ML20080N297
| ML20080N297 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 09/28/1983 |
| From: | Gallagher J PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 GL-82-33, NUDOCS 8310040434 | |
| Download: ML20080N297 (8) | |
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PHILADELPHIA ELECTRIC COMPANY 2301 MARKET STREET P.O. BOX 8699 JOS&PH W. GALL AGH ER stsetnic ruooucv om espaaruser 12151841-5003 September 28, 1983 Docket Nos. 50-277 50-278 Mr. Darrell G. Eisenhut, Director Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C.
20055
Subject:
Safety Parameter Display System Peach Bottom Atomic Power Station
Reference:
Correspondence dated April 15, 1983, S.L. Daltroff, PECo to D.G. Eisenhut, NRC, titled: Response to NUREG-0737, Supplement 1.
Dear Mr. Eisenhut:
This letter transmits a safety analysis describing the basis on which the ' selected parameters f or the Saf ety Parameter Display System (SPDS) are sufficient to assess the safety status of each identified function for a wide range of events.
The safety analysis provides the documentation required by Section 4.2.a of Generic Letter 82-33 (Supplement 1 to NUREG-0737 to all licensees),
and meets our commitment identified in Attachment 1, Page 3 of the referenced letter.
Should you have any questions regarding the enclosed safety analysis, please do not hesitate to contact us.
Very truly yours,
Y k
WCBilm Attachment cc:
A. R. Blough Site Inspector 9
8310040434 830928 PDR ADOCK 05000277 p
A Docket Nos. 50-277 50-278 Peach Bottom Atomic Power Station SPDS Safety Analysis (Parameter Selection)
NUREG-0737, Supplement 1 i
Introduct2cn Supplement 1 of NUREG-0737, " Requirements for Emergency Response C.7pability," contains the basic requirements for the Safety Parameter f>iaplay System (SPDS) for naclear power plants.
In the referenec 2 lett: r, Philadelphia ' Electric Company committed to the Nuc3 car Regulatory. Commission (NRt ) to implement a SPDS at Peach Bottun APS Units 2 and 3.
This letter contained a description of the SPDS, a list of the parameters monitmed, and a comitment to provide a written safety analysis describing the basis on which the selected parameters are sufficient to assess i.
the safety status of each identified function. The purpose of the subject analysis is to meet this NRC commitment.
t References
.I 1.
NUREG-0737, Supplement 1, " Requirements for Emergency Response Capability."
'2.
Letter from S. L. Daltroff of PECo to D. G. Eisenhut of the NRC, dated 4/15/83, " Response to NUREG-0737, Supplement 1."
3.
Letter from T. J. Dente of the BWR Owners' Group to D. G. Eisenhut of the NRC, dated 6/1/82, "NEDO-24934, Emergency Procedure Guidelines, BWR/1-6, Revision 2."
4.
Letter from T. J. Dente of the BWR Owners' Group to D. G.
Eisenhut of the NRC dated 10/4/82, " Errata to BWR Emergency Procedure Guidelines."
5.
Letter from D. G. Eisenhut of the NRC to T. J. Dente of the
{
BWR Owners' Group dated 2/4/83, " Safety Evaluation Report on Emergency Procedure Guidelines, Revision 2, NEDO'24934, June 1982."
SPDS Requirements The requirements of NUREG-0737, Supplement 1 that are applicable to this analysis are:
l.
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 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 and the initial phase of an accident.
(NUREG-0737, Supplement 1, Section 4.1.a)
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2.
The SPDS will continuously display information from which the plant safety status can be readily and reliably assessed by control room personnel who are responsibic for the avoidance of degraded and damaged core events.
(NUREG-0737, Supplement 1, Section 4.1.b) 3.
The minimum information to be provided shall be sufficier.t to l
provide information to p; ant operatora aheat:
(i)
Reactivity control f
=
(ii)
Reactor core cooling and heat removM from the primary system (iii)
Reactor coolant system integrity 3
2 (iv)
Radioactivity control (v)
Containment conditions.
The specific parameters to be displayed shall be determined by the licensee.
(NUREG-0737, Supplement 1, Section 4.1.f)
SPDS Design Analyzed Existing instrumentation in one area of the Peach Bottom control i
room, plus the addition of one parameter, has been selected to serve as the SPDS.
Following the addition of a recorder for reactor pressure, the implementation of the SPDS will be complcte.
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The SPDS uses indicators, recorders, and indicating lights j'
arranged in one section of the control room in the vicinity of the controls used by the operator during abnormal and emergency conditions to avoid a degraded core.
The majority of these displays have been added as a result of other requirements contained in NUREG-0737.
Their location was chosen based on human factors engineering principles and on the understanding of the intent of the SPDS concept that existed at the time the location was chosen.
Utilization of this instrumentation in lieu of an independent display enhances the integration of the SPDS concept into the control. room design.
Additionally, major portions of the instrumentation are safety-grade and are backed up by redundant channels.
The system is available during both normal
- and abnormal conditions.
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. J The systen analyzed includes displays for reactor core cooling and heat removal from the primary systen, reactor coolant system integrity, radioactivity control, and containment canditions.
The variables monitored for each of these functions are shown in Table 1.
The arrangenent of the SPDS and its relationship to the emergency core cooling systems. primary containment isolation systern (PCIS),
centainment atmospheric dilution (CAD) system, and reactor core isolation cooling (RCIC) system is shown ir. Figure 1.
Reference 1 stated that as a result of integrating the SPDS into the control room, it was concluded that displays for reactivity control should not be part of the Peach Bottom SPDS.
This conclusion was based on the human factors consideration of having indicators close to the associated devices used for control.
Analysis for Parameter Selection The basis for SPDS parameter selection is the entry conditions for the upgraded emergency operating procedures.
These new Peach Bottom procedures are designated as the Transient Response Implementation Plan (TRIP) procedures and are in conformance with revision 2 of the BWR Owners' Group Emergency Procedure Guidelines (EPGs).
The emergency operating procedures are symptom oriented, that is the safety status of the plant and correct operator responses are unambiguously defined on the basis of observed symptoms. The EPG development process analyzed a great multitude of conditions with a mixture of qualitative and quantitative analysis to determine that the actions based on the occurring symptoms were appropriate for all events. The analytic basis for the EPGs and the EPG steps were transmitted to the NRC in references 3 and 4.
This information was extensively reviewed by the NRC and a safety evaluation report was issued (reference 5) on EPG revision 2.
This safety evaluation firmly supports the BWR symptomatic approach and the selection of the entry conditions. On this basis, the safety analysis required to establish those parameters which determine the safety status of a BWR plant is already complete.
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j 4 Since the SPDS is to be used by control room personnel who are responsible for the avoidance of degraded and damaged core events, there are two emergency procedures of interest for the purpose of s
determining the SPDS parameters. These are the reactor control prccedure and the containment control procedure. The plant parameters and their limits which trigger entry into these procedures are clearly i
defined and are few in number. The entry conditions for the reactor control procedure ares a.)
Scram condition with reactor power above three percent or unknown.
b.)
Reactor level below -48 inches or unknown.
c.)
Drywell pressure above two psig d.)
Group I containment isolation.
The entry conditions for the containment control procedure are:
i a.)
Torus temperature above 95'F b.)
Torus level outside 14.6 feet to 14.9 feet c.)
Drywell pressure above two psig d.)
Drywell temperature above 145'F.
The Peach Bottom SPDS parameters listed in Table 1 include all of the above entry conditions with the exception of some of the scram initiators.
Scram parameters of reactor water level, reactor pressure, drywell pressure, and main steam line isolation are included in the SPDS to provide information about reactor core cooling and heat removal from the primary system, reactor coolant system integrity, radioactivity control, and containment control.
The list of SPDS parameters should be revised to indicate that these parameters are also used to provide information about. reactivity control.
.The remainder of the automatic scram initiators (ie., turbine control valve closure, turbine stop valve closure, scram discharge volume high level, main steam line high radiation, condenser low vacuum, and neutron monitoring system trips) are used only for indication of reactivity control.
These can be left out of the SPDS ' because of the human factors consideration of locating the reactivity control displays close to the associated devices used for control. In the Peach Bottom control room, alarms for scram conditions, neutron flux indication, control rod position indication, control rod manual controls, scram pushbuttons, f
and standby liquid control system controls are all located on one side of the control room while the SPDS and the controls for the emergency core cooling systems and primary containment isolation system are located on the other side of the room.
It is concluded that it is much I
more effective to have the reactivity control indicators close to j
their associated controls instead of grouped with the SPDS.
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. In summary, the EPGs have been submitted to the NRC and approved.
The Peach Bottom TRIP procedures implement the EPGs on a plant specific basis.
The TRIP procedure entry conditions are symptomatic of emergencies and events which may degrade into emergencies and specify actions appropriate for these conditions.
As such, the TRIP procedure entry conditions are adequate to determine the safety status of the plant. The entry conditions for the reactor control and containment control portions.of the TRIP procedures provide sufficient information to plant operators about:
a.)
Reactivity control b.)
Reactor core cooling and heat removal from the primary system c.)
Reactor coolant system integrity d.)
Radioactivity control e.)
Containment conditions The revised Peach Bottom SPDS parameter list is shown in Table 2 and includes the entry conditions for these functions (except as justified above for those parameters used only for reactivity control). The control room operators will use this information to determine the safety status of the plant and to assess whether abnormal conditions warrant corrective action-to avoid a degraded core.
Conclusion The SPDS has been integrated into the' control room design, di development of symptom-oriented emergency operating procedures, and operator training.
The final SPDS will provide a concise display of critical plant variables to the control room personnel who are responsible for the avoidance of degraded and danaged_ core events to aid them in rapidly and reliably determining the' safety status of the-plant.
This analysis shows the basis en which the SPDS parameters were selected and shows that the selected parameters are sufficient to assess the safety status of each identified function.
Table 2 shows the final list of SPDS parameters and their associated functions.
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FIGURE 1 SPDS ARRANGEMENT 67"
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j 1-2,3,4,5 90" I6 CORE CORE RCIC 11PCI RHR 3 PRAY PCIS SPRAY RHR MSIV ADS CAD CAD i
if Approximate Scale 1" =25" SPDS Indicators 1.
Drywell Temperature Recorder (0 to 240 F).
2.
Suppression Pool Level Recorder (1 to 21 feet).
3.
Suppression Pool Temperature Recorder (30 to 230 F).
4.
Drywell Pressure Recorder (5 to 25 psia and 0 to 225 psig).
5.
Reactor Water Level Recorder (-165 to +50 inches and -325 to 0 inches).
i 6.
Containment Isolation Valve Position Lights (Open/Close).
7.
Reactor Pressure (0 to 1500 psig).
1 1
TABLE 1 SPDS PARAMETERS Function Reactor Reactor Core Coolant Cooling and System Radioactivity Containment Variable Heat Removal Integrity Control Conditions Reactor Water Level I
Riactor Pressure I
Drywell Pressure I
I I
Drywell Temperature I
Suppression Pool Temperature I
Suppression Pool Level I
C:ntainment Isolation ValvePosition(1)
Reactor Lines I
I Other Lines I
(1) Power operated valves except those on systems whose continued operation is essential to the mitigation of a accident.
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TABLE 2 Revised-4PDS ' Parameter List l
Reactor Function l
Reactor Core Coolant 1
Reactivity Cooling And System Radioactivity Containment Control (2)
Heat Removal Integrity Control Conditions I
Variable Reacter Water Level X
X RIactor Pressure X
X Dryw0ll Pressure X
X X
X X
Dryw ll Temperature Suppr:ssion Pool X
Temperature X
Suppr:ssion Pool Level Contcinment Isolation Valva Position (1)
- Rractor Lines X
X X
X
- Other Lines (1) Power operated valves except those on systems whose continued operation is essential to the mitigation of a accident.
(2) Remaining indicators for reactivity control are not part of the SPDS. They are grouped with the associated devices used for control.
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