ML20059F508
| ML20059F508 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 10/27/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20059F491 | List: |
| References | |
| NUDOCS 9311040332 | |
| Download: ML20059F508 (9) | |
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UNITED STATES g3. 13 l., j NUCLEAR REGULATORY COMMISSION
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WASHlhGToN, D.C. 2065tWXM1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS.142 AND 146 TO I
f8CILITY OPERATING LICENSE N05. DPR-24 AND DPR-27 l
WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT. UNIT NOS. 1 AND 2 l
f DOCKET NOS. 50-266 AND 50-301
.l l
1.0 INTRODUCTION
By letter dated June 11,1993 (Reference 1), the Wisconsin Electric Power i
Company (WEPCO), the licensee, pursuant to 10 CFR 50.90, requested an j
amendment to Facility Operating Licenses DPR-24 and DPR-27 for the Point Beach Nuclear Plant (PBNP), Units 1 and 2, respectively. The amendment proposed modification to Technical Specifications (TS), Sections 15.3.1.G.3 and 15.2.3.1.B(4) and (5). The proposed TS revision will reduce the Reactor Coolant System (RCS) total flow rate by 2600 gallons per minute (gpm), and change the T' temperature limit associated with the overtemperature and overpower delta-T setpoint functions, from 573.9 'F to 570 'F in Unit 2.
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As part of the justification to support the decrease in RCS flow rate limit, l
the licensee's submittal included a reference to a justification for continued operation (JCO) regarding the structural integrity of systems and components for operation of Point Beach at a reduced RCS T of 570 'F.
The licensee concluded in their submittal that the reduced RET flow rate and reduced T' setpoint as they relate to changes in primary coolant system temperatures do not significantly affect the structural integrity of the reactor coolant pressure boundary.
In response to questions by the NRC, the JC0 was provided by the licensee.
Conference calls, between WEPCO, Westinghouse and NRC, were held on October 6 and October 7, 1993, regarding questions on the JCO. On October 15, 1993, a meeting was held at the NRC headquarters, where WEPC0 and Westinghouse presented their response to the staff's questions in a draft revision of the JCO. The details of the meeting are documented in a meeting summary dated October 21, 1993. By letter dated October 19, 1993 (Reference 2), WEPC0 submitted clarifying information regarding the proposed amendment, including a copy of the revised JCO.
9311040332 931027
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PDR ADDCK 05000266 P
2.0 BACKGROUND
2.1 Current License Condition The current license condition as stated in the TSs is applicable for both Units 1 and 2 as follows:
(1)
TS 15.2.2, " Safety Limits, Reactor Core" specifies the reactor core safety limits that are used to maintain the integrity of the fuel cladding. The specification states that the combination of thermal power level, coolant pressure, and coolant temperature shall not exceed the limits in Figure 15.2.1-1.
(2)
TS 15.2.3.1.B(4), Section 15.2.3, " Limiting Safety System Settings, Protective Instrumentation," is the overtemperature AT core limit protection setpoint function. TS 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 Figure 15.2.1-1.
(3)
TS 15.3.1.G, " Operational Limitations," specifies the RCS operational limitations for DNB (Departure from Nucleate Boiling)-
related parameters. TS 15.3.1.G.3 specifies that reactor coolant system raw measured total flow rate must be equal to or greater than 181,000 gpm.
2.2 Proposed Changes The proposed change to reduce the RCS measured total flow rate by 2,600 gpm (1.4 %) for Unit 2 requires a change to the Reactor Core Safety Limits graph which in turn causes the Overtemperature and Overpower AT setpoints to be changed. The licensee proposed TS changes which will revise the reactor core safety limits figure, overtemperature AT setpoint, overpower AT setpoint, and the minimum RCS flow rate for Unit 2.
(1)
A new figure is being added to TS 15.2.1, " Safety Limit, Reactor Core," which is applicable to Unit 2.
The title of the existing figure is being modified to indicate it is applicable to only PBNP, Unit 1.
(2)
TS 15.2.3.1.B(4) and (5) is being modified as follows:
T' s
573.9 *F (Unit 1)
T' s
570.0 *F (Unit 2)
(3)
T5 Section 15.3.1.G, " Operational Limitations," is being modified to provide Reactor Coolant System flow limits specific to each j
unit as follows:
Reactor Coolant System raw measured Total Flow Rate:
a.
2 181,800 gpm (Unit 1) b.
2 179,200 gpm (Unit 2) i
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1 1
3-w 3.0 EVALUATION 3.1 TRANSIENT ANALYSIS:
i The licensee proposed to reduce the RCS measured total flow rate limit for Unit 2 by 2,600 gpm. The reduction in flow rate was evaluated by Westinghouse. The evaluation covered the (1) Non Loss of Coolant Accident (Non-LOCA) transient analyses, (2) Loss of Coolant Accident (LOCA) analysis, and (3) Steam Generator Tube Rupture (SGTR) analysis.
3.1.1 Non-LOCA Transient Analyses Evaluation The analyses by Westinghouse used NRC-approved methodologies and included Departure from Nucleate Boiling Ratio (DNBR) design margin.
One percent of the DNBR design margin was allocated to offset the reduction in DNBR margin that would occur at the reduced RCS flow rates.
This one percent reduction in DNBR design margin justifies up to a 2,600 gpm reduction in the RCS total flow rate limit. This reduction in RCS flow limit required a change to the Reactor Core Safety Limits for Unit 2.
The change to the Reactor Core Safety Limits required a change to the Overtemperature and Overpower AT setpoints for Unit 2.
The T' term of these setpoint functions were reduced from 573.9 'F to 570 'F.
The setpoint change is based on WCAP-8745-A (Ref. 3). The reduction of these setpoints provides the appropriate protection against DNB in the core for all of the licensing basis accidents described in the FSAR for Point Beach, Unit 2.
The FSAR Section 14.1.8, " Loss of Reactor Coolant Flow," transient analyses were reanalyzed for the lower RCS flow condition with acceptable DNB results.
An evaluation of the Point Beach FSAR non-LOCA accident analyses that contain non-DNB acceptance criteria was also performed. All acceptance criteria were found to be met with the lower RCS flow.
For the FSAR Section 14.1.8, " Loss of Reactor Coolant Flow," a normal operating pressure of 2,000 psia was used in the analysis to satisfy the RCS pressure limit criteria for this transient.
We have found that the results of the evaluation of the Non-LOCA transient analyses are acceptable as NRC-approved methodologies were used and the results were bounded by the FSAR acceptance criteria.
3.1.2 LOCA Evaluation 3.1.2.1 The Small Break LOCA is presented in Point Beach FSAR Section 14.3.1.
As part of the increased peaking factor change in a previous Unit 2 Amendment (No.123), a Small Break LOCA analysis
was made which supported a RCS flow rate limit as low as 174,000 gpm. This is less than the 179,200 gpm being proposed in the TS change request. Since the Small Break LOCA analysis is not affected by the reduction in the RCS flow limit, we find it to be acceptable.
3.1.2.2 The Large Break LOCA analysis is presented in Point Beach FSAR Section 14.3.2.
The evaluation for the Large Break LOCA was performed by Westinghouse. Their evaluation indicated that the approximately 1.5% reduction in RCS flow was well within the allowed variance for this parameter; and, because there was very little, if any, impact on the transient results, the large Break LOCA results do not change. Since compliance with 10 CFR 50.46 is maintained, we find the results of the Large Break LOCA analysis to be acceptable.
3.1.3 SGTR Evaluation The Steam Generator Tube Rupture analysis is described in the Point Beach FSAR Section 14.2.4.
This analysis is not affected by the RCS flow rate limit reduction because the flow rate used in the analysis is lower than the proposed RCS flow rate limit. The RCS pressure and temperature could also affect this analysis. The analysis is based on an RCS pressure of 2,250 psia and average temperature of 573.9 'F.
However, Unit 2 is operated at 2,000 psia and 570 *F.
The lower pressure would result in a slightly lower mass release and the lower temperature would result in a slightly higher mass release in this analysis. The analysis performed by Westinghouse determined that the pressure effect is greater than the temperature effect and that the off-site radiation doses for the FSAR Section 14.2.4 SGTR analysis remain applicable for Unit 2.
Therefore, we find the results of the SGTR analysis to be acceptable.
3.2 SYSTEM AND COMPONENT INTEGRITY EVALUATIONS:
i Additionally, the effects of reduced RCS flow were assessed for the system and component integrity evaluations.
In October 1992 Westinghouse provided an assessment to the licensee in the form of a JC0 regarding operation of Point Beach at a reduced RCS temperature, based on reduced T operation programs at "other" Westinghouse plants, and using engineering $!dgement to extrapolate those results to Point Beach. On October 19, 1993, the licensee submitted clarifying information, including a revision of the JC0 which provided plant specific information regarding reduced T operation of Point Beach, Units 1 y
and 2.
The October 19, 1993, submittal (Reference 2) indicated that both Units 1 and 2 have been operating at reduced T of 570 *F for the past 21 years. As such, the licensee evaluated the eEects of the reduced RCS average temperature on the structural and pressure boundary integrity of the piping systems, components, and their supports. Further, the licensee utilized the results of their Transient and Fatigue Cycle Monitoring Program from 1986 and 1987 to establish which components in the plant are most susceptible to fatigue over plant life. Components evaluated include the control rod drive mechanisms (CRDM), pressurizer, reactor vessel and internals, steam generators and reactor coolant pumps.
' 3.2.1 Reactor Vessel and Internals For the reactor vessel and internals, the original fatigue analyses were reviewed by Westinghouse, and the limiting components were identified to be the reactor vessel closure studs and the safety injection nozzles.
The maximum cumulative usage factor reported in the reactor design i
stress report is 0.79 for the safety injection nozzles on 40-years life of plant operation. The maximum stresses at critical locations were
)
also evaluated for the reduced T condition. Westinghouse indicated that the increase in stress was Usignificant (less than seven percent for the core barrel outlet nozzle). Considering the conservatism in the analysis, as stated by the licensee, that the actual measured fatigue cycle is about half of what was assumed in the fatigue design analyses, and the combination of stresses due to other loading conditions such as LOCA, seismic and pressure differential, we conclude that operation at the reduced T,y has negligible impact on the original stress and fatigue analyses of the reactor vessel and internals.
3.2.2 Reactor Coolant Loop Piping and Supports The Point Beach piping, including the primary loops and Class 1 auxiliary piping systems, was originally designed to the ASME/ ANSI B31.1 Power Piping Code.
Fatigue analyses were not required by the Code.
Westinghouse indicated that static LOCA loads were used for the original piping analyses and that these loads are not affected by the reduced T,y. Westinghouse also indicated that the analysis of the pressurizer surge line, performed in response to NRC Bulletin 88-11 regarding the thermal stratification issue, included the impact of the reduced T operation. Therefore, the original analyses for the piping, incluEng the pressurizer surge line, primary loops and Class 1 auxiliary systems, and pipe supports remain unchanged for the operation of the Units at a T,y of 570 *F.
3.2.3 Control Rod Drive Mechanisms The Point Beach Units used the same model of Control Rod Drive Mechanisms (CRDM) as one of the other Westinghouse plants that was operation. Westinghouse specifically analyzed for reduced Tno,d T"Th indicated that the Point Beach reduce is equivalent to a T reduction from 602 *F to about 598.9 *F.
e similar CRDMs of Ne "other" plant was analyzed for a T reduction from 610 *F to 595 *F, which is bounding for the Point Beac*h operation. Therefore, we conclude that the Point Beach CRDMs will not be affected at a reduced T " of 570 *F.
3.2.4 Pressurizer Westinghouse evaluated the Series 84 Point Beach pressurizer by comparing the design basis thermal analysis with the T,y of 570 *F operation. The evaluation consisted of reviewing the original stress analysis for the most limiting locations (the Spray Nozzle and the
Upper Head to Shell Junction). The licensee stated that the pressurizer spray nozzle was the most limiting component from a fatigue perspective, with the fatigue usage factor predicted to not exceed approximately 0.85 at the end of the current license. The review indicated that the design bases of stress and fatigue analyses for the Series 84 pressurizer were based on a delta-T of 125 *F which envelo)es the Point Beach operating condition with a delta-T of 112.4 *F at tie reduced T,WCS Toperation.
Based on our review, we conclude that operation at an of 570 'F will not have an adverse impact on the structural integrity,,of the pressurizer.
3.2.5 Steam Generator The Series 44 Steam Generators in the Point Beach Units were not specifically analyzed for Point Beach design paramsters, but had been previously analyzed by Westinghouse using design parameters consistent with other Westinghouse plants. The key design input parameters such as T,y and pressure differential at the interface of the primary and the secondary systems for the previously analyzed Series 44 Unit and the Point Beach Series 44 Unit, were summarized in the JCO. The comparison shows that the input conditions of the previous analysis are more severe than the operation of Point Beach at a T of 570 *F.
Based on the above review, Westinghouse determined th'd stresses and fatigue usage factors for the Point Beach Units are within the Code allowable limits.
Based on our review, we concur with Westinghouse's conclusion that the steam generators are acceptable for operation at the reduced T, of 570 *F.
3.2.6 Reactor Coolant Pumps Westinghouse evaluated the adequacy of the Reactor Coolant Pumps by comparing the operating parameters of Point Beach, Units 1 and 2 at the reduced T,'93 RCP for the other Westinghouse plant.of 570
- F, with those used in the design analysis of the sameModel The comparison shows that the delta-T and corresponding design pressure and thermal transients of the design basis analysis envelop those for Point Beach, Units I and 2 at the reduced T condition. Based on the above review, y
we agree with Westinghouse's c,onclusion that the structural integrity of the RCPs is not adversely impacted by the reduced T,y operating conditions.
4.0 EVALUATION OF TECHNICAL SPECIFICATIONS The TSs were changed as a result of the proposed revisions to modify TS Section 15.3.1.G, " Operation Limitations," Specification 3, to reduce the reactor coolant system (RCS) total flow rate limit by 2,600 gallons per minute i
(gpm), change the overtemperature and overpower setpoints, and change the Reactor Core Safety Limits for Unit 2.
(1)
A new figure was added to TS 15.2.1, " Safety Limit, Reactor Core," which is applicable to Unit 2.
The title of the existing figure was modified to indicate it is applicable to only PBNP, Unit 1.
TS 15.2.1 was modified to read:
I
"1.
The combination of thermal power level, coolant pressure, and coolant temperature shall not exceed the limits shown in Figure 15.2.1-1 for Unit I and Figure 15.2.1-2 for Unit 2."
The associated basis was also changed to reflect the revision to TS 15.2.1.
The proposed basis revision is as follows:
"The curves of Figure 15.2.1-1 and 15.2.1-2 are applicable for 14 x 14 0FA. The curves also apply to the reinsertion of previously-depleted 14 x 14 standard fuel assemblies into an 0FA core."
(2)
TS 15.2.3.1.B(4) and (5) modified as follows:
"T' s
573.9 'F (Unit 1)
T' s
570.0 *F (Unit 2)"
The associated basis was also changed to reflect the revision to TS 15.2.3.1.B(4) and (5). The proposed basis revision is as follows:
"With normal axial power d with allowance for errors',istribution, the reactor trip limit,
) is always below the core safety limit as shown on Figure 15.2.1-1 for Unit I and Figure 15.2.1-2 for Unit 2.
If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectorsd')the overtemperature AT setpoint is automatically reduced The overpower, overtemperature, and pressurizer pressure system setpoints include the effect of reduced system pressure operation (including the effects of fuel densification). The setpoints will not exceed the core safety limits as shown in Figure 15.2.1-1 for Unit I and Figure 15.2.1-2 for Unit 2."
(3)
TS Section 15.3.1.G, " Operational Limitations," was modified to provide Reactor Coolant System flow limits specific to each unit as follows:
- 3.
Reactor Coolant System raw measured Total Flow Rate (See Basis):
a.
2 181,800 gpm (Unit 1) b.
2 179,200 gpm (Unit 2)"
l The associated basis was also changed to reflect the revision to TS 15.2.1.G.3.
The proposed basis revision is as follows:
"The reactor coolant system total flow rate for Unit 1 of 181,800 gpm is based on an assumed measurement uncertainty of 2.1 percent over thermal design flow (178,000 gpm).
The reactor coolant e
j
' system total flow rate for Unit 2 of 179,200 gpm is based on an assumed measurement uncertainty of 2.1 percent over thermal design i
flow (175,400 gpm). The raw measured flow is based upon the use of normalized elbow tap differential which is calibrated against a precision flow calorimetric at the beginning of each cycle."
We find the above changes to be acceptable as discussed in the evaluation made in Section 3.0.
1
5.0 STATE CONSULTATION
]
In accordance with the Comission's regulations, the Wisconsin State official was notified of the proposed issuance of the amendment. The State official had no coments.
6.0 ENVIRONMENTAL CONSIDERATION
These amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or change an inspection or surveillance requirement. The staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Comission has previously published a proposed finding that these amendments involve no significant hazards consideration and there has been no public coment on such finding (58 FR 43940). Accordingly, these amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR Section 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assess-ment need be prepared in connection with the issuance of these amendments.
7.0 CONCLUSION
The staff agrees with the licensee's conclusion that component and system stress and fatigue for operation at reduced reactor coolant temperatures will not exceed Code allowable limits over the current licensed life of the facility. The staff also agrees with the licensee that their JC0 is sufficient to justify operation of Unit 2 until December 31, 1996 (which coincides with the planned replacement of the Unit 2 steam generators).
The staff has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Comission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: H. Balukjian Chen-Ih Wu Date: October 27, 1993
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_9-REFERENCES
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- 1. -
Letter,'B. Link, Wisconsin Electric Power Company (WEPCo), to USNRC, dated,. June 11, 1993.
a 2.
Letter, Bob Link, WEPCo, to USNRC, dated October 19, 1993.
3.
WCAP-8745-A, " Design Basis of Over-Temperature-delta-T and Over.
Temperature-delta-T Trip Functions," E11enberger, F._ L'.,
et al., dated September 1986.
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