ML20011D767
| ML20011D767 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 07/01/1989 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML20011D723 | List: |
| References | |
| NUDOCS 8912280383 | |
| Download: ML20011D767 (11) | |
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TABLE OF CONTENTS CHAPTER 18: HUMAN FACTORS ENGINEERING 18.1 CONTROL ROOM DESIGN REVIEW (CRDR) 18.1.1 CRDR BACKGROUND l
18.I.2 FINAL CONTROL' ROOM DESIGN REVIEW
SUMMARY
REPORT 18.1.3 CRDR RESOLUTION 18.2 SAFETY PARAMETER DISPLAY SYSTEM (SPDS) 18.2.1 SPDS BACKGROUND 18.2.2 CRITICAL SAFETY FUNCTIONS (CSF) 18.2.3 SPDS/CSF RESOLUTION 18.3 EMERGENCY RESPONSE CAPABILITY 18.3.1 EMERGENCY. RESPONSE CAPABILITY BACKGROUND 18.3.2 EMERGENCY RESPONSE FACILITIES (ERF)-
18.3.3 ERF RESOLUTION O
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18.1 CONTROL ROOM DESIGN REVIEW (CRDR) 18.1.1 CRDR BACKGROUND As a result of the Three Mile Island (TMI) Accident, NUREG-0660, "TMI Action Plan," was developed. Task I.D.1 of the Plan required Big Rock Point to perform a Detailed Control Room Design Review (DCRDR)-
to identify and correct any design deficiencies. Subsequent to the issuance of NUREG-0660, NUREG-0737 was issued and published in November 1980 and provided a Clarification of TMI Action Plan Require-ments. NUREG-0737 included only those items from NUREG-0660 which were specifically approved by the NRC for implementation. Withiu NUREG-0737, Item I.D.1 CRDR was clarified.
CPCo responded to the NUREG-0737 requirements for CRDR by letters dated December 19, 1980, revised July 9, 1981, and supplemented February 5, 1982.
The May 1984 NUREG-0828, " Integrated Plant Sefety Assessment-Systematic Evaluation Program," Section 5.3.2.3 also addressed CRDR as follows:
The TMI event (NUREG-0660, Item I.D.1) and experience at other nuclear facilities have shown that existing procedures and controls may be inadequate and may hinder the operator's efforts to cope with an accident. The licensee will perform a detailed review of the man-machine interface to ensure that the operator can perform during stressful, accident conditions. As part of this review, the licensee will study control room design and.
operating procedures, including accident walk-throughs as a part i
of the process for determining the need for a safety parameter display system (SPDS). The licensee will also use this review to determine if control panel equipment identification can be improved to the point where plant transients are reduced. The staff requires that the alternate shutdown panel be included in this review.
The schedule for completion of any control room modifications that are determined to be necessary, including retraining to the new procedures (NUREG-0828, Section 5.3.2.2), will_be presented in a summary report of the control room design review by April 1985. The staff agrees that the small size and unique design of Big Rock Point warrant special consideration in a control room design review; however, to ensure that the objectives of the control room design review are satisfied, the staff will require that the licensee submit the results of the evaluation and the basis for any corrective actions for staff review before implementation.
By letter dated February 2, 1984 CPCo committed to developing a detailed plan for conducting a review of the BRP control room. This O.
letter included the CRDR effort as part of the BRP Living Schedule as Issue Number 5.
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By letter dated June 20, 1984 CPCo provided the Control Room Design Review (CRDR) Program Plan. This Plan was revised by letter dated February 27, 1987 by submittal of the BRP Plant Integrated Plan Issue BN-005, Partial Control Room Design Review Sununary Report. As discussed in the CPCo February 4,1987 letter, the report was a
" partial" report and did not include corrective actions for all Human Engineering Deficiencies (HEDs).
By letter dated August 27, 1987 the corrective actions in the form of a final sununary report was provided.
18.1.2 FINAL CONTROL ROOM DESIGN REVIEW
SUMMARY
REPORT For_ complete details on the CRDR refer to the February 27, 1987
" Partial" Summary Report as modified by inserts provided in the August 25, 1987 " Final" Summary Report.- The CRDR Summary Report consists of Eight Volumes as listed below.
Volume I, the " Modified Program Plan," is incorporated by reference as part of this Updated FHSR as are Volume II through VIII.
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CRDR
SUMMARY
REPORT p
TABLE OF CONTENTS Volume I Modified Program Plan Volume II -NRC Audit with responses to Findings - Appendix 1 SPDS Justification - Appendix 2 Volme III Photographs of Control Room before CRDR initiated -
Appendix 3 Volume IV Photographs of Control Room areas reworked as a result of CRDR - Appendix.3 continued Drawings documenting Control Room panels with CRDR Modifications and Human Factors (HF) principles -
Appendix 3 continued Volume V Identified Human Engineering Deficiencies (EDs)_
0001 through 0399 - Appendix 4 Volume VI Identified HEDs 0400 through 0999 - Appendix 4 continued Volume-VII Identified EDs 1000 through 1552 - Appendix 4 continued -
Volume VIII Summary of EDs by Instrument Number and Summary of General HEDs - Appendix 5 Software Databases used for CRDR Process-- Appendix 6' NOTE:
Supporting documentation for the above Volumes are contained in Volumes IX through XIII G and are available at CPCo.
18.1.3 CRDR RESOLUTION Corrective action on the open Human Engineering Deficiencies (EDs) will be implemented as-described in Section 7 of the Program Plan.
Pending NRC review and our implementation of remaining ED resolutions, this submittal completes our actions associated with the NUREG-0737 Supplement 1 item on conduct of a Detailed Control Room Design Review.
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t 18.2 SAFETY PARAMETER DISPLAY SYSTEM (SPDS) 18.2.1 SPDS BACKGROUND Following the TMI accident, the Nuclear Regulatory Commission (NRC) identified the need in NUREG-0660 for licensees to install a system that would display to operating personnel a minimum set of parameters which defined the safety status of the plant. Further guidance on.
the Safety Parameter Display System (SPDS) was provided by the NRC in NUREG-0737 and Supplement 1 to NUREG-0737. Supplement 1 to NUREG-0737 requires licensees to prepare and submit to the NRC a written safety analysis describing the basis on which the selected parameters are sufficient to assess the safety status of each identified function for a wide range of events, which include symptoms of severe accidents.
It further requires licensees to review their proposed SPDS system in accordance with their technical specifications to determine whether or not an unreviewed safety question exists or if a change to technical specifications is required.
Consumers Power Company's initial response to this Post-THI concern was a commitment to include a final position regarding the installation of an SPDS in the Control Room Design Review Summary Report.
It was our intent to show through the Control Room Design Review process that because of the small size of the Big Rock Point control room, e
the simplicity of system designs and the readily available key i
parameter indications to the operators, a separate SPDS may not be necessary or desirable. Our Control Room Design Program Plan described our methodology for the accomplishment of this task.
In January of 1986 the NRC conducted an In-Progress Control Room Design Review Audit. One of the purposes of this audit was to conduct an evaluation of our justification for not-installing an SPDS.
As a result of the audit and discussions with the NRC staff subsequent to the audit, we adopted the position that because of the simplicity and uniqueness of Big Rock Point (BRP) design the existing instru-mentation on the control room panels.with some minor modifications meets the functional requirements of an SPDS.
Consumers Power Company therefore committed to provide an evaluation and justification for this position to the NRC for their review in accordance with the requirements of-Supplement 1 to NUREG-0737.
'L This evaluation was completed and has been incorporated into the Final CRDR Summary Report, Volume II, Appendix 2, and provides a system description, justification and other. relevant information to show that, because of the-small size of the control room, the simplicity of system designs and the readily available key parameter indications 1
to the operators, the existing control room panel instrumentation is
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the functional equivalent of an SPDS. Because our SPDS concept utilizes existing control room' instrumentation, any-modification which may be proposed and/or implemented will be reviewed in accordance N
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with the plant technical specifications to determine whether an unreviewed safety question exists-.-
18.2.2 CRITICAL SAFETY FUNCTIONS (CSF)
Sopplement 1 to NUREG-0737, the clarification of the TMI action plan, lists five required safety functions to be evaluated when developing a Safety Parameter Display System. The five required functions are:
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Reactivity control 2.
Core cooling and heat removal from the primary system 3.
Reactor coolant system integrity 4.
Radioactivity control 5.
Containment conditions At Big Rock Point the following parameters have been selected to monitor the five selected safety function.
Safety Function Parameter Reactivity control Power Range Monitors s
Core cooling and Primary system level heat removal Primary system pressure l
Pressure Vessel Integrity Primary system pressure l-Radioactivity control RGEM monitor High Range Gamma Monitor 1
Containment control Containment pressure Containment level Containment temperature Containment isolation status The Final CRDR Summary Report, Volume II,. Appendix 2, includes a discussion of how the selected parameters monitor the critical functions.
18.2.3 SPDS/CSF RESOLUTION Since Critical Safety Function justification is within the scope of the CRDR for BRP, techniques developed for Human Engineering Discrepancies (HEDs) were applied to.these parameters.
The devices identified to monitor the CSF parameters have been
(N reviewed from both a static and dynamic standpoint as part of the
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CRDR Program as discussed in Section 18.1.
The-results of the CRDR 18.2-2 MIO488-0257A-BX01
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V Program verify.that identified devices are appropriate with exceptions.
Those exceptions are documented in the form of EDs and will be resolved as a part of the ED process described in Section 18.1.3.-
Based upon the Technical Support Center (TSC) being adjacent to the i
Control-Room and the viewing of the CSF devices through the windows j
being approximately the same as that of an Operator seated at a desk in the Control Room, extension of the critical function display into the TSC is not required.
l As discussed in the Final CRDR Summary Report, Volume II, Appendix 2, j
our SPDS (or the critical safety function) concept does not incltde t
additions to the BRP control room.
Should the ED assessment require i
i instruments to be added, relocated, or enhanced, it will be done according to the plant modification program and within the existing technical specifications. At that time the appropriate safety evaluation will be performed.
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18.3 EMERGENCY RESPONSE CAPABILITY 18.3.1 EMERGENCY RESPONSE CAPABILITY BACKGROUND NUREG-0828, May 1984, Integrated Plant Safety Assessment - Systematic Evaluation Program - Final Report, Section 5.4.12 evaluated the BRP Emergency Response Capability as follows:
NUREG-0737 presented NRC guidance on several issues related to emergency response capability:
1.
Item I.C.1, "Short-Term Accident and Procedures Review" 2.
Item I.D.1, " Control Room Design Review" (CRDR)-
3.
Item I.D.2, " Plant Safety Parameter Display Console" (SPDS) 4.
Item III.A.I.2, " Upgrade Emergency Support Facilities (EOF)"-
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Item III.A.2.2, " Meteorological Data" Subparts and interim steps have been completei for some of these items, and other items have been found necessay during the staff's continuing review on the overall issue of emergency response capability.
The NRC issued Generic Letter 82-33, " Supplement to NUREG-0737 -
Requirements for Emergency Response Capability," dated December 17, 1982, to all licensees. That letter provided additional O-clarification regarding safety parameter display systems, detailed control room design review, Regulatory Guide 1.97 (Revision 2),
applications of emergency response facilities, upgrading of emergency operating procedures, emergency response facilities, and meteorological data.
The letter requests licensees to
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prepare and implement emergency operating procedures (NUREG-0828, Section 5.3.2.2)-
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perform a human factors review of the design of the control room and make any modifications shown to be necessary by the review (NUREG-0828, Section 5.3.2.3) 3.
design and install a console in the control room displaying the most important plant safety parameters 4.
provide indications in the control room of Type A, B, C, D and E variables and meteorological variables listed in Regulatory Guide 1.97 (Rev 2) 5.
provide indication in the Technical Support Center (TSC) of essential variables from Regulatory Guide 1.97 (Rev 2)-
6.
provide indication in the Emergency Operations Facility (EOF) 4 of containment conditions and releases of radiation l
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provide adequate staffing to perform emergency response.
The letter also asked the licensees to submit schedules and plans for meeting these requests. Final schedules were to be negotiated with the NRCs project manager for the plant.
The licensee's submittal dated June 1, 1983 responded to Generic Letter 82-33.
The NUREG-0737 Item Numbers I.D.1 and I.D.2 for CRDR and SPDS along l
with the Generic Letter Request Item Number (2) Human Factors Review, and (3) Display of the Most Important Plant Safety Parameters.were addressed in Sections 18.1 and 18.2 of this Updated FHSR.
l Generic Letter.82-33 requested licensees to provide certain instrumentation from Regulatory Guide 1.97, Revision 2, in the control room, the TSC, and the EOF. The generic letter requests that measurement and indication of Type A, B, C, D, and E variables and meteorological variables as specified in the regulatory guide be provided in the control room. The licensee has concluded that no additional instrumentation as-specified in the regulatory 4
guide is necessary for the Big Rock Point control room. The licensee notes that the staff's SER on SEP Topic VII-3, " Systems Required for Safe Shutdowa" (forwarded by letter dated December 17, 1982), concludes that "the present design is an acceptable alternative to current licensing guidelines." On the. basis of its evaluation presented under Topic VII-3, the staff concludes
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that the additional instrumentation requested by the generic letter for the control room is not necessary for Big Rock Point.
l Any additional instrumentation that may be necessary to enhance the operators' ability to follow the course of an accident will evolve from the control room design review as a part of the determination of the need for an SPDS.
The generic letter requests that indication of variables necessary to perform the TSC function be provided-in the TSC. The TSC at Big Rock Point is located directly outside the door of the control room; it includes the hallway along the front of the control room and the Shift Supervisor's' office. Trained personnel in the TSC can, read nearly all of the control room indicators through the windows at the front of the control room. Also, the indicators for the meteorological parameters are lo'cated right l
outside the control room in the TSC.
Therefore, the licensee L
concludes that no additional indicators need to be installed in the TSC to facilitate the function of the TSC. The staff agrees, o
l The generic letter requests that primary indicators of the l
condition of the containment and radioactivity releases'be l
provided in the EOF. The licensee proposes not to provide such indication and believen that all necessary information can be obtained by communications (such as telephone) with the TSC and 18.3-2 P
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control room-In view of the support function of the EOF, the j
staff agrees.
The generic ler.ter also described guidance other than indication of safety parameters for the emergency response facilities (ERFs)
- Technical Sup; ort Center, Operations Support Center (OSC) and Emergency Operations Facility (EOF). The guidance included aspects such as staffing,' communication, security, space, radiation protection, and.deta analysis.
The licensee believes, an the i-basis of information submitted in a letter dated June 1, 1981, i
that the current TEC, OSC, and EOF are adequate. A review team from the Office of Inspection and Enforcement will conduct an
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onsite review of the acceptability of the emergency response-facilities for Big R >ck Point after the licensee informs the staff that the FRFr are complete. The July 1983 cmergency exercise indicated that additional space is required for the TSC.
The July 1983 exercise demonstrated the capability of the existing design to accomplish tl:e emergency functions until the control room design review identifies any corrective actions that may be necessary to enhance thet capability and an exercise is conducted to demonstrate the capability of the completed ERFs. With regard to the space limitations in the TSC, the licensee, in a letter dated November 23, 1983, committed to complete TSC improvements before the 1984' emergency practice drills, including the renovation of the Shift Supervisor's office. When completed,.the new TSC
.g will have approximately 50% more usable space and will be separated from the Shift Supervisor'sc office. This work is scheduled to be completed in May 1984.
Corrective actions.for the Control Room Design Review and Safety Parameter Display System (if any) are addressed in Sections 18.1 and 18.2 above. Modifications to the Technical Support Center were completed via BRP Facility Change FC-576 and were operable May 21, 1984. The NRC, during a subsequent June 27, 1984 Inspection Report, Number 84-03, stated that:
"This exercise utilized the recently revised and enlarged TSC for the first time in a formal emergency preparedness exercise.
These major structural changes definitely helped to improve the operations and communications aspects of the TSC."
18.3.2 EMERGENCY RESPONSE FACILITIES (ED.F)
The Emergency Response Facilities, Equipment, Procedures, and Staffing are all addressed in BRP Volume 9 - Site Emergency Plan which has been incorporated by reference into this Updated FHSR.
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18.3.3 ERF RESOLUTION Adequacy of the BRP Emergency Response Capability will be reviewed by the NRC during subsequent Emergency Response Facility (ERF) Appraisal p
or Emergency Exercise.
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