ML20117K183
| ML20117K183 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 06/04/1996 |
| From: | Link B WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20117K186 | List: |
| References | |
| VPNPD-96-036, VPNPD-96-36, NUDOCS 9606100090 | |
| Download: ML20117K183 (8) | |
Text
l r .
.. : Wisconsin
! Electnc POWER COMPANY l 231 W M.chigan. PO Box 2046. Mdwoukee. WI 53201 2046 (414)221 2345 VPNPD-96-036 10 CFR 50.4 1 10 CFR 50.90 June 4,1996 Document Control Desk US NUCLEAR REGULATORY COMMISSION Mail Station PI-137 Washington, DC 20555 Gentlemen:
DOCKETS 50-266 AND 50-301 TECHNICAL SPECIFICATIONS CHANGE REOUEST 189 TECHNICAL SPECIFICATION MODIFICATIONS TO MAINTAIN SAFETY MARGIN FOR UNIT 2 WITH REPLACEMENT STEAM GENERATORS POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 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 Sections 15.2.1, " Safety Limit, Reactor Core," 15.2.3, " Limiting Safety System Settings, Protective Instmmentation," and 15.3.1.G, " Operational Limitations," to maintain safety margin for Unit 2 with replacement steam generators. We are scheduled to replace the steam generators in Unit 2 during the Fall,1996 refueling outage.
Marked-up Technical Specifications pages, a safety evaluation, and the no significant hazards consideration are encidsed.
DESCRIPTION OF CURRENT LICENSE CONDITION
- 1. TS Section 15.2.1, " Safety Limit, Reactor Core," Specification I states that the combination of thermal power level, coolant pressure, and coolant temperature shall not exceed the limits shown on Figures 15.2.1-1 and 15.2.1-2, " Reactor Core Safety Limits," for Units I and 2 respectively.
Figures 15.2.1-1 and 15.2.1-2 represent the loci of points of thermal power, reactor coolant system pressure, and average temperature for which the calculated departure from nucleate boiling ratio (DNBR) is no less than the design limit DNBR or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.
I 2. TS Section 15.2.3, " Limiting Safety System Settings, Protective Instrumentation," Specification 1.B(4) l gives the Overtemperature Delta T (OTAT) reactor trip setpoint function and parameter values.
Specification 1.B(5) gives the Overpower Delta T (OPAT) reactor trip setpoint function and parameter
! values.
9606100090 960604 M^d PDR ADOCK 05000266 P PDR hUV A suwayofHinasur Enq Oxpvakt f(b
Document Control Desk June 4,1996 l Page 2 1
l' i l ;
- 3. TS Section 15.3.1.G, " Operational Limitations," gives the departure from nucleate boiling (DNB) parameters l which must be observed during rated power operation. i 1 :
! DESCRIPTION OF PROPOSED CHANGES l
- 1. The safety analyses for operation of Unit 2 with the replacement steam generators were performed to encompass operation of Unit 1. As such, existing Figures 15.2.1-1 and 15.2.1-2, " Reactor Core Safety Limits," for Units 1 and 2 respectively, will be combined into one figure. This new figure will be numbered Figure 15.2.1-1. l liowever, in order to eliminate the possibility of an inadvertent reactor trip while adjusting the OTAT/OPAT setpoints, we desire to make the adjustments during the next scheduled refueling entage for each unit. As such, these proposed setpoints will be implemented for Unit 2 during the Fall,1996 refueling outage (U2R22) l and for Unit I during the Spring,1997 refueling outage (U1R24). If approved, this TSCR will be implemented during U2R22 and the changes to the Unit 2 setpoints can be made immediately. From the time this TSCR is implemented until UIR24, however, Unit I will be operating with the existing setpoints. Hence, existing Figure 15.2.1-1 will continue to apply to Unit 1. Therefore, existing Figure 15.2.1-1 will be renumbered to Figure 15.2.1-2 and the following footnote will be added to the bottom of page 15.2.1-1 to clarify the situation:
l
"* Figure 15.2.1-1 applies to Unit 2 following U2R22 and to Unit I following UlR24. Prior to UlR24, l Figure 15.2.1-2 applies to Unit 1."
The following footnote will be added to the bottom of renumbered Figure 15.2.1-2:
" This figure applies to Unit 1 prior to UlR24. Following UlR24, Figure 15.2.1-1 applies to Unit 1."
Existing Figure 15.2.1-2 will be deleted.
The following footnote will be added to the bottom of new Figure 15.2.1-1:
"* This figure applies to Unit 2 following U2R22 and to Unit I following UIR24. Prior to UlR24, Figure 15.2.1-2 applies to Unit 1."
- 2. Certain inputs to the OTAT/OPAT setpoint functions will be revised to maintain safety margin for Unit 2 with the replacement steam generators and to provide for operation at either 2000 psia or 2250 psia primary pressure. The revised setpoint functions will also apply to operation of Unit 1. Also, a typographical error in the OTAT setpoint function will be corrected.
l
. 1 i
e c
l .
- 1. .
l Document Control Desk l l June 4,1996 Page 3 l
As mentioned above, we desire to make adjustments to the OTAT/OPAT setpoint functions during the next scheduled refueling outage for each unit. Since this TSCR will be implemented during U2R22, the changes to the Unit 2 setpoints can be made immediately. From the time this TSCR is implemented until UIR24, l however, Unit I will be operating with the existing setpoints. Therefore, the following footnote will be added ;
to the bottom of page 15.2.3-2 to clarify the situation:
"* These values apply to Unit 2 following U2R22 and to Unit I following UIR24. Prior to UIR24, the values for Unit I are: T's 573.9 F, P' = 2235 psig K si 1.30, K =2 0.0200, and K3 = 0.000791.'"
For the same reasons, the following footnote will be added to the bottom of page 15.2.3-3:
"* These values apply to Unit 2 following U2R22 and to Unit I following UlR24. Prior to UlR24, the values for Unit I are: T's 573.9 F, and K4 s 1.09 of rated power." i
- 3. References to Figures 15.2.1-1 and 15.2.1-2 in the Bases for TS Section 15.2.3 will be modified to delete the association of these figures with a particular unit.
- 4. Revisions to the DNB related parameters in TS Section 15.3.1.G will be made to maintain safety margin for Unit 2 with the replacement steam generators. The revisions will also apply to operation of Unit 1. He
' Bases for this section is being revised to support the proposed revision to the raw measured totrd flow rate limit.
Again, sicce Unit I will be operating with the existing parameters from the time this TSCR is implemented i until UlR24, the following footnote is being added to the bottom of page 15.3.1-19:
These values apply to Unit 2 following U2R22 and to Unit I following UlR24. Prior to UIR24, T., for i Unit I shall be maintained below 578 F."
This footnote does not apply to the changes to reactor coolant system pressure or raw measured total flow 4 rate since these changes maintain the existing Unit I values. l
)'
BASIS A'ND JUSTIFICATION We are planning to replace the steam generators in PBNP Unit 2 during the Fall,1996 refueling outage, j Westinghouse performed the safety analyses and evaluations for operation of Unit 2 with the replacement steam !
generators. The analyses and evaluations were performed to support a full-power reactor :oolant system average coolant temperature (T.3) window between 557 F and 573.9 F. The analyses and evaluations also support
, operation at either 2000 psia or 2250 psia. In order to maintain safety margins under these new constraints, l certain operating limits must be revised. The proposed changes maintain the margin to safe operation of Unit 2 with the replacement steam generators. The safety analyses and evaluations were performed to cover both units.
Thus, the proposed changes will apply to the operation of Unit I as well as Unit 2.
l
i l . .
i Document Control Desk June 4,1996 Page 4 The proposed changes fall within the scope ofitems that can be relocated to a Core Operating Limits Report (COLR). In anticipation of these changes, we submitted TSCR 185, " Core Operating Limits Report," on l December 13,1995. The purpose of this submittal was to implement a COLR at PBNP with a requested NRC l
approval date of March,29,1996. Our intent was to implement the COLR well in advance of the Fall,1996 outage and make the changes required to maintain safety margin for Unit 2 with the replacement steam generators in accordance with our 10CFR50.59 safety evaluation process.
However, in a telephone conference with Mr. Allen Hansen and Ms. Margaret Chatterton of your staff on January 24,1996, we were informed that the March 29,1996 approval date could not be met and agreed on an approval date of June 30,1996. In a telephone conference with Mr. Allen Hansen on March 13,1996, we were infonned that the June 30,1996 approval date could not be met and that review of TSCR 185 may not begin until l September,1996.
l The Fall,1996 refueling outage is currently scheduled from October 5 to December 15,1996. We are not confident that TSCR 185 will be approved in time for us to implement the COLR prior to the startup of Unit 2 in December. Therefore, we are submitting this license amendment request as a contingency to incorporate the changes required to regain operating margin for Unit 2 with the replacement steam generators. We request approval of these changes by September 1,1996. If TSCR 185 is approved prior to September 1,1996, we will withdraw this license amendment % . st. ;
We have determined that the proposed amendments do 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 l individual or cumulative occupational exposure. Therefore, we conclude that the proposed amendments meet the ;
l requirements of 10 CFR 51.22(c)(9) and that an environmental impact statement or negative declaration and I i environmental impact appraisal need not be prepared. 1 i
If you require additional information, please contact us.
Sincerely, l
- Bob Link
- Vice President l Nuclear Power KVA Attaciunents J
cc: NRC Resident inspector NRC Regional Administrator PSCW 1
l .
l t
TECHNICAL SPECIFICATIONS CHANGE REOUEST 189 SAFETY EVALUAT[QN INTRODUCTION Wisconsin Electric Power Company (Licensee) has applied for amendments to Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant Units 1 and 2. The proposed revisions will modify Technical Specification (TS) Sections 15.2.1," Safety Limit, Reactor Core," 15.2.3, " Limiting Safety System Settings, Protective Instrumentation," and 15.3.1.G, " Operational Limitations," to maintain safety margin for Unit 2 with replacement steam generators. We are scheduled to replace the steam generators in Unit 2 during the Fall, 1996 refueling outage.
EVALUATION We are planning to replace the steam generators in PBNP Unit 2 during the Fall,1996 refueling outage. Westinghouse perfonned the safety analyses and evaluations for operation of Unit 2 with the replacement steam generators including; large and small break LOCA, LOCA forces, non-LOCA events, LOCA long and short term mass and energy release, steam generator tube rupture, and systems and components. All analyses and evaluations were performed in accordance with NRC approved methodologies.
l The safety analyses and evaluations examined the operating conditions with the new steam ,
generators including an average coolant temperature (T,yg) window between 557 F and 573.9 F, ,
an operating pressure of either 2000 or 2250 psia, a reduced thermal design flow, a slightly larger i primary volume, and a slightly smaller secondary volume.
In order to maintain safety margins under these new constraints, certain operating limits must be revised. The proposed changes include a revised Reactor Core Safety Limits (RCSL) figure, revised inputs to the overtemperature delta T and overpower delta T (OTAT/OPAT) setpoint functions, implementation of an average coolant temperature (T,,,) operating band, and restoration of the reactor coolant system (RCS) raw measured flow rate limit to the Unit 1 values.
The safety analyses and evaluations reached the following conclusions:
- The changes associated with the replacement steam generators do not result in exceeding any LOCA design or regulatory limits.
. Non-LOCA acceptance criteria continue to be met.
- The containment design pressure limit will continue to be met.
- Steam generator tube rupture doses meet the acceptance criteria of a "small fraction" of 10 CFR 100 limits.
- j 1
\
l In order to maintain one set of safety analyses for both units, the analyses and evaluations for '
operation of Unit 2 with the replacement steam generators were performed to encompass the !
operation of Unit 1. Therefore, the proposed changes apply to the operation of both units and _I maintain the margin of safety for eact.
CONCLUSION The proposed changes were arrived at through safety analyses and evaluations performed in accordance with NRC approved methodologies. The proposed changes will ensure the safe operation of both Unit 2 with replacement steam generators and Unit I with its existing steam generators. Therefore, approval of these changes will ensure and enhance the continued safe operation of Point Beach Nuclear Plant.
i l
i l
TECIINICAL SPECIFICATION CIIANGE REOUEST 189 l NO SIGNIFICANT II AZARDS CONSIDERATION l 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 requirements of 10 CFR 50.92 and the basis for this conclusion are as follows:
l I
- 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.
The proposed change does not involve a change to structures, systems, or components which would affect the probability or consequences of an accident previously evaluated in the PBNP Final Safety Analyses Report (FSAR). The proposed changes maintain the margin to safe operation for Unit 2 with the replacement steam generators. In order to maintain one set of safety analyses for both units, the analyses for operation of Unit 2 with I the replacement steam generators were performed to encompass the operation of Unit 1. )
Therefore, the proposed changes app' M the operr. tion of both units and maintain the l margin of safety for each. The propos.<1 changes do not change, degrade, or preclude the i prevention or mitigation of the consequences of any accident described in the FSAR.
Therefore, the probability or consequences of a previously evaluated accident are not significantly increased as a result of these changes.
- 2. Operation of this facility under the proposed Technical Specifications change will not create ti a pssibility of a new or different kind of accident from any accident previously evaluated. l 1
The proposed changes do not involve a change to the plant design. The proposed changes maintain the margin to safe operation for Unit 2 with the replacement steam generators.
In order to maintain one set of safety analyses for both units, the analyses for operation of Unit 2 with the replacement steam generators were perfomied to encompass the operation of Unit 1. Therefore, the proposed changes apply to the operation of both units and maintain the margin of safety for each. These changes do not affect any of the parameters or conditions that contribute to initiation of any accidents. In addition, the safety functions of safety-related systems and components, which are related to accident mitigation, have not been altered. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.
l l
e 3. Operation of this facility under the proposed Technical Specifications change will not create a significant reduction in a margin of safety.
The proposed changes maintain the margin to safe operation for Unit 2 with the replacement steam generators. In order to maintain one set of safety analyses for both units, the analyses for operation of Unit 2 with the replacement steam generators were performed to encompass the operation of Unit 1. Therefore, the proposed changes apply to the operation of both units and maintain the margin of safety for each. The proposed changes have no affect on the availability, operabiliiy, or performance of the safety-related systems and components described in the Technical Specifications. Therefore, the proposed changes will not create a significant reduction in a margin of safety.
- _ -__ _- _