ML20197A947
| ML20197A947 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 12/27/1983 |
| From: | Brandon R, Gridley R GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML112971110 | List: |
| References | |
| 83NED141, NEDO-30485, NUDOCS 8605120423 | |
| Download: ML20197A947 (16) | |
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- NORTHERN STATES P0 tier COMPANY NEDO-30485 LICENSE AMENDMENT REQUEST DATED MAY 1, 1986 DRF T23-00518 .
EXHIBIT C 83NED141 Class I December 1983 MONTICELLO DESIGN BASIS ACCIDENT CONTAINMENT PRESSURE AND TD'PERATURE RESPONSE FOR FSAR UPDATE Approved by AJ 1
$ J. Brandon, Man
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Application Engine %er ering Approved by: // 12/47/ff R.I/. Gridlef, Manag'er Fuels Services Licensing 8605120423 860501 PDR ADOCK0500g3 P
NUCLEAR ENERGY BUSINESS OPERATIONS + GCNERAL ELECTRIC COMPANY SAN JOSE. CALIFORNIA 95125 GENER AL h ELECTRIC
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Except as otherwise agreed to in writing, neither the General Electric Company nor any of the contributors to this document makes any warranty or representation (express or implied) with respect to the accuracy, completeness, or usefulness of the information contained in this document or that the use of such information may not infringe privately owned rights, nor do they assure any responsibility for liability or damage any kind which may result from the use of any of the information contained in this document.
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TABLE OF CONTENTS'
?.!.51 ABSTRACT v
1.0 INTRODUCTION
1 2.0 ASSUMPTIONS 2 3.0 CONTAINMENT PRESSURE AND TEMPERATURE 5
RESPONSE
4.0 REFERENCES
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t- NEDO-30485 LIST OF TABLES ,
4 Table No. Page 1 Monticello_ Maximum Containment Conditions 7 For a Design Basis Accident LIST OF FIGURES Figure No. Page 1 Monticello Containment Pressure Response 8 2 Monticello Dryvell Temperature Response 9
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3 Monticello Suppression Pool Temperature 10
Response
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s ABSTRACT The containment response to a postulated design basis accident has been determined for the Monticello Nuclear Generating Plant. The assumptions used in the analysis are consistent with the Plant Unique Load Definition and the Safety Analysis Report.
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1.0 INTRODUCTION
The containment response to a postulated loss-of-coolant accident (LOCA) has been determined for the Monticello Nuclear Generating Plant. The LOCA is defined as a double-ended rupture of one of the 28 inch diameter recirculation system pipes.
The major assumptions used in this analysis are listed in section 2.0. They are consistent with the previous Safety Analysis Report (Reference 1) and the Plant Unique Load Definition (PULD) for Monticello (Reference 2).
The results of the analysis are presented in Section 3.0. The .
temperature and pressure curves for both the dryvell and the wetvell are given.
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2.0 ASSUMPTIONS The plant conditions and assumptions used for the containment evaluation in the Monticello FSAR update are given below. The basis for each assumption is also shown.
The purpose of the assumptions is to =axi=1:e the possible long ter= effects for the contain=ent. Otherwise the assumptions that apply to the short ters response (less than 100 seconds) are consistent with the PULD. The long ter= assu=ptions are consistent with the earlier FSAR. .
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NEDO-30485 i.-
ASSUMPTION BASIS .r
- 1. Reactor is operating at 102% of 1. Licensing requirement rated power (1703 MWt).
- 2. Suppression pool and service water 2. Technical Specification temperature are initially 90 F. Maximum.
- 3. Suppression pool vajer volume is 3. Technical Specification initially 68,000 ft Minimum.
4 Dryvell airspace is initially 135 F 4. Nominal temperature, and 20: relative humidity. mini =um humidity
- 5. Weevell airspace is initially 90 F 5. Technical Specification and 100% relative humidity. te=perature, maxi =um
.hu=idity.
- 6. The vetvell airspace is in thermal 6. Bounding for long term equilibrium with the suppression pool. pressure effects
- 7. Dryvell and vetvell airspace are both 7. Maximum during normal initially 1.0 psig. operations.
- 8. Dryvell fan coolers are inactive. 8. Non-safety equipment.
- 9. Control red drive flov is zero. 9. Non-safety equipment.
- 10. Initial downcomer submergence is 10. Technical Specification.
3.58 ft.
- 11. Normal operation of the plant 11. Licensing require =ent.
system is assumed except for a single active failure.
- 12. The May-Witt decay heat curve is used. 12. Accepted by NRC for "
Mark I containment evaluation.
- 13. The RHR pool cooling mode starts at 13. Accepted by NRC for 10 minutes after the line break. Mark I containment evaluation.
- 14. One RHR loop is inoperative. 14. Limiting single
,, failure for contain-ment, i
NEDO-30485 15.
No heat transfer through dryvell &
vetvell valls is used, 15. Conservative for long term temperature
- effects.
16.
Instantaneous guillotine break of recirculation auction line is used. 16. Worst case break.
17.
RER intertie is present during break.
- 17. Monticello config-uration.
18.
The Emergency-Core-Cooling System starts injecting at 30 seconds, 18. Nominal ti=e for assu=ed injection without offsite power.
19.
For first 10 minutes ECCS consists of one core spray and two LPCI pumps. 19. Nominal injection rate.
20.
Pool cooling consists of 2 LPCI pumps with one heat exchanger. 20. Listing single failure of one RHR loop.
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3.0 CONTAINMENT PRESSURE AND TEMPERATURE RESPONSE ,
The containment response to a postulated recirculation line break is illustrated in Figures 1 to 3. To cover the peak pool temperature the analyses are extended out to one million seconds (about 11.6 days).
For this transient the reactor quickly blevs down due to the large break siae. The vessel pressure drops too rapidly for the high pressure coolant injection system to supply any makeup water. At the ti=e of the break it is assumed that the offsite AC power is lost. The e=ergency core cooling system (ECCS) power must, there- ,
fore, be supplied by the diesel generators. This causes a delay of about 30 seconds before rated ECCS flow is available.
Figure 1 shows the expected pressures in the dryvell and pressure suppression cha=ber (vetvell) . After the vessel is flooded. ECCS flow cascades from the break and quenches the steam in the dryvell.
The vetvell-to-dryvell vacuus breakers open and allow air to return to the dryvell. At approximately 105 seconds into the accident the pressures in the two chambers are equal and remain equal throughout the rest of the transient.
The dryvell temperature response is plotted in Figure 2. Figure 3 ,
gives the suppression pool temperature. The vetvell airspace is assumed to be in thermal equilibrium with the suppression pool.
Table 1 lists the maximum pressures and temperatures expected in the containment for this accident. For the dryvell compartment the maxi =um pressure and temperature occur early in the transient during the blevdown phase. At this time the inertial effects of the flow in the vent system cause the dryvell to remain at a much greater pressure than the vetvell. These high pressures are quickly dissipated as can be seen in Figure 1.
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It should be noted that the values in Table 1 are slightly different than those found in NEDO-30477 (Reference 3). This is because the purpose of this report is to maximize the long term temperatures, while the purpose of NEDO-30477 is to maximize short term values. The assu=ption here of mini =um suppression pool mass is the main cause for these differences.
The vetvell experiences its =axi=um pressure and te=perature later in the transient. The suppression pool continues to heat up due to decay heat from the reactor. The heat is transferred to the pool by the steam and ECCS vater coming out of the reactor and flowing through the vent system. Eventually the residual heat re= oval (RHR) system is able to turn the transient around and re=ove the heat from the pool faster than it is added. This turn-around point corresponds to the maxi =um suppression pool te=perature. This transient assu=es that only one loop of the RER is available. If the operator were to use the full RER capacity, then this transient vould turn around sooner, and the =axt=um suppression pool te=pera-ture would be less.
- One option, that the operator has which isn't shown, is to align the RHR in the contain=ent spray mode. This would quench the steam in the contain=ent airspace and rapidly drop the temperature and pressure. It is conservative to neglect this option.
7 The assu=ption that the operator initiates contain=ent cooling at 10 =inutes into the event is also conservative. The high suppression pool te=peratures vould signal the operator that cooling is needed much sooner.
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NEDO-30485 TABLE 1 MONTICELLO MAXIMUM CONTAINMENT CONDITIONS FOR A DESICN 3 ASIS ACCIDENT Maximum Dryvell Pressure = 41.8 psig at 1.2 seconds Maxi =um Dryvell Temperature = 282 F at 2.6 seconds Max 1=um Wetvell Pressure = 32.2 psig at 1100 seconds Maximum Wetvell Temperature = 182 F at 15500 seconds I
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4.0 REFERENCES
- 1. PMonticello Nuclear Generating Plant, Updated Safety Analysis Report " Docket Number 50-263. Rev. 1, October 1982.
- 2. NEDO-24576. Rev.1, " Mark I Containment Program, Plane Unique Load Definition, Monticello Nuclear Power Plant " October 1981.
- 3. NEDO-30477, " Safety Analysis of the RER Intertie Line, Monticello Nuclear Generating Plant," 1983.
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