ML20054F622
| ML20054F622 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 06/11/1982 |
| From: | Kendall R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20054F597 | List: |
| References | |
| NUDOCS 8206170164 | |
| Download: ML20054F622 (20) | |
Text
{{#Wiki_filter:' UNTTED' STATES OF AMERICA NUCLEAR REGULATORY COMMISSION I-BEFORETHEATOMICSAFETYANDLICENSINGjBOARD } s IN THE MATTER OF ILLIN0IS POWER COMPANY DOCKET NO. 50-461 ~ (Clinton Power Station, Unit 1) NRC STAFF TESTIMONY OF RICK KENDALL REGARDING REACTOR VESSEL WATER LEVEL INSTRUMENTATION (Prairie Alliance Contention 6a) Ql. Please state your name and position with the NRC. A1. My name is Rick Kendall. I am employed by the U. S. Nuclear Regulatory Commission as a Reactor Engineer in the Instrumentation and Control Systems Branch of the Division of Systems Integration. A copy of my professional qualifications is attached. Q2. What is the purpose of your testimony? A2. The purpose of this testimony is to respond to Prairie Alliance Contention 6a, which reads as follows: "The design and fabrication of the CPS control room layout and instrumentation have not been modified to meet current regu-latory reouirements in NUREGs-0660, -0694, -0737. Specifically: (a) 8206170164 820611 PDR ADOCK 05000461 [ PDR
The CPS lacks sufficient instrumentation for displaying andrecordingyLe reactor pressure vessel water 1.evel." m. L Q3. What are the specific requirements of NUREGs-0660, -0694, and -0737 regarding instrumentation for displaying and recording reactor pressure vessel water level? A3. There are no specific requirements in these NUREGs concerning the number of reactor pressure vessel water level instrument channels to be displayed or recorded in the control room. The staff's requirements for this instru-mentation are that its design and implementation be in accordance with the guidelines of Regulatory Guide (RG) 1.97 Revision 2 (Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs ConditionsDuringandFollowing'anAccident). RG 1.97, Rev. 2 defines BWR reactor coolant level as a Category 1. Type B variable (i.e., a variable providing information to indicate whether plant safety functions are being accomplished). Category 1 requirements include qualification, single fail-ure, and continuous indication displayed in the control room. '~ Q4. Does the instrumentation used to display and record reactor pressure vessel water level in the control room comply with these requirements? A4. Yes. There are six channels of reactor pressure vessel water level displayed on vertical indicators in the control room. These include one wide range indicator, one fuel zone range indicator, three narrow range indicators, and one shutdown range indicator. There are five channels of reactor pressure ._a ,-,-a---.
l 3-vessel water level recorded in the control room. These include two wide range recorder one fuel zone range recorder, one narrow range recorder, and one upset range recorder. A written description of each-water level range is provided in Section 7.7.1.1.S.1.2 of the Clinton FSAR. The afety grade vessel water level display instrumentation (two wide range recorders, one wide range indicator, one fuel zone' range recorder, and one fuel zone range indicator) is located on the Reactor Core Cooling Systems (RCCS) Benchboard (P601) in the main control room. e w e e N -e e ,w- .w-
[ UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE Tdt ATOMIC SAFETY AND LIC'ENSING BOARD ( g, ) IN THE MATTER OF ILLIN0IS POWER COMPANY DOCKET NO. 50-461 (Clinton Power Station, Unit 1) NRC STAFF TESTIMONY OF RICK KENDALL REGARDING SAFETY RELIEF VALVE POSITION INDICATION (Prairie Alliance Contentico 6c) Q1. Please state your name and position with the NRC. A1. My name is Rick Kendall. I am employed by the U. S. Nuclear Regulatory Commission as a. Reactor Engineer in the Instrumentation and Control Systems Branch of the Division of Systems Integration. A copy of my professional qualifications is attached. Q2. What is the purpose of your testimony? A2. The purpose of this testimony is to respond to Prairie Alliance Contention 6c, which reads as follows: "The design and fabrication of the CPS control room laySUt and instrumentation have not been modified to meet current regulatory requirements in NUREGs-0660, -0694, -0737. Specifically: (c)
~ ' s Direct indication of safety relief valve position should be, butIis not, provided for in the CPS (, instrumentation." Q3. What are the specific requirements of NUREGs-0660, -0694, and -0737 regarding indication of safety relief valve position? A3. The specific requirement (NUREG-0737, Section II.D.3) is that positive indica-tion of safety relief valve position be provided in the control room derived from a reliable valve position detection device or a reliable indication of flow in the discharge pipe. This indication must be seismically and environ-mentally qualified, be safety grade or else powered from a vital instrument bus with a backup means of determining valve position,and provide an alarm to alert the operators of a change in valve position. This indication must be installed prior to fuel loading (see NUREG-0694). Q4. Does the safety relief valve position indication instrumentation provided at Clinton comply with these requirements? A4. Yes. The Clinton safety relief valve position indication system (SRVPIS) ~ design utilizes a peizoelectric accelerometer mounted on the discharge piping of each SRV. The sensor detects valve vibration levels and pro- ~ vides electrical outputs to signal analysis electronics in the main con-trol room. The SRVPIS is powered from a Class 1E bus and is qualified in accordance with the requirements of IEEE Standards 323-1974 and 344-1975.
e. _.. 3-A diverse measurement of SRV opening or l'ong-term leakage is provided through temperature jlements mounted in the'nnowells on. e,ach of the SRV h I-blowdown pipes to the suppression pool. These indicatioris provide confirmation of the SRVPIS readouts. A comon annunciator is provided in the control room as well as individual valve status alam indications e provided by CRT display. The staff's evaluation of item II.D.3 for Clinton. can be found in Section 7.3.3.1 of NUREG-0853 (Safety Evaluation Report re-lated to the operation of Clinton Power Station Unit No.1). e. O h e ? l e O *e e e
~ UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 1 IN THE MATTER OF ILLIN0IS POWER COMPANY DOCKET NO. 50-461 (Clinton Power Station, Unit 11 NRC STAFF TESTIMONY OF RICK KENDALL REGARDING THE SAFETY PARAMETER DISPLAY SYSTEM (Prairie Alliance Contention 6d) Ql. Please state your name and position with the NRC. ~ A1. My name is Rick Kendall. I am employed by the U. S. Nuclear Regulatory Commission as a Reactor Engineer in the Instrumentation and Control Systems Branch of the Division of Systems Integration. A copy of my professional qualifications is attached. Q2. What is the purpose of your testimony? A2. The purpose of this testimony is to respond to Prairie Alliance Contention 6d, which reads as follows: ~ "The design and fabrication of the CPS control room layout and instrumentation have not been modified to meet current regulatory
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2 l.. ", \\ 'r requirements in NUREGs-0660, -0694, -0737.M,pecifically: (d) 3 g* S g o A Safety Parameter Display System should de, tot. sis not, provided for)in the main control room." }' I-6 Q. What are the specific requirements of NUREGs-0660,' -0694, and >f' -0737 regarding the provision of a Safety Parameter Display System (SPDS) e in the main control room? c '/ r;.s i A. Thespecificrequirement(NUREG-0737,SectionI.D.II)isthat a Safety Parameter Display Systen (SPDS) be installed to display to a
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operating personnel a minimum set of parameters' which define' the safety / t o status of the plant. This installation was to be in accordance viith the requirements and schedule provided in NUREG-0696 (Functional Criterfa i a for Emergency Response Facilities - February 1981). NUREG-0696, however, only lists guideline criteria which the staff intends to use in evaluating emergency response facility design andhoesnotsetforth specific requirements which the Applicant must meet. In. addition, neither NUREG-0694 nor NUREG-0696 contain sch'edules for implementation of a SPDS. TheStaffisatpresintfinalizingdesignrequirementsfoEthe i SPDS. However, proposed requirements and implementation schedules for the SPDS are contained in SECY-82-111 (Requirements for Emergency Response Capability). The Staff is presently deciding whether to adopt these proposed requirements as a staff policy. 4 s I 't a E
^ - Q4. Is there a SPDS installed in the Clinton Unit 1 main co trol room and how well does it conford(to the proposed criteria of SECY-82-lll? g (- ,' A4. Yes, there is a SPDS installed at Clinton. As stated in Section 7.5.2.3 ' 'l of NUREG-0853 (Safety Evaluation Report related to the operation of Clinton t Power Station, Unit No.1), the Clinton SPDS is located in the main control room and is designed to attract the attentien of operation personnel when there exists a trend or condition towards degradation in'the safety para-meters of the plant. The SPDS is accessible, visible, and is distinguish-abie from other displays. The SPDS, excluding sensors and transmitters, is a self-checking system which alerts the operators to system malfunctions. The SPDS is adequately isolated from plant safety systems. The following parameters are indicated by the.SPDS: Plant Conditions Primary Variable Core cooling Reactor water level l Reactivity Source range monitor (SRM) log count rate Reactor coolant system integrity Reactor pressure, drywell pressure, reactor pressure vessel (RPV) isolation, safe-ty/ relief valve position l Containment integrity Containment pressure, contain-ment isolation vaTve positions, l suppression pool /wetwell temper-ature, suppression pool /wetwell level, drywell temperature
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The SPDS is not Class lE, single failure proof, nor designed to functjon, during seismic events. This is consistent with the proposed criteria containedinSECY-8 kill. A final evaluation of the C1'iqton SPDS in-{, cluding human factors principles and informe':on to be displayed will be made when the final SPDS design requirements hhve been established. l S 4 8 I l l l l l~ l I t. N l l l m
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD \\ {. ~ ~ IN THE MATTER OF ~ ILLINOIS POWER COMPAN DOCKET NO. 50-461 (Clinton Power Station, Unit 1) NRC STAFF TESTIMONY OF RICK KENDALL REGARDING ACCIDENT MONITORING INSTRUMENTATION (Prairie Alliance Contention 6e) Q1. Please state your name and position with the NRC. A1. My name is Rick Kendall. I am employed by the U. S. Nuclear Regulatory Commission as a Reactor Engineer in the Instrumentation and Control Systems Branch of the Division of Systems Integration. A copy of ray professional qualifications is attached. ~ Q2. What is the purpose of your testimony? ' A2. The purpose of this testimony is to respond to Prairie Alliance Contention 6e, which reads as follows: "The design and fabrication of the CPS control room layout and instrumentation have not been modified to meet current regulatory requirements in NUREGs-0660, -0694, -0737. Specifically: (e) M i l
1 1 l The CPS lacks'$ equate instrumentation for monitoring accident conditions." Q3. What are the specific requirements of NUREGs-0660. -0694, and -0737 regarding e instrumentation for monitoring accident conditions? A3. The specific requirements for accident monitoring instrumentation are contain-ed in Sections II.F.1, II.F.2, and II.F.3 of NUREG-0660. NUREG-0737 pro-vides clarification of items II.F.1 and II.F.2. II.F.1 requires the capabil-ity to detect and measure concentrations of noble gas fission products in plant gaseous effluents during and following an accident, the capability for continuous sampling of plant gaseous effluent for post accident releases of radioactive iodines and particulates, and the installation of containment ~ high 'ange radiation, containment pressure, containment water. level, and contairment hydrogen monitors. II.F.2 requires that an unambiguous, easy-to-interpret, indication of inadequate core cooling (ICC) be provided; and II.F.3 requires that appropriate instrumentation for accident monitoring be provided based on Regulatory Guide (RG) 1.97 Revision 2, (Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident). Requirements II.F.1 and II.F.2 are to be met prior to fuel loading (see NUREG-0694). NUREG-0660 requires compliance with II.F.3 by June 1982, however, the staff has subsequently proposed, as Commission policy, that conformance to RG 1.97 Rev. ? be addressed in the broader context of the requirements for emergency response l
capability. These proposed requirements are listed in SECY-82-lli (R6qu; ire-ments for Emergency Response Capability - March 11,1982). SECY-82-lli proposes that a schedule for implementation of RG 1.97 Rev. 2 requirements I' be worked out on a plant specific basis. Q4. Does the accident monitoring instrumentation provided at Clinton ccmply 8 with these requirements? A4. The accident monitoring instrumentation provided at Clinfon will be required to be in conformance with the above requirements in accordance with the associated schedules. Concerning item II.F.1, the applicant (Illinois Power Company-IP) has committed to install the required instrumentation prior to fuel loading. Based on this commitment, the staff has concluded that the Clinton design complies with the requirements of item II.F.1 (subparts 1 through 6) pending receipt of confirmatory design details for this accident monitoring instrumentation. Item II.F.1 is addressed in Sections 11.5.1,12.3.4.1, and 6.2.7 of NUREG-0853 (Safety Evaluation Report related to the operation of Clinton Power Station, Unit No.1). Concerning item II.F.2, the BWR Owners Group (BWROG), of which Illinois Power Company is a member, originally concluded (letter dated June 30, 1980) that no additional instrumentation (in addition to reactor vessel water level) was needed to monitor ICC. Incore thermocouples, however, are required for BWRs as specified in RG 1.97 Rev. 2. e e
.o The NRC and representatives from.' General Electric Compa'y and the ~ n BWR Owners Group meg on December 17, 1981 and January _27,1982 to' dis-cuss the NRC requirements and the Owners Group position. As a result - of these meetings, an agreement has 'been reached to broaden the issue from the specific requirements for incore thermocouples to that of mon-e itoring ICC. The Owners Group has agreed to actively participate in the analys is of ICC instrumentation requirements and will be., submitting a final report for NRC review in July 1982. Illinois Power Company has committed to participate in this study. It is the staff's expectation that the conclusions reached from that evaluation will be applied to Clinton. Implementation of appropriate ICC monitoring instrumentation based on the staff's review of the BWROG analysis, if necessary, will satisfy the requirements of item II.F.2. This item is addressed in ~ Section 4.4.2 of NUREG-0853. Concerning item II.F.3, the applicant will-be required to comply with the l requirements of RG 1.97 Rev. 2 in accordance with a schedule to be deter-I mined as discussed in the answer to question Q3 above. It should be noted that this review procedure to assure compliance with RG 1.97 Rev. 2 supercedes Section 7.5.3.1 of NUREG-0853 which lists compliance with RG 1.97 Rev. 2 as a license condition. t N e e 1
~- ~ ? , UNITED STATES 1F' AMERICA NUCLEAR REGULATORY-COPNISSION BEFORE Tk ATOMIC SAFETY AND LICENSING BOARD {, ~ IN THE MATTER OF ILLINDIS POWER COMPANY DOCKET NO. 50-461 (Clinton Power Station, Unit 1) NRC STAFF TESTIMONY OF RICK KENDALL REGARDING INSTRUMENTATION LOCATED ON CONTROL ROOM BACK R0W PANELS (Prairie Alliance Contention 6h) Q1. Pleahe state your name and position with the NRC. A1. My name is Rick Kendall. I am employed by the U.S. Nuclear Regulatory Commission as a Reactor Engineer in the Instrumentation and Control Systems Branch of the Division of Systems Integration. A copy of ray professional qualifications is attached. l Q2. What is. the purpose of your testimony? A2. The purpose of this testimony is to respond to Prairie Alliance Contention 6h, which reads as follows: "The design and fabrication of the CPS control room layout and instrumentation have not been modified to meet current regulatory requirements in NUREGs-0660, -0694, -0737. Specifically: (h) l
Not all CPS control panels are completely unobstructed and accessible. It is insufficient to have certain surveillance and monitorin}g actions on back row panels. g (. Moreov'er, there has been no documentation of the criteria used to determine which instruments should be placed on back row panels. ... Q3.. What are the specific requirements of NUREGs-0660, -0694, and -0737 regarding instrumentation located on main control (front row) panels versus back row panels, the' accessibility of these panhls, and the documentation of the criteria used to detennine which instruments may be placed on back row panels? A3. There are no specific requirements contained in these NUREGs concerning the location of instrumentation (i.e., front row versus back row panels), the accessibility of these panels, or the documentation of the criteria \\ used by the applicant to determi,ne which instruments are located on back row panels. NUREG-0737 Section I.D.1, however, does reference NUREG-0700 ~ (Guidelines for Control Room Design Reviews) which states in Section 6.1.1.1 (Accessibility of Instrumentation / Equipment) that operators should not have to leave the primary operating area to attend to instrumentation on back. panels during operational sequences in which continuous monitoring or the timing of control actions may be critical. The staff's requirements for control room instrumentation are found in General Design Criterion (GDC) 13 (Instrumentation and Control) and GDC 19 (Control Room) of Appendix A to 10 CFR Part 50, and Section 4.20 (Information 4 Read-Out) of IEEE Standard 279-1971 (Criteria for Protection Systems for
.~ Nuclear Power ' Generating Stations). These requirements are that a control room be provided from which actions can be taken to operate the plant safely under normal $$nditions and to naintain it in a sk(e condition un conditions. Also, sufficient instrumentation must be provided to monitor variables and systems over their anticipated ranges for normal operation, e anticipated operational occurrences, and for accident conditions to assure adequate safety, including those variables and systens that can affect the fission process, the reactor coolant pressure boundary, and the containment and its associated systens. Controls must be provided to maintain these variables and systems within prescribed operating ranges. Finally, plant protection systems must be designed to provide the operator with accurate, complete, and timely infonnation pertinent to their status and to plant safety. There is no requirement for the applicant to document L the criteria used to determine which instruments should be placed on back row panels. The instrumentation required by GDC 13 nay not be located on \\ back row panels. t Q4. Does the instrumentation provided in the Clinton Unit 1 control room comply. with these requirements? A4. Yes. The Clinto'n control room (or primary operating area) consists of panels
- P680, 601, P678, P870, P800, and P801 (See FSAR Figure 7.5-1).
Each of these panels (front row panels) is unobstructed and accessible by the operator and combined, they make up the control room as required by GDC 19 and include
the instrumentation, controls, and displays required by GDC 13 and Section ~ 4.20ofIEEEStanday279. The staff's evaluation of,the conforma'nce of the Clinton design to these criteria can be found in Section 7 of NUREG-0853 (Safety Evaluation Report related to'the operation of Clinton Power Station, Unit No.1). There are other panels (designated as back row panels) located outside of the primary operating area or control room (although actually located in the same ~ physical boundry containing the control room) which provide some indication and control functions. Those systems controlled or monitored from these panels, however, are either not-essential to the safety or operation of the plant or would only be used or accessed as part of the long term response to certain plant conditions (i.e., the timing of control or monitoring ac,tions at these panels is not critical). Therefore, this design also complies with the guidance given in Section 6.1.1.1 of NUREG-0700. The instrumentation provided on the back row panels does not affect the fission process, the reactor coolant pressure boundry, the integrity of the reactor core, or the centainment and its associated systems. 4 i i
m. s,. -=e" RICK KENDALL DIVISION OF SYSTEMS INTEGRATION U. S. NUCLEAR' REGULATORY' COMMISSION PROFESSIONAL QUALIFICATI0 S I-I have been with the U. S. Nuclear Regulatory Commission since June 1979. I I am a Reactor Engineer (Instrumentation) in the Instrumentation and Control Systems Branch, Division of Systems Integration, Office of Nuclear Reactor Regulation. I serve as a reviewer in the area of nuclear power plant instrumentation and control systems in performing and coordinating reviews and evaluations of those portions of the applications for Construction Pennits and Operating Licenses and submittals regarding proposed modifications in licensed nuclear power plants for which the branch has responsibility to assure public health and safety and pro-tection of the environment. I serve as project leader and coordinator of other reviewers for the resolution of technical issues and licensi pblemsand provide technical assistance and advice in the areas relating to the safety aspects of raactor plant instrumentation and control systems and components. I received a Bachelor of Science degree in Electrical Engineering from the Univerr.ity of Maryland (College Park) in 1979. Previously I had received an Associate of Arts degree in Electron'ic Technology from Montgomery College-(Rockville,Md.). Other educational background includes the following courses: System Reliability Engineering and Risk Assessment - JBF. Associates, Inc.;. -nm-,-- - ~ -, - - nn ..w----
1 4. I Fundamentals of System Grounding Protection - IEEE; Boiling Water Reactor Simulator School - NRC; Pressurized Water Reactor Simulator School - NRC; Pressurized Wa'ter Reactor Technology (2 coq ses) - NRC; Boiling Water Reactor T.echnology - NRC. In 1978 and 1979 I was employed by the University of Maryland Astronomy Department i as an electronic technician with such duties as designing, constructing, and re-t pairing digital systems (and supporting systems) to display and record data received from telescope photomultiplier tubes at the university observatory. I am currently a member of the Institute of Electrical and Electronics Engineers (IEEE). e - --- -}}