ML19321A559
| ML19321A559 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/21/1980 |
| From: | Swartz L NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Sholly S AFFILIATION NOT ASSIGNED |
| Shared Package | |
| ML19321A555 | List: |
| References | |
| NUDOCS 8007230614 | |
| Download: ML19321A559 (33) | |
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'TAFF 7/21/80 s
h UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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METROPOLITAN EDIS0N COMPANY,
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(Three Mile Island Nuclear Station Unit 1)
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NRC STAFF RESPONSE TO INTERROGATORIES SUBMITTED BY STE'/EN C. SHOLLY In accordance with 10 C.F.R. 5 2.72L and 10 C.F.R. 5 2.744, the Staff has responded to interrogatories submitted by 'teven C. Sholly on June 30,1980 relating to the SER.
Each interrogatory is restated and a response provided. Affidavits identifying the individuals who prepared the responses which are not attached will be forwarded at a later date.
Respectfully submitted, e
Lucinda Low Swartz Counsel for NRC Staff Dated at Bethesda, Maryland this 21st day of July,1980 8007230 @ I
O SER-001--4t page B-2 of the SER, the Staff discusses the necessity of submitting new Technical Specifications on matcers which are. issues in the Restart proceeding.
At what point are these new Technical Specifications due to be submitted; and when may intervenors expect to receive copies of these proposed changes?
If this date is expected to occur af ter litigation begins, and any of the proposed Technical Specifications concerns one or core of my contentions, what mechanisms exist to permit litigation on these revised Technical Specifications?
Response
As noted on page B-2 of the Till-l Restart Evaluation (SER), the licensee has proposed revised Technical Specifications for some order items, and the corresponding open items have been resolved.
Draft Technical Specifications are contained in Section 11 of the licensee's Restart Report.
From the status summary in Section B of the SER, it may be seen that Technical Specifications remain to be submitted for items 79-05B-7 of Order Item 2 and 2.1.7.a of Order Item 8.
At this time, the staff does not have the licensee's schedule for submittal of these Technical Specifications.
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SER-002--In the second full paragraph on page Cl-2, it is stated that,
" licensee has identified' the options available to the operator to prevent overfill, and we concur that operator action can be performed in the time available." It is further stated in the fourth full paragraph that operator i
training on this subject has been conducted. With regards to these state-ments:
(a) What is the likelihood that the indications of steam generator overfill will be misinterpreted under the stress conditions of an accident?
(b) What are the results of such misinterpretations in terms of impact on cooling the reactor core?
(c) What would be the impact of instrumentation failures on correct inter-protation on symptoms of steam generator overfill?
Response
(a)
In clarification of the Restart SER (page C8-37), the stean generator level instrumentation currently available to the operator consists of a total of five separate clearly marked gauges for each steam generator.
These gauges identify three steam generator level ranges, two indicate the startup range, two indicate the operating range and one indicates the full rarge of steam generator level which includes the startup and operating ranges.
However, the operator will be observing only three gauges (one of each range) at any one time. He may manually select one of the other redundant startup or operating range indicators as he sees fit. Of the three level indicating ranges, only the full range level gauge will actually indicate i
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O levels as high as those achieved when approaching overfill. A high steam generator level alarm is also provided approximately twenty feet below the overfill point. This alarm will alert the operator to take appro-priate action to avoid overfilling.
From the alarm setpoint, overfill will occur in approximately seven minutes at the maximum expected emergency feedwater flow rate. As indicated in the Restart SER, a minimum of approximately ten minutes for operator action is available when stean generator level begins to increase from the normal operating point when a loss of offsite power occurs.
In addition, plant emergency procedures include caution statements to alert the operator to watch for potential overfill situations. Based on the above, we consider it unlikely that steam generator level indication will be misinterpreted by the control room operator even under stressful conditions, and time is available to take action to prevent overfilling.
(b) We do not expect misinterpretations of steam generator level indication j
because of the measures discussed in item (a) above.
(c) New steam generator level instrumentation which is designed to meet safety grade requirements including redundancy will be installed by approximately mid 1981. Therefore, a total failure of the level indica-tion is not postulated.
In addition, as described on page C8-37 of the Restart SER, the existing steam generator level instrumentation will be i
upgraded prior to restart to also meet the single failure criterion and thereby minimize the likelihood of its total failure.
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l SER-002--In the second full paragraph on page Cl-2, it is stated that,
" licensee has identified the options available to the operator to prevent overfill, and we concur that operator action can be performed in the time available." It is further stated in the fourth full paragraph that operator training on this subject has been conducted. With regards to these statments:
(a) h t is the likelihood that the indications of steam generator overfill will be d sinterpreted under the stress conditions of an accident?
(b) h t cre the results of such misinterpretations in terms of inpact on cooling the reactor core?
l (c) ht would be the inpact of instrmentation failures on correct interpretation of synptcrrs of steam generator
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overfill?
i (d) With regards to operator training on this topic, what audit methods exist to ensure that operator knowledge on the prob 1m of steaa generator overfill and its synptoms and corrective measures r eains at a high degree of efficiency?
(e) ht is the frequency of retraining on this topic?
ANSWER (d) Prior to restart the NRC will administer examinations, that will cover anong other items, the feedwater systen, including the energency feedwater systen.
To maintain a high degree of efficiency in the long term, the licensee will conduct retraining on this topic as part of the operator requalification program noted below.
(e) Requalification prograns require the administration of an annual examination and enrollment in a lecture series in areas where an individual exhibited weaknesses.
In addition, in a letter, dated March 28, 1980, (copy enclosed) we informed licensees to nodify their requal-ification programs to require that cpecific control manipulations during nonaal, abnonnal and emergency conditions be conducted at minimun frequencies.
Among the required exercises which would address operator response on this topic are:
1.
Ioss of normal feedwater or nonnal feedwater systen failure.
2.
loss of all feedwater (normal and energency).
3.
Ioss of instrument air.
4.
loss of electrical power.
As a mininun, iten 2 nust be performed annually and items 1, 3, and 4 nust be performed at least biennially.
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SER-003--In the second full paragraph on page Cl-2, it is stated that Licensee intends to modify the present EFW system by providing a " fully safety grade system" at an unspecified date in the future. With regards to this comit-ment:
(a) Will such a modification be made a requirement by the Staff?
(b) If not, why not? If so, by what date?
(c) Specify the actions which will be taken in the interim to provide a degree of protection equivalent to a safety grade systen.
(d) Explain the basis for concluding that it is acceptable for such modifi-cations to be made in the long term, as opposed to requiring their implementation prior to restart.
Response
(a) Yes. The stafT has required a fully safety grade emergency feedwater (EFW) system at Till-1. Discussion of this requirement is provided on Restart SER pages C8-34 thru C8-38.
(b) The safety grade modification is to be installed within 60 days after receipt of the required equipment as indicated on page C8-37 of the Restart SER.
(c) Actions to be taken to improve emergency feedwater system reliability and provide what the Staff considers adequate protection prior to i
installation of the safety grade modification are identified on page C8-37 of the Restart SER.
(d) The actions indicated by reference in item (c) above constitute the basis for our conclusion on the acceptability of the long tert.; schedule fdr the fully safety grade EFW system modification.
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SER-004-- h second full paragraph on page Cl-6 provides for a daily check of punp and valve alignments in the EFW syste.
% y should this check not be performed on a shift-by-shift basis, and why would this not provide more protection against misa-ligments such as occured at IMI-2 than would a daily check?
ANSWER h daily che.:k of the punp and valve alignments in the EFW systm refers to an actual walk through of the systs.
In addition to this daily check, proper valve and pucp position indication in the control roce is verified on a shift-by-shift basis.
The shift checks of EFW systs position indication are discussed as parts of IE Bulletin 79-05A, item 5 (Page02-5), IE Bulletin 79-05A, its 7 (PageC2-6), and NUREG-0578, item 2.2.1.c (Page 8-55).
4-5_hP,-005--Provide copies of documents (if not already served on the parties) or provide citations which describe the control room annu iciation system for auto start of the AFW system (pages Cl-7 and 8 of SER).
Response
The design of the control room annunciation for automatic initi,ation of the emergency feedwater system is shown in electrical elementary diagrams SS-209-755 and 756 revision IC-1 which were included in Amendment 18 as an attachment to Question 6 of Supplement 1, Part 2 to.the Restart Report.
SER-006--Pages Cl-9 and 10 of the SER discuss an "AC-independent backup air system" for prevention of loss of air to the EFW initiation system. The sec-tion cf the SER in question states that this provision will prevent con-tinuous cycling of the safety valves." Is this a reference to the code safety valves and PORV?
If not, to what safety valves is this statement referrina? If so, is the specified two-hour period adequate to prevent cycling of these valves under conditions which could lead to core uncovery (specify the studies which confirm your position and provide copies if not already served upon the parties)?
Response
As stated on page Cl-9 of the Restart SER, the safety valves referred to are those which provide overpressure protection for the emergency feedwater pump turbine (MS-V22A & B). By providing a backup air supply to the turbine steam supply pressure regulating valve, the supply pressure can be maintained at the normal operating point below the setpoint of the EFW pump turbine safety valves and thereby prevent their lifting.
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SER-007--Referring to Item 7 on page Cl-10, for what length of time are the motor-driven EFW pumps qualified to operate under the assumed environ-mental conditions resulti.ng from a postulated main steam line break in the Intermediate Building?
Resnonse:
The Staff has been made aware of LER 80-012/0IT-0 dated July 11, 1980, which describes certain nonconservatisms in the original Intermediate Builcing main steam line rupture analysis. The LE'R is a result of the licensee's reevaluation of environmental qualification of safety related equipment in response to IE Bulletin 79-01B. As a result of this information, the environ-mental qualification of the emergency feedwater pump motors will be rereviewed as part of the restart effort. Until this review is completed, we cannot satisfactorily reply to the above interrogatory.
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1 SER-008-Referring ti Item Id, Analysis of Small 8reaks, supply copies of the following documents:
(a)
Letter to D.C. Slear, GPU, f rom G.T. Fairburn, B&W, "HPI Flows During Sr 11 Break Transients",16 April 1980.
(M B&W, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant",
and supplements referenced in Item Id discussion.
Response
A copy of the documents referenced in (a) and (b) is enclosed.
Copies cf these dccuments will be na'e rnilfle in the local nub'.ic c:ccunent roers in harrisburn and in York, Per.nsylvania and it the PuMic Docu.cnt Recr in Washington, D.C.
Copies will be provided to the Licensing Board.
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Sholly Interroga< y SER 9 Discussion on page Cl-13 concerning the L0gA analyses for TMI-1 implies that with break sizes of less than 0.01 ft, core uncovery is not predicted and peak cladding temperatures were less than 800 oegrees Fahrenheit. The discussion states that these results are applicable to TMI-l since the operators have the ability to start redundant EFW and HPI pumps from the control room, assuming a failure of the automatic EFW initiation system.
How sensitive is the analysis of core uncovery and fuel cladding temperatures to operator error in initiating manual EFW or HPI flow following ft ~'ure of the automatic initiation system? What operator training on these a,,as has been conducted and what audit procedures exist to ensure that operator awareness of the problem and its symptoms and correction remains at a high degree of efficiency.
Response
The analyses performed by B&W indicate that the core will be sufficiently cooled following a small break LOCA so long as the core remains covered with water.
Provided that the core remains covered, the cladding temperature was found to continuously decrease from its initial value so that the initial value of 720 i would be the peak cladding temperature.
For break sizes of.01 ft2 and smaller, the analyses indicated that Emergency Feedwater (EFW) would be required under certain conditions to depressurize the primary system sufficiently to actuate the High Pressure Injection System. A sensitivity study was performed to evaluate the time available to the operator to actuate HPI or EFW manually to prevent core uncovery (Ref.1). This time was found to be 20 minutes.
For operator action at any time within the 20 minute period, the cladding temperature response would be approximately the same since the core would remain covered. Analyses were not performed assuming delay times longer than 20 minutes.
Longer delay times would result in some core uncovery which would reduce the core heat removal and cause the cladding temperature to increase. Although a temporary core uncovery could still produce peck cladding temperatures less than the 2200 F limit of 10 C.F.R. 50.46, for an extended period of uncovery that limit would be exceeded.
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-2 The TMI-l small break LOCA procedure; require that the operator immediately verify that emergency feedwater is available. Guidance for manually actuating the system if it fails to start automatically is provided in Attachment 4 to the Small Break LOCA Procedures (Ref. 2).
Operator training in small break loss of coolant accidents has been extensive including simulator training. An individual's knowledge and understanding of symptoms and corrective actions required to deal with small break loss of coolant accidents will be explored in depth in the NRC examinations and will be maintained at a high degree of efficiency in requalification programs which require specific annual retraining in this area.
References:
1.
Babcock & Wilcox Report, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant,"
May 7,1979.
2.
Three Mile Island Nuclear Station Unit 1 Emergency Procedure 1202-6B, Loss of Reactor Coolant / Reactor Coolant Pressure (Small Break LOCA)
Causing Automatic High Pressure Injection, Revision 3, June 6,1980.
Sholly Interrogatory SER 10 Page Cl-14 discusses a pending schedule for implementation of automatic closure of the PORV on '. low pressure. Will such a requirement.be implemented by the Licensee before Restart? If not, why not?
Response
The pressurizer PORV at TMI-l is designed to automatically close at reactor system pressures of less than 2450 psig.
In addition, block valve upstream of the PORV is also provided as a backup means of isolation. Following a small break LOCA the operator is instructed to immediately close the block valve.
This action would terminate water loss from the system in the event that the LOCA was produced by a stuck open PORV. As discussed on page Cl-14 of the Staff SER for TMI-l restart, the NRC B&O Task Force recommended that a system which will cause the block valve to close when the RCS pressure has decreased below a preset pressure be installed at all B&W plants. This system is not required to be installed before restart. A detailed evaluation of all accident and transient conditions for which automatic block valve closure would be actuated will be made in the reliability and risk assessment studies described in Task II.C of the NRC Action Plan. Analyses by B&W indicate, however, that the core will be adequately cooled in the event that the PORY sticks open even if the operator fails to close the block valve. The NRC Staff will therefore not require that the automatic closure system be installed unless the studies described show that in the overall plant safety will be increased.
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m SER-Oll--Item le, Retraining of Operators, on page Cl-16, ccntains a two-paragraph evaluation of Licensee's OARP. Such an eval-uation, in my opinion, does not reflect the depth of review required to conclude that the program neets the requiremnt as stated in Item le. The OARP includes training on a ntunber of very inportant subject areas. Sinply describing the number of hours of training, and describing how operators will be tested on the naterials does little to ensure that the prograns's training materials and nethods ensure operator retention of the needed knowledge, nor does it ensure that the necessary training has in fact been incorporated into the progrm. The Essex Cor-poration evaluation of operator training at IMI-2 (NUREG/CR-1270) found serious flaws with previous Licensee operator training programs. An in depth evaluation of Licensee's program for content, accuracy, training nethods, audit procedures, and 1
followup testing and training is necessary to ensure that the program is adequate to assure public health and safety.
Provide documentation which indicates the depth of Staff's review of the OARP. Discuss the inadequacies described in NUREG/CR-1270, pages80-101, with respect to the Licensee's OARP and explain how these inadequacies have been corrected in OARP.
ANSWER In the discussion starting on page C6-5, ' Training of Operating Staff," we noted that the license 9 was required to satisfy several requirenents in the area of licensed operator training.
To verify these requirements had been net, we reviewed selected lesson plans, audited training lectures and their associated quizzes, attended 'IMI-2 related B&W sinulator denonstrations, and reviewed the written examinat ~n administered by the contracted Auditor Group. Based upon our evaluations, we have determined that the licensee has satisfied the training requirements of this Order and has conducted a program which is adequate to give us reasonable assurance that such training will produce safe and conpetent operators.
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The findings nede by the Essex Corporation in NUREG/CR-1270 regarding training are:
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Training The Met. Ed training program was in full compliance with government imposed standards concerning training.
M -2 training was deficient in that it was not directed at the skills and knowledges required of the operators to safety job requirements.
The essential operator skill is to be able to diagnose what is happening in the plant. Be nost effective training method of acquiring this skill is sinulation.
Only'5 percent of training time is used for sinulation training.
Training in eergency procedures was deficient.
Training at M -2 was deficient in its failure to provide for measurement of operator capabilities.
Training at M-2 was deficient in its training of inscructors.
Training at M-2 was deficient in its archaic approach to learning.
Training at M-2 r is deficient in that it was not closely associated with procedures.
Training at M -2 was deficient in ignoring the fact that operators are dealing with a slowly responding j
system.
'Ihe training program at M-2 did not provide for f.y a 5
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formal updating and tpgrading of methods, materials, and course content.
i Training at DiI-2 failed to establish in the crew the readiness necessary for effective and efficient perfor-mance.
Our review of OARP and Section 6 of the Restart Report indicated that:
The prograa is in full conpliance with goverment inposed standards concerning training.
1he training is directed at skills and knowledge required of operators.
Sir.ulator training has been changed to provide operators the opportunity to diagnose what is happening in the plant.
Training in emergency procedures is provided.
Training provides for evaluation of ope: stor cap-abilities.
The training manager and supervisors are required.
to have e.perience in education and training v
disciplines.
Wre nodern training methods are being used.
Training is closely associated with procedures.
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Training enphasizes react.or transients.
Provisions are made for tpdating and upgrading of methods, materials and course content.
Crew training, as w il as individual training, is being conducted.
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SHOLLY INTERROGATORIES ON EMERGENCY PREPAREDNESS DATED 30 JUNE 198 Question SER-012--With regard to Order Item 3, Short Term Emergency Preparedcess, identify the persons on the Staff who performed the review of Licensee's Emergency Plan and Appendices, specifying what areas each of the persons reviewed.
Identify the professional qualifications of each of the reviewers, referencing in particular experience in emergency planning.
If available, provide an approximate number of Staff man-hours which were spent reviewing Licensee's Emergency Plan. Provide Staff inputs into the review (i.e., memos, reports, notes, etc., which deal with Staff review of Licensee's Emergency Plan). Were any consultants external to the NRC used in the evaluation of Licensee's Emergency Plan? If so, provide copies of all documents which contain information on such reviews.
Specify with respect to all consultants their experience in the emergency planning field.
Answer i
Jack W. Roe performed the review of the Licensee's Emergency Plan and Appendices.
His professional qualifications are attached.
Approximately 400 staff man hours have been spent reviewing the Licensee's Emergency Plan.
Staff inputs into the review are attached.
Marvin Smith, Battelle - Pacific Northwest Laboratories j
assisted in the evaluation of the Licensee's Emergency Plan. His professional qualifications are attached.
Copies of documents which contain information on the review from the consultant are attached.
Copies of all the attachments will be placed in the local public document rooms in Harrisburg and Yozic, Pennsylvania and in the Public Document Room in Washinton, D.C.
Copies will also be provided for each.of the Licensing Board i
members.
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4 UNITED STATES OF AMERICA NUCLEAR RH;UIATORY COMISSION BEFORE THE ATOfIC SAFETY AND LICENSING BOARD In the Matter of MEIROPOLITAN EDISON COMPANY, _et _al.
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Docket No. 50-289
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(Three Mile Island, Unit 1)
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AFFIDAVIT OF BRUCE BOGER I, Bruce Boger, being duly sworn, do depose and state:
1.
I am erployed by the Operator Licensing Branch, Office of Nuclear Reactor Regulation of the United States Nuclear Regulatory Comission.
2.
The answers to Sholly Interrogatories SER-002(d), 002(e), 004, 009, and 011 were prepared in part by me.
I certify that the answers given are true and accurate to the best of my krmledge.
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Bruce A. B6ger (/
Subscribed and sworn to before me this 15th day of July 1980 AOL
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.iotar'y Public/ g',
i W COMMrssion expie 3q'
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s PROFESSIONAL QUALIFICATIONS LIST BRUCE A. B0GER Education June 1971 Received BSNE - University of Virginia June 1972 Received MENE - University of Virginia Work Experience June 1972 to Virginia Electric and Power Company June 1977 Surry Nuclear Power Station Assistant Engineer - Performed startup testing on Unit No. 2, Associate Engineer - Reviewed facility design modifications.
Engineer - Assisted the Supervisor-Engineering Services; trained for and received a Senior Reactor Operator License.
Supervisor - Engineering Services - Directed the activities of the onsite engineering s'taff.
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June 1977 to Virginia Electric and Power Company Saptember 1977 Richmond, Virginia Supervisor - Nuclear Engineering Services - Directed the activities of the offsite engineering staff in support of Surry Power Station.
October 1977 to U. S. Nuclear Regulatory Commission Present
'Bethesda, Maryland
~ Reactor Engineer in the Operator Licensing Branch - Administer licensing examinations to nuclear power plant and research reactor personnel.
Professional Affiliations Registered Professional Engineer - State of Virginia Member - American Nuclear Society e
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' UNITED STATES OF A! ERICA NUCLEAR RFEUIATORY 0@HISSION BEFDRE THE AINIC SAFETY AND LICENSING BOARD In the Matter of
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METROPOLITAN EDISON Or SANY, et al.
Docket No. 50-289 (Three Mile Island, d.t 1)
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AFFIDAVIT OF PAUL COLLINS I, Paul Collins, being duly sworn, do depose and state:
1.
I am employed by the Operator Licensing Branch, Office of Nuclear Reactor Regulation of the United States Muclear Regulatory Conmission.
2.
The answers to Sholly Interrogatories SER-002(d), 002(e), 004, 009, and 011 were prepared in part by me.
I certify that the answers given are true and accurate to the best of my knowledge.
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Paul F. Collins Subscribed and sworn to before me this 15th day of July,1980 D/d b
Marilyn dollehsteY Notary Public M S3MMfSSION EXPIRES JULY F,1982
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RESUME Paul F. Collins AEC-NRC Career 1969-Present i
Chief, Operator Licensing Branch 6
During my tenure as Chief, the number of operator and senior operator applications have increased from about 600 in 1969 to over 1300 in 1976.
ne number of examiners has increased from 4 full time and 9 consultant
. examiners to 8 full time and 17 consultant examiners. New procedures and l
practices introduced during this period have resulted in sufficient t
efficiencies to enable this function to be conducted with a minimam increase in staffing and administrative supports.
In addition to administering operator and se'nior operator license examinations the accomplishments of this branch, since I have have been Chief, have included the following:
1.
ne development of minimum training regairements to become eligible to sit for license examinations.
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%e develognent of specifications and acceptance criteria for
- Nuclear PoEer Plant Simulators used in the above training programs.
%ere are presently twelve simulators operating, under construction or planned.
3.
Development of Appendix A, 10 CFR Part 55, Operator Recualification Programs.
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- Faul F. Collins 4.
Assistance in the preparation of KNSH-1130, Utility Staffing-for Nuclear Power P1 ants.
5.
Publication of NUREG-0094, NRC Operator-Licensing Guide.
In addition to supervising and providing the prime inuput to the above, I have been an active member of ANS.
I served on the ANS 3.4 Committee that developed ANSI N546-1976, tedical Certification and Monitoring of Personnel-Reauiring-Ocerator - Licenses for-Nuclear Power Plants:
Presently, I am c member of the Reactor Operations Division Executive Comnittee; [
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In my present position I am the chief spokesman in NRC for all matters concerning operator licensing., As such, I have frequent contacts with middle and top management of utilities, reactor vendors, simulator manufactorers and training companies.
Experience Suanary 1952-1953' U.S. Army, 2nd. Lt. Transporation Corps. Stationed at U.S.
Port of Drbarkation, N.Y.C.
Assigned 'to the Management Division. Conducted efficiency studies and supervised machine accounting office.
1953-1958 Qu$lity Control Supervisor - E.I. duPont de Nemours & Co.
Savannah River Plant, Aiken, S. C.
Responsible for the receipt and inspection of all components used in the five production reactors. Also responsible for the receipt storage and shipnent for processing a'll heavy water used in the reactor e
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m Rsre Paul F. Collins department. Finally, responsible for the shipnent of spent fuel from the reactor department to reprocessing facilities.
Supervised four foremen and 32 operators.
1958-1963 Reactor Shift Supervisor.
Responsible for the routine operation 1
of one of the production reactors. My time was evenly divided between shift operations and performing engineering support to operations.
1963-1965 Instructor-Reactor Department--Operator Training School Taught courses involving all aspects of reactor operations and related basic principles of physics, chemistry and engineering.
1965-1969' Reactor Engineer-operator Licensing Branch-AEC Responsible for administering examinations to operators and senior operators at all types'of power dnd non-power reactor facilities. Became n'R Group Irader in 1967.
1969-Present Chief, Operator Licensing Branch Education BSME Rensselaer Polytechnic Institute 1952 M 7
UNITED STATES OF AMERICA NUCLEAR REGULATORY C0ft11SSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the flatter of
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METROPOLITAN EDIS0N COMPANY, et al.
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Docket No. 50-289
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(Three !!ile Island, Unit 1)
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AFFIDAVIT OF JARED S. WERMIEL I, Jared S. Wemiel, being duly sworn, do depose and state:
1.
I am a member of the Auxiliary Systems Branch in the Division of Systems Integration, Office of Nuclear Reactor Regulation of the United States Nuclear Regulatory Commission.
I am responsible for the auxiliary systems aspects of the safety review of assigned nuclear power plants, including Three Mile Island, Unit i Restart Program.
2.
The answers to Sholly Interrogatories SER-002a, b, c, SER-003, SER-005, SER-006, and SER-007 were prepared by me.
I certify that the answers given are-true and accurate to the best of my knowledge.
Y Jared S. Wemiel Subscribed and sworn to before ma this 16thday of July, 1980 t
o Nct:ry I'ublic
!t' Co'J.iission. expi res: July 1, 1982
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Jared S. Wermiel Professional Qualifications Auxiliary Systems Branch Division of Systems Safety Office of Nuclear Reactor Regulation I am a Reactor Engineer in the Auxiliary Systeris Branch in the Divisio~n of Systems Safety, Office of Nuclear Reactor Regulation, U.S. Nuclear Re5ulatory Comission.
In this position I perform technical reviews, analyses, and evaluations of reactor plant features pursuint to the con-struction and operation of reactors.
I received a Bachelor of Science Degree in Chemical Engineering from Drexel University in 1972. Since 1972 I have taken courses on PWR and BWR System Operation, Reactor Safety,'and Fire Protection.
My experience includes seven years with the Bechtel Power Corporation as a Systems Design Engineer engaged in the design of various nuclear power plant auxiliary and balance of plant systems. These have in-cluded cooling water systems, water treatment systems and fire protec-tion systems.
I joined the Auxiliary Systems Branch of the Comission in March, 1978.
Since joining the Comission I have performed safety evaluations on nuclear power plant auxiliary systems including main and auxiliary feedwater systems for the Virgil C. Sumer Nuclear Station, Palo Verde Nuclear Generating Station Units 4 and 5, Allens Creek Nuclear Generating Station, North Anna Power Station Units 1 and 2, Byron /Braidwood Stations and Enrico Fermi Atomic Power Plant Unit 2.
I have also reviewed various topical reports and provided comments on proposed ANSI Standards dealing with various auxiliarf systems.
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I have responsibility for the review of the following nuclear power plant t
auxiliary systems:
new and spent fuel storage, ipent fuel pool cooling, fuel handling, service water, component cooling water, condensate storage, ultimate heat sink, instrument air, chemical and volume control, main steam isolation valve leakage control, heating ventilating and air conditioning, fire protec-tion, portions of the main steam system, main feedwater, and auxiliary feed-water.
I am a registered Professional Engineer in the State of Maryland.
I am an Associate Member of the American Institute of Chemical Engineers.
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4 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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METROPOLITAN EDISON COMPANY, _et _al.
)
Docket No. 50-289 (Three Mile Island, Unit 1)
)
AFFIDAVIT OF JACK ROE I, Jack Roe, being duly sworn, do depose and state:
1.
I am a Emergency Planning Analyst, Office of Nuclear Reactor Regulation of the United States Nuclear Regulatory Commission.
I am responsible for reviewing the emergency planning of assigned nuclear power plants, including Three Mile Island, Unit 1 Restart Program.
2.
The answer to SHOLLY Interrogatory SER-12 was prepared by me.
I certify that the answers given are true and accurate to the best of my knowledge.
Y Subscribed and sworn to beforemethis[p_dayof On M rs a
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My Conunission expires
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JACK W. ROE, JR.
PROFESSIONAL QUALIFICATIONS OFFICE OF NUCLEAR REACTOR REGULATION U.S. NUCLEAR REGULATORY COMMISSION My name is Jack W. Roe, Jr.
I am an Emergency Preparedness Analyst in the Office of Nuclear Reactor Regulation.
My duties include the review and evalu-ation of nuclear power reactor emergency plans.
I hold a Bachelor of Science degree in Nuclear Science from the U.S. Naval Academy and a Master of Science degree in Nuclear Engineering from the University of Texas.
I joined the NRC in February of 1976 as a Coordinator for Technical Specifications.
From February of 1976 to May 1977 I participated in the development and implemen-tation of standard technical specifications for nuclear power plants, research and test reactors.
In May 1977, I was tran,ferred to the Reactor Safeguards Licensing Branch where my duties were to coordinate and perform reviews of site physical security plans.
In February 1979, I was transferred to the Retr. tor Safeguards Development Branch where mj duties included the evaluation and coor-dination of NRR's reactor safeguards programs and policies.
In September 1979, I was transferred to the Emergency Preparedness Branch where my duties include the review and evaluation of nuclear power plant ' emergency plans.
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Prior to joining the MRC, I spent 8 years in the U.S. Navy, resigning in February,1976, as a Lieutenant. While in the U.S. Navy I served on two nuclear powered submarines.
During my service in the U.S. Navy I received extensive training and experience in the operation and maintenance of naval nuclear power plants.
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEF0PE THE AT0f11C SAFETY AND LICENSING BOARD In the Matter of
)
METROPOLITAN EDIS0N COMPANY, Docket No. 50-289 ET AL.
(Three Mile Island Nuclear Station
)
Unit No. 1)
)
AFFIDAVIT OF HARLEY SILVER I, Harley Silver, being duly sworn, do depose and state:
1.
I am the Pro, ject Manager for the TMI-l Restart proceeding in the Division of Licensing, Office of Nuclear Reactor Regulation of the United States Nuclear Regulatory Commission.
2.
The answers to Sholly Interrogatories SER 001 and 008 were prepared by me.
I certify that the answers given are true and accurate to the best of my knowledge.
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-. W ley fEr Subscribed and sworn to before me tnis 17th day of July,1980.
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My Coramission Expires: July 1,1982
Harley Silver SenicDr_o[e:t :TaTager Div'ision of Pr; ject !*.ar.aoement Office o(Tuclear Reactor Regiilation i
f.iiclear~Regulafory Co r.ission Professional QualTfications I am a Senior Project Manager, responsible for managing the safety review for the Nuclear Regulatory Commission of assigned, plants, including Three Mile Island Unit 2.
I have served in this capacity since October 1973, and had been assigned Three Mile Island Unit 2 from May 1975 until mid-1979.
Since the fall of 1979, I have been assigned as the Project Manager of the TMI-l Restart Program.
I received the degree of Bachelor of Mechanical Engineering from f.e.
York University in 1949 and have subsequently taken graduate level courses in Engineering and Business Administration.
Between 1950 and 1952, I served in the United States Air Force as a First Lieutenant.
From 1952 to 1955 I was employed as a Design Engineer by the M. k'. Kellogg Company.
From 1955 through 1962, I was employed as a Project Engineer by architect-engineering firms, including Hydrocarbon Research, Inc. and Bechtel Associates.
Between 1963 and 1970, I was employed by the Westinghouse Electric Corp.
in both the Astronuclear Laboratory and Weapons Systems Department as, successively, Project Engineer, Supervisor of various design groups, and Manager of Systems Integration.
In 1971, I joined Offshore Power Systems as Manager of Component Engineering, in which capacity I remained until joining the fluclear Regulatory Commission.
I am a Registered Professional Engineer in the State of tiew York (Certificate fiumber 32892).
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