ML20115D344
| ML20115D344 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 01/04/1984 |
| From: | SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY |
| To: | NRC |
| Shared Package | |
| ML20115D289 | List: |
| References | |
| CON-NRC-03-82-096, CON-NRC-3-82-96, RTR-NUREG-0737, RTR-NUREG-737 NUDOCS 8504190032 | |
| Download: ML20115D344 (36) | |
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DATA COLLECTION AND EVALUATION OF THE PROPOSED SAFETY PARAMETER DISPLAY SYSTEM FOR PHILADELPHIA ELECTRIC COMPANY'S PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3 January 4,1984 Prepared by Science Applications International Corporation Under Contract to The United States Nuclear Regulatory Comission Contract NRC-03-82-096 Ogj41hk $
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TABLE OF CONTENTS Section Page INTRODUCTION....................'........
1 REQUIREMENTS............................
1 EVALUATION
SUMMARY
3 EVALUATION DISCUSSION........................
4 1.
SPDS Parameter Selection............
4 2.
SPDS Users...................
6 3.
Human Factors Analysis of the SPDS............
7 o
Individual SPDS Displays...............
7 o
SPDS Panel Layout..................
14 4.
Procedures and Training..................
18 5.
Implementation......................
18 CONCLUSION.............................
19 RE FE REN CE S.............................
20 ATTACHMENT 1: Peach Bottom SPDS Functions and Variables......
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DATA COLLECTION AND EVALUATION 0F THE PROPOSED SAFETY PARAMETER DISPLAY SYSTEM FOR PHILADELPHIA ELECTRIC COMPANY'S PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3 INTRODUCTION The United States Nuclear Regulatory Commission (NRC) and Science Applications International Corporation (SAIC), as' consultants to the NRC, have performed an evaluation of Philadelphia Electric Company's proposed Safety Parameter Display System (SPDS) for Peach Bottom Atomic Power Sta-tion, units 2 and 3.
The evaluation team consisted of members of the NRC, Division of Human Factors Safety. Human Factors Engineering Branch, and Procedures and Systems Review Branch along with consultInts from SAIC. This.
evaluation included a review of the licensee's Safety Analysis Report
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(Reference 1), data collected on-site at the Peach Bottom facility on l
October 16, 1984, and photographs supplied by the licensee.-
Evaluation of the proposed Peach Bottom SPDS was performed with refer-ence to the requirements of Supplement 1 to NUREG-0737 (Reference 2).
Addi-tional guidance was provided by NUREG-0700 (Reference 3), draft NUREG-0835 (Reference 4), and Reg. Guide 1.97 (Reference 5). This report provides the results of the evaluation.
Comments of the NRC staff members responsible for evaluation of the proposed Peach Bottom SPDS have been integrated into this report in order to represent the consolidated observations, conclusions and recommendations of the NRC staff and its consultant (SAIC).
REQUIREMENTS The requirements for the SPDS as stated in Supplement 1 to NUREG-0737 are:
'4.1.a.
The. SPDS should provide a concise display of critical plant variables to the control room operators to aid them in rapidly and reliably determining the safety status of the plant. Although the SPDS will be operated during normal operations as well as during 2
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abnormal conditions, the principal purpose and function of the SPDS is to aid the control room personnel during abnormal and emergency conditions in determining the safety status of the plant and in assessing whether abnormal conditions warrant corrective action by operators to avoid a degraded core.
This can be particularly important during anticipated transients arid the ini-tial phase of an accident.
"4.1.b.
Each operating reactor shall be provided with a Safety Parameter Display System that is located convenient to the control room operators. This system will continuously display information from which the plant safety status can be readily and reliably assessed by control room personnel who are responsible for the avoidance of degraded and damaged core events.
"4.1.c.
The control room instrumentation required (see General Design Criteria 13 and 19 of Appendix A to 10 CFR 50) provides the operators with the information necessary for safe reactor operation under normal, transient' and accident conditions.
The SPDS is used in addition to the basic components and serves to aid and augment these components. Thus, requirements applicable to control room instrumentation are not needed for this augmentation (e.g., GDC 2, 3,4 in Appendix A; 10 CFR Part 100; single-failure requirements). The SPDS need not meet requirements of the single-failure criteria and it need not be qualified to meet Class 1E requirements. The SPDS shall be suitably isolated from electrical or electronic interference with equipment and sensors that are in use for safety systems.
The SPDS need not be seismically 4
qualified, and additional seismically qualified indication is not
. required for the sole purpose of being a backup for SPDS.
Procedures which describe the timely and correct safety status assessment when the SPDS is and is not available, will be developed by the licer.ste in parallel with the SPDS. Furthermore, operators should be trained to respond to accident conditions both with and without the SPDS available.
"4 1 d There is a wide range of useful information that can be provided by various systems.
This information is reflected in such staff 2
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documents as NUREG-0696, NUREG-0835, and Regulatory Guide 1.97.
Prompt implementation of an SPDS can provide an important contribution to plant safety.
The selection of specific information that should be provided for a particular plant shall be based on engineering judgment of individual plant licensees, taking into account the importance of prompt implementation.
"4.1.e.
The SPDS display shall be designed to incorporate accepted human factors principles so that the displayed information can be readily perceived and comprehended by SPDS users.
"4.1.f.
The minimum information to be provided shall be sufficient to provide information to plant operators about:
(1)
Reactivity control (ii)
Reactor core cooling and heat removal from the primary system
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(iii) Reactor coolant system integrity (iv)
Radioactivity control (v)
Containment conditions The specific parameters to be displayed shall be deter, mined by the licensee."
l EVALUATION
SUMMARY
Our evaluation of the proposed Peach Bottom SPDS resulted in the deter-mination that the system does not meet the spirit or'the intent of the SPDS requirement in Supplement 1-to NUREG-0737. The use of existing fixed panel instrumentation which is dispersed throughout the control room does not represent a concise display.
The licensee's rationale for separating reac-tivity monitoring displays in order to locate the displays near their controls suggests a fundamental misunderstanding of the requirements and the basic purpose of the SPDS.
Moreover, none of the displays in the proposed
. system can be rapidly and reliably read by the intended user (shift supervisor) who will be at the operator's desk during emergency operations.
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The reasons for selecting specific variables for assessment are not adequately analyzed and justified. For example, it is not clear why drywell pressure was selected for assessment of reactivity control. Irt addition, the licensee has not developed procedures or training programs for the use of the proposed SPDS.
In short, the proposed system does not meet the requirement. The detai'ed results of our evaluation are included in the following sections of this report.
5 VALUATION DISCUSSION 1.
SPDS Variable Selection Philadelphia Electric Company stated in their submittal to the NRC (Reference 1) dated September 28, 1983, that the " basis for SPDS parameter selection is the entry conditions for the upgraded emergency operating procedures." While entry conditions are important, they are not necessarily adequate for monitoring the course of an abnormal event over extended periods of time.
Also, the licensee omitted certain reactivity monitoring variables for invalid reasons (as discussed further below in connection with SPDS panel layout).
While selection of variables for display is at the discretion of the licensee, it is required that the basis for selection be adequately described. In view of the overall inadequacy of the proposed SPDS, the licensee's description in this regard is not adequate.
The analytic basis for the emergency procedures guidelines (EPGs) and EPG steps was reviewed by the NRC and a Safety Evaluation Report was issued (Reference 6) on EPG revision 2.
On this basis, the analysis requ' ired to l
identify variables which determine the safety status of a boiling water reactor (BWR) was completed prior to the development of the proposed Peach Bottom SPDS.
However, our review of the variables which were selected for the five SPDS critical safety functions (see Exhibit 1), identified a number l
of variable selection issues which are not adequately described.- For example:
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Drywell pressure is not a key variable used by the operators to assess the safety status of reactivity control.
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Function Reactor 5
Reactor Core Coolant Radio-Contain-Reactivity Cooling and System activity ment Variable Control Heat Removal Integrity Control Conditions'
- 1. Reactor Water X
X Level
- 2. Reactor X
X Pressure
- 3. Drywell X
X X
X Pressure
- 4. Drywell X
Temperature
- 5. Suppression X
Pool Temper-ature
- 6. Suppression X
Pool Level
- 7. Group I Con-X X
X tainment
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Isolation Valve Position
- 8. Neutron X
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Exhibit 1 - Peach Bottom SPDS Variable List 4
(keference 1) 5
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o Intermediate and source range neutron monitors should be included in the reactivity control function.
Group 1 isolation valve position is a useful variab1"e for the o
reactor coolant system integrity function. But, if valve position is used as a variable, HPCI injection valves RCIC, injection valves, low pressure injection valves, SRV position and reactor water cleanup should also be included.
o Reactor water level is a key variable in the assessment of reactor coolant system integrity, but it is not included in this SPDS function.
Group 1 isolation valve position is not a key variable used by the o
operators to assess the status of radioactivity control.
Containment radiation, vent stack radiation and off-gas radiation o
are key variables which Jare not included in the radioactivity control function.
The licensee's Safety Analysis Report (Reference 1) provides no task l-analyses of how the variables are used to assess the safety status of each of the five functions for a wide range of. events. Also, Reference 1 indi-cates that two additional functions, reactor pressure control and contain-ment control, are part of the SPDS.
However, these functions were not included in the Function / Variable Table (Exhibit 1).
2.
SPDS Users During our data collection visit to the control room, it was determined that the intended user of the proposed SPDS is the Shift Supervisor.
In emergency operations, the shift supervisor's workstation is at the opera-tor's desk in the middle of the control room (see Attachment 1 for control room diagrams for each function). The role of the shift supervisor is to proceed through the emergency operating procedurer which are laid out on the operator's desk under a sheet of plexiglass.
As the Shift Supervisor goes through the procedures, he checks off the completed steps with a grease pencil.
Since he cannot read any of the displays from the operator's desk.
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he would eithe'r have to go to the SPDS displays. himself, or have his opera-tors provide him with the information.
In both cases, he would be distracted from his defined emergency response functions. Therefore, it is our judgment that the fixed panel instrumentation identified as the proposed SPDS does not satisfy NUREG-0737 Supplement I requirements to provide the shift supervisor with information more readily available than it'already is in the existing control room.
3.
Human Factors Analysis of the SPDS Supplement 1 to NUREG-0737 requires that the SPDS display shall be designed to incorporate accepted human factors-principles so that the.
displayed information can be readily perceived and comprehended by SPDS I
users.
In order to perform a human factors evaluation of the proposed SPDS displays, we used the guidance provided in Section 6 of NUREG-0700 (Reference 3). Our first step was to evaluate the displays individually, then we evaluated the panel layout and control display integration. The results of our analyses are as follows.
Individual SPDS Displays A.
Reactor Water Level Chart Recorder (LR-2-2-3-110B) 1.
Dual scale indicators display water level from 0 to -325 inches and from +50 to -165 inches.
This conforms to NUREG-0700 guide-line 6.5.1.1.a for completeness of reactor water level informa-tion.
f 2.
An entry condition for the reactar,ontrol procedure is reactor water level below -48 inches.
However, the scales on both water l
1evel meters are graduated in 5 inch increments which makes it difficult to read with the accuracy required even if the user is within 4 feet of the displays. This is a deviation from NUREG-0700 guideline 6.5.1.2.a.
which recommends that the scale units should be consistent with the degree of precision and accuracy needed by the operator.
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The chart recorder part of this display is of the' continuous recorder type.
Both pens are dedicated full-time to each water level signal recorded. However, in order to depict both wide and narrow range trending, the chart paper is laid out to alternately reflect the wide and nr.rrow range water level scales. Therefore, the user only has access to scale information for one ' water level signal at a time.
In order for the operator to evaluate the trends for ooth water level signals,-the chart recorder paper has to be removed from the recorder.
Removing the chart recorder paper from the recorder is not a rapid technique for making trend information available to the operator. This is a deviation from the Supplement I to NUREG-0737 requirement which states that the system will continuously display information from which the plant safety status can be_readily and reliably assessed.
B.
Reactor Pressure Meter 1.
The reactor pressure meter range of 0 to 1500 psig is in confonnance with the Reg. Guide l'.97 requirement for a reactor pressure display with a range of 15 psia to 1500 psig. However, this display could not be evaluated because it was not installed at the time of the data collection visit.
C.
Drywell Temperature Recorder (TR-4805) 1.
The range of the drywell temperature recorder is 0 to 240*F. Reg.
Guide 1.97' requires a range of 40F to 440*F. The licensee has provided no justification for the differences in scale ranges.
2.
An en 'v :ondition for the containment control procedure is drywell temperature above 145'F. The. scale on temperature recorder TR-4805 is graduated in 5* increments and is consistent with the degree of accuracy needed by the operator, if the user is standing within 4 feet of the display.
3.
The chart recorder part of this display is of the continuous recorder type.
The recorder has two pens for containment pressure and temperature. Both pens are dedicated full-time to each 8
c channel which inputs to the recorder.
In order to depict pressure (0 to 70 psig) and temperature (0 to 2400F) the chart paper is laid out to alternately reflect pressure and temperature trending.
In addition the chart paper has a third scale (0 to 25 feet) which-is not used.
This is a deviation from NUREG-0700 guideline 6.5.4.1.b which states that the scales printed on the chart paper should be the same as the scales used on the display.
In order to evaluate the trends for either containment pressure or containment temperature, the chart paper has to be removed from the recorder.
Removing the chart paper from the recorder is not a rapid techni-que for making trend information available to the operator. This is a' deviation from the Supplement I to NUREG-0737 requirement
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which states that the system will continuously display information t
from which the plant safety can be readily and reliably assessed.
4.
The containment pressure scale (PR 4805) is not part of the proposed SPDS.
The licensee states that containment pressure is
-displayed on PR-8102B.
Therefore the containment pressure scale on recorder PR-4805 adds unnecessarily redundant and expanded detail which makes the display less than concise. This problem is compounded by the fact that the recorder paper also depicts an unused 0 to 25 feet scale. This means' that two of the three scales depicted on the chart recorder are not part of the proposed SPDS. This is not a concise display.
5.
The labeling on this recorder (CONTAINMENT PRESS. and CONTAINMENT i
TEMP.) is inconsistent with the (DRYWELL PRESSURE) display. Since containment and drywell mean the same thing in the context of the SPDS, they should be labeled consistently. This is a deviation from NUREG-0700 Guideline 6.6.3.3.b which states that labels should be consistent within and across pieces of equipment. In addition, the entry conditions for Peach Bottom containment con-trol procedures refer to "Drywell temperature," not containment temperature.
This is a deviation from NUREG-0700 Guideline 6.6.3.3.c which states that there should be no mismatch between nomenclaiure used in procedures and that printed on the labels.
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D.
Drywell Pressure Recorder (PR-81028)
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This dual scale instrument, displays drywell pressure from 5 to 25 psia on one scale and 0 to 225 psig on the other scale. An entry condition for both reactor control and containment control procedures is drywell pressure above 2 psig. The drywell pressure psig scale cannot be read to an accuracy of 2 psig even if the user is within 4 feet of the display. This is because the psig scale is graduated in 5 pound increments which is too large to accurately identify 2 psig. This is a deviation of NUREG-0700 Guideline 6.5.1.2.a which states that the scale units should be consistent with the degree of precision and accuracy needed by the operator.
The drywell pressure psia scale is graduated in 1 pound increments and can be read to the accuracy required, but the use of this scale would require operator convertion from psig to psia. This is a deviation of NUREG-0700 Guideline 6.5.1.2.b which states that all displays should indicate values in a form innediately usable by the operator without requiring mental conversion.
2.
The Peach Bottom procedures which define the entry conditions for the reactor control procedure and the containment control procedure specify drywell pressure in psig rather than psia. The use of the psia scale is a deviation from NUREG-0700 guidelines 6.5.1.4.e which states that the printed message on the display face should use the same tenns as the procedures in display identification, parameter identification and its units displayed. Though the licensee did not provide an explanation for the use of psia and psig scales, it is our judgment that the psia scale is being used as a substitute for a narrow range psig scale.
3.
The chart recorder part of this dual scale display is of the continuous recorder type. Both pens (psia and psig) are deJ1cated full-time to each channel which inputs to the recorder. However, in order to depict both the psia and psig trending, the chart paper is laid out to alternately reflect the two scales.
In order to evaluate the trends for both scales, the chart paper has to be removed from the display. This is a deviation from the Supplement I to NUREG-0737 requirement which states that the system will 10 e
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continuously display information from which the plant safety status can be readily and reliably assessed.
4.
The-scale progression on the psig chart paper (0-25-50-75-100) is different than the scale on the chart (0-50-100).
This is a deviation from NUREG-0700 Guideline 6.5.4.1.b. which states that the scales printed on the chart paper should be the same as those shown on the display.
E.
Torus Temperature Recorder (TR-81238) 1.
The range of the suppression pool temperature recorder is 300F to 2300F. This conforms to Reg. Guide 1.97 requirement for a sup-pression pool temperature display with a range of 300F to 2300F.
2.
An entry condition for the containment control procedure is i
suppression pool temperature above 950F. The scale on the display is graduated in 50 increments and can be read to the accuracy required, if the user is standing within 4 feet of the display.
3.
There are no units "0F" printed on the recorder scale. This is a deviation from NUREG-0700 Guideline 6.5.1.4.e which states that f
the printed message on the display face should use the same terms l
as the procedures in display identification, parameter identifica-tion, and units displayed.
4.
The licensee refers to suppression pool temperature in their Safety Analysis Report (Ref.1), But the labeling on this display identifies TORUS WATER LEVEL & TEMP. The terms " suppression pool" and " torus" are used to identify the same area within the plant.
However this is a deviation from NUREG-0700 Guideline 6.6.3.3.a which states that a list of standard names, acronyms, abbrevia-tions, and part/ system numbers should be administrative 1y controlled.
1 5.
The chart recorder part of this display.is of the continuous recorder type. It has two pens which continuously record torus i
level and temperature. Both pens are dedicated full-time to each 11
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channel which inputs to the recorder. However, in order to depict both torus level and torus temperature trending, the chart paper is laid out to alternately reflect level and temperatu,re scales.
Therefore, the user only has access to the trending information for one scale at a time.
In order to evaluate the trends for both variables, the chart paper has to be removed from the display.
Removing the chart paper from the display is not a rapid technique for making information available to the operator.
This is a deviation from the Supplement I to NUREG-0737 requirement which stated that the system will continuously display information from which the plant safety can be readily and reliably assessed.
F.
Torus Level Chart Recorder (LR-8123B) 1.
The range of the suppression pool level recorder is 1 to 21 feet, j
But, the scale units " Feet" are not included on the display face j
in deviation of NUREG-0700 Guideline 6.5.1.4.a which states that the printed message on the display face should use the same terms as the procedures in display identification, parameter identification and units displayed.
j 2.
An entry condition for the containment control _ procedure is suppression pool level outside 14.6 feet to 14.9 feet. The scale is graduated in.2 foot increments which can be read to the accuracy required, if the user is within 4 feet of the display.
However, the display does not have operating band markings to assist the operator. This is a deviation of NUREG-0700 Guideline j
6.5.2.3, which states that zone markings should be used to show i
operational implications of various readings such as " operating range " " upper limits," " lower limits " or " danger range."
3.
See Torus level / temperature recorder evaluation comments in l
previous item E.5 for our comments on the graphic recorder part of j
this display.
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Group 1 Containment Isolation Valve Position 1.
The containment isolation valve position lights (open/. closed) do provide rapid and reliable indication of Group 1 isolation.
However, the labels cannot be read from the user's workstation at the operator's desk.
This is a deviation from NUREG-0700 Guideline 6.6.4.1 which states that the speed and accuracy of human performance in identifying controls / displays is influenced
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by the style and size of characters used for label lettering.
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Neutron Flux Chart Recorder (APRM) NR-2-07-046B 1.
The range of the APRM is from 0 to 125% power. This range exceeds the 100% full power requirement in Reg. Guide 1.97.
2.
An entry condition for the reactor control procedure is a scram condition with reactor power above 3% or unknown.
Since the APRM scale is graduated in 55 increments, it is difficult for the operator to read the APRM scale to the accuracy required ospecially at the end of the scale. This is a deviation of NUREG-C700 Guideline 6.5.1.2.a which states that the scale units should be consistent with the degree of precision and accuracy needed by l
the operator.
However, this is a dual scale recorder which has an IRM scale that can be read to the accuracy required, if the user is _within 4 feet of the display. But the IRM was not included in this function.
3.
The IRM-APRM recorders monitor two channels.
Each channel is color coded. For example, red for IRM Channel B or APRM Channel B and black for IRM Channel D or APRM Channel D.
The channel select switch for Channel B (IRM/APRM) is color coded red and the channel select switch for Channel D is color coded black.-
This is consistent color coding; but, both of the display scales are coded red for IRM and black for APRM. This is a deviation from NUREG-0700 Guideline 6.5.1.6 which states that the same qualities which make color useful for coding can, if inconsistently applied, result in unintended confusion and distraction.
In order to 1
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follow the guidance provided in NUREG-0700, the scales should be coded red for Channel 8 IRM/APRM and black for Channel D IRM/APRM.
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4.
The chart recorder part of this display is of the continuous recorder type. The pens for both channels (red and black ink) are dedicated full-time to each channel which inputs to the recorder.
However, in order to depict the IRM and the APRM variables, the chart paper is laid out to alternately reflect each variables.
Therefore the user only has access to trending information for one variable at a time.
In order to evaluate the trends for both variables, the chart paper has to be removed from the display.
This is not a rapid technique for making trend information availa-ble to the operator. This is a deviation from the Supplement I to NUREG-0737 requirement which states that the system will contin-uously display information from which the plant safety can be readily and reliably assessed.
In summary, all.of the displays for the proposed SPDS have specific human engineering discrepancy problems. Trending information for dual scale recorders is only available for half of the time for each scale. Unused and unnecessary scales are used on the containment temperature recorder. Nomen-clature is inconsistent on the containment temperature recorder.
Unit t
conversion from psia to psig is required on the drywell pressure recorder.
Some recorders, such as the reactor water level recorder cannot be read with the precision and accuracy required.
The torus temperature and level j
recorder scales do not have parameter units on scales. Color coding *is used i
inconsistently on the IRM/APRM recorders.
In addition to the component level discrepancies, none of the displays can be read by the intended user during emergency operations.
SPDS Panel Layout The human factors evaluation of the Peach Bottom SPDS concentrated on two main aspects.
First, we evaluated the functional and task grouping of the displays to determine if the displays promoted readily perceived operator comprehension of the current system condition. This evaluation included bo'th the primary and redundant display instruments. Second, we evaluated the panel layout within the context of its use during emergencies 14
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by the shift supervisor. The detailed results of our analyses are provided below.
The seven variables on vertical panel 20C03 are functionally grouped for e
three of the five SPDS function (see Exhibit 2). The three functions are:
reactor core cooling and heat removal, reactor coolant system inte' rity, and g
containment conditions.
The arrangement of the SPDS and its relationship to emergency core cooling systems, primary containment isolation systems, containment atmospheric dilution system, and reactor core isolation cooling system is shown in Exhibit 2.
It is our judgment that the SPDS panel layout on Panel 20C03 only partially confoms to the panel layout requirement 4.1.b of Supplement I to NUREG-0737 and guidance provided in Section 6.8 (Panel Layout) of NUREG-0700. There are several problems with this panel layout arrangement. First, the licensee has misunderstood a fundamental point of the SPDS requirement. That is, the SPDS is to be used for monitoring and assessing system functions, not for controlling process functions.
Therefore, the fact that the displays are close to their controls is not relevant. Second, the use of two panels (20C03 and 20C05A) results in a design that is not integrated. This is a deviation from the Supplement I to NUREG-0737 requirement and guidance in Section 6.1 of NUREG-0835 which sta,t that the displays should be concentrated at a single location specifically designated for SPDS. This is also a deviation from the guidance provided in NUREG-0700 Guideline 6.8.1.lb which states that displays should be assigned to panels in functional groups related to system structure. The grouping should promote easy understanding of the graphic or pictorial display of the system relationships. The third problem is the fact that none of the displays can be read from the user's workstation. The displays are a minimum of fifteen feet from the operator's desk. This does not satisfy Supplement I to NUREG-0737 Requirement 4.1.b which requires a display system which is located convenient to the control room operators..This is also a deviation from NUREG-0700 Guideline 6.1.2.2.e for positioning of displays on stand-up consoles.
The displayed variable for the reactivity control function are located on 2 panels (see Attachment 1) which are separated by approximately 20 feet.
The four IRM-APRMs along with the SRM are located on panel 20C05A while the rest of the displays are located on panel 20C03. The licensee stated in Reference 1 that the reason for not including displays for reactivity 15
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2.
Suppression Pool Level Recorder (1 to 21 feet).,P) c" 3.
Suppression Pool Tesaperature Recorder (30 to 230 j
4.
Drywell Pressure Recorder (5 to 25 psia and 0 to 225 pelg).
Reactor Water Level Recorder 6.
Containement Isolation Valve Position t.lghts(-165,to 450 inches and -325 to 0 inches).
(Open/Close).
7.
Reactor Pressure (0 to 1500 psig).
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control with the other SPDS displays "was based on the human factors consideration of having indicators close to the associated devices used for control." As with the 20C03 panel instruments, the licensee has,not under-stood that the SPDS is for monitoring and assessing, not controlling. Also, the separation of displays is a deviation from Supplement 1 to NUREG-0737 requirement 4.1.b and the concept of a single integrated SPDS display as discussed in NUREG-0835 Section 6.1.
The displayed variables selected by the licensee for radioactivity control are drywell pressure and' Group 1 isolation valve position (see Exhibit 1).
However, based on the licensee's submittal, Reference 1, it is not clear how these parameters will be used to assess radioactivity control.
In addition, containment radiation, vent stack radiation, and off-gas radia-tion levels are not included in the parameters used to assess radioactivity control. These displays are located on panel 20C10, which is approximately 40 feet from the SPDS displays on 20C03 (see Attachment 1).
However, they should be included among displays used for assessment of radioactivity control.
If the SPDS is part of the control board, it must be easily recogniza-ble and readable. The SPDS displays are neither demarcated nor color coded to identify them as SPDS instrumentation. This is a deviation from NUREG-0700 Guideline 6.8.1.3 whi~ch describes several enhancement techniques for setting apart groups of displays.
This applies to both the primary and redundant SPDS instrumentation.
To prevent misleading the control room operator, displayed data should be validated on a "real time" basis where practical.
Each of the Peach Bottom SPDS variables has a primary and a redundant display associated with it. The use of a primary and redundant display is intended to provide the Shift Supervisor with physical redundancy among different critical plant variables. However, the redundant displays identified by the licensee cannot be readily and reliably used by the Shift Supervisor to validate the variable data displayed on the primary display. There are three reasons for this.
First, the redundant displays are not functionally grouped with the primary displays.
For example, primary display for drywell temperature is located on panel 20C03, while the redundant display is located on panel 20C12. The locations of all primary and redundant displays are provided in 17
- to this report.
Second, the redundant displays have the same human engineering discrepancies that were identified on the primary displays. Third, none of the redundant variables can be read l rom
- the Shift Supervisor's workstation during emergency operations.
It is therefore, our judgment that the proposed redundant displays do not satisfy Supplement 1 to NUREG-0737 Requirement 4.1.e.
In summary, the layout of the proposed SPDS on panels 20C03, 20C05A and 20C12 does not constitute a convenient, readily usable display required by Supplement 1,to NUREG-0737.
The primary displays, redundant displays and other suggested displays are spread throughout the control room. The licen-see's argument that the displays should be close to their controls is not relevant since the SPDS is intended for assessment, not control. None of the displays can be read from the operator's desk where the shift supervisor will be located during emergency operations. Further, no detailed task analysis was developed by the licensee to support their case for the proposed SPDS panel layout. Therefore, it is our judgment that the proposed SPDS layout does not satisfy the layout requirement of Supplement I to NUREG-0737.
4.
Procedures and Training The SPDS Supplement I to NUREG-0737 requirement 4.1.c states that procedures which describe the timely and correct safety status assessment when the SPDS is and is not available will be developed by the licensee in parallel with the SPDS. Furthermore, operators should be trained to respond to accident conditions both with and without the SPDS available. To our knowledge, the Peach Bottom SPDS procedures and operator training have not been developed.
5.
Implementation Following the addition of a recorder for reactor pressure, the imple-mentation of the SPDS as interpreted by the licensee will be complete.
Additionally, major portions of the instrumentation are safety-grade and are backed by redundant channels.
The system is available during both normal and abnormal conditions.
It is our judgment that this represents a prompt implementation schedule.
However, prompt implementation of an inadequate 18
SPDS does not constitute a proper response to Supplement 1 to NUREG-0737 requirement 4.1.d.
CONCLUSION Our evaluation of the proposed Peach Bottom SPDS resulted in'the deter-sination that the system meets neither the ~ spirit nor the letter of the SPDS requirements in Surplement I to NUREG-0737.
The variables selected by the licensee to assess the critical safety functions have not been adequately analyzed and justified by the licensee.
The displays, which are dispersed throughout the control room, do not constitute a concise display intended to enhance the operators' ability to avoid degraded and damaged core events.
Several of them cannot be read to the range and accuracy required. None of the displays can be read rapidly and reliably from the shift supervisor's position at the operator's desk. In addition, there are no operating pro-cedures or training programs developed for the proposed system.
In short, the proposed system does not meet the requirement for an SPDS.
e 19 e
e a e s
REFERENCES 1.
Letter from Joseph W. Gallagher, Philadelphia Electric Company, to D.G.
Eisenhut, U.S. Nuclear Regulatory Commission,
Subject:
Peach Bottom Atomic Power Station SPDS Safety Analysis (Parameter Selection) NUREG-0737, Supplement 1. September 28, 1983.
2.
Supplement I to NUREG-0737 - Requirements for Emergency Response Capability (Generic Letter No. 82-33),
U.S.
Nuclear Regulatory Commission December 17, 1982.
3.
NUREG-0700, Guidelines for Control Room Design Reviews, U.S. Nuclear Regulatory Comission, September 1981.
4.
Draft NUREG-0835. Human Factors Review Guidelines for the Safety Parameter Display System Final Report United States Nuclear Regulatory Commission.
5.
Regulatory Guide 1.97, Instrumentation for Light-Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, United States Nuclear Regulatory Commission May 1983.
6.
Letter from D.G. Eisenhut, U.S. Nuclear Regulatory Commission to T.J.
Dente of the BWR Owners' Group dated February 4,1983,
Subject:
Safety Evaluation Report on Emergency Procedure Guidelines, Revision 2. NEDO 24934, June 1982.
i Peach Bottom 1 and 2 SAIC/1-263-07-139-01/02 Contract NRC-03-82-096 l
20
_,7,.-__..
- - -, - - - - - - ~ ~ _ - -
e 9
a ATTACHMENT 1 PEACH BOTTOM SPDS FUNCTIONS AND V ARI AB L ES e
21
O i
?
SUGGESTED SPDS VARIABLES
\\
- CONTAINME T RADIATION
'[/
~ gl
- VENT STACK RADIATION
/
- OFF-GAS RADIATION
- SRM f-
/
N/
/
p
- IRM e
O//' %N, hrT !
o i
i N
l
{,/>/
t i
03 s.
N s SHIFT SUPERVISOR t
I L#k J
~~
l PRIMARY'SPDS VARIABLES
)REDUNDANTSPDSVARIABLES
- 1. REACTOR WATER LEVEL
- 11. REACTOR WATER LEVEL
- 2. REACTOR PRESSURE
[ 2. REACTOR PRESSURE
- 3. DRVWELL PRESSURE
.3.
DRVWELL PRESSURE 4.
DRVWELL TEMPERATURE
- 4. DRYWELL TEMPERATURE
- 5. SUPPRESSION POOL TEMPERATURE
_5.
SUPPRESSION POOL TEMPERATURE
- 6. SUPPRESSION POOL LEVEL
\\
- 6. SUPPRESSION POOL LEVEL
- 7. GROUP 1 CONTAINMENT L 7.
GROUP 1 CONTAINMENT ISOLATION VALVE POSITION
' ISOLATION VALVE POSITION
- 8. NEUTRON FLUX (APRM)
- 8. NEUTRON FLUX (APRM)
OVERVIEW OF PRIMARY, RECUNDANT, AND SUGOESTED SPDS VARIABLES PEACH BOTTOM
--UNIT 2,
SPDS LAVOUT 1.
.: ___i.
g
. g
,us.a
=
._t
-l
\\
- SRti
- 1RM s.
\\/
.l
/
3 L
3 N
g
/
I
~
oa.
9 l
SHIFT SUPERVISOR WORKSTATION
~
URING EMERGENCY lc< A%M E
I
~
PRIMARY t/ARTABLES
- 1. REACTOR WATER LEVEL
- 2. REACTOR PRESSURE REDUNT NT VARIABLES
- 3. DRYWELL PRESSURE
- 1. REACTOR WATER LEVEL
- 4. GROUP 1 CONTAINMENT
- 2. REACTOR PRESSURE ISOLATION VALVE
- 3. DRYWELL PRESSURE POSITION
- 4. GROUP 1 CONTAINMENT
- 5. NEUTRON FLUX (APRM)
ISOLATION VALVE POSITION
- 5. NEUTRON FLUX (APRM)
PEACH BOTTOM
--UNIT 2,
'LAVOUT 2.
.-.,,y
3-._:
- ~
- ..L-.:
L
^
- ~^
^-
- - ~---
^
2 PEACH 80TTOM ATOMIC POWER STATION SPDS VARI *SLES 1.
Reactivity Control A.
Variable: React'or Water Level 1.
Reactor Water Level Chart Recorder (LR-2-2-3-118) i Ranges: 0 to -325 inches of water / +50 to -165 inches of water Location: 20C03-1 (Primary Source of Information)
- 2.
Reactor Water Level Chart Recorder Ranges: 0,to -325 inches of water / +50 to -165 inches of water Location: 20C04-C(RedundantLocationwithRCICSystem) 8.
Variable: Reactor Pressure 1.
Reactor Pressure Chart Recorder Range: 0 to 1500 PSIG Location: 20C03-1 (Primary Source of Information, not installe,d) 2.
Reactor Pressure Chart. Recorder Range: 'O to,1500 PSIG Location: 20C04-C (Redundant Source of Information)
C.
Variable: Drywell Pressure 4
1.
Drywell Pressure Chart Recorder PR-81028 Ra.nges: 0 to 225 PSIG / 0 to 25 PSIA.
Location: 20C03-2 (Primary Source of Information) 2.
Drywell Pressure Chart Recorder PR 8102A Ranges: 0 to 235 PSIG / 0 to 25 PSIA Location: 20C04-C (Redundant Source of Information) 3.
-,-nm-----
--~~n.-__,,,,-,,,-
^ --
. 2._
o.
D.
Variable: Group 1 Containment Isolation Valve Position e
1.
Main Steam Isolation M0-2-80A 1
2.
Main Steam Isolation M0-2-808 3.
Main Steam Isolation M0-2-80C 4.
Main Steam Isolation M0-2-800 5.
Main Steam Isolation M0-2-86A 6.
Main Steam Isolation M0-2-868 7.
Main Steam Isolation M0-2-86C 8.
Main Steam Isolation M0-2-86D Location: 20C03-1 E.
Variable: Neutron Flux (APRM) 1.
IRM-APRM Chart Recorder (NR-2-07-045A). Range: 0-1251 Power Location: 20C05A.
2.
IRM-APRM/RBM Chart Recorder (NR-2-07-46C). Range: 0-125%
Power Location: 20C05A.
3.
IRM-APRM Chart Recorder (NR-2-07-0468), Range: 0-125% Power Location: 20C05A 4.
IRM-APRM/RBMChartRecorder(NR-2-07-046D), Range: 0-125%
Power Location: 20C05A e
4.
t,
.,e
---..,...e.
.-...-...--__..z.-..----..---.-.
_:_ I.'" '
- l"",'.'
[
l" l
'l r
A l,
l l
\\
/\\
i i
. ~.
/
/
,/
/
N/
/
. /..
I i
~
Q 5
1 OQ
,/ -
+
I I
T" SHIFT SUPERVISOR WORKSTATION DURING EMERGENCY
-. * =
.r o -
h :.or.ca N
I
=-
.e r h
l, PRIMARY VARIABLE -- REACTOR WATER LEVEL CHART RECORDER REDUNDANT VARAIBLE -- REACTOR WATER LEVEL CHART RECORDER FUNCTION 2.
REACTOR CORE COOLING AND HEAT REMOVAL I
i PEACH BOTTOM
--UNIT 2,
SPDS iAYOUT 5.
2.
Reactor Core Cooling and Heat Removal A.
Variable: Reactor Water Level 1.
Reactor Water Level Chart Recorder (LR-2-2-3-118)
Ranges: 0 to -325 inches of water / +50 to -165 inches of water Location: 20C03-1 (Primary Source of Information) 2.
Reactor Water Level Chart Recorder Ranges: 0 to -325 inches of water / +50 to -165 inches of water Location: 20C04-C (Redundant Location with RCIC System)
G e
e 0
0 e
6.
r
- l setaa 1
i l
I J,
l l
./\\
l i
I
/
/
~~.
/'
I
,/..
i f
?7 Q
'.'[
9 i
. -i I
i SHIFT SUPERVISOR H
WORKSTATION URING EMERGENCIES M
A~: A % " :.; _
t I ~~.
~
e REDUNDANT VARIABLES PRIMARY VARIABLES
- 1. REACTOR PRESSURE
- 1. REACTOR PRESSURE
._ 2. DRVWELL PRESSURE
- 2. DRYWELL PRESSURE
- 3. GROUP 1 CONTAINMENT
- 3. GROUP 1 CONTAINMENT ISOLATION VALVE ISOLATION VALVE POSITION POSITION i
FUNCTION 3 -- REACTOR COOLANT SYSTEM INTEGRITY PEACH BOTTOM
--UNIT 2,
SPDS LAVOUT 7.
. :... : ^ ~ L.-
~..-......_. _
--^'
^
~-
1 I,
3.
Reactor Coolant System Integrity A.
Verfable: Reactor Pressure i
1.
R6setor Pressure Chart Recorder Range: 0 to 1500 PSIG Location: 20C03-1 (Primary Source of Information. -Not Installed 2.
Reactor Pressure Chart Recorder Range: 0 to 1500 PSIG Location: 20C04-C (Redundant Source of Information) 8.
Variable: Drywell Pressure 1.
Drywell Pressure Chart Recorder PR-81028 kanges: 0 to 225 PSIG / 0 to 25 PSIA.
Location: 20C03-2 (Primary Source of Information) 2.
Drywell Pressure Chart Recorder PR-8102A Ranges: 0 to 225 PSIG / 0 to 25 PSIA.
Location: 20C04-C (Redundant Source of Information) l C.
Variable: Group 1 Containment Isolation Valve Position 1.
Main Steam Isolation M0-2-80A 2.
Main Steam Isolation M0-2-808 3.
Main Steam Isolation NO-2-80C 4.
Main Steam Isolation MO-2-80D 5.
Main Steam Isolation MO-2-86A 6.
Main Steam Isolation' M0-2-868 7.
Main Steam Isolation M0-2-86C 8.
Main Steam Isolation M0-2-86D Locatfon: 20C03-1 i
8.
_. _. ~,,
_ _ n-
, g
,im a y
1 a
/
i
- SUGGESTED VARIABLES I
a l
l
- CONTAIN^1ENT RADTATION y
VENTSTACKRAD1ATION[
/
8
- OFF-GAS RADIATION /
N/
/
7
('
h' m
n 9
t l
f SHIFT SUPERVISOR WORKSTATION DURING EMERGENCIES
~.
5 AN
.c 4 r_
PRIMARY VARIABLES REDUNDANT VARIABLES
- 1. DRYWELL PRESSURE
- 2. GROUP 1 CONTAINMENT I
GROUP 1 CONTAINMENT ISOLATION VALVE ISOLATION VALVE POSITION POSITION FUNCTION 4 -- RADIOACTIVITY CONTROL
~
PEACH BOTTOM
--UNIT 2,
SPDS LAVOUT 9.
.u 3
f 4.
Radioactivity Control A.
Variable: Drywell Pressure 1.
Drywell Pressure Chart Recorder PR-81028 Ranges: 0 to 225 PSIG / 0 to 25 PSIA.
Location: 20C03-2 (Primary Source of Information) 2.
Drywell Pressure Chart Recorder PR-8102A Ranges: 0 to 225 PSIG / 0 to 25 PSIA.
Location: 20C04-C (Redundant Source of Information) 3 Variable: Group I containment Isolation Valve Position 1.
Main Steam Isolation M0-2-80A i
f.
Main Steam Isolation M0-2-808 3.
Main Steam Isolation M0-2-80C 4.
Main Steam Isolation MO-2-800 5.
Main Steam Isolation M0-2-86A 6.
Main Steam Isolation M0-2-868 7.
Main steam Isolation M0-2-86C 8.
Main Steam Isolation M0-2-860 Location: 20C03-1 9
9 0
9 e
10.
..- L..: '. _ :__.
_ ~~--_ _ ~ -
^ - ~ -
~
3
_3
^~
i e
3 fl ~
I
.l l
[
4 m
w.
t
\\
f<
.s..
/
r.
I f
x o
t I !
l 4..
Q.
l t
~ SHIFT SUPERVISOR i WORKSTATION.
DURING' EMERGENCIES
[re,ye h r_
L REDUNDANT VARIABLES PRIMARY VARIABLES
- 1. DRYWELL PRESSURE ~
- 1. DRYWELL PRESSURE p
pp
- 2. DRYWELL TEMPERATURE
(
- 3. SUPPRESSION POOL
- 3. SUPPRESSION POOL
' " " ^
~
- 4. SUPPRESSION POOL LEVEL
- 4. SUPPRESSION POOL LEVEL e
FUNCTION 5 -- CONTAINMENT CONDITIONS PEACH BOTTOM
--UNIT 2,
SPDS LAVOUT u.
.r
-... -.. m.
w..
=:--_.---.__-
5.
Containment Conditions A.
Variable: Reactor Pressure 1.
Reactor Pressure Chart Recorder Range: 0 to 1500 PSIG Location: 20C03-1 (Primary Source of Information. Not Installed) 2.
Reactor Pressure Chart Recorder Range: 0 to 1500 PSIG Location: 20C04-C (Redundant Source of Information)
- 8.. Variable: Drywell Pressure 1.
Drywell Pressure Chart Recorder PR-81028 Ranges: 0 to 225 PSIG / 0 to 25 PSIA.
Location: 20C03-2.(Primary Source of Information) i 2.
Drywell Pressure Chart Recorder PR-8102A Ranges: 0 to 225 PSIG / 0 to 25 PSIA.
Location: 20C04-C (Redundant Source of Information)
C.
Variable: Drywell Temperature 1.
Drywell Temp, era'ture Chart Recorder TR-4805 Range: 0 to 2400F Location: 20C03-3 (Primary Source of Information) 2.
Drywell Temperature Location: Multipoint Digital Display 20C12 (Redundant
. Source of Information).
12.
D.
. Variable: Suppression Pool Level 1.
Torus Water Level Chart Recorder LR-81238 Range: 1 to 21 feet Location: 20C03-2 (Primary Source of Information) 2.
Torus Water Level Chart Recorder LR-8123A Range: 1 to 21 feet Location: 20C04-C O
O e
13.
,_. _,,...., -...,.... _,. -,, - -,. _