ML20236E174
| ML20236E174 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 10/23/1987 |
| From: | Schnell D UNION ELECTRIC CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| ULNRC-1650, NUDOCS 8710290100 | |
| Download: ML20236E174 (12) | |
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Umov Etscracg. USNRc-og $$87 A October .1901 Gratiot Street. St. Lows Donald F. SchneII Vice President U.S. Nuclear Regula tory Commission 1 ATTN:. Document Control Desk 1 Washing ton, DC 20555 Gen tlemen : ULNRC-1650 DOCKET NUMBER 50-483 CALLANAY PLANT RESPONSES TO QUESTIONS ON CALLAWAY PLANT UPRATING Re f ere nce s : 1) ULNRC-1471, dated March 31, 1987 2) ULNRC-1494, dated April 21, 1987 -3) ULNRC-1618,' dated September 18, 1987 4) NRC le tter,' dated September 28, 1987, from T. W. Alexion to D. F. Schnell 5) ULNRC-1470, dated March 31, 1987 1 l 6) ULNRC-1207, dated November 14,'1985' 7) Westinghouse Owners Group Submittal OG-209, dated January 9, 1987. 8) SLNRC 86-1, dated January 8, 1986, Steam Generator Tcbe Rupture Analysis-SNUPPS 9) SLNRC 86-3, dated February 11, 1986, Steam Generator Tube Rupture Analysis-SNUPPS 10) ULNRC-1518, dated May 27, 1987 References 1 through 3 transmitted the license application and additional supporting information for the Callaway plant uprating. This letter transmits Union Electric responses to your 1 requests for additional information transmitted by Reference 4. In the following paragraphs, each item of the NRC request is j repeated and Union Electric's response follows. Item 1 It is not clear f rom the March 31, 1987 submittal,(ULNRC-1471) that the discrepancy between maximum turbine output (3562 MWt) at Callaway and the requested uprated NSSS license power (3579 MWt) would not impact safety. Explain why disparate power 1 ' levels would not be encountered which might increase the likelihood of challenges to safety systems. ) D P
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y '. Union Electric Reponse-Prior to making our response, we feel it is necessary to make the following clarification. Item 1 refers to the maximum warranted' turbine output as being based on a thermal input of 3562 MWt.. The warranted turbine output-is 1186 MWe, whlch is . based on the current thermal input of 3425 MWt. The calculated turbine output at 105% ' of warranted flow is. called the valves-wide-open condition. This is neither warranted nor maximum. .The maximum available thermal input of 3579 MWtLis only 0.5% greater than the calculated VWO rating. Since the ratings are calculated with conservatism, it is our engineering. judgment that the turbine will utilize the f ull thermal input of 3 579. MWt. Fur thermore G.E., the turbine manuf acturer, has indicated that the turbine will operate satisfactorily with a thermal input of 35791MWt. Therefore, we expect no disparity in the reactor or turbine ratings and no increases in the likelihood of challenges to safety systems. Item 2 It is not clear that all FSAR Chapter 15 events have been reanalyzed at the requested uprated power plus uncertainty (2%). It should be explicitly stated that all Chapter 15 events were reanalyzed for this power (3651 MWt), with any exceptions identified and justified. Union Electric Response All FSAR Chapter 15 events, originally performed at 102% power for the SNUPPS FSAR, were reanalyzed or reevaluated at either 1) core power plus 2% uncertainty (3636 MWt), or
- 2) nominal core power (3 56 5 MWt) with an allowance on power determined on a statistical basis and included in the limit DNBR.
The latter approach is the Westinghouse Improved Thermal Design Proc edure (ITDP) and.is described in Westinghouse reports -WCAP-8567 and WCAP-8568, both dated July, 1975. A delineation of which of the above initial condition power values was used in each of the applicable Chapter 15 analyses, is provided by the following references: Reference 5, Attachment 5, Appendix A, Table 15.0-2 Reference 5, Attachment 5, Appendix B, Table 15.6-2 Reference 6, Attachment C, Table 15.0-2 Reference 6, Attachment D,. Table 15.6-2 l l'
n-i . Item _3_ It is not clear that all reanalyses at the requested uprated power (plus 2%) have addressed the impact of VANTAGE-5 fuel. All cases not reanalyzed should be identified, with explicit justification by reference or explanation for each case. Union Electric Response { All Chapter 15 accidents impacted by the physical change to VANTAGE-5 fuel and/or changes to analysis assumptions associated with the transition to VANTAGE-5 were reanalyzed at the uprated l po we r. These reanalyses were addressed in the OFA and VANTAGE-5 submittals. All applicable safety analysis acceptance criteria were shown by the reanalyses to be satisfied. Those Chapter 15 accidents that were not impacted by the above-mentioned parameters were evaluated at the uprated power to ensure that current analyses remain valid. Item 4 It is not clear that the basis for determining auxiliary ~ f eedwa te r (AFW) minimum flow requirement has considered operation at 3579 MWt. If original AFW sizing analyses assumed at least 3579 MWt (plus 2%), this shoud be explicitly stated and referenced. If the AFW sizing basis was reanalyzed for the uprated power, this should be stated and justification should be provided to show the identification of the limiting event had not changed as a result of the higher power of any changes in analytical technique. Union Electric Response The Station Blackout, Loss of Normal Feedwater and Feedline Break transients were each analyzed in suppor t of the Callaway VANTAGE-5 Report (Reference 5). These are the referenced analyses for the Callaway Uprating Report (Re f ere nce 1). The initial power condition for each transient analysis was assumed to be 3579 MWt plus a 2% allowance for calorimetric error. The balance of the transient analysis assumptions are consistent with the uprated NSSS Thermal power level of 3579 MWt. The auxiliary feedwater system configuration and performance assumed in these analyses have not changed from previous licensing basis calculations. As documented in Reference 5, all applicable acceptance criteria are met for the Station Blackout, Loss of Normal Feedwate r and Feedline Break transients. Therefore, examination of these FSAR Chapter 15 events suppor ts the continued acceptability of the existing Callaway auxiliary feedwater system flow requirements for the uprated power level. The Callaway OFA (Refe rence 6) and VANTAGE-5 Repor t (Reference 5) illustrate that the increase in the design basis full rated thermal power does not alter basic transien t behavior. It is, the re fore, confirmed that the cited transients continue to
] _a_ bea!ication.propriate for minimum AFWS flow requirement de sign veri For large and small break LOCA, a total AFWS flow rate of I 910 gpm symmetrically distributed to all four steam generators was assumed. The flow rate of 940 gpm is based on the operation of the turbine driven AFW pump and the assumed failure of both 3 { motor driven AFW pumps. Item 5 Justification.for the overpressure protection (safety valve 1 sizing) at. Callaway for 3425 MWt operation was given by reference to the analyses presented in WCAP-7769. Applicability of WCAP-7769 does not extend above 102% of 3425 MWt (Ref. WCAP-7769, Section 3.1.1). Callaway overpressure protection for operation 3579 MWt should be justified with analyses performed for the at uprated powe r. The justification should consider changes in fuel design and analytical technique. Union Electric Response The following response addresses the NRC question concerning overpressure protection for the callaway Plant at 3579 MWt as clarified by Mr. Frank Orr of the NRC at a 10/6/87 meeting with Union Electric and Westinghouse representatives. Specifically, the NRC reviewer verbally requested additional information on the original sizing calculations for the pressurizer safety valves, and additional data on the required pressurizer safety valve capacity for the ' limiting overpressure protection design basis transient assuming. reactor scram on various reactor trip signals. The pressurizer safety valves were originally sized in 1974. The criteria used was the peak surge rate into the pressurizer during a complete loss of steam flow from 102% of the Engineered Safeguard Design Rating or 3650.7 MWt. No credit was taken for reac tor trips, pre ssurize r or steam ge ne rator PORV's, steam dump, rod control, or pressurizer spray. The peak surge rate for 1:hese conditions was found to be 40.6 ft./sec. for a total safety valve capacity of 1,200,000 lbm/hr. Adequate relieving capacity was confirmed for the uprated conditions by performing the same analysis at 102% of the uprated power (again 3650.7 MWt). This time the peak surge rate into the pressurizer was 38.9 ft./sec. The slight decrease in surge rate is due to the changes in several modeling parameters as a result of VANTAGE-5 fuel. Therefore, the pressurizer saf ety valve capacity was concluded to be adequate for the uprated conditions with VANTAGE-5 fuel. Westinghouse has also per formed the Loss of Load - Turbine Trip (limiting overpressure transient) analysis as described in Re f erence 5, at the uprated powe r conditions, neglecting reac tor trip on turbine trip. In this analysis, the reac tor trip occurs on High Pressurizer Pressure. Additional cases were made assuming reactor scram on four various trip signals. Main feedwater flow was assumed to be zero at the start of the
_5-f ' transient. Steam generatorssafety valve flow rate was assumed to j not exceed nominal plant steam flow at any time during the .) E-Ltransient.= Operation-of' the steam generator relief valves was inot assumed. The - results of Ref erence 5. and the four additional i . cases with respect-to the pressurizer safety valve performance .are' summarized in Attachments 1 and 2 to -this! 1etter. . Item 6 Identify ~ provisions at-Callaway to address potential rod withdrawal events. during mode 4 operation. Union Electric Response The ' analysis of an uncontrolled rod cluster control assembly-bank withdrawal from a subcritical or low power startup condition lis. described in Reference 5, Attachment 5, Appendix A, Section 15.4.1.. The; analysis.repor ted therein bounds the case.of a rod withdrawal event during mode. 4' operation. .The' analyzed case assures a hot zero power temperature of 557 degrees F.which is i more conservative than.that of a lowe r initial system i temperature. In mode 4, the reactor would be tripped by the source' range! high -neutron ' flux reac tor trip which is required to - be' in operation in mode 4 by. Technical Specification 3.3.1, Table 3.3-1, I te m ; 6 '. b. j Item 7 i + Discuss Callaway near term (interim) and long term (commitments) provisions to address loss-of-coolant accidents j during: mode 4 operation. Union Electric Response l The Westinghouse Owne rs Group (WOG) has developed a program to deal with the shutdown loss-of-coolant-accident (LOCA) issue on a gene ric basis..This program was approved by the WOG member utilities in their Fall,1986 meeting. Subsequent to that mee ting, the WOG met with members of the NRC Staff on November 6,1986 to discuss the program.. In Refe rence 7, the WOG submitted a description of the shutdown LOCA program to the NRC, which incorporated the NRC comments on the program received in 'l the November meeting. In' response to the recommendation made by NRC regarding procedural guidance, the WOG developed Interim - Shutdown LOCA Guidance. This guidance provides information to he considered ' in the development of procedure or training for protecting the l reactor core in the event of a LOCA that occurs during shutdown l when automatic safety injection actuation may be unavailable and when emergency core. cooling system equipment is out-of-service due to technical specification requirements relating to cold overpressure protection. Union Electric training programs ' include Shutdown LOCA sce nar ios. Union Electric will continue to l suppor t the WOG e f for ts toward a gene ric resolution and will J s u
25 Z Z 2 2 L. if 1 ] , l ) implement portions of the. final resolution as deemed specifically ) required.for Callaway. 1 - Item'8 ] In a similar plantiuprating, another Westinghouse Plant -(North Anna) ' required-about 20 technical specification changes to ' accomodate the change to' a higher power. :The Callaway submittal-(ULNRC-1471) ~ lists only about four, most' oC them clerical; and - the pages" don't seem consistent with corresponding technical' specification pages of the~ Callaway reload submittal (ULNRC-Provide, additional documentation to identify and justify 1470)'. all..technicalf specification" changes needed to accommodate the ~ powe r uprating; those. changes ;which were previously made for ;the VANTAGE-5 reload but'also needed for uprating should be included
- .and - j ust ified. by re f erence.
1 IUnion Electric Response f . Attachment 3 to this letter provides a comparison ~ between -l i the Nor th Anna Uprating; submittal technical specification changes .and.the.Callaway submittals. In the first column, the North Anna ( changes'have.been-grouped so that more than one change may fall-under the listed category. The second column gives the location for similar changes in the Callaway-submittal. For convenience a column is also provided to give the justifications for the Callaway changes. On October 22, 1987, during a conversation with Mr. Frank Orr of the NRC, Union Electric was requested to provide information on steam generator tube rupture (SGTR) analyses at' i the uprated power level. The Callaway SGTR analyses have been l l evaluated to -the proposed uprated power level. These analyses have been submitted to the NRC in References 8, 9, a nd 10. The l SGTR-Stuck Atmospheric Relief Valve scenario was evaluated at } 102% of the uprated power as indicated in Table 3-2 of Reference 8.. The SGTR-Overfill Scenario was also evaluated at the uprated ] power-condition as noted in Table 1 of Reference 10. Note that ( Reference 10 represents a reanalysis, per NRC request, of the l Overfill Scenario as presented in Reference 8. If there are any further questions or additional information j l is. required, please contact us. 1 Very truly yours, / Donal F. Schnell i DJW/ma t Attachment x_-._.l_--._----.-
.f' ,\\ ~ I' STATE OF MISSOURI -) ,), S S-CITY OF'ST. LOUIS ) 1cRobert J. -Schukai~, of lawf ul age, being first duly sworn ~ upon oath says that he is General Manager-Engineering.(Nuclear) for Union. Electric Company; that he has read the' foregoing document and 7 L 'knows'the content.thereof; that he'has executed the'same_for and on behalf of said: company with. full power and' authority to do so;'and-that the facts therein stated are true.and correct.to the best'of his knowledge, information and belief.. By C l' A Robe O Schukai General Manager-Engineering' Nuclear-SUBSCRIBED and sworn to before me this 88Alday of Md$/au',198 7, / &f BARBARAbfAF[ 8[ NOTARY PUBUC, STATE OF MISSOURI MY COMMISSION EXPIRES APRIL 22, 1983 ST. LOUIS COUNTY.
.x. hs' ,f 'l pp c. (c'c: Gerald;Charnoff, Esq.. Shaw, Pittman, Potts'&.Trowbridge ' 2 3 0 0 N..' S tree t ', 'N.W. Washington, D.C. 20037
- Dr..J. O.'Cermak-CFA,.Inc.
' 4 Professi,onal.. Drive. (Suite.110) 'Gaithersburg, MD 20879-W. L.L*Forney Chief,jReactor. Project Branch 1' 'U.S. Nuclear-Regulatory Commission Region;III 799, Roosevelt' Road.. ' Glen Ellyn,LIllinoisL60137 2 Bruce.'L5ttle ~ Callaway'. Resident. Of fic'e'- 'U;S. Nuclear _ Regulatory Commission' o -RR$1 Steedman,IMissouri 65077 7 Tom /Alexlon.(2) Office of Nuclear Reactor. Regulation N .U.S. ' Nuclear Regul'atory ' Commission. Mail Stop 316' 7920 Norfolk Avenue Bethesda,-MD' 20014
- Ron.Kucera, Deputy Director
~ ' Department.of Natural Resources P.O.: Box-176 Jeffe'rson City,..MO 65102- --Manager, Electric Department Missouri Public Service Commission .P.O. Box 360
- Jefferson City, MO 65102 Alm _-m.___.j___..._,__.,
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