ML17249A973
| ML17249A973 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 12/31/1979 |
| From: | ROCHESTER GAS & ELECTRIC CORP. |
| To: | |
| Shared Package | |
| ML17249A450 | List: |
| References | |
| TASK-03-11, TASK-3-11, TASK-RR NUDOCS 8001040678 | |
| Download: ML17249A973 (70) | |
Text
DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTAL QUALIFICATION OF EQUIPMENT FOR SPACES/SYSTEMS WHICH MAY -BE USED IN POST ACCIDENT OPERATIONS OUTSIDE CONTAINMENT AT R. E.
GINNA NUCLEAR POWER PLANT ROCHESTER GAS AND ELECTRIC CORPORATION
- DECEMBER, 1979 8001040
TABLE OF CONTENTS 1.0 Introduction 2.0 Summary 3.0 Source Terms and Calculation Methodology 4.0 Review of Areas Requiring Access for Post-Accident Operations 4.1 Area A 4.2 Area B
4.3 Area C
4.4 Area D
4.4.1 Isotopic Analysis 4.4.2 Boron Analysis 4.4.3 Primary Gas Analysis 4.5 Area E
4.6 Area F
4.7 Area G
4.8 Area H
4.9 Area I 4.10 Area J 4.11 Area K
4.12 Areas L and M
4.13 Area N
4.14 Control Room 5.0 Review of Radiation Qualification of Safety Equipment 5.1 Identification and Evaluation of Equipment 5.2 Recommendations 6.0 References
TABLE OF CONTENTS (Cont.)
7.0 Appendix 1:
Liquid and Gaseous Sample Collection and Preparation 7.1 Sample Collection (Nuclear Sample Ro'om) 7.1.1 Liquid,(0 and Radionuclide)
Sample Collection 7.1.2 H2 and Radioactive Sample Collection 7.2 Containment Air Sample Collection 7.3 Sample Preparation (Primary Chemistry Laboratory and Count Room) 7.3.1 Isotopic Analysis of Primary Coolant 7.3.2 Boron Analysis In Primary Chemistry Laboratory 7.3.3 Primary Gas Analysis In Primary Chemistry Laboratory 7.3.4 Hydrogen Analysis In Primary Chemistry Laboratory
0
LIST OF TABLES 3-1 Shielding Source Terms (T=O) 4-1 Access Routes to Vital Areas 4-2 Exposure Rates for Vital Areas as a Function of Time 4-3 Vital Area Radiation Dose Summary 5-1 Equipment Qualification
LIST OF FIGURES 4-1 Reactor Containment Vessel Plan - Above Intermediate Floor 4-2 Reactor Containment Vessel Plan Above Operating Floor 4-3 Auxiliary Building Plan - Basement Floor 4-4 Auxiliary Building Plan - Intermediate Floor 4-5 Auxiliary Building Plan - Operating Floor 4-6 Turbine Building Plan - Operating Floor
INTRODUCTION This report has been prepared in response to the September 13, 1979 letter from Darrel G. Eisenhut of the NRC to all operating nuclear power plants.
This letter presented an implementation schedule for the recommendations presented in "THI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations" (NUREG-0578).
The requirements defined in the September 13th letter were subsequently clarified in a letter from Harold R. Denton to all operating nuclear power plants dated October 30, 1979.
Item 2.1.6.b of NUREG-0578 requires a design review to determine whether post-accident radiation fields unduly limit personnel access to areas necessary for mitigation of,'r recovery from an accident; or unduly degrade the proper operation of safety equipment.
Corrective actions for problems identified as a result of the review are also to be determined.
This report presents the results of such a review for the R. E. Ginna Nuclear Power Plant.
The review was based on the following guidelines:
a)
The post-accident dose rate in areas requiring continuous occupancy should not exceed 15 mr/hr.
b)
The post-accident dose rate in areas which do not require continuous occupancy should be such that the dose to an individual during a required access period is less than 5 rem whole body or its equivalent.
c)
The integrated dose to safety equipment as a result of the accident should be less than the dose for which the equipment has been qualified to ensure that the capability of the equip-ment to perform its safety function has not been degraded.
d)
The minimum radioactive source term used in the evaluation should be equivalent to the source terms recommended in Regulatory Guides 1.3 and 1.4.
SUNDRY This report provides a review of the R. E. Ginna Nuclear Power Plant to determine whether post-accident radiation fields unduly limit personnel access to areas necessary for mitigation of or recovery from an accident or unduly degrade the proper operation of safety equipment.
This review was accomplished using the recom-mendations and guidelines of NUREG-0578 and subsequent clarifications as provided by the NRC.
This evaluation identified the most critical areas requiring personnel access following the onset of extreme accident conditions following a postulated major release of radioactivity into the Containment Building.
Consideration was given to areas where predetermined post-accident functions would be performed (Nuclear Sample Room, Chemistry Laboratory, Count Room, Control Room and air sample penetrations) and to areas where personnel could be called upon to execute certain accident-mitigating or short-term recovery tasks (hydrogen recombiner
- panel, radwaste panel and 'building ventilation filters).
Potential radiation exposures were determined for each task and included addi-tional exposure due to accessing the areas considered.
The review has identified certain areas which may require modifications to existing procedures, shielding, and/or relocation of equipment.
These are discussed in detail in Sections 4.0 and 5.0 and additionally in the RG&E response to Item 2.1.8.a.
The qualification of safety-related equipment has also been reviewed under the assumed accident condition guidance provided in NUREG-0578.
The results of that review are presented in Section 5.2.
We have concluded that the actions which may be taken under the con-servative post-accident conditions assumed in this analysis, can be accomplished without causing undue radiation exposures to personnel.
Where procedural changes and equipment modifications are needed to conform with the dose criteria provided in NUREG-0578, we have identified those changes which are considered most practicable.
It is recognized that in addition to such changes, administrative con-trols would be exercised to further reduce doses to individuals to the lowest practical extent.
We have not identified any safety-related component which would be likely subject to performance degradation, even under the highly conservative accident conditions assumed.
SOURCE TERHS AND CALCULATIONALIIETHODOLOGY SOURCE TERHS The activity assumed for liquid source term calculation is based on 100% of the noble gas inventory, 50', of the halogen core inventory and 1', of all other nuclides in the core inventory.
The activity assumed for gaseous source term calculation is based on 100~ of the noble gas core inventory and 25', of the halogen core inventory.
This approach is, it should be noted, quite conservative.
For ezample, it is impossible to have, the entire noble gas inventory of the core simultaneously in both the liquid and the gas sources.
Two liquid source terms were used in the evaluation.
For systems which contain post accident recirculation fluid, the source term was based on diluting the liquid inventory discussed in the previous paragraph with 303,800 gallons of fluid filling the containment sump from the refueling water storage tank, both accumulators and a
boric acid tank.
These systems include the following:
Residual Heat Removal Containment Spray Recirculation High Pressure Injection Nuclear Sampling (RHR Process Fluid)
For systems which can contain fluid from the reactor coolant system but do not take suction on the containment
- sump, the source term was based on diluting the liquid inventory with the'6,600 gallons in the reactor coolant system.
This source was used for the nuclear sampling system (reactor coolant fluid) and the portion of the chemical and volume control system associated with reactor coolant degassing.
Gaseous source terms were determined for containment and for the waste gas system.
The containment airborne source term was based on diluting the gaseous inventory discussed previously with the air contained in the containment free volume (970,000 cubic feet).
The waste gas system source term was determined for a reactor coolant degassing operation by calculating the quantity of activity entering the volume control tank (VCT) via the normal letdown path and assuming that a quantity equal to the stripping fraction (based upon a stripping efficiency of 1.0) enters the vapor space and is immediately purged to the vent header system (Ref. 1).
The stripping fraction is defined as:
CR - CL CR where:
SF = stripping fraction CR = concentration entering VCT in liquid CL = concentration leaving VCT in liquid Table 3-1 provides the shielding source terms for liquid and gaseous radioactivity which were calculated using the assumptions above.
3.2 1'IETHODOLOGY 3.2.1 Calculation of Dose Rates Dose rates for the areas of interest in this review were calculated by determining the potential contributing sources at a representative location and using the appropriate source term from Table 3-1 adjusted for decay as required.
The dose rate at the representative location was used as the general area dose rate for the area and chosen to envelope all critical locations of interest.
The SDC Code (Ref. 2)
"7-was used in performing the dose rate calculations.
Energy groups required as input to the code were determined using the gamma ray energy and intensity data (Refs.
3 and 4) for the nuclides in Table 3-1
~
3.2.2 Calculation of Doses to Personnel Durin Post Accident Access to Vital Areas Personnel doses received in performing a specific task in a given vital area were calculated as the sum of the doses received during travel to and from the vital area and the dose received while perform-ing the given operation in the vital area.
The doses received during travel were determined by calculating dose rates at selected locations (or at a single location if the dose rate along the travel route is relatively uniform) along the travel route using the methodology discussed in Section 3.2.1 and multiplying the dose rates by the appropriate travel times.
Doses received while performing a given operation were determined by multiplying the dose rate for the given area by the time to perform the operation.
Dose rates for the given vital area were calculated using the methodology discussed in Section 3.2.1.
Travel times were based upon an assumed access rate of 200 feet per minute.
Time estimates for performing indicated tasks were based upon experience gained during a decade of plant operation at Ginna Station.
3.2.3 Calculation of Inte rated Doses to Safet E ui ment The integrated dose to each piece of safety-related equipment was determined by integrating the appropriate dose rate over the time period that it is required to be available to perform its safety
function.
Dose rates were calculated using the methodology discussed in Section 3.2.1.
TABLE 3-R. E.
GINNA Shielding Source Terms (T=O)
~Isoto e Br-84 Kr-87 Te-133 Cs-134 Cs-136 Cs-137 Ba-139 Br-83 Kr-83m Kr-85m Kr-85 Kr-88 Rb-88 Rb-89 Sr-89 Sr-90 Y-90 Sr-92 Y-92 Sr-93 Y-93 Liquid Source Activity (Ci) 3.9+
6 1.9+
7 2.3 + 5 8.9+
4 2.5+ 4 3.7 + 4 7.9 + 5
~
1.7 + 6 3.3 + 6 1.0 + 7 3.9 + 5 2.9 + 7 2.8 + 5 3.6 + 5 3.9 + 5 2.6 + 4 2.6 + 4 4.8+ 5 5.2+
5 5.8 + 5 5.8 + 5 (2)
Gaseous Source Activity (Ci) 2.0+
7 1.9 + 7 8.4 + 5 3.3+
6 1.0+
7 3.9 + 5 2.9 + 7 Containment Sump Concentration
( Ci/cc) 3.5+ 3 1.7+ 4 2.1+
2 8.0 + 1 2.2 + 1 3.3 + 1 7.0 + 2 1.5 + 3 3.0 + 3 8.9+
3 3 ~ 5 + 2 2.6+ 4 2.5 + 2 3.2 + 2 3.5 + 2 2.3 + 1 2.3 + 1 4.3+
2 4.7 + 2 5.2 + 2 5.2 + 2 Reactor Coolant Concentration
( Ci/cc) 2.2 + 4 1.1 + 5 1.3 + 3 5.0+ 2 1.4 + 2 2.1 + 2 4.4+ 3 9.5+ 3 1.9 + 4 5.6 + 4 2.2+
3 1.6 + 5 1.6 + 3 2.0 + 3 2.2 + 3 1.5 + 2 1.5 + 2 2.7+ 3 3.0 + 3 3.3 + 3 3.3 + 3 Con tainmen t Airborne Concentration (pci/cc 7.2+ 1 7.0 + 2 3.0 + 1 1.2 + 2 3.6+
2 1.4+ 1 1.0+
3 Waste Gas Concentration
( Ci/cc 3.4 + 3 6.7 + 5 1.9 + 2 1.1+ 5 2.6+
5 5.9 + 3 8 '
+ 5
~Isoso e
Mo-99 Tc-99m Ru-103 Rh-103m RU-106 Rh-106 Te-132 I-132 Te-134 I-134 Xe-138 Cs-138 Ba-140 La-140 Ce-143 Pr-143 Ce-144 Pr-144 Sr-91 Y-91m Y-91 Zr-95 Nb-95m Nb-95 Zr-97 Liquid Source Activity (ci) 7.9 +
6.8+ 5 5.8+
5 5.8 + 5 2.2 + 5 2.2 + 5 5.8 + 5 3.1 + 7 8.4 +
4.7 + 7 7.9 + 7 7.9+ 5 7.3+ 5 7.9+ 5 6.3 + 5 6.3+
5 4.8 + 5 6.3+
5 4.8 + 5 5.0 +
6.8+
5 6.8+
5 6.8+
5 (2)
Gaseous Source Activity Ci) 1.5 + 7 2.3+
7 7.9 + 7 TABLE 3-1
(
Containment Sump Concentration pci/cc 7.0 + 2 6.1+
2 5.2+
2 5.2 + 2 1.9 + 2 1.9 + 2 5.2 + 2 2.7+
4 7.5 + 2 4.2 + 4 7.0 + 4 7.0 + 2 6.6+
2 7.0 + 2 5.6 + 2 5.6 + 2 4.3 + 2 5.6 + 2 4.3 + 2 4.5+
2 6.1+
2 6.1 + 2 6.1+
2 Reactor Coolant Concentration (pci/cc 4.4 + 3 3.9 + 3 3.-3 3. 2 1. 2 1. 2 3. 3 + 3 1. 8 + 5 4. 8 + 3 2.6+ 5 4.4 + 5 4.4 + 3 4.2 + 3 4.4+ 3 3.6+
3 3.6 + 3 2.7 + 3 3.6 + 3 2.7 2.8 3.9+ 3 3.9+ 3 3.9 Containment Airborne Concentration Ci/cc 5.6 + 2 8.5 + 2 2.9 + 3 Waste Gas Concentration
( Ci/cc 2.7+
4 4.0 + 4 3.4+
6
~Isoeo e
Nb-97m Nb-97 Ru-105 Rh-105m Rh-105 Te-131 I-131 Xe-131m I-133 Xe-133m Xe-133 I-135 Xe-135m Xe-135 Ba-141 La-141 Ce-141 Liquid")
Source Activity (Ci) 6.8 + 5 4.5 + 5 4.5 + 5 2.9+ 5 3.8+
5 2.2 +
7 3.4 + 5 4.1 + 7 2.0+
6 7.9 + 7 3.6 + 7 2.2+
7 1.5+
7 6.8+
5 6.8+ 5 7.3 + 5 (2)
Gaseous Source Activity Ci 1.1 + 7 3.4 + 5 2.0 + 7 2.0+
6 7.'9 + 7 1.8+
7 2.2 + 7 1.5 + 7 TABLE 3-1 (
Containment Sump Concentration (pCi/cc) 6.1 +
2'.0
+ 2 4.0 + 2 2.8 + 2 3.4+ 2 1.9+
4 3.0+
2 3.7 + 4 1.8 + 3 7.0 + 4 3.2+ 4 2.0 + 4 1.4 + 4 6.1+
2 6.1 + 2 6.6 + 2 Reactor Coolant Concentration (pci/cc 3.9 + 3 2.6 + 3 2.6 + 3 1.6+ 3 2.1 + 3 1.2 + 5 1.9 + 3 2.3 + 5 1.1 + 4 4.4 + 5 2.0 + 5 1.2 + 5 8.6 + 4 3.9 + 3 3.9+ 3 4.2+
3 Con tainment Airborne Concentration Ci/cc 3.9 + 2 1.2 + 1 7.4+ 2 7.2 + 1 2.9 + 3 6.6+ 2 8.0 + 2 5.5 + 2 Waste Gas Concentration (pci/cc 1.9+ 4 5.8 +
3.5 +
3.7+
4 1.4 + 6 3.1 +.4 9.1+ 5 2.6 +
(1) Based on 100% noble gas core inventory, 50% halogen core inventory, and 1% of all others core inventory.
(2) Based on 100% noble gas core inventory and 25% halogen core inventory.
REVIEW OF AREAS RE UIRING ACCESS FOR POST-ACCIDENT 'OPERATIONS AREA A:
H dro en Recombiner Control Panel Area "A" as identified on Figure 4-1 is the general location of the hydrogen recombiner control panel on the 253'levation of the Inter-mediate Building.
Post-accident hydrogen formation in the containment vessel may require operation of the hydrogen recombiner system.
The system is started and controlled by operator action at the recombiner panel.
This task would require that an individual spend approximately 30 minutes at the control panel.
Access to the control panel could be required shortly after an accident.
The exposure rate at one hour after the accident in the general area II of "A" would come from the direct dose from the Containment Building (2.2 R/hr) and from sample lines approximately 30 ft. away (0.95 R/hr) totaling about 3.1 R/hr.
The estimated dose to an individual in this area for the necessary occupancy would be 1.6 rem. If occupancy was required at one day after the accident, the exposure rate is estimated to be 0.12 R/hr which results in a 30-minute dose of 0.06 rem.
The access route to area "A" is route "1" described in Table 4-1.
The estimated dose to an individual as he traverses this route is 0.08 rem after at one hour the accident and is less than 0.005 rem at one day.
It should'be noted that the Chemical Drain Tank, located in Area "A",
could receive small quantities of discarded primary coolant samples from the Nuclear Sample Room sink under the existing sample disposal piping arrangement'.
The most significant quantity of sample fluid would be from. the dissolved 02 analysis which is a radiologically
13-impractical test to perform under the assumed accident conditions, and therefore was not addressed in this analysis (see Section 4.3).
While the total calculated dose from performing the described task is less than 3 rem shortly after the accident, recommendations for sample line and Nuclear Sample Room shielding in the sections which follow would provide further dose reduction, particularly during access to and away from this area.
4.2 AREA B:
Post-Accident Containment Air Sam le Penetration No.
203 Area "B" is the general location of containment penetration 0203 in the Intermediate Building, 271'levation (See igure 4-2).
After an
- accident, a containment air sample could be collected at this loca-tion, in addition to alternative sampling penetrations which are later addressed in Areas "F" and "J".
Each post-accident containment air sample penetration contains a
sample inlet line and a return line to the Containment Building.
Both lines have manual valves locked closed.
Upon opening the
- valves, a sample rig would be attached to the lines consisting of
/
- tubing, a gas collection bulb and a vacuum pump.
The sample is collected by circulating containment air through the collection bulb for approximately 5 minutes.
The time required to collect the air sample has been conservatively estimated to be 15 minutes.
The radiation level at one hour in Area "B" will result from the direct dose from the containment (9.5 R/hr), from the nuclear sample lines carrying liquid samples (3 ft. away 60 R/hr), from the 35 ml gas sample bulb (1 ft. away - 1.2 R/hr) and the containment air sample lines (3 ft. distance - 0 '
R/hr).
The summation of these various radiation sources (at one hour after the accident) is 70 R/hr.
The resulting dose is estimated to be 17.8 rem for collecting an air sample at this location.
These radiation levels drop significantly with time (See Table 4-2).
The exposure rate decreases to 7.8 R/hr in one day.
The access route to Area "B" is designated as route 52 in Table 4-2, and would entail some exposure contributed by direct radiation from the Containment Building, sample piping and sample coolers located in the Nuclear Sample Room.
This access route is not anticipated to add significantly to the radiation dose of the individual collecting the sample (0.08 rem).
The gas sample would be placed in a shielded con-tainer for transport to the Primary Chemistry Laboratory.
The dose to an individual performing this task could be reduced by several means.
First, because of the high specific activity, the volume of the air sample collection container can be decreased.
Lead shielding can also be used around the sample container; one inch of lead around the sample would decrease the radiation levels at one hour from 1.2 R/hr to 0.32 R/hr, a reduction factor of about 4.
The 15-minute exposure period assumed for performing this task is likely to be conservative by as much as a factor of 2, and considers the fact that work would be performed in respiratory protection equipment.
The individual taking the sample could readily leave the high dose rate area while the sample was being collected, as is normal practice.
Therefore, utilizing a reduced sample size, sample container shielding and shortened stay-time, this task could be accomplished with a dose of approximately 5 rem.
It is important to note, however, that under the accident conditions assumed in this analysis, priority should be given to the collection and analysis of a primary coolant sample and to onsite and offsite radiological monitoring efforts.
High airborne levels inside con-tainment would be deduced from containment area radiation and airborne radioactivity monitors.
Therefore, containment air sampling, while important, would be less critical after the onset of an accident than the other types of sampling.
Should a containment air sample be deemed necessary, alternative sampling locations are available in the plant, and could afford further dose reduction
[Area "F"].
4.3 AREA C:
Nuclear Sam le Room The Nuclear Sample Room, Intermediate Building, elev.
271 (Figure 4-2) has been designated as Area "C".
Primary system liquid and gas samples would be needed following an accident to assess the extent of fuel damage and to evaluate primary coolant chemistry.
Potential sample sources could include:
The "B" primary loop hot leg; the pressurizer liquid and steam spaces; and the residual heat removal loop.
These samples are collected under a ventilated hood in the Nuclear Sample Room.
Liquid sample lines discharge to a sample sink for collection in a suitable sample collection bottle, and gas samples are collected in a stainless steel pressure vessel which is installed with quick-disconnect fittings.
Samples would be collected and transferred to the adjacent Primary Chemistry Laboratory through a
passbox.
The sample size for normal operations is 250 ml.
Under the accident conditions being considered, the sample size will be reduced sig-nificantly.
However, shielding of the sample will be required to minimize the radiation dose to workers.
This evaluation assumes the first primary system sample will be taken within one hour after an accident, in accordance with NRC guidelines.
At this time, a 75 to 100 ml. sample (75 ml. indicated henceforth) gives an exposure rate at 1 ft. from the container of 600 R/hr.
With 4inches of lead shielding, the exposure rate is reduced to 1.4 R/hr.
Sample lines in the area contribute 60 R/hr three feet away at one hour following the accident.
In addition, the dose rate from a contaminated sample cooler (located behind a 6-inch concrete shield wall in the Nuclear Sample Room) would be 30 R/hr. It has been estimated that sample collection takes a worker 6 minutes resulting in 70 rem with no shielding or 9 rems when the sample bulb is shielded.
The same operation one day after the accident will result in 8.7 and 1.1 rem, for samples without shielding and with shielding, respectively.
Since access to the Nuclear Sample Room may be required within one hour after an accident, consideration must be given not only to shielding the sample collection vessel but also the sample lines, and providing additional shielding around the sample coolers.
One inch of lead shielding will reduce radiation levels from exposed sample piping and the coolers by a factor of 7 thus resulting in acceptable personnel doses.
Actual piping locations
'and available space in the nuclear sample room need to be reviewed to utilize shielding to the extent practicable.
Dissolved oxygen analysis is currently performed using a leuco reagent color test, under the ventilated hood in the Nuclear Sample Room.
Although the test for oxygen could be performed following the accident, several radiological considerations would argue against attempting the analysis in this manner shortly after an accident.
First, it is
17 unlikely that significant dissolved oxygen levels will occur, in the I
coolant due to the presence of hydrogen.
- Second, the leuco test pro-cedure involves relatively large volumes of primary sample which can-not be significantly diluted or reduced.
This could involve signifi-cantly higher direct radiation dose rates from the sample bottle (greater than 1000 R/hr at 1 ft.) and disposal problems from the flushed sample liquid into the sample sink.
In addition, significant quantities of primary gas will evolve using this method and could aggravate airborne radioactivity levels in surrounding work areas.
It is recommended therefore that an alternative means be investigated for performing dissolved oxygen analysis.
The access route to the Nuclear Sample
- Room, Route 52 (See Table 4-1) will add approximately 0.08 rem to the worker dose if the task is done during the first hour after the accident.
Utilization of this access route at later times will not add significantly to worker dose.
It is seen therefore that sampling of primary coolant can be performed within acceptable dose levels, if measures are taken to provide shielding around the sample container, sample lines and sample coolers inside the Nuclear Sample Room.
An alternative approach to adding shielding to the sample coolers could be eventual relocation in a more remote or better shielded portion of the'lant.
4.4 AREA D:
Primar Chemistr Laborator The Chemistry Laboratory in the Service Building has been designated as Area "D".
Radiological and chemical analyses are performed in the Primary Chemistry Laboratory once samples are transferred from the Nuclear Sample Room.
A series of procedures are then utilized
18-for drawing liquid and gas aliquots for counting, and for performing additional required tests including boron, pH, conductivity, hydrogen and other trace chemicals.
Occupancy in this area is assumed to be primarily frequent after an accident.
Radiation levels in Area "D" are primarily attributed to sample lines in the Nuclear Sample Room.
A one foot shield wall separates Areas "C" and "D".
Radiation levels next to the wall separating these areas will read about 0.2 R/hr within one hour after the accident, 0.03 R/hr after one day, and about 0.002 R/hr after two days.
General areas in the Chemistry Laboratory will read about
.03 R/hr after one hour, about 0.004 R/hr after one day and about 0.002 R/hr after two days.
4.F 1 Isotopic Analysis
~
~
Primary coolant sample processing results in high localized doses for the workers involved.
(See 7.0 Appendix for Sample Preparation Time Study.)
Handling the 75 ml. primary coolant sample and sub-sequent diluted samples results in 6.7 rem (40 seconds exposure to the unshielded 75 ml. primary coolant sample) and 0.2 rem (from handling 1 ml. of the primary coolant sample diluted to 50 ml.).
The 75 ml. sample could be shielded while in the sample hood.
The only unshielded time would be during the transfer from the passbox to the sample hood.
This transfer could be done in 10 sec.
using an extension tool to keep the sample vessel at least 6 feet from the worker.
Under these conditions, the exposure rate from the sample vessel would be 19 R/hr at one hour after the accident result-ing in a dose to the worker of less than 0.1 rem.
In the radio-
19 chemistry sample hood with four inches of lead, the sample vessel will read 1.4 R/hr and, therefore, result in about 0.01 rem to the worker.
The total dose to the worker for processing the sample for isotopic analysis, utilizing shielding to the extent possible, is about 0.3 rem.
4.4.2 Boron Analysis For boron analysis the 75 ml. primary coolant sample would be shielded while in the sample hood.
By providing shielding around the titra-tion rig, the total dose to the worker should not exceed the dose of 0.3 rem anticipated for the isotopic analysis.
(See 7.0 Appendix on Sample Preparation Time Study.)
4.4.3 Primary Gas Analysis
~
~
The primary gas analysis involves a transfer rig mounted on the wall next to the sample passbox.
(See 7.0 Appendix on Sample Prepara-tion Time Study.)
Gases from the 75 ml. primary coolant sample vessel are drawn into a gas collection container from which samples are taken for hydrogen and isotopic analyses.
As indicated previously the radiation level from the 75 ml. primary coolant sample vessel is about 600 R/hr at 1 foot at one hour after the accident.
In review-ing the procedures, it has been determined that an individual would be located approximately an arm's length from the sample for about 4
minutes which would results in a dose to that worker of 40 rem.
Utilizing a shadow shield of 2 inches of lead, the exposure rate would be reduced to 20 R/hr and the dose to the worker to about 1.3 rems. If this task were done at one day post accident, the exposure level, using the 2 inch lead shield, would be 2.7 R/hr and the dose to the worker about 0.2 rem.
Although it is apparent that the required handling and analysis of samples can be performed in the Primary Chemistry Laboratory within dose guidelines, several procedural and shielding modifications should be considered.
Recommended shielding additions to the sample lines and coolers in the Nuclear Room will also reduce dose levels considerably in the Primary Chemistry Laboratory.
Shielded sample containers are also necessary, both for temporary sample storage during procressing and for receiving discarded samples.
Disposal of samples into the sample sink, as is normal practice, may pose immediate airborne contamination problems in the Chemistry Laboratory.
All handling of open sample containers should be done behind shield-'ng under the Laboratory's ventilation hood.
This would also apply to use of the gas partitioner (primary H analysis) as some volume of the injected sample will diffuse into the Chemistry Laboratory atmosphere.
Portable shadow shields should be considered for work in close proximity to any exposed sample container (e.g., primary gas sample bomb, boron analysis).
AREA E:
Count Room The Count Room, designated as Area "E", would be primarily affected by post-accident radiation sources in the adjacent Nuclear Sample Room.
Two feet of concrete shielding separate these two rooms.
The Count Room is equipped with a gamma pulse-height analyzer system and counting safe for isotopic analysis of primary liquid and gas
- samples, and plant effluent releases.
Additional counting equipment is also available in the Count Room (e.g.,
GN and proportional counters) for analyzing general plant samples for airborne and surface contamina-tion.
21 Radiation levels from sample lines and sample coolers in the adjacent Nuclear Sample Room would approach 0.09 R/hr one hour after the accident within three feet of the shield wall and about 0.01 R/hr in general areas within the Count Room.
A relatively minor dose rate contribution, 0.002 R/hr, would be caused by direct radiation from the Containment Building.
By the end of the first day after an
- accident, these levels are decreased by over a factor of 10.
4.6 A combination of two alternatives should be considered for post-accident in-plant sample counting.
First, provisions should be made to mobilize count room equipment in the event actual direct radiation background and airborne contamination levels prohibit adequate counting capability.
There appear to be no major problems in accom-plishing this as Ginna Station.
Host of the count room equipment could be relocated in a lower background area onsite (e.g.,
ware-house) within the first two hours following an accident.
In addi-
- tion, a trailer now equipped for routine environmental counting, could be utilized if necessary for sample analysis.
In addition, recommended shielding of Nuclear Sample Room sample lines and coolers will reduce dose rates in the Count Room.
Although certain sensitive counting equipment may still be adversely affected by elevated background levels, the 4-inch thick counting safe housing the GeLi detector would allow gamma isotopic analysis to be performed.
AREA F:
Post-Accident Containment Air Sam le Penetration No.
305 Area "F" is in proximity to post-accident containment air sample lines at Penetration 305 located on the 253 ft. elevation of the Intermediate Building (See Figure 4-1).
For Area "F" liquid sample
- 22 "
lines will not be a problem and the direct dose from containment will be less since this penetration is below the operating floor.
Table 4-2 gives the dose rates in this area as a function of time.
As indicated for Area "B", the task of collecting an air sample from containment is conservatively assumed to require about 15 minutes of worker exposure time.
Without shielding the gas sample bulb, the worker would receive about 0.9 rems collecting the sample.
By using one inch of lead around the sample collection bomb, the dose to the worker would be reduced to 0.7 rems.
Reduction in the sample size and of the length of the sampling lines can also be used to decrease the dose to the worker.
I The access route utilized to get to Area "F" will be 44 as described in Table 4-1.
The radiation dose accessing Area "F" will be small (0.01 rem), provided the sample collection vessel is adequately shielded.
4.7 AREA G:
Radwaste Control Panel Area "G", the radwaste control panel, is located on the basement floor of the Auxiliary Building (See Figure 4-3).
Immediate post-accident access to the waste disposal panel and gas analyzer is not required; however, certain longer term actions may make access necessary to this area.
The waste disposal panel contains pressure gauges for the tanks using cover gas and also for the gas decay tanks and vent header.
Alarms for tank and vent header pressure and gas analyzer oxygen are locally indicated with a waste disposal panel alarm given on the Main Control Board.
All gas system manual operations and releases are controlled locally at the waste panel.
I'n addition, various equipment associated with the liquid waste system are manually controlled at this location.
23-Before recirculation through the.RHR pumps (at least 30 minutes after the accident) the radiation level in Area "G" will be back-ground.
During recirculation, RHR piping, safety injection pumps and pipes and containment spray pumps and pipes will make radiation levels on the basement floor of the Auxiliary Building very high.
It is anticipated than when occupancy is required at the radwaste control panel, an individual would spend 2 minutes in this area once per shift.
From Table 4-2, the radiation levels at this location are estimated to be 3,000 R/hr one hour after the accident.
A two-minute occupancy in this area would result in a 100 rem dose.
After one, three and 10 days, the dose to a worker in two minutes would be 10 rems, 5 rems and 2 rems, respectively.
Access to the radwaste control panel is via route 55 (Table 4-1).
The dose received along this route is estimated to be 0.35 rem during the first hour of the accident and less than 30 mrem 1 day following the accident.
Because of the magnitude of these levels and the extent to which shielding would be required to reduce these levels, it appears probable that a modification to the plant procedures and controls should be investigated to transfer certain radwaste equipment indi-cations and control functions from the radwaste control panel to another location in the plant.
AREA H:
Safe ards Bus 16
, Area "H" is on the intermediate level of the Auxiliary Building around Safeguard Bus 516 (Figure 4-4). It is not expected that any immediate access would be required to the areas of the safeguards buses or their associated motor control centers.
Radiation levels in Area "H" will be from the Safety Injection pumps and associated piping directly below on the basement floor.
The shine through the 18" concrete floor has been estimated to be about 60 R/hr at one hour after the accident.
This radiation level drops to 2.4 R/hr at one day and 0.4 R/hr at 10 days (See Table 4-2).
In the event the CVCS system is used to strip gas from the primary
- coolant, the waste gas piping from the volume control tank to the gas compressor and into the waste gas decay tanks will be con-taminated.
The waste gas decay tanks, once filled with accident source activity, will also contribute to the radiation level in Area "H", although to a lesser extent.
Radiation levels 30 feet from the 1" waste gas piping have been estimated to be 100 R/hr at one hour after the accident.
After one
- day, these levels decrease to 5 R/hr, approximately double the radiation level coming from the floor below.
The waste gas decay tanks have been estimated to contribute 2.8 R/hr to Area "H" at one hour into the accident and 0.09 R/hr after one day.
The waste gas piping is one inch in diameter and could be locally shielded.
One inch of lead shielding would decrease the radiation level by a factor of about 4, and 2 inches of lead by a factor of 10.
Access to Area "H" at one hour after an accident is assumed to be via route 57 (Table 4-1) and will result in an additional 4.4 rem for ingressing and egressing the area.
The majority of this dose comes from the waste gas piping (4 rem). If no waste gas stripping is in progress, then the access dose would be only 0.4 rem. If the waste gas piping is contaminated with stripped gases, shielding with the equivalent of 1" of lead on waste gas piping and 1/2" of lead on RHR sample lines results in a dose to individuals accessing Area "H" of just over 1.5 rem. If access to this area is required at 1 day post-accident, then the radiation dose for ingress and egress would be 0.14 rem and 0.007 rem for unshielded and shielded waste gas and RHR sample piping, respectively.
Area I:
Safe ards Bus 14 The safeguards bus 514 is on the 271'levation.
It has been designated as Area "I" (Figure 4-5) and is directly over Area "H".
Occupancy in this area will be the same as explained for Area "H".
The radiation level in Area "I" will be essentially from the direct dose from containment.
At one hour, the radiation level has been estimated to be about 5 R/hr, after one shift 0.7 R/hr and only 0.04 R/hr after one day.
Access to Area "I" is via route 46 (Table 4-1) and will add 0.35 rem to the worker's dose at one hour following an accident and only 0.005 rem after one day.
AREA J:
Post-Accident Containment Air Sam le Penetration Area "J", defined in Figure 4-4, provides another location where a
containment air sample can be collected.
The task performed at this
location would be the same as in Areas "B" and "F".
Area "J" radia-tion level will be from the direct dose from containment and from two 6" diameter safety injection pipes running several feet below the floor slab.
ll The exposure rate from these sources is estimated to be 35 R/hr at one hour post accident dropping to 1.2 R/hr in one day.
The 35 ml.
sample vessel will not add a'significant amount to the worker's dose.
At one hour, the sample vessel will be reading about 1R/hr and at 1 day 0.13 R/hr.
The dose to the worker at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after an accident (assuming 15 minutes occupancy) would be about 9 rem, at one day about 0.3 rem and at 30 days, 0.01 rem.
Access to Area "J" will be via route 58 (Table 4-1) and will add 3
rem to personnel doses at one hour after the accident (waste gas piping contaminated and not shielded; if not contaminated, then the dose would be 0.4 rem.
One inch of lead shielding around contaminated waste gas piping and 1/2 inch lead shielding on RHR sampling lines.
would reduce the access dose to 1 rem.
Access doses at 1 day would be 0.1 rem if waste gas stripping is in progress and is negligible without gas stripping.
It is clear that access to Area "J" for containment air sampling shortly after an accident would be less preferable due to high direct radiation levels and probable airborne contamination than would be Area "F".
Nevertheless, the area could be accessed after 1
day with proper coordination of waste gas stripping operations to minimize dose.
Also, sample container shielding recommendations previously made for sample transport should also be considered.
27 AREA K:
Auxiliar Buildin HVAC Filters
. Area "K", the Auxiliary Building HVAC filters, 253 ft. elevation (Figure 4-4) will have the same exposure rates as Area "H".
Should access be required in Area "K" to change filters, it has been estimated that it would take approximately 20 man-hours to complete this task.
Table 4-3 provides area dose rates and access doses similar to those calculated for Area "H" following a postulated accident.
At one hour following the onset of the accident, dose rates would be prohibitive for the filter change task to be com-pleted.
However, it is likely that such a task would not be under-taken until later recovery operations were begun. If it is assumed that Area "K" filters are changed 10 days following the accident, then direct radiation levels would result in personnel exposures of less than 5 rem (assuming 4 workers available. for this task).
This also accounts for the likelihood that the contaminated filters themselves would be a direct source of radiation, on the order of 1 R/hr (ref.
5).
Several means are available to reduce exposure in the event such a task needs to be performed.
For instance, new filters can be unboxed in a lower dose area (Aux. Bldg. operating floor) and lowered to the 253 ft. elevation by crane through the floor access hatch.
Also, spent filters could be shielded with lead blankets and left on the 253 ft. elevation for decay prior to further handling or removal.
Hodifications to waste gas piping involving shielding or relocation could also further reduce worker dose in the event primary coolant degassing operations are coincident with filter changes.
AREAS L AND N:
S ent Fuel Pool and Auxiliar Buildin HVAC Filters
'reas "L" and "H" are filter units for the Spent Fuel Pool and Auxiliary Building HVAC systems (Figures 4-5 and 4-2, respectively).
These areas will be affected by radiation levels from containment and from sample lines running along the west and south walls. It is not anticipated that access will be required in these areas im-mediately after an accident, but should conditions dictate, access may be needed to change out filters.
The exposure rates for Areas "L" and "H" are provided in Table 4-2 as a, function of time following the assumed accident.
Access to these areas is via route 52 (Table 4-1) and would result in an approximate 0.08 rem dose one hour after an accident.
After one day the access dose becomes negligible (less than 0.01 rem)
~
If one assumes the same exposure times used for Area "K" (e.g.
20 man-hours at 10 days following an accident),
both filter change tasks would be able to be performed within dose levels of less than 5 rem.
AREA N Area "N", as identified on Figure 4-1, includes the area for the Control Access HEPA and charcoal filters on the 352 ft. elevation.
Access would be required in this area only to change filters, and would be via route 59 (Table 4-1).
The radiation level in Area "N" will be the summation of the direct dose from the containment and from primary coolant sample lines approximately 30 feet away.
The radiation exposure rates are given in Table 4.2 as a function of time.
The dose rate in this area decreases from 1.7 R/hr after the
29-first hour to slightly more than 0.1 R/hr one day following an accident.
It is seen that even with an assumed task performance time of 20 man-hours, the task can be accomplished within acceptable dose guidelines after one day.
CONTROL ROOH Continuous occupancy is required in the Control Room after an accident.
The Control Room is located on the 289 ft. elevation as shown in Figure 4-6.
The direct radiation from the Containment Building will result in radiation levels in the Control Room of approximately 0.018 R/hr at one hour after an accident, quickly dropping off to 0.008 R/hr at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and to near background levels shortly thereafter.
The integrated dose (6 months) to operators in the control room (assuming 100~ occupancy) due to direct radiation from containment has been calculated to be 0.2 rems.
The majority of this dose comes within a few hours after the accidents TABLE 4-1 ACCESS ROUTES TO VITAL AREAS Access Route Descri tion 1.
To H2 Recombiner Panel From Service Building (S.B.),
271 elev. to Intermediate Building (I.B.) South Side; down staircase next to personnel hatch to the 253'lev.
2.
To Air Sample Penetration 203 and Nuclear Sample Rm.
From ST B. 271 elev. to I.B. then south by nuclear sample room and containment penetrations.
3.
Primary Chemistry Lab.
Count Room 4.
Air Sample Penetration and 305 In S.B.
S.B.
271 elev. to Turbine Building (T.B.)
to staircase in NH I.B. down to 253'lev.
T.B. East to double door to I.B.
5.
Radwaste Panel S.B.,
271 elev. to I.B. South by Nuclear Sample Room, along SFP to Auxiliary Building (A.B.) to East staircase down to 236 elev.
6.
Bus 14 Same as "5" but stays on 271'o column lines 8a-9a and L-N.
7.
Bus 16 and HVAC Filters Same as "5" but utilizes Hest staircase in A.B. to the 253 elev. to column lines 8a-9a, L-N.
8.
Air Sample Penetration (behind RHST)
Same as "7" except to column lines 7a-8a, L-N ~
9.
Control Access Filters Same as "2" but utilizes stairs in SW I.B.
to go to the 253 ft. elevation.
31 TABLE 4-2
,, EXPOSURE RATES FOR VITAL AREAS AS A FUNCTION OF TIME EXPOSURE RATE (R/hr)
VITAL AREA B
(unshielded air sample)
(shielded air sample 1" Pb)
C (unshielded liquid sample)
(shielded liquid sample)
D (next to wall)
(general area)
E (hear'all)
(general area)
F (unshielded air sample)
(shielded air sample 1" Pb)
G (excluding waste gas system)
(including waste gas system unshielded pipes)
(including waste gas system 1" Pb on pipes)
J (unshielded air sample)
(shielded air sample Pb)
K (excluding waste gas system)
(including waste gas system unshielded pipes)
(including waste gas system 1" Pb on pipes) 1 HOUR 3.1 71 70 690 91 0.22 0.03 0.09 0.01 3.6 2.7 3100 60 180 93 5.4 35 34 60 180 93 8.4 1
DAY 0.12 7.8 7.7 87 1.1 0.03 0.004 0.003 (2) 0.17 0.07 340 2.4 7.4 2.7 0.04 2.6
'.4 2.4 7.4 2.7 0.11 30 DAYS 0.01 0.47 0.46 5.8 0.48 (2)
(2)
(2)
(2)
(2)
(2) 21 0.16 0.28
- 0. 14 0.10 0.10 0.16 0.28 0.14 0.004 6
MONTHS 4.4 N
Control Room
~
~
tes:
1) not calculated 2) negligible
(<0.001 R/hr) 1.7 0.018 0.11 (2) 0.01 (2)
(2)
32 TABLE 4-3 VITAL AREA RADIATION DOSE
SUMMARY
. Time After Area Accident
~cccu auc Dose In Area (rem)
Access Dose To And From (rem)
Total (rem)
A 1 hr.
1 day 1 hr.
1 day 1 hr.
1 day 30 min.
15 min.
6 min.
1.6 0.06 17.81 2.01 9.01 0.08 neg.
0.08 neg.
0.08 neg.
1.7 0.06 17.9 2.0 9.1 D
1 hr.
1 day 1 hr.
1 day 1 hr.
1 day Frequent Frequent 15 min.
0.032
. 0042 012'032 0 71 0.071 0.05 neg.
- 0. 05 neg.
0.01 nega N/A N/A N/A N/A 0.71 0 ~ 07 H
K
'GM 1 hr.
1 day 1 hr.
1 day 1 hr.
1 day 1 hr.
1 day 1 hr 1 day 1 hr 1 day 2 min.
15 min.
100 10 602 42 5.42
- 0. 042 8.81 0 31 602 2 42 8.42 0.12 0.35
<0.03 0.44 neg4 0.4 neg 0.44 neg4 0.44 neg4 0.08 neg.
100 10 N/A N/A N/A N/A 9.2 0.3 N/A N/A N/A N/A
TABLE 4-3 (Cont. )
Area N
Time After Accident 1 hr.
1 day@
~cccu acc Dose In Area (rem) 172 0.12 Access Dose To
.And From (rem) 0.08 neg.
Total (rem)
N/A N/A Control Room 1 hr.
4 hr.
>8 hr.
Continuous 0.0182 0.0082 Bkg2 neg.
neg.
neg.
N/A N/A N/A 1
Shielding utilized. around sample vessel.
r 2
General area dose rates (R/hr.).
3 General areas in Chemistry Laboratory.
See Sections 4.4.1, 4.4.2, and 4.4 '
and Appendix (7.0) for sample processing.
4 No waste gas stripping in progress.
P 5.0 REVIEW OF RADIATION UALIFICATION OF SAFETY E UIPHENT 5.1 IDENTIFICATION AND EVALUATION OF EQUIPHENT Table 5-1 identifies the location of the safety-related equipment considered in the review.
The equipment considered is consistent with SEP environmental qualification and safe shutdown reviews previously submitted to the Commission.
Table 5-1 also presents the required time each item of equipment has to be capable of performing its safety function; the integrated dose for each given time period and location,.and the radiation qualification of the equipment.
In-tegrated doses were not calculated for components which perform their safety function prior to post-accident recirculation or which are located in areas remote from the systems which may contain radio-active material after an accident.
These components would not have any substantial increase in integrated exposure as a result of the NUREG-0578 requirements.
5.2 RECOHHENDATIONS A comparison of the calculated and qualified integrated doses in Table 5-1 indicates several components which could potentially be exposed to high radiation levels that do not have qualification values available.
These include CV pressure transmitters (PT-945 and 946) and the RHR, containment
- spray, and safety injection pumps.
Evalua-tion of these components should be made to determine the appropriate qualified radiation dose.
TABLE R.
E.
GINNA Equipnent Qualification E ui neat Name Identification Number Location
~Bfldin Rl u.
Col. Lines Tine After Accident Safety Function Re uired Required Operating Duration Calculated Integrated
~no*a Rais Qualified Integrated
~De Rads (r i 6.)
NaOH Addition Valves Feeduater Isolation Valves Hain Stean Isolation Valves BA Supply to SI Pumps RWST Supply to SI Pumps RWST Isolation Valves Aux. PeedD>ater to S/G Service Water to Aux.
FW Pumps CV Sump Isolation Valve RHR Pump Discharge Containment Spray Discharge RWST to RHR Suction Standby Aux.
FW Punps Pd!R Pumps Containment Spray Pumps CCW Pumps Aux.
FW Pumps SI Pumps Service Water Pumps CV Press.
Transnitter CV Press.
Transmitter CV Press.
Transmitter RWST Level BAST Level Transmitter A Steanline Press.
Transn.
B - Steanline Press.
Transn.
Batteries V836 A, B
V4269, 4270 V3516, 3517 V826 A, B, C, D
V825 A, B
V896 A, B
V4007>
4008 V4027> 4028 V850 A> B V857 A, B, C
'860 A, B, C, D
V856 IC + ID 1A + 1B 1A +.18 1A + 1B lA + 1B lA, 1B, 1C 1A, 1B> 1C, 1D PT 947, 948 PT 945, 946 PT - 949, 950 LT - 920> LIC - 921 LT 102, 106, 171, PT - 468> 469, 482 PT - 478, 479, 483 A + B AUXe Turbine IB AUxe AUXe Aux.
IB AUXe AUXe Auxe AUXe Aux. Add AUXe AUXe AUxe IB AUX'creen House AUXe IB AUXe 172
- Aux, IB
~
IB Control 236 271 278 252 236 236 256 256 219 236 236 )
236 272 219 236 271 252 236 253 272 256 256 236 272 256 253 252 8A+9A, L+Nl 4+5, F+E 4+6, F+P5 9A+10A, L+Nl 8A+9A, L+Nl 8A+9A, L+Nl 3+4,
%.H 3+4, G!H 5A+6A, N+Nl 8A+9A, L+Nl 8A+9A, L+Nl 6A+7A, N+Nl 9A+11A, Q!R 5A+6A, NNI 8A+9A'+Nl 6A+8A, Q!.N 3+4>
G+H 8A+9A, L+Nl 5+6, A+C 3+4, J+K 5A+7A, L+N 4+5,
%-H 7A+8A, Ll+Nl 9A+11A, L+Nl 5+6, F5+Pl 3+4, G+H 11+13, P+G2
< 2 nin.
5 secs 5 sec.
1 nin.
< 10 min.
<.30 min.
< 1 min.
< 1 min.
< 30 min.
< 30 nin.
< 1 min.
< 1 nin.
r.10 nin.
< 1 nin.
< 1 nin.
Continuous Operation
< 1 min.
< 30 sec.
Continuous Operation 30 min.
5 sec.
5 sec.
<10 nin.
~10 nin.
30 min.
LT LT LT (1)
LT (1)
LT (1)
LT (1)
LT LT (1)
LT (1)
LT LT LT (1)
LT Continuous Operation Conti o s Op rati.on Continuo s Operation OO M (elo in.
LT Continuous Operation LT Continuous Operation LT Continuous Operation LT Continuous Operation LT (1)
Continuous Operation LT (2)
(2)
(2)
(2)
(2)
(2)
(3)
(3) 2.8 + 6 2.8 + 6 2.8 + 6 4.8 + 4 (3)
.2.8 + 6 2.8+ 6 (3)
(3) 2.8 + 6 (3)
(3)
<1.0 + 3 (3)
(1)
(1)
(3)
(3)
(3)
N/A (5)
N/A N/A N/A N/A N/A 2.0 + 8 2.0 + 8 2.0 + 8 2.0 + 8 2.0 + 8 2.0 + 8 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A 3.0 + 4 3.0+ 4 3.0 + 4 3.0 + 4 N/A
TABLE 5-1 E ui ment Name Identification Number
~Bdldis Kiev.
Col. Lines Time After Accident Safety Function Re uired Required Calculated Qualified Operating Integrated Integrated D s si
~Do Rods
~Doss Rods Diesel Generators I and C Cabinets Reactor Trip Bieakers Safeguard Bus Safeguard Bus A+ B 16 14 DG Bldg.
Control IB Auxo Aux.
252 272 256 256 272 ll+13,.A+A2 12+13, FIB 6+7, F+G
'ga+9a, L+Nl Sa+9a, L+Nl
< 30 sec.
LT
< 30 sec.
Continuous Operation Continuous Operation seconds LT (1)
LT (1)
Continuous Operation LT (3)
(3)
(3)
<1.0 + 3 1.0+
2 N/A N/A N/A N/A N/A
- 1) Long term integrated exposure is calculated for a period of six months considering direct gamma radiation.
- 2) Safety function performed prior to post-accident recirculation phase, therefore, no increase in total dose from present considerations.
- 3) Located in area uith a negligible increase in total dose from present considerations.
- 4) Dose calculated for time period 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- 5) N/A - not available
- 37 REFERENCES 1.
Radiation Analysis Design Manual, South Carolina Electric and Gas Co., Virgil C.
Summer.
2.
E.
D. Arnold and B. F. Maskervitz, "SDC - A Shield Design Cal-culation Code for Fuel Handling Facilities,"ORNL-3041, March 1966.
3.
C., M. Lederer.,
J.
M. Hollander, and J. Perlman, "Table of Isotopes," Sixth Edition, John Wiley and Sons, Ines
, 1967.
4.
A. Tobias, "Data for the Calculatoin of Gamma Radiation Spectra and Beta Hearing from Fission Products '(Revision 3), RD/B/M2669, June 1973.
5.
Personal communication, E. E. Gutwein to C. Montgomery, Gilbert Associates, Inc.,
December 18, 1979.
6.
Letter from L. D. White, Jr. to D. L. Ziemann, "Environmental Qualification of Electrical Equipment R. E. Ginna Nuclear Power Plant",
December 1, 1978.
7.0 APPENDIX 1:
LI UID AND GASEOUS SM1PLE COLLECTION AND PREPARATION 7.1 SM1PLE COLLECTION:
NUCLEAR SAHPLE ROOII (RGB Procedure S-5) 7.1.1 Liquid (0 and Radionuclide)
Sample Collection Task Time Worker Location o
Set up Purge System o
Purge o
Sample flow to sample sink o
Sample flush to BOD bottle (0
Leuco test) o Sample for radioactivity
- analysis, 100 ml.
(dumps bottle, rinse, fillbottle) o Place sample in pass box o
Realign and close valves TOTALS 2 min.
15 min.
30 sec.
5 min.
2 min.
30 sec.
1 min.
6 min.
20 min.
7.1.2 H2 and Radioactive Sample Collection Task Time Worker Location o
Vent and Drain System o
Insert Samples Sample bomb in quick release fittings o
Open isolation valves o
Start Purge o
Purge o
Close isolation valves o
Vent and Drain System to Sink o
Remove sample bomb o
Transfer Sample Bomb to Pass Window o
Realign Valves
~ 5 min.
15 min.
3 min.
15 min.
1 min.
TOTALS 9.25 15.0
- 39 CONTAINHENT AIR SANPLE COLLECTION:
I.B. 253 ELEV., NORTH SIDE (RG&E PROCEDURE RD-1.1)
Task o
Line up valves and connect 35 cc sample collection bulb o
Purge sample through container o
Realign valves and disconnect sample container o
Transfer sample to shielded cask o
Carry shielded sample back to the H.P. Lab.
and Count Room o
Transfer sample to count safe o
Count Sample Time 2 min.
5 min.
1 min.
30 sec.
2 min.
30 sec. 200 sec.
Worker Location 3
7.3 SAlIPLE PREPARATION (PRINARY CHENISTRY LABORATORY AND COUNT ROON) 7.3.1 Isotopic Analysis of Primary Coolant (RG&E Procedure PC-5)
Task o
Remove 100 ml sample from pass box to hood; transfer 20-30 mls of 100 ml sample to beaker and add acid and boil-ing beads, place remaining 70 to 80 ml sample behind lead shield.
GAI to specify amount of shielding required
- curve showing mR/hr vs.
Pb thickness o
Boil Sample; Cool in Water Bath o
Pipe 1 ml to 50 ml volumetric bottle and dilute to 50 mls and mix.
o Transfer 50 ml. diluted sample to count room and place in steel count-ing safe; Note:
4" of steel surround detector; count room has 2
ft. shield walls.
Time 30 sec.
10 sec.
7 min.
1 min.
30 sec.
Worker Location
I 1
~
Task
= o Count Sample-(If conditions in the count room are such that have a high background or airborne contamination, the counting trailer for environmental samples can be used.)
The sample is counted 15 minutes after collection and again 1 week later.
Time 200 sec.
Worker Location Time Task Boron Analysis In Primary Chemistry Laboratory (RG&E Procedure PC-14)
Worker Location o
10 ml of primary coolant sample 30 sec.
is placed in a beaker and moved to the titration rig.
o Hannitol is added to the beaker o
Automatic titration of NaOH to sample along with pH deter-mination o
Presently sample is dumped down sink in Chemistry Laboratory presenting potential airborne problems for accident level specific activities.
10 sec.
2 min.
Primary Gas Analysis In Primary Chemistry Laboratory (RG&E Procedure PC-4)
Task o
Connect 100 cc steel sample bomb in transfer rig mounted on wall next to sample pass window.
Transfer bomb is 300 cc glass cylinder with valves.
o Transfer 100 cc sample to 300 cc glass collection bulb (purge) o Technician checks transfer operation.
Time 2 min.
10 min.
5 sec.
Worker Location
k f
0 Task Time Worker Location sec.
30 sec. 200 sec.
1 min.
. o With syringe remove 5 cc from 30 glass collection bulbs to counting vial.
o Transfer vial to Count Room and place in counting safe.
o Count Sample o
Remove liquid from transfer rig lines.
Hydrogen Analysis In Primary Chemistry Laboratory (RG&E Procedure PC-4)
Task o
After gas sample has purged 10 min.
(see DC-2) draw off 1 cc in syringe from 300 cc glass collection bulb.
o Transfer gas to partitioner (gas chromatograph) o Reading from G.C. gives hydrogen, nitrogen and oxygen.
o The gas then dissipates from G.C. to laboratory.
Time 30 sec.
30 sec.
3 min.
Worker Location 1Inside Nuclear Sample Room 2Outside Nuclear Sample Room in Low Radiation Area 3Within Arm's Reach of Unshielded Sample 10 ft. Away from Unshielded Sample 5Sample Shielded
4 0
V N
i~
ACC Cflf 8
COIL Jt]
r.,
~~~
h IC I
P%
~u-P5/
1-~P i/~
. J" C
I II
~llI
~ CCJ CC
'. ~
j(
~
~
~'
)
t Oa IPC
'K
~
~
I CC PC I
AC f
C
~
~
J.-- =.
I 1
~
1 O IV333 IAPPP P I4
~
I 1.'o1, h.V
[IMO a
~C)
(~4 a ~V a
C 'OOOV A IIV~
/VvA Aprr H lfiI1 Ipoc r p O
PP~
~O
~CCCIIICILP PP II <<1
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