ML19257A292

Forwards Supplementary Response to short-term TMI Lessons Learned Task Force Requirements & Provides Schedule of Implementation of Commitments.Seven Oversize Drawings Encl

ML19257A292
Person / Time
Site:	Point Beach
Issue date:	12/31/1979
From:	Burstein S WISCONSIN ELECTRIC POWER CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8001030581
Download: ML19257A292 (215)

Text

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Wisconsin Electnc ma come 231 W. MICHIGAN, P.O. BOX 2046. MILWAUKEE. WI 53201 December 31, 1979 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. NUCLEAR REGULATORY COMMISSION Washington, D. C. 20555

Dear Mr. Denton:

DOCKETS 50-266 AND 50-301 IMPLEMENTATION OF NUREG-0578 POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 On October 20 and November 27, 1979, we submitted detailed responses to the September 13 and October 30, 1979 NRC letters which discussed the Staff positions, clarifications, and implementation schedules of the NUREG-0578 Short Term Lessons Learned requirements.

In our submittals, we provided information regarding exis ting plant procedures, systems and operations, as well as commitments for changes to meet the NUREG-0578 requirements.

These commitments have been discussed with the NRC Staff on a number of occasions. Based on these discussions, certain modifi-cations and clarifications of our commitments were provided in our letter of December 17, 1979.

Attached is a complete description which documents our implementation of these commitments. To the best of our knowledge and belief, we have met all the January 1, 1980 schedular requirements as described in the attachment.

Very truly yours,

/\ s O Sol Burstein Exe tive Vice President

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1 2.1.1 EMERGENCY POWER SUPPLY REQUIREMENTS FOR THE PRESSURIZER HEATERS, POWER-OPERATED RELIEF VALVES AND BLOCK VALVES, AND PRESSURIZER LEVEL INDICATORS IN .PWRs Pressurizer Heater Power Supply A. Status The current Point Beach Nuclear Plant (PBNP) pressurizer heater power supply, as described in detail below, meets all of the criteria stated in the Staff position and clarification. Original plant design includes appropriate

, equipment for this purpose. Procedures have been developed and implemented which ensure the proper use of this equipment under accident conditions.

B. Discussion The Westinghouse Owners Group has established that pressurizer heater capacity of 100 kW is necessary for proper pressure control in the establishment of natural circulation for a twc-loop plant. The Point Beach Nuclear Plant (PBNP) design has four back-up heater groups, plus a control group. Two of the back-up heater groups , 200 kW each, are powered from the safeguards power supply; one from the B03, 480 volt -

bus and the 3D diesel; and one from the B04, 480 volt bus and the 4D diesel. The centrol power for the heater breakers is also supplied from the respective vital bus

, control power, the "A" battery supplying the B03 bus control power, and the "B" battery supplying the B04 bus control power. The power supply breakers and control power that tie the heater. groups with the emergency buses are qualified to the same safety standards as.the remainder of the vital safety-related breakers at PBNP.

The back-up heaters supplied by vital power supply are stripped by a safety injection signal in the PBNP design.

In order to restore the heaters, the safety injection signal must be reset, and the non-safeguards . lockout relays must be reset (both from the main control room) .

The revisions to the Emergency Operating Procedures have been completed and the pressurizer heaters have been added to the specific steps where applicable. The specific use of the pressurizer heaters comes about on a small break which has been isolated and the plant is establishing a more normal condition of operation in anticipation of securing the high head safety injection pumps if the proper conditions can be reached. The low head safety injection pumps have already been secured at this time.

Since the load on the emergency diesels have been reduced by about 300 KW by the securing of the low head safety injection pumps and the .high head safety injection pumps 1667 066 2.1.1-1

are at minimum load, thus also reducing the diesel loading, the effect of loading a pressurizer heater group

• (200 KW) on the emergency diesels does not appear to be significant. Additional loads, however, such as charging pumps, are needed and added to the diesel as diesel capacity' permits. An Operations Group special order has been prepared and issued to give guidance to the operators and to insure diesel capacity is not exceeded at any time during the conduct of the emergency procedure.

Power Supply for Pressurizer Relief and Block Valves and Pressurizer Level Indicators A. , Status The current PBNP pressurizer relief and block valves' and

' pressurizer level indicators' power supplies, as described below, meet all of the criteria stated in the Staff position and clarification. Original plant design includes appropriate equipment for this purpose.

B. Discussion A review of the PBNP power supplies for pressurizer power-operated relief valves (PORVs) , block valves, and level indication instrument channels has been completed. The motive and control components of the pressurizer PORV and block valves are powered from engineered safeguards buses, vital instrument (battery / inverter) buses, or directly from a battery. The power supply breakers and control power that tie the PORV and block valves with the emergency buses are qualified to the same safety standards as the remainder o# the vital safety-related breakers at PBNP.

Two out of three of the pressurizer level indication instrument channels and the cold calibration channel are powered frcm the vital instrument buses. The third channel is presently powered by an AC/AC~ motor-generator set from a non-safeguards bus. This condition does not, however, affect the operator's ability to obtain reliable pressurizer level indication using the remaining redundant channels following a loss of off-site power.

The' attached Figure 2.1.1-1 and Tables 2.1.1-1 through -3.

provide the pressurizer relief, solenoid and block valve configuration and power supply distribution for the.

pressurizer relief and block valves, air compressors, and pressurizer level instrumentation channels, respectively.

Both the red and blue instrument buses are vital instrument grade buses. The yellow and white instrument buses are non-vital. The white bus is supplied from the respective unit B02 AC/AC motor-generator set which serves to eliminate short term transients (power' spikes) on the white bus.

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Nos. 1 and 2 batteries are shared between units; the inverters are unique to each unit. Also attached is Figure 2.1.1-2 (previously supplied ~to the NRC with other submittals) which is a simplified electrical distribution diagram for the PBNP and combines Figures 8.2-1 and 8.2-2 of the Final Facility Description and Safety Analysis Repor.t (FFDSAR) . Prefix numbers in this Figure and the tables refer to unit designation (e.g.,

l.:2)-PCV-431C indicates the same configuration applies to the valve for both Units 1 and 2) .

The power operated relief valves, 1(2)PCV-430 and 1.( 2 ) PCV-4 31C , are air. operated valves which close on loss of air pressure to. the valve actuator diaphragm.

Instrument air system pressure is applied to the power operated relief valve actuator through solenoid valves 1( 2) SOV-4 3 0 and 1(2) SOV-431C, respectively. The solenoid valve is actuated by two pressurizer pressure bistable /

relay combinations. Both bistables must actuate to operate the solenoid valve allowing the relief valve to open. If a pressurizer pressure channel fails low (Channel 429, 430, 431, or 449), the power operated relief valve associated with that channel cannot open.

Instrument air is supplied 'through containment penetrations from either of two instrument air compressors, K2A and K23. This is backed up by the service air system which is supplied by either of two service air compressors (K3A and K3B). Instrument air pressure at 90 psig is normally supplied by the instrument air compressors.

If the instrument air system falls to 80 psig, service air is then supplied to the instrument air system. This is sufficient pressure to operate air operated valves supplied by the instrument air system. The instrument air compressors are stripped on a safety injection coincident with a loss of AC. They must be manually restarted.

i The instrument air headers inside containment are isolated when a containment isolation signal is initiated for each unit. In the case of automatic safety injection, this isolation remains until safety injection is reset and the instrument air system is unisolated. The instrument air system may be unisolated before safety injection is reset as long as the control switch for the instrument air containment isolation valve is manually held in the "open" position. A backup pressurized nitrogen gas bottle is provided to supply nitrogen gas'to operate the power operated relief valves in the event of instrument air system pressure loss when low temperature-overpressure protection is required during plant shutdowns.

I667 068 2.1.1-3

The current PBNP power supply configuration for the pres-surizer heater, PORVs, block valvss, and level instrumentation, therefore. provides the operator with the capability to maintain press.ure control for the reactor coolant system and monitor pressurizer level when off-site power is not available.

1667 069

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TABLE 2.1.1-1 PBNP PRESSURIZER POWER OPERATED RELIEF AND BLOCK VALVE POWER SUPPLIES ITEM POWER SUPPLY -

VALVE APPLICABILITY Distable/ Relay 1(2)PCV 1(2)MOV 1(2)S03 Combinations 430 431C 515 516 430 431C Channel 429 No. 1 battery / inverter for X Dackup the unit, red instrument to ch.431 bus AC Channel 430 No. 1 battery / inverter from Unit 1 Unit 2, red instrument bus AC No. 2 battery / inverter from Unit 2 Unit 1, blue instrument bus AC ,

Channel 431 No. 2 battery / inverter for X the unit, blue instrument '

bus AC Channel 449 1(2) BO4 ESP bus, yellow Backup to X 3nstrument bus AC unit's Ch. 430

. Motive Power '

-- 1(2)BO3 ESF bus X Ch 1(2)UO4 ESF bus X CN

J Control Power 1(2) BO3 ESP bus X Ef} 1(2)BO4 ESP bus X CZ) ,

12SV DC No. 1 bdttery X No. 2 battery X

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TABLE 2.1.1-2 PBNP AIR COMPRESSOR POWER SUPPLIES AIR SUPPLY ENGINEERED SAFEGUARDS BUS

, COMPRESSOR SYSTEM 1B03 2B03 1304 2B04 K2A . Instrument X K2B Instrument X

'K3A Service .X .

K3B Service X O

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TABLE 2.1.1-3 PBNP PRESSURIZER LEVEL INSTRUMENTATION POWER SUPPLIES PRESSURE LEVEL CHANNEL POWER SUPPLY l (2) L42 6 . No. 1 battery / inverter for the unit, red instrument bus 1(2) L427 1(2) B02 AC/AC motor-generator set nonsafe -

guards bus, white instrument bus l(2) L428 No. 2 battery / inverter for the unit, blue instrument bus -

1(2) L433 No. 1 battery / inverter for the unit, red (Cold Calibration) instrument bus -

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2.1.2 Performance Testing for PWR Relief and Safety Valves A. Status Wisconsin Electric Power Company will be participating in the EPRI valve testing program which is responsive to the requiremcnts presented in NUREG-0578, Item 2.1.2, which recommended in part, " commit to provide performance verification by full scale prototypical testing for all relief and safety valves. Test conditions shall include two-phase slug-flow and subcooled liquid flow calculated to occur for design basic transients and accidents".

. The EPRI program, " Program Plan for the Performance Verification of PWR Safety / Relief Valves and Systems, December 13, 1979", was submitted to the NRC by Mr. William J. Cahill, Jr., Chairman of the EPRI Safety and Analysis Task Force, by letter dated December 17, 1979. The EPRI Program Plan provides for completion of the essential portions of the test program by July, 1981.

B. Discussion We believe that our previous comments, restated here, on the need for such testing are pertinent and of sufficient substance to merit NRC Staff consideration in its ongoing review of lessons learned from Three Mile Island.

We understand that the NRC Staff position in respect to performance testing for safety and relief valves is based principally., if not exclusively, on the failure of the power-operated relief valve at TMI to reclose.

Such a failure to reclose of a RCS relief or safety valve, of course, results in the loss of integrity of the primary coolant pressure boundary. All plants have been analyced for a complete range of break si:es and no license has been issued where the LOCA analysis was deemed inadequate. Additional analyses have been performed since the TMI accident, both independent of and in response to NUREG-0578, Item 2.1.9. We fail to see in what manner a stuck-open relief valve is different from a small break LOCA. We believe that a relief valve testing program will require very large expenditures and diversion of scarce personnel from efforts which may have a safety impact.

The current design basis and Code specifications in this area require redundant Code safety valves, nonisolatable, on any pressurized system as a means of overpressure prctection. In addition, the discharge frcm the Code safety valves must be unrestricted in its capability to ultimately relieve to the atmosphere.

2.1.2-1

To reduce the number of challenges to the Code safety valves because they may not reclose and cannot be

- isolated, the design of Point Beach and other PWR primary coolant systems also includes one or more power operated relief valves (PORV) . Automatic control systems are provided to operate the PORVs and provide relief of RCS pressure at a setpoint below that of the Code safety valves. Manual actuation is also provided.

It is our understanding that, during pcwer operation, this design has eliminated all challenges to the Code safety valves and that, in fact, no Code safety valve has yet been required to open on a Westinghouse PWR plant.

Additionally, the basic design of the inverted U-tube steam generators with a large secondary side water inventory (thirty minutes or greater to dryout without AFW) and safety grade auxiliary feedwater system provide substantial post-accident heat removal capability.

This serves to limit the primary side heatup and pressure transient and severely restricts even the challenges to the PORVs.

Since the PORVs are expected to open, the design also provides block valves in series to insure isolation of the RCS should the PORVs fail open. The operation of the PORV and block valve is ensured by separatio.n of power supplies and actuation logic as described in Item 2.1.1. This design is a proven, accepted practice in pressurized systems, - both nuclear and non-nuclear, and meets all Code and NRC licensing requirements.

The transient and accident analyses specifically exclude both the protection (relief) and centrol (relief-isolation). capabilities of the PORV and block valves.

Instead, the overpressure conditions analyzed must use the Code safety valve setpoints as the upper bounds of the calculations. Conversely, the LOCA analyses must consider a range of breaks which cover an assumed PORV or safety valve opening. Thus, the function of reclosure is strictly an' operational benefit without unanalyzed safety concerns. It should be noted that while the pressurizer relief tank (PRT) provides for the suppression of PORV or safety valve releases and eventual unrestricted release to containment via the tank rupture disk, containment pressure. calculations must assume direct release to the containment for maxi-mizing pressure. Piping system integrity downstream of the safety valves is, thus, not necessary.

1667 076

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2.1.3 INFORMATION TO AID OPERATORS IN ACCIDENT DIAGNOSIS AND CONTROL 2.1.3.a Direct Indication of Power-Operated Relief Valves and Safety Valve Position for PWRs and BWRs

. A. Stacus The power operated relief valve indication, as described below, meets the NRC requirements for direct indication.

.The stem-mounted limit switches from which the indication is derived are safety-grade, seismically qualified, and environmentally qualified as discussed below. Direct indication of position for the Code safety valves has been provided by an acoustic flow monitoring system with sensors for each valve. For each unit, an alarm and annunciator has been provided in thd control room.

B. Discussion The basic operational and safety requirement is that the operator be able to determine the condition of the plant and take appropriate action in response to an accident condition. The Licensee continues to be committed to providing sufficient and necessary instrumentation in th'e control room such that the operator has this accident assessment capability.

The present RCS instrumentation provides the operator with redundant, safety grade instrumentation which permits the determination of the existence of a LCCA regardless of whether the LOCA is due to an open Code safety or relief valve or valves.

The Westinghouse Owners Group, of which Wisconsin Electric Power Company is a member, is performing transient and accident analyses and evaluations which address small break loss-of-coolant accidents generically and independent of location. These results, revised proced6res, and implementation schedule are addressed in Item 2.1.9.

The analyses consider PORV and Code safety. valve releases without the need to identify specifically the individual component condition. In addition, see Sections 4.2.2 and 5.3.4 of the Wisconsin Electric Power Company TMI Accident Review Task Force report previously submitted.

The Point Beach Nuclear Plant pressurizer power-operated relief valves (PORVs) for both units have direct, positive indication in the control room derived from seismically qualified, control grade stem-mounted limit switches. This direction indication has been upgraded 2.1.3-l667 077

in Unit 1 during the recent November 1979 refueling outage by replacing'these switch'es with safety grade

-(seismically and environmentally qualified) switches.

A similar modification will be performed on Unit 2 during its refueling in Spring 1980. Besides actuating valve position lights, the switch outputs are used in conjunction with signals from'the associated safety valves to actuate an alarm and light an annunciator for each unit in the control room. This modification was completed prior to January 1, 1980. Additionally, temperature indication is provided for the commor PORV discharge header.

The position' indication for pressurizer Code safety valves has previously been provided by separate temperature indication in each discharge header piping and pressure, temperature, and level indication in the pressurizer relief tank. Additional direct indication of Code safety valve position has been provided by adding an acoustic flow sensor to the piping immediately downstream of each of the pressurizer Code safety valves.

The acoustic sensors provide inputs to a microprocessor based monitoring system. This monitoring system provides two levels of alarm for each channel and separate alarms for each unit. Alarm outputs from the acoustic monitoring system are combined with. associated power operated relief valve indication to actuate an alarm and light an annunciator for each uni't in the control room.

The acoustic monitoring system was installed with available components prior to January 1, 1980, as described below.

1. The transducers are located directly downstream of each Code safety relief valve in their permanent locations.

2. LOCA qualified cable supplied by the system vendor has been temporarily run from the transducers to the inside containment penetration and' spliced to a TRIAX penetration cable using an approved splicing technique.

3. TWINAX cable qualified to the FR1 flame test has

. been permanently run from the outside containment penetration to the cable spreading room via instrument cable trays. Channel separation was maintained.

A similar splice was made at the penetration.

4. Because the final system equipment was not available, the preamplifiers to be located outside of containment were not installed, instead a special amplifier is used in the main electronics package. The amplifiers and preamplifiers will be installed in the system after they are received'.

2.1.3-2

5. The main electronics package is located in the cable spreading room. A temporary cabinet was used to house the equipment. A seismic cabinet and foundation will be provided at a later date. A local display is available on the main electronics package which will allow monitoring of Unit 1 or Unit 2 Code safety valves. An alarm output for each valve has been connected to the appropriate unit's " relief valve not shut alarm".

6. Remote display devices will be installed in an approp-riate control room location when the display cable has been installed between the control room and the electronics package.

7. Qaalification of the electronics package, display and display cable e.ad preamplifiers must be determined at e later date.

A schemanic description of the currently installed system is provided in the attached Figure 2.1.3-1. The alarm system which includes PORV and Code safety valves is shown in Figure 2.1.3-2. The acoustic monitoring system is presently classified as control grade. It is expected that it will be type tested by Westinghouse during 1980 ,

and, as a result, be classified as safety grade.

C. Schedule The replacement of the stem-mounted limit switches on the Unit 2 pressurizer power operated relief valves will be completed in April.of 1980. The equipment utilized in the acoustic monitoring system will be upgraded to its final configuration by the replacement of the temporary components noted with the final components. This will be accomplished as soon as practicable after component delivery.

I667 079 2.1.3-3

2.1.3.b Instrumentation for Detection of Inadequate Core Cooling in PWRs and BWRs I. Procedures and Description of Existing Instrumentation _

A. Status The Westinghouse Owners' Group, of which Wisconsin Electric Power Company is a member, has performed analyses as required by Item 2.1.9 to study the effects of inadequate core cooling. These analyses were provided to the NRC Bulletins and Orders Task Force for review on October 31, 1979. As part of the submittal made by the Owners' Group, an " Instruction to Restore Core Cooling during a small LOCA" was included. This instruction provides the basis for procedure changes and operator training required to recognize the existence of inadequate core cooling and restore core cooling based on existing instrumentation.

The primary instrumentation being used for the detection of inadequate core cooling is the existing incore thermocouple system.

B. Discussion The procedure effort for detection of inadequate core cooling has been broken down into two areas. The first area is a procedure revision to better define the proper boundaries for establishing and monitoring natural circulation in a loss of AC power situation.

The second is to establish procedural guidelines for possible inadequate core cooling during a small loss of coolant accident. The first effort mentioned above has been completed and is now included in the emergency procedures. The second area is still under review by the Westinghouse Owners' Group. A preliminary procedure has been produced by the Owners' Group and has been promulgated as a special order to the licensed operators and other personnel at PBNP. The final and permanent procedure revisions in this area will be done upon receipt of the completed effort by the Owners' Group.

II. Subcooling Meter A. Status The installation of RCS saturation condition monitoring and alarm has been implemented prior to January 1, 1980, as described below. Other means of indicating subcooling are presently under review. We have concluded that, based 2.1.3-4

r upon an evaluation of the whole scope of long term instrumentation requirements and the NRC clarifications, we will not have an upgraded subcooling monitoring system until later in 1980'than the July 1 date previously noted.

B. Discussion Wisconsin Electric Power Company, in its Task Force review of the accident at Three Mile Island, determined that an indication of subcooling in the primary system would be beneficial to the operator and made a recom-mendation that this.information should be provided to the operators at Poin.t Beach.

This is currently being provided by the on-line process computers for each unit. Each -computer is programmed to determine'the degree of subcooling and to alarm when there is less than 50*F subcooling margin in the core. Display of the subcooling monitoring information is available to the operator on demand.

The inputs used for this determination are primary system wide range pressure and average incore thermocouple temperature. The Chromel-Alumel thermocouple calibrati.on being used is based on National Bureau of Standards Circular 561. This provides for utilization of the full thermocouple range. A check of the calibration was performed over the range of 335 F to 2290*F and the accuracy is within N1%. A revised version of the Information Required on the Subccoling Meter is given in Table 2.1.3-1 with changes noted by bars in the margin.

C. Schedule An upgraded subcooling monitoring system, independent on the on-line process ccmputers, will be installed by January 1, 1981.

III. Design and Installation of New Instrumentation A. Status The conceptual design for the combined reactor vessel level instrumentation and reactor vessel head venting system (Item 2.1.9) is provided a; described below.

The Westinghouse Owners' Group program for analyses and procedures to address inadequate core cooling will provide the basis for any additional new instrumentation, although no need has yet been identified.

I667 081 2.1.3-5

B. Discussion

- The submittal referenced in I. above described time capabilitie.s of the core exit thermocouples in deter-mining.the existence of inadequate core cooling conditions and their superiority in some instances to the loop RTDs for measuring true core conditions. Other means of determining the approach to or existence of inadequate core cooling could be:

1. Reactor Vessel water level

2. Incore detectors

3. Excore detectors

4. Reactor coolant pump motor currents

5. Steam generator pressure A discussion of the possible use of these measurements are addressed below.

The use of incore movable detectors to determine the existence of inadequate core cooling conditions appears doubtful. The detectors could be driven in to the tops of the incore thimbles, which are located at the top of the core, following an accident in which concern for inadequate core cooling. exists. The problem comes in the lack of sensitivity of the detectors to very low

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neutron levels and changes that would occur due to core uncovery. Gamma detectors could perhaps be employed, but they suffer from similar sensitivity problems, and the fact that gamma levels in the fuel region change insignificantly between the covered and uncovered condition.

As a result, it does not appear worthwhile to pursue incore movable detectors as a means of determining inadequate core cooling conditions.

The use of excore detectors has been mentioned as a possibility in responding to core uncovery. The only detectors which would have the required sensitivity are the source range monitors. Since the intermediate and power range monitors are not sensitive enough to the low level changes resulting from vessel voiding, the use of the source range monitors will be investigated further as part of the more indepth study of inadequate core cooling being performed by the Westinghouse Owners' Group.

However, their use is probably limited to those instances when significant voiding exists in the down-ccmer region, since normally water in the downcomer would eftactively shield the detectors from the core region whether voids existed or not.

Reactor coolant pump motor current, which could ce indicative of core voiding, is inappropriate for a e reliable means of determining inadequate core cooling, since a loss of off-site power or pump trip would make this method unreliable.

1667 ogg 2.1.3-6

Steam generator pressure indication, which exists, is useful in the case where heat transfer from primary to secondary is interrupted due to loss of natural circulation.

This, however, does not satisfy requirements to indicate the approach to inadequate core cooling, nor does it indicate the true condition of the core.

Reactor vessel water level determination is the most promising of the items discussed to provide additional capability of determining the approach to and the existence of inadequate ccre cooling. Several systems for determining water level are under review by the

, Westinghouse Owners' Group. A conceptual design of one system is given below:

Vessel Level System Description Af ter examining many different methods and principles for determining the water level in the reactor vessel, a basic delta pressure measurement from the bottom of the vessel to the top of the vessel appears to provide the most meaningful and reliable information to the operator. One of the reasons for choosing this system is that the sources of potential errors are better known for this system than for any other new or untested system.

The conceptual design for the reactor vessel level

. instrumentation is given in Figure 2.1.3-3. This system would utilize the current penetrations to the RCS in conjunction with a reactor vessel head venting system.

The proposed leve.1 indication system covers the full range from normal water level to below the bottom of the core. The description below covers the system for each unit at PBNP.

The level indication system consists of two separate, redundant channels of differential pressure (d/p) instrumentation. These channels would be powered from the vital buses for each unit. Indication and alarm would be provided to the operator in the control room.

The upper pressure taps would be the existing reactor vessel vent line and a spare control rod drive penetration.

The lower pressure taps would be via an incore instru-mentation guide tube conduit. A single conduit would be used if the detailed engineering evaluation determined that such a use would render that incore flux thimble location unusuable. If the incore flux thimble location can still be used, then additional independence may be provided by using two separate locations for the lower pressure tap. This is reflected in the conceptual design shown.

i667 083 2.1.3-7

The protected or environmentally qualified d/p

. transmitters would be located such that with appropriate isolation, instrument lines, vents, and drains, they would be accessible for testinn and calibration at power, if necessary. Inclusion of temperature instrumentation on the sealed reference legs will be based upon the need for density compensation. Also considered will be the desirability of guide tube blowdown capability for venting and flushing.

The behavior of the signal generated by this level instrument under normal and accident conditions is being evaluated. The usefulness of this instrument to provide an unambiguous indication of inadequate core cooling is being evaluated as part of Item 2.1.9. The potential errors and accuracy of a final system configuration are being evaluated to assess its usefulness to provide information to the operator for proper operation of a vessel venting system and for normal water level control during periods when the primary system is open and a water level may exist in the vessel. The connection of the level system to the vessel head is designed to be compatible with the heat vent system. Opcration of a single vent path using the head vent system allo.ws vessel level to be indicated en the cessel level instrument connected to~ the other vent path. Valve and instrumentation power supplies are arranged in this manner to allow for proper operation of the system should a single power supply be lost.

C. Schedule NRC proposal review is required before implementation per the NRC clarification schedule of October 30, 1979.

Detailed design will be provided as soon as.Possible in 1980. Work is proceeding with the assumption that the conceptual design given here is acceptable. Implementation of vessel level indication will be completed by January 1, 1981, assuming valve and instrument delivery schedules allow for installation of the complete system by that date.

~

l667 084 2.1.3-8

r TAB LE 2.1. 3-1 INFORMATION REQUIRED ON THE SUBCOOLING METER Display Information Displayed (T-Tsat, Tsat, T, Tsat, Press as. selected Press, etc.)

Display Type ( Analog, Digital, ERT) Digital Continuous or on Demand On Demand Single or Redundant Display Single (each unit individually) l Location of Display Control Room Reactor Operators Console Alarms (include setpoints) Alarm typewriter, <50*F margin to Tsat setpoint l Overall uncertainty (*F, PSI) (15*F, 115 psig)

Range of Display 0 to 9999.9*F Qualifications (seismic, Original Plant Design environmental, IEEE323)

Calculator Type (process computer, dedicated PRODAC 250 Computer digital, or analog calculator)

If process computer is used, 99.90% Unit l' specify available~ (% of time) 98.78% Unit 2 (May 15, 1979 through December 10, 19 79 base period used)

Single or redundant calculators Single (each unit individually) l Selection Logic (highest T., Single Pressure Channel - Wide lowest press) Range (each unit)  !

Average of Incore Thermocouples Qualifications (seismic, Original Plant Design environmental IEEE32 3)

Calculational Technique __

Steam Tables (Steam Tables, Functional Fit, Ranges) 1667 085

TABLE 2.1.3-1 (Cont'd)

Input Temperature (RTDs or T/Cs) Incore Thermocouples (Chromel-Alumel)

Temperature (number of sensors and All Available Incore Thermocouples locations) (Max. '39 each unit) i Range of. Temperature Sensors 0 - 2300*F per NBS Circular 561 calibration Uncertainty

• of Temperature iS*F @ 620'F (75% x temp) 1 Sensors ('F at 1)

Qualifications (Seismic, Original Plant Design Environmental, IEEE323)

Pressure (specify instrument used) Foxboro Model 611 Pressure Transmitter Pressure (number of sensors and 1 Sensor-Reactor Coolant Rot locations) Leg (each unit) )

Range of Pressure Sensors 0 - 3000 psig Uncertainty

• of pressure 0.5% of full scale sensors (PSI at 1) (115 psig)

' Qualifications (Seismic, Original P.lant Design Environmental, IEEE323)

Backup Capability 1

Availability of Temp. and Press Located in Control Room -

Wide range pressure meter i Incore thermocouple readout with manual selection Availability of Steam Tables, etc. Located in Control Room -

Steam Tables Saturation Curve (P vs T)

Training of Operators Provided as part of cre-and post-TMI training and retraining programs.

Procedures EOP 1, 2, and 3. Revisions include subcooling implications and natural circulation initiation and monitoring I

(EOP 7A).

~

• Uncertainties must address conditions of forced flow and natural circulation.

~~

1667 086

y <v

^

XXX C XX+ PREAMP ^

(See flote 3) 3 .

l To

• AMPLIFIER i PRT $

XXX A TWillAX [ PREAMP ^

MICR0P CESSOR

[ (See flote 3) (includes a Cable (See Hote 1 ' readout display)

& 2) TRIAX *\

- - s Cable pnit

5 - ' TWIllAX

_J2 out-y - Cable put

-[ -

(See flotes 1 -

[ Unit 2

& 3)

1 PCV 434 1 PCV 435

} Input Channel Separation Main- % '

- - - _~

,_ _ _ Cab _]e_ Speading_Roon1_ tained 6-Lead Twisted Pair Cable  ;

- To Contro for Alpha-Numeric Display (Remote)

Room (See Containnent flote 4) 2-Lead Cable for y Annunciator Alarm

-v (Main Control Board 1C04)

P ressurizer Key XXX - TWINAX Cable to TRIAX Cable Splice

-O- - Containnent TRIAX Cable Penetration

-M - Strap-on Acoustic Monitor Transducer Notes: 1) Cable inside containment is to be LOCA qualified. Cable outside containment is to be FR1 flane-test qualified.

~

2) Containment TWIflAX Cable was temporarily installed at power. Permanent installation is to be acconplished at first convenient outage. TWINAX-TRIAX splice is to be LOCA qualified on pennanent installation.

- 3) TWillAX Cable outside containnent is a permanent installation. PREAMP will not os be provided initially. Therefore, TWIllAX installation provides for this.

CX TRI AX-TWIllAX splice is to be FR1 flane-test qualified.

N

4) Two-lead annunciator cable was installed prior to Jan.1,1980. Six-lead display cable is to be installed after Jan.1,1980.

]

N FIGURE ' l.3-1 PRESSURIZER C08" S FETY VALVE POSITI0tt MurilTORING SYSTEM (PBflP Unit !)

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. ' 11 hen Valve is not Shut Annunciator On TRAltl B E ' '

DC POWER When Any Set of Alarm.

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2.1.4 CONTAINMENT ISOLATION PROVISIONS FOR PWRs AND BWRs A. Status All of the NRC position and clarification items are satisfied by the current Point Beach Nuclear ~ Plant Containment Isolation and Containment Ventilation Isolation System actuation and reset features.

B. Discussion Information pertaining to the Containment Isolation System utilized at the Point Beach Nuclear Plant has

- been described in detail to the NRC Staff in Wisconsin Electric Power Company's response to Item 9 of Bulletin 79-06A dated April 27, 1979, and additional information provided in a response dated August 23, 1979. That submittal demonstrates that all of the NRC position and clarification items are satisfied by the current Point Beach Nuclear Plant Containment Isolation System actuation and reset features.

Containment Isolation and Containment Ventilation Isolation are safeguards circuits. As such, they are redundant with the redundant trains separated, testable on line, and capable of automatic and manual initiation. The automat'ic input circuits employ two-out-of-three logic with at least two of the inputs receiving power from battery powered inverters. The output circuits are DC powered from safeguards batteries.

Manual Containment Isolation initiates both Containment Isolation and Containment Ventilation Isolation. Contain-ment Isolation is automatically initiated whenever Safety Injection is automatically initiated. This occurs under the following conditions: Containment High Pressure, Pressurizer High Pressure, and Steam Line Low-Low Pressure in either steam line. Containment Ventilation Isolation is

, automatically initiated upon initiation of Safety Injection (automatic or manual), High Containment Radiation (either of two monitors - particulate or gaseous), or Manual Spray.

Resetting of Safety Injection, Containment Ventilation Isolation, or Containment Isolation will not automatically open any of the fluid paths to or from containment which are isolated upon receipt of the initiating signal. The valves must be individually opened manually by deliberate operator action. Containment Ventilation Isolation valves are currently locked shut and may only be opened during a cold shutdown. Additionally, resetting of Safety Injection does not reset Containment Ventilation Isolation or Containment Isolation. Resetting of Containment Isolation or Containment Ventilation Isolation can occur only if the initiating signal is no longer present.

2.1.4-1 1667 090

Additionally, liquid release via the containment sump cannot occur inadvertently since one of the two isolation valves in each drain line is normally closed and each valve receives a Containment Isolation signal upon SI actuation. Also, there is no automatic actuation of the containment sump. An operator is required to hold the valve control' switch in the open position (spring return to the closed position) while gravity draining of liquid from the containment occurs.

A review of the systems inside containment was performed following the outline developed by the Westinghouse Owners' Group. Systems are listed in Table 2.1.4-1 as essential, desirable, or non-essential following a design basis accident. This listing also identifies the normal function, post-LOCA function, and post-LOCA condition for each penetration.

1667. 09I 2.1.4-2

~ .. - - - - . . - . .

TABLE 2.1.4-1 POIllT IIEACil NUCLEAR PLAtlT -

UNIT 1 AND UNIT 2 CONTAINMEllT PENETRATION CLASSIFICATIOrl, ,

FUNCTIOt3 AND POST-LOCA CONDITION POST-LOCA NORMAL FUNCTION POST-LOCA FUt3CTION CONDITION PEN I CLASSIFICATION 1 NE Supply steam to turbine Steam generator isolation Closed NE Supply steam to turbine Steam generator isolation Closed 2

NE Supply normal feedwater to Steam generator isolation Closed 3

steam generator NE Supply normal feedwater to Steam generator isolation Closed 4

steam generator 5 E Supply auxiliary feedwater Steam generator isolation Open to steam generator 6 E Supply auxiliary feedwater Steam generator isolation Open to steam generator 7 D Remove residual heat during Inside reactor coolant loop Closed shutdown. RilR suction. isolation 8 E Removal residual heat during Safety injection long term Open shutdown. RilR discharge. post-LOCA cooling 9 NE Drain R.C. drain tank Containment isolation Closed 1eactor coolant system Containment isolation Closed 10 D letdown to purification system D Reactor coolant. pump seal Containment isolation closed 11 water return Supply demineralized water Containment isolation Closed

- 12a NE os during shutdown Test connection Containment isolation Closed 12b NE 12c NE Vent R.C. drain tank Containment isolation Closed CD None Safety injection to cold Open m 13 E -

leg rN)

Supply nitrogen to PRT Containment isolation Closed 14a NE E Measure containment pressure Measure containment pressure Scaled instrument 14b provides barrier

i POST-LOCA ,

PEN I CLASSIFICATION tlORf4AI. FullCTION POST-LOCA FUNCTION CONDITION 14c NE Supply nitrogan to safety Containment isolation Closed injection accumulators 15 E Supply CCN to RCP Supply CCW to RCP Open 16 E Supply CCH to RCP Supply CCW to RCP Open 17 E 'RCP CCW return RCP CCU return Open .

18 E RCP CCW return RCP CCN return _

Open 19 NE CCH supply to excess Containment isolation Closed letdown IIX 20 NB Excess letdown llX return Containment isolation Closed 21 Spare 22 E tione Safety injection to Open reactor vessel 23 Spare 24 Spare i,

25a Spare 25b Spare  :

25c E None - Uni t 1 Post-accident containment Closed l air. supply l 26 D RCS makeup RCS makeup Open - check valves inside containmentj provides barriers -

27 E None Safety injection to cold leg Open 28a E Sample reactor coolant system Containment isolation Closed 28b NE Sample pressurizer liquid Conta'inment isolation Closed Ch 28c NE Sample pressurizer gas Containment isolation Closed Ch y 28d Spare 29a E Supply seal water to RCP Supply seal water to RCP Open 29b E Supply seal water to RCP Supply seal water to RCP Open w ,

30a NE Test connection Containment isolation Closed -

i

POST-LOCA -

CLASSIFICATION NONMA1: FUNCTION POST-LOCA FUNCTION CON DITIOli PEN I _

30b Spare 30c NE Provide cooling wpter to PRT Containment isolation Closed 3] a E Measure containment pressure Measure containment pressure Scaled instrument provides barrier 31b NE None Containment atmosphere Closed sample 31c E None Containment atmosphere Closed exhaust 32a E Measure containment pressure Measure. containment pressure Sealed instrument provides barrier 32b NE Te.st of safety injection Containment isolation Closed system 32c NE Auxiliary charging Containment isolat. ion Closed 33a D Supply instrument air Containment isolation Closed 33b D Supply instrument air Containment isolation Closed ,

33c NE Supply service air Containment isolation Closed 34a NE Sample PRT Containment isolation Closed D Sample steam generator Steam generator isolation Closed 34h 34c o Sample steam generator Steam generator isolation Closed 34d NE Sample RCDT Containment isolation Closed 35 E Cool containment atmosphere Cool containment atmosphere Open Cool containment atmosphere Cool containment atmosphere Open

{

Ch 36 37 E

E Cool reactor cavity Cool reactor cavity Open N Cool reactor cavity Cool reactor cavity Open 38 E C 39 E Cool containment atmosphere Cool containment atmosphere Open

< Cool containment atmosphere Cool containment atmosphere Open A 40 E 41 Spare 42a Spare 42b Spare

?

POST-LOCA PEN I CLASSIFICATION NORMAL. FUNCTION POST-LOCA FUNCTION CONDITION 42c Spare - Unit 1 i b None , Unit 2 Post-accident containment Closed air supply 43 E Containment cooling return Containment cooling return Open j

44 E Containment cooling return Containment cooling return Open ,

45 E Cavity cooling return Cavity cooling return Open 46 E Cavity cooling return Cavity cooling return Open 47 E Containment cooling return Containment cooling return Open

! 48 E Containment cooling return Containment cooling return Open.

49 Spare 50 D Steam generator blowdown Steam generator isolation Closed l

i 51 D Steam generator blowdown Steam generator isolation Closed 52 NE Auxiliary steam to heaters Containment isolation Closed -

during shutdown 53 NE Condensate return from heaters Containment isolation Closed

54 E None - Containment spray Cool containment atmosphere Open 55 E None - Containment spray Cool containment atmosphere Open 56 NE Containment test Containment isolation Closed 57 NE Vent steam generator Steam generator isolation Closed t'

58 NE Vent steam generator steam generator isolation Closed 59 Spare 60 Spare ,

61 Spare ,

62 Spare 63 Spare

64 _. Spare 65 CT' Spare CN '

66 ~4 Spare CD

sO LJ1

i POST-LOCA .

f POST-LOCA FUNCTION CONDITIOt1 pet 1 l CLASSIFICATION NORMAIr FUtJCTIOt3 i 67 Spare 68 Spare ,

69 E None - Sump B to RIIR Long term cooling Closed - Open during recirculati 70 E None - Sump B to 111111 Long term cooling Closed - open during recirculati 71 IJE Drain Sump A Containment isolation Closed i V1 NE Containment purge exhaust Containment isolation Closed V2 11 8 Supply air for containment Containment isolation Closed purge i Rll and E Monitor containment atmosphere Containment isolation Closed H12 for radiation R11 and E Monitor containment atmosphere Containment isolation Closed n12 for radiation i

1

• 4 i

C I.A S S I F I C A T I OtJ S : E = Essential i

NE = lion-Essential D= Desirable i

ON N

-4 0%

~

2.1.5 POST-ACCIDENT HYDROGEN CONTROL SYSTEMS FOR PWR AND

,BWR CONTAINMENTS 2.1.5.a Dedicated Penetrations for External Recombiners or Post-Accident Purge Systems A. _

Status The Point Beach Nuclear Plant purge system has been reviewed and determined to provide the required flow capability,' isolation, and single-failure proof operation. Hydrogen recombiners are not a part of the design or licensing basis for the plant.

B. Discussion Appendix D to the Point Beach Nuclear Plant (PBNP) Final Facility Description and Safety Analysis Report (FFDS AR) contains the PBNP post-LOCA containment purging analysis and a description of the post-accident containment ventilation system (PACVS). For this. analysis, the required flow rate of the purge system is 240 cfm, assuming the system is operated one hour per day. with the activity release based upon TID-14844. A design review calculation has been performed which verifies the original design with respect to the flow capacity of the purge system.

The PBNP system includes the following features: a three-inch exhaust pipe from a plenum near the top of the containment structure, a manual flow control valve and manual isolation valve outside containment but upstream of the pipe's entrance into the primary auxiliary building exhaust system, a 3/4-inch sample connection originating from the top of the containment, and a two-inch return line to containment designed to be connected to the service air system in the primary auxiliary building, if required. Operation of the system is specified in EOP-llA, Post-Accident Ventilation System.

The PBNP post-acciden,t purge system design is in accordance with the containment design criteria in effect at the time PBNP was licensed prior to both General Design Criterion No. 54 and No. 56. The as-built system has been reviewed and found to conform to the containnent design criteria approved for PBNP. Redundant isolation is provided with two manual valves (locked closed) in series in each line.

1667 097 2.1.5-1

2.1.5.c Capability to Install Hydrogen Recombiner at Each Light Water Nuclear Power Plant This item is not applicable to the Point Beach Nuclear Plant since hydrogen recombiners are not a part of the design or licensing basis for the plant.

1667 098

\ l '

2.1.5-2

2.1.6.a Integrity of Systems Cutside Containment Likely to Contain Radioactive Materials A. Status An initial program to identify the systems outside containment that could contain radioactive material has been completed. An initial testing of these systems, both liquid and gaseous, has also been completed. A program for future leak testing, leakage reduction, and preventive maintenance has been developed and is outlined in Section 2.1.6.a of the attached Addendmn.

B. Discussion An initial program to identify the systems outside containment that could carry radioactive materials has been completed. The following systems have been identified for testing, based on their possible use during the course of an accident:.

1. High and low head safety injection including the cross connection from low head safety injection to high head safety injection pump suction for long term recirculation;

2. Containment spray including the cross connection from low head safety injection to the spray pumps' suction;

3. Liquid sampling system;

4. Chemical and volume control system as follows:

a. Liquid - Letdown, letdown purificatio'n ion exchangers, volume control tank, charging pumps, and return to reactor coolant. In addition, the lines to and including one chemical and volume control system holdup tank;

b. Gaseous - Letdown gas stripper, cryogenic gas compressors; letdown gas decay tanks,

~

and volume control tank;

5. Waste gas vent header piping, including the vol'ume control tank relief line piping, the gaseous space in the chemical and volume control system holdup tank, and the waste gas compressors.

i667 099 2.1.6-1

6. The normal containment atmosphere mampling system;

7. Post-accident containment ventilation system;

8. The auxiliary building liquid drain piping to and including the waste holdup tank.

The systems which have been excluded from testing on the basis that they are not required for use during an accident are as follows:

. l'. Waste liquid processing system (after the waste holdup tank) ;

2. Boric acid recycle systems (after the chemical and volume control system holdup tank);

3. Waste gas systems (after the waste gas compressors) ; and

4. Cryogenic system.

The initial testing of the vent header, the holdup tank gas space, the waste gas compressors and associated -

valving, the volume control tank relief lines, and the cryogenic gas compressor to volume control tank lines has been completed with helium. Four identified leaks

- were repaired. The liquid lines have been tested at operational pressures using waste, and visual inspection was used for leak detection.

The design problem identified in the NRC letter of October 17, 1979, has been included in the analysis, and the testing of the volume control relief line to the vent header was included in the initial helium leak test program. However, the Point Beach Nuclear Plant design does not provide relief from these systems to a vented tank; accordingly, the North Anna situation does not exist at Point Beach Nuclear Plant.

A program for future leak detection, leakage reduction, and preventive maintenance has been developed as described in Section 2.1.6.a of the Addendum attached herewith. The detailed implementation of this program will be developed in a timely manner to provide for continued future leak testing, leakage reduction, and preventive maintenance on a refueling cycle basis.

1667 100 2.1.6-2

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2.1.6.b Design Review of Plant Shielding of Spaces for Post-Accident Operations A. Statua System identification, accessibility requirements, and occupancy times were defined, and a design review of plant shielding was completed. Appropriate modifications will be implemented by January 1, 1981.

B. Discussion An identification of appropriate systems for consideration in the design review was based on a determination of the required use of those systems during and immediately following a postulated accident. Equipment accessibility, area accessibility, access routes, and occupancy times were defined. System tracing was carried out to complete modelling of appropriate areas and spaces. Radiation level calculations using the source term required by NRC were performed. Since the results of this analysis are intended to be used in the development of plant modifications by January 1, 1981, it should be noted that the analysis assumed the presence of other modifications which are also required'to be implemented by January 1, 1981. Accordingly, the reactor vessel venting system was assumed to be in place for purposes of this analysis.

In the identification of systems to be used immediately following an accident, this assumption allowed the exclusion of the charging and letdown systems and the gas strippers. The charging system is assumed to be used, but without the operation of the letdown system, it would not be carrying highly radioactive water. The letdown system and the gas strippers would not be used, since hydrogen can be removed frcm the reactor coolant system by way of the reactor vessel venting system.

Hydrogen thus released to the containment would then be purged from the containment by the Post-Accident Containment Ventilation System (PACVS), as described in Appendix D of the Point Beach Nuclear Plant Final Facility Description and Safety Analysis Report (FFDSAR).

Wi~th this assumption, the results of the detailed design review of plant shielding are provided in Section 2.1.6.b of the Addendum attached herewith.

A number of problem areas with respect to personnel access have been identified in the analysis. The conceptual design of several alternative modifications to resolve these problem areas are discussed in the Addendum. Selection of one or more of the alternatives 1667 101 2.1.6-3

is proceeding in a timely manner to allow implementation of permanent plant modifications by January 1, 1981.

C. Schedule System identification, access requirements, and a detailed design review of plant shielding have been completed.

Several alternative modifications have been identified, and appropriate modifications will be implemented by January 1, 1981, 1667 102 6

2.1.6-4

2.1.7 IMPROVED AUXILIARY FEEDWATER SYSTEM RELIABILITY FOR PWRs 2.1.7.a Automatic Initiation of the Auxiliary Feedwater System for PWRs A. Status The Point Beach Nuclear Plant (PBNP) auxiliary feedwater system is a safety-grade system which provides for auto-matic initiation and is designed to meet single-failure criteria. The capability also exists to initiate manually auxiliary feedwater from the control room.

Testability of the initiating signals and circuits exist to the same degree as for other safety-grade systems.

The initiating signals and circuits are powered from the emergenc:, buses.

B. Discussion The auxiliary feedwater system design, instrumentation, actuation, and operation has been the subject of ongoing discussion with the NRC beginning with IE Bulletin 79-06B.

Extensive, detailed information on the PBNP auxiliary feedwater system was provided to the NRC (Messrs. Trammell and LeFave - NRC, Asselin-Sandia Labs) in a May 11, 1979 meeting, as documented by the May 15, 1979' letter from Mr. Trammell to Wisconsin Electric Power Company.

. This and other additional information provided informally to the NRC are included in a NRC Staff report to the Commission which is currently in draft form. A separate response was prepared in reply to the September 21, 1979 letter from Mr. D. Eisenhut to Mr. Sol Burstein entitled, "NRC Requirements for Auxiliary Feedwater Systems at Point Beach". This response was dated October 29, 1979.

Additional detailed information, which is also provided below, was submitted dated December 17, 1979, in response to the December 10, 1979 telecopied request from Mr. A.

Schwencer. The following drawings which were provided with that response are referenced here:

Bechtel Drawing 4995466 Sheets 369 370 372 812 813 814 818

• 6 i667 103 868 1523 1532 Westinghouse Drawing 883D195 Sheets 14 20 2.1.7-1 m

The Auxiliary Feedwater System contains two motor-driven

, auxiliary feedwater pumps which are shared between units and two steam turbine driven auxiliary feedwater pumps, one for each unit. The following addresses each item in the NRC position directly:

1. The Point Beach AFW System design provides for auto-matic initiation from a variety of sources. The auxiliary feedwater system can be initiated manually or automatically. Automatic ini'tiation of steam turbine driven pump auxiliary feedwater occurs if both steam generators reach a low-low water level or if an undervoltage condition occurs on both 4,160 volt buses feeding the main feedwater pumps. The latter initiation is anticipatory and not protection grade. It assumes that the loss of both main feed-water pumps will eventually result in loss of steam generator levels.

Automatic initiation of auxiliary feedwater from motor driven pumps occurs if.both main feedwater pumps trip or if there is a low-low water level in either steam generator. A Safety' Injection signal will trip the main feedwater pumps and then start the motor driven auxiliary feedwater pumps through the safeguards sequencer.

Once started, the auxiliary feedwater pumps will continue to operate until manually turned off.

Recovery of electrical power, steam generator water level, or resetting of Safety Injection does not turn off the pumps. This description of conditions, which cause automatic AFW initiation, can also be found on page 10.2-12 of the Point' Beach FFDSAR.

The elementary diagrams which show the AFW pump start circuits have been provided previbusly.

2. Each of the steam generators at Point Beach can be supplied with AFW from either a motor driven or steam turbine driven pump. In addition, the start initiating circuits for each of the pumps are redundant. One exception is the automatic start of the motor driven AFW pumps caused by the trip of both main feed pumps. A failure in this circuit could delay the automatic start of one motor driven AFW pump until the associated steam generator level reaches the low-low setpoint.

3. The Point Beach design includes provisions for testing the AFW initiating signals and circuits from the system monitoring devices through the actuation of the logic relays which make up the start circuits.

These tests are performed on a monthly basis. Once a year, during the refueling shutdown, the AFW pump start circuits associated with safety injection are tested by the Safety Injection / Loss of Offsite Power 2.1.7-2 1667 104

2.1.7 IMPROVED AUXILIARY FEEDWATER SYSTEM RELIABILITY FOR PWRs 2.1.7.a Automatic Initiation of the Auxiliary Feedwater System for PWRs A. Status The Point Beach Nuclear Plant (PBNP) auxiliary feedwater system is a safety-grade system which provides for auto-matic initiation and is designed to meet single-failure criteria. The capability also e::ists to initiate

. manually auxiliary feedwater from the control room.

Testability of the initiating signals and circuits exist to the same degree as for other safety-grade systems.

The initiating signals and circuits are powered from the emergency buses.

B. Discussion The auxiliary feedwater system design, instrumentation, actuation, and operation has been the subject of ongoing discussion with the NRC beginning with IE Bulletin 79-06B.

Extensive, detailed information on the PBNP auxiliary feedwater system was provided to the NRC (Messrs. Trammell and LeFave - NRC, Asselin-Sandia Labs) in a May 11, 1979 meeting, as documented by the May 15, 1979 letter from Mr. Trammell to Wisconsin Electric Power Company.

. This and other additional information provided informally to the NRC are included in a NRC Staff report to the Commission which is currently in draft form. A separate

, response was prepared in reply to the September 21, 1979 letter from Mr. D. Eisenhut to Mr. Sol Burstein entitled, "NRC Requirements for Auxiliary Feedwater Systems at Point Beach". This response was dated October 29, 1979.

Additicnal detailed information, which is also provided below, was submitted dated December 17, 1979, in response to the December 10, 1979 telecopied request from Mr. A.

Schwencer. The following drawings which were provided with that response are referenced here:

Bechtel Drawing 499B466 Sheets 369 370 372 812 813 614 818 Sl3 1667 105 868 1523 1532 Westinghouse Drawing 883D195 Sheets 14 20 2.1.7-1

The Auxiliary Feedwater System'contains two motor-driven auxiliary feedwater pumps which are shared between units

'and two steam turbine driven auxiliary feedwater pumps, one for each unit. The fellowing addresses each item in the NRC position directly:

1. The Point Beach AFW System design provides for auto-matic initiation from a variety of sources. The auxiliary feedwater system can be initiated manually or automatically. Automatic initiation of steam turbine driven pump auxiliary feedwater occurs if both steam generators reach a low-low water level or if an undervoltage condition occurs on both 4,160 volt buses feeding the main feedwater pumps. The latter initiation is anticipatory and not protection grade. It assumes that the loss of both main feed-water pumps will eventually result in loss of steam generator levels.

Automatic initiation of auxiliary feedwater from motor driven pumps occurs if both main feedwater pumps trip or if there is a low-low water level in either steam generator. A Safety-Injection signal will trip the main feedwater pumps and then start the motor driven auxiliary feedwater pumps through the safeguards sequencer.

Once started, the auxiliary feedwater pumps will continue to operate until manually turned off.

Recovery of electrical power, steam generator water level, or resetting of Safety Injection does not turn off the pumps. This description of conditions, which cause automatic AFW initiation, can also be found on page 10.2-12 of the Point Beach FFDSAR.

The. elementary diagrams which show the AFW pump start circuits have been provided previously.

2. Each of the steam generators at Point Beach can be supplied with AFW from either a motor driven or steam turbine driven pump. In addition, the start initiating circuits for each of the pumps are redundant. One exception is the automatic start of the motor driven AFW pumps caused by the trip of both main feed pumps. A failure -in this circuit could delay the automatic start of one motor driven AFW pump until the associated steam generator level reaches the low-low setpoint.

3. The Point Beach design includes provisions for testing the AFW initiating signals and circuits from the system monitoring devices through the actuation of the logic relays which make up the start circuits.

These tests are performed on a monthly basis. Once a year, during the refueling shutdown, the AFW pump start circuits associated with safety injection are tested by the Safety Injection / Loss of Offsite Power

2. '-

1667 106

Test. This Test includes running the pumps after sequencing onto the diesel generators.

4. All of the initiating signals and circuits are powered from either the emergency buses or the batteries.

5. All of the AFW' pumps and motor operated valves are manually operable from the control room. The control switches are shown on the previously provided elementary wiring diagrams. A single failure in the manual portions of the circuits would not result in the loss of system function.

6. The AC motor driven AFW pumps are automatically loaded onto the emergency buses. Each is powered from a separate bus. The power supplies to the motor operated valves are not stripped from the emergency buses and, therefore, do not need to be automatically loaded on.

7. The capability to operate the AFW pumps and valves manually from the control room is always retained.

A failure in the automatic start circuit to a pump will not prevent manual actuation of the AFW system.

The AFW motor operated valves are not operated automatically. The normal valve lineup is shown on FFDSAR Figure 10.2-5 and on TMI Accident Review Task Force Report Figure 3.2-12.

I667 107 t

9 2.1.7-3

-.. . . - - . . - - - . - - . - - . . . . . - ~ - . . _ . . . . . . . . - .

2.1.7.b Auxiliarv Feedwater Flow Indication to Steam Generators for PWRs A. Status Indication of auxiliary feedwater flow to the steam generators at Point Beach consists of locally mounted flow meters on the discharge of each. pump and flow transmitters connected to the same orifices which provide pump flow indication in the control room. A detailed description of the current system is provided below. The operator also has pump discharge pressure and steam generator level indication in the control room. This configuration of flow, pressure, and level provide the capability in the control room to ascertain the actual performance of the auxiliary feedwater system.

B. Discussion Auxiliary feedwater flow indication was previously only provided locally at the output of each auxiliary feed-Water pump. Flow transmitters were connected to the existing system flow orifices to provide control room indication of the output flow from each pump. Because of equipment availability, control grade transmitters were used as described below. In other respects, such as testing capability, separation, and receiving power from safeguard buses, the system meets safety grade criteria. When safety grade transmitters become avail-able, they will be installed to upgrade the system.

This will provide for indication in the control room of the auxiliary feedwater flow from each pump which is reactor protection system grade. The steam generator level indication (also reactor protection system grade) includes redundant channels and provides a diverse method of verifying flow. The operator also has pump discharge pressure (non-redundant and not safety grade) available to determine pump operation.

Due to equipment delivery delays, the flow indication circuits, as shown in Figure 2.1.7-1, are configured as follows:

1. Foxboro differential pressure transmitters (original plant equipment removed from service, except one channel) have been temporarily installed on each of the steam and motor driven auxiliary feed pump discharge flow orifices. Seismic mounting has been provided.

2. Field instrumentation cable from each transmitter to the red or blue Unit 1 or 2 protection rack has been installed permanently. This cable meets 2.1.7-4

IEEE 383-1974 standards. Channel separation was maintained between each of the cables installed.

3. Circuitry in each protection rack used has been installed which allows injection of a test differential pressure signal into the feed flow tset jack. (The transmitter is switched out of the circuit for testing.)

Test points on the input and cutput of the square root converter and the isolation amplifier allow the calibration of these devices to be checked with the test signal input. The wiring of the protection rack circuitry was completed using wire of the grade used in the original installation.

4. The square root converter on order with the Foxboro Company has not been received. A temporary jumper has been installed in the protection rack circuitry bypassing this device. A signal proportional to the differential pressure seen by the transmitter is presently transmitted directly to the flow indicators.

Until the square root converters are received and installed, the indicators will read as follows:

Ao Indicator Scale Flow 200 in H2O 100% 400 gym 100 in H2O 70.7% 282.8 gym Motor Driven 50 in H2O 50.0% 200 gpm Feed Pumps 25 in H2O 35.4% 141.6 gpm Ao Indicator Scale Flow 200 in H2O 100% 200 gpm 100 in H2O 70.7% -141.4 gpm Steam Driven 50 in H2O 50.0% 100 gpm Feed Pumps 25 in H2O 35.4% 70.8 gpm

5. The isolation amplifiers and power supplies being used currently are original plant equipment removed from service (Foxboro). These will be replaced when new Fexboro equipment is received.

6. Permanent field instrumentation cables between the protection racks and the main control board indicators have not been installed. Temporary spare instrument cables are being utilized until this cable installation can be ccmpleted.

7. The control board indicators are located in a temporary location on control board IC04. The location will be changed to the appropriate control board when the permanent cables (Item No. 6 above) are installed and plant operating conditions allow the control board cutouts to be made.

2.1.7-5 fbb[ fQg

C. Schedule The following changes will be made to the existing feed-water flow indication circuits.

1. Permanent field cable will be installed between the protection racks and control board indicators. This cable will meet IEEE 383-1974.

2. When new Foxboro E13 differential' pressure nuclear qualified transmitters are received, they will be installed in the place of the temporary transmitters being used. These transmitters will be qualified seismic to ANSI 344-1971. Anticipated delivery is

. January 15, 1980.

3. The square root converters will be installed in each flow indication circuit when they are received from Foxboro. Anticipated delivery is January 15, 1980.

These devices are being supplied with documentation which certifies that they have been built to the same specifications as similar origina), plan t equipment.

4. The isolation amplifiers and power supplies Will be replaced with new Foxboro equipment af ter delivery.

The new devices will have documentation as in Item 3.

Isolation amplifiers are expected January 15, 1980, and power supplies are expected in April, 1980.

5. The control board indicators will be moved to their final location after the permanent field cable is installed and the main control board cutouts can be made (to be scheduled uring unit cutages) .

Additional flow instrumentation will be installed to indicate the auxiliary feedwater flow going to each steam generator. This instrumentation will be safety grade and installed by January 1, 1981. The design of this system will meet the requirements as clarified in the NRC October 30, 1979 letter.

1667 110 2.1.7-6

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2.1.8.a Post-Accident Sampling A. ' Status A review of sampling capabilities under accident conditions at Point Beach Nuclear Plant has been completed. Interim modifications have been implemented and additional procedures have been developed. These interim modifications have been found to be adequate on a permanent basis to meet the January 1, 1981, requirements.

B. Discussion An analysis of sampling capabilities has been performed.

Further details of the analysis are presented in Section 2.1.8.a of the Addendum attached herewith. Plant procedures on Sampling and Analysis of High Level Reactor Coolant (HPl7. 6. 2 ) and on Contai nment Atmospheric Testing in Accident Situations (HP17.6.3) have been developed and will receive Manager's Supervisory Staff review and approval on or before January 4, 1980. Interim modifi-cations to facilitate the obtaining of a primary coolant sample during an accident have been implemented. Final equipment checkout and initial training will be completed by January 11, 1980. These modifications are described in Figures 2.1.8-1 through 2.1.8-7. Further details are presented in Section.2.1.8.a of the Addendum attached herewith. Interim modifications to facilitate the obtaining of a containment atmosphere sample during an accident have been implemented. Final equipment checkout and. initial training will be completed by January 11, 1980. These modifications are described in Figure 2.1.8-8.

Further details are presented in Section 2.1.8.a of the Addendum attached herewith.

As a result of the analysis of sampling capabilities, these interim modifications have been determined to be adequate on a permanent basis to meet the January 1, 1981, requirements. However, the conceptual designs of a number of alternative modifications have been developed as described in Section 2.1.8.a of the Addendum attached herewith. Evaluation of the feasibility or need of any further alternative modifications will be done during the forthcoming year.

C. Schedule All 2.1.8.a items are on schedule. Further details regarding implementation dates are provided in the discussion above. The presently installed modifications fulfill the January 1, 1981," requirements.

2.1.8-1

2.1.8.b Increased Range of Radiation Effluent Monitors A. Status Interim procedures have been developed to enable the determination of radioactivity concentrations in effluents at the high levels presented in Table 2.1.8.b.2 of the attachment to the NRC letter dated October 30, 1979. In addition, high range monitoring equi.pment has been ordered and is expected to be installed by January 31, 1980. Permanent provisions for monitoring high levels of radioactivity in effluents will be implemented by January 1, 1981. High range in-containment radiation level monitors will be installed by January 1, 1981.

B. Discussion An analysis of the existing radiation monitoring system at Point Beach Nuclear Plant h+; been performed. 'It has been determined that the existing low level effluent instrumentation cannot measure the high levels of radioactivity presented in Table 2.1.8.b.2 of the attachment to the NRC letter dated October 30, 1979.

Accordingly, interim methods have been developed to permit the estimation of high level release rates based -

on direct radiation measurements ?btained at each existing vent stack. The figures showing the relationship between direct radiation measurement and release rates are

. provided in Section 2.1.8.b of the attached Addendum.

These relationships permit the estimation of release rates based on radiation readings obtained from direct contact monitoring instruments.

However, because of personnel dose and availability considerations, we believe that additional installed monitoring instrumentation is required for a more satisfactory interim implementation of high level monitoring capability. Accordingly, high range level monitors have been purchased from the Eberline Instrument Corporation as described in Section 2.1.8.b of the attached Addendum. These monitors are expected to be delivered to Point Beach Nuclear Plant on January 2, 1980; installation of these devices, including control room readouts, will be completed by January 31, 1980.

Use of these instruments will be based on the same relationships between direct radiation readings and release rate as are being presently relied upon for application with portable survey instruments from January 1, 1980, through January 31, 1960. Further conceptual details of these installations are provided in Section 2.1.8.b of the attached Addendum.

Interim measurement of radiciodines and particulates in effluent releases will be accomplished with existing

't 2.1.8-2 l66[ ll}

off-line sampling equipment. Analyses will be performed with the additional upgraded methodology set forth in Health Physics Procedure HPl7.6.1, as also referenced in our response to Item 2.1.8.c.

Permanent upgraded high range effluent monitors will be provided by January 1, 1981.

In-containment radiation level monitors with a maximum range of 108 RADS /hr. will be provided by January 1, 1981. Further discussion is provided in the Addendum.

C. Schedule Interim methodology relying on direct radiation measure-ments taken on vent stacks has been provided. Handheld ins._cments will be relied upon for obtaining these measurements until January 31, 1980. Additional high range monitors have been ordered, are expected to be delivered by January 2, 1980, and will be installed with readout in the control room by January 31, 1980.

Permanent high range effluent monitoring instrumentation and high range in-containment monitors will be provided by January 1, 1981.

1667 114 2.1.8-3

2.1.8.c Improved In-Plant Iodine Instrumentation A. Status Upgraded provisions and improved procedures for the determination of airborne radiciodine under accident conditions at Point. Beach Nuclear Plant have been completed as described below.

B. Discussion Procedures have been reviewed and upgraded to address

. improved sampling of airborne radiciodine under accident conditions. Plant Health Physics Procedure HPl7.6.1 has been written to address appropriate precautions regarding such sampling. This procedure will receive final review and approval by the Manager's Supervisory Staff on or before January 4, 1980. In addition, a reciprocal agreement has been executed between Point Beach Nuclear Plant and Kewaunee Nuclear Plant to provide backup analytical laboratory services for both facilities.

C. Schedule This item is complete except for final Manager's Super-visory Staff review and approval, which will take place on or before January 4, 1980.

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l667 115 2.1.8-4

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1667 123

2.1.9 Transient and Accident Analyses A. Status Analyses of small break loss-of-coolant accidents, symptoms of inadequate core cooling and required actions to restore core cooling, and analysis of transient and accident scenarios including operator actions not

. previously analyzed are being performed on a generic basis by the Westinghouse Owners' Group, of which Wisconsin Electric Power Company is a member. The small break analyses have been completed and were reported in WCAP-9600, which was submitted to the Bulletins and Orders Task Force by the Owners' Group on June 29, 1979.

Incorporated in that report were guidelines that were developed as a result of small break analyses. These guidelines have been reviewed and approved by the Bulletins and Orders Task Force and have b'een presented to the Owners' Group utility representati'es in a seminar held on October 16-19, 1979. Following this seminar, we have developed plant specific procedures and trained our personnel on the new procedures. Revised procedures and training are in place in accordance with the requirements in Enclosure 6 to Mr. Eisenhut's letter of September 13, 1979, and Enclosure 2 to Mr. Denton's letter of October 30, 1979.

The work required to address the other two areas, inadequate core cooling and other transient and accident scenarios, has been performed in conjunction with schedules and requirements established by the Bulletins and Orders Task Force. Analysis related to the definition of inadequate core cooling and guidelines for recognizing the symptoms of inadequate core cooling based on existing plant instrumentation and for restoring core cooling following a small break LOCA were submitted on October 31, 1979. This analysis is a less detailed analysis than was originally proposed, and will be followed with a more extensive and detailed analysis which will be available during the first quarter of 1980. The guidelines will be s

in place during the first quarter of 1980. .The guidelines will be in place during January, 1980, as required by the Bulletins and Orders Task Force.

With respect to the other transients and accidents contained in Chapter 14 of the Point Beach FSAR, the Westinghouse Owners' Group will perform an evaluation of the actions which occur during ari event by constructing sequence of event trees for each of the non-LOCA and LOCA transients.

From these event trees a list of decision points for operator action will be prepared, along with a list of information available to the operator at each decision point. Following this, criteria will be set for credible misoperation, and time available for operator decisions will be qualitatively assessed. The information developed will then be used to test Abnormal (AOP) and Emergency (EOP) Operating Procedures againsr the event 2.1.9-1 lff/ l

sequences and detarmine if inadequacies exist in the

~

AOPs and EOPs. '- ce results of this study will be provided to the c letins and Orders Task Force by March 31, 1980, a, required by D. G. Eisenhut's letter dated September 13, 1979, and H. Denton's letter dated October 30, 1979.

The Owners' Group has also provided test predictions analysis of the LOFT L3-1 nuclear small break experiment.

This analysis tras provided on December 13, 1979, in accordance with the schedule established mutually with the Bulletins and Orders Task Force.

1667 125 1

2.1.9-2

~

d Containment Pressure, Water Level, and Hydrogen Monitors A. Status The commitment by the Wisconsin Electric Power Company to provide containment high range pressure, water level, and hydrogen monitors, for the Point Beach Nuclear Plant, remains unchanged.

B. Discussion Relative to the implementation of containment water

. level, the equivaledt capacity of the wide range level instrument for Point Beach Nuclear Plant will be in excess of 350,000 gallons. The sources are identified as follows:

Source Maximum Volume (Gallonst Refueling Water Storage Tank 285,000 Boric Acid Storage 5,000 Accumulators (2) 17,000 Reactor Coolant System 37,000 (above the nozzles on the vessel)

. Containment Spray Additive Tank 3,000 TOTAL 347,000 gallons Since the sump is the lower portion of the containment building proper, only a wide range instrument is presently contemplated. The current level indication is discreet at 3', 5', 7', and 9' which exceeds the range of interest (7 feet 2350,000 gallons). By limiting the initial range setting to 7' for this instrument and allowing upward recalibration following an accident, maximum accuracy will be maintained over the range of most interest.

C. Schedule A consultant has been retained to assist in meeting the scheduled date of January 1,. 1981, for completion.

January 1, 1981 will be the target date for completion of required plant modifications consistent with procure-ment of equipment.

1667 126 ACRS-1

a Reactor Coolant System Venting A. S.tatus Reactor coolant system venting capability veyond that available with the current penetrations to the Point Beach Nuclear Plant primary system will be provided.

Implementation of reactor vessel head venting modifi-cations to the plant will occur after the required proposed review by the NRC of the conceptual design described below.

B. Discussion The Wisconsin Electric Three Mile Island Accident Review

. Task Force discusses Reactor Coolant System Venting in Section 5.2.1. This report was attached to our October 20, 1979 submittal.

The reactor vessel head venting system conceptual design, being developed along with vessel level indication (Item 2.1. 3.b) , includes the capability to vent to both the containment directly and the pressurizer relief tank (PRT) as evaluated in the report. ~ The conceptual design, as shown in Figure 2.1.3-3, provides for redundancy in the venting and isolation capability through th'e use of these two vent paths, eacn with series isolation valves. The system consists of four normally closed Safety Class 2 " fail closed" isolation valves. The system is designed such that any single active failure will not prevent vessel gas venting nor prevent venting isolation. The valves will be remotely operated from the main control room. Indications of flow through each vent path will be provided to the operator. The system will be designed to meet ANSI Safety Class 2 specifications up to and including the isolation valves. The remainder of the system (containment vent line and connect' ion to the

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PRT) will be designed consistent with the PRT connection line. The power supplied to actuate each vent path will be provided by separate safeguards buses, which will also be the opposite of the bus supplying the associated reactor vessel level indication instrument.

Included in each reactor coolant system penetration connection will be an orifice, appropriately sized to provide both the required flow capabilities for venting and the level instrumentation response and a limitation on primary coolant loss below that size " corresponding to the definition of a LOCA (10 CFR 50 Appendix A)".

3 q The reactor vessel head venting system isolation valves

~- will be supported from the seismic support platform.

The system can be disconnected downstream of the second isolation valve to accommodate refueling. In this manner, the necessary flanged connections will be outside of the reactor coolant pressure boundary. All piping and valves upstream of the flanges will remain integral with the reactor vessel head at all times.

1667 1.27 ACRS-2

F

'. s All items in the NRC position and clarification will be considered in the final detailed engineering design of the system.

C. Schedule January 1, 1981 is the target date for completion of required plant modifications consistent with procurement of equipment. Plant shutdowns will, of course, be required for any vessel head penetration or vessel instrument modifications.

i667 128 4

ACRS-3

~

2.2.1 IMPROVED REACTOR OPERATIONS COMMAND FUNCTION 2.2.1.a Shift Supervisor's Responsibilities A. Status

. The required review of the responsibilities and authority of the Shift Supervisor has been conducted as detailed below. Each item of the NRC position and clarification has been addressed. A special directive has been issued by the Executive Vice President responsible which emphasizes the primary management responsibility of the shif t supervisor for safe operation of the plant under all conditions on his shift and that clearly establishes his command duties.

B. Discussion Point Beach Nuclear Plant (PBNP) operations are carried out in accordance with the Administrative Control Policies and Procedures Manual (QA Volume 1) , which specifies the responsibility and authority of each key position in the plant, including that of the Shift Supervisor. The responsibility and authority of the Shift Supervisor are specified in Procedure PBNP 4.2 of QA Volume 1, with the latest revision dated December 14, 1979. A copy of the PBNP Administrative Control Policies and Procedures Manual is located at the NRC I&E offices in Glen Ellyn, Illinois, and updates are routinely provided directly to the NRC Compliance Inspector.

The following addresses each item of the NRC position specifically:

1. The highest level of corporate management at Wisconsin Electric Power Company which is respon-sible for defining and assigning the Shift Supervisor duties is the Manager, Nuclear Operations, Point Beach Nuclear Plant. We believe that this level of management has both the required direct knowledge and credibility to adequately inform the Shift Supervisors of their responsibilities and duties and the capability to monitor the resulting performance of these personnel. The safe operation of the plant under all con.ditions has long been the Wisconsin Electric Power Company's primary policy as established by management and carried out by all operating personnel. This is evidenced by the plant's record and the high standards of performance which have been maintained during its entire operating history. The existing pre-TMI plant policies and 1667 129 2.2.1-1

. . - - . - - - , -- . - .. -. --._ -.. . ~ - -. - .

procedures have resulted in_ periodic reviews of the authority and responsibilities of key plant personnel, including Shift Supervisors, and will continue to do so. .The NRC is appraised of these changes as previously stated.

We believe that these policies, procedures, and performance warrant consideration by the NRC as meeting the intent and implementation directives of this item and its clarification. We propose to continue Point Beach Nuclear Plant operations with this policy. We have, in addition, issued a special directive, dated December 31, 1979, signed by the Executive Vice President responsible, which clearly reaffirms the policy, recognizes specific review responsibilities, and reiterates the safety objective. This directive was issued to all Duty Shift Supervisors and other personnel whc have been defined as having access to the control room in an "off-normal" or transient situation (see Item

2. 2. 2.a, Control Room Access) . Additionally, it has been posted in the plant to inform all other operational and support personnel.

2. Plant procedures have been reviewed to assure the responsibilities and authority of the Shif t Supervisor. Operating Supervisor and the control operators are properly defined and that the command and responsibility of the Shift Supervisor is properly defined.

' The PBNP policies have always insisted that the Shift. Supervisor be in overall command (see PBNP 4.2). To insure this policy, the on-shift Operating Supervisor is the individual who is normally the direct supervisor of the control room operators during normal and emergency conditions. This allows the Shift Supervisor to remain removed'from the " individual operation" and keep his total involvement on the overall plant.

Revisions to the PBNP Administrative Control Policies

. and Procedures Manual have been made to remind and reenforce plant staff of these concerns.

The lines of authority of the Shift Supervisor were reaffirmed by the October 8, 1979 memo, G. A. Reed to Duty and Call Superintendents and Duty Shift Supervisors as quoted below:

. . . To reaffirm that line of authority, the Duty Shift Supervisor is the licensed person in charge of all shift operations and all orders are to be passed through or given by him or his delegated subordinate

( 193 in appropriate license level.

1667 130 2.2.1-2

"The Superintendent - operations (a licensed SRO) may override the orders of the Duty Shift Supervisor, but must do so only through the Duty Shift Supervisor chain of command. Any Duty and Call Superintendent (licensed or unlicensed) may counsel, assist and influence a Duty Shift Supervisor and will confer with him on Technical Specification-related items and reportable events as per QA Volume I, PBNP 3.13."

' " Successiono 'f authority and responsibility for "off normal" operations and any other control manipulations passes to the operating Supervisor if the Duty Shift Supervisor becomes suddenly dis bled or otherwise incapitated."

The responsibilities cf the operating Supervisor are as outlined in PBNP 4.3. These responsibilities specify his command function during emergency periods over the control room operators.

3. The SRO training program at PBNP has always and will continue to emphasize and reenforce the responsibility for safe operation and the proper. command function during normal operation and emergencies.

4. A review has been made of the administrative functions of the Shift Supervisor. It is felt that our present policies are proper'and complete. The policy at PBNP has always been to eliminate, to the greatest degree possible, the day-to-day administrative duties from the Shift Supervisor on watch. In this regard, for example, scheduling of overtime, review of surveillance testing, etc., are accomplished by the Operations Group day shift personnel. Areas which are necessary, however, to the proper knowledge of the plant, such as assignment of isolation boundary prior to maintenance and review of radiation work permits prior to commencement of work are still properly under the recognizance of the Shift Supervisor.

1667 131 2.2.1-3

. ~ .. . . . ..- --- - - . x-.-

2.2.1.b Shift Technical Advisor A. Status Duty and Call Technical Advisors (DCTAs) have been designated from among the degreed engineering and scientific personnel at PBNP for continuous accident response coverage beginning January 1, 1980. A DCTA will be available on site at all times and capable of being in the control room or the Technical Support Center within ten minutes of being notified of an "off-normal" event. The DCTA has no line authority and no authority to. perform or direct the manipulation of controls. The DCTA role is advisory to the Shift Supervisor and the Duty and Call Superintendent. The DCTA will also provide operating experience assessments and perform routine engineering evaluations of plant operations with emphasis on safety aspects. A single DCTA will' serve both Point Beach units.

B. Discussion The primary responsibility of the Duty and Call Technical Advisor will be to be available to the Shift Supervisor, when called, to support the diagnosis of off-normal events, and to advise the Shift Supervisor of actions to terminate or mitigate the consequences of such events.

Such events may include, but are not limited to, technical problems, reportable occurrences, or other significant operating events.

A Duty and Call Technical Advisor will be available onsite at all times and capable of reporting within ten minutes to the Shift Supervisor in the control room during off-normal plant conditions. The Shift Supervisor or the Duty and Call Superintendent may choose to direct the Duty and Call Technical Advisor to perform his advisory role from either the control room or the Onsite Technical Support Center, or they may direct the Duty and Call Technical Advisor to serve as a liaison between technical support personnel manning the Onsite Technical Support Center and the control room. The Duty and Call Technical Advisor will report to his regularly assigned plant group head during normal plant operations

'when not on duty assignment. During backshifts or holidays, the Duty and Call Technical Advisor, when not in the plant, will be housed onsite in the Energy Information Center (temporary ~ Operational Support Center) ready to respond.

Duty and Call Technical Advisors will have available at least two systems of communication to the control

' room. Four communication systems will be provided:

normal telephone, intra-plant paging, telephone " beeper" system, and a portable radio.

1667 132 2.2.1-4

The Duty and Call Technical Advisor position requires a knowledge of basic fundamentals in the areas of reactor physics, chemistry, materials, fluid mechanics, and heat transfer, in addition to the specifics of transient analysis and other safety analyses as applied to the Point Beach Nuclear Plant. A knowledge of Point Beach plant design, operating, and emergency procedures similar to that obtained by a licensed reactor operator is also required. An engineering or other scientific degree is desirable, but may be waived if sufficient nuclear-related experience has been obtained. The assigned Duty and Call Superintendent and Duty and Call Technical-Advisor will complement one another to provide total coverage of plant design, construction and operation for normal, transient, or accident conditions. The qualification basis for the Duty and Call Technical Advisor is discussed further below.

The complement of Duty and Call Technical Advisors will perform the operating experience assessment function as part of their normal day shift duties. On a continuing basis, they will be routed information and will be involved in the following areas of Point Beach operations and other reactor technical aspects through participation in or review of:

a. Safety significant modification requests,

b. Operating procedures, major maintenance procedures, I&C procedures, and emergency procedures and plans,

c. Point Beach Nuclear Plant Licenseee Event Reports and Significant Operating Events, and other applicable Licensee Event Reports,

d. Point Beach Nuclear Plant periodic reports to federal and state regulatory bodies,

e. NPRDS reports of component failures,

f. Manager's Supervisory Staff Meeting Minutes, g., Safety-related maintenance requests, and

h. Refueling and other major outage planning documents.

The technical qualifications for the Duty and Call Technical Advisor can be met in the short term by upgrading the current plant Technical Assistants.

The Technical Assistants are professional-level staff with B.S. or M.S. degrees in engineering or the sciences and are assigned to work in specific areas of the plant. Several of the Staff, because of their educaticn and rotated assignments in the past at Point Beach, have already achieved a majority of all the 2.2.1-s 1667 133

requirements for a Technical Advisor. Further, continued work on the day shift will afford.them the opportunity to be involved in the overall plant activities, meetings, training sessions, and development that only occur on the day shift. The need to keep abreast of the technology, in order to contribute as a Technical Advisor, is recognized as an overriding reason to keep these

- personnel on the day shift, rather than semi-isolated on shift work. In addition, shift work is usually unacceptable to professional personnel.

The long-term complement of Duty and Call Technical Advisors is envisioned to consist of professional-level Nuclear Plant Engineers and as alternate members experienced supervisors who have had at least ten years of plant operation as licensed. Senior Reactor Operators (SRO) . Additional training of the experienced SRO-type personnel.to meet the qualifications of the Technical Advisor position will be given as required on an individual basis. A college degree in engineering or science, therefore, may not be absolutely required for this position, but the fundamentals of an engineering degree needed to understand plant behavior during transients and accidents will be required. Additional training will stress these fundamentals as weil as -

plant behavior during transients and accidents.

C. Schedule Additional training will be provided for the DCTAs so that they will be more fully qualified for accident and operating experience assessment by January 1, 1981.

1667 134 2.2.1-6

t 2.2.1.c Shift and gelief Turnover Proce'dures A. Status The shift and relief turnover procedures have been reviewed and modified to i'nclude the checklists as discussed below.

These modified procedures address all of the items in the NRC position and are in use by plant personnel.

B. Discussion Control room personnel (Shift Supervisor, Operating Super-visor, and Control Operators) utilize relief and turnover procedures which include review of control room logs, plant and equipment status, and work activities. These procedures have been reviewed, in accordance with the intent -f the NRC position, and checklists were added where needed.

The Auxiliary Operator logs are reviewed by the Auxiliary Operators coming on watch prior to shift relief and additional signatures have been added.to include the off-going watchstander. All major equipment out of service, both safety-related and.non-safety-related, for.a unit in operation is listed on the status board for that unit in the main control roo~m. In all cases, the equipment moves in and out of service via the orderly command function of the Shift Supervisors. This responsibility does not reside with the Auxiliary Operator. The existing surveillance system already includes periodic verification of the position of valves in all safety-related systems.

A check of the effectiveness of the turnover procedure has been added to the Operations Group. surveillance system to provide a means of continuing evaluation of the turnover documents. .

1667 135 2.2.1-7

2.2.2 IMPROVED IN-PLANT EMERGENCY PROCEDURES AND PREPARATIONS 2.2.2.a Control Room Access A. Status We currently limit and will continue to limit control room access at Point Beach Nuclear Plant. The recognition by the NRC of the desirability of such limitation is commendable.

B. Discussion By special memorandum issued October 8, 1979, and revised December 26, 1979, from the Manager, Nuclear Operations, to the Point Beach Nuclear Plant Duty and Call Superintendents and the Duty Shift Supervisors, an administrative procedure has been implemented establishing the authority and responsibility of the person in charge of the control room to limit access during an abnormal operational transient or an accident to those individuals responsible for the direct operation of the Plant, to Technical Advisors, and to designated NRC personnel. The number of NRC personnel to be granted access will be limited and controlled as agreed between the senior Wisconsin Electric official -

onsite and the senior NRC person onsite. This special memorandum on control room access policy has been incorporated into the Administrative Control Policies and

. Procedures Manual (QA Volume 1) .

Current lines of control room authority and line of succession are alr.eady included in the Point Beach Nuclear Plant Administrative Control Policies and Procedures Manual (QA Volume 1) and in the regulations contained in 10 CFR 55. Those provisions have been reaffirmed in the memorandum referenced above. No further procedural modification is necessary to implement these considerations.

t

!667 136 2.2.2-1

2.2.2.b Technical Support Center I. Temporary Technical Support Center A. Status The conference room of the existing office area. at Point Beach _ Nuclear Plant has been designated and equipped as the Temporary Technical Support Center.. Appropriate revisions to the Emergency Plan have been made; copies of the revision have been forwarded to NRC by letter dated December 6, 1979.

B. Discussion The Temporary Technical Support Center (TTSC) is the normal reporting point for designated plant and Company off-site personnel and the Duty & Call Superintendent.

During backshifts, weekends and holidays, it is expected that the Duty & Call Superintendent (DCS) will be the first person to arrive at the TTSC (see Emergency Plan).

Additionally, any authorized NRC or other authorized non-Company personnel reporting in to support the plant organization in an emergency will report to the TTSC for instructions and bri~efing.

The TTSC has seating capacity for 24 people, 40 square feet of table top writing surface, and 36 square feet of blackboard surface. The room has an NRC " hot line" phone, one speaker phone plant extension, and one other additional handset plant extension. The TTSC has a two-channel Gai-tronics speaker and paging handset. Additional communi-cation in the form of two-way radio units on two different frequencies can also be made available if required for local communicaticns. The TTSC is also equi,pped with an FFDSAR, an Emergency Plan manual, Technical Specifications, and QA Manual (Volume 1).

In addition to the TTSC, the rest of the plant office is designated as supportive to the TTSC for other communi-cations, work space, restroom facilities, light food facilities, records, and records reproduction. Specifically, th'e following exists in the plant office area:

1. Six typewriters including four IBM Correcting Selectrics, one IBM magcard typewriter, and one IBM OS/6 Model 450 word processing unit.

2. One Xerox 700 reproduction machine.

3. .Two Bell & Howell Filemaster planetary camera units.

4. One microfilm reader (a second reader /printep is on order for 01/80). l66[ l}7 2.2.2-2

t .

5. Nine telephone extensions.

6. One telephone central switchboard with connections to three outside lines and three tie-lines to corporate headquarters in Milwaukee, one tie-line to Appleton, and one computer-use line.

7. Four Gai-tronics units.

8. All up-to-date Plant technical records since initial operation, including all operating logs, central file documents, and health physics and

. chemistry records.

In addition, the office maintains up-to-date copies of 10 CFR, NRC Regulatory Guides, Wisconsin Department of Industry, Labor, and Human Relations regulations, ANSI standards, and other reference material in the Plant library.

9. Access via microfilm to all Plant drawings, including reproduction. Additionally, many Plant drawings are still maintained in paper form.

10. All spare sets of FFDSARs, Technical Specifications,-

Emergency Plans, QA Volume l's, equipment technical manuals, all Plant major procedures, and many minor and non-nuclear safety related procedure master

. copies.

11. Two restrooms (male and female).

12. Health physics equipment locker outfitted with emergency monitoring equipment.

13. Miscellaneous office supplies.

The TTSC is equipped and located to " provide a place in close communication with the control room so as to have sufficient knowledge of current and projected plant status for orderly implementation of emergency procedures".

The PBNP Emergency Plan has been revised to address the TTSC and its use.

II. Permanent Technical Support Center A. Status Insufficient space is available at Point Beach Nuclear Plant for the conversion of an existing facility as the permanent Technical Support Center. New construction is, 1667 138 2.2.2-3

therefore planned immediately adjacent to the existing office area at the southeast corner of the Plant complex.

Conceptual design has been completed, and detailed design and engineering has begun. Selection of readout parameters and design of the display instrumentation will begin in January, 1980.

B. Discussion The permanent Technical Support Center will be located at the southeast corner of the existing building complex, immediately adjacent to and connected with the existing office area. The location for the Technical Support Center is shown in Figure 2.2.2-1. The existing office area will be reorganized and, together with the new Technical Support Center building, will orovide additional office space for plant personnel and resident NRC inspectors. The Technical Support Center will contain a readout room, restroom facilities, health physics instruments and supplies for use in the event of an accident, records and appropriate documents,*and a briefing room which will serve as the Operations Support Center.

The Technical Support Center will accommodate more than the required minimum of 25 persons. Appropriate radiation and airborne radioactivity monitoring instrumentation will be provided. Air filtration and shielding will also be provided to maintain radiologically acceptable habitability during an accident. Communications will include, as a mj-imum, similar facilities to those described for the temporary or interim Technical Support Center.

C. Schedule As committed in our letter of November 27, 1979, we are proceeding as rapidly as possible on the permanent Technical Support Center. However, this construction will require certain building permits and may also require Public Service Commission of Wisconsin authorization.

~

'Accordingly, we believe that a completion date of January 1, 1981, may be unrealistic. Conceptual planning has been completed and detailed engineering and design has begun. A construction schedule will be

provided as soon as it is available.

1667 139 c ,

2.2.2-4

?

j .

2.2.2.c Operations Support Center I. Temporary Operations Support Center A. Status The existing, onsite Energy Information Center (formerly called the Information and Training Building) has been designated as the temporary Operations Support Center.

Appropriate revisions to the Emergency Plan have been made; copies of the revisions have been forwarded to the NRC by letter dated December 6, 1979.

- B. ' Discussion The temporary Operations Support Center will serve as a personnel staging area in the event of an accident.

The building has an auditorium capable of seating about 75, a conference room with a capacity of about 75, restrooms, and a kitchen. There is an outside telephone line, a plant extension (w ith access to corporate tie lines), and the NRC h'ealth physics dial phone which has not yet been activated by NRC.

II. Permanent Operations Support Center A. Status The permanent Operations Support Center will be a designated

- portion of the permanent Technical Support Center. Refer to the discussion under 2.2.2.b.

B. Discussion The permanent Operations Support Center will serve as a personnel staging area in the event of an accident.

Refer to 2.2.2.b.

C.. Schedule Refer to 2.2.2.b.

1667 140 2.2.2-5

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0 Auxiliary Building roki Pumphouse 7, ,

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Plot Plan With Technical Support Center Figure 2.2.2 -1 .

P00R ORGM

ADDENDUM JANUARY 1, 1980 IMPLEMENTATION STATUS FOR NUREG-0578 ITEMS WISCONSIN ELECTRIC POWER COMPANY DOCKETS 50-266 AND 50-301 POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 .

SECTIONS 2.1.6.a 2.1.6.b 2.1.8.a 2.1.8.b 1667 142

2.1.6 POST-ACCIDENT CONTROL OF RADIATION IN SYSTEMS OUTSIDE CONTAINMENT 2.1.6.a INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT LIKELY TO CONTAIN RADIOACTIVE MATERIALS (ENGINEERED SAFETY SYSTEMS AND AUXILIARY SYSTEMS)

LEAKAGE REDUCTION AND PREVENTIVE MAINTENANCE PROGRAM TABLE OF CONTENTS

~

1.0 Commitment 2.0 Program Scope Definition 2.1 Included Systems 2.2 Excluded Systems 3.0 Program Objectives 3.1 Detection and Quantification of Leakage 3.2 Initiation of Corrective Action for Leakage Reduction 3.3 Preventive Maintenance 4.0 Program Implementation Requirements 4.1 System Analyses 4.2 Program Administration 5.0 References 1667 143 2.1.6.a - 1

LIST OF TABLES Table 1 Review of Systems Penetrating Containment Included in/ Excluded from Leakage Reduction and Preventive Maintenance Program 1667 144 2.1.6.a - 2

Leakage Reduction and Preventive Maintenance Program 1.0 COMMITMENT Wisconsin Electric Power Company (WEPCo) is committed to providing a leakage reduction and a preventive maintenance program description in response to NUREG 0578 Item 2.1.6.a for Units 1 and 2 of the Point Beach Nuclear Plant. Inasmuch as these programs are closely related to one another, it was determined that a single program approach would be most effective in satisfying the NUREG requirements. This document provides a description of the resultant Leakage Reduction and Preventive Maintenance Program.

2.0 PROGRAM SCOPE DEFINITION The Leakage Reduction and Preventive Maintenance Program described in this document is applicable to Units 1 and 2 of the Point Beach Nuclear Plant. The clarification of NUREG 0578 Item 2.1.6.a requires that " Licensees shall, by January 1, 1980, provide a summary description of their program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident... (and also) include a list.of those systems which are excluded from this program." Paragraphs 2.1 and 2.2 below identify those systems to be included in the Leakage Reduction and Preventive Maintenance Program and those to be excluded, respectively.

2.1 INCLUDED SYSTEMS The selection of systems which have been included in the Leakage Reduction and Preventive Maintenance Program was made on the following bases:

2.1.6.a - 3 lg 7

a. The operation of the system is needed for control, mitigation, monitoring, or recovery from an ,

accident; and

b. The system is in communication with liquids and/or gases contained in the reactor coolant system, containment recirculation sump, or containment atmosphere.

The following systems were identified:

Safety Injection, High and Low Head Containment Spray Reactor Coolant Liquid Sampling Contair. ment Atmosphere Sampling ,,

Hydrogen Control (Post Accident Containment Vent)

Chemical and Volume Control (including letdown, holdup, and reactor coolant purification)

Letdown Gas Purification Liquid Waste Disposal (including station drainage sumps and waste holdup)

Waste Gas Vent Header (including gas compressors) 2.2 EXCLUDED SYSTEMS A review of all piping systems penetrating the containment was performed. First, the containment penetrations associated with systems satisfying the two bases for inclusion into the Leakage Reduction and Preventive Maintenance Program were identified.

The portion of those systems associated with the remaining containment penetrations were then reviewed to determine their eligibility for exclusion from the program on one or both of the following bases:

2.1.6.a - 4 667 fNb

a. The system is isolated post-accident and is'not required to operate; or

b. The system may operate post-accident but is not in communication with liquids and/or gases contained in the reactor coolant system, containment recirculation sump, or containment atmosphere.

Results of this review are contained in Table 1.

The above review was necessary but not sufficient proof that the systems eligible for exclusion on the above bases would in fact be allowed to be excluded. The systems which have been included in the program must also be reviewed to identify interfacing systems and then to determine the nature of each interface. If the interface is such that the liquid or gas contained by the interfacing system is in direct communication with the liquid or gas contained by the included system, then the affected portion of the interfacing system must be treated as a part of the included system with which it interfaces or be separately treated as another included system. If direct communication with an interfacing systems is blocked by a closed valve, the system is excluded. However, the valve becomes part of the included system's leakage boundary and a potential leakage path. If direct communication with an interfacing system is blocked by the tubes of a heat exchanger, the interfacing system is excluded.

The tubes become part of the system leakage boundary, but are not considered a potential leakage path. Some specific systems excluded which are not associated with a containment penetration which, however, are peripheral to included systems are as follows:

1. Waste Liquid Processing System (after the Waste Holdup Tank);

I667 147 2.1.6.a - 5

2. Boric Acid Recycle System (after the Chemical and Volume and Control System Holdup Tanks);

3. Waste Gas System (after the Waste Gas Compressors);

and

4. Cryogenic System.

The portion of the systems review described above, which is needed to make a final determination of systems to be included and excluded from the Leakage Reduction and Preventive

~

Maintenance Program, will be part of actual program implementation and will be derived from the system analyses described in Section 4.1.

3.0 PROGRAM OBJECTIVES The position stated by NUREG 0578 Item 2.1.6.a requires that licensees " Establish and implement a program of preventive maintenance to reduce leakage to as-low-as-practical levels.

This program shall include periodic integrated leak tests at intervals not to exceed each refueling cycle." The Clarification adds: " Consider in your program to reduce leakage potential release paths due to design and operator deficiencies..."

3,1 DETECTION AND QUANTIFICATION OF LEAKAGE Detection and quantification of leakage is one objective of the Leakage Reduction and Preventive Maintenance Program. To meet this objective, a system-oriented leakage test / inspection program is to be developed and implemented. The systems scope for the test / inspection program is defined by paragraph 2.1. As part of the program, test procedures will be developed which will govern 1667 148 2.1.6.a - 6

c the testing of system components. The components to be tested will have been determined by system analysis to be part of the system's post-accident operating boundary. The system analysis will also determine post-accident operating pressures, testable system configurations, and required test pressures.

3.1.1 LIQUID FILLED SYSTEMS In the case of liquid filled systems, test pressures will approach as near as reasonably practical or exceed the associated peak post-accident operating pressures. However, it is a goal of

~

the program to test systems as near as possible to the normal system operating configurations, using the system pumps to develop test pressures, whenever possible. When this "open system" method is employed, leakage detection and quantification is achieved primarily by visual observation and collection of leakage as it escapes the system boundary. Detection and quantification of leakage through valves into interfacing systems is required only if that system or the affected portion is excluded from the Leakage Reduction and preventive Maintenance program. Techniques required to detect and quantify through leakage are developed in the test procedures on a case-basis since through leakage detection will tend to vary as a function of the specific system configuration being considered.

When system pumps cannot be used to develop the required test pressure, a test configuration will be developed so as to permit the use of a test pressure source. Detection and quantification methods needed are the same as for the open-system method described above. In addition, an integrated system leakage rate is also readily obtained by measuring the rate at which the test pump displaces fluid into the system from a reference volume (i.e., the test pump's fluid source). It may seem that the detection and quantification of integrated system leakage by this i667 149 2.1.6.a - 7

" hydrostatic test" method is superior to the "open-system" method used in conjunction with visual inspection and leakage collections. However, the " hydrostatic test" method is not the preferred test method for the following reasons:

a. It causes greater disruption to the normal operating configuration than the "open-system" method;

b. It requires more equipment (i . e. tes t pump, reference volume, water, motive power); and

c. It does not allow for testing system pump seal leakof f with the system pump (s) in operation.

3.1.2 GAS FILLED SYSTEMS Several methods of gas system leak detection and quantification will be incorporated into the test procedures. Some of these methods have been successf'211y used in containment isolation valve local leak rate testing (10 CFR50 Appendix J). These methods involve pressurizing (with air or nitrogen) a closed test volume to the required test pressure and proceding with one of the 3 tests described below.

a. Decay Test Method: Measure test volume pressure decay and temperature changes vs. time. Determine leakage rate using ideal gas relationships. The volume of the gas space under test must be known to determine the leakage rate.

b. Reference Volume Method: Connect an air or nitrogen pressurized tank (the reference volume) to the test volume through a gas type pressure regulator. The volume of the tank must be known; but the volume of i667 150 2.1.s.a - a

the gas space in the test volume need not be known.

The initial pressure in the reference volume must be initially higher than in the test volume. Pressure decay and temperature of the reference volume is measured while the pressure in the test volume is held constant by the regulator and the temperature is measured. Temperature in the test volume must be held constant if the volume of the gas space in the test volume is unknown. The leakage rate is determined using ideal gas relationships.

c. Flow Meter Method: Connect an air or nitrogen source to the test volume through a pressure regulator and gas flow meter. Regulate flow into the test volume at test pressure. Correct flowmeter readings for gas temperature and pressure.

An "open system" test using helium gas detection methods may also be used. This involves the injection of helium into a flowing system and a check of system components with a helium detection instrument. During initial system leak testing, which has been completed, this method was successful in detecting and locating leakage sources but did not quantify individual component or integrated system leakage rates. This method can be used as is or in conjunction with any of the other 3 methods described to aid in leak detection.

3.2 INITIATION OF CORRECTIVE ACTION FOR LEAKAGE REDUCTION 3.2.1 DESIGN AND OPERATIONAL DEFICIENCES The system analyses which are t- be performed for the systems defined in Paragraph 2.1 ideur.ify the components which are a part of each system's post-accident leakage boundary. This is to be 1667 151 2.1.6.a - 9

accomplished by the development of special system flow diagrams.

These diagrams, used in conjunction with detailed knowledge of system post-accident operating requirements, will facilitate recognition of system design and operational deficiencies and the impact of system configuration changes upon the integrity of the system. Items that are identified during system analyses that would appear to require re-design or operating procedural changes will oe recommended for further review. Changes made to existing designs or procedures as a result of this review process would be governed by existing administrative controls.

3.2.2 REPAIRS Initiating system repairs will be based upon engineering review of the leakage rate data obtained from the system tests /

~

inspections. The only applicable criteria for initiating repairs is the "as-low-as-practical" goal established by the NUREG.

Engineering review will entail an examination of each individual component for which the leakage detection methods described in 3.1.1 and 3.1.2 identify measurable leakage. The review will consider the following factors in determining if repair is necessary to achieve "as-low-as practical" leakage for each compone n t:

a. The improvement the repair is expected to achieve;

b. The time, materials, and costs required for repair and retest;

c. The effect that removing the component frem service

- for repair has upon the operability of the system or unit; and 1667 152 2.1.6.a - 10

d. The location of the leaking component relative to areas or access routes required for post-accident operation.

3.3 PREVENTIVE MAINTENANCE Preventive maintenance (P-M) intervals are based upon leakage and repair record trends derived from the system tests / inspections and the engineering reviews of leaking components. The goal again is to achieve "as-low-as-practical" system leakages.

Initially the P-M schedule would only include the items requiring repair as a result of an unsatisfactory initial test result. New items will be introduced into the P-M program as a result of subsequen'; tests in the same fashion. Components already established in the P-M program which are indicated by any of the subsequent tests to require repair prior to the next test date will be repaired immediately. In addition, the maintenance interval for that item will be changed to correspond to the period between the date of the last test the component failed and the date of the last test the component passed. When introducing a new component to the P-M program for which an identical component operated under similar conditions is already entered, the new component will adopt the maintenance interval of the identical item.

Using these rules will establish components in need of P-M on the schedule and rapidly force convergence of the required maintenance intervals to optimum levels. Components failing consecutive tests or otherwise requiring maintenance at what is considered too short an interval will be examined for maintenance / design problems and corrective action will be initiated.

1667 153 2.1. 6 . a - 11

4.0 PROGRAM IMPLEMENTATION REQUIREMENTS 4.1 SYSTEM ANALYSES For each system identified in Paragraph 2.1, a system analysis is prepared consisting of the following items:

a. Description of System Operating Modes;

b. Description of System Surveillance Test Modes that may be encountered during power operation;

c. Analysis of System Leakage Boundaries and Operating Parameters While Processing Highly Radioactive Reactor Coolant Liquid; and

d. Analysis of System Leakage Test / Inspection Requirements Listing:

Obj ectives Reactor Operating Modes Allowed During the Test Testing Intervals Test Description 4.2 PROGRAM ADMINISTRATION 4.2.1 RESPONSIBILITY The administration of the Leakage Reduction and Preventive Maintenance Program at the Point Beach Nuclear Plant is the responsibility of the Superintendent - Operations.

1667 154 2.1.6.a - 12

4.2.2 SYSTEM ANALYSES A consultant has been retained for performit.g the required systems analyses as described in 4.1.

4.2.3 LEAKAGE REDUCTION AND PREVENTIVE MAINTENANCE PROGRAM ADMINISTRATIVE INSTRUCTIONS / PROCEDURES The required instructions / procedures will describe program scope, the leakage test / inspection program, the engineering review function for corrective action initiation, and the engineering

~

review function for the preventive maintenance program.

4.2.4 LEAKAGE TEST / INSPECTION PROGRAM PROCEDURES The required procedures will be developed from the system analyses performed per 4.1. The procedures will provide the detailed information required to establish the proper system configu ration, to test the requir'ed system components, to detect and quantify leakage, and to document and report the results.

5.0 REFERENCES

NRC Documents

a. H.R. Der ton (NRC) to All Nuclear Power Plants letter, dated October 30, 1970, " Discussion of Lessons Learned Short Term Requirements"

b. Office of Nuclear Reactor Regulation, USNRC,

. NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations" 1667 155 2.1.6.a - 13

WEPCo Documents

a. S. Burstein (WEPCo) to H.R. Denton (NRC) letter, dated October 20, 1979 " Dockets 50-266 and 50-301 Implementation of NUREG-0578 Point Beach Nuclear Plant, Units 1 and 2"

b. S. Burstein (WEPCo) to H.R. Denton (NRC) letter, dated November 27, 1979 " Dockets 50-266 and 50-301 Implementation of NUREG-0578 Point Beach Nuclear Plant, Unit 1 and 2" Table 1 References

a. Point Beach Plant FFDSAR, Section 5.0: Containment System

b. WEPCo Inter-office Memorandum, R.E. Link to Glenn A.

Reed: " Containment Isolation Valves and Penetrations," July 20, 1979

c. Point Beach Nuclear Plant - Emergency Operating Procedure - 1A, Major, Rev. 16, 12/15/79: "Large Loss of Reactor Coolant" 1667 156 2.1.6.a - 14

TABLE 1 POINT BEACH NUCLEAR POWER PLANT UNITS 1 AND 2 REVIEW OF SYSTEMS PENETRATING CONTAINMENT INCLUDED IN/ EXCLUDED FROM LEAKAGE REDUCTION AND PREVENTIVE MAINTENANCE PROGRAM (1) (2) (3) (4) (5) (6)

POST PENETRATION ACCIDENT SYSTEM INCLUDED /

NUMBER DESCRIPTION OPERATION ISOLATED INTERFACE EXCLUDED 1 M.S. Loop A P SGS E 2 M.S. Loop B P SGS E 3 FW SGS E 4 FW SGS E 5 AFW R SGS E 6 AFW R SGS E 7 RHR Suction R RCL/CL I 8 RER Disch. R RCL/CL I 9 R.C.D.T. Drain CI RCL E 10 RC Letdown R CI RCL I 11 R.C.P Seal H2O Ret. CI RCL E 12a Demin. H 2O M E 12b Test conn. M E 12c N2 to RCDT CI RCL/RCG E RCDT Vent CI CL/CG E 13 SI to Cold Leg R RCL/CL I 14a N2 PRT CI E 14b Cont. Press R CL CG E Xmitter 14c N2 to Accum. CI E 15 CCW to RCPA P CL E 16 CCW to RCPB P CL E 17 CCW from RCPA P CL E 18 CCW from RCPB P CL E 19 CCW to Excess  ? CL E LD Ex 20 CCW from Excess  ? CI/CL E LD Ex 21 Spare X E 22 SI to Rx Vessel R RCL/CL I 23 Spare X E 24 Spare X E 25a Spare X E 25b Spare X E 1667 157 2.1.6.a - 15

TABLE 1 (Continued)

(1) (2) (3) (4) (5) (6)

POST PENETRATION ACCIDENT SYSTEM INCLUDED /

NUMBER DESCRIPTION OPERATION ISOLATED INTERFACE EXCLUDED 25c Post Acc. Cont. P M CG I vent Supply (Unit 1 only) 26 Charging P RCL I 27 SI to Rx Vessel R RCL I 28a RCS Hot Leg P CI RCL/RCG I Sample 28b Pzr Liquid Sample P CI RCL/RCG E 28c Pzr Steam Sample P CI RCL/RCG E 28d Dwt. Tester CL/M E 29a Seal H2O to P RCL I RCP A 29b Seal H2O to P RCL I RCP B 30a Test Conn. M E 30b Spare X E 30c Rx Makeup to PRT CI E 31a Cont. Press R CL CG E Xmitter 31b Post Acc. Vent P M CG I System Sample 31c Post Acc. Vent P M CG I System Exhaust 32a Cont. Press R M CG E Xmitter 32b SI Test Line M RCL/CL E 32c Aux Charging CI RCL/CL E Line 33a Inst. Air to P CI E Cont.

33b Inst. Air to P CI E Cont.

33c Service Air to M E Cont.

34a PRT Sample to CI RCL/RCG E Gas Analyzer 34b RCDT Sample to CI RCL/RCG E Gas Analyzer 35 Cont Coolers R CL E Ex-15A Supply 1667 158 2.1.6.a - 16

TABLE 1 (Continued)

(1) (2) (3) (4) (5) (6)

POST PENETRATION ACCIDENT SYSTEM INCLUDED /

NUMBER DESCRIPTION OPERATION ISOLATED INTERFACE EXCLUDED 36 Cont. Cooler R CL E Ex-15C Supply 37 Cavity Cooler R CL E Ex-30A Supply 30 Cavity Cooler R CL E Ex-30B Supply 39 Cont. Cooler R CL E

- Ex-15B Supply 40 Cont. Cooler R CL E Ex-15D Supply 41 Spare X E 42a Spare X E 426 Spare X E 42c Post Acc. Cont. R M CG I Vent System Supply (Unit 2 only) 43 Cont. Cooler R CL E Hx-15A Return 44 Cont. Cooler R CL E Ex-15C Return 45 - Cavity Cooler R CL E Ex-30A Return 46 Cavity Cooler R CL E Ex-30B Return 47 Cont. Cooler R CL E Ex-15B Return 48 Cont. Cooler R CL E Ex-15D Return 49 Spare X E 50 SGBD P SG E 51 SGBD P SG E 52 Aux. Steam to CL E Heaters 53 Cond. Return CL E From Heaters 54 Cont. Spray R CL I 55 Cont. Spray R CL I 56 Cont. Test Conn. M E 57 SG "A" Vent SG E l667 l59 2.1. 6. a - 17

TABLE 1 (Continued)

(1) (2) (3) (4) (5) (6)

POST PENETRATION ACCIDENT SYSTEM INCLUDED /

NUMBER DESCRIPTION OPERATION ISOLATED INTERFACE EXCLUDED 58 SG "B" Vent SG E 59 Spare X E 60 Spare X E 61 Spare X E 62 Spare X E 63 Spare X E

- 64 Spare X E 65 Spare X E 66 Spare X E 67 Spare X E 68 Spares X E 69 RHR Recire. R CL I Sump Line A 70 RHR Recire. R CL I Sump Line B 71 Cont. Sump to P CI CL E Aux. Bldg.

V1 Purge Exh. P CVI CG E V2 Surge Supply P CVI CG E Rll/R12 Air Particulate P CI CG I and Rad. Gas Mon. Supply Rll/R12 Air Particulate P CI CG I and Rad. Gas Mon. Return 1667 160 2.1.6.a - 18

NOTES FOR TABLE 1 Column (3) Post-Accident Operation P = Possible, depending on nature of accident R = Required for Post-Accident Operation

= Does not operate Post-Accident Column (4) Isolation SGS = Steam Generator System CL = Closed Loop CI = Containment Isolation CVI = Containment Vent Isolation M = Manual Locked Closed X = Fixed Isolation (e.g. spare)

= Not Isolated when system is operated Column (5) System Interface RCL = Reactor Coolant Liquid RCG = Reactor Coolant Gas CL = Containment Liquid CG = Containment Gas

= System Isolated from these interfaces Column (6) Included / Excluded I= Included in Leakage Reduction and Pr'aventive Maintenance Program F = Ercludad frem Leakage Reduction and Preventive Maintenance Program 1667 I61 2.1.6.a - 19

2.1.6 POST-ACCIDENT COMTROL OF RADIATION IN SYSTEMS OUTSIDE CONTAINMENT 2.1.6.b DESIGN REVIEW OF PLANT SHIELDING OF SPACES FOR POST ACCIDENT OPERATIONS TABLE OF CONTENTS 1.0 Plant Radiation Shielding Study Time Periods 2.0 Summary of Plant Radiation Shielding Study 3.0 Identification of Systems, Vital Areas and Occupancy Times 3.1 Identification of Systems 3.2 Vital Areas and Occupancy Times 4.0 Radiation Mapping of Chemistry Laboratory and Sampling S tation 5.0 Descriptions of Calculational Methods 3.1 Description of Computer Codes 5.2 Use of Computer Codes 5.3 Usage of Charts -

6.0 Dose Reduction Due to Time Decay i667 162 2.1.6.b - 1

7.0 Dose Reduction Alternatives 8.0 Abbreviations and Definitions of Terms 9.0 References 1667 163 2.1.6.b - 2

. . - . . ~ ~ . . . . . . . . - - - . . . - . . _ . . . . ... .

LIST OF FIGURES Figure 1 Access Areas, Access Routes, and Target Locations -

Elevation 46'-0" Unit 1 Figure 2 Access Areas, Access Routes, and Target Locations -

Elevation 26'-0" Unit 1 Figure 3 Access Areas, Access Routes, and Target Locations -

Elevation 26'-0" Unit 2 Figure 4 Access Areas, Access Routes, and Target Locations -

Elevation 8'-0" Unit 1 Figure 5 Access Areas, Access Routes, and Target Locations -

Elevation 8'-0" Unit 2 Figure 6 Tabulated Target Dose Rates Figure 7 Tabulated Corridor Dose Rates Figure S Dose Rate vs. Shield Thickness for Incremental Pipe sizes Figure 9 Dose Rate vs. Shield Thickness for Incremental Tubing Sizes Figure 10 Dose Correction Factor for Pipe Length and Target Distance - Unshielded Figure 11 Dose Correction Factor for Pipe Length and Target Distance - 1 ft. Shield Figure 12 Dose Correction Factor for Pipe Length and Target Distance - 3 ft. Shield Figure 13 Dose Correction Factor for Pipe Length and Target Distance - 5 ft. Shield Figure 14 Unit 1 Chemistry Laboratory and Sample Room Tabulated Dose Rates Figure 15 Unit 1 Chemistry Lab and Sample Station Radiation Mapping.

2.1.6.b - 3 1667 164

LIST OF FIGURES (Continued)

Figure 16 Dose Decay Chart Figure 17 Photon Spectrum as a Function of Time for Heavy Metals and Their Daughters Figure 18 Photon Spectrum as a Function of Time for Fission Products 1667 165 2.1.6.b - 4

1.0 PLANT RADIATION BUILDING STUDY TIME PERIODS For the purposes of this study, the post-accident period can be considered in three distinctive phases. Phase A is the period after accident initiation but prior to the initiation of RHR, CS or SI recirculation from the containment sump. During this phase it is assumed the containment has been isolated except for the injection water of the three mentioned systems, and the plant systems outside containment are not contaminated. Phase B is the period after recirculation initiation and extends to the time when cleanup starts. Phase C is the period after cleanup starts.

Phase B was chosen as having the greatest impact on the plant systems and accessible areas and therefore is the phase modeled for the study. This is based on the above definition of Phase A, and an estimated Phase C start of 30 days. With a 30 day decay period the dose levels are reduced by two orders of magnitude (a factor of 1/100).

Phase B was estimated to initiate at approximately one hour after accident initiation; this was based on time required to deplete the RWS Tank volume. For the purposes of this shielding study all dose levels are taken at one hour into the accident.

2.0

SUMMARY

OF PLANT RADIATION SHIELDING STUDY Using the calculational methods described in Section 5.0, dose rates and total dose per access were determined for selected

~

vital areas and access routes in ea n PBNP Unit. A conservative, non-mechanistic source term was assumed as required by NRC for the Short Term Lessons Learned Implementation. The occupancy times were estimated based on the time required to perform plant operations. The id'entification of the vital areas occupancy times, access paths, and transit times are identified in Section 2.1.6.6 - 5 1667 166

3.2. The results of the plant radiation shielding study for the Phase B perioo of the accident duration are shown in Figure 6.

The criteria for dose rates and for accessibility to vital areas are based on GDC 19 which limits the dose to an operator to 5 REM whole body or its equivalent to any part of the body during the course of the accident.

Unit 1 Accident The results of the radiation shielding study for PBNP shows that the Auxiliary building on the 8 ft elevation may be inaccessible south of column line 13, except for the CCW pump area which may be accessible on a restricted basis only.

On the 26 ft elevation of the Auxiliary building, the area south of column line 10 may be accessible on a limited basis only.

TJis area includes the chemistry lab, sample room and counting room. The area between column lines 10 and 13 may be accessible on a limited basis; however, there are hot spots due to direct radiation from the pipe tunnel area and through the open hatch in the floor.

The 46 ft elevation of the Auxiliary building would not pose a significant access problem.

The facade area surrounding the containment may be accessible on a limited basis. Dose levels from the containment and from the pipe penetration area (8 ft and 26 ft elevation) may limit -

occupancy times in the f acade area.

Specific results of target areas identified in Unit 1 are shown on Figure 6.

1667 167 2.1.6.b - 6

Unit 2 Accident The results of the radiation shielding study show that the Auxiliary b'uilding on the 8 f t elevation may be inaccessible north of column line 10.

On the 26 ft elevation of the Auxiliary building the area south of column line 10 would be accessible. The area north of column line 10 may be accessible on a limited access basis only.

The 46 f t elevation of the Auxiliary building would not pose a s.ignificant access problem.

The facade area surrounding the containment may be accessible on a limited basis except near the pipe penetrations on the 8 f t and ,

26 f t levels. Access to the facade area may also be restricted by dose levels from the containment.

Specific results of target areas identified in Unit 2 are shown on Figure 6.

3.0 IDENTIFICATION OF SYSTEMS , VITAL AREAS AND OCCUP ANCY TIMES 3.1 IDENTIFICATICN Of SYSTEMS As previously discussed, the post-accident period was considered in three distinct phases, and the systems of interest can be grouped and either considered unique to a phase or considered to '

be of concern during more than one phase. The following paragraphs discuss each phase and address the systems of interest to each individual phase.

1667 I68 2.1.6.6 - 7

3" . l .1 PHASE A: 0-1 HOUR AFTER INITIATION OF THE ACCIDENT During Phase A appropriate portions of the Residual Heat Removal (RER), Safety Injection (SI), and Containment Spray (CS ) Systems would be operational depending upon the seriousness of the accident. However, since these systems will be injecting non-contaminated water into the reactor coolant system (RCS) from storage tanks there will be no radioactive water (reactor coolant or contaminated injection water) being circulated throughout the auxiliary building. Thus, there is no shielding or accessibility problem regarding the auxiliary building for the duration of Phase A.

3.1.2 PHASE 3: 1 HOUR TO 30 DAYS AFTER INITIATION OF THE ACCIDENT Upon initiation of Phase B, the RHR, SI, and CS systems will begin the recirculation phase from the containment sump. At this time radioactive reactor coolant or injection water will begin to be continuously circulated throughout the applicable portions of the RHR, SI, and CS systems. Since these systems are located in the auxiliary building and will be circulating highly radioactive water, access to the auxiliary building will be limited to that which is absolutely necessary. Accessibility limitations are discussed in later portions of this section.

The shielding analysis considered the effects of operation of the RHR, SI, and CS systems in their recirculation mode during this phase of the accident. Detailed pipe routings of these systems -

were prepared. Areas required to be accessible and the corresponding occupancy times were determined. Estimated dose rates for 2ach area were calculated.

1667 169 2.1.6.b - 8

The following is a description of the flow paths of the RER, SI, and CS systems operating in the recirculation mode.- Applicable Unit 1 and Unit 2 equipment, line, and val've numbers are noted for clarity. The flow path of the recirculation mode of either the Unit 1 or Unit 2 RHR, SI, and CS systems begins within the reactor containment building.at Containment Sump "3" with the radioactive water being introduced to the suctions of the residual heat removal pumps (Unit 1 Train A: 1-P10A; Unit 1 -

Train B: 1-P10B; Unit 2 - Train A: 2-P10A; Unit 2 - Train 3:

2-P10B) through con;qcutive lin9s numbered 10-SI-151R, 10-AC-60lR-1, and 10-AC-60lR-3. From the RER pumps the radioactive coolant is directed to the RER heat exchangers (7 Tit 1 - Train A: 1-HX11 ; Unit 1 - Train B: 1-HX11B; Unit 2-Train A: 2-HX11 ; Unit 2 - Train B: 2-dX113) through line number 8-AC-60lR-4. The discharge of the RHR heat exchangers, travels through line numbered 8-AC-601R-6 to a point at which line number 6-AC-60lR-6 introduces water into the suction lines of both the safety injection pumps and the containment spray pumps. The Unit 1 and Unit 2 Train A and Train B cross-tie lines numbered 8-AC-60lR-4 are also of concern. Radioactive water is introduced to the suctions of the safety injection pumps (Unit 1 - Train A:

1-P155 ; Unit 1 - Train B: 1-P15B; Unit 2- Train A: 2-P15A; Unit 2 - Train B: 2P15B) through consecutive lines numbered 6-SI-151R-5 and 6-SI-151R-2. From the discharge of the SI pumps, the radioactive fluid travels through consecutive line_s numbered 6-SI-150lR-1 and 4-SI-150lR-2 (Train B - Core Deluge Injection Line) and 4-SI-150lR-3 (Train A- Cold Leg Injection Line) into the reactor containment building. The shielding terminates at the point where these latter lines penetrate the reactor containment building. Radioactive water is introduced to the suctions of the containment spray pumps (Unit 1 - Train A:

1-P14A; Unit 1 - Train B: 1-P14B; Unit 2 - train A: 2-P14A;. Unit 2 - Train B: 2-P1413 ) through lines numbered 6-SI-151R-3. From the discharge of the containment spray pumps, the radioactive 2.1.6.b - 9 l 7 1

fluid travels through lines numbered 6-SI-30lR-1 into the reactor containment building. Additionally, the containment spray pump recirculation line could be contaminated (lines numbered 2 -

SI-30lR-3).

3.1.3 PHASE C: 30 DAYS AFTER INITIATION OF THE ACCIDENT UNTIL CLEANUP IS COMPLETE Several systems have been identified as cleanup systems. These systems are as follows:

a. Reactor coolant purification using normal letdown flow from the reactor coolant system to the Chemical and Volume Control System demineralizers;

b. Reactor coolant purification using the letdown gas stripper system;

c. Waste Gas System (necessary when operating the letdown gas stripper system or liquid holdup tanks);

d. Liquid holdup tank operation;

e. Operation of the Blowdown Evaporator in its waste mode; and

f. Post-Accident Containment Venting System including its sampling capability. _

These systems were not included in the shielding study because their operation was not necessary until 30 days after initiation of the accident. Areas associated with these systems that are required for access have been reviewed from a worst case basis by considering that access is required during the recirculation mode of the RER, CS, and SI systems.

2.1.6.5 - 10 ibb[ l[}

3.2 VITAL AREAS CCCUPANCY TIMES AND ACCESS ROUTES Table 1 is a listing of vital areas by "T" number, estimated occupancy time, access route by "C" number, and estimated transit time. The transit times were based en the average man walking at a pace of 4 ft per second.

The vital areas and access route locations are identified on Figures 1 through 5. Figure 6 is a tabulated sammary of the dose levels for the vital areas; Figure 7 is a tabulated summary of the dose levels in the access corridors.

1667 172 2.1.6.6 - 11

TABLE 1 Target Occupancy Time

• Access Route (s) /

4 Description (minutes) Transit Time (sec)

T-01 Unit one Elevator 5 - 10 C-1/4 T-02 1B31 Motor Control Center 5 - 10 C-27/22 T-03 2B31 Motor Control Center 5 - 10 C-27/22,C-28/40 T-04 CCW Heat Exchangers 30 C-27/22,C-29/24 T-05 Gas Compressors 15 - 30 C-30/32 T-06 Letdown Gas Stripper 5 - 10 C-30/8,C-31/12 Valving T-07 Entrance Unit one Pipeway #1 5 - 10 C-1/5,C-2/25 T-08 1832 Motor Control Center 5 - 10 C-1/5,C-2/25,C-3/26 T-09 SI & CS Pump Area 5 - 10 C-1/5,C-2/25, C-3/26,C-4/20 T-10 CCW Pumps 5 - 10 C-5/14 T-ll 2332 Motor Control Center 5 - 10 - C-5/14,C-6/10 T-12 Entrance Unit Two 5 - 10 C-5/14,C-6/10, Pipeway 94 C-7/20 T-13 Cryogenics Control Panel 5 - 10 C-5/14,C-6/10, C-7/10,C-8/S T-14 B33 and B43 Motor Control 5 - 10 C-9/8 rente-T-15 Hold-Up Tank Recire. Pump 5 - 10 C-9/18 T-16 Cas Decay Tank Valve Gallery 5 - 10 C-14/76 T-17 Unit One Sample Room 5 - 10 C-14/66,C-15/18 T-18 Unit one VCT Valve Gallery 5 - 10 C-14/66,C-15/14 C-16/4 T-19 B.A. Transfer Pumps and 5 - 10 C-14/66,C-15/8 Outside Letdown Filter 1 C-16/6,C-17/15 T-20 C59 and Central Stairway ** 20 C-18/8 T-21 Unit Two Letdown Filter 5 - 10 C-18/40,C-19/0 (Valve Gallery) ( Alt: C-14/66, C-15/9,C-10/16, C-ll/16,C-12/8 2.1.6.6 - 12 1667 173

Target Occupancy Time

• Acces s . Route (s ) /

4 Description (minutes) Transit Time (sec) ***

T-22 Unit Two VCT Valve 5 - 10 C-18/8,C-19/40, Gallery C-20/8 (Alt. C-14/66, C-15/9,C-10/16, C-ll/16,C-12/8, C-13/3 T-23 Unit Two Sample Room 5 - 10 C-18/8,C-19/40, C-20/8,C-21/24 (Alt. C-14/66, C-15/9,C-10/16, C-ll/16,C-12/8, C-13/12 T-24 Unit One 42 Pipeway 5 - 10 C-18/8,C-22/16 P.A.V T-25 Demineralizer valve 5 - 10 C-18/8,C-22/16, Gallery C-23/12 T-26 Unit Two $3 Pipeway 5 - 10 C-18/8,C-26/36 P.A.V T-27 HUT Vent Valves 5 - 10 C-25/10

• Occupancy time is the time required at station (target point) to perform the designated plant operation

• • C-59 requires access during first hour

• • • Transit time includes time to and from the target 1667 174 2.1.6.b - 13

4.0 RADIATION MAPPING OF CHEMISTRY LABORATORY AND S AMPLING STATION Unit 1 Accident:

For a mapping of the Chemistry Laboratory, Sample Room and access corridors, see Figure 15. Figure 14 is a summary listing of dose levels for selected target points.

A review of the data indicates these areas are accessible on a limited basis.

Unit 2 Accident:

A summary listing of dose levels in Unit 2 is provided in Figure 14. The Unit 2 sample area is accessible; however, the dose levels in the access corridors are high (2-27R/hr); therefore accessing and occupancy times are limited.

5.0 DESCRIPTION

S OF CALCULATIONAL METHODS

5.1 DESCRIPTION

OF COMPUTER CODES The two computer codes used in this study were ORIGEN and ISOSHLD. ORIGEN computed the radioactive source terms used by ISOSHLD to compute the dose rates from piping and various pieces of equipment.

ORIGEN is a versatile point depletion code which solves the equations of radioactive growth and decay for large numbers of isotopes with arbitrary coupling. The code uses the matrix exponential method to solve a large system of coupled, linear, first-order ordinary dif f erential equations with constant coefficients. The general nature of the matrix exponential 2.1.6.b - 14

method permits the treatment of complex decay and transmutation schemes. An extensive library of nuclear data has been compiled, including half-lives and decay schemes, neutron absorption cross sections, fission yields, disintegration energies, and multigroup photon release data. ORIGEN has been used to compute the radioactivity of fission products and fuel materials that were assumed to be released f rom the reactor core to become the source terms for the dose rate calculations. The code also permits the simulation of chemical processing so that the noble gases halogens, and other fission products may be partitioned as required by the Nuclear Regulatory Commission's guidance on source terms.

ISOSHLD calculates the decay gamma-ray dose at the exterior of a shielded radiation source. The source may be one of a number of common geometric shapes. The code calculates shield region mass attenuation coefficients, buildup factors, and other basic data necessary to solve the specific problem.

ISCSHLD performs kernel integration for common geometric shapes.

The " standard" point attenuation kernel is numerically integrated over the source volume for up to 2s source energy groups.

Buildup is considered characteristic of the last shield region (or a dif f erent specified region) but dependent on the total number of mean free paths from source to dose point, and is obtained by interpolation on ef f ective atomic number f rom a table of point isotopic buildup factor data. Mixed mass attenuation coefficients are obtained from a library of basic data using code _

input material density specifications. The source strength may be specified as any of three inputs: 1) as the emissions from a selection of fission products irradiated under specific conditions, 2) the curies of particular fission and/or activation products, or 3) a number of photons per second of energy 2 specified by input. The source strength was specified by the 2.1.e.b - is 1667 176

later method for this study since the ORIGEN output provided this information.

ISOSHLD was used to perform a parametric study of dose rate as a f unction of pipe diameter , pipe length, distance from source to receptor, amount of intervening concrete shielding, and time after accident. This data was compiled in the following manner:

a. A plot of dose rate f rom a " infinite" length of pipe 10 feet from the receptor verses shield thickness over the range of zero to five feet of concrete. A family of curves was produced for pipe / tube diameters ranging from 0.25 inches to 14 inches. These plots are shown in Figures 8 and 9.

b. A family of curves giving the correction factor for the actual length of pipe and the actual distance to the receptor was then generated. The correction factor was plotted verses the pipe length of up to 40 feet.

Separate curves were produced for distances from receptor to the pipe over the range of 6 inches to 20 feet. These plots are shown in Figures 10, 11, 12, and 13.

c. The above data was based on the ource term I hour af ter the accident. Table factors due to radioactive decay gives the correction for various times up to 30 days after the accident (see Section 6.0) . .

5.2 USE OF COMPUTER CODES ORIGEN was used to estimate the radioactive inventory of the reactor core at the end of a fuel cycle. The approach assumed that all the fuel in the core (48.6 MTU) was irradiated under the 2.1.6.b - 16 1667 177

same conditions for the same length of time, i.e. 31.23 MW/MTU for 1057 days to give a burnup of 33,000 MWD /MTU. After irradiation, the nuclides were partitioned by chemical type; then the gamma production rates were calculated after periods of radioactive decay of up to 30 days. The resulting source strengths are shown in Figures 17 and 18.

5.3 US AGE OF CHARTS The Figure 8 through 13 charts were developed as a convenient and reliable method of estimating dose levels for a given target point. For personnel corridors target points are considered to be 3 Ft from the floor.

To use the charts, one first defines the size of the pipe, and shield thickness. The dose rate for an infinite length of pipe or tubing at 10 Ft target distance is read from Figure 8 or 9.

Next a correction f actor for actual pipe length and actual target distance is selected from Figure 10, 11, 12 or 13 depending on shield thickness. This correction factor is then multiplied by the dose rate determined from Figure 8 or 9 to give the dose rate for the target point. For multiple source targets , the dose rate is additive, therefore this process is completed for each significant source to yield total dose rates.

6.0 DOSE REDUCTICN DUE TO TIME DECAY Figure 16 is a dose correction factor for extended periods into .

the accident. From the curve at the selected period into the accident, a normalized dose is read. This is actually the fractional dose remaining at the given time period. The normalized dose is then multiplied by the dose value derived in section 6.0 for the dose at the given time period. For the purposes of this study, all dose rates have been assumed to occur at one hour after the accident.

2 .1. o. .o-le 1667 178

7.0 DOSE REDUCTION ALTERNATIVES The results,which have been obtained in the shielding study indicate the existence of several potential problem areas (r ef er to Figures 6 and 7) .

These results suggest the following potential dose reduction alternatives:

a. Delaying access until necessary to allow dose reduction due to radioactive decay;

b. Providing access pathways other than those considered;

c. Localized shielding of radioactive piping;

d. Shielding general areas of radioactive piping;

e. Remote control of components rather than manual control;

f. Reroute of the radioactive piping of the safety injection pump discharge lines; and 9 Providing an electrically-driven shielded vehicle for personnel transport within certian areas of the Auxiliary Building. .

Considerations involved in the selection of one or more of these alternatives include the need for piping or structural analysis ,

time duration studies for access to radioactive areas, shielding design and installation, and system or equipment design modifications, i667 l79 2.1.6.b - 18

8.0 ABBREVIATIONS AND DEFINITIONS OF TERMS CCW - Component Cooling Water System CS -

Containment Spray System Dose - The energy absorbed by a specified area, volume, or the whole body Dose Rate - Rate at which energy is being absorbed Erg - A unit of energy in the centimeter, gram, second (CGS) system of units (1 Mev = 1.602 x 10-6 gag) ,

mR/hr - Rem x 10-3 per hour R- Rem R/hr - Rem per hour Rad - The unit of absorbed dose, which is 100 ergs per gram in any medium.

Radiation - Radiation is the transmission of energy through space. In the case of nuclear radiation the energy is generated from the decay of unstable atomic nuclei.

Radioactive- This refers to a pipe containing radioactive Pipe material.

Rem - A rad multiplied by a factor which reflects the (Roentgen damage to tissue by different types of radiation.

Equivalent This term applies to body tissue only. For gammas Man) this factor is 1, for fast neutrons this factor is usually 10, for other types of radiation this factor can be as high as 20.

RHR - Residual Heat Removal System Roentgen - Is that amount of X-or gamma radiation which produces 2.08 x 10 9 ion pairs in one cubic -

centimeter of dry air.

RWS Tank - Reactor Water S torage Tank SI - Safety Injection System 1667 180 2.1.6.6 - 19

9.0 REFERENCES

NRC Documents

a. Harold Denton (NRC)/All Nuclear Power Plants letter dated October 30, 1979, " Discussion of Lessons Learned Short Term Requirements".

b. Office of Nuclear Reactor Regulation, USNRC , NUREG-0 57 8, "TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations".

WEPCO Documents

a. Burstein (WEPCo) /Denton (NRC) letter dated October 20, 1979, " Dockets 50-266 and 50-301 Implementation of NUREG-0578, Point Beach Nuclear Plant Units 1 and 2" .

b. Burstein (WEPCo)/Denton (NRC) letter dated November 27, 1979, " Dockets 50-266 and 50-301 Implementation of NUREG

- 0578, Point Beach Nuclear Plant, Units 1 and 2".

1667 18i 2.1.6.b - 20

3k*d t

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TABULATED TARGET DOSE RATES UNIT 1 ACCIDENT Sheet 1 of 3 Target Total Accumulated Target Dose Rate Dose Per Access No. Description (R/hr) (Rem)

T-00 Control Room Negligible Negligible T-01 Unit One Elevator (Elev. 8') 0.8 0.135 T-02 1B31 Motor Control Center 0.2 0.034 T-03 2B31 Motor Control Center 0.14 0.025 T-04 CCW Heac Exchangers 0.16 0.082 T-05 Gas Compressors 0.16 0.084 T-06 Letdown Gas Stripper Valving 0.32 0.055 T-07 Entrance Unit 1 Pipeway #1 2820. 483.

T-08 1B32 Motor Control Center 1400. 430.

T-09 SI & CS Pump Area 7300. 1272.

T-10 CCW Pumps 1.96 52.32 T-ll 2B32 Motor Control Center Negligible 0.024 T-12 Entrance Unit 2 Pipeway 44 Negligible 0.024 T-13 Cryogenics Control Panel Negligible 0.024 T-14 B33 and B43 Motor Control Center 0.96 0.164 T-15 Hold-up Tank Recirc. Pump 0.96 0.164 T-16 Gas Decay Tank Valve Gallery 6.6 1.1 T-17 Unit 1 Sample Room 25. 4.2 T-18 Unit 1 VCT Valve Gallery 3.3 0.58 T-19 B.A. Transfer Pumps and Outside Letdown Filter 1 2.16 0.449 T-20 C59 and Central Stairway 625. 208.

T-21 Unit 2 Letdown Filter (Valve Gallery) Negligible 1.123 T-22 Unit 2 VCT Valve Gallery Negligible 1.123 -

T-23 Unit 2 Sample Room Negligible 1.123 T-24 Unit 1 #2 Pipeway PAV 4310. 727.1 T-25 Demineralizer Valve Gallery Negligible 18.14 T-26 Unit 2 #3 Pipeway PAV Negligible 19.6 T- 7. / Hut Vent valves 180. 30.5 2.1.6.b Figure 6 lhf[ lgg

TABULATED TARGET DOSE RATES UNIT 2 ACCIDENT Sheet 2 of 3 Target Total Accumulated Target Dose Rate Dose Per Access No. Description (R/hr) (Rem)

T-00 Control Room Negligible Negligible T-01 Unit One Elevator (Elev 8 ' ) Negligible Negligible T-02 1331 Motor Control Center 0.2 0.034 T-03 2331 Motor Control Center 0.14 0.025 T-04 CCW Heat Exchangers 0.16 0.082 T-05 Gas Compressors 0.16 0.084 T-06 Letdown Gas Stripper Valving 0.32 0.055 T-07 Entrance Unit 1 Pipeway #1 Negligible Negligible T-08 1932 Motor Control Center 2.9 0.5 T-09 SI & CS Pump Area 9700. 1672.

T-10 CCW Pumps 2200. 370.7 T-ll 2B32 Motor Control Center 760, 136.14 T-12 Entrance Unit 2 Pipeway 44 2000. 353.3 T-13 Cryogenics Control Panel 5. 15.53 T-14 B33 and B43 Motor Control Center 1670, 279.2 T-15 Hold-up Tank Recire. Pump 0.96 0.164 T-16 Gas Decay Tank Valve Gallery Negligible Negligible T-17 Unit 1 Sample Rcom Negligible Negligible T-18 Unit 1 VCT Valve Gallery Negligible Negligible T-19 B.A. Transfer Pumps and Outside Letdown Filter 1 2.16 0.449 T-20 C59 and Central Stairway 625. 208.

T-21 Unit 2 Letdown Filter (Valve Gallery) 5. 1.123 T-22 Unit 2 VCT Valve Gallery 9.5 1.99 T-23 Unit 2 Sample Room 25. 4.6 T-24 Unit 1 #2 Pipeway PAV Negligible 0.076 T-25 Demineralizer Valve 0.36 18.2 Gallery T-26 Unit 2 93 Pipeway PAV 3960. 679.7 T-27 Hut vent valves 140. 23.8 2.1.6.b i667 190 Figure 6

TABULATED TARGET DCSE RATES Sheet 3 of 3 Dose Rate Criteria - Accessibility is based on GDC 19 which limits the dose to an operator to 5 REM whole body or its equivalent to any part of the body during the course of the accident.

Occupancy times are as shown in Section 3.2.

1667 191 2.1.6.b Figure 6

TABULATED CORRIDOR DOSE RATES UNIT 1 ACCIDENT Sheet 1 of 2 Corridor No. Elevation Dose Rate (R/hr)

C01 46'/8' 2.

CO2 8' l.4 C03 8' 4000.

C04 8' 4000.

C05 8' 1.96 C06 8' l.96 C07 8' Negligible C08 8' Negligible C09 8' O.8 C10 26' 24.5 C11 26' 5.

C12 26' 27.

Cl3 26' 7.

C14 26' ,. Negligible Cl3 26' O.002 C16 26' O.6 C17 26' 2.09 C18 26' 35.

C19 26' 28.3 C20 26' Negligible C21 26' Negligible C22 26' 1960.

C23 26' 2800.

C24 26' 1960.

C25 26' 180.

C26 ,

26' 1950.

C27 46' Negligible C28 46' O.08 C29 46' O.08 C30 46' O.4 C31 46' O.4 2.1.6.b Figure 7 1667 192

TABULATED CORRIDOR DC3E RATES UNIT 2 ACCIDENT Sheet 2 of 2 Corridor No. Elevation Dose Rate (R/hr)

C01 46'/8' Negligible CO2 8' Negligible C03 8' O.15 C04 8' 4000.

C05 8' 960.

C06 8' 2000.

C07 8' 2000.

C08 8' 5.

C09 8' 800.

C10 26' 24.5 Cll 26' 5.

C12 26' 27.

C13 26' 7.

C14 26' Negligible C15 26' Negligible C16 26' Negligible C17 26' 2.09 C18 26' 35.

C19 26' 28.3 C20 26' 2.

C21 26' 8.

C22 26' 1960.

C23 26' 2800.

C24 26' 1960.

C25 26' 180.

C26 26' 1950.

C27 46' O.2 C28 46' O.08 C29 46' O.08 C30 46' O.4 C31 46' ,

0.4 2.1.6.b Figure 7 1667 193

. _ . _ . .._ , _ , . _ _ _ . . _ ~ . -

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9 DOSE RATE VS. SHIELD TH!CKNESS FOR INCREMENTAL TUBING SIZES l

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DOSE CORRECTION FACTOR FOR PIPE LENGTH AND TARGET DISTANCE - UNSHIELDED Unshielded -

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DOSE CORRECTION FACTOR FOR PI?E LENGTH AND TARGET DISTANCE - 3 FT. SEIELD 3 FP Ghidd 24 -

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DOSE CORRECTION FACTOR FOR PIPE LENGTH AND TARGET DISTANCE - 5 FT. SHIELD 5 ft 3 h. L e I d 2.%

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2.1.s.3 1667 199 Figure 13

Ul CHEMISTRY LAB AND SAMPLE ROOM TABULATED DOSE RATES (See Figure 15 for Locations)

TARGET NO. DOSE RATE (R/hr) TARGET NO. DOSE RATE (R/hr)

S-01 9.0 C-14a 0.32 S-02 11.0 C-14b 2.6 S-03 5.0 T-16 6.6 S-04 9.0 T-17 25.0 S-05 6.0 T-18 3.3 5-10 2.0 S-11 0.006 S-C7 6.6 S-08 7.6 S-09 7.2 U2 SAMPLE ROOM AND ACCESS TABULATED DOSE RATES (See Figures 2&3 for Locations)

TARGET NO. DOSE RATE (R/hr) TARGET NO. DOSE RATE (R/hr)

T-23 25.0 C-14a 0.32 C-10 24.5 C-14b 2.6 C-11 5.0 C-12 27.0 C-13 7.0 2.1.6.b Figure 14 1667 200

UNIT I CHEMISTRY LAB AND SAMPLE STATION RADIATION MAPPING

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NORMALIZED DOSE

& 1667 202 mest ura 1 a tvisio n s  ! ev Wrvt4 u (s e CIS #3 M0" C ##P ' "'*

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DOSE DECAY CHART (Dose Rate as a Function of Time)

Sheet 2 of 2 Time Relative Dose Rate Normalized Dose Rate 1 HR 1.39E6 1.

2 HR 1.01E6 0.727 3 HR 8.04E5 0.578 8 HR 4.44E5 0.319 1 DAY 2.21ES 0.159 3 DAYS 1.08E5 0.0777 10 DAYS 4. 91E 4 0.0353 20 DAYS 2.40E4 0.0173 30 DAYS 1.43E4 0.0203

/

l667 203 2.1.6.b Fi' Jure 16

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2.1.8 INSTRUMENTATION TO FOLLOW THE COURSE OF AN ACCIDENT 2.1.8.a IMPROVED POST-ACCIDENT SAMPLING CAPABILITY TABLE OF CONTENTS 1.0 Commitment 2.0 Summary of Implementation 2.1 Reactor Coolant Sampling 2.2 Containment Atmosphere fampling I.3 Post-Accident Analysi.s 2.4 Training 3.0 Design Review of Reactor Coolant Sampling and Laboratory Analysis 3.1 Design Review of Reactor Coolant Sampling 3.2 Design Review of Existing Reactor Coolant Laboratory Analysis 3.3 Interim Methods for Sampling Reactor Coolant 3.4 Interim Methods for Reactor Coolant Laboratory Analysis 1667 206 2.1.8.a - 1

4.0 Design Review of Containment Atmosphere Sample System and Laboratory Analysis 4.1 Design Review of Existing Containment Atmosphere Sample System 4.2 Design Review of Existing Containment Atmosphere Laboratory Analysis 4.3 Interim Methods for Containment Atmosphere Sampling 4.4 Interim Containment Atmosphere Laboratory Analysis 5.0 Alternatives 5.1 Alternatives foc Reactor Coolant Sampling 5.2 Alternatives for Containment Atmosphere Sampling 5.3 Alternatives for Laboratory Analysis 6.0 References i667 207 2.1.8.a - 2

List of Figures Figure 2.1.8.a-1, Sample Bomb Curie Content vs Contact Reading - One RCS Dilution, No Recirculation (Unit One and Two) Figure 2.1.8.a-2, Sample Bomb Curie Content vs Contact Reading - SIS Dilution and Recirculation (Unit One) Figure 2.1.8.a-3, sample Bomb Curie Content vs Contact Reading - SIS Dilution and Recirculation (Unit Two) Figure 2.1.8.a-4, Conceptual Design for Post-Accident Sample Bomb Figure 2.1.8.a-5, Conceptual Flow Diagram for Remote Dilution and Sample Preparation Figure 2.1.8.a-6, Containment Gas Sampling System / 1667 208 . 2.1.8.a - 3

1.0 COMMITMENT Wisconsin Electric Power Company has analyzed the capability for taking samples under accident conditions at Point Beach Nuclear Plant, her implemented certain interim improvements for obtaining samples under accident conditions, and will implement permanent upgrading of sampling facilities by January 1, 1981. 1667 209 2.1.8.a - 4

2.0

SUMMARY

OF IMPLEMENTATION 2.1 REACTOR COOLANT SAMPLING A design review of the existing sampling system and procedures has been completed, and an alternative location for obtaining a reactor coolant hot leg sample containing post-accident source terms has been installed. The location selected is accessible for a length of time sufficient for the technician to obtain the sample while receiving a whole body exposure of less than 900 mR. A shielded sample bomb with provision for remote handling has been fabricated. Procedures for post-accident sampling which include provisions for the use of portable survey instruments to permit rapid assessment of high exposure rates and sample activity have been developed. Operations at the sample station and transport of the sample have been analyzed considering the dose contributions from all potential post-accident sources. Sufficient shielding and procedural precautions are in place to assure that radiation dose criteria are met. 2.2 CONTAINMENT ATMOSPHERE S AMPLING A design review of the existing containment atmosphere sampling system and procedures has been completed. A modification has been installed to allow sampling of the containment atmosphere under postulated accident conditions. A shielded carrier for the sample has been fabricated. A technician can obtain and deliver the containment atmosphere sample to the Chemistry- Laboratory while receiving a whole body radiation exposure of less than 350 mR. Sufficient shielding and precedural precautions including the use of portable instruments to assess high radiation are in place to assure that radiation dose criteria are met. 1667 210 2.1.8.a - 5

2.3 POST-ACCIDENT ANALYSIS A design review of the existing analytical equipment and procedures has been made, and facilities at Point Beach Nuclear Plant are adequate for analyzing post-accident samples. Shielding and handling equipment for the new sample bomb has been fabricated. Access time in the Chemistry Laboratory is limited by high background radiation, but sufficient time is available to perform all required analyses without exceeding the radiation dose criteria. Procedures for analyzing post-accident samples have been prepared and approved. Background radiation levels in the counting room will be controlled by additional shielding if background levels interfere with accurate measurement. An agreement between the Point Beach Nuclear Power Plant and the Kewaunee Nuclear Power Plant has been implemented to provide backup analytical facilities for both plants in the event the normal plant laboratory facilities become unuseable.

                    /

2.4 TRAINING The new sampling equipment at Point Beach Nuclear Plant will be checked out and initial training in the post-accident sampling and analysis procedures will be completed by January 11, 1980. Further training and exercises in the use of the new sample stations and sample bomb may further improve performance of post-accident sampling functions. 1667 211 2.1.8.a - 6

3.0 DESIGN REVIEW OF REACTOR COOLANT SAMPLING AND LABORATORY ANALYSIS 3.1 DESIGN REVIEW OF EXISTING REACTOR COOLANT SAMPLING 3.1.1 FUNCTIONAL DESCRIPTION OF EXISTING REACTOR COOLANT SLMPLING SYSTEM The existing reactor coolant sampling system provides for sampling reactor coolant from three locations: the pressurizer steam space, the pressurizer liquid space, and the hot leg. In addition, the reactor coolant sampling system can be used to obtain samples from the residual heat removal loop and from four locations in the chemical and volume control system. All samples are normally obtained in the sample room. The reactor coolant sampling system containment penetrations are located in pipeway #1 for Unit 1 and in pipeway 44 for Unit 2. The sample lines exit the pipeways and are routed through the Auxiliary Building along column line L. The sample lines penetrate the floor of the sample room at elevation 26' and enter the sample heat exchangers located on the back wall of the sample room. Cooling water to the heat exchangers is supplied from the component cooling water system. From the heat exchangers, the sample lines are routed to the sample panel which houses the valves that allow the technician to select the desired sample flow path. Samples are normally collected in the sample sink adjacent to the sample panel. The sample sink has a hood exhaust fan which draws air through the sampling area and discharges it to the auxiliary building exhaust fan suction so that any potential airborne contaminants are filtered and monitored prior to release. 1667 212 2.1.8.a - 7

Pressurized samples are collected in stainless steel collection vessels (bombs) which are equipped with isolation valves and Swagelok connections. Each sample point is located in the sample sink and is equipped with a bypass line so that the sample line can be flushed prior to drawing a sample, without the sample bomb being in place. After the sample line is flushed to ensure a representative sample and the sample bomb is in place, flow is directed through the sample bomb. After the bomb is flushed and filled, it is isolated, disconnected, and transported to the chemistry laboratory for analysis. Depressurized samples may also be collected in sample bottles in the sample sink. The sample obtained in the sample bomb contains approximately 85 ml of pressurized reactor coolant which is sufficient to perform all of the normal analyses. Degassing, dilution and preparation of the pressurized sample take place in the chemistry laboratory. The sample lines discharge into the sample sink which drains to the equipment drain header and then to the liquid radwaste system. Flushing flow and excess sample can also be directed to the Volume Control Tank. 3.1.2 ACCESSIBILITY OF THE S AMPLE RCCM IN POST-ACCIDENT ENVIRONMENT The reactor coolant sample flow is isolated at the containment isolation valves on a containment isolation signal, which is concurrent with the safety injection actuation signal; thus the sample rcom remains accessible immediately after the start of the accident sequence. However, if the containment isolation signal is reset so that a sample may be taken and sample flow is started through the sample rcom, the dose rate in the sample room would be about 25 R/hr using the NRC source term levels of activity and i667 213 2.1.8.a - 8

assuming that the source term has been diluted by one reactor coolant volume and that sample flow is directed to the Volume Control Tank. Activity levels in the Unit 1 sample room would be higher because the tubing runs in that sample room are longer. If recirculation is assumed to occur, the lowest dose rate in the sample room would be about 3.5 R/hr from the sample lines only; this reduction is due to the additional dilution from the RWST. Additional exposure would result f rom safety injection lines containing source term material from the containment sample. The exposure in the vicinity of the sample panel and sample hood would be much greater. Thus, the sample room cannot be used to obtain a source term sample within one hour, and the existing sampling facilities will have to be modified to allow for collection of the required samples. If a sample were collected in the sample room, however, the existing drainage (to volume Control Tank) and ventilation (out the auxiliary building exhaust stack through HEPA and charcoal filters) are adequate for handling pos t-accident source terms. 3.2 DESIGN REVIEW OF REACTOR CCOLANT LABORATORY ANALYSIS 3.2.1 FUNCTIONAL DESCRIPTICN OF REACTOR CCOLANT LABORATORY ANALYSIS All reactor coolant analyses are performed in the Chemistry Laboratory and in the adjacent Counting Room. The laboratory is equipped with three sample hoods which exhaust to the auxiliary building exhaust f an suction; any airborne contaminants are filtered and monitored prior to release. All drains in the laboratory are routed to the equipment drain header and then to the liquid radwaste sys tem. In addition to regular laboratory 1667 214 2.1.8.a - 9

supplies and equipment, the laboratory also has supplies of shielding material for shielding sources and sensitive ins t r uments . The following analyses are routinely performed in the Chemistry Laboratory. 3.2.1.1 REACTOR COOLANT ISOTOPIC ANALYSIS A 5 mi sample of reactor coolant is diluted to 50 ml and 1 ml of the dilute sample is pipetted into a test tube. The sample in the test tube is then counted using either a Canberra Model 8100 Multichannel Analyzer with a Canberra GeLi Detector or a Packard Model 970 Multichannel Analyzer with an Ortec GeLi Detector. The test tube is placed in contact with the GeLi Detector. 3.2.1.2 REACTOR COOLANT pH A sample of reactor coolant (approximately 50 ml) is placed in a beaker and a pH probe is introduced to the sample. Several models of pH meters are available to perform this analysis. 3.2.1.3 REACTOR CCOLANT GASEOUS ACTIVITY A pressurized sample of reactor coolant is obtained in an 85 mi sample bomb. The pressurized sample bomb is arranged vertically with an evacuated gas sample collection vessel above it. The gas sample collection vessel has a capacity of approximately 318 ml and is equipped with a septum closure for extracting a sample. The sample bomb and the evacuated gas sample collection vessel are connected by tygon tubing with an isolation valve The isolation valve is opened slowly and the pressure in tne sample bomb is relieved to the gas sample collection vessel. When the pressure has equalized, air is admitted to the bottom of the sample bomb to scrub the remaining dissolved gasses from the 1667 215 2.1.8.a - 10

coolant. Any liquid which has entered into the gas sample collection vessel is drained back into the sample bomb. A small (1/2 cc) sample of gas is obtained with a hypodermic syringe by inserting the needle into the septum closure. The sample is injected into a 5 cc vial which is counted on one of the multichannel analyzers. The vial is placed in contact with the GeLi detector. Counting efficiencies have been determined for these geometries. 3.2.1.4 REACTOR COOLANT DISSOLVED HYDROGEN CONCENTRATION The gas that remains in the gas sample collection vessel frcn the preparation of the gaseous activity determination is carried across the laboratory in the gas sample collection vessel and the vessel is attached to a Fisher-Ham!.lton gas partitioner chromatograph Model 29. The entire contents of the vessel are analyzed and the results are recorded on a strip recorder. 3.2.1.5 REACTOR COOLANT BORON CONCENTRATION A sample of reactor coolant (2-100 ml depending on estimated boron concentration) is placed in a beaker. A small amount (approximately 1 gm) of mannitol is added tc the coolant. A pH probe is placed in the beaker and the solution is titrated with a NaOH standard to a pH of 8.5. 3.2.1.6 REACTOR COOLANT CHLORIDE ION CONCENTRATION A relatively large sample (50 ml or 100 ml) is added to a flask containing a f ew drops of Bromphenol blue-mixed indicator. pH is adjusted to approximately 3.1 with nitric acid. The solution is titrated with mercurous nitrate to the first permanent violet tinge. A chloride electrode is also available to perform this analysis. 1667 216 2.1.8.a - 11

3.2.1.7 REACTOR COOLANT DISSOLVED OXYGEN CONCENTRATION Dissolved oxygen concentration during normal operations is determined using the standard indigo carmine test which is used for detecting very low concentrations of dissolved oxygen. For higher dissolved oxygen concentrations, the Winkler test is used. 3.2.2 POST-ACCIDENT ACCESSIBILITY OF THE CHEMISTRY LABORATORY AND COUNTING ROOM Prior to recirculation, the Chemistry Laboratory and Counting Room remain available for unlimited acceca with very low levels of radiation. The principal contribution of this radiation is from the containment. During this time, the only restrictive radiological hazards presented by analysis would be in the handling and disposal of source term material. The equipment in the Chemistry Laboratory is adequate for handling and disposing of source term material. After recircult.cion, the radiation levels will increase significantly (to approximately 9 R/hr) in the Chemistry Laboratory due to the contribution from safety injection suction lines which are supplied from the containment sump by the residual heat removal pumps. The radiation levels in the Counting Room also increase (to approximately 6 R/hr in the center of the rocm). Hcwever, the analytical equipment to be used f or pos t-accident sampling is located immediately adjacent to the west wall of the Counting Room. This two foot thick concrete wall provides sufficient shielding to reduce the

 . background in the vicinity of the analytical equipment to approximately 6 mR/hr. With this background, the analytical equipment will be able to analyze the isetcpic content of the coolant and gaseous activity samples.

I667 217 2.1.8.a - 12

3.2.3 SPECIFIC ACTIVITY OF POST-ACCIDENT SAMPLES The specific activity of the post-accident reactor coolant sample has been evaluated using two dilutions. In the first case it has been assumed that the event initiating fuel damage was not a loss of coolant accident of sufficient severity to require recirculation from the containment sump within 1 hour. In this case, the dilution was assumed to be one reactor coolant volume (approximately 45,200 gallons) . The resulting specific activity is 2.12 Ci/ml assuming that 100% of the noble gasses, 50% of the halogens, and 1% of the remaining fission products enter the coolant. In the second case, it has been assumed that the initiating event was a loss of coolant accident of suf ficient severity to require recirculation. The additional contribution to dilution from the Refueling Water Storage Tank and the Safety Injection Accumulators recults in a total dilution of 347,000 gallons. If recirculation is assumed to have occured, the specific activity of the coolant is reduced to 0.276 Ci/ml. Thus, in the case when recirculation occurs, the exposure that would result from handling and analyzing the sample is reduced, while the exposure due to background contributions is increased. 3.3 INTERIM METHODS FOR SAMPLING REACTCR COOLANT Because the sampling room will not be accessible for long term habitabiliry if reictor coolant containing source term _ contaminants is allowed tc flow through the sample lines, it is essential that the sample station be relocated. A location on the outsid? wall of the sample room near the entrance has been selected. This location is easily accessible frcm the radiation control area entry point for both units. The outside wall of the i667 218 2.1.8.a - 13

sample rocm is poured concrete and concrete block filled with concrete and is 18" thick so that it provides shielding frem sample lines inside the sample room. As presently configured, most of the sample lines are in the back of the sample room, as are the sample heat exchangers. Thus the primary sources of radiation are also more distant from the new sampling station. New pipe routing for the accident sample location has been kept as short as possible and has been routed along the south wall to minimize its contribution at the new sample station. Sampling will be accomplished with a shielded sample bomb constructed of stainless steel tubing and suitable valving to allow the collection of a small (approximately 15 ml) sample. The sample size was selected to be as small as possible to reduce exposure and facilitate handling of the bcmb while yet being large enough to accomplish all of the required reactor coolant analyses with a single trip to the sampling station. To accommodate the new sampling station, a sample line which contains reactor coolant hot leg sample has been run from the discharge of the sample heat exchanger along the south wall of the sample room to the east wall. The sample line penetrates the east wall near the entrance to the sample room. A return line is routed f rom the east wall along the south wall and connects with the sample return line to the Volume Control Tank. The sample supply line and return line penetrations have been located so that the sample inlet is directly above the sample return. The sample lines are equipped with Swagelok fittings to allow rapid connection and disconnection and to minimize leakage. Valves have been installed to allow recirculation with the sample bomb installed and ensure that sample flow is forced through the bomb when the sample is to be collected. Demineralized water for flushing has also been installed and the valving arrangement allcws complete flushing of the sample supply and sample return 1667 219 2.1.8.a - 14

lines. Flushing the sample lines will reduce radiation exposure and the potential for contamination in removing the sample bomb and in reinstalling it if subsequent samples are required. The sample bomb is constructed of stainless steel tubing. The sample bomb is isolated on both sides by a double valve arrangement which ensures that no leakage of the sample will occur during removal and tr,insport of the bomb. The bomb will be connected to the sample lines with Swagelok fittings. The entire sample bomb including valves and fittings is shielded with approximately 1-3/4" of lead. To ensure that the fittings are properly and quickly aligned, a mounting bracket which acts as a keyway and which holds the sample bomb in place at the wall has been installed. This mounting bracket allows the bomb to remain unattended while it is being flushed and filled, further reducing e xpos ure in drawing the sample. The sample bomb valving arrangement allows operation of all valves at the post-accident sampling station to be operated by reach rods so that all sampling manipulations can be accomplished at a distance from the bomb, greatly reducing the exposure to the extremities and further reducing the whole body exposure. A correlation between contact reading on the sample bomb and radioactive material content is provided in Figures 1, 2, and 3. The shielded sample bomb will be transported to and from the sample location on a standard industrial four wheel cart. The bomb will be placed on the end of the cart to maximize the distance between the sample and the technician transporting it. The sample bomb will be delivered to the Chemistry Laboratory. The radiation exposure to the whole body of an individual sampling reactor coolant and transporting it in the post-accident environment will be approximately 650 mR for Unit 1 and 1667 220 2.1=8.a - 15

approximately 900 ma for Unit 2. This estimate of exposure is based on sampling one hour after the initiating event has occurred and assumes that recirculation has occurred and that the sample can be obtained in 5 minutes and transported at 4 feet per second. The higher exposure for Unit 2 results primarily f rom the longer transit times and the higher exposure levels in the transit path. Two people will be required for post-accident sampling and transport. The radiation exposure to the hands (extremities) will be less restrictive than the exposure to the whole body because of the shielded bomb. The dose rate on contact with the bomb is approximately 160 R/hr if it is assumed that the bomb contains source term contaminants diluted only by one reactor coolant volume. The bomb will only be touched when it is removed from the sample rack and when it is placed in the sample hood in the Chemistry Laboratory. It is estimated these manipulations will take less than two minutes. All valve manipulations will be performed by reach rod to minimize exposure. These estimated exposures are based on conservative assumptions and can be further reduced by ensuring that the personnel obtaining the sample are familiar with post-accident sampling procedures and techniques. 3.4 INTERIM METHODS FOR REACTOR COOLANT LABORATORY ANALYSIS 3.4.1 FUNCTIONAL DESCRIPTION OF PCST-ACCIDENT REACTOR CCOLANT ANALYSIS METHCDS The reactor coolant sample which containn source term contaminants can be analyzed using existing laboratory methods and analytical equipment if additional shielding and handling apparatus are made available in the vicinity of the Chemistry Labo r ator y. The shielding and apparatus will be installed in the sample hood prior to obtaining a post-accident sample so that it 1667 221 2.1.8.a - 16

_ . . . . . .. _ _. _ _ _ _ ._ ._. _. .a will be in place before the sample is delivered to the Chemistry Laboratory. The total liquid volume used for these analyses is 15 ml. These analyses can be performed within one hour. The following analyses will be performed. 3.4.1.1 REACTOR COOLANT DISSOLVED GASSES The shielded sample bomb will be placed on the ramped sample bomb holder and attached to an evacuated gas sample collection cessel and to a regulated argon gas supply. The isolation valves between the pressurized sample bomb and the evacuated gas sample collection vessel will be slowly opened to relieve the pressure in the sample bomb. After the preasure has equalized, the regulated supply of argon gas will be admitted to the bottom of the sample bomb to scrub remaining dissolved gasses from the coolant. The gas sample collection vessel and the argon gas supply will be isolated and a 1/2 cc sample of gas will be withdrawn, using a shielded hypodermic syringe. This sample will be injected into the Fisher-Hamilton gas partitioner chromatograph, which will analyze the gas and record oxygen and hydrogen concentrations. The minimum sensitivity of the gas partieioner chromatograph is 0.05% by volume. This corresponds to a minimum level of detection of approximately 2.5 ppm for oxygen and 150 ppb for hydrogen. Hydrogen concentration is monitored due to its potential for creating explosive mixtures. The lower explosive limit of hydrogen in air is 4 percent. Oxygen concentration in this low range will make virtually no contribution to supporting ccmbustion, but dissolved oxygen concentration in the range of parts per million will contribute - to an environment conducive to stress corrosion cracking of stainless steel. Consequently, the range of detection using this method is suitable to providing meaningful post-accident monitoring. 1667 222 2.1.8.a - 17

3.4.1.2 REACTOR COOLANT RADIOACTIVE GASSES A small (1/2 cc) sample of the gas extracted for dissolved gas concentration analysis will be extractnd using a shielded (lead foil wrapped) hypodermic syringe and will be injected into a 4 liter sealed flask which has a septum closure and contains air at atmospheric pressure. After allowing the gas sample to diffuse in the dilution volume, a 1/2 cc sample will be withdrawn from the 4 liter fiask using a hypodermic syringe. The contents of this sample will be injected into a 5 mi serum bottle or a 1 liter polyethylene bottle if further dilution is necessary. The contents of the serum bottle will be analyzed in a multichannel analyzer usirj a geomtery for which counting efficiencies have been determined. The specific activity of the dilute sample in the 5 mi serum bottle (approximately 6.6 uCi/cc if all noble gases and one half of the halogens are assumed to be released by the degassing process) is suitable for accurate analysis in existing equipment. 3.4.1.3 REACTOR COOLANT pH The reactor coolant liquid remaining in the sample bomb will be drained into a collection beaker, and 5 ml of the sample will be pumped into another beaker using a peristaltic pump. The beaker will be on a magnetic stirrer. A pH microprobe, which has been previously standardized, will be inserted into the beaker and will give a direct pH reading.

                                                                                         ~

3.4.1.4 REACTOR COOLANT BORON CONCENTRATION The same 5 mi sample used for pH analysis will be used for baron analysis, a stir bar and approximately 1 gm of mannitol will be added to the 5 mi sample using grip tongs. The sample will be 1667 223 2.1. 8. a - 18

stirred with the magnetic stirrer. The solution will be titrated using a piston burette containing NaOH to the pH endpoint of 8.5. Thi. method has the same accuracy as boron analysis performed during operation. 3.4.1.5 REACTOR COOLANT CHLORIDE ICN CONCENTRATION The remaining 10 ml of reactor coolant will be pumped by tre peristaltic purup into another beaker containing the chloride ion probe. The millivolt reading will be converted to chloride ion concentration. The lower limit of detection for this method is approximately 20 parts per billion and is well below any chloride ion concentration which would be reasonably considered harmful in a post-accident environment. 3.4.1.6 REACTOR COOLANT ISCTOPIC ANALYSIS From the beaker in which the chloride ion analysis was performed, a 0.5 mi sample will be obtained using a capillary tube and hypodermic syringe. This will be injected into a 1000 ml beaker containing demineralized water. The 1000 mi beaker will be stirred using a magnetic stirrer. A second dilution in a 1000 ml beaker will be performed if necessary. A one mi sample will be withdrawn and injected into a 5 mi serum bottle which will be counted on the multichannel analyzer. Since gaseous activity has been removed from the coolant, the specific activity of this doubly dilute solution is approximately 0.07 uCi/ml and is in the range of accurate analysis of the multichannel analyzer system. 3.4.2 ALTERNATIVES FOR REACTOR COOLANT LABORATORY ANALYSES While the capability exists to perform all analyses during one hour, it is not anticipated that all of the analyses will have to be performed. The chloride ion concentration determination will not be made unless there is reason to believe that large scale i667 224 2.1.8.a - 19

contamination of the water injected into the reactor coolant has occurred. This is considered unlikely because Point Beach Nuclear Plant is located on Lake Michigan and is remote from water containing high levels of chloride ions such as might be found in a seawater-cooled plant. An agreement has been implemented between the Kewaunee Nuclear Plant and Point Beach Nuclear Plant to provide backup analytical laboratory service for both facilities. The Kewaunee Nuclear Plant is located within 5 miles of the Point Beach Nuclear Plant. This proximity provides additional assurance that required analyses can be performed within one hour. 1667 225 s 2.1.8.a - 20

4.0 DESIGN REVIEW OF CONTAINMENT ATMOSPHERE S AMPLE SYSTEM AND LABORATORY ANALYSIS 4.1 DESIGN REVIEW OF EXISTING CONTAINMENT ATMOSPHERE S AMPLE SYSTEM 4.1.1 FUNCTIONAL DESCRIPTION OF EXISTING CONTAINMENT ATMOSPHERE SAMPLE SYSTEM The containment atmosphere is currently monitored continuously by the remote containment atmosphere sampling system and is grab sampled prior to any containment purge. The remote containment atmosphere sampling system supply lita exits the containment at the bottom of the personnel access hatch at the 66' level and descends to the sampling cubicle at the 52' level in the facade area adjacent to the containment. The line penetrates the rear of the cubicle. Immediately inside the - cubicle is the containment isolation valve. After the containment isolation valve, the sample line runs into a manifold which allows selection of the sampling point by the operation of solenoid valves. The unit sample points available are containment atmosphere, facade atmosphere (for background determination or purging), and the containment purge stack downstream of the purge exhaust filters. The particulate monitor, R-ll, counts activity deposited on a moving filter paper. The gas monitor, R-12, is aligned in series with R-ll and samples gaseous activity in a volume chamber. A constant flow (10 cfm) pump provides the ficw required for sampling and can - discharge back into the containment or to the containment purge exhaust stack. 1667 226 2.1.8.a - 21

The containment isolation valves for the containment atmosphere sampling system close on containment isolation signal, which is concurrent with the safety injection signal. Containment atmosphere air particulate , iodine, and gaseous activities are determined prior to containment purges by sampling the containment air inside the containment using a portable sample pump to draw the sample over a filter and a charcoal canister. The filter and charcoal canister are counted and the particulate, iodine, and total gaseous activities are determined. 4.1.2 EXISTING SYSTEMS USE IN POST-ACCIDENT ENVIRONMENT The installed containment atmosphere sampling system will not provide usef ul inf ormation f or post-accident monitoring. It is a sensitive instrument designed to detect very low levels of activity, and the detectors would be saturat i by the activity and background levels in the post-accident environment if the source term is calculated from a release of 100% of the noble gasses and 25% of the halogens. Manual sampling using filters and cartridges would result in high exposure to individuals. There would be no access to containment, and no convenient connection for portable air sampler use. Consequently, the existing systems will not provide meaningful pos t-accident information assuming the NRC source term, and modifications will be provided. - 1667 227 2.1.8.a - 22

4.2 DESIGN RE'J EW OF EXISTING CONTAINMENT ATMOSPHERE LABORATORY ANALYSIS For normal operations, the only instruments used for analysis of the containment atmosphere are the installed monitor and the Health Physics Station counting equipment which is used for counting filters and cartridges. However , for atmosphere samples that are suitably prepared, the analytical equipment which is now used for the gaseous analysis of reactor coolant can be used to determine the isotopic and chemical composition of the containment atmosphere. 4.3 INTERIM METHODS FOR CONTAINMENT ATMOSPEERE S AMPLING The existing containment atmosphere sampling system has been mcdified so that a containment air sample can be obtained. A sampling line downstream of the containment isolation valve and suitable valving has been installed so that sample flow can be diverted around R-ll and R-12 and drawn through the existing sample pump. The discharge of the sample pump is aligned to return the sample flow to containment. The bypass line has been equipped with a septum closure nipple so that a small (approximately 1 cc) sample of contaiment atmosphere can be obtained with a shielded (wrapped in lead foil) hypodermic syringe. The nipple will be located so that the sample can be drawn without entering the existing R-11/R-12 cubicle. Access to and from the sample point will be from the turbine building to reduce exposure. The hypodermic syringe will be transported in a shielded carrier to ensure that exposure to the extremities of - the technician obtaining the sample is minimized. The sample will be delivered to the Chemistry Laboratory for analysis. The estimated exposure for obtaining the sample and delivering it to the laboratory is about 350 mR. 1667 228 2.1.8.a - 23

4.4 INTERIM CONTAINMENT ATMCSPHERE LABORATORY ANALYSIS The contaiment atmosphere sample containing source term contaminants can be analyzed using existing laboratory methods and analytical equipment. The only additional precautions that must be taken are to ensure that the sample is shielded during analysis. When the 1 cc containment atmosphere sample is delivered to the Chemistry Laboratory, 1/2 cc will be injected into the Fisher Hamilton gas partitioner chromatograph which will analyze the gas and record oxygen and hydrogen concentrations. The remaining 1/2 cc will be injected into a 4 liter sealed flask which has a septum closure and contains air at atmospheric pressure. After allowing the gas sample to diffuse in the dilution volume, a 1/2 cc sample will be withdrawn from the 4 liter flask using a hypodermic syringe. The contents of this

    > sample will be inj ected into a 5 mi serum vial or a 1 liter polyethylene bottle if further dilution is necessary.                  The contents of the serum vial will be analyzed using a geometry for which counting efficiencies have been determined. The specific activity of the dilute sample in the 4 liter sealed flask is approximately 1.16 uCi/cc and is suitable f or accurate analysis in existing equipment.

These analyses can be completed with existing equipment within one hour of the delivery of the sample syringe to the Chemistry _ Laboratory. An agreement has been implemented between Kewaunee Nuclear Plant and Point Beach Nuclear Plant to provide backup analytical laboratory services for both facilities. The Kewaunee Nuclear Plant is located within 5 miles of the Point Beach Nuclear Plant. This proximity provides additional assurance that the required analyses can be performed within one hour. 2.1.8.a - 24 1667 229

    .0              ALTERNATIVES Point Beach Nuclear Plant posseses the capability to obtain and analyze the required samples which contain source term contaminants using the interim methods. However, the feasibility of and need for further improvements will be examined within the next year.             Some of these alternatives c e described below.

5.1 ALTERNATIVES FOR REACTOR COOLANT SAMPLING Alternative I Refinements of the Interim Post-Accident Sampling Method Specific refinements to be considered include improved design of the sample bomb, a mechanized and/or shielded transport system, relocation of the post-accident sample station to an area of lower exposure rate, and installation of permanent shielding. One refinement to allow improved flushing of the system is shown in Figure 4. Alternative II Installation of a Remote Dilution and Sampling Process System A sample could be drawn, diluted and prepared in a remotely operated and shielded loc ation. A conceptual flow diagram is shown in Figure 5. Alternative III Installation of In Line Analytical Equipment - Scme of the required analyses can be performed in line. The availability of analytical equipment suitable for use in high radiation areas may preclude some analyses such as chloride ion concentration. Equipment which has been evaluated includes a 1667 230 2.1.8.a - 25

boronometer, an automatic chloride analyzer, a fluoroborate ion electrode system to determine boron concentration, automatic degassing and automatic gas sampling. 5.2 ALTERNATIVES FOR CONTAINMENT ATMOSPRERE SAMPLING With the interim sampling nipple installed, the contaiment atmosphere sample can be readily obtained. Refinements to the interim sampling system will be considered including remote operation of the isolation valves, remote monitoring of the radiation levels of the sample line, and shielding to reduce exposure to the technician obtaining the sample. These refinements are shown in Figure 6. 5.3 ALTERNATIVES FOR LABORATORY ANALYSIS 5.3.1 ALTE RNATIVES FOR CHEMICAL ANALYSIS The highest exposure during analysis is when performing the chemical analysis of the reactor coolant sample. Most of the exposure is due to the fact that if recirculation is assumed to have occured, the dose rate in the laboratory area near the sample hood is approximately 9 R/hr. This limits the amount of time the analyst has for performing the required analyses. The analyses can be performed within the guidelines of GCC-19, and familiarity with the post-accident procedures can further reduce e xpos ur e . However, because of the reciprocal agreement between Point Beach Nuclear Plant and Kewaunee Nuclear Plant, it is _ unlikely that further improvements in laboratory facilities will be required. i667 231 2.1.8.a - 26

                                                                   /

5.3.2 ALTERNATIVES FOR ISOTCPIC ANALYSIS The high levels of background in the Counting R.:om make isotopic analysis difficult. The existing equipment can be used in the existing location if the background is subtracted f rom the sample count. To reduce uncertainty, the use of other analytical equipment is desirable. However, because of the reciprocal agreement between Point Beach Nuclear Plant and the Kewaunee Nuclear Plant, it is unlikely that further improvements in laboratory facilities will be needed. 1667 232 2.1.8.a - 27

6.0 REFERENCES

NRC Documents H. R. Denton (NRC) to All Nuclear Power Plants letter, dated October 30, 1979, " Discussion of Lessons Learned Short Term Requirements" Of fice of Nuclear Reactor Regulation, USNRC, NUREG-0 578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations" WEPCo Documents S. Burstein (WEPCo) to H. R. Denton (NRC) letter dated October 20, 1979, " Dockets 50-266 and 50-301 Implementation of NUREG-0578, Point Beach Nuclear Plant Units 1 and 2" S. Burstein (WEPCo) to H. R. Denton (NRC) letter dated November 27, 1979, " Dockets 50-266 and 50-301 Implementation of NUREG-0578, Point Beach Nuclear Plant U..its 1 and 2 1667 233 2.1.8.a - 28

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2 2.1.8 INSTRUMENTATION TO FOLLOW THE COURSE OF AN ACCIDENT 2.1.8.b INCREASED RANGE OF RADIATION MONITORS TABLE OF CONTENTS 1.0 Scope of Requirements of NUREG-0579, Section 2.1.8(b): Increased Range of Radiation Monitors 2.0 Summary of Methods Used to Meet Section 2.1.8(b) Requirements for Monitoring High-Level Gaseous Effluents 2.1 Specific Requirements for Increasing the Range of Radiation Moni: ors Which Must be Provided by January 1, 1980 2.2 Specific Requirements for Increasing the Range of Radiation Monitors Which Must be Provided by January 1, 1981 3.0 Description and Design Review of Existing Plant Monitoring Equipment For Vents Producing Gaseous Effluents 3.1 Introduction 3.2 Auxiliary Building vent Stack 3.3 Drumming Area Ventilation Stack 3.4 Containment Purge Exhaust Stacks 3.5 Combined Air Ejector Decay Duct Exhaust 3.6 Gas Stripper Exhaust 1667 241 2.1.8.b - 1

e TABLE OF CONTENTS (Continued) 4.0 Gaseous Effluent Monitoring: Design Modifications 4.1 Introduction 4.2 Auxiliary Building Vent Stack 4.3 Drumming Area Vent Stack 4.4 Containment Purge Vent Stacks 4.5 Combined Air Ejector Decay Duct Effluent Vent 4.6 Gas Stripper Effluent Vent 5.0 References 1667 242 2.1.8.6 - 2

1 TABLES

1. High Range Noble Gaseous Effluent Radiation Monitoring Equipment 1667 243 _

2.1.8.b - 3

FIGURES

1. Radiciodine Plant Vent Particulate Sampling Detail

2. Auxiliary Building Vent Radiation Detector Installation Detail

3. Auxiliary Building Vent Radiation Detector Shield Cave

4. Auxiliary Building Stack mrad /hr to Ci/sec Conversion Chart

5. Drumming Area Vent Radiation Detector Installation

6. Lead Brick Shielding for Radiation Detector Gaseous Effluent Vents

7. Drumming Area Rad /hr to Ci/sec Conversion Chart

8. Containment Purge Exhaust Stack Radiation Detector Installation

9. Containment Purge Exhaust (1 Fan) Rad /hr to Ci/sec Conversion Chart

10. Containment Purge Exhaust (2 Fans) Rad /hr to Ci/sec Conversion Chart

11. Combined Air Ejector Decay Duct Rad /hr to Ci/sec Conversion Chart

12. Gas Stripper Stack Rad /hr to Ci/sec Conversion Chart 1667 244 2.1.8.b - 4

                                   -.    - - - - . . . - - .  . . - - - - - - - - - - - .~ . . --

. . ~ _ _ - ... r ATTACHMENTS I. Procedure Guidelines for Use in Conducting Noble Gas Effluent Measurement II. Procedure Guidelines for Use in Conducting Measurement of Radiciodine and Particulate from Gaseous Effluents 1667 245 2.1.8.b - 5

d 1.0 SCOPE OF RECUIREMENTS OF NUREG-0578, SECTION 2.1.8(b) INCREASED RANGE OF RADIATION MONITORS Radioactive gaseous effluents are generally low level releases to the surrounding plant atmosphere which are detectable by existing low range monitors at Point Beach Nuclear Plant. Tha sources of these effluents, typically, are: the auxiliary cuilding, waste processing area, the containment, gas stripper, and combined air ejector decay duct. NUREG-0578 is requiring an evaluation of the ranges of existing low-level detection systems because they may not be capable of detecting the high-level releases from vents which exhaust air from areas with systems containing post-accident liquids and gases, assuming source terms postulated by the NRC. The requirements of NUREG-0578, Section 2.1.8.b are:

a. Determine release rates which are possible during accident conditions using source terms in NUREG-0578;

b. Evaluate existing gaseous monitoring systems to determine if their range is adequate to monitor postulated releases (noble gases and iodine); and

c. If the existing gaseous ef fluent systems are inadequate for monitoring postulated accident releases, it is required that methods and procedures be deve:.oped to allow interim monitoring of high level gaseous releases by January 1, 1980. A permanently installed method for monitoring high-level releases must be developed and implemented by January 1, 1981.

More details on the specific requirements for NUREG-0578, Section 2.1.8.6 are contained in Section 2.1 and 2.2. 1667 246 2.1.8.6 - 6

2.O

SUMMARY

OF METHODS USED TO MEET SECTION 2.1.8.b REQUIREMENTS FOR MONITORING HIGH-LEVEL GASEOUS EFFLUENTS Each of the gaseous effluent paths to the atmosphere from Point Beach Nuclear Plant were reviewed to determine the release rates which Nould exist using source terms in NUREG-0578. The existing Point Beach Nuclear Plant gaseous effluent instrumentation was evaluated to determine its adequacy to monitor the postulated release rates, including gaseous radiciodine and particulate releases. The result of this evaluation, which included studies into the radiation level possible for background levels, indicates that mod fications must be made to the existing gaseous monitoring systems. These modifications are:

a. Installation of high-range detectors on all gaseous effluent vent ducts to allow monitoring of expected releases;

b. Installation of shielding around each detector to mimimize the effects of background levels expected during an accident on the reading of the detector;

c. Installation of remote sampling lines on each vent, along with adequate shielding, to allow an operator to obtain vent gases and remotely analyze them for radiciodine and particulate content; and

d. Modifications to existing procedures to instruct personnel in the case that use of the high-range detectors and sampling lines are required.

Although the installation of new equipment is not required by the NRC for interim compliance, it has been concluded that new equipment is required, based on personnel exposure and availability considerations. 1667 247 2.1.8.b - 7

The above modifications are being completed as expeditiously as possible depending on equipment deliveries to the plant site. The high-range radiation monitoring equipment, listed in Table 1, has been purchased frca Eberline and will be installed within 30 days after receipt of the materials. In addition to these steps, plans have been made to establish a design and procure equipment necessary to meet the January 1, 1981 requirements for permanent modifications. 2.1 SPECIFIC REQUIREMENTS FOR INCREASING THE RANGE OF RADIATION MONITORS WHICH MUST BE PROVIDED BY JANUARY 1, 1980 There are two basic requirements in NUREG-0578 Section 2.1.8.b. These are: increasing the range of existing radiation monitoring equipment, if necessary, to monitor noble gas effluents and adding to the capability to sample radiciodine and particulates from the high-level gases. The requirements will be satisfied by the Point Beach Nuclear Plant,and modifications which are being made to the existing vents are described in Section 4.0. Drawings are also contained in that section which depict the installation of equipment for the vent stacks. 2.1.1 REQUIREMENTS FOR RADIOLOGICAL NOBLE GAS EFFLUENT MONITORS 2.1.1.1 GENERAL Until final implementation in January 1, 1981, all operating reactors must provide, by January 1, 1980, an interim method for quantifying high-level releases which meets the requirements of Table 1. Methods are to be developed to quantify release rates of up to 10,000 u Ci/sec. for noble gases from all potential release points, e.g., auxiliary building, radwaste building, fuel 1667 248 2.1.8.b - 8

       . L

• handling building, reactor building, waste gas decay tank .

releases, main condenser air ejector, and other areas that communicate directly with cystems which may contain reactor coolant or containment gases. At point Beach Nuclear plant these requirements are satisfied by the. installation of fixed, high-range detectors with readouts located in the control room. 2.1.1.2 SYSTEM DESCRIPTION The. licensee shall provide the following information on his methods to quantify gaseous release radioactivity from the plant during an accident:

a. Instrumentation to be used including range sensitivity, energy dependence, calibration ,

frequency and technique

          ,                  b. Monitoring / sampling locations, including methods to assure representative measurement nf background radiation correction

c. A description of methods being employed to facilitate access to radiation readings. For January 1, 1980, control room readout is preferred; however, if impractical, in-situ readings by an individual with verbal communication with the control room is acceptable based on a criterion for obtaining radiation readings at least every fifteen minutes during an accident.

d. Source of power to be used. If normal AC power is used, an alternate backup power supply should be 1667 249 2.1.8.b - 9

                                                                                             ~

provided. If DC power is used, this source should be capable of running the readouts for seven consecutive days. At Point Beach Nuclear Plant these requirements are satisfied by the following:

a. An Eberline RMS-II radiation monitoring system is being installed on each plant vent as expeditiously as possible. The ranges of each instrument and other data are described in Table 1 and are further discussed in Sections 3.0 and 4.0. The sensors and readouts will be powered from the instrument bus which supplies control room instrumentation and will insure continuous operation should any loss of offsite power condition occur.

b. Monitoring or sample locations are described in detail in Section 4.0. Drawings showing the location of eagh monitor are also included. The monitors have been located to minimize any containment " radiation shine" and also are encased by lead shielding to minimize any background dose effect on the radiation detector.

c. All indications will be available to the operator in the control room or in an area where radiation doses are not a factor.

d. Source of power to be used will be as described in a., above.

1667 250 2.1.8.b - 10

p 2.1.1.3 PROCEDURES Procedures shall be prepared for conducting all aspects of the measurement or analysis, including:

a. Procedures for minimizing occupational exposures

b. Calculational methods for converting instrument readings to release rates based on exhaust air flow and taking into consideration radionuclide spectrum distribution as a function of time after shutdown.

The above requirements for preparing procedures are being satisfied at Point Beach Nuclear Plant by the guideline procedures for noble gas monitoring contained in Attachment I of this document. As the devices are remote-readout, procedures for minimizing occupational exposure are not required. The curves for converting instrument readings to release rates are depicted in Figures 4, 7, and 9 through 12, one for each vent stack. 2.1.2 REQUIREMENTS FOR THE MONITORING OF HIGH-LEVEL RADIOIODINE AND PARTICULATE GASEOUS EFFLUENTS 2.1.2.1 SYSTEM DESCRIPTION The licensee should provide a system / method description including the following:

a. Instrumentation to be used for analysis of the -

sampling media with discussion of methods used to correct for potentially interfering background levels of radioactivity;

b. Monitoring / sampling locations;

                                                         -                          1667 251 2.1.8.b - 11

c. Method to be used for retrieval and handling of sampling media to minimize occupational exposure;

d. Method to be used for data analysis of individual radionuclides in the presence of high levels of radioactive noble gases; and

e. If normal AC power is used for sample collection and analysis equipment, an alternate backup power supply should be provided. If DC power is used, the source should be capable of providing continuous readout for seven consecutive days.

At Point Beach Nuclear Plant, provisions for monitoring high-level radiciodines and particulates in gaseous effluents include the following:

a. The sample will be collected from a vent stack

                    ,         sampling location typically shown in Figure 1. The charcoal canister will be flushed with a purge to eliminate the noble gases such that the radiciodine and particulates are available for analysis. A gamma spectrometric analysis will be conducted to detect radiciodine. Particulates will be detected using standard techniques;

b. Each monitoring or sampling vent stack location is described in detail in Section 4.0 along with a typical iodine and particulate remote sampling line design;

c. When sampling is necessary, the operator will use a portable sample pump to withdraw a sample through the charcoal canister at the plant vent sampling 1667 252 2.1.8.6 - 12

location. A lead enclosure will be used to handle the canister for removal to the sample analysis location;

d. Data analysis methodology will be as described in a., above; and

e. Normal AC power will be used for sample collection and analysis. DC power or alternate backup power supplies are available for the portable pump.

2.1.2.2 PROCEDURES Procedures for conducting all aspects of the measurement analysis shall be provided including methods for:

a. Minimizing occupational exposure;

b. Determining release rates.

Point Beach Nuclear Plant is meeting this requirement by providing guideline procedures which are contained in Attachment II. They discuss the general methods for obtaining a sample and analyzing it for radioiodine and particulates. 2.2 SPECIFIC REQUIREMENTS FOR INCREASING THE RANGE OF RADIATION MONITORS WHICH MUST BE IMPLEMENTED BY JANUARY 1, 1981 Modifications which involve installation and redesign of the equipment are required by January 1, 1981. This section describes specific requirements which must be met by January 1, 1981, along with the description of the plans to meet those requirements. 1667 253 2.1.8.6 - 13

2.2.1 HIGH-RANGE NOBLE GAS EFFLUENT MONITORING By January 1, 1981, high-range noble gas effluent monitors are to be provideo for each release path. The noble gas effluent monitors should meet the requirements of Table 2.1.8.b.2, NUREG-0578. At Point Beach Nuclear Plant, fixed, high-range noble gas effluent radiation gas monitors are being installed on each plant vent stack as expeditiously as possible to meet interim requirements. In addition to these monitors, sampling points are being provided to collect plant radiciodine and particulates. High-range noble gas effluent monitors to meet tne requirements of Table 2.1.8.b.2, NUREG-0578, will be fully installed by January 1, 1981. 2.2.2 RADIOIODINE AND PARTICULATE MONITORING Means are required to continuously sample and provide analysis of sampling media for radiciodine and particulate levels. Point Beach Nuclear Plant will implement provisions for continuously sampling and providing analysis of the high-level particulate and radiciodines from each vent. 2.2.3 CONTAINMENT RADIATION MONITORS Two radiation monitoring systems in containment shall be provided to meet the requirements of Table 2.1.8.b.3, NUREG-0578. Point Beach Nuclear Plant will install high-range radiation monitors at a suitable point to enable monitoring of containment radiation by January 1, 1981. 1667 254 2.1.8.b - 14

3.0 DESCRIPTION

AND DESIGN EVALUATION OF EXISTING PLANT , MONITORING EQUIPMENT FOR VENTS PRODUCING GASEOUS EFFLUENTS

3.1 INTRODUCTION

$ Gaseous ef fluents at Point Beach Nuclear Plant flow from stack vents supplied by ventilation fans which take suction from air in the auxiliary building, the drumming area located in the waste management building, and both containments which exhaust to the purge stacks for Unit 1 and 2. The remainder of the gaseous effluents res?lt from the letdown gas stripper vent and the combined air ejector decay duct. All of these vents currently have low-level radiation monitors installed. The following specific information for eacb plant vent and monitor is provided in this section:

a. Description of monitoring equipment currently installed, including types and ranges;

b. Description of the existing gaseous effluent vents including size, flow rate, and the general areas served by each; and

c. Evaluation of each gaseous ef fluent monitor with respect to the requirements for monitoring the high-level gaseous effluents.

3.2 AUXILIARY BUILDING VENTILATION STACK The auxiliary building ventilation stack consists of a 54" di ame ter , circular , sheet steel ventilation duct with a flow rate of approximately 61,400 CFM. The auxiliary building 1667 255 2.1.8.b - 15

                                                     ~
   ' ventilation system consists of two intake fans located at the 8.0'   elevation which exhaust through the ventilation stack.              The ventilation stack exhaust is located above the containment facade.

3.2.1 EXISTING GASEOUS EFFLUENT MCNITORING ECUIPMENT The monitoring equipment, designated channel R-14, for the auxiliary building ventilation stack is a Tracer Lab, Model MD-12C beta-gamma G.M. tube detector with a check source. The log-rate meter display unit is located in the control rocm for . 'use by the' operator in monitoring release rates. The range of this detector is 5 x 10-8 to 3 x 10-5 Rad /hr or 5 x 10-7 to 3 x 10-4 u Ci/cc. 3.2.2 EVALUATION OF' INSTALLED SYSTEM The existing auxiliary building vent stack monitor was designed to detect relatively low-level gaseous effluent radiation levels. In order to monitor the source term por.tulated in NUREG-0578, a higher range monitor must be considered. Studies using that source term indicate that a monitor with a range of O.01 to 100 Rad /hr or 3.85 x 10-5 to .385 Ci/cc should be used. 3.3 DRUMMING AREA VENTILATION STACK The' drumming area ventilation stack consists of a 46 inch diameter, circular, sheet steel stack which starts at the 65' elevation in the solic waste processing area. The flow rate is 43,100 CFM and all air from the spent fuel pit and waste - handling equipment is exhausted by the stack. 3.3.1 EXISTING GASEOUS EFFLUENT MONITORING ECUIPME'NT The existing equipment used for monitoring levels of gaseous effluents for the drumming area consists of a detector similar to 1667 256 2.1.8.6 - 16

the g-m tube assembly used in the auxiliary building vent stack monitor. The existing R-21 monitor consists of four, thin-walled, self-quenching type GM tubes operating in parallel and located inside the vent. The range of the monitor is 4.5 x 10-8 to 8.9 x 106 Rad /hr or 5 x 10-7 to 10-4 u Ci/cc with the display readout located in the control room. 3.3.2 EVALUATION OF INSTALLED SYSTEM The maximum levels of radiation which could exist in the drumming area ventilation stack were established using the source term in NUREG-0578. The existing system is not adequate for monitoring these higher levels. Based on the maximum levels, the range of the drumming area ventilation stack monitor should be 0.01 to 100 Rad /hr or 1.1 x 10-7 Ci/cc to 1.1 x 10-3 Ci/cc. 3.4 CONTAINMENT PURGE EXHAUST STACKS The Unit 1 and 2 containment purge exhaust stacks consist of 36" diameter, circular, steel ducts with a purging flow rate of 12,500 CFM with one fan and 25,000 CFM with two fans operating. The stacks are based at elevation 66' and exhaust containment atmosphere which has been first passed through HEPA and charcoal filters. Air flow exists only when containment purge or continuous venting is in process. 3.4.1 EXISTING GASEOUS EFFLUENT MONITORING EQUIPMENT Radiation monitoring for the purge ducts consists of iso-kinetic probe / monitoring systems located at elevation 162'. These probes will obtain samples of any gas passing through the purge ducts to allow monitoring of the containment purge gas and particulate levels. A 10 CFM pump forces the gas sample to flow from the 1667 257 2.1.8.b - 17

iso-kinetic probe down to the monitor where it is analyzed for particulates. The range of the particulate detector (R-ll channel) is 7.3 x 10-8 to 7.3 x 10-5 Rad /hr or 10-6 to 10-3 u Ci/cc. The sample continues through a gas analyzer (R-12 channel) which has measuring range of 7.3 x 10-11 to 7.3 x 10-8 Rad /hr or 10-9 to 10-6 u Ci/cc. . 3.4.2 EVALUATION OF INSTALLED SYSTEM Studies which evaluate the levels resulting from post-accident containment atmosphere being exhausted up a purge vent duct indicate that the existing monitoring system is inadequate to monitor those higher levels. The studies further indicate that the monitor range should be 1-10,000 Rad /hr or 1.4 x 10-5 to. 1.4 x 10-1 Ci/cc. . 3.5 COMBINED AIR EJECTOR DECAY DUCT EXHAUST The combined air ejector decay duct exhaust consists of a 36" diameter. circular, thin sheet steel duct and has a typical flow rate of 25 CFM. The purpose of the decay duct is-to allow the gaseous effluent products produced from the Unit 1 and 2 can' denser air ejectors time to decay (approximately 3-4 hours) prior to venting them through the auxiliary building vent stack. 3.5.1 EXISTING GASEOUS EFFLUENT MONITORING EQUIPMENT The monitoring equipment (Channel CR-9) for the combined air ejector decay duct is a scintillation detector. The display unit is located in the control room. The detector will monitor a range of 1.4 x 10-24 to 1.4 x 10-10 Rad /hr or 10-15 to 10-1 u Ci/cc. 1667 258 2.1.8.6 - 18

e

   -#   ==. m                   -%.

3.5.2 EVALUATION OF INSTALLED. SYSTEM The combined air ejector decay duct exhaus't monitor was designed to detect the extremely low readings of expected releases and is

                       ~

inadequate to detect higher level releases. Calculations indicate that, based on the source term in NURIG-0578, the range of the instrument should be upgraded to 1 to 10 Rad /hr or 1.4 x 10-8 to 1.4 x 10-3 C1/cc. 3.6 GAS STRIPPER EXHAUST The gas stripper building exhaust flows into a 36" diameter, circular, thin sheet metal duct with a typical flow rate of 13,000 CFM. The stack is located in the gas stripper building and vents that area through the Unit 2 containment purge vent. 3.6.1 EXISTING GASEOUS EFFLUENT MONITORING EQUIPMENT The monitoring equipment consists of a beta-gamma G.M. tube detector located in the vent duct. The display unit is located in the control rocm for use by the operator in monitoring release rares. The monitoring equipment channel (GW-ll2) for the gas stripper exhaust stack has a range of 7.3 x 10-8 to 1.45 x 10-4 Rad /hr or 10-6 to 2 x 10-3 u Ci/cc. 3.6.2 EVALUATION.0F INSTALLED SYSTEM The existing system was designed to detect relatively low level releases. Studies using the source term of NURIG-0578 indicate that the rang.e should be upgraded.to 1 to 10 Rad /hr or 1.37 x 10-8 to 1.37 x 10-4 Ci/cc. 1667 259 2.1.8.b - 19

            ~

4.0 GASEOUS EFFLUENT MONITORING: DESIGN MODIFICATIONS

4.1 INTRODUCTION

An evaluation of the ranges of installed gaseous effluent monitoring system (s) compared to the ranges of radiation levels expected from the concentrations postulated by the NRC has resulted in the finding that the existing monitoring systems must be supplemented. Two specific types of changes are being made as required to meet the NUREG-0578, January 1, 1980 commitments, described in Section 2.1 of this document. These are:

a. Installation of high-range, shielded radiation detectors with a display unit in the control room for monitoring noble gas effluents;

b. Installation of remote sampling lines on each stack for detecting radioiodines and particulates.

Alternative methods were investigated for monitoring high-level noble gases. These alternatives consisted of:

a. Using portable monitors;

b. Permanently installing portable monitors with local readout; and

c. Installing fixed detectors with remote readout.

Access in many areas of the plant may be limited due to high radiation levels. For these and other reasons (e.g., manpower limitations, cost, flexibility, and ease of use), fixed detectors with remote readouts in the control room are being installed 1667 260 2.1.8.b - 20

within 30 days after receipt of equipment. Table 1 lists th-equipment which is being installed for noble gas ef fluent monitoring. Alternatives were also examined for monitoring radiciodine and particulates as required in NUREG-0578. Among those alternatives investigated were:

a. On-line radiciodine and particulate monitoring; and

b. Installation of sampling lines with a charcoal cartridge and particulate filter.

The alternative of using an on-line radiciodine and particulates monitor was evaluated and eliminated because no existing monitor is available which allows removal of the charcoal canister without unduly exposing the operator. The other alternative, that of installing sampling points for use with a pump, charcoal cartridge and particulate filter paper is the most feasible design. A typical vent stack installation diagram for the radiciodine and particulate sampling is shown in Figure 1. The radiciodine and particulate sampling system consists of bypass valves, a portable pump, and quick disconnects for attaching the pump and filter. The portable pump will be used to extract samples from the duct and pass it through a charcoal cartridge and filter paper assembly. The system provides flow indication and allows purging of the sample lines. Shielding is also provided which will reduce operator exposure frem the ducts. - 1667 261 2.1.8.b - 21

4.2 AUXILIARY BUILDING VENT STACK ' 4.2.1 NOBLE GAS EFFLUENT MONITOR A radiation monitor with a range of 0.01 to 100 Rad /hr. will be installed to monitor a sampling line from the auxiliary building vent stack The radiation monitor will be installed as shown in Figure 2. The monitor will be shielded from background levels by lead bricks as depicted in Figure 3. A conversion table has bee;t developed to correlate the instrument readings in Rad /hr to release rates of Ci/sec and is shown in Figure 4. 4.2.2 RADIOIODINE/ PARTICULATE MONITOR A sampling line will be installed as shown in Figure 1. Collection of the sample will be as described in the guideline procedure in Attachment II. 4.3 DRUMMING AREA VENT STACK 4.3.1 NOBLE GAS EFFLUENT MONITOR A radiation monitoring element is being installed on the stack with a range of 0.01 to 100 Rad /hr. The monitor will be installed as depicted in Figure 5. Lead bricks will also be installed as shown in Figure 6. A correlation table has also been developed to convert instrument readings in Rad /hr to - release rates of Ci/sec and is shown in Figure 7. 4.3.2 RADIOIODINE AND PARTICULATE MONITOR The radioiodine and particulate sampling lines will be installed as shown in Figure 1. I667 262 2.1.8.6 - 22

4.4 , CONT 5INMENT PURGE VENT STACK 4.4.1 NOBLE GAS EFFLUENT MONITOR A radiation detector is. being installed on the containment purge vent stacks with a range of 1-10,000 Rad /hr. The radiation detector is installed in direct contact with the containment purge vent stack as shown in Figure 8. A shielding wall will be provided as shown in Figure 6. A conversion table has been developed to convert instrument readings in Rad /hr to release rates in Ci/sec. When one fan is operated,.. Figure 9 should be used; for two fans operational, Figure 10 should be used. 4.4.2 RADIOIODINE AND PARTICULATE MONITOR The containment purge vent stacks will have remote sampling . line (s) installed to sample radioi~odines on a charcoal cartridge and filter paper, respectively. Alsamplinglineisshown typically in Figure 1. Sampling is conducted as described in the guideline procedures in Attachment II, 4.5 COMBINED AIRiEJECTOR DECAY DUCT EFFLUENT VENT

                                                           ~

4.5.1 NOBLE GAS EFFLUENT MONITOR The ccmbined air ejector decay duct requires a monitor with a range of 1-10 Rad /hr. A radiation monitor will be installed as shown typically in Figure 5. The radiation monitor will allow high ranges of decay duct air and ejector gas to be monitored and displayed in the 1667 263 2.1.8.b - 23

control room. A conversion table has been developed to convert instrument readings in Rad /hr to release rates in Ci/sec and is shown in Figure 10. 4.5.2 RADIOIODINE AND PARTICULATE MONITOR In' addition to the radiation monitoring element, radiciodine and particulate sampling provisions will be installed on the vent as typically shown in Figure 1. 4.6 GAS STRIPPER EFFLUENT BUILDING VENT 4.6.1 NOBLE GAS EFFLUENT MONITOR The range for the gas stripper radiation monitor is 1 to 10 Rad /hr. A high-range radiation monitor will be installed mounted on the gas stripper vent duct with shielding as shown in Figures 5 and 6, respectively. A graph has been developed to convert . instrument readings of Rad /hr to release rates in Ci/sec and is shown in Figure 12. 4.6.2 RADIOIODINE AND PARTICULATE MONITOR A radiciodine and particulate remote sampling line, shown in Figure 1, is being installed. 1667 264 2.1.8.b - 24

           . - . .     . - - -    ~ . . . ..         . - - -      .-. .             .       . - -

5.0 REFERENCES

1. Drawings

a. Heating, ventilation and air conditioning drawings for auxiliary building, containment and drumming area. Drawing Nos. M-ll4, 116, 118, 119, 120 to 133, M-145 to 149.

b. General plan views: M1-9.

c. Miscellaneous drawings:

E83D195-electrical one-line containment isolation signal path M-2215--P&ID: heating and ventilation M-2097--containment pos t-accident sampling sys tem G-276-P--noble gas system F-2069P--let-down gas stripper G-276-P--cryogenic noble gas removal C-342-344--construction details of various plant vents

2. FSAR Sections

a. Section 9.6.3: auxiliary building ventilation system

b. 11.2: radiation Ionitoring system 1667 265 2.1.8.b - 25

. - . . - - - - - .... M~.. - . - - - n - , . -..

3. Radiation monitoring system: Operating and Maintenance Manual, Tracer Laboratories.

4. ISOSHLD Computer Runs, EDS Version 1, 12/05/79

a. 12/13/79 Run, 20.45.28, Various Plant Stacks and 3/8" Sample Pipe

b. 12/20/79 Run, 13.09.31, 3/8" Tube, 1/2" Tube, 5/8" Tube

5. NRC Documents

a. Harold Denton (NRC)/All Nuclear Power Plants, letter dated October 30, 1979, " Division of Lessons Learned Short Term Requirements"

b. Office Nuclear Reactor Regulation, USNRC, NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations"

6. WEPCO DOCUMENTS

a. Burstein (WEPCO)/Denton (NRC) letter dated October 20, 1979, " Dockets 50-266 and S0-301 Implementation of NUREG-0578, Point Beach Nuclear Plant Units 1 and 2"

                                                                                    ~

b. Burstein (WEPCO)/Denton (NRC) letter dated November 27, 1979, " Dockets 50-266 and 50-301 Implementation of NUREG-0578, Point Beach Nuclear Plant, Units 1 and 2" 1667 266 2.1.8.6 - 26

TABLE 1 HIGH-RANGE NOBLE GASEOUS TFFLUENT RADIATION MONITORING EQUIPMENT VINDOR: Eberline P.O. Box 2108 Santa Fe, New Mexico 87501 SYSTEM MODEL: RMS II HIGH ELECTRONIC DETECTOR I. CHANNEL RANGE CHANNEL $ MODEL QUANTITY

1. Auxilary 0.01-100 ECl-4 DAl-4 1 Building Rad /hr Vent

2. Drumming 0.01-100 ECl-4 DAl-4 1 Area Rad /hr Vent

3. Containment 1-10,000 ECl-6 DAl-5 2 Purge' Vent Ra'd/hr Stacks

4. Containme.: 0.1 to 1000 ECl-5 DAl-5 2 Gas Monitor Rad /hr

5. Combined 1 to 10 ECl-3 DAl-4 1 Air Ejector Rad /hr Decay Duct 1667 267 2.1.8.b - 27

TABLE 1 (Con t ' d ) HIGH ELECTRONIC DETECTOR CHANNEL RANGE CHANNEL 4 MODEL QUANTITY 6." Gas stripper 1 to 10 ECl-3 DAl-4 1 Stack Rad /hr

     .       7. Reactor             0.1 to 1000'       ECl-5                   DAl-5                     2 Coolant             Rad /hr Monitor Total                                                                                10 II. OTHER EQUIPMENT Quantity 1.

Racks for Supporting Modules 2 Eberline RMS 2 - II

2. Coax Cable Belden 9773: (Used Where as Detector to Module distance necessary is greater than 2000')

NOTES:

1. Cabinet for installation of racks to be located in control room is required.

1667 268 2.1.8.b - 28

s ATTACHMENT I PROCEDURE GUIDELINES FOR USE IN CONDUCTING NOBLE GASEOUS EFFLUENT MEASUREMENT 1667 269 2.1.8.6 - 29

PROCEDURE GUIDELINES FOR USE IN CONDUCTING ' NOBLE GASEOUS EFFLUENT MEASUREMENT The following information provides general guidelines which are being used in developing detailed procedures for the operation of the newly-installed noble gas effluent monitoring equipment. Table 1 contains a description of the ranges, type of equipment, and other data for the radiation monitoring detectors being installed. The following guidelines provide data for the use of those detectors in determining vent release rates in Ci/sec.

1. Assuming that an accident situation exists, the operator would observe the gaseous ef fluent monitor to obtain levels in Rad /hr for the plant vents.

2. Conversion to a release rate of Ci/see would be performed by utilization of conversion curves corresponding to the particular plant vent of interest. Background levels are assumed to be neglible due to the shielding installed around the detectors.

3. The following is a tabulation of the conversion charts to be used with the corresponding plant vent / stack:

o Auxiliary Building Vent Stack - Figure 4 - o Drumming Area Vent - Figure 7 o Containment Purge Vent Stack - Figures 9 & 10 o Combined Air Ejector Decay Duct - Figure 11 o Gas Stripper Stack - Figure 12 1667 270 2.1.8.b - 30

ATTACHMENT II PROCEDURE GUIDELINES FOR USE IN CONDUCTING MEASUREMENT OF RADIOIODINE AND PARTICULATES FROM GASEOUS EFFLUENTS 1667 271 2.1.8.5 - 31

ATTACEMENT II PROCEDURE GUIDELINES FOR USE IN CONDUCTING MEASUREMENT OF RADIOIODINES AND PARTICULATES FROM GASEOUS EFFLUENTS Each plant vent is equipped with a remote sampling line to enable radioiodines and particulates to be collected on a charcoal canister and filter paper. The procedure guidelines described here are used during a postulated accident to collect and analyre radiciodine and particulate samples.

1. In an accident situation, the operator utilizes the iodine and particulate remote sampling line shown typically in Figure 1 for each plant vent / stack.

2. The operator connects the portable sampling pump / charcoal cartridge and filter paper assembly at the sampling point via " quick disconnects" on the sample lines. Shield walls protect the operator from excess radiation exposures.

3. The operator opens the bypass valve to purge the sample line. The bypass valve is then closed and the sample valves opened.

4. The pump is run and a flow indicator observed until six volumes of sample are run through the charcoal cartridge and filter paper. Sample valves are then -

closed and the portable pump and cartridge filter and paper assembly is removed. 1667 272 2.1.8.6 - 32

5. The charcoal cartridge and filter paper assembly is placed in a lead sample enclosure assembly and transported to the analysis area.

6. The charcoal cartridge and filter paper samples are placed in a shielded container and the noble gases are allowed to defuse. A gamma spectrometer is then used to analyze the gas for iodine content.

7. The filter paper is counted for particulate levels using standard techniques.

8. The remainder of the vent stacks are sampled using the procedures in 1 through 7.

I667 273 2.1.8.b - 33

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