NPL-99-0369, Application for Amends to Licenses DPR-24 & DPR-27 to Reflect Required Changes to TS as Result of Using Upgraded Fuel at plant.Non-proprietary,WCAP-14788 & Proprietary WCAP- WCAP-14787 Repts Encl.Proprietary Rept Withheld
| ML20196F289 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 06/22/1999 |
| From: | Reddemann M WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20137U998 | List: |
| References | |
| NPL-99-0369, NPL-99-369, NUDOCS 9906290174 | |
| Download: ML20196F289 (8) | |
Text
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h MARK E. REDDEMANN gg gg Sits Vee President A WISCONSIN ENEROY COMPANY Point Beach Nuclear Plant 6610 Nuclear Rd.
Two Rivers, WI 54241 Phone 920 755-6527 r
NPL 99-0369 10 CFR 50.4 10 CFR 50.90
_' June 22,1999 10 CFR 51.22 Document Control Desk U. S. NUCLEAR REGULATORY COMMISSION Mail Stop PI-137 L Washington,DC 20555 E Ladies / Gentlemen:
DOCKETS 50-266 AND 50-301 TECHNICAL SPECIFICATION CHANGE REOUEST 210 AMENDMENT TO FACILITY OPER ATING LICENSES TO REFLECT REOUIRED CHANGES TO THE TECHNICAL SPECIFICATIONS AS A RESULT OF USING UPGR ADED FUEL POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 In accordance with the requirements of 10 CFR 50.4 and 10 CFR 50.90, Wisconsin Electric Power Company (WE), licensee for the Point Beach Nuclear Plant (PBNP), proposes to amend Facility Operating Licenses DPR-24 and DPR-27 for PBNP Units 1 and 2, respectively. The proposed changes provide for the design and operation of the Point Beach fuel cycles with
- upgraded Westinghouse fuel features discussed t>elow, and at higher core power peaking factors than are currently permitted by the plant Technical Specifications.
Included in Attachment I to this letter is a description of the proposed Technical Specification changes and supponing information. Included in Attachment 2 to this letteris a safety evaluation of the proposed changes. Included in Attachment 3 to this letter is a no significant hazards detennination. Included for information in Attachment 4 to this letter is a draft of the PENP FSAR Chapter 14 " Safety Analysis" changes required as a result of the analyses performed for the upgraded fuel. Included in Attachment 5 to this letter is " Westinghouse Revised Thermal
. Design Procedure (RTDP) Instrument Uncenainty Methodology for WE Point Beach Units 1 and 2" (WCAP-14787 (Proprietary) and WCAP-14788 (Non-Proprietary)) used in the analyses performed for the new fuel. Included in Attachment 6 to this letter are the marked-up Technical f[
Specifications indicating the proposed changes.
Attachment 5 of this submittal (RTDP Instrument Uncertainty Methodology) contains two parts - 'P/
(pan A and part B). Pan A of Attachment 5 contains information proprietary to Westinghouse.
Part B of the same attachment contains the non-proprietary version of the WCAP. Accordingly, included in Attachment 5 is a Westinghouse Application for Withholding Proprietary Information from Public Disclosure and an accompanying Affidavit signed by Westinghouse, the owner of the information.
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9906290174 990622 l PDR ADOCK 05000266!
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P NPL 99-0369 June 22,1999 Page 2 I
The Affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.790 of the Commission's regulations. WE requests that the Westinghouse proprietary information be withheld from public disclosure in accordance with 10 CFR 2.790. . Correspondence regarding the proprietary aspects of this Westinghouse report should reference CAW-991335 and be addressed to N. J. Liparulo, Manager of Equipment Design and Regule ary Engineering, Westinghouse Electric Company LLC, P. O. Box 355, Pittsburgh, Pennsylvania 15230-0355.
Discussion PBNP plans to refuel and operate, commencing with Unit 2 Refueling Outage 24 (currently scheduled to begin September,2000), with upgraded Westinghouse fuel features. The first reload with the upgraded fuel for Unit I would commence with Unit 1 Refueling Outage 26 (cunently scheduled to begin April,2001). The upgraded fuel is 0.422 inch outer diameter (OD),
14x14, VANTAGE + fuel with PERFORMANCE + features; hereafter referred to as 422V+. A detailed discussion of the upgraded fuel is included in Attachment 2 of this letter. A summary of the major fuel design parameter changes are provided below.
Summary ofmajorfueldesign changes:
- Advanced zirconium alloy material ZlRLO* cladding with zirconium oxide coated lower section e ZlRLO* instrumentation and guide thimble tubes e ZIRLO* grids with enhancements in the mixing vane pattern
- 0.422 inch rod diameter vs. 0.400 inch for Optimized Fuel Assembly (OFA) fuel currently used The use of the new ZIRLO* fuel assembly material has been evaluated and it has been determined that it is compatible with the PBNP reactor coolant system chemistry parameters. A more detailed discussion on the ZIRLO* alloy and the 422V+ fuel design is provided in Attachment 2.
Incorporating this fuel into future PBNP cores provides many advantages. The major advantages of the 422V+ fuel are provided below.
Advantages offuelupgrade:
- Increased safety margin (peak clad temperature (PCT) margin and departure fram nucleate boiling ratio (DNBR) margin)
- Reduction in the number of required feed fuel assemblies thereby reducing new fuel costs and costs associated with dry fuel storage e Improved fuel assembly components e Improved economics in the suppon of extended fuel cycles
NFL 99-0369 June 22,1999 Page 3 e Improved structural integrity, which reduces bowing and twisting and enhances fuel moves l Technical Specification changes are required to incorporate the 422V+ fuel assemblies into the PBNP cores. A summary of the r quired Technical Specification changes is provided below and are discussedin detailin Attachment 1 of this letter.
Summary of Required Technical Specification Changes:
- Changes to the reactor core safety limit curves to reflect transition cores and full 422V+ cores l
. Changes in the OTAT and OPAT reactor trip setting limits
. RCS Tug range change to 558.1 F to 574.0 F from the cunent 557 F to 573.9 F e RCS flow measurement uncertainty increase to 2.4% from the current 2.1% and !'
corresponding increase in RCS raw measured total flow rate to 182,400 gpm from the current 181,000 gpm a Full power F% peaking factor design limit will increase to 1.77 from the current 1.70
- Maximum Fo(Z) peaking factor limit will increase to 2.60 from the current 2.50 and the K(Z) envelope will be modified a Changes to reflect ZIRLO* material e Changes to reflect transition cores and full cores of 422V+ fuel assemblies in the core
. Restrictions on primary system pressure (2250 psia) for cores containing 422V+ fuel assemblies
. Restdctions on storing fuel in the new fuel vault storage cells e Conesponding Basis section changes to reflect the above changes Implementation of the proposed Technical Specification changes bound operation with reactor cores containing a full core of OFA fuel assemblies, transition cores containing both burned OFA and 422V+ fuel assemblies, and cores containing a full core of 422V+ fuel assemblies.
The loading schedules for incorporation of the upgraded fuel dictate that the PBNP units operate at different operating pressures for a currently scheduled period of approximately five months (the current plan is for Unit 1 to remain on line and operate at 2000 psia until it is shutdown and loaded with 422V+ fuel dudng U1R26, while Unit 2 is loaded with 422V+ fuel and operating at 2250 psia after U2R24). In addition, the analyses done to support the 422V+ fuel require operation at 2250 psia primary system pressure with any combination of 422V+ and burned OFA fuel. To reflect this situation the proposed Technical Specification changes were clearly annotated (where appropriate and necessary) with contingencies for the type of fuel assemblies contained in the reactors, beceuse the Technical Specifications will need to apply to both Units.
1 j The analyses performed by Westinghouse for incorporation of the 422V+ fuel at PBNP was done at increased pdmary system pressure and up-rated thermal power conditions (i.e. nominal primary system pressure of 2250 psia at a thermal reactor power level of 1650 MWt). However, after the initial loading of the 422V+ fuel for each unit, PBNP will operate the reactor cores at the increased primary system pressure of 2250 psia, but not at the up-rated thermal power i
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NPL 99-0369 )
June 22,1999 Page 4 conditions of 1650 MWt (due to additional modifications and analyses required to operate at the up-rated thermal)ower conditions). A separate License Amendment Request will be submitted at a later date when PBNP pursues operating the reactor cores at up-rated thermal power 4 conditions. Westinghouse has confirmed that the analyses performed for the 422V+ fuel at up-rated conditions bound the conditions that PBNP currently intends to operate the reactor cores under (i.e. RCS pressure of 2250 psia at the currently licensed thermal reactor power level of 1518.5 MWt).
This Technical Specification Change Request is supported by the results of the reanalysis of the PBNP FSAR Chapter 14 design basis accidents and events. A detailed discussion of the i analyses performed for incorporation of the 422V+ fuel at PBNP is included in Attachment 2, )
and the resulta It required FS AR (Chapter 14) safety analysis changes are included in Attachme it
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4 for information. !
Westinghouse has performed the large break Loss Of Coolant Accident (LOCA) analysis using ,
I the Best Estimate LOCA (BELOCA) Methodology described under WCAP 14449-P
" Application of Best Estimate Large Break LOCA Methodology to Westinghouse PWR's with Upper Plenum Injection." This WCAP was approved by the NRC for referencing in Licensing Submittals by NRC letter and SER dated May 21,1999 (TAC NO. M94035). Westinghouse is currently incorporating conditions and documentation requirements imposed by the NRC into the WCAP and willissue WCAP 14449-P-A WRC approved) shortly. A summary discussion of the BELOCA methodology is included in Att - at 2, and the Chapter 14 FS AR changes required as a result of the analyses are included in Attachment 4.
Storage of the 422V+ Fueh Storage of the 422V+ upgraded fuel at PBNP was reviewed with respect to criticality effects, the heat transfer capability of the spent fuel pool cooling system, gamma heating effects, and structural loading. Spent fuel pool criticality analyses for the 422V+ fuel were performed by Westinghouse. The analyses concluded that the spent fuel pool La remains below the 0.95 limit for 422V+ fuel. Westinghouse also performed criticality calculations for the new fuel vault that demonstrate that new fuel assemblies with the maximum enrichment of 5.0 w/o U235 and a minimum M 321.25x IFB A rods can utilize all available new fuel storage cells. Accordingly, this restriction on storing new fuel in the new fuel vault is being added to TS 15.5.4.2 " Fuel Storage"(see Attachment 1). The new f uel vault La remains below the 0.95 limit for the fully flooded condition, and below the 0.98 limit for the optimum moderation condition with these restrictions in place. Therefore, the storage of 422V+ fuel meets the required criteria for spent fuel and new fuel storage.
The effect of storage of 422V+ fuel on spent fuel pool cooling is negligible. The decay heat load ,
will not change due to the change in fuel design. Use of 422V+ fuel will not significantly or I adversely affect the natural circulation cooling process in the spent fuel pool. The spent fuel pool heat removal capability will continue to be verified greater than the decay heat load prior to movement of fuel to the spent fuel pool Gamma heating from 422V+ fuel would not result in m
NPL 99-0369 June 22,1999 Page 5 boiling at the spent fuel pool storage location, or unacceptable temperatures elsewhere in the pool. Therefore, the existing spent fuel pool cooling system heat removal capability will not be affected by the 422V+ fuel.
Spent and new fuel pool rack loading, spent fuel pool structure loading, and fuel assembly drop impact loadings all assume a certain weight for the fuel assembly. The standard fuel assembly is the heaviest assembly, followed by the 422V+ fuel and then the OFA fuel. Previous analyses assumed weights consistent with standard fuel assemblies in loading calculations. Since 422V+
fuel weighs less than the weights used in previous analyses, existing structural evaluations will bound the use of the 422V+ fuel.
In conclusion, the applicable aspects of storing the 422V+ fuel assemblies have been reviewed and it has been determined that the applicable acceptance criteria continue to be met and that the applicable analyses remain bounding.
Increased Safety Afargin in PCT and DNBR:
Current Afargins: l In the cunent Point Beach Unit I and 2 designs, the OFA fuel has 8.6% generic margin between
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the design limit DNBR and the safety analysis limit DNBR. Of the 8.6% generic margin,2.6%
of this is reserved for a rod bow penalty. This leaves 6% margin to cover operational occurrences and minor variations of reload parameters for the various transient analyses, which has been sufficient to cover issues during past cycles of operations. The Large Break Loss of Coolant Accident (LBLOCA) transient Peak Cladding Temperature (PCT) for Unit 2 (which is limiting of the two units) currently has a 15 F margin to the 10 CFR 50.46 criterion limit of 2200 F.
Increase in Afargins with 422V+ Fuel:
The limiting transients for Peak Cladding Temperature (PCT) and Departure from Nucleate Boiling (DNB) for the Fuel Upgrade accident analyses were performed at the uprated power level l of 1650 MWt with increased peaking factors. With a full core of the 422V+ fuel and accounting for the uprated power level and increase in peaking factors, the generic DNBR margin between the design limit DNBR and the safety analysis limit DNBR is 8.8%. Again,2.6% of this margin is reserved for the rod bow penalty which leaves 6.2% margin to cover operational occurrences and minor variations of reload parameters for the various transient analyses.
However, since the Point Beach units will not be uprated with the fuel upgrade, additional DNBR margin will be available (approximately 15%). This will provide about 21% margin for use during the transition cycles from OFA to 422V+. Note that the above discussion of DNBR ,
margin does not address transition core penalties; the discussion concentrates on margins that l will exist following the transition to cores containing only 422V+ assemblies. The transition core penalties have also been addressed for the transition cycles and sufficient margin exists to cover these penalties. A more detailed discussion of the transition core penalties is provided in Section 4.6 and Table 4-3 of Atte.chment 2.
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NPL 99-0369 June 22,1999 Page 6 1-With a full core of the 422V+ fuel and accounting for the uprated power level and increase in peaking factors, tile PCT margin for the BELOCA analysis is 72 F from the 2200 F 10 CFR 50.46 criterion limit. The fuel upgrade BELOCA analysis in combination with the 422V+ fuel product provides an increase in margin of 57 F over the margin that currently exists for the OFA fuel.
In conclusion, implementation of the 422V+ fuel design at PBNP allows for increased safety margins in both PCT and DNBR. A more detailed discussion of the PCT and DNBR margins is included in Attachment 2.
As indicated above, detailed analyses and evaluations were performed to support the loading of 422V+ fuel into the PBNP reactor cores. The resuhs of the analyses / evaluations lead to the following general conclusions summarized below.
Analyses / Evaluation
Conclusions:
- The Westinghouse fuel assemblies containing 14x14 422V+ upgraded fuel features for PBNP are mechanically compatible with the current 14x14 OFA fuel assemblies, control rods, peripheral power suppression assemblies, flux detectors used in the instrumentation tubes, fuel handling equipment and reactor internals interfaces, e The stmetural integrity of the 14x14 422V+ fuel assembly features has been evaluated for seismic /LOCA loadings for PBNP. Evaluation of the 422V+ fuel assembly components stresses and grid impact forces due to postulated faulted condition accidents verified that the fuel assembly design is stmeturally acceptable with regards to grid crush. These faulted condition loads include seismic and LOCA forces.
- Changes in the nuclear characteristics due to the transition to 14x14 422V+ fuel assembly features are addressed in Attachment 2 of this submittal. Changes in the nuclear characteristics in equilibrium cycles of the 422V+ fuel assembly will be within the range normally seen for OFA fuel from cycle-to-cycle due to fuel management.
- The reload 14x14 422V + fuel assemblies are hydraulically compatible with the 14x14 OFA fuel assemblies.
- The change in the design full power F"as limit from 1.70 to 1.77 is supponed by analyses summarized in Attachment 2 of this submittal. This change in F"an is only applicable to the new design (422V+). The 14x14 OFA design will retain an F"as limit of 1.70.
- The change in the maximum Fo(Z) limit from 2.50 to 2.60 and moditication to the K(Z) envelope is supported by analyses summarized in Attachment 2 of this submittal. This change in Fn(Z)is only applicable to the new design (422V+). The 14x14 OFA design will retain an Fo(Z) limit of 2.50.
- The core design and analysis results documented in this submittal show the core's capability to operate safely with the uprated power conditions and bound the conditions that PBNP currently intends to operate the reactor cores under (i.e. RCS pressure of 2250 psia at the cunently licensed thermal reactor power level of 1518.3 MWt).
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NPL 99-0369 l
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It has been determined that the proposed changes meet the categorical exclusion criteria of 10 CFR 51.22(c)(9) in that they: (1) Involve no significant hazards consideration;(2) Do not result in a significant change in the types or significant increase in the amounts of any effluents released eff-site; and (3) Do not result in an significant increase in individual or cumulative radiation exposure. Therefore,in accordance with 10 CFR 51.22(b), an environmental assessment or impact statement need not be prepared.
WE respectfully requests that the NRC begin review of this amendment submittal as soon as possible and requests approval of it prior to January 15,2000. Approval of this amendment prior to January 15,2000 will ensure enough lead time for Westinghouse fuel assembly manufacture and fabrication. Should you have any questions on this submittal or require additional I information, please contact me. I 1
Sincerely, Os p N
.Tark E. eddemann Site Vice President Point Beach Nuclear Plant Sunscribed to or, this # jndday swornof before h me ,1999
'_p/9 8W 'b NotaryPublic, State of Wisconsin V-l My Commission expires on ///// /Mi .
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MAW / tat Attachments cc: NRC Regional Administrator l NRC Resident Inspector l NRC Project Manager l PSCW 1
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- bec: M. 'A. Wiesneth W. J. Hennessy J. A. Kos F. Cayia*
M. B. Sellman* J. R. Anderson
- C. R. Peterson* J. G. Schweitzer*
T. G. Staskal* J. K. Ettien* A. B, Beach
- T. A.Trochil(3)*
M. F. Baumann* R. C. Amundson* R. K. Hanneman* B. J. Onesti (OSRC)*
J. A. Palmer
- R. R. Grigg* G. M. Krieser* S. G. Cartwright*
File M. E. Reddemann/R. G. Mende* *w/o attachment I
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