ML19031A658
| ML19031A658 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 03/30/1997 |
| From: | Librizzi F Public Service Electric & Gas Co |
| To: | Lear G Office of Nuclear Reactor Regulation |
| References | |
| Download: ML19031A658 (8) | |
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Publi-e-Service Electric-and Gas Company 80 Park Place Newark, N.J. 07101 Phone 201 /622-7000
- ~~.. #:-
March 30~ 1977
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Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Attention:
Mr. George E. Lear, Chief Operating Reactors Branch 3 Division of Operating Reactors Gentlemen:
REFUELING ACCIDENT INSIDE CONTAINMENT NO. 1 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-272 PSE&G was requested by your letter of January 18, 1977 to provide an evaluation of the potential consequences of a refueling accident inside the No. 1 Unit Containment Building of the Salem Nuclear Generating Station.
Your request specified that the evaluation should consist of two parts: (1) a conservative analysis using parameters (e.g. maximum allowable valve closure times) as limited by the technical specifications and (2) an analysis using para-meters associated with current known facility operating conditions (e.g., actual valve closure times).
An evaluation of a postulated refueling accident inside the con-tainment was performed for two cases and is provided in Enclosure
- 1.
The assumed parameters for both cases are identical, with the exception of the source terms.
Case 1 utilizes Regulatory Guide 1.25 calculated source terms, while Case 2 utilizes "expected 11 fuel rod gap activities.
Maximum allowable isolation valve closure times, as limited by the technical specifications, were utilized in both cases analyzed, since the difference between actual and maxi-mum allowable closure times is relatively short and has been deter-mined to be insignificant.
Supplemental information is provided in Enclosure 2 in response to guidance received from pr~vious discussions with members of your staff.
The Energy People 95-2001 300M 8*75 3/30/77 This evaluation clearly indicates that potential site boundary radiation exposures resulting from the postulated refueling accident are well within 10 CFR Part 100 guidelines.
Should you have any additional concerns regarding this evaluation, please do not hesistate to contact us.
Very truly yours, F. ~~
General Manager - Electric Production 3Cl 63/64
ENCLOSURE l REFUELING ACCIDENT INSIDE CONTAINMENT NO. 1 UNI'r SALEM NUCLEAR GENERATING STATION (All Doses in rem)
DOSE CASE 1 Minimum Exclusion Distance Whole Body Thyroid Low Population Whole Body Thyroid Assumptions Both Cases 5.20E-3 3.22E-l Zone Distance 4.77E-4 2.95E-2 Containment Isolation Time: 12 seconds CASE 2 6o55E-4 4.13£-2 5.94E-5 3.78E-3 10 sec. Sample Travel Time (Time for activity to travel from isokinetic sampler in the plant vent to the radiation monitors RllA-Particulate, Rl2A-Radiogas and Rl2B-Iodine) 2 sec. Technical Specifications Isolation Time Activity Release Assumes Mixing with~n <3ontaiTull.ent (Free Volume of Containment 2.62 x 10 ft )
No Charcoal Filtration Maximum Purge Rate: 35,000 CFM Radiation Monitor in Plant Vent Sampling Mode 100 hro Hold-up Time Before Fuel Transfer (Technical Specification Requirement)
Meteorology as Calculated by NRC Staff in Salem Safety Evaluation Report, Section 2.3, October 11, 1974 Minimum Exclusion Distance - 1,270 meters Low Population Zone Distance -
8,000 meters Semi-infinite Cloud Dose Model as Defined in Regul~tory Guide 1. 25.
,_.;/'"
Cont.
.. Case 1 Fuel rod gap activities calculated using assumptions provided in Regulatory Guide 1.25 (e.g. Axial Peaking Factor of 1.72, fuel rod ~ 23 feet below pool surface, ratio of gap activity to total: Kr-85.3:1, all other noble gases.1:1, iodines.1:1, 17 x 17 array, end of fuel cycle) other assumptions and a table of the calculated activities provide in Appendix I of the Salem FSAR.
Case 2 Expected maximum fuel rod gap activities (e.g. Axial Peaking Factor 1.47, fuel rod~ 23 feet below pool surface, fraction of noble gases and iodines in gap listed in Salem FSAR, Table 14. 3-2C, 17 x 17 array, end of fuel cycle, other assumptions listed in Table 14.3-2C of the Salem FSAR).
RFY:peg
. 3/23/77 lCl 07/08
_ENCLOSURE 2 A
INFORMATI NEEDED TO-EVACOATECONTAINMENl..,SOLATION CAPABILITY 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. 1)
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 cl ass~ redundancy, pmver sources, and technical specifi ca ti on 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.l) 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 pa~t of the Containment Purge and Pressure-Vacuum Relief System.
These valves are designated as lVCl, 1VC2, 1VC3, 1VC4, 1VC5 and iVC6 on Figure 5.3-1 of the Salem FSAR.
The Pressure-Vacuum Re"lief Sys tern 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 atmosphere 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:
Background (CPM)
Alarm Setpoint (CPM)
Sensitivity
( CPM/uci I cc)
Particulate 1000 7000 4.4xl0 11 Monitor Type Gas 839 30,000 6
- 2. 1 x 10 Iodine 180/Mi n 15,000 3.0 x 109 The particulate, gas and iodine monitors are designed and qualified for Seismic Cl ass I service.
The containment and pl ant 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 will also automatically initiate isolation. Additionally, two source range neutron flux monitors are
- A.1)
Cont'd.
required to be in service during the refueling operation, and the operator is provided vJitn control room indication of the two area radiation monitors located inside the Containment Building on Elevation 130 1
The Technical Specifications require 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 time 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 receipt of an isolation signal.
Q.3)
Provide 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 flm\\ls.
Conservatively assume that the FHA is a puff release closest to an exhaust gri 11.
A.3)
It was 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 V\\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' 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, hovJever, 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 + 10 + 2) seconds elapse from the time that the 11 puff 11 enters the Containment fan Coolers until the containment isolation valves are completely closed.
The plant vent continues upward along the side of the Containment Bu"ilding above Elevation 195' 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 fro~ the time where the 11 puff 11 enters the conta"inment fan cooler inlet inside of the Contain-ment until it is discharged from the plant vent.
Since purge flow from the Conta.inmflnt ~s 35~000 CFM~ the maximum VQlume of radioactive air released is 5075 ft.~ due to purge fl6w and 3923 ft.~ due to the amount of air still in the ventilation ductvwrk 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
3 -
A. 3)
Cont 1 d.
This analysis is based on average velocity; *the maximum peak centerline velocity is equal to the average velocity since turbulent flow is present throughouf 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 arrangem.ent 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 207622-A-8851 207623-A-885 l 207624-A-8851 207635-A-8826 207636-A-8826 207637-A-8826 223101-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' Reactor Containment -
Ventilation~ Plan El. 130 1 & Above Reactor Containment - Ventilation - Plan Below El. 130 1
Reactor Containment - Ventilation - Sections Auxi 1 i ary 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 Q.5)
If the summation of the instrument response time (question l.e) and valve closure time (question 2) is greater than the gas transit time (question 3),
provide an analysis as to the volume and amount of radioactive exhaust air
\\'>'hich could be released.
Your res'ponse should include the following:
A.)
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 1>Jater temperature on air flow trajectories.
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 only one value for velocity is given since in turbulent flow, the maximum peak velocity equals the average velocity.
- A.5)
Cont'd, Air Cross Transport From
- To Sect. Area Lenqth Flow Rate Velocity Time FC
- Ring Duct 29.4 ft2 195.9 ft 55,000 ft3/min 1870.7 ft/min 6.28 *sec Disch.
Ring Duct
- 36" 36 11 63 11 x50 11 63 11 x50 11 54 11 x60 11 54 11 x60 11
-* 84"x63" 84"x63"- 72"x90 72 11 x90 11 99 11 x72" 92 11 x72 11
- l 02 11 x72" 7.07 ft2
- 21. 9 ft2 22.5 ft2 36.8 ft2 45 ft2 49.5 ft2 51 ft2 26 ft 58 ft 20 ft 18 ft 34 ft 20 ft 63 ft 35,000 ft3/min 4950 ft/min 0.315 35,000 ft3/min 1598 ft/min
- 2. 18 35,000 ft3/min 1555 ft/min
- 0. 771 95,000 ft3/min 2580 ft/min 0.418 95,000 ft3/min 2110 ft/min 0.967 114,490 ft3 /min 2320 ft/min 0.517 114,490 ft3/min 2240 ft/min 1.684 1 time for radioactive particle to reach monitor tubing after entering FCU'l3.132 El.195'-102"x72 11 51 n 2 118 ft 114,490 n 3/min 2240 ft/min 3.15 Total time for radioactive particle to reach plant vent discharge 16.28 sec sec sec sec sec sec sec sec sec sec l*Je 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 calculations 3 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 following 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 filters do not have any pmver source, they are supplied by safety related equiµment.
/\\lso the Technical Specification requires periodic testing to assure operability of the filters.
=13 Se<
= 3 Sec:
.