ML19257C349
| ML19257C349 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 01/18/1980 |
| From: | Weiss E UNION OF CONCERNED SCIENTISTS |
| To: | NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| Shared Package | |
| ML19257C350 | List: |
| References | |
| RTR-NUREG-0578, RTR-NUREG-0585, RTR-NUREG-578, RTR-NUREG-585 NUDOCS 8001280475 | |
| Download: ML19257C349 (17) | |
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NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD I #>
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In the Matter of )
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FETROPOLITAN EDISON COMPANY , et al., ) Docket No. 50-289
) (Restart)
(T hree Mile Island Nuclear Station )
Unit No. 1) )
)
UNION OF CONCERNED SCIENTISTS RESPONSE TO FIRST SET OF STAFF INTERROGATORIES General Interrogatories
- 1. At the present time, UCS intends to present Robert D.
Pollard to testify on the issues covered by all UCS contentions.
No other witnesses have been retained at this time. A statement of Mr. Pollard's professional qualifications is attached.
- 2. The following is a reasonable descriotion of the testi-many UCS intends to present on each UCS contention:
Contention No. 1 Mr. Pollard will describe the design of the olant to the extent needed to demonstrate that, under the. current design, it is necessary to rely on natural circulation to renove decay heat. He will describe the TMI-2 accident sequence to the extent necessary to demonstrate that natural circulation was inadequate to remove decay heat and that it was necessary to resort to forced cooling. He will then analyze the appropriate short and long term measures recommended by the staff to 1819 101 8001960
A demonstrate that implementation of those measures will not provide a reliable method of forced cooling. We will also present testimony on the design modifications needed to assure a reliable method of forced or " enhanced" cooling.
These would include a) classifying the reactor coolant pumps as safety-grade b) redesigning the residual heat removal system to operate at the design pressure of the primary system and increasing the heat removal capacity of the rhr and c ) increasing the capacity and radiation shielding for the storage of the radioactive water bled from the primary system when the emergency core cooling system in the bleed and feed mode or d) some combination of the above.
Mr. Pollard sill testify that the short and long term lessons learned documents reveal no effort to systematically consider the benefits of improving the reliability of the various methods of forced or " enhanced" cooling described above. These methods provide needed diversity and were both needed and used during the TMI-2 accident. On the contrary, the staff appears to have considered only relatively minor and inexpensive modifications to the natural circulation mode.
The formation of voids precipitated the loss of natural circulation. Yet the staff has not identified sufficiently reliable means of preventing the formation of voids.
In addition, there are other accident sequences which do not begin with void formation but which would also require 1819 102
4 forced cooling, such as a sequence involving physical blockage of any part of the core.
Documents: The Staff's Interrogatories Nos. 2 and 3 ask for, respectively, the documents that will be relied on in the testimony and the documents relied on in answering the interro-gatories. At the present time, it is not oossible to distin-quish between the two. As a general matter, we rely uoon NUREGS 0578, 0585 and 0623, The TMI Restart Report and FSAR's for TMI Uni ts 1 and 2. Documents more soecifically relied on are identified in the answer to each interrogatory.
Knowledce of the Ans we r to the Interrocatorv: T he Staff's Interrogatory No. 4 asks for the identification of individuals having knowledge which served as the basis for the answer. In each case this is Robert D. Pollard, Union of Conserned Scientists, and Ellyn R. Weiss.
Contention No. 2 The testimony on Contention No. 2 will be combined with that on Contention No. 1. See above.
Contention No. 3 Mr. Pollard will describe the importance of the pressurizer heaters and associated controls with reference to the design of the plant and the accident at TMI-2. He will discuss the general design criteria cited in the contention and exclain why they are important to safety. He will note that, while the staff clearly appears to recognize the imoortance of assuring 1819 103
the reliability of this equipment, they fail to require con-formance with the very criteria established by the NRC as necessary to assure the reliability of safety-related equip-me n t . This is not a matter of judgment which can be decided on an ad hoc basis , but is a question of established regula-tory policy that must be complied with.
Contention No. 4 The testimony will note that the staff has recognized the importance of assuring a reliable source of power to the pressurizer heaters and has proposed to do so by adding the he a te rs to presently existing on-site emergency power supplies (diesel cenerators). The testimony will consist of an evalua-tion of the means by which the Applicant proposes to demonstrate reliability of the emergency power supplies af ter addition of the pressurizer heaters and a discussion of the testing orogram needed to demonstrate adequate reliability.
Contention No. 5 Mr. Pollard's testimony will discuss how the valves in question and their instruments and controls can cause or aggra-vate a LOCA and are essential to mitigating the consequences of accidents. The testimony will discuss the logical and tech-nical inconsistency of the staff's position which, while insisting that the power supply to the valves be safety-grade, does not require the valves themselves to be safety-grade.
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The NRC regulations require that " structures, systems and components important to safety" meet the variety of design criteria which have been developed to assure the reliability of such " safety-grade" equipment. These valvms are important to safety within the meaning of the regulations. This is a matter of clearly-established regulatory policy, not a matter of " judgment" which can be waived on an ad hoc basis.
Contention No. 6 The testimony will note the staff's finding in NUREG 0578 that there have been a number of instances of improper opera-tion of relief and safety valves , but that inadequate qualifi-cation testing prevents determination of whether that improper operation resulted from inadequate qualification or from a basic unreliability of the design. The testimony will consist of an evaluation of the licensee's proposed qualification test-ing program and the test results.
Contention No. 7 Mr. Pollard will discuss the design of the plant for the purpose of explaining why a direct measurement of reactor water level is the " direct measure of the desired variable" required by the Commission's regulations. Mr. Pollard served on the IEEE Committee with jurisdiction over IEEE 279. He will explain the basis for incorporating this standard into the regulations. Given that the major purpose of the plant's 1819 105
safety systems is to ensure that the core is adequately covered with water, it is almost inconceivab.~.e that no instrumentation exists to directly measure that variable.
The testimony will discuss the ways in which the absence of a direct measurement of reactor water level contributed to the TMI accident. The staff, instead of requiring compliance with their regulations, proposes that the licensee develop procedures to be used by the operator to recognize inadequate core cooling with currently available instruments. In other words, the staff seeks to permit the operator to use an indirect measurement of the " desired variable" rather than require installation of a means for direct measurerent.
Contention No. 8 The testimony will discuss the short and long tern neasures proposed by the staff which are directed toward mitigating the consequences of a stuck-open relief valve. These measures ae insufficient to assure that the consequences of any other small LOCA will not exceed the criteria for ECCS performance.
For example, the consequences of a pipe break upstream of the relief valve would be unaffected by the staff's proposal to require safety-rel.ated power supplies to the relief valve and associated block valve.
Contention No. 9 Mr. Pollard was the principal author of Regulatory Guide 1.47. His testimony will discuss actual reactor experiences 1819 106
_7_
and other f actors involved in the development of the regulatory pos i tion . He will then discuss the var ious means which are feasible for use in conforming with Reg. C' _ 1 !7 or provi-ding equivalent protection. The testimony will oclineate the ways in which the absence of a means of complying with Reg. Guido 1.47 can lead to serious accidents such as the one which occurred at TMI-2.
Contention No. 10 As noted in the response with resoect to UCS Contentior No. 7, above, Mr. Pollard served on the IEEE ffommittee with jurisdiction over IEEE 279. He will explain the bases for this portion of the standard. He will discuss the ways in which the core cooling and containment isolation systems (and their essential auxiliary supporting systems ) can be prematurely shut off. For example, the operators can shut off not only the ECCS, as they did during the TMI-2 accident, but also the auxi-liary feedwater system and containment isolation systems.
In the case of the ECCS at TMI-2, it is quite clear that the safety function was not permitted to go to completion. Thus, the design violates the commission's regulations. This is not a matter of judgment, but of minimum safety requirements.
Further, Mr. Pollard's testimony will discuss the practices of the staff which permitted this violation of basic safety requirements. Specifically, by defining " initiation" of the safety function as "comoletion" of the safety function, the staff approved a design where the operator can shut off the 18'9 107
_8_
ECCS pump the instant it starts, thus totally negating the purpose of the requirement.
Contention No. 11 Mr. Pollard's testimony will discuss the purpose and need for hydrogen control, the bases for the current regulation and the fact that conformance with these regulations does not protect public health and safety. He will discuss the accident sequence at TMI-2, including analysis of the ways in which the accident demonstrated that the current criterion of 5% fuel cladding reaction is inappropriate, the failure of Metropolitan Edison to do sufficient preoperational testing to detect the excessive flow rate of gases through the recombiner and the inadequate design basis of the radiation shielding of the hydrogen recombiner system. The testimony will conclude that it is not pos sible to justify the conservatism of any criterion less than 100% fuel cladding reaction.
Contention No. 12 The testimony will discuss the basis for GDC 4 and how it interrelates with other requirements o f 10 CFR P ar t 50 such as the single failure criterion. It will also discuss the important characteristics of an adequate environmental qualification program, such as 1) specification of the accident environment 2) de termination of the time equipment must operate in that environment 3) the problems that can affect the validity 1819 108
4 of the actual qualification method (e.g. difficulties in simula-ting the effects of aging over a 40-year period. )
The testimony will relate the above factors to the actual deficiencies disclosed by the TMI-2 accident. We intend to analyze in detail Met Ed's responses to the IE Circulars and Bulletins relating to environmental qualifications for the purpose of demonstrating that the environmental qualification critera applied to TMI-l are inadequate to protect the public and that the staff's response to the accident demonstrates a continuing failure to learn the lessons and to implement the necessary changes. It will be UCS's position that the criteria for environmental qualification detailed in Regulatory Guide 1.89 or equivalent are the minimum necessary to protect public health and safety.
Contention No. 13 The testimony will discuss the ?!RC's policies with respect to the philosophy that there are possible accidents for which no protection for the public is required on the alleged basis that their probability is so low as to be " incredible," i.e.,
the so-called " Class 9" accidents. The testimony will note that since the TMI-2 accident fell within the " incredible" category, the staff's method of analysis is prima facie invalid. For a detailed discussion of UCS's position, see
" Union of Concerned Scientists Request for Reconsideration or, in the Alternative , For Certification," January 7, 1980.
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Contention No. 13 Mr. Pollard's testimony will discuss the regulatory signi-ficance of the classification of equipment as either safety or non-safety related and the inadequacies in the methods used by the staff in an attempt to consider the effects on safety systems of f ailures of "non-safety" components. He will then discuss the ways in which components classified as non-safety caused or aggravated or were used to nitigate the TMI-2 accident. For specific examples, see Contention No. 14. The testimony will go on to identify the requirements that must be applied to components which play a role in core cooling in order to protect public health and safety.
Soecific Interrogatories Question 9-1:
Identify the specific systems within the core cooling and containment isolation sys tems as to which it is alleged that there is inadequate provision to inform the operator that such a system has been deliberately disabled.
Answer Other than providing a general list of plant systens, such as engineered safeguards actuation system, reactor protec-tion sys tem, auxiliary feedwater systen, containment ventila-tion system, service water system, and containment isolation system, we cannot yet answer this question. Ne have asked the staff to inform us whether the plant conforms wi'h Regula-tory Guide 1.47 and, if not, in what ways it differs from the 1819 110
requirement of that Regulatory Guide. See, UCS Interrogatories to the NRC S taf f , regarding Contention 9 (i.e., Interrogatorica 83-91). When we receive that answer, with supporting documenta-tion , we will begin a more detailed analysis of the plant design.
Question 9-2:
With respect to the specific systems identified in response to Ques tion No. 9-1, explain the bases for believing that deliberate disabling is feasible and credible.
Answer The bases for believing that deliberate disabling is feasible and credible fall into four categories:
- 1. Mr. Pollard has extensive experience as a member of the staff. He is fully aware of the s taff's propensity to avoid f acina the question of whether new regulatory guides developed as a result of actual reactor experience, should be backfit. He also reviewed the Babcock & Wilcox - designed Oconee plants, which were licensed at appro-ximately the same time as TMI-1. Thu s , he is aware that deliberate disabling is feasi-ble and credible.
- 2. The accident at TMI-2, which was licen-sed four years after TMI-1, showed that deliberate disabling is feasible and credi-ble for that plant in the case of the auxiliary feedwater system. Familiarity with the design approach of B&W and the review practices of the staff is another basis for believing that the deliberate disabling of other safety systems is fea-sible and credible.
- 3. The study and work that went into evaluating reactor experience and deve-loping the Regulatory Guide itself esta-blishes that deliberate disabling of safety systems is feasible and credible.
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- 4. It is normal procedure to deliberately disable certain safety systems for neriodic tes ting and maintenance. For example, the method of calibrating many of the instru-ments which supply input signals to start core cooling or initiate containment isc-lation requires that the system be deli-berately disabled. Also, maintenance on components in fluid systems commonly requires that valves be closed to isolate the .omponent, thereby deliberately disabling the safety system.
Question 9-3:
With respect to the specific systems identified in response to Question No. 9-1, explain the deficiencies in the present information provided to the operator with respect to that sys tem.
Answer:
As of now, TMI-l has no automatic indication what-soever to inform the operator of the operability status of core cooling and containment isolation systems. The indica-tion presently available, in at least some cases, tells the operator the status of individual components, for example, the position of valves. However, knowing the cosition of the valves cannot be relied upon to inform the operator of the operating s tatus of the system. In the language of the Regulatory Guide, the present deficiency is that the infor-mation is at the component level rather than the system level.
The other basic deficiency is that the component level indication in auxiliary supporting systems (such as croling water for pumps, emergency electric power supplies for core 1819 112
cooling and containment isolation) provides no information whatsoever about the effect that disabling of the supporting system has on the core cooling and containment isolation.
In other words , even if the operator recognizes that the auxilary supporting system is disabled, he cannot be relied upon to recognize that this will also disable the affected safety system.
Question 10-1:
Identify the specific systems within the core cooling and containment isolation system as to which it is alleged that the operator can prevent the comoletion of a safety func tion .
Answer:
See the answer to 9-1. Again, we are awaiting answers to UC - Interrogatories92-100.
Question 12-1:
Provide the definition used by UCS for " equipment important to safety. "
pngwer: .
"E quipment important to safety" is defined in the same manner as used by NRC in the phrase in GDC 4: " structures, systems and components important to safety." We note that we have asked the staff to tell us which " structures systems and components" within the containment and auxiliary build-1819 ii3
ings are covered by GDC 4 (See UCS Interrogatory #113 ) .
Question 13-1:
Identify the particular element (s) of the staf5's method of determining which accidents fall within the design basis accidents which UCS alleges are faulty.
Answer:
The staff has no technical basis for making the determina-tion that certain accidents are of such low probability as to be " incredible." For a more detailed discussion, see the attached letter and Draft NRC Statement of Policy Concerning Reactor Safety Study and Notice of Intention to Promu2 gate Regulations, dated November 1, 1978.
Question 13-2:
Identify any modifications or additions to the Staff's method of determining design basis accidents which UCS would recommend.
Answer:
The modification that UCS would recommend is that design basis accidents be selected without regard to probability at least so long as the state-of-the-art does not permit a rational, technically justifiable determination of the actual orobability of high consequence accidents.
Question 13-3: )8}h }lk Describe any method or methods for determining design basis accidents which UCS would support.
Answer:
UCS would support a method which provided protection against all accidents whose consequences could exceed the limits specified in 10 CFR Part 100. UCS could not support any method based on probability of occurrence because there is no known method of determining the actual probability of any accident sequence with sufficiently narrow error bands to justify using the present " al l-or-no thi ng " approach.
Question 14-1:
Identify those systems or components within the core cooling system which UCS believes can either cause or aggravate an accident or can be called upon to mitigate an accident. If you can provide only a partial answer, please do so.
Answer:
To date, UCS has identified the following such systems or components: pressurizer heaters, pressurizer relief valve (and associated block valve ) , pressurizer safety valve, reactor coolant pu mps , auxiliary feedwater pumps ,
main feedwater pumps, boric acid transfer sys tems , chemical and volume control systems, condenser bypass (steam dump) system and all auxiliary supporting systems necessary for all of the above to function.
Question 14-2: )b If UCS feels it cannot identify all of the systems or ownponenbs described in question 14-1 above, describe the
me thod ( s ) acceptable to UCS for identification of such systems or components.
Answer :
UCS believes that systems or components that can directly or indirectly affect temperature, pressure, and/or flow in the reactor coolant system and/or reactivity of the core can either cause or aggravate an accident or be called upon to mitigate an accident. Therefore, the staff should apply this criteria in answering UCS interrogatory #156.
DATED: January 18, 1980 iB19 116
9 ROBERT D. POLLARD QUALIFICATIONS Mr. Pollard is presently employed as a nuclear safety expert with the Union of Concerned Scientists, a non-profit coalition of scientists, engineers and other orofessionals supported by over 80,000 public sponsors.
Mr. Pollard's formal education in nuclear design began in May, 1959, when he was selected to serve as an electronics technician in the nuclear power program of the U.S. Navy.
After completing the required training, he became an instruc-tor responsible for teaching naval personnel both the theore-tical and practical aspects of operation, maintenance and repair for nuclear propulsion plants. From February, 1964 to April, 1965, he served as senior reactor operator, supervis-ing the reactor control division of the U.S.S. Sargo, a nuclear-powered submarine.
After his honorable discharge in 1965, Mr. Pollard attended Syracuse University, where he received the degree of Bachelor of Science magna cum laude in Electrical Engi-neering in June, 1969.
In July, 1969, Mr. Pollard was hired by the Atomic Energy Commission (AEC), and continued as a technical exoart with the AEC and its successor the United States Nuclear Regulatory Commission (NRC) until February, 1976. After joining the AEC, he studied advanced electrical and nuclear engineering at the Graduate School of the University of New Mexico in Albuquerque. He subsequently advanced to the cositions of Reactor Engineer (I ns trumenta tion ) and Project Manager with AEC/NRC.
As a Reactor Engineer , Mr. Pollard was primarily respon-sible for performing detailed technical reviews analyzing and evaluating the adequacy of the design of reactor protec-tion sys tems , control systems and emergency electrical power systems in proposed nuclear facilities. In September 1974, he was promoted to the position of Project Manager and became -esponsible for planning and coordinating all aspects of the tsign and safety reviews of applications for licenses to cons t ruct and operate several commercial nuclear power plants. He served as Project Manager for the review of a number of nuclear power plants including: Indian Point, Uni t 3, Comanche Peak, Units 1 and 2, and Catawba, Units 1 and 2. While with NRC , Mr. Pollard also served on the standards group, participating in developing standards and safety guides, and as a member of IEEE Committees.
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