ML19305C443

Forwards Responses to Requests & Questions Raised During NRC 800225 & 26 Site Visit to Review Util Implementation of Lessons Learned Task Force short-term Recommendations

ML19305C443
Person / Time
Site:	Davis Besse
Issue date:	03/21/1980
From:	Crouse R TOLEDO EDISON CO.
To:	Reid R Office of Nuclear Reactor Regulation
References
RTR-NUREG-0578, RTR-NUREG-578 TAC-12453, NUDOCS 8003280560
Download: ML19305C443 (48)

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TOLEDO

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Docket No. 50-346 w.a License No. NPF-3 Serial No. 601 March 21, 1979 Director of Nuclear Reactor Regulation Attention:

Mr. Robert N. Reid, Chief Operating Reactors Branch No. 4 Division of Operating Reactors United States Nuclear Regulatory Commission Washington, D.C.

20555

Dear Mr. Reid:

On February 25 and 26, 1980 Mr. C. Long of your staff led an NRC team visit to the Davis-Besse Nuclear Power Station Unit 1 facility to review the Toledo Edison implementation of short-term recommendations of your Lessons Learned Task Force. These recommendations were documented in your letter of October 30, 1979. Attachments A-E of this letter respond to requests and questions raised during the site visit, Very truly yours,

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THE TOLECO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO. OHIO 43652 8 0 032 80 F6 o

ag Docket No. 50-346 License No. NFP-3 Serial No. 601 March 21, 1980 Attachment A DAVIS-BESSE UNIT 1 NRC LESSONS LEARNED SITE VISIT DOCUMENTATION FEBRUARY 25-26, 1980 The response to questions raised during the site visit are organized by paragraph numbers as referenced in NUREG 0578 "TMI2 Lessons Learned Task Force Status Reports and Short Term Recommendations".

2.1.1 1.

Document magnitude of heaters needed to maintain natural circula-tion.

(Include startup test data to support conservatism.)

Response

The proper operation and kilowatt (KW) output of the essential pressurizer heaters required to maintain natural circulation has been addressed during startup testing per TP 600.13 " Pressurizer Operational and Spray Flow Test".

This test procedure is described under the Test Abstract 600-13 in Chapter 14 of the DB-1 Final Safety Analysis Report (FSAR).

It demonstrated that the essential heater KW output did not exceed 126 KW.

Section 7 of TP 600.13 verified heater demand to compensate for ambient heat losses and pressurizer mini spray flow.

Total heater output required for this compensation was 81.79 KW.

2.

Document time when above heaters must be placed on emergency diesel.

(Related to NC)

Response

Babcock & Wilcox has provided information indicating as a minimum two hours is acceptable to energizing the p;:essurizer heaters.

3.

Document that the capacity of emerrancy diesels is adequate to take the heaters with LOCA loads includ3d.

Response

The capacity of the installed 2600 KW emergency diesel generators to provide emergency power to the pressurizer heaters with loss of cooling accident (LOCA) loads included has been addressed during startup tercing per TP 0310.02, " Integrated Safety Features Actuation System (SFAS)

Test".

This test procedure, described under the Test Abstract 310-02 in Chapter 14 of the DB-1 FSAR, demonstrated that the LOCA loads were less than the total rating of the emergency diesel generators. Section 7 of TP 0310.02 verified the sequencer function of the SFAS on a loss of 1

power. During sequencing operation of loads, the initial in rush of

p.

A-2 pump starts went to a maximum of 2250 KW, with a total load of 1800 KW during operation with all LOCA loads connected. Manual operation of these heaters are addressed in SP 1103.05 " Pressurizer Operation".

2.1.3.a 1.

Document qualification program schedule for preamplifiers associ-ated with the acoustic valve position indicating monitoring system.

Response

The current projected schedule for environmental qualification of the TEC supplied preampliers is October, 1980. This program will use proto-type testing information.

2.1.3.b 1.

Document range of temperature inputs to sub-cooled meter.

Response

The reactor coolant saturation meter will initially utilize a non-safety grade hot leg temperature range of 120 F to 920 F.

This temperature range will be upgraded as described in 2 below.

2.

Document when temperature inputs and power supply to the sub-cooled meter will be safety grade.

Response

The schedule for safety grade qualification of the reactor coolant saturation meter is dependent on the development, purchase, and installa-tion of a qualified millivolt to volt converter. This is currently projected to be available at DB-1 by December,1980, and installed during the first extended outage after receipt.

2.1.4 1.

Document list of essential /non-essential systems and incident level for isolating non-essential syscems.

Response

Attachment B provides a list of mechanical penetrations, their associ-ated systems and the safety feature system actuation incident level for which non-essential penetrations are isolateJ.

2.

Define essential /non-essential systems for containment penetrations.

Response

Attachment B provides a list of mechanical penetrations and defines them as essential or non-essential.

o A-3 3.

Document the change of the component cooling water (CCW) system isolation to incident level 3 or 2.

(Discuss why 24 psig is needed for CCW.)

Response

In review of the discussions between Mr. C. Long of your staff and Mr. F. Miller of my staff, Toledo Edison has identified a condition during the DB-1 analyzed event of a main steam line break (MSLB) inside the containment vessel where the incident level 3 is still desired.

During this MSLB the containment vessel would exceed 4 psig but not 24 psig.

It is desirable to maintain component cooling water to the reactor coolant pump seals to continue the capability for forced circulation in the reactor coolant system. For this reason Toledo Edison will not be proposing a change in safety features actuation signal incident level for the component cooling water system.

2.1.5.c 1.

Document that the H dilution system is operated entirely from the 2

Control Room and does not require changing procedures to operate the system or shielding.

Response

A check of equipment locations and the system operacing procedure SP 1104.55,

" Containment Hydrogen Dilution and Hydrogen Purge" verifies that all required equipment for system operation during an accident condition can be operated from the Control Room.

2.1.6.a 1.

Provide a summary description of the leak reduction program as it is operating at the plant. How will leakage be kept ALAP including periodic tests, use of walkthroughs, area monitors and effluent monitors.

Response

Attachment C provides a summary description of the leakage reduction program. The leakage reduction tests are scheduled to be repeated at least every refueling outage.

In addition to these tests, the DB-1 Chemistry and Health Physics Department records daily the radiation levels from four radiation monitors. These include the station vent monitors (2), the fuel handling area exhaust monitor and the radwaste area exhaust monitor.

The results of these daily readings are trended and would reveal whether area radiation is increasing within the station or in a particular area.

If this logging shows a trend, further investi-gations are made as to the sources. Visual indications of leakage are also converted into corrective maintenance work orders.

o A-4 2.

Provide measured leakage from systems listed in your 1/17/80 letter.

Provide schedule to complete any testing not done and give system.

Response

Attachment C provides a summary of the measured leakage results for the initial effort of the leakage reduction program.

The only test not com-pleted at this time is ST 5051.04, "ECCS Refueling Surveillance Test",

which checks the leakage for the high pressure injection (HPI) system and the " piggyback" piping between the decay heat and HPI System. This testing is planned to take place during the upcoming refueling outage scheduled to commence April 10, 1980.

3.

Provide response to our request in 10/30/79 related to our letter regarding North Anna and Related Incidents dated 10/17/79.

Response

In comparing the North Anna incident to Davis-Besse, the makeup tank relief valve (PSV 1893) would provide the flow path as stated in the North Anna NRC letter. At Davis-Besse, this relief valve discharges to the reactor coolant drafn tank which in turn discharges to the gaseous radwaste system or to the Clean Waste Receiver Tanks in the Clean Rad-waste System. The receiver tanks have a rupture disc on them. Consid-ering this flow path, an incident such as the North Anna incident, would not occur unless the rupture disc on the Clean Waste Receiver Tank would be ruptured.

2.1.6.b 1.

Document effect of having radioactivity in the makeup and purifi-cation system on access in plant.

Response

Toledo Edison has provided containment isolation signals to the makeup and purification system. This system is not required under post-accident conditions. As a recult a detailed plant shielding review was conducted without considering these systems in service.

This was filed with the NRC by Toledo Edison on January 31, 1980 (Serial No. 585).

This evaluation has been updated and provides decay curves and time motion studies for areas requiring post-accident access. This revised evaluation is provided as Attachment D to this letter. Although non-essential in the post accident condition, at your staff's request Toledo Edison has completed a review of the effects on plant accessibility as a result of contaminated letdown, makeup, purification and gaseous radwaste systems. The results indicate no additional ef fetes on plant accessibility required in a post-accident condition. A discussion of this review is provided as Attachment E to this letter.

2.

Document effect of Auxiliary Building radiation levels on environ-mental qualification of ESF equipment. Provide schedule (3/31/80).

e A-5

Response

Toledo Edison will provide a response by March 31, 1980.

3.

Provide Radiation Zone Maps, decay curves and time motion study.

Response

This information is provided in Attachment D to this letter.

2.1.7.b 1.

Describe the safety grade flow instrumentation for Auxiliary Feed-water flow.

Response

Toledo Edison will provide one safety grade flow instrument for each auxiliary feedwater train at DB-1.

This will be a differential pressure device set across an orifice located in each auxiliary feedwater line.

These will be downstream of any crossconnect to ensure indication of auxiliary feedwater flow delivered to each steam generator.

The power supplies, instrumentation and displays will meet safety grade require-ments identified in the DB-1 FSAR. As identified in Toledo Edison's response of November 21, 1979 (Serial No. 559) Davis-Besse's existing safety grade steam generator level indication is utilized to meet the redundancy requirements for indication of flow to each steam generator.

These are currently scheduled for installation during the spring refuel-ing outage of 1981.

2.1.8.a i

1.

Document that procedures for taking prompt samples and analyzing i

samples on-site with present situation (high radiation samples) have been reviewed and revised if necessary.

(Both containment air and RCS samples)

Response

AD 1850.04 " Post Accident Radiological Sampling and Counting" provides directions on how to collect highly radioactive samples from the reactor i

coolant system and containment.

This procedure has efficient methods for analyzing the samples.

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2.

Document the review of the radiation counting facility for back-ground (noble gas - purge) and alternative sites available to handle problems.

(Lobby and water treatment lab sites)

Response

If the airborne radioactivity in the counting room is sufficiently low that personnel can enter the area, gamma spectroscopy counting can be performed at this location. This can be accomplished because the count-ing cave for the Ge(Li) detector is equipped to be purged with bottled I

l' l

a A-6 air.

In the event the gamma spectroscopy equipment has to be removed from the plant to the water plant lab, interference from airborne radio-activity should not effect counting highly radioactive samples.

3.

Describe the proposal for long term containment air / liquid sample analysis capability and schedule.

Response

This response will be provided with Item 2.1.6.b(2) on March 31, 1980.

2.1.8.b 1.

Document you have interim procedures for quantifying noble gas /

radioiodine releases.

Response

AD 1827.10, " Emergency Off-Site Dose Estimates" is used to quantify noble gas /radioiodine releases.

2.

Document what will be done to quantify releases from steam dump valves / auxiliary feedwater steam turbine. Schedule.

Res ponse EP 1202.57 " Steam Generator Leak" is used to quantify releases from the atmospheric relief valves / auxiliary feedwater steam system.

In addition, Toledo Edison is currently having developed a methodology to provide quantification of steam system releases utilizing the cur-rently installed steam line radiation monitors.

This methodology for the currently installed equipments will be available prior to startup af ter the upcoming refueling outage currently scheduled for April 10, 1980.

3.

Provide answers to NRC 10/30/79 letter questions on Pages 33-35.

Response

The item numbers below correspond to those on Pages 33-35 of the referenced letter.

1.a.1 The instrument used for interim monitoring of noble gases is an RD-17D probe at5 ached t a RM-16 count rate meter which has a range up to 1x10 pCi/cc.

The calibration is based on the average energy of fission gases.

The calibration frequency is semiannually. The RD-17D probe is calibrated using a radiation source and the RM-16 using a pulser.

1

e A-7 11 The sampling location is in the non-rad supply air and exhuast equipment room.

In order to correct for background radiation, a second RD-17D probe reading is used.

This probe is adjacent to the probe used for messuring the noble gases in the sample tube and is separated by a lead brick wall.

iii The interim noble gas sampler is located on the same elevation as the Control Room, within a one minute walking distance.

iv Radiation reaoings are continuously read out on the RM-16 rate meter.

v The noble gas monitor can be operated with normal AC power or battery which will be operable for 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br />.

2.A.I.a The station vent interim sampler assembly has provisions for collecting particulate and radioiodine samples.

Interference from noble gases for iodine analyses is reduced by the use of silver zeolite cartridges.

b The sampling location is in the non-rad supply air and exhaust equipment room.

c Radiation fields, where the interim sampler is located, will be sufficiently low that individuals will have access to collect the filters.

In order to minimize high radiation from accumulating on the filters, the collection time will be reduced.

d Interference from high levels of radioactive noble gases is reduced by purging the Ge(Li) detector cave with bottled air if the equipment is used in the normal counting room.

If the counting has to be set up outside of the station, the noble gases are not expected to interfer with samples which are highly radioactive.

e In case of loss of normal AC power to the motor for the sample collection pump, the motor will be supplied with power from a i

diesel generator. This change will be made during the Spring, 1980 refueling outage. Analysis equipment requires AC power, however, the equipment can be transported to a location where AC power is available, if necessary.

J 4.

Document position on installing high activity effluent monitors /in-containment monitors.

Response

This response will be provided with 2.1.6.b(2) on March 31, 1980 2.1.8.c 1.

Document the portable instrumentation available, where located (TSC, CR), proper training and procedures completed.

e A-8

Response

Portable instrumentation available for the Control Room and interim Technical Support Center will include a radiation survey meter, air sampler and a scaler-counter system to analyze air filters.

Procedures are available for the use of the equipment and Health Physics personnel have been trained. This equipment will be kept in the instrument calibra-tion room.

In addition to the above, the computer alarm printouts will provide the TSC with an on-going chronology of the event. The approximately 320C alarms, and monitored contacts provide current status of plant problems which can be correlated with the data, or further investigated via phone communications.

It is felt from the station's experience in the investigation of unit trips, which utilize much the same information, thac sufficient data will be available to effectively analyze and assess an event.

2.2.1.b 1.

Document the long term training program for the Shif t Technical Advisor.

Response

The long term training program for Shift Technical Advisors at this time is not completely formulated. The total program will be somewhat depen-dent upon the qualifications of the five people entering the program and their previous experience. As a minimum, the program will consist of the following:

a)

A six-week Pressurized Water Reactor Technology Course taught by the Training Department. The PWR Technology course compre-hensively covers all systems of the plant.

b)

A qualification maual with qualification checkoffs required to be obtained from licensed instructors or operators which will focus on the safety systems in particular.

c)

A fourteen-day course taught by Babcock & Wilcox specifically designed for training of Shif t Technical Advisors.

It will cover the accident analysis, thermo-hydraulic analysis of the reactor coolant system, and will include approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> on the B&W simulator, d)

Training on the Station's Technical Specifications.

2.

Document how the operational experience assessment function has been implemented.

A-9

Response

The operational experience assessment function is required per AD 1839.04,

" Shift Technical Advisor". This administrative procedure specifies the requirements in detail of what reports and industry information must be reviewed and assessed to determine its safety significance as it applies to Davis-Besse. The results of this review are put on a required read-ing list for the Shift Technical Advisors for review.

The required reading list is also outlined in AD 1839.04.

1.

Document lack of present capability of plant computer to display data to interim TSC and plant personnel walking specific data to interim TSC.

Response

The station computer currently does not have the availability to allow additional access capability for a direct interim Technical Support Center (TSC) interface.

Efforts are underway to provide additional interface capability in conjunction with the development of data acqui-sition and display system for the long term TSC.

For the interim Technical Support Center, data flow will be provided by phone communications and hard copy computer data.

A communicator is specified by the current DB-1 Emergency Plan to be stationed in the control room upon activation of the interim Technical Support Center (TSC) and Emergency Control Center (ECC).

This individual will relay plant status information on a real time basis to the on-site emergency organization for analysis. Communication of technical data will be over a secure phone network connecting the interim TSC and the l

ECC with the Control Room. The communicator will additionally provide situational reports on operator actions and intended activity which would not be otherwise available.

Hard copy information will be provided by the process computer which is supplied from an uninterruptable pover supply.

The computer Sequence of Events monitor review provides the interim TSC with the trip sequence in milliseconds of 121 key input points which flag actuation of the Reactor Protection System, Safety Features Actuation System, Steam and Feedwater Rupture Control System, and turbine generator or grid problems.

A Post-Trip Review provides the interim TSC with data on 63 key parameters following a turbine or reactor trip. The printout will contain values from 15 minutes before and after the event. The parameters, as deter-j mined by Toledo Edison and B&W, are automatically saved and printed upon

. demand, and will provide sufficient information to permit an engineering analysis of the transient.

-~

A-10 In addition to the above, the computer alarm printouts will provide the interim TSC with an on-going chronology of the event.

The approximately 3200 alarms, and monitored contracts provide current status of plant problems which can be correlated with the data, or further investigated via phone communications.

It is felt from the station's experience in the investigation of unit trips, which utilize much the same information, that sufficient data will be available to effectively analyze and assess an event.

2.

Document long term plans for TSC.

Response

The basic concept for the long term Technical Support Center and other emergency operation facilities remains unchanged from Toledo Edison's letter of December 27, 1979 (Serial No. 571). However, more details including its revised on-site location and justifying criteria will be provided in a separate letter by March 27, 1980.

dh c/5-13 l

1

Docket No. 50-346 License No. NPF-3 Serial No. 601 March 21, 1980 Attachment B LIST OF ESSENTIAL AND NON-ESSENTIAL CONTAINMENT PENETRATIONS 4

Note:

Penetration numbers correspond to those of j

Table 6-8 (pages 6-49) of Davis-Besse Nuclear Power Station Unit 1, Final Safety Analysis Report (FSAR).

i i

i 1

h 9

i

E = Essential DB-1 CONTAINMENT PENETRATIONS Page 1 of 7, NE = Non-Essential NA = Not Applicable BF = Blind Flange SA = Safety Features Actuation Signal Containment Panstration Flow Signal Normal Isolation Number Service Direction (Incident Level)

Valve Position Position Designation i

1 Pressurizer Sample Line Out SA (2)

Closed Closed NE 2

Steam Generator Secondary Out SA (2)

Open Closed NE Water Sample Line 3

Component Cooling Water In SA (4)

Open Closed NE Inlet Line 4

Component cooling Water Out SA (4)

Open Closed NE Outlet Line 5,6,7 Containment Air Cooling Units In SA (2)

Open Open E

Service Water Inlet Lines 8 A-J Containment Vessel Vacuum In SA (2)

Open Closed NE Breakers 9, 10, 11 Containment Air Cooling Units Out SA (2)

Open Open E

Service Water Outlet Lines 12 Component Cooling Water In SA (4)

Open Closed NE Supply to Control Rod Drive Mechanisms 13 Containment Vessel Normal Out SA (2)

Open Closed NE Sump Drain 14 Letdown Line to Purification Out SA (2)

Open Closed NE Demineralizers

E = Essential DB-1 CONTAINMENT PENETRATIONS Page 2 of 7 NE = Non-Essential NA = Not Applicable BF = Blind Flange SA = Safety Features Actuation Signal Containment Pznetration Flow Signal Normal Isolation Number Service Direction (Incident Level)

Valve Position Position Designation 15 Spare 16 Containment Vessel Equipment Out SA (2)

Open Closed NE Vent Header 17 Containment Vessel Leak Test In Local Locked Locked Inlet Line (BF)

Manual Closed Closed 18 Steam Cenerator Secondary Out SA (2)

Open Closed NE Water Sample Line 19 liigh Pressure Injection /

In SA (2)

Closed Open E

Make Up Line.

SA (2) Note (b)

Open Closed NE 20, 22 liigh Pressure Injection Lines In SA (2)

Closed Open E

21 Demineralized Water Supply In SA (2)

Open Closed NE Line 23, 24 Fuel Fransfer Tube In, Out Manual Closed Closed NE (BF) 25, 26 Containment Spray Lines In SA (2)/ Manual Closed Open E

27, 28 Low Pressure Injection Lines In Remote Open Open E

Manual 29 Decay Heat Pump Suction Line Out Remote Manual Closed Closed E

and Manual Note (a)

E = Essential DB-1 CONTAINMENT PENETRATIONS Page 3 of 7 NE = Non-Essential NA = Not Applicable BF = Blind Flange SA = Safety Features Actuation Signal Containment Pcnetration Flow Signal Normal Isolation Number Service Direction (Incident Level)

Valve Position Position Designation 30, 31 Containment Vessel Emergency Out SA (2)

Closed Closed E

i Sump Recirculation Lines (5)

Closed Open E

32 Reactor Coolant System Drain Out SA (2)

Open Closed NE Line to R.C. Drain Tank 33 Containment Vessel Purge In SA (1)

Closed Closed NE Inlet Line 1

34 Containment Vessel Purge Out SA (1)

Closed Closed NE Outlet Line 35, 36 Auxiliary Feed Water Lines In Remote Open Open E

Manual 37 38 Main Feedwater Lines In SA (4)

Open Closed NE 0

39, 40 Main Steam Lines Out SA (4)

Open Closed NE 41 Pressurizer Quench Tank In SA (2)

Open Closed NE Circulating Inlet Line 42-B Containment Vessel Air Sample In SA (1)

Open Closed NE t

Return 42-A Service Air Supply Line In SA (2)

Open Closed NE 43-A Instrument Air Supply Line In SA (2)

Open Closed NE 6

E = Essential DB-1 CONTAINMENT PENETRATIONS Page 4 of 7 NE = Non-Essential NA = Not Applicable BF = Blind Flange SA = Safety Features Actuation Signal Containment Penstration Flow Signal Normal Isolation Number Service Direction (Incident Level)

Valve Position Position Designation 43-B Containment Vessel Air In SA (1)

Open Closed E

Sample Return 44-A Core Flooding Tank Fill and In SA (2)

Closed Closed NE Nitrogen Supply Lines 44-B Pressurizer Quench Tank In SA (2)

Open Closed NE Nitrogen Supply Line 45 Spare 46 Spare 47-A Core Flooding Tank Sample Line Out SA (2)/ Remote Closed Closed NE Manual 47-B Core Flooding Tank Vent Line Out SA (2)/ Remote Closed Closed NE Manual 48 Pressurizer Quench Tank Out SA (2)

Open Closed NE Circulating Outlet Line 49 Refueling Canal Fill Line In/Out Manual Locked Locked NE Closed Closed 50 High Pressure Injection Line In SA (2)

Closed Open E

51 Hydrogen Purge System Exhaust Out SA (2)

Closed Closed E

E = Essential DB-1 CONTAINMENT PENETRATIONS Page 5 of 7 NE = Non-Essential NA = Not Applicable BF = Blind Flange SA = Safety Features Actuation Signal Containment Penetration Flow Signal Normal Isolation Number Service Direction (Incident Level)

Valve Position Position uestgiation z

52, 53 Reactor Coolant Pump Seal In SA (2)

Open Closed NE 54, 55 Water Supply Note (b) 56 Reactor Coolant Pump Seal Out SA (2)

Open Closed NE Water Return Note (b) 57, 58 Steam Generator Drain Lines Out Remote Manual &

Closed Closed NE Local Manual 59 Secondary Side Chemical In/Out Manual Closed Closed NE Cleaning 60, 61, 62 Spare 63, 64, Spare 65, 66 67 Hydrogen Dilution System In SA (2)

Closed Closed E

Supply 68-A Pressurizer Quench Tank Sample Out SA (2)

Closed Closed NE 68-B Containment Air Sample Out SA (1)

Open Closed E

69 Hydrogen Dilution System In SA (2)

Closed Closed E

g.. ~.

70 Spare 71-A Containment Pressure Sensor Out Remote Open Open E

1 Manual

E = Essential DB-1 CONTAINMENT PENETRATIONS Page 6 of 7,

NE = Non-Essential NA = Not Applicable BF = Blind Flange SA = Safety Features Actuation Signal Containment Pen:tration Flow Signal Normal Isolation Number Service Direction (Incident Level)

Valve Position Position Designation 71-B Containment Air Sample Out SA (1)

Open Closed E

71-C Core Flooding Tank N Fill In SA (2)

Closed Closed NE 2

Line 72-A Containment Pressure Sensor Out Remote Open Open E

Manual 72-B Spare 72-C Containment Pressure Out Remote Open Open E

Differential Transmitter Manual 73-A Containment Pressure Sensor Out Remote Open Open E

Manual 73-B Containment Air Sample Out SA (1)

Open Closed E

73-C Containment Pressure Out Remote Open Open E

Differential Transmitter Manual 74-A Containment Pressure Sensor Out Remote Open Open E

Manual 74-B Containment Air Sample Out SA (1)

Open Closed E

74-C Pressurizer Auxiliary Spray In Remote Closed Closed E

Manual Note (a)

E = Essential DB-1 CONTAINMENT PENETRATIONS Page 7 of 7 NE = Non-Essential NA = Not Applicable BP = Blind Flange SA = Safety Features Actuation Signal Containment Penetration Flow Signal Normal Isolation Number Service Direction (Incident Level)

Valve Position Position Designation 75, 76, 77 Spare 78, 79 80 Emergency Lock See Chapter 3 of DB-1 FSAR 81 Personnel Lock for description and arrangement.

82 Equipment Hatch

\\

Notes ott Containment Penetration Table -

(a) Boron dilution flow path after 7 days.

(b) Toledo Edison requested a Technical Specification change on March 23, 1979 to change to Incident Level 3.

Jh c/4-10

Docket No. 50-346 License No. NPF-3

SUMMARY

DESCRIPTION OF THE DB-1 Serial No. 601 LEAKAGE REDUCTION PROGRAM STATUS March 21, 1980 Attachment C The initial testing effort of the Leakage Reduction Program at Davis-Besse has been completed. The only system not tested is the HPI system and associated " piggyback" piping. This system will be tested during the refueling outage as part of an ECCS Refueling Surveillance Test, ST 5051.04. The reactor coolant system, makeup, letdown, seal injection, ceal return, low pressure injection and containment spray systems are all covered by existing surveillance test procedures (ST) and no special testing was done on these systems. The results of a recent performance of the existing ST's and the results of Special Leakage Reduction testing on other systems are detailed below.

1.

Reactor coolant system (RCS), makeup, letdown, seal injection, and seal return systems - These systems are leak tested by ST 5042.01 and ST 5042.02. A recent performance of ST 5042.01 was 1/13/80. This test verifies that seal return, controlled leakage, is less than 10 GPM. This is not a measure of any system leakage to the building atmosphere. Controlled leakage was 3.38 when tested on 1/13/80. Actual total leakage to atmosphere from these systems is measured daily by ST 5042.02, RCS Water Inventory Balance. As of 1/17/80, the total calculated leakage was.1540 GPM. This number varies slightly from day to day, and is attributed to reactor coolant pump (RCP) seal leak off. RCP seal leak off is periodically collected and measured for use as a known system leak rate for comparison to the daily calculated value. Seal leak off was last measured during the most recent plant shutdown and was found to be

.2352 GPM, very close to the daily calculation.

2.

Low pressure injection and containment spray systems - These systems are tested for leakage by ST 5051.03. Containment spray was tested on 12/18/79 and low pressure injection on 12/20/79. No collectable leakage was found on either system.

3.

Primary sampling system - The high pressure portion of this system was tested by PT 5164.04 on 1/24/80. Pressurized from the RCS to normal operating pressure, four leaks were found.

Packing leaks were found on SS24, SS23, and RC53. There was also leakage out a cap downstream of RC54, indicating both RC54 leaking by and cap leakage. The only collectable and measurable leakage was on SS24 at about one drop every eleven minutes. Leakage from RC54 out its cap could not be collected due to location and interferences. Packing leakage on SS25 and RC53 was indicated by crystallized boric acid around the valve steams, and no current water leakage was observed on these.

C-2 4.

Waste gas system - Tested by PT 5172.00 in three sections:

a.

Waste gas surge tank and low pressure piping (up to the waste gas compressor suction isolation valves and tank inlet piping back to the first isolation valve). This section was tested on 1/29/80, pressurized to 7 psig from the station nitrogen header. Inspecting piping and valves with a sonic leack detector, no leakage to the building atmosphere was found. The amount of nitrogen used from a nitrogen bottle to3 mafntain pressure for one hour showed a consumption of 5.1 ft per hour. Since no atmospheric leakage was found, this nitrogen losa rate is attributed to boundary valve leakage into other closed systems.

b.

Waste gas compressors and associated piping (from the comprersor suction isolation valves to the waste gas decay tanks' inlet isolation valves). This section was pressurized to 140 psig from a nitrogen bottle, and several problems were discovered. High nitrogen consumption was traced to compressor 1-1 discharge relief valve leaking by.

Leakage through the relief valve made it impossible to maintain test pressure for

, a full hour, but all accessible piping was inspected at the required pressure of 140 psig and no leaks to the building atmosphere were found.

In the test of compressor 1-2 and it's piping the only leakage to atmosphere en this section was at the closure gasket on the suction filter of compressor 1-2, and this leakage was so low that nitrogen consumption over an hour's time could not be measured.

c.

Waste gas decay tanks (from inlet isolation valves to the first isolation valve on any tank outlet). These tanks were tested on 1/30/80, by pressurizing them to approximately 99 psig from the station nitrogen header. Tank 1-1 was empty thestartofthetestandwasfullypressurizeginabout90 at minutes. Leakage over an hour's time was 23.7 ft per hour.

No leakage to atmosphere was found and the nitrogen used is attributed to a combination of boundary valve leakage into other closed systems and makeup for the volume change as the tank contents cooled. Decay tank 1-2 was pressurized by transferring the contenta of tank 1-1 into it using the waste gas compressors. No leakage was found using sonic detectors and no measurable amount of nitrogen was used to maintain pressure during the test.

Decay tank 1-3 was pressurized from an initial pressure of 62psigto99psigusingthestationnitrogenheager. Nitrogen added to maintain pressure was measured at 5.7 ft per hour.

No leakage to the building atmosphere was found, and nitrogen consumption is attributed to boundary valve leakage into other systems and minor volume changes due to suspected slow cooling of the tank contents.

i C-3 5.

Reactor coolant drain tank system - The reactor coolant drain tank and associated piping was tested by PT 5164.05 on 1/31/80.

The system was tested to 7 psig and no leaks were found. No t

nitrogen was needed to maintain pressure for an hour, indicating no boundary valve leakage.

6.

Work requests have been submitted to maintenance for repair of the leaks and other problems discovered during testing. As of 2/5/80, work has commenced on correcting the problems.

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i Dock.at No. 50-346

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Licensa No. NPF-3 i

Serial No. 601 March 21, 1980 l

Attachment D l

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2 i

DESIGN REVIEW OF PLANT SHIELDING FOR 1

i SPACES OUTSIDE CONTAINMENT WHICH MAY BE USED IN POST ACCIDENT j

OPERATIONS FOR DAVIS-BESSE UNIT 1 REVISION 1 1

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Table of Contents i

I.

Introduction II.

Scope of Design Review A.

Systems Engineering Methodology B.

Shielding Design Review Methodology C.

Personnel Exposure 1.imits and Methodology III. Results of Review Appendix A - Accident Radiation Zone Maps Appendix B - Decay Curves Appendix C - Time and Motion Study to Deterrine Operator Radiation 1

Doses during Post Accident Radiological Sampling 1

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INTRODUCTION This report describes the design review of plant shielding of spaces for post accident operations for Davis-3 esse Uni: 1.

Systems required to process pri=ary reactor coolant outside the containment during post accident conditions were selec:ed for evaluation.

Large radiation sources were postulated to be present in :he selected systems.

Areas which are vital for post accident occupancy were evaluated to determine if access and perfor=ance of required operator activities might be unduly impaired due to the presence of the postulated radia:1on source in these systens.

As a byproduct of this review, a number of radiation zone maps and associated curves hcve been produced which will alert operational personnel to potential radiation levels in the plant at various times following the accident. These maps and curves are included in Appendices A and B, respectively.

II.

SCOPE OF DESIGN REVIEW A.

Systems Engineering Me:bodology 1.

Selection of Syste=s for Shielding Review The criteria applied in selection of plant systems used in the shielding review results in several classifications of systems selected for various reasons as discussed below.

Catecorv A (Recirculation Syste=s)

The first group of systems are those syste=s required by plant design to mitigate a design basis loss of coolant accident and which =ight contain highly radioactive sources in excess of the curren design basis.

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The emergency core cooling systems were designed to mitigate the consequences of a loss of coolant accident and prevent extensive core da= age as recuired by 10 CFR 50 Appendix K.

Nevertheless, for the purposes of this study these i

systems were postulated to coniain significan: additional sources of radio-activity above and beyond the original plant design basis for the purposes of the required shielding review. These plant systems are designed to be used following a loss of coolant accident. Operators are trained to respond to a loss of coolant accident by ucing the emergency core coot.ing systems.

Procedures are written to specify the proper use of these systems by plant operators following an accident.

In summary, the emergency core cooling s stems are designed and expected to operate following a loss of coolant accident to prevent significant core damage.

If a significant radioactive source is to be considered above and beyond the normal plant design basis as has been required by NRC, then a

?

first priori:7 safety concern is to ensure that operation of these systa=s containing a significant source will not adversely i= pac: operator functions required outside the contai= ment. Therefore, the following syste=s have been selected to ensure that this first priority safety concern is adequately addressed by the existing plant shielding design:

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o Those portions of the containment spray systems used to recirculate water from the containment vessel emergency sump back into.the containment vessel.

o Those portions of the decay heat removal system used to recirculate water from the containment vessel emergency su=p back into the containment vessel.

o Those portions of the high pressure injection system used to recirculate water from the low pressure injection system back into the containment vessel.

Categorv B (Extensions of Containment Atmosphere)

In addition to systems listed above, there are other systems or portions of systems which would contain radioactivity by virture of their connection to the containment following an accident. Proper operation of the emergency core cooling system would prevent extensive core damage and mean that these systems would not be expected to contain the significant radioactive sources required oy this special analysis. Nevertheless, such sources have been pestulated in the following systems for this study:

o Those portions of the containment ventilation systems external to the containment up to the first closed isolation valve which could contain the atmosphere from the containment.

o Those portions of the sampling system used to obtain a containment atmosphere sample.

Category C (Liould Samples)

Lessons Learned Task 2.1.8 requires that certain post accident liquid samples be obtained from the reactor coolant system.

Those portions of the sampling system which must be used to meet the intent of Task 2.1.8 were selected for this shielding review.

Categerv D (Latdown)

There is no reason to operate additional systems while the reactor coolant system is contaminated to the significant levels required as part of the shielding design review.

Since there is no systems engineering logic for mechanistically analyzing either the letdown and makeup system or the waste gas system, an arbitrary, nonsechanistic assumption has been selected to be responsive to the NUREG 0578 requirement. Therefore, the following portion of the letdown system has been selected for analysis:

o That portion of the letdown system from the reactor coolant system past the failed fuel detector up to the inlet valves to the purification dominera11:ers.

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Quantification of Potential Radioactive Source Release Fractions The following release fractions were used as a basis for determining the concentrations f or the shielding review:

o Source A:

Containment atmosphere: 100% noble gases, 25 halogens o Source B: Reactor coolant: 100% noble gases, 50% halogens, 1% solids o Source C:

Containment sump liquid: 50% halogens, 1% solids The above release fractions were applied to the total curies available for the particular chemical species (i.e. noble gas, halogens, or solid) for an equilibrium fission product inventory for a light water reactor.

The Regulatory Daide 1.7, solids release fraction,1%, was used in this review. No noble gases were included in the containment sump liquid (Source C) consistent with Regulatory Guide 1.7.

Cursory analyses have indicated that the halogens dominate all shielding requirements and that i

contributions to the total dose rates from noble gases are negligible for the purposes of a shielding design review.

3.

Source Term Models Section 2 above outlines the assumptions used for release fractions for the shielding design review.

These release fractions are, however, only the first step in modeling the source terms for the activity concentrations in the systems under review.

The important modeling parameters, decay time and dilution volume, obviously also affect any shielding analysis. The following sections outline the rationale for the selection of values for l

these key parameters.

a.

Decay Time For the first stage of the shielding design review process, no decay time credit was used with the above releases. The primary reason for this was to develop a set of accident radiation zone maps normalized to no decay (refer to Appendix A) that could be used as a tool by the plant staf f along with a set of decay curves (refer to Appendix B) to quantitatively assess the in-plant dose status quickly following any abnormal occurrence.

For identifying problem areas, however, the following decay times were used in assessing anticipated potential personnel radiation exposure due to those operator actions that may be required post LOCA.

For enalyses of personnel exposures in vital areas outside the control room, radioactive decay of one hour was considered. This is consistent with the NRC requirement to obtain a reactor coolant sample one hour after en accident, which is the only short term evolution outside the control room.

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. Additional decay time was also allowed for the review of all those ECCS systems previously outlined in Section II that are used to recirculate water from the containment sump back into the containment. That decay time was 40 minutes which is consistent with the minimum time for initiation of recirculation as per FSAR Section 15.4.6.4.

b.

Dilution Volume The volume used for dilution is important, affecting the calculations of dose rate in a linear fashion.

The following dilution volumes were used with the release fractions and decay times listed above to arrive at the final source terms for the shielding reviews:

o Source A: Containment free volume. The volume occupied by the ECCS water was neglected.

o Source B: Reactor coolant system volume based on reactor coolant density at the operating temperature and pressure.

o Source C: The volume of water present at the time of recirculation (Reactor coolant system + borated water storage tank +

core flood tanks).

1 c.

Sources Used in Piping and Equipment for Each System Under Review In defining the limits of the connected piping subject to contamination listed below, normally shut valves were assumed to remain shut.

o Containment spray system - At the initiation of recirculation, Source C was used.

o Righ pressure injection system - At the initiation of recirev tion, Source C was used.

This assumes that the HPI and LPI are cross connected.

o Decay heat removal system - Source C was used for the sump recirculation mode.

o Sampling systems - The sources used in the shielding design review for sampling systems were as follows:

Containment air sample -

Source A Reactor coolant sample -

Source 3 o Letdown system - The liquid source was Source B.

B.

The Shielding Design Review Methodology 1.

Analytical Shielding Techniques

'The previous sections outlined the rationale and assumptions for the selection of the systems that would undergo a shielding design review as well as the formulation of the sources for those systems. The next step in the review process was to use those sources along with standard point kernel shielding analytical techniques to estimate dose rates from those selected systems.

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' l For compartments containing the systems under review, estimates were made for a genermi area dose rate rather than to superimpose the maximum dose rate at contact with the surfaces of all individual components of that system in the compartment.

The maximum area dose rates indicated on the accident radiation zone maps (Appendix A) are based on a whole body dose to the head of a 6 foot tall person in the normally (not an accident condition) accessible (by walking, not climbing on pipes and equipment) portions of the compartment under review. For corridors outside compartments, reviews were done to check the Jose rate transmitted into the corridor through the walls of ad-jacent compartments. Checks were also made for any piping or equipment that could directly contribute to corridor dose rates, i.e. piping that may be run l

directly in the corridor or equipment / piping in a compartment that could shine directly into corridors with no attenuation through compartment valls..The i

source term from containment vessel leakage which is contained within the negative pressure boundary was considered in the dose rates within and adjacent to the negative pressure boundary.

1 2.

Accident Radiation Zone Maps one of the two principal products of this review process is the series of accident radiation zone maps (refer to Appendix A) for Davis-Besse Unit 1.

These zone maps represent the correlation of the dose rates as estimated above with the required operator actions and resultant necessary accessibility to vital areas.

By using these zone maps along with the decay curves (refer to Appendix B) potential problem areas were identified as noted in Section III.

The zone boundaries were formulated based on the following rationale:

i Zone Designation Rationale D, Zone Dose Rate Limits (Rem /hr) i A-I The first zone is consistent 0 da D c,. 0.015 with the personnel radiation exposure guidelines of Section II.C.2 for vital areas requiring continuous occupancy.

A-II The second zone is consistent 0.015 mg D gi 0.1 j

with the personnel radiation exposure guideline of Section II.C.2 for vital areas requiring infrequent access or corridors to these areas.

Such zones involve no time and motion evaluations.

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A-III The third zone is consisten: with 0.100 4. D 6 5 the personnel radiation exposure guidelines of Section II.C.2.

Zones in this range required that a time and motion study be done to ensure that integrated exposure was not greater :han 5 Rem as given in General Design Criteria 19, of 10 CFR 50, Appendix A.

"he subsequent zones were selected by grouping : hem by powers of 10 so that rapid assessment of addi:ional shielding naasures could be used via " tenth value layers" of cot =non shielding sa:erials.

Zone Designation D, Zone Dose Rate Limits (Re:/hr)

A-IV 5 4 D d 50 A-V

% D f 500 A-VI 50J 4 D g 5000 A-VII 5000 4 D f 50,000 A "III 50,0004D f,500,000 Note: These zone designations should not be confused vi:h those used for the normal plant operation zone maps found in Chapter 12 of :he Final Safety Analysis Report (FSAR).

C.

Personnel Exposure Limits and Methodology 1.

Access Those operator actions required pos: LOCA were reviewed to ensure that first priori:y safety actions can be achieved in the postulated radiation fields.

This review ensures that access is available and required operator actions can be achieved except as noted in See:1on III.

In addition, since other unspecified operator actions =ay be desired, the evolutions involved in a normal shutdown sequence were reviewed as well as the general provisions for occupancy or access to the following key areas:

o Control room System control panels o

o Baergency power supplies i

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Personnel Radiation Exposure Guidelines The general basis for personnel radiation exposure guidelines was 10 CTR 50, Appendix A, GDC 19.

The following additional radiation 11=1: guidelines were used to evaluate occupancy and accessibility of key plant areas.

General area dose ra:es were used rather :han maxi =u= surface dose rates.

Contributions from all 5 - rces were considered.

Vital areas requiring continuous occupancy Vital areas such as the control room and the onsite technical support center vere verified to ensure the direct dose ra:e was less than 15 =r/hr at all times.

Vital areas requiring infrequent access.

For these areas the dose rate was verified to be less than 5 R/hr when required for access.

For dose rates greater than 100 mr/hr, a =an-rem calculation including time and motion analysis was performed to insure that the integrated exposure for an operator action did not exceed 5 res as required by GDC 19.

For dose ra:es less than 100 =r/hr, a =an-rem calculation was not necessary.

!!!. Results of Review A review was cade of the required operator tasks outside the control room following a loss of coolant accident to see if the doses received by plant personnel vould exceed allevable limits. The basis for this review was emergency procedure EP 1202.06, Loss of Reactor Coolant and Reactor Coolant Pressure.

All required operator actions were found to be accomplished from vi:hin :he control room with the exception of aligning power to cotor operated valves and verifying boron dilution flow race.

These actions can be delayed until seven days af:er the start of emergency operations.

These actions are:

I 1.

Observe the auxiliary spray flow on flew indicator FI 4999 located in room 227 on Elevation 565.

2.

Align emergency power from motor control centars MCCEll3 in room 304 on Elevation 585 and MCCFllA in room 427 on eleva: ion 603.

3ased on the seven day decay ti=e the dose received in performing the above actions and traveling to and from the areas was well below the allowable limits.

Depending on the reactor coolant syste= leak rate, the HPI pumps =ay have to be opera:ed in the " piggy back" mode of opera: ion with the decay heat pc=ps.

In order to accomplish this evolution manual valves DH63 and DH64 must be opened. These valves are located in Rooms 115 and 105, respectively.

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• During the first refueling outage, as required by Operating License condition 2.C. (3)(g), DE63 and DH64 will be modified to motor operated valves with control and position indication in the control room. This modification is covered by Fccility Change Raquest 79-078.

Ti:ne and motion analyses (refer to Appendix C) have been performed to insure that the integrated exposure to an operator while obtaining post accident samples will not exceed 5 rem as required by GDC 19.

The post accident samples considered were the reactor coolant system, containment vessel atmosphere, and the station vent.

The zone maps (Appendix A) and the decay curves (Appendix B) can be used to determine expected dose rates in various areas of the plant at any time after an accident.

l l

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Rav. 1

1 Appendix A Accident Radiation Zone Maps The following equipment location drawings have been marked to show the accident radiation zones as described in Section II.B.2.

M-120 M-121, Revision 1 M-122 M-123 M-124 M-125 Revision 1

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Decay Curves Figures lA, 1B, 2, and 3 corresponding to sources A, B, and C outlined in Section II.A.2 cre a set of curves developed for the shielding design review to serve as generic tools to estimate transient decay credit. These generic curves were generated rather than developing a parametric set of curves for each source that would account explicitiv for the effects of self-attenuation in the source material on actual dose rates. The primary r.ssumption in the application of the curves was that the dose rate from a source was directly proportional to the total gamma-ray energy relase rate (in Mev/sec) from the source in question, i.e.,

Source A, 3, or C.

Therefore, the decay curves are in a similar manner as those from the work of Lurie, E al.pese curves were developed note properly fission product energy release rate curves.

for the Sandia Laboratory research directed by Bonzon, g al.'

All curves have been normalized to the initial energy relase rate for the source in question.

I. A. Luri, D. H. Houston, and J. A. Naber, " Definition of Loss-of-Coolant N

Accident Radiation Source: Summary and Conclusions", SAND 78-0091, May 1978.

L. L. Bonzon, K. T. Gillen, E. A. Salazar, " Qualification Testing Evaluation Program Light Water Reactor Safety Research Quarterly Report: October-December 1978", STREG/CR-0813 or SAND 78-0761, June 1979.

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Appendix C Time and Motion Study to Determine Operator Radiation Doses During Accident Radiological Sampling Revision 1 i

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Time and Motion Study to Determine Operator Radiation Doses During Post Accident Radiological Sampling Operator radiation doses received during post accident radiological sampling of the reactor coolant system (RCS), containment vessel atmcsphere, and the station ven:

are shown in Tables 1, 2 and 3, respectively.

Each table is broken down into various evolutions required for obtaining the samples, with the assumed dose rate and exposure time given for each evolution. For this study, acceptable radiation doses :o an operator are based on 10CFR Part 50, Appendix A, General Design Criter-ion (GDC) 19 limits of 5 Rem shole body.

The evolutions are based on Davis-Besse Nuclear Power Station Unit No.1 Administra-tive Procedure AD 1850.04, Post Accident Radiological Sampling and Counting, excep:

for the station vent sample. The station vent sampling evolutions are based on discussions with D. Briden and R. Sund of TECo.

The dose rates are based on the " Design Review of Plant Shielding for Space: Outside Containment which may be used in Post Accident Operations for Davis-Besse Unit 1" (at: ached to BT-9510), except for Room 515.

For Room 515 new dose rates have been calcula:ed assuming a containment vessel leak rate of 0.5% per day for the firs:

day and 0.25% per day after the first day.

In order to align the temporary station vent sample station, entry into Room 515 is required.

This system should be aligned within 12 minutes af ter receiving an SFAS Lacident level 1 signal. This is necessary so that the radioiodine adsorbed in Emergency Ventilation System (EVS) filters will not cause excessive dose ra:es.

(Of :he two EVS filters, filter No.1-1 is significantly closer to the radia: ion monitor where the temporary system must be aligned).

The time of 12 minutes is very conservative, since it was assumed that all of the containment vessel source is released to the inside of the containment vessel at the time of the event, and since a containment vessel accident leakage rate of 0.5% per day for the first day was used in the calculation.

The value of 0.5% per day is conservative with respect to

he measured leakage rate.

Figure 1, Room 515 Radiation Dose Rate versus Time, shows the radiation dose rate which would be expected.if both EVS f ans are maintained on af ter an accident.

All samples can be taken with operators receiving radiation doses within the guide-lines of CDC 19.

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Table 1 Time and Motion Study for RCS I.iquid Sample Accident Dose Rate Exposure Dose, ares (1)

Evolution Radiarion Zone Rem /hr Time, see.

1) Move sample cylinder A-III 1.11 30 9

into Elevator No. 3 in g ridor on Elev.

603

2) Dose in elevator A-I 14 Negligible i

3) Move sample cylinder A-IV 10.0 30 83 fro = elevator to corridor on Elev. 545

4) Move sample cylinder A-II 0.1 14 1

into position 50 ft.

I away

5) Attach sample cylinder A-III 5.0 60 83 I)

6) Draw sa=ple A-II 0.1 300 8

7) Disconnect sa=ple A-III 5.0 10 14 cylinde.

8) Move sample cylinder A-II 0.211 14 1

to elevator 50 ft.

(plus 0.111 away Rem / hour from sample cylinder)

9) Move sample cylinder A-IV 11.0 30 92 into elevator (plus 1.0 rem /

hour from sample cylinder)

10) Dose in elevator A-I 1.0 14 4

(plus 1.0 rem /

hour from sample cylinder)

11) Unioad sample cylinder A-III 2.11 30 18 from elevator (plus 1.0 rem /

hour from sample cylinder)

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Table 1 (cont'd)

Accident Dose Rate Exposure Dose, stem (1)

Evolution Radiation Zone Rem /hr Time, sec.

)

l

12) Move to Elevator No. 2 (4) 0.111 30 1

(plus 0.111 rea/ hour from sample cylinder (5))

13) Move sample cylinder (4) 1.0 30 8

into Elevator No. 2 (plus 1.0 rem /

hour from sample cylinder)

14) Dose in elevator (4) 1.0 5

1 (plus 1.0 rem /

hour from sample cylinder)

15) Unload sample cylinder (4) 1.0 30 8

from elevator (plus 1.0 rem /

hour fro = sample cylinder)

16) Move sa=ple to counting (4) 0.11 58 2

facility in office (plus 0.111 building lobby rem / hour from sample cylinder)

Totals 699 333 seconds trem Notes 1 - Fractions of aram rounded up to whole mram.

2 - Time spent waiting for elevator is spant in stairway where the zone would be an A-I.

3 - Personnel wait in Zone A-II.

6 - No accident zone designated. Check normal plant operation zone map.

5 - When moving sample cylinder, assume that personnel stay 3 feet away from sample cylinder, except on the elevator.

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-a' Table 2 Time and Motion Study for Containment Vessel Atmosphere Sample Accident Dose Rate Exposure Evolution Radiation Zone Rem /hr

, ine, sec.

Dose, arem (1)

1) Mave sample bomb A-III 1.11 30 9

into Elevator No. 3 in corridor in Elev.

603 5

Negligible

2) Dose in elevator A-1

3) Move sample bomb A-III 0.25 30, 2

from elevator to corridor in Elev. 585

4) Move through Zone A-IV 50.0 20 277 A-IV to sa:ple station

5) Attach sample bomb A-III 5.0 300 417 and take sample

6) Move through A-IV A-IV 50.1 20 278 to elevator (plus 0.111 res/hr fro:

sa=ple benb) (2)

7) Move sa:ple bomb A-III 1.25 30 10 into Elevator No. 3 (plus 1.0 re=/

hour from sa=ple bomb) (3)

8) Dose in Elevator A-I 1.0 5

1 (plus 1.0 rem /

hour from sample bomb) (3)

9) Unioad sample beab A-III 2.11 30 18 from elevator (plus 1.0 rem /

hour from sample bomb) (3)

'.111 30 1

10) Move to Elevator No. 2 (4)

O (plus 0.111 rem / hour from sample bomb) (2)

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Table 2 (cont'd) 1 Accident Dose Race Exposure

)

Evolution Radiation Zone Rem /hr Ti=e, see.

Dose, mrem

11) Load sample bomb into (4) 1.0 30 8

Elevator No. 2 (plus 1.0 rem /

hour from sample bomb) (3)

12) Dose in Elevator No. 2 (4) 1.0 5

1 (plus 1.0 rem /

hour from sample bomb) (3)

13) Unload sample bomb (4) 1.0 30 8

(plus 1.0 rem /

hour from sa=ple bomb) (3)

14) Move sample to counting (4) 0.111

.~ S 2

facility in office (plus 0.111 rem / hour fro:

sample bomb) (2)

Totals 623 1032 seconds are Notes 1-Fractions of arem ro inded up to whole arem.

2-3 feet from sample bomb.

3-At surface of sample bomb.

4-No accident zone designated. Check normal plant operation zone map.

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Table 3 Time and Motion Study for Station Vent Radiation Level Accident Dose Rate Exposure Evolution Radiation Zone Rem /hr Time, sec.

Dose, mrem A.

Align Temporary Sample Station (only one time af ter an accident) 1)

Walk through A-III 5

13 13 Heater Bay to Room 515 2)

Align temporary A-VI 100 60 1667 sample station at RE-2024 (RE-2025) in Room 515(1) 3)

Walk through Heater A-III 5

13 13 Bay to leave Room 5 15 Totals 86 1693 B.

Obtain Station Vent Sample in Room 516 1)

Enter Room 516 A-III 5

300 417(2) and take sample Notes 1 - This evolution should be performed within 12 minutes af ter the accident.

See text.

2 - Each time a sample is taken in Room 516.

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. Qoeket No. 50-346 Licensa No. NPF-3 Scrial No. 601 March 21, 1980 Attachment E

' EFFECTS ON PLANT ACCESSIBILITY AS A RESULT OF CONTAMINATING THE LETDOWN, MAKEUP AND PURIFICATION AND GASEOUS RADWASTE SYSTEMS WITH A TID 14844 SOURCE TERM Summary Previously reported (I) access for reactor coolant system, containment vessel atmosphere and station vent sampling has not changed as a result of contaminating the letdown, makeup and purification, and gaseous radwaste systems.

Control room radiation dose rates are increased by only a fraction of a mrem / hour as a result of contaminating these systems. The radwaste control panel, C1702, in Room 110 can be used on a limited basis to control the gaseous radwaste system.

Systems Reviewed The letdown system was previously reviewed (I) up to the inlet valves of the purification demineralizers. This review covered the highly contaminated makeup and purification system including the purification demineralizers, makeup tank, makeup pumps, interconnecting piping and branch piping up to the first normally closed valve.

As a result of using the makeup and purification system, the gaseous radwaste system would also be highly contaminated. This review covered the gaseous radwaste syatem from the makeup tank and the pressurizer through the containment vent header to the waste gas surge tank, the waste gas compressors, the waste gas decay tanks, interconnecting piping and branch piping up to the first normally closed valve.

Source Terms The makeup and purification systen was assumed to be contaminated with these release fractions:

100% noble gases, 50% halogens, 1% solids up to the makeup tank where the noble gases were assumed to be removed. The gaseous radwaste system was assumed to be contaminated with 100% noble gases.

Methodology The same methodology used in the previous report II) were used for this review.

(1) Design Review of Plant Shielding for Spaces Outside Containment which may be used in Post Accident Operations for Davis-Besse Unit 1.

a e

.