ML19309A709
| ML19309A709 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 02/25/1980 |
| From: | Sholly S AFFILIATION NOT ASSIGNED |
| To: | METROPOLITAN EDISON CO. |
| Shared Package | |
| ML19309A704 | List: |
| References | |
| NUDOCS 8004010044 | |
| Download: ML19309A709 (36) | |
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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD f
<a o In the Matter of , )
Doc t" 0-289 METROPOLITAN EDISON COMPANY es a (Three Mlle Island, Unit 1) )
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INTERVENOR STEVEN C. SHOLLY THERD SEI OF INTERROGATORIES AND DOCUMENT REOUESTS TO LICENSEE Steven C. Sholly, Intervenor, hereby flies the following interrogatories and document requests with Licensee pursuant ;
to 10 CFR 2.740 and 10 CFR 2.741. Interrogatories are to be answered fully, in writing, and under oath by any officers or employees of the Licensee who have personal ' nowledste thereof. The answer to each interrogatory should contain i
the names and identification of the persons supplying the answers. In addition, all documents relied upon in supplying the answers are to be referenced according to title, authors, source of document, and page numbers utilized in supplying the answers.
Interrogatories and document requests are arranged according to the Contention numbers as accepted by the Board. Contention numbers appear first, followed by the interrogatory or document request number within the particular contention.
Interrogatories and document requests are considered l to be continuing and are to be supplemented as required by 800401
appilcable NRC regulations.
16-007--WLth regards to qualificattens of a witness to permit review of Licensee's Security Plan, what qualifications would be acceptable to Licensee?
16-008--With reference to ALAB-410 (5 NRC 1398--1977) at pages 1409 through 1410, supply the following information about Licensee's Security Plans
- a. Design criteria for all hardware which is depended upon by Licensee to prevent internal sabotage
- b. Vendor hardware specifications with regards to performance of all hardware uttitzed by Licensee to prevent internal sabotages
- c. Design criteria and performance specifications of all communications systems utt11 zed by Licensee in protecting against internal sabotage or to request assistance of outside agencies in preventtng acts of internal sabotage;
- d. Design criteria for all tamper indlcattng seals in Type I Vltal Areas:
- e. Types of locks uttilzed in preventing access i to Type I Vital Areas
- f. Maximum and minimum staffing requirements for security force on all shifts, including the numbers of Licensee employees and contract security personnels
- g. A general description of any arrangements for offsite response untts which may be called upon to assist in dealing with an act or threat of internal sabotage: .
- h. Descriptions of any and all audit methods used
i to evaluate the effectiveness of security forces at TMI: and, L. A description of routine testing and inspection methods used by the security force at TMI.
16-009--Provide names and last known addresses for all employees of Licensee engaged in work on the security force at TMI, whether at Unit 1 or 2 or both, who have resigned, been asked to resign, or quit since the accident at Unit 2 on 28 March 1979'.'
16-010--Provide a listing of any and all attempted acts of sabotage, suspected attempted acts of sabotage, threatened acts of sabotage, and thefts or suspected thefts of security-related documents with respect to both Units 1 and 2 at TMI. This listing shall be arranged in chronological order and shall include the date of occurrence, time of day, a description of the incident, and a description of Licensee's response to the incident.
The listing sr.all be confined to acts perpetrated or threatened by employees of Licensee or employees of Licensee contractors.
16-011--Specify, according to the requirements of 10 CFR 73.55(a),
how Licensee's physical protection system and security organization protect against sabotage by members of
the site security force, whether employed by Licensee or a contractor oganLzation, Lncluding Gregg Security.
16-012--Specify the date by which Licensee will implement the physical search requirements of 10 CFR 73.55(d)(1).
16-013--State unequivocally that all unoccupied vital areas are locked and protected by an active intrusion alarm system as required by 10 CFR 73.55(d)(7). If Licensee cannot make such a statement, explain why not in detail. -
16-014--Spectfy any and all occasions on which Ltcensee has been cited for violation of NRC security regulations.
Describe the violations , including date, time, and pertinent facts surrounding the Lncident, including Licensee's response to the citation and any and all commitments made by Licensee to change or upgrade security plans and/or capabilities in response to each incident. Cite the NRC regulations which were violated in each instance.
16-015--Provide evidence that Licensee's training program for security force personnel includes the following features l
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- a. Training in tactics and force which might be utilized by adversary groups:
- b. Training in the recognition of sabotage-related devices and equipment that might be used against Licensee's facility
- c. Training in protected area security procedures and practices and the vulnerability of such practices and procedures to being rendered ineffective by insider actions:
- d. Training in the vulnerabilities and consequences of industrial sabotages
- e. Training in access control systems and their operations, including limitations
- f. Training in contraband control techniques and detection. systems:
- g. Training in security system operations after component failures :
- h. Training in the use of and defense against incapactating agents:
L. Training in response to hostage situations
- j. Training in the response to confirmed attempted sabotages and,
- k. General recognition in all security-related training of the special problems posed by insider-initiated sabotage, including problems posed by intimate familiarity with security procedures and systems and means for their disabling.
Where such training does not exist, explain why and whether l l
such training will be implemented and when. Each of these areas of knowledge is taken from Appendix B to 10 CFR 73.
l 15-005--Has Licensee or its contractors ever asked TMI operators to evaluate the design adequacy of the control room?
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If so, provide copies of all documents relating thereto.
If not, why not?
15-006--To what extent do plant operators play a role in the development of operating procedures, including procedures for emergency situations? Do operators actually write such procedures? What is the extent of their input into plant operating procedures?
15-007--Specify the environmental condLttons present in the Untt I control room, including temperature, humidity, noise level, lighting requirements , and provlsions for protection agalnst contaminants (radioactive or otherwise) . .
SpecLfy how these environmental conditions are maintained.
15-008--What is LLcensee's polley regarding ' housekeeping duties in the control room?
15-009--What are LLcensee's policies regarding operator conduct and dress while on 'uty?
15-010--Specify the bases upon which the control room for Unit I was designed. Provide descriptions for all factual bases and assumptions uttitzed in developing the design for the Unit I control room.
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15-011--Describe in detail how the control room panels are i layed ou': in functionally demarcated areas.
15-012--Detail all instances in which operators have resorted to their own labelling in the control room of Unit I due to inadequacies in the original labelling.
15-013--For panels in the control room having similar controls for a number of different features , discuss how these controls are differentiated other than by labelling.
Discuss the extent to which large numbers of similar controls on any given panel gives rise to operator error because of the absence of differences in appearance, shape, color, and/or texture.
15-014--Discuss precautions taken to ensure that controls for protection systems in the Unit 1 control room cannot be accidently manipulated.
15-015--Discuss how operators may determine if an annunciator lamp is burned out.
15-016--Discuss how meters used in the control room are coded !
l to permit easy determination by operators of whether the meters indicate normal, mtrginal, or out-of-tolerance l l
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P 15-017--Discuss the extent to which analog trend recorders are utilized in the Unit I control room.
15-018--Identify and provide the location within the Unit I control room of all chart recorders which record more six (6) functions.
15-019--How many annunciator lights are there in the Unit I control room? How many separate meters? How many control switches? How pany strip chart recorders?
15-020--How many alarms which appear in the Unit I control room refer an operator to a remote (out-of-control-room) location, requiring the operator to leave his station or send an auxiliary operator to answer the j alarm or assess the cause for the alarm or correct the alarm-causing situation. Identify each alarm and the name ani location of the remote station to which the operator is referred by the alarm.
15-021--For all the following code colors used in the Unit I control room, list each separate meaning which this color can have (i.e., valve open, flow stopped, etc.):
- a. Red d. White
- b. Green e. Blue
- c. Amber f. Flashing lights of any color
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15-022--To what extent has Licensee used " task analysis" as described in EPRI NP-309, " Human Factors Review of Nuclear Power Plant Control Room Design," in evaluating the design of the Unit I control room?
~15-023--Identify by name, title, organization, and professional qualifications of any and all experts in the fleid of human factors engineering who participated in the design or review of the design of the Unit 1 control room, or any portion thereof.
15-024--Specify each instance of where procedures specify one measurement unit and the measurement unit on the display in the control room is in another type -E unit.
15-025--How much time is spent by operators during watch turnover to ensure that the oncoming shift is adequately appraised of plant conditions? Are operators paid for time so spent?
15-026--What is the basis on which the composition of reactor operator shifts is determined? How does this basis ensure an adequate mix of talent, technical knowledge, and operating experience on all shifts?
15-027--What restrictions exist on the amount of overtime which reactor operators can be required to work in any given two-week periodi
.15-028--To what extent does the simulator on which TMI operators are trained differ from the control room of Unit I?
With respect to these differences, how does the training of Unit 1 operators ensure that these differences are communicated to the operators? How does operator training ensure that nagative transfer of learning does not occur with respect to differences in design of the simulator and the Unit I control room?
15-029--To what extent does the Unit 1 control room make use of computer-based graphic display to inform operators of system status and parameters? i l
15-030--To what extent do control room operating procedures !
make use of fault tree analysts to aid operators in assessing accident conditions and ensuring that proposed actions will not d'egrade the existing situation? If fault tree analysis is not utilized, explain why not.
l 15-031--Is the window separating the shift supervisor's office in the Unit 1 control room from the control room area bullet-resisting as per 10 CFR 73.55(c)(6)? If not , why not?
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- 15-032--Identify and supply copies of any and all LL;ensee Event Reports (LER's) and NRC Inspection Reports which identify as a causitive factor human error, operator error, or other similar cause.
15-033--Describe the extent of information available to the reactor operators at the operator's desk in the Unit I control room. Identify procedures , manufacturers manuals, and Licensee practices and procedures which are not available immediately at the operator's desk.
15-034--What means will be utilized in the Unit I control room to assure that operators know the status of the PORV on the pressurizer, i.e., whether the PORV is open or closed? Is this means single-failure proof? Is the pressurizer qualified as " safety-grade"? If not , why not?
15-035--On what panel in the Unit 1 control room are the following features located:
- a. PORV status
- b. Status panel for ACDT (reactor coolant drain tank) ,
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- c. Emergency (auxiliary) feedwater controls
- d. Display for reactor coolant pump vibration and eccentricity I
- e. Pressurizer level indication
- f. Rea ctor coolant pump seal pressure
- g. Reactor coolant pump seal temperature
- h. Reactor coolant pump controls
- 1. Borated water storage tank controls
- j. High pressure injection controls
- k. Low pressure injection controls
- 1. Decay heat indicators
- m. Decay heat pump controls
- n. ECCS actuation control
- o. ECCS status panel
- p. Letdown controls
- q. Intermediate closed cooling pump controls 15-036--Provide assurance that the procedures associated with pressurizer PORV and associated block valve reflect
. the same label nomenclature as the labels on the control room panels. l 15-037--Provide assurance that reactor operators will not bypass engineered safeguards following a turbine trip, unless the reactor is in a stable conditions under which the l
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core will remain covered.
15-038--Provide assurance that the mindset problem associated l l
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I with avolding solid water conditions in the pressurizer will be avolded by retraining which reactor operators will have undergone by the time at which Licensee proposes restart.
15-039--Provide evidence that Licensee will provide a direct indication of emergency feedwater flow to the steam generators to the Unit I control room panels.
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15-040--Provide evidence that Licensee will provide an alarm or other appropriate indtcation that the emergency feedwater system is misaligned or otherwise inoperative.
15-041--Provide evidence that operator training programs develop l appropriate visual search strategies for operators to ensure effective response to emergency conditions.
15-042--Provide evidence that operator tralning has included training in natural circulation, including the parameters j under which it is possible to place the reactor in a l l
natural circulation mode. l l
15-043--Provide appropriate documentattor that would provide assurance that visual aculty has been properly considered in the design of the Unit I control room. l l
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l 15-044--Provide the factual basis for Licensee's operator training program, t.e., on what bases are specific training tasks determined to be appropriate? Provide documentation to support the incorporation of human factors engineering considerations in Licensee'c operator training program.
Are operator functions in the control room translated into specific tasks and subsequently into quantifiable performance requirements against which operator performance can be assessed? If not, why not? If so, provide details.
15-045--Describe the extent to which video monttors in the control room at Unit 1 permit reactor operators to monitor critical areas within the plant. If such monitoring capability is not yet available, descrlbe when it will be available or explain why it will not be made available befora restart.
15-046--How many CRT displays exist in the D'itn 1 control room?
What types of information can be displayed on these CRT's?
Does the CRT display system have computer-based graphics capability, and, if so, to what extent will this be used in the Unit 1 control room?
4 15-047--Describe the measures taken to assure that critical controls are not subject to unintentional or accidental
initiation while operators are manipulating other controls or attempting to monitor meters, displays, and charts. Identify any and all instances in which protective systems and engineered safeguards systems have unintentionally or accidently been lattiated.
15-048--Discuss the extent to which inadvertent or accidental initiation of protective systems or engineered safety features systems may cause such systems to be challenged more frequently than their design basis.
15-049--Describe Licensee policies and procedures regarding the replacement of burned-out alarm and control indicator lights in the control room. If such Ilghts are not replaced immediately upon discovery of their burned-out conditions, provide reasons therefore and explain how such a practice protects the public health and safety.
15-050--To what extent does the alarm / annunciator display system in Unit 1 make use of auditory discrimination and/or prioritization systems to assist operators in locating alarming indicators and assessing possible patterns in alarm conditions? If such methods are not used, explain l why not. If they are used, explain ho's they work and provide the design basis for such methods.
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15-051--Is there an annunciator for reactor trip in the Unit I control room? If so, where is it located? If not, explain why not.
15-052--Explain the extent to which extinguished indicator and/or alarm lamps are utilized by reactor operators as positive indicators ' of system status . Explain, if such positive indication is based upon extinguished lamps, how this impacts on the availability of critical information for reactor operators during emergency situations when accurate information is an abso' lute necessity.
15-053--To what extent must critical operating parameters be inferred from changes in associated parameters, rather than such parameters being available directly at the control panel. For each such instance, discuss how l
such a procedure may induce operator error based upon erroneous interpretation of associated parameters , and how reactor operator training has been arranged to assure that operators correctly interpret such information. l 15-054--Discuss how operators determine that the primary system has reached saturation conditions. If done manually, I explain why this function cannot be monitored by the Unit I computer,-thus assuring that when saturation conditions occur, operators are immediately aware of the condition.
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15-055--Explain why there is not a consistent practice in the placement of labelling at control / display locations in the Unit I control room, L .c. , why some labels are placed above and others are placed below. Discuss the extent to which such inconsistent practices may cause confusion among operators and why a consistent labelling practice in this regard would not reduce the chances for operator error, and thus~ provide more protection for public health and safety.
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15-056--During what year did prcliminary work begin on the design of the Unit I control room? Identify by name, title, and organization who worked on the initial design of the Unit I control room. Similarly, identify who was responsible for the final design of the Unit I contro1~ room.
15-057--What formal steps has Licensee management taken to ensure that the capabilities and limitations, both physical and mental, of reactor operators are taken into account in the design and revlow of the design of the Unit 1 control room? Is there a formal plan for periodic review of control room design which reviews operator concerns about the control room design? If so, identify and ciscuss. If not, explain why not.
15-058--What mechanism exists within Licensee organization to review and approve proposed changes to control room design? Besides this mechanism, what other rouces are called upon to participate in such reviews?
15-059--Identify by name, title, and organization, who among the staffs of Licensee, tPe Unit 1 architect-engineer, and Babcock and Wilcox which participated in the design or review of design of the Unit 1 control room had formal training and education in human factors engineering.
For each such person, l'dentify the extent to which he or she participated in the design or review of the design of the Unit I control room. Identify recommendations regarding control room design made by such individuals and explain whether or not their advice was followed, and if it was not, explain why not.
15-060--To what extent during the design of the Unit I control room were mockups utilized to assess the design of the Unit I control room?
15-061--Identify any mechanisms within Licensee's organization which provide a systematic review of operator performance and provide suggestions for improvements in control room design, operating procedures, and training programs.
15-062--NUREG/CR-1270, volume I, at pages 72-76, discusses several procedures used at TMI-2 which this report charactertzes as " Emergency Procedures" and Ldentlfles specific inadequacles in these procedures which cause problems for reactor operators in tmplementing these procedures. The procedures identLf ted in the referenced section of NUREG/CR-1270, volume I, are s ,
Loss of Reactor Collant/ Reactor Collant System Pressure (EP 2202-1.3)
Pressurtzer Operation (UP 2103-1.3)
Post-Accident Hydrogen Control (AP 2203-2.6)
Pressurtzer System Failure (EP 2202-1.5)
Do the slmtlar procedures at Unit I suffer from the same or slmtlar inadequacles? Explain your answers.
15-063--ANSI standard N18.7, " Administrative Controls and Quality Assurance for the Operattonal Phase of Nuclear Power Plants" (1976), defLnes requirements for the preparatton of instructions and procedures . To what extent have Unit 1 procedures been evaluated against this standard?
Who performed the evaluatLons? When were they performed 7 Are revtsed procedures evaluated against the standard, and Lf so by whom?
15-064--NUREG/CR-1270, volume I, at pages 77-78, Ldentify a major problem with procedures at TMI-2 in that operator access to procedures depends heavily on operator memory
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of what is in each procedure and to what situations 1 :
each procedure applies. The report suggests that a "dectston ald" is needed which Ls separate from the
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procedures themselves, but which Ldentifles for the 1
,- operators the procedures which should be in effect for
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a given situation. Does Ltcensee plan any such deciston aLd for use by Unit 1 operators? If so, describe the t
'.c saltent features and whether Lt has received NRC
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. approval (and Lf so, by whom in hRC). If not, justify v
. the exclusion of such an operator ald, addressing the need evidenced for such an aLd durin6 the TMI-2 accident on 3/28/79.
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15-P65--Discuss the extent to which TMI-1 Emergency Procedures address the problems posed by multiple failures Ln determining whleh procedures should be followed.
j 15-066--NUREG/CR-1270. volume I, at page 78 states , " Presently
,' at TMI there is no formal method for getting operator p
inputs to procedure updates, or even for havin8 users
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I Ldenttfy the problems in using procedures." The
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document further states that , "The fact that operators are not formally in the loop to update procedures reflects the general attttude of Met. Ed. toward control room operators and senior reactor operators."
In view of the fact that Lt ts 1Lkely that no one could 1
__ _ _ _ _ o - - _- o e know more about the limLtations and problems associated with the procedures than the people who have to use them (L.e., control room operators and sentor reactor operators).
explatn why these persons should not be formally included within the procedure review and update mechantsm at TMI. In addttlon, explain how Met Ed management has changed Lts vtew of control room operators and senior reactor operators and how this change has been communtcated' to those Lnvolved.
15-067--According to NUREG/CR-1270, volume I, at pages 78-79, the detection and Lsolation of ooerational problems at a nuclear power plant are diagnostle activttles which cannot rely on memory or Entuitive knowledge of the plant, or even training (since sLgntftcant periods of time can lapse between tratntnB for a fault s Ltuation and the actual occurrence of that fault in operatLons).
What ts left is the use of emergency procedures. Ltcensee's Training Divtston member Mr. Beers stated in testimony before the President's Commission on the AccLdent at Three Mlle Island that emergency procedures are written as guldelines only and that operators should primartly rely on their training. If operator tralning does not include tralning on a spectfle fault which then occurs during operations, even the training has ilmited usefulness.
Detall the mean res whLch Licensee has taken to assure
O that both operator training and emergency procedures have been improved such that there Ls a reasonable assurance that operators will perform the correct steps in the correct sequences during emergency situations in order to protect the public health and safety.
15-068--According to NUREG/CR-1270, volume I, at page 79, TMI-2 emergency procedures regarding override of HPI (High Pressure Injection using ECCS) did not provide clear guidance on what conditions HPI could or should be by-passed and Lf so, to what degree the pumps should be throttled. Inasmuch as HPI is part of the Engineered Safeguards system at TMI, and as such is a primary means of protecting reactor integrity under conditions of a LOCA, adequate procedures governing the use of the HPI system are essential to the protection of public health and safety. Provide the procedures uttitzed as described herein, or describe the conditions under which HPI may be over-rldden or throttled back. Provide suf ficient detall in the ,
1 description such that there is clearly enough guldance for operators so that only under the proper conditions i ts such a vital system over-rldden or throttled.
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- {, 15-069--Desertbe the physical and medical standards which l, control room operators and sentor reactor operators
}je must meet, especially with respect to Ilmitations
', imposed by the design and layout of the TMI-1 control k
i room as it now extsts. Include in your description any standards on the following which operator or senior operator candidates must meet and maintain compliance withs
[ a. Vistal aculty requirements
- b. Color bitndness tests
, c. Heartn6 tests
,'- d. Psychological evaluations
- e. Evidence of dru8 or alcohol use
- f. Hetght and welght limits
- g. History of falnting, setzures, and cardto-vascular problems 15-070--According to NUREG/CR-1270, volume I, at pages 90-2, simulator training accounted for only 5% of the time that the crew involved in the TMI-2 accident spent
- - in tralning. The report potnts out that a signtftcant degree of training in procedural skills acquisttton and control skills acquisition are need d by reactor operators, and that a high degree of fliellty in simulatton-with respect to console configuratton ana system response is required for such skill acquisttlan. The report
further asserts that acquisttlan of procedural and control skills cannot be accomplished with the B &.W simulatton faciltty due to a lack of fidelity with the TMI-2 control room. In the light of these facts, discuss in detall the presence or lack of fidelity of the TMI-1 control room with the B & W simulator and the degree to which any dissimilarttles may impact on the acquLsttlon of procedural and control skills-(for a discusston of what these skills entall, see the referenced pages of NUREG/CR-1270, volume I) .
s 15-071--NUREG/CR-1270, volume I, at paSes 91-92, states that
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the B & W simulator is more suttable for the development of diagnostic skills than for the development of procedural and control skt11s. The report further states that the shift supervisoc for TMI-2 at the time of the accident had undergone requalification training on the B & W simulator which consisted of 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> of simulator training in which a total of 19 different evaluations and emergencies were simulated.
Of these 19,14 were performed only once and only one was performed as often as three times. The report suggested that this indicates that the simulator is being uttilzed to illustrate what a selected emergency looks like in terms of display readtngs and plant reactions, rather than allowing operators to acquire
skills through practt'ce in responding to faults and formulating hypotheses concerning what is happening in the plant. To what extent does the cited simulator tralning represent the simulator training whLch is normally given to reactor operators, senior reactor operators, and shLft supervisors for TMI-17 In addition, what chan6es in simulator tralning has LLcensee implemented since the TMI-2 accident to assure that sufficient training on the simulator takes place? To what degree is such training limLted by the use of the B & W simulator by other operators from utilltles other than the LLcensee?
15-072--To what degree is evaluatLon of simulator training based upon object Lve, quantitative measurements of operator performance, as opposed to subjective opinions of the instructor 7 15-073--To what degree do l'ho tests and quizzes given reactor operators at TMI-1 by the LLeensee's Training Division suffer from the following inadequactes , all of which were noted in the training of TMI-2 operators in NUREG/CR-1270, volume 1, et pages95-96i
- a. Tests and gutzzes fall to reflect job requirements and are not based upon specific trainLng objectives.
- b. Tests and quizzes are not performance based and criterton referenced.
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- c. Tests and quizzes should be methods of providing feedback to operators as to their performance strengths and weaknesses, but are not.
- d. Tests and quizzes fall to measure operator capability to follow procedures, diagnose transients and identify causitive factors, anticipate response of slowly reacting systems, and understand what is going on in the plant during emergency conditions.
15-074--To what extent does the Operator Accelerated Retraining Pro 6 ram (OARP) discussed in the Restart Report, Section 6, lack any or all of the following features:
- a. A formal, quantitative method for evaluating the effectiveness of the ~ course materials ,
including the accuracy and currency of the materials, and the adequacy and appropriateness of visual and audio materials utilized in the program.
- b. A formal method for updating and upgrading the training methods, materials, techniques, and content.
- c. Specific selection criteria for instructors which emphasize instructional skill as well as plant control skills,
- d. Task specific analysis of operator duties in the control room to assure that the OARP addresses each necessary task.
- e. Sufficient simulator training which will allow operators to develop necessary skills , as opposed to use of the simulator as a demonstration tool.
15-075--To what extent does Licensee's review of LER's , Abnormal Occurrence Reports, and other operational data relevant to control room operations incorporate human factors engineering perspectives, so as to ensure that the under-
lying causes of personnel error are determined and the cause source corrected so as to prevent future occurrences of the same or similar errors 7 15-076--Identify by name, title, and position within Licensee organization, any and all persons on the PORC with formal training in the fleid of human factors engineering.
17-001--In reference to scenario "B" in Contention #17 as admitted by the Board, could the diesel generator therein described have been placed in an operable condition, given the facts of the conditions as they existed on 28 March 1979 at Unit 2 of TMIT If so, how--be specific. If so, could this have been done without significant risk to the health and safety of the person or persons involved in placing the generator in an operable condition 7 i
17-002--In reference to scenario "B" in Contention #17 as admitted by the Board, if a total offsite power loss had occurred before the diesel generator referenced in Contention #17 had been placed in an operable status ,
could the diesel generator have been placed in an operable condition before significant core melting would have )
occurred, given the situation as occurred on 28 March 1979 at TMI-27 l
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4 17-003--What is Licensee's opinion regarding the impact on the sequence of events at Unit 2 and their subsequent
," resolution" if personn,el from Unit I had not been available for assistance?
17-004--Given the description of secnario "F" in Contention #17 as accepted by the Board and given the sequence of events as they transpired on 28 March 1979 at TMI-2, would the venting described in scenario "F" have required any type of protective action in order to protect the pubile health and safety? If so, describe. If so, also describe whether or not such protective actions could have been implemented in sufficient time to be of use in reducing public exposure to radiation.
17-005--Assuming that the condition of the Unit 2 reactor had been as described in scenario "C" in Contention #17 as accepted by the Board, to what extent would the total amount of radiation released to the containment as well as to the auxiliary building (from all pathways) have been increased beyond what was so released during the I Unit 2 accident on 28 March 19797 Specify in your answer any increases in the following isotopes
- a. Iodine-131 , c. Cesium-137
- b. Strontium-90 d. Krypton-85 l
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17-006--Inasmuch as the Battelle, Columbus Laboratories report on the TMI accident and alternative sequences (NUREG/CR-1219) identifies as Case 8 " loss of all AC electric power",
and inasmuch as this proposed case falls within the scope of scenario "B" in Contention #17, and inasmuch as the Battelle report predicts complete core meltdown by 2.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> if power is not restored, explain why the Licensee should not be required to install meltdown mitigation features at IMI-1 prior to restart to ensure sufficient time for evacuation in the event of a complete core meltdown.
17-007--Given the facts as described in Interrogatory 17-006 (above), reconcile Licensee's Emergency Plan with the fact of potential complete core meltdown by 2.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> into this particular accident sequence, i.e. , how does this fact change the validity of the assumptions and bases utilized in determining the sizes of Licensee's proposed EPZ for plume exposure?
01-010--Explain why the proposed high-radiation containment isolation signal cannot be made to comply with the single-failure criterion.
01-011--In reference to Licensee's proposed use of Reactor Trip in place of HPI initiation as a diverse containment isolation signal, is there any condition under which
~_
HPI injection could occur in the absence of Reactor Trip or after the Reactor Trip signal has been cleared?
If so, identify and describe each such condition.
01-012--By what date will Licensee have developed the by-pass and over-ride procedures for containment isolation?
Will these procedures be submitted to NRC for review and approval prior to proposed restart?
04-012--Provide a copy of Licensee's REMP, including locations of all sampling devices.
04-013--Identify any and all changes made to Licensee's REMP since the Unit 2 accident in March 1978.
05-007--Does Licensee have installed at Unit 1 or plan to install effluent radiation monitors capable of remaining on-scale during the highest release period which would be associated with a complete core melt accident? If not , explain why not. If so, specify each such monitor, including its location, model and manufacturer, and operating range.
09-004--Does Licensee plan to ut,ilize radiation monitoring devices which are capable of transmitting radiation monitoring readings directly to the Unit 1 control room
from remote locations off-site? If so, specify the number, model and manufacturer, and locations for such devices. If not, explain why such devices will not provide substantial protection of public health and safety by providing rapid and direct measurement of radiation leve1s in the environment, measurements which could be utilized in providing information to off-site authorities on which protec'tive action decisions could be made.
09-005--Does Licensee have or plan to have before restart the Atmospheric Release Advisory Capability System (ARAC)7 If not , explain why this system will not provide substantial additional protection of public health and safety, particularly in view of the potential for personnel error in calculating off-site radiation dose rates.
09-006--Has Licensee ever requested that the NRC install ARAC at TMI-17 If so, provide documentation of this request, including NRC response. If not, why not?
11-006--How many hydrogen recombiners of the type Licensee intends to install at Unit I would be require.d to successfully recombine the amount of hydrogen generated during the IMI-2 accident?
11-007--In reference to Licensee's answer to Interrcgatory 11-006, what is Licensee's judgment as to any potential impact on containment integrity which might result from the installation of that number of hydrogen recombiners?
11-008--Has Licensee investigated alternative means of controlling hydrogen gas concentrations in the Unit I containment (other than hydrogen recombiners and venting)? If so, ;
i specify the methds investigated, who performed the investigations, and the results of these investigations , l l
If not, explain why not. 1 13-010--Does Licensee's proposed new computer for Unit I have i
the capability of determining position on fault-trees, and displaying such data to plant operators? If not, is the computer capable of being so modified as to provide this capability?
13-011--Does the new computer proposed for Unit I have the capability of providing CRT display output to several CRT's simultaneously, each displaying different data? I 13-012--Does the new computer proposed for Unit I have the capability of providing hard copy of graphs of
! plant operating parameters?
13-013--Does the new computer proposed for Unit I have the I canability to crowlde dame Ro a mamaRm-eff-mies
location 7 If not, why not? Has either the NRC or the Commonwealth of Pennsylvania ever requested that computer display information be transmitted directly to them from either the Unit 1 or 2 control room? If so, provide documentation regarding such requests, including Licensee's responses to such requests.
14-003--List and describe each and every change in personnel at the management or supervisory level, directly affecting the operation of Unit 1, which have been made since the 28th of March 1979. For each new person hired by Licensee, describe'his new position, professional qualifications , and experience in the nuclear power fleid.
14-004--Inasmuch as NRC's I & E Special Review Group has cited quality assurance / quality control as being the " master control system available to management to assure that ,
l all management control systems are operable and effective ;
in terms of .providing safety", describe each and every change made to Licensee's QA/QC program at Unit I since the Unit 2 accident on 28 March 1979. l l
14-005--Describe the mechanisms within Licensee organization which assure that Licensee management reviews work performed by its personnel to verify that its directives
o . .
r and policies are effectively carried out on a timely basis.
14-006--Identify and describe any and all new technical capabilities which have been added by Licensee since the Unit 2 accident, specifying under what terms these capabilities ;
are to be provided and within what time frame they ;
I are available in the event of a serious accident at J l
Unit 1.
1 l
Respectfully submitted, !
h I V ;
Steven C Sholly W l 304 Souti Market StEeet l Mechanicsuurg, PA 17055 h--717-766-1857 w--717-566-3237 DATED: 25 February 1980 l l 1
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of MEIROPOLITAN EDISON COMPANY D ck No -289 (Three Mlle Island, Unit 1)
CERTIFICATE OF SERVICE I hereby certify that single coples of INTERVENOR STEVEN C. SHOLLY FOURIH SET OF INTERROGATORIES AND DOCLMENT REQUESTS TO LICENSEE were served upon the persons on this service ilst by depostt in the United States Mall, postage prepald, on this 25th of February 1980.
. //1 .
bl Steven C. Sholly Ivan W. Smith, Esq. Docketing and Service Section Chairman, Atomic Safety and Office of the Secretary Ltcensing Board U.S. Nuclear Regulatory U.S. Nuclear Regulatory Commission CommissLon Washington, D.C. 20555 Washington, D.C. 20555 John A. Levin, Esq.
Dr. Walter H. Jordan Assistant Counsel Atomic Safety and Licensing Pennsylvanta Pubile Board Panel Uttllty Commission 881 West Guter Drive P. O. Box 3265 Oak Ridge, TN 37830 Harrisburg, PA 17120 Dr. Linda W. LLttle Karin W. Carter, Esq.
Atomic Safety and Licensing Assistant Attorney General Board Panel Commonwealth of Pennsylvanta 5000 Hermitage Drive 505 Executtve House Raleigh, NC 27612 P.O. Box 2357 Harrisburg, PA 17120 -
James A. Tourte11otte, Esq.
Office of the Executive Legal DLrector (OELD)
U.S. Nuclear Regulatory Commisston Washington, D.C. 20555
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John E. Minnich Holly S. Keck Chairman Legislation Chairman Dauphin County Board of ANGRY Commissioners 245 West Philadelphia Street Dauphin County Courthouse York, PA 17404 Front and Market Streets Harrisburg, PA 17101 Robert Q. Pollard Chesapeake Energy Alliance Walter W. Cohen, Esq. 609 Montpelier Street Consumer Advocate Baltimore, MD 21218 Department of Justice l 14th Floor, Strawberry Square Chauncey Kepford !
Harrisburg, PA 17110 Environmental Coalition on l Nuclear Power {
Jordan D. Cunningham, Esq. '33 Orlando Avenue i Attorney for Newberry cate College, PA 16801 1 Township TMI Steering Committee Marvin-I. Lewis 2320 North Second Street 6504 Bradford Terrace Harrisburg, PA 17110 PhiladelphLa, PA 19149 Theodore A. Adler, Esq. Marjorie M. Aamodt Attorney for TMIA R.D. # 5 Widoff, Reager, Selkowitz, Coatesville, PA 19320 and Adler P.O. Box 1547 George F. Trowbridge, Esq.
Harrisburg, PA 17105 Attorney for Licensee Shaw, Pittman, Potts, &
Ellyn Weiss, Esq. Trowbridge Sheldon Harmon, & Weiss 1800 M Street, NW Attorney for UCS Washington, D.C. 20036 Suite 506 1725 I Street, NW Washington, D.C. 20006 Karen Sheldon, Esq.
Sheldon, Harmon & Weiss Attorney for PANE !
Suite 506 l 1725 I Street, NW l l
Washington, D.C. 20006 l t
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