ML18053A679
| ML18053A679 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 12/01/1988 |
| From: | Smedley R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML18053A680 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.05, TASK-TM TAC-49700, NUDOCS 8812060069 | |
| Download: ML18053A679 (11) | |
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consumers Power POW ERi Nii MICHlliAN"S PROliRESS General Offices: 1945 West Parnell Road, Jackson, Ml 49201 * (517) 788-0550 December 1, 1988 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT -
NUREG-0737, ITEM II.K.3.5 REACTOR COOLANT PUMP TRIP (TAC NO 49700) -
ADDITIONAL INFORMATION NRC letter dated April 15, 1988 requested additional information regarding Reactor Coolant Pump (RCP) Trip setpoint$, instruments, instrument uncertain-ties and procedures involved in the Palisades implementation of the Trip 2/Leave 2 trip strategy during transients be submitted within 60 days.
Consumers Power Company submittals dated June 10 and September 23, 1988 requested additional response time to allow our consideration of CE Owners Group studies regarding instrument inaccuracy.
The requested additional information is provided below.
NRC Question:
"l.
Consumers Power's letter of May 29, 1987 stated that reactor coolant system pressure was used to trip the first set of pumps.
It also stated that secondary pressure and containment radiation and secondary system radiation were used to distinguish steam line breaks and tube ruptures from small break LOCAs.
However, the letter did not provide the plant specific setpoints for these parameters.
Identify the setpoints used to determine when to trip the first and second set of pumps."
CPCo Response:
The first two Primary Coolant Pumps (PCPs) are tripped in Emergency Operating Procedure (EOP) 1.0, "Standard Post Trip Actions".
The setpoint for tripping these first two RCPs is pressurizer pressure at or below 1300 psia. _The valu~ of 1300 psia is based on information from Appendix 1 of CEN-268, i.e. following a small break LOCA, Primary Coolant System (PCS) pressure stabilizes at a pressure above the steam generator secondary side pressure.
CEN-268 expresses the nominal setpoint for tripping the first two PCPs of a 2700 MWt class plant at 1210 psia.
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Nuclear Regulatory Commission Palisades Plant Rx Coolant Pump Additional Info December 1, 1988 2
the 91 psia error of Enclosure 2 (for PI-0104) is included this results in:
1210 psia + 91
~ 1300 psia.
The error for PI-0104 is used since this is the digital readout instrument which is most visible to the operators in the Control Room and is most frequently used.
The other pressurizer pressure instrument errors are only 29 psia higher than PI-0104.
Since the calculation for these errors are conservative, it is felt that the 1300 psia is a valid setpoint regardless of which pressuri~er pressure instrument is used.
Palisades Technical Specification 3.1.7 requires secondary steam safety valve maximum relief setting of 1940 psia.
Assuming a 3% relief setpoint accuracy results in:
1040 psia + (0.03 x 1040) = 1071.2 psia.
Hence, as required by CEN-268, the small break LOCA pressure plateau is below the 1300 psia trip setpoint for the first two PCPs.
The second two PCPs may be tripped in EOP 1.0 if pressurizer pressure falls below the "minimum pressure for PCP operation" curve (refer to of Enclosure 1).
This curve is primarily based on net positive suction head for PCP operation.
For large break LOCA events this curve would obviously be exceeded and the second two PCPs would be tripped.
In other events (e.g. steam generator tube rupture) the tripping of the second two PCPs would be performed if pressurizer pressure fell below the curve (based on existing PCS temperature).
LOCA VS. NON-LOCA EVENT A principle decision for tripping the second two PCPs outside of EOP 1.0 is the diagnosis of the event (i.e. LOCA versus non-LOCA events).
In the Palisades EOPs the operators perform this diagnosis using the "Event Diagnostic Flow Chart" (refer to Attachment 1 of Enclosure 1). As can be seen on this attachment (Page 4 of 8), the diagnosis of LOCA events is initially determined by the trending of pressurizer level and pressurizer pressure.
If either of these parameters is trending in the non-conservative direction, then a LOCA event may be in progress.
The value of 1650 psia is above the automatic safety injection actuation system setpoint of 1605 psia.
On page 6 in Attachment 1 to EOP 1.0, the operators are asked if PCS subcooling is rising or if either steam generator is less than 700 psia.
An affirmative answer to either of these questions indicates an excess steam demand event is in progress (as opposed to a LOCA event).
The value of 700 psia is based upon recent analysis (reference Enclosure 2) which indicates that the accuracy of the steam generator (S/G) pressure instruments is 90 psia during small break LOCA events.
The nominal post-trip steam generator pressure is expected in the range of 700 psia to 900 psia (reference Enclosure 1, Step 9c).
If a LOCA event was in progress, a negative response should have occurred 0Cl188-0207-NL02
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Nuclear Regulatory Commission Palisades Plant Rx Coolant Pump Additional Info December 1, 1988 3
from the previous questions.
In this case, the operators are asked if containment pressure is greater than 1 psig or increasing.
A positive response to this question indicates a LOCA inside containment and the LOCA procedure is considered.
A negative answer indicates either a S/G tube rupture event or a LOCA outside Containment.
In the later case the operators are asked if steam plant radiation monitors activity are high.
This is a qualitative type of a question which calls for the operators to make a judgement that these monitors are indicating higher than "normal."
In general the operators would key off of alarms on these monitors to answer the question positively.
These monitors are located outside of the Containment Building and their accuracies are described in Enclosure 3.
A negative response to this question would result in the LOCA procedure being considered.
In EOP 1.0, the operators are asked one last question regarding diagnosis of one event being apparent (reference Page 8 of 8 of Attachment 1 to ).
If the operator does not feel that only one event is in progress, then the Functional Recovery Procedure is implemented.
If only one event is considered in progress, the.n the appropriate EOP which was previously considered would be implemented.
Tripping of the second two PCPs in EOPs which are entered after EOP 1.0 falls into two strategy categories:
LOCA and non-LOCA.
The LOCA strategy procedures are EOP 4.0, '~oss of Coolant Accident Recovery" and EOP 9.0, "Functional Recovery Procedure." In both of these procedures the second two PCPs are tripped if pressurizer pressure falls below 1300 psia (see previous discussion for development of this setpoint).
The non-LOCA strategy procedures are EOP 5.0, "S/G Tube Rupture" and EOP 6.0, "Excess Steam Demand Event."
In both of these procedures, the second two PCPs are tripped if Pressurizer pressure falls below the minimum pressure for PCP operation curve (see previous discussion on this item).
Therefore, the PCP trip strategy follows the guidance found in the Com-bustion Engineering Emergency Procedure Guidelines, CEN-152 (which used CEN-268 as a basis for its development).
NRC Question:
"2.
The letter of May 29, 1987 provided preliminary estim~tes of the instru-ment uncertainties for normal and accident conditions.
Provide the final uncertainties for review.
Also, for the setpoints identified in response to question 1, discuss how the effects of instrument uncertainty were included in determining the setpoints for primary pressure, secondary pressure, and containment and secondary reactivity."
OC1188-0207-NL02
0 Nuclear Regulatory Commission Palisades Plant Rx Coolant Pump Additional Info December 1, 1988 CPCo Response:
Final instrument uncertainties are included in Enclosures 2, 3, and 4.
The effects of the instrument uncertainties on the setpoints used in response to Question 1 are described in the response to Question 1.
NRC Question:
4 "3.
Consumers Power did not provide sufficient information in its May 29, 1987 response to GL 86-06 item 3 to determine how the uncertainties in the generic analysis presented in CEN-268 affect the results as they apply to Palisades.
Therefore, identify the Palisades plant specific features not representative of the reference plant used in the analyses presented in CEN-268 to determine the setpoints for tripping the first and second sets of pumps.
At a minimum, Consumers Power should discuss core power; decay heat; HPIS capacity; steam generator tube areas; and setpoints for reactor trip, safety injection, and accumulator injection.
Show that the values used in the generic analysis are either representative of those at Palisades or conservative.
If a reference plant parameter is not representative for Palisades, discuss how this was considered in determining the plant specific setpoints."
CPCo Response:
The Palisades plant specific parameters and those irt generic analysis CEN-268 are summarized on Table 1.
The following discussion presents our evaluation of how Palisades plant specific features would affect the analysis results presented in CEN-268 to determine required setpoints for tripping the first and second set of pumps.
The evaluation was performed only in consideration of LOCA events, both for best estimate* and licensing cases, our primary concern being to conservatively identify LOCA events and trip all four pumps.
The benefits to other events in tripping the first and second set of pumps at higher or lower setpoints were not considered.
The comparison indicates that, for the most part, the parameters in CEN-268 are representative of the Palisades plant or conservative.
The generic (CEN-268) analysis is considered conservative in comparison to Palisades when the values used in the generic analysis would have a smaller safety margin than woul9 the values used at Palisades.
Major differences are discussed below.
The Palisades plant normal operating power level, PCS Pressure, and PCS Temperature are lower, prior to event initiation, than values used in the generic analysis.
Since the power level is much lower, the Decay Heat OC1188-0207-NL02
Nuclear Regulatory Commission Palisades Plant Rx Coolant Pump Additional Info December 1, 1988 5
generated after trip, in both best estimate and licensing analysis cases, will be much lower.
Based upon engineering principles the lower initial values of RCS pressure and temperature would cause the RCS pressure to drop to lower values upon initiation of a LOCA condition and subsequently recover to a lower pressure plateau.
Hence, the recommended setpoint is conservative.
The difference in Charging Pump capacity is not considered to be signif-icant, and would not substantially affect results of the LOCA cases.
The Palisades HPSI capacity is greater than that modeled in CEN-268 and is, therefore, conservative.
Figure 2 indicates that the CEN-268 case may have modeled partial LPSI injection at an RCS pressure of 209 psia.
The Palisades LPSI injection is initiated at 193 psia.
However, since the RCS pressure in analyzed cases does not reach 209 psia, LPSI injection would not occur in these events.
The generic analysis assumes a steam generator heat transfer area degraded by 50%.
The Palisades plant specific area is based upon the number of steam generator tubes that have been plugged.
The current reduction in area due to tube plugging in both steam generators results in a degradation of 24%.
The generic analysis is therefore conservative.
Setpoints used in the analysis for steam generator secondary side relief valves, for reactor trip, for safety injection, and for accumulator injection are either representative of those at Palisades or conservative.
Temperature of SI and charging water used in the analysis is non-conservative in comparison to temperatures used in Palisades plant specific analyses.
It is estimated that using a temperature of 100 deg. F for SI and charging would be equal to less than a 10% loss in heat removal capability of.
injected and make-up water.
However, the generic analysis decay heat generated is approximately 10% higher and offsets the temperature difference.
More importantly, the Palisades plant specific HPSI injection flowrate is considerably higher than the generic analysis and would provide a greater mass for heat removal and core re-flood.
We estimate that Palisades plant specific analysis would demonstrate less core uncovery and/or uncovery for a shorter time period.
OC1188-0207-NL02
Nuclear Regulatory Commission Palisades Plant Rx Coolant Pump Additional Info December 1, 1988 TABLE 1 COMPARISON OF REFERENCE PARAMETERS GENERIC ANALYSIS IN CEN-268 vs PALISADES PLANT SPECIFIC Reference Parameter Core Power - nominal
- event initiation Decay Heat - Appendix K
- best estimate Initial PCS Pressure Initial Tcold Make up flow
- one charging pump HPIS capacity - Appendix K-one pump Best estimate
- one pump analyses
- two pump analyses Steam Generator tube areas -
Setpoints: Steam Generator SV Reactor Trip - LOCA Safety Injection Accumulator Injection SI and Charging water temp.
OC1188-0207-NL02 CEN-268 2700 Mwt 2754 Mwt
- 1. 2 x 1971 ANS 1.0 x 1979 ANS 2250 psia 550 deg F 44 gpm Figure 1 Figure 2 Figure 2 Palisades 2530 Mwt 2580 Mwt same same 2100 psia 544 deg F 34 gpm Figure 3 Figure 3 Figure 3 See discussion 1000-1050 psia 985-1025 psia 1750 psia 1750 psia 1°600 psia 1605 psia 200 psig 200 psig 70 deg F 100 deg F 6
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Nuclear Regulatory Commission Palisades Plant Rx Coolant Pump Additional Info December 1, 1988 NRC Question:
"4. Consumers Power did not identify which emergency operating procedures (EOPs) require the use of RCP pump trip guidelines in its response to GL 86-06, item 4.
Identify the EOPs (EOP title and number) to complete the review of this item."
CPCo Response:
NRC The following Emergency Operating Procedures (EOPs) require the use of the RCP pump trip guidelines:
EOP Number EOP Title EOP 1.0 Standard Post Trip Actions EOP 4~0 Loss of Coolant Accident Recovery EOP 5.0 Steam Generator Tube Rupture EOP 6.0 Excess Steam Demand Event EOP 9.0 Functional Recovery Procedure Question:
10 "5.
Consumers Power indicated in its May 29, 1987 response to GL 86-06 that one of the parameters used to determine whether the second set of pumps should be tripped was containment radiation.
In CEN-268, Rev 1, and in CEN-268, Supplement 1, Rev 1, it was stated that a containment radiation alarm was no longer recommended for use in the T2/12 pump trip strategy.
This was because containment radiation was determi~ed not to have satis-factory sensitivity.
In light of the recommendations in CEN-268, Rev 1, justify continued use of this parameter in the Palisades pump trip strategy.
Provide information to show the Palisades containment radiation alarms are sufficiently sensitive to perform the function assigned them in the Palisades plant specific T2/12 pump trip strategy."
CPCo Response:
Palisades originally implemented its upgraded EOPs (in response to NUREG 0737 Supplement 1) based on Combustion Engineering Emergency Procedure OC1188-0207-NL02
Nuclear Regulatory Commission Palisades Plant Rx Coolant Pump Additional Info December 1, 1988 11 Guidelines, CEN-152 Revision 2.
In this initial upgrade the Event Diag-nostic Flow Chart for EOP 1.0 included a decision point concerning Con-tainment Radiation which was designed to separate LOCA from non-LOCA events.
This has subsequently been changed such that this decision point is not used to differentiate between LOCA and non-LOCA events.
This decision point continues to be used prior to considering the Excess Steam Demand Event procedure since analyses has shown that these monitors will indicate high during high energy line break scenarios inside Containment.
In such events the operators are directed to the Functional Recovery Procedure (EOP 9.0) since implementation of the Excess Steam Demand Event procedure (EOP 6.0) would eventually lead to the use of EOP 9.0 when operators detected the abnormal monitor response (i.e. if these monitors were found to be alarming in EOP 6.0, then EOP 9.0 would be implemented).
Richard W Smedley Staff Licensing Engineer CC Administrator, Region III, NRC NRC Resident Inspector..... Palisades Enclosures :
EOP 1.0, Standard Post Trip Actions :
EA-RJC-88-01, Calculations of Loop Accuracies for Rosemount Transmitters During a Small Break LOCA :
EA-A-PAL-88-020, Calculations of Loop Accuracies During Various Accident Scenarios - Radiation Monitoring :
EA-RJC-88-03, Instrument Loop Accuracy Calculations - PCS Temperature Loops OC1188-0207-NL02