ML20044D371
| ML20044D371 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 05/10/1993 |
| From: | Barth C NRC OFFICE OF THE GENERAL COUNSEL (OGC) |
| To: | Chilk S NRC OFFICE OF THE SECRETARY (SECY) |
| References | |
| CON-#293-13958 DCOM, NUDOCS 9305190058 | |
| Download: ML20044D371 (10) | |
Text
{{#Wiki_filter:/3 . DOCKET ME **2 -h m (7-L enoo.a urit eAc. 'o, UNITED STATES NUCLEAR REGULATORY COMMISSION DOCdiED ~ -{ ,I W ASHINGTON, D. C. 20555 V5 HEC k*****/ May 10, 1993 '93 iM 12 Td0:56
- i-Samuel J. Chilk, Secretary Office of the Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555 In the Matter of SACRAMENTO MUNICIPAL UTILITY DISTRICT-(Rancho Seco Nuclear Generating Station) 40cket No. 50-312-OLA l
Dear Mr. Chilk:
In my letter to you of March 30, 1993, I stated that the licensee informed us that it was reevaluating the information previously supplied to the NRC staff. The staff relied on this information in preparing its March 26, 1993, filing "NRC Staff's Support.of Licensee's Motion for Reconsideration" with the Commission in the Rancho Seco proceeding. In the April 1,1993, letter from the licensee (Enclosure 1), it submitted-sufficient additional clarifying information to enable the staff. to determine - that the conclusion in its March 26, 1993, filing need not be amended. The staff assessment of the need for electrical power for spent fuel pool cooling i at Rancho Seco based on. the latest licensee submittal is provided in a Sincerely, l ? Charles A. Barth Counsel for NRC Staff
Enclosures:
I 1. SMUD to NRC letter, 4/1/93 -2. NRC Staff Assessment of Loss of Offsite Power-l cc: Service List j e i 9305190058 930510 j PDR ADDCK 05000312 i G PDR 6O i l
iI Mr. James R. Shetler Rancho Seco Nuclear Generating Station Docket No. 50-312 CC" Mr. S. David Freeman, General Manager Mr. Thomas D. Murphy Sacramento Municipal Utility District Atomic Safety and Licensing Board 6201 S. Street Panel P. O. Box 15830 U.S. Nuclear Regulatory Commission Sacramento, California 95813 Washington, D.C. 20555 Thomas A. Baxter, Esq. Mr. John Bartus Shaw, Pittman, Potts & Trowbridge Ms. JoAnne Scott 2300 N. Street, N.W. Federal Energy Regulatory Washington, D.C. 20037 Commission 825 North Capitol Street, N.E. Mr. Jerry Delezinski Washington, D.C. 20425 Licensing Supervisor Sacramento Municipal Utility District Ms. Helen Hubbard Rancho Seco Nuclear Generating Station P. O. Box 63 14440 Twin Cities Road Sunol, California 94586 Herald, California 95638-9799 Fnvironmental Conservation Mr. Robert B. Borsum, Licensing Organization Representative Suite 320 B & W Nuclear Technologies 101 First Street Nuclear Power Divisior. Los Altos, California 94022 1700 Rockville Pike - Suite 525 Rockville, Maryland 20852 Ms. Jan Schori, General Counsel Sacramento Municipal Utility Regional Administrator, Region V District U.S. Nuclear Regulatory Commission 6201 S. Street 1450 Maria Lane, Suite 210 P. O. Box 15830 Walnut Creek, California 94596 Sacramento, California 95813 Sacramento County James P. McGranery, Jr., Esq. Board of Supervisors Dow, Lohnes & Albertson 700 H. Street, Suite 2450 Attorneys At Law Sacramento, California 95814 Suite 500 1255 Twenty-Third Street, N.W. Mr. Leo Fassler Washington, D.C. 20037-1194 Assistant General Manager and Chief Operations Officer Mr. John Hickman Sacramento Municipal Utility District Senior Health Physicist 6201 S. Street Environmental Radioactive P. O. 15830 Management Unit Sacramento, California 95852-1830 Environmental Management Branch State Department of Health Mr. Charles Bechhoefer, Chairman Services Atomic Safety and Licensing Board 714 P. Street, Room 616 Panel Sacramento, California 95814 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Mr. Richard F. Cole Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555
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SMUD l 5 SACRAMENTO MUNietPAL UTIUTY DISTRICT.~. P. O. Box 1583o. Sacramento CA 95B52183o.19161452 3211 i AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA j DAGM/NUC 93-079 .j April 1, 1993 U. S. Nuclear Regulatory Commission j! uttn: Document Control Desk Washington, D. C. 20555 ) i i Docket No. 50-312 l Rancho Seco Nuclear Station License No. DPR-54 j CLARIFICATION OF TI1E PERMANENTLY DEFUELED TECI1NICAL SPECIFICATION LOOP AND SFP DECAY HEAT { ANALYSES ) 9
References:
J l 1. J. Sheller (SM UD) to S. Weiss (NRC) letter DAGM NUC 91-183, dated November 19,1991, Proposed Amendment No.182, j Revision 3, the Permanently Defueled Technical Specifications l 2. D. Keuter (SM UD) to S. Weiss (NRC) letter AGM/NUC 91-081, i 2 dated May 20,1991 Proposed Decommissioning Plan 3. J. Sheller (SMUD) to T. E. Murley (NRC) letter DAGM/NUC-91-136, dated October 21,1991, Supplement to Rancho Seco Environmental Report - Post Operating License Stage .l ] Attention: Seymour Weiss f Based on our discussions with your staff regarding the Loss Of Off-site Power 1 (LOOP) and Spent Fuel Pool (SFP) decay heat analyses presented in the safety i analysis for the Permanently Defueled Technical Specifications (PDTS), we are providing the attached, modified LOOP and SFP Decay Heat Load safety analysis summaries. These summaries clarify the conclusions made in j Reference 1. We are also providing additional supplemental information, based. on our calculations and analyses performed for Reference 1. ~9304090227 3040'1' ' a PDR A 05000312-3d i l PDR TV l P DISTRfCT HEADQUARTERS O 6201 S Street. Sacramento _CA 95817-1899
Seymour Weiss DAGM NUC 93-079 l l i Refer to the LOOP and SFP Decay Heat Load analyses that begin at the bottom of page 2 of 71 and the middle of page 3 of 71 of Attachment II to Reference 1, l respectively. Attached is the revised wording for these two analyses. The revised [ t wording is shown in italics. The PDTS safety analysis (Reference 1) provided information based on difTerent assumptions and analyses than those used in the Rancho Seco Decommissioning Plan (DP) and Environmental Report Supplement (ERS)(Reference 2 and 3). The DP and ERS assumptions and conclusions continue to be valid. We base our conclusions, as stated in the DP and ERS, on early, conservative District analyses. The PDTS analyses incorporate newer empirical SFP heat-up information that demonstrates the SFP can not reach boiling when assummg evaporative cooling. Members of your staff requiring additional information or clarification may contact Jerry Delezenski at (916) 452-3211, extension 4914. Sincerely, ^ G ,e ,/, ,4,.-, S / M $.: x / - James R. Shetler f Deputy Assistant General Manager Nuclear i Attachment cc: J. B. Martin, NRC, Walnut Creek S. Brown, NRC, Rockville I t i c
ATTACHMENT Revised LOOP and SFP Decay Heat Load Analyses (2 Pages) i b I 4 t ? I b
LOOP During normal power operations and post accident conditions, it is imperative that electrical power be available to support equipment needed to operate the plant and mitigate the consequences of design basisaccidents associated uith poner operations. In the PDM, a LOOP would result in the loss of SFP cooling. However, there is adequate time available See +hMFP 4esay-heet-lead-waluanca bhw) to take corrective action without a safety consequence, even in the event of an extended LOOP condition (i.e., a totalLOOPfora prolonged period oftime wellin excess ofseveralhours). As ofJune 7,1991, an extrapolation ofDistrict calculation Z-SFC-M2533 shous that foran initialSFP level of23 feet 3 inches, it would take a minimum of15 days to boildoun the SFP to the top ofthe spent fuelassemblies. Inis amount oftime does not consider the time to heat up the SFP nater from the maximum allowed normal operating temperature (140 *F) to the boUing point (212 *F). Rancho Seco has six off-site power transmission lines that are tied to the SMUD and PG&Ee/ectrica/ grids. SMUD has the capability to receive power directly from the District's hydroelectric or other electricalgenerating units in less than eight hours. A District evaluation of the off-site electrical grid for Rancho Seco, performed pursuant to 10 CFR 50.63, Station Blackout, verified the stability of the Western grid. The probability of a LOOP at Rancho Seco, as evaluated in accordance with the guidelines of Regulatory Guide.1.155, is less than once per 20 years. Therefore, the emergency diesel generators are not required to ensure power availability to support SFP cooling equipment in the event of a LOOP. An ahernate power supply can be made available well within the minimum time required to take corrective action to restore SFP cooling. A simpleaddition of nater to the SFP, sis the dieseldriven Erepumpper theplant Loss ofSFP Cooling or Lese 1 recovery procedure, would greatly utend the time a vailable to recoserfrom a LOOP condition (see also SAR page 13 fora destnplion ofthe a vailable SFP water make-up methods). In addition, ifa LOOP event occurs at Rancho Seco, it is expected to be ofa short duration (on the order ofa few hours). This conclusion is basedon the oft-site a/cpouerreliability analysis summarizedin the aboreparagraph. A short duration L OOP uillha ve a neghgible impact on an anal) sis which looks at the long term e/Tects oflosing theprimary spent fuelpoolcooling s) stem. Fora short duration LOOP condition coupled uith an extendedloss oftheprimary SFP cooling system, District calculation Z-SFC-M2SS4 (summarized belowin the SFP Decay Heat Load analysis) concludes that for any initial SFP level 223 feet 3 incLes, it takes a minimum of 250 hours before the SFP bulk coolant temperature can reach a steady state value of approximately 185 *F, asof November 1,1991. Thus, based on this District calculation, SFP water will not boil following an extendedloss of the primary SFP cooling system that may or may not include a short duration LOOP event.
SFP Decav Heat Load The controls required to protect the spent fuel in the PDM are predicated primarily on the level of decay heat in the SFP. The District calculated (SMUD Calculation Z-SFC-M2551) the decay heat load for the SFP in the defueled condition using the methodology described in ANSI /ANS 5.1-1979, Branch Technical Position (BTP) ASB 9-2, andempirica/SFPheat-up data. The decay heat load in the SFP, as a function of calendar date, is shown in Table 1 (see page 54 of this SAR). Also, this table provides the calculated SFP water heat-up rate following loss of the primarySFP cooling system nith the fue/ storage building (FSB) ientilation s) stem operating. The FSB ientilation system consists ofone oftuv udundant A uxiliary Building exhaust fans and a FSB supply fan. Specifically, Table I provides = a functiereef-edeMudate, the r SFP decay heat load in BTU /hr, calculated using BTP ASB 9-2 andmodified using empirica1SFPheat-up data. Table 1 also presents the time (in hours) it takes following the loss of primary SFP cooling, as a function ofinitial SFP temperature, for the SFP temperature to reach 180*F and/or a steady state temperature, glien initial SFP levels of 23 feet 3 inches and 37 feet. This District SFP heat-up rate evaluation concludes the SFP will not boil during the PDM if primarySFP cooling is lost at either of the two minimum allowed SFP levels uith the FSB ientilation system operatinge The maximum steady state SFP { temperature that can be reached from these conditions is approximately 185'F. l During the defueled condition, the normal operating SFP temperature is maintained below 90"F. A SFP operating temperature near70"F is not uncommon. At an initial SFP temperature of 90*F and 140*F,if the pnmary SFP cooling systemis lost, a minimum of 350 and 250 hours, respectively, is available for operators to take corrective action to restore the primarySFP cooling system ptior to exceeding 180*F when the initial SFP level is 223 feet 3 inches. Again, the maximum steady state SFP temperature that can be reached under these conditions is approximately 185*F. No safety implications exist at a 23 feet 3 inch SFP level as long as personnel exposure is monitored and maintained as low as is reasonably achievable. The ruptured fuel assembly event does not require a specific minimum SFP water level to mitigate its consequences. Requiring a minimum water level of 23 feet 3 inches in the SFP provides adequate shielding above active fuel to protect i individuals in the Fuel Storage Building w hen fuel handling operations are not in progress. Also, a minimum of A7 feet of water is maintained in the SFP during fuel handling operations for shiMing and worker safety concerns only, not for accident mitigation concerns. Based on this same District calculation summarizedabovefortheSFPDecay
- Heat Load, seieralhundredhours are available before significant evaporative losses of SFP water intentory can occur due to a loss ofprimary SFP cooling.
Approximately 1 foot ofSFP nater would evaporate every 70 hours if the SFP uater uasat thesteadystate temperature ofapproximately 180*E A simple i addition of water to the SFP would make up any evaporative water losses and extend the time to implement corrective actions to restore primarySFP 1 cooling, if necessary. Specification D3/4.1 requires operators to take immediate I actions to restore the SFP level if the level drops below the minimum allowed level. The evaluation on page 15 of this SAR lists several options available to l operators for providing make-up water to the SFP. i i f
ENCLOSURE 2 r NRC STAFF ASSESSMENT OF THE LOSS OF 0FFSITE POWER The release of radioactive materials into the environs and the resulting public radiation exposure are the primary impacts of a decommissioning-related accident. The radionuclide inventory at Rancho Seco Nuclear Generating Station (RSNGS) is less than the radionuclide inventory at the reference PWR evaluated in the NRC report " Technology, Safety, and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station," NUREG/CR-0130. The accidents postulated for SAFSTOR at RSNGS are bounded by the accidents postulated in NUREG/CR-0130 which formed the basis of the NRC re art " Final a i Generic Environmental Impact Statement on Decommissioning of Nuclear facilities," NUREG-0586. In addition, during the Custodial-SAFSTOR period, fuel-handling accidents and complete loss of offsite power continue to be possible at RSNGS. These accidents were addressed in Chapter 14 of the licensee USAR, and the Licensee Proposed Amendment No.182, Revision 3 and its two supplements. The short-lived radionuclides in the spent fuel at RSNGS have undergone substantial radioactive decay since the plant shut down on June 7, 1989. Therefore, the dose impact for a fuel handling accident at RSNGS is now significantly less than the dose for an operating plant. The radionuclide of concern for a fuel handling accident is Kr-85. In the current plant condition, a 2-hour integrated total body dose attributed to the maximally exposed individual is 0.013 rem. This dose is a small fraction of the 10 CFR Part 100 accident dose limit of 25 rem. The 13-mrem whole-body dose is approximately 1.3 percent of i the 1000-mrem protective action guideline recommended in the EPA report, "A Manual of Protective Action for Nuclear Incidents," EPA 520/1-75-001. During a complete loss of offsite power, control of the spent fuel decay heat i load, and thus the protection of the spent fuel integrity, is the primary consideration. The controls required to protect the spent fuel are based on anticipated decay heat generated by the spent fuel stored in the SFP. A complete loss of offsite power would result in loss of the spent fuel cooling system (SFC)andofthespentfuelbuildingventilationsystem. SMUD analyzed the effect of a complete loss of offsite power on the SFP. This analysis assumed the most limiting initial conditions allowed by the RSNGS technical specifications; i.e., the SFP water level as low as 23 feet 3 inches and the SFP bulk water temperature as high as 140 degrees Fahrenheit. The normal operating conditions of the SFP are the water level above 37 feet and the bulk water temperature below 90 degrees Fahrenheit. The results of the analysis for an initial temperature of 140 degrees and the initial water level of 23 feet and 3 inches indicated that it would take over 15 days, after bulk boiling begins, for boiling to reduce the level of water in the SFP to the top ' Sacramento Municipal District Utility letter to NRC, dated April 1, 1993. l i 1
" of the spent fuel assemblies. This analysis did not take credit for any convective cooling by the ventilation system. This period of 15 days did not include the time required to raise the SFP bulk water temperature from its initial temperature of 140 degrees Fahrenheit to 212 degrees Fahrenheit, the boiling point of water at standard conditions. Boiling of the water will not damage the fuel assemblies, because these fuel assemblies are designed to function at temperatures much higher than 212 degrees Fahrenheit. The staff, using very conservative assumptions, calculated that it would take at least 3 days to increase the SFP bulk temperature from 140 degrees Fahrenheit to 212 degrees Fahrenheit, with the initial SFP level at 23 feet and 3 inches. The assumptions the steff made to calculate SFP heat-up time to reach the bulk boiling temperature were the following: (1) The energy addition rate to the SFP water was assumed to be ( 1.68E6 BTU /HR. Thisenergyratpwasbasedonthedecayenergyin i the fuel as of November 1, 1991. The current decay energy of the fuel is actually less than the value used because of the additional 1\\ year-period the fuel has been stored in the SFP. The use of the November 1,1991, decay rate artificially increased the amount of energy assumed to be added to the SFP water. This assumption was conservative because the net effect was to increase the rate the StP temperature was raised. j (2) The boundary of the water volume in the SFP was assumed to be perfectly insulated. That is, ng energy was allowed to be transferred from the SFP water volume to (or through) the walls, floor, or surface of the SFP. The energy that would have escaped from the SFP water was assumed artificially to be absorbed by the SFP water, increasing the SFP water temperature. This assumption was conservative because the net effect was to increase the rate the SFP temperature was raised. (3) No credit was given for the energy absorption capability of the metal located in the SFP (fuel storage racks). This metal would act as a heat sink during a SFP water heat-up evolution (from 140 degrees Fahrenheit to 212 degrees Fahrenheit). The energy that would be absorbed by the metal was artificially assumed to be absorbed by the SFP water and increased the SFP water temperature. j This assumption was conservative because the net effect was to increase the rate the SFP temperature.was raised. (4) The volume of the fuel and fuel racks were assumed to displace 20 percent of the water in the SFP. This displaced volume was not included in the calculation, reducing the mass of SFP water available to absorb the energy added to the SFP water. This assumption increased the amount of energy absorbed per unit mass of SFP water, which increased the rate the SFP water temperature was raised. l 2Sacramento Municipal District Utility letter to NRC, dated November 19, 1991, Attachment II at page 54. i
r 4 (5) The loss of water mass from the SFP due to evaporation during the heat-up evolution was not included in this calculation. It was the technical judgment of the staff that the added complexity to model this evaporation phenomenon (mass and associated energy loss) was not necessary because the net effect on the heat-up time would be minimal due to the counteracting effects of mass loss and associated energy loss. With each unit of SFP water mass lost due to evaporation, an associated amount of energy will also be removed from the remaining SFP water volume. The net effect of mass and energy removal would be to reduce the rate at which the SFP water temperature would rise. Additionally, the licensee calculated that approximately 1 foot of SFP water level would evaporate every 70 hours if the SFP water temperature was at a steady-state temperature of 180 degrees Fahrenheit. This steady-state temperature is approximately the mean 1 temperature for the heat-up evolution considered. This further supports the decision not to include the evaporation phenomenon because of the minimal effect on the time to raise the SFP water temperature from 140 degrees Fahrenheit to 212 degrees Fahrenheit. SMUD also has procedures that address loss of offsite power to RSNGS. SMUD has equipment onsite (a diesel powered fire pump) which may be used to add water to the SFP, if necessary, even during periods when offsite power is ~ unavailable. The 18-day period provides ample time to take corrective action even with a complete loss of offsite power. The staff has concluded that the complete loss of offsite power to RSNGS during decommissioning will not significantly impact the health and safety of the public. The staff bases its conclusion on the considerable length of time available for SMUD to implement its loss of offsite power procedures to either restore offsite power or take corrective measures such as adding water to the SFP with existing equipment which does not need offsite power to function. i b L P t l l 1 I}}