ML19297G756
| ML19297G756 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 05/20/1985 |
| From: | Ferrell C Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19297G757 | List: |
| References | |
| OL, NUDOCS 8505220027 | |
| Download: ML19297G756 (8) | |
Text
IN-PROGRESS AUDIT OF THE DETAILED CONTROL ROOM DESIGN REVIEW FOR ALABAMA POWER COMPANY'S JOSEPH M. FARLEY NUCLEAR PLANT UNITS 1 AND 2 12 APRIL 1985 Prepared by:
Science Applications Internationa Corporation 1710 Goodridge Drive McLean, Virginia 22102 Prepared for:
U.S. Nuclear Regulatory Commission Washington, D.C. 20B55 Contract NRC-03-82-096 8504240027 850418 C '-
ADOCK 05000348 CF
4 FOREWORD This report documents ar. in-progress audit of the Detailed Control Room Design Review (DCRDR) beirg conducted by Alabama Power Company (APCo) for its Joseph M. Farley Nuclear Plant, Units 1 and 2.
The audit was conducted by a tea m comprised of one representative of the U.S.
NRC, one representative from Science Applications International Corporation (S AIC),
and one representative from Comex Corporation (a subcontractor to SAIC).
SAIC's participation was provided under Contract NRC-03-82-096, Technical Assistance in Support of Reactor Licensing Actions:
Progra m III.'
SAIC had previously provided to the NRC and evaluation of APCo's Program Plan.
TABLE OF CONTENTS Section Pace Background..............................
I Discussion..............................
3 Func ti on a nd Ta s k Ana lysi s......................
4 A.
Phase 1 - Identification of Information and Control Requirements 4
B.
Phase 2 - Evaluation of Existing Control Room..
7 Summary and Conclusion......
8 Re f e r e n c e s..............................
10 Appendices..............................
11
In-Progress Audit of the Detailed Control Room Design Review for Alabama Power Company's Joseph M. Farley Nuclear Power Plant Units 1 and 2 This report documents the findings of the United States Nuclear Regula-tory Commission (U5NRC) aucit team during the in-progress audit of the Detailed Control Room Design Review (DCRDR) for Alabama Power Company's
( APCo's) Joseph M. Farley Nuclear Plant Units 1 and 2.
The DCRDR audit was conducted March 11 through March 13, 1985.
The audit team was comprised of one representative from the USNRC Human Factors Engineering Branch, one consultant from Science Applications International Corporation (SAIC), and one consultant from Comex Corporation.
The audit was conducted on-site at APCo's corporate headquarters in Birmingham, Alabama.
This report was prepared by SAIC with the aid of input from Comex Corporation. Comments orovided by the NRC audit team leader have been integrated into this report in order that the report represents the consolioated observations, recom-mendations and conclusions of the NRC audit team members.
A list of persons present during the meetings is included as Appendix 1 to this report.
BACKGROUND Item I.[L1, " Control Room Design Reviews," of Section I.IL, " Control Room Design," of the NRC Action Plan (NUREG-0660), developed as a result of the TMI-2 accident (Reference 1), states that the operating licensees and applicants for operating licenses will be required to perform a Detailed Control Room Design Review (DCRDR) to identify and correct design discrepcn-cies. Supplement I to NUREG-0737 (Reference 2), dated December 17, 1982, confirmed and clarified the DCRDR requirement in NUREG-0660. As a result of Supplement I to NUREG-0737, each applicant or licensee is required 20 conduct its DCRDR on a schedule negotiate. with NRC.
NUREG-0700 (Reference 3) describes four phases of the DCRDR to be performed by the applicant and licensee.
The phases are:
1.
Planning 2.
Review I
3.
Assessment and Implementation 4.
Reporting.
NUREG-0800 (Reference A) Appendix A to Standard Review Plan, Section 18.1 provides the necessary criteria for evaluating each phase.
As a requirement of Supplement I to NUREG-0737, the applicants and licensees are required to submit a Program Plan that describes how the following elements of the DCRDR will be accomplished.
1.
Establishment of a qualified multidisciplinary review team.
~
2.
r unction and task analyses to identify control room operator tasks and information and control requirements during emergency operations.
3.
A comparison of display and control requirements with a control room inventory.
4 A control room survey to identify deviations from accepted human factors principles.
5.
Assessment of human engineering discrepancies (HEDs) to determine which HEDs are significant and should be corrected.
6.
Selection of design improvements.
~
7.
Verification that improvements will provide the necessary corrections.
8.
Verification that improvements will not introduce new HEDs.
9.
Coordination of control room improvements with changes from other p ro g ra ms such as SPDS, operator training, Reg. Guide 1.97 (Reference 5) instrumentation, and upgrade of emergency operating procedures.
2
The NRC requires each.,splicant and licensee to submit a Summary Report at the end of the DCRDR. The report should describe the proposed control room changes, implementation schedules, and provide justification for leaving safety-significant HEDs uncorrected or partially corrected.
DISCUSSION APCo submitted the DCRDR Program Plan for Joseph M. Farley Nuclear Plant Units 1 and 2 (Reference 6) by letter dated October 31, 1983 (Reference 7). SAIC assisted the USNRC staff in their review of the APCo Program Plan, and submitted an evaluation of the Farley Program Plan (Reference 8) to the staff, dated December 16, 1983.
USNRC comments on the APCo Program Plan (Reference 9), dated March 9,1984, were forwarded to APCo.
One of the conclusions of the Program Plan review was that APCo had not submitted sufficient information for the staff to perform an adequate evaluation of the following areas:
(1) function and task analysis; (2) the control room inventory; (3) the essessment of human engineering discrepan-cies (HEDs); (4) selection of design improvements; (5) verification that corrections resolve HEDs without introducing new HEDs; and (6) coordination of the DCRDR with other programs.
In order to gain greater assurance that APCo was committed to conducting an effective DCRDR, the NRC recommended that a meeting be held at its main offices in Bethesda.
That meeting took place on July 25, 1984.
SAIC issued a summary of the meeting to the NRC (Reference 10). This was subsequently decketed by the NRC in a letter to APCo from the NRC (Reference 11).
One of the results of that meeting was that the NRC decided to conduct an in-progress audit of the licensee's task analysis.
This in-progress audit report addresses the single topic of that audit, namely task analysis.
The NRC audit team's review and evaluation of the DCRDR task analysis was accomplished through briefings by APCo personnel as well as by Westinghouse personnel, discussions with the licensee's human factors consultant, and reviews of completed documentation.
Following the discussion sessions, an exit briefing was held on March 13, 1984, in which the NRC audit team presented the licensee with the principal audit findings.
3
The audit commenced with a discussion of the personnel participating in the DCRDR task analysis effort and the anticipated schedule for the work for the overall DCRDR effort.
APCo has engaged the services of the Westinghouse Company as well as the human factors services of Search Technology, Inc.
Milestone dates were: (1) May 1,1985 - Westinghouse to submit the task analysis (" front-end" portion) final draft to APCo for review; (2) May 31 -
APCo operations personnel complete review of this dra ft; (3) June / July -
utilize questionnaires for operations personnel; (4) late summer /early fall
- conduct control room surveys; (5) first quarter of 1986 - APCo assessment of HEDs and integration of DCRDR with five-year plant modification plan; and (6) September / October 1986 - APCo to submit the DCRDR Summary Report.
FUNCTION AND TASK AMALYSIS The Westinghouse approach to the task analysis program consisted of two phases: (1) identification of information and control requirements; and (2) evaluation of the existing control room.
Their intent was to take a
" fresh look" at the control room to establish an analytical basis for what is and what should be in the control room.
As will be described, this
" fresh lool" is strictly independent of the existing control room, and is also primarily a systems engineering approach.
Phase 1 - Identification of Information and Control Requirements APCo's task analysis is based on Revision 1 of the Westinghouse Owner's Group (WOG) symptom-oriented Emergency Response Guidelines, Function Restoration Guidelines, status trees, and critical safety functions, as well as procedures branching from main emergency procedures.
These procedures and background documents have been approved by the NRC as providing an adequate basis for generically identifying information and control necds (Reference 12).
Farley personnel have developed the plant-specific E0Ps in accordance with their Procedures Generation Package.
The scope of the event sequences and emergency operating procedures selected for task analysis is complete.
In fact, the scope not only exceeds the minimum sequences recom-mended by NUREG-0700, but it is also the complete set of emergency sequences.
(See Appendices 2 and 3 for lists of selected sequences.)
4
Westinghouse personr.el, located at Westinghouse headquarters in Pennsylvania, are taking E0Ps and developing tasks and subtasks for each of the Farley plant-specific procedural steps.
Thus Westinghouse has adopted the " top-down" approach as recommended by NUREG-0700. Also important to note is the fact that this work is being done independently of the control room, and also.r. dependently of Farley operations personnel.
This latter point, however, has led to a weakness in the overall approach, namely that a portion of the DCRDR team did not appear to be completely conversant with differences between the generic E0Ps and the Farley plant-specific E0Ps.
Several differences uncovered during the audit could not be explained.
One example was the DCRDR team's removing a
" caution" note from the generic E0Ps when going to the plant-specific E0Ps.
Another example was the resequencing of step 6 of FF EEP-1 from step 12 of the loss of secondary containment El.
The NRC audit team felt that this weakness would possibly limit a thorough integration and validation of the E0Ps on the front end of the task analysis.
Validation of the E0Ps is performed during the walk-through of the control room; however, the validity of the information and control requirements is based on having a validated plant-specific E0P set.
To date, it did not appear as though APCo has exercised the portion of their Program Plan (Reference 6) which provides for an " emergency procedures specialist" to provide input to the DCRDR team concerning operator tasks and instrumentation requirements of the plant-specific E0Ps.
Although not required by Supplement I to NUREG-0737, Westinghouse will conduct a complete task analysis of abnormal operating procedure AOP 28.0 (the hot shutdown panel).
The task analysis process includes sequentially the " top-down" identi-fication of the event sequences, the plant-specific E0Ps, the systems, operator functions, operator tasks, and operator task requirements. The latter are systematically identified and documented on task analysis work-sheets. The following data ere recorded on the worksheets: operator func-tion, task step (including all cautions and notes), task purpose, and task actions.
The information and control requirements for the task actions are identified and include all requirements to perform the task action.
For instrumentation this includes numerical values, units of measurement, 5
characteristics, and precision. For controls this includes both type and position of control.
The NRC audit team noted several potential problem areas associated with the information that is c ag recorded on the task analysis worksheets.
The specificity of informational requirements was not recorded in sufficient detail to indicate whether one particular indication might be sufficient or whether multiple indications would be used.
For instance, the information requirement for a pump was " pump running." This could be indicated directly by flow indication or discharge pressure; or indirectly by motor amps, breaker shut, motor run light, water level decreasing / increasing, etc.
The same informational requirements were used for a motor-driven pump as for a turbine-driven pump, although the running indications could be very difter-ent.
This potential problem area also pointed out the need for more flexi-bility in considering alternatives to what type of instrument is appro-priate. Human factors specialists could play an important role in this area.
Another potential problem area was the unnecessarily restrictive preci-sion level that was applied to the values of some parameters.
For example, the total feed flow to intact steam generators should be " greater than 377 gpm" for step seven of task analysis worksheet (TAW) EEP-3. Westinghouse determined this value of 377 gpm from a set point study.
It includes the accuracy of the instrument. The precision associated with this value is i I gpm.
It may have been desirable from a practical point of view for APCo to have rounded this number up to 380 gpm because this would have enveloped the system requirement that flow be greater than 377 gpm.
However, due to the restrictive precision level of this particular example, it will probably be necessary to write a HED on the associated flow gage in the control room, although the gage scaling may be suitable for the required reading precision (i.e., 380).
Because such restrictive precision is used only as a result of a " rule-of-thumb" for completion of the TAWS, APCo may discover that it spends a lot of time writing HEDs and then redetermining a more realistic precision level in order to justify not replacing instruments unnecessarily throughout the control room.
Another potential problem area was the lack of detail for control requirements.
Only " type" and " position" were recorded in the TAW.
The 6
type included " discrete" or " variable" and the " position" included "stop,"
"st a rt, "t hrottl ed," "modul a te," e tc.
There was no determination of what type of discrete switch (i.e., detented versus spring-loaded, momentary contact positioning) might be appropriate for the intended control function.
In addition, there was no determination of the " ate, gain, and response characteristics of the needed control.
There was also no determination as to whether multiple control functions woulo be controlled from one or multiple controls.
Although no single industry standard in this area exists, some consideration for the type of control seems appropriate in the
" front end" of a task analysis.
Thi; p:rticular potential problem area pointed out the need for the participation of additional human factors expertise.
At the request of the NRC audit tea m, the licensee's contractor, Westinghouse, conducted a Paase I table-top task analysis of Farley Nuclear Plant Function Restoration Procedure FNP-1-FRP-C.1, " Response to Inadequate Core Cooling," and the Critical Sa fety Tree, FNP-2 CSF-0.2, " Core Cooling."
The methods appeared to be adequate to insure that the required information and control display requirements were identified.
However, the methods used would not ensure that the best indication of a specific parameter was identified. This NRC concern was discussed previously, and the example given was the case of " pump running."
Phase 2 - Evaluation of Existino Control Room Phase 2 of the task analysis had not been started at the time of the audit. The actual control room and/or the simulator (which is basically the same as the control room) will be used for this evaluation.
Phase 2 consists of two steps to ensure that the operator information and control capability requirement can be met by the displays and controls that are in the control room.
The first step is an engineering identification to verify that displays and controls exist in the control room and meet the requirements identified on the TAWS.
Documentation will be recorded on task analysis worksheets, sheet 2.
The use of engineering staff to verify display and control availability is adequate.
However, the NRC audit team expressed a concern that the lack of human factor 7
participation, especially with regard to the suitability aspects of the verification, would reduce the effectiveness of this process.
The second step in the evaluation of the existing control room is to validate that the operators can effectively use the existing controls and displays to accomplish the task analy. sis program. For this step, all the major disciplines contributing to the DCRDR (i.e., human factors, systems engineering, and operations) will participate. The method to be used for the validation step will be an operator walk-through performed in the actual control room or the plant simulator. Documentation will be recorded on task analysls worksheets, sheet 2.
SUMKARY AND CONCLUSIONS In sLmmary, the purpose of the in-progres udit was to assess the compliance of APCo's task analysis process with its DCRDR Program Plan and the requir tments of NUREG-0737, Supplement 1.
At the time of the Farley audit, the DCRDR team had partially completed Phase 1, or the identification of information and control requirements for the task analysis.
Phase 2, the evaluatica of the existing control room, had not been started.
Following a brief entrance meeting with the APCo DCRDR personnel and the contractors from Westinghouse and Search Technology, Inc., in which scheduling and milestones were discussed, the NRC audit team embarked on an agenda that addressed only the systems function and task analysis requirement of Supplement I to NUREG-0737. The NRC audit team arrived at the following conclusions:
1.
The scope of emergency operating procedures that will be used in the task analysis process is excellent and complete.
2.
The method for identifying individual tasks and subtasks from the procedural steps is good.
3.
The use of the Westinghouse Owner's Group generic background information to determine task actions was good.
8
4.
Group synergism between hu. nan factors consultants, operations personnel, E0P writers, and task analysis contractors is not as strong as it could be considering the excellent capabilities of the individual participants.
5.
The degree of specificity of details defining informational and control requirements in the TAWS could be improved in the front end of the task analysis.
In this regard, more information on alternative equipment, as well as more details on control require-ments, would be desirable.
This process should be documented.
6.
Plant-specific information should be better integrated into the task analysis process with regard to the determination of precision.
If this is not done. APCo may write a large number of HEDs which will require assessment and ultimate down-grade to "no-action status" with a written justification for the down-grade.
7.
Procedures personnel should participate in the validation of dif-ferences between generic guidelines and plant-specific procedures.
8.
Reactor operations personnel should be better integrated into the DCRDR as described in the Program Plan.
9
References 1.
NUREG-0660, Vol.1. "NRC Action Plan Developed as a Result of the TMI-2 Accident," U.S. Nuclear Regulatory Commission, May 1980; Revision 1, August 1980.
2.
NUREG-0737, Supplement 1. " Clarification of TMI Action Plan Require-ments," U.S. Nuclear Regulatory Commission, December 1982.
3.
NUREG-0700, " Guidelines for Control Room Design Reviews," U.S.
Nuclear Regulatory Commission, September 1981.
4 NUREG-0800, " Standard Review Plan," Section 18.1, " Control Room," and Appendix A. " Evaluation Criteria for Detailed Control Room Design Reviews (DCRDR)," September 1984.
5.
Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Follow-ing an Accident," U.S. Nuclear Regulatory Commission, May 1983.
6.
Alabama Power Company Control Room Design Review Program Plan, dated October 1983, Attachment to Reference 7.
7.
Letter from F.L. Clayton, Alabama Power Company, to S. A. Varga, NRC, dated October 31, 1983, submitting Control Room Design Review Program Plan for Joseph M. Farley Nuclear Plant, Units 1 and 2.
8.
"DCRDR Program Plan Evaluation for Joseph M. Farley Nuclear Plant Units 1 and 2," Science Applications International Corporation, December 16, 1983.
9.
" Review of Joseph M. Farley Nuclear Plant, Units 1 and 2, Detailed Control Room Design Review Pregram Plan Submittal," U.S. NRC, March 9, 1984.
10.
" Minutes Meeting Between NRC and Alabama Power Company on the DCRDR for Farley Nuclear Plant, Units 1 and 2," SAIC, August 3,1984.
11.
" Summary of Meeting Held on July 25, 1984 by NRC and APCo Representa-tives to Discuss the Detailed Control Room Design Reviews (DRCDR),"
Docket Nos. 50-348, 50-364, August 16, 1984.
12.
" Meeting Summary - Task Analysis Requirements of Supplement I to NUREG-0737, March 29,1984 Meeting with Westinghouse Owner's Group Procedures Subcommittee and Other Interested Persons," Memo to D.L. Ziemann, NRC from H.B. Clayton, NRC, April 5,1984 10
Appendices 1.
List of Attendees 2.
Event Sequences Selected for Task Analysis 3.
Emergency Operating Procedures Selected for Task Analysis 4.
Representative Draft Task Analysis Worksheets 11
Attendance - 3/11/85 Name Affiliation Jack J. Kramer USNRC/DHF5/HFEB John R. Stokley SAIC/NRC Mark Good Comex/NRC Ralph Surman Westinghouse Churck Sterrett Westinghouse Russell Oft Westinghouse Michael E. Maddox Search Technology Doug M. Kinney Alabama Power John Garlington APCo 12
EVENT SEQUENCES SELECTED FOR TASK ANALYSIS s
SPURIOUS SAFETY INJECTION s
LOSS OF REACTOR COOLANT (SMALL BREAK 1 INCH DIAMETER)
- s LOSS OF REACTOR COOLANT (SMALL BREAK 4 INCH DIAMETER) -
e LOSS OF REACTOR COOLANT (LARGE BREAK) e LOSS OF SECONDARY COOLANT s
COMBINED LOSS OF REACTOR AND SECONDARY COOLANT e
STEAM GENERATOR TUBE RUPTURE (DESIGN BASIS)
- e STEAM GENERATOR TUBE RUPTURE (MULTIPLE RUPTURES)
- e INADEQUATE CORE COOLING (RESULTING FROM FAILURES IN EMERGENCY CORE COOLING SYSTEM) e INADEQUATE CORE COOLING (RESULTING FROM LOSS OF SECONDARY HEAT SINK) e PRESSURIZED THERMAL SHDCK s
CONTAINMENT HIGH PRESSURE
- EVENT SEQUENCES RECOMMENDED IN NUREG-0700 13
EMERGENCY OPERATING PROCEDURES SELECTED FOR TASK ANALYSIS EMERGENCY EVENT PROCEDURES (EEP) e REACTOR TRIP OR SAFETY INJECTION EEP-0 e
LOSS OF REACTOR OR SECONDARY COOLANT EEP-1 e
51 TERMINATION ESP-1.1 e-POST-LOCA C00LDOL'N AND DEPRESSURIZATION ESP-1.2 e
TRANSFER TO COLD LEG RECIRCULATION ESP-1.3 e
TRANSTER TO HOT LEG RECIRCULATION ESP-1.4 e
FAULTED STEAM GENERATOR ISOLATION EEP-2 e
STEAM GENERATOR TUBE RUPTURE EEP-3 14
EMERGENCY OPERATING PROCEDURES SELECTED FOR TASK ANALYSIS (CONT.)
STATUS TREES AND FUNCTION RESTORATION PROCEDURES FRP) e FRP-5.1 RESPONSE TO NUCLEAR POWER GENERATION /ATWS e
FRP-5.2 RESPONSE TO LOSS OF CORE SHUTDOWN e
FRP-C.1 RESPONSE TO INADEQUATE CORE COOLING e
FRP-C.2 RESPONSE TO DEGRADED CORE COOLING e
FRP-C.3 RESPONSE TO SATURATED CORE CONDITIONS e
FRP-P.1 RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITION s
FRP-P.2 RESPONSE TO ANTICIPATED PRESSURIZED THERMAL SHOCK CONDITION e
FRP-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK e
FRP-H.2 RESPONSE TO STEAM GENERATOR OVERPRESSURE s
FRP-H.3 RESPONSE TO STEAM GENERATOR HIGH LEVEL e
FRP-H.4 RESPONSE TO LOSS OF NORMAL STEAM RELEASE CAPABILITIES e
FRP-H.5 RESPONSE TO STEAM GENERATOR LOW LEVEL e
FRP-Z.1 RESPONSE TO HIGH CONTAINMENT PRESSURE e
FRP-Z.2 RESPONSE TO CONTAINMENT FLOODING e
FRP-Z.3 RESPONSE TO HIGH CONTAIMENT RADIATION LEVEL 15
EMERGENCY OPERATING PROCEDURES SELECTED FOR TASK ANALYSIS (CONT.)
STATUS TREES AND FUNCTION RESTORATION PROCEDURES (FRP) (CONT.)
e FRP-1.1 KESPONSE TO HIGH PRE 55UR12ER LEVEL e
FRP-I.2 RESPONSE TO LOW PRESSURIZER LEVEL e
FRP-I.3 RESPONSE TO VOIDS IN REACTOR VESSEL 16
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