ML20236B325
| ML20236B325 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 03/10/1989 |
| From: | Fay C WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| CON-NRC-89-033, CON-NRC-89-33 VPNPD-89-144, NUDOCS 8903210022 | |
| Download: ML20236B325 (3) | |
Text
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Electnc.
POWER COMPAth',
l-231 W Mchigan, PO Box 2046, Mitwoukee,WI 53201 '
(414)221 2345
-VPNPD-89-144 l
. March'10, 1989
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q Document Control Desk U.S.
NUCLEAR REGULATORY COMMISSION i
Mail Station Pl-137
-Washington, D.C.
20555 Gentlemen:
DOCKETS 50-266'AND 50-301
' STEAM GENERATOR TUBE RUPTURE ANALYSIS INFORMATION TECHNICAL SPECIFICATION CHANGE REQUEST 127 INCREASED ALLOWABLE CORE POWER PEAKING FACTORS j
POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 j
l The steam generator tube rupture (SGTR) analysis was submitted as a part of.our Technical Specification Change Request No. 127 to increase the allowable core power peaking factors at Point Beach Nuclear Plant (PBNP).
Mr. Warren H.
Swenson's letter to us dated February 22, 1989, requested that Wisconsin Electric provide justification for the.30 minute safety injection (SI) termination time used as an input assumption in our SGTR analysis.
The request for additional information was based on the assumption that the SI terminationftime was not a part of our original licensing basis.
This letter responds to that request.
i The 30 minute assumption is not a change from our current Final Safety Analysis Report (FSAR) analysis.
Onlyethe number of SI pumps running during this time period is changed to be consistent er9 with the current Emergency Operating Procedures at PBNP.
Section 14.2.4.4 of the FSAR states that the pressure between the ruptured 1 steam generator and the primary will be equalized within 30 o
minutes for the current SGTR analysis.
SI termination is included SQ@
in the 30 minutes required to equalize pressure.
The major mo difference between our current analysis and the analysis performed
$8 to encompass the peaking factor changes (other than the peaking GO factor increase itself) lies in the assumption regarding SI pump
+ct operation.
While the current analysis assumes that one SI pump
$8 will be shut down by the operators early in the transient, leaving 13 @
one pump' running for the first 30 minutes, the reanalysis assumes N
that both SI pumps will run until 30 minutes.
This change in the
.$x assumption regarding the number of SI pumps running reflects a I
yf@g. change in our SGTR recovery procedures, which are now based on the Westinghouse Owners Group Emergency Response Guidelines.
8 da m_
Document Control Desk March 10, 1989 Page 2 The 30 minute assumption is not changed in the increased peaking factor reanalysis because we believe that it is an appropriate time period.
The four main actions in a SGTR response are identification and isolation of the affected steam generator, cooldown of the RCS, depressurization of the RCS, and SI termi-nation.
The training department at Point Beach Nuclear Plant (PBNP) has estimated the time required to complete each of these actions based on observed operator response during training at the Kewaunee Nuclear Plant simulator (a PBNP-specific simulator is not yet operational).
The time required to identify and isolate a steam generator containing a tube break of the size assumed in the analysis is estimated to be 15 to 20 minutes.
Cooldown of the reactor is then initiated at the maxinum achievable cooldown rate.
Reducing the core exit temperature by the approximately 50 F required by procedure typically takes 10 minutes or less.
Since the cooldown and the break flow will also depressurize the primary l
system, any additional required depressurization to equalize l
primary and ruptured SG pressure can be completed within about one minute.
Therefore, for the tube rupture assumed in the analysis, the operators require approximately 30 minutes to reach the point at which SI can be terminated.
Emergency Operating Procedure (EOP) 3.0, Steam Generator Tube Rupture, contains a caution which instructs the operators to secure SI flow once the SI termination criteria are satisfied, so that the faulted steam generator will not be overfilled.
The EOP, however, currently instructs the operators to continue SI flow for 11 minutes to flush the injection lines with water from the refueling water storage tank (RWST) to remove the high concen-tration boric acid taken from the boric acid storage tank (BAST).
During a design basis tube rupture, SI suction will be transferred early in the transient from the BAST to the RWST, which contains low concentration boric acid.
This precludes the need to sepa-rately flush the lines, as they will be flushed as a part of the normal injection.
Since the separate SI flush would be unnecessary in the event of a design-basis SGTR and delay in SI termination is undesirable from SG overfill and radiological release considerations, EOP 3.0 will be revised to require a determination of the need for flushing based on RWST level.
The 11 minute flush will be required only when an insufficient volume of RWST water has not been injected into the RCS.
When the volume of water injected from the RWST is greater then the volume required to flush the SI system, no separate flush will be required.
In a design basis tube rupture, sufficient RWST water volume will be injected into the RCS such that a flush will not be required and SI flow will be terminated as soon as the SI termination criteria are met.
p-A s
Document Control Desk March 10, 1989 Page 3 As discussed above, we expect the SI termination criteria to be met within approximately 30 minutes.
Since the operators will be instructed in the reviccd EOP to secure SI flow without a flush once the termination criterla are satisfied, the SI flow is expected to be secured within about 30 minutes as assumed in both the current and new SGTR analyses.
The proposed revision to EOP 3.0 will be implemented prior to the implementation of the requested Technical Specification change.
Therefore, we believe the 30 minute SI termination time used in the analysis of the steam generator tube rupture event is appropriate.
Please contact us should you have any questions regarding this information.
Very truly yours,
'v0
-w C.
W.
ay Vice President Nuclear Power Copies to NRC Regional Administrator, Region III NRC Resident Inspector
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