ML20155F137

Forwards Meeting Summary & Slides Presented at 860227 Meeting W/Epri to Improve Technical Interchange W/Industry. EPRI Programs Address Plant Design,Const,Operation & Life Extension

ML20155F137
Person / Time
Issue date:	04/11/1986
From:	Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:	Palladino NRC COMMISSION (OCM)
References
NUDOCS 8604210351
Download: ML20155F137 (165)

Text

.

APR 111996 MEMORANDUM FOR:

Chairman Palladino FROM:

Victor Stello, Jr.

-. Executive Director for Operations

SUBJECT:

MEETING WITH ELECTRIC POWER RESEARCH INSTITUTE NUCLEAR POWER PROGRAMS In order to further improve our technical interchange with the industry, NRR met with representatives of the Electric Power Research Institute (EPRI) on February 27, 1986. The EPRI programs address many substantive issues in the areas of_ nuclear plant design, construction, operation and life extension.

EPRI's budget for its nuclear power program is approximately $60 million in 1986. However, many of the associated programs are jointly funded and managed. resulting in a very economical and effective use of these resources.

Since the scope of the information discussed in this meetin'gsbears on the activities of the NRC, I am providing a copy of the slides'for your information.

Original signed by, Victor Stello Victor Stello, Jr.

Executive Director for Operations

Enclosures:

1. Meeting Summary

2. Copy of slides Distribution:

cc: Commissioner Roberts Central Files T. Rehm Comissioner Asselstine NRC PDR J. Sniezek Comissioner Bernthal ED0 Rdg. File Commissioner Zech TAMB Rdg. File SECY PPAS OPE V. Stello OGC H. Denton J. Funches

Contact:

T. Speis, NRR T. Speis 49-27517 M. Williams J. Roe

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MEMORANDUM FOR:

Chairman Palladino FROM:

Victor Stello,.Jr.

Executive Director for_ Operations

SUBJECT:

MEETING WITH ELECTRIC POWER RESEARC

-INSTITUTE NUCLEAR POWER PROGRAMS In order to further improve our technical interc nge'with the industry, NRR met with representatives of the Electric Power esearch Institute (EPRI) on

~.

-February 27, 1986. The EPRI programs addres many substantive issues in the areas of nuclear plant design, construction operation and life extension.

EPRI's budget-for its nuclear power progr is approximately $60 million in 1986.' However many of the associated ograms are jointly funded and managed', resulting in a very economica and effective use of these resources.

SincethescopeoftheinformationdJcussedinthismeetingbearsonthe activities-of the NRC, I am providing a copy of the slides for your information.

/

/

/

/

/

Victor Stello, Jr.

Executive Director for Operations

Enclosures:

1. Meeting Summary

2. Copy of slides Distribution:

cc: Commissioner oberts Central Files T. Rehm Commissioner Asselstine NRC PDR J. Sniezek Commission Bernthal E00 Rdg. File Commissio r Zech TAMB Rdg. File SECY PPAS OPE V. Stello OGC H. Denton J. Funches

Contact:

/T. Speis, NRR T. Spets

/49-27517 M. Williams J. Roe

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ELECTRIC POWER RESEARCH INSTITUTE MEETING

SUMMARY

On ' February 27, 1986, NRR met with EPRI, represented by the Department Directors of Nuclear Power Programs (NPD) to discuss the current and future research.

EPRI provided an overview of their nuclear power program followed by a fairly detailed discussion of several major projects. Some of the more significant products from their work are the development of guidelines for steam generator chemistry control, nondestructive examination, BWR pipe crack guidelines on inspection and chemistry control and remote examination and sampling by robots. This summary report of the meeting is provided to you for your information.

For further information consult the detailed EPRI slides contained in Enclosure 2.

EPRI gave a presentation that was divided into four major divisions which paralleled the organization of EPRI NPD. These divisions are:

1.

Nuclear Safety Analysis Center (NSAC) 2.

Engineering and Operations Center (E&O) 3.

Safety Technology (ST) 4.

Systems and Material Department (S&M)

The NSAC was organized at EPRI by the utilities as a result of the TMI event.

NSAC International Program has 13 Associate Members. Programs are focused on Generic Safety Issues. Some are conducted using the matrix management system that involves the other departinents in the following areas:

1.

Pressurized Thermal Shock 2.

Steam Generator Integrity 3.

Decay Heat Removal 4.

Piping Integrity 5.

Technical Specification Improvement 6.

Diesel Generator Reliability The results of the NSAC work is integrated toward resolution of safety issues. Approximately 32 individual issues are currently being tracked in the Generic Safety Analyses Program. The issues have been prioritized in four levels and only a few can be worked on in light of budget limitations. The top priority items are:

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Steam Generator Integrity 2.

Environmental Qualification of Equipment 3.

ATWS 4

Seismic. Design Criteria 5.

Station Blackout 6.

Hydrogen Control Measures

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The Engineering and Operations Department is concentrating its research in the areas of:

1.

Nuclear Plant Life Extension 2.

Nuclear Plant Constructibility 3.

Plant Availability 4.

Low Level Waste and Coolant Technology The Nuclear Plant Life extension work is using the Surry 1 and Monticello plants as models. The work is concentrated in a detailed assessment of the major components, systems and structures. The study is identifying the utility actions that are recommended to establish a basis for plant life extension. These actions are identifying maintenance requirements along with necessary plant equipment records and implementation of material sampling plans which would show the effects of aging on material properties. Aging specimens have been installed in 7 plants at this time.

In the area of nuclear plant constructibility EPRI is researching ways to reduce plant construction costs for new plants and modifications to operating plants. The work is centering on using computer-aided engineering techniques, providing program management for the Nuclear Construction Issues Group (NCIG) and conducting workshops on plant lay-up and equipment preservation. As a tool for developing better plant layout EPRI has developed a three dimensional computer model for the Midland plant.

The 1985 key accomplishments for the E&O Department in these areas were:

1.

Surry 1 and Monticello Pilot Project on plant life extension under way.

2.

Developed interim guidelines for application of computer-aided engineering.

3.

Completed Midland three dimensional construction model.

Provided pro 4.

Group (NCIG) gram management for Nuclear Construction Issues 5.

Conduct workshop on plant lay-up and equipment preservation.

6.

Published guidelines /databook on qualification on mechanical equipment.

7.

Conducted seminars on maintaining equipment Agingspecimens(cable, devices, lubricants) qualification.

8.

installed in 7 plants.

The E&O is studying the minimization of waste by various techniques and evaluation of the impact of these techniques.

EPRI has developed a computer program to perform radwaste volume reduction economic analysis.

In the area of Coolant Technology the work is centered in Cobalt 60 control through zinc infection, surface modification, prefilming of surfaces and cobalt alloy replacement.

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The Maintenance and Equipment Application Center (MEAC) has developed two robots to assist the utilities in inspection and limited sampling of areas within a plant that would be too dangerous for humans to enter or work continuously. These robots are the Surveyor Mobil Robot and the Surbot Mobile Robot. Additional MEAC has developed and demonstrated an underwater inspection vehicle. The MEAC provides training courses on pressure boundry bolting and diesel generator maintenance. Current R&D emphasis is in the areas of Plant Life Extension, Decontamination, Robotic Devices, Equipment Qualification, Computer Generated Displays and Computer Aided Construction.

The Safety Technology Department is active in the areas of:

1.

Source Term 2.

Seismicity Resolution 3.

Codes and Standards on Structural Response 4.

Integration of PRA on major issues 5.

Digital Systems Implementation The Safety Technology Department has recently focused heavily on source term and seismic technology. There is growing emphasis on safety control development. There is continuing work on a variety of computer codes for safety analysis, fuel management and near-real-time plant monitoring.

All of the work requiring experimental input is geared toward large scale demonstration experiments and tests that are frequently co-sponsored by government, industry or foreign organizations.

The seismic work involves the EUS seismicity study and large installation in i

Taiwan that consists of large scale concrete containment type structure to l

measure the effects of earthquakes. This information will be used to validate the current methods that are in use for predicting structural responses.

Considerable work has been done in the areas of internal pressure effects on concrete containment walls that have liners attached. Work has been done and is still ongoing that will be used in producing computer codes including one that can be used to predict the transoort and deposition of sediment in the coolant piping. The source term experimental work is nearing completion and the next task will focus on integration of information to support implementation of the technology in a meaningful way.

The Systems and Materials. Department is active in the following areas:

1.

LWR Fuel and Spent Fuel Storage 2.

Component Reliability 3.

Corrosion Control l

4.

Advanced LWR 5.

BWR Owners l

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' c The Systems and Material Department LWR fuel group is developing a technical basis for designing LWR fuel for higher burnup. Limited work is being done on zircaloy corrosion and fission gas release. Additionally, the group is nearly finished with a computer code that.will provide the utilities-with the capability to evaluate fuel designs and support licensing of reloads. Final version is scheduled for release on March 31, 1986. The group is also developing and transferring to field practice qualified NDE equipment and procedures. The group is conducting a test program structured to increase the allowable pipe stresses and determining the feasibility of using simplified design methods.

In the area of pressurized thermal shock, the Department is working to develop correct elastic models for crack arrest in the ductile regions of the pressure vessel. A significant result of this work will be additional life for neutron damaged vessel steel. Corrosion fatigue crack growth has been studied extensively during the past five years. _The Department is studying the relationships between hydrogen water' chemistry and intergranular stress corrosion and how this corrosion can be mitigated. The Department has demonstrated the feasibility of an on-line, in-plant corrosion cracking sensor. Work that has been done by EPRI shows that the current seismic piping design is very conservative.

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r AGENDA EPRI R&D BRIEFING FOR NRC MANAGEMENT EPRI Washington DC Office 1800 Massachusetts Ave.-NW Suite 700 Washington, DC, 20036 FEBRUARY 27, 1986 8:30 a.m.

CONTINENTAL BREAKFAST 9:00 WELCOME AND INTRODUCTIONS Bill Rasin, Duke Power Company Chairman, Safety Technology Task' Force (for Ed Kintner, GPU Chairman, Nuclear Power Divisional Committee)

John Taylor, Vice President - Nuclear EPRI Harold Denton, Director of Regu~iation NRC 9:15 NUCLEAR POWER DIVISION PROGRAM John Taylor Overview 9:30 NUCLEAR SAFETY ANALYSIS CENTER (NSAC)

Jeff Jeffries, Carolina Power & Light Company Chairman, NSAC Task Force A. David Rossin, Director, NSAC Overview Issue Prioritization Decay Heat Removal PTS Code Acceptance Station Blackout 10:00 BREAK

' AGENDA EPRI R&D BRIEFING FOR NRC MANAGEMENT FEBRUARY 27, 1986 PAGE 2 10:15 a.m.

ENGINEERING & OPERATIONS DEPARTMENT (E&O)

Bud Fay, Wisconsin Electric & Power Company Chairman, E&O Task Force c

Don Rubio, Director E&O Department Overview Nuclear Plant Life Extension & Constructability Plant Availability Low-Level Waste and Coolant Technology 11:00 SAFETY TECHNOLOGY DEPARTMENT (ST)

Bill Rasin, Duke Power Company Chairman, ST Task Force Walt Loewenstein, Director, ST Department Overview Source Term Research Seismic Center Safety Control Testing to Support Analyses Disciplined Software 12:00 Noon LUNCH 12:45 p.m.

SYSTEMS & MATERIALS DEPARTMENT (S&M)

Lou DelGeorge, Commonwealth Edison Company Chairman, S&M Task Force Karl Stahlkopf, Director, S&M Department Overview Hydrogen Water Chemistry BWR Water Chemistry and Impurities l

NDE Center / Ultrasonic Inspection Seismic Response of Piping and Supports l

2:00 OISCUSSION John Taylor Harold Denton 2:45 A0JOURNMENT

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i EPRI R&D BRIEFING FOR

'M, NRC MANAGEMENT 9

FEBRUARY 27, 1986 F

G NUCLEAR POWER DIVISION ELECTRIC POWER RESEARCH INSTITUTE WASHINGTON, DC

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EPRI OVERVIEW 0F EPRI'S NUCLEAR POWER PROGRAM ELECTRIC POWER RESEARCH INSTITUTE NUCLEAR POWER DIVISION JOHN J. TAYLOR EPRI R&D BRIEFING FOR V.S. NUCLEAR REGULATORY COMMISSION MANAGEMENT FEBRUARY 27. 1986 l

i Nuclear Power Division l

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EPRI OVERVIEW 0F EPRI'S NUCLEAR POWER PROGRAM UTLINE INTRODUCTION EXAMPLES OF PRODUCTS FROM 1985 STATilS OF NPD R&D NEW MAJOR PROJECTS GUIDELINES

SUMMARY

Nuclear Power Diviolon

EPRI OVERVIEW OF EPRI'S NUCLEAR POWER PROGRAM SOURCE TERM

+

LARGE TEST PROGRAMS

+

WORLDWIDE TECHNOLOGY INTEGRATION

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IDCOR INTERFACE SEISMIC

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SEISMIC MARGINS

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LOTUNG TEST PIPING INTEGRITY

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SE!SMIC

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BLOWDOWN CONTROL TECHNOLOGY

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SIMULATORS

+

A1

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DIGITAL CONTROL i

Nuclear Powa-Division

EPRI OVERVIEW OF EPRI'S NUCLEAR POWER PROGRAM TESTING TO SUPPORT ANALYSIS

+

MIST

)

DECAY HEAT REMOVAL TECHNICAL SPECIFICATION IMPROVEMENT STATION BLACK 0UT SOFTWARE LWR FUEL & SPENT FUEL STORAGE

+

HIGH BURNUP FUEL l

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l Nuclear Power Olvision w

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EPRI OVERVIEW OF EPRI'S NUCLEAR POWER PROGRAM MAINTAINABILITY

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IMPROVE U.S. PLANT AVAILABILITY

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PROVIDE TECHNICAL INITIATIVES TO CONTAIN 0 & M COST ESCALATION

+

REDUCE MAINTENANCE RELATED ABNORMAL EVENTS

+

5 YEAR, 56.9 MILLION EFFORT ROBOTICS

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ESTABLISH UTILITY ADVISORY PANEL

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DEVELOP GUIDELINES FOR POWER PLANT ROBOTIC AIDS

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DEVELOP ROBOTS OPTIMIZED TO ASSIST IN PLANT OPERATION, SURVEILLANCE AND MAINTENANCE 6+ YEAR, $12.9 MILLION EFFORT

+

Nuclear Power Division l

EPRI OVERVIEW 0F EPRI'S NUCLEAR POWER PROGRAM PLANT LIFE EXTENSION

+

PILOT PLANT EVALUATION

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GENERIC BWR & PWR EVALUATIONS

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CRITICAL STRUCTURES, COMPONENTS &

SYSTEMS

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RECORDS REQUIREMENTS i,},,u(, d A

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LIFE EXTENSION GUIDELINES

+

5 YEAR, 57 MILLION EFFORT l

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EPRI OVERVIEW 0F EPRl'S NUCLEAR POWER PROGRAM COMPONENT RELIABILITY

+

NDE R&D - INSPECTION & DETECTION

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DYNAMIC PLUS STATIC P! PING LOADS

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CRACK GROWTH CORROSION CONTROL

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HYDROGEN WATER CHEMISTRY STEAM GENERATOR LONG TERM EFFECTIVENESS OF CORRECTIVE

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MEASURES

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NEWER PROBLEMS NOT COMPLETELY ADDRESSED

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INCENTIVE: COSTS OF LOST CAPACITY HIGH

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5 YEAR, 530 MILLION EFFORT

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INTERACTIONS WITH NRC, INPO, DOE Nuclear Power Division

- ~ - - - -

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EPRI OVERVIEW 0F EPRI'S NUCLEAR POWER PROGRAM LOW LEVEL WASTE

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WASTE & VOLUME REDUCTION

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RADWASTE DIRECT ASSAY

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LLW DISPOSAL TECHNOLOGIES COOLANT TECHNOLOGY

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CHEMICAL CONTROL

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RADIATION CONTROL NUCLEAR PLANT CONSTRUCTABILITY

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EQUIPMENT QUALIFICATION DATA BASE

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CAE GUIDELINES h

d Nuclear Power Division j

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EPRI n

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STEAM GENERATOR GUIDELINES lE k

E STEAM GENERATOR REFERENCE BOOK

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COMPENDIUM OF OPERATING EXPERIENCE, LABORATORY DATA, AND GUIDELINES

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RECOMMENDS CORRECTIVE ACTION FOR S/G CORROSION

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RECOMMENDS REPLACEMENT S/G AND B-0-P DESIGN FEATURES

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E SECONDARY WATER CHEMISTRY GUIDELINES k

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WITHOUT REGULATION, ADOPTED BY U.S. &

FOREIGN UTILITIES

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ESSENTIALLY ELIMINATES DENTING AS A CO:lCERN NDE GUIDELINES a

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EPRI NDE CENTER CLASSES TO TRAIN EDDY g"r CURRENT ANALYSTS k

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EPRI NDE CENTER CLASSES ON PLANNING UTILITY 151 PROGRAMS AND MONITORING VENDORS (STARTS APRIL 1986) h kr E

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Nuclea r Power Division U-h

EPRI OVERVIEW OF EPRl'S NUCLEAR POWER PROGRAM BWR PIPE CRACK GUIDELINES NDE INSPECTION GUIDELINES (COORDINATION PLAN - EPRI, BWROG & NRC -

LEADS TO MUTUALLY SATISFACTORY CREDIBLE INSPECTION)

BWR WATER CHEMISTRY GUIDELINES (WITHOUT REGULAT!uNS THESE GUIDELINES ARE WIDELY ADOPTED)

HYDROGEN WATER CHEMISTRY IMPLEMENTATION GUIDELINES (COMMON BASE FOR DESIGNING FACILITIES FOR STORING AND INJECTING WATER INTO A BWR)

Nuclear Power Division

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NUCl. EAR SAFETY ANALYSIS CENTER NRC BRIEFING A. DAVID ROSSIN FEBRUARY 27, 1986

NSAC REPORTS

-1985-REPORT NUMBER TITLE 75 FRACTURE EVALUATION OF A W REACTOR.DURING A PTS TRANSIENT 83 BRUNSWICK DECAY HEAT REMOVAL PROBABILISTIC SAFETY STUDY 84 ZION RESIDUAL HEAT REMOVAL SYSTEM PROBABILISTIC RISK ASSESSMENT 85 LOSSES OF 0FF-SITE POWER AT U.S. NUCLEAR POWER PLANTS-ALL YEARS THROUGH 1984-86 REALISTIC ECCS EVALUATION METHODOLOGY FOR ADVANCED LWRs 87 PLANT SPECIFIC COMPARED TO GENERIC ASSESSMENT OF STATION BLACK 0UT 88 RHR EXPERIENCE REVIEW AND SAFETY ANALYSIS 89 FRACTURE EVALUATION OF A C-E REACTOR DURING A PTS TRANSIENT 90 DEVELOPING A LIVING SCHEDULE FUNDAMENTAL CONCEPT 91 A PARAMETRIC STUDY OF AN ANTICIPATED TRANSIENT WITHOUT SCRAM IN A W FOUR-LOOP PLANT 92 SETPOINT RELAXATION ANALYSIS FOR SCRAM REDUCTION USING RETRAN-02 ON C-E ANALOG PWRs 93 SETPOINT RELAXAi!0N ANALYSIS FOR SCRAM REDUCTION USING RETRAN-02 ON C-E DIGITAL PWRs

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NSAC REPORTS

-1985-(CONTINUED)

REPORT NUMBER TITLE 94 REDUCING SCRAM BY MODIFYING REACTOR SETPOINTS FOR A W 4-LOOP PLANT

..P-3975 ANALYSIS OF THE HYDROGEN BURN IN THE TMI-2 CONTAINMENT NP-3938 A FULL RANGE DRIFT FLUX CORRELATION FOR VERTICAL FLOW NP-4146 EPRI'S R&D CONTRIBUTIONS TO THE TECHNICAL BASIS FCR REVISION OF-ECCS RULES NSAC REPORTS

-1986-95 GENERIC SAFETY ISSUE TRACKING AND EVALUATION

SUMMARY

DESCRIPTIONS-1985 (IN PRESS) 96 EFFECT OF DIESEL START TIME ON BWR/6 PEAK CLAD TEMPERATURE-LICENSING BASIS CALCULATIONS

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NSAC INTERNATIONAL ASSOCIATES 13 ASSOCIATES NSAC ASSOCIATES (DOMESTIC)

AMERICAN ELECTRIC POWER FLORIDA POWER CORPORATION NEW YORK POWER AUTHORITY DAIRYLAND TOLEDO EDISON NEBRASKA PUBLIC POWER DISTRICT OMAHA PUBLIC POWER DISTRICT (J0INED EPRI)

ALL PENNSYLVANIA AND NEW JERSEY NUCLEAR UTILITIES TREATED LIKE MEMBERS OR ASSOCIATES DETROIT EDISON (1986)

CONSUMERS POWER (1986)

GULF STATES (1986)

MIDDLE SOUTH (1986)

LONG ISLAND LIGHTING (1985/86)

L.

COMBINED ALL NSAC ACTIVITIES INT 8 ONE RESEARCH PROJECT RP-2420 GENERIC SAFETY ANALYSIS ALWR SUPPORT NUMARC SUPPORT MATRIX MANAGEMENT (IN-HOUSE RESEARCH COSTS)-

REGULATORY SAFETY ASSESSMENT HYDROGEN CONTROL OWNERS GROUP (N0 EPRI BASE PROGRAM FUNDING IN 1986) r l

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4 NUCLEAR POWER-DIVISION PROGRAMS UNDER NSAC MATRIX MANAGEMENT MATRIX MANAGER PRESSURIZED THERMAL CHEXAL SHOCX (COMPLETED)

STEAM GENERATOR INTEGRITY LANG

. DECAY HEAT REMOVAL VINE PIPING INTEGRITY CHEXAL TECHNICAL SPEC. IMPROVEMENT POWER DIESEL GENERAT,0R WYCKOFF PERFORMANCE

i SCRAM ON LOW SG LEVEL SUME MILITARY PLANTS HAVE NOT USED S/G LEVEL SYSTEMS HISTORICALLY S/G LEVEL SCRAM WAS A COST REDUC TRADE-0FF MADE RATHER THAN TO INCORPORATE MOR PRESSURIZER SAFETY VALVE CAPACITY INDICAT TRANSIENT ANALYSIS BETTER ANALYSIS TOOLS AVAILABLE NOW SCRAM ON TURBINE TRIP CAN WE RAISE THE SCRAM SETPOINT ABOV LEVEL?

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CONCLUSIONS SG LEVEL SCRAM 1.

IN ALL CASES WHERE REACTOR PROTECTION IS REQUIRED BECAUSE OF LOSS OF THE S/G HEAT SINK. ADEQUATE PROTECTION.IS PROVIDED BY SCRAMS OTHER THAN S/G 2.

W PLANT STEAM FLOW / FEED MISMATCH SCRAM IS NO 3.

W PLANT LOW-LOW SG LEVEL CAN BE REMOVED FOR REAC POWER BELOW 50%.

TURBINE TRIP 4.

W PLANT WITH 40% STEAM DUMP CAN INCREASE SCRA FROM 10% TO 50% POWER ON TURBINE TRIP.

5.

CE PLANT WITH 40% STEAM DUMP CAN INCREASE SCR ON TURBINE TRIP FROM 15% TO AT LEAST 35%.

A HIGHER POWER LEVEL PROBABLY CAN BE DEMONSTRATED THRO ANALYSIS.

t.

L CONCLUSION ALL NSAC WORK INTEGRATED TOWARD RESOLUTION OF SAFETY ISSUES EXAMPLE:

STATION BLACK 0UT

- LOSS OF 0FF-SITE POWER

- DIESEL GENERATOR RELIABILITY

- DIESEL GENERATOR START TIME

- DC POWER RELIABILITY WORKSHOP

- DECAY HEAT REMOVAL

- ENGINEERING REVIEW OF DESIGN CHANGES

- PRA ON'DHR (BASED ON OCONEE PRA EXPERIENCE)

- COST / BENEFIT ANALYSIS; ON-SITE AVERTED COSTS 9

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- GENERIC SAFETY ANALYSIS PROGRAM

,y W. LAYMAN-J.CHA0 B.-CHEXAL J. LANG~

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ABOUT 30 ISSUES ARE BEING TRACKED ANALYSIS UNDERWAY ON THE FOLLOWING ISSUES:

SHUTDOWN DECAY HEAT REMOVAL SCRAM REDUCTION STATION BLACK 0UT REDUCTION OF PIPING RESTRAINTS TECHNICAL SPECIFICATION IMPROVEMENT ANALYSES COMPLETED FOR THE FOLLOWING ISSUES:

PRESSURE VESSEL THERMAL SHOCK BWR ATWS STEAM GENERATOR SAFETY ISSUES LOSS OF.0FFSIGHT POWER DATA BASE DHR RISK EVALUATION USING PRA (WITH REGULATORY TECHNICAL ASSESSMENT GROUP)

7-n_

OVERALL INDUSTRY PR!0RITY REACTOR VESSEL THERMAL SHOCK A

STEAM GENERATOR INTEGRITY AAA S'HUTDOWN HEAT REMOVAL SYSTEMS AAA-ENVIRONMENTAL QUALIFICATION OF EQUIPMENT AA EMERGENCY PLANNING AA SINGLE FAILURE CRITERIA A

ATWS AAA RELIABILITY OF VITAL EQUIPMENT AA NEAR TERM REGULATORY REQUIREMENT PRIORITIZATION A

SITING CRITERIA A

WATER HAMMER A

CONTAINMENT INTEGRITY AA FRACTURE TOUGHNESS OF S/G AND RCP SUPPORTS INACTIVE SYSTEMS INTERACTIONS AA-SEISMIC DESIGN CRITERIA AAA

OVERALL INDUSTRY PRIORITY,

. CONTAINMENT EMERGENCY SUMP PERFORMANCE A

STATION BLACX0VT AAA SEISMIC QUALIFICATION OF EQUIPMENT AA

- SAFETY IMPLICATIONS OF CONTROL SYSTEMS AA HYDROGEN CONTROL MEASURES AND BURN EFFECTS AAA DC POWER RELIABILITY AA

- I&C SYSTEM RELIABILITY AA PLANT-AIR SYSTEMS AN0MAllES AA LIGHTNING PROTECTION

,A

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IN-CORE INSTRUMENTATION TUBE RUPTURE INACTIVE TWO-PHASE NATURAL CIRCULATION A

HP INJECTION SYSTEMS AA INADVERTENT CRITICALITY INACTIVE HIGH STRENGTH BOLTING DEGRADATION INACTIVE PIPE CRACKS IN BOILING WATER REACTORS AA P! PING INTEGRITY AA TECHNICAL SPECIFICATIONS AA e

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RESIDUAL HEAT REMOVAL EXPERIENCE REVIEW AND SAFETY ANALYSIS-PWRs:

NSAC-52 BWRs:

NSAC-88 ACTUAL LOSSES OR DEGRADATIONS OF OPERATING RHR SYSTEMS IN SHUTDOWN COOLING MODE:

PWRs:

96 EVENTS OVER FIVE YEARS BWRs:

90 EVENTS OVER SEVEN YEARS MAJOR SAFETY BENEFIT IN COLD SHUTDOWN COMES FROM MAINTAINING ADEQUATE COOLANT INVENTORY IN REACTOR (RECOGNITION, PREVENTION, AND RECOVERY ARE ALL IMPORTANT).

SHUTDOWN' REACTOR SAFETY IS ENSURED BEST BY TRAINING, PROCEDURES, MANAGEMENT AND ADMINISTRATIVE CONTROLS WHICH ADDRESS POTENTIAL CHALLENGES AND MAKE EFFECTIVE USE OF ALL.AVAILABLE EQUIPMENT.

MAJOR DESIGN CHANGES ARE NOT NECESSARY TO COPE WITH CONCERNS RAISED BY OPERATING EXPERIENCE ANALYSIS.

MAJOR CONCLUSIONS AND RECOMMENDATIONS OF NSAC-52 FOR PWRs ARE CONSISTENT WITH RECENT INP0 SOER AND AEOD CASE STUDY.

INP0 IS DEVELOPING SOER BASED ON NSAC-88 FOR BWRs.

MANY AREAS OF POTENTIAL SAFETY CONCERN HAVE ALREADY BEEN

' ANALYZED.

OVER 100 NRC AND INP0 EVENT REPORTS HAVE BEEN WRITTEN ON EVENTS IN NSAC-52 AND NSAC-88.

PLANT SPECIFIC PRAs 0F ZION (NSAC-84) AND BRUNSWICK (NSAC-83) VALIDATED CONCLUSIONS OF OPERATING EXPERIENCE STUDIES.

PRINCIPLE CONCLUSIONS CONCERNING RISK OF CORE OR PLANT DAMAGE DURING SHUTDOWN:

RHR WEAKNESSES OBSERVED IN OPERATING EXPERIENCE EVALUATED BY PRA TO RESULT IN ACCEPTABLY SMALL RISK OF CORE MELT, RISK OF PLANT DAMAGE FURTHER REDUCED BY ENSURING ADEQUATE AVAILABILITY OF DHR SYSTEMS, INVENTORY ADDITION SYSTEMS, AND KEY SUPPORT SYSTEMS DURING SHUTDOWN.

THIS INCLUDES MAINTAINING MAIN CONDENSER (AND STEAM GENERATORS ON PWRs) AVAILABLE DURING HOT SHUTDOWN.

PRA SENSITIVITY STUDIES SHOWED PROCEDURES. TRAINING.

ADMINISTRATIVE AND MAINTENANCE IMPROVEMENTS ARE VERY IMPORTANT.

PLANT INCIDENT ASSISTANCE

-- PLANT:

RANCHO SECO DATE:

DECEMBER 26, 1985 CONTACT:

RON COLOMBO PHONE CALL ACTION:

LAYMAN, ROSSIN TO PLANT 12/26 FOLLOW-UP:

PTS CALCULATION REPORT TO SMUD l-7-86 CONCLUSION:

C00LDOWN RATE SIMILAR TO " LIGHT BULB" EVENT, BUT SHORTER DURATION AND LESS TEMPERATURE CHANGE CONTACT ARRANGED FOR ICS CONSULTATION

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04:10:00 20:00 30:00 40:00 04:50:00 TIME RCS-COLD TEMP

EPRI ASSESSMENT OF RANCHO SECO REACTOR VESSEL TRANSIENT 12/26/85

- 12/31 RECEIVED CONFIRMATION OF TRANSIENT INFORMATION AND REACTOR VESSEL FLUENCE FROM SMUD 1/6 COMPLETED ANALYf.IS REPORT AND MAILED FINAL REPORT TO SMUD CONCLUSION RANCHO SECO REACTOR VESSEL BELTLINE REGION HAS ADEQUATE STRUCTURAL INTEGRITY FOR RETURN TO SERVICE WITHOUT FURTHER EVALUATION.

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DRAFThSMESECTIONXIAPPENDIXFOREVALUATIONOF UNANTICIPATED OPERATIONAL TRANSIENTS ALLOWS DEMONSTRATION OF ADEQUATE STRUCTURAL INTEGRITY OF THE REACTOR VESSEL BELTLINE WITHOUT DOING FURTHER INTEGRITY ANALYSES AS LONG AS 1.

REACTOR COOLANT PRESSURE HAS NOT EXCEEDED DESIGN PRESSURE.

T -RTNDTS HAS NOT BEEN LESS THAN 55'F DURING THE 2.

C TRANSIENT.

RTNDTS = RTNDTo + a RTNDT + 2.

NDT = [CF] F (. 8.1 LOG F) (REG GUIDE 1.99) a RT CF FROM SECY-82-465 i

ASME SECTION XI APPENDIX FOR EVALUATION OF UNANTICIPATED TRANSIENTS 1/83 GRIESBACH AND MARSTON AS ASME SECTION XI MEMBERS GOT COMMITTEE TO ESTABLISH A WORXING GROUP DN OPERATIONAL' TRANSIENTS.

GRIESBACH HAS BEEN THE CHAIRMAN OF THE WORKING GROUP SINCE 1/83.

83-85 VIRTUALLY ALL OF THE FUNDING FOR THE WORKING GROUP TECHNICAL WORK WAS FROM EPRI.

2/86 THE DRAFT ASME SECTION XI APPENDIX FOR EVALUATION OF UNANTICIPATED OPERATIONAL TRANSIENTS WILL GO TO THE SECTION XI SUBCOMMITTEE FOR ITS APPROVAL, NRC TECHNICAL PERSONNEL HAVE BEEN INVOLVED IN THE PREPARATION AND REVIEW 0F THE DRAFT AND HAVE ENDORSED IT.

4 1.0F 2 WHAT DETERMINES THE RISK FROM STATION BLACK 0UT

-FOUR IMPORTANT' FACTORS-8 0FF-SITE POWER UNRELIABILITY.

IS OF DECLINING IMPORTANCE BECAUSE OVERALL UNRELIABILITY IS EXTREMELY LOW PLANT SPECIFIC UNRELIABILITY IS DIFFICULT TO DETERMINE 8

EMERGENCY ON-SITE POWER REDUNDANCY AND UNRELIABILITY THESE ARE GENERALLY RECOGNIZED AS THE BACKBONE OF A NUCLEAR PLANTS SHUTDOWN POWER RELIABILITY 8

OTHER RECOGNIZED BUT NOT FULLY QUALIFIED AC POWER SOURC THESE PLAY A MAJOR ROLE OUTSIDE THE U.S.

THEY MUST BE COVERED BY OPERATING AND SURVEILLANCE TESTING PROCEDURES DO NOT HAVE TO BE SEISMICALLY QUALIFIED 4

0 4

- - -e-

2 0F 2 WHAT DETERMINES THE RISK FROM STATION BLACK 0UT

-FOUR IMPORTANT FACTORS-O COPING DURATION IS THE PERIOD DURING WHICH THE PLANT CAN PREVENT CORE UNC0VERY FROM THE TIME WHEN OFF-SITE AND EMERGENCY ON-SITE AC POWER IS LOST OVERSEAS UTILITIES AND REGULATORS GENERALLY ACCEPT THAT RECOGNIZED BUT NOT QUALIFIED AC POWER SOURCES PROVIDE COPING TIME HOW TO INCLUDE SUCH SOURCES IN COPING DURATION MUST YET BE WORKED OUT BY U.S. NUCLEAR INDUSTRY AND NRC.

1 0F 2-i EMERGENCY DIESEL GENERATOR NOTES 3

FROM OVERSEAS O

GERMANS HAVE 4 EDGs PER NUCLEAR UNIT a

ONE IS SUFFICIENT IF NO LOCA HAVE 2 ADDITIONAL BACXUP EDGs FOR AFW BUT WITHOUT CHARGING, BATTERIES WILL LAST ONLY 30 MINUTE 8

IN ADDITION TO EDGs, SWEDISH PLANTS HAVE PERMANENTLY INSTALLED BACKUP GAS TURBINE GENERATORS THAT CAN IN 3 MINUTES 8

FRENCH HAVE ONE PORTABLE GAS TURBINE GENERATOR NUCLEAR UNITS AND FACILITIES FOR CONNECTING IN 2 HO 9

FRENCH HAVE INSTALLED ONE 140 KW STEAM TURBINE NUCLEAR UNIT TO:

PROVIDE SEAL INJECTION WATER IN PRIMARY SYSTEM CHARGE BATTERIES BUT WITHOUT CHARGING, BATTERIES WILL LAST ONLY ONE HOUR 8

FINNISH AND SWEDISH EDG UNRELIABILITY IS 1.2%

I

,,-,n-.-

1 2 0F 2 EMERGENCY DIESEL GENERATOR NOTES FROM OVERSEAS S

GERMANS REPORT THEIR UNRELIABILITY TO BE 0.8%

I FOR PLANNED STARTS, SWISS BRING EDGs TO SPEED OVER A 10 MINUTE PERIOD AND THEN LOAD OVER 4 MINUTES I

BELGIUM HAS DROPPED BACK TO 2 FAST STARTS PER YEAR 0

SWEDEN HAS BACKFIT A SOFT START DEVICE ONTO ITS NUCLEAR PLAN EDGs

-FOR PLANNED STARTS THEY ARE BROUGHT TO SPEED IN 30 SECONDS (VS. U.S. 10 SECONDS)

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- U.S. DIESEL GENERATOR' RELIABILITY SURVEY PRELIMINARY DATA 1983 AND 1984 67% UTILITIES

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UNRELIABILITY ~1.3%

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e EPRI ACTIVITIES ON REQUIRED EMERGENCY DIESEL GENERATOR START TIME BINDI CHEXAL NUCLEAR POWER DIVISION e

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CURRENT DIESEL START TIME REQUIREMENT AND ITS BASIS REQUIREMENT PER NRC REG. GUIDE 1.108. EMERGENCY DIESEL GENERATOR MUST REACH RATED SPEED AND. VOLTAGE WITHIN 10 SECONDS BASIS HYPOTHETICAL LARGE BREAK LOCA COINCIDENT LOSS OF 0FF-SITE POWER FAILURE OF ONE OF THE EDGs TO' SUPPLY POWER APPENDIX K CONSERVATISM

- - - - - - - - - - ' - - ^ - ' -

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l PROBLEM 5

L 5

FAST STARTS, FAST LOADING, AND A LARGE NUMBER OF TESTS

-s CONTRIBUTE TO STESS AND WEAR ON EMERGENCY DIESEL 3

GENERATORS.

FAST STARTING IS ACCEPTED WORLDWIDE AS A

]

MAJOR CAUSE OF EDG DEGRADATION.

]

NRC IS ENCOURAGING A REDUCTION IN THE NUMBER OF FAST 2

STARTS, E

[

OVERSEAS MOST COUNTRIES ARE SLOWING DOWN THE RATE AT q

WHICH THEY START AND LOAD THEIR DIESELS.

j ONE U.S. DIESEL MANUFACTURER HAS STATED THAT IN 1

NUCLEAR POWER PLANT SERVICE, A DIESEL'S WEAR IS

"]

DEPENDENT PRIMARILY ON THE NUMBER OF FAST STARTS.

]

eof HAS PERFORMED ANALYSES'THAT SHOW THAT FAST START i

MECHANICAL STRESSES ARE 21% HIGHER THAN THE STRESSES 3

AT RATED SPEED j

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RESOLUTION USE OF REALISTIC APPROACH TO ECCS LICENSING BASED ON N l

PROPOSED DOCUMENT SECY 83-472.

l AVAILABLE IN GE SAFER /GESTR LOCA PACKAGE LICENSED FOR L.

ON JET PUMP BWRS.

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CONCLUSIONS THIS STUDY HIGHLIGHTS THE USE OF A REALISTIC APPROACH TO ECCS LICENSING (ACCEPTED BY NRC) TO REDUCE DIESEL DEGRADATION AND HELP RESOLVE THE STATION BLACX0VT ISSUE.

OTHER AREAS WHERE OPERATIONAL FLEX 1BILITY CAN BE IMPROVED USING THE SAME APPROACH ARE:

IMPROVED FUEL UTILIZATION RELAXATION OF PERFORMANCE REQUIREMENTS IMPROVEMENTS IN TECH SPECS ALLOW PLANT POWER INCREASE e

O

c.

k LOSSES OF ALL OFF-SITE POWER

[-

AT U.S. NUCLEAR PLANTS

-1965-NUMBER OF EVENTS IN 63.9 TOTAL SITE LOSSES PER CALENDAR YEARS SITE YEAR Ia. Longer than 30 minutes 2

0.031 Ib. Less than 30 minutes 4

0.063 Total 6

0.094

-II.

0 III.

1 0.016 IV.

1 0.016 L

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LOSSES OF ALL OFF-SITE POWER AT U.S. NUCLEAR PLANTS Three Most Recent Years (1983, 1984, 1985)

NUMBER OF EVENTS IN 172.9 TOTAL SITE LOSSES PER~

CALENDAR YEARS SITE YEAR Ia. Longer than 30 minutes 4

0.023 Ib. Less than 30 minutes 9

0.052 Total 13 0.075 II.

3 0.017 III.

8 0.046 IV.

8 0.046 D

N

LOSSES OF ALL OFF-SITE POWER AT U.S. NUCLEAR' PLANTS

-ALL YEARS THROUGH 1985-NUMBER OF EVENTS IN 664.9 TOTAL SITE LOSSES PER-CALENDAR YEARS SITE YEAR Ia.. Longer than 30 minutes 25 0.038 Ib. Less than 30 minutes 30 0.045 Total 55 0.083 II.

7 0.011' III.

34 0.051 IV.

12 0.018

f 4

I LOSS ~0F ALL OFF-SITE POWER COMPARISONS

-LOSSES PER SITE YEAR-LATEST YEAR 1984 1985 la Langer than 30 minutes 0.000 0.031

-Ib-Less than 30 minutes 0.089 0.063 Total 0.089' O.094 3 MOST RECENT YEARS Thru Thru 1984 1985 la-0.013 0.023 lb 0.031 0.052 Total 0.044 0.075 ALL YEARS Thru Thru 1984 1985 la 0.038 0.038 lb 0.043 0.045 Total 0.081 0.083

t-E LOSS OF OFF-SITE POWER AT U.S. NUCLEAR PLANTS

.- ALL YEARS.THROUGH 1985 -

THE FIFTEEN OUTAGES LONGER THAN ONE HOUR Duration Plant hours: minutes Year Where caused by Severe' Weather Arkansas Nuclear One 1.29 1978 i

Dresden 4:00 1965 Tornadol Farley 2:45 1983 Ft.-St. Vrain 1:45 1983 Snow & Ice 2

Indian Point 6:28 1977 Lightning 1:45 1980 Lightning Millstone 5:00 1976 Hurricane 3

5:30 1985 Hurricane 4

1-Pilgrim 2:40 1977 Snow & Ice l

8:54 1978 Snow & Ice & Salt Spray l

Prairie Island 1:02 1980 i

St. Lucie 2:50 1977 i

Turkey Point 1:02 1977 2:00 1977 2:05 1985 1At the time of this event Oresden had 5 transmission lines all on one right-of-way. Today it has 12 transmission lines located on two right-of-ways.

2Line arrangement entering plant has been improved and it is now a technical specification that at least one of three gas turbines is operable.

3As a precautionary step, both units were shutdown in advance of peak storm conditions-units were shutdown when power was lost.

4Have equipment to wash insulators while energized.

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LOSS ~0F OFF-SITE POWER-AT U.S. NUCLEAR PLANTS

- ALL YEARS THROUGH 1985 -

~

SOME INTERESTING OBSERVATIONS.

Losses'of off-site power longer than:

0 minutes 27 of 65 sites 30 minutes 14 of 65 sites I hour 10 of 65 sites 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 7 of 65 sites 38 of 65 sites have never lost all off-site power.

51 of 65 sites have not had a loss longer than 30 minutes.

The median duration of all-losses is 1/2 hour.

There were only 4 losses of all off-site power longer than 30 minutes in the last 3 years.

.-r-r

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y

' WEATHER-RELATED LOSSES OF OFF-SITE POWER

- ALL YEARS THROUGH 1985 -

. - Total weather-related losses of off-site power and unit trip 15 events in 664.9 site years 0.023 events / site year Seven weather events have been substantial and 8 minor.

One of the weather-related losses occurred in 1985 and one in 1983. The others occurred a number of years ago: two in 1980, one in 1965, and the remainder in the 1970's.

The small number of weather losses is the result of correcting identified weaknesses.

Median duration for weather-related losses 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Longest duration 8:54 in 1978 Corrective actions are expected to prevent recurrence.

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-l EN90NEERONG 2 OPERA 700XS DEPART'M ENT' NRC REVIEW DEPARTMENT DIRECTOR: A. RUBIO FEBRUARY 27,1986

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

m o

ENG0NEEADMG is CPERA700XS DEMT, NUCLEAR PLANT LIFE EXTENSION & CONSTRUCTIBILITY PLANT AVAILABILITY LOW LEVEL WASTE & COOLANT TECHNOLOGY 4

MCJC:. EAR P:.AXT :.0FE EXTENSDON AND CONS 7EUC70B0:.07Y OBJECTIVES:

- EXTEND OPERATING LIFETIME BEYOND 40 YEAR LICENSED TERM

- REDUCE CONSTRUCTION COSTS FOR OPERATING PLANT MODIFICATIONS

- ESTABLISH AND MAINTAIN EQUIPMENT QUAllFICATION

NUCLEAR PLANT LDFE EXTENSDON 1866 %BY ACCOM P;.08M MENTS SURRY-1 AND MONTICELLO PILOT PROJECTS:

- DETAILED ASSESSMENT OF MAJOR COMPONENTS, SYSTEMS &

STRUCTURES

- IDENTIFICATION OF UTILITY ACTIONS REQUIRED; MAINTENANCE, RECORDS, MATERIAL SAMPLING PLANS

- LIFE EXTENSION COSTS; MAINTENANCE, REPLACEMENTS, REPAIRS, INSPECTIONS

- LIFE EXTENSION IMPLEMENTATION & MANAGEMENT PLAN STATUS: ~ 50% COMPLETE

NUCLEAR PLANT CONSTRUCTOBOLDTV ISSUE:

- EXCESSIVE PLANT CONSTRUCTION COSTS; NEW T

CONSTRUCTION-OPERATING PLANT MODIFICATIONS SOLUTIONS (PARTIA'L):

- RESOLUTION OF KEY QUALITY ASSURANCE ISSUES

- INCREASED APPLICATION OF COMPUTER-AIDED ENGINEERING; IMPROVE PLANT INFORMATION MANAGEMENT

- IMPROVED DESIGN / CONSTRUCTION PRACTICES N

MUCLEAR PLANT C0fMSTRUCTOBOLUTV 1965 %EY ACCOM P.0$M MEXTS DEVELOPED INTERIM GUIDELINES FOR APPLICATION OF COMPUTER-AIDED ENGINEERING COMPLETED MIDLAND 5-D CONSTRUCTION COMPUTER MODEL EPR~l TO PROVIDE PROGRAM MANAGEMENT FOR NUCLEAR CONSTRUCTION ISSUES GROUP (NClG)

CONDUCTED WORKSHOP ON PLANT LAYUP & EQUIPMENT PRESERVATION 4

?

!XTERDM GUDDHLDX HS FOR CAI APFLDCATOX DESDGE/00XSTRUCTD@X/0PERATD0X$

NEED : REDUCE COSTS

- AUTOMATE PLANT WORK ACTIVITIES

- IMPROVED PLANT INFORMATION ACCESS AND CONTROL SOLUTION: COMPUTER-AIDED ENGINEERING

- INTERACTIVE COMPUTER APPLICATIONS

- COMMON SUPPORTING PLANT DATA BASE FURTHER NEEDS:

- SYSTEMATIC DEFINITION OF PLANT DATA STRUCTURE

- GUIDELINES FOR USEAGE

- INDUSTRY ACCEPTANCE N

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EGUD P M EN T @UA LD FDCA700N - E N VD RON M EM TAL

%BY 1925 ACCOM P:.08}iMENTS

- PUBLISHED GUIDELINES / DATA BANK ON QUALIFICATION OF MECHANICAL EQUIPMENT

- CONDUCTED SEMINARS ON MAINTAINING EQUIPMENT QUALIFICATION

- AGING SPECIMENS (CABLE, DEVICES, LUBRICANTS)

INSTALLED IN 7 PLANTS O

em.-.

EQUDPXEXT QUALOFOCA700X DEGRADED CORE

- PUBLISHED EQUIPMENT DATA FROM FULL SCALE HYDROGEN BURN TESTS AT NTS

- REPORT REVIEWED BY NRC SENIOR ADVISORY RGVIEW PANEL PRI0R TO PUBLICATION NRC TO ISSUE POSITION ON RULEMAKING FOR LARGE DRY CONTAINMENTS BY MID-1986 i

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ROW L:EVEL WAS'FE

.a WASTE MINIMIZATION:

DRY WASTE REDUCTION WET WASTE REDUCTION ADVANCED VOLUME REDUCTION TECHN.

REGULATORY CONCERNS / IMPACT:

MONITORING DISPOSAL TECHNOLOGY STABILIZATION & STORAGE

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aOW EEVEL WASTE TOPIC:

Economics of Radwaste Volume Reduction. UTILITY NEED: Ability to perform side by side cost comparisons of available in plant and mobile volume reduction technology. APPROACH: Develop user friendly IBM compatible program to perform radwaste volume reduction economic analysis. e 0 _ __ -,_.w.., -,y- , _ _ - _.__. _. __ _ _ _r._-,,-_-._

~

OW lEEV

EL WASTE TOPIC:

Radwaste Package Direct Assay Tech. N UTILITY NEED: Capability to accurately identify and quantify radionuclides in the waste burial container. APPROACH: Develop and demonstrate methods which would perform direct gamma & TRU radionuclide assay of waste containers. 4 e ..,p.,._,..w. .__._-,__....____,.__,,,-g7. m.7-.-_,y,y7__

~ LOW :GVEL WASTE TOPIC: LIW Disposal Technologies UTILITY NEED: Ability to influence & respond to state disposal agencies' alternatives to shallow land burial. APPROACH: Develop generic design. Assess cost and performance features for disposal technology alternatives. O f I W .,---,v w_, ,,,,_.w. _--_-__y. y...__,_.._

COOLANT TEO]Sn40 LOGY CHEMICAL CONTROL: ADVANCED MONITORING TECHNIQUES WATER QUALITY IMPROVEMENTS PRIMARY CHEMISTRY GUIDELINES RADIATION CONTROL: ZINC INJECTION PASSIVATION SURFACE MODIFICATION PREFILMING COBALT REPLACEMENT LOMI DECONTAMINATION i l e 0

SULFATE VALUES FOR S/G A JLINE 19,1984 THROLIGH JUNE 21,1984 .W ,P n \\ de, l l .= i I w fr E~ to 25 30 35 40 45 50 35 40 45 70 ?$ SO 35

• 0 f10TEt AT 32. 64. AND 31 HOUR.

P C T3 &ERE PLACED trf 3ERV!CE I 4 L SULFATE VALUES FOR S/G A NOVEMBER 8,1984 THROUGH NOVEMBER 11,1984 W SS-I P IM S &_;^,-- : 0, 9* 10 E0 30 40 50 40 70 60 80 100 s f tt'E IN HCWS it0TE: af SJ HoLits pro al H')JR3. PIE!J P8ECCAT3 8.EPE PLHCED fft jlSy[CE

COOLGTT TECENOLOGY TOPIC: Water Quality Improvements UTILITY NEED: Water Chemistry controlled particularly for impurities. APPROACH: Develop robust configurations for Condensate Polishers. Provide generic cost / benefit evaluations for retrofit options. I

COOLANT TECHNOLOGY TOPIC: PWR Primary Water Chemistry Guidelines UTILITY NEED: Guidelines have proved a powerful aid in improving plant water chemistry. APPROACH: Committee of 12 Utility & 3 NSSS vendor Representatives prepared guidelines using format of PWR Secondary and BWR water chemistry guidelines. O O

P k [ V I f [ t 2.0 I 1.8 1.6 NWC g 1.4 8 S"1.2 HWC 1.0

• 0.8 8

0.6 l 0.4 f NWC + Zn 0.2 0 o,o t I O A l l l 1 l 0 0.4 0.8 1.2 1.6 2.0 2.4 Exposure Time (10 h) 3 The effect of zine on cobatt. deposition on 316 SS In normal water chemistry and hydrogen water chemistry.

==m.

COOEANT ".fECHNOLOGT1 TOPIC: Surface Modification /Co-60 Reduction UTILITY NEED: Reduce Radiation Field buildup on BWR RCS piping and SG Channel Heads. APPROACH: Field test electropolishing techniques and demonstrate no adverse effects. 5 P

COOLANT TECEE40 LOGY TOPIC: Prefilming of BWR Piping. UTILITY NEED: Reduce rapid contamination of new plants & replaced piping in old plants. APPROACH: Demonstrate benefit of prefilming to give adherent oxide film on surfaces.

COOLANT TECHNOLOGY TOPIC: Cobalt Replacement UTILITY NEED: Minimize cobalt release by wear and corrosion to reduce inventory available for activation to Co-60. APPROACH: Qualify cobalt-free wear-resistant alloys for use in components and as a substitute for cobalt-base hardfacing alloys. i

COOLANT TECPMOLOGY TOPIC: LOMI Chemien1 Decontamination Process UTILITY NEED: Low cost, effective, corrosion-free decontamination method. APPROACH: Demonstrated no corrosion problems. LOMI modified to reduce radwaste generated.

PWE EADEATION MELD CONTEOL TEC15E6iiEQUES

~ BEFORE POWER Electropolish channel heads. Extended ASCENSION: hot functional tests with good water quality. DURING pH Control. Hydrogen Peroxide OPERATION: addition at shutdown. REFUELING: Use Zircaloy grids in replacment fuel. MAINTENANCE: Replace wearing valves with cobalt-free alternatives (especially CVCS flow -controllers). Valve Maintenance procedures to remove debris. SPEC. REPAIRS Chemical decontamination. & MAINTENANCE: Electropolish channel heads. Use low-cobalt for replacement tubing. PLANT LIFE Complete primary system EXTENSION: decontamination. l en l

ETiB EADHTION MELD CONTEOL TECEMZQUES BEFORE POWER ASCENSION: RepInce control blade cobalt alloys. Control oxygen at 200-400-ppb during hot functional tests. DURING Minimize feedwater iron input. OPERATION: Maintain low reactor water conductivity Zine Injection Process. REFUELING: Low cobalt materials in replacement fuel. Cobalt-free control blades. MAINTENANCE: Replace wearing valves with cobalt-free alternatives (es'pecially feedwater flow controllers). Valve Maintenance procedures to remove debris. SPEC. REPAIRS Chemical decontamination. & MAINTENANCE: Electropolish replacement piping. Air oxidation or water prefilming of pipe. PIANT LIFE Complete Re tetor Coolant System EXTENSION: decontamin ation. ~ -g

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}fe R, Ae Ce WOEESEO:PS/SEMINAES/homETENGS

• Instrument Improvement - March
• Computer Modeling, Data Management, and Video Applications - June
• Dose Reduction and Personi1el

~ Factors - September

• Bolting and Bolted.

Connections - November

O Ee 2 A C. '2EO.MO: LOGY MANSFEB

• Improved BWR CRD Handling System Design Review / Bid Evaluation
• Underwater Inspection Vehicle Demonstrated at Catawba Four Utilities Interested
• Surveyor Mobile Robot Lab tests / Field Hardening Demonstrated at ICUEE Show t

O e

E,1 Ae Ce "1.sufiEMO: LOGY ".:2ANSFE:E(Cont)

• Surbot Mobile Robot Simulated-Plant Evaluation
• Training Courseware 3 Video Tapes - Pressure Boundary Bolting Available from MEAC i

l f

EUMAN ENGETBEMEG:MAETfAliNAEEltTg GUIDEMNES UTILITY NEED Incorporate improved maintenance as a plant design goal SOLUTION . Develop human engineering guidelines for designing new or backfitting existing plants STATUS Guidelines available (NP-4350) i e n

EsA"F e'. MESS MAEAGEMEEF :PEOGEAM GbroWLiWES UTILITY NEED Worker productivity, health 8c safety challenged by conditions leading to heat ~ stress SOLUTION Develop comprehensive guidelines for managing heat stress STATUS Guideline available (NP-4453) ~

DL4 GEOM 20/TEOUELESECOTETG ein0L48 SO3TWKM UTILITY NEED Improve the diagnostic and troubleshooting skills of plant personnel SOLUTION Develop and evaluate prototype program for training these skills STATUS Diesel generator course. ware tested w g7- .,.-m_ rm. 9 9, r 7,-_.

PEEF01BEETm 3EFA AC@MenTHON GUla3ErMES w

UTILITY NEED Insufficient data to locate chronic " lost megawatts" SOLUTION Develop guidelines for a structured approach to heat rate increase diagnosis STATUS Guidelines available (NP-3915) 4

MOBII,E SUEvts1RAMCE ECEOT COW &w,Eun :EATEDWAIRE UTILITY NEED Reduce human exposure to hostile environments for radiation, visual, and environmental surveying SOLUTION 4 Develop untethered, remotely operated robotic system STATUS SURVEYOR commercially available e

V. M. O. EIDP:E0VElG NTS Commemf*I hmd.wan UTILITY NEED Failure of motor operated valves is a major cause of abnormal events SOLUTION Develop microprocessor control and diagnostic system STATUS Prototype tested. Field test ongoing e n.-..

) VALVE :PAcln0RTG EWEOVEMENTS CommemfLai FamiWan T UTILITY NEED Packing leaks increase maintenance costs and radiation levels SOLUTION Develop improved graphite packing and guidelines for " live loading" STATUS Prototype tested. Field test ongoing

.n0 ELD EMDEHD EenmisUMENTS Guid.eHmes Be CommercimL E:mreiware. UTILITY NEED Instrument & electrical events cause plant trips & 30% of LERs SOLUTION Develop Guidelines Demonstrate with hardware STATUS Guidelines developed. Module tested a ,n w-n-.n

,NL40E01R12Y VJ0GE.ITOE MORfHTOEIDhTG GU.ID:ELIHS UTILITY NEED Assistance in evaluating plant monitoring program alternatives SOLUTION Develop Guidelines giving technical and skills requirements and corporate goals STATUS Interim Guidelines (NP-4346) 4 I

ENG0XEERONG ss OPERA 700XS DEPT. CURRENT R&D EMPHASIS - PLANT LIFE EXTENSION - SURFACE ELECTR0 POLISHING & PASSIVATION - DECONTAMINATION - ROBOTICS DEVICES --- RADIATION CONTROL - MAINTAINABILITY GUIDELINES - EQUIPMENT QUALIFICATION - COMPUTER GENERATED DISPLAYS - COMPUTER AIDED CONSTRUCTION O

.e e SAFETY TECHNOLOGY DEPARTMENT W. B. L0EWENSTEIN RISK ASSESSMENT

1. WALL SOURCE TERM R. V0 GEL ANALYTICAL METHODS & VERIFICATION R. BREEN SAFETY CONTROL & TESTING R. DUFFEY

-WBL:AV/3932ST6s

PROGRAM 1310 RISK ASSESSMENT l WALL PROBABILISTIC ANALYSIS D. WORLEDGE & APPLICATION PLANT & GE0 TECHNICAL ENGINEERING R. KASSAWARA CIVIL ENGINEERING H. TANG SEISMOLOGY C. STEPP WBL:AV/3932ST68

l. PROGRAM 1315 SOURCE TERM R. V0 GEL AEROSOL TECHNOLOGY F. RAHN LARGE SCALE EXPERIMENTS FISSION PRODUCT. BEHAVIOR R. RITZMAN TMI-2 DATA USE G. THOMAS CODE DEVELOPMENT. BWR POWER LEVEL R. SEHGAL IDCOR COORDINATION M. LEVERETT WBL:AV/3932ST68

PROGRAM 1320 ANALYTICAL METHODS & VERIFICATION R. BREEN REACTOR PERFORMANCE

0. OZER THERMAL PERFORMANCE L. AGEE

. METHODS SUPPORT & ENHANCEMENT R. BREEN WBL:AV/3932ST68

e-t- P L PROGRAM 1330 SAFETY CONTROL & TESTING R. DUFFEY SAFETY CONTROL SYSTEMS B. SUN SAFETY MARGINS & TESTING J.-P. SURSOCK WBL:AV/3932ST68

m = SAFETY TECHNOLOGY OVERVIEW ? SAFETY RESEARCH IN TRANSITION L FROM: ACCIDENT DESCRIPTION e TO: ACCIDENT PREVENTION & ^ ACCIDENT ACCOMODATION VIA: CONSOLIDATION & CONVERGENCE OF RESULTS w E r E I t E WBL:AV/3999ST60

SAFETY TECHNOLOGY PRIORITIES BASE PROGRAM SOURCE TERM SEISHIC SAFETY CONTROL TESTING TO SUPPORT ANALYSIS DISCIPLINED SOFTWARE i l WBL:AV/39995T6D

SAFETY TECHNOLOGY OVERVIEW HIGHLIGHTS FOR 1985 (1) SOURCE TERM e MAJOR LARGE TEST PROGRAMS WINDING DOWN EXPL0lTATION & CONSOLIDATION OF RESULTS e e IDCOR DIALOGUE ESTABLISHED e IDCOR METHODOLOGY VERIFICATION e OVERALL INTEGRAT100N IS NEW FOCUS SElSMIC e SEISMICITY OWNERS GROUP PROCEEDING WELL e SEISMIC MARGINS IN HAND e LOTUNG TEST READY e PIPING RESPONSE WORK DEPLOYED ~ l

SAFETY TECHNOLOGY OVERVIEW HIGHLIGHTS FOR 1985 (II) SAFETY CONTROL & TESTING e MIST PROJECT PROCEEDING WELL l e SIMULATOR QUAllFICATION GROWTH e Al SPECULATIONS PROMISING e OPERATION AIDS BEING TESTED & DEPLOYED e DIGITAL FEEDWATER CONTROL ON SITE e SGTR CLOSURE PROMISED 1 PRA e APPLICATIONS FOR RELIABILITY WITH GROWING INTEREST e COMMON MODE STUDY PROMISING e HUMAN RELIABILITY MODELING PROGRESS SOFTWARE e MMS COMMERCIAllZATION SUCCESSFUL e VIPRE-SER BEING DEVELOPED e RETRAN CONSORTIUM INITIATED e ATHOS/PORTH0S VERY SUCCESSFUL e RASP EVOLUTION CCNTINUING e TECHNOLOGY TRANSFER IN CORE MONITORING e

SAFETY TECHNOLOGY OVERVIEW EMPHASIS FOR 1986 IS ON DEPLOYMENT & CLOSURE SOURCE TERM CONSOLIDATION & CLOSURE ^ PROBABILITICS FOCUS FOR CLOSURE SEISMIC ESSENTIAL COMPLETION OF OWNERS' GROUP SCOPE METHODOLOGY FOR MARGINS DEVELOPED CODES & STANDARDS WORK ESSENTIAL PRA DEPLOY APPLICATIONS FOR RELIABILITY EVALUATIONS SOFTWARE DE-EMPHASIZE ENHANCEMENT & MAINTENANCE EMPHASIZE A) COMMERCIALIZATION AND/OR B) USER SUPPORTED ACTivlTIES SAFETY CONTROL & TESTING EMPHASIZE DEPLOYMENT & USE OF PRODUCTS EXPLORE DIGITAL TECHNOLOGY FOR INTEGRATED SYSTEMS MIST SAFETY DEFINITION

COMPLETION / CLOSURE i r------------------------------------------- I I I I I I l I I I I I I B I I l I I I i l 8 I I I I I I I I I l I I I I i l ORIGINAL I i 8 8 l l SCOPE l l 1 I l I I I 1 l l POTENTI AL FOR ENHANCED SCOPE l l I I I l 1 I l DEVELOPMENT FOR ASSURING ENHANCED l 8 SCOPE I I I l POTENTIAL FOR RE-ENHANCED SCOPE B I i.___ DEVELOPMENT FOR ASSURING RE-ENHANCED SCOPE

SAFETY TECHNOLOGY OVERVIEW 1986 THRUSTS DEPLOYMENT OF RESEARCH RESULTS SOURCE TERM ACTION SEISMICITY RESOLUTION CODES & STANDARDS ON STRUCTURAL RESPONSE SOFTWARE COMMERCIALIZATION OR USER SUPPORTED SOFTWARE ACTIVITIES INTEGRATION OF PRA ON MAJOR ISSUES (E.G. SOURCE TERM) DIGITAL SYSTEMS IMPLEMENTATION CLOSURE ON B&W SAFETY TESTING WBL:AV/3999ST6D

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RAFT CODE VERSUS EXPERIMENTS AEROSOL TOTAL MASS AEROSOL-TEST PARTICLE SIZE (uM) DEPOSITED (AG) gg NUMBER CODE MEASUREMENT CALCULATION MEASUREMENT MARVIKEN 2A -5 -6 8.4 3.7 1 -7 -6 14.5 18.,0 2B -9 18.1 37.1 7 -11 31.3 22.5 % DEPOSITION CALCULATION MEASUREMENT ANL HOT 2 -0.2 0.2-1 20 18 TUBE 4 -0.8 0.1-1 11 13 AEROSOL CONCENTRATION (PART/CM3) CALCULATION MEASUREMENT 10 -107 STEP 1 -0.5 -0.2 107 6 RLR3997ST5A

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h. I+ 4 \\ ) rs EPRD SYSTEMS & MATERDALS DEPARTMENT KARL E. STAHLECPF 1 I L I I. s..

Systems & Materials Depastnient K. E. StaNkopf I l l l mR Fuel & Spent Component RehaWisty Corrosion Control Advanced LWR 6A'22 SPE 23 spp pg D Fra%hn* G.Dau* H Jones *- D.NaNe* 3051 Entended fuel Life 3025 NDE Center 3c38 Ptars Materials 3068 Regulasory Statulizasion

• S. gen
• G. Dau*
• W thlanen*
• J Gdman 3052 FuelRe' tmhey 3026 Inspection & Detection

- L. Nelson

• 3062 Utddy Requirements a

+ D. Frankhn*

• M. Sehravesh

- D Ca h mati p %g - S. Liu -W Civids 3053 Basse Fuel Research - M Avioll 3063 SmallLWRs

• A. Mactuels 3033 Hydrogen Water 3027 Structural hasws Chenustry
• W Sugnet 3054 Core Components
• D. Horns

+ R Jones

• 3064 Amernatsve Concepts
• J Santucci

- S. Tagant - L. Nelson *

• D. Notde*

- T. Gnesbach 3022 Spene FuelStorage . H Wdhams - ft. Larnbest. +R.L e

• BWR Owners Group R Jones
• i L _.

W SySlemS and Materials Department Structure

• Dual Capacity

X:@X BURXUP FOR LWfR FUEL 03JECTHVE o DIVELOP A VECHOSICAL BA292 F02 U2]M9 A%D LICEM23 DSS STAMDARD-DI2!@% LM FUEL TO HIGH EURMUPS O a

F - ----~~---- - - - - ~ ~ - - - ~ - - ~ - - - - - - - - - - - - - - ~. en X0@X BURXUP FOR LWR FUEL ((ccat) BENEFIT l o REDUCED FUEL-CYCLE CCSTS u o ALLOWS ECCMCMICAL IMPLEMENTATICX OF LONS CYCLES. WHECH IMPROVE CAPACITY FACTORD o EPRI CCDITRACTCES (C-E AND W) CBTAIMED MEC APPROVAL CD1 HiSH-BUMUP TCPDCAL REPORTS BASED CM EPRS DATA j

MUGX BUEXUP F@M LWM FUEL (cont) STATUS e EPR] C@MPLETED LA27 0F SURVEILLAMCE PR@@ RAM 2 EV COMPLEi!M3 MOT CELL EEAMSMA700%2 0F PWR FUEL AFTER 55 @Wd/9 o LIESTED W@%E CM SELECTiiD PHEMGMEMA CDS@@iM2: - ZSICALCY CO2EC200M - FE2230%-SA2 RELEASE

6 a a e 4 en e G G E 10 Original NRC re# ease estirnate ( i 5 l Measured re4 ease from standarc PWR fuel O l 0 W E 2 4 M M 1 Sumus (GWdttu) ~ t l l l

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p E@H BURMUP FOR LWR FURL (ccen.) APPLICATICM e 522EMTIALLY ALL U.S. PWAS EFTEMDiXS BUnMUP e ABOUT MALF CF U.S. BM20 EXTEMDIMS 202MUP e SALTIMC23 SA2 2: ELECTR!C DCCUMEMT2 "F1227-023" TOTAL 2222F172 CF $42 MILLICM UP TC 1EEE FC2 EACM CP THE TEC CALVE 27 CLIFF 2 UMET2, BASED CDS THE RW2MLT2 CF RPE@$-1 4

d g a EBC@RE FUIL-PIMFDMMANCE CODE OBJECTEVE e PROVIDI UTELETEEE WITH CAPAEILITY TO - EVALUATE FUEL BE2i3MS - SUPPORT LECEMalMS OF RELOADS

ESC @RE FUBL PERFORMANCE CODH ((ccat) STATUS o PRERELEASE VER250M RELEASED IM 1988 o FiMAL VER2iCM COMPLETED: TO BE RELEA8ED BY SCFTWARE CENTER MARCH 31,1952. o UTILITY GR@UP FOR RESULATORY ACTICM (USRA) EEIN3 FORMED 70 SUPPORT SUBMITTAL TO MRC O

ESCORE VERSION 1 ~ PREDICTED VERSUS MEASURED TEMPERATURES ~ 3.5 a J 8 3.0 g a g g 2.0 a",k[r 1.5 a o j g a y 1.0 A I o 0.5 1 0 O 1 2 3 MEASURED, *F (Thousands)

ESC @RE APPLICAT00% o 15 UTIL! TIES CBTAEXED PRERELEASE VERS CM o 2@ UTIL3 tie 2 EXTERESTED IN FOXAL VEM20CM AMD GIMERAL U2IR2 GROUP o 11 UTILITIES INTEND TO JOEM USRA o 1 UTILITY 20HEDULED TO USE E200RE EM 1253 EM MMO LICE %2]M9 APPLICATSCM I i e v

axsprevaox me envaevaox (xen me) OBJECTIVE DEVELOP AND TRANSFER TO FIELD PRACTICE QUALIFIED, IMPROVED NDE EQUIPMENT AND PROCEDURES SCCPE o DETECTION AND SIZINS OF ISSCC IN BWR PIPING o PWR PIPENS o VESSELINSPECTION o STEAM SENERATOR INSPECTION o BOLTINS i o TUR!MME t w -w- -,-- --, - -----,,, -w,- + e-

POP 0X@ AXD FDTT0X@ DYNAMDC MLDAB0LOTY OBJECTIVE o TO DEVELOP AM IMPROVED, REALISTIC, AND DEFENSIBLE SET OF DESIGN RULES WHICH CAN BE ADOPTED BY THE ASME CODE FOR TREATMENT OF DYNAMIC PLUS STATIC PIPING LCADS a M U

R:PDN@ AXD A1nr0M@ DYMAKC FMELDAK STATUS o PROJECT STARTED MARCH 15,1955 AND EXPECTED TO LAST THREE YEARS o THREE TYPES CF TEST 8 PLAMNED; MATERIALS. PIPE COMPONENTS, PIPIMS SYSTEMS o AS OF JANUARY 1,1988, 8 0F 48 COMPCNENT TESTS ARE COMPLETE h

noa e e A c C ~ 9 0 s 0 a,z b

: m* +

g w l ~ r> l+- \\' T o..; 1 0 i- / q "" > 3 /:2 ( "b t! s 3e ^ Al / '11!**4 .5 E H_ tilii l i [ r LlH g C5 .-IS * * << o 8 d _>cs g, 3 g~o( O Y Y 0 _**5 4-E -* 3o / a R32 om y22 12 'W w y E v) 0 m W. C 0 0o ,,,,,5 ,0 e_( o o ,,3p9 oo -y o .=.

i 1 PDP:NG AND FTT NG DYNAM:C RELDAB:LTY SIGNIFICANCE o EARLY COMPONENT TEST RESULTS SUGSEST EXCESSIVE MARSON OF 15 TO 25 TIMES SSE REQUIRED FOR PIPE FAILURE (MARSIN OF 1.5 IS CODE BASIS) o ALLOWABLE STRESS CAN DE SIGNIFICANTLY INCREASED TO REDUCI USE CF SMUBBERS o SIMPLIFIED DESISM METHODS MAY BE FEASIBLE - - - -,. -., - -,..., - - ~. - -. - -... -.. - - - - - -

m t PRESSUR ZHD TMIMMAL SKOCX CBJECTIVE o TO CCRRECT ELASTIC MODELS FOR CRACK A* REST IN THE DUCTILE REGICMS CF NUCLEAR PRESSURE VESSELS O M M

^ PRES $UROZED THERMAL SM@CM STATUS o THREE YEAR PROGRAM COMPLETED DECEMBER 31,1985 - $1.4M o OUR EXPERIMENTS HAVE SHOWN 5E% HIGHER CRACK ARREST TOUGHMESS THAN THAT USED FOR NRC CRITERIA DEVELOPMENT l i m

PNESSUR ZHD THIRMAL SMCCM SISMIFICAMCE o ADDITIONAL LIFE FOR NEUTRON DAMAGED VESSEL STEEL e.e O

PMEDUC700X @F CRACX @R@WirX 0X MIACTCM PRISSURI VHSSIL (MPV) STEEL 2 EACE@RCUD0D e RELDAELE PREDUCTCC00 CF RPW FG.AM EENAVi@R 12 A90 DEPCRTAGST ELEME907 CF PLAS07 2AFETV AMC LEFE EIITES022@C0 EVALUATi@SS2 o LWR SERWDCE E90ViR000MED072 ACCELERATE CRAC% @mCMVM RATE 2 090 RPV MATERIALS 9

4 i l ? / Linearinterpolation is i recommended to account for 7 / / ratio dependence of water Pp/ S 10-2 environment curves, for 0.25 < R < 0.65 for shallow sloce:

• da/dN' = (2.13 x 109 C: AK'8 Q = 3.75R + 0.06
• O

/ 4 ~ R = K,n/Km, ni /y m N l 'k E / Determine the AK at which P ^ the law changes by S calculation of the / + Z intersection of the two / ,g cumes 10~3 Surface flaws b 4/ 3 (water reactor d envircnment) f Subsurface flaws applicable for (air environment) R S 0.25 - / i' da/dN = .0 4.77 x 10~'O x AKs.72s -] 8 0.25 < R < 0.65 - j R 2: 0.65 d R = K gK*" / I x Linearinterpolation is m 9 recommended to account .f f ~/ '~6 l for R ratio dependence M of water environment 9 d 10 f* curves, for 0.25 < R < 0.65 g for steep slope: u6 [ _k da/dN = (1.48.10~")C, AKsm f Q, = 26.9 R - 5.725 y R = K gK. y/ m m i i i i i il, i i ,,,,,,i 10 10' 102 l Stress lntensity Factor Range (AK MPa O) i sm s Figure 4. Revised Appendix A crack growth reference curves from the 19CO addenda. (Si edition) l

• P

-y-V$ '5g-se --g% -e-e, -M---

• + + *

= E i ? PRID:CT00X @F @ RACE @R@WTX DX 1 MIACTOM PAISSURH VHSSIL (MPV) STIMLS F EACE22@UMD (seaW) 6 e @@MM@@ECM FAT 0@ME CRACE @%@TJVM MA@ EEEM @7030ED ? EETEM20VELY DUROM@ THE PA@T MWE YEAR @ 5 e EELEAELE FRED CTDVE EETM@@@ AME E@ta EEDE@ E DEVEL@ PED A%D PE@@@@2D F@2 ADCPTD@M BY 0%DU@TEV@@DE@ L

PREDUCT:0X @F CRACX @R@W7M DX REACTOR PRES $URE VHSSEL (MPV) STHILS STATUS .o SYSTEMATIC EEASUMEEIMTS OF CMC % SECMTM RATES MAVE BEEDS MA@E FORE A WDCE RASC@E OF RELEVAMT CCG0DITICM@. METALLUR@DCAL VARIABLE 2, EMWORCCG-EEGOTAL C090@070C002,27252222. A00@ LCADGBS@ EATE2 ALL HAVE IDSPCMTA007 EFFECT2 CGS CRACI 92@ETH RATES o EPEI DATA AMD CTHER IBSDU2TRY CATA MAVE EEE90 C@LLECTED AMD MADE ACCE22EELE TM105@M EEEAC (EPRI DATABASE FOR E00VDE@%EEMYALLY A22E2TED c CIACEDES) l

PRED:C70@X @F CHACX @ROUTM DX RIACTOR PRESSURE VHSSEL (RPV) ST!3LS 27ATU2(scr.Wd) e PAMALLEL RESERACM HAS DEFGMED @@%2@@0@% C2ACX0%@ MECDMM02M2 AMD DEVELC@ED PHY20CALLY- @A2ED MCDEL2 @F THE PR@@E2222 A e ALL RAW DATA ARE @EIMS REAGUALYZED 150 LE@M7 @F THE FMV20CALLY-@ASED M@ EEL 2 U20%@ C@MPUTER T@@l2 @EWELC@ED FOR TM!@ PUEF@@E N l

SA53381,0.025%S,288'C BWR WATER, R = 0.7,0.8 10' E ASME XI R > 0.65 U ~ 10 2 y_ = h 0 O

s. 10' go 3

\\__ ASME XI air 7 4 T 7 I O E O 1. A 10" :::c 0.1 v 5 0.03 5 0 01 Q ~ 0.001 0 _C 0.00025 0 10 5 l l i i t i i i 10 100 AK (MPa /?ii) g-s 10** Y 10** 1 6 10"' 1 0 f.i eI 9 10-, at 8 sa o so.. ~,......! ,.,,,,, il ........I i......I 10-' 10 10-' 10-' 10** 10'8 Base Mate. E.2 x 10-'aK88 Frecuency (mm/s)

=

N J

4 PREDUC70@X @F CRACX @M@MTX OX MIACTOM PRESSURI VHSSEL (MPV) STIUi!L2 FLA%@/UTELEZA70@G6 e @ATA AMALV@0@ MAS RE@ULTED DM A FR@F@@ED REYD@0@DS T@ THE A@ME @@@E CROTERBA FCR EVALUATBCC0 @F FLAN @ EEF@@2D T@ ADR e A FUDS @AM2007 ALLY MEN EVALUATI@M LSETMC@, TERME @ "73ME @@MABCS ADSALY20@" MA2 RE@ULTED FRC00 THE FMY@3CALLY-EA@ED MO@2L @F THE @@RAC@C@Cd CRA@%DGS@ PR@@22@ 100 UATER. REAGCALV@E2 @F ALL @ATA EM THE "TEME D@MAGGS"l@ DM FR@@RE22. TREAL AFFLB@AT0@SC@ MAYE E2200 @@GS@UCTED u 4 ..---.n-, ,a n-

e = -1 PRRDBCT@N @F CRAC% @R@WnrM BN 3 MBACTOM PRESSURI VHSSEL (MPV) 27I2LS f?.? PLAZ@/UTELEZA70@GS (eed'd) LP:.:.1 cls e m@2ARCE c@MYaus@ 000 THE EFFECT@ CF EETAL-M LUR@CCAL @ULFUR @95 CRACE @%@MVM RATE @. HU@MER . :l;. RAVE 2 Am A2@@@DATE3 M97H MU@MEE @ULF22 LEV 2i@, i TVPCCAL CF DCE22VCC REACT @R PR0!@@UME V3@@2L2 i ~ N! @@ @L@22 MAMUFACTURE. DC@CMGM000A72@ CRSTEM0A i V Am zucas to Av@ac UzcUm ps99AL71s@ m FLAW i EVALUA70@G02. 1 if.: i w; $'.Tk ? dE: L-A = y.:;1? 4 .y.,.. 4

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f 1 b h i XYDACGHX WATHA CX HMBSTRY v 3 i BACXGROUND h o INTERGRANULAR STRESS CORRCSICE CRACXiNG (OGBCC) 0F EWR PIPING MAS HAD SIGN 1FICANT IMPACT ON PLANT AVAILABILITY i o NUMERCUS REMEDIES HAVE BiEi!N DEVELOPED-ALTERNATE ALLOYS AND STRESS IMPROVEMENT TREATMENTS W c LAECRATORY STUDIES AND LIMITED IN-REACTOR TESTS w HAVE INDICATED THAT ORYGEN SUPPRESSION V!A HYDROGEN INJECTION CAN PREVENT iGSCC AS LONG AS WATER PURSTY la HIGH ? Y b E ~ E E E

i l l l l oxygen comew or conomen poesnnes ab l iner no l Normai '8# s w a range - E i icsec l 90**** l l ram own w m _ a W^ rat. IGsCC Nfy rooma P*acamary system range Irnourtry concentraten or concuctrwry Figure 2-1 Schematic Surmary of the Results of Laboratory Studies of Effect of Impurities on SCC of Sensitized Austenttic Stainless Steels

l ) ) i ~ \\ I. 1 l DEPENDENCE OF IGSCC IN SENSITIZED TYPE 304 STAINLESS STEEL ON ECP Crack Growth Rate (mm/s) 10~5 Fitzpatrick o a y a CERT C Dresden-2 C o 4 CERT 10 + C e Lab CERT U C Ringhals 1 l 0

• U CERT C

Forsmark 1 NoIGSCC Crack Growth I below -0.23 Dresden 2 { O Crack Growth 10'7 3 p 1w, e - 0.4 -0.2 0 0.2 0.4

• "ff' Corrosion Potential (v,she) ems.n e.=

e

l h XYDRG@IX WATER dXEXB2TRY BACXGROUND (cont'd) EPRI PROJECT WITH CCHMONWEALTH EDISCN CC., GENERAL ELECTRIC AND APPLIED PROCESS TECHNOLOGY To o DEHCMSTRATE ABILITY 70 CFERATE A EWR CCNTIM-UCUSLY WITHIN HWC SPECIFICATION o VERIFY THAT IGSCC WILL MCT INITHATE AND EXISTING FLAWS DO NOT PRCPAGATE IN HWC o-DETERMINE THE EFFECT OF HWC CN CTHER STRUCTURAL HATERIALS AND CN THE FUEL o DEFINE THE RADICLCGICAL IMPACT OF HWC e H

xvencesx wAvsnexmsrnv(xwe) 1985 ACCCHPL!SHMENTS(1) - o DOCUMENTED EFFECTS OF CPERATING DRESDEN-2 CN HWC DURING CYCLE 9 - WATER CHEMISTRY ~ -IN PLANT MATERIALS l= STING - RADICLCGiCAL REPACT - FUEL SURVEILLANCE 4 I e-w r

xvanceux wAunexnBww(xwe) CYCLE 9 FUEL SURVE!LLANCE RESULTS o F00LSIDE AND HOT CELL EXAMINATIONS o CXiDATION AND HYDRIDIN9 BEHAVICR - No HEASURAELE EFFECTS OF HWC o CRUD CHARACsiR!SiiCS - ECHE D!FFERENCES CCHPARED TO NWC - NO IMPACT ON FUEL PERFORMANCE AT DRESDEN-2 - WILL NEED TO BE ASSESSED AT OTHER PLANTS O l i l

xvonosex wxrenexerernvowe) 1935 ACCOMPL!SHMENTS(2) o DEMONSTRATED THE EFFEchVENESS OF HWC IN HITIGATING ISSCC AT DRESDEN-2 DURING THE FIRST S HONTHS OF CYCLE 10 - WATER CHEMISTRY AND ECP MEASUREHEhTS - ON-LINE CRACK GROWTH MONITOR (CAVS) g a W t s w-- ,_,e-- -e a .n-m --,- - - -, - - - - - - - - - - r n,,,-,-m-, e

d I DRESDEN 2 HWC CAVS 0.780 ftpfffirfffffffffr i f(( I l ((ffff[ffff JJJd)Jjh - 200 pbb o2 LAB TEST I f-[g-[J.dJlh bJj j JJJdJN [ffI j' L da/dt = 1.7-2.8 X 10-5 g,,4, 0.740 (149-245 mils /fpy) CHACK ~ LENGTH ~-------8""------. 0.720 D2HWC daldt = 5.4 X 10-7 in.Ar * ~ (4.7 mils /fpy) 0.700

• EXCLUDING INTERRRUPTIONS l

l l i I i 0.6800 0.4 0.8 1.2 1.6 2.0 3 ON LINE TEST TIME thrs X 10 ) I P S

xvomeesx wxrsm exsrerav(xwe) 1985 ACCOMPLISHMENTE(5) o DEVELOPED GENERIC GUIDELINES FOR HWC IMPLEMENTATiCN - DEFINES ASPECTS OF IMPLEMENTATION THAT CAN PROCEED UNDER 1CCFR 50.59 AND CITES Pi!RTINil!NT CCDES AND STANDARDS - PROVIDES CONSERVATIVE ANALYSIS PROCEDURES FOR ASPECTS OF IMPLEMENTATION THAT NEED ANALYSIS (E.G., SITING OF CRYCGENIC STORAGE TANK) - NRC REVIEW IS IN PROCESS M ,--r --,ea v --e-

xvncesx wenexmmy(xwe) 1985 ACC0HPL!SHMENTS(4) o ANALYZED THE RESULTS OF SHORT-TERN HYDROGEN INJECTION TESTS AT THREE U.S. PLANTS - CONTROL OF IGSCC BY HWC IS POS$1 ELE AT ALL THREE PLANTS - EACH PLANT RESPONDS DIFFERENTLY TO HYDROGEN ~ INJECTION - VAR! ABILITY OF R + PONSES CANNOT EE EXPLAINil!D .EY DESIGN DIFFERENCES - THE KINETICS OF THE RADIOLYTic DECCHFCSITION AND RECCHEINTAION RF.ACTICMS INVOLVED IN HWC NEED TO EE UNDEssTOCD 1 i

Recirc Oxygen Concentration (ppb) I e PB-3 e PIL 100.0 v FITZ A DRE-2 u 10.0 =- A = 2 1.0 ^ I I l-I i-0.2 0.6 1.0 1.4 1.8 2.2 2.6 Feedwater Hydrogen Concentration (ppm) EP9tI7487 I i ) 1 ~

\\ l 1 l Normalized MSLAM Actisity 6.0 5.0 L' y _. _-i u 4.0 3.0 a 8 B3 2.0 e plL u v FITZ 1.0 '---- A ORE 2 0.0 l l l l l l 0.0 0.4 0.8 1.2 1.6 2.0 2.4 Feedwater Hydrogen Concentration (ppm) um.n L

i ~ XYDE@@2X WATER CXEM137RY

SUMMARY

OF STATUS AND PLANS ) o FIRST CYCLE OF HWC 0FERATION AT DRESDEN-2 WAS CCHPLETED 10/84 (CYCLE 9). CUTCCHE WAS .FAVCRABLE o SECOND CYCLE (CYCLE 10)IS IN PROGRESS-BEGAN F 4/25. RESULTS TO DATE CONTINUE POSITIVE o NEED FOR A THIRD CYCLE TO CONFORM SATISFACTORY FUEL PERFORMANCE WILL BE ASSESSED DURING 1833 e SCCPE OF WATER CHEMISTRY WORK WILL BE EXPANDED TO ADDRESS PLANT-TC-PLANT VARIABILITY t a 4 -r --e -r- -.... - _ -. _.-_.. ~,. - e

PLANT MATEMBALS PIMFORMANCE 1@@@ ACC@MPLD2MMEMT2 e @EMCM27 RATED THE FEA20@0LOTY CF AGS CDS-LIDSE 090-@LAGST C@22C20008 CRACCIDOS@ SEM2@E WMCCM CAGS E0000702 THE A@@RE220VESE22 CF PLAGST CC@LASSV2 TOTAIC STRUCTURAL MATE 20ALS e DOCUDSEDSTED THE EFFECT2 07 A EAM25 CF EM2 WATER CMEMD272Y 100@U207522 CSS E@2CC CF 27ABCSLE22 27 EELS AMD l@ESTDFDED AGS APPECACH TOTAE@ DGSMD25702 DEVELOPMEST 4 ene O . -w

l d osygen conesne or common poisneel a incremens Normal 'e's se range - P soscc possee rene sm % vc., 'nem no rose inscc m regen Pvwi onmary sysernranes lenounty concentracon or conducevwy Figure 2 1 Schematic Surnary of the Results of Laborato y Studies of Effect of Impurities on SCC of Sensitized Austenitic Stainless Steels

ee

• a s

Crack Crowth Rates of Types 304 and 316NG Stainless Steels in 200 ppb Cxygen Water at 550'r Cantaining sodium sulfate = 60 e* 1 C 40 -E

s:

wg 20 = = 0 Alloy Type 304 316NG 304 316MG 304 316NG CCtECTIVITY < 0.1 0.4s 1.0 (us/en) U e r-w--+- w-- -'---r----- =v'" ^ =- ' " ---}}