ML19031A495
| ML19031A495 | |
| Person / Time | |
|---|---|
| Site: | Salem  (DPR-070) |
| Issue date: | 04/28/1977 |
| From: | Mittl R Public Service Electric & Gas Co |
| To: | Parr O Office of Nuclear Reactor Regulation |
| References | |
| Download: ML19031A495 (7) | |
Text
{{#Wiki_filter:Public Service Electric and Gas Company 80 Park Place Newark, N.J. 07101 Phone 201 /622-7000 April 28, 1977 File Cy;-" .....;:/.. :.,;.*~ '" --L-!t Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention~ Mr. Olan D. Parr, Chief Gentlemen: Light Water Reactors Branch III Division of Project Management REFUELING ACCIDENT INSIDE CONTAINMENT NO. 2 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-311 PSE&G was requested by your letter of March 16, 1977 to pro-vide an evaluation of the potential consequences of a refueling accident inside the No. 2 Unit Containment Building of the Salem Nuclear Generating Station. Your request ~pecified that our evaluation should utilize assumptions comparable to those given in Regulatory Guide 1. 25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors," assuming the worst single failure. This evaluation is provided in Enclosure 1 and clearly indicates that potential site boundary radiation exposures resulting from the postulated refueling accident are well within 10CFR Part 100 Guidelines. provides responses to your specific questions con-cerning containment isolation capability during a refueling accident. It should be noted that these responses address the Salem No. 1 Unit (Docket No. 50-272) and were submitted on March 30, 1977. This information has been provided since the No. 2 Unit equipment involved is similar to that installed in The Energy People 95-2001 300M 8-75
Director of Nuclear Reactor Regulation 4/28/77 No. 1 Unit and it is therefore expected that the Technical Specification requirements will also be similar. The enclosed drawings, however, are specifically for the No. 2 Unit. Should you have any additional concerns regarding this evalua-tion, please do not hesitate to contact us. Encl. lFl 13/14 fJ;;;rs, R. L. Mi~il General Manager - Licensing and Environment Engineering and Construction
ENCLOSURE 1 REFUELING ACCIDENT INSIDE CONTAINMENT NO. 2 UNIT SALEM NUCLEAR GENERATING STATION (All Doses in rem) Minimum Exclusion Distance 1270 meters Whole Body Dose .98 Thyroid Dose 59.9 Assumptions Low Population Zone Distance 8045 meters .089 5.49 100 hr. Hold-up time Before Fuel Transfer (Proposed Technical Specification Requirement) Meteorology as Calculated by NRC Staff in Salem Safety Evaluation Report, Section 2.3, October 11, 1974 Semi-infinite Cloud Dose Model as Defined in Regulator Guide 1. 25. No isolation 3 Breathing Rate = 3.47E-4 m /sec No credit for charcoal filtration Iodine DF = 100 Fuel rod gap activities calculated using assumptions provided in Regulatory Guide 1.25: a) Axial peaking factor of 1.72 b) Fuel Rod > 23 feet below pool surface c) Ratio of gap activity to total KR85. 3: 1 All other noble gaseous.1:1 Iodines.1:1 d) 1 7 x 1 7 array e) End of fuel cycle (650 days @ 3558 MWt) f) Highest rated discharged assembly g) Release of all gap activity in damaged assembly(other assumptions and a table of the calculated activities provided in Appendix!of the Salem FSAR) RFY:pd 2-22-77 2Dl 45
e ENCLOSURE 2 A INFORMATION NEEDED TOEVALUAITCONTAINMENT~OLATION CAPAeILITY DURING REFUELING ACCIDENT The following are the responses to USNRC questions pertaining to the evaluation of a fuel handling accident (FHA) inside the Containment Building. Q.l) Describe all instrumentation which would detect a fuel handling accident (FHA) inside containment. Your responses should include the following information: a) Instrumentation function, e.g., close containment isolation valves; b) Type of instruments and setpoints, e.g., mr/hr~ and normal background reading; c) Safety class, redundancy, power*sources, and technical specification requirements; d) A description of instrument response following a FHA taking into account instrument location; e) Response time for the instrument to signal containment isolation after the FHA. A. 1) A fuel handling accident (FHA) inside the Containment would be detected by the containment and plant vent radiation monitors. A high radiation signal from any of these monitors will initiate automatic closure of the contain-ment isolation valves, which are part of the Containment Purge and Pressure-Vacuum Relief System. These valves are designated as lVCl, 1VC2, 1VC3, 1VC4, lVC5 and lVC6 on Figure 5.3-1 of the Salem FSAR. The Pressure-Vacuum Relief System serves to limit differentials between the Containment Building pressure and atmospheric pressure, whereas the Containment Purge System serves to supply fresh air to and vent the Containment atmosphere. The radiation monitors can monitor either the Containment atm~sphere or the plant vent; automatic closure of the Containment isolation valves will occur when a high radiation alarm is received from the selected source, although the plant vent is monitored whenever the Containment is being purged. The Containment/Plant Vent Monitoring/Sampling System consists of three (3) separate radiation monitors--a particulate monitor, a gaseous monitor, and an iodine monitor. The pertinent information associated with each of these radiation monitors is as follows: Monitor T:r~e Particulate Gas Iodine Background (CPM) 1000 839 180/Mi n Alarm Setpoint ( CPM) 7000 30,000 15,000 Sensitivity 4.4 x 1011 2.1 x 10 6 3.0 x 10 9 ( CPM/uci I cc) ~ The particulate, gas and iodine monitors are designed and qualified for Seismic Class I service. The containment and plant v~nt radiation monitors and the sampling system are connected to vital power sources. Although the system is not redundant, assurance of function is provided in that the system logic is designed such that any one of the three (3) monitors will -initiate isolation. Loss of power to any monitor v1ill also c:.utomatically initiate isolation. Additionally, two source range neutron flux monitors are A.l) Cont'd. required to be in service during the refueling operation, and the operator is provided with control room indication of the two ar~a radiation monitors located inside the Containment Building on Elevation 130 1 The Technical Specifications req0ire that the Containment Purge and Pressure-Vacuum Relief Isolation System be operational during refueling operations. Q.2) Describe the response of the containment isolation valves following the FHA. Include valve closure times including expected valve closure time as well as Technical Specification requirements. A.2) The response time for Containment Building ventilation isolation has been determined during pre-operational testing. The results obtained do* not differ significantly from Technical Specification requirements. The protection logic response ti~e is in the order of 0.02 seconds. The response time for the initiation of ventilation isolation after a high radiation alarm is 0.10 seconds. The isolation valves close within two (2) seconds of re~eipt of an isolation signal. Q.3) Ptovide the transit time from the point where a monitor can respond to a release from the FHA to the inboard.isolation valve based on the maximum air velocity (peak centerline velocity) at maximum exhaust flow. Also include the transit time based on average velocity and normally expected air flows. Conservatively assume that the FHA is a puff release closest to an exhaust gri 11. A.3) It vrns assumed that the fuel handling accident (FHA) was a 11 puff 11 release as close as possible to an exhaust grill, i.e., a Containment Fan Cooler Unit intake. It \\\\lould take 13 seconds for the 11 puff 11 of radioactivity to travel from the Containment Fan Cooler Unit inlet through the ventilation ductwork and to Elevation 195 1 in the plant vent where the radiation monitoring sampling line inlet is located; this time lapse has been calculated assuming that only the Fan Cooler Unit furthest from the Containment Purge line inlet is operating, however, the results of the analysis are independent of both this assumption and of the mass transit from pool surface to the Fan Cooler Unit duct inlet. It has also been calculated *that it would take approximately 10 seconds for the radioactive 11 puff 11 to reach the radiation monitor from the sampling line inlet~ and approximately 2 seconds (see response to Question 2) for the con-tainment isolation valves to close. Therefore~ it has been estimated that 25 ( 13 + l 0 + 2) seconds el apse from the time that the 11 puff 11 enters the Containment Fan Coolers unti1 the containment isolation valves are completely c 'losed. The plant vent continues upward along the side of the Containment Building above Elevation 195 1 where the radiation monitor inlet tubing is located. It will take approximately 3 seconds for the gas to go from Elevation 195 1 to the plant vent discharge. Therefore, 16 (13 + 3) seconds elapse from the time where the 11 puff 11 enters the containment fan cooler inlet inside of the Contain-ment until it is discharged from the plant vent. Since purge flow from the Containm~nt is 35,000 CFM, the maximum vslume of radioactive air released is 5075 ft.~ due to purge flow and 3923 ft. due to the amount of air still in the ventilation ductwork up to the plant vent discharge after the isolation valves are closed. Therefore the total amount of radioactive air released to the atmosphere is 5075 + 3932 = 9007 ft.J
I 3 - Cont'd. This analysis is based on average velocity; the maximum peak centerline velocity is equal to the average velocity since turbulent flow is present throughout the ventilation ductwork. Q.4) Provide drawings of the containment which clearly show the location of the radiation monitors relative to the ventilation exhaust system including all exhaust inlets and duct arrangement up to the outboard isolation valves. Q.4) The following is a list of drawings that show the portion of the ventilation system which is pertinent to our analysis: A.4) Drawing Number Title Q.5) 207622-A-8851 207623-A-8851 207624-A-8851 207635-A-8826 2076 36-A-8826 207637-A-8826 2231 Ol-A-8989 223102-A-8989 223103-A-8989 223104-A-8989 239634-A*-l 520 Auxiliary Building - Ventilation Ducts - El. 100' Auxiliary Building - Ventilation Ducts - El. 122 1 Auxiliary Building - Ventilation Equip. & Ducts - El. 122 1 Reactor Containment - Ventilation - Plan El. 130' & Above Reactor Containment - Ventilation - Plan Below El. 130 1 Reactor Containment - Ventilation - Sections Auxiliary Building - Pl ant Vent - Sheet 1 Auxiliary Building - Plant Vent - Sheet 2 Auxiliary Building - Plant Vent - Sheet 3 Auxiliary Building - Plant Vent - Sheet 4 North Penetration Area - El. 78 1 , 100 1 130 1 and Roof Above 100 1 - External Tubing - Radiation Monitoring If the summation of the instrument response time (question l.e) and valve closure time (question 2) is greater th.an the gas transit time (question 3), provide an analysis as to the volume and amount of radioactive exhaust air which could be released. Your response should include the following: a) Duct sizes b) Maximum (peak) air velocity c) Average air velocity d) Containment isolation valve closure characteristics e) Exhaust system flow rates f) Methodology used to calculate gas transit times from the pool surface to the inlet to the exhaust system g) Air velocity profiles over the pool surface. You should consider the effects of pool water temperature on air flow trajectories. A.) The following is a.tabulation of the duct sizes, velocities and flow rates, etc. for that portion of the ventilation system pertinent to the analysis of
- a. FHA inside the Containment Building.
Note that on*ly one value for velor:ity is given since in turbulent flow, the maximum peak velocity equals the average velocity.
I l-A.5) Cont 1d. Air Cross Transport From - To Sect. Area Length Flow Rate Velocit,y: Time FC - Ring Duct 29.4 ft2 195.9 ft 55,000 ft3/min 1870.7 ft/min 6.28 sec Disch. Ring f t2 35,000 ft3/min Duct - 36 11 7.07 26 ft 4950 ft/min 0.315 sec 36 11 - 63 11 x50 11
- 21. 9 ft2 58 ft 35,000 ft3/min 1598 ft/min
- 2. 18 sec 63 11 x50 11
- 54. II x60 11 22.5 ft2 20 ft 35,000 ft3/min 1555 ft/min
- 0. 771 sec 54 11 x60 11
- 84 11 x63 11 36.8 f t 2 18 ft 3 95,000 ft /min 2580 ft/min 0.418 sec 84 11 x63 11 72 11 x90 45 ft2 34 ft 95,000 ft 3/min 2110 ft/min 0.967 sec 72 11 x90 11 99 11 x72 11 49.5 ft2 20 ft 114,490 ft3/min 2320 ft/min 0.517 sec 92 11 x72 11 -102 11 x?2 11 51 ft2 63 ft 114,490 ft3 /min 2240 ft/min l.684 sec Total time for radioactive particle to reach monitor tubing after entering FCU'l3.132 sec El.195'-102 11 x72 11 51 ft2 118 ft 114,490 ft 3/min 2240 ft/min
- 3. 15 sec Total time for radioactive particle to reach p 1 ant vent discharge 16.28 sec.
\\~e have not considered the gas transit time from the pool surface to the inlet to the exhaust system nor have we considered the air velocity profiles over the pool surfaces. It is our opinion that this analysis is not required since our system design is such that the time for a puff of radioactive air to go from the pool to the exhaust system is common to both the monitoring system and the purge system. Our calculations3 as seen in answer to Question No. 3 above, estimate that a maximum 9007 ft. of radioactive air is released to the atmosphere if a FHA occurs inside the Containment. Q.6) Describe any charcoal filters which would mitigate the consequence of the FHA. If so, include the follm'ling information: type (e.g., kidney), redundancy, power sources, safety grade, technical specification requirements. A.6) Charcoal filters are provided in both the Containment Iodine Removal System ~ (subsystem of the Containment Ventilation System) and the Auxiliary Building Exhaust System. The Containment Iodine Removal System is contained within the Containment Building and is a recirculation system. The Auxiliary Building system when used in conjunction with the Containment Purge System is a once-through system. Both systems use pleated bed adsorber cells which consist of 1-inch thick charcoal beds. The charcoal filters are designed to Seismic Class I criteria. Although the filte1~s do not have any power source, they are supplied by safety related equipment. i\\lso the Technical Specification requires periodic testing to assure operability of the filters.
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