ML20138D728
| ML20138D728 | |
| Person / Time | |
|---|---|
| Site: | Satsop |
| Issue date: | 11/30/1985 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20138D699 | List: |
| References | |
| NUDOCS 8512130290 | |
| Download: ML20138D728 (300) | |
Text
{{#Wiki_filter:. u s DRAFT SAFETY EVALUATION REPORT WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO. 3 DOCKET NO. 50-508 _ . _ u..- November 1985 4
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8512130290 PDR 851119ADOCK 05000508 PDR E
4 TABLE OF CONTENTS Page ABSTRACT......................................................... iii 1 INTRODUCTION AND GENERAL DESCRIPTION........................ 1-1 1.1 Introduction........................................... 1-1 1.2 Standard Design........................................ 1-4 1.2.1 References to the CESSAR SER.................... 1-5 d % 1.3 General Plant Description.............................. 1-5 1.4 Comparison with Similar Facility Designs. . . . . . . . . . . . . . . 1-6 1.5 Identi fication of Agents and Contractors. . . . . . . . . . . . . . . 1-6 , 2 SITE CHARACTERISTICS........................................ 2-1 2.1 Geography and Demography............................... 2-1 2.1.1 Site Location and Description................... 2-1
- 2.1.2 Exclusion Area Authority and Control............ 2-2 2.1.3 Population Distribution.........................
2-3 2.1.4 Conclusion...................................... 2-5 2.2 Nearby Industrial, Transportation, and Military Fa c i 1 1 t i e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-5. . j 2.2.1 Transportation Routes........................... 2-5 2.2.2 Nearby Facilities............................... 2-7 2.2.3 Conclusion Regarding the E.aluation of Potential Accidents......................... ... 2-9 4 i
.... i WNP-3 DSER TC
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I l TABLE OF CONTENTS (Continued) l Pagg 1 2.3 Meteorology............................................ 2-9 2.3.1 Regional Climatology............................ 2-9 l . 2.3.2 Local Meteorology............................... 2-12 j 2.3.3 Onsite Meteorological Measurements Program...... 2-13
- 2.3.4 Short-Term (Accident) Diffusion Estimates....... 2-15 2.3.5 Long-Term (Routine) Diffusion Estimates......... 2-16 r
2.4 Hydrologic Engineering................................. 2-17 l
, 2.4.1 Hydrologic Description.......................... 2-17 2.4.2 F1oods.................... ..................... 2-19 2.4.3 Probable Maximum Flood on Streams and Rivers.... 2-22 2.4.4 Potential Dam Fa11ures.......................... 2-24 2.4.5 Probable Maximum Surge and Seiche Flooding...... 2-25 2.4.6 Probable Maximum Tsunami Flooding............... 2-25 2.4.7 Ice Effects..................................... 2-26 2.4.8 Cooling Water Canals and Reservoirs............. 2-26 2.4.9 Channel Diversions.............................. 2-26 2.4.10 Flood Protection Requirements................... 2-27 2.4.11 Cooling Water Supp1y............................ 2-28 2.4.12 Groundwater..................................... 2-30 2.4.14 Technical Specifications and Emergency Operation Requirement........................... 2-35 2.4.15 Conclusions..................................... 2-35 2.5 Geology and Seismology................................. 2-36 '
2.5.1 Basic Geologic and Seismic Information. . . . . . . . . . 2-38 2.5.2 Vibrato ry Ground Motion . . . . . . . . . . . . . . . . . . . . . . . . . 2-43 2.5.3 Surface Faulting................................ 2-45 11 WNP-3 DSER TC
TABLE OF CONTENTS (Continued) Page 2.5.4 Stability of Subsurfaca Materials and Foundations..................................... 2-52 2.5.5 Stability of S1 opes............................. 2-67 2.5.6 Embankments and Dams............................ 2-70 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS.... 3-1 3.1 General................................................ 3-1 3.2 Classification of Structures, Systems, and Components.. 3-1 3.2.1 Seismic Classification.......................... 3-1 3.2.2 System Quality Group Classification............. 3-1 3.3 Wind and Tornado Criteria and Loadings................. 3-1 3.3.1 Wind Design Criteria............................ 3-1 3.3.2 Tornado Desi gn Cri teri a. . . . . . . . . . . . . . . . . . . . . . . . . 3-2 3.4 Water Level (Flood) Design............................. 3-4 3.4.1 Flood Protection................................ 3-4 3.4.2 Water Level (Flood) Design Procedures........... 3-6 3.5 Missile Protection..................................... 3-7
. 3.5.1 Missile Selection and Description............... 3-7 3.5.2 Structures, Systems, and Components To Be Protected from Externally Generated Missiles.... 3-13 3.5.3 Barrier Design Procedures....................... 3-15
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t TABLE OF CONTENTS (Continued) Pagg 3.6 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping.............. ....... 3-16 3.6.1 Plant' Design for Protection Against Postulated Piping Failures in Fluid Systems Outside Containment..................................... 3-16 3.6.2 Determination of Rupture Locations and
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Dynamic Effects Associated with the Postulated Rupture of Piping............................... 3-19 3.7 Seismic Des 1gn......................................... 3-19 3.7.1 Seismic Input................................... 3-19 3.7.2 Seismic System Analysis......................... 3-20 3.7.3 Seismic Subsystem Analysis...................... 3-20
- 3.7.4 Seismic Instrumentation Program................. 3-21 1
3.8 Design of Seismic Category I Structures................ 3-22 3.8.1 Concrete Containment............................ 3-22 3.8.2 Steel Containment............................... 3-22 3.8.3 Concrete and Structural Steel Internal Structures...................................... 3-23 3.8.4 Other Seismic Category I Structures............. 3-24 3.8.5 Foundations..................................... 3-25
. 3.8.6 Structural Audit................................ 3-26 3.11 Environmental Qualification of Electrical Equipment Important to Safety and Safety-Related Mechanical Equipment.............................................. 3-26 3.11.1 Introduction................................... 3-26
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t 4 4, i .. j TABLE OF CONTENTS (Continued) > i
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} 3.11.2 Background..................................... 3-27 ! c { 3.11.3 Completeness of the Environmental Qualification l Program........................................ 3-28 f 1 i 1 4 REACT 0R..................................................... 1 i 1 1 4.1 Introduction........................................... .4-1
- 4.2 Fuel 0esign............................................ 4-1 4.3 Nuclear 0esign.
- ....................................... 4-3 I i 4.4 Thermal-Hydraulic 0esign............................... 4-3 ,
) ' 4.4.1 Loose Parts Monitoring System................... 4-3
, 4.4.2 Inadequate Core Cooling Requirements
- (SRP-4.4-Section II.9).......................... 4-4 4.4.3 Plant Specific Information
? (SRP 4.4-Secti,on 11.1).......................... 4-4 . l 4.6 Functional Design of Reactivity Control System ........ 4-5 1 L__ - . 5 --- REACTOR C00LANT1YSTEM.~7.- J. .- . ;. - . .--- rr. cr. c . . . . . . . . . . . . . . 5-1 - 1 5.1 S umma ry De s c r i p t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 l 5.2 Integrity of Reactor Coolant Pressure Boundary. . . . . . . . . 5-1 f 5.2.4 Reactor Coolant Pressure Boundary Inservice 4 l Inspection and Testing.......................... 5-1 1 5.2.5 Reactor Coolant Pressure Boundary Leakage 4 04tection....................................... 5-5 r ! 5.3 Reactor Vessel......................................... 5-9 i 5.3.1 Reactor Vessel Materials (Fracture Toughness)... 5-9 i 5.3.2 Pressure-Temperature Limits..................... 5-12
- 5.3.3 Reactor Vessel Integri ty. . . . . . . . . . . . . . . . . . . . . . . . 5-13 y WNP-3 DSER TC
a TABLE OF CONTENTS (Continued) P.agg 5.4 Component and Subsystem Design......................... 5-15 f 5.4.1 Reactor Coolant Pump Flywheel Integrity......... 5-15 5.4.2 Steam Generator................................. 5-17 6 ENGINEERED SAFETY FEATURES.................................. 6-1 6.1 Materials..........................'.................... 6-1 1 6.1.1 Post-Accident Emergency Cooling Water Chemistry....................................... 6-1 6.1.2 Organic Materials............................... 6-2 6.2 Containment Systems.................................... 6-3 1 6.2.1 Containment Functional Design................... 6-4 6.2.2 Conta i nment Heat Removal Systems. . . . . . . . . . . . . . . . 6-10 6.2.3 Secondary Containment Functional Design......... 6-12 - 6.2.4 Containment Isolation System.................... 6-14 6.2.5 Combustible Gas Control System.................. 6-19' ] 6.2.6 Containment Leakage Testing Program............. 6-20 6.2.7 Fracture Prevention of Containment Pressure Boundary........................................ 6-23 , i 6.4 Control Room Habitability.............................. 6-25 6.5 Engineered-Safety-Feature Atmosphere Cleanup System......................................... 6-26 6.5.2 Containment Spray as a Fission Product Removal System.......................................... 6-26 4 vi WNP-3 DSER TC
TABLE OF CONTENTS (Continued)
. P_ag 6.6 Inservice Inspection of Class 2 and 3 Components....... 6-26 6.6.1 Compliance with the SRP......................... 6-26 6.6.2 Examination Requirements........................ 6-28 6.6.3 Evaluation of Compliance with 10 CFR 50.55a(g).. 6-28 6.6.4 Conclusions..................................... 6-29 j 6.6.5 References...................................... 6-30 9 AUXILIARY SYSTEMS........................................... 9-1 9.1 Fuel Storage and Handling ............................. 9-2 ,
9.1.1 New Fuel Storage ............................... 9-2 9.1.2 Spent Fuel Storage ............................. 9-4 9.1.3 Spent Fuel Pool Cooling and Cleanup System ..... 9-7 9.1.4 Light Load Handling Systems (Related to Refueling) 9-9 9.1.5 Overhead Heavy Load Handling System ............ 9-12 9.2 Water Systems ......................................... 9-14 9.2.1 Station Service Water System ................... 9-14 9.2.2 Reactor Auxiliary' Cooling Water Systems ........ 9-16 _ 9.2.3 Demineralized Water Makeup System .............. 9-23 9.2.4 Potable and Sanitary Water Systems . . . . . . . . . . . . . 9-24 9.2.5 Ultimate Heat Sink ............................. 9-25 l 9.3 Process Auxiliaries ................................... 9-28 9.3.1 Compressed Air System .......................... 9-28 , 9.3.2 Process and Post-Accident Sampling System ...... 9-28 9.3.3 Equipment and Floor Draining System ............ 9-38 4 9.3.4 Chemical and Volume Control System ............. 9-38 9.5 O the r Aux i l i a ry Sys tems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-39 9.5.1 Fire Protection Program ........................ 9-39 10 STEAM AND POWER CONVERSION SYSTEM .......................... 10-1 2 . 10.3 Main Steam Supply System .............................. 10-1 10.3.5 Secondary Water Chemistry ..................... 10-1 yli WNP-3 OSER TC
TABLE OF CONTENTS (Continued) Pa21 13.4 Other Features......................................... 10-3 10.4.6 Condensate Cleanup System...................... 10-3 10.4.7 Condensate and Feedwater System................ 10-4 10.4.8 Steam Generator Blowdown System................ 10-4 12 RADIATION PROTECTION........................................ 12-1 12.1 Ensuring that Occupational Radiation Ooses are ALARA... 12-1 12.1.1 Policy Considerations.......................... 12-2 12.1.2 Design Considerations.......................... 12-2 12.1.3 Operational Considerations..................... 12-3 12.2 Radiation Sources...................................... 12-4 _ 12.2.1 Contained and Airborne Sources................. 12-4 12.3 Radiation Protection Design Features................... 12-6 12.3.1 Facility Design Features....................... 12-6 12.3.2 Shielding...................................... 12-7 12.3.3 Vent 11ation.................................... 12-10 12.3.4 Area Radiation and Airborne Radioactivity Monitoring Instrumentation..................... 12-11 12.4 Dose Assessment........................................ 12-13' 12.5 Operational Radiation Protection Program............... 12-14 12.5.1 Organization................................... 12-14 12.5.2 Equipment, Instrumentation, and Facilities..... 12-16 12.5.3 Procedures..................................... 12-17 viii WNP-3 DSER TC
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l . l r l l TABLE OF CONTENTS (Continued) l l l P_ggg l l 13 CONDUCT OF OPERATIONS....................................... 13-1 I 13.1 Organizational Structure of Applicant.................. 13-1 13.2 Training............................................... 13-1 13.2.1 Licensed Operator Training Program............. 13-1 13.2.2 Training for Nonlicensed Plant Staff........... 13-13 13.3 Emergency Planning..................................... 13-13 l 13.6 Industrial Security.................................... 13-14 1 13.6.1 Introduction................................... 13-14 13.6.2 Physical Security Organization................. 13-15 13.6.3 Physical Barriers.............................. 13-15 13.6.4 Identification of Vital Areas.................. 13-16 _l 13.6.5 Access Requirements............................ 13-16 r 13.6.6 Detection Aids................................. 13-17 ; 13.6.7 Communications................................. 13-18 -- 13.6.8 Test and Maintenance Requirements.............. 13-18 13.6.9 Response Requirements.......................... 13-19 13.6.10 Employee Screening Program..................... 13-20 14 I N I T I A L T E ST P R0G RAM . . . . . . . . . . . . . . . . . . . . . . ; . . . . . . . . . . . . . . . . . 14-1 15 ACCIDENT ANALYSES........................................... 15-1 15.4 Reactivity and Power Distribution Anomalies............ 15-1 15.4.1 CEA Wi thdrawal f rom Low Power. . . . . . . . . . . . . . . . . . 15-1 15.4.2 CEA Withdrawal from Full Power................. 15-1 1 __ 1x WNP-3 DSER TC
TABLE OF CONTENTS (Continued) P_agg l 15.4.3 CEA Misoperation Events........................ 15-1 15.4.7 Fuel Misloading Event.......................... 15-1 l 15.4.8 CEA Ejection Event............................. 15-1 l I l 15.x Radiological Consequences of Design Basis Accident..... 15-1 1 15.x.1 Loss-of-Coolant Accident....................... 15-2 l 15.7 Radioactive Releases from a Subsystem or Component..... 15-5 15.7.4 Fuel Handling Accident......................... 15-5 18 HUMAN FACTOR S ENG I NE ER I NG . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18-1 l t l l 1 I i i 1 1 l l l i l i l l t
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1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT j l 1.1 Introduction This report is a Draft Safety Evaluation Report (DSER) on the application for an operating license (OL) for Washington Public Power Supply System (WPPSS) Nuclear Project No. 3 (WNP-3). In March 1974, WPPSS filed an application with the U.S. Nuclear Regulatory Commission (NRC) for permits to construct and oper-ate the proposed WNP-3 plant. The plant is located in southeastern Grays Harbor - County, Washington, approximately 1 mile south of the Chehalis River near its confluence with the Satsop River. The largest cities within 50 miles of the site are Olympia and Aberdeen. Olympia, the state capital, is 26 miles east of the site; Aberdeen is 16 miles west of the site. Following reviews by the staff and the Advisory Committee on Reactor Safeguards (ACRS), public hearings were held before an Atomic Safety and Licensing Board. A construction permit (CP) for WNP-3 (CPPR-154) was issued on April 11, 1978. In response to the OL application for WNP-3, the NRC staff performed an accep-tance review and, on August 20, 1982, issued a letter accepting the application. _ Information received on the WNP-3 OL application was docketed on August 22, 1982. On July 8, 1983, the WPPSS Executive Board adopted a resolution calling for an immediate construction delay of WNP-3 until an ensured source of funding for continued construction could be obtained. The applicant informed the staff that, as of September 30, 1983, construction of WNP-3 was about 75% complete. By letter dated November 18, 1983, WPPSS informed the staff that the projected fuel load date ranges from June 1987 to December 1989. A detailed implementa-tion plan for construction delay at WNP-3 was submitted to the staff on September 15, 1983. - Before issuing an OL for a nuclear power plant, the NRC staff is required to conduct a review of the offects of the plant on public health and safety. The staff safety review of WNP-3 has been based on NUREG-0800, " Standard Review Plan for the Review of Safety. Analysis Reports for Nuclear Power Reactors, LWR 1-1 WNP-3 DSER SEC 1
Edition" (SRP). In general, an audit review of each of the areas listed in the Areas of Review section of the SRP was performed according to the guidelines provided in the Review Procedures portion of the SRP. Exceptions to this prac-tice are noted in the applicable sections of this report. This DSER summarizes the results of the staff's radiological safety review of WNP-3 and delineates the scope of the technical details considered in evaluating the radiological safety aspects of its proposed operation. The design of the station was reviewed against Federal regulations, CP criteria, and the S?.P, except where noted otherwise. The SRP covers a variety of site conditions and plant designs. Each SRP section is written to provide the complete procedure
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and all acceptance criteria for all the areas of review pertinent to the section. However, for any given application, the staff may select and emphasize p'artic-ular aspects of each SRP section as appropriate for the application. In some cases, the major portion of the review of a plant feature may be done on a gen-eric basis, with the designer of that feature, rather than in the context of reviews of particular applications from utilities. In other cases, a plant feature may be sufficiently similar to that of a previous plant so that a de novo review of the feature is not needed. - During the . course of its review, the staff held a number of meetings with repre-sentatives of the applicant to discuss the design, construction, and proposed - operation of the plant. The staff requested additional information, which the applicant provided in amendments to the Final Safety Analysis Report (FSAR). This information is available to the public for review at the NRC Public Docu-ment Room at 1717 H Street, N.W., Washington D.C. 20555 and at the Local Public Document Room at the W. H. Memorial Library, 125 Main Street, South, Montesano, Washington 28563. . Following the incident at the Three Mile Island Unit 2 Nuclear Plant (TMI-2), the Commission paused in its ?' censing activities to assess the impact of the ' incident. During this pause, the recommendations of several groups established to investigate the lessons learned from THI-2 became available. All available recommendations were correlated and assimilated in a "TMI Action Plan," pub-lished as NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Acci-dent." Additional guidance relating to implementation of the Action Plan is in
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NUREG-0737, " Clarification of TMI Action Plan Requirements," and in Supplement 1 1-2 WNP-3 DSER SEC 1
to NUREG-0737. Licensing requirements based on the lessons learned from the THI-2 incident have been established to provide additional safety margins. These have been incorporated into the design and operation of WNP-3. As part of its review of the application against the NRC regulations, the staff will ask,the applicant to certify that WNP-3 meets the applicable requirements of Parts 20, 50, 51, and 100 of Title 10 of the Code of Federal Regulations (CFR). Following the applicant's response to this request, the staff will l address its findings in this area in the Safety Evaluation Report (SER). In accordance with the provisions of the National Environmental Policy Act
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(NEPA) of 1969, a Final Environmental Statement (FES) that sets forth the envi-ronmental considerations related to the proposed construction and operation of l WNP-3 was published in May 1985 (NUREG-1033). The review and evaluation of WNP-3 for an OL is only one of many stages at which the staff reviews the design, construction, and operating features of the fact-lity. The facility design was extensively reviewed before the applicant was granted a CP for the facility. Construction of the facility has been monitored - in accordance with a detailed monitoring and inspection program at the OL stage. The NRC staff has reviewed the final design of the facility to determine that the Commission's regulations have been met. If an OL is granted, WNP-3 must be : operated in accordance with the terms of the OL and the Commission's regulations, and the facility will be subject to the staff's continuing inspection program. In addition to the NRC staff review, the ACRS will review the application and will meet with both the applicant and the staff to discuss the final design and proposed operation of the plant. The Committee's report to the Chairman of the NRC will be included in a supplement to the SER. The NRC Project Manager assigned to the OL application for WNP-3 is Mr. Braj K. S i n'gh . Mr. Singh may be contacted by calling (301) 492-8423 or by writing: Mr. Braj K. Singh Division of Licensing U.S. Nuclear Regulatory Commissier. Washington, D.C. 20555 1-3 WNP-3 DSER SEC 1
1.2 Standard Design The FSAR submitted with the WNP-3 application describes the design of the balance-of plant (BOP) structures, systems, and components and incorporates, by reference, the Combustion Engineering report, " Combustion Engineering Standard Safety Analysis Report" (CESSAR FSAR). CESSAR describes the design of the System 80 nuclear steam supply system (NSSS). The initial Commission policy statement on standardization of nuclear power plants was issued on April 28, 1972. It provided the impetus for both industry and the Commission to initiate active planning in their respective areas in order to realize the be'nefits of standardization while maintaining protection ~
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of the health and safety of the public and of the environment. In a sub' sequent statement issued on March 5, 1973, the Commission announced its intent to imple-ment a standardization policy for nuclear power plants. WASH-1341, " Program-matic Information for the Licensing of Standardized Nuclear Power Plants," was issued on August 20, 1974. Amendment I to WASH-1341, dealing with " options" and " overlaps," was issued on January 16, 1975. The regulations governing the submittal and review of standard designs under the " reference system" option - are found in Appendix 0 to 10 CFR 50, " Domestic Licensing of Production and Utilization Facilities," and 10 CFR 2.110. CESSAR was submitted by Combustica Engineering in the form of an application for a Final Design Approval (FDA) from the Commission and in response to Op-tion 1 of the Commissions' standardization policy. Option 1 allows for the review of a " reference system" that involves an entire facility design or major fraction of a design outside the context of a license application. On Decem-ber 21, 1979, the application for CESSAR was docketed.
. The staff evaluation of CESSAR is presented in NUREG-0852, " Safety Evaluation Report Related to the Final Design Approval of the Combustion Engineering Stan-dard Nuclear Steam Supply System (CESSAR)," hereinafter referred to as the CESSAR SER. The CESSAR SER has been incorporated by reference into this SER.
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I l l 1.2.1 References to the CESSAR SER The WNP-3 NSSS design description and evaluation are incorporated into the WNP-3 l j SER by reference to the appropriate CESSAR SER chapter, section, table, or [ figure. For example, the phrase "the staff evaluation is presented in the CESSAR l SER, Section 9.3.5" is used to incorporate the identified CESSAR SER section f into this SER. Any other chapter, section, table, or figure referenced in this j SER refers to a chapter, section, table or figure of this SER. Each reference l to a CESSAR SER section also incorporates all figures, tables, and appendices referred to in the incorporated CESSAR SER section. All topical reports referred to in the CESSAR SER section are also incorporated in the WNP-3 SER. t 1.3 General Plant Description i 1 The System 80 nuclear steam supply system, as described in the CESSAR SER, will l incorporate a pressurized-water reactor and a two-loop reactor coolant system f with a net electrical output of approximately 1240 MW. Each loop of the reactor ! coolant system will consist of an outlet pipe (hot leg), one steam generator, f l two inlet pipes (cold legs) and two reactor coolant pumps, one in each cold _: leg. An electrically heated pressurizer will be connected to one loop and will ; i establish and maintain the reactor coolant pressure, t t The nuclear steam supply system will be housed in a containment structure. The ! containment will consist of a steel containment vessel surrounded by a reinforced j concrete shield building. The containment vessel is a cylindrical vessel with i l a hemispherical dome and ellipsoidal bottom. The shield building is a medium-leakage concrete structure that surrounds the steel containment vessel. ' l The steam and power conversion system will be designed to remove heat from the I reactor coolant in the two steam generators and convert it to electrical energy. The excess heat removed by the condenser will be discharged to cooling towers i through the circulating water system. The plant will be capable of being supplied electrical power from two indepen-dent offsite power sources and will also ce provided with an independent and i redundant onsite emergency power supply system which 15 capable of supplying _ l electric power to the engineered safety features. 1-5 WNP-3 DSER SEC 1
l l l Before an OL is issued, the staff will review the final design to determine j that all of the Commission's safety requirements have been met. The facility l may then be operated only in accordance with the terms of the operating license and the Commission's regulations under the continued surveillance of the Commis-j sion's staff.
- j. 1.4 Comparison with Similar Facility Desians A comparison of the CESSAR design with a similar facility design is presented i in the CESSAR SER, Section 1.3.
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1.5 Identification of Agents and Contractors WPPSS will act as project manager and operating agent for the participants and will have full authority and responsibility to engineer, design, construct, operate, and maintain WNP-3. 1 The applicant has retained EBASCO to perform architectural, engineering, and construction services. Combustion Engineering has been contracted to design, _ manufacture, and deliver the nuclear steam supply system and nuclear fuel for the initial core. In addition, Combustion Engineering will furnish technical assistance for erection, initial fuel loading, testing, and initial startup of -, the nuclear steam supply system. The turbine generators will be manufactured by Westinghouse. The applicant also will use consultants, as required, in specialized areas. i
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l 1 l l 2 SITE CHARACTERISTICS 2.1 Geography and Demography The July 1981 edition of " Standard Review Plan for the review of Safety Analysis Reports for Nuclear Power Plants," (SRP, NUREG-0800) includes Chapter 2, Site Characteristics. The Washington Nuclear Project 3 (WNP-3) was reviewed in l accordance with Section 2.1.1, 2.1.2, and 2.1.3. The results of this review are contained in Section 2.1 of this safety Evaluation Report. - - l 2.1.1 Site Location and Description The site for WNP-3, which was formerly proposed as a two unit site, consists , of 2450 acres located in southeastern Grays Harbor-County in the western part of the state of Washington. Figure 2.1 shows the site location including the low population zone (LPZ), and the surrounding area within 4 miles. The principal plant features in conjunction with the exclusion area and site boundary lines (including land ownership) are shown in Figure 2.2. The area within 50 miles of the site is shown in Figure 2.3. Figure 2.4 shows the site __ location relative to various transportation routes and transmission lines. These include the Chehalts and Satsop rivers, the Union Pacific and Burlington Northern railroad lines, an eight inch natural gas pipeline, the BPA transmis-
- sion lines, and U.S. Highway 12, as well as other local roads. As shown in Figures 2.1, 2.2, and 2.4, WNP-3 is located about 1.25 miles south of the Chehalis River near its junction with the Satsop River. The site is situated on a ridge above the Chehalis River in an area that is generally rural in nature and sparsely populated but extensively used for timber production, with some scattered farmlands in the fertile river bottom lands. The city of.
Olympia, the capitol of Washington, is about 26 miles east and the cities of ! Aberdeen-Hoquiam, Washington are approximately 20 miles west of the site. Tacoma, Washington is the largest populated area in the vicinity and is located 50 miles northeast. The coordinates of WND-3 are 46' 57' 34" north latitude l and 123* 28' 15" west longitude. The Universal Transverse Mercator (UTM)
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j coordinates are 5,200,546 meters north and 464,176 meters east, in zone 10. l 2-1 WNP-3 DSER SEC 2
2.1.2 Exclusion Area Authority and Control The exclusion area for the WNP-3 site is defined as a circular area with a 4300 foot (1310) radius measured from the center of the reactor containmert building. The applicant owns 1120 acres of the site in which the exclusion area is located. The rest of the site area is owned by private corporations or individuals, but under the control of the Washington Public Power Supply System (WPPSS) by agreement with the owners. These agreements provide ease-ments to the land. All of the mineral rights within the designated exclusion area will be either owned or leased by the applicant. Activities that occur within the exclusion area that are unrelated to plant operations are those associated with tree farming, and transmission line maintenance. Approximately 30 persons and associated temporary structures and facilities may be located within the exclusion area from time to time in order to conduct timber farming activities. These activities are subject to prior planning and approval by the applicant. In addition, by agreement with WPPSS, personnel from the Bonneville Power Administration (BPA) will be conducting maintenance in the BFA transmis-sion corridor falling within the exclusion area boundary. There are no resi-dents living within the exclusion area and there are no major highways, railways, or waterways traversing the area. However, there is some vehicular traffic on the access roads leading to the plant and on portions of the BPS corridor. Easements for these portions of the exclusion area are being negotiated with - Grays Harbor County and the BPA. The applicant does not anticipate any problems in obtaining the necessary authorization for controlling this area. Keyes Road will be abandoned and the applicant expects to provide the county with an appropriate right-of-way to run another road through the site. The applicant is making arrangements to control and, if necessary, evacuate the exclusion area in the event of an emergency. Section 13.3 of this report will provide more details on these arrangements when they are finalized. We conclude that when the negotiations to acquire the appropriate easements, ' and subsequently obtain full control are completed, the applicant will have the authority and will be able to control all activity unrelated to plant operations within the exclusion area as required by 10 CFR Part 100. We also conclude that the activities unrelateo to piant operation within the exclusion area will not interfere with normal pla~. cperation. 2-2 WNP-3 USER SEC 2
l 2.1.3 Population Distribution l The resident population in the vicinity of the WNP-3 site is shown as a function l of distance in the table below. The year 2030 is the nearest census year to I the end of plant life. l The closest resident lives about one mile (1609 meters) from the site. The j nearest communities in the vicinity of the site with a population of more than 1000 persons are the cities of Elma, located about four miles northeast with a l population of 2720 in 1980, and Montesano, which is about six miles west-northwest with a 1980 population of 3247. The closest large communities nearby
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are the cities of Aberdeen-Hoquiam, located approximately 20 miles west of the' site. This combined area had a populattun of 35,170 in 1980. The populhtion within five miles of the site in 1980 was 5867, and within ten miles it was 15,165. As indicated in Table 2.1, the population within five miles of the site is expected to increase by about 3600 persons during the life of the plant. The applicant reported that there were 514,954 people living within 50 l miles of the site in 1980, and this number is expected to increase to 646,145 l l by 1990. By the year 2030 the populaticn within 50 miles of WNP-3 is expected _. to reach 1,012,502. Olympia, the capitol of Washington, is about 26 miles east of the site and had a population of 27,447 in 1980. The largest city near the ! site is Tacoma, Washington, located about 50 miles northeast, with a 1980 - population of 158,501. The applicant predicts a population growth rate of about 1.36% per year for the area within 50 miles of the site during the life of the plant as compared to a growth rate of about 0.87% per year predicted by ! the U.S. Bureau of Economic Analysis (BEA) for this area. The staff made an 1
- independent assessment of the population within 50 miles of the plant for the year 2030, based on Bureau of the Census data, and estimated a growth rate of about 0.80% per year. Comparing the staff's estimate and the BEA predictions
. Indicates that the appItcant's projections are c1nservative.
The applicant has designated a low population zono (LPZ) for the site which is a circular area with a three mile (4827 meter) radius measured from the midpoint of the centerline between Unit 3 and the previously planned Unit 5. Except for the Chehalls and Satsop Rivers, the LPZ consists mostly of wooded areas and 2-3 WNP-3 DSER SEC 2
some agricultural land much the same as the rest of the area in the general vicinity around the WNP-3 site. There is very little transient population within the LPZ. There is some limited recreational activity on the rivers, including hunting and fishing, and there are several tree farming companies scattered throughout the area. According to the applicant, about 906 residents were living within the LPZ in 1980, and the population within the LPZ is expected to increase to about 1500 during the life of the plant. Except for the travelers using U.S. Highway 12, which is three miles north of the plant, going from the Puget Sound area to the Pacific Coast region, there is no significant amount of transient popula- ,
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tion within the ten miles surrounding the site. There are essentially no [ migrant workers within the ten mile radius around the plant. Logging opera-l tions vary from time to time throughout this area employing about ten persons per operation. On the average,120 persons per year are employed in logging activities at twelve different locations. Also within the ten mile radius, there are four companies, generally associated with lumber products, that employ about 130 persons. During peak periods this total may rise to about 185. The applicant estimates that as many as 128 fishermen may be seen in the area on a good day, and during the hunting season a total of 2700 may utilize the woods and river tributaries around WNP-3. There are six sc'hools within ten miles of the site. One is located in Satsup, two in Elma, and three in Montesano. -- Approximately 3550 students and staff occupy these facilities during the school - year. Four nursing homes, the Elma youth home, and the county jail, all within ten miles of the plant, averaged about 356 persons in 1981. Although most of ' these people (fishermen, hunters, students, loggers and other employees) are transit..y some of them reside within the ten mile radius and are included in the population statistics given above. Section 13.3 of this report provides a discussion of the Emergency Preparedness plans for protecting the public in this area. The nearest densely populated
. center, of about 25,000 or more persons, as defined by 10 CFR Part 100, is the combined area of Aberdeen-Hoquiam. The applicant has indicated that the center's nearest boundary is approximately 16 miles west of the site. This distance is at least one and one-thiro times the distance to the LPZ outer radius, as required by 10 CFR Part 100. .
2-4 WNP-3 DSER SEC 2
! 2.1.4 Conclusion Pending the applicant's completion of all ownership and easement transactions, we conclude that the exclusion area, low population zone and population center l distance' meets the criteria of 10 CFR Part 100 and will be acceptable. This finding is based on the 10 CFR Part 100 definitions of the exclusion area, the low population zone and the population center distance, our analysis of the onsite meteorological data from which the relative concentration factors (X/Q) l l were calculated (see'Section 2.3 of this report), and the calculated potential radiological dose consequences of design basis accidents (see Section 15.0 of this report). I 2.2 Nearby Industrial. Transportation, and Military Facilities l 2.2.1 Transportation Routes There are no major highways, railroads, or waterways traversing the WNP-3 exclusion area. A transmission corridor owned by the Bonneville Power Adminis-tration (BPA) penetrates the exclusion area but is limited to BPA maintenance i vehicles only. Public access to the exclusion area may be attained by means of Keyes Road and Workman Creek Road. These roads are used primarily for local traffic, including tree farming activities. The only major highway nearby is - U.S. 12, a divided four-lane highway running in an east-west direction approxi-mately three miles north of the plant. This highway is classified as a " scenic and recreational highway" used principally by motorists traveling between U.S. 1 which runs in a north-south direction along the Pacific Ocean, and I-5 which also runs north-south in the Puget Sound area. Because of the separation dis-tance involved and the topography of the area, accidents occurring on U.S. 12 that may present a hazardous materials problem do not pose a threat to the safo ,
. operation of the plant.
A single railroad track maintained and used by the Union Pacific Railroad, and a rail line owned by the Burlington Northern Ratiroad are located about 1.25 miles and three miles north of the plant, respectively. Both rail lines run parallel to the Chehalis River and U.S. 12 in a valley at an elevation about I - 2-5 WNP-3 DSER SEC 2 l
350 feet below and separated from WNP-3 by dense forest which is prevalent in the area. Rail traffic along the closest rail line averages about two trains daily, involving mostly shipments of lumber and related products and infre- . quently shipments of caustic soda, chlorine and propane. The applicant has identified several types of potential accidents involving hazardous material shipments on the clasest railroad line (7000 feet) to the plant. The accidents include: (a) the detonation of a box car loaded with 132,000 lbs of TNT, (b) the rupture, spill, and ignition of a tank car filled with 85 tons of l propane, and (c) the rupture and spill of a tank car loaded with 90 tons of liquid chlorine. The above events were assumed to occur on the railroad at a point nearest to the plant. l With respect to TNT, the applicant's analysis indicates that the overpressure j from detonating 132,000 lbs of TNT at 7000 feet from the plant will not exceed j the criteria in Regulatory Guide 1.91. Our review supports this finding and we concur with the applicant that TNT explosions on the railroad do not pose a l significant hazard to the plant. The applicant's analysis of an 85 ton propane tank car spill indicates that _ flammable concentrations would not reach the plant area. Since the railroad is about 350 feet below plant grade elevation, the heavier than air propane cloud would not tend to flow toward the plant. Detonation of 85 tons of - l propane at the railroad would produce a peak reflected overpressure of less l than 1.0 psi at the plant. Hence we find that the potential propane tank car l accident does not pose a significant hazard to the plant. l Potential chlorine spills on the railroad have led the applicant to provide chlorine detectors for the control room. A more detailed discussion of chlo-rine protection is given in Section 6.4 of this report. 1 l In view of the above, the staff concludes that the potential risks associated ! dith the postulated accidents on the railroad line near the WNp-3 site aae sufficiently low and do not exceed the criteria specified in Regulatory Guides 1.91 and 1.78. I i 2-6 WNP-3 DSER SEC 2
In addition to the above, the applicant also has identified a number of hazard-ous gas sources on site. .Two of these have some potential for control room operator incapacitation. Specifically, there is a 15000 gallon tank of aqueous ammonia, and four 1 ton tanks of liquid sulfur dioxide. The staff is currently reviewing the control room habitability system with respect to these gases, and is currently considering this to be an open item. The Chehalis River, located about 1.25 miles from the plant, and just north of the Union Pacific track, is used principally by small pleasure and fishing boats. Because the river is not commercially navigabla in this area (river mile 21), making barge traffic impractical, hazards to the plant from accidents
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on the river are non ' existent. River traffic does not pose a hazard to the
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l intake structure, because make-up water is supplied by a Ranney well collection system (which obtains water from about 80 feet below the surface of the river), and there is no potential for highway or railroad chemical spills flowing into the river and entering the make-up water system. 2.2.2 Nearby Facilities There is no extensive industrial activity around the WNp-3 site. The Thiokol Corporation, located about five miles east-northeast of the plant, stores materials such as methanol, metallic sodium, and nitrogen, which are used in - l the production of a bleaching agent for the pulp and paper industry. The Western Washington Paper Co. supplies gases for industrial and medical applica-tions. The gases, stored five miles northeast of the site, include acetylene, hydrogen, propane, argon, helium and nitrogen. There are also several gasoline and diesel fuel distributors that have storage facilities about five miles from the site. Tree farming is the main industry in this area and there is no significant amount of hazardous material ineIved in this type of operation. Because of the quantities and distances ins ,1ved, the materials stored at these l'ocations do not pose a hazard to the plant. There are no large industrial expansion programs planned for this area in the foreseeable future. There are no military bases, bombing ranges, munitions plants, missile instal-lations, or major airports within the generai vicinity of the WNP-3 Site. There are two grass airstrips and a s=a'.*, airport within five miles of WNP-3. 2-7 WNP-3 DSER SEC 2
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One of the airstrips, which hasn't t5een used ir years, s located 1.2 miles north-northwest. The other is about 4.2 mile 6 east-northeast of the site and is used ver9 infrequently. TheairstripsareJntendedforprivateuse,usually
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' crop dus, ting. Elmaafrport,theclosestactiYeairportinthearea,islocated three miles northeast of the site. It has a 2100 foot paved runway and can accommodate light single and twin engine fixed wing aircraft and helicopters that weigh under 12,500'peunds. The airport was formerly unattended, so no D' recordswerekeptontheusheithad. It'was estipat6d that there were be-
- tween 4000 and 1:'0QO.take-off and landing operations in 1980. A minor expan-sion program has been proposed for the Elma airport. The largest airport near theplantisBokermanField,anditislocatedabouth22mileswest.
g ~ Although
, there were ns jcheduled cemmercial fif$ts prior to 1980, commuter service is ~
proposed and(K is projected that there will be about 125,000 operations'at this airport (y the year 2000. f Two low-level, Federal Airways (V-204 and V-P,7) are located in the airspace around WMp-3. There are about five commerc hi' flights per day on route V-204 I which pass over the site at a minimum altitude of 4500 feet MSL. Approximately seven commercial flights per day use routb V-27 which passes within eleven _. miles'of the plant at a minteum altitude of 3200 feet MSL. Ft. Lewis Army Base and McChord Air Force Base are located 42 and 48 miles east-northeast of the
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site, respectively. There Ee one or two military helicopter flights per month - '~ that use the V-204 routo. , Aircraft froiMcChord AFB (about 280 flights $er year)$se a predetermined route which passes within four miles of the plant. No ordinance is carried on the miittary aircraft flying in this area. Ths applicant is analyzing the effects on plant etreciures assuming an aircraft crashes into WNP-3. The staff will r.eview the analysis when- it is Ebeitted.and evaluate the potential -
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risks to tno safe ~opeastion of the plant; '
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NoLPGorLNG11hN,arelocatedwithinfivemt$softheplant,butthereis one natural. gas pipeline nearby. +.lisaneightinchlinethat is buried at a minimum depth of. 30 inches and ha(a,naximune operating pressure of 305 psig. This line runs in an, cast west direction aoout five miles north of the site.
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2-8 WNP-3 DSER SEC 2
-_-__-_--__--____--____-__---_---_____.--._-.__.___-_____.-___---_---__.le__-
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Two lines, four inches in diameter, are tapped off of the main eight inch gas line and serve the towns of Elma and Montesano. There are no plans to transport other products in these pipelines or to up grade the system. Because of the distance involved these gas lines do not present a hazard to the WNP-3 facilities. 2.2.3 Conclusion Regarding the Evaluation of Potential Accidents On the basis of the information provided by the applicant, and our review based upon criteria in 10 CFR Part 50, Appendix A, GDC 4, and in Standard Review Plan Section 2.2.3, we have determined, subject to a satisfactory conclusion pending our evaluation of the applicant's aircraft analysis, that the WNP-3 site will be adequately protected and can be operated with an accept-able degree of safety considering the activities at nearby transportation, industrial, and military facilities. 2 2.3 Meteorology Evaluation of regional and local climatological information, including extremes of climate and severe weather occurrences which may affect the design and __ siting of a nuclear plant, is required to assure that the plant can be designed and operated within the requirements of Commission regulations. Information concerning atmospheric diffusion characteristics of a nuclear power plant site -- is required for a determination that radioactive effluents from postulated acci-dental releases, as well as routine operational releases, are within Commission
- guidelines. Sections 2.3.1 through 2.3.5 have been prepared in accordance with the review procedures described in the Standard Re.'tew Plan (NUREG-0800), util-tzing information presented in Section 2.3, of the FSAR, responses to requests for additional information, and generally available reference materials as de-scribed in the appropriate sections of the Standard Review Plan.
2.3.1 Regional Climatology The plant is located in southwest Washington on a ridge near the eastern shore of the Pacific Ocean and in a maritime type of climate. r
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t i . Maritime air masses dominate the region year round. The mean annual temperature in the area is about 10*C (50 F) ranging from about 2.9'C (37.2 F) in January to about 17.6*C (63.6 F) in July. Annual precipitation in the area is about 1270 mm (50 inches). ThemovabentofweathersystemsfromthePacificOce,anoverthesitemaintains considerable cloudiness and nearly constant rainfall due to lifting of this moist air ove'r the mountains in tSis area of Washington. Severe weather phenoment 'which affect the site area include about 5 thunderstorms per year which can be expected on about 5 days each year. Thunderstorms occur primarily in spring and summer. Considering the small frequency of thunderstorms, the
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applicant has estimated the number of lightning strikes to the sguare ' kilometer containing the plant to be 1.1 per year. Hail does not usually occur with the thunderstorms and is not a significant phenomena. - Tornadoes are not common in the region. For a two degree latitude-longitude - 2
" square", 33127 square kilom,eters (17791 miles ) containing the site, 5 tornadoes were reported for the period 1954-1981. Using an observed tornado path area of
.088 square Km (.034 sq. miles), the computed probability of occurrence for r tor- _.
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nado at the plant site is about .4.9 x 10 ' per year. The applicant has followed the recommendations of Regulatory Guide 1.76, " Design Basis Tornado for Nuclear Power Plants," for this region of the country. The applicant's design basis tor- " nado for category I structures has a,107m/s (240 mph), rotational velocity with a translational velocity of;27m/s S 0 mph), a total pressure drop of 2.25 psi and an average rate of pressure drop of 1,.2 psi in 1 second. The tornado winds andtotalpressuredropareconsgtentwithRegulatoryGuide1.76forthesite area. $ . High wind speed occurrences in the area are usually associated with severe . thunderstorms and intense extratropical cyclones. The highest " fastest mile" wind speed reported at Olympia, WA was.27m/s (60 mph) in November 1958. -The ~ applicant has selected an operating basis wind speed (defined as the " fastest mile" wind speed at a height of 9.1m (30 fee,t) with a return period of 100 years) to be 46.9 m/s (105 mph) for consideration in' plant design. , b 2-10 WNp-3 DSER SEC 2
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1 Y , [ . 4 Since the ultimate heat sink for the plant is a dry cooling tower, meteorologi-cal conditions related to extreme temperature and wind speed are relevant to determination of the adequacy of the tower to perform its function for a 30 day period. A maximum hourly temperature of 38.6'C (101.5"F) was used for design of the plant ultimate heat sink, dry cooling towers. This temperature was statistically determined to be the 40 year maximum hourly average temperature by. correlating short term site data with the same period at Elma, WA. Based
- upon this statistical analysis the maximum hourly average temperature in 40 years at Elma, WA of 40.6'C (105'F) was converted to a maximum hourly average temperature'onsite for the 40 year period of 38.6'C (101.5'C). This estimate of the long-term maximum temperature is reasonable and is therefore a conserva-tive basis for the design of the dry cooling towers thereby sat'isfying the intent of Regulatory Guide 1.27 for 30 day cooling capability. Heavy snowfall
! is not common in the region; roof loads may accumulate due to a wintertime precipitation comprised primarily of a mixture of snow and rain. Maximum l ] monthly snowfall observed at Olympia was 526 mm (20.5 inches) in January 1972 l
which was also the maximum snowfall in a 24-hour period. Ice storms, which can plug drains and scuppers as well as disrupt offsite power, are relatively infrequent. The estimate of th'e snowpack based on ANSI 58.1-1982, extrapolated
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from the 50 year return period in the standard to a 100 year return period, produces a weight of less than 15 psf. This snowpack weight, when added to the weight produced by the 48-hour probable maximum winter precipitation produces a " load of less than the design snowload of 80 psf, which was utilized in the design combined snow and ice load of Category I structures. During the 5 year i period 1960-1964, about 41 atmospheric stagnation cases defined as persisting for two days or longer totaling at lust 164 days were reported in the area. As discussed above, the staff has reviewed available information relative to the regional meteorological conditions of importance to the safe design and siting of this plant in accordance with the criteria contained in Section 2.3.1
- of the Standard Review Plan. Based on this review, the staff concludes that the applicant has' identified appropriate regional meteorological conditions for consideration in the design and siting of this plant. The applicant has met the. requirements of 10 CFR Part 100.10 and 10 CFR Part 50, Appendix A, General Design Criterion 2. The design basis tornaco characteristics selected by the 4 . _
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- b applicant conform to the position set forth in Regulatory Guide 1.76, and, therefore, meet the requirement of 10 CFR Part 50, Appendix A, General Design Criterion 4 to determine an acceptable design basis tornado for missile generation.
2.3.2 Local Meteorology 1 Climatological data from Olympia, WA, and nearby climatological cooperative stations and available onsite data have been used to assess local meteor-ological characteristics of the plant site. - Extreme temperatures of -22.2*C (-8'F) and 40.6*C (105'F) have been reported
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in the area. The applicant has considered a summer outdoor des ~ign temperature ~ of 28.3*C (83*F) and a winter minimum temperature of 4.4*C (40*F) in the' design of all heating, ventilation and air conditioning (HVAC) systems to maintain a 23.9'C (75'F) control room area temperature during normal or accident conditions.
-Regional analyses in NUREG/CR-1390, " Probability Estimates of Temperature Extremes for the Contiguous United. States" show that an ambient temperature of 35'C (95'F) will be exceeded for at least one hour every two years, on the average, and that an ambient temperature of about 42.2*C (108'F) will be _
exceeded at least one hour every 100 years, on the average. Also, an ambient i temperature of less than -8.9'C (16*F) is expected to occur for at least one hour every two years, on the average, and an ambient temperature of less than -#
-22'C (-8'F) is expected to occur for at least one hour every 100 years, on the average. Further justification of the adequacy of the ambient extreme tempera-tures considered by the applicant for the design of HVAC systems protecting safety-related auxiliary systems and components is required. This will be an open issue only if exceedence of extreme design temperatures for the HVAC system results in failure or malfunction of Category I auxiliary systems and components, which is being evaluated by the Auxiliary Systems Branch.
Precipitation is observed throughout the year, ranging from over 203 mm (8 inches) _in December to less than 25 mm (1 inch) in July. Maximum and minimum monthly i amounts of precipitation observed at Olympia have been 504 mm (19.8 inches) in January 1953 and 0 mm (0 inches) in August 1946, respectively. The maximum amount of precipitation in a 24 hour period at Olympia was 125 mm (4.93 inches) in February 1951. - _ 2-12 WNP-3 DSER SEC 2
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Average annual precipitation at Olympia is about 1290 mm (51 inches) and onsite precipitation measurements for the 2 year period 1970-1981 presented by the applicant indicate annual precipitation of about 1600 mm (63 inches). These differences can be attributed to the different periods of record and primarily terrain differences between the two locations. Wind data taken from the 10 meter level of the onsite meteorological tower for a 2 year period (October 1979 - September 1981), as summarized by the applicant, indicate prevailing winds from the southwest (20%) with a secondary peak frequency from the northeast (8.0%). The mean annual wind speed observed at the 10 meter level of the onsite meteorological tower for the period 1970-1981 was
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about 1.6 m/sec (4 mph), with calm conditions (defined as wind ' speeds less than the starting threshold of the anemometer) occurring almost 9.4% of the time. Atmospheric stability assessments, based on vertical temperature difference measurements for the 2 year period (1979-1981), have been summarized by the applicant for the 10 meter (30 feet) - 60 meter (197 feet) layer. Unstable conditions (indicating rapid diffusion rates) do not occur very frequently. Neutral and stable conditions predominate and occur 99% of the time, reflecting _ the cloudy and rainy conditions existing in the region. As discussed above, the staff has reviewed available information relative to ~e local meteorological conditions of importance to the safe design and siting of this plant in accordance with the criteria contained in Section 2.3.2 of the Standard Review Plan. The staff concludes that, with the exception of the design basis temperatures for HVAC systems, the applicant has identified and considered appropriate local meteorological conditions in the design and siting of this plant, and, therefore, meets the requirements of 10 CFR Part 100.10 and 10 CFR Part 50, Appendix A, General Design Criterion 2. 2.3.3 Onsite Meteorological Measurements Program The onsite pre-operational meteorological measurements program was initiated at the WNP-3 site in 1973 and ended in 1975. They re-commenced in 1979 and ended in 1981. Measurements were maoe on a tower extending 60 meters (197 feet) above grade. The tower is located about 1207 (3/4 mile) northwest of the . 2-13 WNP-3 DSER SEC 2
.- . _ . - . - - . - .- . . . . - . . - . _ - .. -- . - ~ ..- . --
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- plant structures. The following meteorological measurements were made on the tower
- wind speed and direction at the 10 meters, and 60 meters levels; f
vertical temperature gradient between the 60 meters and 20 meter levels. l Ambient temperature was measured at the 10 meter level and dewpoint at the j 60 meters level. Precipitation was measured at ground level near the tower. i ~ i A digital data acquisition system, backed up by analog strip charts, wa's used
; to record meteorological data. Daily checks and quarterly calibration were done on the equipment. The joint data recovery for wind speed, and wind direction at the 10 meter level, and atmospheric stability (defined by the j vertical temperature difference between the 60 meter and 10 meter levels for
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j the 2 year period October 1979 - September 1981 presented in the FSAR wa's in ~
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I excess of 95%. i
- The meteorological measurements system complies with the accuracy specifications
) in Regulatory Guide 1.23, "Onsite Meteorologia1 Programs." The representative-a ness of the 2 year period of onsite data to long term conditions was determined by comparisons of data from the site to measurements at Olympia and nearby
- locations. These comparisons indicate that reasonable estimates of atmospheric _
j dispersion for accidental and routine releases of radioactive effluents can be f made from the onsite data record. a The meteorological measurements and data collection program has been terminated after the decision to postpone or possibly cancel completion of Units 3 and 5 was made by WPPSS. 4 If the project is resumed, the meteorological program on the 60 meter tower will be re-activated and will serve as the operational meteorological monitor-ing system.as well as for emergency preparedness requirements. The meteor-
~ ological measurements will be available in the control room as well as in the emergency operation facility (EOF) and technical support center (TSC).
- The meteorological program described above meets the criteria for meteorological measurements during plant operation and as part of the emergency response capa-
) bil'i ty. Any meteorology measurement upgraces.must be completed in accordance with the schedule of NUREG-0737, III.A.2, " Clarification of TMI Action Plan . Requirements," and its supplement, and a post implementation staff ' review will 2-14 WNP-3 DSER SEC,2 i
be conducted. The incorporation of current meteorological information into a . real-time atmospheric dispersion model for dose assessments will also be l l considered as part of the upgraded capability. ' The staff has reviewed the onsite meteorological measurements system in accor-dance with the criteria contained in Section 2.3.3 of the Standard Review Plan. The meteorological measurements program has provided data to represent onsite meteorological conditions as required in 10 CFR Part 100,10. The staff concludes that the historical site data provide a reasonable basis for making assessments of atmospheric dispersion conditions for estimating consequences of design basis accident and routine releases from the plant. 2.3.4 Short-Term (Accident) Diffusion Estimates To audit the applicant's assessments, the staff has performed an independent assessment of short-term (less than 30 days) accidental releases from buildings and vents using the direction-dependent atmospheric dispersion model described in Regulatory Guide 1.145, " Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," with consideration of increased __ lateral dispersicn during stable conditions accompanied by low wind speeds. Two years (October 1979 - September 1981) of onsite data available to the staff on magnetic tape, which had 95% data recovery, were used for this evaluation. - Wind speed and wind direction data were based on measurements at the 10 meter level and atmospheric stability was defined by the vertical temperature gradient measured between the 60 meter and 10 meter levels. A ground-level release with a building wake factor, cA, of 2133 meter
- was assumed. The relative concen-
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tration (X/Q) for the 0-2 hour time period was determined to be 4.1 x 10
- sec/m' at an exclusion area boundary distance of 1311 meter (.81 miles) in the south sector. The X/Q values for appropriate time periods at the outer boun-dary of the low population zone 4827 meter (3 miles) are:
Time Period X/Q(sec/m}- 0-8 hours 6.0 x 10 5
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8-24 hours 4.0 x 10 5
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1-4 days 1.6 x 10 5
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4-30 days 4.3 x 10 5 _ 2-15 WNP-3 DSER SEC 2
i The applicant has calculated a higher X/Q value for the 0-2 hour time period at the cxclusion area and low population zone boundary than that calculated by the staff. These differences may be attributed to slightly different use of the models and the data by the staff and the applicant. Based on the above staff and applicant evaluations performed in accordance with the criteria contained in Section 2.3.4 of the Standard Review Plan, the staff concludes that the applicant has conservatively considered atmospheric disper-sion conditions at the exclusion area and low population zone boundaries for assessments of the consequences of radioactive releases for design basis acci-dents in accordance with the requirements of 10 CFR Part 100.11. The atmospheric dispersion estimate's provided above which were independently calculated by the' staff have been used by the staff in an independent assessment of the consequen-ces of radioactive releases for design basis accidents. 2.3.5 Long-Term (Routine) Diffusion Estimates To audit the applicant's estimates, the staff performed an independent calcula-tion of annual average relative concentration (X/Q) and relative deposition (D/Q) m values. Annual average relative concentration (X/Q) and relative deposition (D/Q) -- values at specific receptor points and in arrays to 80 Km (50 miles) for use in population dose assessment were based on the straight-line gaussian atmos-pheric dispersion model, described in Regulatory Guide 1.111, modified to reflect spatial and temporal variations in airflow as described in NUREG/CR 2919. Continuous and periodic releases through the plant vents were considered as ground level releases. The results of the evaluation were published in the DES. The staff compared its long-term diffusion estimates with those provided by the applicant and determined the analyses were in general agreement. The staff concludes that the applicant has considered representative atmospheric dispersion estimates for demonstrating compliance with the numerical guides for doses contained in 10 CFR Part 50, Appenoix I. 2-16 WNP-3 DSER SEC 2
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. i 2.4 Hydrologic Engineering !
The staff has reviewed the hydrologic engineering aspects of the applicant's design, design criteria, and design bases for safety-related facilities at the Washington Public Power Supply System Nuclear Project No. 3 (WNP-3). The acceptance criteria used as a basis for staff evaluations are set forth in SRP 2.4.1 through 2.4.14 (NUREG-0800). These acceptance criteria include the applicable GDC reactor site criteria (10 CFR 100), and standards for protection against radiation (10 CFR 20, Appendix B, Table II). Guidelines for implemen-tation of the requirements of the acceptance criteria are provided in RGs, ANSI standards, and Branch Technical Positions (BTPs) identified in SRP 2.4.1 through 2.4.14. Conformance to the acceptance criteria provides the ba'ses for concluding that the site and facilities meet the requirements of 10 CFR 20, 50, and'100 with respect to hydrologic engineering. 2.4.1 Hydrologic Description WNP-3 is located in Satsop, Washington, approximately 1.4 miles south of the Chehalis River near the confluence of the Satsop River. The site is about _ 26 miles west of Olympia and about 16 miles east of Aberdeen, Washington. As shown on Figure 2.4.1, WNP-3 is situated on a ridge between Workman Creek and the Chehalis River. " The Chehalts River which heads in the Willapa hills in southwest Washington, flows generally eastward to the city of Chehalis where it changes its course abruptly to the north. About 10 miles north of C'hehalis, near Grand Mound, the river flows northwesterly to Elma, then west to Grays Harbor at Aberdeen. The river and its tributaries have a drainage area of about 2,115 mi8 The drainage area at the site, including the Satsop River, is about 1,765 mi2 The average annual flow at this location is about 6,820 cfs. The Chehalis River basin is shown on Figure 2.4.2. - The major tributaries of the Chehalis River in the vicinity of the site are the Satsop and Wynocchee Rivers. The Satsop River has a drainage area of about 300 mi2 and an average annual flow of aoout 2030 cfs. The Wynocchee River has a drainage area of about 100 mi* and an average annual flow of about 1200 cfs. _ 2-17 WNP-3 DSER SEC 2
Both of these tributaries rise on the south side of the Olympic Mountains and flow southward to their confluences with the Chehalis River. A number of smali tributaries to the :outh of the Chehalis River head in the hills sur-rounding the site. These include: Elizabeth Creek, Hyatt Creek, Fuller Creek, Purgatory Creek and Workman Creek. All of these streams are relatively short, intermittant streams, originating at elevations between 300 to 400 feet above mean sea level (MSL). As shown on Figure 2.4.2 there are two dams and associated reservoirs on the tributaries of the Chehalis River. The Wynocchee dam and lake, which is a Corps of Engineers project, provides water supply for industry and agriculture The lake also offers recreation opportunities -
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and storage for flood control. for the public. The Skookumchuck dam and reservoir project is operated 'by the Pacific Power and Light Company (PP&Ls). The reservoir provides make-up water I for PP&L's Centralia Steam Electric Station. The water resources of the Chehalis River valley include both surface and ground supplies. Within 5 miles of the plant, surface water permits have been granted by the Washington State Department of Ecology to about 78 users. Most surface -_ water is used for irrigation, with the remainder for domestic use, livestock j watering, fish propagation, fire protection and industrial use. Except for a single domestic water user located within a mile downstream of the plant, -~ there are no other known users of Chehalis River water for domestic purposes between the plant and Grays Harbor. Groundwater in the Chehalis River valley is obtained from shallow wells which tap the alluvial aquifer and is used mostly for drinking and irrigation. There are 45 known wells within 2 miles of the plant. Five major municipal water systems within 20 miles of the site are served partially or totally by
. groundwater.
The applicant has provided hydrologic descriptions of the plant site and vicinity. The staff has reviewed the applicant's information in accordance with procedures in SRP 2.4.1. The staff concludes that the requirements of GDC 2 ; and 10 CFR Part 100, with respect to general hydrologic descriptions, have been l met. -
'2-18 WNP-3 DSER SEC 2
2.4.2 Floods 2.4.2.1 Flood Design Considerations Five potential sources of site flooding were considered by the applicant: (1) intense local precipitation on the plant yard; (2) floods on the Chehalis River; ' (3) dam failures; (4) surges and seiches; (5) tsunamis. The staff has reviewed the material presented by the applicant in accordance with procedures in SRP 2.4.2 and concludes that in addition to these five flooding sources, other sources of potential flooding of the plant site are the small creeks near the site. Since the applicant did not address potential ' flooding from these small creeks, the staff made an independent evaluation as described in Section 2.4.3. 2.4.2.2 Effects of Local Intense Precipitation' 3 At WNP-3, a site drainage system consisting of catch basins, drain pipes, and ~; - ditches has been provided to carry surface runoff south to Workman Creek and north to the Chehalis River via Fuller and Purgatory Creeks. The drainage sys-tem is designed for a 100 year recurrence storm with pipes flowing half full. The intensity of this storm is 2.9 inches per hour. This is less than the probable maximum precipitation (PMP) so during a PMP event, some water could pond on the site. PMP is the estimated depth of precipitation (rainfall) for which there is virtually no risk of exceeding. At WNP-3, the PMP values used by the ~ ~ applicant for durations of 1 through 6, 12, 24, and 48 hours are as follows: M w ! 2-19 WNP-3 DSER SEC 2 l
l Duration Incremental PMP Total PMP (hours) (inches) (inches) 1 4.32 4.32 2 1.92 6.24 3 2.24 8.48 6 6.24 14.72 12 5.44 20.26 24 5.12 25.28 48 3.00 28.28 These PMP values were determined from Technical Paper 40 (U.S. Weather Bureau, 1963). Although Technical Paper 40 does not present PMP estimates, it does - - give a method for determining PMP values from 100 year rainfall values which e given in the paper. The applicant states that the 100 year rainfall values used to obtain the PMP values above, are more conservative than those given in NOAA Atlas No. 2, Volume IX-Washington (NOAA, 1973). In addition, these PMP values are also more conservative than those given in Hydrometeorological Report 43 (U.S. Weather Bureau, 1966). The staff has reviewed the three PMP references used by the applicant and agrees that Technical Paper 40 results in more conservative PMP estimates than the other two references. The staff concludes that the PMP values used . by the applicant are appropriate for the evaluation of site and roof drainage. The minimum elevation of openings to safety-related structures is 390.5 feet MSL. The elevation of the paved road that surrounds the plant is 390.0 feet and the grassed areas within this road are at elevation 389.5 feet MSL. Because the grassed areas are 1 feet lower than openings to safety-related structures, water could pond I feet deep in these areas before any structure or equipment would be affected. Using the PMP rainfall values tabulated above, the applicant estimated the maximum depth of ponding in the grassed areas would be 2.0 inches assuming that the site drainage system is functioning as designed. The applicant states in the FSAR, that in the unlikely event thst 1 y drain is completely blocked, rainwater could pond to elevation 390 feet MSL, which is the elevation of the
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2-20 WNP-3 DSFR 9Fr ?
! crown of the surrounding access road. Subsequent runoff would overflow the I road to lower-lying areas. Since exterior door openings are at elevation j 390.5 feet MSL, the applicant concluded that water from local intense precipita-tion would not enter safety-related buildings. The staff has reviewed the material presented by the applicant and concludes that the applicant has not provided sufficient information to support its conclusion that local floods will r.at enter safety-related buildings. From topographic maps provided by the applicant, it appears that because the areas to the south, east and west slope upward from the plant; water could pond to a higher elevation than the top of the access road, possibly entering safety-
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related structures. The staff will require that the applicant identify th? areas where water will pond before overflowing the access road and the areas where water will drain to once it overflows the road. In identifying these areas, obstructions to flow such as temporary and permanent buildings, trailers, sheds, etc., should also be shown. It should also be assumed that all of the site drainage system is blocked. Once flow areas have been identified, the applicant should provide - assurances that the flow areas are of sufficient capacity to prevent water , from ponding to excessive levels. The applicant should also discuss whether the fence which surrounds the site will adversely obstruct flows. " In discussing the effects of local intense precipitation on roofs of safety-related buildings, the applicant stated that the roof drainage system is designed for a 100 year recurrence storm at 50 percent capacity. With the exception of a roof section adjacent to the steam tunnel, safety-related buildings have parapets that are 12 inches above the roof high points and 18 inches above the low points. Thus the average depth of ponding would be
. about 15 inches. The applicant states that roofs are designed to support the load induced by up to 15.4 inches of water.
The roof section adjacent to the steam tunnel at elevation 417.5 feet MSL is surrounded by walls to elevation 443.5 feet MSL or higher. The appli. cant stated that this roof is capable of supporting the load induced by a 48 hour PMP plus the 100 year snow pack. This is equal to 28.28 inches of rainfall - plus a snow-depth water equivalent of 3.78 inches or 32.06 inches. 2-21 WNP-3 DSER SEC 2
The applicant does not state whether the surrounding walls are equipped with scuppers or other means of limiting water depths. Therefore, it is possible that water could pond to a much greater depth than 32.06 inches because the walls are at least 26 feet high (443.5 feet-417.5 feet). The applicant should thus consider rainfall for durations greater than 48 hours. The effect of a 26 foot depth of water should be addressed unless it can be demonstrated that water cannot physically pond to this depth. Alternately, the applicant should consider putting scuppers (or other devices) in walls to limit water depths l to those that can be supported safely by the roof section adjacent to the steam tunnel.
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The staff has reviewed the material presented by the applicant in the FSAR, using the procedures described in SRP Section 2.4.2. Based on this review, the staff concludes that the applicant has not provided sufficient information to support its conclusions that ponded water will not enter safety-related buildings or that the roof section adjacent to the steam tunnel is capable of supporting potential rainfall loads. Thus the staff cannot conclude at this time, that the plant meets the requirements of GDC 2 with respect to flooding by intense local precipitation. m The staff, however, does conclude that during a PMP event, water levels on roofs of safety-related structures will remain at or below the levels - determined by the applicant except for the roof adjacent to the steam tunnel. 2.4.3 Probable Maximum Flood on Streams and Rivers The Probable Maximum Flood (PMF) is defined as the hypothetical precipitation-induced flood that is considered to be the most severe reasonably possible. Severe rainfall storms in western Washington occur mostly in the winter months when there is snow on the ground. Consequently, the applicant estimated the PMF for the Chehalis River based on PMP and snowmelt over the drainage basin. As a first step in estimating the PMF, the applicant subdivided the Chehalis River drainage basin into three subbasins ano developed individual unit
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l 2-22 WNP-3 DSER SEC 2
hydrographs for each subbasin using Corps of Engineers procedures. A flood hydrograph was then developed for each subbasin using the Corps of Engineers flood hydrograph computer program, HEC-1. To determine a PMF for each subbasin, each hydrograph was then increased by the average annual river flow to account for base flow conditions. The three individual PMF's were then routed, where appropriate, and combined at the site to form a single PMF. To account for potential antecedent flood conditions as recommended in RG 1.59, the applicant assumed that a storm equal to 50 percent of the PMP would occur three days prior to the PMP storm. The antecedent flood resulting from these conditions was combined with the PMF at the site. The resultant hydrograph had a peak discharge of 353,000 cfs. PMF water levels in the Chehalis River adjacent to the site were determined by means of the Corps of Engineers HEC-2, " Water Surface Profiles" computer program. The PMF discharge (353,000 cfs) stillwater level was estimated to be 53.1 feet MSL. The applicant determined that coincident wind-wave activity would result in a maximum wave runup, including wind setup, of 23.1 feet. Adding this to the PMF stillwater level resulted in a maximum flood level of 76.2 feet MSL. Since the plant is at elevation 390 feet MSL, the applicant concluded that no safety- c_ related structures would be affected by a PMF on the Chehalis River. The applicant did not address the potential for flooding from the small creeks -' near the site; however, based upon the following observations, the staff con-cluded that floods on these small creeks will not affect the safety of the plant.
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Workman Creek which runs south of the plant has a stream bed elevation which is more than 200 feet lower than the plant grade elevation. Because the drainage area of this creek is small, less than 10 miles ,2 the staff concludes that a PMF . would not rise 200 feet in the creek. In addition to Workman Creek, there are two other creeks in the vicinity, Fuller Creek and Purgatory Creek. Both of these have drainage areas of less than one mile 2and each creek flows away from the plant. The staff thus concludes that floods on these creeks will not affect the safe operation of WNP-3. O mm 2-23 WNP-3 DSER SEC 2
At the CP stage, the staff reviewed the applicant's analyses and the effects of coincident wind wave activities. The staff concurred then with the applicant's analyses and concluded that there is no potential danger to safety-related structures due to the PMF with coincident waves. The staff has reviewed the FSAR material presented by the applicant in accordance with procedures described in SRP 2.4.2 and 2.4.3. Based on this review, the staff concludes th.u. the plant meets the guidelines of RG 1.59, " Design Basis Floods for Nuclear Power Plants", and the requirements of GOC 2 with respect to flooding from the Chehalis River and the small creeks adjacent to the site. 2.4.4 Potential Dam Failures As shown in Figure 2.4.2, the only dam located upstream of the plant is ' Skookumchuck Dam. The applicant estimated the effect of a failure of this dam-coincident with a Standard Project Flood (SPF) equal to one half of the PMF. It was assumed that the reservoir would be full at the time the dam failed. Using Corps of Engineers procedures, the applicant estimated a dam failure hydro-graph with a peak discharge of 260,000 cfs. This hydrograph was routed down-stream and combined with SPF's from the Chehalis~ River sub-basins. The resultant < peak flow at the site was determined to be 182,000 cfs. Using-the same proce-dures as were used to determine PMF levels, the applicant determined the flood level at the site due to failure of Skookumchuck dam would be 39.6 feet MSL. -- Since this elevation is less than the PMF level, the applicant concluded that failure of Skookumchuck Dam will not affect the safety of the plant. The staff reviewed the applicant's dam failure analysis at the CP stage and concluded that there would be no flooding of the plant due to das failures. The staff has reviewed Section 2.4.4 of the FSAR, using the procedures described in SRP 2.4.4. The staff concurs that conservative procedures have been used and that potential dam failures pose no threat to the plant. Thus the staff concludes that the plant meets the requirements of GDC 2 and 10 CFR 100, Appendix A, with respect to flooding by dam failures. r o
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2-24 WNP-3 DSER SEC 2
2.4.5 Probable Maximum Surge and Seiche Flooding There are.no historical seiche records at Grays Harbor, the largest body of
, water near the site, and the probability of seiche occurrence at Grays Harbor is extremely remote due to the shallow tidal flats. The plant site located 1 at River Mile 21 and elevation 390 feet MSL is not susceptible to surge or
- seiche flooding, i
l 2.4.6 Probable Maximum Tsunami Flooding ] WNP-3 is located about 30 miles inland from the Pacific Ocean with all safety-l related equipment'at an elevation of 390 feet MSL or higher. The maximum histo-
! rical tsunami wave height originating within 1000 miles of the site was'32.8 feet '
i 1 and occurred in Cook Inlet, Alaska, in 1901. The most damaging tsunami of local . 1 origin, for the Washington coast area, was generated by the Alaskan Earthquake of 1964 (epicenter more than 1000 miles from the site), and caused minor damage i at the Ocean Shores Development which is located on the spit that protects the i Grays Harbor entrance. The elevation of the highest wave was 13.3 feet MSL;
; causing a break in the. sand dune dike and the deposition of winter storm debris __
along the spit. There was no overtopping of the spit and no flooding resulted i at Aberdeen. 4 S: f The applicant estimated that a probable maximum tsunami (PNT) approaching the I Chehalis River through Grays Harbor would result in only a 3.5 foot increase in I water level at the mouth of the Chehalis River. The effects of the PMT would be reduced to negligible heights at the plant site because of attenuation in the river channel.
- The staff has reviewed the applicant's tabulation of historical tsunami and
. Its estimate of the PMT at the site. Based on procedures in SRP 2.4.6, the i staff concludes that no credible tsunami event could threaten the plant ~Thus, the requirements of GDC 2 as it relates to structures,. systems and components important to safety being designed to withstand the effects of tsunami, have l been met.
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2.4.7 Ice Effects The Pacific ocean, which is about 30 miles west of the site, greatly influences the climate in the WNP-3 area. The ocean acts to moderate the seasonal and daily variability in climate throughout the year such that winters are warmer than at other locations at similar latitudes. Because of this, there are no conditions which might produce a permanent ice cover or ice jam on the Chehalis River. In addition, because of the large difference in elevation between the Chehalis River and the plant (Section 2.4.3), even if ice jams did form, floods resulting from these jams would not affect the safe operation of the plant. Water required for normal operation of WNP-3 will be supplied from groun'dwater infiltration-type structures (Ranney Wells). Therefore, potential icing will 4 not affect the normal plant water supply. Emergency safe shutdown and cool i down of WNP-3 can be accomplished using the ultimate heat sink which consists of dry cooling towers located adjacent to the reactor auxiliary building. Make-up water is not required for the dry cooling towers and during periods of low temperature the design of the towers prevents freezing of the tower or pipelines. __ The staff has reviewed the information provided by the applicant concerning ice effects, in accordance with procedures in SRP 2.4.7. The staff concludes ~= 1 that icing will not affect the safe operation of the plant. 2.4.8 Cooling Water Canals and Reservoirs 4 There are no safety-related or other cooling water canals or reservoirs 4
- associated with WNP-3.
. 2.4.9 Channel Diversions The Chehalis is a meandering river that shows a number of former channel
- locations, oxbows and sloughs in the vicinity of the plant (see Figure 2.4.1).
As described in Section 2.4.11.1, the source of makeup water for the WNP-3 cooling tower is the alluvial aquifer that underlies the Chehalis River flood- _ plain. Recharge to the aquifer occurs all across the river valley as well as
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2-26 WNP-3 DSER SEC 2
in the river channel and from the 70 to 80 inches of annual rainfall in the area. The aquifer reacts much like a reservoir by accepting and storing sur-face inflow during periods of high river flows and high rainfall and releasing the stored water when rainfall and river flows aren't as plentiful. Since the aquifer is recharged by both rainfall and the river across the entire valley it is possible for the river channel to meander considerably before the makeup water capability would be affected. Water for plant use is withdrawn from the aquifer by means of two Ranney wells
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located as shown on Figure 2.4.1. As part of the design to place Ranney col-lectors in the floodplain, the adjacent river banks are being stabilized to minimize erosion. However, because the Chehalis is a meandering river, display-ing oxbows and sloughs, diversion affecting the Ranney wells is still considered possible. Regardless of the availability of the Ranney wells, the safety of the plant will nut be jeopardized because, as described in Section 2.4.11.2, emergency cooling of the plant can be accomplished using the ultimate heat sink, which consists of a dry cooling tower. The staff thus concludes that potential channel diversions, although remote, __ present no safety-related hazard to the plant and that the requirements of 10 CFR Part 100, relative to channel diversions, have been met.
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2.4.10 Flood Protection Requirements As described in Section 2.4.3, the staff concluded that the plant is located considerably higher than any credible flood in the Chehalis River. However, in Section 2.4.2, the staff concluded that the applicant had not provided sufficient information to support its conclusion that local irtense precipita-tion will not enter safety related buildings. Additionally, it is not evident
. that a roof section adjacent to the steam tunnel has been designed to support potential ponded rainfall. The applicant will be required to provide additional information in support of its conclusions. Resolution of these issues will be addressed in a supplement to this SER.
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I 2.4.11 Cooling Water Supply 2.4.11.1 Normal Water Supply Under normal operating conditions, waste heat will be dissipated to the atmosp-here by a natural draft cooling tower. Makeup water to replace the water lost by evaporation, blowdown and drift, will be supplied to the cooling tower by two Ranney wells located in the alluvial aquifer which underlies the Chehalis River valley (see Figure 2.4.1). To prevent excessive buildup of dissolved solids in the cooling system, a certain amount of cooling water must be contin-uously discharged to the Chehalis River after first being cooled down by a
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supplemental cooling facility. The blowdown discharge will be' diluted in the~ river through the use of a submergad multiport diffuser. The maximum makeup water requirement for WNP-3 is approximately 18,000 gpm (40.0 cfs), and a single Ranney well is capable of supplying this amount on a ' continuous basis. The capability of the Ranney wells to supply this quantity of water is independent of low flows in the Chehalis River. The State Energy Facility Site Evaluation Council (EFSEC), however, has administratively estab- - lished that plant makeup withdrawal (except for a hot-standby maintenance flow of 2 cfs) must cease when the daily river flow goes below 550 cfs. Additionally, plant withdrawal may not exceed the difference between the river flow and -- 550 cfs. The long-term annual average flow of the Chehalis River at the site is estimated to be about 6820 cfs. The estimated average monthly flows vary from 730 cfs in August to 14,900 cfs in January. The minimum and maximum historical flows at the site are estimated to be about 400 and 97,100 cfs, respectively. Because of the water withdrawal limitation established by the EFSEC, the plant . will have to be shut down whenever the daily river flow goes below 550 cfs. The applicant estimates that, on the average, this will occur about four' days a year. The applicant has completed a contract with the city of Aberdeen to purchase releases of 62 cfs of flow from the Wynocchee Reservoir to supplement the
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2-28 WNP-3 DSER SEC 2
Chehalis River during low flow periods. This water is to be used to mitigate adverse impacts associated with the consumptive use of river water. The staff has reviewed the material presented by the applicant and concludes that neither drought periods nor the conditions established by the EFSEC, re-garding withdrawal of water for plant use, will unduly restrict the availability of cooling water for normal operations as required by GDC 44. 2.4.11.2 Emergency Water Supply Emergency safe shutdown and cooldown of WNP-3 can be accomplished using the
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ultimate heat sink which consists of a dry cooling tower located adjacent to ~ the reactor auxiliary building. The UHS is also used for heat rejection'during normal operation and shutdown. The UHS is comprised of two independent 100 per-cent capacity trains. Each train has a cooling tower of the aircooled heat exchange type, with Component Cooling Water System (CCWS) fluid passing through the tube side and air over the exterior surface of the tubes. The cooling tower fans cycle on and off and through low and high speeds automatically dur-ing all operating modes to maintain the desired temperature. __ The UHS is capable of operating under a range of heat loads during normal and emergency conditions. With a dry bulb ambient temperature of 101.5*F, each -' train has a design heat rejection capability of 180 x 10' Btu /hr with 11,000 gpm CCWS flow entering the cooling tower at 153*F and leaving at 120 F. During normal operation, the UHS in conjunction with the CCWS heat exchanger can reject the normal heat loads while maintaining CCWS temperatures at or below 95*F. For accident conditions the UHS is designed to reject the maximum accident heat loads without the need for additional cooling by the CCWS heat exchanger. By operating the fans at full speed, one cooling tower can maintain the CCWS temperature at or below 120*F during accident conditions. As suggested in R.G. 1.27, the applicant analyzed the 30-day period following a design basis accident. This analysis showed that each cooling tower is cap-able of dissipating the maximum heat loao following a LOCA. By maintaining the temperature of the water exiting the cooling tower at 120 F or less, the maximum _ 2-29 WNP-3 DSER SEC 2
i . heat rejected during the 30 day analysis, was 133 x 10' Btu /hr. Since the , cooling tower has a design heat rejection capability of 180 x 108 Btu /hr, the applicant concluded that the UHS is capable of providing adequate cooling for l at least 30 days. The applicant thus concluded that the UHS meets the recom-mendations set forth in R.G. 1.27 and thus the requirements of GDC 44 with respect to thermal aspects of the heat transfer system. The staff has not completed its review of the UHS so a conclusion on its acceptability cannot be made at this time. -
! 2.4.12 Groundwater 2.4.12.1 Groundwater Conditions I
The plant area is underlain by the Astoria Formation which is a thick (2500 to 3000 feet) deposit of relatively impermeable tertiary sandstones. This formation
! dips northward toward the Chehalis River and forms the most extensive geologic unit at the site. All Category I structures are founded in the Astoria sand-stone approximately 2000 feet above its base. North of the plant and south of _
the Chehalis River, pleistocene terrace deposits overlie the Astoria Formation. 1 These deposits consist mostly of sands, gravels and silts. Further north, the i Chehalis River valley is underlain by alluvial materials. The Ranney wells --- which supply the cooling water required for normal operation of the plant are founded in these alluvial materials. In the site area, groundwater is found in the Astoria Formation, the pleistocene . deposits and in the alluvial materials in the Chehalis River floodplain. The Astoria Formation has very low permeability and permits only small amounts of i . recharge and minimal groundwater movement. Because of this, it is not a produc-tive groundwater source. The groundwater table beneath the plant site area follows the ground topography and is parallel to the weathered and unweathered zones of the Astoria sandstone. The groundwater slopes northward toward the Chehalis River. Prior to construction of the plant, the groundwater level was - at an elevation of about 380 feet MSL.
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Groundwater also occurs in a discontinuous manner in the pleistocene terrace deposits. Recharge is derived from infiltration of rainfall on the areas above the terrace levels and infiltration from the Chehalis River. There are no known major aquifers within these deposits and only three domestic wells tap the terrace groundwater in small perched aquifers. The only satisfactory source of groundwater in the site vicinity occurs in the alluvial aquifer that underlies the Chehalis River valley. This aquifer extends downward from about 10 feet below the ground surface to about 165 feet. The high permeability and transmissivity coefficients of this unconfined aquifer indicate that the aquifer reacts much like a reservoir and a hydraulic conduit. Recharge to the aquifer occurs all across the river valley asew'll as in the ~ - river channels from the 70 to 80 inches of annual precipitation in the area and from the high surface inflow from the widespread Chehalis River basin. The aquifer accepts surface water for storage during these periods until the under-ground storage is full. The perme.able aquifer discharges readily into streams and rivers during periods of low flow. The alluvial aquifer is limited hori-zontally by tertiary sandstone sediments on the south side of the river and by by the southern edge of the Olympic Mountains on the north side of the valley. _ The aquifer extends two miles across the Chehalis River valley, about 14 miles downstream to Grays Harbor and about 15 miles upstream to the eastern limit of Grays Harbor County. As described in Section 2.4.11.1, this aquifer will be - used to supply makeup water to the plant. 2.4.12.2 Dewatering System A permanent groundwater drainage system (GWDS) that operates solely by gravity has been installed around the WNP-3 reactor auxiliary building (RAB). The GWDS l consists of vertical 6 inch diameter half-round drain pipes spaced at 8.5 feet intervals around the RAB at the interface between the rock and exterior concrete walls. The vertical pipes, which drain the surrounding rock, rise from the base of the foundation mat to elevation 390 feet MSL except at the west side of the RAB where the vertical drain pipes extend into the turbine building, four feet above the floor elevation of 390 feet MSL. This elevation is above the highest level (3 feet) that tne applicant has estimated cir-culating water could rise, in the event cf a Circulating Water System break _ inside the building.
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2-31 WNP-3 DSER SEC 2
Thus any water released into the Turbine Building would be prevented from directly entering the GWDS. Groundwater that seeps into the vertical drain pipes is conveyed to a 8 inch diameter horizontal drain pipe located along the periphery of the mat. Collec-ted groundwater is then routed to a 6 foot diameter drainage tun mi that drains into a small tributary of Workman Creek south of the plant. In aucition, 8 inch diameter perforated undermat drains have been placed diagonally beneath the foundation mat. These undermat drains are also connected to the drainage tun-nels. Manholes are provided at each corner of the RAB to allow for periodic inspection and cleaning of the GWOS. . The GWDS is not classified as a seismic Category I system except for the man-holes at the corners of the RAB and the upper 5 feet of the extended vertical drain pipes at the west side wall of the RAB. The manholes are seismic Cate-gory I to provide access to the horizontal drain pipe and to the drainage tun-nel in the event of an earthquake. The upper portions of the vertical drain pipes are seismically qualified to resist the passive pressure of the sandstone on the embedded portion of the pipes and to resist the peak seismic accelera- _ tion of the RAB at grade elevation. The applicant has stated that in the unlikely event of a complete blockage of - the GWDS, the walls and foundation mat of the RAB are designed to withstand the resulting hydrostatic load of a groundwater level at elevation 365 feet MSL. This elevation is 39 feet above the bottom of the mat, elevation 326 feet MSL, and 24.5 feet below the plant grade elevation of 389.5 ft MSL. The applicant states that in the event of a complete blockage of the GWDS, there would be sufficient time to repair the system before the surrounding groundwater would rise to an elevation of 365 feet MSL. This time was estimated to be a minimum of 115 days. Using the procedures in SRP 2.4.12 including Branch Technical Position (BTB) HGEB-1, the staff has reviewed the information provided by the applicant in the FSAR. The staff concludes that there is inadequate information with which to assess the applicant's conclusion that in the event of a complete failure of the dewatering system there will be sufficient time to repair the system before the surrounding groundwater would rise tc an unacceptable level. The staff . _ 2-32 WNP-3 DSER SEC 2
Y has asked the applicant to provide additional information concerning the dewatering system. Until the staff receives the requested information, it cannot determine whether the plant design meets the criteria of BTP HGEB1 of SRP 2.4.12 or the requirements of GDC 4 with respect to the dewatering system. The applicant considered a break in the underground Circulating Water System (CWS) pipe ahd its effect on the ability of the dewatering system to maintain water levels below elevation 365 feet MSL. The applicant stated that if a major break in the CWS pipe was to occur, water would be forced upward out of the trench in which it lies, and would be drained away from the surface by storm
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drains. The staff has reviewed the information provided by the appitcant according to procedures described in BTP HGEB1 in SRP 2.12, and concludes that insufficient information has been provided by the applicant, concerning the effect of postu-lated pipe breaks on the dewatering system. The staff has submitted questions to the appiteant on the subject of pipe breaks. The staff will complete its review pending receipt of responses from the applicant concerning the effect __ of pipe breaks on the dewatering system. In Section 2.4.4 of the CP-SER, the staff stated that-the applicant had committed --- to monitor groundwater levels at the walls of the reactor auxiliary butiding and to radiologically monitor discharges through the groundwater drainage sys-tem. The applicant proposed to do this by installing one piezometer at each reactor auxiliary building wall, located between the manholes and 10 feet from the wall. This instrumentation was to provide an alarm in the control room if a specified level was exceeded. The monitoring program proposed by the appli-cant in the FSAR is considerably different than that proposed at the CP stage.
. The applicant did not indicate in the FSAR if any part or all of the CP stage monitoring program will be used during plant operation; instead, it stated that' inservice surveillance of the vertical drains, horizontal headers,'and the drain tunnels will be made at 90 day intervals during the wet season. The vertical !
drains will be' tested by dropping a light down the drain to the horizontal- ' header and observing the light from the manholes through the header. The hori-l zontal headers'will be inspected by shining a light at one end of each header _ _l
'l 2-33 WNP-3 DSER SEC 2 i
and observing it from the other end. The drainage tunnel will be inspected by walking along it from the manhole to the concrete plug and looking through the concrete plug to daylight. The staff agrees that this surveillance program will effectively indicate whether the dewatering system is functioning and whether there is any blockage. However, it will not be effective if there is standing water in the vertical drains. The applicant should explain if any part or all of the monitoring pro-gram committed to in the CP-SER is to be used to monitor the performance of the dewatering system during operation. If it is not to be used, the applicant
- should explain why not. In any event, the applicant should describe the proce-
~^
dures to be used to monitor groundwater levels in the event that there is stand-ing water in the vertical drains. The staff has reviewed the information provided by the applicant and concludes that it cannot complete its review of the dewatering system because the applicant has not provided sufficient information to assess the following: (1) The time for groundwater levels to rebound to the hydrostatic design levc1 _- of 326 feet MSL in the event of a complete failure of the dewatering system. (2) The adequacy of the in-service monitoring program. -"- (3) The effects of pipe-breaks on the dewatering system. The staff has submitted questions to the applicant and will complete its review pending receipt of responses to those questions. 2.4.13 Accidental Release of Liquid Effluents (To be provided later) l _l _i 2-34 WNP-3 DSER SEC 2 l
2.4.14 Technical Specifications and Emergency Operation Requirement Because the staff has not yet completed its Hydrologic Engineering review, the need for technical specifications and/or emergency operation requirements has not been determined at this time. 2.4.15 Conclusions According to procedures outlined in the SRP, the staff has reviewed the design of WNP-3 with regard to hydrologically and hydraulically related plant safety features. On the basis of this review, the staff concludes that large-scale river or stream floods do not pose a threat to the safe operation of the~ plant ~ or the integrity of the site. The staff, however, is unable to conclude'that local flooding will not threaten the plant. The staff concludes that WNP-3 meets the requirements of GDC 2 with respect to potential flood hazards except for the outstanding item concerning local flooding. The staff has reviewed the availability of water for normal cooling purposes during diminished flow periods in the Chehalis River and the conditions imposed _ by the EFSEC. The staff concludes that there is sufficient water available to maintain safe plant operation over any reasonable drought period as required by GDC 44 with respect to normal cooling water availability. -- The staff has reviewed the information on the dewatering system presented by the applicant and concludes that there is insufficient information concerning the time it would take for the groundwater level to rise to an unacceptable level in the event of a total failure of the dewatering system. There is also not enough information concerning the effect of pipe breaks on the dewatering system nor on the proposed in-service monitoring program. The staff will com-
. plete its review of the dewatering system once the applicant addresses the concerns stated in this draft SER and provides responses to the questions which' have already been provided to the applicant.
In addition, the staff has not completed its review of the performance of the ultimate heat sink. This review is being conducted by Argonne National Laboratories and the results will be presente,d in the SER. . _l 2-35 WNP-3 DSER SEC 2
.. <- .~ .
p- w ; - y , sS 2.5 Geology and Seismstogy Thegeol.ogyandsetAmologyoftheWNP.-3sitewerereviewedduringtheearly
*andmiddle1970'sduringtheConsfrud$ionPermit(CP) review.
? - ?, -
As a result of the CP reytaw the NRC staff concluded that:
', .y . ,,
(1) the inferred large deep seated fault, blocks that have been associated
~/
with large earthquakes in thi southeia part of the Puget Sound are not present inffhe site area; s cd-s e, ~< An m (2) movement of faults in the site vicinity most likely ceased'in the late ~ - Tertiary, more than 2 million years before present (mybp),-and are there-fore not capable yithin the meaning of Appendix A, 10 CFR 100; _
^
/
(3) there are no known structures # in the immediate site vicinity that could beexpectedtolocalize'earthquakesthefd;
; /s (4) the applicant's assessment of the'possible volcanic risks in the site _
degion are adequate and that a problem of this type does not exist at the sit.e; and /-
. ,- a (5)' the safe shutdown earthquake (SSE) with a maximum acceleration of 0.32, andtheop'erationb(sisearthquakeof0..~qareconservativewhenapplied to the foundation level. /
d - During coastruction of the, facility numerous minor faults were found in the excavation. The applicant investigated these faults and determined that they were at least 630,000 yearsoldbutmostlikely'morethan2millionyears old. The staff reviewed the appl'icant's data'and made several visits to the
' ' /-
site to, examine the faulis'and c'oncluded that the faults were not capable.' In-1974 through 1976, as a risult of licensing. activities for the Skagit x Nuclear Power Plant, studies Ee're initiated concerning the 1872 Earthquake
~
(MMI IX, magnitude 7.0). New data from- tnese studies raised the question as to whether or'not 'an event of that size could occur at the WNP-3 site. The . _ applicant investigated that earthquake, mostly in regard to its Hanford sites,
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?
~
/ 5
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j 2-36 WNP-3 DSER SEC 2
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+
.. f . -_h
1 i and determined that it was related to tectonic structure within a broad epicen-tral zone and therefore could not occur at the site. The NRC staff reviewed i that data and data compiled by a panel of experts formed by Northwest US Util-ities, and that of another panel of USGS and NOAA experts, and concluded that the 1872 Earthquake was centered in the region between Entiat, Washington and Chilliwack, British Columbia (NRC, 1978) and s,hould not be expected to recur at the site. The most recent staff and USGS discussion of this earthquake can be found in the WNP-2 SSER. 1 L j On May 18, 1980, after several weeks of resurgent activity, Mount St. Helens erupted violently sending large quantities of ash several hundred miles down-wind to the east. The NRC requested the applicant to reassess'the volcanic
~ -
hazards to the site based on the new data. The applicant concluded, bas'ed on that assessment, that the maximum potential ashfall that could be expected from such an eruption from the closest volcanoes during the worst meteorological conditions, would result in a maximum of 1.75 inches of ash at the site. They ' stated that the plant design could accommodate that kind of ashfall. The NRC staff reviewed the applicant's data and USGS data collected with partial NRC l funding and concurred with the applicant's conclusion, but requested additional _ supporting data. The staff has completed its review of the FSAR. It has held several meetings -' with its advisors, the U.S. Geological Survey and its geological consultant, Dr. David Slemmons, two technical meetings with the applicant and its consul-tants, and conducted a geological reconnaissance of the site and region around 4 the site. On April 28, 1983 we transmitted questions, including those of our advisors to the applicant. Because of the June 1983 postponement of the WNP-3 site construction, those questions or outstanding issues have not been responded to. These open topics will be presented in Sections 2.5.2, 2.5.2,
. and 2.5.3.
Because of the extensive geologic and. seismic information (mostly about subduction zones) that has come out since completion of the CP review, new staff concerns have arisen; however, the following CP conclusions of the staff
- are still valid
~
i 2-37 WNP-3 DSER SEC 2 l
f} (1) the inferred large deep-seated fault blocks that have been associated with large earthquak'es in the southern part of Puget Sound are not present in the site area; (2) movement on mapped faults in the ' site vicinity, including those in the excav'ation are ancient and are not capable; and (3) the volcanic hazard to the site has been adequately addressed even in light of the recent eruption of Mt. St. Helens and has been appropriately j considered in the design.'- .3
~
Based on new data since the CP, the adequacy of the SSE is in question for.the~ following reasons which reflect our general concerns: (1) Thepossibilityofalargeorgreatearthquakeonasubduction$one beneath the site; (2) the.possibilityofundcognizedlowanglethrustfaultsinthesite vicinity that could cause large close-in earthquakes or surface faulting _. a') the site.
. 3 These issues will be addressed in greater detail in the following sections. "
/ .
2.5.1 Basic Geologic and Seismic Infor7ation fx 2.5.1.1 ~ Regional Geology The WNP-3 site is located in the' Pacific Border Physiorca.6 hic Province of Washington State, about two miles south of the town of Satsop and 16 miles . . e'ast of the city of Aberdeen. The site areo lies in the Chehalis Lowlands, which comprise a physiographic zone separatin.g the northern termination of 'tinIe - Oregon Coast Range from the Olympic Mountains. ' l t The site and its environs are largelf underlain by Cenozoics ' trata. Palative to more northern areas of the region, rocks of"the site area are not highly i- < , deformed. Igneous rocks of Mesozoic and Cenozoic age, however, are more . _ 2.-38 WNP-3 DSER SEC 2 j _ _ _ _ _ _ }
h r _. l i abundant than either sedimentary or metamorphic units thoughout the region. : The nearest outcrops to the site of Mesozoic and Paleozoic rocks (metamorphic, igneous, and sedimentary types) are found in the highly deformed area some miles to the north and northwest of the proposed plant area. Lithologically, the Cenozoic strata consists predominantly of marine clastic sediments deposited on a basement of Eocene oceanic basalts. The tectonic history of the site region is complex, with castward and westward directed low-angle thrusts, grabens, granitic plutons, and stratovolcanoes beiag best displayed and developed in the Northern Cascades. In the Northern Cascades, the Paleozoic Era is characterized by metamorphic and eugeosynclinal rocks. Eugeosynclinal sediments, granitic plutons, low-angle t'hrusts, and grabens were formed throughout the Mesozoic Era. During Cenozoic time,'the formation of grabens, granitic plutons and basalt flows predominated tectonic activity. These events were folicwed by several orogenic periods which caused folding and faulting of the older formed rocks and general uplift of the region, and the stratovolcanoes of the Cascade range began to form. The structural features that were formed during these orogentes, and the region, were subse-quently eroded during the Quarternary to produce the present day topography. __ While it appears that the last major period of deformation in the region ended in the Late Tertiary (Pliocene), evidence from Pleistocene deposits in the coastal areas west of the site, from 1100 year old fault dates in the Puget -"- Sound area to the north, and from three active stratovolcanoes in the central part of the state to the east of the site, show that tectonism continued on a more miner scale through the Pleistocene into the Holocene. The tectonic deformation of Western North America appears to be intimately related to the interaction of two major lithospheric plates, the North American Plate and the Pacific Plate. The interaction is principally along two major . transcurrent faults, the San Andreas Fault in California and the Queen Charlotte Fault off Western Canada. However, in the area between Cape Mendocino in' northern California and the southern extent of the Queen Charlotte Fault off the western tip of Vancouver Island, the two major plates named above are separated from one another by the small Juan de Fuca Plate.
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2-39 WNP-3 DSER SEC 2
l The interaction between the Juan de Fuca Plate and the North American Plate is not presently understood. The magnetic anomaly pattern east and west of the Juan de Fuca Ridge indicates that part of the Juan de Fuca Plate has been subducted beneath the North American Plate. Also, the chain of stratovolcanoes which forms the axis of the Cascade Mountains is believed to have been produced . by magma from a subducting plate (Atwater,1970). Several other types of data 1 indicate that an episode of late Cenozoic subduction occurred in this region of western North America. Seismic reflection surveys off the coast show a sediment-filled trench at the base of the Continental slope (Hays and Ewing, 1970). Anomalously high gravity values on Vancouver Island are suggestive of a remnant subducting slab beneath the region (Stacey, 1973). Seismic wave velocities indicate that a high velocity slab exists beneath the Puget Sound Basin (McKenlie and Julian,1971; Crosson,1972) which is indicative of a subducting lithospheric plate. j .
- The applicant has thoroughly reviewed the above-mentioned items and other types of data related to the current interaction of the lithospheric plate boundaries, including studies of plate kinematics (Silver,1971; Atawater,1970).
! While the available data are not clearly definitive, the applicant concludes c that the data tends to support the interpretation that subduction is no longer occurring along the Juan de Fuca-North American Plate boundary or is occurring asseismically, c Available evidence examined during the CP review indicated that subduction along the Juan de Fuca Plate-North American Plate boundary was not currently occurring. In particular, earthquake activity indicative of a Benioff zone (a characteristic of subducting plates) was absent in this region. Also, the orientation of the present regional stress field was inconsistent with active subduction. Analysis of earthquake source mechanisms showed that the maximum principal stress is north-south compressional and the minimum principal stress varies from east-west to nearly vertical (Dehlinger and Couch,1969; Couch and - McFarlane,1971; Crosson,1972; Malone, et al.,1975). , l New information has been developed since publication of the CP SER and the l FSAR, however, which may require a mocilication of tne above conclusions. This new information may indicate that succuction is continuing and that the _
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2-40 WNP-3 DSER SEC 2
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two plates may be coupled. That information and NRC staff's concerns are l presented in Sections 2.5.2 and 2.5.3. Numerous reverse faults of a generally northwest or northeast trend, marking elongated basement uplifts, occur throughout the basaltic rocks of the region. , These structural features are cut by east to northeast trending normal faults bounding areas showing different amplitudes of folding. Some of these faults significantly displace Tertiary strata in the region. The above described faults are thought to be the result of northeast compression of the crust, which was recurrent several times throughout the early Tertiary, until at least the middle Miocene. The basaltic basement complex shows the highest degree of faulting, with the intensity of faalting declining with the dec~reasing ages of-
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i the overlying rock units. I A line of stratovolcanoes extends along the Cascade Mountains from northern California to southern British Columbia. Eight of the volcanoes are within 200 miles of the Satsep site, the nearest being Mt. Rainier and Mt. St. Helens, each about 80 miles away. All of the volcanoes are believed to have been l active within the past 15,000 years and three of them, Mt. St. Helens, Mt. __ Rainier, and Mt. Baker are considered active at the present time. . Prior to 1980 Mt. Rainier had received the most study. The studies show that --- it has been intermittently active during the last 10,000 years. This activity has been mainly of pyroclastic type, but includes at least one flow which extended nine miles from the mountain. Three of the tephra eruptions deposited about one inch of material up to 25 miles east of the mountain. The last major eruption occurred about 2000 years ago, but minor eruptive activity occurred 120 years and 150 years ago.
. In addition to the eruptions of tephra, numerous mud flows have occurred at Mt. Rainter. The largest of these, the Osceola mud flow, occurred 5700 years ;
ago. It extended about 70 miles down-valley from the volcano. None of the l river valleys which could be potential mud flow pathways pass near the Satsop < site. We conclude, therefore, that no mud flow hazard exists at the site. 2 MD 2-41 WNP-3 DSER SEC 2
-. c . . - - - -.
A reassessment of the volcanic hazard was made after the May 18, 1980 eruption i of Mt. St. Helens. It was found that downwind of the prevailing winds from the volcano at about 80 miles (plant's distance) there was an accumulation of 6 inches of tephra. The applicant reduced that value of 1.75 inches because the WNP-3 plant is upwind from the nearest Cascade volcanoes. This is a rea-sonable assumption but we require more data about the maximum thickness of tephra landfall and maximum rate of ash fall to support it. . In summary, it can be said that, while the geologic conditions of the Satsop site and its environs are very complex, and the area is still tectonically active, based on our review of the applicant's work to date, there are no - known faults or other structures in the immediate vicinity of t'he site which - could be expected to localize earthquakes; however, because of recent findings about the tectonics of the region, we require additional information to support that conclusion. The outstanding items concerning faulting in the region are discussed more fully in Section 2.5.3. ' l 2.5.1.2 Site Geology 1 _ The WNP-3 site is located on a ridge in the Willapa Hills, 1 mile south of the intersection of the Satsop and Chehalis Rivers. The site elevation was +595 MSL prior to excavation. Elevations rise to +1,768 MSL at Minot Peak, 4 miles to ---
- the south. The floodplain of the Chehalis River Valley is about 1 mile wide i
and has a general elevation of 425 MSL. Drainage patterns in the site area form ) a modified dendritic pattern that is structurally controlled to some extent by the regional Tertiary folding and jointing. Slopes are generally moderate, but range from nearly flat to vertical. The abundant weathering profiles, relict erosion _ surfaces and Pleistocene terraces in the area were used extensively to determine an upper limit to areal tectonic events. The site vicinity is underlain by Quarternary deposits which consist of weathered gravels'of the Wedekind Creek and Logan Hill formations of early to middle Pleistocene age; glacio-fluvial sands, silts, and gravels of middle to late Pleistocene age; loess of la+e Pleistocene age; colluvium and landslide deposits of late Pleistocene to Holocene age; and Holocene colluvium.
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2-42 WNP-3 DSER SEC 2
l i
. 1 Approximately 15,000 feet of Tertiary rocks are present in the site vicinity, .
the oldest of which is the middle Eocene Crescent formation, a submarine basalt. Late Eocene Skookumchuck and McIntosh formations siltstones, tuffs, breccias, and sandstones overlie the basalt. The late Eocene to early Miocene Lincoln Creek formation of tuffaceous siltstone overlies the older four formations and is overlain by early to middle Miocene sandstones of the Astoria formation. The uppermost rocks in the site area are siltstones, sandstones, and conglom-erates of the Montesano formation of late Miocene to early Pleistocene age. The plant site is founded on massively bedded sandstone of the Astoria formation. Structurally the sita is located on the nose of a broad poorly defined anticline, which is an extension of one of the areas several uplif ts, the 'Minot Peak
~ - ~
uplift. Typical of other anticlines in the region, the Minot Peak uplif't has the basaltic basement rocks exposed in its core. Several significant faults (some with several thousand feet of displacement) in the site area can be shown by various means (e.g., terrace dating, saprolitization rates, erosion rates) to be associated with deformations no younger in age than Middle Quaternary (more than 630,000 years ago). Thus, they are not considered to be capable faults within the meaning of 10 CFR Part 100, Appendix A. _ Numerous landslides have been mapped on the site locality. Many of these, though not most commonly, have been identified in the Astoria formation, which -= is the foundation bedrock. These slides in the Astoria formation are related to slippage along weathered siltstone interbeds. Based on a detailed investiga-tion of local landslides, the applicant determined the geologic and geomorphic conditions necessary for sliding to occur: strong weathering of the Astoria rock, the presence of siltstone beds in the Astoria, topographic slopes inclined in the direction of bedding dip, and undercutting of bedding beneath dip slopes. Site investigations showed that these conditions do not exist at the site. The
. staff concludes that landsliding does not represent a problem at the site.
l 2.5.2 Vibratory Ground Motion l As a result of regional and site investigations performed by the applicant and others since the issuance of the CP-SER for WNP-3 in February 1975, the knowledge 9 He 2-43 WNP-3 DSER SEC 2
~ of the area has been greatly enhanced. The applicant has, and is continuing to undertake numerous studies and investigations that will provide an extensive amount of new information and interpretation. The staff anticipates that our l review of this new information will lead to an understanding and resolution of l many issues relating to the site vibratory ground motion determination. The increasing amount of new information, however, may require the reinterpreta-tion of some previous positions of the staff, the USGS, and the applicant. Presently the open seismological items have been transmitted in the form of questions (0230.1 through 0230.6) to tFa applicant. The applicant and the staff have met to discuss these open issues, and it is anticipated that the applicant will undertake a rigorous program of investigations t'o collect'the ~
~
information which will allow the staff to resolve the open issues. A suinmary of these issues follows. The most significant seismologic issues involves the seismogenic potential of the subducting Juan de Fuca plate beneath WNP-3. The staff concluded in the CP-SER for WNP-3 in February 1976 that "while the available data are not clearly definitive, we believe that they terd to support the interpretation _ that subduction is no longer occurring along the Juan de Fuca - North American Plate boundary." Since that time addit'ional recordings of small earthquakes have revealed an inclined zone of seismicity dipping to the east-northest --- (Crosson,1980). In addition, based upon the work of Ruff and Kanamori (1980) and Kanamori (1983) regarding the seismogenic potential of subduction zones, a number of questions regarding the Juan de Fuca zone have been raised. It is the applicant's position as discussed in FSAR sections 2.5.1.1.4.2 and 2.5.2.4.2.2, that the interface between the Juan de Fuca and North American plates will not i be the location of a large magnitude earthquake. The staff has indicated via
- the review questions that the applicant must document in greater detail their position.
In particular the staff has requested that the applicant document the following information regarding the Juan de Fuca plate. This includes the applicability of Kanamori (1983) relationship, and examples of aseismic subduction zones which share the same ' characteristics witn tne Juan de Fuca zone. The magnitude of the largest shock in the plate or along the plate interface that could occur _ 2-44 WNP-3 DSER SEC 2
without exceeding the SSE and ground motion attenuation from subduction zones that can be used for the WNP-3 site will also be documented. The magnitude of the maximum credible earthquake on the subduction zone, along with estimates of vertical and horizontal response spectra, depth and configuration of the subducting plate based upon earthquake locations cross-sections, fault planc solutions, and historic earthquake re-locations will also be provided by the applicant and reviewed by the staff. The staff has also requested that the applicant calculate site specific response spectra for the maximum historical earthquake, not associated with known geo-logic structure, in the tectonic province of the site, and for the maximum earthquake on the Olympia Lineament. The applicant has also be'en asked to estimate the annual exceedance probability of the SSE using all possible' seismologic source including the subduction zone. The staff., the USGS, and Dr. Slemmons will undertake and participate in meetings and probably several site visits to review the applicant's additional information and field investigations. Upon the applicant's submission and the staff's review of the new information, the staff will issue its Final SER. This SER _ will discuss in detati all the relevant geologic and seismic issues including the regional and site geology, c'apable faulting, seismicity, operating and safe shutdown earthquakes, and the vibratory ground motion. Reports by the USGS and " Dr. Slemmons will be incorporated as appendices and will be discussed in the SER. 2.5.3 Surface Faulting The applicant has determined that the structural geology of the site and regions around the site is characterized by large uplifts and faults and folds
. related to those uplifts that were formed by regional northeast directed compression during the Tertiary period. Three of these uplifts are present within the site vicinity, the Minot Peak uplift, the Blu'e Mountain uplift, and the Black Hills uplift. The site is located on an anticline which is the northern extension of the Minot Peak uplift. All of the upiffts are bounded primarily on the southwest sides and southeast sides by high angle faults that strike north-northwest, and east-northeast, respectively, with offsets ranging. _
_ 2-45 WNP-3 OSER SEC 2
b i + from several thousand feet to several hundred feet. The closest faults of this
- kind to the site are the Weikswood fault on the southwest side of the Minot Peak .
l uplift and the Gibson Creek fault on the southeast side of the uplift. Offsets j on both faults exceed 2000 feet. The Weikswood fault is approximately 1 mile l south of the site at its closest approach, and the Gibson Creek fault is about 5 3/4 miles south of the site. The applicant investigated all of the faults in the site vicinity by means of j a literature search, mapping, borings, trenching, and remote sensing techniques. l The applicant determined an upper limit of age of last movement on the faults I by analyzing cross-cutting relationships between faults and stratigraphic
* -^ ! contacts, relict erosion surfaces, Quarternary deposits, paleos'ois and weath ~
ering profiles. By determining the ages of these features the applicant'was q able to show an upper limit of movement on these faults of at least 630,000 [ years before present and more likely 2 million years before present. The staff
} has reviewed the data that is the basis for the conclusion and concludes that the faults mapped in the site vicinity are not capable within the meaning of f Appendix A. Numerous minor faults were encountered in excavations for the j plant. Most of these faults are northwest to northeast striking reverse faults.
l; The applicant has made a good case in the FSAR for relating these faults to the regional faults and to the Late Tertiary northeast directed compression. NRC staff geologists examined these faults on several occasions. The NRC concludes -- , that the faults mapped on and around the site are not capable (Appendix A). 1 On the other hand, considerable new geological information regarding the
. tectonics of the site region has been developed since the FSAR was published.
a Although we hold to our position that the faults in the site locality are not j capable, some of the new data raises some concern.- For example, it is not clear what happens to the faults at depth. If they are indeed related to j Late Tertiary tectonics which are no longer in existence that is one thing, j but if they are tied to large eastward dipping thrust faults that flatten i downward (eastward), which are related to an active subduction tectonic style of the Juan de Fuca plate, then additional analyses and possibly investiga-l tions, will have to be carried out. A major northwest-trending fault in the Humptulips River area (Tabor and Cady, 1976) is a possible fault of this kind. + It projects northwestward under Quaternary ceposits to an outcrop of steeply . _ [ ! 2-46 WNP-3 DSER SEC 2-
dipping Pleistocene deposits (op. cit) on the west Fork of the Humptulips River. The capability of this fault may be important to the site in light of the following. Offshore studies by Silver (1972) and Snavely and Wagner (1982) indicate a subduction tectonic style characterized by eastward (landward) dipping thrust faults that generally steepen westward (upwards) and that have offset seliments as young as Quaternary. Considering this structural framework, we have asked the applicant to evaluate the possibility that the Humptulips fault, if capable, extends southeastward as a continuous fault or fault zone along the steepened west limb of the Wynocchee anticline (Rau, 1976) and on into the less well-defined Melbourne anticline (Gower and Pease, 1965) or alternatively to the southeast of these structures. We have requested the applicant to determine whether or not the Humptulips fault is tihroughgoing and
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capable, and, if so to evaluate the effects on the site. Recent seismic reflection, remote sensing, and geophysical data covering the area has been gathered that post dates the FSAR publication and therefore has not been evaluated with respect to the site. We have recommended that the applicant assess these data with respect to the site. e Many of the natural drainage features in the site vicinity occur along projec-tions of mapped faults although the faults are shown to terminate away from the stream valleys but along projections of their trends. Also many drainages are --- oriented in a pattern that is paralleled to the north-northwest and northeast striking fault pattern, yet the streams are not considered to be fault controlled by the applicant. Evidence that supports the conclusion that the drainage features are not fault controlled is needed before the staff can complete its review. The applicant has dismissed offset magnetic anomalies KK and HH on the Juan de . Foca plate as probably due to episodic jumping of short transform faults connecting offset segments of the spreading ridge as suggested by Hey (1977). (FSAR 2.5-44). Provided that successive jumps are in the same direction and occur after equal increments of spreading, the jumps should produce a V-shaped , wake consisting of a pair of lineaments intersecting at the ridge. Although I KK seems to form such a wake, mirrored in the Pacific plate, HH is less convinc- ) 1 ingly matched (c.f. Barr, 1974 and Elevers and others, 1973). Considering the 2-47 WNP-3 DSER SEC 2
difficulty of identifying the mirror image of HH, the applicant has been
. requested to evaluate the hypothesis that HH is a fault as suggested by Pavoni (1966), and that the on-shore subcrustal extension of HH could be the source of deep-seated major earthquakes in the Puget Sound region (Fox,1983), and to evaluate the response at the site of a major earthquake on fault HH.
C emen. 4 m g 2-48 WNP-3 DSER SEC 2
REFERENCES
- 1. Atwater, T., 1970, Implications of Plate Tectonics for the Cenozoic Tectonic Evolution of Western North America; GSA, 81, pp 3513-3535.
- 2. Barr, S. M.,1974, Sea Mount formed near the crest of Juan de Fuca Ridge, NE Pacific Ocean; Marine Geology, Vol. 17, p 1-19.
- 3. Couch, R. W. and MacFarlane, W. T. ,1971, A Fault Plane Solution of the October 1969 Mt. Rainier Earthquake and Tectonic Movements in the Pacific Northwest Derived from Fault Plane and First Motion Studies; EOS, AGU Trans, 52, 428 p.
- 4. Crosson, R. S., 1980, Review of Seismicity in the Puget Sound Region from 1970 through 1978; Report presented at the Puget Sound Earthquake Hazards Workshop, University of Washington, Maple Valley Center, Lake Wilderness, October 1980.
- 5. Crosson, R. S., 1972, Small Earthquakes, Structure, and Tectonics of the Puget Sound Region BSSA, 62, 5, pp 1133-1171.
m
- 6. Dehlinger, P. and Couch, R. W., 1969, Synthesis of Geophysical Results in the Juan de Fuca and Gorda Ridge Areas (NE Pacific Ocean), EOS, AGU Trans, 50, 186 p.
- 7. Elvers, D., Srivastava, S. P., Potter, K., Moorley, J., Sdidel, D., 1973, A symetric spreading across the Juan de Fuca and Gorda rises as obtained from a detained magnetic survey; Earth and Planetary Sciences Letters, vol. 20 p. 211-219.
- 8. Fox, K. F. , Jr. ,1983, Northeast-trending subcrustal fault transects western Washington; U.S. Geological Survey Open-File Report 83-398.
- 9. Gower, H. P., and Pease, H. , Jr. ,1965, Geology of the Monteseno Quadrangle, Washington; U.S. Geological Survey GQ Map 374. .
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- 10. Hayes, D. E. and Ewing, M., 1970, Pacific Boundary Structure; in The Sea, A. E. Maxwell (ed), 4, Part II, pp 29-72.
- 11. Hey, R., 1977, A new class of "pseudofaults" and their bearing on plate tectonics: a propagating rift model; Earth and Planetary Sciences Letters, v. 37, p. 321-325.
- 12. Kanamori, H.,1983, Global Seismicity; Preprint, California Institute of Technology.
- 13. Malone, S. D., Roths, G. H., and Smith, S. W., 1975, Detatis of micro-
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earthquake swarms in the Columbia Basin, Washington, BSSA,'65, 4, pp 855-865.
- 14. McKenzie, D. and Julian, B.,1971, Puget Sound, Washington - Earthquake and the Mantle Structure Beneath the Northwestern United States; GSA, 82, PP 3519-3524.
- 15. Pavoni, N., 1966, Tectonic interpretation of the magnetic anomalies _
southwest of Vancouver Island; Pure and applied geophysics, v. 63, P. 172-178.
- 16. Rau, W. W., 1967, Geology of the Wynocchee Valley Quadrangle, Washington; Washington State Division of Mines and Geology Bulletin no. 46, 51 p.
- 17. Ruff, L., and H. Kanamort, 1980, Seismicity and the Subduction Process; Physics of the Earth and Planetary Interiors, v 23, p. 240.
- 18. Silver, E. A., 1972, Pleistocene Tectonic Accretion of the Continental Slope off Washington; Marine Geology, v. 13, p. 239-249.
C
- 19. Silver, E. A., 1971c, Small Plate Tectonics in the Northeastern Pacific; GAS, 82, pp 3491-3496. .
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- 20. Snavely, P. D. , Jr., and Wagner, H. ,1982, Geologic cross section across the continental margin off Greys Harbor, southwestern Washington; U.S.
Geological Survey Open-File Report 82-459, 11 p.
- 21. Stacey, R. A., 1973, Gravity Anomalies, Crustal Structure, and Plate Techtonics in the Canadian Cordillera; Can Jour Earth Sci., 10, pp 615-628.
- 22. Tabor, R. W., and Cady, W. M., 1978, Geologic map of the Olympic Peninsula, Washington; U.S. Geological Survey Miscellaneous Field Investigations Man I-993.
- 23. US NRC, 1983, Geosciences Review Questions, Memorandum from R. E. Jackson te G. W. Knighton, Supply System Nuclear Project No. 3, April 28, 1983.
- 24. UN NRC, 1982, Safety Evaluation Report related to'the Operation of WPPSS Nuclear Project No. 2, docket No. 50-397, Supplement No. 1, NUREG-0892, August, 1982.
e 4 j i J 4
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2.5.4 Stability of Subsurface Materials and Foundations 2.5.4.1 Site Conditions 2.5.4.1.1 General Plant Description The WNP-3 plant site is located on a ridge at the northern edge of Willapa Hills, in the southeastern portion of Grays Harbor County in the State of Washington. The site is approximately 16 miles east of Aberdeen and approx-imately three miles south of the town of Satsop. Prior to the start of the plant construction, the ridge at the plant location was at an elevation of approximately 480 feet above mean sea level. During early construction,'the ~ general plant grade was excavated to an elevation of approximately 390 feet. The common foundation mat for the WNP-3 reactor building and reactor auxiliary building is supported on essentially fresh sandstone at an elevation of approxi-mately 326 feet. All other Seismic Category I structures are founded at plant grade on weathered sandstone. These structures include UHS Dry cooling Tower Train A and Train B, Dry Cooling Tower Control Building, Condensate and Refuel- _ ing Water Storage Tank Enclosure, two Diesel Oil Storage Tank Enclosures and Manholes for the gravity drainage system. About 1000 feet northeast of the powerblock there is a cooling tower which is used for normal cooling operations. -- It is not a seismic Category I structure. The ultimate heat sink function for WNP-3 plant is performed by two dry cooling towers; no makeup water is required for safety-related cooling of the plant. There are permanent natural slopes in the north-south direction and man-made excavation slopes in the east west direction whose failure could affect the safe operation of -the plant. The natural rock slope south of the power block dips at an average slope of 3 horizontal to 1 vertical (3:1), with a maximum slope of 1-1/2:1. The crest of the natural slope is more than 250 feet from the edge of the powerblock. The man-made rock slope east of the powerblock is about 180 feet high, rising at an average slope of 4-1/2:1, with a maximum slope of 3:1. The toe of the slope to the east is more than 350 feet from the edge of the dry cooling towers. O w AM 2-52 WNP-3 DSER SEC 2
2.5.4.1.2 Subsurface Investigation The subsurface investigation program at the site consisted of drilling and trenching. A total of 95 borings were drilled at the site and soil and rock samples were recovered. Most of the borings were drilled with mud and tricone bit. Rock cores were obtained by us'.ng NX double tube coring barrels and i diamond bits. Core recovery and rock quality designation (RQD) were recorded. Piezometers were installed in 52 boreholes after completion of the drilling. In addition, 65 trenches of various lengths and depths were excavated. The geophysical investigations consisted of seismic refraction surveys and cross-hole shear wave velocity measurements. The geophysical s'urveys show
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good spatial correlation with the test boring results.
! 2.5.4.1.3 Subsurface Materials 2.5.4.1.3.1 Investigation of Subsurface Materials Properties ine piant area is unuerlain oy asturla Ivien.Livii, e mariastone sei ves less __
degrees of alteration including residual soil, weathered sandstone, fresh (unweathered) sandstone and tuff. This formation also contains some siltstone strata and several tuff beds. Fresh sandstone is differentiated from weathered -e-sandstone on the basis of color change. Fresh sandstone is light to dark gray, and weathered sandstone is yellowish-brown from the oxidation of iron minerals in the sandstone. Joints in sandstone show two dominant trends; one set ranges in strike from N34*E to N65'E and dips generally from 50*-70' SE and the second j set strikes about N43'W and dips nearly vertical. The general spacing of
; joints within a joint set is approximately 1 to 5 feet and the distance between I joint sets is 40 to 100 feet.
I . During construction all residual soils and most of the tuff were excavated from the plant area, and the plant grade was established at elevation 390 feet above mean sea level on the exposed weathered sandstone surface. At the location of WNP-3 site, the weathered sandstone extends to a depth ranging from 40 to 60 feet below final plant grade level and is underlain by fresh sand-i stone. A 7 to 11 foot thick tuff bed exists in the vicinity of the northern. _ 2-53 WNP-3 0$ER SEC 2
edge of the plant location at 0 to 15 feet below the finished plant grade. The general subsurface profiles under the plant location are shown on FSAR Figures 2.5-72, 2.5-75, and 2.5-76. The Reactor Buildings and Reactor Auxiliary Buildings are founded on a common mat over essentially fresh sandstone at Elevation 326. Based on subsurface exploration results (RQO values ranging from 90 to 100% and core recovery of 85 to 100%) the applicant has concluded and the staff concurs that there'are no zones of alteration or irregular weathering or zones of structural weakness below Elevation 326. The unconfined compression strengths of intact fresh standstone' core samples ~
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range from 300 psi to 850 psi. The test values showed tangent modulus values at 50% ultimate strength for this material to range from 1 x 10' to 2.5 x 105 psi. The strength and compressiblity results for fresh sandstone are listed in Table 2.5-16 of the FSAR. We find these test results to be generally inconsistent with the higher values reported in the literature (e.g., Founda-tion Engineering Handbook by Winterkorn and Fang) for sandstone. However, sinea thasa valinas retult in conservative analyses, we find them to be accept- _ able. The applicant has, also, listed on Table 2.5-16 and used in his analysis of some structures, Poisson's ratio value of 0.35 to 0.50 for fresh sandstone. We find this value of Poisson's ratio for rock material to be too high. The -- applicant should either provide further substantiation of his reasons for using this range of values or use an appropriate value of Poisson's ratio in his analysis. The permeability of the fresh sandstone was determined by means of in-situ packer tests in two borings. In addition, field falling head and rising head permeability tests were conducted. The test results generally agreed with the in-situ packer test results. The highest permeability recorded in fresh
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sandstone was 7.5 x 10 ' cm/sec. We find this value of permeability for fresh sandstone to be reasonable. The applicant determined the dynamic properties of the fresh sandstone from laboratory sonic velocity measurements ano from fielo seismic refraction surveys and cross-hole shear wave velocity measurements. The P-wave velocity. _ 2-54 WNP-3 DSER SEC 2
values recorded ranged from about 6000 to 8000 ft/sec, and the S-wave velocity values ranged from about 2500 to 4000 ft/sec. These results are shown on . Table 2.5-16 of the FSAR. The staft .tnds these results to be reasonable and acceptable. The foundations of Category I structures other than the Reactor Building and the Reactor Auxiliary Building are founded at or slightly below grade level (El. 390) and rest on weathered sandstone. The weathered sandstone has low to moderate hardness and is found by the applicant not to have any zones of structural weakness at the location of the plant site (as evidenced by core recovery of 85 to 100 percent and RQD of 90 to 100 percent).
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The results of strength and compressibility tests on weathered sandstone are summarized in Table 2.5-15 of the FSAR. The unconfined compression strength of test samples are shown to range from 350 to 800 psi and the tangent modulus values at 50% ultimate strength range from 2.5 x 105 to 7.0 x 105 psi. We find these test results to be inconsistent with those reported in the litera-ture. However, since the values result in conservative analyses, we find them to be acceptable. The field permeability tests indicate that the highest per- _ meability coefficient recorded for weathered sandstone is 7.5 x 10
- cm/see which is the same as for the fresh sandstone. We find these permeability test results to be reasonable and acceptable. e The dynamic properties for weathered sandstone were investigated by the appli-cant through field stesmic refraction surveys and cross hole testing. These tests were supplemented by laboratory sonic velocity measurements on represent-ative weathered sandstone samples. The field and laboratory dynamic test results are given in Table 2.5-15 of the FSAR. The results indicate that the P-wave velocity values range from about 5000 to 9000 ft/sec, and the S wave velocity
. values for weathered sandstone range from about 2,300 to 4,000 ft/sec. We find j these values to be reasonable. The tuff beds below plant grade are composed of coarse to fine grained material and are about 7 to 11 feet thick. Based on strength testing, the applicant G m 1 2-55 WNP-3 DSER SEC 2
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t determined that the relative hardness and engineering properties of the tuff do l not differ significantly from those of the fresh sandstone. The applicant also j - did not find any evidence of bedding planes between the ruff layers and the [ sandstone.
- 2.5.4.1.3.2 Design Values of Subsurface Materials I
Fresh and Weathered Sandstone I I In Section 2.5.4.11 of the WNP-3 FSAR, the applicant has stated that representa-i ! tive values of compressive strengths of the fresh and weathered sandstone cores I were based on the statistical method using results of the laborhtory compression i tests; however, the details of the method are not given. The applicant has also stated that he used reduction factors (RF) to arrive at the design compres-e sive strengths for the two materials. Although the procedure to compute the RF value is briefly described, the computed values of (RF) and the corresponding final design values of the compressive strength are not given. In view of the j wide variation in the results of the test data shown in Tables 2.5-15 and 2.5-16, l the staff requires the applicant to provide, (1) the various steps used in e s statistical analyses to arrive at the representative compressive strengths , along with assumptions and the results of the analyses, (ii) the values of ] (Vp)1 and the computed values of (RF) along with the method of deriving values " of (V p)f and (V p)3 in these computations, and (iii) the computed values of design compressive strengths. i
- The' applicant stated in the FSAR that a statistical method was used to calcu-late values of the representative tangent moduli from unconfined compression test results on selected core specimens; however, the detatis of these analyses j are not given. Reduction factors (RF), similar to those used in computing the
- design compressive strengths were used to compute the design elastic moduli.
l Again, the details are not given and the procedures are not justified. The j staff requires that the applicant explain in detail the statistical analyses, justify the bases for assumptions and provide the results of analyses. The , computed values of design elastic moduli for fresh and weathered sandstone
- ' materials should also be presented.
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~---,,------,,.n----.- n . . . , ~ . ~ . , - , - - , _ - ~ , . - - - - , . ..--.- _,.- - --.--.--
In describing the procedure for arriving at the design values for Poisson's
, ratio, the applicant has stated in the-FSAR that the values were " selected based on the comparative method using results of laboratory compression tests on selected core specimens." The staff requires the applicant to provide details of the so called ' comparative method' and to justify this method's use in selecting design Poisson's ratio values for the fresh and weathered sandstone materials. Although the range is given, the applicant should document and justify the values of the design Poisson's ratio used for these materials in the various plant designs and analyses.
In response to staff questions 241.10 and 241.23, the applicant stated that a selected value of modulus of subgrade reaction of 500 lb/in* wa's used in' the - static analysis (using MSC/NASTRAN computer program) for the Tank Enclos~ure Structure. In analytically deriving'the value of this modulus of subgrade
, reaction the applicant used a shear modulus value of 330 ksi corresponding to
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strain value of 10 2 in./in. in his. dynamic shear modulus versus strain curve. Further, the expression used for modulus determination is indirectly derived from an equation given by Barkan for vibratory loads and for the purpose of machine foundations design. We find.that the applicant's use of the subject _ equation, the assumptions made in utilizing the equation and the resulting value of the modulus used in MSC/NASTRAN need further justification. e The applicant has selected design permeability values of 2 x 10 cm/see and 6 x 10 cm/sec for fresh and weathered sandstone materials, respectively. These values are very close to the maximum recorded permeability measurements in the field and laboratory for these materials and, therefore, are reasonable and acceptable values to be used in design and analysis. Dynamic shear wave velocities of .3,800 ft/sec for the fresh sandstone and
. 3,200 ft/sec for the weathered sandstone have been selected by the applicant to be used in the design and analysis of the plant. Little or no explanation is given in the FSAR in support of the selection of the values. In view of the wide variation, exhibited by test results, in the. values of S-wave velocity data presented in FSAR Appendix 2.5 D Tables 2.5-15 and 2.5-16, the applicant needs to further justify the use of these valu'es for design.
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The applicant performed one strain-controlled cyclic triaxial test to establish the design values and shape of shear modulus versus strain curve for fresh sandstone. To account for end effects and non-uniform strains within the test specimen, the average axial strain was obtained by dividing the measured axial strain by a correction factor of 4. Values of shear modulus were obtained at four different strain levels. Damping ratio for fresh sandstone was calcu-
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lated for one strain level of 3 x 10 8 Based on the results of measurements and interpretation of the test data, the applicant found that the shape of the shear modulus versus strain curve is different and steeper than the published curves for rock (e.g., Schnabel et al.1972) and the calculated damping at 3 x 10 2 strain is also higher than corresponding values in the published curves. We find the applicant has not provided bases for selecting a damping ~ ratic versus strain curve and has not sufficiently justified deviations from the published dynamic properties of fresh sandstone. Moreover, the applicant's results are based on only one dynamic test on fresh sandstone. Only a few different strain values were used to measure the shear modulus and damping. The correction factors used to divide measured axial strain have not been properly justified. For these reasons, applicant needs to provide further bases for selecting the shear modulus and damping ratio versus strain curves _ shown in Figures 2.5.121 and 2.5-122 of the FSAR. For the weathered sandstone material, the applicant has used the same dynamic -- shear modulus versus strain curve as for the fresh sandstone; however, no explanation for this assumption is provided. A curve for weathered sandstone damping ratio versus strain is not given. The applicant should justify the design curves for dynamic properties of weathered sandstone and provide the bases for assumptions used in deriving the design curves. Tuff Beds Since the relative hardness and engineet ag properties of the tuff beds below grade were found by the appitcant to be s*milar to those of the weathered sandstone, the static and dynamic design values for the tuff were selected as identical to those of the weathered sandstone. We find this assumption to be reasonable and acceptable. 4 m
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Residual Soils Although all residual soils were excavated from underneath the location of the plant structures, their engineering properties were established by the applicant for use in the evaluation of stability of man-made slopes east of the WNP-3 plant structures. Based on interpretation of laboratory test results of seventeen unconsolidated-undrained triaxial tests, six consolidated-undrained triaxial tests and three unconfined compression tests, the applicant established three sets of strength parameters for the residual soils: maximum strength parameters of c = 1000
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psf, 0 = 40*, minimum strength parameters of c = 750 psf, 0 = 26* and average ~ strength parameters of c = 1000 psf, p = 30'. The applicant used all th'ree sets in slope stability evaluation. We consider the design e and 0 values for residual soils used by the applicant in analyses to be reasonable and the approach of using an appropriate variation of solid properties in the slope stability analyses to be acceptable. 2.5.4.1.4 Groundwater Conditions _ The pre-construction groundwater conditions at the site were determined using borehole piezometer readings taken from April 1973 to September 1974 during the -- subsurface exploration. At that time, the groundwater levels in the WNP-3 powerblock area ranged from El 385 feet to El 411 feet. During construction, the plant level was excavated to an elevation of 390 feet and the bottom of the common mat foundation was excavated to an approximate elevation of 326 feet. For this situation the temporary ct oundwater flows from relatively impermeable sandstone (permeability coeffic.ient of approximately . 2 x 10
- cm/sec) were handled with a drainage system within the common mat excavation. Groundwater collected was drained from the excavation through a gravity flow to the slope south of the plant location. The groundwater condi-tions were monitored from October 1977 to December 1979 using piezometers around the WNP-3 excavation; the recorded levels during this period ranged from El 330 feet to 390 feet as shown on FSAR Figure 3.4.1-5.
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The plant, has a permanently lowered watertable level to an elevation below the foundation mat level (below 326 feet) along the exterior faces of the cate-gory structures by means of a gravity drainage system. The drainage system is not Category I, as the walls and the mat of the Reactor Auxiliary Building are designed to withstand full hydrostatic loads (water table at El 365) that may be caused by a complete blockage of the drainage system. The staff's evaluation of groundwater is presented in Section 2.4 of this SER. 2.5.4.2 Excavation and Backfill All residual soil was excavated from the plant area. The excavation into rock for the powerblock area extended about 64 feet below the final grade level (390 feet above mean sea level). It penetrated the overlying weathered sandstone for a depth of about 60 feet and exposed the fresh sandstone surface at elevation 326 feet. Vertical cuts were made in sandstone formations. The vertical rock sides were cleaned by air jetting and protected against weathering by short-creting over welded wire fabric. The bottom of excavation was covered by _ concrete mud mat. The Category I structures other than the powerblock structures are located -- slightly below plant grade on weathered sandstone. All cuts for these struc-tures were vertical. No backfill was required beneath or around seismic Category I structures since they were placed directly against fresh or weath-ered sandstone. Class Al Structural Fill, consisting of a well graded sand and gravel having a maximum size of 6 inches and a maximum of 15 percent passing the number 200 sieve, was used to backfill beneath, around and above seismic Category I buried pipe. When used as a bedding material for the pipes, backfill with the maximum particle size of 3/4 inch was used. In the FSAR Section 2.5.4.2.6, the applicant has stated that the Class Al structural fill was compacted to a specification of at least 95 percent of the maximum modified Proctor density (ASTM 1557-78). The in place density tests . 2-60 WNP-3 DSER SEC 2
were performed in each lift. The results indicated that less than 10 percent of the tested densities fell below 90 percent of the specified density. The l staff finds these as placed densities of the Class Al structural fill beneath, around and above seismic Category I buried piping to be reasonable and acceptable. In response to the staff question 241.22, the applicant has stated that there are other areas (including areas under and adjacent to Diesel Generator Fuel Oil Storage and Transfer System, and Class IE Duct Lines from Reactor Auxiliary
- Building to Dry Cooling Tower and Refueling Water Storage Tank Area) where placement of Class I structural fill has not been completed. The applicant has committed to provide for staff review the results of all Category I field density and moisture content tests performed under and adjacent to safety
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related structures as construction progresses. We find this commitment to be acceptable. The staff will review and evaluate the information when provided l by the applicant. The applicant used soil-cement (concrete sand mixed with 10 percent Type II Portland cement and 10.4% 2% moisture) compacted to a specification of 95 percent Standard Proctor Density (ASTM 0558-57) as a backfill in the construc- _ tion access ramps adjacent to the Reactor Auxiliary Building. In these areas
- the in place density test results showed almost 100% compliance with the specified compaction requirement. We find these results to be acceptable. --
2.5.4.3 Response of Rock to Seismic Loading l In Sections 2.5.2 and 2.5.4.7 of the WNP-3 FSAR, the applicant has stated that since the shear wave velocities in the underlying sandstone at the plant site is greater than 3000 ft/sec, there will be no amplification or modification of the input acceleration time histories at the plant site and the design earth-quake would be defined by Regulatory Guide 1.60 response spectra anchored to the maximum design ground acceleration (0.329) at the site. We concur with the applicant's assessment, given in FSAR Section 2.5.4.7, that the response of rock to seismic loading would not result in a modification of the input motion and find it acceptable. However, in Sections 3.7 of the same FSAR, the l applicant stated that he used a deconvolution analysis through 570 ft of rock column which resulted in a substantial mcdtf' cation of the design motion and . .
.- 2-61 WNP-3 DSER SEC 2
resulted in base slab response spectra lower than R.G. 160 spectra in frequency range of interest. Thus, the information presented in the Section 3.7 of the FSAR is inconsistent with the information presented in Section 2.5.4.7. The procedure given in FSAR Section 3.7 for evaluating response of rock to seismic loading is not acceptable to the staff because input motion would not be substantially altered within the firm rock surrounding structures. In response to staff questions 241.9 and 241.24, the applicant informed the staff that the depth of rock for deconvolution was based on the results of a sensitivity study in which the depth of rock column was gradually increased to determine the lower boundary of the analytical model until no difference in response of the building could be detected. However, the analy'tical parameters and the results of this study have not been provided to the staff for review. During an audit of the applicant's calculations on September 26 to September 30, 1983, the staff'was verbally informed by the applicant's A/E (Ebasco) that the l calculations and results pertaining to the said sensitivity study were not saved by the applicant's A/E. Based on a review of the information provided in the FSAR Sections 2.5.2, 2.5.4.7, and 3.7, the applicant's response to our 0's and our audit findings, we conclude that the procedure used by the applicant in determining the response of the rock using deconvolution through 570 foot rock column is not acceptable, an assessment of the results of its application to structural seismic design and analyses is presented in Section 3.7 of this SER. During the audit on September 26 - 30, 1983, the staff found that, in the NASTRAN computer program input for the soil-structure interaction analysis of the Reactor Auxiliary Building (RAB), a value of Poisson's ratio of 0.5385 was used. This is inconsistent with the design parameters shown on Table 2.5-16
. of the FSAR. Moreover, a value of Poisson's ratio greater than 0.5 is not theoretically possible. The applicant should provide justification of this issue and assess the impact on analysis.
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2.5.4.4 Foundation Stability The Category I structures on a common mat (Reactor Building and Reactor Auxiliary Building / Fuel Handling Building) are supported on firm, fresh sandstone at a depth of 65 feet below finished plant grade (El 390 feet). The foundations for other Category I Structures, (Refueling Water Storage and Condensate Storage , Tank Enclosure Structure Dry Cooling Tower Train A Structure and Control Building, Dry Cooling Tower Train B Structure, and Two Diesel Oil storage Tank Enclosures) rest on weathered sandstone slightly below El 390. The Category I buried pipelines are located on Class Al structural fill overlying either sandstone or soil-cement underlain by sandstone. Bearing Capacity In FSAR Section 2.5.4.10.1, the applicant has stated that Terzaghi's bearing capacity formula (Modified for rock) was used to compute the ultimate bearing capacities for the Category I foundation mat. We find this state-of-the-art procedure to be acceptable for static bearing capacity calculations. However, the applicant has not provided any information about the procedure and assump-tions for dynamic bearing capacity calc.iations. The staff requires this - information along with the bases for arriving at the dynamic loads used in these analyses. < For the Reactor Auxiliary Building mat foundation static bearing capacity calculations, the appitcant neglected the effect of cohesion and used a rupture angle of 20'. We find these rock properties assumptions to be reasonable and acceptable. In view of the foundation being supported on firm fresh sandstone 64 feet below grade, sufficient margin of safety against bearing capacity failure exists. However, we find that applicant's calculation results given in the
. FSAR Section 2.5.4.11.7 are inconsistent with those given in response to staff Question 241.'3; the minimum factor of safety in the FSAR is stated to be 34, whereas it is stated to be 6.3 in response to Question 241.3. The applicant needs to appropriately amend his submittals to make them consistent and correct.
The applicant should also provide for staff review, adequate information on the methods, assumptions and results of dynamic bearing capacity calculations. 2-63 WNP-3 0$ER SEC 2
'The applicanc has provided the results ok static bearing capacity calculations for the Condensate / Refueling Water Tank foundation and Dry Cooling Tower and Control Buildin's Foundation in response to staff Question 241.3. The values given for the factors of safety of the two buildings are 20.0 and 23.8, respect-ively, which we find to be adequate. However, for these computations, as well as for the dynamic bearing capacity calculations, the applicant has not provided the procedure and the rock properties used in the analysis. We require the applicant to provide the necessary details of the procedures used for static and dynamic bearing capacity calculations and the assumptions made in these analyses for staff review.
l l ~ ~ ! The applica.at haf not provided any information on the static and dynamic bearing ' capacity calculation procedures, assumptions aad results of analyses for the following seismic Category I structures: Dry Cooling Tower Train B l Structure, Two Diesel Oil Storage Tank E.1 closure Structures and Category I Drainage Manholes. The necessary bearing capacity calculation results for these structures along with the procedures and assumptions should be provided for staff review. ' l . The attached Iable 2.2, provided by the applicant in response to staff question 241.3, shows pertinent details cIf static and dynamic loads on three
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seismic Category I strue.tures. The staff requires the applicant to modify - this table to make fi. consistent with the information provided in the FSAR. In addition, the applicant should include information on other seismic Cate-gory I structures to complete this Table,'and submit the amended Table for staff review. f Settlement
. In response to staff question 241.4, the applicant informed us that, since all seismic Category I structures were founded directly on either fresh or weathered sandstone, settlenent or rebound was not considered by the applicant to be a factor in the design of the plant. The applicint did not and does not have any settlement moniCoring of the plant foundations. Also, no information has been submitted for staff review on the potential or, actual differential settlements between plant foundations and buried piping.67 duct run penetrations.
F 2-64 , WNP-3 DSER SEC 2
! L
Using Boussinesq equation, the applicant has computed the value of estimated post-construction total settlement of the Reactor Auxiliary Building- mat to be less than half inch. However, allowable settlements are not given in the FSAR. The applicant has also not provided for staff review estimated and allowable settlement values for other seismic Category I structures. We do not agree with the applicant that construction and post-construction settlements for rock supported structures need not be considered in the design of structures, piping, and duct run penetrations because the stresses induced due to differential settlements may be significant. We require the applicant
- to provide for staff review the values of allowable differential settlements that the Category I buildings can withstand (in combination with other appro-priate loads) and still meet code allowable stresses of the FSAR. In addition, since the applicant is not monitoring the actual total and differential settle-ments between various Category I foundations, he should assume a minimum of one-half inch differential settlement to check the design of piping running between rock supported structures. Piping and duct run penetrations should i
also be assessed for a minimum of one-half inch differential settlement. The results of these confirmatory analyses should be provided for staff review. . P Lateral Pressures 1 ~ The exterior walls of the Category I structures were placed directly against the vertically excavated rock face. The applicant computed the static lateral pressures resulting from (a) hydrostatic pressure due to possible failure of the groundwater drainage system around Category I structure (b) long term creep of sandstone causing active lateral pressures on exterior walls, and (c) the effect of the adjacent building surcharge at the ground surface causing lateral pressure on embedded walls. The applicant has stated in the FSAR that
. the Category I structure walls are designed for full hydrostatic pressure up to an elevation of 365 feet, incorporating a coefficient of active earth pressure -
of 0.22, corresponding to 9 = 40', and using Boussinesq stress distribution to evaluate the effect of surcharge on the embedded exterior walls. We find these computational procedures and assumptions to be reasonable and acceptable. The 1 -
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2-65 WNP-3 OSER SEC 2
. _ __. ~__ .. _ _ _ - _ _ _ _ _ ._- _ __- _ ~ _ _ . . _ . _ -
applicant stated in response to staff Question No. 241.17 that total static ' , lateral pressure was determined to be 2 kips /ft2 . The appl'icant has, however, not provided the distribution of static lateral pressure along the depth of the walls. We require that the applicant submit this information for review. The applicant computed the dynamic lateral pressure based on the effects of rock-structJre interaction analysis. The dynamic lateral pressure obtained from this analysis and used in design was 10.27 kips /ft*; the distribution of l this pressure with depth of wall was not provided by'the applicant for review. , This information is needed to complete our SER. 1 o l I As stated earlier in Section 2.5.4.3 of this SER, it is the staff s position ~ that the modification of rock response due to deconvolution is not acceptable. Since the applicant used this procedure to compute seismic lateral earth pressureshthestaffrequestedtheapplicant(Question 241.18) tore-compute these pressures without using deconvolution and by utilizing the state-of-the ' art Seeci and Whitman (1976) approach. The applicant has responded to this question (informally received by the staff on September 2'3,1983). The appli-cant's computations using the Seed and Whitman approach and incorporating j the effect of' static lateral pressure, hydrostatic pressure'and'effect of, surcharge load show the total dynamic lateral pressures to be 4.7 kips /ft". This valuelis less than one-half of the value used by the applicant in the design. .These computation results are reasonable and acceptable to the staff. l The applicant should, however, provide a comparison of the dynamic lateral pressure distribution along the depth of the seismic C'ategory I walls using the above mentioned two approaches (rock-interaction analysis and Seed and Whitman pro edure),for staff review. ' '
. Liquefaction Potential
~
There is no potential for liquefaction of sandstone that supports structures,. systems a$d components. The compacted Class Al backfill 'placed under, around and over,the seismic Category I buried piping. is well graded, has a maximum , particle size of 6 inches and has been compacted to 95 percent modified Proctor density.,The-soilcementbackfill-placeoatthelocationofthel construction ramps (at 95% Standard Proctor density) nas unconfined compressive strength of.
.-,s I' '
2-66 WNP-3.DSER SEC 2 w u
*.,i*
. l I
s e approximately 600 lb/in8 , which is similar to that of the sandstone. The applicant has not considered the liquefaction potential of Class Al structural fill or the soil cement backfill. We consider this approach to be reasonable, because as a result of the high compaction and compressive strength of these materials they can be considered to be not susceptible to liquefaction. 2.5.4.5 Conclusion Based on the applicant's design criteria and construction reports and on the results of the applicant's site investigations, laboratory and field tests, and analysis, the staff has concluded that the site and plant foundations will be
~
adequate to safely support the WPPSS Nuclear Project No. 3 (WNP-3) in accordance with the requirements of Appendix A to 10 CFR Part 100, pending satisfactory resolution of the open and confirmatory items identified above. 2.5.5 Stability of Slopes The WNP-3 plant site is surrounded by natural rock slopes gently dipping to the south of the plant and cut rock slopes rising to the east of the site. The natural rock slope to the south of the plant has an average slope away from the plant of 3 horizontal:1 vertical with a maximum slope of 1.5H:1V. - The crest of the slope is more than 250 feet away from the edge of the power-block structures. A typical cross-section through the slope is shown on FSAR Figure 2.5-110. The material forming the slope essentially consists of 6 weathered and fresh sandstone. The presence of a thin bed (<<10 feet) of residual soil near the toe of the slope was neglected by the applicant in the stability analysis of the natural slopes.
. Typical cross-sections analyzed for stability analyses of cut rock slopes east of the plant are shown on FSAR figures 2.5-111 and 2.5-112. The slope generally rises at an average slope of 4.5H:1V with maximum slope of 3:5H:1V. The toe of the slope is more than 350 feet away from the edge of the Category I structures.
The slope consists of weathered sandstone and residual soil. 2-67 WNP-3 DSER SEC 2
l As evidenced by the results of the site exploration, the applicant has deter-mined that the sandstone at the plant site is massive, without continuous joints, seams or layers of weaker material. The applicant determined the static strength parameters of the weathered and fresh sandstone on the basis of 13 untaxial compression tests. Based on these test results, a cohesion of 23 kips /ft2and an angle of internal friction of 0* were selected for analyses. We find these rock strength parameters to be reasonable and acceptab2: Shear strength parameters for residual soil to be used in stability analyses of cut slopes were obtained from seventeen unconsolidated-undrained triaxial tests, six consolidated-undrained triaxial tests and three unconfined compres-sion tests. Based on these test results, the applicant selected the following
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properties for residual soil: Angle of internal Cohesion, C friction, / High 1000 lb/ft 2 40* Average 1000 lb/ft2 30' Low 750 lb/ft 2 26* a We consider the applicant's use of these residual soil properties to be reasonable and acceptable. Natural Slope The static stability of the natural rock slope has been investigated by the applicant using the Simplified Bishop Method of Slices and the Sliding Wedge Method of Analysis. Two diff. cent groundwater conditions, viz., (1) normal groundwater level elevation of 320 feet (with drainage system operating)'and -~ (ii) groundwater level elevation of 365 feet (with drainage system blocked), are considered. The result of the applicant's analyses indicate that the natural slope has a minimum factor of safety of 5.5 for Slip Circle Method of Analysis and minimum factor of 7.6 for tne Sliding Wedge Method of Analysis. ! < -. 2-60 WNP-3 DSER SEC 2
Based on these results, the staff concludes that for static design loads, the natural slope south of the plant is stable. . The applicant has made a seismic stability evaluation of the natural slope for SSE condition using Slip Circle and Sliding Wedge analysis approaches. In these analyses, a horizontal seismic coefficient of 0.32 and a vertical seismic coefficient of 0.22 were used. The applicant's results for these analyses indicate minimuu factors of safety of 2.0 for Slip Circle and 2.64 for dynamic Wedge Method of Analysis. We consider that the margin of safety is adequate and acceptable. Cut Slopes
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The applicant has analyzed the cut rock slopes east of the plant structures using Bishop's Slip Circle Method and the Sliding Wedge Method of Analysis for static and dynamic cases. Two cross-sections shown on FSAR figures 2.5-111 and 2.5-112 have been analyzed. For seismic stability analyses, a seismic coefficient corresponding to 0.32 horizontal and 0.22 vertical were used for SSE. _ The following minimum factors of safety were computed by the Applicant from these analyses of cut slopes: - Factor of safety Method of analysis Static case Dynamic case Slip circle 3.36 1.45 Wedge method 3.13 1.40' We find these factors of safety for stability of cut rock slopes to be acceptable. The staff concludes that the natural and man-made slopes around the plant site , have been analyzed by the applicant in ar. appropriate and reasonable manner, l - l-l 2-62 WNP-3 DSER SEC 2 i
and, based on the results of analyses presented by the applicant, the staff concludes these slopes have an adequate margin of safety, and meet the require-ments of 10 CFR 100. The natural and cut slopes are, therefore, acceptable. 2.5.6 Embankments and Dams
; There are no embankments or dans associated with the WNP-3 plant used for plant flood protection or for impounding cooling water required for operation of the plant.
4 l 1 i I O . 2-70 WNP-3 DSER SEC 2
Table 2.1 Resident population versus distance year 0-1 0- 2 0-3 0-4 0-5 0-10 miler miles miles miles miles miles 1980 15 109 906 3061 5867 15165 1990 16 119 1000 3363 6519 16915 2030 20 ifit 1506 4989 10130 27103 1 3 a as 2-71 WNP-3 DSER SEC 2
e t i s f o
)
m ( s . e l i m 4 n i h t i w a e r a d n a e n o z n o i A t l a u p o p w o L 1 2 e r u g i F
?O . ' &7w 8le $n F9
I - 7 d E ? w 8 . 9 M n N Figur.e 2.2 Principal plant features in relation to exclusion area and property lines
l N I N 4 4 t z . m. w o ' e m
- o e
m n N Figure 2.3 Area within 50 miles of site e
T v'1 E ? w 8 . E vs E ~ Figure 2.4 Transportation routes and pipelines within 5 miles (_ m) of site Q
Table 2.2 Description of Allowable i Foundation Foundation foundation base Foundation bearing Factor of Category I dimensions elevations and depth to loading capacity safety structure (as-built) (MSL) fresh sandstone Static Dynamic Static Dynamic Static Dynamic Condensate / 171'-0 x Top El 390.00' 5'-0 thick mat 2.5 6.0 50.0 50.0 20.0 8.3 refueling 66'-0 on weathered water tank sandstone 40' to foundation fresh sandstone Dry cooling 85'-0 x Top El 390.00' 4'-0 thick mat 2.1 4.6 50.0 50.0 23.8 10.5 tower and 272'-0 on weathered control b1dg. sandstone 35' to fresh sandstone
)) Shield b1dg. 298'-0 x Top El 315.00' 9'-0 thick mat 13.6 17.7 85.0 85.0 6.3 4.8 o' containment, 310'-0 located in internal fresh sandstone structure, fuel handling b1dg. on common mat E
7 w C3 23 n ro
?
______m
. i I
3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 3.1 General 3.2 Classification of Structures, Systems, and Components 3.2.1 Seismic Classification 3.2.2 System Quality Group Classification - 3.3 Wind and Tornado Criteria and Loadings 3.3.1 Wind Design Criteria All Category I structures exposed to wind forces were designed to withstand the effects of the design wind. The design wind specified has a velocity of 105 mph based on a recurrence of 100 years. The procedures that were used to transform the wind velocity into pressure loadings on structures and the associated vertical distribution of wind pres-sures and gust factors are in accordance with ASCE Paper 3269 and ANSI - A58.1-1972. These documents are acceptable to the staff. The staff concludes that the plant design is acceptable and meets the require-ments of General Design Criterion 2. This conclusion is based on the following: 1 The applicant has met the requirements of GDC 2 with respect to the capability of the structures to withstand design wind loading so that their design reflects.. (1) appropriate consideration for the most severe wind recorded for the site l- with an appropriate margin; i l 3-1 WNP-3 DSER SEC 3
(2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena; and (3) the importance of the safety function to be performed. The applicant has met these requirements by using ANSI A58.1 and ASCE paper No. 3269, which the staff has reviewed and found acceptable, to transform the wind velocity into an effective pressure on structures and for selecting pressure coefficients corresponding to the structures geometry and physical configuration. The applicant has designed the plant structures with sufficient' margin to prevent structural damage during the most severe wind loadings that have been determined appropriate for the site so that the requirements of Item I listed above are met. In addition, the design of seismic Category 1 structures, as required by Item 2 listed above, has included in an acceptable manner load combination which occur as a result of the most severe wind load and the loads resulting from normal and accident conditions. The procedures used to determine the loadings on structures induced by the design wind specified for the plant are acceptable since these procedures have been used in the design of conventional structures and proven to provide a < conservative basis which together with other engineering design considerations assures that the structures will withstand such environmental forces. The use of these procedures provides reasonable assurance that in the event of design basis winds, the structural integrity of the plant structures that have to be designed for the design wind will not be impaired and, in consequence, safety-related systems and components located within these structures are adequately protected and will perform their intended safety functions if needed, thus , satisfying the requirement of Item 3 listed above. 3.3.2 Tornado Design Criteria All Category I structures exposed to tornado forces and needed for the safe shutdown of the plant were designed to resist a tornado of 240 mph tangential wind velocity and a 60 mph translational wind velocity. The simultaneous 3-2 WNP-3 DSER SEC 3
atmospheric pressure drop was assumed to be 2.25 psi in 1.9 seconds. Tornado missiles are also considered in the design as discussed in Section 3.5 of this report. We conclude that the plant design is acceptable and meets the recommendations of Standard Review Plan 3.3.2 and the requirements of General Design Criterion 2. This conclusion is based on the following: The applicant has met the recommendations of Standard Review Plan 3.3.2 and the requirements of GDC 2 with respect to the structure capability to withstand design tornado wind loading and tornado missiles so that their design reflects (1) appropriate consideration for the most severe tornado recorded for the site with an appropriate margin; (2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena; and (3) the importance of the safety function to be performed. The applicant has met these requirements.by using ANSI A58.1 and ASCE Paper No. 3269, which the staff has reviewed and found acceptable, to transform the wind velocity generated by the tornado into an effective pressure on structures and for selecting pressure coefficients corresponding to the structures geometry and physical configuration. The applicant has designed the plant structures with sufficient margin to prevent structural damage during the most severe tornado loadings that have been determined appropritte for the site so that the requirements of Item 1 . listed above are met. In addition, the design of seismic Category 1 structures, as required by Item 2 listed above, has included in an acceptable mannew 'oad combinations which occur as a result of the most severe tornado wind load and the loads resulting from normal and accident conditions. l The procedures utilized to determine the loadings on structures induced by the j design basis tornado specified for the plant are acceptable since these I l
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3-3 WNP-3 DSER SEC 3 1
( procedures have been used in the design of conventional structures and proven l to provide a conservative basis which together with other engineering design considerations assures that the structures will withstand such environmental forces. The use of these procedures provides reasonable assurance that in the event of design basis tornado, the structural integrity of the plant structures that have to be designed for tornadoes will not be impaired and, in consequence, safety-related systems and components located within these structures will be adequately protected and may be expected to perform necessary safety functions as required, thus satisfying the requirement of item 3 listed above. 3.4 Water Level (Flood) Design 3.4.1 Flood Protection s The design of the facility for flood protection was reviewed in accordance with Section 3.4.1 of the Standard Review Plan (SRP), NUREG-0800. An audit review of each of the areas listed in the " Areas of Review" portion of the SRP section . was performed according to the guidelines provided in the " Review Procedures" portion of the SRP section. Conformance with the acceptance criteria formed the basis for the staff's evaluation of the design of the facility for flood protection with respect to the applicable regulations of 10 CFR 50. To ensure conformance with the requirements of General Design Criterion 2,
" Design Bases for Protection Against Natural Phenomena," the staff reviewed the overall plant flood protection design including all systems and components whose failure due to flooding could prevent safe shutdown of the plant or result in the uncontrolled release of significant radioactivity. The applicant has provided protection from inundation and the static and dynamic effects of l flooding for safety related structures, systems, and components by providing - " hardened protection" in accordance with the guidelines of Regulatory Guide 1.59, " Design Basis Floods for Nuclear Power Plants." The plant site is a " dry site" as defined in Regulatory Guide 1.102, " Flood Protection for Nuclear Power Plants," Position C.1.
3-4 WNP-3 DSER SEC 3
The source of flooding at the site is the probable maximum flood (PMF) in the Chehalis River. The water level at the site vicinity resulting frem the PMF in the river is 76.2 ft MSL, approximately 313 ft below plant grade. Because all safety-related systems and components are located at the plant grade that is well above the highest PMF level, they are not subjected to flooding concerns . resulting from the PMF. Refer to Section 2.4.2 of this SER for further discus-sion of site flooding caused by local intense precipitation. The reactor auxiliary building (RAB), fuel handling building, and reactor build-ing are protected against flooding as a result of groundwater seepage by the installation of a permanent groundwater drainage system (GWDS). The GWDS per-manently lowers the groundwater in the vicinity of the plant. ' Watertight seals are also provided on all below grade penetrations of the RAB to further limit groundwater seepage into the building. The dry cooling towers and refueling
, water stcrage tank structures are at plant grade and thus are not susceptible to flooding as a result of groundwate seepage.
The GWDS is not classified as seismic Category I. The applicant has stated that this classification is adequate since a failure of the GWDS (clogging of the drain pipes) during a seismic event would not cause an appreciable rise in the groundwater level for a minimum of 115 days. In addition, the GWDS will be inspectable to ensure proper functioning at any time, including after an earth- - quake. Refer to Section 2.4.12 of this SER for further discussion regarding the GWDS. Within safety-related plant structures, protection against flooding from fail-ures in fluid piping systems as identified in the guidelines of Branch Technical Position ASB 3-1, " Protection Against Postulated Piping Failures in Fluid Systems Outside Containment," is provided by equipment location and drainage as described
. under Sections 3.6.1 and 9.3.3 of this SER.
On the basis of its review of the design criteria and bases, and the safety classification of safety-related systems, structures, and components necessary for a safe plant shutdown during and following flood conditions, the staff concludes that the design of the facility for flood protection conforms to the
- requirements of General Design Criterion 2 with respect to protection against l
natural phenomena and conforms to the guidelines of Regulatory Guides 1.59 and 3-5 WNP-3 DSER SEC 3 l
f 1.102 concerning flood protection. Therefore, the flood protection design j meets the acceptance criteria of SRP Section 3.4.1 and is acceptable. The l staff further concludes that the CESSAR interface requirements are satisfied by i-i the above described design. 3.4.2 Water Level (Flood) Design Procedures ' f The design flood level 'resulting from the most unfavorable condition or combina-t j tion of conditions that produce the maximum water level at the site is discussed j in Section 2.4, Hydrology. The hydrostatic effect of the flood was considered j in the design of all Category I struct'ures exposed to the water head. With the exception noted at the end of this section we conclude-that the plant
; flood structural design procedures are acceptable and meet the recommendations j of Standard Review Plan 3.4.2 and the requirements of General Design Criterion 2. l This conclusion is based on the following:
The applicant has met the recommendations of Standard Review Plan 3.4.2 and j the requirements of GOC 2 with respect to the capability to withstand the effects of the floor or highest groundwater level so that their design reflects
- (1) appropriate consideration for the most severe flood recorded for the site with an appropriate margin; (2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena; and I (3). the importance of the safety M ction to be performed.
The applicant has designed the plant structures with sufficient margin to prevent structural damage during the most severe flood or groundwater and the ~' I associated dynamic effects that have been determined appropriate for the site
, so that the requirements of Item 1 listed above are met. In addition, the 8
design of seismic Category I structures, as required by Item 2 listed above,. l has included-in an acceptable manner load combinations which occur as a result l of the most severe flood or groundwater-related loads and the loads resulting=_ I from normal and accident conditions. f 3-6. WNP-3 DSER SEC 3 4 _ -_ _ . , _ _ _ _ . - _ _ _ . . _ . __,_._ _
1 I The procedures used to determine the loadings on seismic Category I structures induced by the design flood or. highest groundwater level specified for the plant are acceptable since these procedures have been used in the design of I conventional structures and proven to provide a conservative basis which together with other engineering design considerations assures that the struc-tures will withstand such environmental forces. The use of these procedures provides reasonable assurance that in the event of , floods or high groundwater, the structural integrity of the plant seismic 1 Category I structures will not be impaired and, in consequence, seismic Cate-4 gory I systems and components located within these structures will be adequately
~~
j protected and may be expec'ted to perform necessary safety functions, as required,
- thus satisfying requirement of item 3 listed above.
l The applicant has used PVC as a Waterstop material in the RAB (question 220.11 f and reply). Since such material is subject to radiation deterioration and { attendant production of destructive chemical reactions, the staff will require additional studies to show that these materials, as used, will not be hazardous ] to the plant structures and will perform their intended functions throughout ] . j the lifetime of the plant. It was noted that the reply to question 220.11 is not complete. The applicants answer was compared with information shown on drawing WPSS-3240, G-2520-51, "RAB Internal Structures - Sheet 1." In that J drawing reference is made to silicone rubber joint sealer, non-specific material premolded joint filler, polyethylene and epoxy grout. A complete discussion of i this issue, which demonstrates that all of the organic materials used in water-i stop and structural joint filler applications are satisfactory to resist chemical and radiation deterioration, is required. 3.5 Missile Protection 3.5.1 Missile Selection and Description -- 3.5.1.1 Internally Generated Missiles (Outside Containment) The design of the facility for providing protection from internally generated j missiles (outside containment) was reviewed in accordance with Se'ction 3.5.1.1_
. 3-7 WNP-3 DSER SEC 3
l of the Standard Review Plan (SRP), NUREG-0800. An audit review of each of the areas listed in the " Areas of Review" portion of the SRP section was performed according to the guidelines provided in the " Review Procedures" portion of the SRP section. Conformance with the acceptance criteria except as noted below formed the basis for the staff's evaluation of the design of the facility for providing protection from internally generated missile outside containment with respect to the applicable regulations of 10 CFR 50. The acceptance criteria for the design of the facility for providing missile protection includes meeting Regulatory Guide 1.115, " Protection Against Low-Trajectory Turbine Missiles." The review of turbine missiles is discussed
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separately in Section 3.5.1.3. General Design Criterion 4, " Environmental and Missile Design Bases," requires protection of plant structures, systems, and components, whose failure could lead to offsite radiological consequences or that are required for safe plant shutdown, against postulated missiles associated with plant operation. The missiles considered in this evaluation include those missiles generated by rotating or pressurized (high-energy fluid system) equipment. Protection is provided by any one or a combination of compartmentalization, barriers, separation, orientation, and equipment design. The primary means of providing protection to safety-related equipment from damage resulting from internally generated missiles is through the use of plant physical arrangement. Safety-related systems and components of safety-related systems are physically separated from their redundant components. The applicant has provided an evaluation of potential missile sources from rotating equipment failures and pressurized component failures. The potential
. missiles resulting from this analysis are valves in high energy systems. The applicant's evaluation has verified that plant design features such as walls or' 4
separation of redundant systems will prevent these missiles from causing adverse effect on safety-related systems and components. Other missile sources are precluded by the design of the equipment itself. The staff concurs with the applicant's assumptions and evaluation for potential missiles outside con-tainment. Protection of safety-related equipment and stored fuel from_the _ effects of turbine missiles _is discussed in Section 3.5.1.3 of this SER. 3-8 WNP-3 DSER SEC 3
l I The potential sources of missiles which were evaluated in the fuel handling building are considered to be generated from failure of either a pressurized component or a rotating component. There are no high energy systems located j within the fuel handling building and therefore missiles from pressurized com-ponents are not postulated. The only rotating pieces of equipment in the fuel handling building are the component cooling water pumps, fuel pool cooling pumps, and the fuel pool cleanup pumps. All of these pumps and their motors are located at elevations below the spent fuel pool and are separated by seismic Category I barriers which prevent any missiles from penetrating the spent fuel pool. In addition, the s.taff requested the applicant to provide assur'ance that turbine
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driven pumps would not become a source of missiles or that missiles from the pump tur-bine could not damage safety-related equipment. There are two types of turbine driven pumps at the plant, the steam generator feedwater pumps (nonsafety-related) and the auxiliary feedwater pumps (safety-related). The steam gen-erator feedwater pumps incorporate redundant overspeed protection devices and both the turbine and pump casings are designed of sufficient strength to prevent the release of missiles generated by failure of the rotor or impeller. In the unlikely event that a missile penetrated the casing, the steam generator feed-water pumps are oriented such that the path of the missile would be away from - ! safety-related components. Further, each of the two trains of the auxiliary feedwater system is located in a separate concrete cubicle containing one motor driven and one turbine driven pump. Thus, the plant design incorporates phy-sical separation of trains A and B components with sufficient redundancy to ensure safe shutdown of the plant. The staff concludes that the above described design satisfies CESSAR interface requirements. [However, the applicant's analysis does not address pressurized tanks and gas cylinders (;t275 psig) from becoming potential missiles. Therefore~, the staff cannot conclude that the design is in conformance with the requirements of General Design Criterion 4 as it relates to protection against internally generated missiles until the applicant provides additional information in this
' regard. Resolution of this item will be reported in the final SER.]
3-9 WNP-3 DSER SEC 3 L
3.5.1.2 Internally Generated Missiles (Inside Containment) The design of the facility for providing protection from internally generated missiles inside containment was reviewed in accordance with Section 3.5.1.2 of the Standard Review Plan (SRP), NUREG-0800. An audit review of each of the areas listed in the " Areas of Review" portion of the SRP ,section was performed according to the guidelines provided in the " Review Procedures" portion of the SRP section. Conformance with the acceptance criteria formed the basis for the staff's evaluation of the design of the facility for providing protection from internally generated missiles with respect to the applicable regulations of 10 CFR 50. ' All plant structures, systems and components (SSC) inside containment whose failure could lead to offsite radiological consequences or that are required for safe plant shutdown must be protected against the effects of internally generated missiles in accordance with the requirements of General Design Cri-terion 4, " Environmental and Missile Design Bases." Potential missiles that could be generated inside containment are from failures of rotating components, pressurized components (high-energy fluid system) failures and gravitational effects. - ! With regard to potential missiles from pressurized high-energy systems inside l the containment, the applicant has analyzed the primary missiles that can be generated in the reactor vessel head area. The missiles considered in this i context were the closure head nut, closure head nut and stud and the control i rod drive assembly. The applicant's analysis verified that structures and
- shields provide protection for safety-related equipment from the above primary
, missiles. Also, potential gravitational missiles inside the containment result- ! ing from seismic events are prevented by either designing the structures, systems and components located inside the containment as seismic Category I or by l designing them to withstand seismic Category I loads without falling. With regard to potential missile sources from rotating equipment, the applicant ! has verified that all HVAC rotating equipment located inside containment is designed to withstand the' impact of self generated missiles such as fans or l , impeller blades by fabricating the equipment housing with sufficient material _ 4 thickness. Also, either duct reinforcement or missile barriers have been j # 3-10 WNP-3 DSER SEC 3 i.
-.~ .. - . .~.. - -. -. -- . . . - ~. .-.. - - . . - ..
provided at the discharge of the fans to contain the generated missiles and additionally prevent the generation of secondary missiles outside the HVAC rotating equipment housing. For a discussion of compliance with the criteria of Regulatory Guide 1.14, " Reactor Coolant Pump Flywheel Integrity," as it relates to potential missile sources, refer to Section 5.4.1.1 of this SER and of the CESSAR SER. The applicant has stated that temperature sensors or other detectors installed on pipes or in wells, nuts, bolts, studs, and combinations thereof contribute insignificantly to missile hazards due to the low amount of stored energy. [However, the applicant has not provided specific information regarding protec-
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tion against other potential primary system high-energy missile sources identi-fled in CESSAR FSAR Table 3.5-1 and in the CESSAR SER Section 3.5.1.2. Addi-tionally, the applicant has not provided information on secondary missiles generated by the impact of primary missiles associated with high-energy systems. Therefore, the staff cannot conclude that the WNP-3 design is in conformance with the requirements of General Design Criterion 4 as it relates to protection against internally generated missiles inside the containment. Resolution of this item will be reported in the final SER.] . 3.5.1.3 Turbine Missiles We have reviewed the WNP Unit 3 facility with regard to the turbine missile issue and conclude that the probability of unacceptable damage to safety-related systems and components due to turbine missiles is acceptably low (i.e., less than 10 ' per year) provided that the turbine missile generation probability is maintained to be 10
- per reactor year or less for the life of the olant by an acceptable maintenance program. In reaching this conclusion, the staff has factored into consideration the favorable orientation of the turbine generator.
The staff considers the turbine missile issue as a confirmatory item if the applicant agrees to: (1) submit for NRC approval, within three years of obtaining an operating J license, a turbine system maintenance program based on the manufacturer's calculations of missile generation probabilities, or 3-11 WNP-3 DSER SEC 3
(2) volumetrically inspect all low pressure turbine rotors at the second refueling outage and every other (alternate) refueling outage thereafter
- until a maintenance program is approved by the staff, and conduct turbine steam valve maintenance, (following initiation of power output) in accordance with present NRC recommendations as stated in SRP Section 10.2 of NUREG-0800.
3.5.1.4 Missiles Generated by Natural Phenomena The tornado missile spectrum was reviewed in accordance with Section 3.5.1.4 of the Standard Review Plan (SRP), NUREG-0800. An audit review of each of the areas listed in the " Areas of Review" portion of the SRP section was performed according to the guidelines provided in the " Review Procedures"-portion of the SRP section except as noted below. Conformance with the acceptance criteria formed the basis for the staff's evaluation of the tornado missile spectrum with respect to the applicable regulations of 10 CFR 50. The portions of the " Review Procedures" concerning the probability per year of damage to safety-related systems due to missiles was not used in the staff's review. The staff's review for this section of the SRP is concerned with establishing the missile spectrum, not with calculating the probability of damage. General Design Criterion 2, " Design Bases for Protection Against Natural Phenom-ena," requires that structures, systems, and components important to safety be designed to withstand the effects of natural phenomena, and General Design Cri-terion 4, " Environmental and Missile Design Bases," requires that these same plant features be protected against missiles. The missiles generated by natural phenomena of concern are those resulting from tornadoes. The applicant has
. identified a spectrum of missiles for a tornado Region III site as identified in Regulatory Guide 1.76, " Design Basis Tornado for Nuclear Power Plants," Po-sitions C.1 and C.2. The spectrum includes the weight, velocity, kinetic energy,
- impact area, and height in accordance with current tornado missile criteria.
The staff has reviewed this spectrum and concludes that it is representative of missiles at the site and is, therefore, acceptable. Discussion of the protec-tion (barriers and structures) afforded to safety-related equipment from the-
~
identified tornado missiles including compliance with the guidelines of
-. 3-12 WNP-3 DSER SEC 3 l
Regulatory Guide 1.117, " Tornado Design Classification," is provided in Sec-tion 3.5.2 of this SER. Discussion of the adequacy of barriers and structures designed to withstand the effects of the identified tornado missiles is provided in Section 3.5.3 of this SER. On the basis of its review of the tornado missile spectrum, the staff concludes that the spectrum was properly selected and meets the requirements of General Design Criteria 2 and 4 with respect to protection against natural phenomena and missiles and the guidelines of Regulatnry Guide 1.76 with respect to identi-fication of missiles generated by natural phenomena and is, therefore, accept-able. The tornado missile spectrum meets the acceptance criteria of SRP Sec-
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tion 3.5.1.4. The staff further concludes that the above described design satisfies the CESSAR interface requirements. 3.5.2 Structures, Systems, and Components to be Protected from Externally Generated Missiles . The design of the facility for providing protection from tornado generated mis-siles was reviewed in accordance with Section 3.5.2 of the Standard Review Plan . (SRP),NUREG-0800. An audit review of each of' the areas listed in the " Areas of Review" portion of the SRP section was performed according to the guidelines provided in the " Review Procedures" portion.of the SRP section. Conformance < with the acceptance criteria formed the basis for the staff's evaluation of the design of the facility for providing protection from tornado generated missiles with respect to the applicable regulations of 10 CFR 50. General Design Criterion (GDC) 2, " Design Basis for Protection Against Natural Phenomena," requires that all structures, systems, and components important to safety be protected from the effects of natural phenomena, and GDC 4, " Environ-mental and Missile Design Bases," requires tha t all structures, systems, and components important ta safety be protected from the effects of externally gen-erated missiles. The h P-3 site is located in tornado Region III as identified in Regulatory Guide 1.76, " Design Basis Tornado for Nuclear Power Plants." The tornado missile spectrum is discussed in Section 3.5.1.4 of this SER. Protec-tion from low-trajectory turbine missiles including compliance with RG 1.115, i
" Protection Against Low-Trajectory Turbine Missiles", is discussed in Sec .
tion 3.5.1.3 of this SER. i
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3-13 WNP-3 DSER SEC 3 l
l The applicant has identified all safety-related structures, systems, and com-l ponents requiring protection from externally generated missiles. All safety-related structures are designed to withstand postulated tornado generated mis-i siles without damage to safety related equipment. Safety-related systen s and j components and stored fuel and spent fuel pool are located within tornado-missile protected structures or are provided with tornado missile barriers. The two dry cooling towers which constitute the ultimate heat sink for WNP-3 are enclosed in structures designed to prevent tornado and missile impact damage to any vital component of the towers. The cooling tower fans, partic-ularly, are protected from tornado generated missiles by missile grating. Therefore, the staff concludes that the guidelines of Regulatory Guides 1.13,
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" Spent Fuel Storage Facility Design Basis," 1.27, " Ultimate Hea't Sink for Nuclear Power Plants," and 1.117, " Tornado Design Classification," concerning tornado missile protection for stored fuel, ultimate heat sink and the spent fuel pool are met. With regard to HVAC openings, the outside air HVAC intakes for the control room, the fuel building, diesel generator (DG) area, and the electrical equipment and battery rooms are all protected from tornado-missiles by protective missile grating. Also, the component cooling water system dry cooling towers electrical equipment room outside air HVAC intake and exhaust openings are protected against tornado-missiles by missile grating. idition-ally, the applicant states that the DG combustion air intake opening is protec-ted from external missiles by shield bars and that both the normal and emer-gency combustion air exhaust path openings are protected against externally generated missiles. [However, the applicant has not provided assurance that HVAC exhaust openings such as those for the control room, fuel building, ECCS area / fuel building emergency filtration system, diesel generator area, and the electrical equipment and battery rooms are protected from tornado-missiles.
Also, the staff is unable to conclude that possible damage to the DG combustion air exhaust assembly (for example, the silencer) from tornado-missile, such as
. crimping will not disable the DG. Therefore, the staff cannot conclude that the requirements of GDCs 2 and 4 with respect to missile protection and the guidelines of RG 1.117 concerning tornado-missile protection for safety-related structures, systems and components are met. Resolution of this concern will be reported in the final SER.]
[On the basis of the preceding, the staff concludes that, except as noted above,
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j the applicant's list of safety-related structures, systems and components to be
- _ 3-14 WNP-3 DSER SEC 3
.. -- . . - . . . - . - . - - . =..- . --- _ - _ . . . --...-.~.-..:
protected from externally generated missiles and the provisions in the plant design providing this protection are in accordance with the requirements of GDCs 2 and 4 with respect to missile protection and the guidelines of Regula-tory Guides 1.13, 1.27, and 1.117 as they relate to tornado missile protection for safety-related structures, systems and components including stored fuel and ultimate heat sink. The staff therefore concludes that the design meets the ac:eptance criteria of SRP Section 3.5.2 except as noted above. Also, the staff cannot conclude that the design meets the intent of CESSAR interface requirements. Resolution of the concerns identified above will be reported in the final SER.] 4 3.5.3 Barrier Design Procedures The plant Category I structures, systems and components are shielded from, or designed for, various postulated missiles. Missiles considered in the design of structures include tornado generated missiles and various containment internal missiles, such as those associated with a loss-of-coolant accident. Information has been provided indicating that the procedures that were used in the design of the structures, shields and barriers to resist the effect of missiles are adequate. The analysis of structures, shields and barriers to determine the effects of missile impact was accomplished in two steps. In the < first step, the potential damage that could be done by the missile in the immediate vicinity of impact was investigated. This was accomplished by estimating the depth of penetration o'f the missile into the impacted structure. Furthermore, secondary missiles are prevented by fixing the target thickness well above that determined for penetration. In the second step of the analysis, the overall structural response of the target when impacted by a missile is determined using established methods of impactive analysis. The equivalent
. loads of missile impact, whether the missile is environmentally generated or aceidentally generated within.the plant, are combined with other applicable loads as is discussed in Section 3.8 of this report.
We conclude that the barrier design is acceptable and meets the recommendations of Standard Review Plan 3.5.3 and the requirements of General Design Criteria 2 and 4 with respect to the capabilities of the structures, shields, and barriers to provide sufficient protection to equipment that must withstand the effects _. 3-15 WNP-3 DSER SEC 3
F f of natural phenomena (tornado missiles) and environmental effects including the effects of missiles, pipe whipping and ifischarging fluids. This conclusion is based on the following: The procedures' utilized to determine the ef'fects ard loadings on seismic Category I structures and missile shields ano barriers induced by design basis e missiles selected for the plant are acceptable since these procedures provide a conservative basis for engineering design to assure that the structure or barriers are adequately resistant to and'will withstand the effects of such forces. s, The use of these procedu'res provides reasonable asurance that in the event of
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design basis missiles striking seismic Category I structures or otner missile shields and barriers, the structural integrity of the structures, shields and barriers will not be impaired or degraded to an extent that will result in a loss of required protection. Seismic Category I systems and components protected 4 by these structures are, therefore, adequately protected against the effects of missiles and will perform their inte$ led safoty function, if needed. Conform-ance with these procedures is.an acceptable basis for satisfying in pa't the requirements of General Design Criteria 2 and 4. 3.6 PEctection Against Dynamic Effects Associated with the Postolated Rupture - of Piping 3.6.1 Plant Design for Protection Against Postulated Piping Failures in Fluid' Systems Outside Containment The design of the facility for providing protection against postulated piping
- ~ ,.. failures outside containment was reviewed in.accordance with Section 3.6.1 of the Standard Review Plan (SRP), NUREG-3800. An audit rev'few of each of the areas listed in the." Areas of Review" portion of the SRP section was performed according to the guidelines provided in the " Review Procedures" portion of the SRP section. Conformance with the acceptance criteria formni the, basis for thestaff'sevaluationofthedesignofthe'facilityfor$rovidingprotection against postulated piping failures outsice containment with respect to the applicable regulations of 10 SFR'60.
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3-16' WNP-3 DSER SEC 3
===-..-.w--:..= : === = : = : =.===-..: . = - -
- ==.
1 s
,The staff's guidelines for meeting the requirements of General Design Crite-
[ rion 4, " Environmental and Missile Design Bases," concerning protection against {. postulated piping failure in high-energy and moderate-energy fluid systems l outside containment are contained in Branch Technical Position (BTP) ASB 3-1,
" Protection Against Postulated Failures'in Fluid Systems Outside Containment."
The applicant has. identified high- and moderat.e-energy piping systems in accor-dance with these guidelines and has also identified those systems requiring pro-
, tection from postulated piping failures (refer to Section 3.6.1 of the CESSAR
- SER for a discussion of the high- and moderate-energy fluid systems outside containment which are in the CESSAR scope).
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- The plant design accomodates the effects of postulated pipe breaks and cracks j including pipe whip, jet impingement and environmental effects. The means used j to protect essential (safety-related) systems and components include physical
! separation, enclosure within suitably designed structures, pipe whip restraints, and equipment shields. To be consistent with BTP ASB 3-1, the applicant has , utilized separation as the primary means of protection, and where separation ! was not feasible, one of the other acceptable methods of protection was used. I j The plant design includes the ability to sustain a high-energy pipe break acci- l ,i dent coincident with a single active failure and retain the capability for safe
- cold shutdown. For postulated pipe failures, the resulting effect will not -
cause the loss of function of power supplies or controls and instrumentation needed to complete a safety action and will not preclude the habitability of
.the control room as indicated in BTP ASB 3-1. .
- The applicant has also analyzed the effects of moderate-energy line breaks out-l side containment on safety-related systems by postulating cracks in moderate j energy lines at any location. For moderate-energy essential system piping cracks
! . in other than dual purpose moderate-energy essential systems which. satisfy the guidelines of BTP ASB 3-1, Position 3.b.(3), a single active failure in the ! redundant train or trains of the essential system was also considered and it j was shown that safe shutdown will not be affected or the functional capability- 1 of the. essential systems will not be compromised. -[However, the staff cannot l accept the applicant's assumption that a seismic event concurrent with a crack l in non-seismically designed moderate-energy piping is.not a credible event since ( 8 3-17' WNP-3 DSER SEC 3 I _ . _ . _ , . _ _ _ . . . . _ _ , _ . _ . , _ _ _ _ _ _ _ . , _ _ _ _ _ . _ , , - , , , . _ . . , _ _ . _ , __, . . , _ - .
. u 1
l
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the seismic event by itself can cause a pipe break in .a non-seismically designed
/
piping system. Also, it cannot grant credit for mitigation of flooding con-l sequences resulting from postulated seismically induced pipe failures by non-seismically designed systems, components or equipment such as floor drainage ! systems,. sump pumps etc. Therefore, the staff cannot conclude thnt; moderate- '
- energy systems have been designed to meet the intent of the guidelines set forth in BTP ASB 3-1. Resolution of this concern will be reported in the final SER.]
. s+ >
< The main steam and feedwater systems up to the first restraint outside contain-ment are classified as part of the break exclusion (BEX) boundar'y as defined in
, item B.1.6 of BTP MEB 3-1,'"Eostulated Breaks and Leakagt Locations in Fluid j System Piping Outside Containment." At the staff's request, th'e applicant pro- ,
vided the results of a subcompartment analysis of a nonmechanistic break in
- theselinestodeterminetheenvironmentaleffectsinthecompartdenttihousing the main steam and feedwater lines. The applicant determined that the struc-tural integrity of the applicable steam tunnel (there are 2 steam tunnels) which houses the BEX portion of these lines will not be affected by the pressure increase from the resulting blowdown. Thetunnelisgentedto'relievethepres-sure effects.'V Main steam isolation and feedwater isolation valves (MSIVs and FWIVs) functional capability will be maintained by assuring.that they are envi-ronmentally qualified to conservat.ive bounding conditions determined by the analysis. The staff cokcurs with this analysis. Environmental qualification <
of essectial auxiliary feedwater (AFW) system pumps and flow control / isolation valvesandAFWturbinesteamsupplyvalves$ndessentialequipmentlocatedin l the steam tunnel including the MSIVs and FWIVs and the atmospheric dump valves
- - is discussed in Section 3.11 of this SER. [The applicant has not provided a pressure and, environmental analysis for the other subcompartments outside con-tainment which house high-energy piping (the CVCS charging and letdown, steam i generator (SG) blowdown and auxiliary steam lines). The staff evaluation of i 4 . the results of the analysis to assure that safd ~related equipment is protected from the postulated failure in these pipiag n s -is will be provided in the final SER.]+ ,
ll I [0n the ba' sis of its review described above, the staff cannot conclude that the applicanthsadequatelydesignedandprotectedareasandsystemsrequiredfor i safe plant shutdown following postulated failures in-high- and moderate-energy.
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3 1.' 3-18 _V[P-3DSERSEC3
piping outside containment as required by GDC 4 until it has completed its review of the subcompartment pressure and environmental analysis for the CVCS charging and letdown lines, SG blowdown lines and auxiliary steam lines, and until its concern relating to moderate-energy piping identified above is re-solved. The resolution of all these concerns will be discussed in the final SER. CESSAR interface requireme,:ts (refer to CESSAR SER, Seqtion 3.6.1) specify that safety-related equipment must be protected from the effects of high- and . moderate-energy pipe failures. Therefore, the staff cannot conclude that these requirements have been met by the applicant until the review of applicant's responses to the concerns identified above has been completed.]
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3.6.2 Determination of Rupture Locations and Dynamic Effects Associ~ated with the Postulated Rupture of Piping 3., 7 Seismic Design 3.7.1 Seismic Input The peak accelerations associated with SSE were selected based on the seismicity - evaluation described in Section 2.5 of the FSAR. The earthquake on which SSE is modeled has a Richter magnitude of 7-1/2, and originates at a distance of approxi-mately 22 miles from the site. The peak horizontal *baserock acceleration at the < site associated with this earthquake is 0.32 g (SSE). The duration of strong motions for this earthquake is estimated to be approximately 30 seconds. The operating basis earthquake (OBE) is chosen to be 1/2 SSE or 0.16 g. Vertical accelerations are 2/3 of horizontal, that is, 0.22 g SSE and 0.11 g 08E. Horizontal design response spectra conform to the recommendations of Regulatory Guides 1.60 and 1.61. -The vertical design response spectra.does not comply
. with the recommendations of Regulatory Guide 1.60 in the 33 to 50 Hz range.
As discussed in DSER Section 2.5.2, the Geosciences Branch is reviewing the above seismic input assumptions and will report conclusions in the SER. Although the duration of_the earthquake record was to be 30 seconds it.was found that the duration of earthquake was truncated to 20 seconds in the horizontal direction (Structural Design Audit Finding #5). This matter is - under investigation at this writing.
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3-19 WNP-3 DSER SEC 3 ' l
The horizontal and vertical acceleration time histories were derived by applying a deconvolution methodology to a finite element model of the rock site. The staff disagrees with the results of this analysis (question 220.13, and Struc-tural Design Audit Finding #1). Essentially, the applicant has allowed a significant reduction in the earthquake excitation at the base mat based on his deconvolution analysis of a rock site. This issue remains unresolved as of
- this writing.
3.7.2 Seismic System Analysis 3.7.3 Seismic Subsystem Analysis The scope of review of the Seismic System and Subsystem Analysis for the plant included the seismic analysis methods for all Category I structures, systems and components. It included review of procedures for modeling, seismic soil-structure interaction, development of floor response spectra, inclusion of torsional effects, evaluation of Category I structure overturning, and determi-nation of composite damping. The review has included design criteria and procedures for evaluation of interaction of non-Category I structures and piping, with Category I structures and piping and effects of parameter varia-tions on floor response spectra. The review has also included criteria and seismic analysis procedures for reactor internals and Category I buried piping outside the containment. The system and subsystem analyses were performed by the applicant on an elastic bases. Modal time history methods form the bases for the analyses of all major Category I structures. Response spectra methods were used in the design of Category I systems and components.
. The finite element approach is used to evaluate soil-structure interaction and structure to structure interaction effects from seismic excitation.
As noted above in Section 3.7.1 of this report, the seismic acceleration input at the base mat used by the applicant is in question. (See questions 220.15 l i I T~~~ 3-20 WNP-3 DSER SEC 3
.-. . _ _ _ = .
and 220.16). In addition several issues regarding the construction and use of 't floor response spectra are also unresolved at this time. These issues are: (1) Derivation of floor response spectra without consideration of out-of plane acceleration factors (See question 220.18 and Audit Funding #17). It is noted that the torsional-effects analysis described in paragraph 3.7.2.11 of the FSAR was not made available to the audit team at the Structural Design Audit. It is hereby requested that the applicant make this analysis available to the staff as part of his resolution of this issue. (2) Actual application of methods for peak-broadening and smoothing of response spectra are not in accordance with methods recommended by Regulatory Guide 1.122 (to which the applicant is committed). See question 220.22, 220.23, and Structural Design Audit Findings #11 and #19. 3.7.4 Seismic Instrumentation Program The type, number, location and utilization of strong motion accelerographs to record seismic events and to provide data on the frequency, amplitude and phase relationship of the seismic response of the containment structure were compared with Regulatory Guide 1.12 requirements. Supporting instrumentation is to be installed on Category I structures, :ystems and components in order to provide data for the verification of the seismic responses determined analytically for such Category I items. The ranges of the types of instrumentation as well as the readout locations have not been provided. A seismic surveillance scheme as outlined in the SRP was not provided although it is said to be incorporated in the technical specifications for the plant (See question 220.19). The technical specifica-tions for the plant were not available, as of this writing, in order that the surveillance scheme could be verified.
- c_ _ ~ . 3-21 WNP-3 DSER SEC 3 .
I 1 1 3.8 Design of Seismic Category I Structures 3.8.1 Concrete Containment 3.8.2 Steel Containment The containment consists of a free-standing steel shell located within a reinforced concrete shield building. These are founded on a common mat with (but separated by seismic gaps above the mat) a reactor auxiliary building. The containment was designed, fabricated, constructed and tested as a Class MC vessel in accordance with Subsection NE of the ASME Boiler and Pressure Vessel
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Code, Section III. Loads include an appropriate combination of' dead and live loads; thermal loads; seismic and loss-of-coolant accident-induced loads including pressure and jet forces. The analysis of the containment was based.on elastic thin shell theory. The allowable stress and strain limits are in accordance with those delineated in the applicable sections of Subsection NE of the ASME Code, Section III, for the various loading conditions. . The following issues are unresolved at this time: . 4 (1) Question 220.25, which is a request for drawings, remains unanswered. (2) Question 220.26 which reque:ts validation of computer programs used in the containment design remains unanswered. An advance copy of the applicants proposed answer was provided to the staff in reply to portion (c) of Structural Design Audit Finding #4. However, the staff considers this proposed answer to be incomplete and lacking in specific details. (3) Portions (a) and (b) of Structural Design Audit Finding #4 regarding compliance with ASME code requirements and buckling analysis remain unresolved. (4) The answer to question 220.28 is not responsive. Neither a complete description nor results of calculations for the containment static analy-l sis was provided.
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3-22 WNP-3 DSER SEC 3 t
(5) Question 220.27 was a staff request for an ultimate capacity analysis of the containment in accordance with the criteria contained in the SRP. The answer provided by the applicant indicated that the ultimate capacity analysis would not be prepared. (6) According to Table 1.8-3 of the FSAR a Design Report meeting the guidelines set forth in Appendix C to SRP Section 3.8.4 was prepared for the contain-ment. The staff was to examine the design report at the Structural Design Audit but did not do so. Therefore, the applicant is requested to make the Design Report available for staff review. This comment also applies to SER sections 3.8.3, 3.8.4 and 3.8.5. (7) Question 220.30 regarding use of plastic filler materials remains unan-swered. 3.8.3 Concrete and Structural Steel Internal Structures 4 The containment interior structures consist of walls, compartments and floors. The major code used in the design of concrete internal structures is ACI 318-71. For steel internal structures the AISC Specification, " Specification for the Design, Fabrication and Erection of Structural Steel for Building," is used. (For equipment supports, Subsection NF of the ASME Code is used.) 4 The containment concrete and steel internal structures were designed to resist various combinations of dead and live loads, accident induced loads, including pressure and jet loads, and seismic loads. The load combinations used cover
' those cases likely to occur and include all loads which may act simultaneously.
The design and analysis procedures that were used for the internal structures are the same as those on previously licensed applications and, in general, are in accordance with procedures delineated in the ACI 318-71 Codes and in the AISC Specification for. concrete and steel structures, respectively. l The containment internal structures were designed and proportioned to remain within limits established by the Regulatory staff under the various load combinations. .These limits are, in general, based on the ACI 318-71 Code and on the AISC Specification for concrete and steel structures, respectively, modified as appropriate for load combinations that are considered extreme. , 3-23 WNP-3 DSER SEC 3
The material of construction, their fabrication, construction and installation, are in accordance with the ACI 318-71 Code and AISC Specification for concrete and steel structures, respectively. The response to question 220.35 is unsatisfactory. Differences between the application of ACI 318-71 and the staffs referenced design standard consisting of Reg. Guide 1.142 and ACI 349 were not documented. All that is stated in Table 1.8-3 is that'ACI 318-71 was used in accordance with PSAR commitments. This is not considered to satisfy the applicant's obligation to document deviations from NUREG 0800. This comment applies to SER sections 3.8.4 and 3.8.5 as well. 3.8.4 Other Seismic Category I Structures Category I structures other than containment and its interior structures are all of structural steel and concrete. The structural components consist of slabs, walls, beams and columns. The major code used in the design of concrete Category I structures is the ACI 318-71, " Building Code Requirements for Reinforced Concrete." For steel Category I structures, the AISC, " Specification . for the Design, Fabrication and Erection of Structural Steel for Buildings," is used. The concrete and steel Category I structures were designed to resist various combinations of dead loads; live loads, environmental loads including winds, tornadoes, OBE and SSE; and loads generated by postulated ruptures of high energy pipes such as reaction and jet impingement forces, compartment pressures, and impact effects of whipping pipes. The design and analysis procedures ~ that were used for these Category I struc- . tures are the same as those approved on previously licensed applications and, in general, are in accordance with procedures delineated in the ACI 318-71 - Codes and in the AISC Specification for concrete and steel structures, respectively. The various Category I structures are designed and proportioned to remain within limits established by the Regulatory staff under the various load . _ 3-24 WNP-3 D$ER SEC 3
combinations. These limits are, in general, based on the ACI 318-71 Code and not the AISC Specification.for concrete and steel structures, respectively, modified as appropriate for load combinations that are considered extreme. The materials of construction, their fabrication, construction and installation are in accordance with the ACI 318-71 Code and the AISC Specification for concrete and steel structures, respet.tively. With respect to Other Category I structures, the following issues remain outstanding: (1) Structural Design' Audit Finding #18 regarding the arbitrary reduction of seismic accelerations obtained from the NASTRAN analysis of the shield building remains unresolved at this time. (2) Structural Design Audit Finding #21 regarding use of negative signs on load factors in the shield building analysis is not resolved. (3) Structural Design Audit Finding #22 regarding computer printouts with compressive stresses undefined in the output is not resolved. (4) Structural Design Audit Finding #23 regarding verification of the static vs. dynamic analysis of the dry cooling tower is not resolved. 3.8.5 Foundations Foundations of Category I structures are described in Section 3.8.5 of the SAR. Primarily, these foundations are reinforced concrete of the mat type. The major code used in the design of these concrete mat foundations is ACI f 318-71. These concrete foundations-have been designed to resist various combination of dead loads, live loads, environmental loads including winds, tornadoes, OBE and SSE, and loads generated by postulated ruptures of high energy pipes. The design and analysis procedures that were used for these Category I founda-
- tions are the same as those approved on previously licensed applications and, -
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3-25 WNP-3 DSER SEC 3
l i in general, are in accordance with procedures delineated in the ACI 318-71 Code. The various Category I foundations were designed and proportioned to remain within limits established by the Regulatory staff under the various load combinations. These limits are, in general, based on the ACI 349 Code modified as appropriate for load combinations that are considered extreme. The materi-als of construction, their fabrication, construction and installation, are in accordance with the ACI 318-71 Code. 3.8.6 Structural Audit Structural Design Audit Finding #24 regarding foundation uplift in the stabil-ity analysis remains outstanding. General Comments (Section 3.8) (1) Structural Design Audit Finding #25 regarding the applicants documentation of load and load combination values in the various analyses remains unresolved. (2) The staff has not completed its review of all of the answers provided by the applicant to staff questions. Further issues may therefore become apparent as these reviews are completed. < 3.11 Environmental Qualification of Electrical Equipment Important to Safety and Safety-Related Mechanical Equipment 3.11.1 Introduction Equipment that is used to perform a necessary safety function must be demon-
. strated to be capable of maintaining functional operability under all service conditions postulated to occur during its installed life for the time it'is required to operate. This requirement, which is embodied in GDC 1 and 4 and in Sections III, XI, and XVII of Appendix B to 10 CFR 50, is applicable to equip-ment-located inside as well as outside containment. More detailed requirements anS guidance relating to the methods and procedures for demonstrating this capability for electrical equipment have been set fo.rth in 10 CFR 50.49, 3-26 WNP-3 DSER SEC 3
4 .
" Environmental Qual.ification of Electric Equipment Important to Safety for
- Nuclear Power Plants," and NUREG-0588, " Interim Staff Position on Environmental I Qualification' of Safety Related Electrical Equipment." NUREG-0588 supplements A-IEEE Standard 323 and rarious NRC Regulatory Guides and industry standards.
i l 3.11.2 Background l l NUREG-0588 was issued in December 1979 to promote a more orderly and systematic { implementation of electrical equipment qualification programs by industry and i to provide guidance to the NRC staff for use in ongoing licensing reviews. The > 4 positions contained in this section provide guidance on (1) how to establish l environmental service conditions, (2) how to select methods tha't are considered appropriate for qualifying equipment in different areas of the plant, and (3) other factors such as margin, aging, and documentation. l - In February 1980, the NRC requested the Washington Public Power Supply System I to review and evaluate +he environmental qualification documentation for each a item of safety-related electric equipment which could be exposed to a harsh environment and to identify- the degree to which their qualification program complies with the staff positions described in NUREG-0588. IE Bulletin 79-01B, ,
- " Environmental Qualification of Class IE Equipment," issued January 14, 1980, and its supplements dated February 29, September 30, and October 24,^1980, l
[ established environmental qualification requirements for operating reactors. , This bulletin and its supplements were provided to the applicant for consider-ation in his review. { A final rule on environmental qualification of electric equipment important to j safety for nuclear power plants became effective on February 22, 1983. This-rule, Section 50.49 of 10 CFR 50, specifies the requirements to be met for-
. demonstrating the environmental qualification of electrical-equipment important to safety located in a harsh environment. In'accordance with this rule,- ~
- equipment for WNP-3 may be qualified to the criteria specified in Category I of NUREG-0588.
4 1 The qualification requirements for mechanical equipment are principally con-i tained in Appendices A and B of 10 CFR 50. The qualification methods defined
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! .in NUREG-0588 can also be applied to mechanical equipment.
. 3-27 WNP-3 DSER SEC 3
In response to the above requirements, the applicant has provided some prelimi-nary equipment qualification information in Section 3.11 of the FSAR and a submittal dated April 28, 1981. 3.11.3 Completeness of the Environmental Qualification Program In order to demonstrate compliance with the final rule, 10 CFR 50.49, the following information must be submitted by the applicant before an operating license can be granted. In accordance*with the scope defined in 10 CFR 50.49, provide: (1) A list of all nonsafety-related electrical equipment located in a harsh environment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions by the safety-related equipment. A description of the method used to identify this ' equipment must be included. The nonsafety-related equipment identified must be included in the environmental qualification program. (2) A statement that all safety-related electric equipment in a harsh environ-ment, as defined in the scope of 10 CFR 50.49, is included in the equipment qualification program (including equipment required for MELB, fuel handling < accident,etc.). (3) A list of all Category 1 and 2 post-accident monitoring equipment currently installed, or to be installed before plant operation, in response to Regulatory Guide 1.97, Revision 2. The equipment identified must be included in the environmental qualification program. Also provide information demonstrating qualification of all equipment in a harsh environment within the scope of 10 CFR 50.49, or provide justification for interim operation pending completion of qualification as required by 10 CFR 50.49. This material should be submitted to allow sufficient time for staff review and approval before issuance of an operating license. 1 Although there are no detailed requirements for mechanical equipment, GOC 1
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and 4; Sections III and XVII of Appendix B to 10 CFR 50; and SRP 3.11, l 3-28 WNP-3 OSER SEC 3 ,
Revision 2, contain the following requirements and guidance related to equipment qualification:
- Components shall ce designed to be compatible with the postulated environ-mental conditions, including those associated with LOCAs.
- Measures shall be established for the selection and review for suitability of application of materials, parts, and equipment that are essential to safety-related functions.
Design control measures shall be established for verifying the adequacy of design.
- Equipment qualification records shall be maintained and shall include the results of tests and materials analyses.
In order to demonstrate compliance with General Design Criterion 4 of Appendix A to 10 CFR Part 50 for mechanical equipment, the staff requires that the applicant perform a review and evaluation that includes the followinn: . (1) Identification of safety related mechanical equipment located in harsh environmental areas, including required operating time. (2) Identification of the nonmetallic subcomponents of this equipment. (3) Identification of the environmental conditions for which this equipment ! must be qualified. The environments defined in the electrical equipment program are also applicable to mechanical equipment. (4) Identification of nonmetallic material capabilities. (5) Evaluation of environmental effects. The list of equipment identified should be submitted. From this list the staff will select approximately three items of mechanical equipment for which documentation of their environmental qualification should be provided for review. Also, the results of the review should be provided for all 3-29 WNP-3 DSER SEC 3
w . .. . . , -
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4 mechanical equipment in harsh environment areas and corrective actions identified. Justification for interim operation must be submitted prior to fuel load for any mechanical equipment whose qualification cannot be established. For mechanical equipment, the staff review will concentrate on materials which are sensitive to environmental effects for example; seals, gaskets, lubricants, fluids for hydraulic systems, diaphragms, etc. Additionally, all safety-related equipment should be subjected to a maintenance, 2 surveillance and periodic testing program in accordance with Reg. Guide 1.33, to detect any age-related degradation that could affect the qua'lification of the equipment and to maintain the equipment in a qualified condition, j Upon receipt of a final submittal, the staff will review the environmental
] qualification program for compliance and request any additional information needed to establish its acceptability. The staff will then perform an audit review of electrical equipment environmental qualification files and associated 1 installed equipment. Following this audit, an SER supplement will be prepared 4
documenting the results of.the review and evaluation. Prior to granting of an i operating license, the staff must be able to conclude that full compliance with
! 10 CFR 50.49 and all applicable rules and regulations has been demonstrated.
t 4 l 1 e I e i
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3-30 WNP-3 DSER SEC 3
4 REACTOR 4.1 Introduction 4.2 Fuel Design WNP-3 is a CESSAR System 80 plant. We have reviewed the CESSAR fuel system and concluded (Rubenstein, June 8, 1983) that the CESSAR fuel system has been '1 designed so that (a) the fuel system will not be damaged as a result of normal operation and anticipated operational occurrences, (b) fuel damage during postulated accidents would not be severe enough to prevent control rod insertion when it is required, and (c) core coolability will always be maintained, even after severe postulated accidents, and thereby meets the related requirements of 10 CFR Part 50.46; 10 CFR Part 50, Appendix A, General Design Criteria 10, 27, and 35, 10 CFR Part 50, Appendix K, and 10 CFR Part 100. This conclusion is based on the following: (1) Combustion Engineering has provided sufficient evidence that these design objectives will be met based on operating experience, prototype testing, and analytical predictions. Those analytical predictions dealing with control rod ejection and fuel densification have been performed in accor-dance with the guidance of Regulatory Guide 1.77 and with an acceptable alternative of Regulatory Guide 1.126 (Ref. 56). (2) Combustion Engineering has provided for testing and inspection of new fuel to ensure that it is within design tolerances at the time of core loading. The NRC will require (Rubenstein, February 17, 1983) that applicants (a) make commitments to perform CEA reactivity checks and post-irradiation surveillance to detect anomalies or confirm that the *
- fuel has performed as expected and (b) provide assurance of adequate shoulder gap clearance.
1 4-1 WNP-3 DSER SEC 4
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(3) Combustion Engineering has described methods of adequately predicting fuel rod failures during postulated accidents so that radioactivity releases are not underestimated and thereby meets the related requirements
'of 10 CFR Part 100. In meeting these requirements, C-E has (a) used the fission product release assumptions of Regulatory Guides 1.25 (Ref. 57) and 1.77, and an acceptable (more conservative) alternative to 1.4 (Ref.
58), and (b) performed the analysis for fuel rod failures for the rod ejec-tion accident in accordance with the guidelines of Regulatory Guide 1.77. On the basis of the NRC's review of the fuel system design, the NRC concluded (Rubenstein, June 8, 1983) that the CESSAR fuel system design has met all the requirements of the applicable regulations, regulatory guides, and current regulatory positions ~. All applicants referencing the CESSAR FSAR (including the WPPSS application for WNP-3) must supply the following applicant-specific information (Rubenstein, June 8, 1983): (1) A CEA surveillance program (see paragraphs 4.2.1.1(j) and 4.2.3.1(j)). - In Q490.2 (Berlinger, April 14,1983), WPPSS was asked to provide this information. No response has been received. (2) A fuel assembly loads analysis due to combined seismic and LOCA forces (see paragraph 4.2.3.3(d)). In Q490.2, WPPSS was asked to provide this information. We have not yet received a response. (3) A commitment to perform a general fuel surveillance program (see paragraph 4.2.4.3). In Q490.2, WPPSS was asked to provide this information. We 4 have not yet received a response. (4) Certification that the mechanical fracturing analysis result conforms to the acceptance criterion (see. paragraph 4.2.3.2(g)). In Q490.2, WPPSS was asked to provide this information. We have not yet received a response. l
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l l 4-2 WNP-3 DSER SEC 4
In addition, the following license condition will be required to address the concern discussed in Section 4.2.3.1(g) related to axial growth.
"The licensee shall confirm that adequate shoulder gap clearance will be maintained during the first two refueling outages (Cycles 1/2 and 2/3).
This may be done either by analysis or hardware modification and shall be based on measurements taken on a sufficient number of fuel assemblies irradiated in'WNP-3." 4.3 Nuclear Design The Nuclear design of WNP-3 is identical to the corresponding item in the CESSAR (Combustion Engineering Standard Safety Analysis Report). 4.4 Thermal-Hydraulic Design The CESSAR SER discusses all items in this section except for the staff review of: (1) the loose parts monitoring system; (2) the instrumentation for detection of inadequate core cooling requirements as described in item II.F.2 of NUREG-0737 and; (3) the plant-specific information on the Core Protection Calculators (CPCs) and statistical combination of uncertainties (SCU). 4.4.1 Loose Parts Monitoring System The acceptance criteria for the Loose Parts Monitoring System (LPMS) are set forth in Regulatory Guide 1.133, Revision 1, " Loose-Part Detection Program for the Primary System of Light Water Reactors," May 1981. The applicant has pro-vided a description of the LPMS which will be used at WNP-3. The design will include pairs of pieso-electric accelerometers located at the following natural collection regions: (1) hot leg nozzles adjacent to steam generator inlet plenum; (2) incore instrument nozzle located on the vessel bottom beam, adja-cent to vessel inlet plenum; (3) Control Element Drive Mechanism (CEDM) nozzle located on vessel head. The system will provide eight channels with the capa-bility of monitoring and detecting loose parts impacting within 3 feet of a sensor having a kinetic energy of 0.5 ft-lb and weighing from 0.25 to 30 lb. The system will remain functional following an Operating Basis Earthquake (OBE).
- 4-3 WNP-3 OSER SEC 4
. l l
i In response to the staff's request for additional information on the LPMS system, the applicant responded (Letter from G. C. Sorensen to G. W. Knighton, September 2, 1983) that the LPMS will be in full compliance with the requirements of Regulatory Guide 1.133. This will include the submission of a training program for plant personnel prior to power operation which will address Technical Specifications, purpose and operation of the LPMS system. The applicant has committed to pro-vide Technical Specifications for the LPMS in accordance with Position C.5 of Regulatory Guide 1.133 which will specify the limiting conditions for operation and the surveillance requirements. We will also require the applicant to com-mit to provide prior to power operation a final design report showing confor-mance with positions C.1 through C.6 of Regulatory Guide 1.133 which contains the following: (1) An evaluation of the LPMS for conformance to Regulatory Guide 1.133. (2) A description of the system hardware, operation and implementation of the loose parts detection program after start-up testing. This should also include the baseline data and alarm settings. (3) A description and evaluation of diagnostic procedures used to confirm the presence of a loose part. A sample table of contents of the LPMS description is enclosed in Table 4-1. 4.4.2 Inadequate Core Cooling Requirements (SRP-4.4-Section 11.9) The applicant has not responded to questions asked relative to inadequate core cooling requirements as specified in Item II.F.2 of NUREG-0737 and this is therefore an open item. 4.4.3 Plant Specific Information (SRP 4.4-Section II.1) The applicant has not responded to a question asked relative to plant-specific information on: (1) the application of statistical combination of uncertainties (SCU) and plant-specific instrument uncertainties; and (2) digital core pro-tection calculator (CPC) including the Reactor Power Cutback System (RPCS) for
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4-4 WNP-3 DSER SEC 4
1 which infc sation is needed on the plant-specific data base constants, soft-ware impl Mentation testing and effects of SCU on DNBR. This is therefore an 4 open item. 4.6 Functional Design of Reactivity Control Sy_ stems The functional design of reactivity control systen.3 is within the scope of , CESsag, Rcyc;. t.o Sec.t. ion 4.6 of the CESSAR SER for this discussion. ) I 1 s 4 j 4 l i i G 4-5 WNP-3 DSER SEC 4
. . . ._ =___ _ _. . __ ._ _
Table 4.1 Loose Part Detection Program Description I. System Description A. Scale piping diagram showing LPM sensor locations. B. Sensor specifications (type, manufacturer, sensitivity, temperature rating,etc.). C. Sensor mounting details (drawing and procedure). D. Preamplifier or line driver (type, manufacturer, location and specifications). Functional description of LPtis.
- F.
Theory of operation, detection logic, alarm display. 1.
- 2. Data recorder specifications (No. of channels, length of
. recording, frequency range, and conditions under which recording is initiated).
II. Operational Procedures A. System Calibration Procedures and Results
- 1. Initial and subsequent calibrations
- 2. Functional check, as defined in Regulatory Guide 1.133
- 3. Channel check, as defined in Regulatory Guide 1.133
- 8. Plant Operator Instructions for Use of LPMS
- 1. Procedures for routine operation
- 2. Procedures to be used following indication of a loose part
- a. Method to confirn existence of loose part
- b. Method of diagnose a loose part (size and location)
III. Evaluation for Conformance to Regulatory Guide 1.133 ! A. Loose Part Detection Program ) 8. Loose Part Detection System I i l 4-6 WNP-3 DSER SEC 4
- - -. , . , ~ - - , - . , ,m-.., , - ..-., -- ,,, -. -.-,, ,,--c n n-,. ---- , , n,. ,, - .
5 REACTOR COOLANT SYSTEM 5.1 Summary Description 5.2 Integrity of Reactor Coolant Pressure Boundary 5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing This section was prepared with the technical assistance of DOE co.ntractors from the Idaho National Engineering Laboratory. ~ 5.2.4.1 Compliance with the Standard Review Plans a The July 1981 Edition of the " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants" (NUREG-0800) includes Section 5.2.4,
" Reactor Coolant Pressure Boundary Inservice Inspection and Testing." The i
Washington Nuclear Project No. 3 (WNP-3) review is continuing because the Applicant has not submitted a Preservice Inspection (PSI) Program and has not j completed the PSI examinations. In FSAR Table 1.8-3, the Applicant has commit-3 ted to comply with the Standard Review Plan (SRP) 5.2.4 acceptance criteria. The staff review to date was conducted in accordance with SRP Section 5.2.4 l except as discussed below, i Paragraph II.3, " Acceptance Criteria, Examination Categories and Methods," will be reviewed when the complete PSI Program has been received from the Applicant. Paragraph II.4, " Acceptance Criteria, Inspection Intervals," has not been. reviewed because this area' applies only to inservice inspections (ISI), not to PSI. This subject will be addressed during review of the ISI program after licensing.
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i . i 5-1 WNP-3 DSER SEC 5
Paragraph II.5, " Acceptance Criteria, Evaluation of Examination Results," has been reviewed. The Applicant committed in the FSAR to incorporate ASME Code Section XI, Article IWB-3000, " Standards for Examination Evaluation," into the PSI Program. However, ongoing NRC generic activities and research projects indicate that the presently specified ASME Code procedures may not always be capable of detecting the acceptable size flaws specified in the IWB-3000 acceptance standards. For example, ASME Code procedures specified for volu-metric examination of reactor vessels, bolts and studs, and piping have not proven to be capable of detecting the acceptable size flaws in all cases. The staff will continue to evaluate the development of new or improved proce-dures and will require that these improved procedures be made a part of the inservice examination requirements. The Applicant's repair pro'cedures based on ASME Code Section XI, Article IWB-4000, " Repair Procedures," have not been reviewed. Repairs are not generally necessary in the PSI program. This subject will be addressed during the staff review of the ISI program. , Paragraph II.7, " Acceptance Criteria, Code Exemptions," will be reviewed when the completed PSI Program Plan is submitted by the Applicant. Paragraph II.8, " Acceptance Criteria, Relief Requests," has not been completed because the Applicant has not identified all limitations to examination. Specific areas where ASME Code examination requirements cannot be met will be identified as performance of the PSI progresses. The complete evaluation of the PSI program will be presented in a supplement to this Safety Evaluation Report (SER) after the Applicant submits the required examination information, identifies all plant-specific areas where ASME Code Section XI requirements cannot be met, and provides a supporting technical justification. I i 5.2.4.2 Examination Requirements General Design Criterion 32, " Inspection of Reactor Coolcat Pressure Boundary,"- Appendix A of 10 CFR Part 50 requires, in part, that components which are part of the reactor coolant pressure boundary be designed to permit periodic examina-tion and testing of important areas and features to assess their structural l and leak-tight integrity. To ensure that no deleterious defects develop during service, selected welds and weld heat-affected-zones (HAZ) will be examined . 5-2 WNP-3 DSER SEC 5 I
- . . . , , --,r. - , .--,,.a- - -- -- -
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periodically. The design of the ASME Code Class 1 and 2 components of the reactor coolant pressure boundary incorporates provisions for access for inservice examinations, as required by Paragraph IWA-1500 of Section XI of the ASME Code. Section 50.55a(g), 10 CFR Part 50, defines the detailed requirements for the preservice and inservice programs. Based upon the construction permit date of April 11, 1978, this section of the regulations requires that a preservice inspection program be developed and implemented using at least the Edition and Addenda of Section XI of the ASME Code applied to the construction of the particular components. The components (including supports) may meet requirements set forth in subsequent editions of this Code and Addenda which are incorporated by reference in 10'CFR 50.55a(b) subject to the limitations and modifications listed therein. The initial ISI program must comply with the requirements of the latest Edition
'and Addenda of Section XI of the ASME Code in effect twelve months prior to the date of issuance of the operating license, subject to the limitations and modifications listed in Section 50.55a(b) of 10 CFR Part 50.
5.2.4.3 Evaluation of Compliance with 10 CFR 50.55a(g) Review has been completed on the information presented in the FSAR through < Amendment 3 dated April 1983. The preservice examination on the pipi.1g and components, except NSSS components, will be examined in accordance wit i the requirements of the 1977 Edition of ASME Code Section XI with Addenda u ough Summer 1978. The NSS components will be examined in accordance with the requirements of the 1974 Edition of ASME Code Section XI with Addenda through Summer 1975 except that the steam generator tubing will be examined in accordance with ASME Code Section XI 1980 Edition with Addenda through Winter 1980. - " The Preservice Inspection (PSI) Program for systems and components withir the - reactor coolant pressure boundary has not been received. However, the Applicant has stated in the FSAR that these systems and components will be examined per
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5-3 WNP-3 DSER SEC 5
the applicable Code requirements. Based on the review of the FSAR, the staff has established technical positions that should be included in the PSI Program. The Applicant has committed to identify all plant-specific areas where the Code requirements cannot be met after the examinations are performed and provide a supporting technical justification for requesting relief. The SER input will be completed after the Applicant: (1) Dockets a camplete and acceptable PSI Program, (2) Submits the requested additional information regarding the PSI /ISI program, and (3) Submits all relief requests with a supporting technical justification. The staff considers the review of the. PSI Program an open issue subject to the Applicant providing an acceptable response to the above requirements. The initial Inservice Inspection Program has not been submitted by the Applica-t. This program will be evaluated after the applicable ASME Code Edition and Addenda can be determined based on Section 50.55a(b) of 10 C'FR Part 50, but before inservice inspection commences during the first refueling outage. 5.2.4.4 Conclusions The conduct of periodic examinations and hydrostatic testing of pressure-retaining components of the reactor coolant pressure boundary, in accordance with the requirements of Section XI of the ASME Code and 10 CFR Part 50, will provide reasonable assurance that structural degradation or loss of leak-tight . integrity occurring during service will be detected in time to permit correc-tive action before the safety functions of a component are comprised. Compli-ance with the preservice and inservice examinations required by the Code and 10 CFR Part 50 constitutes an acceptable basis for satisfying the inspection requirements of General Design Criterion 32. 5-4 WNP-3 DSER SEC 5
5.2.4.5 References
- 1. NUREG-0800, Standard Review Plans, Section 5.2.4, " Reactor Coolant Boundary Inservice Inspection and Testing," July 1981.
- 2. Code of Federal Regulations, Volume 10, Part 50.
- 3. American Society of Mechanical Engineers Bciler and Pressure Vessel Cede, Section XI, Division 1.
1974 Edition, through Summer 1975 Addenda 1977 Edition, through Summer 1978 Addenda 1980 Edition, through Winter 1980 Addenda 5.2.5 Reactor Coolant Pressure Boundary Leakage Detection The reactor coolant pressure boundary (RCPB) leakage detection systems were reviewed in accordance with Section 5.2.5 of the Standard Review Plan (SRP), NUREG-0800. An audit review of each of the areas listed in the " Areas of Review" portion of the SRP section was performed according to the guidelines provided in the " Review Procedures" portion of the SRP section. Conformance with the acceptance criteria formed the basis for the staff's evaluation of the reactor coolant pressure boundary leakage detection systems with respect to the applicable regulations of 10 CFR 50. A limited amount of leakage is to be expected from components forming the reac-tor coolant pressure boundary (RCPB). Means are provided for detecting and identifyirg *h.s leakage in accordance with the requirements of General Design Criterton (GL,C) 3C, " Quality of Reactor Coolant Pressure Boundary." Leakage is,clas- . sified into two types - identified and unidentified. Components such as valve stem packing, pump shaft seals, and flanges are not completely leak tight. Since this leakage is expected, it is considered as identified leakage and is monitored, limited, and separated from other leakage (unidentified) by directing it to closed systems as identified in the guidelines of Position C.1 of Regula-tory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems." Refer to CESSAR SER Section 5.2.5 for discussion on the sources, disposition, and indication of identified leakage. ~ 5-5 WNP-3 DSER SEC 5
j e
- Unidentified leakage, which includes steam generator tube sheet and intersys-
- tem leakage, is monitored by several devices as identified in the guidelines of 7 i Positions C.1, C.3 and C.4 of Regulatory Guide 1.45. Steam generator tube leak- !
age is monitored by the condenser air removal system radiation monitors, steam , ] generator blowdown system radiation monitors, or routine steam generator water j samples. The method of detection of intersystem leakage depends on the partic- ) ular interfacing system. Leakage of reactor coolant through the safety injec-
- tion system can'be identified by high pressure alarms in the control room. In ,
i the event seat leakage takes place past two shutdown cooling isolation valves, j the leakage will pressurize the shutdown cooling lines and lift the two relief j valves. The discharge from the relief valves is directed to the safety injec- . l tion system recirculation sump and monitored as an unidentified leakage source. ; 1 . The means of detecting intersystem leakage of primary coolant to the component
! cooling water system through the letdown heat exchanger, reactor coolant pump seal heat exchanger and thermal barriers is as follows. Heat exchanger leaks will produce inleakage of reactor coolant and fission products into the cooling
) water. Such inleakage will increase the radioactivity content of the cooling
- water. The increase will be detected by the component cooling water system
{ radiation monitors located in the recirculation lines across the component cool-
- ing water pumps of each train. Leakage of reactor coolant also increases the j inventory in the component cooling water system, causing an increase in the l
a surge tank level which would result in a high level alarm in the main control i room. i ! Leakage to the primary rector containment from unidentified sources is collected i and the flow rate monitored with an accuracy of I gpm or_better. Indication i of unidentified leakage into the containment is monitored by four independent methods: j (1)' Sump Level and Flow Monitoring
- Unidentified leakage inside the containment including condensate from the
] containment fan coolers will flow to the containment drain sump. Leakage , from the reactor coolant system (RCS) will result in either an increase in i humidity.in containment (which will cause condensation on the fan cooler 5 5-6 WNP-3 DSER SEC 5 i
. . - - - . _ , , _ - . , _ . _ _ , , , . - - , , - . . _ , . . _ _ . . , - , , , _ _ . , _ , _ . , _ , , , , , _ _ _ . _ _ - , , , _ _ . _ . , _ . . , _ _ , . . . _ _ . . , - ~ . . _ , , . _ _
_ - _a . . - _ _ _ _ , _ . . . . . _ _ . _ _ _ .._ _ : _ -. ._ _ .. coils) or water on the floor. Thus, RCS unidentified leakage will pass to -l the containment sump. All flow entering the sump is routed first to a mea-surement tank. The tank is fitted with a level transmitter that sends a signal proportional to the tank level to the main centrol room. An alarm J occurs whenever the equivalent of one gpm in one hour is exceeded as pre-scribed by Regulatory Guide 1.45, Positions C.2 and C.S. Sump level and
! flow monitoring equipment will remain functional after being subject to an SSE.
i Unidentified leakage inside the reactor cavity will be collected in the reactor cavity sump and will be pumped directly into the measurement tank at the containment drain sump. Pump start alarm, sump lev'el alarm, and
; flood detection alarms are provided for the reactor cavity area to alert l the operator in case of any leakage into the area.
(2) Airborne p' articulate Radioactivity Monitoring j The containment atmosphere is monitored for radioactive particulates by j the containment atmosphere / containment purge airborne radiation monitors. l These monitors are a pair of identical and redundant units. The particu-I late channel in each monitor is capable of detecting the airborne radio-3 active particulates resulting from an increase of one gpm in the leakage - l rate from the primary coolant pressure boundary into the containment atmo-sphere within one hour. In addition, the particulate filter tape and the l' downstream iodine filters may be removed for laboratory analysis. The monitoring equipment used for leakage detection has been designed to re-j main functional following an SSE as indicated in guidelines of Regulatory Guide 1.45, Position C.6.
. (3) Airborne Gaseous Radioactivity Monitoring The containment atmosphere is monitored for radioactive gases by the containment / atmosphere purge airborne radiation monitors. These monitors are a pair of identical and redundant units. The gaseous channel in each monitor is capable of detecting the airborne radioactive gases resulting from an increase of one gpm in the leakage rate from the primary coolant
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5-7 WNP-3 DSER SEC 5
- . : -...--.a-.-... .- --. - . - = , . . . - . -. . _ . . . . - .
[ pressure boundary into the containment attrosphere within one. hour. The i 4 airborne gas monitoring equipment used for leakage detection;has been de-
~
signed to remain functional following an SSE. 4
. (4) Reactor Coolant _ Inventory Monitoring s
Abnormal leakage from the reactor coolant system is also detected through j measurement of the net amount of makeup flow to the system (refer to CESSAR j SER Section 9.3.4 and Section 5.2.5 for further discussion).
]
- As. described above, the RCPB leakage flow and radioactive monitors within
, containment are seismi Category I, testable, and may be calibrated as identified in the guidelines of Positions C.6, C.7, and C.8 of Regulatory
; Guide 1.45. Further,.their Jccuracy meets the guidelines of Pnsition C.5
; of Regulatory Guide 1.4ru ..
/ ) %
Additional soufces of indicatic,n of unidentified leakage include contain-
; ment pressure, temperature and hucidity indicators, pressuri:er level s
~
l indicators, and low pressure safety injection header pressure. Technical
; specifications will include limiting conditions for identified ana unidenti- ] fied leakage and will also address availability of the various leakage de-l tection systems to assure adequate coverage of all times as indicated in -
l Regulatory Guide 1.45, Posittori C.9. i I .
; On the basis of tne preceding, the staff concludes that the RCPB leakage ; detection systems are diverse and p'rovide reasonable assurance that primary j system leakage (both identified and unidentified) will be detected. The Estem: meet the requirements of GDC 30 with respect to provisions for RCPB leak detection and identification, and the guidelines of Regulatory Guide 1.45
- with respect to RCPB leakage detection system design. The staff therefore con-cludes that the design is acceptable and that it meets the accep'.ance criteria q of SRP Section 5.2.5. It further concludes that the CESSAR interface require-a ments,asdiscussedinCESSAROy.Section5.2.5,aresatisfiedbytheabove-
'I described design. i' 8 WNP-3 DSER SEC 5 n ab
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'] ; 5.3.1 Reactor Vessel Materials (Fracture Toughness) 4 i-j In this section we have reviewed the fracture toughness of ferritic reactor r j vessel and reactor coolant pressure boundary materials, and the materials
{ surveillance program' for the reactor vessel beltline. The acceptance criteria i~, and references which are the basis for this evaluation are set forth in Para- , d graph 11.3.a of Standard Review Plan (SRP) Section 5.2.3 and Paragraphs II.5, [ II.6 and II.7 (Appendices G and H, 10 CFR Part 50) of SRP Section 5.3.1 in i NUREG-0800 Rev. - 1, dated July 1981. A discussion of this review follows.
- i 5:e
-i GDC 31, " Fracture Prevention of Reactor Coolant Pressure Boundary," of Appen-dix A ts 10 CFR Part 50, requires that the reactor coolant pressure boundary be D
4 designed with sufficient margin to assure that when stressed under operating, J maintenance, testing, and anticipated transient conditions, the boundary behaves in a non-brittle manner and the probability of rapidly propagating l, fracture is minimized. GDC 32, " Inspection of Reactor Coolant Pressure Boundary," h of Appendix A to 10 CFR Part 50, requires, in part, that the reactor coolant jj pressure boundary be designed to permit an appropriate material surveillance
- .i program for the reactor pressure vessel.
io :- ._
,! The staff reviews the materials selection, toughness requirements, and extent j of materials testing conducted by the applicant in accordance with the above i criteria subject to the rules and requirements of 10 CFR Part 50, Paragraph !! 50.55a, " Codes and Standards"; 10 CFR Part 50, Appendix G, " Fracture Toughness
- h Requirements"; and 10 CFR Part 50, Appendix H, " Reactor, Vessel Material Surveil-1ance Program Requirements."
[{
- i l (1). Compliance to Section 50.55(a), 10 CFR 50
$l. The Edition and Addenda of the ASME Code which are' applicable to the design
' and fabrication of any reactor vessel are specified in Section 50.55a of 10 CFR
.ll',! Part 50 and are based on the construction permit date. The construction , [- permits for WNP-3 was isswa on April 11, 1978. Based on the construction-l permit date,'Section 50.55a of 10 CFR Part 50 requires that the WNP-3 reactor , r
~. 5-9 WNP-3 DSER SEC 5 1
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vessel meet, as a minimum, the requirements of the 1971 edition of the ASME
, Code, Summer 1972 Addenda. The WNP-3 FSAR states that the reactor vessels were designed, fabricated, tested, inspected, and stamped according to the 1971 ASME Code, Summer 1973 addenda. Therefore, the applicant complied with the explicit ; requirements of Section 50.55a,10 CFR Part 50.
(2) Compliance to Appendix G, 10 CFR 50 The staff review of this section is discussed in CESSAR SER Section 5.3.1. In j that SER, the staff determined that the requirements of Appendix G,10 CFR 50
, had been met, with the exception of Sections III.B.5 and III.C.2. However, subsequent to that review, Appendix G, 10 CFR 50 was revised. The revised
. Appendix G, 10 CFR 50 was published on May 27, 1980 and became effective
'i July 26, 1983. The applicant must review the material test program and pres-
. sure-temperature limits to determine whether the test program and pressure-temperature limits comply with'tbe explicit requirements of Appendix G, 10 CFR 50. If they do not, the applicant must propose an alternative test program and pressure-temperature limits which will provide a safety margin equivalent to that required by Appendix G, 10 CFR 50.
(3) Compliance to Appendix H, 10 CFR 50 *
?
The toughness properties of the reactor vessel beltline materials must be
; monitored throughout the service life of WNP-3 by a material surveillance program that meets the requirements of Appendix H, 10 CFR 50. CESSAR indicates that a(D System 80 reactor vessel material surveillance programs will satisfy the requirements of Appendix H,10 CFRL50. Since each reactor vessel material surveillance program is designed based on the actual , reactor 4 vessel beltline ,
~
4 properties, the staff in its SER for CESSAR indicated that the actual reactor ni
;j vesse,1 beltline fracture toughness properties and surveill,ance data must be 1
l reviewed for each reference plant. The applicant has provided actual reactor vessel beltline fracture toughness properties for WNP-3, but has not' reported . 1 the materials or withdrawal schedule for the WNP-3 surveillance capsules. A revised Appendix H, 10 CFR 50 was published on May 27, 1983 and became effective j on July 26, 1983. As a result, t,he capsule withdrawal schedule requirements I were revised to require that the applicant's withdrawal schedule be submitted _ i m I
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5-10 WNP-3DSE}SEC,5 1 e
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I for approval and include its technical justification. Until the materials and withdrawal schedules for the WNP-3 surveillance capsules have been identified, we cannot complete our review of the applicant's compliance to Appendix H, 10 CFR 50. (4) Conclusions of Complic ee to Appendices G and H. 10 CFR 50 Appendix G, " Protection Against Nonductile Failure," Section III of the ASME Boiler and Pressure Vessel Code, will be used, together with the fracture toughness test results required by Appendices G and H, 10 CFR Part 50, to j ' calculate the reactor coolant pressure boundary pressure-temperature limitations for WNP-3.
-! The fracture toughness tests required by the ASME Code and Appendix G of 10 CFR Part 50 will provide reasonable assurance that adequate safety margins against the possibility of nonductile behavior or rapidly propagating fracture can be established for all pressure-retaining components of the reactor coolant boundary. The use of Appendix G, Section III of the ASME Code, as a guide in establishing safe operating procedures, and the use of the results of the j fracture toughness tests performed in accordance with the ASME Code and NRC regulations, will provide adequate safety margins during operating, testing, maintenance, and anticipated transfent conditions. Compliance with these Code -
, provisions and NRC regulations constitutes an acceptable basis for satisfying the fracture toughness requirements of GDC 31.
The materials surveillance program, required by Appendix H, 10 CFR Part 50,
- j will provide information on material properties and the effects of irradiation 0 on material properties so that changes in fracture toughness of material in the WNP-3 reactor vessel beltline caused by exposure to neutron radiation can J be properly assessed and adequate safety margins against the possibility of vessel failure can be provided.
Compliance with ASTM E-185-73 and Apper. dix H 10 CFR Part 50, assures that the surveillance program constitutes an acceptable basis for monitoring radiation-induced changes in the fracture toughness of the reactor vessel material and satisfies the materials surveillance requirements of GDC 31 and GDC 32. . 5-11 WNP-3 DSER SEC 5
, . . . _ . . _ . ..___.___.-.__u..._.._____ ___u ._ _ _ _ __ _ . _. ~ . _.... - _... a -
4 5.3.2 Pressure-Temperature Limits i In this section we review the applicant's pressure-temperature limits for operation of their reactor vessels. The acceptance criteria and list of j references which are the basis for this evaluation are set forth in the Standard j Review Plan (SRP) Section 5.3.2 of NUREG-0880 Rev.1, dated July 1981. A
. discussion of this review follows.
Appendix G, " Fracture Toughness Requirements," and Appendix H, " Reactor Vessel i Material Surveillance Program Requirements," 10 CFR Part 50, describe the i conditions that require pressure-temperature limits for the reactor coolant pressure boundary and provide the general bases for these limit's. These j appendices specifically require that pressure-temperature limits must provide safety margins for the reactor coolant pressure boundary at least as great as the safety margins recommended in the ASME Boiler and Pressure Vessel Code, Section III, Appendix G, " Protection Against Nonductile Failure." Appendix G, 10 CFR Part 50 requires additional safety margins whenever the reactor core is critical, except for low-level physics tests. l The following pressure-temperature limits imposed on the reactor coolant pressure boundary during operation and tests are reviewed to ensure that they provide adequate safety margins against nonductile behavior or rapidly propagat- -
; ing failure of ferritic components as required by GDC 31:
(1) Preservice hydrostatic tests, l (2) Inservice leak and hydrostatic tests, I (3) Heatup and cooldown operations, and (4) Core operation. l . The applicant has not provided actual pressure-temperature limits for WNP-3, but has referenced the methodology outlined in the CESSAR FSAR, which the
! staff found to be acceptable (see the CESSAR SER, Section 5.3.2). However, 4
pressure-temperature limits depend upon the fracture toughness properties of i the ferritic reactor pressure vessel materials. Until the applicant submits actual WNP-3 pressure-temperature limit curves and provides the information i _
~
l 5-12 WNP-3 DSER SEC 5 e I
. - . . . . . . - - - . . - . . .- . . . . - . . . - , . . . ~ . . - -..-.=. =
I l requested in Section 5.3.1 of this SER, the staff will not be able to complete l its review of the WNP-3 pressure-temperature limits. 1 The pressure-temperature limits to be imposed on the reactor coolant system
; for all operating and testing conditions to ensure adequate safety margins . 'j against nonductile or rapidly propagating failure must be in conformance with ] j established criteria, codes, and standards acceptable to the staff. The use of operating limits based on these criteria, as defined by applicable regula-d tions, codes, and standards, will provide reasonable assurance that nonductile j or rapidly propagating failure will not occur, and will constitute an acceptable basis for satisfying the applicable requirements of GOC 31.
- i 2
i' 5.3.3 Reactor Vessel Integrity The staff has reviewed the FSAR sections related to the reactor vessel integ-rity of WNP-3. Although most areas are reviewed separately, reactor vessel integrity is of such importance that a special summary review of all factors relating to reactor vessel integrity is warranted. l The staff has reviewed the information in each area to ensure that it is complete with respect to the BOP scope and to ensure that no inconsistencies l exist that would reduce the certainty of vessel integrity. In addition, the - 3 staff has reviewed the CESSAR FSAR sections related to reactor vessel integrity to ensure that it is complete with respect to the NSSS scope and to ensure that i all interface requirements have been met by the applicant. A discussion of j this review is contained in the CESSAR SER (NUREG-0852). The areas reviewed
; are:
1 (1) Design (Section 5.3.1)
- 1 j (2) Materials of construction (Section 5.3.1)-
] (3) Fabrication methods (Section 5.3.1)
~
d (4) Operating conditions (Section 5.3.2) l f
-j l
The acceptance criteria and references which are the basis for the evaluation f i are set forth in Paragraphs II.1, II.6 and II.7 (Appendices G and H, 10 CFR i i l ~'.~- 5-13 WNP-3 DSER.SEC 5 4
+
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Part 50) of Standard Review Plan (SRP) Section 5.3.3 in NUREG-0880 Rev. 1, 4 dated July 1981. I i Until the applicant suppifes the information necessary to complete our evalua-tion of compliance to Appendices G and H,10 CFR 50, and reactor vessel pres- {.
, ! sure-temperature limits, we cannot complete our evaluation of the strucutral j integrity of the WNP-3 reactor vessel.
MATERIALS ENGINEERING BRANCH
-; MATERIALS APPLICATION SECTION j-Question 251.1 J:
Appendices G and H. 10 CFR Part 50 wer,e revised in the Federal Register on May 27, 1983 and became effective on July 26, 1983. i
, ..2
- a. Identify ferritic reactor coolant pressure boundary materials that do not !
I comply with the fracture toughness requirements of Section 50.55a and Appendices G and H of 10 CFR Part 50.
,a
; b. For materials that cannot meet the fracture toughness requiremects of
-i
'j Section 50.55a and Appendices G and H.of 10 CFR Part 50, provide alterna- -
] tive fracture toughness data and analyses to demonstrate their equivalence >
[{ to the requirements of 10 CFR Part 50. 4 4 I c. To demonstrate conformance to Appendices G and H, 10 CFR Part 50; (1) Provide pressure temperature limit curves for hydrostatic pressure j and leak tests, heat-up cooldown and core operations.
. .i q
i (2) Identify the withdrawal schedule, lead factor, test samples and - 1
.j materials in the Reactor Vessel Materials Surveillance Program.
d (3) Provide technical justification for the capsule withdrawal schedule. s
' ' ~ ~ '
, 5-14 WNP-3 DSER SEC 5
..y_,. . .. . .
7.,_ y ~ y . 7,,. e ,
. , . , , ,, ., m , , - . - - - ~
. _ _ _ . - _ . . . - - - . . .- ~ - - . ~ -...-. .- . . .. ~ - 2. . a i
5 j (4) Report the projected peak end of life neutron fluence (E > 1 MeV) at the inside surface and 3/4 T location of the reactor vessel. j Question 251.1 i j Are all weld materials, which were used in fabrication of the reactor vessel beltline, identified in FSAR Tables 5.2-4e and 5.2-4F? Is the copper chemical composition for these beltline weld materials less than the copper chemical a composition for the limiting beltline plate (M-4305-6)? For all beltline weld j metals, which have a copper chemical composition greater than the limiting I beltline plate, identify the weld metal (heat and lot of flux and wire), its >;j location in the beltline, its copper and nickel chemical composition and j provide its Charpy V-notch test data. 1 a
; 5.4 Component arid Subsystem Design 5.4.1 Reactor Coolant Pump Flywheel Integrity 5.4.1.5 Pump Flywheel Integrity
{ The safety objective of this review is to assure that the integrity of the 'l primary reactor coolant pump flywheel is maintained to prevent failure at - normal operating speeds and speeds that might be reached under accident condi-f tions and thus preclude the generation of missiles. a i ~j The basis for review is outlined in Standard Review Plan (SRP), Section 5.4.1.1 i
; and the Regulatory Guide 1.14, which describes and recommends a method accept-able to the NRC staff in implementing General Design Criterion 4, " Environmental $ and-Missile Design Bases," of Appendix A of 10 CFR Part 50 with regard to mini- . . mizing the potential for failure of flywheels of the reactor coolant pump.
(1) Materials and Fabrication Flywheels are fabricated from ASTM, A-543, Grade B, Class 2 plate. The material is produced by a process that minimizes flaws and improves fracture toughness properties. The materials as well as finished flywheels are subjected to 100 _
]
~ ~
5-15 WNP-3 DSER SEC 5
- s .
__.._._.m__a -- l i 1 l percent volumetric ultrasonic inspection using procedures and acceptance
., standards specified in Section III of the ASME Code.
l The nil-ductility transition temperature (NDTT) of the flywheel material is
, obtained by two drop weight tests (DWT) which exhibit "no break" performance j at -5*F in accordance with ASTM E-208. The Charpy V-notch energy level is at least 50 foot pounds in the WR orientation at 70*F. Hence, the RT f 10*F NOT 'j can be assumed. The above drop weight tests also demonstrate that the NDTT of it the material is no higher than 10*F. ; (2) Design Basis j
j The calculated stresses at the operating speed, due to centrifugal forces and l the interference fit on the shaft, are within the Regulatory Guide limits. The pump runs at 1190 rpm and may operate briefly at overspeed of 110 percent during the loss of outside load. The design speed is 125 percent of the i operating speed. The flywheels are also tested at 125 percent of the maximum a 4 synchronous speed of the motor. The combined stresses at the design overspeed,
; due to interference fit and centrifugal forces, are within the Regulatory Guide j limit.
I j The flywheels can be inspected by removing the cover. Hence, any crack that - f develops can be noticed. The critical crack length at the key-ways, where the stress concentration is high, is about 6 inches at the design overspeed. I 1 (3) Evaluation q We have reviewed the material fabrication, design and inspection aspects of j the pump flywheels for compliance with the Regulatory Guide 1.14. We have
.f
. concluded that the structural integrity of the flywheels is adequate to with-
; stand the forces imposed by overspeed transients without the loss of function, -
] and the integrity will be verified periodically by inspection to assure that the integrity is maintained. 3
~
j 5-16 WNP-3 DSER SEC 5 i 1 ,.
;- ,m ,_ - -
, - ~ - - - - - . - - - - - - - - - --
_ . u _ _._ . :_ _ -u . . . _ _ , _ . . _ ._ __ _._ . ~ . _ u _. . _ u. , um m. .I t i 5.4.2 Steam Generator 5.4.2.2 Steam Generator Tube Inse:vice Inspection
- 1,
,j This section was prepared with the technical assistance of DOE contractors
. from the Idaho National Engineering Laboratory.
2 t
,;l 5.4.2.2.1 Compliance with the Standard Review Plan L ,
.]
j The July 1981 edition of the " Standard Review Plan for the Review of Safety
, Analysis Reports for Nuclear Power Plants" (NUREG-0800) includes Section 5.4.2.2, " Steam Generator Tube Inservice Inspection." The FSAR was reviewed in q accordance with this section of the Standard Review Plan (SRP). The results of 1 1 this review are summarized below.
i. q The SRP. Acceptance Criteria recommend that the Applicant perform examinations
;1 based on Regulatory Guide 1.83 and the applicable Standard Technical Specifica-
] tions. The FSAR Table 1.8-3 states that compliance with Section 5.4.2.2 of l NUREG-0800 is under review and a compliance statement will be provided in a li subsequent amendment.
]
l, 5.4.2.2.2 Evaluation of the Inspection Program -
.g a
j General Design Criterion 32, " Inspection of Reactor Coolant Pressure Boundary," rj Appenc"x A of 10 CFR Part 50 requires, in part, that components which are part
.y of th. reactor coolant pressure boundary be designed to permit periodic examina-k Fh tion und-testing of important areas and features to assess their structural and 1 i leak-tight integrity. The steam generators have been designed to meet the ASME ] Boiler and Pressure Vessel Code requirements for Class 1 and 2 components.
j . Provisions also have been made to permit inservice inspection of the Class 1
] and 2 components, including individual steam generator tubes. The design sj aspects that provide access for examination and the preposed inspection program Il must comply with the requirements of Section XI of the ASME Code with respect to the examination methods to be used, provisions for a baseline examination, 1
17 WNP-3 DSER SEC 5 t
,, .- , . , , ~ . . - , - -.--,....,,-n -, - , - - , . . . . -..- -. _-
= . -_. .
.-- a _ .. - - - - _. - . .:.2.a . w . -. a an f
- j selection and sampling of tubes, inspection intervals, and actions to be taken in the event that defects are identified.
The proposed inspection program must also follow the recommendations of Regula-2 tory Guide 1.83, Revision 1, " Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes," and NUREG-0212, Revision 2, " Standard Technical Specifi-
]!! cations for Combustion Engineering Pressurized Water Reactors."
j The Applicant, in Chapter 16 of the FSAR, has committed to develop the Technical Specifications using the guidance in the latest revision of NUREG-0212 as well as the approved CESSAR-F Technical Specifications. Based on the above, the staff
}4 considers the preservice examination of the steam generators an'open issue subject .I
- j to the Applicant docketing an inspection program that complies with the latest q revision of NUREG-0212.
I 1 5.4.2.2.3 Conclusions i Conformance with Regulatory Guide 1.83, NUREG-0212, and the inspection require-
]
.j ments of Section XI of the ASME Code constitutes an acceptable basis for meeting, in part, the requirements of General Design Criterion 32. 5.4.2.2.4 References - 4 1. NUREG-0800, Standard Review Plans, Section 5.2.4, " Reactor Coolant Boundary 3 Inservice Inspection and Testing," Section 5.4.2.2, " Steam Generator Tube
! Inservice Inspection," July 1981.
y i
; 2. Code of Federal Regulations, Volume 10, Part 50.
- 3. American Society of Mechanical Engineers Boiler and Pressure Vessel Code,
; Section XI, Division 1, 1980 Edition through Winter 1980 Addenda.
- 4. NUREG-0212, Revision 2, " Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors."
l
^
_ 5-18 WNP-3 DSER SEC 5 i L- _ . _ _ _ _ _ ..
_=__ _ ;. .
.;,.w.- . u. :. . - . --- a.a. .w a
- i 3 5. Regulatory Guide 1.83, Revision 1, " Inservice Inspection of Pressurized j
.4 Water Steam Generator Tubes."
i j (Editor's Note: See end of 6.6.5 for continuation of 5.4.2 discussion.)
.1j
-q
-4 5.4.11 Pressurizer Relief Tank (Reactor Drain Tank)
.1 1
j The reactor drain tank is within the scope of CESSAR. Refer to Section 5.4.11 (d s1 of the CESSAR SER for this discussion. Vf-
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. 5-19 WNP-3 DSER SEC 5
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6 ENGINEERED SAFETY FEATURES d 6.1 Materials
~
U 6.1.1 Post-Accident Emergency Cooling Water Chemistry h b Introduction q,D This review is related to providing and maintaining the* proper pH of the con-i y tainment sump water and recirculated containment spray water following a design basis accident to reduce the likelihood of stress corrosion cracking of austeni-j tic stainless steel.
!j .l The applicant will use borated water with a concentration of 4000 - 4400 ppm boron from the refueling water storage tank during the initial injection phase of containment spray. The borated water will be mixed with a 40 percent by <- weight sodium hydroxide solution from the chemical storage ta'nk.
T1 M The resulting solution will have a pH greater than 7, and will drain to the f _ 0] containment sump. Mixing is achieved as the solution is continuously recircu-m ] lated from the sump to the containment spray nozzles during the recirculation J, phase of containment spray. d u Evaluation
.1 di d The post-accident emergency cooling water chemistry has been reviewed in accor-N, dance with Section 6.l.'1 of Standard Review Plant (NUREG-0800, July 1981).
h n We evaluated the pH of the water (mixture of refueling water storage tank and (j sodium hydroxide solution) in the containment sump. We verified by independent
?.2 calculations that sufficient sodium hydroxide is available to raise the contain- )j ment sump water pH above the minimum 7.0 level to reduce the probability of L- stress-corrosion cracking of austenitic stainless steel components. The l]
'l
~
_. 5-1 WNP-3 DSER SEC 6
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. u. . -. w . . a. .. c.~an wc Si> l
.i 1
I
] removal effectiveness of the chemical additive for fission products in contain- <
] j ment is reviewed in Section 6.5.2. We will review the surveillance require-ments in the plant Technical Specifications to verify that sufficient sodium j hydroxide is maintained in the containment spray additive tank.
i' Conclusion
~;
. On the basis of the above evaluation, we conclude that the postaccident emer-gency cooling water chemistry meet's the minimum pH acceptance criterion of Stan-m
}d dard Review Plan Section 6.1.1, the positions of Branch Technical Pcsition MTEB 61, the requirements of General Design Criterion 14 of Appendix A to 10 CFR 50,
?]
and the CESSAR interface requirements, and is therefore acceptable.
)l,,
6.1.2 Organic Materials Li Introduction
- s
, This evaluation is conducted to verify that protective coatings applied inside
;~ containment meet the testing requirements of ANSI N101.2, " Protective Coatings y (Paints) for Light Water Reactor Containment Facilities," American National 1 Standards Institute (1972), and the quality assurance guidelines of RG 1.54
" Quality Assurance Requirements for Protective Coatings Applied to Water Cooled -
Nuclear Power Plants." Compliance with these requirements provides assurance jj that the protective coatings will not fail under design basis accident condi-y tions and generate significant quantities of solid debris that would adversely Q affect the engineered safety features. k: < l Evaluation g
".3 We have reviewed the organic materials in accordance with SRP 6.1.2 i
M (NUREG-0800). In the FSAR, the applicant states that the coating system used q
.' on exposed surfaces inside the containment have been qualified in accordance with ANSI N101.2. The applicant also states that the protective coating system
~
for the containment are applied in accordance with RG 1.54. I j _ 6-2 WNP-3 DSER SEC 6 1 1 l__.-__..__._____ _ _ _ - - . - -
m.c =.= =. . . . a.NuOhw w a.A i . l t j The applicant meets the positions of RG 1.54 and the testing requirements of 4 ANSI'N101.2. These measures demonstrate their suitability to withstand a postu-3 lated design-basis accident environment. i
't y The consequences of solid debris that can potentially be formed from unquali-fied paints are reviewed in Section 6.2.2. The control of combustible gases
- f that can potentially be generated from the organic materials and from qualified 6 and unqualified paints is reviewed in Section 6.2.5.
g b 'g.ij Conclusions N . il
.] On the basis of the above evaluation, we conclude that the orga'nic materials
,] meet the testing requirements of ANSI N101.2 and the positions of RG 1.54 and ] are, therefore, acceptable.
'j ;j 6.2 Containment Systems '
i { The containment systems for the WNP, Unit 3 include dual containment structures, j containment heat removal systems, a shield building ventilation system, a con-
]
tainment isolation system and a containment combustible gas control system.
.j The primary and secondary containment structures and their associated systems f.l all function to prevent or control the release of radioactive fission products -
3 g which might be released following a postulated loss-of-coolant accident (LOCA), secondary system pipe rupture, or any other accident releasing radioactive mate-rial into the containment atmosphere. k:? 9 [. The staff has reviewed the applicant's design, design bases, and safety analy- l ses for the containment and the containment systems provided in the FSAR. The acceptance criteria used as the basis for our evaluation are contained in Sec- i
} . tions 6.2.1, " Containment Functional Design," 6.2.2, " Containment Heat Removal ] Systems," 6.2.3, " Secondary Containment Functional Design," 6.2.4, " Containment-My Isolation System," 6.2.5, " Combustible Gas Control In Containment," and 6.2.6, J " Containment Leakage Testing," of the Standard Review Plan (SRP), NUREG-0800, l l dated July 1981. These acceptance criteria include the applicable General De-a i sign Criteria (Appendix A of 10 CFR Part 50), Regulatory Guides, Branch Techni-
- cal Positions, and industry codes and standards as specified in the above cited
~
l sections of the SRP. The results of the staff review are discussed below.
~
- 6-3 WNP-3 DSER SEC 6 d
4
'_ __ _, .__. - _ _ _ ,,y __ ._ s,. .. , , , _ . .
_ - a__ - - - _ _ _ _ _ = &. .m a. a_ i m . . u _.-. .._,__ _..a 5 l 6.2.1 Containment Functional Design j a 6.2.1.1 Containment Structure h] The reactor containment is a free-standing steel structure with a net free volume 1
] of 3,218,000 ft .8 The containment structure houses the nuclear steam supply d system including the reactor, reactor coolant pumps, pressurizer and steam gene-N
- ,3 rators, as well as certain components of the plant's engineered safety feature tj systems. The structure is designed for an internal pressure of 44 psig and a a
'j .
temperature of 367*F. The reactor containment is completely enclosed by a shield
] building with an annulus region between the structures.
1 "1
.] Maximum Pressure and Temperature Analysis b
3 The applicaat has analyzed the containment pressure and temperature responses for postulated reactor coolant system and secondary system pipe ruptures to establish the containment design bases and the conditions for environmental j qualification of safety related equipment in the containment. The most limit-j ing single active failure, from the standpoint of predicting the highest con-tainment pressure and temperature, was assumed in the containment analyses. 1] j The applicant's postulated spectrum of breaks in the reactor coolant system - (i.e., loss-of-coolant accidents), as described in FSAR Table 6.2.1-1, include
$ double ended slot breaks in the hot leg and in the cold leg at the reactor cool-h, ant pump suction and discharge. For the cold leg breaks both minimum and maxi-F mum emergency core cooling system (ECCS) flow cases were considered. The reactor w
A coolant system pipe break spectrum is based on that prescribed in CESSAR Sec-y tion 6.2.1.3, which the staff found acceptable in the CESSAR SER of November, 1981. Q The design basis LOCA at WNP-3 was determined to be a hot leg slot break of f 19.2 ft 2; the failure of one train of the containment spray system was the worst d single active failure. d 3 j The spectrum of breaks postulated for the secondary system includes slot breaks
+
11 in the main steam line at five different power levels, from 0% to 102% of full power. Slot breaks were determined to be more severe than guillotine breaks. The maximum break area that allowed a pure steam blowdown was chosen as the L l L
'~ -
6-4 WNP-3 DSER SEC 6 L l>
__ a_ .u . . 2_ _._ _. a - 2 t..=_: .e a _.. _. _ m . ._ _ .._ . ._:._.-
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i f most limiting break for each power level. Main feedwater line breaks were not
.j included because they result in two phase blowdowns and thus are not as severe o
j as main steam breaks. Again, the spectrum of main steam line breaks (MSLB) analyzed for WNP-3 is the same as that prescribed in CESSAR Section 6.2.1.4,
]
.I which the staff found acceptable in the CESSAR SER of November, 1981. The MSLB 1 resulting in the highest containment pressure was found to be the fou--square-
.i j) foot slot break at 0". power with a failure of one spray train. The peak contain-
'j ment temperature of 367'F was calculated to occur following a 8.78 ft 8 slot break in the main steam line at 102*. power, with a failure of one spray train.
, The applicant has performed the containment pressure and temperature analyses using the CONTEMPT-LT computer mode. Initial conditions and in~put data, includ-
; ing passive and active heat sink parameters, were conservatively chosen to pra-j duce the highest containment pressure. The highest containment pressure was i calculated to be 39.4 psig (versus the containment design pressure of 44 psig),
l which occurred for the design basis LOCA identified above. For the long term containment pressure response, a double ended slot break at l the pump suction (with minimum safety injection) was analyzed. The analysis showed the containment pressure would be reduced to approximately 17 psig; 1.e., less thin 50". of the peak calculated value (38.2 psig), in 24 hours, in accor-t dance with staff guidelines. The design basis main steam line break was analyzed - to establish the peak containment temperature to be used in developing the tem-j perature profiles for environmentally qualifying safety-related equipment lo-T cated in containment. The peak temperature was calculated to be 367 F. e We have 'eviewed the applicaat's selection of initial conditions, input para-y] meters, and analytical assumptions and find them to be acceptable and in con-
] formance with staff guidelines. Staff confirmatory analyses were performed for C.j 7) 1 . the design basis reactor coolant system break and the design basis main steam j line break using the CONTEMPT-LT/28 computer code. The results of the confir-j matory analysis are in close agreement with the applicant's results, and confirm i their acceptability.
Based on our review of the applicant's containment functional analysis, as discussed above we conclude that the applicant has satisfactorily demonstrated 6-5 WNP-3 DSER SEC 6 i.!
sw.. - =- ' RM . u. 2 :.:. _ x.-. --. - . . = .- a a a ..= -a j t i 1 1 l the adequacy o'f the containment functional design, and has appropriately deter-
]
i mined the containment temperatures and pressures to which safety-related equip-s ment in containment must be environmentally qualified. j Protection Against Damage From External Pressure 1 1 j To demonstrate the adequacy of the containment against the maximum expected i external pressure, the applicant has analyzed the consequences of a postulated
) inadvertent actuation of the containment heat removal system during normal plant operation. The operation of two spray trains and two containment fan coolers was assumed to occur during normal operation. One of the two vacuum breakers j was assumed failed. The applicant calculated a maximum pressurs differential of 0.58 pounds per square inch, which is less than the containment vessel design 3
external (differential) pressure of 0.7 psid. The initial conditions and assump-q tions used in the analysis were chosen to maximize the differential pressure
; load on the containment. Based on our reveiw of the applicant's analysis we i find the containment design has sufficient margin to accommodate the maximum postulated external load. '1.
d 6 ' 1.2 Suocompartment Analysis Subcompartment analyses were performed to determine the acceptability of the -
-l design differential pressure loac'i gs on containment internal structures from high-energy line rupture accidents. The applicant's subcompartment analyses included the reactor cavity, and pressurizer and steam generator compartments, where high energy line ruptures were postulated to occur. A spectrum of pipe breaks was analyzed by the applicant to determine the limiting break that re-suited in peak loads on each of the subcompartment walls.
-] The reactor coolant systeri (RCS) break 1;, considered in the WNP-3 subcompartment analysis (except for one), and the mass and energy release data, were obtained
']j from CESSAR Section 6.2.1.2. We find this approach acceptable based on the
~
i staff CESSAR SER. The one break that differs from th'se o in CESSAR is the guil-i lotine break of the discharge leg in the reactor coolant system; a 350-in 2 break size is specified in CESSAR, where as the break size has been reduced to 100-in2 in the WNP-3 safety analysis report. The inherent stiffness of the system, .
~
2
~'
j 6-6 WNP-3 DSER SEC 6 fj if
- , 'P
- i 2 . . . m a_ . .__ _m .. a. _ . ;_ ____ . .... d_ u .
- A q
together with pipe whip restraints, limits the postulated pipe rupture to this ^ ( break area. The 100-in2 break mass and energy data were calculated based on I l the methodology described in CESSAR. The staff has reviewed the applicant's t analysis and find the mass and energy release data acceptable, contingent upon _ the acceptability of the limiting pipe break size (see SER Chapter 3.0). f1 {lu The applicant used in the RELAP-4 M006 computer program to analyze the pressure i transients in the reactor cavity, and the steam generator and pressurizer com-partments. Separate discussions for each subcompartment are presented below: Reactor Cavity Analysis The reactor activity is a heavily reinforced concrete structure that performs
- the dual function of providing reactor vessel support and radiation shielding.
The reactor cavity is Essentially a cylindrical annular air space between the Y reactor vessel and the primary shield wall. The cavity is bounded at the top and bottom by a neutron shield. The major vent paths for the reactor cavity { are the six piping penetrations (two hot legs and four cold legs) through the primary shield wall to the steam generator compartments. j The applicant postulated 100 in8 , discharge and hot leg guillotine breaks; the
] design basis break was found to occur in the discharge leg. The peak differen- -
fj tial pressure across the reactor cavity wall was calculated by the applicant to ^j . be 29.6 psid, versus a design value of 211.4 psid.
- 1
<- The applicant chose to use the reactor cavity nodalization sensitivity study of
,' Carolina Power & Light on their Shearon Harris Plant (Docket Nos. 50-400, -401, e -402,-403) as a basis for verification of the WNP-3 reactor cavity nodali-3 zation scheme. It was concluded in the Shearon Harris nodalization study that
.dq . subcompartment nodalization models were determined principally by physical flow q restrictions within each compartment. These flow restrictions consider the ,y presence of steel and concrete obstructions, doorways, vent shafts, grating, l reactor coolant pumps, piping, the steam generator, the pressurizer, the reactor 4 vessel, and the reactor cavity missile and neutron shields. The subcompartment models in WNP-3 take into account all physical flow obstructions. All assump-i tions utilized by the applicant in the reactor cavity subcompartment analysis j ' 6-7 WNP-3 DSER SEC 6
~I
- L. , _ .. . _ _ _ __
vda m m._. . . _._ . _ _ _ . m _ a . __ _ n .. m .__ a _.... _ _ _ i
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have been reviewed and found to be appropriate. In addition, the staff perfor-med a confirmatory analysis using the COMPARE-MOD 1A computer code, and the 2 same nodal model as the applicant. Although the staff's analysis predicted a j higher peak differential pressure (34.6 psid), the reactor cavity design is adequate for the differential pressure loads from the worst postulated pipe f rupture within the reactor cavity. t d The applicant has not provided in the FSAR an analysis of the forces and moments s 1 on the reactor vessel due to the differential pressure across the vessel caused by a reactor coolant system break within the reactor cavity. The applicant has indicated that the methodology presented in CESSAR Chapter 3 was used for the 2 force / moment analysis. However, it is not clear that the generic analysis in
,l the CESSAR is applicable to WNP-3. This matter will remain an open item until further justification or analysis is provided by the applicant.
.j Steam Generator Subcompartment (SGS) Analysis 4
1 1 The walls of the steam generator comp rtment are constructed of reinforced con-4 4 crete. The applicant considered a spectrum of primary coolant system pipe breaks j q for resultant load impact on the SGC walls. A steam line break was not postu-y lated because the routing of this line does not pass through the steam generator d.
.j.
compartment. The staff has reviewed the spectrum of breaks postulated by the - j applicant and finds it acceptable. The SGC has been nodalized as 20 volumes.
'j The staff has accepted this nodalization based on the above cited Shearon Harris i
subcompartment nodalization sensitivity study. fj For the spectrum of breaks analyzed for the steam generator compartment, the applicant's analysis shows the design basis break is the 592-in 2 guillotine j break in the suction leg. The analysis results a peak differential pressure of
.l 6
. 13.1 psid on the steam generator subcompartment walls versus the design value j of 25.9 psid. Our confirmatory calculation, using the COMPARE-MOD 1A computer code, confirms the acceptability of the applicant's results.
]
q Based on our review of the applicant's analysis and the results of our confirm-
; atory analysis, we find the design basis of the steam generator subcompartment j walls is adequate. .
J . 6-8 WNP-3 DSER SEC 6 I f_
n w.c * -,-a-- - .-. - x ._. - := . . - . ~
..-..a--=:....~.2.
.J
) -
f Pressurizer Subcompartment Analysis The pipe breaks considered for the pressurizer subcompartment include surge line, spray line, and safety relief valve line breaks. The most limiting case
,, is the double-ended guillotine break in the surge line. Based on our review of j these breaks, we find the applicant's choice acceptable for the pressurizer
.3 compartment analysis.
1 ,j The pressurizer compartment nodalization consisted of 13 volumes. The applicant calculated a peak differential pressure of 24.3 psid across the walls of the
,j subcompartment, compared to the design value of 84 psig. We have performed a 'j confirmatory analysis using the COMPARE-MOD 1A computer code an'd our results j confirm the acceptability of the applicant's results'. We therefore find the i, applicant's pressurizer compartment analysis to be acceptable.
9 6.2.1.3 Mass and Energy Release Analyses for Postulated Loss-Of-Coolant
'i Accidents
.i
; For the containment functional analysis, the applicant obtained mass and energy
,;. li release data for postulated loss-of-coolant accidents from CESSAR Section 6.2.1.3.
] We find this approach acceptable based on the staff findings in the CESSAR SER dated November,1981.
" J:1 -
]
6.2.1.4 Mass and Energy Release Analysis for Postulated Secondary System d Pipe Breaks Inside Containment
;I
]j The applicant used the mass and energy release data provided in CESSAR, Section 6.2.1.4 for main steam line breaks. The CESSAR MSLB mass and energy release data may be used if certain interface requirements (for example, main steam and
] . feedwater isolation valve closure times and maximum steam line and feedwater j line volumes) are met by the applicant. have confirmed that the WNP-3 MSLB
,i analysis satisfies these interface requirements and, therefore, conclude that use of the CESSAR MSLB mass and energy release data is acceptable.
; However, the applicant has not addressed the concerns of IE Bulletin 80-04, i regarding the impact of runout flow from the feedwater system. We have requested
!j lj _.=. 6-9 WNP-3 DSER SEC 6 l li ) l :1,
~ w .w- :a: w.- . :== =_ .= u =.-..: . w -. 2- . c - . .c a : u.axw 1, .
f b
; additional information in this regard; this matter will remain an open item l until we can review the applicant's response. I 1
i j 6.2.1.5 Minimum Containment Pressure Analysis for Performance Capability ,j Studies on the Emergency Core Cooling System (ECCS) Ti . J. j The applicant has not adequately demonstrated the applicability of the CESSAR $q minimum containment pressure analysis to the WNP-3 design. Therefore, a CESSAR interface requirement has not been met. The applicant has indicated that a
- j WNP-3 plant-specific analysis will be performed. This matter will remain an
,g open item until-we have had the opportunity to review the applicant's analysis. - t i 6.2.1.6 Summary and Conclusions d We have evaluated the WNP-3 containment functional capability with respect to the requirements of General Design criteria 16 and 50'of Appendix A to 10 CFR Part 50. We have found the applicant'.s analyses of the dynamic pressure loads that act on the containment vessel and subcompartment structures from postulated
- , pipe breaks acceptable, with the following qualifications:
- 1. ;.
'] -(1) The forces and moments acting on the reactor vessel as a result of the
- design basis reactor coolant system break within the reactor cavity are -
needed to complete our review. The guidelines of SRP 6.2.1.2 and Section
- i 3.2 of NUREG-0609 should be followed.
'4 .1 J
fj (2) A minimum containment pressure analysis that is applicable to the WNP-3 d plant is needed to complete our review. r.3 a (3) The applicant's response to IE Bu11etin-80-04 is needed to complete our j . review. g _. j; .- 6.2.2 Containment Heat Removal Systems il s. The function of the containment heat removal systems is to remove heat from the I
.}
y containment atmosphere to limit, reduce and maintain at acceptably low levels , 'a t both the containment pressure and temperature following a LOCA or secondary _ j d system pipe rupture. -The containment spray system (CSS) is the only active W
~
6-10 WNP-3 DSER SEC 6 ' ~
= .~.- ----= - w. ~ . : . . - . = : . .u - .- :,-
i 3 .
! containment heat removal system at WNP-3 and also serves as a fission product .j removal and control system (see SER Section 6.5).
The CSS consists of two redundant and independent 100% - capacity trains, each containing a containment spray pump, a shutdown cooling heat exchanger, a spray q header, and associated valves, piping and instruments. Each of the two contain- .j ment spray pumps has a design flow rate of 5000 gpm of water at a head of
.j 645 feet during the injection mode, and a flow rate of 6000 gpm at a head of 600 feet during the recirculation mode. The CSS is automatically started by j the containment spray actuation signal (CSAS) which is initiated by high-high j containment pressure. The CSAS may also be initiated manually in the control
'l room. Upon receipt of a CSAS the containment spray pumps are started, the sys-
.i tem isolation valves are opened, and the borated water from the refueling water I
storage tank (RWST) flows into the containment. Full spray flow from the nozzles
.! is established in about 50 seconds after receipt of the CSAS, assuming loss of offsite power; this satisfies the CESSAR interface requirement of establishing
- full spray flow in less than 58 seconds. When the water level in the RWST reaches a specified low level, a recirculation actuation signal (RAS) automatically l' realigns containment spray pump suction from the RWST to the safety injection system (SIS) recirculation sump. The operator must then verify that the appro-priate amount of water has been discharged into the containment, the flow path C from the SIS recirculation sump to the containment spray pumps is open, and the -
minimum flow lines are isolated. Finally, the operator will manually close the RWST isolation valves from the control room to complete the transition from the injection mode to the recirculation mode. g
.i
-1 The CSS is designed to Quality Group B and Seismic Category I requirements. 1 The applicant has also provided a failure mode and effects analysis (FMEA) and other information demonstrating the ability of the CSS to function following i postulated single active failures. I 7 i We have reviewed the applicant's net positive suction head (NPSH) calculations and find that sufficient NPSH will be available for the CSS pumps during both the injection and the recirculation modes of operation. The applicant's evalua-tion of the available NPSH is consistent with the guidelines of Regulatory i Guide 1.1, Rev. O, " Net Positive Suction Head for Emergency Core Cooling and .
~
Containment Heat Removal System Pump," and is acceptable. i 6-11 WNP-3 DSER SEC 6 f 1 1 1
- ~ ~ _m we .-w -
y ry,, _o~ a m y- a w gr ,= y*.-,-qe-e.- y.,-,m pyrta-- ,
7 _ . 2 m . ._. u ___ s _ _ m - ,3 ey e .
- Ji ij Regulatory Guide 1.82, " Sumps for Emergency Core' Cooling and Containment Spray !
.t
.j . System," provides guidelines to be met by reactor building sumps that are de-14 signed to be sources of water for the ECCS and the CSS following a LOCA. The j guidelines address redundancy, location, and arrangement of sumps as well as ,
j s provisions to screen out debris and ensure adequate pump performance. The ap- %: plicant's sump design conforms to Regulatory Guide 1.82, revision 0, and we
- ] find the design acceptable.
5
- . ) Periodic testing and inspection of the system actiw components, i.e., pumps, ,
- d valves, etc., will be performed in accordance with the in-service inspection 1
- .;j requirements of the ASME Code Section XI to assure the operability and perfor-
[j mance of the system. d , M e Based on our review, we conclude that the containment spray system satisfies
.{
,.] the requirements of General Design Criteria 38, 39, 40 and the provisions of
- ; Regulatory Guides 1.1 and 1.82, Rev. O, and, therefore, is acceptable.
'l
* ^t 6.2.3 Secondary Containment Functional Design vt P j; The secondary containment encompasses the annual space between the concrete
..i d shield building and the steel primary containment vessel, the ECCS and mechant- 'q j cal penetration areas, and the fuel handling building. The shield building is -- a seismic Category I structure which provides biological shielding, controlled ,j release of airborne radioactive materials following an accident, and environ- . /t mental protection. The shield building ventilation. system (S8VS) is an engi-p neered safety feature designed to maintain a' negative pressure in the annulus [q following a LOCA and to filter the airborne fission products which'may leak Il from the primary containment to the annulus following a LOCA. (The submittals do not fully describe the fission product removal mechanisms as required in a ':d, ! discussion of the modeling of design basis accidents.) During normal operation, y the annular subatmospheric pressure of minus 10 inches (water gage) is main- ] tained by. the annulus vacuum maintenance systems (AVMS). The regions compris- ,1
] ing the secondary containment,.other than the annulus, are designed to maintain a subatmospheric pressure by the ECCS area / fuel handling building (FHB) filtered
[.f1 exhaust system following a design basis accident. ,I - <3
~~ '
6-12 WNP-3 DSER SEC 6 m.. 1 1
( = : x w = : - : - . - .z _ a = = - -. a .. .2. w = ~ ..- ~ e . = . - . . ~ - . . . . . . h . s ,] The SBVS consists of two independent 100% capacity trains; each train is actu-j ated by a separate channel of containment isolation actuation signal (CIAS) and
.; all redundant active components are powered from separate ESF bases. Each train includes one full capacity exhaust fan, a filter train (including a demister, c'. electric heating coil, prefilters, HEPA filter, charcoal absorbers, and after- ~. I HEPA filter), ductwork, valves, and instrumentation and controls. A failure 3 modes and effects analysis was performed by the applicant to show the system -f meets the single failure criterion. All components and ductwork are designed to meet seismic Category I requirements.
': .?
'i .
The applicant has analyzed the performance of the SBVS using the WATEMPT com-puter code. The results of the analysis indicate a negative pressure relative i to the outside atmosphere can be maintained in the annulus throughout the tran- -] 1 sient following a LOCA, thus ensuring no primary containment out-leakage escapes unfiltered directly through the shield building. However, Appendix 6.2A to the
; WNP-3 FSAR does not have sufficient information concerning the WATEMPT. codes to
{ permit a complete evaluation. We will conclude on the applicant's method of i analysis of the post-LOCA SBVS performance after we have had an opportunity to f review the additional information we have requested from the applicant.
- i Preoperational testing of the SBVS will verify the system performance capabil-
)
ity to achieve and maintain a negative annulus pressure. Periodic testing and - l inspection of the SBVS will be included in the plant Technical Specifications. 1 i The applicant has also identified systems for which through-line leakage fol-
, lowing a LOCA could result in containment bypass leakage. The applicant has .f committed to perform local leak rate tests on the potential bypass leak paths j in accordance with the requirements of Appendix J to 10 CFR Part 50. The total j ,
potential bypass leakage rate will be limited to 22 percent of the design leak ] *
. rate of the containment (0.2 weight percent of the internal net free volume per day at a pressure of 39.4 psig), or 0.044 w/o per day. $i With the exception of the need for additional information concerning the post-j LOCA shield building annulus pressure transient analysis the staff concludes ,
,j that the secondary containment systems meet the requirements of GDC 41, 42, and .! 43 of Appendix A, 10 CFR Part 50, and therefore, are acceptable. I i -i .
- ]
6-13 WNP-3 DSER SEC 6 i i> i
( -
=w. a.w.u . :. i ~.w i. - - , . . . - .
- .d. 1 i
i - i 6.2.4 Containment Isolation System The function of the containment isolation system (CIS) is to allow the normal j or emergency passage of fluids through the containment boundary while preserv-i j ing the ability of the boundary to minimize the release of fission products ~ Tj that may result from a postulated accident. This section, therefore, is con-1 g
,2 cerned with the isolation of fluid systems which penetrate the containment boun-
- dary, including the design and testing requirements for isolation barriers and l
actuators. The isolation barriers include valves, closed piping systems, and
; blind flanges. In general, for each penetration at least two barriers are re- ; quired between the containment atmosphere or the reactor coolant system and the j outside atmosphere so that failure of a single barrier does not' prevent iso-
[ lation. l
.j All non-essential systems are those systems either automatically isolated by l one of the actuation signals (CIAS, SIAS, MSIS) discussed below or normally .i locked closed. The containment isolation actuation signal (CIAS) and safety j injecticn actuation signal (SIAS) are initiated by high containment pressure or low pressurizer pressure signals. The main steam isolation signal (MSIS) is initiated by high containmeht pressure low steam generator pressure, or high steam generator water level. These signals can also be initiated manually from the control room. For the containment purge and vent system, the isolation -
valves are also isolated by the containment high radiation signal in addition
- to the CIAS. We, there, conclude that the containment isolation signals provide acceptable diversity.
s We have reviewed the applicant's designation of essential systems. The essen-tial systems do not require automatic isolation, and if the automatic isolation exists,-the systems are equipped with override features for remote manual opera-1 tion from the control room. Those systems or portions of systems classified by 1 q the applicant as essential include the high pressure safety injection system, j containment spray system, auxiliary feedwater system, main steam and feedwater j isolation system, chemical volume and control system (CVCS) charging and let-j down lines, CVCS reactor coolant pump (RCP) seal injection lines', RCP component cooling water lines, instrument and control air system, hydrogen purge system,
.I and RCP seal bleed off lines. These systems are considered important to
' 1 6-14 WNP-3 DSER SEC 6 i]
l . __ 1
.g . L a .= -== u- r. - . . . :- - - - . . w . , = .a . -w :- l
- I l
j post-accident safe shutdown and valuable in accident mitigation, particularly ' in the event' of a small break LOCA or a secondary system rupture, and, therefore, j their classification as essential is acceptable. Provisions are made to allow
] the operator in the control room to detect leakage from remote manually con-trolled systems. These provisions include instrumentation to measure radiation l
j levels, flow rates, pressure and sump water levels in the safety equipment area and the penetration area. f We have reviewed the closure times for the containment isolation valves. Most valves close in 10 seconds or less. In particular, the containment purge and I vent systems are designed to close in 5 seconds (except for isolation valve
.j 2PV-8019 on penetration 80, which will be discussed below). We' conclude that ] the containment isolation valve closure times, with the exception of valve 2PV-B019, are acceptable.
i j We have reviewed the containment purge and vent systems against the provisions i of Branch Technical Position (BTP) CSB 6-4, " Containment Purging During Normal
; Plant Operations." The 48-inch containment purge valves will be sealed closed ; during normal operation and be verified to remain closed at least every 31 days; j requirements for this will be included in the plant Technical Specifications. ; Furthermore, as a result of our study of valve leakage due to seal deteriora-j tion, leakage integrity tests of the purge and vent system isolation valves -
M must be conducted periodically; i.e., over and above the leak testing require-1
! ments of Appendix J. This requirement, together with the test frequency will be included in the plant Technical Specifications. W conclude that the 48-inch containment purge system design satisfies the provis ons of BTP CSB 6-4 and j that operation of the system as proposed; i.e., only during shutdown and refue-i ling, is acceptable.
x a ] . The containment vent system, consisting of two 8-inch containment penetrations _'{ (P-80 and P-81), is designed to close following receipt of a CIAS or high radi -
- j ation signal. The containment vent exhaust penetration, P-81, has two isolation 1 valves which are designed to close in less than 5 seconds and fail closed on loss of operating power; we have found this penetration design acceptable.
However, the containment vent make-up penetration (P-80) is equippe,d with a motor operated isolation valve (2PV-B019) and a check valve (2PV-V021). The _ i
+
j 6-15 WNP-3 DSER SEC 6 E_f m_.- r , ,,v. . ,- ~ -. . _ - _ _ --. -- - - - _ - - - - - - - - - - - - - - -
m u- .
. .__ m m .i i! -
q motor-operated valve has a closure time of 10 seconds and is designed to fail
.i "as is" (FAI). The FAI design is not consistent with the statement in FSAR
. Section 6.2.4.2 that the isolation valves on both the containment vent and purge
~j systems are designed to fail close. Also, the 10-second valve closure time j differs from the 5-second assumption used in the radiological dose analysis
<; (FSAR Section 9.4.6.6.6). The check valve (2PV-V021) in the containment vent a
y make up line is not an appropriate type of containment isolation valve for use i in a line which directly connects the containment atmosphere to the outside [i environment. It is the staff's position that the containment isolation valves ut in line P-80 should be automatic, power operated valves having less than 5 1
} second closure times, and should fail closed on loss of operating power. This
] matter will remain an open item pending receipt of additional t'nformation re-
] garding the applicant's plans for complying with the staff position.
- i
. ate that the applicant has not provided an analysis of the reduction in the containment pressure resulting from the partial loss of containment atmosphere
] through purge and vent system isolation valves,_which may be open at the onset
] of a LOCA, and the consequent impact on the minimum ECCS backpressure deter-
]
~
mination. This analysis is called for in BTP CSB 6-4. Therefore, this matter , will remain an open item pending receipt of an appropriate analysis from the j applicant. ij '-
] Our review has confirmed that the WNP-3 containment isolation system meets the
! explicit requirements of GDC 54, 55, 56, and 57, except as discussed below.
,1 i
t The isolation provisions for the chemical and volume control system (CVCS) charg-tj ing line (penetration 41) and the CVCS reactor coolant pump (RCP) seal _ injection 1 line (penetration 93) conform to GDC 55 except that the power-operated isolation 4 valves outside the containment are normally open and are closed by remote-d . manual control from the control room. As discussed above, the CVCS charging i line and the CVCS reactor coolant pump seal injection line are essential lines - that have an impact on plant safety. The isolation valves will also be subject to-Type C leak testing. Based on the guidance provided in SRP 6.2.4, we find
'F -the use of remote-manual instead of automatic isolation valves acceptable. We, require, however, that Class 1E emergency power be provided to valves 2CH-VQ040 J
and 2CH-VQ005 in these lines; because of the above this is an open issue. _ I j Z. 6-16 WNP-3 DSER SEC 6
, _ _ _ _ _ , - ~ - .. . -_ _ . .,._ _ . . _ _ _ - _ . _ . . . _ _ . _ ._ __..___
u -
,m - -
nm y: = = - -w A =. _ - t I 4 l The safety injection system (SIS) recirculation sump discharge lines (penetra-tions 23 and 24) have only one isolation valve in each line, outside contain-ment. If ' isolation valves were provided inside containment, they would be sub-j merged following an accident. Since these. lines have an important safety func-
'I tion, system reliability is greater with only one isolationi /alve in each line.
'j Also, the SIS is a closed engineered safety feature system outside containment I
.) whose integrity is appropriately maintained throughout plant life. Based on
- Tj the provisions of-SRP 6.2.4, we find the single isolation valve design in the dl , SIS recirculation. sump, discharge lipes to be acceptable. We will, however, require the applicant to discuss the adequacy.of the criteria used in the de- 'i sign of the piping between the containment and the isolation valve, and the
'l 1 valve itself, and the leakage control provisions on the penetration er in the
-1 q penetration area. ,
; , The reactor building vacuum relief isolation valves (penetrations 65 and 75)
] are normally closed and would'only open in response to a high vacuum signal in
- j the reactor building. The inboard check valves ensure that flow would always beint'olhecontainment. Basedonthesystemdesignandoperatingrequirements j.
we find the containment isolation provisions for the reactor building vacuum relief lines acceptable, p We have reviewed information provided by the applicant to demonstrate complf- - [.] ance with the provisions of NUREG-0737 Item II.E.4.2, " Containment Isolation Dependabili ty. As previously described, the applicant has complied with the N] y , provisions regarding diversity in parameters sensed for initiation of contain- .. ment isolation, identification of essential and non-essential systems, automatic isolation of nones'sential systems, and closure of containment purge and vent
. isolation valves on a high radiation signal. In addition, the FSAR states that
} all power-operated isolation valves will remain in thetc accident position after h an accidsnt signal clears \u)1til_ deliberate operator action ks taken to reopen i ,
j f the valves. The containment setpoint pressure that initiates containment ' iso-il lation'should be reduced to the minimum value ceapatible with' normal operating j ceditions. The containment setpoint pressure and the justification for it d should be provided by the applicant; this information will be reviewed by the 1 . I staff in conjunction with the d6yelopment of the plant technical specifications. Finally, the applicant has conitted to keep the 48-inch containment purge valves ~ 9
~~
Lj m g *j m 6-17 WNP-3 DSER SEC 6
, , - , - . , .,e.y- , ,,.-r yy+ < ,..,ye--e.-- -,~-,ye ---.w,,
= . .m. a .. . . - ._ : - - _=- ..
~
G- d I
-l 1
j closed during the operational conditions of power operation, startup, hot stand-j by, and hot shutdown and verify that the valves are closed at least~ every 31 days. This is acceptable, except that the purge valves shoulc be sealed closed
'i 1 (either electrically or mechanically) in accordance with SRP 6.2.4. Except for lj ,
the two issues identified above, namely, justification for the containment iso-lation setpoint pressure and a commitment to seal close the purge valves, we
} conclude that the applicant has complied with the provisions of NUREG-0737, Item II.E.4.2.
-1 l 1
.] The containment isolation system meets the provisions of Regulatory Guide 1.26,
-:j " Quality Group Classifications and Standards for Water , Steam ,' and Radioactive-d Waste-Containing Components of Nuclear Power Plants," 1.29, "Se'ismic Design Classification," and 1.141, " Containment Isolation Provisions for Fluid Systems."
]
j The containment isolation valves are designed in accordance wtih ASME i, Section III Class II requirements and Quality Class I. We conclude that the
.; WNP-3 containment isolation system meets the requirements of GDC 54, 55, and 57, and NUREG-0737 Item II.E.4.2, and conforms to SRP 6.4.2 and CSB BTP 6-4, with the exception of the six issues summarized below:
i
-1 d 1. The applicant should upgrade the isolation provisions for the containment n'
vent make up line (P-80); it is the NRC position that redundant power- [ operated, automatic isolation valves should be provided, which close in - 3
,! less than 5 seconds and fail closed upon loss of power to the valve opera- 'j tor; D
st
- 2. The applicant should provide an analysis of the effect purge system opera-
"}
.i tion at the time of a LOCA on the minimum containment pressure analysis for ECCS evaluation.
. 3. It is the NRC position that the applicant provide Class IE emergency power j to valves 2CH-VQ040 and 2CH-VQ005 in the reactor coolant pump seal injec ~
, [j tion and chemical and volume control paths; gj 1 4 4. The applicant should confirm the design adequacy of the piping between the containment and the isolation valve, and the valve, in the SIS recircu-
; lation sump discharge lines (penetrations 23 and 24), and the affiliated j .
leakage control provisions.
'I 6-18 hNP-3 DSER SEC 6 i '! 4
-.a.-.-.--- -~ a = a a . _ . . = - - ..c.-. L i ,
J . . < 1 l 't ji 5. The applicant should justify that the containment isolation setpofnt pres-sure is the minimum value compatible with normal operating conditions.
. ' ,6. ,The applicant should commit to seal close the 48-inch purge valves do,ing j ' operating modes requiring containment integrity, i* ,
l j 6.2.5 Combustible Gas Control System * '
$a k Following a LOCA, hydrogen may accumulate within the containment as a result .l of: 1) hydrogen dissolved in reactor coolant system; 2) metal-water reaction -) between the zirconium fuel cladding and the reactor coolant; 3) corrosion of metals by emergency core coolant and containment spray solution'; s and 4) radio- I c.j lytic decomposition of the post-accident emergency cooling water. The applicant has provided a combustible gas control system (CGCS) to monitor ar.d control the ]
j hydrogen concentration in containment following a LOCA. The CGCS includes the j containment hydrogen analyzers, the containment., hydrogen recombiners, and the, c s ' yj containment hydrogen purge system. - B n , The hydrogen analyzer system consists of two redundant subsystems, each of which j can take samples from six locations within containment and one location in the 1 shield building annulus. The hydrogen recombiner system consists of two sta-
; stionary 100% capacity thermal (electrical) recombiners located within the con- -
1 1 tainment. Both the hydrogen analyzer system and the hydrogen recombiner system fj are designed to Safety Class 2 and Seismic Category I' standards, and are powered a j from Class IE power sources. The recombiner will oe. started manually from the l control room by the operator upon indication of a hydrogen concentration of a
- i greater than 3.0 volume percent. i j
, Each of the two Westinghouse electric, hydrogen recombiners is capable of proces-d sing 100 scfm of containment atmosphere for po.st-accident hydrogen control. The ],~ staff has reviewed tests that were conducted for a full-scale prototype and;a ~
p.roduction recombiner. The tests consisted of proof-of prin'ciple testing, test- ;
; ing on a prototype recombiner, environmental qualification testing, and func- I 'j tional tests for a production recombiner'. (These tests are described in WCAP-7820 i and its supplements.) The results of these tests demonstrate that the recombiner I
r is capable of controlling the hydrogen in a post-LOCA containment env'iionment. L _ 3 ! [ l: 6-19 h WNP-3 DSER SEC 6 f il . . t -
_~ . _ . _ - a_s.a ;
.:_.:a : -
.,a n.
} ,
t
. .A purge system has been provided, in addition to the hydrogen recombiner system, j -
in accordance with Section 50.44 of 10 CFR Part 50. The purge system consists
'of two 100 percent capacity exhaust trains and a single nakeup train.
f]f The applicant has analyzed the production and accumulation of hydrogen within the containment from the sources discussed above. SRP 6.2.5 recommends that
% the analysis of hydrogen production should be based on the parameters listed in M} Table 1 of Regulatory Guide 1.7 for the purpose of establishing the design N. basis for combustible gas control systems. The applicant has been requested to s
fj confirm that their analysis is in conformance with RG 1.7. The applicant's h*j' analysis shows that one electric hydrogen.recombiner actuated at a containment h hydrogen concentration of'3.0 volume percent is capable of limit.ing the hydro-gen concentration in containment to below the R.G. 1.7 lower flammability limit dis of 4.0 volume percent. The applicant should discuss the emergency procedures
'] that will be in effect to guide the operator in actuating the hydrogen analyzer.
.a y
) -
The applicant has evaluated the possibility for pocketing of hydrogen in the
. containment following a LOCA and concluded that pocketing of hydrogen is not f
8 very likely. This finding is based on the open, internal design of the contain-ment, the low hydrogen generation rates from the various potential sources, and
% the effectiveness of hydrogen mass transport by convection. Based on our review y
of the applicant's rationale, we agree with the applicant's-conclusion that -- % b pockets of flammable hydrogen are not likely to form. P.
'O 5 We conclude that the CGCS satisfies the design and performance requirements of hj Section 50.44 of 10 CFR Part 50, " Standards for Combustible Gas Control Systems w $ in Light Water Cooled Power Reactors," the guidelines of Regulatory Guide 1.7 M and the requirements of GDC 41, 42, 43, and'50, and is acceptable, provided the e
M applicant justifies the hydrogen production analysis, and adequate emergency
~
procedures are in place to guide the operator in actuating the hydrogen analy-S zers and hydrogen recombiners.
~. ~9 ?!
j} 6.2.6~ Containment Leakage Testing Program 9 Q The containment design includes the provisions and features necessary to satisfy the testing requirements of Appendix'J to 10 CFR Part 50. The design of the .
~
LM ,[2 I 6-20 WNP-3 DSER SEC 6 , )j' { s- w= ~ _ - - . - - . . - ,,. . -_ -,- .- _
.v .. - -
_ _ = . .
. ~ . \
l , it I
~!
containment penetration and isolation valves permits periodic leakage rate tes-I ting at the pressure specified in Appendix J to 10 CFR Part 50. Included are l those penetrations that have resilient seals and expansion bellows; i.e., air j locks, emergency hatches, and electrical penetrations. 14
.t ,
The containment leakage testing program complies with the requirements of
@l e
Appendix J to 10 CFR Part 50. Such compliance provides adequate assurance that y
.q h containment leaktight integrity can be verified throughout service lifetime and
- !j that the leakage rates will be periodically checked during service on a timely il basis to maintain such leakages within the specified limits of the Technical
- 4 11 Specifications. The plant's Technical Specifications will contain appropriate
.a q surveillance requirements for containment leak testing, including test frequen-1 j cies.
d
, Maintaining containment leakage rates within such limits provides reasonable assurance that, in the event of any radioactivity releases within the contain-
]
d ment, the loss of the containment atmosphere through leak paths will not be in excess of acceptable limits specified for the site; i.e., the resultant dose
} will be within 10 CFR Part 100 guidelines in the event of a design basis LOCA.
u 4 We conclude that the applicant's program complies with the requirements of
'1 Appendix J and with the requirements of GDC 52, 53, and 54, and therefore, is --
[ acceptable. j < Outstanding issues, Containment Systems Branch O,.. h (1) The forces and moments acting on the reactor vessel following the design ij basis reactor coolant system break within the reactor cavity is required. ( This information may be either plant (WNP-3) specific or shown to be eppli-cable to WNP-3. The guidelines of SRP 6.2.1.2 and Section 3.2 of NUREG-0609 should be followed. a N? M (2) A minimum containment pressure analysis to support emergency core cooling system capability studies that is a'pplicable to WNP-3 should be provided. i4 (3) The applicant's response to IE Bulletin 80-04, Main Steam Line Break with.
~
, Continued Feedwater Addition, is required. il . 7 6-21 WNP-3 DSER SEC 6
w -
.. - . w- . ua m w:a t
[ (4) Additional information is required concerning the applicant's shield
- buiding pressure response analysis. In particular, a more detailed descrip-i tion of the WATEMPT code.is needed, including the assumptions made in the i analysis, to determine the conservatism in the results. -1 il J.1 (5) The applicant should upgrade the isolation provisions for the containment 4
fj vent make up line (P-80). It is the staff's position that the containment j isolation valves in line P-80 should be automatic, power operated valves
- lf having less than 5-second closure times, and should fail close upon loss
.]A of power to the valve operators.
Ci i (6) The applicant should commit to seal close the 48-inch purg'e valves either Q electrically or mechanically. q (7) The applicant should justify that the setpoint pressure for containment l1) isolation is the minimum value compatible with normal cperation conditions, g (8) The applicant should provide an analysis of the effect of purge system Hl.j operation at the time of a LOCA on the minimum containment pressure analy-
-:j sis for the ECCS performance evaluation.
.j 2]
?1 (9) It is the staff position that the applicant provide Class IE emergency -
g power to valves 2CH-VQ040 and 2CH-VQ005 in the reactor coolant pump seal injection and chemical and volume control paths. J (10) With regard to the SIS recirculation sump discharge lines (penetration 23
] and 24), the applicant should discuss the adequacy of the criteria used in
]q the design of the piping between the containment and the isolation valve, $ and.the valve itself, and the leakage control provisions on the penetration fj . or serving the penetration area. c l (11) The applicant should discuss the basis for the hydrogen production analy-sis and justify any deviations from RG 1.7. J (12) The applicant should discuss the emergency procedures that will be in effect 3 to guide the operator in actuating the hydrogen analyzers end hydrogen j recombiners.
~
i 6-22 WNP-3 DSER SEC 6
-___ -__ __ . . _ . . . . _ _ . ~ -- ___ - _ _ . - .- _ _. ---
. .. 1
- =: -
- . _ _ _ . _ _ m _ .Mhd y
J i j i si j 6.2.7 Fracture Prevention of Containment Pressure Boundary 3 '! 4*
] Our safety evaluation review 0- Asses the ferritic materials in the Washington ]j5 Public Power Supply $ystem Nuclear Project No. 3 (WNP-3) containment system that constitute the containment pressure boundary to determine if the material 4 fracture toughness is in compliance with the requirements of General Design
- Criterion 51, " Fracture Prevention of Containment Pressure Boundary."
rv . . ,[q GDC 51 requires that under operating, maintenance, testing and postulated acci-(g dent conditions, (1) the ferritic materials of the containment pressure boundary a, behave in a nonbrittle manner and (2) the probability of rapidly propagating I) ji fracture is minimized. 1 a The WNP-3 containment system includes a ferritic steel containment vessel d enclosed by a' reinforced concrete shield building. The ferritic materials of
} the containment pressure boundary which are considered in our assessment are j those which have been applied in the fabrication of the containment vessel, N equipment hatch, personnel locks, penetrations and fluid system components M including the valves required to isolate the system. These components are the d parts of the containment system which are not backed by concrete and must sus-g tain loads during the performance of the containment function under the con-ditions cited by GDC 51. -
q j We have determined that the fracture toughness requirements contained in ASME Code editions and addenda typical of those used in the design of the WNP-3 con-g tainment may not ensure compliance with GDC 51 for all areas of the containment pressure boundary. We have elected to apply in our licensing reviews of fer-q ritic containment pressure boundary materials the criteria for Class 2 compo-j nents identified in the Summer 1977 Addenda of Section III of the ASME Code. f) Because the fracture toughness criteria that have been applied in construction ?
- d typically differ in Code classification and Code edition and addenda, we have
] chosen the criteria in the Summer 1977 Addenda of Section III of the Code to i provide a uniform review, consistent with the safety function of the containment 3 pressure boundary materials. Therefore, we will review the materials of the q. components of th'e WNP-3 containment pressure boundary according to the fracture toughness requirements of the Summer 1977 Addenda of Section III for Class 2
]j components.
- 1
} 6-23 WNP-3 DSER SEC 6 m
y= w_;_.= s = .sL22 :,,_u .
'i
- a dy 9 .
]
] Considered in our review will be components of the containment system which are i load bearing and provide a pressure boundary in the performance of the contain-
't j ment function under operating, maintenance, testing and postulated accident cy conditions as addressed in GDC 51. These components are the containment ves-
] sel, equipment hatch, personnel airlocks, penetrations and elements of specific 63 containment penetrating systems.
$r.
d c4 In some cases, materials will not have been fracture toughness tested or will y have been inappropriately tested. Generally, those materials will not h've a lq:o been fracture toughness tested because the ASME Code edition and addenda in y effect at the time the components were ordered did not require that they be tested. Our assessment of the fracture toughness of materials'that were not J i, fracture toughness tested or were inappropriately tested is based on the metal-Q lurgical characterization of these materials and fracture toughness data pre-4 sented in NUREG-0577, " Potential for Low Fracture Toughness and Lamellar Tearing
'j on PWR Steam Generator and Reactor Coolant Pump Supports," USNRC, October 1979 . and ASME Code Section III, Summer 1977 Addenda, Subsection NC.
U
,j The metallurgical characterization of these materials, with respect to their c; fracture toughness, will be developed from a review of how these materials were al j fabricated and what thermal history they experienced during fabrication. The
] metallurgical characterization of these materials, when correlated with the -
'j
; data presented in NUREG-0577 and the Summer 1977 Addenda of the ASME Code Sec-
] tion III, provides the technical basis for our evaluation of compliance with
-}
the Code requirements for materials that were not fracture toughness tested. n. H q Based on our review of the available fracture toughness data and materials fabri-y cation histories, and the use of correlations between metallurgical characte-8 ristics and material fracture toughness, we then will conclude that the fer-k y
. ritic components in the WNP-3 containment pressure boundary meet the fracture
] toughness requirements that are specified for Class 2 components by the 1977 (q Addenda of Section III of the ASME Code. Compliance with these Code require-d ments provides reasonable assurance that the WNP-3 reactor containment pressure 7 boundary will behave in a nonbrittle manner, that the probability of rapidly i i propagating fracture will be minimized, and that the requirements of GDC 51 are
'. ; satisfied. , .
.i
<j 6-24 WNP-3 DSER SEC 6 ~j
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; 6.4 Control Room Habitability 1 _ - -
j The requirements for the protection of the control room personnel under acci-dent conditions are specified in General Design Criterion (GDC) 19. The appli-l y cant proposes to meet these requirements by incorporating shielding and emer-j gency ventilation systems in the control room design and by having an adequate - [ supply of self-contained breathing apparatus available in the control room for [.d the emergency team. The applicant has stated in the FSAR that there is redun- , q^ 1
,1
.2 dancy in the emergency systems and that the control room emergency ventilation
] systems are in general conformance with Regulatory Guide 1.52. This conformance y is evaluated in Section 6.5.1 of this report.
,1 .
j The control room heating, ventilating and air conditioning (HVAC) system is 1-ij designed to automatically transfer to the emergency isolation mode of operation
~l upon receipt of a high radiation or chlorine signal from the outside air intake duct detectors or a containment isolation actuation signal. In the emergency isoittion mode, the control room air is to be recirculated at the rate of 6000
]
? cubic feet per minute (cfm) through 99% efficient ESF grade, charcoal filtration d units. In the event of a radiation release, the operator would override the 3 isolation mode and manually initiate the emergency pressurization mode of oper-ation. Each emergency filter train can supply a maximum of approximately 1000 cfm of filtered outside air for pressurization. The transfer to the emergency -
a] S modes of operation may also be manually initiated from the control room. d .. The staff has evaluated the habitability of the control room in accordance with a SRP Section 6.4 and Regulatory Guides 1.78 and 1.95. The applicant's estimated y concentration levels of S0, in the control room following an accidental release 2 are much higher than the guideline values of Regulatory Guide 1.78. The staff h will, therefore, require additional analysis, including consideration of' the g basis for the 13% S0 2 release fraction, outside air intake rate, isolation time, da concentration of 502 inside and outside the control room as a function of time, j basis for meteorological assumptions, emergency procedures, locations of all j onsite toxic gas release points relative to normal and emergency control room
) air intakes and other data specified in Table C.3 of Regulatory Guide 1.78. In addition, the applicant will be required to comply with Regulatory Guide 1.78 r (Regulatory Position C.3) and SRP Section 6.4 (page 12) which call for l-i 6-'25 WNP-3 DSER SEC 6
hi;u _ n _ __.u: a .a.l
\ ..
n
,j J; instrumentation readout and alarm in the control room and quick acting automatic . isolation, or provide adequate justification for any alternatives.
N The staff also determined that the following areas relating to the control room d :1 dose need to be addressed by the applicant: n; ' [. (1) justification for the location of the normal and emergency air intakes g relative to major radiation release points; and
'[.j (2) measures the applicant will take (emergency procedures) to assure that the j control room operator will manually switch the emergency ventilation sys-
) tem to the pressurization mode in the event of a radiation release.
i [ Based upon the foregoing, the applicant has not demonstrated that the control
]ij room habitability system will adequately protect the control room operators in accordance with the requirements of NUREG-0737, Item III.D.3.4, and 10 CFR Part 50, Appendix A, GDC 19.
6.5 Engineered-Safety-Feature Atmosphere Cleanup System L 6.5.2 Containment Spray as a Fission Product Removal System c1 d q FSAR Question 450.~1 requested additional information concerning the design and - y operation of the containment spray system. -Because the applicant has not sup-plied the requested information, the review of the containment spray system s cannot be completed. Therefore, the containment spray system design will be 51 considered as an open item. 9 p y 6.6 Inservice Inspection of Class 2 and 3 Components M
.; This section was prepared with the technical assistance of DOE contractors from the Idaho National Engineering Laboratory.
q l H 6.6.1 Compliance with the SRP li The July 1981 Edition of the " Standard Review Plan for the Review of Safety N Analysis Reports for Nuclear Power Plants" (NUREG-0800) includes Section 6.6,. .
; 17 6-26 WNP-3 DSER SEC 6 1
- .. . . .:.. .,nw N:z
. c., _ . _ _ _ _.._
l l .
't j " Inservice Inspection of Class 2 and 3 Components." The review is continuing I
because the applicant has not submitted a Preservice Inspection (PSI) Program l and has not complete'd the PSI examinations. The FSAR Table 1.8-3 states that j compliance with Section 6.6 of NUREG-0800 is under review and a compliance state-j ment will be provided in a sub uquent amendment. The staff review to date was
.) conducted in accordance with Standard Review Plan Section 6.6 except as dis-cussed below. '
j li - j Paragraph II.3, " Acceptance Criteria, Examination Categories and Methods," will .$ be reviewed when the completed PSI Program has been received. l . i q Paragraph II.4, " Acceptance Criteria, Inspection Intervals," ha's not been re- ~ .] viewed because this area applies only to inservice inspection (ISI) not to PSI. d This subject will be addressed during review of the ISI program after licensing. il
)
j Paragraph II.5, " Acceptance Criteria, Evaluation of Examination Results," has d been reviewed. The Applicant committed in the FSAR to incorporate ASME Code i Section XI, Articles IWC-3000 and IWD-3000, " Standards for Examination Evalu- . ation," into the PSI program. However, ongoing NRC generic activities and re' search projects indicate that the presently specified ASME Code procedures may not always be capable of detecting - the acceptable size flaws specified in these standards. For example, ASME Code . procedures specified for volumetric examination of vessels, bolts and studs, and piping have not proven to be capable of detecting acceptable size flaws in all cases. The staff will continue to evaluate the development of new or improved procedures and will require that these improved procedures be made a part of the inservice examination requirements. The Applicant's repair pro-f 3, cedures based on ASME Code Section XI, Articles IWC-4000 and IWD-4000, " Repair
, Procedures," have not been reviewed. Repairs are not generally necessary in d the PSI program. This subject will be addressed during review of the ISI pro-1 y gram. . .
.d li Paragraph II.7, " Acceptance Criteria, Augmented ISI to Protect Against Postu-j lated Piping Failures," has not been completed because this subject has not yet l l, been addressed in the applicant's PSI program. The applicant's augmented ISI
~
l j program will be reviewed after it is submitted. l! - 6-27 WNP-3 DSER SEC 6 Lj El !L _ _ . __- _ _ _ _ -
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} l Paragraph II.8, " Acceptance Criteria, Code Exemptions," will be reviewed for i compliance to IWC-1220 when the Applicant's PSI Program has been received.
Paragraph II.9, " Acceptance Criteria, Relief Requests," has not been completed '
.)
because the applicant has not identified the limitations to examination.
; Specific areas where ASME Code examination requirements cannot be met will be -! identified as the PSI progresses. The complete evaluation of the PSI program wil be presented in a supplement to the Safety Evaluation Report (SER) after l the Applicant submits the required examination information and identifies all plant-specific areas where ASME Code Section XI requirements cannot be met and 3
provides a supporting technical justification. 3 6.6.2 Examination Requirements I t d General Design Criteria 36, 39, 42, and 45, Appendix A of 10 CFR Part 50 l I requires, in part, that the Class 2 and 3 components be designed to permit i appropriate periodic inspection of important components to ensure system inte-I grity and capability. Section 50.55a(g) of 10 CFR Part 50 defines the detailed 1 requirements for the preservice and inservice inspection programs.
] Based upon the construction permit date of April 11, 1978, this section of the
] regulations requires that a preservice inspection program for Class 2 and 3 components be developed a'nd implemented using at least the Edition and Addenda -
l; of Section XI of the ASME Code applied to the construction of the particular j components. The components (including supports) may meet the requirements set [j forth in subsequent editions of this Code and Addenda.which are incorporated by fj reference in 10 CFR 50.55a(b) subject to the limitations and modifications -{j listed therein.
,p $l 1 The initial ISI program must comply with the requirements of the latest Edition
' -Q . and Addenda of Section XI of the ASME Code in effect twelve months prior to the
] date of issuance of the operating license, subject to the limitations and modifi-4 cations listed in Section 50.55a(b) of 10 CFR Part 50.
! -; E ii 6.6.3 Evaluation of Compliance with 10 CFR 50.55a(g) '7 . i Review has been completed on the information presented in the FSAR through
, Amendment 3 dated April 1983. The Class 2 and Class 3 piping and components
- v. i fj 6-28 WNP-3 OSER SEC 6
?
,y _ . _ . .
dL. diu....= _v x- .: a.m. a:.. a -...= = .--_. = a a l 4
.i
>l will receive preservice examinations in accordance with the requirements of the q 1977 Edition of ASME Code Section XI with Addenda through Summer 1978. The
. secondary side of the steam generators will be examined in accordance with the requirements of the 1974 Edition of ASME Code Section XI with Addenda through d Summer 1975.
I 2 j The Preservice Inspection (PSI) Program for the Class 2 and 3 components has j not been received. However, the applicant has stated in the FSAR that these
.d components will be examined per the applicable Code requirements. Based on the j review of the FSAR, the staff has established technical positions that should
) be included in the PSI Program. The applicant has committed to identify all
] plant-specific areas where the Code requirements cannot be met'after the exami-tj nations are performed and provide a supporting technical justification for j requesting relief. The SER will be completed after the Applicant:
0 j (1) Dockets a complete and acceptable PSI Program, l 4 (2) Submits the requested additional information regarding the PSI /ISI program, 1 and ij (3) Submits all relief requests with a supporting technical justification, y
~
d j The staff considers the review of the PSI Program an open issue subject to the Applicant providing an acceptable response to the above requirements. The initial Inservice Inspection Program has not been submitted by the appli- ,1 cant. This program will be evaluated after the applicable ASME Code Edition J and Addenda can be determined based on Section 50.55a(b) of 10 CFR Part 50, but j before inservice inspection commences during the first refueling outage. 6.6.4 Conclusions
;j '1' Compliance with the preservice and inservice inspections required by the ASME
,, Code and 10 CFR Part 50 constitutes an acceptable basis for satisfying applic- , j able requirements of General Design Criteria 36, 39, 42, and 45.
! \
, "] 6-29 WNP-3 DSER SEC 6 i e______________________.__.____-.
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- i 6.6.5 References q .
- 1. NUREG-0800, Standard Review Plan, Section 6.6, " Inservice Inspection of Class 2 and 3 Components," July 1981. .
- ] .
.j 2. Code of Federal Regulations, Volume 10, Part 50.
11 Kf
'} 3. American Society of Mechanical Engineers Boiler and Pressure Vessel Code, d Section XI, Division 1.
1
-:)
j 1974 Edition through Summer 1975 Addenda ~
? '
a 1977 Edition through Summer 1978 Addenda I Review of the FSAR and Technical Positions Regarding the l Preservice(PSI)/ Inservice (ISI) Inspection Programs i I Question 250.1 For completion of SER Sections 5.2.4 and 6.6, the staff requires that the PSI Program Plan be submitted for review. The PSI Program should include reference to the ASME Code Section XI Edition and Addenda that will be used for the selec-
; tion of components for examinations, lists of the components subject to exami- -
nation, a description of the components exempt from examination by the applic-able Code, and the examination isometric drawings. Paragraph 50.55a(b)(2)(iv) requires that ASME Code Class 2 piping welds in the Residual Heat Removal (RHR) Systems, Emergency Core Cooling (ECCS) Systems, and Containment Heat Removal (CHR) Systems shall be examined. These systems should not be completely exempted from preservice volumetric examination based on Sec-tion XI exclusion criteria contained in IWC-1220. To satisfy the inspection requirements of General Design Criteria 36, 39, 42, and 45, the Preservice q Inspection Program must include volumetric examination of a representative
} sample of wolds in the RHR, ECCS and CHR Systems.
1 J _1 -
-1 -
6-30 WNP-3 DSER SEC 6 i - 1 1 .-
s _ _w '
..u _.~ _ _._ ._ _ . .- - . 2 _ m .m . :_ = r:,
J ~N 1 Question 250.2 1 u
] Plans for preservice examination of the reactor pressure vessel welds should j address the degree of compliance with Regulatory Guide 1.150. In FSAR j Section 1.8, Table 1.8-1, the Applicant indicates exceptions to Regulatory Cj Guide 1.150. List the exceptions being taken and discuss the degree of compli- .I ance and the qualification of procedures to be used to assure finding service-
"i . induced flaws on the inside surface. j Question 250.3 Describe the measures taken to ensure that austenitic stainless' steel piping welds are examined using effective techniques and the methods of assuring ade-quate examination sensitivity over the required examination volume. Discuss
, the preservice examination criteria used to record, report, and plot geometric
$ or metallurgical ultrasonic indications in the piping systems to assure corre-
. lation of baseline data with inservice inspection results.
, s4 The ASME Code, Section XI, 1977 Edition with Addenda through Summer 1978 and
-{ 1980 Edition specifies the use of Appendix III of Section XI for ferritic piping
.l welds. If this requirement is not applicable (for example, for austenitic piping welds), ultrasonic examination is required by Section XI to be conducted in -
] accordance with the applicable requirements of Article 5 of Section V, as amended by IWA-2232. A technical justification is required if any alternatives are used. If Section XI, Appendix III, Supplement 7, will be used for the exami-
^
nation of austenitic piping welds, discuss the following: l1
'l (1) All modifications permitted by Supplement 7 >
- l. . (2) Methods of qualifying the procedure for examination through the weld (if i
4 complete examination is to be considered for examination conducted with b only one side access).
]
A I When using either Article 5 of Section V or Appendix III of Section XI'for 'i ,i examination of either ferritic or austenitic piping welds, the following should lj be incorporated. l .- 6-31 WNP-3 DSER SEC 6 l
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; -(1) Any crack-like indication, regardless of ultrasonic amplitude, discovered d during examination of piping welds or adjacent base metal materials should be recorded and investigated by a Level II or Level III examiner to the j extent necessary to determine the shape, identity, and location of the j reflector.
4 j (2) The Owner should evaluate and take corrective action for the disposition y of any indication investigated and found to be other than geometrical or j metallurgical in nature. b ' d Question 250.4 J - i
] All preservice examination requirements defined in Section XI of the ASME Code that have been determined to be impractical must be identified and a supporting d
) technical justification for requests for relief must be provided. The relief
- request submittal should include at least the following information:
(1) For ASME Code Class 1 and 2 components, provide a table similar to IWB-2500 j and IWC-2500 confirming that either the Section XI preservice examination
'j was performed on the component or relief is requested. .. (2) Where relief is requested for pressure retaining welds in the reactor ves- -
a i sel, identify the specific welds that did not receive a 100% preservice
! ultrasonic examination, and indicate the extent of the examination that
; was performed.
-i
] (3) Where relief is requested for piping system welds (Examination Category j 8-J, C-F, and C-G), provide a list of the specific welds that did not
] receive a complete Section XI preservice examination including drawing or isometric identification number, system, weld number, and physical con-q figuration (e.g., pipe-to-nozzle weld, etc.). Indicate the extent of the
] preservice examination-that was performed. When the voluinetric examina-J tion was performed frr one side of the weld, discuss whether the entire l l weld volume and the heat affected zone (HAZ) and base metal on the far side of the weld were examined. State the primary reason that a specific
, j examination is impractical (e.g., support of component restricts access, t . !q j ~ 6-32 WNP-3 DSER SEC 6 q q .
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a 4 fitting prevents adequate ultrasonic coupling on one side, component-to-
.: component welds prevent ultrasonic examination, etc.). Indicate any alter-
-l' native or supplemental examinations performed and methods of fabrication
.4
.-j examination.
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6-33 WNP-3 DSER SEC 6 t 9
.-.=.-a.-..=.=.-
9 AUXILIARY SYSTEMS , 4 j The staff has reviewed the design of the auxiliary systems necessary for safe j reactor operation, shutdown, fuel storage, or whose failure might affect plant j safety, including their safety-related objectives and the manner in which jj these objectives are achieved.
'c] ) The design for auxiliary systems for WNP-3 is the responsibility of the applicant. 't However, since WNP-3 is a CESSAR reference plant, the applicant is required tc.
incorporate the interfaces identified in CESSAR for the various auxiliary system designs. Refer to the corresponding sections of the CESSAR SER for a discussion l1 of staff's evaluation of the CESSAR interfaces. The auxiliary systems necessary for safe reactor operation or shutdown include
] the component cooling water system, the ultimate heat sink, the condensate stor-l age facility, the auxiliary feedwater system, the essential chilled water system, essential portions of the compressed air, equipment and floor drainage, and cliemical and volume control systems, and the heating, ventilation and air con-j ditioning (HVAC) systems for the control room, ESF systems and essential portions j of the Reactor Auxiliary Building (RA8).
i
-l j The auxiliary systems necessary to assure the safety of the fuel storage facility j include new fuel storage, spent fuel storage, the fuel pool cooling and cleanup
- . system, fuel handling systems and the HVAC system for the essential portions of the fuel building.
1 l' The staff has also reviewed other auxiliary systems to verify that their fail-ure will not prevent safe shutdown of the plant or result in unacceptable
] release of radioactivity to the environment. These systems include the demin-eralized water makeup system, potable and sanit'ary water system, service water ]
,j system, circulating water system, plant makeup water system, nonessential j chilled water system, nonessential portions of the compressed air, equipment and floor drainage and chemical and volume control systems, and the HVAC 1
~; )! WNP-3 DSER SEC 9 9-1
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I i i 1 systems for the turbine building and nonessential portions of the RAB and the j fuel building. I j None of the above mentioned systems performing a safety function or utilized for maintaining the safety of the fuel storage facility is shared between WNP-3
.i and any other WNP unit. Therefore, the requirements of General Design Crite-rion 5, " Sharing of Structures, Systems, and Components," which concerns the
] 1 capability to maintain safe operation of multiple units when essential systems are shared, are not applicable.
}
) 9.1 Fuel Storage and Handling 1
9.1.1 New Fuel Storage 1
] The new fuel storage facility was reviewed in accordance with Section 9.1.1 of the Standard Review Plan (SRP), NUREG-0800. An audit review of each of the areas listed in the " Areas of Review" portion of the SRP section was performed
'.; according to the guidelines provided in the " Review Procedures" portion of the
; SRP section. Conformance with the acceptance criteria except as noted below.
; formed the basis for the staff's evaluation of the new fuel storage facility
,i t with respect to the applicable regulations of 10 CFR 50. i
! The acceptance criteria for the new fuel storage facility include compliance with guidelines of ANS 57.1, " Design Requirements of Light-Water Reactor Fuel j Handling System," and ANS 57 3, " Design Requirements for New LWR Storage Facili-l ties." The guidelines contained in the " Review Procedures" were used in lieu .1 of ANS 57.1 and ANS 57.3.
j a
.j The new fuel storage facility provides dry storage for a maximum of 90 fuel assemblies (more than one-third of a core load) and includes the new fuel assem-
- bly storage racks, and the concrete storage cavity that contains the storage
, racks.
l
! The fuel handling building which houses the facility is designed to seismic lI Category I criteria as are the storage racks and cavity. This building is also
; designed against flooding and tornado missiles (refer to Sections 3.4.1 and ;
WNP-3 D5ER SEC 9 9-2 L
-- u ' _ m .a._.s _ _ n. - t 1 wc __uua n waa.a x - :.:
!I I Lj '1 . ii ,1 3.5.2 of this SER). Thus, the requirements of General Design Criterion 2, "De-sign Bases for Protection Against Natural Phenomena," and the guidelines of Regu- !] latory Guide 1.29, " Seismic Design Classification," Position C.1 are satisfied. i sj li- The storage cavity housing the new fuel storage racks is not located in the l il. 11 vicinity of any moderate or high-energy lines or rotating machinery. Therefore, 7 physical protection by means of separation is provided for the new fuel from
, internally generated missiles and the effects of pipe breaks (refer to Sec-
! .} tions 3.5.1.1 and 3.6.1 of this SER). Thus, the requirements of General Design Criterion 4 " Environmental and Missile Design Bases," are satisfied. H 3 The facility is designed to store unirradiated, low emission, fuel assemblies. U Accidental damage to the fuel would release relatively minor amounts of radio- f !l, g activity that would be accommodated by the fuel building ventilation system. j The racks can withstand the maximum uplift forces exerted by the fuel handling i j 1 machine. Thus, the requirements of General Design Criterion 61, " Fuel Storage 1.! and Handling and Radioactivity Control," are satisfied. Li [: The new fuel storage racks are designed to store the fuel assemblies in an array I with a minimum center-to-center spacing which is sufficient to maintain a K,ff jj of 0.95 or less in flooded condition. The racks are also designed to maintain .j a K,ff of 0 98 or less under optimum moderation (foam, small droplets, spray, _ (! or fogging). The racks themselves are designed to preclude the inadvertent ' l- placement of a fuel assembly in other than the prescribed spacing. Thus, the ] requirements of General Design Criterion 62, " Prevention of Criticality in Fuel [j Storage and Handling," are satisfied. !4 l Based on the staff's review, it concludes that the new fuel storage facility is g [} in conformance with the requirements of General Design Criteria 2, 4, 61, and 62, ; l-I as they relate to new fuel protection against natural phenomena, missiles, pipe > t! g break effects, radiation protection and prevention of criticality, and the. [. guidelines of Regulatory Guide 1.29 relating to seismic classification. The ' L staff, therefore, concludes that the design is acceptable and meets the accep- !i tance criteria of SRP Section 9.1.1. The staff further concludes that the CESSAR interface requirements are satisfied by the above described design. J WNP-3 DSER SEC 9 9-3 t ' . . . _ . _ _ _ _ _ _ . - - ~ _ . _ _ . , _ _ . . _ . _ . _ _ . _ _ . _ . _ . - - . _ _ .- _ _,-.- _ -
- w. a...u u._. n._m .--- -
_ _ ...__ a a __ - . c a _ :- - %s 1I 9.1.2 Spent Fuel Storage ~ The spent fuel storage facility was reviewed in accordance with Section 9.1.2 i of the Standard Review Plan (SRP), NUREG-0800. An audit review of each of the l areas listed in the " Areas of Review" portion of the SRP section was performed [ according to the guidelines provided in the " Review Procedures" portion of the t SRP section. Conformance with the acceptance criteria, except as noted below, h formed the basis for the staff's evaluation of the spent fuel storage facility .; with respect to the applicable regulations of 10 CFR 50. ?
- y The acceptance criteria for the spent fuel storage facility include compliance with various portions of the guidelines of AN5 57.2, " Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations."
}
J The guidelines contained in the " Review Procedures" were used in lieu of ANS 57.2. l Additionally, the acceptance criteria include Regulatory Guide 1.115, " Protection I f Against Low Trajectory Turbine Missiles." Turbine missiles are evaluated sepa-
- i rately in Section 3.5.1.3 of this SER.
') + .! Protection against damage to stored irradiated fuel due to failure of light / heavy
- J lo-J handling systems is discussed in Sections 9 1.'4 and 9.1.5 of this report.
A L$ ,1 The spent fuel storage facility provides underwater storage for 1120 fuel assem- _.
] blies or for at least 10 normal refuelings plus one full core off-load. The i facility, located in the fuel handling building, includes the spent fuel storage fl racks and the stainless steel lined concrete pool that contains the storage
+- racks. A i .] The structure housing the facility (the fuel handling building) including the i' spent fuel pool, storage racks, and gates is designed to seismic Category I ,, criteria. The spent fuel pool liner plate is designed to stay in place in an ! y SSE, thus precluding potential mechanical damage to the spent fuel or damage - resulting from overheating due to blocking of cooling water flow paths. The fuel handling butiding is also designed against flooding and tornado missiles
- H (refer to Sections 3.4.1 and 3.5.2 of this SER). Thus, the requirements of L General Design Criterion I., " Design Bases for. Protection Against Natural Phenom-
- ena," and the guidelines of Regulatory Guides 1.13, " Spent Fuel Storage Facility
- WNP-3 DSER SEC 9 9-4 l-
l
%G _. w . _. ._ . ._ . .. L .. m 2 .c 4. . .l a _ _. . .m 1. .. .~ . .m _ . _ . . ML -- -
3- -. I Design Basis," Position C.3, 1.29, " Seismic Design Classification," Positions C.1
)i and C.2, and 1.117, " Tornado Design Classification," Positions C.1 through C.3 i are satisfied for the facility.
1
- j The fuel pool is not located in the vicinity of any high-energy lines or Ii rotating machinery. Therefore, physical protection by means of separation is
} provided for the spent fuel from internally generated missiles and the effects of pipe breaks (refer to Sections 3.5.1.1 and 3.6.1 of this SER). Thus, the '
requirements of General Design Criterion 4, " Environmental and Missile Design l Bases," and the guidelines of Regulatory Guide 1.13, Position C.3 are
! satisfied.
l
] Each spent fuel assembly will be stored in a stainless steel can. The assembled 'l storage cans are formed into rack assemblies with the cans oriented such that .l alternate storage cans are positioned with neutron absorbing B C plates located 4
I i in the gaps on the outside of the cans in the north-south and east-west direc-
; tions respectively. This will ensure neutron absorber plates between adjacent storage locations. The spacing and design of the racks are such that the effec-tive multiplication factor (K,ff) for new or spent fuel stored within the rack,
! will not exceed 0.95 under all conditions including fuel handling accidents.
The rack arrays have a center-to-center spacing of 11.12 inches. The storage
}l racks are designed such that a fuel assembly cannot be inadvertently positioned -
j in other than a prescribed storage position. The racks can withstand the impact j of a dropped fuel assembly without unacceptable damage to the fuel and can with-
'i stand the maximum uplift forces exerted by the fuel handling machine. Thus, .j the requirements of General Design Criteria 61, " Fuel Storage and Handling and j Radioactivity Control, " and 62 " Prevention of Criticality in Fuel Storage and .j Handling," and the guidelines of Regulatory Guide 1.13 concerning fuel storage I facility design are satisfied.
1 I
.I The design of the spent fuel pool includes a fuel pool liner leakage system to
- 1 o detect and limit leakage of the pool liner welds, a pool water level and tem-1 I
.i perature monitoring and alarm system, and radiation monitoring and alarm systems with annunciation in the control room. These features satisfy the requirements
- ! of GDC 63," Monitoring Fuel and Waste Storage."
i l t1 WNP-3 DSER SEC 9 9-5 !j
,,.-,..,.,..,-r -- . . - -
, - - , ,4 -- - -
, , , - - - , - . .- - , . - - - . ,,,_r.,,-,.. ,
L ,- -
- m. -
_ =u. - . _ _ - -
, Based on the staff's review, it concludes that the spent fuel storage facility .f _ is in conformance with the requirements of General Design Criteria 2, 4, 61, 62, and 63 as they relate to protection against natural phenomena, missiles,
- j pipe break effects, radiation protection, prevention of criticality, and
[] ' monitoring provisions, and the guidelines of Regulatory Guides 1.13, 1.29 and 1.117 concerning the facility's design, seismic classification and protection
,- against tornado missiles. The staff, therefore, concludes that the spent fuel storage facility design meets the acceptance criteria of SRP Section 9.1 2 and
- .j is acceptable. The staff further concludes that the CESSAR interface require-ments are satisfied by the above described design
-i Spent Fuel Pool Materials -
., Introduction i' Nuclear reactor plants include storage facilities for the wet storage of spent fuel assemblies. The safety function of the spent fuel pool and storage racks is to maintain the spent fuel assemblies in a subcritical array during all credible storage conditions. We have reviewed the compatibility and chemical stability of the materials (except the fuel assemblies) wetted by the pool
- q water.
Evaluation The information provided in the FSAR was not sufficient for us to complete our evaluation. The applicant provided additional information by letters dated July 15 and September 2, 1983. The information provided in the applicant's a responses is insufficient for the completion of our evaluation. To complete our review, we need the following information; i lj (1) Identify and list by either brand name, generic name (i.e. , S.S. Type 304,. lq 316) or industry specification all the materials used for fabrication of
.l the high density spent fuel storage racks and all other structaral compo-
}} nents wetted by cooling water, except the fuel assembifes, including the i neutron poison material, rack leveling feet and rack frame. ,1
- I i
~1 WNP-3 DSER SEC 9 9-6 i
'I 1 -_ _ . _ _ _ _ . _ . . _ _ . _ . _ . . _ _ . _ _ _ . . _ _ _ _ . . _ _ _ _ _ _ . _ . . _ .
~ --
w
= , :a - . e_- ab l'
q l . l l j (2) Provide test or operating data which provides assurance that the neutron poison material will not degrade during the lifetime of the spent fuel ! I storage pool. (3) Provide a description of any materials monitoring program for the pool.
.! In particular provide information on the frequency of inspection and type N of samples used in the monitoring program. .i (4) Show that no buildup of gases will occur in the cavities containing the poison materials a
j 9.1.3 Spent Fuel Pool Cooling and Cleanup System - i
-t i The spent fuel pool cooling and cleanup system was reviewed in accordance with $, Section 9.1.3 of the Standard Review Plan (SRP), NUREG-0800. An audit review of each of the areas listed in the " Areas of Review" portion of the SRP section 2
I was performed according to the guidelines provided in the " Review Procedures" -f portion of the SRP section. Conformance with the acceptance criteria,.except [ as noted below, formed the basis for the staff's evaluation of the spent fuel j pool cooling and cleanup system with respect to the applicable regulations of 10 CFR 50. t _ .f The acceptance criteria for the spent fuel pool cooling and cleanup system ll include compliance with the guidelines of Regulatory Guide 1.52, " Design, j[ Testing and Maintenance Criteria for Post Accident Engineered-Safety-Feature
- Atmosphere Cleanup System Air Filtration and Absorption Units of Light-Water
- . Cooled Nuclear Power Plants," and 10 CFR Part 20 based on Regulatory Guide 8.8, C "Information Relevant to Ensuring That Occupational Radiation Exposures at l Nuclear Power Stations Will Be As Low As Is Reasonably Achievable." Compliance l with the guidelines of Regulatory Guides 1.52 and 8.8 is discussed separately j in Sections 9.4.2 and 12.1 of this SER, respectively.
i. '4 The spent fuel pool cooling and cleanup system is designed to maintain water { quality and clarity and to remove decay heat generated by spent fuel pool ll assemblies ii the pool. The system includes all components and piping from
] inlet to exit from the st: rage pools, piping used for fuel pool makeup, and the 1 ! cleanup filter /demineralizers to the point of discharge to the radw&ste system.
(l WNP-3 DSER SEC 9 9-7 o o, l ., kI *
- _ ua~.w.~ . a ... A a : . . . - . - . u ~ ...~ .-..- a O2 Li
] The design consists of two independent, full-capacity essential fuel pool cooling trains, each with a fuel pool cooling pump and a fuel pool cooling heat
,t i] exchanger and two separate nonessential purification trains, each with a fuel
.] pool cleanup pump, a cleanup basket strainer, a fuel pool filter, and a fuel
}
storage tank. The line from the RWST to the fuel pool is seismic Category I. j t Thus, the requirements of General Design Criterion 61. " Fuel Storage and Hand-ling and Radioactivity Control," and the guidelines of Regulatory Guide 1.13, 3 Position C.6 concerning fuel pool design are met. The system incorporates control room alarmed pool water temperature and building radiation level monitoring systems. Local pool water level alarms are provided
]. in the fuel handling building. Thus, the requirements of General Design Cri-1 terion 63, " Monitoring Fuel and Waste Storage," are satisfied.
l l Based on the staff's review, it concludes that the fuel pool cooling and cleanup l system is in conformancc with the requirements of General Design Criteria 2, 4, i 44, 45, 46, 61, and 63 r elating to protection against natural phenomena, mis-
] siles and environmental effects, cooling water capability, inservice inspection, functional testing, fuel cooling and radiation protection, and monitoring provi-
~1 sions, and the guidelines of Regulatory Guides 1.13 and 1.29, and Branch Techni-
] cal Position 9.2 relating to the systems design, seismic classification, and de-j '
sign decay heat loads. The staff therefore, concludes that the spent fuel pool ; cooling and cleanup system meets the acceptance criteria of SRP Section 9.1 3 i and is acceptable. The staff further conclude that the CESSAR interface require-y ments are satisfied by the above described design. Spent Fuel Pool Cleanup System Chemistry - Introduction i
,, The spent fuel pool cleanup system is designed to maintain optical clarity and I
q to remove corrosion products, fission products, and impurities from the spent j fuel pool water. Water purity and clarity in the spent fuel pool, refueling ! l pool, and refueling canal are maintained by filtering and demineralizing the k pool water through a mixed bed demineralizer. The pool cleanup system consists : of two parallel trains of cleanup equipment. The cleanup loop is normally run WNP-3 DSER SEC 9 9-8
w.._.-. . -. . - .-. .- - ..u .a . , ~ . . . .. - . u - a. waemaws I 1 j on an intermittent basis which is determined by the chemistry conditions of the spent fuel pool water. j Evaluation
.i The information provided by the applicant was not sufficient for us to complete 4
1 our evaluation. The applicant has not provided the additional information on q the spent fuel pool cleanup system. q , To complete our review, we need the following information: 1)l. I Describe the samples and instrumentation and their frequency of measurement
;j that will be performed to monitor the Spent Fuel Pool water purity and need for 'j ion exchanger resin and filter replacement. State the chemical and radio- ] chemical limits to be used in monitoring the spent fuel pool water and for ini- ] tiating corrective action. Provide the basis for establishing these limits.
Your response should consider variables such as: gross gamma and iodine activ-ity, demineralizer and/or filter differential pressure, demineralizer decon-j tamination factor, pH and crud level. 1 j 9.1. 4 Light Load Handling Systems (Related to Refueling) f
.} - ] The light load handling system was reviewed in accordance with Section 9.1.4 of j the Standard Review Plan (SRP) NUREG-0800. An audit review of each of the ] areas listed in the " Areas of Review" portion of the SRP section was performed according to the guidelines provided in the " Review Procedures" portion of the . SRP section. Conformance with the acceptance criteria except as noted below , formed the basis for staff's evaluation of the light load handling system with i respect to the applicable regulations of 10 CFR Part 50.
l j The acceptance criteria for the light load handling systems include meeting the-
] guidelines of ANS 57.1, " Design Requirements for LWR Fuel Handling Systems."
4l The guidelines contained in the " Review Procedures" portion of SRP Section 9.1.4 l were used in lieu of ANS 57.1. , i
! WNP-3 DSER SEC 9 9-9 I
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.t j Any load of weight less than one fuel assembly and its associated handling tool is defined as a light load. The light load handling system is related to re-fueling and consists of those components and equipment used in handling new fuel from the receiving station to the loading of spent fuel into the shipping cask.
The system includes the equipment designed to facilitate the periodic refueling l of the reactor and includes the refueling machine, the control element assembly 1 (CEA) change platform, the fuel transfer system, the spent fuel handling machine, j- new fuel elevator and associated handling tools and devices. The handling of fuel during refueling is controlled by a series of interlocks to ensure that fuel handling procedures are maintained.
- d The fuel handling system with the exception of new fuel handling crane, cask A
handling crane, containment polar crane and the system arrangement are in the CESSAR scope. Refer to Section 9.1.4 of the CESSAR SER for further discussion of the fuel handling system design. Also, refer to Section 9.1.5 of this SER for discussion of the cask handling and containment polar cranes which are used for overhead heavy load handling. The 10-ton new fuel handling crane is used to transfer new fuel from the ship-ping container to the new fuel storage racks. Additionally, during refueling,
, the crane is used to move each fuel assembly from the new fuel storage rack to the new fuel elevator. The associated hoist is equipped with a load sensing .
device which prevents the hoist from subjecting a fuel assembly to tensile loads greater than 5,000 pounds. The physical arrangement of the crane pre-cludes its travel over the spent fuel pool. Also, the crane is designed to prevent its falling into the new fuel handling area in an earthquake. Addi-d tional design features including interlocks and limit switches are provided to assure safe handling of loads and limit the potential for a load drop. The light load handling system is housed within the fuel building, containment and containment shield building annulus which are seismic Category I, flood-and tornado protected structures (refer to Sections 3.4.1 and 3.5.2 of this SER). Although fuel handling system components are not required to function following a safe shutdown earthquake, the equipment utilized in the fuel han-dling system is designed such that the combined dead loads, live loads and 4 l - 1 l' WNP-3 DSER SEC 9 0-10 1 l
- :.m.. x- -
-.Y- -L -.-w.-.. .c-
- - . - L: ?
i
-4 seismic loads will not cause any portion of the equipment to disengage from l their supports and fall. Therefore, they will not fail in a manner that re-sults in unacceptable consequences such as fuel damage or damage to safety-related equipment. The design thus satisfies the requirements of General j Design Criterion (GDC) 2, " Design Bases for Protection Against Natural Phe-j nomena," and the guidelines of Regulatory Guides (RGS) 1.29, " Seismic Design ] Classifications" (Position C.2) and 1.13, " Spent Fuel Storage Facility Design Bases" (Positions C.1 and C.6).
a j [The applicant has not provided an analysis concerning the specific criterion S relating to the maximum kinetic energy of any load lighter than a fuel assembly d 1 as identified in item'6 of the " Review Procedures" portion of SRP Section 9.1.4. 4
,1 i
The staff is, therefore, unable to conclude that when light loads are dropped
] over the irradiated fuel in the spent fuel pool or reactor vessel from their } maximum normal elevation, they will not result in greater fuel damage than that ] assumed for a dropped fuel assembly in the design basis fuel-handling accident.
j Consequently, the staff cannot conclude that the design meets the requirements of GDC 61, " Fuel Storage and Handling and Radioactivity Control," and 62, " Pre-j} vention of Criticality in Fuel Storage and Handling" and the guidelines of
) ~
RG 1.13, Position C.3 fr. this regard.]
^^
[ Based on the above, the staff concludes that except as noted above, the light - load handling system is in conformance with the requirements of GDC 2, 61 and l' 62 as they relate to protection against natural phenomena, safe fuel handling including prevention of criticality, and the guidelines of RG 1.13, Post-g tions C.1, C.3 and C.6 and 1.29, Position C.2, with respect to overhead crane interlocks, prevention of unacceptable releases in fuel handling accidents, and maintaining the plant in a safe condition following a seismic event. The staff . further concludes that the system design meets the CESSAR interface require-
,. ments. The light load handling system meets the acceptance criteria of SRP Section 9.1.4, except as noted above. The staff will report resolution of its O concerns identified above in the final SER.]
i iI l l WNP-3 DSER SEC 9 9-11 Ii l L,~ .
.,. - - - ~
,_ _ _ _ . . - - . . -~,
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1 - - 9.1.5 Overhead Heavy Load Handling System j ' The overhead heavy load handling systems were reviewed in accordance with Sec-tion 9.1.5 of the Standard Review Plan (SRP), NUREG-0800. An audit review of each of the areas listed.in the " Areas of Review" portion of toe SRP section was performed according t.o the guidelines prov'ided in the " Review Procedures" 1 portion of the SAhsecti'on. Conformance with the accepta'nce criteria except as
} noted below formed the ba[is for staff's waluation of the overhead heavy load handling system with respect to the applicable regulations of 10 CFR 50.
j The acceptant.e criteria for the overhead heavy load ' handling system include
.1
~
meetingthegejdelinesofANSI/ANS57e1and57.2. The guidelines contained in L' NUREG-0+y2 and the " Review Procedures" portion of SRP Section 9.1.5 were used in lieu of ANI/ANS 5/.1 andl7.2. m e 1 . % s Thboverhead'hdavy'Icadhandlingsystemconsistsofcomponentsandequipment usedtomoveloa'dsweichingmor$thanonefuelassemblyanditsassociatedhan-
. . s .. .
d1ing device z The~: equipment includes the cask handling crane, the containment polar crane and the new fuel handling crane. The fuel building cask handling crane with main and auxiliary hooks of 150 tons and 10 tons capacity respec-tively is used for handling a spent fuel shipping Sash The containment polar crane with main and auxiliary hooks of 250 tons and 25 tons capacity respec- - tively is useds to.more the reactor vessel head, head equipment support struc-
. ~
ture, and reactor intu r.als t The new fuel handling crane which is used both for light load hanlling and'ov'er(ead heavy load handling is discussed in Sec-
; tion 9.1.hafthisSER'.} '
l' [ 'O '
, ,c The overhead heavy load handling system lfs houyed
- within the fuel building and
.n containwn+, which are seismic Category g I, flood- and tornado protected struc-j tures (re.fer to Sections 3.4.1 and 3.5).qf thjf 3ER). Although.the fuel han-
' ~
d1ing system comporgnts are not required to function following an SSE, the cask ~ q handlingandcontaiN1fntpola'cranesargdesignedtoseismicCategoryIcrite-r .] > ria so that they willanot fail in a m@iner which resul_ts in unacceptable conse-i quences such as fuel damage or damage"to jafety-related equipment.1 Therefore, l1 , . 9;
) '
1 1 N. ) i g n - i, i
,; p q.g WNP-1 DSER SEC 9 T
. 'b' \ '4 9-12 ,
l-q - l ,O 'i 7- >
- _ _ . . .w:, i =..- x
. u.am uw-w n.n Y Nm .au.Lw_.-na ...1.==:
1 the design satisfies the requirements of General Design Criterion (GCD) 2, "De- 'j sign Bases for Protection Against Natural Phenomena," and the guidelines of j Regulatory Guides (RGS) 1.13, " Spent Fuel Storage Facility Design Bases," Posi-l- tions C.1. and C.6, and 1.29, "Sei mic Design Classifications," Position C.2. U , ;4 s The physical length of the rails through which the spent fuel cask handling crane travels limits the crane's approach to the spent fuel pool storace area such that either dropping or tipping of-the spent fuel cask into the spent fuel pool is not physically possible. _4 f The cask is transported along a prescribed path and interlocks prevent the cask from being Iffted in excess of 30 feet above any structural surface to minimize [.ij , the possibility of damage to spent fuel in the event, the cask is dropped. A y dropped cask cannot, therefore, result in fuel damage in excess of that assumed y in the design basis fuel handling accident, or damage safety-related equipment.
- } Thus, the requirements OF GDC 4, " Environmental and Missile Design Bases" and P 61, " Fuel Storage and Handling and Radioactivity Control" and the guidelines of
- RG 1.13, Positions C.3 and C.5 are satisfied for handling of the spant fuel
[ cask. ? L The containment polar crane is utilized to lift the reactor vessel closure head h and other equipment during refueling. It is also used to life equipment during _ I maintenance as needed. Refer to Section 9.1.4 of the CESSAR SER for a discus- ' h; sion of the reactor vessel closure head load drop analysis. Administrative con-r trols are provided to limit the maximum lift height for the reactor vessel clo-sure head to 17 feet above the reactor vessel closure flange thus satisfying the y CESSAR interface requirement in this area. The polar crane design also includes j various interlocks and limit switches to assure proper load handling. [The ap- } pitcant has not discussed whether movement of loads by the polar crane over ex-posed fuel in the reactor vessel will be prevented once the reactor vessel head [.1 and the upper guide structure are removed. Therefore, the staff is unable to conclude that the design of the polar crane includes protective devices against [' l' t potential damage to the spent fuel within the reactor due to dropped loads. The L A staff will report resolution of the concern in the final SER.] } p ! I' . l WNP-3 DSER SEC 9 9-13 l l?, O- _ l
lY ' wQ. = - L-~.--w-.w..=- = - u-- -
- na l
1 I I i !i- *
,S Additional criteria regarding the safe handling of heavy loads is contained in -1 .; 1 1 NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants." HUREG-0612 was . l. transmitted to the applicant for action by NRC generic letters dated December 22,
. 1980 and February 3,1981. NUREG-0612 resolved Generic Task A-36 and provideds g guidelines for necessary changes to ensure safe handling of heavy loads once the -l ~ plant becomes og rational. Specifically, Enclosure 2 attached to the December 22, ) 1980 generic letter identified a number of interim measures dealing with safe load paths, procedures,'perator o training, and crane inspeci.fons,3 testing and maintenance. Subsequently, in May 1983, the staff requested the applicant tv g;. provide-specific information relating to NUREG-0612. [By letter dated March 7,' M 1984, the applicant responded that they had no schedule for providing a response relating to compliancejwith the guidelines of,NUREG-0612. Therefore, the staff
- % n cannot conclude that the overhead heavy load handling system design meets the f,- requirements of GOC 4 and 61 until it completed its evaluation o'f the applicant's y response in this regard.J
'i
]. Qu [ Based on the above, the sYaff concludes that except as noted above, the over-head heavy load handling system is in conformance with the requirements of
~
l vj GDC 2,'4 and 61 as they relate to protection against natural phenomena, missile } s protection, safe handling of the spent fuel cask and the guidelines of RG 1.13, )l)) Positions C.1, C.3, C.5 and C.6, and 1.29, Position C.2 with respect to over-ff head crane interlocks and maintaining plant safety in a seismic event. The - Q staff'cannot conclude' that the system is acceptable with respect to NUREG-0612 issues. The staff further concludes that the syst .s design as described above {} g meets the CESSAR interface requirements. The overhead heavy load handling sys- [ temmeet,stheacceptancecriteriaofSRPSyction9.1.5,exceptasnotedabove, g The staff will report resolution of the concerns : identified above in the final SER.] y
}. ? 9.2 Water Systems >
i ? 9.2.1 Station Service Water System i #
,9
.(-
h The station service water system (SWS) was reviewed in accordance with Sec- , tion 9.2.1 of the Standard Review Plan (SRP), NUREG-0800. An audit review'of , i each of the areas listed in the " Areas of Review" portion of the SRP section was performed according to the guidelines provided in the " Review Procedures" Fj !
- u' .
WNP-3 DSER SEC 9 a 9-14 g' pt [*_
- _.* NN 'I, TIT @ % DINE % Y l -4 *D 'f'?* ' N I '@ @ E * @E N TN Wk 'V '**W N T4INM iINN N %" "IIDFW- 'k '
YTY N" Y #T #
ana _ - .a._ w -
- - . . . . . _ . = . . . _ .a_. . a i
1 I - i portion of the SRP section. Conformance with the acceptance criteria formed jl the basis for staff's evaluation of the SWS with respect to the applicable l
] regulations of 10 CFR Part 50.
The nonsafety-related (nonseismic Category I, Quality Group D) SWS consisting of two subsystems, the turbine building subsystec and the component cooling water system (CCWS) heat exchanger subsystem,'provides cooling water for removal of heat from plan't auxiliary equipment associated with the power conversion system and the CCWS heat exchangers during normal plant operation. See Sec- -fj tion 9.2.2.1 of this SER for discussion of the CCWS. The SWS also supplies
-l cooling water to the Integrated Leak Test system compressors during plant shut-
]I down. Service water drawn from the closed loop circulating water system (CWS),
i after absorption of heat, is discharged back into the CWS for cooling and recy-h cling as part of the CWS cooling tower recirculation load. See Section 10.4.5
.l of this SER for further discussion of the CWS.
1
,! The SWS for WNP-3 consists of two 100 percent capacity trains each with a par-allel mounted centrifugal pump. These pumps take suction from the condenser inlet to provide cooling water to the plant auxiliaries mentioned above via the l turbine building subsystem. The CCWS heat exchangers cooling subsystem is .! supplied with wator directly from the CWS without additional pumping by the SWS ~q 3 pumps. During plant shutdown or loss of main circulating water pumps, water is _
ll supplied to the SWS pumps and the CCWS heat exchangers by a motor driven auxil-j iary circulating water pump. The SWS utilizes chlorinated circulating water, j for control of slime, algae and bacterial growth in order to reduce organic l fouling of the SWS equipment and the cooled components. The SWS was evaluated and found to provide no functions necessary for achieving cf safe reactor shutdown conditions or for accident prevention or mitigation. ] Failure of the system will not affect safcty-related system functions. Specif-
.j ically, failure of the system will not affect the safety-related CCWS, since
} .the SWS connection to the CCWS heat exchangers can be manually isolated by . safety-related gate valves. Additionally, during an accident, if the SWS be- ,j ; comes unavailable, decay heat removal will be accomplished directly from the ] CCWS to the ultimate heat sink (see Sections 9.2.2.1 and 9.2.5 of this SER for
~
'l WNP-3 DSER SEC 9 9-15 d q
- n
......--.~...v._.-..ra_.--<a....- -. . . - . . . Lw a:O Oi 4
1 J . 1 J' further discussion). Protection from flooding for safety-related systems due to failure of the SWS has been provided (see Section 10.4.5 of this SER for 1 further discussion). Thus, the requirements of General Design Criterion 1 (GDC) 2 " Design Bases for Protection Against Natural Phenomena," and the ._) guidelines of Regulatory Guide (RG) 1.29, " Seismic Design Classification," d Position C.2 are met. ^ Because the SWS is not shared with another unit and serves no safety function, ^ the requirements of GDC 4, " Environmental and Missile Design Bases," GDC 5,
" Sharing of Structures, Systems, and Components," GDC 44, " Cooling Water,"
. GDC 45, " Inspection of Cooling Water System," and GDC 46, " Testing of Cooling
] Water System" are not applicable. '
au
.! Based on the above, the staff concludes that the station service water system l
i meets the requirements of GDC 2 with respect to the need for protection against natural phenomena and the guidelines of RG 1.29, Position C.2.with respect to
, seismic classification, and is, therefore, acceptable. The system meets the H applicable acceptance criteria of SPR Section 9.2.1.
I 1 i 9.2.2 Reactor Auxiliary Cooling Water Systems - The cooling water systems for reactor auxiliaries were reviewed in accordance - J with Section 9.2.2 of the Standard Review Plan (SRP), NUREG-0800. An audit review of each of the areas listed in the " Areas of Review" portion of the SRP j section was performed, according to the guidelines provided in the " Review Pro-d cedures" portion of the SRP section. Conformance with the acceptance criteria g formed the basis for staff's evaluation of the cooling water systems for reac-tor auxiliaries with respect to the applicable regulations of 10 CFR 50. 2 The reactor auxiliary cooling water systems consist of the component cooling
? water system (CCWS) and the essential and nonessential chilled water systems.
E These systems are used to provide cooling water for heat removal from reactor plant components. Portions of the CCWS and the chilled water system are 7 safety related. l3 1 . WNP-3 DSER SEC'9 9-16 ,j
-t 1
-1 _ _____ - __
4 , 1
- nu : - . _ ;MJ. 2 _ L _. _wu 6&J' G; L,i.._ J ,; u , . _L:JJ52Gj:;Q l l
4 i
~i 9.2.2.1 Component Cooling Water System (FSAR Section 9.2.2)
The component cooling water system (CCWS) is a closed loop cooling water system that removes heat from the reactor auxiliaries which it serves during power
- l4 operation and during normal and emergency shutdown. Heat is rejected from the CCWS either directly to the atmosphere via the dry cooling towers (the ultimate
.9 ,d heat sink), or indirectly to the atmosphere via the CCWS heat exchangers and y ~ the circulating water system when the service water system is available (see >g Sections 9.2.1, 9.2.5 and 10.4.5 of this SER for further discussion). .n q% The CCWS consists of two separate, redundant, essential (safety-related) full b capacity closed cooling trains which serve redundant safety-related components. 1 -6 Each train consists of an essential supply header system, two 50 percent capac-j f ty pumps, one full capacity heat exchanger, one 100 percent capacity dry cool- I '] ing tower, one surge tank, and associated piping, valves, controls and instrumentation. Each essential supply header system has the capability of q p supplying the cooling water for essential components during normal plant opera-q tion, normal shutdown and post-accident conditions. Nonessential (nonsafety- ! ,j related) headers which serve nonessential loads are connected to both the essential headers during normal operation. A nonessential, normally isolated 3 chemical addition tank connected to the nonessential header is also provided 1 Q for maintaining proper CCWS water quality. During normal operation, component - cooling water is pumped by the CCWS pumps (one per train) through the load, the dry cooling towers, the shell side of the CCWS heat exchangers and back to the [ pumps. The surge tank provided on the suction side of the CCWS pumps of each q train automatically accommodates expansion and contraction due to system ther-mal changes and provides adequate NPSH. On annunciation of a low level alarm !IC in the control room, makeup water to the surge tank is provided from the seis-mic Category I condensate storage tank by the condensate transfer pumps via a '] manually opened valve (normally closed) on the CCWS makeup supply line. A safety j injection actuation signal (SIAS) or a rapid rate of surge tank water level - j.;l decrease signal, such as would result from a moderate energy line crack or l} failure in the nonessential portion of the CCWS automatically isolates by valve ] closures the nonessential supply headers from each of the two essential supply 3 headers and also the essential supply headers from each other. The system is 1 WNP-3 DSER SEC 9 9-17 f a
..i Y a - - -.w.w... . - a ~~ a wa. .:.... LL . a~ne-l
) designed such that no single valve failing to close causes simultaneous loss of both the essential CCWS trains. Each essential train is provided with a con-
'l tinually operating safety-related radiation monitor in the return header of the CCWS pumps which alarms in the control room on detection of high radiation. i l The CCWS supplies cooling water to safety-related and nonsafety-related reactor
. , auxiliaries during normal operation and to safety-related components during - postulated accident and emergency conditions. The safety-related components il served by the CCWS are the shutdown heat exchangers, containment spray pumps d
j and motor air coolers, diesel generator coolers, HVAC water chillers and essen- ~l tial chilled water system water chiller condensers. The nonsafety-related com-
.1 -
c.] ponents served by the system include the fuel pool cooling heat exchangers, l floor drain waste evaporator, letdown heat exchanger, chemical . addition tank,
~l boric acid concentrator, volume reduction package, the four reactor coolant 1
l pumps (RCP) seal, oil and motor coolers, steam generator blowdown heat exchang-ers, inorganic chemical waste evaporators, gas stripper, waste gas compressors, compressor, primary sample coolers, secondary sample coolers and secondary par-ticulate compressor. ]! -
-I g The' safety-realted components in the CCW system are physically separated and "1 protected from the effects of high-energy line breaks. The safety-related CCWS b components and piping are seismic Category I and Quality Group C and are _
located within seismic Category I structures. The system is capable of with-N standing the effects of extreme wind, flood, tornado and tornado missiles. N Adequate ' isolation is provided between safety-related (seismic Category I) por- ,. tions of the CCWS and nonsafety-related (nonseismic Category I) portions as Q discussed above. Thus, the requirements of General Design Criterion (GDC) 2, 16 " Design Bases for Protection Against Natural Phenomena," and the guidelines of ] Regulatory Guide (RG) 1.29, " Seismic Design Classification," Positions C.1 and L~ C.2 are met. Lj l} The system is designed to meet the single failure criterion with two separate i redundant essential cooling trains to serve those components essential for j achieving safe shutdown. During normal operation, one pump per train will op-f erate which is sufficient to meet the heat removal requirements for the plant i j WNP-3 DSER SEC 9 9-18 4
. _ __ _ ~_m _._ . _ - - . _ . _ . . . _ , , , _ . . , - _ . . _ . . . . , , , . - - -. - - ._._.-
U N - .. m . .... ~ = : ... - , - : ~ =: a L .. x . .n.<.-.- .a-..-..= p -
'l i .]
auxiliaries. During an emergency situation, the SIAS will activate the spare CCWS pumps and both the cooling trains will operate although one cooling train is sufficient to supply cooling water to the needed essential components. Dur-
- 3 ing a normal plant shutdown, both CCWS trains will operate to achieve reactor cooldown within 271s hours. The components of the essential cooling trains are i powered by their respective ESF buses and on loss of offsite power are automat-ically loaded on the diesel generators. During an emergency, the nonsafety-related components of the CCWS are automatically isolated by redundant safety-related (seismic Category I) valves (one per train) powered from separate emer-mj
.m j gency power supplies. Thus, the requirements of GDC 44, " Cooling Water" are satisfied.
i! li Loss of CCW flow to the reactor coolant pumps (RCPs).due to inadvertent closure of the containment isolation valves or single failure in the common supply and i return line can result in a multiple locked rotor condition or unacceptable RCP l seal failure. In response to this concern, the applicant referenced CESSAR j FSAR Section 5.4.1.3 which discusses the CESSAR system 80 RCP test (Topical i Report CENPD-201A). The test results demonstrated satisfactory operation of i i the KSB RCPs for 30 minutes without cooling water flow to the seal coolers and l motor coolers in that the pump bearing temperature remained within prescribed
]
( limits and excessive seal leakage did not occur. The CCWS design includes re-dundant safety grade flow switches, one on each cooling water line to the RCPs - lj which alarm loss of cooling water flow to the RCPs in the control room and ] thereby alerts the operator of this condition. The staff concludes that the 30 minute time period is adequate for the operator to either restore cooling water flow or trip the RCPs, and, therefore the applicant's design meets the staff's position on this issue. The containment supply and return headers are cross connected so that if one
~
CCWS pump fails, the RCPs can be supplied with sufficient cooling water by the
- j unaffected CCWS pumps. The applicant has also demonstrated that the system can
'N withstand a loss of offsite power without damage to RCP pump seals in accor-dance with Item II.K.3.25 of NUREG-0737 since the CCWS pumps are powered from j ~ onsite emergency sources. WNP-3 DSER SEC 9 9-19 L.-= , . . . . . . , .
AL s a d v.., L _v.- w _ __. x. u .wuu w
- 1 1
j During normal operation, the CCWS is in either continuous or intermittent oper-ation. Availability of pumps not running will be ensured by periodic tests and inspection per plant Technical Specifications. The system components are lo-cated in accessible areas to permit inservice inspection as required. Thus, j the requirements of GDC 45, " Inspection of Cooling Water System," and 46, l ] " Testing of Cooling Water System" are satisfied. N 0 . j Based on the above, the staff concludes that the component cooling water system j meets the requirements of GDC 2, 44, 45, and 46 with respect to protection *
?.i against natural phenomena, decay heat removal capability, inservice inspection, j and functional testing, and the guidelines of RG 1.29, Positions C.1 and C.2 j w
. ith respect to the system's seismic classification. The CCWS meets the accep-i tance criteria of SRP Section 9.2.2, and is, therefore, acceptable. The staff
)1 further concludes that the CESSAR interface requirements are satisfied by the above described design.
e 9.2.2.2 Essential and Nonessential Chilled Water Systems (FSAR Section 9.2.9) 1 The essential chilled water system (ECWS) is a closed loop system that trans-fers heat from essential air cooling units in the Reactor Auxiliary Building (RAB) and the control room to the component cooling water systems (CCWS) via the condensers of the ECWS chillers. - The ECWS consists of two independent 100 percent capacity essential trains, each having a packaged water chiller, expansion tank, chilled water pump, independent chilled water distribution loop, and associated piping valves and
. instrumentation. The supply and return headers of the essential trains are also connected to the nonessential chilled water system (NECWS) via normally closed safety related motor-operated isolation valves. During normal and acci-dent conditions, the system provides chilled water to the redundant safety-t related cooling coils of the air handling units located in the control room, -
and electrical equipment and battery rooms. Additionally, during normal and
! accident conditions, the system provides chilled water to the redundant safety-j related unit coolers located in the ECCS area rooms as needed. The ECCS area rooms include the HPCI pump rooms, LPCI pump rooms, containment spray pump 4
~
2
] WNP-3 DSER SEC 9 9-20 i-m _ .m . , . . , . ...-.-...,,..--_,._._m.._ , ~ ...-. , _ - ~ _ _ _ .- - -. _ ._ _
.v ._ww. ~ . N'
..=. ~.u 2 . x - =;
.1-rooms, auxiliary feedwater pump rooms, CCWS pump and heat exchanger rooms, ECCS shutdown heat exchanger rooms and the ECCS penetration area rooms.
4 The ECWS is located in a seismic Category I, missile, flood and tornado pro-tected structure (i.e., the RA8). The components of the ECWS including its distribution piping are designed to seismic Category I, Quality Group C re-quirements. Adequate isolation is provided between the seismic Category I ECWS and interconnections to the nonseismic Category I NECWS as described below. 3 Makeup water is available to the system from the seismic Category I condensate storage tank. Thus, the ECWS meets the requirements of General Design
-4 Criterion 2, " Design Bases for Protection Against Natural Phenomena" and the l ]
guidelines of Regulatory Guide 1.29, " Seismic Design Classifications," Positions C.1 and C.2. One train of the ECWS normally operates with the other on standby. Each train d is powered from the emergency bus associated with the equipment it cools. A
, containment isolation actuation signal automatically starts both the trains of the ECWS, isolates each train from the other and also from the NECWS by ensur-ing closure of the safety-related isolation valves if they are not already in a closed position. Additionally, an abnormal rate of level charge in the ECWS expansion tanks will cause isolation of the ECWS from the NECWS such as may result from a break in a nonseismic Category I line. Both the essential trains _
!] also start following a loss of offsite power. Since either train of the ECWS can supply adequate chilled water to the safety-related cooling coils of the air handling units and the applicable unit coolers in the ECCS area during power. operation, or to achieve safe shutdown following an accident, the system p meets the single failure criterion. As mentioned above the nonessential chilled water system is normally isolated from the essential trains of the ECWS ? by seismic Category I isolation valves. [However, FSAR Section 9.2.2, "Compo-nent Cooling Water System," states that both the CCWS subsystems (i.e., one ,y CCWS pump per subsystem) will be operating during normal operation including j . normal shutdown. Further, FSAR Section 9.4.3, " Auxiliary and Radwaste Area , Ventilation System," states that the unit coolers located in the CCWS pump and ,] heat exchanger rooms are interlocked with their associated pumps and their .! cooling water supply from the associated ECWS subsystems when these pumps are WNP-3 DSER SEC 9 9-21 _ , _ , _ -.__m,.,_,._-._...-,_ . . _ - . . . - - , _. . _ _ _ . - ._ - _ ._.
~ - .. . . - -. --
- .. a n : w w . m u a ~ . _ :..._u ..a. w._..i.=z_.w_:a i
i - 3 operating. These statements appear to be in confifct since it is not clear how cooling can be provided to both CCWS pump rooms when only one ECWS train is in ! i 4 operation and the other is in a standby condition during normal operation as n .! stated in FSAR Section 9.2.9. The staff is, therefore, unable to conclude that l the design of the ECWS meets the requirements of General Design Criterien 44,
" Cooling Water." The staff will report resolution of the above concern in the final SER.]
- q -
f j During normal operation, all portions of at least one train of the ECWS are in ]1 . continuous or intermittent operation, and the operating train can be alternated between the two trains to equalize running time. Availability of the ECWS will
'i be ensured by periodic tests and inspection per plant Technical Specifications.
j The system components are accessible to permit periodic inservice inspection as 1 required. Thus, the ECWS meets the requirements of General Design Criterion 45,
" Inspection of Cooling Water System," and 46, " Testing of Cooling Water System."
ii 1 The nonessential chilled water system (NECWS) is a closed loop system which consists of three nonessential chillers and its own distribution system. The NECWS supplies chilled water during normal operation to the nonsafety-related i containment fan coolers, control element drive mechanism (CEDM) fan coolers, if radwaste and solid waste control room air handling units, and the steam tunnel l air handling units. Additionally, during normal operation, the operator can -- ]1 provide chilled water from the NECWS chillers to the safety-related coolers in d the RAB by opening the normally closed seismic Category I isolation valves
] between the NECWS and the ECWS. Further, in event of a loss of offsite power,
{l the ECWS can be used to supply cooling to the non-safety related containment 3 fan coolers, and CEDM coolers normally supplied by the NECWS by opening the J normally closed interconnecting isolation valves. Since the system is not nec- .] , D. i essary for plant shutdown or to prevent the release of radioactive material, it 1 is not safety related and is not designed to seismic Category I requirements. Thus, the requirements of General Design Criteria 44, 45 and 46 are not appli-i.] cable to the NECWS. l. ,j [ Based on the above, the staff concludes that the essential chilled water sys-O tem meets the requirements of General Design Criterion 2 regarding protection 1i i4 - t .
.1
~
i WNP-3 DSER SEC 9 9-22
. . _ _ . _ . _ . _ ..-_.m.,_,__.-_, __ . . _ - . . _ - , ~ _ _ -. _ ._
-..=_= w h ~ a - - . . - . - - . . - . .-. . - -. :-...
i
'l i
, against natural phenomena, and the guidelines of Regulatory Guide 1.29, j Position C.1 and C.2 as applicable. The staff also concludes that except as
}
noted above, the ECWS meets the requirements of General Design Criteria 44, 45 l
} and 46 as they relate to cooling water system design, periodic inspection, and l l testint1 The staff, therefore, concludes that except as noted above, the es- ]j sential chilled water system meets the acceptance criteria of SRP Section 9.2.2.
The staff will report resolution of the concern identified above in the final
).; SER.]
1
.1 9.2.3 Demineralized Water Makeup System 1
l l The demineralized water makeup system (DWMS) was reviewed in accordance with
.. j. Section 9.2.3 of the Standard Review Plan (SRP), NUREG-0800. An audit review j of each of the areas listed in the " Areas of Review portion of the SRP section
,j . was performed according to the guidelines provided in the " Review Procedures" l portion of the SRP section. Conformance with the acceptance criteria formed the basis for our evaluation of the demineralized water makeup system with re-
'i spect to the applicable regulations of 10 CFR Part 50.
The nonsafety-related (Quality Group D, nonseismic Category I) DWMS consisting of two independent but interconnected demineralizer trains processes raw water from the nonsafety related plant makeup water system (MWS) to provide a contin- ._ uous source of demineralized water to the demineralized water storage tank
') (DWST) and the deionized water storage tank. The DWST, in turn, provides makeup water to the reactor makeup water storage tank, refueling water storage j] tank, and the condensate storage tank (CST). Additionally, the DWMS provides 3 demineralized water for miscellaneous uses such as decontamination stations, -l '
laboratories and washing of floors. j The system is located in the water treatment building, while the demineralized j water storage tank and associated valves and piping are located in the yard. The system performs no safety-related function. Protection from flooding for , safety-related equipment resulting from failure of the DWMS is discussed in Section 9.3.3 of this SER. Adequate isolation is provided at DWMS connections
.1 I
L: 1: t d WNP-3 DSER SEC 9 9-23
O i: - L.cc _,z- .. _ ._. _ _ _ . _ - __ _ n. ._ . z. . . _. . a . -- :_- _, a i j to safety-related systems. The system is capable of fulfilling the normal op- l
, erating requirements of the plant for acceptable makeup water quality and quan-tity, which includes the makeup to the safety-related CST as mentioned above.
f Refer to Section 9.2.6 for further discussion on makeup to the CST. The system j can be remote manually, semi-automatically, or fully automatically operated as j needed from a local control panel in the water treatment building. The system
] includes check valves to prevent contamination of the DWMS at each point of dis-
-] -4 charge from the system by backflow from the systems which it supplies. The sys-tem includes local and control room instrumentation and alarms which annunciate
< abnormal conditions in order to prevent delivery of off-specification water.
.j Failure of the DWMS does not affect any safety-related equipment or the capa-bility to safely shut down the plant. Thus, the requirements of General Design Criterion 2, " Design Bases for Protection Against Natural Phenomena," and the j guidelines of Regulatory Guide (RG) 1.29, " Seismic Design Classification," Posi-tion C.2 are met. i
.I Based on our review, we conclude that the DWMS meets the requirements of GDC 2 4
with respect to the need for protection against natural phenomena as its fail-ure does not affect safety system functions, and meets the guidance of RG 1.29, Position C.2 concerning its seismic classification and is, therefore, accept-able. The demineralized water makeup system meets the acceptance criteria of SRP Section 9.2.3. - 9.2.4 Potable and Sanitary Water Systems The potable and sanitary water systems were reviewed in accordance with Section 9.2.4 of the Standard Review Plan (SRP), NUREG-0800. An audit review 1 of each of the areas listed in the " Areas of Review" portion of the SRP section was performed according to the guidelines provided in the " Review Procedures" portion of the SRP section. Conformance with the acceptance criteria' formed 4 the basis for staff's evaluation of the potable and sanitary water systems with-respect to the applicable regulations of 10 CFR Part 50. j y J J M WNP-3 DSER SEC 9 9-24 L -. -- - - - - - - --
e a a- =- - . ~ = ~ ~ = " * " " i d i-j The nonsafety-related (Quality Group D and nonseismic Category I) potable and sanitary water systems are designed to supply treated water for human consump-1 tion and sanitary purposes. The source of potable water for use by plant per-sonnel is the plant makeup water pre-treated with sodium hypochlorite. The
! plant makeup water is supplied by the groundwater collection well system. The .- l j potable and sanitary water systems are not connected to any system which is a
- '} potential source of radiological contamination. Additionally, the piping de-
] sign for the system includes check valves which will limit backflow of impuri-J ties into the system. Thus, the requirements of General Design Criterion i (GDC) 60 with regard to prevention of inadvertent contamination from radioactive
.l
.i material are met.
Protection from flooding for safety-related equipment resulting from failure of h the potable and sanitary water systems is discussed in Section 9.3.3 of this
-! SER. The systems serve no safety function, and the failure of the systems do not affect plant safety.
4 Based on the above, the staff concludes that the potable and sanitary water systems meet the requirements of GDC 60 with respect to prevention of inadver-l tent contamination from radioactive material and are, therefore acceptable. j The potable and sanitary water systems meet the acceptance criteria of SRP Sec-4
] tion 9.2.4.
1 9.2.5 Ultimate Heat Sink u
- ] The ultimate heat sink (UHS) was reviewed in accordance with Section 9.2.5 of l;j the Standard Review Plan (SRP), NUREG-0800. An audit review of each of the h' areas listed in the " Areas of Review" portion of the SRP section was performed
. according.to the guidelines provided in the " Review Procedures" portion of the
, 4 SRP section. Conformance with the acceptance criteria formed the basis for our evaluation of the UHS with respect to the applicable regulations of 10 CFR Part 50.
'i i
i I I 1 . 1 I l i WNP-3 DSER SEC 9 9-25 l
} ' t. . . . _ _ _ . . . . --- - -
7 - - - - . 7 , - l
...- - -, w .=- . --.:- . .: .
- --. - w .. u. .. .w 2 -.8.ea
'l f
- I The UHS provides heat rejection from the component cooling water system (CCWS, described in Section 9.2.2.1 of this SER) to the atmosphere during normal oper-j ation, plant shutdown and accident conditions. During normal operation and j refueling, the system operates in conjunction with the CCWS heat, exchangers j which reject heat through the service water system (SWS) and circulating water
'l system (CWS) to the atmosphere via a natural draft wet cooling tower. See 1
; Sections 9.2.1 and 10.4.5 of this SER for further discussion. During accident
] conditions, however, the UHS can maintain the CCWS at design temperatures by l
sufficient heat rejection without the need for supplemental cooling by the CCWS heat exchangers utilizing the SWS. i The UHS consists of two independent 100 percent capacity trains. Each train t has a cooling tower of the air-cooled heat exchanger type (dry. cooling tower) with the CCWS water flowing through the tube side and air passing over the ex-
} terior extended surface of the tubes. The system is designed to seismic j Category I, Quality Group C requirements. Both the trains are separated by a i concrete wall and by their respective electrical equipment rooms. Each cooling
.t . j tower train is divided into 10 cells. Each cell is separated by a concrete { wall and includes tube bundles, heat retention doors and induced draft fans which are controlled as needed to maintain proper CCWS temperature. Concrete
] structures provide seismic Category I structural support as well as missile j protection for the heat transfer equipment. As stated in Section 3.5.2 of this -
] SER, the two dry cooling towers are enclosed in structures designed to prevent
, tornado missile impact damage. Additionally, the cooling tower fans are pro-q tected form tornado generated missiles by missile grating.
.; 1
,l The UHS is designed to withstand severe natural phenomena and site related j events without loss of safety function. The design of the dry cooling towers ;
provides protection against earthquake (SSE), extreme winds and tornado load- '
! ings, flood, extreme temperatures, and volcanic ash fall. The system is also i separated from high-energy piping systems. The cooling tower fans are powered from redundant Class IE essential buses. Thus, the requirements of General y Design Criterion (GOC) 2, " Design Bases for Protection Against Natural Phenomena,"
i and the guidelines of Regulatory Guide (RG) 1.27, " Ultimate Heat Sink for i Nuclear' Power Plants," Positions C.2 and C.3, regarding UHS protection against 4 1 A L WNP-3 DSER SEC 9 9-26 !^ l l , . . . _ . . _ _ . _ . ._. . .
)
u - - 2 m a. - - a _. ~ s . - 1 .2 u _.co r 3 natural phenomena and RG 1.29, " Seismic Design Classification," Position C.1,
, regarding seismic design classification, are met.
)
4
) The UHS is designed to remove sufficient decay heat directly from the CCWS un-
'd- > der the worst postulated site meteorological condition, i.e., maximum ambient
- 1
] dry bulb temperature, for achieving a safe shutdown following a design basis 1 accident (LOCA) and loss of offsite power assuming a coincident single failure.
See Section 2.4 of this SER for further discussion on site meteorology. The
;j applicant has used BTP ASB 9-2 assumptions to calculate the heat input to the UHS due to fission product and heavy element decay following a design basis ac-i cident (LOCA). The applicant's analysis showed that the maximum heat rejection
,d rate requirement by the UHS occurs approximately 4 hours after the design basis accident (LOCA) occurs and that thereafter the heat rejection rate requirement decreases as the reactor is shut down. Each dry cooling tower is sized to re-ject heat at a rate in excess of that required under the worst site meteorologi-cal condition. Therefore, the above analysis demonstrates that a single. tower . ]y is sufficient to assure post accident shutdown for at least 30 days, and main-
'{ tain the CCWS temperature at its design value without the need for supplemental
{ cooling. Makeup water to the CCWS to account for normal system losses due to
.- l leakage is provided from the seismic Category I condensate storage tank. Based i on the above, the staff concludes that the UHS is capable of providing sufficient cooling for at least 30 days under normal shutdown and accident conditions and
[l _ q maintaining the CCWS temperature within design limits. Thus, the requirements ] of GDC 44, " Cooling Water," and the guidelines of RG 1.27, Positions C.1, C.2 j and C.3 regarding the UHS' ability to remove sufficient heat and maintain the CCWS at proper temperature under all modes of plant operation are met. 1 k The dry cooling tower fans will be periodically tested in accordance with plant ] Technical Specifications. The UHS is accessible to permit inservice inspection (! I as required. Thus, the requirements of GDC 45, " Inspection of Cooling Water System," and 46, " Testing of Cooling Water System," are met.
- 4
']:j_ Based on the above, the staff concludes that the UHS meets the requirements of I GDC 2, 44, 45, and 46 with respect to protection against natural phenomena, l decay heat removal capability, inservice inspection and functional testing, and l - ! WNP-3 DSER SEC 9 9-27 t j l
~ ,._ 7 y -, ..
-...--.~...,7.- , .
. ... , j y . ; .7 .
_.t -.7 - ;. -~ , q ), .;., g7 ,
A.N.:- =.. x~.A:-..w-- = = -.-. . . .A . w=~a= I q the guidelines of RG 1.29, Position C.1 and RG 1.27, Positions C.1, C.2, and j C.3 and BTP ASB 9-2 with respect-to design capability, seismic classification, and capability to remove sufficient decay heat to maintain plant safety and is,
! therefore, acceptable. The UHS ~ meets the acceptance criteria of SRP Section 9.2.5.
- 9.3 Process Auxiliaries ,
( 9.3.1 Compressed Air System i 3 9.3.2 Process and Post-Accident Sampling System j A. Process Sampling System
.j Introduction i
~
The process sampling system is designed to provide representative liquid and gaseous samples drawn from the primary and secondary coolant systems, the associated auxiliary system process streams, and the spent fuel pool cleanup system. Provisions are made to assure that representative samples are obtained ,] from well mixed streams or volumes of effluent by the selection of proper d sampling procedures. In the event of an accident, all sample lines which pass through the containment are automatically isolated by two fail-closed solenoid _ operated valves. Evaluation The information provided by the applicant has been reviewed in accordance with Section 9.3.2 of the Standard Review Plan (NUREG-0800, July 1981) i The process sampling system includes piping and other components associated j with the system from the point of sample withdrawal from a fluid system up to y the analyzing station, sampling station, or local sampling point. Our review j included the provisions proposed to sample all principal fluid process streams
! associated with plant operation and the applicant's proposed design of these 1
s
, WNP-3 DSER SEC 9 9-28 4
lw ...,,-.,..m.c- s .*a- <=*-.w.e e m -+ + ww < --v " * ~
* * ' " * ' " ' ' '< 't
"*TY"
..----:. -~ .-.2 - ~ .. . -. . .. = .: = ~
b i z systems, including the location of sampling points, as shown on piping and instrumentation diagrams. t We determined that the proposed process sampling system meets (1) the require- l 3 ments of General Design Criterion 13 in Appendix A to 10 CFR Part 50 to monitor
.j variables that can affect the fission process for normal operation, anticipated operational occurrences, and accident conditions, by sampling the reactor cool-i j'
ant, the ECCS core flooding tank, the refueling water storage tank, the boric acid mix tank, and the boron injection tank for baron concentration; (2) the j requirements of General Design Criterion 14 in Appendix A to 10 CFR Part 50, to j assure a low probability of abnormal leakage, rapidly propagating failure, and
! gross rupture, by sampling the reactor coolant and the secondary coolant for chemical impurities that can affect the reactor coolant pressure boundary mate- 'l rial integrity; (3) the requirements of General Design Criterion 26 in Appen-i dix A to 10 CFR Part 50 to control the rate of reactivity changes, by sampling the reactor coolant, the refueling water storage tank, and the boric acid mix tank for boron concentration; and (4) the requirements of General Design Crite-ria 63 and 64 in Appendix A to 10 CFR Part 50 to monitor for radioactivity that may be released from normal operations, including anticipated operational occur-j rences, and from postulated accidents, by sampling the reactor coolant, the i
l pressurizer tank, the steam generator blowdown, the secondary coolant condens-3 ate treatment waste, the sump inside containment, the containment atmosphere, _ i the spent fuel pool, the gaseous radwaste storage tank for radioactivity, and the CESSAR interface requirements discussed in the CESSAR SER. i We further determined that the proposed process sampling system meets (a) the il standards of ANSI N13.11969 for obtaining airborne radioactive samples; (b) the requirements of 10 CFR Part 20, 20.1(c) and regulatory positions 2.d(2), 2.f(3), -] 2.f(8), and 2.i(6) of Regulatory Guide 8.8, Revision 3, "Information Relevant j to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations j Will Be As Low As Is Reasonably Achievable," to maintain radiation exposures to- { as low as is reasonably achievable, by providing (1) ventilation systems and
.] gaseous radwaste treatment system to contain airborne radioactive materials; i (2) liquid radwaste treatment system to contain radioactive material in fluids; 1 (3) spent fuel Pool cleanup system to remove radioactive contaminants in the i
Ii WNP-3 DSER SEC 9 9-29 ,1 1- . , . . . - - , . ..
, _ , , ~ .
, - z. - _ . . .2.. w . 2.- : = : a = a. aw . ~.x.. u.a . w a l.
Ei 1 -
)
.] spent fuel pool water; and (4) remotely operated containment isolation valves 'i to limit reactor coolant loss in the event of rupture of a sampling line; (c) the requirements of General Design Criterion 50 in Appendix A to 10 CFR Part 50 j to control the release of radioactive materials to the environment by providing il - isolation valves that will fail in the closed position; and (d) regulatory posi-1 tions C.1, C.2, and C.3 of Regulatory Guide 1.26, Revision 3, " Quality Group Classifications and Standards for Water-Steam, and Radioactive-Waste-Containing j Components'of Nuclear Power Plants," and C.1, C.2, C.3, and C.4 of Regulatory Guide 1.29, Revision 3, " Seismic Design Classification," by designing the sampl-g ing lines and components of the process sampling system to conform to the clas-i] sification of the system to which each sampling line and component is connected, fj and thus meets the quality standards requirements of General Design Criterion 1
.d and the seismic requirements of General Design Criterion 2.
' ]{ i} Conclusion On the basis of the above evaluation, we conclude that the proposed process i 'q samp'ing system meets the relevant requirements of 10 CFR Part 20, 9 20.1(c), ' j> General Design Criteria 1, 2, 13, 14, 26, 60, 63, and 64 in Appendix A to j 10 CFR Part 50, and the appropriate sections in Regulatory Guides 8.8, j 1.26 and 1.29 and therefore is acceptable. B. Post-Accident Sampling System (NUREG-0737, II.B.3)
-i y Introduction
,] j Subsequent to the TMI-2 incident, the need was recognized for an improved post-accident sampling system (PASS) to determine the extent of core degradation '.4I j following a severe reactor accident. Criteria for an acceptable sampling and , 91 anaiysis system are specified in NuREG-0737, Item II.s.3. The system should q~ have the capability to obtain and quantitatively analyze reactor coolant and containment atmosphere samples without radiation exposure,to aoy individual ] exceeding 5 rem to the whole body or 75 rem to the extremities (GDC-19) during i 3 and following an accident in which there is core degradation. Materials to be i analyzed'and quantified include certain radionuclides that are indicators of
~
WNP-3 DSER SEC 9 9-30 i.I
- )
i-
- . _ ..=.~.=m_,.. _
= _. -=_m
._ ..--.-.u,.-, .. . -.-.a.- . .:.-, . - . . - . . . . .w. au :. .a l
severity of core damage (e.g., noble gases,-isotopes of iodine and cesium, and
.j nonvolatile isotopes), hydrogen in the containment atmosphere and total dis-j solved gases or hydrogen, boron, and chloride in reactor coolant samples.
I 4 1 To comply with NUREG-0737, Item II.B.3, the applicant should (1) review and mod-l ify his sampling, chemical analysis, and radionuclide determination capabilities 4 as necessary and (2) provide the staff with information pertaining to system i design, analytical capabilities and procedures in sufficient detail to demon-i strate that the criteria are met.
- ?
Evaluation i By letter dated April 1, 1983, the applicant provided information on the PASS.
, Criterion (1):
f The applicant shall have the capability to promptly obtain reactor coolant samples and containment atmosphere samples. The combined time allotted for sampling and analysis should be three hours or less from the time a decision is made to take a sample.
; The applicant has designated that the static uninterruptable power supply be _
I available for the post accident sampling system in the event of a loss of off-
, site power. Information was not provided on the capability to obtain and j analyze reactor coolant and containment atmosphere samples within three hours .j from the time a decision is made to take a sample. We find the applicant 7j partially meets Criterion (1).
I Criterion (2): , N j The applicant shall establish an onsite radiological and chemical 7i analysis capability to provide, within the three hour time frame j established above, quantification of the following: i l 1 I
. WNP-3 DSER SEC 9 9-31 i.3 i L ____ _ _ _ . _____._ _ _- _ . - -- - -.. -
a u N.=-.. .. w.a. - = a -. :- --- w s =- :---
" - A"=
4 1 'j (a) Certain radionuclides in the reactor coolant and containment j atmosphere that may be indicators of the degree of core i damage (e.g., noble gases, fodines and cesiums and non-volatile isotopes); 9] I
- t y
(b) hydrogen levels in the containment atmosphere; d ] .n (c) dissolved gases (e.g., H2 ), chloride (time allotted for f] analysis subject to discussion below), and boron concen-
. tration of liquids; 1
(d) alternatively, have inline monitoring capabilities to perform q all or part of the above analyses. a1 ,j The PASS provides diluted and undiluted liquid and gaseous samples for grab sample analysis, and'inline monitoring of hydrogen in the containment atmosphere. ij We find that these provisions partially meet Criterion (2). The applicant should provide a procedure to estimate the extent of core damage based on radio-j nuclide concentrations and taking into consideration other physical parameters such as core temperature data and sample location. Also, information on an j;i onsite radiological and chemical analysis capability must be provided. _. g]j Criterion (3):
.i 2
.]
a-Reactor coolant and containment atmosphere sampling during post-a accident conditions shall not require an isolated auxiliary system
].i (e.g., the letdown system, reactor water cleanup system) to be ^1 placed in operation in order to use the sampling system.
k.l ., j, Reactor coolant and containment atmosphere sampling during postaccident condi-tions does not require an isolated auxiliary system to be placed in operation
] ,
in order to perform the sampling function. The applicant's proposal to meet
'j Criterion (3) is acceptable since PASS sampling is performed without requiring -i; -
y
'{ .
q
~
WNP-3 DSER SEC 9 9-32 3 'i
. , n, n. -- - - 7- 3_ _ . 3 ;; ; ;~; 3 _; ~ -
. A . i 2 _ _ _. a - e___._i -
--_ _ _ ._; .wo m_la
.j i
i operation of an isolated auxiliary system and all PASS valves which are not ,1, accessible after an accident are environmentally qualified. ' 'a d j Criterion (4): l Pressurized reactor coolant samples are not required if the applicant q] can quantify the amount of dissolved gases with unpressurized reactor j coolant samples. The measurement of either total dissolved gases or j H2 gas in reactor coolant samples is considered adequate. Measuring si the 02 concentration is recommended but is not mandatory. q i l Pressurized reactor coolant samples are cooled and degassed to obtain represent-ative dissolved hydrogen and oxygen samples at the PASS sampling station. We
] have determined that these provisions meet Criterion (4) of Item II.B.3 in
; NUREG-0737 and are, therefore. acceptable.
.i I
1 Criterion (5): 4 nj The time for a chloride analysis to be performed is dependent upon two factors: (a) if the plant's coolant water is seawater or l brackish water, and (b) if there is only a single barrier between
]. primary containment systems and the cooling water. Under both of _ -1 the above conditions the applicant shall provide for a chloride (j analysis within 24 hours of the sample being taken. For all other . .) cases, the applicant shall provide for the analysis to be completed within 4 days. The chloride analysis does not have to be done onsite --4
'j The applicant proposes that chloride analysis will be done at an offsite labora- i j tory to meet the four day requirement. However, to comply with Criterion (5), ij specific arrangements need to be made with the offsite laboratory and a licensed j shipping container needs to be available for transporting the sample. 1
. .j i
1 'd WNP-3 DSER SEC 9 9-33 I b . 1 3 i
- . _ . - - . . , , . g,m p.,.,; . ,, ,. _ y. y. . l
. . . . ~ . - . . . . , . . - , . - . . , v n e ., .. .- -, .
~
. d = __ --
. _ _ _ = _.m.______.__._u__ .- dE.un._ __. ._m. -
+
' i, j Criterion (6):
^1 l
The design basis for plant equipment for reactor coolant and contain-
- ] l j ..
ment atmosphere sampling and analysis must assume that it is possible ' 'a-to obtain and analyze a sample without radiation exposures to any
- .0 individual exceeding the criteria of GDC-19 (Appendix A, 10 CFR 1
?] Part 50) (i.e. , 5 rem whole body, 75 rem extremities). (Note that il the design and operational review criterion was changed from the l
,q operational limits of 10 CFR Part 20 (NUREG-0578) to the GDC-19 lj criterion (October 30, 1979 letter from H. R. Denton to all (.)A licensees.)) ,j
- The applicant has performed a shielding analysis on operator exposure while
,s obtaining and transporting a PASS sample. This evaluation does not meet Criterion (16) which requires a manrem exposure estimate based on person motion study for sampling, transport, and analysis of all required parameters.
Criterion (7): .t The analysis of primary coolant samples for boron is required for ,j PWRs. (Note that Rev. 2 of Regulatory Guide 1.97 specifies the ' -] need for primary coolant boron analysis capability at BWR plants.) _
?
J The applicant will have the capability to analyze reactor coolant samples for
;j boron. Boron accuracy is discussed in Criterion (10). This provision meets .j the recommendations of Regulatory Guide 1.97, Rev. 2 and Criterion (7) and is,
,N therefore, acceptable. ii
-, Criterion (8):
1
.; If inline monitoring is used for any sampling and analytical -
capability specified herein, the applicant shall provide backup sampling through grab samples, and shall demonstrate the capa-bility of analyzing the samples. Established planning for +1 i e ,j WNP-3 DSER SEC 9 9-34
.I L, - ._ ,,.m. .-_ m , _ ;. ._ ,. , . m mm .m m . _ _ _
,.:. _ -. a .- . .... _.. - -. -
=- :x . . . ----- = - - - - - " e u h i
-l 4
I i i j analysis at offsite facilities is acceptable. Equipment pro-vided for backup sampling shall be capable of providing at least one sample per day for 7 days following onset of the accident and at
] least one sample per week until the accident condition no longer exists.
I h Reactor coolant dissolved oxygen and containment. hydrogen concentration measure-i ments will be performed inline. The PASS has the capability of providing '
, backup samples of diluted gas, diluted liquid and undiluted liquid. We find j these provisions meet Criterion (8) and are, therefore, acceptable.
3 Criterien (9): I
- i The applicant's radiological and chemical sample analysis capability a shall include provisions to:
l A d l (a) Identify and quantify the isotopes of the nuclide categories discussed - above to levels corresponding to the source term given in Regulatory
;j Guides 1.3 or 1.4 and 1.7. Where necessary and practicable, the i ability to dilute samples to provide capability for measurement and reduction of Personnel exposure should be provided. Sensitivity of l onsite liquid sample analysis capability should be such as to permit measurement of nuclide concentration in the range from approximately _
1 pCi/g to 10 Ci/g. ,.g a j (b) Restrict background levels of radiation in the radiological and
- 8 chemical analysis facility from sources such that the sample analysis 9
will provide results with an acceptably small error (approximately a q factor of 2). This can be accomplished through the use of sufficient Gj shielding around samples and outside sources, and by the use of a j ventilation system design which will control the presence of airborne radioactivity.
.i i
j The radionuclides in both the primary coolant and the containment atmosphere Lj will ha idanc1fied and quantified. Provisions are available for diluted reactor l1i 1 !j WNP-3 DSER SEC 9 9-35 yy_ .-
Z... i .:._ _ 1 .__ . _w.au '
~
a- <
- i. .-=
..___m__. m._,__z ^
i ; l
- l l
j 1 i coolant samples to minimize personnel exposure. The PASS can perform radioiso- l j tope analyses at the levels corresponding to the source term given in Regulatory l Guides 1.4, Rev. 2, and 1.7. Radiation background levels will be restricted by shielding. We find these provisions partially meet Criterion (9). The appli- . cant should determine whether radiochemical analysis results can be obtained f.1 ,3 _ within an acceptably small error (approximately a factor of 2). Also, informa-il tion on the ventilation system design provisions to control airborne radio-h t c activity should be provided. Criterion (10): 4 cli
.1 Accuracy, ranage, and sensitivity shall be adequate to provide pertinent data to the operator in order to describe radiological and chemical
{ status of the reactor coolant systems.
).l 4
) The applicant has not provided sufficient information for our review to deter-
"J mine compliance with the requirements of Criterion (10).
.d q
Criterion (11): N I In the design of the postaccident sampling and analysis capability,
.. consideration should be given to the following items: t 7 (a) Provisions for purging sample lines, for reducing plateout in sample g ifne, for minimizing sample loss or distortion, for preventing block-
.g age of sample lines by loose material in the RCS or containment, for F appropriate disposal of the samples, and for flow restrictions to
- M
$ limit reactor coolant loss from a rupture of the sample line. The !j postaccident reactor coolant and containment atmosphere samples should j be representative of the reactor coolant in the core area and the containment atmosphere following a transient or accident. The sample-4 lines should be as short as possible to minimize the volume of fluid sj to be taken from containment. The residues of sample collection l
,q 4 should be returned to containment or to a closed system.
.g ) t l
} WNP-3 DSER SEC 9 9-36 l 1 n n~~_,y ,n_ - ,, y-r 3 ;;~ - 7;===:z__,2- ,L I
- N A ..~ N .-. = .- -....- ---- - .- - - - . = - - - - - - - - - - - - - -- -- - x = " L- -1 l l l
s (b) The ventilation exhaust from the sampling station should be filtered j with charcoal adsorbers and high-efficiency particulate air (HEPA)
]; filters.
l a The applicant has addressed provisions for purging to ensure samples are j representative, size of sample line to limit reactor coolant loss from a rupture j of the sample line. and ventilation exhaust from PASS filtered through charcoal j adsorbers and HEPA filters. To limit iodine plateout, the containment air j sample line is heat traced. We determined that these provisions meet Crit-
.! erion (11) of Item II.B.3 of NUREG-0737, and are, therefore, acceptable, q
f 7 Conclusion - 1 q
-1 '1 On the basis of the above evaluation, we conclude that the post-accident
- - sampling system meets five of the eleven criteria of NUREG-0737, Item II.B.3.
Additional information is needed to complete our review of the six remaining criteria: q i j Criterion (1): j Provide information on the capability to promptly obtain and analyze
- reactor coolant and containment atmosphere samples within three hours _
from the time a decision is made to take a sample. j Criterion (2): 1 1
] Provide information on onsite radiological anc. chemical analysis capability.
l Provide a procedure to estimate the extent of core damage. i
] Criterion (5):
j Make specific arrangements for chloride analysis at an offsite laboratoty <4
~i and also for a licensed shipping container. }
e l ,1 WNP-3 DSER SEC 9 9-37 _ ~7 _ _
_ _ . . . _ -- = . ~ . - - - - - - - - - - ~ ~ - - - - - " ' " " l Criterion (6): Perform a personmotion study for man-rem exposures accumulated during
.] sampling, transport, and analysis of all required parameters. !
'l ; Criterion (9):
; 1 l
Discuss whether radiochemical analysis results can be obtained within
; an acceptably small error (approximately a factor of 2). Discuss the j ventilation system design relative to airborne radioactivity control.
J Criterion (10):
- I
{ Discuss the accuracy, range, and sensitivity of the PASS parameters to ensure an adequate description of the radiological and chemical status of the reactor coolant system. 9.3.3 Equipment and Floor Drainage System - s _. . -
, 9.3.4 Chemical and Volume Control System 4
Our evaluation of the Chemical and Volume Control System (CVCS) proposed for -
] use in WNP-3 is presented in Section 9.3.4 of the CESSAR SER. In that evalua- ,l tion, we conclude that the CESSAR CVCS is acceptable, provided the CVCS inter-j face requirements for balance of plant (BOP) are adequate.
CESSAR identifies interface requirements for the CVCS with the 80P in Sec-
.: tion 9.3.4, which include normal and emergency power; protection from natural ] phenomena such as floods, winds, tornadoes, and earthquakes; protection from j pipe failure and missiles; separation of components; thermal ifmitations; 1
t inspection and testing; materials compatibility; system / component arrangements;- I
! radwaste management; overpressure protection; refueling water tank design param-eters; alternate source of borated water from the spent fuel pool; fire protec-tion; operating temperature ranges; environmental control; and mechanical interaction between components.
j WNP-3 DSER SEC 9 9-38 l
v: . _ 5
- " " - ' " ~ ~ " ~ ' " ~'^~
___-.~....=-=a~="--
'i
; 9.5 Other Auxiliary Systems
'i l 9.5.1 Fire Protection Program
.. j
- j -
~
Al Wehave$eviewedthefireprotectionprogramforconformancewithSRPSec-1 tion 9.5.1, Fire Protection. The SRP contains, in BTP CMEB 9.5.1, the technical j requirements of Appendix A.to BTP 9.5.1 and Appendix R to 10 CFR 50. The applicant's Fire Hazards Analysis, transmitted by letter dated October 22, 1982 and amended in Revision 2 of the FSAR, was in response to.the staff's request that he evaluate his fire protection program against the guidelines of 3 Appendix R to BTP ASB 9.5.1, " Guidelines for Fire Protection for Nuclear Power
- .4
<! Plants."
-l
(' The applicant has provided an evaluation of the plant fire protection program against the guidelines of BTP CMED 9.5.1 (NUREG-0800, July 1981 which includes Appendix R to 10 CFR 50. We will require that such an evaluation be performed.
. We have reviewed the automatic and manually operated water fire suppression
- k. systems, the fire detection systems, fire barriers, fire doors and dampers, _
fire protection administrative controls, and the fire brigade size and training. 2 The objective of this review is to ensure that in the event of a fire, personnel 3 and plant equipment would be adequate to safely shutdown the reactor, to main-M tain the plant in a safe shutdown condition, and to minimize the release of h radioactive material in the environment. As part of its reviews, we will visit
.+
, the plant site to examine the relationship of safety-related components, systems, and structures in specific plant. areas to both combustible materials and to
' }t
;i associated fire detection and suppression systems.
Our consultant, Gage Babcock & Associates, Inc., participated in the review of
; the fire protection program.
.i
[ t .' a l l WNP-3 DSER SEC 9 9-39 lQ
~--..-..,- . - - . - . . . , . - , , , - . - - , , . , - - - -. -
.t
' ~
W~_ _ +-- . A=m.- m u:
..j .
) .S n . .
1 s q , j 9. 5.1. 2 Fire Protee. tion Program Requirements ,, ,
- 4 ,
[,
- Fire Prote:: tion *Prograg 1 n % ,
, s, ,
j The fire protection p'rogram esl$6b11shes policy for the protection of structures, a - u. s systems and c'omponents importantV co safety. The applicant has not committed ga to conform to the technical re,uirepats s for fire prctection programs in BTP 3 CMEB 9.5.1, Section C.1. We will require that the applicant commit to conform 0 / t M . to the above guidelines, y p/ i f)
^
[ . + Fire Hazards Analysis %" ,a w e
.a .
, ; D ,,
^ < ,
llrj The applicant pronded a fire hazard analysis with the FSAR. The analyses y identified the fi}e' areas of the plant, and for each fire' area specified the
.'! , combustible materials present, identified safety related systems, determined ,a s
the consequences of a fire od safe shutdown capability, and summarized-available 4 11 . Ii fire protection. Our' evaluabga of the identified fire hazards is contained 1 in the balance of this report.
- WP;
,i*
3 \ 'S.g y . GDC 3, Appendix A to 10 CFf: Part.50 requfies tiat " Fire fighting systems shall
, s y'i bedetsignedtoyssqrethatruptureorinadveg.entgoerationdoesnotsignifi-
[.[ , ], . cantly impair tN5btfetygapabi){ty of those structures, systems and components." v . g 4 To satisfy this requirement the applicant has designed componants required for
']
hot shutdown.so that rupturior' r ' inadvertent operation of fire suppression sys-W
;r tads will not adversely affect the operability of these components. Where 3
N .3 necessary, apg' 1riata protection is provided to prevent impingement of water
'j.g .
spray on cca:oncats , .. rguired. for hot shutdown. Redundant trains of components 1C
~
that are susceptible do damage from yater spray are physically separated so
'.y that manual fire suppression activities will not adversely,. affect the operabil-e ityofcomNnenfo.potfrvolvedinthepostulatedfire. However, we are con-
. . . . . , , n g
t s 3 cernedthatthesghanismb_ywhichQirefkghtinghosestreamssystemsmaycause Q the simultaneous f4 l d eingthedesign.'Weh}iu?.egfredur.dantordiversetrainshavenotbeencon ill(fiquire,thattheapplicantidentifysuchmechanismsthat xt w %- - l b ' wer,e, considered in his firh hazards analysis and. the peasures taken to preclude
.j s 4 pr l.1 h a. ,' ', -
,.] * ~ e WNP-3 DSER SEC 9 9-40 s$'
*e-- e+*ece- .m -- -a av.e = y== sM*-PW'** e--"
+*----^^"*-%."^ ' * " ^ " + " ' " *
- 7d=_
,,,_g_'__,_ __ __,[ ,..__...AU l
. ~ . - . . -w~..-..=_.~.--- . . - . . - - . . ~ . . . . u.:.:. A :i 1
the fire or fire suppressant induced failure of redundant or diverse safety 1 trains. 1 1 j Implementation of Fire Protection Program -- 1 d -1 J The licensee has not indicated when the fire protection program will be opera- ~l i [ tional. We will require that the fire protection program be operational at or before initial fuel loading.
?
95.1.3 Administrative Controls a
;j The administrative controls for fire protection consist of the fire protection ]
El organization, the controls over combustibles and ignition sources, and the pre-fire plans and procedures for fighting fires. In the fire hazards analysis,
.',i the applicant did not provide information for us to independently verify j compliance with our guidelines. We will require that the applicant provide
- ) details showing compliance with the guidelines in BTP CHEB 9.51, Item C.2 1
'~ regarding administrative controls. d
] Fire Brigade and Fire Brigade Training The applicant, has not provided a description of the plant fire brigade, includ- _
ing equipment, to verify the guidelines contained in BTP CMEB 9.51, Item C.3. We will require that the applicant provide details showing compliance with the guidelines in BTP CMEB 9.51, Item C.3 in the establishment and training of the fire brigade. i.t fi;J 9.5.1.4 General Plant Guidelines 3 Building Design Qa y Fire areas are defined by walls and floor / ceiling assemblies that have a fire q resistance rating of 3 hours. Some waTls and floor / ceilings are not fire rated. 4 They are delineated in Amendment 2 of the applicant's fire hazard analysis. We j are concerned that all fire barriers which separate redundant shutdown-related
.1 i
4 Jj WNP-3 DSER SEC 9 9-41 1 1
- , , _ . . ..n . .. n . . .-- ..v. - -
l'
~~
-u - -_
L_ _A?l:2.;.=_L,=.n lnL ,2ikiil.,5E j divisions are not 3-hour fire, rated as determined by the test method identified in ASTM E119. We are also concerned that such barriers are not complete, in that they may contain unprotected openings which may act as an avenue for
, vertical and horizontal fire spread. We will require that all walls and floor / ~
j ceiling assemblies which define fire areas and which separate redundant shut- .
~
, down systems be tested by an independent fire authority and meet the acceptance
.; criteria of ASTM E119 in accordance with Section C.5.a(1) of BTP CMEB 9.51.
- [] ,} s Piping, conduit, cable tray, and bus duct penetrations of fire rated barriers y- are provided with penetration seals that have been successfully tested in
,j accordance with ASTM E119, by an independent laboratory in the configurations Jj which are typical of what is found throughout the plant. Openings inside conduit ,1 are provided with penetration seals either at the barrier or at both ends' of the conduit on either side of the barrier. This configuration conforms with ] Section C.S.a of BTP CMEB 9.51 and is, therefore, acceptable.
.i j Door openings in fire rated barriers are, for the most part, equipped with labeled fire doors. In th'e fire hazards analysis, the applicant indicated that certain multifunction doors, such as bullet proof, watertight and pressure resistive doors are not fire rated but are;of " certified fire resistive [ l construction". Weareconcernedthatthesddoorsmaynotbeabletowithstand 73 anticipated fire exposures in the plant and maintain their integrityc ;We*will-
-o ._
require that the applicant provide compensatory fire protection for these doors a to meet Section C.5.a(5) of BTP CMEB 9.51. ' 3 J.. Ventilation ducts that penetrate all fire barriers are provided with fire '
}; dampers commensurate with the fire rating of the barrier. The fire dampers are Q U.L. labeled and installed according to the manufacturer's directions and NFPA e Standards Nos. 80 and 90A.1
% 3 j- '
,s J( We conclude that the fire dampers will be po side & in accordance with guidelines-of BTP CMEB 9.51, Section .C.5.a(5) and are, therefore, acceptable, p 1 f
b Walls and structural materials throughout the plant are noncombustible. Mate-Li rials that are used as interior finish, including thermal l'nsulation, radiation [ , , a '6
$l WNP-3 DSER SEC 9 9-42
,3 i .i J n --- - .- . -., = - w =- - - . , . . t
=s.<.nw
- .w a .-- . -
a:-.= = u:- u . a. .xL.a _.a. _,w ~
~I
.)
j shielding and sound proofing have all been tested in accordance with the method
.1 i in ASTM E84 in the configuration in which they are found in the plant. They have been demonstrated to exhibit flame spread, smoke and fuel contribution 0;
ratings of 50 or less by an independent laboratory. We find this acceptable. -
.]
q. cj All oil filled transformers are located outside in the transformer yard. ~ j . .I They are separated from safety-related structures by more than 75 feet. The
/3 4
transformers are installed over gravel filled pits, designed to collect all oil
- s
~* $
- from a potential spill. We find this acceptable. The applicant has not pro-
]e q vided information in the fire hazards analysis regarding transformers inside ~l fire areas containing safety-related systems. We will require that such trans-formers be of the dry type or insulated and cooled with noncombustible liquid in accordance with Section C.5.a(12) of BTP CME 8 9.51.: a Safc Shutdown Capability
}
The information provided by the applicant is insufficient to verify compliance j with our guidelines. We will require the applicant to provide a safe shutdown j analysis in accordance with the guidelines of Section C.S.b of BTP CMEB 9 51. i i 1 Alternate Shutdown Capability Aq The applicant has not completed a fire hazards analysis of the consequences N of a fire in areas, such as the control room, which contain redundant shutdown
, systems and which require the provision of an alternate shutdown capability. We M
will require that an alternate shutdown capability comply with the guidelines b contained in Section C.S.c of BTP CMEB 9.51. Control of Combustibles . ,j. E The applicant has stated in the fire hazard analysis that safety-related sys-tems have been isolated or separated from combustible materials to the extent possible, that the use of piastic materials has been minimized, and that stor-a age of flammable materials complies with the requirements of NFPA 30. The staff j finds this acceptable. 5 - l WNP-3 DSER SEC 9 9-43 i
?
- r_-_ r: . _ - - -- -
=- -- -._:- -
- - - ~ ~
~ ' ~
= .- .L r <:.2.2-.. - .. L.G.O
.j l -
The applicant has not provided information in the fire hazards analysis, regard-l ing the routing of hydrogen lines, to enable us to verify compliance with our j guidelines. We will require that hydrogen lines in safety related areas be j either designed to seismic Class I requirements, or sleeved and vented to the ~ j outside, or equipped with excess flow valves in accordance with Section C.5.d(5) , M of BTP CMEB 9.51. c1 sj .d Electrical Cable Construction, Cable Trays, and Penetrations d 1 d Cable trays are constructed of galvanized steel. Fire breaks, consisting of , fire resistive material are provided in horizontal and vertical cable trays )) ~ .i to prevent the propagation of fire. We find this acceptable. 1_ The power, control and instrumentation cables used in the plant has been tested i and has passed the acceptance criteria of IEEE-38374. We find this acceptable. Safety related cable trays outside the cable spreading room are not provided l with continuous line-type heat detectors. The applicant has not justified
- i this deviation.
.q 4 .) In addition, the applicant has not provided sufficient information for us to [ verify compliance with Section C.S.e.(2) of BTP CMEB 9.51. We are concerned _
] that safety related cable trays which are not accessible for manual fire d fighting may be prone to fire damage. We will require that safety related cable trays which are not accessible be protected in accordance with Sec-tion C.5.e of BTP CMEB 9.51.
a.< Q -. Ventilation
+;
l!
-3 There are no ventilation systems in the plant designed specifically to exhaust
)
-1 smoke or other products of combustion. ' Normal plant ventilation systems will
] be utilized for that purpose. Portable smoke ejectors are provided to facilitate ng
/ the removal of the products of combustion should normal ventilation systems
..] be unavailable because of damper closures or other failures. Because the normal
'$ ventilation system is capable of being realigned so as to shut down the supply 4
i WNP-3 DSER SEC 9 9-44 I i
-wa . , . - .. . -e, m.. 7v-. 4. -e --*,.,e7w-..7.p, . . . . . - . - - - - - -
_ ...==x
- . .: . - = .w . _ . . . . . ~ . .. . . .- . -a. _ ww.._ t l
l . l l air system which maintaining the return air system in an exhaust mode, we find d this acceptable. The power supply and controls for the ventilation systems are located outside the fire areas served by the systems. 1; _ ] Stairwells are designed to reduce smoke infiltration during a fire. Charcoal , 1 filters have been protected in accordance with Regulatory Guide 1.52. This ~ t
]
includes the provision of manual fire suppression system for nonsafety filters. ,.j We find this acceptable. 'j Lighting and Communication 1 h Redundant AC emergency 1ighting is provided in areas where safety-related
] functions are performed, in access routes to these areas, and for emergency evacuation. In addition, emergency DC lighting from the 125-volt station batteries provided emergency lighting for the control room and remote shutdown ; rooms. We are concerned that under this arrangement, a fire in one area may j result in damaged circuits for emergency lighting in other fire areas. We will j require that fixed, self-contained emergency lighting units, with individual, .} 8-hour capacity power supplies, be installed in accordance with Section C.S.g l of BTP CMEB 9.51. "1 d
j Emergency plant communications is provided via voice powered head sets located _ pl at preselected stations. However, the applicant has not provided sufficient
]
information for us to verify compliance with our guidelines. We will require that a communications capability, independent of normal plant systems, be pro-5.: vided for use by the fire brigade and other operations personnel as delineated R in Section C.5.g of BTP CHEB 9.51. Fire Detection and Suppression 1 Fire Detection A fire detection system is provided for all areas containing safety and safe shutdown-related systems and in areas which present a potential fire exposure to such systems. The detection systems and all signaling circuits, such as L WNP-3 DSER SEC 9 9-45 i g r.. , s ~--..e-~ ;, - .m e - w 3+- %.%. %,,~ v, c.-.- + , >% = - . e%+-e-,.e,- - -- r - ,w
..= .. J u -~ ~ . L i x . w .s u - _ . .._, ~. ....L:. .L ~ . z. _._..: _ w .
1
+ .
t water flow alarms, are Class A supervised as defined in NFPA Standard 720. The systems provide alarm and trouble indication to the control room even j under single break or ground fault conditions.
.$ ~~- ; All fire detection devices and associated equipment are either UL listed or FM j
approved. The fire detection systems are designed and installed in accordance
'I with NFPA Standards Nos. 72A and 720. We find this acceptable. }
The applicant has not provided sufficient information for us to verify that the primary and secondary power supplies for fire detection and signaling systems complies with our guidelines. We will require that such systems comply with Section C.6 a(6) of BTP CMEB 9.51. Fire Protection Water Supply System t The water supply system for fire fighting consists of two fire pumps: one is 1
! electric motor driven and the other is diesel engine driven. We are concerned j that the fire pumps and their controllers are not listed by an independent 1] testing laboratory for the intended use. We will require that the fire pumps 4 and controllers be UL listed or FM approved in accordance with Section C.6.b. . of BTP CMEB 9.51.
The two fire pumps are located in separate cubicles of the fire pumphouse and
, are completely separated by 3-hour fire rated walls. Each pump has a separate r
suction, and discharges through independent underground connections into the main fire water piping distribution system. Each fire pump has a rated capacity
.' of 3500 gpm at 125 psi.
Fire protection water for the plant is taken from wells at the makeup pumphouse and pumped by the makeup booster pump into two 300,000 gallon fire water stor-
) age tanks which are reserved exclusively for fire fighting. The fire pumps can draw suction from either or both tanks. The interconnecting piping is
'j arranged such that a leak in one tank will not cause drainage of both the water i storage tanks. We find this acceptable.
)
l
.i ) .
i l WNP-3 DSER SEC 9 9-46 ,4 !-L__, ., _
j ad AMLaL L.J _. L.~ u.. . : .= = ..= 2 . i = = = -~ '
.i 1-i i
'l
; ;' The greatest water demand for the fixed fire suppression systems is 2000 gpm, q~ and coupled with 1000 gpa for hose streams, creates a total water demand of i 3000 gps. We find that the water supply system can deliver the required water demand with one fire pump out of service.
7! .& An 18,000 gallon auxiliary fire water storage tank is provided as a water supply ~
]ij for standpipe and hose systems protecting equipment required for safe shutdown in the event of a Safe Shutdown Earthquake (SSE). The tank is located on the roof of the Reactor Auxiliary Building. The tank is designed in accordance with
] ASME Section VIII, is seismically designed and supported, and is sized to supply j two 75 gpa hose streams for two hours. This conforms with Section C.6.c(4) of BTP CMEB 9.51 and is, therefore, acceptable. -
~
i
- j The fire protection water distribution system consists of an underground 12-inch
[ pipe loop around the main plant building. The loop provides two separate flow paths to the internal distribution system in the Fuel Handling, Reactor Auxil-
- ] fary and Turbine Buf1 dings. The water pressure in the distribution system is maintained at approximately 120 psi by an electrically driven jockey pump. We
]j find this acceptable.
'] Yard hydrants are provided at intervals of approximately 250 feet along the j fire main loop. The lateral to each hydrant is provided with a curb box valve . d to facilitate hydrant maintenance and repairs without shutting down any part of
) 1 the water distribution system. Hose houses are installed adjacent to each
- t hydrant and are equipped with a compliment of fire fighting equipment in accord-i1
.it ance with NFPA Standard No. 24. We find this acceptable.
l t ] All sectional and isolation valves in the fire water distribution system are 1 either post indicator valves (PIV) for underground piping or outside screw and ] yoke (OS&Y) valves for interior building piping. Supervision has not been pro- !j vided for all valves in the fire protection water supply system in accordance - 'I with NFPA Standards Nos. 26 and 13. To meet out guidelines in Section C.6.c
;j of BTP CMEB 9.51, we will require the applicant to provide locks or electronic a tamper switches for all PIV and OS&Y type control and sectional water supply l valves.
I J
]
1 l WNP-3 DSER SEC 9 9-47 !i a !h. l
.; .~, =. 73,, . = ; - - _
=._ _: ._. .: - :=
s w ::=. . - . -- .. -. ;. . -. x-~ . .: . .- . u_ .x am I
'I Sprinkler and Standpipe Systems The sprinkler systems and standpipe hose systems are independently connected to the looped yard main or from the internal cross connections through build- ~~
i ings so that no single failure in the water supply system will impair both the I
; primary and backup fire protection system.
~ The wet pipe sprinkler systems, preaction sprinkler system, multicycle sprinkler systems and water spray systems are designed and installed in accordance with
] the appropriate provisions of NFPA Standards Nos. 13 and 15. The areas equipped ll with water suppression systems are identified in Appendix 9.5A1 of the appli- ' ) 1 cant's fire hazards analysis.
Interior manual hose stations are provided and equipped so as to reach any loca-
! tion inside a building with at least one effective hose stream. Individual j standpipes ~are 21/2, 3 or 4 inches in diameter for multiple hose connections
} and 21/2 inches for single hose connections. Hydraulic calculations have i
j verified that the 21/2 and 3 inch diameter piping is capable of supplying at least 100 gpm at 65 psi at the hydraulically most remote outlet. This conforms to the requirements of NFPA Standard No.14 and is, therefore, an acceptable deviation from Section C.6.c(4) of BTP CMEB 9.51. J; The standpipe and hose systems installed in the Reactor Building, Reactor Auxi-a i liary Building and Fuel Handling Building are designed to supply water for i manual hose use in areas within hose reach of equipment required to be opera-tional in the event of an SSE. The piping system serving such hose stations have been analyzed for SSE loading and have been provided with supports to assure system integrity. The seismically analyzed standpipe system is supplied from the dedicated 18,000 gallon capacity Auxiliary Fire Water Tank. This conforms with the guidelines of Section C.6.c(4) of BTP CMEB 9.51 and is, therefore, acceptable. i 4 WNP-3 DSER SEC 9 9-48
_ . . - . _. w -.:. ..u..c.. -
- ._ .-_.- - .: _ : z. a ; a~ .:,: . . .. w i
l Foam Suppression Systems l Foam fire suppression systems utilizing a 3 percent solution of Agueous Film j Forming Foam (AFFF) are provided for the protection of the Diesel Fuel Oil Stor- ~ q age Tank areas. The systems are designed and installed in accordance with NFPA .
;i Standard No. 118. We find this acceptable. -1 ji i
Portable Fire Extinduishers
-t lj Portable fire extinguishers are provided throughout the plant in conformance with NFPA Standard No. 10. The extinguisher types used are: drv chemical; 3]
j carbon dioxide; halon; and water. All portable fire extingui ners are either
; UL listed or FM approved. This complies with Section C.6.f of BTP CMEB 9.51 i and is, therefore, acceptable.
I s
.j 9. 5.1. 5 Fire Protection of Specific Plant Areas t
j A. Containment
.i l .f.1 The applicant has not identified the location and relative position of redundant q shutdown systems within containment. Therefore, sufficient information has ]: not been provided to independently verify compliance with our guidelines. We _ ] will require that fire protection for redundant shutdown systems in containment 2
comply with Section C.7.a of BTP CMEB 9.51. i -lj The reactor coolant pumps are protected by automatic multicycle sprinkler
- ; systems in lieu of an oil collection system. This system does not provide an equivalent level of protection. We will require that the reactor coolant j pumps be equipped with an oil collection system in accordance with Section C.7.a
+ .i of BTP CMEB 9.51. A 1 j , B ., Control Room 1
? The control room is separated from adjacent plant areas by 3-hour fire rated
; walls, floor and ceiling. A computer tape storage room and supervisor's office,
, i i - 4 y l WNP-3 DSER SEC 9 9-49 1
- . _ ~ . _. - . -_ .. . .- ,_- -. __
. w :.a = :..:-.= ._ __ ..a : - x . w . : :: .~.= wa ::
-l.
- 1 j
4 t which is not separated from the control room by 1-hour fire walls, are not
- 1- equipped with an automatic fire suppression system, and therefore, are not t consistent with Section C.7.b of BTP CMEB 9.51. We will require that
.} peripheral rooms in the control room are protected in accordance with
,j Section C.7.b of BTP CME 8 9.51. M ) 'j Ionization-type smoke detectors are located throughout the control room at the
- {
,1 ceiling. Fire detectors have not been provided in control room cabinets and , l
.)4 consoles, and therefore, are not consistent with Section C.7.b of BTP CME 8 9.51.
j We will require that the applicant proviae cabinet mounted fire detectors in H accordance with Section C.7.b of BTP C'4EB 9.51. ,1 i si The applicant has not completed a fita hazards analysis of the. consequences of .a
, :6 1
a fire in the control room on redundant shutdown systems. We will report on { this-issue in a subsequent SER. 1
- ,s j Cable Spreading Rooms
'j .
- Cable Vault "A" contains control and instrument cables associated with Safety 4
-l Trains A and C. Cable Vault "B" contains control and instrument cables 1 associated with Safety Trains B and D. The vaults are separated from each other M.,! and from adjoining plant areas by walls, floors and ceilings of 3-hour fire _
resistive construction. All openings in the fire barriers are protected by q either 3-hour rated fire doors, dampers or penetration seals. The cable vaults ] are equipped with preaction type sprinkler systems and areawide smoke detection systems. We find this acceptable. The applicant has not provide sufficient information for us to verify compliance with our guidelines concerning the d, establishment of aisleways between cable tray stacks. We will require that ,d aisle separation between tray stacks be at least 3 feet wide and 8 feet high y in accordance with Section C.7.c of BTP CMEB 9.51. .d ,A
- i i:
,i !, 4 l.. i WNP-3 DSER SEC 9 9-50 l !~ l j _ - . _ . , . . - . . _ . . . . - - - _ . . . _ _ _ . _ . . . . _ . . . _ _ , _ . - . . , , ~ . . . . . . .
.u- .
.=.=ww.._..___..-.--_..w__
-l
~
. 9.5.1 8 Summary of Deviations from CMEB 9.51 l The technical requirements of Appendix R to 10 CFR 50 and Appendix A to BTP 1
.! ASB 9.51 have been inicluded in BTP CMEB 9.51. Listed below are the deviations l .
from the guidelines of BTP CMEB 9.51 that have been identified and approved: . n.- h
] 1. Diameter of standpipe supply piping, Section 9.5.1.4 n.i i
a Conclusion
.G, The following are the open fire protection items:
-f j The applicant has been informed of the necessity to resolve all open items so that all fire protection features can be implemented prior to fuel loading.
l- We will report our review of these unresolved items in a subsequent safety I evaluation report.
.l 4
b
- i
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q 1 fl l
- 1
} WNP-3 DSER SEC 9 9-51 i
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f il i i I f 10 STEAM AND POWER CONVERSION SYSTEM
- 4 10.3 Main Steam Supply System ~~
i .i.j 10.3.5 Secondary Water Chemistry i
.j Introduction
.}
-l In late 1975, we incorporated provisions into the Standard Technical Specifica-j tions that required limiting conditions for operation and surveillance require-
.] ments for secondary water chemistry parameters. The Technical Specifications ~] for 11 pressurized water reactor plants that have been issued an operating license from 1974 until 1979 contain either these provisions or a requirements y to establish these provisions after baseline chemistry conditions have been
~l determined. The intent of the provisions was to provide added assurance that the operators of newly licensed plants would properly monitor and control sec-t t
ondary water chemistry to limit corrosion of steam generator components such
] ;
as tubes and tube support plates. 1 j In a number of instances, the Technical Specifications have significantly { restricted the operational flexibility of some plants with little or no benefit with regard to limiting degradation of steam generator tube and the tube support .j plates. Based on this experience and the knowledge gained in recent years, we ,I have concluded that Technical Specification limits are not the most effective -I a way of assuring that steam generator degradation will be minimized. Due to the complexity of the corrosion phenoT.ena involved and the state-of-the-art as it exists today, we are of the opinion that, in lieu of specifying limit-ing conditions in the Technical Specificattor., a more effective approach would
.; be to specify a Technical Specification that required the implementation of a 4 secondary water chemistry monitoring and control program containing appropriate procedures and administrative controls. This has been the approach for control of secondary water programs since 1979.
6 10-1 WNP-3 DSER SEC 10 I ,1 u w_ , yee se .pg .+ pe e. = * =- .~n=ese-,y-. -se,..w %vev=p per-,, . 7gs - . .. e
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..= _
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+
i .I q The required program and procedures are to be developed by applicants with input from their reactor vendor or other consultants, to account for site and plant specific factors that affect water chemistry conditions in the steam generators. In our view, plant operation following such procedures would provide assurance j that licensees would devote proper attention to controlling secondary water chemistry, while also providing the needed flexibility to allow them to deal effectively with an offnormal condition that might arise.
'i l Evaluation l
In the FSAR the applicant provided details of a secondary water' chemistry moni-
] toring and control program. The information provided in the FSAR was not suf-
{ ficient for us to complete our evaluation. The applicant provided additional l information by letters dated July 15 and September 2, 1983. The information y provided by the applicant is inadequate to complete our evaluation. To complete our evaluation, we need the following information; i A summary of operative procedures to be used for the steam generator secondary j water chemistry control and monitoring program, addressing the following: l 1 1. Identify the sampling schedule for the critical chemical and other param- - eters and the control points or limits for these parameters for each opera-l ting mode of the plant, i.e., dry layup, cold shutdown, hot standby / shut-down, and power operation, j 2. Identify the procedures used to measure the values of the critical param-
.{ eters, i.e., standard identifiable procedures and/or instruments.
-4 l- 3. Identify the sampling points, considering as a minimum the steam generator-1 blowdown, the hot well discharge, the feedwater, and the demineralizer [j effluent. We recommend a process flow chart similar to that in EPRI NP-2704-SR "PWR Secondary Water Chemistry Guidelines."
]
- 4. State the procedures for recording and management of data, defining i corrective actions for various out-of-specification parameters.
10-2 WNP-3 OSER SEC 10 a
-4
v_-- w =.-___.a.=-x- sa 1 .
?
o i q 5. Identify (a) the authority responsible for interpreting the data and j { initiating action and (b) the sequence and timing of administrative events li required to initiate corrective action. w 1 10.4 Other Features .
~~
n yi 10.4.6 Condensate Cleanup System 1
;i
[] Introduction
-3
.j The purpose of the Condensate Cleanup System is to remove dissolved and suspen-j de'd solids from the condensate in order to maintain a high qual ~ity of the feed-water being supplied to the steam generators under all normal plant conditions j (startup, shutdown, hot standby, power operation). This is accomplished by
. directing the full flow of condensate to a set of mixed bed demineralizer units.
Since the demineralizers need periodic resin regeneration, spare units are pro-Vided in the system to replace units take~n out of service. The system provides final polishing of the secondary cycle condensate water. l Evaluation The condensate cleanup system is designed to assist in the control of the sec- -- ondary side water chemistry and is part of the total control system. The condensate cleanup system includes all components and equipment necessary
., for the removal of dissolved and suspended impurities which may be present in a
j the condensate. J lj We have reviewed the CCS equipment design, materials and system operation in
. accordance with Section 10.4.6 of Standard Review Plan, NUREG-0800.
The system meets the requirements for condensate cleanup capacity, provides f effluent of the required purity, and contains adequate instrumentation to moni-tor the effectiveness of the system. The system is connected tn radioactive waste disposal systems to allow disposal of spent resin or regenerant solutions when required. We have reviewed the sampling equipment, sampling locations, _ i - 10 WNP-3 DSER SEC 10 1 i I
.- - - . _ _ _ - - - . - - . . _ . . ~
a.-..-..-.--. u .- - . ~ . .... a - ~ ~ .w ..- .... _ , . . . . . w . ._ : - a li i
.1 i
j and instrumentation to monitor and control the CCS process parameters. On the j basis of this review, we find that the instrumentation and sampling equipment provided is adequate to monitor and control process parameters.
~
4 Based on our review of the applicant's proposed design criteria and design ;
] bases for the condensate cleanup system and the requirements for operation of
] 3 the system, we conclude that the design of the condensate cleanup system and supporting systems meets our guidelines and is, therefore, acceptable. The j secondary water chemistry monitoring and control program is evaluated in Sec-
,i tion 10.3.5.
,y Conclusion Based on the foregoing evaluation, we conclude that the concensate cleanup system meets our guidelines, and is therefore, acceptable, i . i. i 10.4.7 Condensate and Feedwater System .: ._
]i 10.4.8 Steam Generator Blowdown System j Introduction 1 ] The Steam Generator Blowdown System (BDS) is designed to maintain the specified 1 water chemistry in the steam generators during all operational modes. The
[ system continuously removes particulate impurities from the blowdown flow by directing the flow through electromagnetic filters and demineralizers before 3 J returning to the condenser. i 4 Evaluation i The steam generator blowdown system (SGBS) controls the concentration of chem-ical impurities and radioactive materials in the secondary coolant. The scope
~
of review of the SGBS included piping and instrumentation diagrams, seismic and quality group classifications, design process parameters, and instrumenta-tion and process controls. The review has included the applicant's evaluation
; of the proposed system operation and the applicant's estimate of the controll .
ing process parameters. j ~5 10-4 WNP-3 DSER SEC 10 i 3e .. A.h w m m .2. , _ . _.m. . _ _ . - ......m.- .
. _ . - .___c__..._..__m..__m.m._._-.u_.._ -
-- _ e A l i
i, The steam generator blowdown system is monitored continuously for radiation in f the secondary side of the steam generator. Radioactive blowdown is handled i i 3 routinely in the demineralizer system and the electromagnetic filters. Back-
..~
; wash fluids are handled in Secondary Particulate Waste System and the Radwaste j System.
1
;j The steam generator blowdown system from steam generator nozzles to the BEX area il is designed to seismic Category 1 and ASME III Class 2 requirements up to last
?) isolation valve and downstream up to BEX area. The steam generator blowdown system from BEX area to Seismic Interface Restraint System is designed to i seismic Category 1 and ANSI B31.1 requirements. Thus, the SGBS meets the l quality standards requirements of General Design Criterion 1 an'd the seismic
-j requirements of General Design Criterion 2.
~3 The secondary water chemistry monitoring and control program is evaluated in l Section 10.3.5.
?
We have reviewed the SGBDS in accordance with Section 10.4.8 of Standard Review
-; Plan, NUREG-0800.
s j Instrumentation and automatic controls are provided to monitor and control
-l the operation of the blowdown system, with provision for sampling of the blow- -
.1
-j down, in conformance with the guidelines of Branch Technical Position MTEB 53.
1 q Conclusion
.r, a
as Based on the foregoing evaluation, we conclude that the proposed steam generator
.; blowdown system meets our guidelines and is, therefore, acceptable.
1 5 o I k
~
ir - WNP-3 DSER SEC 10
' 10-5 1
A-t <4
- 2. : - -
w .- . . am ' ~ ^ : :.a a c1 1 m 7 j 12 RADIATION PROTECTION '- 9
.l :
j The staff has evaluated the proposed radiation protection program presented in ~
- {j Chapter 12 of the FSAR against the review guidelines, and criteria set forth j in the Standard Review Plan (SRP), NUREG-0800, Section 12.' The radiation U protection measures at.WNP-3 are intended to ensure that internal and external radiation dose to plant personnel and contractors, due to plant conditions, l including anticipated operational occurrences, will be within applicable limits N of 10 CFR 20, and will be as low as is reasonably achievable (ALARA).
71
.i.l; d The basis of the staff's acceptance of the WNP-3 radiation protection program is that doses to personnel will be maintained within the limits of 10 CFR 20, g " Standards for Protection Against Radiation." The applicant's radiation d protection design and program features are consistent with the guidelines of j Regulatory Guide 8.8, "Information Relevant To Ensuring That Occupational j Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable" (Rev. 3). On the basis of this review, the, staff concludes that the radiation protection mea.9res incorporated in the design and the proposed radiation protection program will provide a reasonable assurance that occupational doses ,
f will be maintained ALARA and below the limits of 10 CFR 20 both during plant
; operation and during decommissioning, d
12.1 Ensuring that Occupational Radiation Doses are ALARA '4, % The staff has audited the policy considerations, design considerations, and d 1 operational considerations contained in the WNP-3 FSAR against the criteria set forth in NUREG-0800, Section 12.1. The staff review consisted of ensuring j that the applicant had either committed to following the criteria of the j regulatory guides and staff positions referenced in NUREG-0800 (SRP) Sec-
; tion 12.1 or provided acceptable alternatives. In addition, the staff selec-
.] tively reviewed the applicant's FSAR against the acceptance criteria of the SRP using the review procedures in the SRP. This selective review found the
]
1; plant acceptable in these areas. Details of the review follow. j i 12-1 WNP-3 OSER SEC 12
'l - . . . - - _ , , , . , , . . . , , ,
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! l a 12.1.1 Policy Considerations 1
1 i The applicant provides a management commitment to ensure that WNP-3 will be designed, constructed, and operated in a manner consistent with Regulatory tj Guides 8.8 and 1.8 " Personnel Selection and Training" (Rev. 1). The applicant :
- e
,4 has committed to implement a radiation protection program in accordance with j Regulatory Guide 8.10 " Operating Philosophy for Maintaining Occupational i Radiation Exposures As Low As Is Reasonably Achievable." The overall ALARA j responsibility, upper management direction and support lies with the Director j of Support Services. The Plant Manager is responsible for the radiological j safety of all in plant personnel and the implementation of ALARA Policy by his staff. The Plant Health Physics / Chemistry Manager is responsib'le for develop-M ment of good procedures and radiation protection practices, including preplan-
!)a ning, use of equipment and work techniques. 3
, In addition, line supervisors are also responsible for maintaining plant doses ALARA. The ALARA philosophy was applied during the initial design of the
] plant. Since then, the applicant has continued to review, update, and modify 1
'J the plant design and construction phases. The plant's staff - Health Physics
] and Chemistry, periodically review, update, and modify plant design features and maintenance features as appropriate, using dose data and experience gained A from operating nuclear power plants. This'is done to ensure that occupational -
1 1 doses will be kept ALARA in accordance with Regulatory Guide 8.8 and NUREG-0800 f criteria, i 1 1 12.1.2 Design Considerations
,q d
(' The objective of plant radiation protection design is to maintain individual
.h personnel doses ALARA as well as the collective doses of all personnel and
,h . within the limits of 10 CFR Part 20. i l d The applicant, using feedback information from operating plants and following I i guidelines of Regulatory Guide 8.8, has incorporated facility and equipment design j 4 improvements at St. Lucie 1 and 2, and Waterford 3, which plants are similar in
- j design to WNP-3, to satisfy the plant's radiation protection design objectives.
~~
. 12-2 WNP-3 OSER SEC 12
g
._ _ _ _ = _ __: :- _ - au 1 .
i. J j Examples of these design features include: (1) Packaging of units, skid mounted for free access and quick removal to a
]
Ji low radiation area for maintenance or repair. 1 .
@A (2) Most pumps are flanged to facilitate ease of removal to a low radiation area, pump casings are provided with drain connections and pumps are 3 equipped with mechanical seals for greater reliability and reduction in d ;j servicing time.
7 (3) Ion exchangers are designed for complete drainage, spent resin is removed
;l y remotely by hydraulically flushing the resin to the solid ' waste management h
r, system. (4) Absorber (Charcoal) Beds are completely drainable, with single point
..1 ei connections for charcoal removal and equipped with accessories to provide 1 a safe and rapid removal of contaminated charcoal.
( (5) Tanks are designed to be isolated for maintenance and are complet.ely ([j drainable, and flushable, with a minimum of crevices to avoid accumulation of radioactive crud. c; f.5! These design considerations conform with the guidelines of Regulatory Guide 8.8 and NUREG-0800 and are acceptable. t [ 12.1.3 Operational Considerations b:1 l
} The WNP-3 operational considerations included the development of a radiological
[ training program using the guidelines of Regulatory Guide 8.27, " Radiation
. Protection Training for Personnel at Light-Water Cooled Nuclear Power Plants" and 8.29 " Instruction Concerning Risk From Occupational Radiation Exposure," a radiation zoning and access control system, and general guidelines for workers 'q performing maintenance in high radiation areas. These operational considerations j are to ensure that operating and maintenance personnel will follow specific j plans and procedures in order to ensure that ALARA goals are achieved in the ] ;
operation of the plant. High radiation exposure operations are to be preplanned 1
- - 12-3 WNP-3 DSER SEC 12 1
1 l
. ._ , _ _ _ ~ .~ , __ __ _.
.u. : . ~a. w w q
d and carried out by personnel trained in radiation protection and using proper d equipment. During such activities, personnel will be monitored for exposure to 1
.] radiation and contamination. Upon completion of major maintenance jobs, _
~
personnel radiation exposures will be evaluated and compared with predicted person-rem exposures. The results are used to make changes in future job i
$f procedures and techniques. The plant's health physics management will periodi-g cally review radiation dose trends to determine major problem areas and to Q determine which worker groups are accumulating the highest dose. Plant person- ).l4 nel will use these findings to recommend design modifications or changes in ] plant procedures. The operational considerations conform to Regulatory Guides 8.8 (Rev. 3), and to NUREG-0800 and are acceptaLle. ] , ~.l g' The staff concludes that the policy, design and operational considerations at WNP-3 are adequate to ensure that occupational radiation exposures will be ALARA in accordance with Regulatory Guides 8.8 and 8.10 and meet the criteria of NUREG-0800 and are acceptable.
12.2 Radiation Sources The staff has audited the contained sources and airborne radioactive material
, source terms provided in Section 12.2 and Chapter 11 of the WNP-3 FSAR against $l the criteria set forth in Section 12.2 of NUREG-0800. These source terms are -
n
, used as inputs for dose assessment and for the design of the shielding and E ventilation systems. The staff review consisted of ensuring that the applicant had either committed to following the criteria of the regulatory guides and staff positions referenced in Section 12.2 of NUREG-0800 or provided acceptable y alternatives. In addition, the staff selectively compared source terms for }j specific systems used by the applicant against those used for plants of similar design. This selective review found the plant's source terms equivalent to those used at other plants. Details of the review follow.
6
@ 12.2.1 Contained and Airborne Sources 8
Inside the containment during power operation, the greatest potential for 3 personnel dose during operation is due to nitrogen-16, noble gases, and neu- 'l trans. Outside the containment and after shutdown inside the containment i 4 { ' ( 12-4 WNP-3 OSER SEC 12 t i ! 1
.m - . .,
- a. X '
& - A
_ + . - ., ma.m NGggdm l -
.)-
~the primary sources of personnel exposures are fission products from fuel d cladding defects and activation products, including activated corrosion pro-
]' ducts. Almost all of the airborne radioactivity within the plant is due to j equipment leakage. The fission product source terms are based on 1% fuel
/ cladding defect at full power operation. The coolant and corrosion activation ;
product source terms are based on operating experience and reactors of similar
; design; allowances are included for the buildup of activated corrosion products.
3 ~ Neutron and prompt gamma source terms are based on reactor core physics calcula-h' r:1 tions and operating experience from reactors of similar' design. The source A;) terms presented are comparable to estimates by other applicants with similar
] .;
design and are acceptable.
,1 3 The applicant has provided a tabulation of the maximum expected. radioactive airborne concentrations in equipment cubicles,, corridors, and operating areas, from equipment leakage. The bases of these leakage calculations are in accor-E ' dance with Regulatory Guide 1.112, " Calculations of Releases of Radioactive g Materials in Gaseous and Liquid Effluents From Light-Water-Cooled Power ]
Reactors," and are acceptable. S. J
- The WNP-3 FSAR shows maximum expected radioactive airborne concentrations q in some plant areas in excess of maximum permissible concentration as
] defined in 10 CFR 20.203(d)(1)(11). The applicant should resolve this -
] discrepancy and, until resolution, this item will remain open (471.12).
The ventilation system will be designed to provide sufficient vol_ume changes y per hour in occupied areas which may contain significant airborne activity to [ maintain exposure to personnel ALARA. Air will be rcuted from areas of low M potential airborne contamination to areas of increasing potential airborne (i5 contamination. The resulting estimated airborne radioactivity concentrations in-frequently occupied areas will be a small fraction of 10 CFR 20.103 limits
,. and are acceptable. In accordance with NUREG-0800, the source terms used to
(* : develop these airborne concentration values are comparable to estimates by , other applicants with similar design and are acceptable. ' i-t - . t. L* 12-5 WNP-3 DSER SEC 12 i-i ,.&r:7. n: =z==:yv::v~ cm= a: ::::=- == m=
~
..=; -. = m .~ M E Rhia m 'G I .
1 l : . 12.3 Radiation protection Design Features
]
~!
1 The staff has audited the facility design features, shielding, ventilation, w j and radiation and airborne monitoring instrumentation contained in the WNP-3 - ai FSAR against the criteria set forth in NUREG-0800, Section 12.3. The staff j A
} review consisted of ensuring that the applicant had either committed to follow-h ing the criteria of the regulatory guides and staff positions referenced in $ Section 12.3 of NUREG-0800, or provided acceptable alternatives. In addition,
[j the staff selectively reviewed the applicant's FSAR against the specific areas 1 of review and review procedures identified in NUREG-0800. This review found
.j
] the plant acceptable in these areas. Details of the review follow.
] 4 12.3.1 Facility Design Features
- :l l The applicant has provided evidence that the dose accumulating . unctions performed by workers have been considered in the plant design. Features have 3 been included in the design to help maintain doses ALARA in the performance of
} those functions. These features will facilitate access to work areas, reduce
] or allow the reduction of source intensity, reduce the time required in the
];j radiation fields, and provide for portable shielding and remote-handling tools.
The applicant's facility design features are consistent with the guidance of
.]a Regulatory Guide 8.8 (Rev. 3) and NUREG-0800. Therefore, the staff concludes -
that the facility design features are acceptable. j The applicant has provided five radiation zones as a basis for classifying
; occupancy and access restrictions for various areas within the plant. On this .. )
basis, maximum design dose rates are established for each zone and used as
@1! input for shielding of the respective zones. The areas that will have to be
- i. occupied on a predictable basis during normal operations and anticipated
. occurren m s are zones so that exposures are below the limits of 10 CFR 20, and will be ALARA. The zoning system and access control features also meet the 1 posting entry requirements of 10 CFR 20.203 or standard NRC Technical Specifica- ]a tions, and are consistent with Regulatory Guide 8.8.
' l. lt i 12-6' WNP-3 DSER SEC 12 l l, 3
..~.n + , . , ,..e,.. ..,
-- .n - , - .v--. ,
. % .2 , m c ., g .3. c e c . n m , m _ , , _ _ __-
4
.d '
1 . j
] Several features are included in the plant design and operational program to minimize the buildup of activated corrosion products, a major contribution to
.] .
! occupational doses. Examples include:
y 23 (1) Most pumps are provided with drain connections to facilitate decontamination. ; 4..i j (2) Ion exchangers are designed for complete drainage and are designed with a dA minimum of crevices to reduce accumulation of radioactive crud.
!)
1
) (3) A steam generator drain pump is provided to accomplish a more complete ,
and rapid drainage. 7[).
]
i (4) An electromagnetic filter is used to remove radioactive corrosion products from coolant. 1 i The applicant's corrosion product control features are consistent with the j guidance of Regulatory Guide 8.8 (Rev. 3) and NUREG-0800 and are acceptable. J 3 j The design features incorporated by the applicant for maintaining occupational radiation doses ALARA during plant operation and maintenance will also serve to maintain radiation doses ALARA during decommissioning operations and are, therefore, acceptable. V 4 Jn 12.3.2 Shielding R
~. The objectiva of the plant's radiation shielding is to provide protection .,; against radiation for personnel, both inside and outside the plant, during fd normal operation, including anticipated operational occurrences and during reactor accidents. The shielding was designed.to meet the requirements of the h radiation dose rate zone system discussed above. The following are several II of the shielding design features incorporated into WNP-3.
3. j e (1) Reduction of neutron activation of equipment, piping, supports and other 1 materials by providing suitable shielding around the reactor vessel, and
] '
to minimize radiation streaming into the reactor cavity and general containment spaces. i
~%L7 12-7 WNP-3 DSER SEC 12 'l . . . , . _ . . _ _ . _... .. . .
~
_ _ . . _ _% v- 1.iiLai.i h a..c.s___..____ _u..nu a ei i
] (2) Shielding is provided for all equipment anticipated to contain radiation l sources.
0.i
.; (3) Shielding discontinuities such as shield plugs, hatch covers, shield -j doors to high radiation areas are provided with offsets to reduce radia- ;
j- tion streaming. j (4) Access labyrinths are provided for areas containing high level radiation
- sources to preclude a direct radiation path from the equipment to acces-i sible areas.
_i These shielding techniques are designed to maintain personnel radiation exposures
]j ALARA. Therefore, the staff concludes that the shielding design objectives are acceptable.
'] i The applicant's shielding-design methods included the use of standard computer codes. The applicant also used shielding information from operating nuclear j plants as input data for the shield design calculations. The staff concludes j that the shielding-design methodology presented is acceptable. a . d
.j The fuel transfer tube shield structure is a combination of concrete, steel j and lead. The design objective of the shield is to completely enclose the fuel -
transfer tube by shielding materials to prevent inadvertent exposure of person-nel to this high radiation source. The access opening is covered by a one foot thick shield of lead shot. The expected dose rate limits are 5 mrem /hr in the most likely occupied areas and 25 mrem /hr in areas of infrequent occupancy and
;q narrow gap areas.
3.] The applicant stated that portable continuous radiation monitoring equip-ment with local audible and visaal alarms will be provided if conditions i 4 11, warrant. It is our position that all accessible portions of the spent-j fuel system must be clearly posted with signs stating that potentially l lethal radiation fields are possible during fuel transfer. If other than
- ! permanent shielding is used, local audible and visible alarming radiation monitors must be installed as required by NUREG-0800. Use of portable radiation alarm monitors to be installed "if conditions warrant" is not .
'~
._ 12-8 WNP-3 DSER SEC 12 3
. . . _ . . . _ - . . _ . . - -, s -. _
, . _ . . . . . . _ . . . . _ . 7.,
g -- 21m.uACcmLJ _w ' - m._.__ _. a - O l _b . s ca-- a
-4 1 i
j acceptable. The applicant should state in Subsection 12.3.2.3.5 of the j FSAR that the access to fuel transfer tube will be in compliance with l NUREG-0800. Until then this remains an open item (471.26) (new item).
= , a, -
_j In accordance with the criteria of Item II.B.2, NUREG-0737, " Clarification of ;
] TMI Action Plan Requirements" the applicant has performed a. design review of .d station shielding to allow access to plant areas after an accident.
1 4 The systems designed to function after an accident include: Safety Injection, i Shutdown Cooling, CVCS, Containment Spray and Recirculation, Sampling, Gaseous Radwaste, Shield Building Ventilation and Control Room Air Air-conditioning systems. The dose rate calculations were performed for the areas of the above systems by using well known computer codes in superimposing the effects of all sources toobtainthemaximumexpecteddoseratethroughoutthypant. The radiation environment was evaluated for 1, 2, 4, 8, 12 hours, I' day, I week, 1, 3, 6 j months and 1 year following the reactor shutdown followik a LOCA with signiff- ,j cant core damage. Dose rate zone maps were provided for each relevant area.
, Vital areas requiring accessibility following an accident are identified with 1
respect to location, oCCJpancy requirements, and maximum dose levels. Vital - areas include: Control Room, TSC, Sampling room, Hot Lab, Health Physics
.t .: Office, and Counting room.
a b d The shielding design review by the applicant showed the less-than-15 mrems h criterion is met by WNP-3 for vital areas requiring extended or continuous j occupancy. Additionally, GOC 19 limits are met for those vital areas requiring i only infrequent access. d; . j On the basis of its review, the staff has concluded that the applicant has l performed a radiation and shielding design review for vital areas access in 0 accordance with Item II.B.2 of NUREG-0737. t i I 12-9 WNP-3 DSER SEC 12 l ( (
_=- c5 ' _ sw .__. m. L - _ . . . ._ u- . ,., ._ u _ u
)
1 i l
. i 4'
12.3.3 Ventilation j; The ventilation systems at WNP-3 are designed to protect personnel and equip- _
'; ment from extreme thermal environmental conditions and ensure that plant 't ] personnel are not inadvertently exposed to airborne contaminants exceeding j .j those given in 10 CFR 20.103. The applicant intends to maintain personnel d exposures ALARA by:
()i (1) Maintaining airflow from areas of potentially low airborne contamination d to areas of higher potential concentrations; a li 1 _j (2) Ensuring negative or positive pressures to prevent exfiltration or infil-tration of potential contamiaants, respectively; and 4 g (3) locating ventilation systems intakes so that-intake of potentially contam-inated air from other building exhaust points is minimized.
]
j The design criteria are in accordance with the guidalines of Regulatory U Guide 8.8 (Rev. 3). Some examples of exposure reduction features in the a f ventilation system are listed below. (1) Adequate space is provided around the venti.' tion fans and filter units - 4 to allow rapid servicing and replacement of sections and filters. (2) Pipe equipment vents directly to the appropriate radwaste subsystem for
; treatment thus preventing spread of contamination.
'.{
'f f,I (3) Welded seams are used throughout the duct work on contaminated systems to d the extent possible to reduca system leakage.
,j (4) Use of filters that can be easily maintained for containing radioactivity
) so they will not create additional radiation hazards to personnel in j normally occupied area. .i i
I g 12-10 WNP-3 DSER SEC 12 1 1 1 _ _ , . . , . . _ , - .m. . . _ _ _
}? .
- v x .
.uc.. ._.= = =. _: . . uaah y ,
t i j 12.3.4 Area Radiation and Airborne Radioactivity Monitoring Instrumentation j ' i .j 12.3.4.1 Area Radiation Monitoring Instrumentation
; The applicant's area radiation monitoring system is designed to: ;
a (1) Monitor the radiation levels in areas where radiation levels could become i significant and where personnel may be present; E
-l (2) Alarm when the radiation levels exceed preset levels to warn of excessive
.] radiation l'evels; and y .-
' (3) Provide a continuous record of radiation levels at key locations through-j out the plant.
1 In order to meet these objectives, the applicant plans to use 73 area monitors located in areas where personnel may be present and where radiation levels .j could become significant. The area radiation monitoring system is equipped with local and remote audio and visual alarms and a facility for central { recording. 1 The applicant should inform the NRC whether or not post-accident radiation - monitors for containment and sampling area will be provided in compliance
.g with Regulatory Guide 1.97 or should provide a basis for an acceptable .l deviation (471.1).
The applicant has provided area radiation monitors around the fuel storage 1 area to meet the requirements of 10 CFR 70.24. g .
- As per SRP, NUREG-0800 the applicant should commit to the implementation
-j of Regulatory Guide 8.12. " Criticality Accident Alarm Systems" and N16.2-1969 or provide a description of their alternative approach. Until then, this remains an open item (471.23).
To meet the criteria of the TMI Action Plan Item II.F.1.3, applicant has committed to installing two high-range gamma monitors at WNP-3. .
- ~_ 12-11 WNP-3 DSER SEC 12 me rrW e e- p b -w ,* r - y w g
.- - _ _ -e .,- . _ ,_ __ - .. . - - - . .
p A .- . . e .. -uw=w g! ;, g g . _ = - j' : -x_- x..a - - - .
/ .
q - ,j ,. ' q, q - - ., 4 . g, N
- The range 'of the monitors doet not comply with Table 11.F.1-3 of NUREG-0737,
.t.
- Q thus Table 12.3Miof the FSAR should be revised to show a range of i
1 R/hr to 100 h(gamma only). ' Until such revkMon this remains open , j item (471.27),jnty,ita). N 1- - i A q Inaddition,thelaoplicanthasprovidedplantla ugdrawingsshowingthe
] location of the high-range monitors, pl N
j;, 12.3.4.2 Airborne]ndioactivityMonitorin'gInstrumentation, .~ N ., \ ,) The design object,ives of the. applicant's airborne radioactivity monitoring L s , . 'j system are: *4-e . W ' s j 'i '
. 'k 1 :! w 0..
.9'+! p(h) ( To assist in ma htaining"decupational radiation exposyre to airborne h 4; contaminantsALARAI ,, M
%,0w 1
i1' (2) To check on, the integrity of systems containing radioactivity which are U beingmonitorekandi, d 4 .-
- 3- e g 1 ,
p t . (3) To,Jarn of 'tr. advertent releah of airborn,e'radioactihit) to prevent over-exposiktfvf personnel. yl' T :
^
. , > s. u 1% In order to meet theise objectives, the applicant plans to use ventilation duct ' s s monitors in key locations throughout the plant and portable continuous air y monitors. The ventilation system monitors will be used to provide representa-
- l m- tive air concentrations and rapid Qdication of' abnormal conditions at fixed i locations such'as exhaust ducts.from areas in.which-the airborn'a radioactivity
- t
,, ,4 cou.ld increase,f$d in t,which, personnel normally have access,' consolidated ,, s
- y vjntiMtion exhausted from the plant, and air intake ducts to the control room elnl . 'for post-accident habitability monitoring purposes. Portable or mobile air '
+ g 4 monitors can be relocated,to virtually any @ catfu of the plant, and can be
- connedted to t'Na> ~plar[t rhf ,sf tion, montyring coinmun! cation's.
s sy' tem through the
-t' spare junction b, oxes located thropghout tile plant.
( 's 4 [0J , s _W l' N . i
% g . .. ,
! Allradiationmonitorswillbehehiodicallycalibratedwithstandardsources
- .l traceable'totheNBg , i .u . ,
,s , .
'e7
.\
,s '1 .
;3 12-12 \ WNP-3 DSER SEC 12 s s, s are
- _ ?DT_'_T_ TOL __2_YX_'IS L " ' ' " ~ " ~ ~ ^ ~ * ~ ~ *'#~~~~**'"~'"~~~% ^ ~ ' ' * * '
^__2:1_ _ L 2_ __r____r_ :__e w n~_ .._
. w : --- .- .x.: a . = - .u. w . -.= -. . ~--a.= x = .-- - .= -. = a :=: w =-s The applicant should describe how the radioactive airborne monitoring system will detect the MPC-hours of radioactivity (particulates, iodine, and noble gases) from any compartment which could contain airborne radio- .i 1 activity and which could be occupied by personnel as per-SRP, NUREG-0800, 1 4 Section 12.3-12.4, 4.b.1. (471.21). ;
i 12.4 Dose Assessment q- The staff has audited the applicant's dose assessment for the WNP-3 provided j in Section 12.4 of the FSAR, against the criteria set forth in NUREG-0800, ',j Section 12 3-12.4. This review consisted of ensuring that the appifcant had
.i either committed to following the criteria of the regulatory guides and staff 1 positions referenced in Section 12.3-12.4 of NUREG-0800, or provided acceptable q! alternatives. In addition, the staff selectively compared the dose assessment made by the applicant for specific functions against those made for other '! plants of similar design. This selective review found the plant's dose assess- '
1 i ment equivalent to those of other plants. Details of the review follow. l The applicant has performed an assessment of the doses that will be received
- i. by plant and contractor personnel. This dose assessment is based on occupancy, factors, expected dose rates, expected airborne radioactivity concentrations, and historical information from operation BWR power plants. The dose assess- -
4 ment includes a breakdown of the annual person-rem doses associated with major i functions such as routine operations, routine maintenance, inservice inspections,
; special maintenance, radwaste processing, refueling, and health physics. The applicant estimated the total annual collective dose to plant personnel and contractors to be 440 person-rems. This estimate is consistent with the acceptance criteria in NUREG-0800, that is, using the methods outlined in Regulatory Guide 8.19.
Current 1'y, operating BWRs average 740 person-rems per unit annually, with y particular plants experiencing an average lifetime annual dose as high as 1850
] person-rems. These dose average are based on widely varying yearly doses for BWRs. The staff finds the bases for the WNP-3 exposure estimate acceptable. I v ?
a
. 12-13 WNP-3 DSER SEC 12 f
- , eme -n..n..,., . . - - . ~ . ,v- ..n.y v, ..w~ . n - m r .: ,
i
.u .x - .. u --. . a. u. .-- . . .~ a L. . . a -. w .. : ~ w .--.~ .a a.- . . ~
- I l -
The dose breakdown for refueling work in Table 12.4-8 does not agree with j dose for refueling work given in Table 12.4-2. The applicant should ,} resolve the discrepancy (471.14).
'l
, ji The applicant has provided a tabulation of the maximum expected, radioactive j' y airborne concentrations, as well as estimates of the inhalation dose equivalent jj rates to plant personnel. The dose equivalent rates are derived from the
] airborne radioactivity source terms given in Chapter 11 of,the FSAR. The Ed applicant's assumptions and models on which his internal and submersion dose 1
21 estimates are based for occupational exposures are consistent with those of the 1 ~
] .;
staff and are acceptable. ,q The staff concludes that the applicant's dose Jssessments for contained sources q and airborne radioactive material are comparable to estimates by other applicants
- with similar design and are acceptable.
1 o , .j ' 12.5 Operational Radiation Protectico Program l 1 ThestaffhasauditedtheorganizItion,equipmentinstrumentation, facilities, j and procedures for radiation protection contained in the WNP-3 FSAR against J the criteria of NUREG-CB00, Section 12.5. The plant's health physics program 1 objectives are to provide reasonable assurance that the limits of 10 CFR 20
~
j! are not exceeced, to further reduce unavoidable exposures, and to ensure that individual and total person-rem occupational radiation doses are maintained
;j ALARA. The, staff review consisted of ensuring that the applicant had either
-] _ committed to following the criteria of the regulatory guides a'nd< staff positions
]
J referenced in Section 12.5 of NUREG-0800 or provided acceptable alternatives and selectively compared the applicant's FSAR against the acceptance criteria o e
] of the SRP using review procedures in NUREG-0800. This selective review found the plant acceptable in these areas. Details of the review follow.
l
':j ;
1 j 12.5.1 Organization
'i I The Health Physics / Chemistry manager at WNP-3, in conjunction with line super-
] visors, is responsible for implementing and enforcing the plant's health
! physics program. However, the ultimate responsibility of the health physics..
~
j program lies with the Plant Manager.
't , . _ _ _ 12-14 WNP-3 DSER SEC 12 M.___.. . _ _ . 1., __.,;_._ ,.,_,. . .
- ,. . 3, y, ,
. . . ___ a. : _ . . n ___ _ __2-- -..____1._.___._._ . _ _ _ _ ... ...; .. .._ ._ , _ . .__.cu . _u i
1-1 The WNP-3 station radiation protection organization has been evaluated in accordance with the position of NUREG-0731, " Criteria for Utility Management j and Technical Organization," Regulatory Guide 8.8 (Section C.1.b(21,31)) and
._ ~
d NUREG-0800. i i The paragraphs below present an evaluation of how the health physics organiza-
..! tion for WNP-3 compares with the various staff positions concerning plant 'l organization and management criteria.
t j (1) The organization description for WNP-3 shows that the Health Physics / f Chemistry Manager reports to the Plant Manager. This satisfies the l requirements of Regulatory Guide 8.8 and is acceptable. j (2) The health physics and chemistry function at WNP-3 are separated into a
; Health Physics Section and Chemistry Section that are supervised by a
. Health Physics Supervisor and a Chemistry Supervisor. This satisfies the l recommendation of NUREG-0731 and is acceptable.
i A j (3) The applicant has shown that the qualifications of Health Physics / Chemistry j - Manager meets the requirements of Regulatory Guide 1.8, " Personnel Selection 1 and Trainincj," and is acceptable as the station's RPM.
- l l'
(4) The applicant has committed to using the criteria of ANSI 3.1, December
] 1979 draft, in selecting the individual temporarily filling the Health i Physics / Chemistry Manager's position. This satisfies the requirement of NUREG-0731 and is acceptable.
(5) The applicant has proposed to train health physics technicians in accor-dance with the criteria of ANSI /ANS 3.1-1978, which requires one year of related technical training and two years of experience. This is equivalent j to the criteria of ANSI 18.1 which also reoutres such training and experi-1 ence and therefore is acceptable. . t
] (6) The applicant has committed to having at least one health physics tech-l nician on site at all times. This satisfies the criteria of NUREG-0731 f and is acceptable.
~ ~
l 12-15 WNP-3 DSER SEC 12 'l ' 1 L - .. . . -
.= . ::_; - 2.- . :.. - m _ . _ ..- - . - ... . w w . ~ 1' a ,
J 4 i j WNP-3 has, shown that the current health physics organization meets staff j1 criteria as stated in NUREG-0731, Regulatory Guide 8.8 and NUREG-0800 for an
,; acceptable radiation protection organization.
q y 4 g. c- 12.5.2 Equipment, Instrumentation, and Facilities j I)d ej LThe radiation protection features at WNP-3 station include a counting room, a 'j- radiochemistry laboratory, a conventional chemistry laboratory, health physics .d offices, calibration facility, laundry fac"ity, personnel decontamination, M i and dry cleaning area. These facilities are sufficient to mainta'in occupational g radiation exposure Al. ARA and are consistent with the provisions of Regulatory i Guide 8.8. Equipment to be used for radiation protection purposes includes y portable radiation survey instruments, personnel monitoring equipment, fixed j'? and portable area and airborne radioactivity monitors, laboratory equipment,
~1 4
air samplers, respiratory protection equipment, and protective clothing.
.f In order to meet the criteria of III.D.3.3 of NUREG-0737, the applicant has j, committed to having the post-accident-capability to sample and determine 4
airborne radiotodine concentrations by using portable air samplers and using J silver zeolite as a sample medium. If entrapped noble gases interfere with the i l 1-radioiodine analysis, clean air or nitrogen flushing will be performed. , hj - Q- i The Regulatory Guide 1.70 and the SRP (NUREG-0800) state that the descrip-D,i tion of health physics instrumentation should include the instruments ' 6i jj sensitivity. The applicant provided (in Table 12.5-1) the type of radia- 'j tion the instruments detect and not the instrument's sensitivity. The applicant should provide the instrument's sensitivity (471.6). 14
.i g Low background counting facilities for postaccident analysis are available.
[ The use of sampling equipment and analysis systems for the determination of radiciocine during an accident situation has been incorporated into the WNP-3 j station's training program. 'd d IJ The postaccident radioiodine sampling and analysis provisions described for
- the WNP-3 station satisfactorily meet the staff's position as outlined in
~ ~
NUREG-0737 ,and are acceptable. l .. 12-16 WNP-3 DSER SEC 12 r [] _
-~ . W e.
. y. *7 n+
sm - . - y mme m_ _ y -_ w,.. j - - 1+y - - . - - ~ . - -
p.- L- . . .-_- . .. m- . .- . . :. . . . . ~ .-w.. . ..a...= .aw-.: A-8 i i . 12.5.3 Procedures 1 All station persorinel entering controlled radiation areas will be assigned . t thermoluminescent dosimeter (TLD) badges and pocket dosimeters. Special
; neutron surveys will be provided when plant personnel enter neutron areas when -
i required by 10 CFR 20. Whole-body counts of all plant personnel will be
*j conducted on a scheduled basis and other bioassays will be provided when deemed 2 necessary by the plant's health physics staff, using the guidance of Regulatory i
- t Guide 8.9 and 8.26. All radiation exposure information will be processed and 1
; recorded in accordance with 10 CFR 20.
1 l '! Maintenance, repair, surveillance, and refueling procedures and methods used i
} by the applicant are reviewed to ensure that all plant radiation protection 1
j procedures, and practices, and criteria have been considered, to ensure that occupational radiation exposures will be ALARA and in accordance with Regulatory Guide 8.8. Procedures are also developed to meet the requirements of Regulatory Guide 1.33, " Quality Assurance Program Requirements (Operation)." 4 Based on the information presented in the FSAR and the applicant's responses j to the staff's questions, the staff concludes that the applicant intends to implement a radiation protection program that will maintain inplant radiation y exposures within the applicable limits of 10 CFR Part 20 and will maintain - exposures ALARA in accordance with Regulatory Guide 8.8. i i I a 4
~ .
G 12-17 WNP-3 DSER SEC 12 g
- . . . _ . .- _...a____ _ __ __ _.._m.m__.- m, a ^
13 CONDUCT OF OPERATIONS %
- ~
~
13.1 Organizational Structure of Applicant 1-1 13.2 Training y 4 The applicant's training programs for licensed reactor operators and nonlicensed plant staff were reviewed according to SRP 13.2 (NUREG-0800). The staff accept-ance criteria included applicable porti6ns of 10 CFR Parts 19,.50 and 55, and Regulatory Guide 1.S, as well as the TMI Action Plan (NUREG-0737) and H. R. Deriten's a letter of March 28, 1980 to all power reactor applicants and licensees. 13.2.1 Licensed Operator Training Program 2 A training program for WNP-3 licensed reactor operators has been implemented to develop and maintain an organization fully qualified to operate the plant and maintain the plant safety. The initial a'nd requalification programs, q which are designed to meet the requirements of 10 CFR Parts 50 and 55, and TMI
, Action Plan related requirements, are based on the individual employee's level _
i of education, experience and skills as well as on the level of assigned respon-sibility and intended position. l} 13.2.1.1 Initial Training Program d f_j The initial training program for personnel who will be licensed consists of the following discrete segments: (1) Academic and Nuclear Plant Fundamental This training course will be approximately twenty weeks in length and is j designed to provide individuals with basic knowledge in science and technology j of power plant operations. The major areas to be covered are mathematics, j physics, basic nuclear physics, reactor theory, radiation protection, enemistry,
.j _ .) ^
13-1 WNP-3 DSER SEC 13 ,j - 'i l __m.,__r.--_.,-----m..._-.y..--.~t-.--wm-n--., , , . - - v--. - - , ,--- - - .-
,.__._._._im.____. .. ..m.-.._.um...._____.__w__ ., 11 ,
i instrumentation and control, health physics, electrical theory, transient analysis, fluid flow, thermodynamics and heat transfer. The extent of partici-pation will be determined by the individual's experience and education level. An integral part of the fundamental training program is reactor startup experi-
,; ence. This is a one-week training course conducted at a research reactor for cold license candidates only.
j l With respect to instruction in the topics of fluid flow, thermodynamics and heat j transfer, we require the applicant to provide a program in accordance with the l guidelines as outlined in Enclosure 2 of H. R. Denton's March 28, 1980 letter. We will review the program when it is received and report our findings in the final SER. H (2) Plant Systems - Classroom This training course is designed to provide cold license candidates with an in-depth study of the WNP-3 systems and equipment. The course consists of approxi-mately nine weeks of classroom lectures on nuclear steam supply system and balance of the plant system design, components and operation; instrumentation, control and electrical system design and operation; safety analysis and techni-
.; cal specifications; and operating and emergency experience. Effectiveness of this training will be monitored through examinations. -
i i In addition to the above topics, we require (as specified in Enclosure 1 of q H. R. Denton's March 28, 1980 letter) the applicant to modify the program j to provide training in the use of installed plant systems to control or miti-j gate an accident in which the core is severely damaged. We will review this
'i modification, when it is received, and report our findings in the final SER.
M i I . (3) Plant Systems - Observation i This segment of the training program consists of four weeks of plant observation. 4 The major objective of this program is to familiarize each cold license candi-date with the daily routine involved in the operation of a PWR of similar design.
~ )
j 13-2 WNP-3 DSER SEC 13 1 m s _ _ _ _..- . . m , , , , , . _ ~_ _
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t (4) Simulator Training
'5 i
.j This training program will be approximately ten weeks in length on the WNP-3
!j simulator, and will consist of classroom lectures, simulator control' room lec-
'l tures and demonstration, and simulator control room exercises. The training ;
will include, but not be limited to the following: A
= Classroom sessions consisting of lectures, seminars, and examinations.
']
Demonstrations of how to control individual systems and the integrated Il plant. 1 i j
- Practice of normal and emergency plant operations, including recognition of emergency conditions and response to malfunctions by the license candidate.
l
.l
] =
Exercises during which the license candidates operate the simulator with-out instructor assistance and receive evaluation of ability to safely and j efficiently operate the plant.
..~ i
'] Simulator sessions will also include all the control manipulations as listed
,;.1 j in Section 13.2.1.2.1(2)(b) of this report. At the conclusions of the simula- -
,.; tor training phase, each candidate will be given examinations to determine his
-l q
9 ability to control the operations of the plant in a safe and competent manner. i .,#4 N (5) Onsite Experience y 4 Training in the form of practical work assignments at the WNP-3 will be provided l for approximately 26 weeks. Work assignments may include: plant operating
~
q;I . procedure preparation and verification, pr'eoperational testing of plant systems, participation in hot functional testing program, low power physics and escala-
}l]j tion to power test programs, and preparing and providing instruction on plant ;f systems. Emphasis shall be on the license candidate gaining thorough knowledge of WNP-3.
I 1 -
.- 13-3 WNP-3 DSER SEC 13
~ e - . , - , . . . _ , _ , . .
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- - - . - - ; - - . -w, -- = =
e 4 j (6) Senior Operator and Shift Manager Duties t
-; This training phase, designed to train senior reactor operators and shift mana-4 gers, is approximately one to three weeks in duration. The program consists of:
l leadership, communication, motivation of personnel, problem analysis, decision ; analysis, command responsibility and limits, and administrative requirements for
>I j the particular SRO position.
d 1 j (7) License Review Training i This is an approximately four week course designed to improve the weak areas 3 brought out from a comprehensive examination and to bring the license candidates to a peak knowledge level for the NRC examinations. .t
] 1 (8) Training Program Evaluation i
j The performance of employees participating in the cold license training program .1 are monitored and evaluated throughout the program. Frequent examinations are j given to license candidates in order to determine the effectiveness of the i training and knowledge of the trainees. b - a . j Based on our review, we find that the applicant's initial training program - If conforms to the requrements of the applicable portions of 10 CFR Parts 50 and '} 55, and follows the guidance given in Regulatory Guide 1.8. However, as noted in the above Sections 13.2.1.1(1) and (2) of this report, the applicant's initial
] training program does not satisfy the requirements as outlined in H. R. Denton's- .] March 28, 1980 letter to all power reactor applicants and licensees. Thus, we ;! have not been able to conclude that the applicant's initial training program for 1
j reactor operators and senior reactor operators is acceptable, s1 - 4 13.2.1.2 Licensed Operator Requalification and Replacement Training Programs i Following the initial licensing of cold license candidates, requalification and
! replacement training programs will be implemented to maintain and demonstrate the continued competence and the level of proficiency of all licensed personnel.
e
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.i '. 13.2.1.2.1 Requalification Training Program d
1
,j A requalification training program conducted by the applicant for all licensed
$ reactor operators and senior reactor operators will be implemented shortly after
,d receipt of cold licenses. This program will be conducted on a repetitive two- ' l
] year cycle anif will consist of the following: ~$
] (1) Lecture Series 1,
] The applicant has indicated that at least six pre planned requalification train-
}.:j ' ing lectures will be scheduled throughout the year. Lecture subjects and content will be based on the results of the annual examination' administered to e' licensed reactor operators and senior reactor operators. However, the content of the examination described in the FSAR by the applicant does not cover all the i subjects as listed in Appendix A of 10 CFR Part 55. We will require the content
, of the annual examinations to be modified to include the following subjects as d listed in Appendix A of 10 CFR Part 55 as well as in Enclosure 1 of H. R. Denton's A March 28, 1980 letter:
s. Theory and principles of operation [
- General and specific plant operating characteristics -
x f r.'4 - Plant instrumentation and control systems 1 41 q
- Plant protection systems y,;
e O) d' = Engineered safety systems 4 Normal, abnormal, and emergency operating procedures j
- Radiation control and safety
- j 1
J
= Technical specifications I
Applicable portions of Title 10, Charter I, Code of Federal Regulations . t h. 13-5 WNP-3 DSER SEC 13
- \ l w - - - - . ._ _ _ .1
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Heat transfer, fluid flow, thermodynamics and mitigation of accidents
.l involving a degraded core.
1
-i ,
9 The annual written examination results will indicate the scope and depth of {.j training needed by each individual in the above areas. ,_- 1:1 4 We will review the applicant's modification to include instruction of the above subjects in the Lecture Series and report our findings in the final SER. T T lI (2) On-the-Job Training
- )
3 The on-the-job training portion of the requalification program 'will consist of
,; t the following:
3 (a) Simulator Training I lj Each licensed reactor operator will spend 40 hours annually in a simulator TJ training program. The program will contain instruction in facility changes,
' 'A recognition of emergency conditions, and operating experience at WNP-3 and I?; similar plants.
- l<.'
v Lt.( I.: , (b) Control Manipulations -
,y ,N Gj Licensed reactor operators will manipulate and direct or evaluate the activi-jI ties of those manipulating the station controls through the following reactivity A
c' changes during the term of their licenses. The asterisked items will be pj El performed annually and all other items will be performed on a two year cycle. .D..
% ** Plant or reactor start-ups to include a range that reactivity feedback
. from nuclear heat addition is noticeable and heat up rate is established J
- Plant shutdown
- i
** Manual control of steam generators and/or feedwater during start-up and shutdown j -
s, n f 13-6 WNP-3 DS$R SEC 13 a
'5 -.
cx ~w.a =c:. = x-w.u- .: a wa ~. -. . . =: :- . .. - ~ ----- . .--a 4 j
- Boration and/or dilution during power operation
'l
, ** Any significant (210*.') power changes due to manual changes in control rod ,
; position or boron concentration C1 ** Loss of coolant including:
l y; (.] 1. Significant PWR steam generator tube leaks d 2. Inside primary containment
- 3. Large and small, including leak-rate determination d 4. Saturated Reactor Coolant response (PWR)
!! 1 j
- Loss of instrument air a
4
'j
- Loss of electrical power (and/or degraded power sources) a>
1
** Loss of core coolant flow / natural circulation i
- Loss of condenser vacuum
- i M
- Loss of service water if required for safety ,
.i -
d
- Loss of shutdown cooling
-)
- l Loss of component cooling system or cooling to an individual component
( 1 Loss of normal feedwater or normal feedwater system failure m B
** Loss of all feedwater (normal and emergency) q 'l 9d
- Loss of protective systems channel 1
Mispositioned control rod or rods (or rod drops) 2 J
- Inability to drive control rods
}
1 i - 13-7 WNP-3 DS$R SEC 13 a I t 4._ _ __
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i 1 i
- Conditions requiring use of emergency boration of standby liquid control l system a
1 e
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- Fuel cladding failure or high activity in reactor coolant or offgas
-j .
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- Turbine or generator trip
?
- Malfunction of automatic control system (s) which affect reactivity
?
- f.
- Malfunction of reactor coolant pressure / volume control system
' il. 1
- Reactor trip C)
- 4 Main steam line break (inside or outside containment)
]
Nuclear Instrumentation failure (s) Ti hj The above control manipulations will be performed on the simulator and/or the 9 plant. (c) Knowledge of Facility Design, Procedure, and License Changes N - The applicant has not addressed the instructions of procedure changes and facil-1 ity license changes. As described in Appendix A of 10 CFR Part 55, we require
. .a 9 the applicant to provide a training program to ensure that each licensed reac-tor operator and senior reactor operator will be cognizant of facility design r changes, procedure changes, and facility license changes. We will review the Et w
applicant's modification of the program to incude instructions of these subjects 99 and report our findings in the final SER. 2y1 4 [ (d) Knowledge of Abnormal and Emergency Procedures A
'j In order to ensure a continuing awareness of the actions and responses necessary during abnormal and emergency situations, as described in the Appendix A of
; 10 CFR Part 55, we require the applicant to provide a program to ensure that, t
j each licensed reactor operator and senior reactor operator will review the g I 13-8 WNP-3 DSER SEC 13
'i )
I
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l contents of all abnormal and emergency procedures on a regularly scheduled j basis. We will review this program when it is received and report our findings
$ in the final SER.
i -
,j aj (3) Evaluation ;
4 3 1 The evaluation program for licensed personnel includes the following: =.1 A ' j (a) Annual Written Examination ] An annual written examination will be given to each licensed reactor operator 7 and senior reactor operator. The examination will contain the categories as described under Lecture Series. The applicant has indicated that a grade of A less than 70% in any category shall require accelerated requalification in that
- 1 category. A grade of less than 75% overall requires accelerated requalifica-
] tion in all categories graded less than 75%.
j As specified in H. R. Denton's March 28, 1980 letter, we require the above q criteria for accelerated requalification to be modified to be consistent with q the new passing grade for issuance of a license; 80% overall and 70% each 2 category. We will review the applicant's modification to the above criteria and report our findings in the final SER. - 3 c$ (b) Annual Oral Examination and performance Observation j An annual oral requalification examination will be given. In addition, each M licensed reactor operator and senior reactor operator will be evaluated annually on his performance. The evaluation will include observation of performance j during actual or simulated plant conditions. Any individual given an unsatis- 'ld factory overall evaluation will require accelerated requalification. 3 1 j (4) Accelerated Requalification
- 3 l Individuals requiring accelerated requalification as a result of annual examina-
! tion will not perform licensed-duties until successfully completing the program. ! Accelerated requalification will be given in the categories required or areas .
- l. -
~ '
13-9 WNP-3 DSER SEC 13 4 _1 _ _ _ _ _ _ _ - -
~
. .- . : x .2. . . . : a. hb.h.~.u w a - ==a identified in the written or oral examination. Successful completion of the 3 program will be measured by a reexamination of the individual categories, repeat-
~l ing an entire written examination or repeating the oral examination. Success-ful completion of an accelerated requalification program will be by grade .'
criteria identified in the written and oral sections above. ' 13.2.1.2.2 Replacement Training Program m Replacement training will be conducted to fill vacancies and will prepare indi-
- viduals for increased responsibility in the supervisory, technical or operating
'i staff. Replacement personnel will receive training comparable to that received
'j-.
.t 4
by the initial staff. This will ensure that the required level of proficiency j is maintained. 1,
; As noted in the above Sections 13.2.1.2(1), 13.2.1.2.1(2)(c) and (d), and j
13.2.1.2.1(3)(c) of this report, we find that the applicant's requalification i and replacement training programs do not satisfy the requirements specified in i Appendix A of 10 CFR Part 55 and in the letter from H. R. Denton to all power reactor reactor applicants and licensees dated March 28, 1980. Therefore, we
}
j have not been able to conclude that the applicant's requalification and replace-ment training programs for reactor operators and senior reactor operators are acceptable. , .m 4 9 13.2.1.2.3 TMI Related Requirements for New Operating License I.A.2.1 Immediate Upgrading Reactor Operator and Senior Reactor Operator Training and Qualifications
- g. ,
g p The applicant has established a program intended to assure that all reactor 1 operator and senior reactor operator license candidates have the prescribed experience, qualification and training. 4 't The applicant has indicated that certifications completed pursuant to Section 55.10(a)(6) and 55.33a(4) and (5) of 10 CFR Part 55 shall be made by the Plant
, Manager. However, as specified in Enclosure 1 of H. R. Denton's March 28, 1980
! letter, we require that all licensed operator candidates will be certified _
n j , 1 13-10 WNP-3 DSER SEC 13
'?
t
. _ , . . - - _ ~ _ - _ _ _ - - . _ , . _ . _ - . , _ _ , . . - . , _.__--.-..,m-.. ._.m.,, ,,m._- . . . , . - - ,
_ _ . ._ Qww ___:.___ :n _.n._zi a*' - u.u 2 __. . a i ._ u:&. _. l . 1 1 l l competent to take the NRC license examinations by the highest level of corpor-
] ate management for plant operation (for example, Vice President for Operations) fj prior to application for the examinations.
I j As*an operating license applicant, WNP-3 is not subject to the one year of ; i
-j experience requirements for cold license SR0 candidates. However, after one j year of station operation, individuals applying for an SR0 license will be i required to comply with the one year experience requirement for hot license SR0' applicants, unless previously experienced in an equivalent position at i another nuclear plant or at a military propulsion reactor. The experience
'; of license applicants in the latter category will be documented by the ap'licant p on a case-by-case basis in sufficient detail so that the staff'can make a find- 'j ing regarding equivalency. SRO license applicants who possess a degree in engineering or applicable sciences are considered to meet the one year expri-
.j ence requirements as an RO provided they: (1) satisfy the requirements set 'l forth in Sections A.1.a and A.2 of Enclosure 1 to the letter from H. R. Denton l to all power reactor applicants and licensees, dated March 28, 1980, and f (2) have participated in a training program equivalent to that of a cold senior j reactor operator applicant. The applicant has not committed to comply with the l above requirements. ] 'i ii Also, the requirement for three months on-shift experience for control room -
$q operators an SRO candidates as an extra person on shift is not required for
-l ,j cold license candidates and, hence, is not applicable to WNP-3. However, WNP-3 will comply with this requirement for hot license candidats after three months (n of station operation.
_] The' applicant's training program includes topics in heat tycuaier, fluid flow,
- q and thermodynamics. However, the applicant has not provided a program for the
. instruction of these topics in accordarce with Enclosure 2 of H. R. Denton's j March 28, 1980 letter. We require the applicant tu provide this program for j review, and we will report our findings in the final SER.
1 Reactor and plant transient training is primarily performed by each license i applicant at a simulator facility and includes classroom discussions of typical i
; transients as well as demonstration of casualty and transient response on the .
3~ _ _ . _ _ 13-11 WNP-3 OSER SEC 13 i
=~ .n.-,-,.- r, ., - p.~,n,,. c.e..- p- n , -. , m .n- . .~, , , -
L a .; a .: = : a w _ : s S $ A .: a . a. _ . N .. ~ - ~.:.u.-..u.-.- = w
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j - 1 l simulator. This knowledge is tested in-depth during the certification examina-tion given by the training facility. Based on our review, we have not been able d)
~
to conclude that the applicant of WNP-3 has satisfied the requirements of this
; item of the TMI Action Plan.
a
']
3 I.A.2.3 Administration of Training Program
- }
j As specified in Enclosure 1 of H. R. Denton's March 28, 1980 letter, we require j that all instructors who teach systems, integrated responses, transient and 1 a simulator courses shall be SRO certified and will continue to participate in appropriate requalification programs. Vendor-supplied instructors who teach q the above subjects shall also be similarly certified. Othermelmbersoftheper-manent or nonpermanent training staff who are responsible for teaching technical subjects, such as reactor theory, heat transfer, fluid mechanics, thermody-namics, tealth physics, chemistry, and instrumentation are not expected to have an RO or SR0 license. Guest lecturers considered to be used on a limited bases shall be monitored by a qualified instructor. These guest lecturers are exempt from the SRO criterion. 1 Based on our review, we find that the applicant of the WNP-3 has not committed to comply with the above requirements of this item of the TMI Action Plan. II.B.4 Training for Mitigating Core Damage q} " As specified in Enclosure 3 of H. R. Denton's March 28, 1980 letter, we require that shift technical advisors and personnel in the operating chain up to and including the plant manager will receive training for mitigating core damage. si Managers and technicians in the instrumentation and control, health physics and chemistry departments will receive-mitigating core damage training commen-
]
- ].j
. . surate with their responsibilities.
m . . Based on our review, we find that the applicant has not complied with the above requirements. In addition, the applicant has not provided, for NRC review, a j training program for mitigating core damage in accordance with the guidance as
- specified in Enclosure 3 of H. R. Denton's March 28, 1980 letter. We will i review this training program when it is received and will report our findings .
; in the final SER.
~
j . . _ 13-12 WNP-3 DSER SEC 13 L q M .._- . - _
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Y.C - _. _& U _ _ _ . _. . .u _ u u .u a-t
'i 13.2.2 Training for Nonlicensed Plant Staff I
j The applicant has described in the FSAR the details of the training given to I ' nonlicensed plant personnel. The training program for nonlicensed personnel f will provide training for maintenance personnel, equipment operators, health ;
- 1. physics and chemistry technicans, management and supervisory personnel, techni-
.j j cal personnel and training instructors. .) T 4 All permanently employed plant personnel will participate in a general employee 4
j training program consisting of, but not limited to radiological health and j safety, quality assurance, industrial safety, plant security, station emergency plan, fire protection and other appropriate plant plans and procedures. f j The applicant has not provided a training program for the Shift Technical
! Advisors (STA). We require the applicant to provide for our review a training program for the STA in accordance with the guidance as specified in NUREG-0737, ! Appendix C. We will report the results of our review in the final SER. ]
j
.4 t
The fire protection training program includes classroom instructions and train-1 ing in fire fighting equipment use, strategies, techniques and periodic drills. We conclude that the applicant's fire protection training program conforms to
'l the guidance given in the Standard Review Plan, Section 13.2.2.II.C.A and is - -l ] acceptable. ] ] Based on our review, we find that the training given for nonlicensed plant staff personnel meets the requirements of 10 CFR Part 19 and Part 50 and follows the fl guidance given in Regulatory Guide 1.8. Therefore, we conclude that the appli-
- j cant's training program for nonlicensed plant staff, with the exception of the
] STA training program, is acceptable.
t 13.3 Emergency Planning
-1 ] The applicant h_as submitted emergency plans required by the upgraded regulations ] on emergency planning that were published in the Federal Register on August 19, 1980, and became effective on November 3, 1980. The regulations contain a i
I 13-1"3 WNP-3 DSER SEC 13 i
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-1 revised Appendix E to 10 CFR 50, " Emergency Planning and Preparedness for Pro- 'l i duction and Utilization Facilities," which establishes minimum requirements for
.I j an acceptable state of onsite emergency preparedness, and a new 10 CFR 50.47,
" Emergency Plans," which specifies standards that must be met- for both onsite
.;, and offsite emergency response. This latter section incorporates the joint -
. NRC/ Federal Emergency Management Agency (FEMA) standards for use in evaluating
] State and local radiological emergency plans and preparedness. The applicant's Emergency Plan, Revision 0 dated April 12, 1982, is in the
! process of review by the NRC. Following this review, requests for additional information may be generated. When this information is submitted, the NRC j staff will review the information and make a finding on its ade'quacy.
NRC and FEMA have agreed that FEMA will make a finding and determination as to ij the adequacy of State and local government emergency response plans. NRC will determine the adequacy of the applicant's emergency response plans with respect f
- j. ' j to the standards listed in 10 CFR 50.47(b), the requirements of Appendix E to 1 10 CFR 50, and the guidance in NUREG-0654, " Criteria for Preparation and Evalua-tion of Radiological Emergency Response Plans and Preparedness in Support of
- j Nuclear Power Plants," dated November 1980. After the above determinations are
~
made by NRC and FEMA, the NRC staff will make a finding in the licensing process as to the overall and integrated state of preparedness. In accordance with - 10 CFR 50.47(a), a full power operating ifcense will not be issued unless the
] NRC staff's overall finding is such that the state of onsite and offsite emer- ;l gency preparedness provides reasonable assurance that adequate protei:tive mea-sures can and will be taken in the event of a radiological emergency. The NRC
] -1 ~ staff will provide its evaluation in a supplement to this report. j 13.6 Industrial Security 1 .
# 13.6.1. Introduction -
,n4 The Washington Public Power Supply System has filed with the Nuclear Regulatory y Commission for the WNP-3 site a Physical Security Plan, Safeguards Contingency j Plan, and a Security Training and Qualification Plan. O
~~
13-14 WNP-3 DSER SEC 13
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,J 1 This Safety Evaluation Report (SER) summarizes how the applicant has provided
,j. .
- 4 for meeting the requirements of 10 CFR Part 73. The SER is composed of a' KJ basic analysis that is available for public review, and a protected Appendix.
d , 1 ,![ 13.6.2 Physical Security Organization j
] To satisfy the requirements of 10 CFR 73.55(b) the Washington Public Power j Supply System has provided a physical security organization that includes a i '
1', j Shift Supervisor who is onsite at all times with the authority to direct the
'] physical protection activities. To implement the commitments made in the
.j ,
. physical security plan, training and qualification plan, and the safeguards
'] contingency plan, written security procedures specifying the duties of the i} security organization members have been developed and are.available for
.j inspection. The training program and critical security tasks and duties for
;- the security organization personnel are defined in the "WNP-3 Security Person-nel Training and Qualification Plan" which meets the requirements of 10 CFR
.] Part 73, Appendix B for the training, equipping and qualification of the
- i security organization members. The physical security plan and the training 1
program provide commitments that preclude the assignment of any individual to a security related duty or task prior to the individual being trained, equipped and qualified to perform the assigned duty in accordance with the approved M guard training and qualification plan. - W -d 13.6.3 Physical Barriers
.d) 1
' 7] In meeting the requirements of 10 CFR 73.55(c) the applicant has provided a
.I
'; protected area barrier which meets the definition in 10 CFR 73.2(f)(1). An j isolation zone, to permit observation of activities along the barrier, of at i least 20 feet is provided on both sides of the barrier with the exception of the locations listed in the Appendix. The staff has reviewed those locations and determined that the security measures in place are satisfactory and con- , tinue to meet the requirements of 10 CFR 73.55(c).
]
i Illumination of 0.2 foot-candles is maintain'ed for the isolation zones, pro- ,j tected area barrier, and external portions of the protected area. In areas S t, d _ . . 13-15
~
WNP-3 DSER SEC 13
,1
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I j' where illumination of 0.2 foot-candles cannot be maintained, special procedures
- j are applied as described in the Appendix.
i
>I ,
s 13.6.4 Identification of Vital Areas -
] The Appendix contains a discussion of the applicant's program and identifies ,j, those areas and equipments determined to be vital.
.i oi 1 Vital equipment is located within vital areas which are located within the pro-tected area and which requires passage through at least two barriers, as defined d in 10 CFR 73.2(f)(1) and (2), to gain access to the vital equipment. Vital area 4 j~ barriers are separated from the protected area barrier.
- j J The control room and central alarm station are provided with bullet-resistant j walls, doors, ceilings, floors, and windows. Based on these findings and the
; j analysis set forth in paragraph D of the Appendix, the staff has concluded that the applicant's program for identification and protection of vital equipment j satisfies the regulatory intent. However, this program is subject to onsite j validation by the staff in the future, and to subsequeat changes if found to be .'j necessary.
lj a
! 13.6.5 Access Requirements -
L.) !J In accordance with 10 CFR 73.55(d) all points of personnel and vehicle access to l, the protected area are controlled. The individual responsible for controlling the final point of access into the protected area is located in a bullet-resistant
; structure. As part of the access control program, vehicles (except under emer-
,} . gency conditions), personnel, packages, and materials entering the protected , l-j area are searched for explosives, firearms and ince "tiary devices by electronic j ~ search equipment and/or physical search. q UI 4 4 Vehicles admitted to the protected area, except licensee designated vehicles, M~ , are controlled by escorts. Licensee designated vehicles are limited to on-site ! station functions and remain in the protected area except for operational main-j tenance, repair, security and emergency purposes. Positive control over the ; vehicles is maintained by personnel authorized to use the vehicles or by the _
~
13-16 WNP-3 DSER SEC 13 J. .. . .- .. . ~ - - .
a_. :. - . . a. - _ = . - . - ~ a . L ... . 2 . u _ . = k- l
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l l
; escort personnel. A picture badge / key card system, utilizing encoded information,
, identifies individuals that are authorized unescorted access to protected and vital areas, and is used to control access to these areas. Individuals not
] authorized unescorted access are issued non picture badges that indicate an
] escort is required. Access authorizations are limited to those individuals who ;
j have a need for access to perform their duties.
.1 1 Unoccupied vital areas are locked and alarmed. During periods of refueling or 'l major maintenance, access to the reactor containment is positively controlled n
j by a member of the security organization to assure that only authorized indivi-duals and materials are permitted to enter. In addition, all doors and personnel / i equipment hatches into the reactor containment are locked and alarmed. Keys, j locks combinat' ions and related equipment are changed on an annual basis. In j addition, when an individual's access authorization has been t,erminated due to j the lack of reliability or trustworthiness, or for poor work performance, the j keys, locks, cobbinations and related equipment to which that person had access 1 are changed. 1 i 13.6.6 Detection Aids
] In satisfying the requirements of 10 CFR 73.55(e) the applicant has installed } intrusion detection systems at the protected area barrier, at entrances to vital -
j i areas, and at all emergency exits. The applicant has exceeded the regulation 1 by providing two separate and dissimilar perimeter intrusion detection systems j at the protected area bar-ter. Alarms from the intrusion detection system annunciate within the continuously manned central alarm station and a secondary i alarm station located within the protected area. The central alarm station is a y located such that the interior of the station is not visible from outside the perimeter of the protected area. In addition, the central station is constructed
- Ji so that walls, floors, ceilings, and doors are bullet-resistant. The alarm j stations are located and designed in such a manner so a single act cannot-inter ~
j dict the capability of calling for assistance or responding to alarms. The ! H central alarm station contains no other functions or duties that would interfere l
] with its alarm response function. The intrusion detection system transmission 1 lines and associated alarm annunciation hardware are self-checking and tamper- -j indicating. Alarm annunciators indicate the type of alarm and its location when Z_ 13-17 WNP-3 DSER SEC 13
_ re .--=~-- - ..,~,w. .....m. , ,-.f...,, , , . _ _ , ,, , , , , _ , , ,
; - - - w - .a.=: a.:..::. =. w 1m-1
~
i a 1
.I t
1 activated. An automatic indication of when the alarm system is on standby power is provided in the central alarm station. ~~
$ 13.6.7 Communications -
i
~
1 i As required in 10 CFR 73.55(f) the applicant has provided for the capability of j continuous communications between the central and secondary alarm station j operators, guards, watchmen, and armed response personnel through the use of a conventional telephone system, and a security radio system. In addition, direct q communication with the local law enforcement authorities is maintained through
-, the use of a conventional telephone system and two-way FM radio links.
j . 1n All non portable communication links, except the conventional telephone system,
;] are provided with an uninterruptible emergency power source.
i 13.6.8 Test and Maintenance Requirements In meeting the requirements of 10 CFR 73.55(g) the applicant has established a
.f program for the testing and maintenance of all intrusion alarms, emergency alarms, ,j communication equipment, physical barriers and other security related devices and ,] equipment. Equipment or devices that do not meet the design performance criteria
,! or have failed to otherwise operate will be compensated for by appropriate com- - pensatory measures as defined in the "WNP-3 Physical Security Plan" and in site procedures. The compensatory measures defined in these plans will assure that '~! the effectiveness of the security system is not reduced by failures or other contingencies affecting the operation of the security related equipment or
;j structures. Intrusion detection systems are tested for proper performance at j the beginning and end of any period they are used for security. Such testing d will be conducted at least once every seven days.
1 9 Communication systems for onsite communications are tested at the beginning j of each security shift. Offsite communications are tested at least once each day. 4 a i Audits of the security program are conducted once every 12 months by personnel independent of site security management and supervision. The audits, focusing.
.o l
13-18 WNP-3 DSER SEC 13 d g , v~ - - - -.s%-_- 4 -
*..-.e# js .#- . m. .mmm., _ .-- -- -~m._ a-- . - - --
- m. -.
. m. u , _ . _ . ,_ .
s 'k e o t.
.}:
1t - on the offectiveness of the physical protection provided by the onsite security j organization implementing the approved security program plans, include, but are
.s
;j not limited to: a review of the security procedures and practices; system testing j and maintenance programs; and local law enforcement assistance agreements. A
- l report is prepared documenting audit findings and recommendations and is submitted .
~
6 to the Plant Management.
,fj q
j 13.6.9 Response Requirements a In meeting the requirements of 10 CFR 73.55(h) the applicant has provided for
.1 .
Nj armed responders immediately available for response duties on all shifts con-f sistent with the requirements of the regulations. Consideratio'ns used in sup-
;} port of this number are attached (see Appendix). In addition, liaison with local
?J enforcement. authorities to provide additional response support in the event of sA j security events has been established and documented. ,y The applicant's safeguards contingency plan for dealing with thefts, threats,
..{ and radiological sabotage events satisfies the requirements of 10 CFR Part 73,
,j Appendix C. The Plan identifies appropriate security events which could initi-
- N
,3 ate a radiological sabotage event and identifies the applicant's preplanning,
'q response resources, safeguards contingency participants and coordination activi-
- N ties for each identified event. Through this plan, upon the detection of ab- -'
y .j normal presence or activities within the protected or vital areas, response i-I activities using the available resources would be initiated. The response activities and objectives include the neutralization of the existing threat by
, l requiring the response force members to interpose themselves between the adver-il .
sary and their objective, instructions to use force commensurate with that used s ij by the adversary, and authority to request sufficient assistance from the local
- 1 j
law enforcement authorities to maintain control over the situation. 4 *
.jr To assist in the assessment / response activities a closed circuit television --
f _ system, providing the capability to observe the entire protected area perimeter, A isolation zones and a majority of the protected area, is provided to the security Q organization. .f i ! .f;) T - 13-19 WNP-3 DSER SEC~13 ,l a ._ m___r._,-,_- -m ,, m._ a _ - - -- m
- - 1
. .. _ _ .s a_ m
.1 ,
I t~ i r .
) 13.6.10 Employee Screening Program l
j In meeting the requirements of 10 CFR 73.55(a) to protect against the design basis threat as stated in 10 CFR 73.1(a)(1)(fi), the Washington Public Power Supply System has provided an employee screening program. Personnel who suc- i
. .1 cessfully complete the employee screening program or its equivalent may be granted -l unescorted access to protected and vital areas at the WNP-3 site. All other .1 personnel requiring access to the site are escorted by persons authorized and .. l .~ t '1 trained for escort duties and who have successfully completed the employee i screening program.
I
.l
- l The employee screening program is based upon accepted industry ' standards and
] includes a background investigation, a psychological evaluation, and a continuing observation program. In addition, the applicant may recognize the screening
.j 1 program of other nuclear utilities or contractors based upon a comparability review. The plan also provides for a " grandfather clause" exclusion which j allows recognition of a certain period of trustworthy service with the utility a or contractor, as being equivalent to the overall employee screening program.
. The staff has reviewed the applicant's screening program against the accepted industry standards (ANSI N18.17 1973) and has determined that the Washington
! Public Power Supply System's program is acceptable.
i j
.5 2.s
~1
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I b i - o d 13-20 WNP-3 DSER SEC 13 l , t k __ ____ . _ _ _
..=-.. - = . = - . - s..-- -.. .a--.-.=...- - - -.=.w . a m uu l
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l:-
-I 1 14. INITIAL TEST PROGRAM ,
q The initial test program encompasses the scope of events that commences with y . t completion of system construction and terminates with the completion of power
,j ascension testing. The initial test program is conducted in three separate and sequential subprograms: system lineup testing, the preoperational test program and the startup test program. At the conclusion of these subprograms, j a unit is ready for normal power operation. The system lineup testing, preoper-j ational, and startup test subprograms are accomplished in the following five
_j distinct and sequential major phases: i
- Phase I - System Lineup Testing 3
Phase II - Preoperational Testing
)
Phase III - Fuel Loading and Post Core Hot Functional Testing 1
- Phase IV - Initial Criticality and Low Power Physics Testing j = Phase V - Power Ascension Testing 4
-)
i System lineup testing includes cleaning and flushing of piping systems and
,i equipment; electrical equipment checks such as insulation resistance measure- - ; ments. phase verification, continuity checks, voltage measurements, grounding 3 checks, circuit breaker operation and relay operation; initial operation of li motors and valves; calibration of instruments; and adjustments of relief and n safety valves.
The preoperational test program commences with system / component turnover from the construction activity to the operations activity, and terminates with the beginning of unit fuel loading. These tests demonstrate, to the extent practi-1 cable, the capability of structures, systems,' and components to meet perfor- _ M mance requirements, and to satisfy design requirements. To the extent practi-
- ).
cable, the objectives of the preor.erational test program are to:
'l 'j (1) Document the performance and operability of equipment and systems m
14-1 WNP-3 DSER SEC 14 1 [} - ,., _ . .
(.- .u ..:.- :- - -. - - . - - - - .. --.;.c..=.-..,..~ . : w..uc-. .. . -=.
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I i - t t (2) Provide base-line test and operating data on equipment and systems for i
.j future reference i
t (3) Operate new equipment for a sufficient time period so that design, j manufacturing, and installation defects can be detected and corrected ; 1 l (4) Ensure that plant systems operate together on an integrated basis
.]
(5) Familiarize the plant operating, technical, and maintenance personnel with
; the plant operation (6) Confirm the adequacy of plant operating and emergency operiting ] procedures.
l
! The startup test program commences with fuel loading, and terminates with the
, completion of power ascension testing. These tests confirm the design bases and
; demonstrate, to the extent practicable, that the plant operates and responds
! to transients as designed. Startup testing is sequenced to ensure that the safety of the plant is not dependent upon the performance of untested struc-
; tures, systems, or components.
i j The objectives of the startup test program are to: -
~1 1
(1) Accomplish a controlled, orderly, and safe initial core loading
.4
,j (2) Accomplish a controlled, orderly, and safe initial criticality i
.l (3) Conduct low power testing sufficient to ensure that design parameters are 1
satisfied, and that safety analysis assumptions are correct or conservative
.3 tl s
- (4) Perform a controlled, orderly, and safe power ascension with requisite i testing, terminating at plant rated conditions.
Y The applicant has made extensive reference to the Combustion Engineering Stan-
! dard Safety Analysis Report (CESSAR). Preoperational and startup test abstracts f pertinent to Combustion Engineering's " System 80" Nuclear Steam Supply System _
l _ f 14-2 WNP-3 DSER'SEC 14
! 1 j
- - - - - -__ A - ~.a.La _ = w. _ p.a.. a =. .w w w l ,
4 j' ' fp (NSSS) are contained in CESSAR, as ar'e other NSSS-related topics (e.g., system descriptions, accident analysis, and Standard Technical Specifications). CESSAR l was evaluated in NUREG-0852. 1 , I j Our review of Chapter 14 or the applicant's FSAR concentrated on the administra- ;
~
tion of the test program and the completeness of%the preoperational and startup
} tests. The Safety Evaluation Report which was issued at the completion of the O, Construction Permit review was s=
reexamined to determine the principal design M criteria for the plant and to' identify any specific concerns or unique design j features that would warrant special test consideration. , Chapters I through 12 J of the FSAR were reviewed for familiarization with the facility design and
. s .
nomenclature. Chapter 15 was reviewed to identify assumptions' pertaining to
-l
- performance characteristics that should be verified by, testing and to identify l all structures, systems, components and design features that were assunied to
! function (either explicitly or implicitly) in the accident analyses. Licensee Event Report Summaries for operating reactors of similar design were reviewed i
to identify potentially serious events and chronic or generic problems that
; might warrant special test consideration. Standard Technical Specifications were reviewed to identify all structures, systeas, and components that would be j relied upon for establishing conformance with safety limits or limiting condi- ] tions for operation. Post-TMI related testing requirements were reviewed in i
conformance with: NUREG-0660, "NRC Action Plan Developed as a Result of the -
) '
THI-2 Accident;" NUREG-0694, "TMI-Related Requirements for New Operating Licenses;" and NUREG-0737, " Clarification of TMI Action Plan Requirements."
) And finally, Startup Test Reports for other Combustion Engineering reactor j plants were reviewed to identify problem areas that should be emphasized in
- 'j the initial test program.
I i l In determining the acceptability of the applicant's test program, the criteria
- of the Standard Review Plan, NUREG-0800, Section 14.2 were used. Our review included verification of the following features of the initial test program:
(1) The applicant plans to develop test procedures u:,ing input from the NSSS l vendor, the architect-engineer, the applicant's engineering staff, and other equipment suppliers and contractors. Operating experiences at simi-g lar plants are being factored into the development of the test procedures I I i -. 14-3 WNP-3 DSER SEC 14 i L_. _ , __m ____m._, ,_ _ , ,~ ~_ .
u =- -- . _ u we . ,_ :- = ..a .,. w w m i: - 4 r 4 l o (2) The applicant plans to conduct tests using approved test procedures.
-p Administrative controls cover (a) the completion of test prerequisites,
] (b) the completion of necessary data sheets and other documentation, and (c) the review and approval of modifications to test procedures. The ',
j applicant has stated that administrative procedures also cover implementa- ; tion of modifications or repair requirements. identified as being required
- j. by the tests and any necessary retesting.
Gj (3) The applicant plans to review the results of each test for technical i]j adequacy and completeness by review groups that include the NSSS vendor and architect-engineer as appropriate. Preoperational test results will ,l be reviewed prior to fuel loading and the startup from eac'h test condition j or power level will be reviewed prior to proceeding to the.next test condi-1 tion or power level. , l (4) The applicant plans to use normal plant operating and emergency operating j procedures in performing the initial test program, thereby verifying the correctness of the procedures to the extent practicable. i! f (5) The schedule for conducting the initial test program a' lows adequate time j to conduct all preoperational and startup tests. Preaperational test pro-j cedures will be available for NRC Regional Administrator review at least -
)i 60 days prior to scheduled implementation. Startur, test procedures will l be available for review not less than 60 days prior to the scheduled fuel loading date.
y i% (6) A description of each test is presented in Chapter 14 of the FSAR. We
.q t verified that there are test descriptions for those structures, systems, 3 components, and design features that: (a) will be used for shutdown and J
g cooldown of the reactor under normal, transient, and accident conditions
.l and for maintaining the reactor in a safe condition for an extended shut-j down period; (b) will be used for establishing conformance with safety limits or limiting conditions for operation that will be included in the
,) facility technical specifications; (c) are classified as engineered safety il features or will be relied on to support or ensure the operations of engi-1 neered safety features within design limits; (d) are assumed to function .
(l C . 14-4 WNP-3 OSER SEC 14 Ji
.,--._. .- n , - - . . .
..=--=,.,n.,, .n- , = _ , .. .,
_z. -J - a -- -. - _ . . w. a _ . x .aw a a -.- 4 f or for which credit is taken in the accident analysis of the facility, as described in the FSAR; or (e) will be used to process, store, cor, trol, or
] limit the release of radioactive materials.
(7) The test objectives, prerequisites, test methods, and acce,:ance criteria ;. u: for each test description are in sufficient detail to establish that the 'M functional adequacy of the structures, systems, cc.aponents, and design I features will be demonstrated. a
] (8) The test program conforms with Regulatory Guide 1.68, "Preoperational and j
- Initial Startup Test Programs for Water-Cooled Power Reactors," Revision 2.
Exceptions have been identified, and technical justificati6n provided. q The applicant made a number of changes to the initial test program because of C. the staff's comments. Examples of these changes are:
;)
j (1) Testing.was added to more accurately determine the Reactor Protection and j Engieered Safety Feature System trip response times. '). (2) Testing was added to verify steam generator safety relief valve operability.
~l (3) Testing was modified to verify that no blockages exist in the containment - ] spray nozzle flow path.
4 7 (4) Natural circulation tests were expanded and will be repeated for training 1-1 purposes to comply with TMI-2 Action Plan Item I.G.1 for low power training and testing.
-} (5) Testing was added to provide improved assurance of proper auxiliary feed-water system performance.
k d (6) Testing was added to verify the heat removal capacity of ECCS coolers
-1 .i during post-accident conditions.
(7) Testing was added to verify the capabilities of the emergency lighting system. _
-t i 14-5 WNP-3 DSER SEC 14 1
i 1
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j i
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(8) Testing was added for the loose parts monitoring system.
~i' (9) Testing was added to demonstrate operability and perform leak tests of f; sectionalizing devices in the spent fuel stordge pool. - 'j; .
3 q (10) Testing was added.to demonstrate that, for hot containment penetrations
,?) where coolers are not used, concrete temperatures do not exceed design ;j limits.
s.
..t l The following items remain unresolved:
- s d,
j Request for Additional Information (RAI) Question Number Question '} 640.01 Tests which may be waived or rescheduled, conditional on I the results of earlier tests, should be identified. 1
~I.
640.03 FSAR Section 14.2.11 should state that fuel loading and }j startup test procedures should be available at least 60 H days prior to fuel loading. 640.04 Individual test descriptions should be expanded to indicate -
., the sources of acceptance criteria.
d 4 640,08
] FSAR Subsection 14.2.12.2.23 (ECCS Area Ventilation) should i state that data will be extrapolated te verify design heet s removal capability as stated in response to this item.
i 640.09 (1) FSAR Subsection 14.2.7 should state the level of con-
- formance with the testing requirements of Regulatory 4 Guide 1.95, " Protection of Nuclear Power Plant Control --
1
,d - Room Operators Against an Accidental Chlorine Release."
A d., (2) FSAR Subsection 14.2.12.2.48 (Control Room Leak Rate
; Test) should state conformance with Regulatory Guide i
i 1.95 acceptance criteria, or the testing described in. this test abstract should be referenced by," or 14-6 WNP-3 DSER SEC 14 m q .es**'waeep&*up eepesy - 9 dp 7pmW gg- _,,e.g
'=f _e.te'@
y e,A g M Mgirht @.gy%=ey- W 1 m'Gr3'- L =^^ " ' "^-
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- - . .c,._u_ w . ___- u -mm,a _. -.i. . - . - -...a__, - ,
i incorporated into, FSAR Subsection 14.2.12.16 (Control y 1 Room HVAC). W . 640.11 (1) FSAR Subsection 14.2.12.2.8 (Diesel Generators g ' and Auxiliary Systems) should be modified to ; Q demonstrate that testing is conducted in accord-d ance with Regulatory Guide 1.108, Positions C.2.a 3 (1-9) and C.2.b. <Q
;A N 640.12 (1) The Essential 125V DC Preoperational Test descrip-4- tion should be clarified to ensure that testing 1 will verify the ability of equipnient to start and
< ;2 _{q operate with batteries at minimum voltage. . N.g Mt (2) The Standby Transformers Preoperational Test jj . description should be revised to demonstrate
; proper operation of transformer cooling under
$ rated load, or how part-load test data will be n]{
+ .
used to verify full-load capability. l&
] 640.14 (1) Provide the following test abstracts, modify
, existing test abstracts, or provide justification -
, .78 q for exception with the following positions of k Regulatory Guide 1.68, Appendix A:
cd
'a (hf 1.b (1) Service Water System n?
1.n (16) Heating systems for the ]j? refueling water storage tank
>wn 4
(, 5.n Obtain baseline data for the
~
loose parts monitoring system M a
?l 14A FSAR Subsection 14.2.7 (Conformance of Test Programs with Regulatory Guides) states that the WNP-3 test
, program " generally conforms to the requirements of" _
!M n
]
~~
.14-7 WNP-3 DSER SEC 14 d I J i i _ _. . , . , - - _ - . _ ,
,<- .~. -. - aa.a.as- > .w ..a.- - u. :.u. x . t..~ .
the listed regulatory guides. Either state that the WNP-3 test program conforms to the applicable guides, or provide justification for any exceptions taken.
,1 148 FSAR Subsections 14.2.12.5 and 14.2.12.8 state that ; .j -
additional tests will be provided at a later date. Either provide these additional tests, or modify i these sections accordingly. .-Q 9 I Based on our review of the FSAR Section 14.2 as amended through Amendment 5, 3 we have concluded that (with the exception of the items described above) the f initial plant test program is acceptable and meets the requirements of 10 CFR Part 50, 650.34(b)(6)(fii) that requires inclusion of. plans for.preoperational testing and initial operations in the FSAR and 10 CFR Part 50, Appendix B, Sec-q tion XI that requires a test program to assure that all testing required to
;. demonstrate that structures, systems, and components will perform satisfactorily ] in service is identified and performed in accordance with written test proce-dures which incorporate the requirements and acceptance limits contained in 1 appIicable design documents. We have further concluded that if acceptable responses to the above items are made, then the initial test program described in the application will meet the acceptance criteria of Section 14.2 of the ;j Standard Review Plan, NUREG-0800, and the successful completion of the test - ~] program will demonstrate the functional adequacy of plant structures, systems, si and components. .j' This review and evaluation was performed with the assistance of Battelle Pacific 5d Northwest Laboratories' personnel. ~
Future changes to this approved test program should be submitted to the staff j with justification tar the changes for review by the staff. If a change relates $ to individual test descriptions, the justification should consider the safety-j related categories enumerated above pertaining to the criteria of Standard Review Plan 14.2. di a a . y -
'i s 14-8 WNP-3 DSER SEC 14
[ q A
, _._,,._m._,,_..._.
~. - , ,
am... ~. w-. :a - o :. =: . . .a a w ch I 1 i
.1
- 1
,1; g 15 ACCIDENT ANALYSES ,
, 15.4 Reactivity and Power Distribution Anomalies - '1 ij d The input for the following items are identical to the corresponding items in h
7:1 the CESSAR FSAR: g 15.4.1 CEA Withdrawal from Low Power
- 15.4.2 CEA Withdrawal from Full Power 15.4.3 CEA Misoperation Events 15.4.7 Fuel Misloading Event A
si 15.4.8 CEA Ejection Event d, J 15.x Radiological Consequences of Design Basis Accident
.]
The WNP-3 FSAR references the CESSAR FSAR in areas pertaining to the NSSS. Consequently, the staff has not analyzed certain accidents but has determined whether or not the interface requirements for CESSAR as specified below have been satisfied for WNP-3. The accidents which have not been specifically ana-g lyzed include 's A Q *- Steam Line Break Accidents
- = Reactor Coolant Pump Locked Rotor Steam Generator Tube Rupture d
- Rod Ejection Accident and a
j
- Small Line~ Break Accident y
For those accidents stated above, the CESSAR SER has established the foll.owing j site-related interface requirements for CESSAR reference plants on the basis
. of the analyses described in the CESSAR SER and its supplements:
(1) Coolant activity: 0.1 uC1/gm dose equivalent I-131 (DEI-131) for secondary
' coolant maximum fission product concentration; 1.0 uti/gm DEI-131 for
'i
;q-l ;
. _ 15-1 WNP-3 DSER SEC 15 n,,,.n-r,_,7,.,~y--.,..e._---~..~._-_,--. - - _ ~ ~
s m, nau - - ~ ~
- v .....m . . . ,3gMfQiG_3,_ ;f:
1 j primary coolant maximum equilibrium fission product concentration and a i spike limit of 60 uCi/gm DEI-131. (2) Steam generator tube leakage: 1.0 gallon per minute (gpm) primary to
~
secondary. - j' (3) Maximum condensate storage tank iodine concentration: 0.1 uCi/lb, DEI-131. (4) Atmospheric dispersion factors (X/Q, sec/m3 ) equal to or less than: 2.5E-3* (~ ,s (2.0E-3 for radiological dose evaluation of Reactor Coolant Pump Lockea g Rotor Accident) for 0-2 hour post accident X/Q at the Exclusion Area Bound-
? ary; 1.0E-4 for the .0-8 hour X/Q at the outer boundary of the Low Population 1
j Zone; 2.8E-5 for the 1-4 day X/Q at the outer boundary of the Low Population t ii Zone; and 8.3E-6 for 4-30 day X/Q at the outer boundary of the Low Popula-llj tion Zone. H The atmospheric dispersion parameters are given in Section 2.3 of this report, id. Based on these values, the staff concludes that the above meteorological param-
] eters envelope those of the WNP-3 site and therefore, the accident interface requirements on the meteorological parameters are met.
a The staff will also ensure that the technical specification interface require-
-1 ments related to coolant activity, steam generator tube leakage and condensate storage tank iodine concentration levels identified above will be included in the WNP-3 plant technical specifications. -
/: u 15.x.1 Loss-of-Coolant Accident i The applicant has selected and analyzed a hypothetical design basis LOCA and M p; has shown that the distances to the Exclusion Area and to the Low Population b g Zone Boundaries are sufficient to provide reasonable
- 2.5E-3 = 2.5 x 10 8 as-
{ surance that the radiological consequences of such an accident are within the f guidelines in 10 CFR 100.1 (a)(1) and (2). The analysis has included the fol-
-l lowing sources and radioactivity transport paths to the atmosphere:
(1) contribution from contair.ent leakage; (2) contribution from post-LOCA leakage from ESF systems outside containment; l and f (3) contribution from containment purge. . i 15-2 WNP-3 DSER SEC 15 1
~i 1 -
~ _ _ _ __ _ _.. _ .. _ _ _ _. ._ _ _.- ......_ _ - . . _ . . ~ . , _ _
u .w a ~ ' - . w:c % .x a .. .~.. .., .. . .-~ lq .
-i .
d The staff review has confirmed the applicant's findings based on the following:
.3 1 (1) The applicant's provisions for a design of the containment system
?) and the shield building ventilation system (SBVS) are acceptable as l identified in Section 6 of this report. ;
~
(2) The staff's independent analysis of the radiological consequences of a hypothetical design-basis LOCA as described below. hj (a) Containment Leakage' Contribution
] The WNP-3 plant includes a double containment design to collect q
g and filter the leakage of fission products from a postulated
.j design-basis LOCA. The double containment consists of a free- .i standing steel primary containment vessel surrounded by a rein-d forced concrete shield building. The reinforced concrete auxil-iary building is also a part of the secondary containment barrier.
Leakage from the primary containment which enters the secondary a containment is treated by either the SBVS or the EA/FHBFES before its release to the atmosphere. Both of these systems are engi-neered safety features (see SER Sections 6.5.1 and 9.4 for a description of these systems). Another ESF in the primary con-tainment is the containment spray system with a sodium hydroxide il additive to enhance the removal of iodine in the containment - following a LOCA (see SER Section 6.5.2 for a system description). a d The principal assumptions employed in the staff analysis are JJ
*s listed in Table 15.2. The dose model and dose conversion ,saram-llf aters are consistent with those given in Regulatory Guide 1.4,
,}/] " Assumptions Used for Evaluating the Potential Radiological Con-g sequences of a Loss-of-Coolant Accident for Pressurized Water
. Reactors."
r[j j In the analysis of the design basis LOCA, the primary contain-d ment was assumed to leak at the design leak rate of 0.5 percent per day for the first 24 hours following 'the accident and at j 0.25 percent per day thereaf ter. The applicant established to j the satisfaction of the staff that the shield building annulus . 1 1
' ' ~
1 15-3 WNP-3 DSER SEC 15 l i 4
...----.c.cy- - e.--wg...+.yp,q. e-ws,ee--.e. v ee + w- .w- w-,
-% w 7 -.4-----=
_w = xn ui , l n0.?si2._t . u zz . . ~. ax.- d.. g -
'1 i
- i
;j and auxiliary building would not experience a period of post 1
fj accident positive pressure and, hence, no exfiltration. There-A fore, no leakage into the secondary containments was assumed to 3 bypass an ESF gas treatment system throughout the course of the I..I accident. ; 71 _ 9 - d Forty perceat of the leakage from the primary containment was d
- -t assumed to enter the shield building annulus were it was pos-
] tulated to go directly to the intake of the SBVS. Thirty-eight N percent of the leakage from the primary containment was assumed FJ to enter the auxiliary buildina tihere the staff also assumed .;; that it went directly to the intakes of th'e EA/FHBFES. Because j neither of these systems contain any recirculation capabilities, l, no credit for holdup or mixing in either the shield building 3j annulus or the auxiliary building was assumed. The remaining twenty-two percent was assumed to bypass the secondary containntent I
and go directly to the environment. In response to FSAR question j 450.7, the applicant has committed to incorporate these leakage pathway fractions into the plant technical specification for containment leakage. Incorporation of these fractions into the technical specifications will assure that the containment is tested in a manner consistent with this LOCA evaluation. - h (b) post-LOCA Leakage from ESF Systems Outside Containment da As part of the LOCA, the staff also evaluated the consequences q of leakage of containment sump water that is circulated by the
- n, ECCS after the postulated accident. During the recirculation f mode, the sump water is circulated outside containment to the j auxiliary building. If a significant leak should develop, a h .
fraction of the iodine in the water could become airborne in the q auxiliary building and exit to the atmosphere. Because the ECCS - ] areas in the auxiliary building of WNP-3 are served by an ESF atmospheric cleanup system (the EA/FHBFES), doses from passive si i failures were not considered as specified in SRP Section 15.6.5,
'. !, Appendix B. The staff evaluated the potential radiological con- I sequences from normal ECCS component leakage by assuming a tota) ! - l -j %, -;- ~
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ECCS component leak rate of I gpm. The resulting radiological consequences were only 6 Rem to the thyroid at the Exclusion
'; Area Boundary and 10 Rem to the thyroid at the Low Population 1 ., Zone. -
9 t : a Staff Conclusions 1 M The staff's calculated thyroid and whole body doses from the hypothetical LOCA i are listed in Table 15.1. The staff concludes that the distances to the Exclu-1 y sion Area and to the Low Population Zone Boundaries of the WNP-3 site, in con-Jj junction with the ESF's of the WNP-3 design, are sufficient to provide reason-4 able assurance that the total radiological consequences of a postulated LOCA ] will be within the guidelines set forth in 10 CFR 100. The conclusion is based i - upon the staff review of the applicant's analyses,.and on an independent ana-Q lysis performed by the staff to verify that the total calculated doses are with-l in the guidelines. N ,j 15.7 Radioactive Releases from a Subsystem or Component r; 9 15.7.4 Fuel Handling Accident
.fi
[ For the analysis of the fuel-handling accident in the fuel pool, the staff as-
. sumed that a fuel assembly was dropped in the fuel pool during refueling opera- -
tions'and that the equivalent of all of the fuel rods in the dropped assembly 1 were damaged, thereby releasing the volative fission gases from the fuel rod
- 1. gaps into the pool. The radiation monitors located'above the pool would rapidly
- . detect the release of activity from the pool and initiate the ESF grade
, EA/FHBFES. The radioactive material that would escape from the fuel pool was .s f.} assumed to be released to the environment as a " puff" release, with the iodine . activity reduced by filtration through the EA/FHBFES. The radiological conse-quences following the postulated accident are shown in Table 15.1 and the as-m ] sumptions and parameters used in the analysis are shown in Table 15.3. The Ej dose model and dose conversion factors employed in the analysis were the same
'.l as those given in Regulatory Guide 1.25, " Assumptions Used for Evaluating the j Potential Radiological Consequences of a Fuel Handling Accident in the Fuel i Handling and Storage Facility for Boiling and Pressurized Water Reactors." 1 O
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The staff has also evaluated the consequences of a fuel-handling accident inside primary containment. The applicant states that at all times during refueling 1
- operations the containment will be ventilated to the atmosphere through the d =
, reactor building purge system. -
j' . In response to FSAR question 450.10, the applicant ha demonstrated that a re-
] ] lease from a fuel handling accident can be rapidly de a:ted and the containment ]') isolated prior to the release of any material from tb c- + x.ent..
ll 3 The staff finds that the applicant has provided an adequate system to mitigate 1 the radiological consequences of a postulated fuel-handling accident inside the containment and in the spent fuel pool area. The staff concludes that the fuel-
] handling area ventilation system meets the relevant requirements of GDC 61.
- i The staff further concludes that the distances to the Exclusion Area and the
,f Low Population Zone Boundaries, in conjunction with the operation of the dose
!l mitigating ESF and implementation of plant procedures, are sufficient to provide reasonable assurance that the calculated offsite radiological consequences of a postulated fuel-handling accident are well within the 10 CFR 100 exposure
] guidelines. -
')
The staf'f's conclusion is based on (1) the staff's determination'that the design
- c features and plant procedures at WNP-3 meet the requirements of GDC 61 with -
respect to radioactivity control; (2) the staff review of the applicant's as-
; sumptions and analyses of the radiological consequences from the fuel-handling ) accident; (3) the staff's independent analyses using the assumptions in Regula-
-] ~ tory Guide 1.25, Sections C.1.a through C.I.k, and (4) the WNP-3 Technical Spec-ifications relating to fuel-handling and ventilation system operation.
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l Table 15.1 Radiological consequences of design-basis accidents I
.l 1
i i Exclusion Area low Population
! Boundary (Rem) Zone Boundary (Rem) =
i Postulated Accident Thyroid Whole Body Thyroid Whole Body
.b ;
~
il
'l Loss of coolant accident
.} (LOCA)
, Containment leakage 0-2 hr 194 16 - -
0-8 hr - - 76 4.9
; 8-24 hr - -
33 1.2
; 24-96 hr - -
27 0.3 l 96-720 hr _ 19 0.1 Total Containment leakage 194 16 155 6.4 i j ECCS component leakage 5.7 0.01 9.6 0.01 t 1 j Total (LOCA) 200 16 160 6.4 i j Fuel-handling accident '1.4 0.5 0.2 0.1 i in fuel-handling area
.i l
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, Table 15.2 Assumptions used to calculate loss-of-coolant accident doses ] Power level, Mwt 4100 .
- s .
) Operating time, years 3 .
4 . Rj Fraction of core inventory available for leakage, % M Iodines 25 d Noble Gases 100 9, ij Initial iodine composition in containment, % a Elemental 91.0
.j Organic 4.0 Particulate 5.0
. j Primary containment volumes, cu f t
!! Sprayed 2.8E6*
1.y Unsprayed 4.2E5
$ Primary containment leakage rate, %/ day
' E! 0-24 hours after accident 0.5
'i after 24 hours 0.25 O Containment leakage pathways, %
Reactor building 40 Auxiliary building 38
- Bypass 22
.4
-1 Containment spray system effectiveness, inverse hr Elemental iodine removal coefficient 10
, Organic iodine removal coefficient 0 -
,, Particulate iodine removal coefficient 0.45 1
Atmospheric diffusion factors (second per cubic meter)
.. 0-2 hour Exclusion Area Boundary 4.1E-4 d) 0-8 hour Low Population Zone Boundary 6.0E-5
'4 8-24 hour Low Population Zone Boundary 4.0E-5 fj ~
1-4 day Low Population Zone Boundary 1.6E-5 j 4-30 day Low Population Zone Boundary 4.3E-6 , c
- 2.8E6 = 2.8 x 10' = 2,800,000
.i q
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1 - j Table 15.3 Assumptions used for estimating ?.he radiological ,! consequences following a postulated fuel handling accident
.l l Parameter and unit of measure Quantity . ,
I
-j ;
Power level, Hwt 4100 - Number of fuel assemblies damaged 1
, Total number of fuel assemblies in core 217 Peaking factor of damaged rod 1.65
- Shutdown time, hr 72.0
- Inventory released from damaged rods (iodines and noble
^1 gases),% 10 i Pool decontamination factors
. Iodine 100 N} Noble gases -
1 1 Iodine fractions released from pool, % . Elemental .75 Organic 25 Iodine removal efficiencies, % Elemental 99 l Organic 99
! X/Q values, sec/ cubic meter
-1 0-2 hour EAB 4.1E-4 j 0-8 hour LPZ 6.0E-5
$6 aa 3
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/;t 31
!j 18 HUMAN FACTORS ENGINEERING .
2 . cl .I
.j s], Position i ,
3 Licensees and applicants for operating licenses shall conduct a Detailed Con-31
- ,j v.
i trol Room Design Review (DCRDR). The objective is to " improve the ability of
^
- . -} nuclear power plant control room operators to prevent accidents or cope with u accidents if they occur by improving the information provided to them" (NUREG-0660, Item I.0). The need to conduct a DCRDR was confirmed in NUREG-0737 and Supplement 1 to NUREG-6
- J. DCRDR requirements in Supplement I to d NUREG-0737 replaced those in earlier documents. Supplement I to NUREG-0737 j' requires each applicant or licensee to conduct a DCRDR on a schedule negotiated 4'd with the Nuclear Regulatory Commission (NRC).
4 i
.}
NUREG-0700 describes four phases of the DCRDR and provides applicants and licen-l sees with guidelines for its conduct.
.c c) li- The phases are ifj (1) planning (2) review
'] i
- -)' (3) assessment and implementation iU (4) reporting
(
- l Criteria for evaluating each phase are contained in Appendix A to the Standard 1
! ).: Review Plan (SRP), Section 18.1 of NUREG-0800. '
- c. ;
.s
]1 A Program Plan is to be submitted within two months of the start of the DCRDR. _. .j Consistent with the requirements of Supplement 1 to NUREG-0737, the Program 1, Plan shall describe how the following elements of the DCRDR will be accomplished: ' J' f. (1) ' Establishment of a qualified multidisciplinary review team.
.I m
1 . I l 18-1 WNP-3 DSER SEC 18 t
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.i (2) Function and task analyses to identify control room operator tasks and information and control requirements during emergency operations.
l
.j (3) A comparison of display and control requirements with a control room 1 inventory. ~ ,i (4) A control room survey to identify deviations from accepted human factors principles. ~i o (5) Assessment of human engineering discrepancies (HEDs) to determine which l .j HEDs are significant and should be corrected.
- j (6) Selection of design improvements.
'l .
{ (7) Verification that selected design improvements will provide the necessary
'l correction. '
4 i (8) Verification that improvements will not introduce new HEDs.
,} (9) Coordination of control room improvements with changes from other programs ,j such as SPDS, operator training, Reg. Guide 1.97 instrumentation and up- ; graded emergency operating procedures. - ] A Summary Report is to be submitted at the end of the DCRDR. As a minimum it j shall 1 $l (1) outline proposed control room changes
~l
-4
~j (2) o wline proposed schedules for implementation i .
~1 3 (3) provide summary justification for HEDs with safety significance to be left -
j 4 uncorrected or partially corrected
'l ] The NRC will evaluate the organization, process, and results of the DCRDR.
j Evaluation will include review of required documentation (Program Plan and ] Summary Report) and may also include reviews of additional documentation,
. .T ' 18-2 WNP-3 DSER SEC 18 1.c
,_ u.- -. w a - :- =- - -
a =.a.L . 2 - z . .. 2==- -. = -c l i A ' ]I briefings, discussions, and on-site audits. In progress audits may be conducted j
" after submission of the Program Plan but prior to submission of the Summary Re-Tj port. Pre-implementation audits may be conducted after submission of the Sum-
.j mary Report. Evaluation will be in accordance with the requirements of Supple-
! ment I to NUREG-0737. Additional guidance for the evaluation is provided by ; ,
NUREG-0700 and NUREG-0800, Appendix A to SRP Section 18.1. Results of the NRC evaluation of a DCRDR will be documented in a Safety Evaluation Report (SER) or ' ]i SER Supplement. t NUREG-0737 Supplement 1 also requires that each operating reactor be provided with a safety parameter display system (SPDS) that is located convenient to the 11 control room operators. This system will continuously display information from
]
j with the plant safety status can be readily and reliably assessed. The princi-pal purpose and function of the SPDS is to aid the control room personnel during i abnormal and emergency conditions in determining the safety status of the plant 1 l and in assessing whether abnormal conditions warrant corrective action by opera-l tors to avoid a degraded core. A written SPDS safety analysis shall be prepared . t describing the basis on which the selected parameters are sufficient to assess t the safety status of each identified function for a wide range of events, which "i include symptoms of severe accidents. The applicant's safety analysis and SPDS
-) implementation plant will be reviewed by the NRC staff to confirm: (1) the ade-quacy of the parameters selected to be displayed to detect critical safety -
functions; (2) that means are provided to assure that the data displayed are valid; (3) the adequacy of the design and installation of the system from a human factors perspective; and, (4) the adequacy of the verification and valida-
] tion (V&V) program to assure a highly reliable to SPDS.
l g Status I As requested by Generic Letter 82-33, in its letter of April 14, 1983, the j Washington Public Power Supply System (Supply System) submitted a proposed j schedule for activities required by Supplement I to NUREG-0737. By letter dated July 12, 1983, the Supply System submitted its " Control Room Design Review Program Plan of WPPSS Nuclear Project 3." The staff did not complete i its review of the Program Plan. I ,l \
] The staff review will be resumed when the project is reactivated. -
I +l 18-3 WNP-3 DSER SEC 18 l )i _.~ . _ _ _ . _ _ . _ _ , . _ . . _ , _ - . _ _
u- - a _:: ~ . . . .- _ .2. j. -w - a.a t i REFERENCES l
<i
-; 1. NUREG-0660, Volume 1, May 1980; NRC Action Plan Developed as a Result of f the TMI-2 Accident. -
! 2 j 2. NUREG-0737, November 1980; Clarification of TMI Action Plan Requirements.
.i j 3. Supplement 1 to NUREG-0737, December 1982; Requirements for Emergency d . Response Capability (Generic Letter 82-33).
l j 4. NUREG-0700, September 1981; Guidelines for Control Room Design Reviews.
- 5. Letter to G. W. Knighton, NRC, from G. D. Bouchey, Washington Public Power I Supply System,
Subject:
Nuclear Project 3 Response to Generic Letter t j No. 82-33 Rcquirements for Emergency Response Capability, dated April 14 1983.
) 6. Letter to G. W. Knighton, NRC, from G. D. Bouchey, Washington Public Power Supply System,
Subject:
Nuclear Project 3 Control Room Design Review Program Plan, Dated July 12, 1983. i l
^l -l:
I l
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; 18-4 WNP-3 DSER SEC 18 1
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