ML19341C577
| ML19341C577 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Callaway  |
| Issue date: | 02/24/1981 |
| From: | Petrick N STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| SLNRC-81-011, SLNRC-81-11, NUDOCS 8103030770 | |
| Download: ML19341C577 (32) | |
Text
{{#Wiki_filter:.. e'N i g\\7 s ff gI /I SNUPPS j ]/ /J h Standardised Nuclear Unit ,pf Power Mont System q IS8/ g 9 k s. b~1 Nicleofas A. Petrick s ch.h. ch.try R e cd8% Enecutive Director i Roshwille. Meryland 20060 \\ (301)seewoo1o 4 7 T;7 j February 24, 1981 SLNRC 81- 011 FILE: 0541 SUBJ: NRC Request for Additional Information - SNUPPS FSAR l Mr. Harold R. Denton, Director -l Office of Nuclear Reactor Regulation i U. S. Nuclear Regulatory Comission Washington, D. C. 20555 4 I Docket Nos: STN 50-482, STN 50-483, and STN 50-486 i
Reference:
NRC (Tedesco) letter to J. K. Bryan and G. L. Koester dated January 7,1981 1
Dear Mr. Denton:
The referenced letter requested additional information regarding the SNUPPS FSAR. The enclosure to this letter provides the requested information. This information will be incorporated into the next j
revision to the FSAR.
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Ver truly yours, b RN-
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450.09 The following information is currently missing (15.6.3) from the Callaway FSAR and is needed to ccmplete our review.
For the steam generator tube rupture accident provide the following figures:
(1)
SGTR break flow rate vs Time (2)
SGTR integrated tube leak mass vs Time (3)
Primary system pressure vs Time (4)
Secondary system pressure vs Time (5)
PORV flow rate vs Time (6)
MS Safety valve flow rate per steamline vs Time (7)
Atmospheric dump valve flow rate vs time (8)
Steam generator steaming rate vs Time (9)
Reactor coolant temperature vs Time (10) Feedwater flow rate into the steam generators vs Time (11) Water level in the affected steam generator relative to the top of the tube bundle vs Time.
Also, provide the mass of secondary coolant in a steam generator.
RESPONSE
Refer to revised Section 15.6.3. In view of the fact that the steam generator tube material is Inconel-600 and is a highly ductile material, it is considered that the assumption of a complete severance is somewhat con-servative. The more probable mode of tube failure would be one or more minor leaks of undetermined origin. Activity in the steam and power conversion system is subject to continual surveillance, and an accumulation of minor leaks which exceed the limits established in the Technical Specifications is not permitted during the unit operation. The recovery sequence for a SGTR is discussed in Section 15.6.3.2.
The operator is expected to determine that a SGTR has occurred and to identify and isolate the affected steam generator on a restricted time scale to minimize contamination of the secondary system and ensure termination of radioactive release to the atmosphere from the affected unit. The recovery procedure can be carried cut on a time scale which ensures that break flow to the secondary system is terminated before water level in the affected steam generator rises into the main steam pipe. Sufficient indications and controls are provided to enable the operator to carry out these functions satisfactorily. Consideration of the indications provided at the control board, together with the magnitude of the break flow, leads to the conclusion that the accident diagnostics and isolation procedure can be completed within 30 minutes of accident initiation for the design basis event. Assuming normal operation of the various plant control systems, the following sequence of events is initiated by a tube rupture: a. Pressurit - low pressure and low level alarms are Octuated aim 'rging pump flow increases in an attempt to ma pressurizer level. On the sec-ondary side, tnet, ' steam flow /feedwater flow ' edwater flow to the affected mismatch before trip a. steam generator is reduceo ' to the additional break flow, which is now bein,
- ptied to that unit.
b. Decrease in RCS pressure (Figure 15.6-3a) due to continued loss of reactor coolant inventory leads to a reactor trip signal generated by low pressurizer pressure overtemperature at. Resultant plant cool-down (Figure 15.6-3B) following reactor trip leads to a rapid change of pressurizer level (Figure 15.6-3E), and the safety injection signal, initiated by low pressurizer pressure, follows soon after the reactor trip. The safety injection signal automatically terminates normal feedwater supply and initiates auxiliary feedwater addition. c. The steam generator blowdown liquid monitor and/or the condenser offgas radiation monitor will alarm, indicating a sharp increase in radioactivity in the secondary system, and will automatically terminate steam generator blowdown. d. The reactor trip automatically trips the turbine, and if offsite power is available the steam dump valves open, permitting steam dump to the condenser. In the event of a coincident station blackout (loss of offsite power), as assumed in the transients presented in this section, the steam dump valves would auto-matically close to protect the condenser. The steam generator pressure (Figure 15.6-3C) would rapidly increase, resulting in steam discharge to the atmo-sphere through the steam generator safety power-
operated relief valves. In Figure 15.6-3F, the steam flow is presented as a function of time. The flow is constant initially until reactor trip, followed by turbine trip, which results in a large decrease in flow, but a rapid increase in steam pressure to the safety / relief valve setpoints. ,Following reactor trip, the continued action of e. auxiliary feedwater supply and berated safety in-jection flow (supplied from the refueling water storage tank) provide a heat sink which absorbs the decay heat. f. Safety injection flow results in increasing the pres-surizer water level (Figure 15.6-3E); the rate of which depends upon the amount of operating auxiliary equipment. 15.6.3.2 Analysis of Effects and Consequences Method of Analysis Mass snd energy balance calculations are performed to determine primary-to-secondary mass release and to determine the amount of steam vented from each of the steam generators, using LOFTRAN (Ref. 1). In estimating the mass transfer from the RCS through the broken tube, the following assumptions are made: a. Reactor trip occurs automatically as a result of low pressurizer pressure or overtemperature At. Loss of offsite power occurs at reactor trip. b. Following the initiation of the safety injection signal, two centrifugal charging pumps are actuated and are assumed in the analyses to continue to deliver flow for 30 minutes. c. After reactor trip, the break flow reaches equilibrium when incoming safety injection flow is balanced by outgoing break flow, as shown in Figure 15.6-3. The rasultant break flow is assumed to persist from plant trip until 30 minutes after the accident. No operator actions are assumed. d., The steam generators are controlled at the power-operated relief valve setting. The operator identifies the accident type and termin-e. ates break flow to the affected steam generator within 30 minutes of accident initiation. The above assumptions, suitably conservative for the design basis tube rupture, are made to maximize doses and do not explicitly model operator actionr for recovery.
Key Recovery Sequence The recovery sequence to be followed consists of the following major operator actions: a. Identification of the faulted steam generator b. Isolation of the faulted steam generator Assuring subcooling of the RCS fluid to approximately c. 50 F below no load temperature d. Controlled depressurization of the RCS to a value equal to the faulted steam generator pressure Subsequent termination of safety injection flow e. Results Figure 15.6-3 illustrates the flow rate that would result through the ruptured steam generator tube. The previous assumptions lead to an estimate of 89,240 pounds for the total amount of reactor coolant transferred to the secondary side of the faulted steam generator as a result of a tube rupture accident. The integrated steam flow is 61,700 pound released through the safety / power-operated relief valves. The folloving is a list of figures of pertinent time dependent parameters: Figure 15.6-3a - Core Pressure Figure 15.6-3b - Reactor Coolant System Temperature Figure 15.6-3c - Steam Generator Pressure (Faulted Steam Generator) Figure 15.6-3d - Steam Generator Temperature (Faulted Steam Generator) Figure 15.6-3e - Pressurizer Water Volume Figure 15.6-3f - Steam Generator Flow (Faulted Steam Generator) Figure 15.6-3g - Feedwater Flow to Faulted Steam Generator Figure 15.6-3h - Faulted Steam Generator Safety / Relief Valve Flow Rate Figure 15.6-3i - {aultedSteamGeneratorBreakFlowRate Figure 15.6-3j - Steam Generator Mass Figure 15.6-3k - Faulted Steam Generator Liquid Volume
The DNB calculations performed with LOFTRAN (Ref.1) indicate that DNB limits are met. In Table 15.6-1, the sequence of events are presented. These events are the normal plant response to the normal plant setpoints. Loss of offsite power at reactor trip and no operator actions were assumed. 15.6.3.3 ' Radiological Consequences 15.6.3.3.1 Method of Analysis 15.6.3.3.1.1 Physical Modr' The evaluation of the radiological consequences due to a pos-tulated steam generator tube rupture (SGTR) assumes a complete severance of a single steam generator tube while the reactor is operating at full rated power and a coincident loss of offsite power. Occurrence of the accident leads to an increase in contamination of the secondary system due to reactor coolant leakage through the tube break. A reactor trip occurs auto-matically, as a result of low pressurizer pressure. The reactor trip will automatically trip the turbine. The resulting sharp increase in radioactivity in the secondary system will be detected by radiation monitors (refer to Section 11.5), which will automatically terminate steam generator b. The noble gas activity in the reactor coolant is based on 1-percent failed fuel, as provided in Table 11.1-5. c. The secondary coolant activity is based on the dose equivalent of 0.1 pCi/gm of I-131. The following assumptions and parameters are used to calculate the activity released and the offsite doses following an SGTR: The total amount of discharge of reactor coolant into a. the.econdary system through the rupture is 89,240 l s pounds. b. It is assumed that 17 percent (Ref. 25) of the reactor coolant that flows to the affected steam generator flashes to steam and is immediately released to the environment. No credit has been taken for scrubbing. c. The 1-gpm primary-to-secondary leak is assumed to occur to the unaffected steam generators. d. All noble gas activity in the reactor coolant which is transported to the secondary system via the tube rupture and the primary-to-secondary leakage is assumed to be immediately released to the environment.
e. At 30 minutes after the accident, it is assumed that the RCS and steam generator pressures are equalized and below the steam generator relief valve set pressure. Thus, the affected steam generator has been totally isolated, and all primary-to-secondary leakage has been terminated to the unaffected steam generators. f. The iodine partition faction between the liquid and steam in the steam generator is assumed to be 0.01. g. The steam releases from the steam generators are assumed to be as follows: Affected steam generator (0-30 mins) 61,700 lbs Unaffected steam generator (0-2 hrs) 454,000 lbs Unaffected steam generaeor (2-8 hrs) 1,200,000 lbs h. Offsite power is lost. i. Eight hours after the accident, the RHR system is assumed to be in operation to cool down the plant. Thus, no additional steam release is assumed. j. Radioactive decay during the release of activity is conservatively ignored for all isotopes, except for the following which have extremely short half-lives: Kr-89, Xe-137, Xe-135m, and Xc-138. No decay during transit or ground deposition was considered. 18. Letter NS-TMA-2014, dated December 11, 1978, Anderson, T. M. (Westinghouse) to Tedesco R. L. (NRC). 19. Johnson, W. J. and Thompson, C. M., " Westinghouse Emergency Core Cooling System Evaluation Model - Modified October 1975 Version," WCAP-9168 (Froprietary) and WCAP-9169 (Non-Proprietary), September 1977. 20. " Westinghouse ECCS Evaluation Model Sensitivity Studies," WCAP-8341 (Proprietary) and WCAP-8342 (Non-Proprietary), July 1974. 21. Salvatori, R., " Westinghouse ECCS - Plan Sensitivity Studies, WCAP-8340 (Proprietary) and WCAP-8356 (Non-Proprietary), July 1974. 22. Johnson, W. J., Massie, H. W., and Thompson, C. M., " Westinghouse ECCS-Four Loop Plant (17x17) Sensitivity Studies," WCAP-8565-P-A (Proprietary) and WCAP-8566-A (Non-Proprietary), July 1975. 23. Letter NS-TMA-2030, dated February 12, 1979, Anderson, T. M. (Westinghouse) to Denton, H. R. (NRC). 24. DiNunno, J. J., et al., " Calculation of Distance Factors for Power and Test Reactor Sites," TID-14844, Division of Licensing and Regulation, AEC, Washington, D.C., 1962.
25. USNRC NUREG-0409, " Iodine Behavior in a PWR Cooling System Following a Postulated Steam Generator Tube Rupture Accident," by Postma A. K. and Tam P. S., dated January 1978. TABLE 15.6-1 TIME SEQUENCE OF EVENTS FOR INCIDENT WHICH RESULTS IN A DECREASE IN REACTOR COOLANT INVENTORY Time Accident Event (sec) Inadvertent opening of a pressurizer safety valve Safety valve opens fully 0.0 Overtemperature AT reactor trip setpoint reached 13.8 Minimum DNBR occurs 15.0 Rods begin to drop 15.8 Steam generator tube rupture Tube rupture occurs 0.0 Reactor trip signal 253.1 Rod motion 255.1 Feedwater terminated 255.1 Steam generator safety / relief valves opened 258.0 Safety injection signal 434.6 Safety injection 459.6 Auxiliary feedwater injection 496.0 Operstor takes actions to isolate and cooldown 1800.0
TABLE 15.6-4 PARAMETER USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A STEAM GENERATOR TUBE RUPTURE (SGTR) I. Source Data
- a. ' Core power level, MWt 3,565 b.
Steam generator tube 1 leakage, gpm c. Reactor coolant iodine activity: 1. Case 1 Initial activity equal to dose equivalent of 1.0 pCi/gm of I-131 with an assumed iodine spike that increases the rate of iodine release into the reactor coolant by a factor of 500 2. Case 2 ?.n assumed pre-accident iodine spike, which has resulted in the dose equivalent of 60 pCi/gm of I-131 d. Reactor coolant noble gas Based on 1-percent failed activity, both cases fuel as provided in Table 13.1-5 e. Secondary system initial Dose equivalent of activity 0.1 pCi/gm of I-131 f. Reactor coolant mass, lbs 5.3E+5 g. Steam generator mass l'.089E+5 (each), lbs h. Offsite power Lost i. Primary-to-secondary 30 minutes leakage duration II. Atmospheric Dispersion Factors See Table 15A-2 III. Activity Release Jata a. Affected steam generator 1. Reactor coolant dis-charged to steam gener-ator, lbs 89,240 i m
TABLE 15.6-4 (Sheet 2) 2. Flashed reactor coolant, 17 percent 3. Iodine partition factor 1.0 for flashed fraction of reactor coolant 4. Total steam release, 61,700 lbs 5. Iodine partition factor 0.01 for the nonflashed fraction of reactor coolant that mixes with the initial iodine activity in the steam generator 6. Isolation time, mins 30 b. Unaffected steam generators 1. Primary-to-secondary 250 leakage, lbs 2. Flashed reactor coolant, 0 percent 3. Feedwater flow rate, lbs 0-2 hours 441,000 2-8 hours 1,253,000 4. Total steam release, lbs 0-2 hours 454,000 2-8 hours 1,200,000 5. Iodine partition factor 0.01 6. Isolation time, hrs 8 c. Activity released to the environment
- 1.
Case 1 \\ Isotope 0-2 hr (Cil 0-8 hr (Ci) I-131 6.53E+1 6.58E+1 I-132 1.12E+2 1.12E+2 I-133 1.28E+2 1.29E+2 I-134 1.34E+2 1.34E+2 I-135 1.07E+2 1.07E+2 Xe-131m 8'.10E+0 8.10E+0 Xe-133m 4.44E+1 4.44E+1 Xe-133 2.21E+3 2.21E+3 Xe-135m 2.70E+0 2.70E+0 Xe-135 1.34E+2 1.34E+2 Xe-138 8.20E+0 8.20E+0 Kr-83m 1.02E+1 1.02E+1 Kr-85m 4.78E+1 4.78E+1
l l l TABLE 15.6-5 j y l RADIOLOGICAL CONSEQUENCES OF A t l STEAM GENERATOR TUBE RUPTURE
- Doses (rem)
Callaway Wolf creek 1. Case 1 Exclusion Area Boundary (0-2 hr) Thyroid, rem 9.7E+0 7.3E+0 Whole body, rem 5.6E-2 4.2E-2 Low Population Zone Outer Boundary (duration) Thyroid, rem 1.3E+0 1.0E+0 Whole body, rem 7.3E-3 5.9E-3 2. Case 2 Exclusion Area Boundary (0-2 hr) Thyroid, rem 5.5E+1 4.lE+1 Whole body, rem 9.3E-2 7.0E-2 Low Population Zone Outer Boundary (duration) Thyroid, rem 7.1E+0 5.7E+0 Whole body, rem 1.2E-2 1.0E-2
- The effects of the revised primary-to-secondary mass transfer and the effects of the revised secondary side release on the activity released to the environment are being evaluated.
The cevised activity releases and radiological consequences of a steam generator tube rupture will be provided in a subsequent revision of the FSAR. gmy .g 6+ +me e-
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