ML20235J835
| ML20235J835 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 09/30/1987 |
| From: | Andrews R OMAHA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| TASK-2.K.3.05, TASK-TM GL-86-06, GL-86-6, LIC-87-649, NUDOCS 8710020117 | |
| Download: ML20235J835 (7) | |
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Omaha Public Power District 1623 Harney Omaha. Nebraska 68102 2247 402/536-4000 September 30, 1987 LIC-87-649
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U. S. Nuclear Regulatory Commission l
ATTN:
Document Control Desk Washington, DC 20555
References:
(1)
Docket No. 50-285 (2)
Resolution of TMI Action Item II.K.3.5, " Automatic Trip of Reac-tor Coolant Pumps," NRC Generic Letter No.83-10a, February 8, 1983.
(3)
Resolution of TMI Action Item II.K.3.5, " Automatic Trip of Reac-tor Coolant Pumps," NRC Generic Letter No. 86-06, May 29,1986.
(4)
Combustion Engineering Nuclear Power Systems Division, "Justifi-cation of Trip-two/ Leave-two Reactor Coolant Pump Trip Strategy During Transients," (prepared for the CE Owners Group) Combus-tion Engineering Report CEN-268 (March 1984) and Supplement.
(5)
Letter OPPD (R. L. Andrews) to NRC (A. Bournia) dated July 16, 1987.
(LIC-87-516)
Gentlemen:
SUBJECT:
Response to Generic Letter 86-06 Enclosed is the Omaha Public Power District's plant specific response for Fort Cal-houn Station Unit I as required in Reference 3 above, regarding our trip-two/ leave-two procedures for small break loss of coolant accidents (LOCA).
Please contact us if you have any questions.
l Sincerely, (4ihm.w../~
R. L. Andrews Division Manager Nuclear Production RLA/rh Attachments c:
LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N.W.
Washington, DC 20036 R. D. Martin, NRC Regional Administrator A. Bournia, NRC Project Manager P. H. Harrell, NRC Senior Resident Inspector g
0710020117 870930 g
ADOCK 0500 5
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f ATTACHMENT Response to Generic Letter 86-06 l
The Omaha Public Power District's trip-two/ leave-two (T2/L2) reactor coolant pump (RCP) trip strategy has incorporated the Combustion Engineering Owners Group (CE0G). methodology into the Fort Calhoun Station Unit No.1 Emergency Operating Procedures (EOPs). The CEOG methodology in Reference (4) is applicable to Fort Calhoun Station in resolving the generic issues presented in Appendix A of Generic Letter 86-06.
The specific issue of instrumentation selection and uncertainties is addressed in questions 1 through 4 as found in Reference (3),Section IV, Implementation.
Question 1 Identify the instrumentation to be used to determine the RCP trip setpoints, in-cluding the degree of rodundancy of each parameter signal needed for the criteria chosen.
ER1QQ1112 The Fort Calhoun Station Unit No. 1 Emergency Operating Procedures (EOPs) use the following instrumentation to diagnose RCS depressurization events, to determine wisen to trip RCPs, or as feedback to verify plant status and E0P selection: RCS (i.e., pressurizer) pressure, steam generator pressure, containment pressure, RCS temperature, RCS subcooling, pressurizer level, steam generator level, containment radiation, and secondary plant radiation.
RCS pressure and steam generator pressure are both used quantitatively in the E0Ps to directly implement the RCP trip strategy. The other instrumentation listed is used in a qualitative or trending sense to help implement the strategy.
INSTRUMENTATION USED QUANTITATIVELY The pressurizer pressure signal is used quantitatively for tripping the first two RCPs at a numerical setpoint regardless of the specific event diagnosis. The degree of redundancy for the instrument channel whose ranges are of interest is:
i 2 Channels Loop 105/115 0-2500 PSIA Steam generator pressure is also used in a quantitative (setpoint) sense to dis-tinguish Uncontrolled Heat Extraction events (VHEs) from LOCAs and Steam Generator Tube Rupture events (SGTRs).
The degree of redundancy for steam generator pres-sures for the ranges of interest is as follows:
Steam Generator RC-2A Loop 913 (Channels A,B) 0-1200 PSIA Loop 902 (Channels A,B,C,0) 0-1000 PSIA Steam Generator RC-2B Loop 914 (Channels A,B) 0-1200 PSIA Loop 905 (Channels A,B,C,0) 0-1000 PSIA
INSTRUMENTATION VSED OVALITATIVELY The remainder of the instrumentation is used qualitatively.
This instrumentation falls into two categories:
a) instrumentation used in the event diagnostic pro-cess, and b) instrumentation used in a verification or feedback process.
a)
INSTRUMENTATION VSED IN THE EVENT DIAGNOSTIC PROCESS Pressurizer level (trending) and RCS pressure decreasing rapidly (trending) are used to initially identify any RCS depressurization event.
Containment pressure is used to identify events within containment. The absence of secondary radiation is used to distinguish LOCAs from SGTRs; pressurizer level trending is used to help determine that the event depressurization is not caused by heater or spray malfunction. A Break Identification logic diagram for assisting operators in break identification is attached. The degree of redundancy for each of these parameters for the ranges of inter-est are as follows:
PRESSURIZER PRESSVRE 2 Channels Loop 105/115 0-2500 PSIA PRESSURIZER LEVEL 2 Channels Loop 101 X/Y 0-100% Level CONTAINMENT PRESSURE 2 Channels Loop 785/786
-5 PSIG to 5 PSIG 2 Channels Loop 783/784
-5 PSIG to 195 PSIG SECONDARY RADIATI@
1 Channel, condenser offgas, RM-057 1 Channel, per steam generator blowdown line, RM-054A and RM-0548.
b)
INSTRUMENTATION VSED IN A VERIFICATION OR FEEDBACK SENSE The remaining four parameters (RCS temperature, RCS subcooling margin, con-tainment radiation and S/G level) are used in a verification or feedback sense rather than in an event diagnostic sense.
Consequently, information on the instrumentation channels and the level of redundancy is not includ-ed.
k Figure 4 I I
J BREAK IDENTIFICATION CHART l
PZR Level Changing i
AND l
PZR Pressure Decreasing Rapidly ir S/G Uncontrolled YES RC-2A/28 NO Loss of-Heat.
Pressure Primary Extraction
<800 PSI Coolant i,
YES Containment NO YES /
inment NO Pressure Pressure Increasing Increasing i,
Condenser YES Off-Gas R NO Blowdown lad Monitors Alarming i,
i, i,
i, - -
Break Break LOCA S/G LOCA Inside Outside Inside Tube Outside Containment Containment Containment Rupture Containment i,
i, i,
E0P-05 E0P-03 E0P-04 E0P-03 Uncontrolled Loss of S/G Loss of Heat Coolant Tube Coolant Extraction Accident Rupture Accident I$ SUED E0P-04 SEP 171986 Page 15 of 24 R1 09-17-86 FC/E0P/01
l Ouestion 2 Identify the instrumentation uncertainties for both normal and adverse containment conditions. Describe the basis for the selection of the adverse containment para-meters. Address, as appropriate, local conditions such as fluid jets or pipe whip which might influence the instrumentation reliability.
Response
Instrumentation uncertainties have been calculated based on adverse conditions in the containment.
Local conditions such as fluid jets and pipe whip have not been factored into this evaluation since the level of instrument redundancy and the di-verse physical locations of the various instrument channels provides control room operators with adequate information to diagnose events and to trip RCPs as appro-priate.
l The adverse containment temperatures considered were conservatively based on the containment thermodynamic response to small break LOCAs up to the times when oper-ator actions are directed in the Fort Calhoun E0Ps. Also, generic radiation dose calculations have shown radiation total integrated doses (TID) to be much lower during a small break LOCA without core damage than the TID used in determining the instrument errors for very harsh environments. The errors for temperature and radiation effects during a small break LOCA were used to conservatively calculate the total transmitter channel instrument errors.
For Fort Calhoun Unit No. I the conservative transmitter channel errors corresponding to the small break LOCA temperature and radiation effects are listed below:
Pressurizer pressure:
114.68 PSIA Steam generator pressure:
62.00 PSIA Although numerical errors were not evaluated for parameters that are used in a qualitative sense in the event diagnosis and RCP trip process, the parameters have been evaluated in terms of plant transient behavior and the E0P steps which are based on these parameters to confirm their applicability.
These parameters are:
pressurizer level, containment pressure, and secondary radiation.
Numerical
-(
errors were also not evaluated for parameters that are used in a verification or feedback sense. These parameters are RCS temperature, RCS subcooling, containment radiation, and steam generator level. The calculated errors referenced above have been evaluated in terms of plant transient behavior and the E0P steps which are based on these parameters.
The evaluation has shown that there is sufficient margin to plant safety when the errors are added to or subtracted from the plant data and then compared to the E0P directed operator action values for implementing the RCP trip strategy.
Question 3 4
In addressing the selection of the criterion, consideration of uncertainties asso-ciated with the CE0G supplied analyses values must be provided. These uncertain-i ties include both uncertainties in the computer program results and uncertainties 1
resulting from plant specific features not representative of the CE0G generic data group.
Response
The CE0G supplied analyses of CEN-268 established a thermally conservative value of RCS pressure to use in tripping the first two RCPs. The value thus established for Fort Calhoun Station was 1210 psia.
This value does not address instruments-I tion errors. Omaha Public Power District has implemented this value procedurally I
by using 1350 psia in the E0Ps; plant operators are directed to manually trip two RCPs whenever RCS pressure decreases in an uncontrolled manner to 1350 psia. The allowance made for instrument error is thus established as 1350-1210 - 140 psia.
Calculations corresponding to small break LOCA conditions of the total channel instrument error for wide range pressurizer pressure have shown an error of 114.68 psia. Hence, the setpoint contains an additional margin of 25.32 psia above that required to support the instrument errors. Additionally, it should be mentioned that 1350 psia is below the SIAS setpoint of 1600 psia, so that SIAS can be used by operators as a warning that the first two RCPs may have to be tripped.
Additional allowances to accommodate errors associated with the CE0G supplied value of 1210 psia are not required. The fundamental reason for this is that the value of 1210 psia is not derived from the transient computer analyses, but from a conservative quasi-steady-state energy balance between the primary plant and the steam generators as the so-called RCS pressure plateau is established.
In es-sence, the calculations establish primary pressure relative to secondary pressure such that decay heat removal is established. The accuracy of this calculation is 10 psia with the 1210 psia value being a rounded-up value. Additionally, less conservative calculations have shown that the inherent conservatism in the 1210 psia value is well over 100 psia. Accordingly, further margin to accommodate the data from the CEOG results is not needed and has not been implemented in the Fort Calhoun E0Ps.
Question 4 Identify all plant procedures (except for those concerning normal operations such as normal cooldown) which require RCP trip guidelines.
Reference to the CE0G EPGs is acceptable if endorsed by the licensee.
Include training and procedures which provide direction for use of individual steam generators with and without operat-ing RCPs.
Response
i The RCP trip strategy is implemented in the Fort Calhoun E0Ps as described below.
The implementation strategy cc.mplies with the CE0G generic Emergency Procedure Guidelines, CEN-152, Revision 02. The RCP trip strategy for each E0P is summar-ized below:
E0P Number E0P Name RCP Trio Strateov E0P-01 Reactor Trip Trip two RCPs if RCS pressure 1 1350 psia following SIAS.
E0P-02 Electrical This E0P is directed at situations Emergency where offsite power is not available.
Consequently, specific instructions to trip the RCPs are not required in this E0P.
E0P-03 Loss of Coolant This procedure trips all RCPs when Accident pressure is 11350 psia following SIAS.
E0P-04 Steam Generator Trip two RCPs if RCS pressure 11350 Tube Rupture psia following SIAS.
E0P-05 Uncontrolled Trip two RCPs if RCS pressure i 1350 Heat Extraction psia following SIAS.
E0P-06 Loss of All This procedure first trips two RCPs in Feedwater order to provide forced flow and to decrease the energy input to the RCS.
If feedwater cannot be restored, then all RCPs are tripped to further de-crease the energy input to the RCS.
None of the RCP related steps in this E0P are related to the trip logic asso-ciated with RCS depressurization events; in fact, a loss of all feed-water will eventually pressurize the RCS for this event prior to operator i
actions to initiate once through cool-ing.
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l E0P-20 Functional This procedure trips all RCPs if pres-Recovery surizer pressure s 1350 psia following SIAS. This provides adequate core cool-ing in the case of an undiagnosed LOCA.