ML20036B956

Summary of 930504 & 05 Meetings W/Westinghouse in Monroeville,Pa Re Responses to Staff Request for Addl Info on SSAR Chapter 12,new 10CFR20 Considerations & AP600 Plant Layout for Purposes of Radiation Protection

ML20036B956
Person / Time
Site:	05200003
Issue date:	05/28/1993
From:	Essig T Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 9306070252
Download: ML20036B956 (50)

Text

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t UNITED STATES j

,.j NUCLEAR REGULATORY COMMISSION WASHINoToN. D.C. 20555-0001

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May 28, 1993 Docket No.52-003 APPLICANT: Westinghouse Electric Corporation PROJECT:

Westinghouse AP600

SUBJECT:

PUBLIC MEETING ON MAY 4 AND 5, 1993, TO DISCUSS OCCUPATIONAL RADIATION PROTECTION ASPECTS OF THE AP600 DESIGN A meeting was held on May 4 and 5, 1993, at Westinghouse offices in Monroeville, Pennsylvania, between representatives of the United States.

Nuclear Regulatory Commission (NRC) and the Westinghouse Electric Corporation (WEC). The purpose of the meeting was to discuss Westinghouse responses to the staff's request for additional information (RAI) on standard safety analysis report (SSAR) Chapter 12, new 10 CFR Part 20 considerations, and AP600 plant layout for purposes of radiation protection. provides a list of attendees.

During the meeting, the following specific topics were discussed: High and very high radiation area access control, vital area access (post accident),

shielding of modularized components, area radiation monitors, containment access during power operations, auxiliary building post-accident source term, and Cobalt replacement materials.

In addition, the staff discussed Westing-house's response to SSAR Chapter 12 RAI's (Radiation Protection), and reviewed numerous plant general arrangement / layout drawings and three dimensional models of certain AP600 systems. The meeting agenda is shown in Enclosure 2.

Non-proprietary presentation materials used by Westinghouse during the meeting are contained in Enclosure 3.

The following commitments were made during the meeting:

1.

Westinghouse will make a copy of two proprietary reports (Radiological Design Analysis Manual and ALARA Guidelines) available in its Rockville office for staff review. This action was to be completed by May 18, 1993.

2.

New radiation zone maps which specifically show access control points to the radiologically controlled area, including the spent fuel transfer tube, will be provided to the staff on or before November 5,1993.

3.

Westinghouse will re-review proposed locations for area radiation monitors-(ARMS) to determine if placing an ARM on the refuelling bridge crane and the spent fuel pool crane is warranted. ARM locations will be added to the radiation zone maps discussed in (2).

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Westinghouse will provide an additional set of full-size annotated layout and general arrangement drawings to facilitate the staff's review of SSAR Chapter 12. The staff will provide Westinghouse with a list of specific 4

drawings needed, as well as the needed annotations.

Original Stoned By:

i; Thomas H. Essig, Acting Section Chief i

Standardization Project Directorate i

Associate Director for Advanced Reactors j

and License Renewal Office of Nuclear Reactor. Regulation Enclosures As stated l

cc w/ enclosures:

See next page DJI_STRIBUTION:

Docket File PDST R/F DCrutchfield RBorchardt POR PShea TKenyon RHasselberg BPerch, 8H7 DISTRIBUTION w/o enclosures:

TMurley/FMiraglia WRussell, 12G18 GBagchi, 7H15 TEssig EJordan, MNBB3701 ACRS (11)

JMoore, 15B18 GGrant, 17G21 DCarter, 1004 JWigginton, 10D4 0FC:

LA:PDST:ADAR (A)SC:PDST:@AR DATE:

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05fX)793 05/t'6/93 0FFICIAL RECORD COPY:

DOCUMENT NAME:MTGSUM.TE

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MEETING ATTENDEES MAY 4 AND 5, 1993 HAE

' ORGANIZATION Andrea Sterdis Westinghouse Mike Conner Bechtel Wyman Lee Bechtel Vijay Parameswaran Bechtel Jim Sejvar Westinghouse Tom Essig NRC/NRR/PDST Dan Carter NRC/NRR/PRPB Jim Wigginton NRC/NRR/PRPB Tom Johnson Westinghouse Naum Alper Westinghouse Steve Zawalick Westinghouse Rao Mandava Westinghouse Heinrich Farin Bechtel Ed Cummins Westinghouse Carl Bergmann Westinghouse Dan Trombola Westinghouse Doug Ekroth Westinghouse

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WESTINGHOUSE AP600 RADIATION PROTECTION MEETING AGENDAITEMS

+ HIGH/VERY HIGH RADIATION AREA ACCESS o INCORPORATING NEW PART 20 REQUIREMENTS o ADDRESS VHRA REGULATORY GUIDE 8,38 RECOMMENDATIONS

+ VITAL AREA ACCESS (POST ACCIDENT) o LOCATIONS o EXPECTED DOSES AND STAY TIMES

+ SHIELDING OF IN-CONTAINMENT MODULARIZED COMPONENTS o PERSONNEL EXPOSURE DUE TO PROXIMITY OF COMPONENTS DURING MAINTENANCE OPERATIONS o MOVING OF COMPONENTS TO LOW DOSE AREAS

+ AREA RADIATION MONITORS o NUMBER AND LOCATIONS OF MONITORS o CRITICALITY MONITORING

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+ CONTAINMENT ACCESS REQUIREMENTS DURING POWER OPERATIONS i

o TYPES OF OPERATIONS o LOCATIONS o ENVIRONMENT (RADIATION, INDUSTRIAL HYGIENE

[e.g. HEAT])

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+ RAl's SUBMITTED TO DATE

+ VIEW ALL AVAILABLE AP600 MODELS SHOWING l

0 COMPONENT / SKID SHIELDING o BUILDING LAYOUTS AND COMPONENT PLACEMENT l

+ AUX. BUILDING ACCIDENT SOURCE TERM l

+ COBALT-60 REPLACEMENT MATERIALS (NOREM, i

AND EVERIT 50) REFERENCE EPRI REPORT l

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UNITED STATES NUCLEAR REGULATORY COMMISSION WESTINGHOUSE ENERGY CENTER

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E MAY 4 AND 5,1993

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WESTINGHOUSE ELECTRIC CORPORATION PRESENTATION TO UNITED STATES NUCLEAR REGULATORY COMMISSION WESTINGHOUSE ENERGY CENTER MAY 4 AND 5,1993 t

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M M M W 'M M WM M OU AGENDA Radiation Protection Meeting i

May 4 and 5,1993 l

Location: Westinghouse. Energy Center East i

May 4,1993, 8:30 am. CR 515a I

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Introduction R. Mandava II.

Overview 1.

Plant Arrangement T. Johnson 2.

ALARA J. Sejvar Goals / Regulatory requimments Dose assessment Dose reduction methods / features l

Industry experience III.

Shielding Calculations W. Lee IV.

Radiation Zone Maps M. O'Connor V.

Topics for Discussion 1.

liigit/Very fligh Radiation Area Access M. O'Connor Incorporating new Part 20 requirements Reg. Guide 8.38 recommendations

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May 5,1993, 8:30 am CR 515A 2.

Vital Area Access (Post accident)

Locations and stay time S. Zawalick Expected doses M. O'Connor 3.

Shielding of Incontainment Modularized Components J. Sejvar Personnel exposure due to proximity of components during maintenance operations Moving of components to low dose areas 4.

Contairunent Access Requirements During Power Operation Types of operations / Locations J. Sejvar Environment M. O'Connor 5.

Area Radiation Monitors Number and location of monitors M. O'Connor Criticality monitoring M. O'Connor 6.

RAl's Submitted to Date M. O'Connor 7.

AP600 models T. Johnson Component / Skid shiciding Building layouts and component placement VI.

Summary / Actions

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ALARA C

REVIEW OF MAJOR ALARA CONSIDERATIONS GOALS AND REQUIREMENTS 1

DOSE ASSESSMENT AP600 DESIGN FEATURES l

OPERATING PLANT EXPERIENCE i

DESIGN PHASE SOURCE REDUCTION EFFORTS l

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ALARA SSAR - CHAPTER 12 1

ALARA POLICY AND GENERAL CONSIDERATIONS RADIATION SOURCES

- DESIGN BASIS SOURCE TERMS FOR SHIELDING AND VENTILATION NORMAL OPERATION / REFUELING / DESIGN BASIS ACCIDENTS GENERALLY CONSERVATIVE RADIATION PROTECTION DESIGN FEATURES (REG. GUIDE'8.8)

- SPECIFIC ALARA EQUIPMENT FEATURES

- SHIELDING DESCRIPTION / ZONING DIAGRAMS

- ANALYTICAL METHODS

- VENTILATION CRITERIA / GUIDES (REFER TO SECTIONS 6 & 9)

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DOSE ASSESSMENT

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- PROJECTED ORE BY WORK CATEGORIES (REG. GUIDE 1.89) l l

l HEALTH PHYSICS FACILITIES

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Reactor Operations and Surveillance 14 21 i

Routine Maintenance 12 18 in-Service inspections 17 24 Special Maintenance 16 23 i

Waste Processing 5

8 Refueling 4

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SUMMARY

OF ALARA DESIGN FEATURES FOR AP600 STEAM GENERATOR l

BASIC ENGINEERING DESIGN IMPROVEMENTS (LESSONS LEARNED) l ROBOTICS CONSIDERATIONS l

CANNED MOTOR RCP's LESS MAINTENANCE l

REMOVED TO LOW BACKGROUND WORK AREA REDUCED NUMBER OF VALVES, PUMPS. ETC h

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STAFF, REQUIREMENTS THAT ARE 1/2 OF TODAY'S PLANTS l

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l LAYOUT FEATURES INTEGRATED SHIELDING PROVISIONS FOR MAINTENANCE MINIMAL FIELD-RUN PIPING SHOP FAB VS FIELD FAB ENGINEERED FOR ACCESSIBILITY AND MAINTENANCE PRE-TESTED FOR INCREASED RELIABILITY REDUCED PIPING WELDS REDUCED ISI REQUIREMENTS FEWER CRUD TRAPS

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SOURCE REDUCTION MEASURES

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INC-690 S/G MATERIAL LOW-COBALT IMPROVED CORROSION RESISTANCE ZlRC FUEL ROD GRIDS ALTERNATE MATERIALS FOR HARD-FACING VALVES

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ANNUAL COBALT INPUT (GRAMS /YR) FOR REFERENCE 2-LOOP PLANT AND AP600 TWO-LOOP COMPONENT PLANT AP600 REMARKS REACTOR COOLANT 38.8 11.6 USE OF LOW-COBALT S/G TUBING)

STEAM GENERATOR TUBING IN AP600 (0.015% OF COBALT IN LIEU OF 0.057% FOR TYPICAL)

FUEL ASSEMBLIES 0.2 0.4 AP600 HAS 45 CONTROL (CONTROL RODS)

CONTROL RODS VS. 29 FOR TYPICAL l

MCPS 1.1 4.0 AP600 HAS 4 VS. 2 MCPS

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ANNUAL COBALT INPUT (GRAMS.'YR) FOR REFERENCE 2-LOOP PLANT AND AP600 (CONTINUED)

TWO-LOOP COMPONENT PLANT AP600 REMARKS REACTOR VESSEL 1.4 1.7 AP600 INTERNALS SIMILAR INTERNALS TO 3-LOOP PLANT CRDMs 2.8 7.8 ASSUMES AP600 LOAD FOLLOW, TYPICAL OPERATING PLANT DOES NOT VALVES 5.0 3.0 AP600 HAS 60% OF THE VALVES IN THE TYPICAL PLANT TOTAL 49.4 28.5

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l INDUSTRY / PLANT INPUT UTILITY REQUIREMENTS DOCUMENT-EPRl/ UTILITY CONFORMANCE REVIEWS DIVISION TASK FORCE AP600 MODEL DISPLAYS AT UTILITY SEMINARS DIRECT CONTACT WITH OPERATING PLANT PERSONNEL INPUT FROM OUR SERVICE ORGANIZATIONS / DIVISIONS INTERACTIONS ASSOCIATED WITH INDUSTRY SOURCE REDUCTION TEAMS / PROGRAMS

_ REVIEW OF PLANT OUTAGE ALARA. -REPORTS -

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OPERATING PLANT EXPERIENCE INDUSTRY EXPERIENCE - MAJOR FACTORS FOR LOW ORE LOW SOURCE TERMS SMALL EXPERIENCED STAFF COMPONENT RELIABILITY EXAMPLES:

LOW ORE PLANTS GENERALLY OPERATE WITH SMALL EXPERIENCED STAFF

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PRAIRIE ISLAND LOWEST ORE / PLANT EFFLUENT / WASTE VOLUMES IN INDUSTRY DO NOT OPERATE WITH FUEL DEFECTS

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A Major Advantace of Plant Simplicity is Reduced Radiation Exposure to Workers

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AP600 HAS MANAGEMENT COMMITMENT BASED ON PROVEN COMPONENTS HAS FEWER COMPONENTS AND EQUIPMENT TO MAINTAIN AND INSPECT INCLUDES PROVISIONS FOR STATE-OF-THE-ART ROBOTICS SYSTEMS 1

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CONSTRUCTED OF MATERIALS THAT REDUCE THE AMOUNT OF CORROSION PRODUCT ACTIVITY ORE OBJECTIVE OF 70 MAN-REM PER YEAR - CURRENT ESTIMATE OF l

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AP600 SHIELDING CALCULATIONS j

CALCULATION PROCESS I

OPERATION l

SHUTDOWN CALCULATION RESULTS ZONING SHIELDING RECOMMENDATIONS

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AP600 SHIELDING CALCULATIONS L.2 SHUTDOWN NORMAL OPERATION _

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CALCULATION SHUTDOWN COMPUTER ANALYSIS RADIATION LEVEL 1r OPERATION RADIATION LEVEL V

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DESIGN DATA / ASSUMPTIONS 0.25% FAILED FUEL MAXIMUM ACCUMULATION TIME WITH RADIOACTIVE DECAY 100% RETENTION OF RADIOACTIVITY IN DEMINERALIZERS AND FILTERS DFs AS RECOMMENDED BY NUREG 0017 FOR PROCESS FLOW ACTIVITY MAXIMUM TANK LEVEL MULTIPLE SOURCES

AP600 SHIELDING CALCULATIONS COMPUTER PROGRAMS NE602 (SOURCE 2) 1 1)

RADIOACTIVITY ACCUMULATION WITH DECAY 2)

CALCULATE GAMMA ENERGY FROM ISOTOPIC ACTIVITIES 4

NE650 (CYLSEC/ SHIELD SG)

A POINT KERNEL 3-DIMENSIONAL SHIEl. DING CODE 1)

GAMMA ATTENUATION CALCULATION 2)

PIPING, DUCT AND ELECTRICAL PENETRATION CALC ORIGEN

1) ORNL ISOTOPE GENERATION AND DEPLETION CODE

'" MATRIX-EXPONENTIAL METHOD

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A PROGRAM COMPUTES GAMMA RAY SCATTERING FROM A POINT l

SOURCE TO A SERIES OF POINT DETECTORS ANISN i

A PROGRAM SOLVES THE BOLTZMANN TRANSPORT EQUATION FOR NEUTRON AND/OR GAMMA FLUXES. AS A FUNCTIONS OF ENERGY, ANGLE, AND POSITION IN ONE-DIMENSIONAL SLAB, CYLINDRICAL, OR.

SPHERICAL GEOMETRY

- QAD-CG A POINT-KERNEL. GAMMA AND. NEUTRON. SHIELD PENETRATION CODES WHICH CALCULATES DOSE RATE AND HEATING IN COMPLEX THREE-DIMENSIONAL-GEOMETRY AT SPECIFIED DETECTOR POINTS

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HIGH/VERY HIGH RADIATION AREAS REVISED 10CFR20 REQUIREMENTS INCORPORATED VERY HIGH RADIATION LEVELS ON ZONE DWGS l

DRAFT REG. GUIDE DG 8006 HIGH RAD AREAS (> 100 mr/hr, s;1 r/hr) l ACCESS CONTROL BY AUDIBLE / VISIBLE ALARMS HIGH RAD AREAS (> 1 r/hr)

ACCESS CONTROL BY LOCKED ENTRANCES VERY HIGH RAD AREAS (>500 rad /hr) 1 ACCESS CONTROL BY LOCKED ENTRANCES i

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AP600 APPROACH SHIELDING REQUIREMENTS i

BASED ON DESIGN BASES CONDITIONS 0.25% FAILED FUEL CONSERVATIVE ASSUMPTIONS FOR SYS. OPERATION CONSERVATIVE ASSUMPTIONS OF SOURCE CONFIGURATION ACCESS CONTROL REQUIREMENTS BASED ON EXPECTED DOSE LEVELS OPERATING EXPERIENCE AP600 SOURCE REDUCTION METHODS e

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- Times Design Basis Accident Activities Severe Accident Source Auxiliary Building Requirements to >1 Month Zone Diagrams w/ Routes i

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Post-Accident Vital Areas and Access Times Post-Accident Activity Auxiliary Building Time (in man-minutes)

Main Control Room Access Continuous Post-Accident Sampling 37 / sample Spent Fuel Pool Cooling Make-up 22 Containment Inventory Make-up 43 Temporary Water Hookup to Passive Containment 8

Cooling System Tank Temporary HVAC to Main Control Room and PAMS 17 Cabinets Temporary Power to Class IE Regulating Transformers 33 Temporary Power to Hydrogen Recombiners 26 4

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FREQUENCY AND DEGREE OF UNSCHEDULED REPAIRS SPECIFIC UTILITY PRACTICES AND PROCEDURES 4

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AT-POWER ACCESS OPERATIONS WHICH MAY REQUIRE CONTAINMENT AT-POWER ACCESS SURVEILLANCE ROUTINE PATROLS HEALTH PHYSICS SURVEYS PATROLS TO IDENTIFY SYSTEM LEAKS ELECTRICAL REPAIRS REPAIR / REPLACE TRANSMITTERS VALVE OPERATOR REPAIRS

AT-POWER ACCESS MECHANICAL REPAIRS ISOLATE SYSTEM / COMPONENT LEAKS VALVE ADJUSTMENTS / REPAIRS LEAK TEST PERSONNEL HATCH FAN COOLER REPAIRS RCP OIL ADDITIONS REPAIR IN-CORE DETECTOR DRIVE SYSTEM REPAIR / REPLACE RADIATION MONITORING DETECTORS STRIP CHART RECORDER MAINTENANCE

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80 F BULK TEMPERATURE MAINTAINED < 120 F RADIOLOGICAL VFS MAINTAINS UNLIMITED ACCESS TO GENERAL ACCESS AREAS BASED ON EXPECTED'RCS LEAKAGE NO RESPIRATORS REQUIRED

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i IN AREAS WHICH ARE NORMALLY ACCESSIBLE AND WHERE CHANGES IN NORMAL PLANT OPERATING CONDITIONS CAN CAUSE SIGNIFICANT INCREASES IN EXPOSURE RATES ABOVE THAT EXPECTED FOR THE AREAS IN AREAS WHICH ARE NORMALLY ACCESSIBLE OR OCASSIONALLY. ACCESSIBLE WHERE SIGNIFICANT INCREASES IN EXPOSURE-RATES MIGHT OCCUR DUE TO OPERATIONAL TRANSIENTS OR MAINTENANCE ACTIVITIES CONSIDER ATION IS GIVEN TO THE ENVIRONMENTAL CONDITIONS UNDER WHICH THE DETECTOR ELECTRONICS- ' WOULD BE OPERATING IN TO ENSURE PROPER OPERATION hw m.

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AREA RADIATION MONITORS mesoT No.

AP600 SSAR ANSI /ANS/HPSSC-6.8.1 Comments Table 11.5-2 Table 2 1

Post Accident Primary Sampling Sample Room Station Area 2

Containment Area Containment -

Personnel Hatch Area

[ Operating Deck (Inside Containment) elev.135']

Containment - Refueling Function provided by Platform ARM on Operating Deck

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AP600 SSAR ANSI /ANS/HPSSC-6.8.1 Comments Table 11.5-2 Table 2 Containment - In-Core AP600 has an Instrument Area integrated reactor vessel head package which includes the in-core instruments (top entry), therefore there l

is no in-core seal table area which would need area monitoring.

3 Main Control Control Room Only one ARM is Room Area [3 ea]

necessary 4

Chemistry Radiochemistry Laboratory Area Laboratory 5-Fuel Handling Area Fuel Storage and Cask Handling Areas Liquid Radwaste A common ARM Area provided 6

Gaseous Radwaste Area

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AP600 SSAR ANSI /ANS/HPSSC-6.8.1 Comments Table 11.5-2 Table 2 7

Solid Radwaste Radwaste Control Panel One ARM provided in Area Area Radwaste Control Room 8

Technical Support Provides information Center Area similar to MCR 9

Radwaste Storage Solid Radwaste Storage One ARM in the Area Area storage area corridor 10 Hot Machine Shop An ARM will be added to this area HVAC Filter Area Not provided - No rapid increase anticipated RHR Pump Area Not provided - The area is an access controlled area RHR Heat Exchanger Not provided - The area Area is an access controlled area

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AP600 SSAR ANSl/ANS/HPSSC-6.8.1 Comments Table 11.5-2 Table 2 Drumming Station Not provided - The area Control Panel Area is an access-controlled area Equipment Not provided inside Decontamination Areas Decon areas which have access control 11 Add one ARM to Annex il Staging and Storage Area

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M AREA RADIATION MONITORS CRITICALITY MONITORING 10CFR70 REQUIRES CRITICALITY MONITORS FOR STORAGE OF LARGE QUANTITIES OF SPECIAL NUCLEAR MATERIAL SUCH AS NEW FUEL ASSEMBLIES REGULATORY GUIDE 8.12 INDICATES THAT AN EXEMPTION FROM SEC.70.24 OF 10CFR70 IS APPROPRIATE IF CRITICALITY ANALYSIS OF NUCLEAR FUEL STORAGE AREAS DEMONSTRATES CRITICALITY IS PRACTICALLY IMPOSSIBLE SSAR SECTION 9.1 DESCRIBES THE NEW FUEL RACK DESIGN WHICH PROVIDES MINIMUM SPACING SUFFICIENT TO MAINTAIN REQUIRED DEGREE OF SUBCRITICALITY CRITICALITY ANALYSES PERFORMED IN ACCORDANCE WITH ANSI N16.1 AND SATISFIES THE PROBABILITY LIMITS IDENTIFIED IN ANSI 57.2 AND 57.3 TO DEMONSTRATE CRITICALITY IS PRECLUDED BY FUEL RACK DESIGN.

SSAR SECTION 4.3.2.6.1 DESCRIBES CRITICALITY ANALYSES

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RAI 471.1:

THE RESPONSE TO THIS RAI PROVIDED THE FOLLOWING INFORMATION FOR NRC TO PERFORM CONFIRMATORY SHIELDING CALCULATIONS:

MATERIAL COMPOSITION AND THICKNESS OF SHIELD WALLS AROUND RADIOACTIVE SOURCES DOMINANT RADIATION SOURCE EQUIPMENT NUMBER SOURCE GEOMETRY EFFECTIVE SOURCE DENSITY EQUIPMENT SELF-SHIELDING REFERENCE TO SSAR SOURCE TERM TABLE

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_J RAI 471.2:

THE RESPONSE TO THIS RAI PROVIDED THE FOLLOWING INFORMATION AS REQUIRED BY THE REVISED 10CFR PART20 AND FOR POST ACCIDENT CONDITIONS:

TRAFFIC PATTERNS FOR NORMAL / SHUTDOWN OPERATION IDENTIFICATION OF VERY HIGH RADIATION AREAS TRAFFIC PATTERNS FOR ACCESS TO VITAL AREAS DURING POST ACCIDENT l

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RAI 471.3:

THE RESPONSE TO THIS RAI PROVIDED THE FOLLOWING INFORMATION AS REQUIRED BY THE SRP TO ENSURE THAT PLANT'S VENTILATION FLOW IS SUFFICIENT TO MAINTAIN AIRBORNE RADIOACTIVITY LEVELS ALARA:

EXPECTED PEAK AIRBORNE RADIOACTIVITY CONCENTRATIONS IN ROOMS / AREAS WORK ACTIVITY / DESCRIPTION ESTIMATED MANHOUR OCCUPANCY ESTIMATED AIRBORNE DOSE TO OPERATING PERSONNEL

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