ML20244A826

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Rept on Containment Purge Operational Data Gathering & Evaluation Program & Proposed Containment Purge Criteria
ML20244A826
Person / Time
Site: Clinton Constellation icon.png
Issue date: 03/31/1989
From:
ILLINOIS POWER CO.
To:
Shared Package
ML20244A825 List:
References
NUDOCS 8904180210
Download: ML20244A826 (31)
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CLINTON POWER STATION REPORT ON

(: CONTAINMENT PURGE OPERATIONAL DATA GATHERING AND EVALUATION PROGRAM L.

AND PROPOSED CONTAINMENT PURGE CRITERIA

- March 1989 8904180210 89033 1 PDR ADOCK O I g PNU-

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i CLINTON POWER STATION REPORT ON CONTAINMENT PURGE OPERATIONAL DATA GATHERING AND EVALUATION PROGRAM AND PROPOSED CONTAINMENT PURGE CRITERIA Table of Contents Section Page 1.0 Introduction 3 2.0 Summary 3 3.0 System Description 3 4.0 Data / Operational Experience Summary 4 5.0 Bases for New Containment Purge Criteria 5 6.0 Proposed Containment Purge Criteria 7 7.0 Attachments 8

1. Background of Containment Purge and Access

'lanagement Issue at CPS

2. Results of Containment Purge Operational Data Gathering and Evaluation Program
3. Containment Purge Isolation Test Data
4. Containment Access Management Program Data DLH8/JCA50 Page 2

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CLINTON POWER STATION-REPORT ON CONTAINMENT-PURGE OPERATIONAL DATA GATHERING AND EVALUATION PROGRAM AND PROPOSED CONTAINMENT PURGE CRITERIA.

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'1.0 . Introduction The purpose'of.this report is to present the results of the Containment L Purge Operational Data Gathering and Evaluation Program and the proposed l containment purge criteria to be used for subsequent fuel cycles.

Section 6.2.4.1 of Supplement 5 to the Clinton Safety Evaluation Report (NUREG 0853) (SSER 5) requires Illinois Power' Company-(IP), before startup after the first refueling outage, to-provide the NRC staff with a p reevaluation of the need to use the containment purge systems during-plant operational modes 1, 2, and 3 and to provide criteria for purge.

operation which will be used for the remainder of the plant life.

Further detailed background information on these issues is contained in Attachment-1.

2.0 Summary In order to develop the proposed containment purge criteria, data from three different sources were used. Section 4.1 of this' report discusses data collected during normal plant operation. These data indicate relatively low levels of airborne radioactivity in the containment building atmosphere during normal plant operations with' continuous purge operation. Section.4.2 discusses a special test performed to establish airborne radioactivity buildup and drawdown rates and equilibrium levels while the containment purge systems were isolated, and the aectiveness of the purge systems in removing airborne radioactivity when restarted.

JThe test indicated that with the containment purge. systems isolated, airborne radioactive contamination reached an equilibrium. level of approximately 0.05 of the Maximum Permissible Concentration per 10CFR20 in about 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Section 4.3 presents data from the Containment Access Management Program which show that containment entries are necessary on a daily basis and during all shifts.

Additional factors affecting dose to the workers at Clinton Power Station (CPS) and potential dose to the public used in developing the final proposed containment purge criteria are discussed in Section 5.0 of this report. The proposed criteria are presented in Section 6.0.

3.0 System Description

A schematic of the containment purge systems is shown in Figure 1-1 of Attachment 1. There are two Containment Building-Vent / Purge Systems; the Containment Building Ventilation (CBV) system and the Continuous Containment Purge (CCP) System. The CBV system provides 16,000 cubic feet per minute (CFM) of ventilation and has two 36-inch nominal diameter DLH8/JCA50 Page 3

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. s containment penetrations (one supply and one exhaust). The CCP. system-provides.8,000 CFM air flow and has two 12-inch nominal diameter penetrations (one supply and one exhaust). For bcth systems, normal exhaust flows directly to.the common station He. sting, Ventilation and Air Conditioning (HVAC) vent stack. If radioactivity is detected in the exhaust air, the flow can be diverted by the operator through the high

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efficiency particulate and charcoal filters of the drywell purge system.

If high radiation levels are detected in the exhaust air, the containment is automatically isolated and the Standby Cas Treatment System for the secondary containment is started.

4.0 Data / Operational Experience Summary During the first fuel cycle, the CCP system was used.in all modes of operation; the CBV system was not used in modes 1, 2, or 3. The CCP system was operated almost continuously during plant modes 1, 2, and 3 to support multiple daily personnel entries. The CCP system was only shut down (daring modes 1-3) for relatively short periods. to vent the drywell and to perform valve operability checks on the isolation valves for the-CCP and the CBV systems. The drywell was vented in order to maintain the drywell-to-containment differential pressure within the -0.2 to +1.0 psid range in accordance with the requirements of Technical Specification; 3.6.2.5.

4.1 Containment Purge Operational Data Gathering and Evaluation Program Illinois Power collected pertinent data on the operation of the plant during the first fuel cycle as outlined in Licensing Review Group-(LRG) II Position Paper 4-CSB. This paper was applied to the Clinton licensing docket by IP letter number U-0731, dated September 10, 1984 The Position Paper specified the guidelines for recording data on containment airborne radioactivity, reactor coolant and suppression pool radioactivity, ventilation operation, sump flows, reactor power level, and other data. This data would be used to determine if quantifiable relationships between such parameters

.could be established and utilized in developing permanent purge guidelines.

The data gathered for this program indicate that for the first fuel cycle airborne radioactivity levels in the containment building have been low. During this cycle, fuel reliability performance was very good (see Figure 2-8). The fuel reliability indicator for the first fuel cycle averaged about 20 micro-curies per second at the steam jet air ejector (main condenser off-gas retreatment). The telease rates were expected by design to be 25,000 micro-curies per second for radioactive isotopes other than iodine-131 and 100 micro-curies per second for iodine-131 as indicated in the Updated Safety Analysis Report (USAR). The remainder of the data from this program is presented in Attachment 2.

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,. i e a 8 4.2 Containment Isolation Test l

Since the data taken during normal ventilation operation was l inconclusive because of low radioactivity levels, a test was conducted in which the CBV and CCP systems were isolated for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The purpose of this test was to establish buildup and drawdown rates and equilibrium levels of airborne radioactivity in the containment building. This test showed that with the CBV and

, CCP systems isolated, the airborne radioactivity level built up to l about 0.05 of the Maximum Permissible Concentration (MPC) per i 10CFR20. This data, while reptasentative of the first fuel cycle, may not be representative of future plant conditions. Total activation and fission product inventories have not reached equilibrium levels, and the associated coolant activity is relatively low. The results of this test are provided in Attachment 3.

The purge isolation test also showed that because of the use of instrument / service air in the containment building, it was necessary to exhaust air from the containment building briefly several times a day during the test in order to maintain the containment building-to-secondary containment differential pressure within Technical Specification section 3.6.1.6 limits of -0.25 to +0.25 psid. The proposed operating criteria presented in Section 6.0 of this report take this consideration into account.

4.3. Containment Access Management Program Section 6.2.4.1 of Supplement 2 to NUREG-0853 (SSER 2) required Illinois Power to develop a Containment Access Managament Program to minimize access cime in the containment. This program was submitted to the NRC as IRG II Position Paper 5-CSB and it was applied to the Clinton licennsng docket by IP letter number U-0731, dated September 10, 1984. I*. was summarized in SSER 5 to require plant procedures to consolidate containment entries and to collect and evaluate data during the first fuel cycle to identify improvements to plant procedures or hardware modifications.

The Containment Access Management Program has verified the need for uultiple daily entries into the containment for operation, i surveillance, and maintenance. The actual required man-hours l (approximately 7000 man-hours per year) validated the estimate of  :

8370 man-hours made prior to licensing nnd is of the same magnitude as the access requirements at other BV plants. Additional j details of the required access requirements are contained in ,

Attachment 4. '

5.0 Bases for New Containment Purge Criteria

}

The following factors have been considered in the development of the j proposed containment purge criteria. j I

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5.1 Radiological Dose to Workers Based on the demonstrated need for multiple containment entries into the BWR Mark III containment, the following radiological dose to the workers is predicted for the listed levels of airborne radioactivity buildup.*

Containment Dose Per Dose for Airborne Radio- Year 40 Years activity Level

.05 MPC(cycle 1 buildup) .8 Man-REM 32 Man-REM

.1 MPC 1.6 Man-REM 64 Man-REM

.25 MPC 4 Man-REM 160 Man-REM As this table shows, curtailment of containment purging would cause added dose to the workers, contradictory to the commitment to maintain dose As Low As Reasonably Achievable (ALARA).

5.2 Radon Dose to Workers Although exposure to rr. don is not controlled by the CPS operating license, it has been shown to build up when the containment is isolated (See Figure 3-4 of Attachment 3). Radon produces an alpha dose to'the lungs. Environmental Protection Agency regulatory guidance for control of radon has become more strict. The most efficient way to remove radon is by engineered ventilation systems.

The requirement to maintain radon dose ALARA supports the need for regular purging of the containment building.

5.3 Containment Isolation Requirements Continuous operation of the 12-inch CCP system with the 36-inch CBV system isolated meets the requirements for containment isolation of j Branch Technical Position CSB 6-4 as stated in section 6.2.4.1 of 1 SSER 2. SSER 5, sections 6.2.4.1 and 3.10.3.1 conclude, based on testing and analysis, that the 12-inch valves are capable of closing against design basis accident containment pressure. As discussed in )'

the CPS USAR Appendix D,Section II.E.4.2 these valves are also provided with diverse actuation signals.

In addition, it is noted that previous risk studies of reactor plants (Reactor Risk Reference Document (RUREG-1150), Interim Reliability Evaluation Program (NUREG/CR-2802, NUREG/CR-3085),

Reactor Safety Study Methodology Applications Program (NUREG/CR-1659), and Reactor Safety Study (WASH-1400)) have shown that the probability of severe accidents requiring containment ,

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  • based on 7000 man-hours in containment per year.  ;

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' isolation are very low (on the order of 1E-4* to 1E-6 per reactor j year). From industry data (IDCOR Technical Report 86.3 D1, l

" Individual ~ Plant Evaluation Methodology for Boiling Water . 1 Reactors," Table A.2-1), the probability of f ailure of an air operated containment isolation valve to close:is on the order of 1E-4. The combined probability of even one of the 12-inch j containment isolation valves failing to close when needed to protect -

the public is less than 1E-8. Containment ventilation penetrations are equipped.with redundant valves to provide further assurance of a closure.

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5.4' Reliability of Isolation Valves and Dose to the Public ,

1 Any purge regime that would lead to daily cycling of the 12-inch containment isolation valves would increase the probability of failure of these valves.to close were an accident to ever occur which would require these valves to function. Daily cycling would also lead to significantly increased maintenance' requirements and, thus, increased worker dose.

6.0 Proposed Containment Purge Criteria In order to balance the concerns for the potential dose to the public.

should the 12-inch valves fail to close in the event of an accident and for the projected dose to' CPS workers that would. result'from reduced purge' operation, the following proposed purge criteria will be instituted j at the start of the second fuel cycle.

The following criteria shall apply to operation of the136-inch CBV and 12-inch CCP systems in plant modes 1, 2, and 3. These criteria are flexible enough to support CPS operation yet provide a high degree of protection for the health and safety of the public.

6.1 - Both the CBV and CCP systems must be shut down and their respective containment isolation valves closed while venting the drywell into the containment buildt g to preclude a direct path from the drywell to the environment.

6.2 The 36-inch CBV system shall not be operated beyond the 250 hour0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> per year time limit specified in Technical Specification 3.6.1.8.a.

6.3 The 36-inch CBV system and the 12-inch CCP system shall not be operated simultaneously. The isolation valves for either system shall be verified shut before opening the valves for the other system.

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  • 1E-4 represents abbreviated scientific notation for the value 1x10 or

.0001.

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6.4 LEither the CBV or the CCP system may be operated to maintain the containment building air pressure _within'the limits required by the Technical Specifications, or to reduce airborne radioactivity levels

.in the containment building to maintain worker _ dose ALARA, or for

,. the duration of 'any welding task or any task in which chemicals requiring ventilation are used.

6.5' Continuous operation of the CCP system is allowed except while

. venting the drywell (see criteria 6.1) or while operating the CBV system (criteria 6.3) .

7.0 Attachments Attachment 1 - Background of Containment Purge and Access Management Issue at CPS Figure 1-1 CPS Containment'HVAC Schematic Attachment 2 - Results of Containment Purge Operational Data Gathering and Evaluation Program Figure 2-1 ' CPS Containment Area Radiation Dose Rates Figure 2-2 CPS Containment Airborne Radioactivity - Noble Cases Figure 2-3' CPS Containment Airborne Radioactivity - Particulate Figure 2-4 CPS Reactor and Suppression Pool Water - Cross Gamma and Dose Equivalent Iodine Figure 2-5 CPS Containment Building Air Temperature Figure 2-6 CPS Containment Sump Flows Figure 2-7 CPS Reactor Power Figure 2-8 CPS Main Condenser Off-Gas Radioactivity -

Retreatment Noble Gases Attachmant 3 - Containment Purge Isolation Test Data Figure'3-1 Containment Isolation Test - Radioactivity Buildup -

All Isotopes /All Sample Locations Figure 3-2 Containment Isolation Test - Radioactivity Classes - All Sample Locations Figure 3-3 Containment Isolation Test - Containment and Drywell Pressures a Figure 3-4 Containment Isolation Test - Radon Levels Attachment 4 - Containment Access Management Program Data Figure 4-1 CPS Containment Access.- Average Number of Entries per Day and Average Man-hours per Day By Month ,

Figure 4-2 CPS Containment Access - Man-hours and Number of l Entries by Task Figure 4-3 CPS Containment Access - Entries and Man-hours by Department Figure 4-4 CPS Containment Access - Man-hours in Containment Building by Hour of Day Entry Started DLH8/JCA50 Page 8

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Attachment 1 Background of Containment Purge and Access Management Issue at CPS The Clinton Power Station is a GE BWR/6 with a Mark III containment building.

The Mark III containment was designed to provide a larger containment volume than previous containment designs. However, since equipment for several systems is located in the containment, the Mark III design requires frequent personnel entry for equipment surveillance and maintenance. In order to support personnel entries, some containment purging is required to maintain air quality. To keep the amount of purging required to a minimum, stations l' with the Mark III containment were required to implement the Containment Access Management Program (CAMP) and the Containment Purge Operational Data Gathering and Evaluation Program.

The objectives of these programs were to minimize the time personnel are required to be in the containment building and to rovide criteria for containment building purge operations to be used for the remainder of plant life based on evaluations of first cycle operating data.

The CAMP involves providing plant procedures to consolidate containment entries and collecting and evaluating data to identify improvements to plant procedures for minimizing containment entries. The Licensing Review Group (LRG) II Position Paper 5-CSB provided an estimate of 8370 man-hours per year required in the containment based on a December 5, 1977, General Electric Mark III Containment Dose Reduction Study. The data collected indicates that on a yearly basis CPS has expended fewer man-hours in the containment building during power operation than was predicted; however, many daily entries are still required. The graphs in Attachaent 4 summarize the access requirements.

The original design for the containment building HVAC system at Clinton consisted of 36-inch supply and exhaust-lines and 4-inch bypass lines. It was proposed to use the 36-inch lines for_ continuous ventilation during normal plant operating conditions, and the 4-inch bypass lines for post-LOCA hydrogen control. The NRC staff reviewed this design against the criteria specified in Branch Technical Position (BTP) CSB 6-4 Rev. 2 (7/81), " Containment Purging During Normal Plant Operation," and determined that the continuous use of the 36-inch lines would not meet those criteria. The Clinton SER states, "The staff believes that purging should be minimized during normal reactor operation and should not be relied on for temperature and humidity control.

Therefore, the staff requires the applicant to provide a realistic estimate of the number of hours per year that purging is expected through each purge valvo, and a justification for this use."

In response to those comments IP proposed a redesign of the containment ventilation system. Essentially, 12-inch supply and exhaust lines and penetrations were added along with their own supply and exhaust fans. These components tie into the existing duct work inside containment and can provide continuous 8,000 CFM purging during operation in plant modes 1, 2 and 3. In  ;

addition, mechanical stops were installed on the 36-inch valves to limit them )

to only open 50* (90' is fully open), A schematic of the CBV and CCP systems is shown in Figure 1-1. Operation of the 36-inch lines would be limited to Page 1-1

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  • 4 e 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> a year in plant modes 1, 2, and 3. Also, the 12-inch lines and the 36-inch lines would not be operated simultaneously during Modes 1, 2, and 3.

In Supplement 2 to NUREG-0853 (SSER 2) the NRC staff accepted this design and operating philosophy for purge application through the first refueling outage.

Also, in SSER 2 it was reported that IP had committed to developing and implementing the Containment Access Management Program and the purge data collection program as well as providing interim purge guidelines for the first fuel cycle. The key requirement reported in SSER 2 was as follows:

"The NRC staff requires that the applicant propose the purge criteria to be used for the remainder of the plant life, based on the above cited programs, before startup after the first refueling outage. The applicant has agreed to abide by this requiremen..."

The BWR/6 Licensing Review Group (LRG)-II position papers describing the Containment Purge Operational Data Gathering and Evaluation Program and the Containment Access Management Program were transmitted to the NRC by IP letter number U-0716, dated June 29, 1984. In IP letter number U-0731, dated September 10, 1984, these program position papers were incorporated into Clinton's license application. Also, attached to the September letter was a paper proposing the interim guidelines for operation of the containment purge system up to the first refueling outage.

NRC's Supplement 5 to NUREG-0853 (SSER 5) contains summary descriptions of these programs. The concluding paragraph states, "The staff has reviewed the above programs and finds them acceptable through the first refueling outage.

Before startup after the first refueling outage the applicant shall provide the staff with a reevaluation of the need to use the containment purge systems during operational modes 1 through 3. Criteria shall be provided which will be used for the remainder of the plant life."

These programs also affect Technical Specification 3.6.1.8.a which limits operation of the 36-inch supply and exhaust line valves to less than 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> per year. A footnote to this Technical Specification states that 3 months after completion of the first refueling outage the limit will drop to 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> per year.

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s o Attachment 2 Results of Containment Purge Operational Data Gathering and Evaluation Program Attachment 2 comprises graphs of general operational and radiological data taken during the first fuel cycle in accordance with the Licensing Review Group (LRG) II position paper on Containment Purge Operational Data Gathering and Evaluation Program. Data was taken from startup after the first planned outage which was performed just af ter the warranty run (November 23, 1987) until shutdown fot the first refueling outage (January 2, 1989) This data was collected with the expectation that short-term and long-term containment airborne radioactivity levels could be correlated to certain plant parameters (i.e. reactor coolant iodine concentration, identified and unidentified leak rates, etc.), and conditions. No quantifiable relationships could be established which could replace direct measurement of airborne isotopes. In fact, it was originally believed that iodines would be the most significant isotopes for dose considerations when in reality particulate and noble gases were more dominant during this cycle. The routine station operation daca specified in the LRG II position paper are presented in the following figures; however, it does not appear to be particularly useful for final purge criteria development. Direct airborne radioactivity measurement appears to be the most practical approach.

Breaks in the data during the months of March, April, July, and November 1988 reflect plant outages.

Figure 2-1 is a graph of area radiation dose rates reported by the area radiation monitors located in containment. It is noted that dose rates increased at the end of the cycle. This increase correlates with direct measurements taken to support work planning for refueling. The reason for this increase has not yet been verified, but it apparently is the result of radioactive crud breaking loose and spreading through the piping.

Figure 2-2 and 2-3 are graphs of noble gas and particulate airborne radioactivity measured at the three continuous air monitors (CAMS) located in the containment building. It should be noted that the CAMS are also designed to detect airborne radioactive iodine but that all iodine readings during the first fuel cycle were below the useful range of the instrument.

Figure 2-4 is a plot of gross gamma measured in samples of reactor and suppression pool water and also a plot of dose equivalent iodine determined from samples of reactor water. Suppression pool water quality was maintained through the operation of the suppression pool cleanup system. The reactor coolant iodine concentration was maintained well below the Technical Specification 3.4.5.a limit of 0.2 micro-curies per gram.

Figure 2-5 is a graph of containment building air temperature through the first fuel cycle. This graph contains no significant information with respect to purging, but the extreme heat wave of July-August, 1988, is evident.

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Figure 2-6 is a plot of the flow rates for the containment sumps. It shows

= equipment drains- (identified leakage) sump flow rate, floor drains (unidentified leakage) sump flow rate, and the total of the two. Sump flows ,

are indications of primary system leakage'uad should have a direct. impact on containment atrborne radioactivity concentration.

Figure 2-7 is.a graph of reactor power level through the first fuel cycle.

Figure 2-8 provides main condenser off-gas retreatment noble gas activityLin micro-curies /cc and indicates the good fuel reliability for the first cycle.

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Containment Purge Isolation Test Data The graphs in this attachment depict the data taken from the containment purge isolation test which was performed November 4-7, 1988. This test was conducted to determine the equilibrium conditions that would be reached under long-term isolation conditions. Air samples were taken at the locations shown in the following table.

Radiological Sample Locations Location Location Number (Elev./Az.) Description 1 803' 80' General area 2 778' 45' General area 3 755' 185" General area 4 789' 340' Reactor Water Clean-up "A" Heat exchanger Cubicle Entrance 5 789' 20* Reactor Water Clean-up "B" Heat exchanger Cubicle Entrance 6 778' 240* Reactor Water Clean-up Backwash Receiver Tank Cubicle Entrance Figure 3-1 is a graph showing the buildup of airborne radioactivity in Maximum Permissible Concentration (MPC) from the start of purge isolation and the subsequent drawdown rate from the end of isolation. Although the data are widely scattered, a curve fit to the data shows that activity built up to an average of nearly 0.05 FTC. A bounding curve would yield about .06 MPC. Note that the buildup reaches equilibrium in about 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, yielding a build-up half-life of approximately six hours. Also, once purge was restored, the airborne radioactivity was drawn down quickly to a low level. The drawdown rate was such that after 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />, residual airborne activity was very small, yielding a removal half-life of about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

Figure 3-2 provides a breakdown of activity between particulate, noble gas, and iodine activities. This graph shows that between 80 and 90 percent of the total activity was particulate. The noble gas activity was approximately one-fifth of the total activity. The iodine activity was lower than anticipated.

Figure 3-3 provides the containment and drywell pressure history during the test. As the graph shows, the containment had to be vented approximately every six hours in order to maintain primary-to-secondary containment differential pressure between -0.25 and +0.25 psid as required by Technical Specification 3.6.1.6. The graph also shows that the drywell was vented to the containment during the test in order to maintain the drywell-to-containment differential pressure within the limits of .2 to +1.0 psid as specified by Technical Specification 3.6.2.5. No radiological effect of the drywell venting was noted on any of the containment atmosphere radioactivity samples.

Figure 3-4 is a plot of radon levels as a fraction of the working level during l the test period. The working level is defined in 10CFR20 as a combination of radon and daughter products result 1ng in a certain alpha energy emission level The figure shows that the working level fraction did increase with the purge systems isolated during the test and was reduced after ventilation was restored.

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  • e Attachment 4 Containment Access Management Program Data The graphs and charts in this attachment are based on data taken for the Containment Access Management Program (CAMP). They cover the first fuel cycle for.the Clinton Power Station. Data was taken from startup after the first planned outage which wer performed just af ter the warranty run (November 23, 1987) until shutdown for the firse refueling outage (January 2,1989) for plant modes 1, 2, and 3.

Figure 4-1 is a graph of the average number of entries per day and the average number of man-hours per day for containment building entries for each month of the first fuel cycle. The increased number of entries near the end of the fuel cycle reflect preparation for the refueling outage (i.e. work on fuel handling equipment, ALARA walkdowns for job planning, etc.).

Figure 4-2 is a bar graph of the number of entries and man-hours in the containment by the type of work performed. This graph shows that the two major tasks performed in the containment building are maintenance and surveillance.

The category of "other surveillance" includes fire watch inspections, security inspections, Industrial Safety inspections, NRC inspections, Radiological Protection inspections, Chemistry Sampling, and Operations inspections.

Figure 4-3 is a bar graph of the number of entries and man-hours in the containment by departments. This graph indicates that the Operations and Radiation Protection departments make the most entries. Also indicated is the diversity of skills and work functions which require access to the containment.

Figure 4-4 shows man-hours in the containment building plotted against the time of day the entry started. It indicates that entries are required at all hours of the day. The day shift (0700-1500) was the busiest shift for containment access, accounting for 57 percent of all containment work man-hours. The afternoon shift (1500-2300) accounted for 21 percent of the containment work man-hours. The night shift (2300-0700) accounted for 22 percent of man-hours.

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