A" which is defined to include:
Those generic technical activities judged by the staff to warrant priority attention in terms of manpower and/or funds to attain early resolution.
These matters include those the resolution of which could (1) provide a significant increase in assurance of the health and safety of the public, or (2) have a significant impact upon the reactor licensing process.
1/
Public Service Co. of Oklahoma, et al (Black Fox Station, Units 1 and 2).
Docket Mos. STN 50-556, 50-557 (hereinafter Black Fox Testimony).
The appendix consists of Revision 1 of all Category A Task Action Plans, which originally appeared in NUREG-0410, "NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants," Jan.
1, 1978 (hereinafter entitled MUREG-0510, " Identification of Unresolved Safety Issues Relating to Nuclear Power Plants," (hereinaf ter NUREG-0510), will be referenced often since they contain detailed explanations of'many of the safety problems which the N.9 staff acknowledges to be unresolved.
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4.
S AFETY CONTENTION TWO NRC staff review is inadequate to identify and correct modes of interaction between reactor systems in the Comanche Peak design which can adversely affect the redundance or independence of safety systems.
5.
EXPLANATION As discussed in Task A-17 of the Black Fox Testimony, the present system for staff review divides responsibility among staff units organized by plant systems, such as electrical systems, mechanical systems, etc.
There is no systematic method for review of the unintentional interaction between such systems, although such interactions can be very dangerous.
System interaction is listed as Category A in both NUREG-0410 and NUREG-0510 and is also classified in Appendix C of NUREG-0510 as a " Potential High Risk Item."
Perhaps the classic example of systems interaction, and one which applies almost uniquely to the Westinghouse design was revealed in an incident at the Zion plant in July, 1977.
The Westinghouse design is characterized by the large number and type of interaction between reactor control systems and protection, or safety systems.
At Zion, a series of dummy signals used for testing disabled shared sensors for both protection and control. As a result, the water level in the core was lowered, and at the same time the protection against the accident was disabled.
In other words, the same sequence of events both " caused the transient and paralyzed the safety 23l2.44
provided for that very transient. " !
As an additional example, in the recent Three Mile Island accident, poorly understood interaction between non-safety and safety systems led or contributed to the releases of radioactive water from the containment to the auxiliary building, from which it reached the public.
6.
SAFETY CONTENTION THREE Comanche Peak is not in compliance with the require-ments of General Design Criterion 4 of Appendix A to 10 C.F.R. Part 50, " Environmental and Missile Design Bases. "
Specifically, neither the staff nor the applicant has a reliable method for evaluating or ensuring that Class IE safety-related equip-ment is qualified to withstand and operate in the environmental conditions of the most severe postulated accident.
7.
EXPLANATION GDC 4 requires generally that equipment required to operate in the event of an accident (Class IE safety-related equipment or components) can be qualified to withstand the environment of the accident.
Regulatory Guide 1.89 provides that compliance with IEEE Standard 323-1974 will be one acceptable method of demonstrating compliance with GDC 4.
2/
Memorandum from Stephen H.
Hanauer to E.G.
Case, " Inter-action Between Control System and Protection System," August 18, 1977.
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There are essentially two problems with this insofar as Comanche Peak is concerned.
First, the staff requires conformance to IEEE-323-1974 only for plants for which an SER was issued after July 1,1974.
Although TUGCO made some commitment to meet the 1974 standard, it is not clear, part-icularly with respect to aging, that TUGCO has committed to all of the proper standards.
In addition, even if TUGCO did commit to meet all of IEEE-323-1974, neither the nuclear industry nor the NRC staff has yet established methods in many areas to demonstrate equipment qualification.
Such areas include testing margins, aging effect; on materials and equipment and the adequacy of testing procedures to simulate the worst case environment for the equipment.
In other words, neither the applicant nor the staff adequately understands either the environment which the equipment will be exposed to or its ability to withstand that environment.
This problem is discussed as Task A-24 in the Black Fox Testimony and is still classified in NUREG-0510 as a Class A unresolved safety issue and a potential high risk item.
8.
SAFETY CONTENTION FOUR Comanche Peak is not in compliance with the requirements of General Design Criterion 2 of Appendix A to 10 C.F.R. Part 50,
" Design Bases for Protection Against Natural Phenomena."
Specifically, neither the staff nor the applicant have reliable methods for evaluating and ensuring that structures, systems and components important to safety are designed to withstand 2312 ;46
7.
the ef fects of the appropriately severe earthquake without losing its capability to perform its safety functions.
9.
EXPLANATION The seismic design process used by the staff requires, in brief, identification of the magnitude of the safe shutdown earthquake, determination of Ehe resultant ground motion, building motion and plant equipment motion, and comparison of the resultant loads on safety equipment with the allowable loads.
The NRC staff admits that it can not presently assure that the methods it allows to be used for analysing seismic design result in adequately conservative plant design.
(See Black Fox Testimony, Task A-40).
This unresolved safety problem is also still Category A and a potential high risk item, as classified in NUREG-0510.
For example, the effects of aging and cumulative radiation on the ability of electrical equipment to withstand seismic stresses is not considered.
Batteries are particularly vulnerable since old plates crumble and get weak and spongy.
Yet no tests have been done to determine whether batteries in use for a period of time would still be able to withstand an earthquake.
Nor is the effect of aftershocks considered in the seismic analysis, although such aftershock, even though of less severity than the original quake, could result in additional damage to structures already deformed from the first shock.
Finally, the recent NRC Order requiring the shutdown 2312 347
of five operating plants because of a gross error in seismic load calculations clearly illuminated a grave failure in the system for verifying the adequacy of seismic design.
A large number of complex, proprietary computer codes are used by the Architect-Engineers to compute seismic stresses, yet the NRC staff makes no effort to systematically review the accuracy or validity of the codes.
The following quote from Harold Denton, Director of the Division of Nuclear Reactor Regulation, at an NRC meeting on March 13, 1979, will suffice to make this point:
" Stone and Webster said we will use this code, we said fine, that sounds like an accredited kind of code.
We did not look behind it...."
10.
SAFETY CONTENTION FIVE Comanche Peak does not comply with the requirements of Gencral Design Criterion 3 of Appendix A to 10 C.F.R. Part 50,
" Fire Protection."
Specifically, present fire protection requiremente are not adequate to prevent a fire from disabling the electrical cables for all redundant safety systems.
11.
EXPLANATION In the fire at the Browns Ferry plant in Alabama in 1975, a fire in the cable spreading room destroyed at once the cables for both sets of redundant safety systems.
Since that time, the staff has made some improvements in such areas of administrative controls and control of ignition sources.
However, tests conducted as part of the Fire Protection Research Program have disclosed that the 5-foot ph/cical 2312 548 u
9.
separation requirement of Regulatory Guide 1.75 is still inadequate to prevent the spread of fire from one set of cables to the other.
In addition, tests on mineral wool blankets proposed for fire retardants have shown them to act as wicks in some cases, sprinkler systems have failed, and at least some " fire-retardant" cable coatings have burned.
The fact is that present regulatory practice does not provide the protection against fires for safety systems required by GDC 3.
12.
SAFETY CONTENTION SIX The D. C. power system for the Comanche Peak plant fails to meet the single failure criterion as defined in 10 C.F.R. Part 50 Appendix A.
13.
EXPLANATION Present practice permits a D.C. power supply with two separate divisions which supply power for control and actuation of safety-related systems.
Failure of one D.C.
division of a two-division redundant system causes the plant to shut down and creates the need for the other division to cool off the reactor, destroying the redundancy required by the single failure criterion.
This unresolved safety problem is discussed as Task A-30 in the Black Fox Testimony and NUREG-0410.
14.
SAFETY CONTENTION SEVEN The Comanche Peak design does not provide adequate reliable instrumentation to monitor variables and systems 2312 549
10.
affecting the integrity of the reactor core, the pressure boundary or the containment after an accident, in violation of general Design Criterion 13 of Appendix A of 10 C.F.R. Part 50.
15.
EXPLANATION The accident at Th'ree Mile Island demonstrated graphically the inadequacy of post-accident monitoring, in terms of the parameters monitored, the range and accuracy of the instrumentation, and the qualification of the instrumentation for the accident and post-accident environment.
For example, there is no way to directly measure the water level or ten.perature in the core after an accident.
The only temperature measurements at TMI were from non-safety grade equipment, some of which
" luckily" survived the accident.
It is also possible that failure to believe the instrumentation which was available after the TMI accident exacerbated conditions there.
Even before Three Mile Island, the lack of criteria or guidances to implement the Regulatory Guide on this subject, 1.97, was listed as a Category A unresolved safety problem in NUREG-0410 and the Black Fox case (Task A-34).
This January, in NUREG-0510, it was classified as "Not Directly Relevant to Risk. "
We suspect that this assessment will change in the aftermath of Three Mile Island.
16.
SAFETY CONTENTION EIGHT The Comanche Peak design does not adequately account for failure of passive components in fluid systems important to safety.l} }
it. 17. EXPLANATION The single failure criterion is a benchmark of nuclear power regulation. It is defined in 10C.F.R.Part 50, Appendix A. Basically, it holds that, in assessing whether a system important to safety is adequately redundant, one assumes first the occurrence of an accident and all of the failures caused by the accident. Then, one assumes an additional, unrelated failure - the single failure - and the system must still be able to perform its safety function. The NRC has always recognized that the single failure could occur either in an active or passive component. (An active component is one which requires motion to perform its safety function. A passive component does not, such as a valve which is supposed to be open both during normal and emergency situations). However, the staff has not yet devised a way to apply the single failure criterion to passive components. Thus, " consideration of the need to design against single failures of passive components in system important to safety" is listed as one of those matters which, although the design requirements are not yet suitably defined, must still be considered by applicants. (Introduction, 10 C.F.R. Part 50, Appendix A). However, the fact is that this has not been adequately considered either by applicants or staff. It is an open issue in RESAR-3, the design for this plant. In addition, certain components are incorrectly classified as passive. Any component which is capable of mechanical 2312 351
12. motion should be classified as an active component, rather than only components which must move in order to perform their safety function. 18. SAFETY CONTENTION NINE The Comanche Peak design does not provide adequate equipment outside of the control room to promptly put the reactor in hot shutdown and so maintain it until attaining cold shutdown, also from outside the control room, as required by General Design Criterion 19 of Appendix A to 10 C.F.R. Part 50. 19. EXPLANATION GDC requires that equipment be provided outside the control roon, such that if an accident caused the operators to evacuate the control room, the plant could be brought to safe shutdown from the outside. However, in designing and evaluating the plant, both the applicant and the staff have used the unreasonable assumption that whatever event caused evacuation from the control room did not damage any of the equipment in the control room. The result of this is that if the event, such as a fire for example, did damage equipment in the control room, the plant could well lose shutdown capability. 20. SAFETY CONTENTION TEN Neither the applicants nor the staff have adequately considered the ef fects of aging and cumulative radiation on safety-related equipment which must be seismically and envircn-mentally qualified. 2312 352 w m
13. 21. EXPLANATION Structures, systems and components important to safety must be qualified to demonstrate their ability to with-stand natural forces such as earthquakes and the accident environment and still perform their safety functions. In analyzing the ability of equipment to survive, insufficient account is taken of the affect of aging, which can progressively weaken components. Brand new equipment is tested and no systematic effort made to determine for how long the results are valid. This, although a plant is essentially certified to be safe for its lifetime when it receives an operating license, the Commission cannot say with reasonable assurance that sufficient margin exists to maintain equipment qualification over 30-40 years. 22. SAFETY CONTENTION ELEVEN The Comanche Peak design fails to address the possi-bility of a Class Nine accident. 23. EXPLANATION The environmental report submitted by the Applicant reveals that the plant is designed to withstand Class Eight acci-dents (Section 7.1.8), but that the possibility of a Class Nine accident is considered too remote to address (Section 7.1. 9). The Three Mile Island accident demonstrates the possibility of successive failures of plant systems and a possibility of danger to public health and safety with the occurrence of a Class Nine accident. 2312 353
14. 24. SAFETY CONTENTION TWELVE The Comanche Peak facility is not designed to accomodate a total loss of AC power. 25. EXPLANATION This serious and unresolved safety problem is discussed in Task A-35 of NUREG-0410, and Applicant's FSAR fails to adequately address the safety implications from a total loss of AC power. 26. SAFETY CONTENTION THIRTEEN Applicant lacks the ability to detect inadequately sized flaws within the reactor vessel and pipes within the contain-ment units. 27. EXPLANATION The ability to detect inadequately sized flaws is essential for the assessment of the margin against failure under various plant conditions throughout the full life of Comanche Peak. Without the ability to detect inadequately sized flaws, Applicant cannot guarantee that reactor coolant pressure can remain within safe bounds. This generic problem is discussed in Task A-14 of NUREG-0410. 28. SAFETY CONTENTION FOURTEEN Applicant's FSAR fails to present a means for dealing with pressure transien.; produced by enmponent failure, personnel error, or spurious valve actuation which exceed the pressure temperature limits of the reactor vessel. 2312 J54
15. 29. EXPLANATION This is a generic safety problem for pressurized water reactors presented in Task A-26 of NUREG-0410. As noted by the staff in the Black Fox Testimony previously cited, since 1972 "There have been over thirty reported incidents or pressure transients in pressurized water reactors which exceeded the pressure temperature limits of the reactor vessels involved". 30. SAFETY CONTENTION FIFTEEN The Comanche Peak design fails to adequatcly address the possibility of snubber support malfunctions which is a generic unresolved safety problem according to the NRC staff in Task A-13 of NUREG-0410. 31. SAFETY CONTENTION SIXTEEN The Comanche Peak containmnet units are structurally deficient for the safe operation of the plant. 32. EXPLANATION According to at least one worker " honeycombs" occurred immediately after concrete pours and in sufficient quantity to make 23l2 55
16. the containment units unsafe. " Honeycombs" represent breaks and cracks and improper bonding of concrete. Additionally, a fire occurred on the concrete in Containment Unit Number One while that concrete was in a curing stage. While conrete is curing, it should maintain a temperature of between 50 and 80 degrees. The high temperatures resulting from the fire make the containment structure unsafe. 33. SAFETY CONTENTION SEVENTEEN Walls of a Seismic One category in the control room are unstable and unsafe. 34. EXPLANATION According to workers who constructed walls to the control room, the mortar blocks used during construction were made from mortar that was too old, and when those blocks were put in place they were not properly bonded together and metal wire was not properly placed between every two layers of blocks. The affidavits of workers which further explain and support Contention Sixteen and Contention Seventeen are available to staff inspection upon request. 35. SAFETY CONTENTION EIGHTEEN Construction of the Comanche Peak facility has been - marred by a lack of observance of quality assurance-quality control. 36. EXPLANATION The inspection and enforcement reports document a 2312 356
17. history of problems with welding, and concrete pours at the Comanche Peak site. Falsification of documents took place under the super-vision of Robert W. Hunt Company, and according to recent investi-gative news stories appearing in the Fort Worth Star Telegram falsification of documents and failure to perform core samples tests continues at the Comanche Peak site. A worker who provided information to Intervenors stated that in the four years that he worked at the Comanche Peak site he spent most of his time working around mortar and concrete and that the mortar he worked around never had a QA-QC test performed on it. 37. SAFETY CONTENTION NINETEEN Brown and Root has failed to adequately supervise and guarantee the safe construction of the Comanche Peak facility. 38. EXPLANATION Brown and Root has failed to observe quality assurance-quality control. Concrete too old or too dry for proper bonding was poured in containment units under the direction of Brown and Root supervisors. Breaks and cracks and holes in the concrete in the containment units were covered up by Brown and Root workers under the direction of Brown and Root supervisors who did not properly report structural deficiencies. Mortar Slocks were not properly bonded together and metal wire was not placed between the layers of blocks upon the direction of Brown and Root supervisors. Brown and Root terminated the employment of individuals who ccmplained about incompetence and failure to meet safety regulations. Brown and Root hired unskilled laborers to perform very skilled tasks such as the inspection of welding. 2312 357
18. Affidavits of workers to support this contention are available to staff upon request. 39. SAFETY CONTENTION TWENTY The Comanche' Peak design fails to protect against corrosion which causes a cracking of pipes and a leakage of radio-active water. 40. EXPLANATION The Westinghouse pressurized water reactor is particularly plagued by a chemical reaction involving acids, chlorine and salt which causes a corrosive buildup between pipes and metal fittings which leads to a buildup of pressure and a leakage of radioactive water. The leakage of radioactive water poses a danger in and of itself, but there is also the potential that the plant's cooling system could be impaired with such leakage. NRC staff members Brian Grimes and Darrell Eisenhut recently repo.ted on this problem which has caused extensive pipe damage in at least five nuclear plants and moderate damage in at least ten other plants. The possibility of degradation to the integrity of steam generating tubes due to corrosion induced wastage and vibration induced cracking is discussed in Task A-3 of NUREG-0410. 41. SAFETY CONTENTION TWENTY-ONE The Comanche Peak design is not adequately insured against a water hammer problem which could affect a number of critical safety components. 2312 358 4 i
19. 42. EXPLANATION There exists a serious water hammer problem affecting all pressurized water reactors and involvi:.3 some critical safety components. These unresolved safety related problems have been identified by the NRC staff as high priority matters in both Task A-1 of NUREG-0410 and the Black Fox Testimony of staff witnesses M.B. Aycock, L.P. Crocker, and C.O. Thomas, Jr. 43. SAFETY CONTENTION TWENTY-TWO The Comanche Peak design fails to adequately protect against pipe breaks. 44. EXPLANATION This is a generic unresolved problem with serious safety implications and is discussed by the NRC staff in Task A-18 of NUREG-0410 and the Black Fox Testimony as having high priority. Applicant has the duty to insure against pipe breaks to prevent pipe whip and loss of coolant. There can be no assurance that Comanche Peak is safe for operation in the absence of adequate design criteria for the postulation of pipe breaks and the protection the re f rom. 45. SAFETY CONTENTION TWENTY-THREE The Comanche Peak design does not adequately address the possibility of a steam line break inside containment, nor does it insure the ability of equipment within containment to survive such an event so as to assure safe shutdown of the plant. 2312 359
20. 46. EXPLANATION N Existing main steam line break analyses, both inside and outside containment are seriously inadequate. NRC staff has recognized the seriousness of these problems and addressed them as high priority unresolved safety problems in Tasks A-21 and A-22 of NUREG-1040. This is a generic safety problem affecting all pressurized water reactors of the Comanche Peak type. 47. SAFETY CONTENTION TWENTY-FOUR The Comanche Peak design does not adequately insure the reliable operation of emergency power within the plant. 48. EXPLANATION The NRC staff has recognized that a generic unresolved safety problem arises from the unreliability of emergency on-site diesel generators at pressurized water reactors of the Comanche Peak type. That problem is addressed in Task B-56 of NUREG-0410. Present practice with regard to facilities of the Comanche Peak type permit the connection of non-safety loads in addition to the required safety loads to Class IE power sources so as to significantly affect the reliability of those power sources which are essential to the emergency power within the plant. That generic problem was given high priority by the NRC staff in Task A-25 of NUREG-0410. 49. SAFETY CONTENTION TWENTY-FIVE The Comanche Peak design has not adequately resolved a generic safety. problem for pressurized water reacrors wherein 2312 360
the steam generator and reactor coolant pump support materials are subject to lamellar tearing due to low fracture toughness. 50. EXPLANATION This matter was identified as a high priority safety problem applicable to reactors of the Comanche Peak type by the NRC staff in Task A-12 of NUREG-0410. 51. SAFETY CONTENTION TWENTY-SIX The Comanche Peak design does not adequately insure that safety related water supplies will be available for plant operation in the event of extreme cold weather and ice buildup. 52. EXPLANATION Ice storms have been known to occur in the Dallas-Fort Worth area. This past winter's ice storm incapacitated certain lignite plants of Texas Utilities by freezing water resources. An ice buildup at Comanche Peak could have disastrous consequences for the safe and effective operation of the plant cooling system. 53. SAFETY CONTENTION TWENTY-SEVEN The Comanche Peak design has not given due consideration to sabotage. 54. EXPLANATION There is no assurance that the Comanche Peak reactors and other critical portions of the plant are adequately protected against sabotage. At the time of the construction permit for the Comanche Peak facility, reduction of the vulnerability of the plant to industrial or terrorist sabotage was treated as a plant physical 2312 361
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security function, and not as a plant design requirement.
- However, the NRC staff has categorized sabotage as an unresolved safety problem in Task A-29 of NUREG-0410.
Texas Courts have recognized a very real possibility of sabotage and have permitted sabotage to be a consideration in Texas Utilities condemnation of land in and around the Comanche Peak facility. The Comanche Peak design fails to address the possibility that is recognized by Texas law. 55. SAFETY CONTENTION TWENTY-EIGHT The Comanche Peak design fails to idequately provide for the movement and handling of heavy loads in the vicinity of spent fuel at the facility. 56. EXPLANATION This is a generic, unresolved high priority safety matter reccgnized by the NRC staff in Task A-36 of NUREG-0410. The staff recognized the need for a systematic review of this subject "to assess safety margins and to improve those margins where warranted". Presently there is no assurance that the Comanche Peak design is adequate for the movement of heavy loads around spent fuel. 57. SAFETY CONTENTION TWENTY-NINE Serious errors were made in the design and construction of the reactor vessels used at the Comanche Peak facility. 58. EXPLANATION As reported in a Dallas Times Herald newspaper article of February 28,'1979,. the reactor vessel for Comanche Peak Unit 2 2313 001
would not fit correctly on the supports built for the vessel. The vessel was 45 degrees off it's mark and presents a serious safety problem if not properly corrected. The wide margin of error indicates serious communication problems exist between Gibbs and Hill, Inc., Westinghouse Electric Corporation, Combustion Engineering, and Applicant which could result in serious safety problems. 59. SAFETY CONTENTION THIRTY The Comanche Peak design does not adequately evaluate the potential damage to systems essential to the cooling and safe shutdown of the plant due to turbine missiles. 60. EXPLANATION This is a generic and unresolved serious safety problem which was given high priority by the NRC staff in Task A-32 and A-37 of NUREG-0410. 61. S AFETY CONTENTION THIRTY-ONE No adequate evacuation plans exist for the contingency of a major accident and a release of radiation that could reach the Dallas-Fort Worth area such that Appendix E of 10 CFR Part 50 is not met. 62. EXPLANATION The accident at the Three Mile Island facility demon-strated the inadequacy of communications and evacuation procedures. Applicant's ef forts to meet the requirements of Appendix E are 2313 002 w-
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24. less than those of the Three Mile Island applicant. Specifically, Applicant cannot demonstrate sufficient coordination with state and local officials to accomplish the evacuation of the Dallas-Fort Worth metropolitan area in the event of the worst postulated accident. Additionally, there are insufficient facilities in the Glen Rose area for emergency first-aid treatment. 63. SAFETY CONTENTION THIRTY-TWO The staff and the Applicant have not adequately addressed the long-term effects of low-level radiation emissions, and Applicant has failed to insure that those emissions are as " low as reasonably achieveable". 64. EXPLANATION The 1972 radiation risk estimates of the BEIR Report are no longer acceptable. The. Environmental Protection Agency in 1976, contracted with the National Academy of Sciences to prepare a new BEIR Report. The NAS Study should be completed soon, and Applicant should be required to incorporate design criteria which will adequately protect the public from the effects of low-level radiation. 65. SAFETY CONTENTION THIRTY-THREE The Comanche Peak design does not adequately protect human safety and insure that radiciodine releases are as " low as reasonably achievable". 2313 003 ~
25. 66. EXPLANATION The Comanche Peak facility should be required to incorporate design criteria which will guarantee that there will be no harmful releases as defined by research presently being monitored by the NRC, of radiciodines and x-rays. Research is presently being done with animals to investigate the relative hazard per-unit dose to the thyroid of different short-lived radio-iodines and x-rays and the sensitivity of the thyroid to these exposures is a function of age. (Testimony of Saul Levine, Director, Office of Nuclear Regulatory Research, U. S. Nuclear Regulatory Commission, before the Subcommittee on the Environment and the Atmosphere, House Committee on Science and Technology, June 7, 1978). The possibility remains that Comanche Peak could experience an accident of the kind and magnitude experienced at Three Mile Island. Releases occurred at the Three Mile Island facility which could prove to have a detrimental affect to human health. Some radiologists are now concerned that prolonged exposure to low-level radiation could have a greater adverse affect on human health than larger dosages over shorter periods of time. The Applicant should be required to make an assessment of early mortality following inhalation of radioactive materials from the worse possible accident at the Comanche Peak facility, and incorporate design modifications which will keep emissions as low.as reasonably achieveable. 23)3 004
26. ENVIRONMENTAL CONTENTIONS 67. Ef/IRONMENTAL CONTENTION ONE Operation of the Comanche Peak facility would result in the production of unneeded, unsalable, and uneconomically priced electricity in light of the order of the Texas Public Utility Commission in Docket No.14, and thus Applicant cannot survive the cost-benefit analysis required of 10 CFR. Sections 51.20 (b) and 51.21. 68. EXPLANATION The Texas Public Utility Commission has prohibited utilities in the state of Texas from selling power interstate. Texas Utilities operating companies (TESCO, DP&L, and TP&L) have reserve margins which approximate 50 per cent, and those reserve margins are expected to remain at that level until well past 1981. The companies' own calculations reveal high reserve margins (See TESCO's calculations in Taylor Exhibit WMT-2 of Docket No. 1903 before the Texas Public Utility Commission) and those margins should be higher based upon exaggerations of growth and weather conditions. Applicant has failed to demonstrate within its Environmental Report a need for power in Texas. Operation of the Comancho Peak facility would be counter-productive to national environmental concerns for conservation of recources and efficient utilization of energy. 69. ENVIRONMENTAL CONTENTION TWO Actual demand for electricity and actual reserve margins within the Texas Utilities system constitute a significant change 2313 005 a ao.
in the circumstances used to justify construction of Comanche Peak, and Applicant cannot survive the cost-benefit analysis required by 10 CFR Sections 51.20 (b) and 51.21 and th e change in circumstances. 70. EXPLANATION The cost-benefit analysis called for in 10 CFR Section 51.20 (b) requires the Applicant to quantify the various factors involved "to the fullest extent practicable". The Environmental Report submitted by the Applicant recognizes the gross exaggerations of demand and reserve margins that justified the construction permit for Comanche Peak. The factor used to continue construction and operation of Comanche Peak is avoidance of reliance o.' foreign oil. Texas Utilities has never relied on foreign oil and the requirements of 10 CFR Section 51.20 (b) cannot be met. Texas Utilities over-estimated its demand and need for power and under-estimated its reserve margins in applying for a construction permit. Actual demands and reserve margins within the Texas Utilities system constitute a significant change in circumstances requiring a re-evaluation of the cost-benefit analysis for Comanche Peak. In light of actual demand, actual reserve margins, actual need, actual cost of construction, and the existence and ready availability of fossil-fuel alternatives to nuclear power, the operation of the Comanche Peak facility would unjustifiably and unreasonably waste resources, damage the environment, and pose serious danger to the health and safety of individuals in the surrounding area. 2313 006
71. ENVIRONMENTAL CONTENTION THREE The Applicant has failed to provide a meaningful assessment of the environmental, as well as safety, risks associated with the operation of the Comanche Peak facility. The Applicant has failed to adequately address the mitigation of accident consequences with nuclear plant design other than early fatalities. 72. EXPLANATION The Three Mile Island accident proves that a possible i result of an accident is the prolonged r' 'sase of radiation. Applicant has failed to obtain wind roses for the Glen Rose area and construct a meaningful dispersion model for radioactive material. Applicant needs to predict a " killing distance" for that radioactive material and to assess the total danger to the environment as well as human health and safety. 73. ENVIRON 11 ENTAL CONTENTION FOUR The Applicant has not given adequate attention to the criteria for " mini-decommissionings" or the replacement of major pieces of equipment. 74. EXPLANATION Spent fuel is not the only radioactive item that poses a threat to the environment. Applicant must consider within its cost-benefit analysis the possibility of replacing contaminated pipes and large pieces of machinery and the long-term storage of thosel items;, Applicant has failed to consider the environmental i effects " mini-decommissionings". 2313 007 A
29. 75. ENVIRONMENTAL CONTENTION FIVE The incremental burden of nuclear waste on the environment precludes the Applicant from meeting the cost-benefit analysis required for an operating license. 76. EXPLANATION The possibilities that storage of nuclear waste may affect water supplies or release harmful radiation to the atmosphere due to inadequate containment when coupled with the fact that the Applicant has failed to prove that the benefit of nuclear power f outweighs its cost calls for denial of an operating permit. The disposal of nuclear waste presents serious health, safety, environmental, and moral problems which should mandate the denial of any operating license. While the NRC may not require the Applicant to adequately deal with all the problems associated with the disposal of nuclear waste, the NRC has mandated that Applicants meet a cost-benefit analysis test. When the Applicant can prove no need for power generated by nuclear energy, the burden and cost and environmental adversity posed by nuclear waste clearly outweigh any imagined benefits. 77. ENVIRONMENTAL CONTENTION SIX The Applicant has failed to postulate the possibilities, the effect on the environment, and the cost of " cleanups" which necessarily follow a nuclear accident such that the Applicant cannot survive the cost-benefit analysis required by 10 CFR Sections 3 008 51.20 (b) and 51.21.
78. EXPLANATION While it may not be possible to devise meaningful plans for cleanup of an accident until the accident occurs, there are certain accidents capable of postulation such as the one that occurred at Three Mile Island. The Applicant must incorporate within the cost-benefit analysis the cost associated with recovering from certain accidents and the environmental implications that might be associated with such recovery. The Applicant has failed to incorporate those considerations in its cost-benefit analysis. 79. ENVIRONMENTAL CONTENTION SEVEN The entire Comanche Peak facility will at one point constitute ntclear waste, and the Applicant has not revealed an adequate method for protecting the environment and preventing sabotage after the useful life of the plant, nor has the Applicant fully considered total decommissioning costs within the cost-benefit analysis required by 10 CFR Sections 51.20 (b) and 51.21. 80. EXPLANATION _ Total decommissioning costs in dollars and the incremental burden upon the environment should both be considered in the cost-benefit analysis. 2313 009
31. ACORN MEMBERS WITH INTEREST AFFECTED 81. The NRC staff recommended that ACORN be permitted to intervene, "with reservation that such permission be subject to submittal of the name and address of.one member of the organization, a statement as to the manner in which the individual's interests may be affected and an expressed representation by such individual that ACORN is authorized to represent his or her, interests in this proceeding". Modifications have been made in the previously filed Affidavit of Ruth Martin to comply with that reservation. i The Affidavit of Ruth Martin is attached. In compliance with the recent ruling of the Atomic Safety and Licensing Appeal Board's decision in the Allens Creek proceeding, similar affidavits of other ACORN members will be provided to the NRC staff upon request. (The Allen Creek decision contemplates such a limited showina to the staff on pages 45 and 46 of the April 4,1979 slip opinion). CONCLUSION 82. Intervenors pray that the First Amended Petition for Intervention and the Supplement containing Contentions be 2313 010
32. accepted favorably by the Nuclear Regulatory Commission, Atomic Safety and Licensing Board, and that ACORN and Mary and Clyde Bishop and Oda and William Wood, as representative clients of West Tc as Legal Services, be accepted as Intervenors in this matter and that the foregoing Contentions be accepted for litigation in this matter. Respectfully submitted, 0% t OFFffY M.) GAY Counsel for Interveno WEST TEXAS LEGAL SE? VICES 406 W. T. Waggoner Building 810 Houston Street Fort Worth, Texas 76102 (817) 336-3943 DATE) at Fort Worth, Texas, this 7th day of May, 1979. 23l3 011 }
NUCLEAR REGULATORY COMMISSION IN THE MATTER OF S, DOCKET NO. 50-445 AND 50-445 TEXAS UTILITIES GENERATING COMPANY, ET AL S COMMANCHE PEAK STEAM ELECTRIC APPLICATION FOR ISSUANCE OF STATIONS, UNITS 1 AND 2 S FACILITY OPERATING LICENSE Di 9 AFFI-DAVIT $s?Q3 2 J jg7gA THE STATE OF TEXAS S 6 2 jp COUNTY OF TARRANT S g g k, q>.L RUTH MARTIN, being duly sworn, on oath states the folloring: (1) That she resides at 3920 Schwartz, Fort Worth, Texas; (2) That she resides approximately 35 miles from the Commanche Peak facility; (3) That she personally believes that she would be injured or that her health, safety and property would be adversely affected by radioactive emissions of nuclear waste and normal or accidental occurrences at the Commanche Peak facility; (4) That she is a member of Fort Worth and Texas ACORN; (5) That she is on the Board of Fort Worth and Texas ACORN; (6) That she authorizes ACORN to represent her interest in preventing the granting of an operating permit for Commanche Peak; (7) That ACORN has an interest in intervening befor+ the NRC in that the health and safety and environmt.t and financial security of its members will be injured or otherwise jeopardized if Texas Utilities Generating Company is granted an operating license for Commanche Peak; (3) That the Fort Worth and Dallas ACORN Boards have approved this intervention before the Nuclear Regulatory Commis-sion, and that she is authorized by the Fort Worth Board to sign this Affidavit; 2313 012
~ (91 That she has reviewed the representations contained in the Petition to Intervene regarding membership and in-terests to be affected, and that such representations are true and accurate to the best of her knowledge; (10) That it is in the best interest of the public to have a public hearing on the licensing on Commanche Peak. UN C RUTH MARTIN, Affi nt . SUBSCRIBED AND SWORN TO before me, this the day of I
- p.
,'May, 1979. k cc u Y Y.,a l z-Notary Public in and for Tarrant County, Texas ,,gg,, g.,y My commission expires: p;uG b, I Y ./ 2313 013 1
/t 49l OJ3F,4 W i ( #f t $2 UETED STATES OF AFERICA cf y NUCLEAR REGUIATOW CDMISSICN e BEME THE ATOEC SAFM AND LICENSING BOARD N IN DE MrER CF S TEXAS UTILITIES GDEPATING CCEPAW, ET AL S Docket Nos. 50-445 50-446 (Octnanche Peak Steam Electrim Station, thits 1 and 2) S
- u. MUCATE OF SERVICE I hereby rtify that copies of the " Supplemental Petition and Cbntentions of Interverors, ACOIW, Mary and Clyde Bishcp and Oda and William Wood" in the above captioned proceeding have been served on the following by deposit in the thited States Mail, Certified, Beturn Fbceipt Ibquested, this 7th day of May,1979.
M in%th S. Bowers, Esq., Chairman Richard W. Irwerre, Esq. Atcrttic Safety and Licensing Board Assistant Attorney General U.S. Nuclear Pegulatory Ccmission Environmental Protection Division Washington, D. C. 20555 P.O. Box 12548, Capitol Station Austin, Tucas 78711 Iester Kornblith, Esq., Member Atcznic Safety and Li nsing Board Atomic Safety and Licensing Board Panel U.S. Nuclear Pegulatory Ccmnission U.S. Nuclear Regulatory Carmtission Washington, D. C. 20555 Washington, D. C. 20555 Richard Cole, Esq., Menber Ibcketing and Service Secticn (4) Atcmic Safety and Licensing Board Office of the Secretary U.S. Nuclear Pegulatory Commissian U.S. Nuclear Pegulatory Cdmission Washington, D. C. 20555 Washingten, D. C. 20555 Nicholas S. Peynolds Nancy Ja h l Dievoise and Liber:ran CEUR 1200 17th Street, N.W. 1417 Eighth Avenue Washingtcn, D. C. 20036 Fort Worth, Texas 76104 Mrs. Juanita Ellis President, CME 1426 South Polk Dallas, Te.us 75224 .e 'n, J Geoffrey W. Attorney at a 2313 014 _}}