ML17193B357
| ML17193B357 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 04/28/1981 |
| From: | Kniel K Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| ISSUANCES-SP, NUDOCS 8104300135 | |
| Download: ML17193B357 (19) | |
Text
0
'~
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION THE ATOMIC SAFETY AND LICENSING BOARD 4/28/81 In the Matter of
)
)
)
)
)
Docket Nos. 50-237-SP 6
(~~~~i;~~~1o~~~~49-SP "-: /({lt\\.~~~\\.,. ~
COMMONWEALTH EDISON COMPANY (Dresden Station, Units 2 & 3)
-I f\\ \\' f, ?,
~0.-C z -\\
o.t~" "~
-9 U. s. t4\\l tA"'\\S!t""'"
/
co Affidavit of Karl Kniel on Unresolved Safety Issues ~,
. \\i;,
~-\\\\
My name is Karl Kniel.
I am employed by the U. S. Nuclear Regulato N
~
Commission in the position of Chief, Generic Issues Branch, Division of Safety Technology, Office of Nuclear Reactor Regulation.
A orincipal function of the r,eneric Issues Branch (GIB) is to pursue the resolution of specific generic issues designated as Unresolved Safety Issues (USis).
Each USI is managed both technically and administratively in accordance with an approved action plan describing the issue and the tasks to be performed within a defined budget and schedule. All Task Managers report technically to the GIB.
In my position as Chief, fiIB, I am responsible for supervising the work on technical resolution of USis and it is in that capacity that I have orepared this affidavit.
M.v exoerience and qualifications are further described in the attachment appended to this affidavit.
I certify that the contents of this affidavit are true and correct to the best of my knowledge.
Subscribed and sworn to before me, this.27tf. day of April, 1981.
~L~at;1
.*810.4 a oo/35 #
~I I
,. In its order dated January 26, 1981, the Licensing Board requested the parties to respond to the following Board Question 2:
Based on a review and analysis of the various generic un-resolved safety issues under continuing study, what relevance is there if any to the proposed spent _fuel pool modification?
Further, what is the potential health and safety implication of any relevant issues remaining unresolved?
Identification of Unresolved Safety Issues In responding to these questions, I have interpreted the boards reference to generic unresolved safety issues to mean those issues which the Commission has designated as Unresolved Safety Issues under Section 210 of the Energy Reorganization Act of 1974 as amended (PL 95-209) on December 13, 1977.
It is the NRC's view that the intent of Section 210 was to a~sure that plans were developed and implemented on issues with potentially signi-ficant public safety implications.
In 1978, the NRC undertook a review of over 130 generic issues addressed in the NRC program to determine which issues fit this description and qualify as "Unresolved Safety Issues" for reporting to the Congress. The NRC review included the development of proposals by the NRC staff and review and final approval by the NRC Commissioners.
This review is described in a report NUREG-0510, "Identification of Unresolved Safety Issues Relating to Nuclear Power Plants - A Report to Congress," dated January 1979.
The report provides the following definition of an "Unresolved Safety Issue:
11 "An Unresolved Safety Issue is a matter affecting a number of nuclear power plants that poses important questions concerning the adequacy of existing safety requirements for which a final resolution has not yet been developed and that involves conditions not likely to be acceptable over the lifetime of the plants it affects."
t:
~..
[
r Further the report indicates that in applying this definition, matters that pose "important questions concerning the adequacy of existing Safety requirements II were judged to be those for Whi Ch re sol Ution i S necessary to (1) compensate for a possible major reduction in the degree of protection of the public health and safety, or (2) provide a potentially significant decrease in the risk to the public health and safety. Quite simply, an "Unresolved Safety Issue" is potentially significant from a public standpoint and its resolution is likely to result in NRC action on the affected plants.
All of the issues addressed in the NRC program were systematically evaluated against this definition as described in NUREG-0510.
As a result, seventeen "Unresolved Safety Issues" addressed by twenty-two tasks in the NRC program were identified. The issues are listed below.
- 1.
- 2.
- 3.
- 4.
- 5.
- 6.
- 7.
- 8.
- 9.
- 10.
- 11.
- 12.
- 13.
- 14.
- 15.
- 16.
- 17.
"UNRESOLVED SAFETY ISSUES" (APPLICABLE TASK NOS.).
Waterhammer - (A-1)
Asymmetric Blowdown Loads on the Reactor Coolant System - (A-2)
Pressurized Water Reactor Steam Generator Tube Integrity - (A-3, A-4, A-5)
BWR Mark I and Mark II Pressure Suppression Containments - (A-6, A-7, A-8, A-39)
Anticipated Transients Without Scram - (A-9)
BWR Nozzle Cracking - (A-10)
Reactor Vessel Materials Toughness - (A-11)
Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports - (A-12)
Systems Interaction in Nuclear Power Plants - (A-17)
Environmental Qualification of Safety-Related Electrical Equipment -
(A-24)
Reactor Vessel Pressure Transient Protection - (A-26)
Residual Heat Removal Requirements - (A-31)
Control of Heavy Loads Near Spent Fuel - (A-36)
Seismic.Design Criteria - (A-40)
Pipe Cracks at Boiling Water Reactors - (A-42)
Containment Emergency Sump Reliability - (A-43)
Station Blackout - ( A-44)
~'
- t.
....* New "Unresolved Safety Issues" An in-depth and systematic review of generic safety concerns identified since January 1979 has been performed by the staff to determine if any of these issues should be designated as new "Unresolved Safety Issues.
11 The candidate issues originated from concerns identified in NUREG-0660, 11NRC Action Plan as a Result of the TMI-2 Accident, 11 ACRS reconmendations, abnormal occurrence reports and other operating experience.
The staff's proposed list was reviewed and commented on by the ACRS, the Office of Analysis and Evaluation of Operational Data (AEOD) and the Office of Policy Evaluation. The ACRS and AEOD also proposed that several additional "Unresolved Safety Issues" be considered by the Commission.
The Commission considered the above information and approved the following four new "Unresolved Safety Issues:"
- 18.
- 19.
- 20.
- 21.
Shutdown Decay Heat Removal Requirements - (A-45)
Seismic Qualif1cation of Equipment in Operating Plants - (A-46)
Safety Implications of Cont,rol Systems - (A-47)
Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment - (A-48)
A' description of the above process together with a list of the issues considered is presented in NUREG-0705, "Identification of New Unresolved Safety Issues Relating to Nuclear Power Plants, Special Report to Congress, 11 dated March 1981.
An expanded discussion of each of the new "Unresolved Safety Issues" is also contained in NUREG-0705.
With the exception of Tasks A-9, A-43, A-44, A-45, A-46, A-47, and A-48, Task Action Plans for the generic tasks above are included in NUREG-0649, 11Task Action Plans for Unresolved Safety Issues Related to Nuclear Power Plants.
11 A technical resolution for Task A-9 has been proposed by the NRC staff in Volume 4 of NUREG-0460, issued for corranent.
This served as a basis for the staff's proposal for rulemaking on this issue. The Task Action Plan for Task A-43 was issued in January 1981, and the Task Action Plan for A-44 was issued in July 1980~ Task Action Plans for the newly designated USis A-45, A-46, A-47 and A-48 are being developed.
Each Task Action Plan provides a description of the problem; the staff's approaches to its resolution; a general discussion of the bases upon which continued plant licensing or operation can proceed pending completion of the task; the technical organizations involved in the task and estimates of the manpower required; a description of the interactions with other NRC offices, the Advisory Committee on Reactor Safeguards and outside organizations; estimates of funding required for contractor-supplied technical assistance; prospective dates for completing the tasks; and a description of potential problems that could alter the planned approach or schedule.
In addition to the Task Action Plans, the staff issues NUREG-0606, "Office of Nuclear Reactor Regulation Unresolved Safety Issues Summary, Aqua Book" on a quarterly basis which provides current schedule information for each of the Task Acti-0n Plans.
Progress reports to Congress regarding the resolution of USis have been reported annually in the NRC Annual Report since 1978.
The most recent progress report is presented on pages 42-57 of the NRC 1980 Annual Report.
Discussion of Unresolved Safety Issue Tasks as They Relate to Dresden Unit Nos. 2 and 3 Spent Fuel Pool Modification I have reviewed each of the twenty-one Unresolved Safety Issues listed above by task and addressed their applicability individually to the Dresden Spent Fuel Pool Modification.
As a part of the discussion for each task,*
I have provided a one or two sentence description of each USI.
More detailed descriptions are provided in the Task Action Plan for each
)
USI and in NUREG-0705 for the four new issues.
L.
- A-1 Waterhammer This task is.directed toward icentifyinq potential waterhammer events in various systems with safety significance and providing guidance for the,
licensing review orocess.
The fuel oool cooling system is a low pressure system involving circulation of liquid water at temperatures below the atmospheric boiling point and therefore significant waterhammer events are not likely. Operating experience has not identified waterhammer events in fuel pool cooling systems of safety significance. Resolution of this USI task is not relevant to the Dresden Spent Fuel Pool Modifica-tion.
A-2 Asymmetric Slowdown Loads on the Reactor Coolant System The purpose of this task was to develop a better understanding for the ootential for larqe asymmetric loads resulting from primary system oipe failure in pressurized water reactor plants and then to develop acceptable criteria and ouidelines for evaluatinq plant analyses. A technical resolution has been completed and published as "Asymmetric Slowdown Loads on PWR Primary Systems," NUREG-0609.
This resolved USI is not applicable to Boiling Water Reactors such as the Dresden units, and therefore is not applicable to the Dresden Spent Fuel Pool Modification.
A-3, 4, 5 Westinghouse, Combustion Engineering, and Babcock and Wilcox Steam rienerator Tube Integrity The Drimar.v concern of these Task Action Plans is the capability of degraded tubes to maintain their integrity during normal operation and under postulated accident conditions (LOCA or main steam line break) with adequate safety marqins and the establishment of inspection and plugging criteria needed to provide assurance of such integrity. This USI is not relevant to Boiling Water Reactors, and thus is not applicable to the Dresden Spent Fuel Pool Modification.
' r 6 A-6 Mark I Containment Short-Term Program During large scale testing of the Mark III containment system design, new suppression pool hydrodynamic loads associated with a postulated loss-of-coolant accident were identified which had not been explicitly considered in the original design of the Mark I containment system.
The pur~ose of the Mark I containment short-term program (STP) was to pro-vide assurance that for a 11most probable load 11 the Mark I containment system of each operating BWR facility would maintain its integrity and functional capability during a oostulated LOCA.
A technical resolution, 11Mark I Containment Short Term Program," NUREG-0408, was completed in December 1977 and imolementation for each Mark I operating reactor including Dresden Unit Nos. 2 and 3 is complete.
This resolved USI is not applicable to the Dresden Spent Fuel Pool Modification.
A-7 Mark I Containment Long~Term Program The purpose of the Mark I Containment Long-Term Program was to (1) establish des1gn basis LOCA loads which are appropriate for the life of the facility and (2} restore the originally-intended design safety margins. A technical resolution, "Safety Evaluation Report - Mark I Containment Long-Term Program, 11 NUREG-0661 was issued in July 1980.
On January 13, 1981, the Commission issued orders to twenty-two operating Mark I *plants including Dresden Unit Nos. 2 and 3 to implement the requirements set forth in NUREG-0661.
This implementation is underway.
This resolved USI is not ap~licable to the Dresden S~ent Fuel Pool Modification, as it relates solely to the reactor containment.
A-8 Mark II Containment Pool Dynamic Loads ~ Long-Term Program The purpose of the Mark II Containment Long-Term Program is to assure that all loading conditions have been correctly identified and designed for in the Mark II containment.
The Dresden units do not have Mark II containments, and thus this US! is not applicable to the Dresden Spent Fuel Pool Modification.
A-9 Anticipated Transient Without Scram (ATWS)
The purpose of this task was to determine the probability and potential consequences of an ATWS event and develop an appropriate safety objective.
After many years of discussions and evaluations of vendor models and analyses the staff issued its prooosed resolution of this issue, "Anticipated Transients Without Scram for Light Water Reactors," NUREG-0460, Volume 4, For Comment, March 1980.
The anticipated transient without scram issue has been recommended for rulemaking to the Commission which is currently scheduled for the summer of 1981. This LISI relates to reactor operation, and is not relevant to the Dresden Spent Fuel Pool Modification.
A-10 BWR Feedwater Noztl e *cracking The ouroose of this task was to review the operating experience with cracking in BWR feedwater nozzles and control rod drive return line nozzles and to determine a course of action for resolution of this problem.
The staff issued "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking," NUREG-0619, November 1980, as its techni-cal resolution.
Letters requiring implementation of this resolution were sent to operating BWR plants including Dresden Unit Nos. 2 and 3 on November 13, 1980. All commitment letters from the licensees have been received and are being reviewed by the staff. This resolved USI is not applicable to the Dresden Spent Fuel Pool Modification, as it relates solely to the reactor operation and equipment inside containment.
A-11 Reactor Vessel Materials Toughness Resistance to brittle fracture is described quantitatively by a material property generally denoted as "fracture tou9hness.
11 The objective of Task A-11 is to develop safety criteria to allow a more precise assess-ment of safety margins during normal operating transients and accident conditions in older reactor pressure vessels with marginal fracture toughness.
This USI is not applicable to the Dresden Spent Fuel Pool Modification, as it relates solely to the reactor vessel.
A-12 Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports The purpose of this task was to review the adeauacy of fracture toughness of steam generator and reactor coolant pump supports, with respect to postulated accident loads (especially from a loss-of-coolant accident or earthquake).
The staff's proposed resolution, "Potential for Low Fracture Toughness and Lamellar Tearing on PWR Steam Generator and Reactor Coolant Pump Supports," NUREG-0577, was issued for comment in October 1979.
Although this original study was confined to PWR steam generator and reactor coolant pump suoports the staff is considering that the implementation review may include other safety-related PWR component supports and BWR component supports.
However, the Dresden fuel pool, fuel storage racks, and supports for the fuel pool cooling system are not included within this potentially expanded scope and, therefore, this USI is not applicable to the Dresden Spent Fuel Pool Modification.
A-17 Systems Interaction The concern for systems interaction arises because the design and analysis of systems is frequently assigned to teams with functional engineering specialties such *as civil, electrical, mechanical, or
f j
t 10.;.
nuclear.
The question is whether the work of these functional specialists is sufficiently integrated in their design and analysis activities to enable them to identify adverse interactions between and among systems.
The problem to be resolved by this task is to identify where the present design, analysis, and review procedures may not acceptably account for potential adverse systems interaction and to recommend the regulatory action that should be taken to rectify deficiencies in the procedures.
A number of different studies for potential systems interactions in nuclear power pl ants have been performed as a result of the NRC interest in this subject. These studies have not identified any deficiencies in our review procedures relating to fuel storage pools and ancillary systems.
The Dresden Fuel Storage Pool Modification contemplates only a change in the fuel storage racks. These racks are designed to meet Seismic Category I requirements.
No permanent modifications are being made to any connecting fuel pool service systems. Therefore, the fuel storage modification will not increase the vulnerability of the Dresden Plant to any potential adverse systems interaction.
A-24 Qualification of Class IE Safety-Related Equipment The purpose of this task is to provide a generic approach to the develop-ment of the review of equipment qualification methodology and acceptance criteria. A report "Interim Staff Position on Environmental Oualification of Safety-Related Electrical Equipment, 11 NUREG-0588 was issued for comment, in December 1979, and provided the staff technical resolution of this issue.
The electrical equipment used in systems used in conjunction with operation of the spent fuel pool such as fuel pool cooling and cleanup and fuel element handling machinery are not required to be environmentally
f I
I i: t*
f qualified to other than normal commercial grade requirements. Therefore, this US! is not applicable to the Dresden Spent Fuel Pool Modification.
A-26 Reactor Vessel Transient Protection The issue in this task was the protection of the reactor pressure vessel in Pressurized Water Reactors (PWRs) from transient pressure changes when the vessel is at a relatively low temperature. This work was prompted by concerns related to the safety margins available to vessel
- damage as a result of such events. The technical resolution of this issue was presented in "Reactor Vessel Pressure Transient Protection for Pressurized Water Reactors," NUREG-0224, September 1978.
This resolved USI is not applicable to the Dresden Spent Fuel Pool Modification, as it relates solely to the reactor vessel.
A-31 Residual Heat Removal Requirements The issue in this task was the need to demonstrate the capability to go from hot shutdown to cold shutdown without the use of offsite power.
A technical resolution was incorporated into Standard Review Plan (SRP)
Section 5.4.7 and Branch Technical Position {BTP) RSB 5-1 attached to the SRP.
Additional guidance is pending in proposed Regulatory Guide 1.139. This USI is not applicable to the Dresden Spent Fuel Pool Modification.
A-36 Control of Heavy Loads Near Spent Fuel The concern of this issue is the potential for a heavy object falling onto spent fuel in the storage pool or the reactor core resulting in damage to the fuel that could release radioactivity within the plant and to the environment. The objective of the task was to systematically review NRC requirements, operating facility design, and Technical Specifications to establish the minimum criteria for operating plants that accommodate refueling procedures and reduce the potential and consequences of postulated heavy load drop accidents to an acceptable level.
A technical resolution of Task A-36 has been issued as Control of Heavy Loads at Nuclear Power Plants," NUREG-0612, July 1980.
By letter dated December 22, 1980, Commonwealth Edison Company was requested to provide additional information related to the plant-specific control of heavy loads at Dresden Station. Our request was supplemented by an additional letter dated February 3, 1981.
Commonwealth Edison Company (CECo) response was requested by May 15, 1981.
Upon receipt of CECo's response, the NRC staff will compare the pla~t-specific control of heavy loads at Dresden Station to our generic positions established in NUREfi-0612 and determine whether additional changes are needed.
This subject was covered in testimony during the Dresden Spent Fuel Pool Modification evidentiary hearings.
A-39 Determinati~n.of _SafetQ-Relief,Valve (SRV) Pool Dynamic Loads and Temperature L1m1ts for B R Containment Operation of BWR primary system pressure relief valves can result in hydrodynamic loads on the suppression pool retaining structures or structures located within the pool~ The purpose of this ~ask is to develop acceptable methods for determining these loads and designing for them in Mark I, II and III BWR containments. A technical resolution for Mark I containments such as at Dresden 2 and 3 was issued in NUREfi-0661 (see title listed under A-7J.
As for USI A-7, implementation has been initiated by the January 13, 1981 Commission orders in accordance with the requirements in NUREG-0661.
This resolved USI is not applicable to the Dresden Spent Fuel Pool Modification, as the spent fuel pool is outside containment.
- A-40 Seismic Design Criteria - Short-Term Program The objective of Unresolved Safety Issue A-40 is to investigate selected areas of the seismic design sequence to quantify conservatisms to the extent possible, and to develop recommendations for improvements to the seismic design process.
The seismic design of the new spent fuel storage racks at Dresden is under review by the NRC staff. Our review to date indicates that no major problems exist. In addition, the SEP program is developing a site-specific ground spectra for the Dresden site. The preliminary ground spectra is enveloped by the 0.2g Regulatory Guide 1.60 spectra that is being used for the current SEP seismic re-evaluation.
No significant oroblems have been identified relative to the spent fuel pool modification.
It is not anticipated that the final site specific spectra will be sub-stantially different from the preliminary ground spectra.
In the event that the final spectra is not enveloped by the spectra used by the licensee, further assessment of the seismic design will be required.
Since the A-40 program is intended to quantify conservatisms in seismic design and will not directly relate to the seismic analysi? at Dresden, we do not anticipate that there will be any adverse impact on the Dresden fuel pool storage racks.
The seismic capability of the proposed racks and the spent fuel pool structure, with the proposed additional fuel loading, have been analyzed by the licensee and reviewed by the staff. Based on this review, the staff concluded that the proposed structural, mechanical and material design of the modified spent fuel pool is acceptable.
- A-42 Pipe Cracks at Boiling Water Reactors This task was i~itiated to address the problem of leaks and cracks in the heat-affected zones of welds that join austenitic stainless steel
- Riping in BWRs.
A technical resolution has been issued as "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," NUREG-0313, Revision 1, dated July 1980.
Implementation has been initiated. The concern and resolution is for piping in the primary system pressure boundary and therefore this US! is not applicable to the Dresden Spent Fuel Pool Modification.
A-43 Containment Emergency Sump Performance This issue is directed at assuring satisfactory hydraulic performance of the containment sump during the recirculation phase of the recovery from a loss-of-coolant accident.
The principal concerns are that break initiated debris from insulation could cause blockage of the sump or that adverse flow conditions and vortex formation in the sump could adversely affect the recirculation system performance. This US! is not applicable to the Dresden Spent Fuel Pool Modification.
A-44 Station Blackout Electric power for safety systems at nuclear power plants is supplied by two redundant and independent divisipns.
Each of these divisions includes an offsite alternating current (ac} source and an onsite ac source (usually diesel qenerators) and a direct current (de) source.
The unlikely but possible loss of ac power from both offsite and onsite sources is referred to as station blackout.
The purpose of this task is to evaluate the risks posed by station blackout accidents and assess the effectiveness of safety improvements in reducing those risks.
The principal concern in this task is the cooling of fuel in the reactor subsequent to a station blackout event. Cooling of the fuel in the spent*
fuel pool is much less critical in terms of the time available (many hours to
- days) and means available (permanently installed systems as well as temporary simple portable equipment that can be used for adding water to the pool} to assure cooling of the fuel. This task is therefore directed at the much more sensitive problem of the need to maintain or reestablish cooling for the fuel in the reactor. However, it is expected that any safety requirements that result from resolution of this task would also provide additional assurance of fuel pool cooling with respect to a station blackout event.
A-45 Shutdown Decay Heat Removal Requirements This issue is concerned with the reliability of systems used to remove decay heat following a normal shutdown or shutdown resulting from a transient or accident. This USI will investigate alternative means of decay heat removal or improving reliability of decay heat removal in PWR and BWR fuel cores within the reactor.
This USI is not applicable to the Dresden Spent Fuel Pool Modification.
A-46 Seismic Qualification of Equipment in Operating Plants The design criteria and methods for seismic qualification of mechanical and electrical equipment in nuclear power plants have undergone significant change during the course of the conmercial nuclear power program.
The objective of this task is to establish an explicit set of guidelines that could be used to judge the adequacy of the seismic qualification of mechanical and electrical equioment at all operating plants in lieu of attempting to backfit current desi'gn criteria for new plants.
The electrical and mechanical equipment used in systems to support fuel pool operation, such as fuel pool cooling and cleanup, are not required to be Seismic Category I. The new fuel racks provided in the fuel pool modification are designed to Seismic Category I requirements and have been evaluated by the staff as adequate in its Safety Evaluation Report.
Therefore, this USI is not applicable to the Dresden Spent Fuel Pool Modification.
i
' l
~
- A-47 Safety Implications of Control Systems This issue concerns the potential for accidents or transients being made more severe as a result of control system failures or malfunctions.
The concern in this task is for transients and accidents occurring to the reactor and therefore this USI is not applicable to the Dresden Spent Fuel Pool Modification.
A-48 Hydrogen Control Measures and.Effects of Hydrogen Burns on Safety Equipment This issue is concerned with the potential for large hydrogen releases inside containment during an accident and various means to cope with large releases to the containment such as inerting or controlled burning.
The potential effects of proposed hydrogen control measures on safety including the effects of hydrogen burns on safety-related equipment will also be investigated. This USI is focused on events inside the containment and therefore is not applicable to the Dresden S~ent Fuel Pool ~hich is located outside of the containment.
~.
Karl Kni el Professional. Qualifications I am Chief, Generic Issues Branch, Divisfon of Safety Technology, Office of Nuclear Reactor Regulation, U. S. Nuclear Regulatory Commission.
In this position, I am responsible for supervising the work on technical resolution of. designated 11Unresolved Safety Issues.
11 I was born in Los Angeles, California and attended the Massachusetts Institute of Technology from 1947 to 1952 receiving the degrees of Bachelor of Science and Master of Science in Chemical Engineering.
After graduation, I served with the U. S. Army Chemical Corps as an instructor at the Chemical, Biological, and Radiological Warfare School at Fort Sam Houston Texas.
I have twenty-eight years of experience in engineering and engineering management in private industry and government, most of it being directly related to design, analysis, and safety of nuclear reactors.
The details of this experience are summarized below.
April 1980 to Present: Chief, Generic Issues Branch, Division of Safety Technology, Office of Nuclear Reactor Regulation, U. S. Nuclear Regulatory Commission Plan, organize and manage the activities of the Generic Issues Branch to develop action plans for designated 11Unresolved Safety Issues 11 (USis) leading to the resolution of these issues and implementation in the licensing process; provide technical and program management of assigned USI reviewers within the Office of Nuclear Reactor Regulation (NRR) or other NRC offices and/or direct management of contractor resources; prepare reports on resolution of USls, and Reports to Congress on the status of resolution; initiate and monitor the implementation of resolved USis into the regulatory process by appropriate NRC offices and/or NRR divisions; manage, coordinate or monitor NRR activities on other outstanding generic issues and maintain up-to-date documentation regarding the status and relevance of each generic issue to the licensing process.
June 1978 to April 1980:
Chief, Core Performance Branch, Division of Systems Safety, Office of Nuclear Reactor Regulation, U. S. Nuclear Regulatory Commission Managed the activities of a branch engaged in the review, analysis, and evaluation of calculational methods used by applicants for the licensing of nuclear power plants in the fuel and core design areas of reactor plant engineering. The branch was also responsible for developing and using, in conjunction with consultants, independent calculational methods, including complex computer codes, for analyzing fuel and reactor core performance during steady-state, transient and accident conditions.
- December 1972 to June 1978:
Chief, Light Water Reactors Branch #2, Division of Reactor Licensing, Office of Nuclear Reactor Regulation, U. s.. Atomic Energy Commission, Regulatory Staff and it's successor, U. S. Nuclear Regulatory Cormnission Supervised, performed and coordinated the project management functions associated with the safety review, analysis and evaluation of applications for construction permits and operating licenses for light water power reactor facilities. Also performed those functions for certain government owned propulsion concepts and reactor facilities exempt from licensing.
Supervised reviews of amendments and changes to licenses and technical specifications through full-power.operation.
March 1968 to December 1972:
Project Manager, Division of Reactor Licensing, U. S. Atomic Energy Commission Served as Project Manager for the construction permit review for Point Beach Unit 2.
Served as Project Manager for the operating license review for Indian Point Unit 2 and Point Beach Units 1 and 2.
March 1966 to March 1968:
Nuclear Engineer, Division of Reactor Licensing,
- 0. S. Atomic Energy Cormnission Served as a technical reviewer on several cases including construction permit review for Turkey Point Units 3 and 4 and the operating 1 icense review for the Connecticut Yankee-Hadam Neck Plant.
July 1955 to February 1966:
Nuclear Development Associates and successor company, United Nuclear Corporation Inftial assignments were to a series of reactor conceptual design orojects for high performance mobile reactors performing analysis in heat transfer and reactor physics. Subsequently, I directed and performed reactor and shield analysis for a conceptual design involving nuclear power application to a military vehicle.
As a participant in a four corporation study, I performed analyses of nuclear safety and shielding in a study of nuclear rocket engines.
I performed analyses of reactor physics problems associated with high conversion ratios in light water reactors and evaluated the potential for burnable poisons in extending the core life of light water reactors.
I served as group leader responsible for all shield design and analysis for the Military Compact Reactor.
In this capacity, I directed and performed shield synthesis, nuclear analysis, thermal analysis and mechanical design analysis and provided guidance for experi-mental programs in shield physics and materials.
1953 to.1955:
U. S. Central Intelligence Agency Served as an engineer, providing technical support to unique intelligence requirements..
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of COMMONWEALTH EDISON COMPANY (Dresden Station, Units 2 and 3)
)
)
)
)
)
Docket Nos. 50-237 50-249 (Spent Fuel Pool Modification)
CERTIFICATE OF SERVICE l hereby certify that copies of THE NRC STAFF'S SUBMISSION ON BOARD QUESTION 2 lN *THE ABOVE-CAPTIONED PROCEEDING, CONSISTING OF "AFFIDAVIT OF KARL KNIEL ON UNRESOLVED SAFETY ISSUES" AND NUREG-0705, have been served on the following by deposit in the United States mail, first class or, as indicated by an asterisk, through deposit in the NRC's internal mail system, this 28th day of Apri 1, 1981.
John F. Wolf, Esq., Chairman 3409 Shepherd Street Chevy Chase, Maryland 20015 Dr. Linda W. Little 5000 Hermitage Drive Raleigh, North Carolina 27612 Dr. Forrest J. Remick 305 E. Hamilton Avenue State College, Pa. 16801 Philip P. Steptoe Esq.
Isham, Lincoln and Beale One First National Plaza Chi ca go, 111. 60603 Gary N. Wright Illinois Department of Nuclear Safety 1035 Outer Park Drive, 5th Floor Springfield, Illinois 62704 Mar.v Jo Murray,.Esq *.
Assistant Attorney General Environmental Control Division 188 West Randolph Street, Suite 2315 Chicago, Ill. 60601 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory ColTU'Tiission Washington, D. C. 20555
- Atomic Safety and Licensing Appeal Board Panel U.S. Nuclear Regulatory Commission Washington, D. C. 20555
- Docketing and Service Section U.S. Nuclear Regulatory Co1TD11ission Washington, D. C. 20555
- Charles A. Barth Counsel for NRC Staff