ML20092H007
| ML20092H007 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 09/13/1995 |
| From: | Link B WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| VPNPD-95-072, VPNPD-95-72, NUDOCS 9509200198 | |
| Download: ML20092H007 (5) | |
Text
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'dc Wisconsin (Electnc POWER CO. PANY M
231 W Michigan Po Box 2046. Milwaukee.WI 53201-2046
.-(414)221-2345 i
VPNPD-95-072 10 CFR 50.4 10 CFR 50.90 September 13, 1995 Document Control Desk U.S.-NUCLEAR REGULATORY COMMISSION Mail' Station P1-137 i
- Washington,-DC 20555-
' Gentlemen:.
DOCKETS-50-266 AND'50-301 TECHNICAL SPECIFICATIONS CHANGE REOUEST 181 MODIFICATION TO TS 15.3.1.G.3
-REACTOR COOLANT SYSTEM RAW MEASURED TOTAL FLOW RATE POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 t
In accordance with the requirements of 10 CFR 50.4 and 50.90, Wisconsin Electric Power Company (Licensee) hereby requests amendments to Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant (PBNP) Units 1 and 2, respectively,
'to-incorporate changes to the plant Technical Specifications.
The proposed revisions will modify Technical Specification Section 15.3.1.G, " Operational Limitations," Specification 3.b, to reduce the reactor coolant system'(RCS) raw measured total flow rate limit by 4000 gallons per minute (gpm) for Unit 2.
The annual refueling and maintenance outage for PBNP Unit 2 is scheduled to begin.on October.7, 1995. : Eddy current testing of the steam generator U-tubes is: scheduled to begin on October 17, 1995.
LBased on the eddy current test results from the 1994 refueling outage and industry experience since that time, it is possible that the number of tubes that will have to be removed from service during the upcoming outage could result in the need for a reduction of the current Technical Specification RCS flow rate limit.
We will not know the exact amount of' reduction until the eddy current l
testing is completed.
.We have determined that a 4000 gpm reduction in the RCS raw measured total flow. rate limit will bound any reasonably expected increase in the level of steam generator tube plugging.
We-are submitting this. request at-this time to avoid an exigent condition in the future.
A no significant hazards consideration applicable to a 4000 gpm RCS' flow reduction and related changes is. enclosed.
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Document Control' Desk-7 Septe*ber-13, 1995_.
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. Additional changes to the Technical: Specifications may be required.
i toisupport the reduction in'the RCS flow rate limit including:
Lovertemperature and overpower Delta T; protection, reactor. core' i
safety limits, and rated power.. With the exception'of a change to-1
~
' rated power,'these changes will be similar to those requested in j
.j Technical. Specification Change Request 177'and' approved by.the NRC 1
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_as ' Amendments'- 156/160,- dated October; 28,_1994.-
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Evaluations ~are'being performed to account for the increased plugging levels and the resultant reduction in'RCS flow.
These evaluations are being performed ~for a 30% plugging level in the j
1 tsteam generators. "We will, submit the results of the evaluations j
'and copies:of:the marked-up pages for required Technical 1
Specification changes when they are available.
DESCRIPTION OF CURRENT LICENSE CONDITION i
Specification-15.3.1.G, " Operational Limitations," specifies thesRCS operational limitations'for DNB (Departure from Nucleate Boiling) related parameters.
Specification 15.3.1.G.3' presently
~,
specifies that reactor coolant system raw measured total flow rate j
.must be 2181,800 gpm for Unit 1, and 2174,000 gpm for Unit 2.
Specification 15.2.1, " Safety Limit, Reactor Core," specifies l
the reactor core safety limits that are used to maintain the integrity of the fuel' cladding.
Tne specification states that the E
combination of thermal power level, coolant pressure, and coolant i
temperature shall not' exceed the limits shown on Figure 15.2.1-1 for Unit 1 and on Figure 15.2.1-2 for Unit 2.
i Technical Specification Section 15.2.3, Limiting Safety System Settings, Protective Instrumentation," Specification 15.2.3.1.B(4) is the Overtemperature AT core limit protection setpoint function.
Specification 15.2.3.1.B(5) is the Overpower AT core limit protection setpoint function.
These functions provide setpoints that prevent exceeding the reactor core safety limits shown in Figures 15.2.1-1 and 2..
DESCRIPTION OF PROPOSED CHANGES i
This Technical. Specification Change Request (TSCR) proposes to revise Specification 15.3.1.G.3 as~follows:
i
" 3 '. / Reactor Coolant System raw measured Total Flow Rate i.
(See Basis):
a.
Unit 1.2181,800 gpm b.
Unit 2'2170,000-gpm"
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-Document Control Desk i
Septabber.13, 1995 l
Page 3 4
4 The associated basis will also be changed ~to reflect the' revision' to TS 15.3.1.G.3.
i Changes to Technical Specification 15.2.3.1.B(4), "Overtemperature Delta ~T," Specification 15.2.3.1.B(5), " Overpower Delta T,"
3 j
. Specification 15.1.j, and Figure 15.2.1-2, " Reactor Core Safety Limits, Point Beach. Unit 2," may also be required.
BASIS AND JUSTIFICATION The'4000 gpm reduction in the RCS raw measured total flow rate'
~ limit for Unit 2 will be verified acceptable based on evaluations Lthat will be performed by Westinghouse and reviewed by Wisconsin Electric.
The results of these evaluations will be submitted in
-support of the necessary Technical Specification changes when they' are available.
It is believed that the proposed amendments will not involve a significant hazards consideration, authorize a significant change in the types-or total amounts.of any effluent release, or result in any significant~ increase in individual cnr cumulative occupational exposure.
Therefore, we conclude that the proposed amendments meet
. the requirements of 10 CFR 51.22(c) (9) and that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared.
i Please contact us if there are any questions.
J I
Sincerely,
/W
' L Bob Link 4
Vice President 7
Nuclear Power
!~
KVA/jg Attachment i
cc:
NRC Resident Inspector.
1 NRC Regional' Administrator i
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l TECHNICAL BPECIFICATION CHANGE REOUEST 181 i
l NO SIGNIFICANT HAZARDS CONSIDERATION In accordance with the requirements of 10 CFR 50.91(a), Wisconsin Electric Power Company (Licensee) has evaluated the proposed changes against the standards of 10 CFR 50.92 and has determined that the operation of Point Beach Nuclear Plant, Units 1 and 2, in accordance with the proposed amendments does not present a significant hazards consideration.
The analysis of the l
requirements of 10 CFR 50.92 and the basis for this conclusion are as follows:
1.
Operation of this facility under the proposed Technical Specifications will not create a significant increase in the probability or consequences of an accident previously evaluated.
This proposed change reduces the Unit 2 Reactor Coolant System raw measured total flow rate limit by 4000 gpm.
Related changes to Overtemperature and Overpower Delta T protection, reactor core cafety limits, and rated power may also be required.
Evaluations to be performed by Westinghouse and Wisconsin Electric will verify that all safety analysis requirements will still be met at the reduced flow rate limit without exceeding acceptable limits.
A reduction of the RCS flow limit does not affect any parameters that could affect the probability of an accident.
Therefore, there is no expected increase in the probability or consequences of an accident previously evaluated.
2.
Operation of this facility under the proposed Technical Specifications change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
This proposed change reduces the Unit 2 Reactor Coolant System raw measured total flow rate limit by 4000 gpm.
Related changes to Overtemperature and Overpower Delta T protection, reactor core safety limits, and rated power may also be required.
Evaluations to be performed by Westinghouse and Wisconsin Electric will verify that all safety analysis requirements are still met at the reduced flow rate limit.
There is no physical change to the facility, its systemc, or its operation.
Thus, a new or different kind of accident cannot occur.
3.
Operation of this facility under the proposed Technical Specifications change will not create a significant reduction in a margin of safety.
This proposed change reduces the Unit 2 Reactor Coolant System raw measured total flow rate limit by 4000 gpm.
Evaluations to be performed by Westinghouse and Wisconsin Electric will verify that all acceptable safety analysis requirements are met and margins of safety are maintained at the reduced flow rate limit.
If necessary, modifications
to'the overtemperature and Overpower Delta T function,
- core safety limits, and/or rated power for PBNP Unit 2 will be made to maintain acceptable margins and prevent the possibility of exceeding the core safety limits.
Therefore, this reduction in RCS total flow rate limit does not reduce any existing margin of safety.
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