ML20148T503
| ML20148T503 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 11/30/1978 |
| From: | Sherwood G GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20148T502 | List: |
| References | |
| NUDOCS 7812050221 | |
| Download: ML20148T503 (197) | |
Text
O UNITED STATES 0F AMERICA NUCLEAR REGVLAT0RY C 0 M M I S S I 0"N 1
I In the Matter of )
General Electric Company) Docket No. STN 50-447 Standard Plant )
l AMENDMENT N0. 46 TO APPLICATION FOR EXTENSION OF GESSAR-238 NUCLEAR ISLAND General Electric Company, applicant in the above named proceedings, hereby files Amendment No. 46 to the General Electric Standard Safety Analysis Report (238 GESSAR).
Amendment No. 46 consists of one part entitled Appendix C and amends the 238 GESSAR by providing written assessment of the GESSAR-238 Nuclear O is,emo oesioa e9eimet ceteserv 1. 11. 111 eme iv metters ennroved ro the Staf f since March 1,1974, which is the regulatory requirement cutoff date for the GESSAR-238 Nuclear Island.
We feel Amendment No. 46 fulfills the conditions necessary for granting extension of the GESSAR PDAs.
Respectfully Submitted General Electric Company By: ~
Glenn G. Sherwood, Manager Safety & Licensing Operation Subscribed and sworn to before me the 30th day of November, 1978.
By: vM b . /L
/Y NOTARY PUBLIC a
adbah"WMa C:ccecWAVww.~n m p OFFilCIAL SEAL h Q. !
m RUTHE M. KINNAMON
, NOTARY PUBUC - CAUFORNIA
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SANTA CLARA COUNTY i en m A m n_ ~.s m a ms
NUCLE AR ENERGY BUSINESS GROUP e GENER AL ELECTRIC COMPANY SAN JOSE, CALIF O RNI A 95125 O GEN ER AL h ELECTRIC APPLICABLE TO:
PUBLICATION NO.
ERRMA And ADDENDA T.1. E. NO. g TITLE General Electric Standard 46 go, Safety Analysis Report DATE November 1978 O April 1973 NOTE: Correct all copies of the applicable ISSUE DATE publication as specified below.
REFERENCES PARAG APH L NE) (CO R R E C NS AN ADDITIONS) 1 Appendix C Insert new Appendix C.
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APPENDIX C EXTENSION REVIEW MATTERS FOR PRELIMINARY DESIGN APPROVALS O
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O O O O O APPENDIX C TABLE OF CONTENTS PARAGRAPH C.1 Introduction C.2 Assessments C.2.1 Part I Category I Matters Approved by RRRC and Issued from March 1974 through August 1978 C.2.1.1 Regulatory Guide 1.7 Control of Combustible Gas Concentrations Revision 2 in Containment Following a Loss-of-Coolant Accident C . 2.1. 2 Regulatory Guide 1.9 Selection, Design, and Qualification for Revision 0 Diesel-Generator Units Used as Onsite Electric Power Systems at Nuclear Power n
1 Plants C.2.1.3 Regulatory Guide 1.20 Comprehensive Vibration Assessment Program Revision 2 for Reactor Internals During Preoperational and Initial Startup Testing C.2.1.4 Regulatory Guide 1.28 Quality Assurance Program Requirements Revision 1 (Design and Construction)
C.2.1.5 Regulatory Guide 1.29 Seismic Design Classification Revision 2 C.2.1.6 Regulatory Guide 1.31 Control of Ferrite Content in Stainless Revision 2 Steel Weld Metal d
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C.2.1.7 Regulatory Guide 1.32 Criteria for Safety-Related Electric
- Revision 2 Power Systems for Nuclear ~ Power Plants '
C.2.1.8 Regulatory Guide 1.33 Quality Assurance Program Requirements '
Revision 1 (Operation)
! C.2.1.9 Regulatory Guide 1.35 Inservice Inspection of Ungrouted Revision 2 Tendons in Pre-stressed Concrete l
Containment Structures C.2.1.10 .Pegulatory Guide 1.38 Quality Assurance Requirements for i Revision 2 Packaging, Shipping, Receiving, Storage ;
and liandling of Items for Water-Cooled ,
Nuclear Power Plants i
C.2.1.ll Regulatory Guide 1.39 Housekeeping Requirements for Water-Revision 2 Cooled Nuclear Power Plants C . 2.1.12 Regulatory Guide 1.52 Design, Testing, and Maintenance for
, Revision 2 Engineered Safety Feature Atmosphere T' Cleanup System Air Filtration and 22 Absorption Units of Light Water Cooled Nuclear Power Plants C . 2.1.13 Reguldtory Guide 1.63 Electric Penetration Assemblies in i Revision 1 Containment Structures for Light-Water-Cooled Nuclear Power Plants i
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l C.2.1.14 Regulatory Guide 1.64 Quality Assurance Requirements for the Revision 2 Design of Nuclear Power Plants C . 2.1.15 Regulatory Guide 1.68 Initial Test Programs for Water-Cooled Revision 2 Reactor Power Plants C . 2.1.16 Regulatory Guide 1.68.1 Preoperational and Initial Startup Revision 0
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Testing of Feedwater and Condensate Systems for Boiling Water Reactor Power Plants I E$
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C.2.1.17 Regulatory Guide 1.72 Spray Pond Plastic Piping Revision 1 C.2.1.18 Regulatory Guide 1.84 Code Case Acceptability - ASME Section Revision 12 III Design and Fabrication C.2.1.19 Regulatory Guide 1.85 Code Case Acceptability - ASME Section Revision 12 III Materials :
C . 2.1. 20 . Regulatory Guide 1.90 Inservice Inspection of Pre-stressed Revision 1 Concrete Containment Structures with
< Grouted Tendons
, C .2.1. 21 Regulatory Guide 1.92 Combining Modal Responses and Spatial Revision Components in Seismic Response Analysis C.2.1.22 Regulatory Guide 1.94 Quality Assurance Requirements for Revision 1 Installation, Inspection, and Testing ,
of Structural Concrete and Structural c3 Steel during the Construction Phase of 2, Nuclear Power Plants C . 2.1. 23 Regulatory Guide 1.95 Protection of Nuclear Power Plant Revision 1 Control Room Operators Against an Accidental Chlorine Release I
C.2.1.24 Regulatory Guide 1.99 Effects of Residual Elements on Revision 1 Predicted Radiation Damage to Reactor Vessel Materials C.2.1.25 Regulatory Guide 1.100 Seismic Qualification of Electric Revision 1 Equipment for Nuclear Power Plants C.2.1.26 Regulatory Guide 1.103 Post-Tensioned Pre-stressing Systems Revision 1 for Concrete Reactor Vessels and Containments C . 2.1. 27 Regulatory Guide !.106 Thermal Overload Protection for Electric '
-; Revision 1 Motors on Motor-Operated Valves 8
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O O O O O C.2.1.28 Regulatory Guide 1.107 Qualifications for Cement Grouting Revision 1 for Pre-stressing Tendons in Containment Structures C . 2.1. 29 Regulatory Guide 1.116 Quality Assurance Requirements for Revision 0-R Installation, Inspection, and Testing of Mechanical Equipment and Systems C.2.1.30 Regulatory Guide 1.118 Periodic Testing of Electric Power and Revision 1 Protection Systems C.2.1.31 Regulatory Guide 1.120 Fire Protection Guidelines for Nuclear Revision 1 Power Plants C.2.1.32 Regulatory Guide 1.122 Development of Floor Design. Response Revision 1 Spectra for Seismic Design of Floor-Supported Equipment or Components C.2.1.33 Regulatory Guide 1.123 Quality Assurance Requirements for Revision 1 Control of Procurement of Items and o Services for Nuclear Power Plants C.2.1.34 Regulatory Guide 1.126 An Acceptable Model and Related Revision 1 Statistical Methods for the Analysis of Fuel Densification C.2.1.35 Regulatory Guide 1.128 Installation Design and Installation Revision 0 of Large Lead Storage Batteries for Nuclear Power Plants C.2.1.36 Regulatory Guide 1.129 Maintenance, Testing, and Replacement Revision 0 of Large Lead Storage Batteries for Nuclear Power Plants C.2.l.37 Regulatory Guide 1.131 Qualification Tests of Electric Cables, Revision 0 Field Splices and Connections for Light Water Cooled Nuclear Power Plants C
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O O O O O C.2.1.38 Regulatory Guide 1.132 Site Investigations for Foundations of Revision 0 Nuclear Power Plants C.2.1.39 Regulatory Guide 1.134 Medical Certification and Monitoring Revision 0 of Personnel Requiring Operator Licenses C.2.1.40 Regulatory Guide 1.135 Normal Water Level and Discharge at Revision 0 Nuclear Power Plants C.2.1.41 Regulatory Guide 1.136 Material for Concrete Containments Revision 0 C.21.142 Regulatory Guide 1.137 Fuel Oil Systems for Standby Diesel Revision 0 Generators C.2.1.43 NUREG-0102 Interfaces for Standard Designs (SRP 1.8)
C.2.1.44 Regulatory Guide 1.138 Laboratory Investigation of Soils Revision 0 for Engineering Analysis and Design of Nuclear Power' plants 7
< C.2.1.45 Regulatory Guide 1.XXX Permanent Dewatering Systems Revision 0 C.2.1.46 Regulatory Guide 1.140 Design, Testing, and Maintenance Revision 0 Criteria for Normal Ventilation Exhaust System Air Filtration and Absorption Units of LWR's C . 2.1. 47 Regulatory Guide 1.142 Safety-Related Concrete Structures Revision 0 C.2.1.48 Regulatory Guide 8.19 Occupational Radiation Dose Assess-Revision 0 ment at LWR's - Design Stage Man-Rem Estimates C.2.1.49 RSB 5-2 Reactor Coolant System Overpressure
_. Revision 0 Protection C
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O O O O O C.2.2 Part 2 Category II Matters Approved by RRRC from March 1974 through August 1978 C . 2. 2.1 Regulatory Guide 1.27 Ultimate Heat Sink for Nuclear Power f Revision 2 Plants i i
C.2.2.2 Regulatory Guide 1.52 Design, Testing, and Maintenance Revision 1 Criteria for Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Absorption Units of Light-Water-Cooled Nuclear Power Plants C.2.2.3 Regulatory Guide 1.59 Design Basis Floods for Nuclear Power Revision 2 Plants C.2.2.4 Regulatory Guide 1.91 Evaluation of Explosions Postulated-Revision 1 to Occur on Transportation Routes Near Nuclear Power Plant Sites C.2.2.5 Regulatory Guide 1.97 Instrumentation for Light-Water-Cooled i
Revision 1 Nuclear Power Plants to Assess Plant Il Conditions During and Following an Accident C.2.2.6 Regulatory Guide 1.102 Flood Protection for Nuclear Power Plants Revision 1 C.2.2.7 Regulatory Guide 1.105 Instrument Setpoints Revision 1 C.2.2.8 Regulatory Guide 1.108 Periodic Testing of Diesel Generator Revision 1 Units Used as Onsite Electric Power Systems at Nuclear Power Plants C.2.2.9 Regulatory Guide 1.115 Protection Against Low-Trajectory Revision 1 Turbine Missiles C . 2. 2.10 Regulatory Guide 1.117 Tornado Design Classification
- Revision 1 8
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O O O O O C.2.2.11 Regulatory Guide 1.124 Service Limits and Loading Combinations l Revision 1 for Class 1 Linear Type Component Supports l
( C . 2. 2.12 Regulatory Guide 1.130 Design Limits and Loading Combinations l Revision 0 for Class 1 Plate- and Shell-Type
! Component Supports C . 2. 2.13 Regulatory Guide 1.137 Fuel Oil Systems for Standby Diesel Revision 0 Generators C.2.2.14 Regulatory Guide 8.8 Information Relevant to Ensuring that Revision 2 Occupational Radiation Exposures at Nuclear Power Stations will be as Low as is Reasonably Achievable (Nuclear Power Reactors)
C.2.2.15 Branch Technical Position Guidelines for Fire Protection for ASB 9.5-1 Nuclear Power Plants Under Review and Construction a C.2.2.16 Branch Technical Position Material Selection and Processing
.E MTEB S-7 Guidelines for BWR Coolant Pressure 2 Boundary Piping 8
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O O O O O C.2.3 Part 3 Category III Matters Approved by RRRC from March 1974 through August 1978 C.2.3.1 Regulatory Guide 1.56 Maintenance of Water Purity in Boiling Revision 1 Water Reactors C.2.3.2 Regulatory Guide 1.68.2 Initial Startup Test Program to Demon-Revision 1 strate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants C.2.3.3 Regulatory Guide 1.99 Effects of Residual Elements on Predicted Revision 1 Radiation Damage to Reactor Vessel Materials C.2.3.4 Regulatory. Guide 1.101 Guidance on Being Operator at the Revision 1 Controls of a Nuclear Power Plant C.2.3.5 Regulatory Guide 1.114 Emergency Planning for Nuclear Power Revision 1 Plants
$ C.2.3.6 Regulatory Guide 1.121 Bases for Plugging Degraded PWR Steam
~ Revision 0 Generator Tubes C.2.3.7 Regulatory Guide 1.127 Inspection of Water-Control Structures Revision 1 Associated with Nuclear Power Plants C.2.3.8 SRP 5.4.7 Residual Heat Removal System Revision 1 C.2.3.9 Regulatory Guide 1.141 Containment Isolation Provisions for Revision 0 Fluid Systems C . 2. 3.10 RSB 5-2 Reactor Coolant System Overpressurization Revision 0 Protection d
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O O O O O C.2.4 Part 4 NRR Category IV Matters C.2.3 1 Regulatory Guide 1.12 Instrumentation for Earthquakes Revision 1 C.2.4.2 Regulatory Guide 1.13 Spent Fuel Storage Facility Design Revision 1 Basis C.2.4.3 Regulatory Guide 1.14 Reactor Coolant Pump Flywheel Integrity Revision 1 C.2.4.4 Regulatory Guide 1.75 Physical Independence of Electric Systems Revision 1 C . 2.4. 5 Regulatory Guide 1.76 Design Basis Tornado for Nuclear Power Revision 0 Plants C.2.4.6 Regulatory Guide 1.79 Preoperational Testing of Emergency Revision 1 Core Cooling Systems for Pressurized Water Reactors
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C.2.4.7 Regulatory Guide 1.80 Preoperational Testing of Instrument Revision 0 Air Systems C.2.4.8 Regulatory Guide 1.82 Sumps for Emergency Core Cooling and Revision 0 Containment Spray Systems C.2.4.9 Regulatory Guide 1.83 Inservice Inspection of Pressurized Revision 1 Water Reactor Steam Generator Tubes C . 2.4.10 Regulatory Guide 1.89 Qualification of Class lE Equipment for Nuclear Power Plants
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Revision 0 C.2.4.ll Regulatory Guide 1.93 Availability of Electric Power Sources Revision 0 C.2.4.12 Regulatory Guide 1.104 Overhead Crane Handling Systems for Revision 0 Nuclear Power Plants 8
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O O O O O C . 2. 4.13 SRP 5.4.2.1 BTP MTEB-5-3, Monitoring of Secondary Side Water Chemistry in PWR Steam Generators C.2.4.14 SRP 6.2.1, 6.2.1A, 6.2.1B, BTP CSB-6-1, Minimum Containment Pressure 6.2.1.2. 6.2.1.3, 6.2.1.4, Model for PWR ECCS Performance Evaluation 6.2.1.5 C . 2. 4.15 SRP 6.25 BTP CSB-6-2, Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident C.2.4.16 SRP 6.2.3 BTP CSB-6-3, Determination of Bypass Leakage Path in Dual Containment Plants C.2.4.17 SRP 6.2.4 BTP CSB-6-4, Containment Purging During Normal Plant Operations C.2.4.18 SRP 9.1.4 BTP ASB-9.1, Overhead Handling Systems for Nuclear Power Plants C.2.4.19 SRP 10.4.9 BTP ASB-10.1, Design Guidelines
? for Auxiliary Feedwater System Pump
, Drive and Power Supply Diversity for PWR's C.2.4.20 SRP 3.5.3 Procedures for Composite Section Local Damage Prediction (SRP Section 3.5.3, Par. II.l.C)
C.2.4.21 SRP 3.7.1 Development of Design Time History for Soil-Structure Interaction Analysis (SRP Section 3.7.1, Par. II.2)
C.2.4.22 SRP 3.7.2 Procedures for Seismic Systems Analysis (SRP Section 3.7.2, Par. II) i C.2.4.23 SRP 3.7.3 Procedures for Seismic Subsystem Analysis (SRP Section 3.7.3, Par. II) 4 8
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O O O O O C.2.4.24 SRP 3.8.1 Design and Construction of Concrete Containments (SRP Section 3.8.1, Par. II)
C.2.4.25 SRP 3.8.2 Design and Construction of Steel Containments (SRP Section 3.8.2, Par. II)
C.2.4.26 SRP 3.8.3 Structural Design Criteria for Category I Structures Inside Containment (SRP Section 3.8.3, Par. II)
C.2.4.27 SRP 3.8.4 Structural Design Criteria for Other Seismic Category I Structures (SRP Section 3.8.4, Par. II)
C.2.4.28 SRP 3.8.5 Structural Design Criteria for Founda-tions (SRP Section 3.8.5, Par. II)
C . 2.4. 29 SRP 3.7, 11.2, 11.3, Seismic Design Requirements for Radwaste 11.4 Systems and Their Housing Structures (SRP
? Section 11.2, BTP ETSB 11-1, Par. B.v) 5.
C.2.4.30 SRP 3.3.2 Turnado Load Effect Combinations (SRP Section 3.3.2, Par. II.2.d)
C . 2.4. 31 SRP 3.4.2 Dynamic Effects of Wave Action (SRP Section 3.4.2, Par. II)
C.2.4.32 SRP 10.4.7 Water Hammer for Steam Generators with Preheaters (SRP Section 10.4.7, Par. I.2.b)
C.2.4.33 SRP 4.4 Thermal-Hydraulic Stability (SRP Section 4.4, Par. II.5)
C.2.4.34 SRP 5.2.5 Intersystem Leakage Detection (SRP Section 5.2.5, Par. II.4) and R.G. 1.45 C.2.4.35 SRP 3.2.2 Main Steam Isolation Valve Leakage
- Control System (SRP Section 10.3, Q Par. III.3 and BTP RSB-3.2)
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O O O- O O C.2.4.36 SRP 3.5.3 Ductility of Reinforced Concrete and Steel Structural Elements Subjected to Impactive or Impulsive Loads C.2.4.37 SRP 3.7.1 Response Spectra in Vertical Direction C.2.4.38 SRP 3.8.1, 3.8.2 BWR Mark III Containment Pool Dynamics C.2.4.39 SRP 3.8.4 Air Blast Loads C.2.4.40 SRP 3.5.3 Tornado Missile Impact C.2.4.41 SRP 6.3 Passive Failures During Long-Term Cooling Following LOCA C.2.4.42 SRP 6.3 Control Room Position Indication of Manual (Handwheel) Valves in the ECCS C.2.4.43 SRP 15.1.5 Long-Term Recovery from Steamline n Break: Operator Action to Prevent
-; Overpressurization C.2.4.44 SRP 5.4.6, 5.4.7, 6.3 Pump Operability Requirements C.2.4.45 SRP 3.5.1 Gravity Missiles, Vessel Seal Ring Missiles Inside Containment C.2.4.46 SRP 4.4. Core Thermal-Hydraulic Analysis C.2.4.47 SRP 8.3 Degraded Grid Voltage Conditions C.2.4.48 SRP 6.2.1.2 Asymmetric Loads on Components Located Within Containment Subcompartments C.2.4.49 SRP 6.2.6 Containment Leak Testing Program C
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C.2.4.50 SRP 6.2.1.4 Containment Response Due to Main Steam Line Break and Failure of MSLIV to Close C . 2. 4. 51 SRP 3.6.1, 3.6.2 Main Steam and Feedwater Pipe Failures C.2.4.52 SRP 9.2.2.2 Design Requirements for Cooling Water to Reactor Coolant Pumps C.2.4.53 SRP 10.4.7 Design Guidelines for Water Hammer in Steam Generators with Top Feedring Design (BTP ASB-10.2)
C . 2.4. 54 SRP 3.11 Environmental Control Systems for Safety-Related Equipment
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C.1 Introduction The Commission's August 22, 1978 policy statement on standardization includes a provision which allows a two year extension on Preliminary Pesign approvals previously issued for a three year term.
The policy statement states that each application for PDA extension will be subject to an assessment of the design with respect to Category I, II, III and IV matters approved for implementation since the regulatory requirements cutoff date for the PDA in question.
This document contains the assessment package for the GESSAR-238 Nuclear Island Design and includes the assessments on each Category I, II, III and IV matter approved since the March 1, 1974 cutoff date.
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C-1 113078
C.2 Asres_sments r
. General Electric has assessed the GESSAR design against the positions in the Regulatory Guides and other NRC matters as specified cs a necessary condition for the PDA extension and according to the list in Roger S. Boyd's letter to Glenn G. Sherwood on October 13, 1978, entitled, " Extension Review O Matters For Preliminary Design Approvals." The assessments are in the following sections and each item has been assessed separately. The writeup shows the revision assessed and whether the design is in full compliance or where applicable General Electric's alternate position where the design does not directly comply with the position.
i Most of the'new guides and positions cover issues which are already addressed l
in GESSAR, and therefore the alternate positions have been established. l 1
Using this approach, the assessments include no coninitments beyond the current design basis. '
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C-2 113078
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l APPENDIX C O PART I ASSESSMEN' EGORY I ITEMS AGAINST GES.c ;. EAR ISLAND DESIGN O
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113078
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C.2.1 Category I Matters The NRC definition of Category I matters is as follows:
"This review will determine whether you have clearly delineated the extent to which the de' 1 ready conforms to these matters.
There should be no changes.
Assessment of Category I Matters O
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C-4 113078
l GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION !
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C . 2.1.1 REGULATORY GUIDE: 1.7 REVISION: 2 DATED: Issued in 1978 TITLE: Control of Combustible Gas Concentrations In Containment Following a Loss-of-Coolant Accident
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REGULATORY GUIDE:
This guide describes methods that are acceptable to the NRC staff for implementing the provisions of Criterion 41, " Containment Atmosphere Cleanup," which requires that systems to control hydrogen, oxygen and other substances that may be released into the reactor containment be provided as necessary to control the concentrations of such substances following postulated accidents and ensure that containment integrity is maintained.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
The GESSAR design complies with this guide. The Flammability
/N Control System, Drywell/ Containment Mixing System and the d Containment Atmosphere Monitoring System are provided to meet the 4
monitoring and combustible gas control requirements of this guide.
These systems are designed based on the calculational assumptions contained in this guide. i l
General Electric has analyzed the production and accumulation of hydrogen in the containment due to the metal-water reaction between the fuel cladding and the reactor coolant, and as a result of radio!ytic decomposition of the post-accident emergency cooling water.
GESSAR requires a adundant hydrogen mixing system and redundant i hydrogen combiners to limit the hydrogen concentration within the I containment to 4 volume percent. A backup, controlled purge system is also provided. This system is capable of purging any hydrogen O that might be released from the reactor pressure vessel into the i drywell through the horizontal vents in the drywell wall into the suppression pool with the larger volume.
A thermal recombiner sy. tem is provided to maintain the long term hydrogen concentration to less than 4 volume percent when the mixing system is in operation and the hydrogen concentration in the containment would continue to increase due to radiolysis. The recombiner is designed to seismic Category I criteria and has redundancy in all active components. A mobile combustible gas O. control system is not used in the GESSAR. design, therefore shielding required for a mobile unit is not included.
C-5 113078 i
- 4. - m w GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION O
C . 2.1. 2 REGULATORY GUIDE: 1.9 REVISION: 0* DATED: March 10, 1971 TITLE: SELECTION OF DIESEL GENERATOR SET CAPACITY FOR STANDBY POWER SUPPLIES REGULATORY GUIDE:
This safety guide describes an acceptable basis for the selection of diese: generator sets of sufficient capacity and margin to implement General Design Criterion 17.
A diesel generator set selected for use as a standby power supply should have the capability to (1) start and accelerate a number of large motor loads in rapid succession, and be able to sustain the loss of any such load, and (2) supply continuously the sum of the loads needed to be powered at any one time.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
The Division 1 and Division 2 (emergency) diesel generator set will be selected to assure compliance with Regulatory Guide 1.9.
They will start, accelerate and sustain loads on ESF buses in the pattern required for various modes within the parameter suggested in the safety guide. This will include loads in GESSAR Tables 8.3.1 and 8.3.2 and those balance of plant loads such as service water later determined to be justified for connection to ESF buses.
The HPCS diesel-generator will conform with the intent of Regulatory Guide 1.9, although the starting transient for the single large motor load may cause the voltage or frequency variations to exceed the maximum suggested. The components of this system are supplied as a package and will be tested for performance under typical starting conditions. Effective function of the system is not impaired.
A detailed assessment of this guide against the HPCS diesel is provided in Table 7-2 of Topical Report NEDO 10905, dated May 1973.
O *GE is not aware of a Revision 1 to this guide as of October 1978.
C-6 113078
1 n GESSAR REGULATORY GUIDANCE ASSESSMENT I U PDA EXTENSION C.2.1.3 REGULATORY GUIDE: 1.20 REVISION: 2 DATED: May 1976 TITLE: Comprehensive Vibration Assessment Program for Reactor Internals During Pre-Operational and Startup Testing REGULATORY GUIDE:
Regulatory Guide 1.20 describes a comprehensive vibration assessment program for reactor internals during preoperational and initial startup testing. The vibration assessment program meets ,
the requirements of Criterion 1, " Quality Standardr and Records," I of Appendix A to 10CFR Part 50 and Section 50.34, " Contents of l Applications; Technical Information," of 10CFR Part 50.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
Vibration testing of reactor internals has been per. formed virtually in all GE-BWR plants. At the time of original issue of AEC Regulatory Guide 1.20, test programs for compliance were O instituted. The first BWR/6 plant of each size will be considered a prototype and will be instrumented and subjected to preopera-tional and startup flow testing to demonstrate that flow-induced vibrations similar to those expected during operation will not cause damage.
Subsequent plants which have internals similar to those of the prototypes will also be tested in compliance with the requirements of Regulatory Guide 1.20. General Electric is committed to confirm satisfactory vibration performance of internals in these plants through preoperational flow testing followed by inspection for evidence of excessive vibration.
Extensive vibration measurements in prototype plants together with satisfactory operating experience in 11 BWR/4 plants have established O the adequacy of BWR/4 reactor internal designs. General Electric will continue these test programs for the GESSAR plants to verify structural integrity and to establish the margin of safety.
The GESSAR plant complies with this guide. 4 O
C-7 113078
GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C.2.1.4 REGULATORY GUIDE: 1.28 REVISION: 1 DATED: March, 1978 TITLE: Quality Assurance Program Requirements (Design and Construction) 1 REGULATORY GUIDE:
Regulatory Guide 1.28 endorses ANSI N45.2-1977 as an adequate basis for compl- q with quality assurance program requirements of Appendix B to 10CFR:,, (subject to certain itemized considerations), for the design and construction phases of nuclear power plants.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
General Electric has an implemented QA Program which was developed to satisfy ANSI N45.2 and the various other ANSI N45.2 " daughter" standards.
The GE alternate position for Rev. O of Reg. Guide 1.28 is documented in a letter dated March 15, 1978, from A. Breed to C. J. Heltemes (NRC).
Although GE has not yet formally submitted its alternate position for Reg. Guide 1.28, Rev. 1, to the NRC, our position has been documented in O- a letter from G. G. Sherwood to the USNRC dated May 31, 1978, as follows:
The positions of the regulatory guide are implemented fully except those areas identified below:
o C. Regulatory Position, Item 2.b - It is recommended that the treatment of guidance provided in Section 19, second paragraph, Item (4) of ANSI N45.2-1977 (when to perform audits) remain as guidance rather than be imposed as a requirement.
The following are editorial comments:
o C. Regulatory Position, Item 2.d - The reference should be to the fourth paragraph rather than to the fifth, o C. Regulatory Position, Item 2.e - The reference should be to the fifth paragraph rather than to the sixth.
o C. Regulatory Position, Item 3, third sentence - Revise to read: "---Code covered items such as pumps and valves."
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GESSAR REGULATORY GUIDANCE ASSESSMENTS O PDA EXTENSION l
l C.2.1.5 REGULATORY GUIDE: 1.29 REVISION: 2 DATED: February 1976 TITLE: Seismic Design Classification REGULATORY GUIDE INTENT Regulatory Guide 1.29 describes an acceptable method of identifying and classifying those features of light-water cooled nuclear power plants that should be designed to withstand the effects of SSE.
i EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE Except for those items identified below, the structures, systems and components important to safety that are required to withstand :
the effects of a Safe Shutdown Earthquake, and remain functional have been properly classified as seismic Category I items. All other structures, systems, and components that may be required for operation of the facility are designed to other than seismic I Category I requirements. The exceptions are as follows- I Position C.1(b) Replace with the following:
1 The reactor core and reactor vessel internals which are engineered !
safety features. !
l Position C.1(e) Replace with the following:
A. A suitable interface restraint shall be provided for the Main Steam Line (MSL) and the Main Feedwater Line (MFL) to physically define the seismic category transitional point between steam and feedwater piping in the Auxiliary Build'.g and in the Turbine Building.
B. MSL and MFL piping and other components (including the shutoff valves and branch piping of 2-1/2 in, or larger nominal pipe O up to and including the first valve capable of timely action) that are located between the reactor vessel and the seismic inter-face restraint shall be designated Seismic 1.
C. MSL'and MFL piping and other components outside the seismic
, interface restraint (i.e., generally located in the Standard Plant Turbine Building) are not designated Seismic 1.
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C-9 113078 4
~ - , -m , , - , . ~ v' -. m.., . . . _ - . . , .
, . . . . _ , _ , .--,m ., .~ , .
RG 1.29 Position C.1.(h) replace with the following:
The component cooling water portions of the reactor recirculation pumps are not required to be Seismic Category I since the pumps do not perform a safety function.
Position C.1.(o) replace with the following:
The secondary containment is interpreted to mean the reactor shield O' building, fuel building, and ECCS rooms and their main corridors at basement level in the auxiliary building.
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O C-10 113078
i GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION C.2.1.6 REGULATORY GUIDE: 1.31 REVISION: 2 DATED: May 1977 TITLE: Control of Ferrite Content in Stainless Steel Weld Metal REGULATORY GUIDE: ,'
O Regulatory Guide 1.31 describes an acceptable method of imple-menting requirements with regard to the control of ferrite content in stainless steel weld metal.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE:
GESSAR implements the provisions of this Regulatory Guide. As a result of discussion with the NRC staff, General Electric conducted l a program to demonstrate that the control of weld fil,ler material 1 to contain a minimum of five percent delta ferrite was adequate to I prevent fissuring in production welds. In this program, GE measured !
the ferrite content of production welds in five boiling water reactor plants, encompassing 338 welds. From these data and others O the staff agreed that production testing is not required if, prior to use, the delta ferrite content of each lot and heat of filler metal is verified by making determinations on undiluted weld deposits using magnetic measuring devices. (Reference GESSAR 5.2.5.7)
For the GESSAR plants, all austenitic stainless steel weld filler materials are supplied with a minimum of 5% ferrite (minimum Ferrite Number 5).
The weld metal filler materials must also comply with the chemical analysis requirements and delta ferrite determination requirements of Section III of the American Society of Mecahnical Engineers Boiler and Pressure Vessel Code.
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O C-ll 113078
GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C.2.1.7 REGULATORY GUIDE: 1.32 REVISION: 2 DATED: February 1977 TITLE: Criteris for Safety Related Electric Power Systems for Nuclear.
Power Plants
]
( REGULATORY GUIDE:
Regulatory Guide 1.32 identifies disparity between GDC 17, IEEE-308 R. G. 1.6, IEEE-450 and provides an acceptable supplement to these criteria for offsite power supply, and standby power supply systems.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE:
The onsite standby alternating current power system for the GESSAR design will consist of three independent distribution systems and associated onsite power sources to support the engineered safety features and auxiliary support systems. Preferred power will be supplied by the off-site power systems (Section 8.2 of GESSAR) and, under a condition of a loss of the off-site power source, the class IE portions of each of the three distribution systems will receive O power from its associated on-site diesel generator unit.
The proposed 125 volt direct current power system is in conformance with the requirements of IEEE Standard 308-1974, and with the recommendations for Regulatory Guides 1.6 and 1.32.
Three independent 120 volt alternating current power supply systems will provide power to the four channel solid state protection system. The system is designated the Nuclear System Protection System -- Power Supply System and will be designed in accordance with IEEE Standard 308-1974 and the subject regulatory guide.
The ampere-hour capacity and short time rating of batteries will be in accordance with criteria given in IEEE Standard 308 and will be
(] adequate to supply all electrical loads required until a-c power is restored for the operation of battery chargers. The chargers for the 125-volt d-c systems are connected to 450-volt motor control centers and are capable of carrying the normal direct current load and, at the same time, charging the batteries to a fully charged condition. The sizing of the battery chargers will meet IEEE Standard 308. ,
Compliance with Regulatory Guide 1.93 will be assessed in another section of this report and Regulatory Guide 1.81 was deemed applicable f to multi unit plants and not in the scope of GESSAR.
O Therefore, the BWR/6 GESSAR systems satisfy the provisions of the subject regulatory guide.
C-12 113078
. GESSAR REGULATORY GUIDANCE ASSESSMENT U<-
PDA EXTENSION C . 2.1. 8 REGULATORY GUIDE: 1.33 REVISION: 1 DATED: January 1977 TITLE: Quality Assurance Program Requirements (Operation) l Q REGULATORY GUIDE:
Regulatory Guide 1.33 endorses ANSI N18.7-1976/ANS 3.2 which defines the overall quality assurance program requirements for the operation phase of nuclear power plants. The NRC finds this document to be an adequate basis for the nuclear power plant owner to comply with the quality ::surance program requirements of Appendix B to 10CFR50, subject to certain itemized considerations.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE:
General Electric is committed to the support of nuclear power plant owners to the extent that requirements are contractually delegated to GE. Since this Regulatory Guide applies only to the Owner and is limited to the operation phase, GE has no direct obligation for O its implementation.
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GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION C.2.1.9 REGULATORY GUIDE: 1.35 REVISION: 2 DATED: January 1976 i i
TITLE: Inservice Inspection of Ungrouted Tendons in Pre-Stressed Concrete Containment Structures C)
This regulatory guide is not applicable to GESSAR and therefore was not assessed.
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C-14 113078
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GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION ,
C . 2.1.10 REGULATORY GUIDE: 1.38 REVISION: 2 DATED: May 1977 TITLE: Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants l O !
REGULATORY GUIDE:
Regulatory Guide 1.38 endorses ANSI N45.2.2 '.972 which defines i requirements for packaging, shipping, receiving, storage, and I handling of safety-related items for nuclear power plants. These l requirements deal with the protection and control necessary to assure that the requisite quality of those important parts of the plant are preserved from the time items are fabricated until they are incorporated in the plant.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE:
q General Electric recognizes its responsibilities relative to protection D and preservation of items important to safety and has an implemented QA Program which satisfies the provisions of Regulatory Guice 1.38, Rev. 2. GE has provided alternate acceptable means of implementing certain provisions of Regulatory Guide 1.38. .aese positions are documented in a letter dated March 15, 1978, from A. Breed (GE) to C. J. Heltemes, Jr. (USNRC) and are summarized below:
GE implements the provisions of Regulatory Guide 1.38, Rev. 2, May 1977, including the regulatory position relative to ANSI N45.2.2-1972, and implements the alternate acceptable positions identified below:
- 1. Section 3.7.1(1): GE will use cleated, sheathed boxes up to 1000 lbs. rather than 500 ios. This type of box is e safe for and has been tested for loads up to 1000 lbs.
Other national standards allow this: see Federal Speci-fication PPP-B-601. Special qualification testing may be required for loads above 1000 lbs.
- 2. Section 3.7.2: Skids or runners shall be used on containers with a gross weight of 100 lbs. or more. Skids or runners shall be fabricated from 3x4 inch nominal lumber size minimum and laid flat except where this is impractical because of the small dimensions of the container.
C-15 113078
Regulatory Guide 1.38 - cont'd O eese 2
- 3. Section 4.3.4: Since title to equipment generally changes hands at the time it is moved off the supplier's dock into the carrier, GE will make these. inspections "immediately prior to loading" rather than "after loading," as presently indicated. To have this inspection and possible repair performed after loading presents legal complications, as once the equipment enters the transport vehicle, the O carrier has some responsibility and our customer has ownership.
- 4. Appendix Section A3.4.1(4) and (5): During printing of the standard, a transposition occurred between the last sentence of (4) and (5). The correct requirements are as follows:
(4) "However, preservatives for inaccessible inside l surface of pumps, valves, and pipe for systems '
containing reactor coolant water shall be the water flushable type."
(5) "The name of the preservative used shall be indicated to facilitate touch up."
- 5. Appendix Section A.3.4.2(3): Inert gas blankets are currently used on the reactor pressure vessel (RPV) and on some of the heat exchangers supplied by GE. Provisions are made for measuring and maintaining the RPV blanket pressure within the required range during shipment and storage. Heat exchangers or tanks containing carbon steel which require an inert gas blanket will be inerted prior to shipment. During storage, provision shall be made for measuring and maintaining the inert gas blanket pressure within the required range within each pressurized purged item or container.
- 6. Appendix Section A3.7.1(3) and (4): GE will work to the following requirement in lieu of items (3) and (4):
O Fiberboard boxes shall be securely closed either with a water resistant adhesive applied to the entire area of contact between the flaps, or all seams and joints shall be sealed with not less than 2-inch wide, water-resistant tape.
1 b
C-16 113078
.=.. .- - .. _. - ____ . - ___
O GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION C.2.1.ll REGULATORY GUIDE:-1.39 REVISION: 2 DATED: Sept. 1977 TITLE: Housekeeping Requirements for Water-Cooled Nuclear Power Plants REGULATORY GUIDE:
This guide describes an acceptable method of complying with the Commission's regulations with regard to housekeeping require-ments for the control of work activities, conditions, and environ-ments at water-cooled nuclear power plant sites.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
This Regulatory Guide applies to the owner's responsibilities at the nuclear plant and is outside the scope of GESSAR.
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GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C . 2.1.12 REGULATORY GUIDE: 1.52 REVISION: 2 DATED: April 1978 l TITLE: Design, Testing and Maintenance Criteria for Engineered-Safety Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light Water Cooled Nuclear Power Plants l
Q REGULATORY GUIDE:
This guide describes criteria to be used in the design, maintenance and testing of atmosphere cleanup systems, filtration units and adsorption units.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE:
The Standby Gas Treatment System (SGTS) consists of redundant exhaust fans and filtration trains each consisting of a demester, heat coil, prefilter, HEPA filters and charcoal filter. Redundant components are separated and protected. The Standby Gas Treatment I System and other systems which clean air prior to release to the I environment are designed to function under appropriate environmental condition such as for the hypothetical LOCA. All requirements of l pd this guide are met except that ANSI Standards N509-1975 and N510-1975 were not used as design basis for these systems. Instead, design alternatives presented in ORNL-NSIC-65 " Design, Construction and Testing of High Efficiency of Air Filtration Systems for Nuclear Application," have been used. However, the strict design, testing and quality assurance requirements imposed assure that the air handling systems will provide their required functions.
The sta'ndby gas treatment system (SGTS) will be designed to control exfiltration of potentially contaminated air from the plant following a postulated accident or abnormal occurrence which could result in abnormally high airborne radiation.
All essential equipment of the SGTS that mitigates the consequences of the hypothetical LOCA event and surrounding structures will be seismic Category I design.
/]
The design criteria for sizing the SGTS includes the thermal transient which occurs in the shield building shortly after a LOCA. Following a postulated loss-of-coolant accident the pressure in tne secondary containment volumes could increase due to inleakage and the starting time required for the SGTS. Additionally, the annulus pressure and temperature will increase due to heat transfer through and expansion of the primary containment shell. GESSAR provMes an analysis of the annulus pressure transient which considers tne above phenomena and which indicates that the annulus will be maintained at an O ennropriete neoetive pressure.
C-18 113078
- _ _ _ _ . - _ _ ~ - . - - - - . _ - _ . - . . . - - . - _ . - - _ _ _ _ - _ - . _ - _ . _ _ _ . .
- .O eege 2 The basis for determination of filter surface will comply with Regulatory Guide 1.3.
GESSAR section 6.5.3.1 specifically lists the compliance status of the GESSAR design with respect to Regulatory Guide 1.52. The charcoal filters of the standby gas treatment system have been given credit for 99% efficiency in the removal of all species of iodine during the LOCA.
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s GESSAR REGULATORY GUIDANCE ASSESSMENT t
PDA EXTENSION 9 ' C . 2.1'.13 REGULATORY GUIDE: 1.63 REVISION: 1 DATED: May 1977 TITLE: Electrical Penetrations Assemblies in Containment Structures for Light-Water-Cooled Nuclear Power Plants A
V REGULATORY GUIDE:
Regulatory guide 1.63 describes acceptable methods of design and qualification of containment electrical penetrations. It endorses IEEE-317 - 1976 with specific exceptions.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE:
GE has stated in Section 7.1.2.4 of GESSAR that the design meet the requirements of IEEE dated 317-1972, " Standard for Electrical Penetration Assemblies in Containment Structures for Nuclear Fueled Power Generating Stations."
pd With respect to the drywell penetrations, GE has provided the staff with preliminary design for the drywell penetrations and a description of their qualification program (Section 3.8.6 of GESSAR).
The qualification program for the drywell penetrations includes results from previously conducted tests to demonstrate the sealing capability of the potting material used in the penetrations. Once the penetrations are constructed, the drywell structure proof tests and the periodic drywell leak test will provide continued assurance that the penetrations are' intact.
The GESSAR standard plant design meets the provisions of this guide with paragraph C.2 being interpreted as follows:
- 1. Circuit protection shall comply with the attached table.
O 2. Conformance with IEEE-279 does not require that the circuit protective devices be identified as Class lE unless the load is Class lE. Non IE circuits having redundant circuit protection will maintain independence of the redundant protective elements such that no event causing a need for the protection can disable the protective function. ,
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i l
O cessAR Reco uToRv coloANCe AssessneNT PDA EXTENSION summary Table of Conformance with Regulatory Guide 1.63 for Circuits Penetrating Primary Containment Very Low Currents Involved Q use of Two Interrupting -- No Action Required --
1 Devices in Analysis Conformance by
_, Series Required Inspection Recirc Pumps X Power Circuits on Motor Control Centers X
)
Control Circuits, Alarm, l solenoids, etc. Circuits -
Normally Protected by
]
t small Fuses or Breakers X I
Instrumentation l Circuits, Thermocouples, l Annunciator - All Low Current Level Applications X O
O C-21 113078 '
q GESSAR REGULATORY GUIDANCE ASSESSMENT v PDA EXTENSION C.2.1.14 REGULATORY GUIDE: 1.64 REVISION: 2 DATED: June 1976 TITLE: Quality Assurance Requirements for the Design of Nuclear Power Plants REGULATORY GUIDE:
Regulatory Guide 1.64 endorses ANSI N45.2.11-1974, subject to the comments in section C, " Regulatory Position." The purpose of the standard is to provide requirements and guidance for a quality assurance program for the design of nuclear power plant structures, systems, and components whose satisfactory and reliable performance is required to (1) prevent accidents that could cause undue risk to the health and safety of the public; or (2) to mitigate the conse-quences of such accidents if they were to occur.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE:
Since the original GESSAR issue date, GE has had a QA Program O- implemented in compliance with ANSI N45.2.11 (first as a draft, later as an issued standard). The implemented BWR QA Program is in conformance with Rev. 2 of Regulatory Guide 1.64 except as modified by the GE Alternate Position. This Alternate Position was approved by the NRC September 16, 1977, by letter from C. J. Heltemes, Jr.
The alternate positions are quoted below:
Comply with the provisions of Regulatory Guide 1.64, Rev. 2, June 1976, including the regulatory position relative to ANSI N45.2.11-1974, except for the following modifying provisions to the second paragraph of Section 6.1 of ANSI N45.2.11-1974:
- 1. If in an exceptional circumstance the designor's immediate O' supervisor is the only technically qualified individual available in the organization to perform 6 design verifica-tion by design review, this review may be conducted by the supervisor, providing that:
a, the justification is individually documented and ,
approved in advance by the supervisor's management, and O
C-22
3078
Regulatory Guide 1.64 - cont'd O rege 2
- b. quality assurance audits cover frequency and effective-ness of use of supervisors as design verifiers, to guard against abuse.
- 2. An individual who contributed to a given design may participate in a group verification of that design provided that the individual who contributed to the O design (a) does not verify his contribution to the design, and (b) does not serve as the chairman or leader of the group verification activity.
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.g GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION l C.2.1.15 REGULATORY GUIDE: 1.68 REVISION: 2 DATED: January 1978 TITLE: Initial Test Programs for Water Cooled Reactor Power Plants REGULATORY GUIDE:
This guide describes the general scope and depth of initial test programs acceptable to the NRC staff.
EVALUATION OF GESSAR-(BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE:
- The BWR-6 preoperational and startup test program will comply with ,
the subject guide with minor qualifications. 1 In Appendix A5.T the guide specifies verification of valve setpoint
- and reset pressures during power testing. This is interpreted to be limited to those valves that actuate during a transient.
In Appendix A-Sii the guide specifies a pump trip or control valve closure at 50% and 100% of rated power. However, the test will be done at a power between 50% and 75% of rated and again at 100% of rated power. The flow coastdown rate is not strongly dependent; hence more flexibility in initial power is warranted.
Also, Appendix A.5kk of the guide specifies a loss of feedwater heating at 50% and 100% of rated power. The test will be done at 90% power only because.the loss of feedwater heating results in a power increase and this must necessarily be started below 100%
power to conform to power limitations of the test. The loss of a feedwater heater at low to mid power level results in a mild transient due to reduced initial feedwater temperature and flow. The dynamic response of the plant is verified in the test at the higher power level; hence,.that at 50% of rated power is inappropriate.
Appendix A.l.h(5) indicates " Prompt Relief Trip" as a plant feature and Appendix A.1.h(6) lists " cold water interlocks" whereas none of these are applicable to the GESSAR (BWR-6).
The guide specifies a demonstration that core limits will not be exceeded during or following the exchange of control rod patterns that will be permitted during operation (the demonstration test should be conducted at the highest power level at which control rod pattern exchanges will be allowed during plant operation).
General Electric will comply with the guide via the adherence to generically approved sequence exchange procedures. These generic procedures will assure that core limits will not be exceeded during sequence exchanges at power. A representative sequence exchange
'O uti1izin9 the eneroved sener4c procedvres mer optione11v be 9errormed for the purpose of familiarizing the plant operating and technical staff with the ope. ration of the facility, but will not be required to further qualify the generic procedures.
C-24 113078
Page 2 The RHR system will then be operated in the shutdown cooling modes wherein the RHR heat exchanger is . connected directly into the reactor water circuit to bring the reactor to the cold low pressure condition.
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GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EX1ENSION C.2.1.16 REGULATORY GUIDE: 1.68.1 REVISION: 0 DATED: December 1975 TITLE: PRE 0PERATIONAL AND INITIAL STARTUP TESTING 0F FEEDWATER AND CONDENSATE SYSTEMS FOR BOILING WATER REACTOR POWER PLANTS O REGULATORY GUIDE:
1 This guide describes in detail the type and nature of BWR feedwater and conden- I sate system tests that are acceptable to the staff.
EVALUATION 0F GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE: l The preoperational and startup testing of the feedwater and condensate systems are not within the GE scope of services. Therefore, this guide is not applicable to GESSAR.
However, Chapter 14 of GESSAR states that a comprehensive testing program will be developed to ensure that all nuclear safety related equipment and systems
(' will perform in accordance with their design criteria. As individual systems are completed, they will be tested, reviewed, and approved according to pre-determined and written procedures. In general, all procedures will be developed in accordance with NRC publications, such as Regulatory Guide 1.68.1. l l
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GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION C . 2.1.17 REGULATORY GUIDE: 1.72 REVISION: 1 DATED: January 1978 TITLE: SPRAY POND PIPING MADE FROM FIBERGLASS-REINFORCED THERM 0 SETTING RESIN O REGULATORY GUIDE:
This guide describes a method for implenanting these requirements with regard to the design, fabrication, and testing of fiberglass-reinforced thermosetting resin (RTR) piping for spray pond applications. This guide applies to plants using spray ponds as an ultimate heat sink as opposed to rivers, cooling towers, etc.
EVALUATION OF GESSAR (BWR-6/MK-III) WITH RESPECT TO REGULATORY GUIDE:
The ultimate heat sink is a site unique feature and is not included in the scope of GE supplied equipment or services. Therefore, this guide is not applicable to GESSAR.
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l GESSAR REGULATORY GUIDANCE ASSESSMENTS PDA EXTENSION l
C.2.1.18 REGULATORY GUIDE: 1.84 REVISION: 12 DATED: March 1978 I TITLE: Code Case Acceptability ASME Section III Design and Fabrication REGULATORY GUIDE INTENT:
This guide provides a list of ASME Design and Fabrication Code Cases that have been generically approved by the Regulatory Staff. -
Code Cases on this list may, for design purposes, be used until l appropriately annulled. Annulled cases are considered " active" for equipment that has been contractually committed to fabrica-tion prior to the annullment.
The guide, Revision 6 and later revisions, requires specific ,
approval by the NRC on all ASME Section III, Class 1 Code Cases !
not listed.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE GE meets the requirements of Revision 12 of this guide as appli-i cable. However, commitment to meet a specific revision of this j guide has little significance since the guide is revised as new code cases are issued by the ASME and approved by the NRC. GE's policy.is to request and obtain NRC approval of code cases applied to ASME Class 1 components not listed in previous / current editions of the Regulatory Guide prior to use of the code case.
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q GESSAR REGULATORY GUIDANCE ASSESSMENTS V
PDA EXTENSION C.2.1.19 REGULATORY GUIDE: 1.85 REVISION: 12 DATED: March 1978 TITLE: Code Cases Acceptability ASME Section III Materials
,o REGULATORY GUIDE:
V This guide provides a list of ASME Materials Code Cases that have been generically approved by the Regulatory Staff.
Code Cases on this list may, for design purposes, be used until appropriately annulled. Annulled cases are considered " active" for equipment that has been contractually committed to fabrication prior to the annulment.
The guide, Revision 6 and later revisions, requires specific approval by the NRC on all ASME Section III, Class 1 Code Cases not listed.
p EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
V Identified GE supplied NSSS analysis, design and/or equipment utilized is in compliance with the objectives of the subject Regulatory Guide.
GE meets the requirements of Revision 12 of this guide as appli-cable. However, commitment to meet a specific revision of this guide has little significance since the guide is revised as new code cases are issued by the ASME and approved by the NRC. GE's policy is to request and obtain NRC approval of code cases applied to ASME Class 1 components not listed in the Reg. Guide prior to use of the code case.
o tj C-29 113078
I GESSAR REGULATORY GUIDANCE ASSESSMENT
- ]
PDA EXTENSION C.2.1.20 REGULATORY GUIDE: 1.90 REVISION: 1 DATED: August 1977 TITLE: Inservice Inspection of Pre-Stressed Concrete Containment Structures With Grouted Tendons l
l This regulatory guide is not applicable to GESSAR and therefore was l not assessed. '
e l l
/3
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l GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION C . 2.1. 21 REGULATORY GUIDE: 1.92 REVISION: 1 DATED: February 1976 TITLE: Combination of Modes and Spatial Components in Seismic Response Analysis i
Q REGULATORY GUIDE:
This guide describes methods for combining the individual modal responses from a seismic response dynamic analysis and in combining maximum values (in case of time history dynamic analysis) or the representative maximum values (in case of spectrum dynamic analysis) due to multi-directional seismic impact motions.
EVALUATION OF G?: SAP, (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE:
The GESSAR design is in agreement with the guidelines through the use of an alternate position guide to comply with the regulatory position. !
General Electric uses both the time-history and the response spec-O trum methods. When the time history method is used for combining i
the effects of three-dimensional earthquakes, the vector sum at every step is used to calculate the maximum response.
l When response spectrum method of modal analysis is used, all modes except the closely spaced modes are combined by the square root of l the sum of the squares (SRSS) method. However, for closely spaced modes, use is made of the Double Sum Method given in Section 1.2.3 in the Guide without the absolute sign (i.e.).
N N 1/2 R= I IR R 8 k=1 k=1 k s ks l
where R is the representative maximum value of a particular re-Q sponse of a given element to a given component of an earthquake, R k is the peak value of the response of the element due to the kth mode, and N is the number of significant modes considered in the modal response combination, sR is the peak value of the response of the element attributed to sth mode.
(w[ wj) }2 1
'ks * ~1+[
@k *k
- Es "s} ~
O C-31 113078
l.92 in which w[ = wk El~O3k Ok*Ok* t s *k O where wk and are the modal frequency and the damping ratio in thekthmode,hespectivelyandt is the duration of the earth-3 quake.
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p v
GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION C.2.1.22 REGULATORY GUIDE: 1.94 REVISION: 1 DATED: April 1976 TITLE: Quality Assurance Requirements for Installation, Inspection and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE:
This guide describes requirements for installation, inspection and testing of structural concrete and structural steel during the construction of a nuclear power plant.
GESSAR implements the provisions of this guide.
Section 3.8.3.1.6 of GESSAR covers materials, quality control, and special construction techniques. All concrete materials shall be approved on the basis of conformance to the specifications and standard technical methods of the ASTM, and shall be from sources deter-mined acceptable prior to start of construction. All concrete work will be in accordance with ACI-318-71, " Building Code Requirements Q for Reinforced Concrete".
Reinforcing steel for concrete structures will be deformed bars meeting the requirements of ASTM A 615, Grade 60. Placing and splicing of bars will be in accordance with the requirements of ASME,Section III Div. 2.
The structural steel materials will conform to all applicable requirements of the AISC Manual of Steel Construction, unless otherwise noted and will comply with ASTM specifications.
The following list of codes of practice with indicated exceptions are listed in section 3.8.3.1.6.7 of GESSAR and establish the standards of construction procedure.
f] ACI 214, " Recommended Practice of Evaluation of Compression Test Results of Field Concrete" ACI 306, " Recommended Practice for Cold-Weather Concreting" ACI 318, " Building Code Requirements for Reinforced Concrete" ACI 347, " Recommended Practice for Concrete Formwork" ACI SP-2, " Manual of Concrete Inspection" (a3 C-33 113078
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)
1.94 AST" C94, " Ready-Mixed Concrete" ACI 305, " Recommended Practice for Hot-Weather Concreting" ACI 211.1, " Recommended Practice for Selecting Proportions for Normal Weight Concrete" l ACI 304, " Recommended Practice for Measuring, Mixing, Trans- i O Portia9. ae ei cias coacrete" 1 1
ACI 315, " Manual of Standard Practice for Detailing Reinforced l Concrete Structures" !
ASME, " Boiler and Pressure Vessel Code",Section VIII , :
Subsection B, Requirements Pertaining to Methods of I Fabrication of Pressure Vessels Part UW, Requirements for Pressure Vessels Fabricated by Welding AWS 01.1, " Code for Welding in Building Construction" ASTM, C618, Class F, Specification for Fly Ash for Use as an Admixture in Portland Cement Concrete
' O ANSI N 45.2.5-74, " Supplementary Quality Assurance Requirements for Installation, Inspection, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants" O
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Y l GESSAR REGULATORY GUIDANCE ASSESSMENT O.
PDA EXTENSION i l
C . 2.1. 23 REGULATORY GUIDE: 1.95 REVISION: 1 DATED: January 1977 l l
TITLE: Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release -
REGULATORY GUIDE: 1 This guide describes assumptions acceptable to the Regulatory Staff to be used in assessing the habitability of the control room during and after a postulated exterM1 release of chlorine. It also describes requirements for control room isolation and emergency procedures.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
In the GESSAR design, chlorine is identified by human detection in accordance with Paragraph C.7 of Reg. Guide 1.78. For specific cases in which a plant would be sufficiently close to a railroad or i highway, analysis will be carried out to show whether or not the O chlorine limits stated in Regulatory Guide 1.78 will be exceeded due to a postulated accident. If the limit could be exceeded, chlorine detection devices will be placed in the control room and intake ducts. The plan would also include an automatic insulation system.
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.GESSAR REGULATORY GUIDANCE ASSESSMENTS O PDA EXTENSION :
C . 2.1. 24 REGULATORY GUIDE: 1.99 REVISION: 1 DATED: April 1977 TITLE: Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials REGULATORY GUIDE:
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Regulatory Guide 1.99 describes an acceptable procedure for pre-diction of radiation damage to the beltline of reactor vessels of light water reactors.
EVALVATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE:
GESSAR is in compliance with this Regulatory Guide.
Section 5.4.5 of GESSAR states that a surveillance test program will include the preparation of a series of Charpy V-notch impact specimens and tensile specimens from the base metal of the reactor vessel, weld heat-affected zone metal, and weld metal from a reactor steel joint that simulates a welded joint in the reacter vessel at !
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the core beltline region. !
l The specimens and neutron monitor wires are placed near core midheight adjacent to the reactor vessel wall. The specimens are installed at startup. Selected groups of specimens can be removed at intervals 4
during the reactor lifetime, and the mechanical properties determined.
I GESSAR states that the vessel material surveillance program will meet the material surveillance program and the requirements of ASTM E185. Predictions of neutron radiation damage to the beltline of reactor vessels are based on the following procedures which follow the provisions of Regulatory Guide 1.99.
The adjusted reference temperature will be determined at the 1/4T position of the reactor pressure vessel wall. The curves in Figure 1 O of Regulatory Guide 1.99 have been extrapolated from 50 F to 20 F to provide information at low fluences which produce radiation l induced temperature shifts less than 50 F. General Electric's 1 radiation damage data (NED0-21708) are consistent with the adjustment l curves shown in Regulatory Guide 1.99.
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l GESSAR REGULATORY GUIDANCE ASSESSMENT
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V PDA EXTENSION C . 2.1. 25 REGULATORY GUIDE: 1.100 REVISION: 1 DATED: August 1977 TITLE: Seismic Qualification of Electric Equipment for Nucl. ear Power Plants REGULATORY GUIDE:
This regulatory guide describes guidelines for complying with the Commission's regulations with respect to verifying the adequacy of the seismic design of electrical equipment. This guide requires I that IEEE Standard 344-1975 with supplemental regulatory require- I ments be used. l l
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE: l The requirements of IEEE Standard 344-1975 as well as the supple-mentary regulatory requirements are included in the GESSAR design.
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V General Electric stated in Section 3.10.1.3 of GESSAR that all Category I electrical equipment will be seismically qualified in accordance with the staff position " Electrical and Mechanical Equipment Seismic Qualification Program," dated December 1973.
GESSAR also states that General Electric will perform response type of multi-frequency, multi-axis testing as elaborated in IEEE-344-1975.
Qualification of all Reactor Island equipment in General Electric's engineering scope will be required to have qualification performed in accordance with applicable IEEE standards. These include IEEE-344 (General Seismic) and IEEE-383 for Cable, IEEE-387 for diesel generators, IEEE-344 for motors, and IEEE-317 for primary containment penetrations.
Suppliers of seismic Category I equipment will verify that their o components or systems will not suffer loss of function during or V after seismic loading due to the SSE. The tests will be performed in accordance with IEEE Standard 344-1975. The magnitude and frequency of the SSE loadings which each component will experience is determined by its location within the plant.
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GESSAR REGULATORY GUIDANCE ASSESSMENT g-t PDA EXTENSION C.2.1.26 REGULATORY GUIDE: 1.103 REVISION: 1 DATED: October 1976 j TITLE: Post-Tensioned Pre-stressing Systems for Concrete Reactor Vessels and Containments This regulatory guide is not applicable to GESSAR and therefore was not assessed.
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GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C.2.1.27 REGULATORY. GUIDE: 1.106 REVISION: 1 DATED: March 1977 TITLE: Thermal Overload Protection Valve Motors REGULATORY GUIDE:
The regulatcry guide describes acceptable methods of disabling the !
thermal protection on motor operated valve motors during emergency operation in order to enhance their availability for safety functions. )
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
The standard plant design complies with this guides The GESSAR design has thermal overload protection devices which are active only when the equipment is in the test mode. When the equipment is in the normal mode, the thermal overload protection is bypassed.
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GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C . 2.1. 28 REGULATORY GUIDE: 1.107 REVISION: 1 DATED: February 1977 TITLE: Qualification for Cement Grouting for Pre-stressing Tendons i in Containment Sructures This regulatory guide is not applicable to GESSAR and therefore was not assessed.
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h GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION I
C . 2.1. 29 REGULATORY GUIDE: 1.116 REVISION: 0-R DATED: May 1977 TITLE: Quality Assurance Req'uirements for Installation, Inspection, I and Testing of Mechanical Equipment and Systems !
REGULATORY GUIDE:
Regulatory Guide 1.116 endorses ANSI N45.2.8-1975, subject to the comments in section C, " Regulatory Position." The purpose of the standard is to provide requirements and guidelines to assure that important items of nuclear power plants, including structures, systems, and compor,ents, are installed, inspected, and tested in a manner that will provide adequate confidence that they will perform satisfactorily '.n service.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE:
General Electric has an implemented QA Program which was developed to satisfy the various quality-related ANSI N45.2 series standards, including N45.2.5. The program implements the provisions of Regulatory Guide 1.116 Rev. 0-R. The GE proposed alternate posi-O- tion to Regulatory Position 3 is documented in a letter dated March 15, 1978, from A. Breed (GE) to C. J. Heltemes, Jr. (USNRC).
The alternate position is quoted below I
Regulatory Position 3 - Comply with the ,
implementation of Regulatory Guide 1.68 as !
described in GESSAR, Chapter 14, " Initial Test."
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'GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C .2.1. 30 REGULATORY GUIDE: 1.118 REVISION: 1 DATED: November 1977 TITLE: Periodic Testing of Electric Power and Protection Systems REGULATORY GUIDE:
Regulatory Guide 1.118 describes acceptable methods of periodically testing the functional performance and responses of electric power and protection systems.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
The provisions of this guide are applied to the standard plant.
For example, the reactor protection system design provides assurance that, through redundancy, each channel has sufficient reliability to fulfill the single failure criterion. No single component failure, intentional bypass maintenance operation, calibration operation, or test to verify operational availability will impair the ability of the system to perform its intended safety function.
The reactor protection system includes design features that permit-inservice testing. This ensures the functional reliability of the system should the reactor variable exceed the corrective action setpoint. The system can be tested during operation because the system is arranged in four separately powered divisions and each division has a logic which can produce an automatic trip signal.
The logic scheme is two out of four arrangement. Manual scram testing is performed by operating one of the four manual scram controls. This tests one division. Indicating lights verify that the actuator contacts have opened. Each active component of the emergency core cooling systems provided to operate in a design basis accident is designed to be operable for test purposes during normal operation of the system. The segments are subject to tests to verify the performance of the full operational sequence that O brings each system into operation.
For purpose of practical implementation and clarification of ambiguities, General Electric is implementing the provisions of the regulatory guide in the manner described below.
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R.G. 1.118 O Page 2 Portion C.1 "Means shall be included in the design to facilitate, insofar as practicable and safe, response time testing from sensor input (at the sensor input connection for process instruments) to and including the actuated equipment."
fosition C12 - Subparagraph starting with "When it is not possible....",
add clarifying words to read as follows, "When this is necessary the test installation shall duplicate as nearly as practicable those expected environmental conditions and mechanical configurations of the actual installation which can effect the validity of the response data."
Position C14(b) - This subparagraph is reworded as follows: "Re-moval of fuse or opening a breaker for the purpose of deactivating i
instrumentation or control circuits is permitted only if such actions cause (1) trip of the associated protection system channel 1 (or) the actuation (startup and operation) of the associated Class 1E Load group."
l Position C.16 - Test intervals, both initial and revised, should be such that significant changes in failure rates can be detected.
p Accordingly, Section 6.5.1 should be supplemented by adding an j item (8) to Section 6.5.1 as follows: (8) detection of significant '
increase in failure rates.
The constraints imposed by this paragraph may make it very difficult if not impossible to test HPCS and RCIC as devised without actually
- injecting high pressure cold water into the reactor vessel.
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GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C . 2.1. 31 (1) REGULATORY GUIDE: 1.120 REVISION: 1 DATE: November 1977 (issued for comment only)
(2) BRANCH TECHNICAL POSITION: APPENDIX A of APCSB 9.5-1 of STANDARD REVIEW PLAN DATED: May 1, 1976 O REGULATORY GUIDE AND BRANCH TECHNICAL POSITION:
The purpose of the Fire Protection program is to ensure the capa-bilities to shutdown the reactor and maintain it in a safe shutdown condition and to minimize radioactive releases to the environment in the event of a fire.
The guide addresses fire protection programs for safety-related systems and equipment and for other plant areas containing fire hazards that could adversely affect safety-related systems and supplements Regulatory Guide 1.75, " Physical Separation of Electrical Systems".
O GENERAL ASSESSsENT FOR AeeENDIx ^ OF BTe AeCSB 9.5-1:
The GE Nuclear Island design was evaluated for conformance against Appendix A to BTP, USNRC-APCSB 9.5-1, " Guidelines for Fire Protection for Nuclear Power Plant Docketed Prior to July 1, 1976" and a Fire Hazard analysis was conducted on the GE Nuclear Island design.
These conformance evaluations and alternates and the Fire Hazard analyses report were submitted to the USNRC as follows:
(1) The conformance evaluation and alternates, titled: Response to Appendix A, United States Nuclear Regulatory Commission Auxiliary Power Conversion Systems Branch, Branch Technical Position 9.5-1 of Standard Review Plan Section 9.5-1, " Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976," was sub-t mitted to Roger S. Boyd, Director of Division of Project Management on Docket STN-50-447 submittal MFN 168-77, 183-085-77 from W. D. Gilbert, General Electric Company.
.(2) The Fire Hazcrd Analysis Report titled: TVA STRIDE Fire Hazard Analysis was submitted in May 1977 to USNRC through the TVA Hartsville A and B docket number STN 50-518, STN 50-519, STN 50-520 and STN 50-521. It should be noted that this Fire Hazard Analysis was conducted ia accordance with the USNRC document titled:
" Enclosure 2, Supplemental Guidance on Information Needed for Fire Q Protection Program Evaluation."
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EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE O The GE Power Generation Control Complex design has ALSO been evaluated i against Regulation Guide 1.120 ar.d BTPAPCSB 9.5-1. This evaluation is presented in NED0-10466, Revision 2 and Amendment 1, " Power Generation ,
Control Complex Design Criteria and Safety Evaluation", dated March 1978.
The Control Room panels comply with the requirements of Reg. Guide 1.120 except for the following:
The fourth paragraph of Position 6.b states:
" Smoke detectors should be provided in the control room, cabinets, and consoles. If redundant safe-shutdown equipment is located in the same control room cabinet or console, additional fire protection measures should be provided. Alarm and local indication should be provided in the control room." l Rather than the above, the control room panels will be designed to the separation requirements of Reg. Guide 1.75. That is, redundant Class 1E systems existing in any one bay of a cabinet are separated by a barrier or six-inch distance. Barriers are steel cans around devices and conduit around panel wiring.
Smoke detectors are not provided because the leakage through spaces and around instruments and doors is high enough to allow adequate detection by sensors located in the control room.
Q Regulatory Guide 1.120, Revision 1, is in draft review status. Certain minor differences exist between Appendix A of the DTP APCSB 9.5-1 and Regulatory Guide 1.120 which will not cause additional impact.
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GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C.2.1.32 REGULATORY GUIDE: 1.122 REVISION: 1 DATED: February.1978 TITLE: Development of Floor Design-Spectra for Seismic Design of Floor Supported Equipment in Components O
REGULATORY GUIDE:
This guide describes methods for developing design response spectra at various floors or other equipment support locations of interest from the time-history motions resulting from the dynamic analyses of the supporting structures.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
General Electric complies with all guidelines of the regulatory guide exception position C,2 where instead of using 15 percent in frequency for spectrum broadening, GE uses 110 percent. Justifi-cation to this exception is provided in Section 3.7.2.8 of GESSAR.
O This section illustrates the conservative assumptions that have been included in the calculation of the floor response spectra.
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GESSAR RE60LATORY GUIDANCE ASSESSMENT PDA EXTENSION C . 2.1. 33 REGULATORY GUIDE:'1.123 REVISION: 1 DATED: July 1977 y
TITLE: Quality Assurance Requirements for Control and Procurement of Items and Services for Nuclear Power Plants ,
1 REGULATORY GUIDE:
Regulatory Guide 1.123 endorses ANSI N45.2.13-1976, subjecttothe l comments in Section C, " Regulatory Position." The purpose of the .
standard is to describe requirements and provide guidelines for the )
control of activities to be exercised during procurement of items and services which affect the quality of nuclear power plants.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
General Electric has an implemented QA Program which was developed to satisfy the various quality-related ANSI N45.2 series standards, j including N45.2.13. The program is in compliance with Regulatory l Guide 1.123, Rev. 1.
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GESSAR REGULATORY GUIDANCE ASSESSMENTS O EDA EXTENSION I
C.2.1.34 REGULATORY GUIDE: 1.126 REVISION: 1 DATED: March 1978 TITLE: An Acceptable Model and Related Statistical Methods for the Analysis of Fuel Densification l
REGULATORY GUIDE:
O This guide provides an analytical model and related assumptions and .
procedures that are acceptable to the NRC staff for predicting the effects of fuel densification in light-water-cooled nuclear power ~~
reactors, The guide also describes statistical methods related to product sampling that will provide assurance that this and other approved analytical models will adequately describe the effects of densification for each initial core and reload fuel quantity produced.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE GE is in the process of coming into compliance with this guide.
The current GE methods employed in the analysis of fuel densifi-O cetioa =ev de rouad 4a the ron owia9 refereaces:
- 1. Grossenbacher, RJ; Holtzclaw, KW; et.al., "BWR/4 and BWR/5 Fuel Design," NED0-20944 (Sections 2.1.2.1.0 and 2.3.2.H),
October 1976.
- 2. "BWR/4/5/6 Standard Safety Analysis Report Rev. 2," (Sections 4.2.1.2.1.4 and 4.2.3.2.8).
- 3. " Generic Reload Fuel Application, "NED0-24011 (Section 2.4.2.1), l May 1977.
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GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION O
C.2.1.35 REGULATORY GUIDE: 1.128 REVISION: 0* DATED: April 1977 q TITLE: Intallation Design and Installation of Large Lead Storage Batteries for Nuclear Power Plants REGULATORY GUIDE:
This Regulatory Guide describes an installation design for Class 1E batteries that is in general agreement with IEEE-484-1975. ,
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EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
The 125 volt d-c systems are divided into 4 Class 1E divisions.
Each system has d-c battery, battery charges, and load center distribution panels. These are designed as Class 1E equipment in accordance with applicable classes of IEEE Standard 308-1974. The plant design and layout from these d-c systems will provide physical separation of the equipment, cabling, and instrumentation essential n
U to plant safety. Each system is located in its own ventilated room and all the components are housed in a safety class structure. The ;
battery rooms are independently ventilated to keep the gases produced due to charging of the batteries below an explosure concentration.
Therefore the Class 1E battery installations are in accordance with the regulatory guide with the following interpretation of position C.1.
The area in the immediate vicinity of the battery vent is excluded from the area defined by the phrase".....at any location within the battery area."
Hydrogen detection sensors and alarms are not provided since the Class IE ventilation system whose failure is alarmed and will prevent the accumulation of hydrogen. Fire detection instrumentation is part of the fire protection system rather than the battery system.
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- GE is not aware of a Revision 1 to this guide as of October, 1978.
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GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION C.2.1.36 REGULATORY GUIDE: 1.129 REVISION: 0 DATED: April 1977 TITLE: Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Nuclear Power Plants O ReGutATORv GUIoE:
1 This regulatory guide describes a method acceptable to the NRC staff for performing the maintenance, testing, and replacement of large lead storage batteries for all types of nuclear power plants. l The guide requires the use of IEEE Std. 450-1975 and additional I regulatory requirements. The guide does not include design re-quirements.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO GESSAR:
The maintenance, testing, and replacement of storage batteries are not in the GE. scope of supply. This information will be provided )
by the applicant.
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GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C.2.1.37 REGULATORY GUIDE: 1.131 REVISION: 0 DATED: August 1977 TITLE: Qualification Tests of Electric Cables, Field Splices, and Connections for Light-Water-Cooled Nuclear Power Plants O
V REGULATORY GUIDE:
This regulatory guide describes a method acceptable to the NRC staff for complying with the Commission's regulations with regard to qualification testing of electric cables, field splices, and connections for service in light-water-cooled nuclear power plants to ensure that the cables, field splices, and connections can perform their safety-related functions.
l This guide states that the procedures described in IEEE Standard '
383-1974 are acceptable with the additional requirements stated in I the guide.
Q EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
The GESSAR Design complies with this guide with the'following clarifications:
- 1. Reg. Position C.5 of the guide states that the radiological source terms should be obtained from Reg. Guide 1.89. In order to maintain a consistant approach in tnis area the GE position on Reg. Guide 1.89 is used.
- 2. Reg. Position C.13(i) allows only the qualification test sequence defined in IEEE 383 paragraph 2.3.3. GE may use other tests which are determined to have acceptable results.
- 3. GE uses a more restrictive phrase in place of the word " burn" in paragraph 2.5.5 of IEEE 383 which is referenced in Reg.
Q Position C.13(s). The phrase used by GE is " burn or glow until. cooled.
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GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION C.2.1.38 REGULATORY GUIDE: 1.132 REVISION: 0 DATED: September 1977 TITLE: Site Investigations for Foundations of Nuclear Power Plants REGULATORY GUIDE:
This guide describes programs of site investigations that would bc normally meet the needs for evaluating the safety of the site from l the standpoint of the performance of foundations and earthworks '
under most anticipated loading conditions, including earthquakes. '
EVALUATION OF GESSAR (BWR-6/ Mark III) WITH RESPECT TO REGULATORY GUIDE GESSAR Section 2.5.1 states that detailed geologic and seismic information regarding each site will be gathered as appropriate from published reports, maps, knowledgeable sources, surveys, geophysical investigations, borings, trenches, and other investi-gations and documents.
The collected information will be used for evaluating the suitability of the site for the construction of the nuclear power plant. The
- location of Class 1 structures, planning the extent of investi-(/ gations and backfill, choosing the compaction criteria and other site related decisions will utilize the collected information.
The information related to local and regional geology as it effects seismological investigations, history of earthquakes, correlation of epicenters with geological structures, identification of active faults and surface faulting, will be obtained and presented to the NRC.
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GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION .
C .2.1.39 - REGULATORY GUIDE: 1.134 REVISION: 0 DATED: September 1977 TITLE: Medical . Certification and Monitoring of Personnel Requiring Operator Licenses REGULATORY GUIDE:
This guide describes a method for complying with the Commission's O regulations for evaluation of the medical qualifications of appli-cants for initial or renewal operator or senior operator licenses.
EVALUATION OF GESSAR (BWR-6/ Mark III) WITH RESPECT TO THE REGULATORY GUIDE The medical qualifications for operators will be the responsibility of the owner and therefore is not mentioned in the GESSAR text.
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GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION C.2.1.40 REGULATORY GUIDE: 1.135 REVISION: 0 DATED: September 1977 TITLE: Normal Water Level and Discharge at Nuclear Power Plants REGULATORY GUIDE:
O This guide states requirements for.the analysis of the capability of structures to resist the forces of normal ground water, or from lakes, streams, oceans, etc.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
The design basis flood elevation for GESSAR satisfies the guide-lines of the regulatory guide and provides a conservative basis for engineering design to assure the structure will withstand such hydrology forces.
For flood protection, the design basis flood elevation is approximately one-foot below the plant finished grade elevation including allowance O ror coiacident waves aad resuiteat ruaun. This r1ood protectioa will be verified for each plant because it requires a site unique analyses.
With the proposed plant grade one foot above the elevation of the design basis flood, Category 1 structures, systems and components will be protected from the hydrodynamic phenomena associated with the flood. The hydrostatic effect of the flood will be considered in the design of all Category I structures exposed to the water head. All seismic Category I structures will be designed to remain stable when subjected to either overturning moments or uplift forces of the flood.
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GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION I C.2.1.41 REGULATORY GUIDE: 1.136 REVISION: 0 DATED: November 1977 TITLE: MATERIAL FOR CONCRETE CONTAINMENTS REGULATORY GUIDE:
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This guide provides information regarding the NRC staff's position on the acceptability for NRC licensing actions of Article CC-2000 of the "ASME B&PVC,Section III, Division 2 (Concrete Reactor Vessels and Containments)" relative to material requirements. In those areas where the provisions of the above code are considered insufficient for licensing purposes, the staff has provided supple-mentary guidelines it considers to be acceptable.
GENERAL COMPLIANCE OR ALTERNATE APPROACH ASSESSMENT:
Section 3.8.3.1.6.7 of GESSAR lists the codes of practice which will establish the standards of construction procedure. These are l q also listed on the assessment of Reg. Guide 1.94.
V The GESSAR Design meets the requirements of Position C.1 and C.2.
These requirements are applicable to the Mat foundation of the MK III containment. Positions C.3 and C.4 are not applicable since prestressed concrete is not used.
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GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION C.2.1.42 REGULATORY GUIDE: 1.137 REVISION: 0 DATED: January 1978 TITLE: Fuel Oil Systems for Standby Diesel Generators REGULATORY GUIDE INTENT:
O This regulatory guide describes a method that is acceptable to the NRC staff for complying with the Commission's regulations regarding fuel-oil systems for standby diesel generators. This guide requires the use of ANSI N-195-1976 and additional regulatory requirements.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE: ,
Within the limits of GE design scope the GESSAR design is in compliance. Because this guide has limited application to the GESSAR design because such things as fuel oil quantity, fuel oil storage capacity, and the location of parts of the system are not in General Electric's scope-of-supply and/or service. The GESSAR design complies with applicable requirements in section 9.5. It O
states, "...that the design and construction of the diesel generator fuel oil system #all conform to the IEEE Criteria for Class 1E Electrical Systr.ms for Nuclear Power Generating Stations (IEEE 308 and ASME Code-3, Quality Group C). The miscellaneous equipment will also conform to applicable standards of NEMA, DEMA, ASTM, IEEE, ANSI, API, and NFPA."
The GESSAR design. complies with all applicable requirements,such as; insuring adequate net positive suction head from the day tanks, adequate storage capacity to supply fuel for seven days at maximum post LOCA loads and independent fuel storage for each diesel engine.
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GESSAR 3EGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C.2.1.43 NUREG 0102 - Interfaces for Standard Design (SRP 1.8)
The staff report describes safety-related interfaces, for light water reactors, at the preliminary design stage of review, that should be presented by the reactor vendor in a Nuclear Steam Supply O SSAR.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO NUREG-102 REQUIREMENTS:
Although the GESSAR documents do not include the entire nuclear power plant, they do specifically describe and delineate the safety-related interface requirements imposed on the balance of the nuclear power plant. The general rules for determining interfaces is that boundaries extend to just outside of the building walls. A major factor in the design process is the determination of the exact description of the interface, therefore, parameters such as dimensions and orientation of pipes, type of connections, and pressures at the interface points are established and identified in h Section 1.10 of GESSAR.
For each of the GESSAR systems, General Electric has identified the safety-related interfaces with the balance-of plant. This is accomplished in two ways. There is a summary of B0P interfaces on a system basis (Table 1.10-1) and there is a description of individual GESSAR system interfaces by location within each system on interface location drawings. Each interface location is then identified by a unique alpha numeric designation.
The interface of safety requirements in GESSAR with the balance-of-plant is illustrated in a prepared matrix of the principal categories of interface information. For each of the categories, General Electric Company has provided specific interface requirements, that n are broken down into primary and secondary functions. The primary V safety functions are defined as those which must be satisfied for engineering safeguards operation. These safety functions are further broken down into the initiating safety functions and support-ing functions. Other interfaces which are not required to mitigate the consequences of postulated design accidents are also identified in additional tables. .
In summary, the safety interfaces for the NSSS-80P and for the NSSS-RI are discussed in detail in the NSSS GESSAR docket and are in compliance with NUREG-102.
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C.2.1.44 REGULATORY GUIDE: 1.138 REVISION: 0 DATED: April 1978 TITLE: Laboratory Investigations of Soils for Engineering Analysis and Design of Nuclear Power Plants REGULATORY GUIDE:
This guide describes laboratory investigations and testing practices for determining soil and rock properties and characteristics needed for engineering analysis and design for foundations and earthworks
- for nuclear power plants. Criteria for planning and performing laboratory tests are also given in the regulatory position.
EVALUATION OF GESSAR (BWR/6-MARK III) WITH RESPECT TO REGULATORY GUIDE Site soil evaluation and analysis is a site unique issue and is not within the GE scope of services. Therefore, this guide is not applicable to GESSAR.
GESSAR Section 2.5.1 states that detailed geologic and seismic information regarding each site will be gathered as appropriate from published reports, maps, knowledgeable sources, surveys, geophysical investigation, borings, trenches and other investi-gations and documents. 1 The collected information will be used for evaluating the suita-bility of the site for the construction of the Nuclear Power Plant.
For each site selected analysis will be performed to ensure that the soil bearing pressures are compatible with the site soil conditions.
The location of class 1 structures, planning the extent of excavation and backfill, choosing the compaction criteria and other site related decisions will utilize the collected information.
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GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION I i O C.2.1.45 REGULATORY GUIDE: 1.XXX REVISION: DRAFT DATED: Not Dated ,
TITLE: PERMANENT DEWATERING SYSTEM REGULATORY GUIDE:
The intent of this guide is to describe methods acceptable to the NRC staff for designing a permanent dewatering system.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
The necessity to dewater is a site unique consideration. Therefore this guide is not within the scope of GESSAR.
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GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION C.2.1.46 REGULATORY GUIDE: 1.140 REVISION: 0 DATED: March 1978 TITLE: Design, Testing, and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants O
REGULATORY GUIDE:
This guide presents methods acceptable to the NRC staff for implementing the Commission's regulations in 10 CFR Part 50 and in Appendices A and 1 to 10 CFR 50 with regard to the design, testing, and maintenance criteria for air filtration and adsorption units installed in the normal ventilation exhaust systems of light-water-cooled nuclear power plants. This guide applies only to atmosphere cleanup systems designed to collect airborne radioactive materials during normal plant operation, including anticipated operational occurrences, and addresses the atmosphere cleanup systems, including the various components and ductwork in the normal operating environmer' lhis guide does not apply to post-accident engineered-safety-feature atmos re cleanup systems that are designed to mitigate the consequences of pc' ated accidents. Regulatory O Guide 1.52, " Design, Testing, and Mainte~
Engineered-Safety-Feature Atmosphere CD Criteria for Post-accident system Air Filtration and Adsorption Units of Light-Water-Coole- ar Power Plants," provides guidance for these systems.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
The design as described in GESSAR Section 9.4.5.2.3 includes air filtration and I adsorption units in the Containment Purge, Exhaust and Pressure Control System only. However the need for filtration and adsorption units is expected to be determined on a site unique basis. If filtration and adsorption units are installed in the Containment Purge, Exhaust and Pressure Control System, they will comply with this guide. The determination of whether filtration and O adsorption systems should be included on normal effluent systems is not interpreted to be within the scope of this guide. Therefore, this guide is considered not applicable unless the determination is made that filtration and adsorption are required.
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v PDA EXTENSION C.2.1.47 REGULATORY GUIDE: 1.142 REVISION: 0 DATED: April 1978 TITLE: Safety-Related Concrete Structures for Nuclear Power Plants (Other Than Reactor Vessels and Containments)
O REGULATORY GUIDE:
This guide describes a method acceptable to the NRC staff for complying with the Commission's regulations with regard to safety-related concrete structures (other than reactor vessels and containments) for nuclear power plants.
ACI Standard 349-76, " Code Requirements for Nuclear Safety-Related Structures" is incorporated in this guide by reference.
ACI Standard 349-76 delineates requirements for the sizing of concrete structural systems and elements, requirements for construction details, and specifications and tests for materials.
O EVALUATION OF GESSAR (BWR/6-MARK III) WITH RESPECT TO REGULATORY GUIDE The GESSAR design complies with this guide with the following minor exceptions:
Paragraph C.9.a -
In load combinations (9), (10) and (11), the factor 1.4 for T should be retained. T isthethermaleffectsandloadsduri8 normal operation or sh8tdown conditions based on the most crit cal transient or steady state condition which is determined by rigorous analysis rather than by using values established by engineering practice or established industrial standards from which the Live Loads are selected. Thus, Tg should not have the same factor as L.
O earaorann c 9.c -
The load combination (7) is equivalent to the load combination Abnormal / Extreme environmental in Table Cc i nsel for containment design. (7) is more conservative than that .oc containment design.
There is no reason to make it more conservative by using factor 1.25 for both Pa and Eg.
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O Paragraph C.9.d -
The load combination (10) is equivalent to Eq. (9-2) of ACI 318-71 l with W replaced by 1.1E. ACI 318-71 is a building code for non- l nuclear structures for which the seismic loads are selected from _
industrial standards such as UBC instead of performing more rigorous seismic analyses which are required for nuclear structures.
This means that seismic load considered in combination (10) is more certain than that in ACI 318-71. Thus, no increase in load factor for E is required. Similar reason applies to the load O combinationf2).
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GESSAR - REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C.2.1.48 REGULATORY GUIDE: 8.19 REVISION: 0 DATED: May 1978 .
TITLE: Occupational Radiation Dose Assessment in Light-Water Reactor Power Plants Design Stage Man-Rem Estimates O
REGULATORY GUIDE:
The objective of this guide is to describe a method acceptable to the NRC Staff for performing an assessment of collective occupa-tional radiation dose as part of the process of designing a light-water-cooled power reactor (LWR).
EVALUATION OF GESSAR (BWR/6 MK III) WITH RESPECT TO REGULATORY GUIDE:
Chapter 12 of GESSAR addresses radiation control and provides estimates cf radiation exposure throughout the plant.
l The emphasis in GESSAR is on shielding as a means of maintaining i
'Je-- low radiation dose rates and exposures. Exposures behind shielding, such as during shutdown maintenance, is not treated in detail.
However, tables 12.1.25, 12.1.26, and 12.1.27 give man hour and dose rate data. Exposure data reflects experience from plants !
through early Nine Mile Point, Vermont Yankee and Oyster Creek. '
GESSAR does not include radiation evaluations in the extreme detail that the guide suggests. However as a part of the early phase of the GESSAR design the principles of this guide were used to reduce ;
operational exposure. Through this design process and the corre- '
sponding NRC review ALARA for the GESSAR design was established.
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C.2.1.49 BRANCH TECHNICAL POSITION RSB 5-2 REVISION: 0 DATED: April 1978 '
l O TITLE: Overgressur4zetion erotection of eressurized weter Reectors While Operating at Low Temperatures l
BRANCH TECHNICAL POSITION:
This BTP describes methods acceptable to the NRC staff for avoiding inadvertent overpressurization of a Pressurized Water Reactor primary system at low temperatures.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE This BTP is written specifically for Pressurized Water Reactors and is not applicable to the GESSAR design.
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O APPENDIX C PART II ASSESSMENTS OF CATEGORY II ITEMS Q AGAINST GESPAR-238 NUCLEAR ISLAND DESIGN O
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C.2.2 Category II Matters The NRC definition of Category II matters is as follows:
"This review will define the , extent to which the design conforms, I 1
or provides an acceptable alternative, to these matters. For those O ceses where tne desion is not in substentiel conformence with these matters or acceptable alternatives are not provided you should j demonstrate why conformance is not necessary. The outcome of the staff review may result in additional requir: nents."
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PDA EXTENSION C . 2. 2.1 REGULATORY GUIDE: 1.27 REVISION: 2 DATED: January 197.6 TITLE: Ultimate Heat Sink for Nuclear Power Plants O REGULATORY' GUIDE:
This guide applies to all types of nuclear power plants that use water primarily as the ultimate heat sink. The guide describes a basis that may be used to. implement General. Design Criteria 44 and 2.
EVALUATION OF GESSAR (BWR-6/ Mark III) WITH RESPECT TO REGULATORY GUIDE Section 1.2.2.8.4 of GESSAR states that the ultimate heat sink is the responsibility of the owner. General Electric has no direct responsibility for the implementation of this guide which is not in the GESSAR scope.
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.PDA EXTENSION '
O C.2.2.2 REGULATORY GUIDE: .l.52 REVISION: 1 DATED: July 1976 TITLE: Design, Testing, and Maintenance Criteria for Engineered i Safety Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light Water Cooled Nuclear Power Plants Revision 2 of this guide was assessed in Section C.2.1.2 of this O Appendix and is applicable here..
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C.2.2.3 REGULATORY GUIDE: 1.59 REVISION: 2 DATED: August 1977 1
I TITLE: Design Basis Floods for Nuclear Power Plants !
IO REGULATORY GUIDE: !
l In cases where site flooding is postulated this guide states i requirements for flood protection for safety-related structures, systems and components.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE:
l The BWR/6 Mark III standard plant is designed to be placed above -!
the site DBFL (Design Basis Flood Level) and therefore safety- I related structures, systems and components are not affected by flocding. The determination of the Design Basis Flood is strictly i a Jite-related issue and this aspect of the guide is not applicable ,
.O to G~tSSAR. The structure; of safety significance are designed for a design basis flood as defined in Regulatory Guide 1.59, up I
1 to an elevation of one foot below plant grade including allowance for the effects of_ coincident waves and the resultant ramp as calculated from site unique' parameters.
The probable maximum flood on streams and rivers is defined in and determined by the methods prescribed by the Department of the Army, Office of the Chief Engineers, Civil Engineering Bulletin No. 52-8, March 1965.
The standard plant structures of safety significance are assumed to be located such that failure of existing impoundments and potential future impoundments will not result in flooding in excess of what is described above.
O The GESSAR has been written with a broad spectrum of sites in mind. As such it is not possible to discuss local and regional geology and seismology of a specific site.
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l q GESSAR .GULATORY GUIDANCE ASSESSMENT I PDA EXTENSION C.2.2.4 REGULATORY GUIDE: 1.91 REVISION: 1 DATED: February 1978 l
l TITLE: EVALUATION OF EXPLOSIONS POSTULATED TO OCCUR ON TRANSPORTATION ROUTES ,
NEAR NUCLEAR POWER PLANT SITES l REGULATORY GUIDE:
This guide describes a method for determining distances from the power plant to a railway, highway, or navigable waterway beyond which any explosion that might ,
occur on these transportation routes is not likely to have an adverse effect on I plant operation or prevent a safe shutdown.
EVALUATION OF GESSAR (BWR-6/MK III WITH RESPECT TO REGULATORY GUIDE i
The power plant site and transportation facilities are not within the GE scope of design. Therefore, this guide is the responsibility of the utility appli-cant.
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GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION C.2.2.5 ' REGULATORY GUIDE: 1.97 REVISION: 1 DATED: August 1977 I I
TITLE: Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident G
V REGULATORY GUIDE:
This guide describes a method acceptable to the NRC staff to .
provide instrumentation to monitor plant variables and systems l during and following an accident. 1 EVALUATION 0F GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE:
General Electric has proposed a revised implementation policy to the recommendations in this Regulatory Guide in a letter to Mr. Robert B. Minogue, Director, Office of Standards and Develop-ment., from Glenn G. Sherwood, Manager, Safety and Licensing, dated October 25, 1977. In this letter General Electric urged that the p
guide be applied only to future designs to the extent reasonable and practicable and as justified by a value/ impact assessment. The letter also asked for a revision to the regulatory guide be issued to clarify 'the interpretation of the guide and provide an unam-biguous basis for the specification of technical requirements.
Therefore the assessment statements which follow are not intended to indicate that the GESSAR design is in compliance with the Reglatory Guide as it now stands, but to show that the design is adequate and acceptable from a safety standpoint and that there is no demonstrated deficiency to GDC 13, GDC 19, or GDC 64 which were the basis for the PDA approval.
The Design Basis accidents listed in Chapter 15 of GESSAR have been analyzed, and bounding and limiting conditions established.
Instrumentation is included in the design to provide information about the effects of a transient or accident, such as pressure increase, decreases in reactor coolant inventory, and a release of radioactivity. Section 7.5.2.8.3 of GESSAR describes the design requirements of the post accident tracking instrumentation. The specific information from the instrumentation already available, allows the operator to direct.his attention to the accomplishment of essential. safety actions to protect the public such as reactor shutdown, containment isolation, core cooling, and containment heat removal.. The accomplishment of these essential functions is an appropriate objective explicitly stated in Section B of this guide. ,
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'It is General Electric's opinion that the intent Regulatory
'O Position C.l.b is complied with by Engineered-Safety-Feature (ESF) systems that are automatically actuated and fail-safe in order to preclude the need for operator action since these ESF systems are fully redundant, safety grade, automatically actuated, diverse, sized for maximum credible conditions, and designed to take the best possible corrective action.
With respect to Position C.3,-current instrumentation in conjunc- !
tion with radiological emergency plans are satisfactory to protect the health and safety of-the-public. The extended' ranges for g3' radiation monitors; in effect, duplicate the function of portable instrumentation required by emergency procedures. The respective ranges documented and approved.in GESSAR were determined by .
evaluating the worst case and bounding Design Basis Events by applying NRC conservative technical conditions, and by applying !
appropriate margins.
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GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C.2.2.6 REGULATORY GUIDE: 1.102 REVISION: 1 DATED: September 1976 TITLE: Flood Protection for Nuclear Power Plants
{ REGULATORY GUIDE:
This guide describes types of flood protection guidelines for the safety-related structures, systems, and components identified in regulatory guide 1.29 that must be designed to withstand the Safe Shutdown Earthquake and remain functional.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE:
The GESSAR BWR-6/Mk III design complies with this guide.
The General Electric Standard Plant design (BWR-6/MK III) is based on a " dry site" as. defined in paragraph B1 of the regulatory guide.
For those other structures not covered by GESSAR, the specific plant applicant has the responsibility for implementing the provisions O' of this regulatory guide. j The' design basis flood elevation of GESSAR is approximately one foot below the plant finished grade elevation including allowance l for coincident waves and resultant run-up. This flood protection must be verified on a case-by-case basis. With the proposed plant grade-one foot above the elevation of the design basis flood, Category I structures are designed to withstand the flooding forces.
Category I systems and components are protected from flooding by Category I structures.
The safety design basis of the GESSAR standard plant provides that equipment of safety significance will be unaffected by flooding considerations. The structures of safety significance will be
, located on the site so that (a) the total design is compatible with existing ground water levels up to 2 ft. below grade, (b) the flood level associated with the design bssis flood is at or below an elevation corresponding to approximately 1 ft. below plant grade, and (c) the loading of these structures does not include simultaneous flood levels and seismic events.
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GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C.2.2.7 REGULATORY GUIDE: 1.105 REVISION: 1 DATED: November 1976 TITLE: Instrument Setpoints REGULATORY GUIDE:
This guide describes guidelines for ens ring that the instrument setpoints in systems important to safety initially are within and I remain within the specified limits. l EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE:
The standard plant design complies with this guide except as noted.
Documentation will be prepared during the FSAR stage of plants referencing GESSAR to record appropriate methods for the establish-ment of recommended instrument nominal trip setpoints and technical specification limits in a consistent and repeatable manner.
The setpoints considered which have protective functions are those Os associated with the Reactor Protection Instrumentation, Isolati(;n Instrumentation, Emergency Core Cooling System and Rod Withdrawal Block Instrument. For example, during the establishment of technical specification limits instrument drift, accuracy and calibration are conservatively accounted for.
Position C.4 of the guide discusses environmental qualification and that instruments should not be annealed, stress relieved, work hardened, etc. The environmental qualificaton of instrument: will be in accordance with the GE position on Reg. Guide 1.89. No additional environmental qualification will be done based on Position C.4 of this guide.
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GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C.2.2.8 REGULATORY GUIDE: 1.108 REVISION: 1 DATED: August 1977 TITLE: Periodic Testing of Diesel Generator Units Used as On-Site Electric Power Systems at Nuclear Power Plants REGULATORY GUIDE:
This Regulatory Guide describes Diesel Generator Design features and testing provisions which are acceptable for verification of availability and reliability of Standby Power Sources.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE:
The design of the Standby Diesel Generator systems are in conformance with the subject regulatory guide with the following exception to Paragraph C.1.b(5): The diesel generator surveillance system is not required to have a "first out" annunication feature because annuniciation of individual protective trips give the operator adequate information for correct action.
With regard to the proposed Division 1 and 2 standby power sources, the staff required GE to conform to the position outlined in Standard Review Plan Appendix 7 BTP EICSB2, " Diesel Generator Reliability Qualification Testing". GE has stated its commitment to a qualification program in conformance with this commitment.
GESSAR states that readiness of the diesels is of prime importance and will be demonstrated by periodic testing. The testing program will be designed to test the ability to start the ESF system loads as well as to run under the load long enoirgh to bring all components of the system into equilibrium conditions. Full functional tests of the automatic control circuitry will be conducted on a periodic basis to demonstrate correct operation.
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GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C.2.2.9 REGULATORY GUIDE: 1.115 REVISION: 1 DATED: July 1977 TITLE: Protection Against Low-Trajectory Turbine Missiles
] REGULATORY GUIDE:
This guide describes guidelines for protecting safety-related structures, systems, and components against low-trajectory missiles resulting from turbine failure by appropriate orientation and placement of the turbine generator set.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE:
The main turbine generator is not within the GE scope of supply nor described in GESSAR. The orientation of the main turbine generator is established by the applicant. Therefore, the assessment of the plant relative to this guide must be done on a project unique basis. GE recommends that the main turbine generator be oriented A radially to the reactor building in order to preclude missile U damage.
The subject turbine missiles are treated in Section 3.5.2.1.1 of GESSAR. The implications for potential turbine missiles on plant design, the significance of reported turbine failures at or near design speed and at excessive overspeed are addressed with respect to modern nuclear units and the newer electro-hydraulic control systems. Reported turbine failures at or near design speed and at excessive stresses, are separately addressed under paragraphs 3.5.2.1.1.1 and 3.5.2.1.1.3 respectively. Analysis of the proba-bility of wheel burst failure and the probability of reaching excessive overspeed as they apply to modern nuclear turbines are covered in paragraphs 3.5.2.1.1.2 and 3.5.2.1.1.4 respectively.
Finally paragraph 3.5.2.1.1.5 evaluates the implications of potential turbine missiles on plant design.
The conclusions in GESSAR are stated such that the failure rate for modern nuclear turbines is much lower than that recommended in the Regulatory Guide; however, compliance with the Reg. Guide will be done on a project unique basis as stated above.
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PDA EXTENSION C . 2. 2.10 REGULATORY GUIDE: 1.117 REVISION: 1 DATED: April, 1978 TITLE: Tornado Design Classification Q REGULATORY GUIDE:
This guide describes a method for identifying those structures, systems, and components of light-water-cooled reactors that should be designed to withstand the effects of the Design Basis Tornado (see Regulatory Guide 1.76, " Design Basis Tornado for Nuclear Power Plants") including tornado missiles, and remain functional.
I EVALUATION 0F GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE: 1 The BWR-6/MK III Standard Plant Tornado Design Classification of structures, systems, and components, including their foundations and supports, that are designed or protected to withstand the effects of a Design Basis Tornado per Reg. Guide 1.76 (including
,/'g tornado missiles) without loss of capability to perform their U safety function) are those as listed in the " Appendix" of Regulatory Guide 1.117 except for portions that are not within GE scope of supply.
The General Electric Company modified their application with amendment #43 to incorporate the Staff's position on tornado missile velocities as a design basis for the GESSAR 238 Nuclear Island design.
The analysis of structures, shields, and barriers indicates tornado '
winds could damage non-category I structures. However, ability to shutdown the reactor, integrity of the containment, and capability of the essential heat removal systems are not impaired. All seismic Category 1 systems are protected by being housed in tornado resistant structures. Collapse of non-Category I towers or stacks will not endanger Category I structures since plant arrangement provides sufficient distance between them.
The specific plant applicant is responsible for the design and protection of GE out-of-scope of supply structures, systems, and components.
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GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C.2.2.11 ' REGULATORY GUIDE: 1,124 REVISION: 1 DATED: January.1978 TITLE: Design Limits and Loading Combinations for Class 1 Linear Type Component Supports O REGULATORY GUIDE:
This guide delineates acceptable levels of service limits and combinations of loadings associated with normal operation, postulated accidents, and specified seismic events for the design of Class 1 linear-type component supports as defined in Subsection i NF of Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE:
All Class 1 linear type component supports are designed in accordance with the rules and regulations of the ASME Boiler and Pressure Vessel Code,Section III, Subsection NF. The design requirement in O NF includes the analysis and/or tests to demonstrate that all such component supports will not deform under faulted plant conditions to the extent that would impair the required operability of the supported components to perform a safety function for safe shutdown of the plant. Therefore, the design meets the intent of 10CFR50 General Design Criteria 2.
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I GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION-l C . 2. 2.12 REGULATORY GUIDE: 1.130 REVISION: 0 DATED: ' July 1977 1 TITLE: Design Limits and Loading Combinations for Class 1 Plate-and Shell-Type Component Supports REGULATORY GUIDE:
This guide delineates design limits and combinations of loadings 1 associated with normal operation, postulated accidents, and speci- '
fied seismic events for the design of Class 1 plate-and-shell type .
component supports as defined in Subsection NF of Section III of l the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. i EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
i The GESSAR design fully complies with Subsection NF of Section III I O of the ASME Boiler and Pressure Vend Code. Some of the analytical details of the guide which are in addition to the code are not included in the plate and shell type component support analysis.
However, it is concluded that the conservative ASME Code Analysis demonstrates the adequacy of the design.
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GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C . 2. 2.13 REGULATORY GUIDE: 1.137 REVISION: 0 DATED: January 1978 TITLE: Fuel Oil Systems for Standby Diesel Generators REGULATORY GUIDE INTENT:
This regulatory guide describes a method that is acceptable to the NRC staff for. complying with the Commission's' regulations regarding fuel-oil systems for standby diesel generators. This guide requires the use of ANSI N-195-1976 and additional regulatory requirements.
EVALUATION OF. GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
This Regulatory Guide was assessed as a Category I item in section C.2.1.42.
Paragraph C.2 of this guide which is a Category II item, covers requirements for oil storage and oil storage equipment. Therefore it is out of the scope of GESSAR.
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GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION O
i C . 2. 2.14 REGULATORY GUIDE: 8.8 REVISION: 3 DATED: June 1978 TITLE: Information Relevant to Maintaining Occupational Radiation Exposure As Low As Reasonably Achievable (Nuclear Power Reactors)
Q REGULATORY GUIDE:
This Guide primarily is addressed to assuring that the plant operator carries out his activities so operational radiation exposure is ALARA. However, the system / equipment suppliers are also responsible for designing / arranging so as to achieve ALARA conditions per pertinent parts of Section C.2 EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
The GE design meets the requirements of this guide except as discussed below:
Position C.2.b(1) states: " Exposure of personnel servicing a C specific component (such as a pump, filter, or valve) to radiation from other components containing radioactive material can be reduced by providing shielding between the individual components that constitute substantial radiation sources and the receptor."
Typically, pumps, piping and valves comprise relatively closely arranged systems so that shielding valves one from another, especially where several valves are involved, would not usually be ALARA.
However, consideration is given to arranging valves and equipment for ease of access and rigging so time of exposure is reduced.
Process components such as filters, ion exchanges and concentrators are normally located in separate shielded compartments containing only the equipment and piping.
Regarding crud control, GE is currently participating in or considering O' programs in the area of crud reduction / control and the use of
' alternative materials. These programs are long range but results will be applied in new designs as they are shown to be ALARA and otherwise practical.
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.PDA EXTENSION Q
C . 2. 2.15 BRANCH TECHNICAL POSITION: ASB 9.5-1 DATED: May 1, 1976 TITLE: Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976 The assessment for this item was written with the assessment on Regulatory Guide 1.120, Section C.2.1.31 of this document.
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PDA EXTENSION C . 2.2.16 BRANCH TECHNICAL POSITION MTEB 5-7 DATED: July 1977 ~
TITLE: Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping O
BRANCH TECHNICAL POSITION SCOPE:
Branch Technical Position MTEB 5-7 provides material selection and processing guidelines, and identifies acceptable methods to minimize susceptibility to stress corrosion in BWR pressure boundary piping.
EVALUATION OF GESSAR (BWR-6/MK-III) WITH RESPECT TO BRANCH TECHN!JAL POSITION:
GESSAR is in compliance with the Branch Technical Position.
The requirements of MTEB 5-7 for inservice inspection and leak detection are applied or weld sensitized wrought material will be O eliminated by one of the following techniques:
- 2. Solution heat treat piping after welding.
- 3. Apply a corrosion resistant stai '
ss steel weld overlay (5%
minimum ferrite) in the region wr ch will be heat affected by subsequent welding.
These three techniques comply with the requirements of MTEB 5-7.
The above position supersedes the position taken on limiting sensi-tization in GESSAR Section 5.2.
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1 APPENDIX C PART III ASSESSMENTS OF CATEGORY III ITEMS AGAINST GESSAR-235 NUCLEAR ISLAND DESIGN O
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. SECTION III l
C.2.3 Category III Matters 1 The NRC definition of Category III matters is as follows:
"This review will determine the extent to which the design conforms to l these ma'tters or whether' acceptable alternatives are provided. If the i
design does not conform to the stated Category III requirements or no acceptable alternative has been provided, staff approved revisions to the design will be required."
I Assessment of Category III Matters 1
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C.2.3.1 REGULATORY GUIDE: 1,56 REVISION: 1 DATED: July 1978 -
j TITLE: MAINTENANCE OF WATER PURITY IN BOILING WATER REACTORS 1
Q REGULATORY GUIDE:
- This guide describes a method acceptable to the NRC staff for implementing the general design criteria with regard to minimizing the probability of corrosion-induced failure of the reactor coolant pressure boundary in boiling water reactors (BWRs) by maintaining
, acceptable purity levels in the reactor coolant. It further describes instrumentation acceptable to the NRC staff for determining the ,
condition of the reactor coolant and coolant purification systems in BWRs.
4 EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
The BWR/6 plant is designed so that the water purity requirements of this guide can be met.
Water quality is maintained within required limits by the condensate demineralizer system and the reactor water cleanup system. The oxygen content is maintained within required limits by de-aeration in the main condenser and appropriate operating procedures.
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l l GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C.2.3.2 REGULATORY GUIDE: 1. 68.. REVISION: 1 DATED: July 1978 ._
TITLE: Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water Cooled Nuclear Power Plants REGULATORY GUIDE:
This guide describes an initial startup test program, acceptable to the NRC staff, for demonstrating the capability to shut down the hot operating and cold reactor from_outside the control room by fulfilling three objectives:
- 1. Verification that the nuclear power plant can be safely shut ;
down from outside the control room. l
- 2. Verification that the nuclear power plant can be maintained in the hot shutdown condition from outside the control room.
- 3. Verification that the nuclear power plant has the potential for being safely cooled from hot and cold shutdown conditions from outside the control room.
O EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO RIGULATORY GUIDE:
The Remote Shutdown System is designed with the cam bility to accomplish the objectives of the test program outlined in the Regulatory Guide.
The system described in GESSAR (Section 7.5) prcvides remote con-trol for reactor systems needed to carry out the shutdown function from outside the main control room and bring the reactor to cold condition in an orderly f ashion.
The system provides a backup variation to the normal system used in the main control room permitting the shutdown of the reactor from outside the control room when feedwater is unavailable and normal O' heat sinks are lost (turbine and condenser).
Activation of the relief valves and the Reactor Core Isolation Cooling (RCIC) System will bring the reactor to a hot shutdown condition after scram and isolation. During this phase of shutdown, the suppression pool will be cooled as required by operating the Residual Heat Removal (RHR) System in the suppression pool cooling mode. Reactor pressure will be controlled and core decay and sensible heat rejected to the suppression pool by releasing steam through the relief valves. Reactor water inventory will be maintained O ny tne acic syste . This proceavre wiii cooi tae reector eod reduce its pressure at a controlled rate until reactor pressure becomes low enough to discontinue RCIC operation.
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GESSAR REGULATORY GUIDANCE ASSESSMENTS PDA EXTENSION l
C.2.3.3 ' REGULATORY GUIDE: 1.99 REVISION: 1 DATED: April 1977 i TITLE: Effects of flesidual Elements on Predicted Radiation Damage to Reactor Vessel Materials REGULATORY GUIDE:
Regulatory Guide 1.99 describes an acceptable procedure for pre-diction of radiation damage to the beltline of reactor vessels of light water reactors. '
i EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REGULATORY GUIDE:
GESSAR is in compliance with this Regulatory Guide. I Section 5.4.5 of GESSAR states that a surveillance test program will include the preparation of a series of Charpy V-notch impact specimens and tensile specimens from the base metal of the reactor vessel, weld heat-affected zone metal, and weld 7etal from a reactor ;
steel joint that simulates a welded joint in the reactor vessel at O the core beltline region.
The specimens and neutron monitor wires are placed near core midheight l adjacent to the reactor vessel wall. The specimens are installed at startup. Selected groups of specimens can be removed at intervals during the reactor lifetime, and the mechanical prop;rties determined.
GESSA3 states that the vessel material surveillance program will ,
meet the material surveillance program and the requirements of '
ASTM E185. Predictions of neutron radiation damage to the beltline of reactor vessels are based on the following procedures which follow the provisionsof Regulatory Guide 1.99.
The adjusted reference temperature will be determined at the 1/4T position of the reactor pressure vessel wall.
Q ThecurvesinFigure1 of Regulatory Guide 1.99 have been extrapolated from 50 F to 20 F l to provide information at low fluences which produce radiation '
induced temperature shifts less than 50 F. General Electric's radiation damage data (NE00-21708) are consistent with the adjustment curves shown in Regulatory Guide 1.99.
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O cessAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION l
C.2.3.4 ~ REGULATORY GUIDE: 1.101 REVISION: 1 DATED: March 1977 1
... 1 TITLE: Emergency Planning for Nuclear Power Plants REGULATORY GUIDE:
A This guide provides more complete guidance in developing the V emergency plans required in the FSAR.
EVALUATION OF GESSAR (BWR-6/ Mark III) WITH RESPECT TO REGULATORY GUIDE I
The emergency planni.ig for nuclear power plants must be accom- I plished by the owner and are not included in the GESSAR text.
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O GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION C.2.3.5 REGULATORY GUIDE: 1.114 REVISION: 1 DATED: November 1976 "l-TITLEi Guidance On.Being Operator at the Controls of a Nuclear Power Plant -
REGULATORY GUIDE:
O This guide describes a method acceptable to the NRC staff for complying with the Commission's regulations that require an operator to be present at the controls of a nuclear power plant.
EVALUATION OF GESSAR (BWR-6/ Mark III) WITH RESPECT TO REGULATORY GUIDE The operation of the nuclear plant is the responsibility of the owner. General Electric provides guidance and training but this is not within the scope of GESSAR and therefore is not mentioned.
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GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION C.2.3.6 REGULATORY GUIDE: 1.121 REVISION: 0 DATED:
TITLE: Bases for Plugging Degraded PWR Stream Generator Tubes "
This regulatory guide is not applicable to the BWR, O
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l O GESSAP REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION .
1 C.2.3.7 REGULATORY GUIDE: 1.127 REVISION: 1 DATED: March 1978 TITLE: Inspection of Water-Control Structures Associated With Nuclear Power Plants l
REGULATORY GUIDE:
This guide describes a basis for developing an appropriate in-service inspection and surveillance program for dams, slopes, canals, and other water-control structures associated with emergency cooling water systems as flood protection of nuclear power plants.
EVALUATION OF GESSAR (BWR-6/ Mark III) WITH RESPECT TO REGULATORY GUIDE The water control structures are a site unique feature and are not i included in the scope of GE supplied equipment or services. There-fore this guide is not applicable to GESSAR.
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, GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION 1
C.2.3.8 - REGULATORY GUIDE: 1.139 REVISION: 0 DATED: May 1978 -
SRP 5.4.7 REVISION: 1 DATED: April 1977 l
TITLE: RESIDUAL HEAT REMOVAL (RHR) SYSTEM !
I REGULATORY GUIDE AND STANDARD REVIEW PLAN INTENT:
This document describes a method acceptable to the NRC staff for complying with the Commission's regulations with regard to removal of decay and sensible heat after reactor shutdown.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
The GESSAR design complies with both Reg. Guide 1.139 and SRP 5.4.7 as follows:
- 1. The term " cold shutdown" used in the Guide is interpreted to be the condi-tion achieved when the reactor coolant temperature is less than 212 F. ,
- 2. Tha second sentence in Position C.2.a regarding alarms is deleted.
- 3. The word " diverse" is deleted from the third sentence of Position C.2.a since the extent of diversity is not defined.
- 4. The fifth sentence of Position C.2.a is deleted and the following sentence i is substituted: "The low pressure parts of the RHR system should be 3
capable of being isolated from the RCS by two valves in series. Closure logic and motive power for each valve should be divisionally separated from the other redundant valve."
The RHR system (GESSAR 5.5.7.3) will be designed to meet the following power generation design bases in order to remove the decay and sensible heat after reactor shutdown.
O The system shall have sufficient heat removal capacity to cool down the reactor to 125 F within approximately 20 hrs after shutdown.
Fuel pool connections shall be provided so that RHR heat exchangers can be used to supplement the fuel pool cooling capacity.
The system shall be able to condense reactor steam generated by decay heat and direct the condensate to the suction side of the RCIC pumps.
The RHR system is composed of four subsystems: containment heat removal, low pressure emergency core cooling, shutdown cooling, and steam condensing (via O RCIC) subsysteme.
C-93 113078
I GESLAR O RFG. GUIDE 1.139 (Continued)
The major equipment of the RHR system consists of three independent closed -
loops, two heat exchangers, three main system pumps, and service water pumps.
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Two loops, each consisting of a heat exchanger, main system pumps, and associ-ated piping, are located in separate protected areas of the auxiliary building. -
The third loop, made up of a pump and associated piping, is also located in a separate area of the auxiliary building to minimize the possibility of a single physical event causing the loss of the entire system.
The shutdown cooling and steam condensing (via RCIC) subsystems make use of the )
same hardware, consisting of pumps, piping, heat exchangers, valves, monitors, '
and controls. In the shutdown cooling mode, the RHR system can also be used to l supplement spent fuel pool cooling. The low pressure RHR piping it protected from Reactor Coolant System pressure by isolation valves.
Independent of the event that may necessitate plant shutdown, the reactor is normally brought to 100 psig by using either the main condenser, or when it is not available, the RCIC/HPCI system. Using the ADS valves to control pressure, reactor vessel makeup water is automatically provided via the RCIC/HPCI systems.
Which in this condition, the RHR system is used to maintain the suppression pool temperature within shutdown limits. Since the three systems are divisionally separate, no single failure, together with loss of off-site power, is capable of preventing system pressure from reaching the 100 psig level.
Paragraph 15.1.27.2.2 of GESSAR discusses the results from single failures that could result in a loss of shutdown cooling and that no unique safety actions are required. In these cases,. shutdown cooling is simply re-established using other normal cooling equipment. In cases where one of the RHRs shutdown cooling suction valves cannot be opened, alternate paths are available within the RHR system to accomplish the shutdown cooling function (Figure 15.1.27.1 of GESSAR).
Evaluation of the alternate paths demonstrated that, even under worst case conditions, a cooling path is available to remove residual heat from the core and thus comply fully with GDC 34.
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GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C.2.3.9 ' REGULATORY GUIDE: 1.141 REVISION: 0 DATED: April 1978 . . -
TIT. 'ainment Isolation Provisions for Fluid Systems REGl
. As a method acceptable to the NRC staff for co: che Commission's requirement with respect to con-tai solation of fluid systems. This guide incorporates ANSI N27 , " Containment Isolation P;ovisions for Fluid Systems".
EVALUATION OF GESSAR (BWR/6-MK III) WITH RESPECT TO REG. GUIDE:
The GESSAR design complies with the guide and N271 with the following clarifications:
(1) As explained in Section 6.2.4.3.2 of GESSAR, which is an evaluation of GESSAR Fluid Systrms against criterion 55, at least one isolation valve per p:p:line shall take the position of greatest safety when a single active failure is assumed.
O Some of the isolation valves are motor operated. Upon loss of power to a motor operated valve, the valve will fail "as is".
However, the other isolation valve in a pipeline is powered from a different electrical division and will close (GESSAR Section 6.2.4).
(2) Indicating circuits may be wired to power circuits in order to provide positive indication of availability. This wiring scheme is acceptable because the seismic testing of the finished panels in which the lights and circuits are mounted show that failures of these circuits are unlikely as a result of the seismic 1 event. Also, a single failure of the circuit in one valve is acceptable because the isolation function would still be carried out by the redundant valve, which has been acceptable in the past.
The safety design bases for the main steam and feedwater piping are contained in Section 5.5.9 of GESSAR and follows the guidelines in Reg. Guide 1.26 and 1.29. The inspection and testing of these systems are described in Section 5.5.12.4.
Specific mention is made in GESSAR Section 6.2.1.4.3.3 of the leak tests for valves performing an isolation function in the fluid system lines penetrating the containment.
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1 GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION RSB 5-2 REVISION: 0 C . 2. 3.10 TITLE: Reactor Coolant System Overpressurization Protection -
The assessment of this item can be found in Section C.2.1.49.
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APPENDIX C PART IV O AssessneNis or CAreGoRv Iv itens AGAINST GESSAR-235 NUCLEAR ISLAND DESIGN O
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SECTION IV Category IV Matters
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C.2.4 l l
The NRC definition of Category IV matters is as follows:
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" Category IV matters are those which have not been received by I i
the RRRC, but which the Director, NRR, deems to have sufficient safety attributes to warrant their being addressed during the PDA extension ;
1 review. These matters will be treated identically to the Category II I matters."
O Assessment of Category IV Matters L)
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GESSAR REGULATORY GUIDANCE ASSESSMENT
.O PDA EXTENSION
. C . 2. 4.1 REGULATORY GUIDE:. 1.12 REVISION: 1 DATED: April 1974 TITLE: Instrumentation for Earthquake REGULATORY GUIDE:
This guide sets forth requirements for seismic instrumentation. It
, states that the guidance in ANSI N18.5 is acceptable to the regu-latory staff when supplemented by additional requirements stated in the guide.
EVALUATION OF GESSAR (BWR-6/MR III) WITH RESPECT TO REG. GUIDE:
The installation of the specified seismic instrumentation in the reactor containment structure and at other Category 1 structures is discussed in GESSAR Section 3.7.4. This constitutes an effective system to record data on seismic ground motion as well as the frequency and amplitude relationship of the seismic response of representative major structures and systems. A prompt realization of pertinent data at the control room will provide sufficient information for evaluation of the seismic response in the event of an earthquake. The data obtained from the seismic instrumentation p will be sufficient to enable the operator to assess the plant V event. Provision of such seismic instrumentation complies with Reg. Guide 1.12 except for the minor alternate positions .,hown below which appear to have reasonable justification.
- 1. Regulatory Position Paragraphs C.1.a(1) and C.1.2(2) - Location of sensors directly on reactor vessel and reactor piping is not feasible and not done because of high surface temperatures.
Location of " rigid" supports is considered as satisfying the intent of Regulato y Guide 1.12.
- 2. Regulatory Positior Paragraph C.1.b - This paragraph is not construed to mean that quantitative information is to be provided on a control room front panel, but rather that the information be available promptly upon-receipt of a warning signal or signals. Front panel space is not O available for equipment that is not normally used nor essential to plant operation. Sufficient information is made available to the operator so that he can make a decision regarding continued operation without, leaving the control room.
- 3. Regulatory Position Paragraphs C.1 and C.3 - These para-graphs are in conflict with the Standard Review Plan, which classifies in two categories: (1) up to and in-cluding 0.3g and (2) above 0.3g. The standard 0.3g Q should be embraced in the first category.
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1 GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION C.2.4.2-
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REGULATORY GUIDE: 1.13 REVISION: 1 DATED: December 1975 l
TITLE: Fuel Storage Facility Design Basis O REGULATORY GUIDE:
This guide d'elineates the design criteria that are appropriately applied to the fuel storage facility.
EVALUATION OF GESSAR (BWR/6-Mark III) WITH RESPECT TO REG GUIDL': .
The discussion of spent fuel handling equipment and procedures is contained in Section 9.1.4 of GESSAR. The handling of loads over the spent fuel storage for the small crane is governed by adminis-trative controls. The height of lift of the general purpose crane is limited to six feet above the top of the spent fuel storage rack. Redundant safety interlocks are provided as well as limit
. switches to prevent accidentally running the grapple into the pool walls. The cask crane movement is limited by structural barriers.
The fuel building and fuel storage facilities will be designed to seismic Category 1 requirements. The concrete sides and roof of the fuel building will be designed to prevent tornado-borne missiles from causing unacceptable results. The design also in-cludes the ability to maintain a slight negative pressure in the building with the heating and ventillation system. The refueling platform is designed to seismic Category 1 requirements such that it will not lose its structural integrity following an SSE. There-fore, it is concluded that the GESSAR design follows the guidelines ,
set forth in the Regulatory Guide.
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._ __ - _ . . _ - _ - ._ . _ . _ . . _ _ _ . . . . _ . ~ . _ -_
i GESSAR REGULATORY-GUIDANCE ASSESSMENT POA EXTENSION C.2.4.3 REGULATORY GUIDE: 1.14 REVISION: 1 DATED: August 1975 c_
TITLE: Reactor Coolant Pump Flywheel Integrity Q REGULATORY GUIDE:
This guide describes a method acceptable to the NRC staff of imple-menting requirements with regard to minimizing the potential for '
failures of the flywheels of reactor coolant pump motors.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
BWR's are not equipped with flywheels on the recirculation system pumps or motors. Therefore this guide is not applicable.
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GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION C.2.4.4 REGULATORY GUIDE: 1.75 REVISION: 1 DATED: January 1975 1
TITLE: Physical Independence of Electric Systems '
4 REGULATORY GUIDE:
O This guide sets forth criteria for the separation of circuits and equipment. It states that the guidance in IEEE Std 384-1974 is acceptable to the Regulatory staff when supplemented by additional requirements included in the guide.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
The GESSAR design is in compliance with the Regulatory Position through the incorporation of the alternate approach cited below.
The proposed design criteria for the separation of redundant safety equipment was set forth in Section 3.12 of GESSAR and meets the O General Design Criteria 3, 17 and 21 pertaining to the physical O independence of Class 1E circuits and the regulatory position of i Regulatory Guide 1.75. Where the GESSAR design takes exception with the regulatory position is noted below.
- 1. Position C.1: Interrupting devices actuated only by fault current are not considered to be. isolation devices unless acceptable coordination can be verified by tests.
- 2. Position C.4 is implemented as follows:
" Associated circuits installed in accordance with Section 4.5(1) should be subject to the requirements of Class 1E circuits for cable derating, environmental qualification, flame retardance splicing restrictions, and raceway fill unless it can be demonstrated that Class 1E circuits are not degraded below an O acceptable level by the absence of such requirements."
- 3. Position C.6: Specific submittals of information will be based on NRC requests.
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[ Regulatory Guide 1.75 - continued O
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- 4. Position C.7 non-Class IE instrumentation circuits can be exempted from the provisions of Section 4.6.2 provided they are not routed in the same raceway as power and control cables . . .
or are not routed with associated cables of a redundant
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division.
- 5. Position C.8 is implemented as follows:
r3 "Section 5.1.1.1 should not be construed to imply that ade-U quate separation of redundant circuits can always be achieved within a confined space such as a cable tunnel that is effec-tively unventilated."
- 6. Position C.11 is implemented as follows:
...and should preclude the need to frequently consult reference..."
Certain non-Class 1E loads important to orderly shutdown and surveillance such as emergency lighting are not disconnected upon a LOCA signal.
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GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION
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REGULATORY GUIDE: 1.76 REVISION: 0 DATED: April 1974 l
TITLE: Design Basis Tornado for Nuclear Power Plants REGULATORY GUIDE:
This guide describes a design basis tornado acceptable to the Regulatory staff for each of three regions within the contiguous United States that a nuclear power plant should be designed to withstand without undue risk to the health and safety of the public.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
The GESSAR standard plant is designed to remain in a safe condition in the event the most severe tornado corresponding to Region 1 should occur. Tornado resistant structures are also designed to protect safety related systems and components from missile impact effects from airborne objects and debris transported by tornado O winds.
Section 3.3 GESSAR defines the wind loadings applicable to GESSAR ;
design and states that "in those cases where the plant is located I in the areas where higher design wind velocity is expected the plant structures will be reevaluated." Also "The data supporting the selection of the design velocity and recurrence interval will be presented by the Applicant."
The non-venting structures will be designed for the worst possible combinations of wind velocity and associated atmospheric pressure drop. The venting structures will be designed for the worst probable combination of wind velocity and associated difference of the pressure within the structures.
Tornado winds could damage non-Category I structures. However, ability to shut down the reactor, integrity of the containment, and capability of the essential heat-removal systems are not impaired.
All seismic Category I systems are protected by being housed in tornado-resistant structures. Collapse of non-Category I cooling towers or stacks would not endanger Category I structures since ,
plan; arrangement provides sufficient distance between them.
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GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C.2.4.6 REGULATORY GUIDE: 1.79 REVISION: 1 DATED: September 1975 TITLE: Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors O
REGULATORY GUIDE:
This guide describes a peroperational test program acceptable to the staff specifically for emergency core cooling systems (ECCS) in pressurized water reactor power plants.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
Since this guide describes preoperational test programs for pres-surized water reactor ECCS it is not applicable to GESSAR.
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i GESSAR' REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION C.2.4.7 REGULATORY GUIDE: 1.80 REVISION: 0 DATED: June 1974 m_. ,
TITLE: Preoperational Testing of Instrument Air Systems
.Q REGULATORY GUIDE:
This guide describes a method acceptable to the Regulatory staff for complying with the Commission's regulations with respect to verifying the operability of safety related instrument air systems before placing these systems into service.
EVALUATION 0F GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
The GESSAR design complies with this guide. This guide is ,
applicable only to the portion of the pneumatic Air Systems which '
support the operation of the safety related valves. Applicable preoperational procedures will meet the requirements of this guide.
O Although the specific test procedures are nat outlined in GESSAR, the document states that preoperational tests of the ADS system are
. conducted during the final stages of plant construction prior to initial startup. These tests assure correct functioning of all control, instrumentation, pumps, piping, and valves.
Chapter 14 of GESSAR lists the preoperational tests that are per-formed and states that acceptance criteria for each preoperational test are established by the designer'and are included in the pre-operational test specifications and instructions.
Preoperar,ional test results are reviewed and evaluated to assure compliance with all acceptance criteria.
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GESSAR REGULATORY GUIDANCE ASSESSMENT O PDA EXTENSION C.2.4.8 REGULATORY GUIDE: 1.82 REVISION: 'O DATED: June 1974 ,_
-TITLE: Sump for Emergency Core Cooling and Containment Spray Systems REGULATORY GUIDE:
The sumps or pump intakes that serve the emergency core cooling system (ECCS) and/or'the containment spray system (CSS) following a ,
loss of coolant accident (LOCA) should meet the criteria of the i regulatory guide which are intended to provide optimum use of the available coolant. This guide places requirements on PWR sumps which are provided for the collection of reactor coolant and chemically reactive spray solutions.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
1 This regulatory guide specifically states that 'it is applicable to pressurized water reactors. Therefore the guide is not applicable to GESSAR.
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GESSAR REGULATORY GUIDE ASSESSMENT k^l PDA EXTENSION C.2.4.9 REGULATORY GUIDE: 1.83 REVISION 1 DATED: JULY 1975 TITLE: Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes 0 -
This guide is not applicable to the BWR and was not assessed against GESSAR.
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GESSAR REGULATORY GUIDE ASSESSMENT PDA EXTENSION O
C.2.4.10 REGULATORY GUIDE: 1.89 REVISION: 0 DATED: November 1974 ..
TITLE: Qualification of Class 1E Equipment for Nuclear Power Plants
{ REGULATORY GUIDE:
This regulatory guide describes a method acceptable to the Regulatory staff for complying with the Commission's regulations with regard to qualifying Class 1E equipment and interfaces that are to be used in nuclear power plants and components or equipment of any interface whose failure could adversely affect any Class 1E equipment.
This guide states that the procedures described in IEEE Standard 323-1974 are acceptable with additional regulatory requirements.
~ EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
P The requirements of this guide are used by GE with the following 5
clarifications:
Regulatory Guide 1.7 does not adequately define the radiological source terms used for equipment qualification. The following radiological source term assumptions are used by GE to determine the requirement for the radiation qual:fication of equipment oder LOCA conditions:
- 1. Fission Product Release One hundred percent of the noble gases, fifty percent of the halogens, and one percent of the solid fission products are assumed to be released from the core.
- 2. Distribution of Fission Products Released a) Consistent with Regulatory Guide 1.7 (Safety Guide 7, dated 3/10/71) assumptions, 50% of the halogens and 1% of the solids present in the core are intimately mixed with the coolant water.
b) Consistent with Regulatory Guide 1.3 assumptions, 100% of the noble gases and 25% of the halogens present in the core are uniformly mixed in the primary containment free air volume.
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GESSAR PDA-EXT
- 3. Use of Other IEEE Standards Appropriate constraints on the application of IEEE-323 to the qualification of large electrical and electro-mechanical equipment may be applied in accordance with IEEE Standards for p qualification of such equipment. The standards presently available and which may be used as appropriate are:
a) IEEE-334-1975 " Standard for Type Tests of Continuous Duty Class 1E Motors for Nuclear Power Generating Stations" b) IEEE-382-1972 " Standard for Type Tests of Class 1E Electric Valve Operators for Nuclear Power Generating Stations" c) IEEE-383-1974 " Standard for Type Tests of Class 1E Electric Cables, Field Splices and Connections for Nuclear Power Generating Stations" d) I-EEE-387-1977 " Criteria for Diesel Generator Units Applied as Standby Power Supplies for Nuclear Power O Generating Stations" A detailed implementation program addressing IEEE 323-1974 is currently under way. Methods of implementing IEEE 323-1974 for the GE NSSS equipment will be defined in NEDO 21898
" Environmental Qualification of Class I Electrical Equipment".
Methods of implementing IEEE 323-1974 in the Nuclear Island design will be covered in other documentation.
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C 2.4.11 REGULATORY GUIDE: .1.93 REVISION: 0 DATED: December 1974 TITLE: Availability of Electric Power Sources REGULATORY GUIDE INTENT:
Section 50.36(c)(2), " Limiting Conditions for Operation," of 10 CFR Part 50, " Licensing of Production and Utilization Facilit.ies,"
requires the Technical Specifications to include the limiting conditions for operation (LC0) cand actions required to be taken by the licensee when the LC0 is not met.
This guide describes operating procedures and restrictions accept-able to the Regulatory staff which should be implemented if the available electric power sources are less than the LCO.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG GUIDE:
O The offsite power sources for a plant referencing GESSAR are in the applicant's scope. This guide establishes LC0's based on c~nbina-tions of offsite and onsite power. Therefore, this guide is not considered applicable to GESSAR.
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f GESSAR REGULATORY GUIDANCE ASSESSMENT PDA EXTENSION l
C . 2.4.12 REGULATORY GUIDE: 1.104 ._ REVISION: 0 DATED: February 1976 TITLE: Overhead Crane Handling Systems for Nuclear Power Plants O REGULATORY GUIDE: .
This guide" includes requirements, acceptable.to the NRC, to be placed on o'verhead cranes which handle critical loads such as spent fuel casks or loads which if dropped could damage safety related systems.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REG. GUIDE:
1 Revision 0 of this guide was issued for comment in February 1976. !
It was found to be unusable by GE as well as the Nuclear Industry l in general. Subsequently, a preliminary assessment of the GESSAR design was done against Revision 1 (draft, January 1978). We believe the GESSAR design is in full compliance with this draft.
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GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION C. 2.4.13 SRP SECTION: 5.4.2.1 DATED: 11/24/78 (CATEGORY IV, ITEM B-1)
TITLE: Branch Technical Position 5-3, Monitoring of Secondary Side Water Chemistry in PWR Steam Generators V
This item is not applicable to the BWR and therefore was not assessed against the GESSAR design.
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4 GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION i
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- C.2.4.14 SRP SECTION 6.2.1 ~ DATED
- 11/24/75 6.2.lA (CATEGORY IV, ITEM B-2) 6.2.1B 6.2.1.2 6.2.1.3
)h 6.2.1.4 6.2.1.5 i
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i TITLE: Branch Technical Position CSB-6-1, Minimum Containment j Pressure Model for PWR ECCS Performance Evaluation These items are not applicable to the BWR and therefore were not assessed against the GESSAR design.
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GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION O C . 2.4.15 SRP SECTION: 6.2.5 REVISION: 0 DATED: 11/24/75 CATEGORY IV, ITEM B-3 TITLE: .BTP-CSB-6-2, Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident _
BRIEF DESCRIPTION OF SRP 6.2.5 (BTP-CSB 6-2)
CSB 6-2 describes an acceptable method of implementing control of combustible gas concentrations in BWR/6 Mark III Containment.
EVALUATION OF GESSAR (BWR-6/MX III) WITH RESPECT TO SRP 6.2.5 (CSB 6-2):
The hydrogen concentration will be the limiting parameter, therefore the hydrogen gas may be referred to as the combustible gas. The hydrogen control system for the GESSAR design consists of the following:
- 1) Containment Atmosphere Monitoring (CAM) System
- 2) Hydrogen Mixing System
- 3) Hydrogen Recombiner System
- 4) Drywell Purge Recombiner Backup The CAM system is to be automatically initiated upon detection of loss of coolant accident conditions in order to monitor the hydrogen concentration. The system is classified as a safety system and shall be designed to meet Safety Class 3 requirements. Safety class 3, as applied to the CAM system, requires classification as Seismic Category I, Electrical Class IE and Quality Group C. The CAM system is designed to meet requirements for an engineered safety feature. Therefore, CAM systems satisfy the provision of CSB 6-2.
HYDR 0 GEN MIXING SYSTEM Hydrogen mixing system consists of two air compressors which take p' suction near the operating floor elevation and discharge through the ceiling of the drywell. The compressor head is sufficient to depress the water level inside the drywell weir wall gap and force air to exit the drywell through the submerged suppression pool vents. Any entrained steam in the exit flow will be condensed as the flow bubbles through the suppression pool. The hydrogen _ mixing system meets all requirements for Safety Class 2 equipment. The hydrogen mixing blowers will be installed in the reactor building to provide an engineered safety feature to maintain the containment integrity. The blower is classified as an active component and Seismic Category I, Safety Class 2. Therefore, the hydrogen mixing C' : system complies with CSB 6-2.
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GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION O SRP Section 6.2.5 Page 2 HYDR'0 GEN REC 0MBINER SYSTEM The hydrogen recombiner is an engineered safety feature necessary to maintain the containment in a safe post-accident condition. The components of the system are classified as Safety Class 2, Seismic Category I. The Quality Group Class for components is B. The l
,q GESSAR design satisfies the regulatory requirement for recombiner '!
U flow rates since hydrogen concentration is kept below 4 volume percent. The recombiner may be installed either inside or outside containment. If the recombiner is installed outside containment additional justification of how the design meets this BTP will be provided.
ORYWELL PURGE RECOMBINER BACKUP This system is not required to be redundant, seismically qualified, nor meet the single failure criteria per Regulatory Guide 1 '.
However, the 2-inch drywell vent line whicF 's employed for t.his mode of operation is seismic Category I, sa."ety class 2 since the vent line and valv0s constitute part of the primary containment boundary. The system will not in itself create' safety problems pd which may affect the containment integrity. The purge exhaust shall be treated by .le Standby Gas Treatment System so that the potential iodine release does not exceed the dose limits establisned in 10CFR100. The purge system utilizes the hydrogen mixing compressors.
The major portion of the drywell atnosphere is transferred back to j the containment via the drywell suppression pool vents. In conjunction, 1 the 2-inch drywell J1eed next will be opened for purging to the shield annulus. The recombiner backup purge system satisfies CSB 6-2.
The parameter values listed in Table 1 of CSB 6-2 are used in the General Electric Computer Code, which calculates hydrogen concentration in containment following a loss-of-coolant accident. This code was developed in compliance with SRP 6.2.5.
References:
- 1) Containment Hydrogen Recombiner, General Electric Design Specification, A62-4590, 22A5621.
- 2) NEDE-21501, Engineering Manual Hydrogen Oxygen Concentration Index.
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STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION C.2.4.16 SRP SECTION: 6.2.3 REVISION: 0 DATED: 11/24/75 Category IV, Item B-4 -
TITLE: BTP-CSB-6-3 Determination of Bypass Leakage in Dual Containment plants BRIEF DESCRIPTION OF SRP SECTION 6.2.3 (BTP-CSB-6-3)
O The basis for determining bypass leakage paths can have an impact on site dose (10CFR100). This BTP sets criteria which systems that will contribute / mitigate site dose have to address in order to provide some assurance that there will not be a safety related concern in regards to bypass leakage.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO SRP 6.2.3 (BTP-CSB-6-3)
The leakage from the primary containment following a postulated design basis loss-of-coolant accident (LOCA) is limited such that off-site dose does not exceed the guidelines of Regulatory Guide 1.3.
There are basically four types of potential leakage pathways for a Mark III free standing steel containment: (1) leakage from the O primery coateinmeat directly to the environs < unprocessed); (2) leakage from the primary containment to secondary containment annulus - (mixed and processed); (3) leakage from the primary containment to the secondary containment excluding the annulus (processed); and (4) leakage from the secondary containment to the environs during time periods when the secondary containment pressure is more positive than a -1/4 in, w.g.
The licensing topical report (NEDO 21424) "238 NI Containment Bypass Sealing and Testing Methods," which has been submitted to the NRC for review and approval, addresses all of the specific criteria defined on this BTP.
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GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION O C.2.4.17 SRP SECTION: 6.2.4 REVISION: DATED: 11/24/75 CATEGORY IV, ITEM B-5 TITLE: BTP 6-4, Containment Purging During Normal Operation
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BRIEF DESCRIPTION OF SRP 6.2.4 - BTP 6-4:
This branch technical position pertains to system lines which can provide an open path from the containment to the environs during normal plant operation; e.g., the purge and vent lines of the containment purge system. It supplements the position taken in SRP O- Section 6.2.4 EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO SRP 6.2-4, BTP 6-4:
GESSAR Section 9.4.5.4, " Justification of Drywell Venting During Normal Operation," describes how the GESSAR design meets the BTP.
The containment purge system is designed to prevent any significant release of radioactive material following a loss of coolant or fuel handling accident.
The interconnecting piping of the penetration and the values are Seismic Category I and Quality Group B. The space between the two isolation valves allows a sealing air supply.to the penetration during a LOCA. This ensures that leakage out of the containment O- can be eliminated through the ventilation penetrations. A connection directly to the standby gas treatment system serves the drywell and containment when purging 1 required through the standby gas treatment and the isolation valves, ope.' d ring aormal operation are capable l of full closure in 10 records at peak temperature against the full pressure of the containment air derates. The valves are fail safe l which close on loss of actuating air pressure. Remote manual controls also enable the operator to isolate the containment venti-lation systems.
In the event of a LOCA, the signals for closure actuation include LOCA high drywell pressure and/or low low reactor water level, high airborne radioactivity in the exhaust duct air intake, and reactor vessel high pressure.
O Tne BTe stetes thet en enelys4, to snew tne reduction in contein-ment pressure be done to assure that adequate net positive suction '
head (NPSH) is maintained. This is not done since the GESSAR design does not depend on containment pressure to maintain adequate NPSH. The BTP also states that allowable leak rates should be specified for a spectrum of design basis flows against which the ,
valve must close. This is not done since a positive leakage control system is used between the valves.
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GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION O C.2.4.18 SRP SECTION: 9.1.4 REVISION: 0 DATED: 11/24/75 CATEGORY IV, ITEM B6 TITLE: BPT ASB-9.1, Overload Handling Systems for Nuclear Power Plants BRIEF DESCRIPTION OF SRP 9.1.4, (BTP-ASB-9.1)
This Branch Technical Position describes an acceptable method for designing overhead handling systems when failure could cause damage to spent fuel or safety related systems.
EVALUATION OF GESSAR DESIGN (BWR-6/MK III) WITH RESPECT TO SRP 9.1.4:
The overhead handling systems are designed to comply with the intent of the Branch Technical Position in that no single component failure will cause the load to drop or to swing uncontrollably out of an essentially level attitude. The safety factor of all lifting members is five or better in reference to the ultimate breaking strength of the material.
An example of the approach taken in design is the reactor pressure vessel head strongback which is used for lifting both the pressure vessel head and the drywell head.
The strongback is designed such that one leg of the cruciform will support the rated load and such that no single component failure O will cause the load to drop or swing uncontrollably out of an essentially level attitude. The structure is designed in accordance with "The Manual of Steel Construction" by AISC. All welding is in accordance with ASME Boiler and Pressure Vessel Code Section IX. A safety factor of 5 or greater in reference to the ultimate material strength is used for the design. The completed assembly is proof tested at 125 percent rated load. After the load test, all structural welds are magnetic particle inspected.
The dryer and separator sling is a lifting device used for transporting the steam dryer or the shroud head with the steam separators between the reactor vessel and the storage pools. The sling consists of a cruciform shaped structure which is suspended from a hook box with four wire ropes and turnbuckles. The hook box, with two hook pins, n engages the reactor service crane sister hook. On the end of each V arm of the cruciform is a socket with a pneumatically operated pin for engaging the four lift eyes on the steam dryer or shroud head.
The sling has been designed such that one hook pin and one main beam of the cruciform is capable of carrying the total load and so that no single component failure will cause the load to drop or swing uncontrollably out of an essentially level attitude. The safety factor of all lifting members is 5 or better in reference to the ultimate breaking strength of the material. The structure is designed in accordance with "The Manual of Steel Construction" by Q AISC. The completed assembly is proof tested at 125 percent or greater of rated load and all structural welds are magnetic particle inspected after load test.
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GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION 4
C.2.4.19 SRP SECTION: 10.4.9 REVISION: DATED: 11/24/75 ,_
CATEGORY IV ITEM B-7 TITLE: BTP ASB-10.1, Design Guidelines for Auxiliary Feedwater Systsm Pump Drive and Power Supply Diversity for PWRs (G/ This item is applicable to PWRs and therefore was not assessed against GESSAR.
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GESSAR. STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION O C.2.4.20 SRP SECTION: .3.5.3 REVISION: 0 DATED: 11/24/75 CATEGORY IV, ITEM B-8 TITLE: Procedures For Composite Section Local Damage Prediction (SRP --
Section 3.5.3, Par. II.1.C) l BRIEF DESCRIPTION OF SRP 3.5.3: ;
l
(~ This SRP states that NRC staff's position relating to the procedures to be utilized in the design of seismic Category I structures to withstand the effects of missile impact on composite missile barriers.
EVALUATION OF GESSAR'(BWR-6/MK III) WITH RESPECT TO STANDARD REVIEW PLAN 3.5.3 (FAR. II.1.C)
The GESSAR design complies with this SRP as discussed under SRP 3.5.3, Category IV, Item C-5 (Section C.2.4.40 this appendix).
Missile-resistant barriers and structures are designed to withstand and absorb missile impact loads to prevent damage to the protected structures, systems and components. The principal design features of structures serving primarily as missile-resistant barriers are identi-fied in GESSAR Section 3.5.1. These barriers are single barrier O concrete structures. For example, the shield building constitutes i'
the barrier which protects the containment from postulated external missiles, and the drywell constitutes the barrier which protects the containment from internally generated missiles. These barriers have i been demonstrated to have adequate penetration-resistant characteristics (see GESSAR 3.5.4.1).
There are no composite or multi-element missile barriers that are designed for the sole purpose of repelling postulated missiles.
Therefore,Section II.l.C is not applicable to missile-resistant '
barriers and structures.
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i p .GESSAR STANDARD REVIEW PLANT ASSESSMENT V
PDA EXTENSION ,
C.2.4.21 SRP SECTION: 3.7.1 REVISION: 0 DATED: November 24, 1975 ~~;
(Category IV, Item B-9) j i
TITLE: Develcpment of Time History for Soil-Structure Interaction Analysis BRIEF DESCRIPTION OF SRP SECTION 3.7.1, PARA. II.4 This SRP paragraph describes NRC Staff Acceptance Criteria for developing the time history to be used at the base of the soil structure interaction system.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO SRP 3.7.1, PARA. II.2:
The development of design time history for Soil structure Interaction analysis in GESSAR is consistent with SRP Section 3.7.1.
O The Auxiliary Building, Cable Room, Diesel Generator Building, Radwaste Building and Reactor Building have independent foundations supported on soil, rock or compacted backfill. In all cases the maximum value of embedment is used for the dynamic analyses to determine seismic soil-structure interaction effects. The founda-tion support materials will withstand the pressures imposed by appropriate loading combinations, without failure. ;
This issue is also discussed in GESSAR section 3.7.1.
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1 m GESSAR STANDARD REVIEW PLAN ASSESSMENT l
V 1 PDA EXTENSION C.2.4.22 SRP SECTION: 3.7.2 REVISION: 0 DATED: 11/24/75 _..
Category IV Item B-10 TITLE: Procedures For Seismic System Analysis ,
1 BRIEF DESCRIPTION OF SRP SECTION 3.7.2, PARAGRAPH II This SRP paragraph describes criteria acceptable to the NRC staff for seismic analysis methods including natural frequencies and response loads and procedures for analytical modeling. ,
1 1
EVA UATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO SRP 3.7.2, PARAGRAPH II:
Analysis of seismic Category 1 structures, systems and components is accomplished where applicable, using the response spectrum or time history approach. Either approach utilizes the natural period, l mode shapes and appropriate damping factors of the particular system. Certain pieces of equipment having very high natural be frequencies may be analyzed statically. In some cases, dynamic testing of equipment may be used for seismic qualification.
While combining modal responses, all modes except closely spaced modes are combined by the square root of the sum of squares (SRSS) l method. However, for closely spaced modes, GE does not comply with SRP-3.7.2-II.7. Instead, the closely spaced modes are combined by taking the algebraic sum of such modes or by using the Double Sum Method as given in Reference 1. (These alternate procedures and ,
justification for them are given in GESSAR-251. However, GESSAR-238 l reflects the SRP requirements (i.e.,) combines closely spaced modes !
by absolute summation. Effect of parameter variations on floor '
response spectra is accounted for in the GESSAR by broadening tha spectrum peaks by 110 percent as against the SRP-3.7.2-II.9 regiire-ment of 115 percent.
REFERENCE 1 l
A. K. Singh, S. L. Chu and S. Singh, " Influence of Closely Spaced Modes in Response Spectrum Method of Analysis", ASCE Speciality Conference on Structural Design of Nuclear Plant Facilities, December 1973.
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GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION l
C.2.4.23 SRP SECTION: 3.7.3 REVISION: 0 DATED: November 24, 1975 I (Category IV Item B-11)
TITLE: Procedure for Seismic Subsystem Analysis :
1 Os BRIEF DESCRIPTION OF SRP SECTION 3.7.3, PARA. II- -
This SRP paragraph describes criteria acceptable to the NRC Staff for seismic subsystem analysis such as determination of the number '
of earthquake cycles and analytical procedures for piping systems.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO SRP 3.7.3, PARA. II:
The GESSAR design meets all the requirements of SRP Section 3.7.3, Para. II for seismic subsystem analysis, except for the following:
- a. Only one OBE intensity earthquake is postulated for fatigue O evaluation as against 5 OBE's required in SRP-3.7.3-II-2.
- b. Requirements and exceptions for combining closely spaced modes is the same as in the assessment for SRP 3.7.2.
Section 3.7.3 of GESSAR provides a detailed description of the seismic subsystem analysis.
The seismic analysis of the equipment and piping in NSSS scope essentially meet the intent of the criteria set forth in SRP 3.7.3.
The slight deviations are given below.
(i) All ASME subsection NB components have been analyzed to 60 total maximum stress cycles due to 0BE's. The RPV and internals have been analyzed to 10 total maximum stress cycles due to OBE's.
(ii) Static analysis was employed for seismic analysis of seismic category I RHR heat exchanger. 'Its fundamental natural frequency was computed to be approximately 20 cps.
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.~. _,. - . . - , . - .- - .-.
GESSAR STANDARD REVIEW PLAN' ASSESSMENT PDA EXTENSION O C.2.4.24 SRP SECTION: '3.8.1 REVISION: 0 DATE: 11/24/75 CATEGORY IV, ITEM B-12 TITLE: Concrete Containment .
BRIEF DESCRIPTION OF SRP 3.8.1:
This SRP provides the NRC staff's position on design of concrete containments.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO STANDARD REVIEW PLAN:
This SRP does not apply to the Mark III free-standing metal containment.
Refer to SRP 3.8.2.
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GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION O C.2.4.25 SRP SECTION: 3.8.? REVISION: 0 DATED: 11/24/75 CATEGORY IV, ITEM B-13 TITLE: Design and Construction of Steel Containments (SRP Section 3.8.2, c Par. II)
BRIEF DESCRIPTION OF SRP 3.8.2:
This SRP states the NRC staff's position on steel containments and
.O Other Class MC steel portions of steel containments.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO STANDARD REVIEW PLAN 3.8.2 (PAR. II)
The containment vessel is described in Section 3.8.2 of GESSAR and is a free-standing fixed-end vertical cylindrical steel pressure vessel with an ellipsoidal head and a flat bottom steel liner plate. The cylindrical shell has horizontal external stiffness and is embedded and anchored into the mat foundations.
The design of the containment is based on the natural phenomena postu-lated to occur at the site and the Design Basis Accident which assumes the instantaneous circumferential rupture of a main steam line upstream of the main steam line flow restrictor. These conditions are coupled O with the loss of offsite power and the partial loss of the redundant engineered safety features systems.
The design, materials and fabrication are in accordance with GESSAR Section 3.8 which states that all structures and systems important to safety shall be designed with the applicable codes and standards listed.
The steel containment vessel is classified Class MC in accordance with Sub-Article NA-2130,Section III of the ASME Code. The steel cylindrical shell and dome of the steel containment vessel, including all penetrations and attachments with boundaries defined in GESSAR Section 3.8.2.1.4 is designed and constructed in accordance with Subsection NE Class MC components, including the requirements for quality assurance of Article NA-4000, and inspection requirements of Article NA-5000 of Section III cf the ASME Code.
O Loads and loading combinations are described in Section 3.8.3 of GESSAR and includes the loads and loading combinations listed in paragraph II.3b of the SRP except that the SRP stipulates that: 1 PRV (tension) <.1.5 SRV PRV (compression) < 1.2 (allowable given by NE-3133) and for the GESSAR design:
PRV tension and compression < the greater of 1.2 Sm or Sy O
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GESSAR STANDARD REVIEW PLAN ASSESSMENT l
PDA EXTENSION Q C.2.4.26 SRP SECTION: 3.8.3 REVISION: 0 DATED: 11/24/75 CATEGORY IV, ITEM B-14 TITLE: Structural Design Criteria for Category I Structures _.
Inside Containment (SRP Section 3.8.3, Para. II) "
BRIEF DESCRIPTION OF SRP 3.8.3, PAR. II:
This guideline states the requirements for the design of structures O inside containment such as the drywell, the weir wall, containment pool walls, reactor pedestal, shield wall and pipe support structure, I
and other internal structures.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO SRP 3.8.3, PAR. II:
The structural details of the internal containment structures of the GESSAR design are discussed in Section 3.8.3 of GESSAR and are in compliance with the guidelines of SRP 3.8.3, paragraph II. This section of GESSAR includes a separate writeup for each appropriate internal structure as defined in the SRP for BWR containment internal structures, specifically, the drywell, the weir wall, containment, pool walls, reactor pedestal, shield wall, and pipe support structure, and other internal structures.
O The ePoi4ce8ie codes end stenderds for the des 42n of these interne, structures in the containment follow:
(1) American Concrete Institute - American Society of Mechanical Engineers, " Proposed Standard Code for Concrete Reactor Vessels and Containments," proposed Section III, Division 2. ASME Boiler and Pressure Vessel Code, and applicable portlans of Article cc-3000 for concrete pressure resisting portion of the drywell, containment pools, and weir wall.
(2) American Institute of Steel Construction (AISC), " Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings", 1969, for structural steel framing, plate wall steel elements, and steel other than drywell head and hatch covers.
(3) ASME Boiler and Pressure Vessel Code, Subsection NE of Section III, l
. Division 1, Class MC, governs d2 sign of steel elements serving
-as pressure boundaries, sucn as drywell head, locks and penetrations, but does not govern fabrication, installation and test of these elements, because the drywell is not a containment. ,
(4) American Welding Society Code AWS D12.1-66. This code wil.1 apply to butt welded reinforcement splices.
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GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION A Item B-14 V Page 2 (5) Regulatory Guide 1.15, Testing of Reinforcing Bars for Category I Concrete Structures, AEC. ..
-4 (6) AEC publications TID 7024 and TID 25021, " Nuclear Reactors and Earthquakes" and " Summary of Current Seismic Design Practice for Nuclear Reactor Facilities" are utilized as input for seismic design of all Category I structures.
O (7) Reguietory Guide 1.29, Seismic oesign Ciessificetion AeC.
(8) Plant principal specifications, by structural designer (to be listed and summarized later).
(9) American Association of State Highway Officials (AASHO),
" Standard Specifications for Highway Bridges", tenth edition 1969, governs distribution of concentrated loads on concrete slabs.
(10) American Welding Society (AWS), " Coding for Welding in Building Construction". (AWS D1.0-69).
(11) Regulatory Guide 1.10, Mechanical (Cadwell) Splices in Reinforcing.
O (12) Regoletery Guide 1.55 Concrete eiecement in Cetegory 1 structures.
Loads and loading combinations for the structures are presented in ,
Section 3.8.3.1.3 of GESSAR and is in comp.liance with SRP except that a load factor of 1.25 is not used. In the load combination the alternating components of Safety Relief Valve and seismic response are combined by the Square Root Sum of the Squares (SRSS) method as defined in Section 3.7.3.3 of GESSAR.
The structural design criteria is contained in section 3.8.3.1.5 of .
GESSAR, a d are in compliance with the specification in the SRP.
The con. rete pressure resisting portions of the drywell, containment pool, and weir wall are defined to meet the American Concrete Institute - American Society of Mechanical Engineers, proposed " Code for Concrete Reactor Vessels and Containments," proposed Section III, O Division 2, ASME Boiler and Pressure Vessel Code, applicable portions of Article CC-3000. The appropriate revisions will also apply.
The structured design criteria for the steel portion follow Standard Review Plan 3.8.2 as specified. ;
The GESSAR sections for each of these inside containment. structures also describe the materials, quality control, and construction techniques, which also meet the SRP guidelines. The testing pro-grams are also defined for each component in their respective sections of the GESSAR documentation as defined by the SRP.
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t GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION O C.2.4.27 SRP SECTION: 3.8.4 REVISION: 0 DATED: 11/24/75 ,
CATEGORY IV, ITEM B15 1 TITLE: Structural Design Criteria for Other Seismic Category I -i Structures (SRP Section 3-8.4, Par. II)
BRIEF DESCRIPTION OF SRP 3.8.4:
This SRP states the NRC staff's position on the design of Category I l O- structures other than the containment and its interior structures.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT SRP 3.8.4, Par II:
Other Seismic Category I structures which are part of the Nuclear Island are the shield building, the auxiliary and fuel buildings, control building, diesel generator building, and the radwaste building substructure. Each of these structures are described in Section 3.8.4 l of GESSAR. This section of GESSAR includes statements for Applicable !
Codes, Standards and Specifications and a listing of the Loads and Loading Combinations for each structure with distinct sections for steel members and concrete members as prescribed by the SRP and all the loads are in conpliance with SRP 3.8.4.
The Category I structures within Nuclear Island other than the contain-ment which contain high energy pipes are the shield building. Guard pipes and the steam tunnel walk protect the shield building from impact by the high energy pipes. However, the shield building will be designed to accommodate the guard pipe support forces transmitted to the shield building.
The auxiliary building, steam tunnel, and RHR rooms will be designed to handle the consequences of high energy pipe breaks. The RHR rooms will be designed for 2 psid pressure, the associated temperature rise, and the jet force. Steam generated in the RHR compartment from the pipe break will exit to the steam tunnel through blow out panels.
The design of the other Category I structures comply with this SRP except that a load factor of 1.25 is not used. In the load combina-tion the alternating components of Safety Relief Valve and seismic O. response are combined by the Square Root Sum of the Squares (SRSS) method. This design is discussed further in Attachment A, Section A.8.4 Appendix 3B of GESSAR.
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. GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION C . 2.4. 28 SRP SECTION: 3.8.5 REVISION: 0 DATED: 11/24/75 (Category IV, Item B-16)
TITLE: Structural Design Criteria For Foundations Q BRIEF DESCRIPTION OF SRP 3.8.5, PARA II:
This section of the SRP defines the internal branches of the NRC responsible for the review responsibilities of this SRP section and the applicable review areas, acceptance criteria, review procedures, and evaluation findings statement.
EVALUf. TION OF 238 GESSAR (BWR-6/MK III) WITH RESPECT TO SRP SECTION 3.8.5, PARA. II:
The NRC SRP area of review and the associated NRC SRP acceptance criteria for the areas of review are adequately covered by the 238 GESSAR BWR-6/MK III plant with the exceptions noted below.
II.1 Description of the Foundations A. The requirement of the SRP for the GESSAR are met for the following buildings:
- 1. Reactor Bldg, which includes the Containment Bldg.,
Shield Bldg. , Drywell Structure, Weir Well Structure.
Pedestal and Shield Wall Structures, and other Reactor Bldg. Internal Structures.
- 2. Auxiliary Bldg.
- 3. Fuel Bldg.
- 4. Control Bldg.
O s- oiesei ceaeretor 81d9
- 6. Radwaste Bldg.
- 7. Cooling Water Intake Structure.
B. Exceptions to the SRP for the GESSAR Design:
The Diesel Generator Oil Storage Tank and Vault and the Condensate Storage Tank Retaining Basin are the responsi-bility of the applicant and are to be discussed _in the PSAR of the applicant per GESSAR, pages 3.8-53 and 54.
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GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION C.2.4.29 SRP: 3.7 11.3 REVISION: 0 DATED: 11/24/75 11.2 11.4 CATEGORY IV, ITEM B-17 TITLE: . Seismic Design Requirement for Radwaste System and Their Housing Structures (SRP Section 11.2, BPT-ETSB 11-1, Par B.V)
BRIEF DESCRIPTION OF SRP SECTION 11.2, BTP-ETSB 11-1, PAR. B.V:
This position paper sets forth design guidance for radwaste systems and recommendations are given for provisions to preclude the inadvertent O re'eese or red 4oective meter 4eis ene sets criterie for seismic design requirement of the radwaste system and structures.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO SRP SECTION 11.2, BTP ETSB 11-1, PAR. B.V:
The offgas system is in compliance with the SRP and meets the seismic design requirements for gaseous radioactive waste management systems.
Seismic design criteria for the charcoal adsorbers are contained in the NRC Effluent Treatment Systems Branch Technical Position 11-1 (ETSB BTP 11-1) Revision 1, " Design Guidance for Radioactive Waste Management Systems Installed in Light-Water-Cooled Nuclear Power Reactor Plants." The General Electric Company design contains the p/
L following criteria. The support elements, including the skirts, legs and anchor bolting for the charcoal adsorber tanks of the offgas system shall be designed as follows:
(1) The fundamental frequency of the charcoal adsorber tanks, including the support elements, is greater than 33.
(2) The charcoal adsorber tanks are mounted on the basemat of the building housing the tanks.
(3) The charcoal adsorber tanks, including the support elements, are designed with a horizontal static coefficient equivalent to that of the OBE.
GESSAR (Section 3.6.2.6.2) states that the radwaste building is so designed to preclude accidental release of radioactive material to s the environs. The equipment classifications for the radwaste system are listed in Table 3.2.1, page 3.2-15 of GESSAR and is designated as Safety Class 3 structure, Seismic Category I. The description of the structure is in Section 3.8.4.5.1 of GESSAR and states that the major walls and grade level slabs will be comprised of reinforced concrete approximately two or three feet thick, resulting in a very rigid fix like structure. The substructure up to the top of the slabs will be a Category I structure. A reinforced concrete super-structure is supported on the radwaste substructure but it is a non-category I_ structure; however, the major structural walls and slabs of the superstructure will be designed to resist Category I O seismic loads.
C-131 113078
l GESSAR STANDARD REVIEW PLAN ASSESSMt.?T PDA EXTENSION C.2.4.30 ~ SRP SECTION: 3.3.2 REVISION: 0 DATED: 11/14/75 (Category IV, Item B-18)
TITLE: Tornado Loadings 1
Q BRIEF DESCRIPTION OF SRP SECTION 3.3.2, PARA II.,2.d:
This section of the SRP defines the internal branches of the NRC responsible for the review responsibilities of this SRP section, the applicable review areas relating to the design of structures that must withstand the effects of the design basis tornado (DBT) specified for the plant, the design parameters and procedures relative to structural loads from pressure and missiles, the NRC acceptance criteria and review procedures, and the NRC acceptable evaluation findings statement.
EVALUATION OF 238 GESSAR (BWR-6/ MARK III) WITH RESPECT TO SRP SECTION 3.3.2, PARA II.2.d:
The NRC SRP areas of review and the associated NRC SRP acceptance criteria for the areas of review are adequately covered by the 238 GESSAR BWR-6/ MARK III Os Plant.
The design parameters and procedures for loads applicable to the tornado, l including the tornado wind translational and tangential velocities, the ;
tornado generated pressure differential and its associated time interval, and the spectrum of tornado generated missiles including their characteris-tics are in accordance with this section of the SRP as follows:
- 1. The bases for the selection and values of the parameters are enveloped for the standard plant design in accordance with the SRP Sections 2.3.1, 2.3.2, 3.5.14, and Regulatory Guide 1.76. The enveloping site characteristics were selected after a review of values used in recently licensed plants and using the Region I maximum tornado characteristics from R.G. 1.76. The tornado generated spectrum of missiles is in accordance with SRP Section 3.5.14, Safety Evaluation O Report (sea) NuREG-75/110, page 3-8, Table II, Column C (TVA Horizontal Velocity and Vertical Velocity), and SER NUREG-0124 (Supplement #1' to NUREG-75/110, page 3-2).
The procedures used to transform missile impacts into loads on structures and assess damage to barriers and targets considered local penetration, spalling, scabbing, and overall structural response per SRP 3.5.3. For local damage formulas were used such as the " Modified Petry Formula", "Standord Formula", and " Ballistic Research Laboratories formula" along with full scale missile test data from the EPRI NP-440 Report entitled " Full-Scale Tornado
.O Missile Tests" conducted at Tonapah, Nevada (July 1977). For C-132 113078
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l 4
. GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION SRP SECTION 3.3.2 - Page 2 overall structural response, conservation of momentum methods and other conventional analyses methods are used. Minimum concrete O wall and roof thicknesses are 24 and 21 inches on an interim basis using 3000 psi concrete with a minimum of .2% reinforcing steel EWEF per ACI 349. STRIDE design meets new SRP interim critera except for concrete strength of 4000 psi. However, from quoted excerpts of the " Abstract" of the EPRI-NP-440 report on page V using 3000 psi concrete test panels: " Data from the 18 tests can be used directly for structural design or for validating design and analysis techniques." --- "The results show that a 1500 pound utility pole, 1-inch rod, and 3-inch pipe are ineffective for producing significant local damage even under the improbably severa tornado-missile impact conditions represented by the tests." ----
"Although 12-inch pipes produced craters in the face of the panels, impact tests with these 743 pound missiles showed that 18-inch thick walls are adequate for preventing backface scabbing (secondary missiles) in the highest tornado intensity region of the U.S.,
while 12-inch thick panels are adequate for other regions." ----
O " Contrary to the predictions of conventional structural design methods, which do not account for missile deformation, no overall permanent deflections of the panels were produced by any of the missiles." See the general discussion below entitled " Generic versus Site Unique Design for Site Characteristics."
- 2. The design procedures utilized to transform the tornado parameters into effective loads on structures are in accordance with the SRP Section II.2.a, b, c, and d using the criteria delineated in the American Society of Civil Engineers (ASCE) Paper No. 3269, " Wind Forces on Structures", Trans. of the ASCE, Vol. 126, Part II, 1961.
- 3. The design procedure used to demonstrate that failure of any structure or component not to be designed for tornado loads will not affect the capability of other structures or components to perform necessary O =arety ro"ctio"s eets this sae ee ro11o s:
- a. All " Seismic Category I" systems important to safety are housed in tornado resistant structures,
- b. Those structures not designed for tornado loads that could fail (e.g., cooling towers and stacks) are arranged and located on the site with sufficient distance between them and the
" Seismic Category I" structures. See the discussion below.
entitled " Generic versus Site Unique Design for Site Charac-teristics."
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Q GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION 1 I
SRP SECTION 3.3.2 - Page 3 GENERIC VERSUS SITE UNIQUE DESIGN FOR SITE CHARACTERISTICS:
Where site characteristics / parameters are not enveloped by or fall outside of the design scope of the 238 GESSAR for BWR6/MK III, the O Applicant (0wner) is responsible to present the supporting design criteria and data to substantiate the site unique characteristics / parameters for the NRC SRP review. See the two sections quoted below from the 238 GESSAR for general background information and an understanding of the " Generic Design Concept" versus the " Site Unique Design Concept". !
Quote.
2.0 SITE CHARACTERISTICS - NUCLEAR ISLAND DESIGN ENVELOPE :
I A detailed description of all site characteristics is not practical i for purposes of this report, since this document is not based on a However, it is possible to define an specific site location. j envelope of site related parameters which will blanket the majority of potential reactor sites in the conterminous United States. This envelope of site related parameters establishes the conditions of O'.
phenomena which the generic BWR 6 - Mark III plant (Nuclear Island) is designed to accommodate. These characteristics, which were picked after a review of values used in recently licensed plants, provide the bases for design of the standard plant.
i Variations in chosen site parameters are to be expected. When a specific plant site is selected, a plant unique set of plant design conditions will be established. It is expected that the unique site will have most design conditions lower than and with a limited number higher than the GESSAR envelope. Confirming calculations and analyses will be made with the site unique conditions for tb2 GESSAR buildings using the loading combinations and allowable stresses given in Section 3.8. For the total loading condition on any structure or system, it is anticipated that the GESSAR design will be adequate If the unique site conditions indicate that areas O ef the GESSAR desion ere 4aedeavete, e s4te caiave eesion will he provided. The meteorological assumptions used in the accident analyses are defined in Chapters 2 and 15. If the site unique meteorological conditions vary from the GESSAR assumptions, the accident analyses will be repeated to determine new offsite doses. i For these cases where the doses exceed allowable limits, design modifications and new analyses will be performed to provide a design that meets licensing requirements.
3.5.2.1.2 Environmental Load Generated Missiles. Environmental loads can result from basically four sources: seismic activity;
- high water level; winds; and tornados. Since all seismic Category I C-134 113078 l
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I C GESSAR STANDARD REVIEW PLAN ASSESSMENT l
PDA EXTENSION SRP SECTION 3.3.2 - Page 4 structures and components will be designed to resist seismic activity, i there will not be seismic Category'I structure related missiles as
- c. a result of seismic activity. The same is true of high water level d since all seismic Category I structures will be designed to withstand the design water level for the plant. Of the two remaining environmen-j tal loads, only tornados need be considered since the maximum absolute velocity of the design tornado is' greater than three times that of the maximum wind velocities (Reference is made to Subsec- ;
tions 3.3.2 and 3.3.1, respectively). In other words, potential l missiles resulting from a tornado would be a good deal more energetic than that of the wind.
Seismic Category I structures are designed to resist the design tornado; hence, missiles resulting.from effects of tornadoes on such structures are not a consideration. As discussed in Subsec-tion 3.3.2.3, tornado winds could damage non-Category I structures i thereby producing missiles of a sort. However, integrity of the l containment, capability of the essential heat-removal systems are l not impaired nor is there a risk that offsite exposures would 4 O exceed the 10CFR100 guidelines since such non-Category I structures J
(e.g., cooling towers and stacks) are' arranged with sufficient distance between them and the seismic Category I structures.
There will be in this plant, as in past plants, various objects within and outside the site boundary that could become missiles as a result of the tornado generated winds. The whirlwind characteristics of a tornado make possible the imparting of high velocities to objects that are potential missiles (i.e., already detached),
especially where the objects have high area-to-weight ratios. The area-to-weight ratio factor is important in determining the time duration m ,,hio che object is sustained in the wind field by the strong updrafts prior to becoming airborne. There are a number of objects that can become missiles even though their weight is too great to permit the tornado winds to sustain them for any appreciable O length of time.
For instance, under certain circumstances the aerodynamics of an automobile will cause it to be lifted temporarily; however, its heavy weight and the random orientation of surfaces after once lifted and tumbled cause it to drop quickly, thereafter.
Missile type characteristics and parameters for the above objects are described in Subsection 3.5.3.1 Unquote.
C-135 113078
GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION O C . 2.4. 31 SRP SECTION: 3.4.2 REVISION: 0 DATED: 11/24/75 CATEGORY IV, ITEM B-19 TITLE: Dynamic Effects of Wave Action (SRP Section 3.4.2, Par. II)
BRIEF DESCRIPTION OF SRP 3.4.2, PARAGRAPH II:
For plants where the flood level is higher than the proposed grade around the plant structures, the dynamic phenomena associated with O soch a fioodiao. such es c"rreats, wind weves end their nydrolooicei effects, are to be considered.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO STANDARD REVIEW PLAN:
Section 3.4.1 of GESSAR states that the design basis flood elevation is approximately one foot below the plant finished grade including allowance for coincident waves and the resultant runnings. Since the flood elevation is one foot below the finished plant grade, there will be no dynamic forces due to the flood.
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Q STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION C.2.4.32 SRP SECTION: 10.4.7 DATED: 10/01/75 CATEGORY IV ITEM B-20 TITLE: Water Hammer for Steam Generators with Preheaters O <sar sectioa 10.4.7. per. 1.2.o)
This item is applicable to the steam generator for a PWR and therefore was not assessed against the GESSAR design.
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GESSAR STANDARD REVIEW PLAN ASSESSMENT f]
PDA EXTENSION 1
C.2.4.33 SRP SECTION: 4.4 REVISION: 0 DATED: 1.1/24/75 CATEGORY IV, ITEM B-21 TITLE: Thermal Hydraulic Stability (SRP Section 4.4, par. II.5) o k) BRIEF DESCRIPTION OF SRP 4.4, par. II-5:
This standard review plan defines typical methods to demonstrate that the reactor is free of undamped oscillations or other hydraulic i instabilities for all conditions of steady-state operation, for all '
operational transients, for all load following maneuvers, and for partial loop operation.
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EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO SRP 4.4, PAR. II.5:
GESSAR is in compliance with the requirements of this Standard i Review Plan. The mathematical model representing the core examines !
the linearized reactivity response of a reactor system with density- l
~3 dependent reactivity feedback caused by boiling. In addition, j (V channels, or regions, are examined separately on an axially multi- !
noded basis by grouping various channels that are thermodynamically and hydraulically similar. This interchannel hydrodynamic interaction, or coupling, exists through pressure variations in the inlet plenum, such as can be caused by disturbances in the flow distribution between regions or channels. This approach provides a reasonably j accurate, three-dimensional representation of the reactor's hydro- '
dynamics.
The core model (References 1-6) solves the dynamic equations that represent the reactor core in "e frequency domain. From the solu-tion of these dynamic equations, the reactivity and individual channel hydrodynamic stability of the boiling water reactor is determined for a given reactor flow rate, power distribution, and total power. This gives the most basic understanding of the inherent bq core behavior (and hence the system behavior) and is the principal consideration in evaluating the stable performance of the reactor.
As new experimental or reactor operating data are obtained, the model is refined to improve its capability and accuracy.
The plant model considers the entire reactor system, neutronics, heat transfer, hydraulics, and the basic processes, as well as associated control systems such as flow controller, pressure regulator, feedwater controller, etc. Although, the control systems may be stable when analyzed individually, final control system settings o must be made in conjunction with the operating reactor so that the O
C-138 113078
Page 2 SRP 4.4 O- ITEM 8-21 entire system is stable. The plant model yields results that are essentially equivalent to those achieved with the core model and allows the addition of the controllers, which have adjustable features permitting the attainment.of the desired performance.
The plant model solves the dynamic equations that-present the BWR ,
system in the time domain. The variables, such as steam flow and i O pressure, are represented as . function of time.
of this model is discussed in Reference 7.
The extensiveness l The results show the analytical methods to be an effective and useful design too1, with significant conservatism in its application to boiling water, reactor core evaluation. Neal and Zivi (Reference 8) l further confirm the effective application of essentially the same model to channel and core analysis. Reference 9 discusses plant ;
application of total plant stability, reactor core performance and I channel hydrodynamic performance and in additon gives the analytical '
development of stability theory (Appendix A). ,
l STABILITY ANALYSIS FOR SINGLE RECIRCULATION LOOP OPERATION l 1
The least stable power / flow condition attainable under normal O conditions occurs at natural circulation with the control rods set for rated power and flow. This condition may be reached following i
l the trip of both recirculation pumps. Operation along the minimum forced recirculation line with one pump running at minimum speed is i more stable than operating with natural circulation flow only, but is less stable than operating with both pumps operating at minimum speed. The core stability along the forced circulation, rated rod pattern line for single loop operation is the same as that for both loops operable except that rated power is not attainable. Hence, l the core is limited to maximum power for single pump operation and only manual flow control should be used.
O C-139 113078 s . . _ . , _ _ __ ._ - - - -.-
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Page.3 SRP 4.4
]- ITEM B-21 REFERENCES
- 1. ' KAPL-2170 Hydrodynamic Stability of a Boiling Channel, by A.B. Jones; 2 October 1961.
- 2. KAPL--2208 Hydrodynamic Stability of a Boiling Channel, Part 2, by A.B. Jones; 20 April 1962.
O 3. KAPL-2290 Hydrodynamic Stability of a Boiling Channel Part 3,'by A.B. Jones and D.G. Dight; 28 June 1963.
- 4. KAPL-3070 Hydrodynamic Stability of a Boiling Channel, Part 4, by A.B. Jones; 18 August 1964.
- 5. KAPL-3072 Reactivity Stability of a Boiling Reactor, Part 1, by A.B. Jones and W.M. Yarbrough; 14 September 1964.
- 6. KAPL-3093 Reactivity Stability of a Boiling Reactor, Part 2, by A.B. Jones; 1 March.1965.
- 7. " Analytical Mothods of Plant Transient Evaluations for General Electric Boiling Water Reactor", General Electric Company, BWR Systems Department, February 1973, (NED0-10802).
- 8. Neal, L.G. , and Zivi, S.M. , "The Stability of Boiling Water Reactors and Loops", Nuclear Science and Engineering, 30 P. 25, 1967. l
- 9. Woffinden, F.B., et al, " Stability and Dynamic Performance of the General Electric Boiling Water Reactor", January 1977 (NED0-21506).
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C-140 113078 f
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v GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION l
C.2.4.34'SRP SECTION: 5.2.5 REVISION: 0 DATED: 11/24/75 (Category IV, Item B-22)
TITLE: Intersystem Leakage O BRIEF DESCRIPTION OF SRP SECTION 5.2.5, PARA. II.4:
Provisions should be made.to monitor systems connected to the RCPB
, for signs of intersystem leakage. Detection methods include radio-activity monitoring and indicators to show abnormal water levels or flow in the potentially affected systems and unaccountable increases in reactor coolant make-up flow.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO SRP 5.2.5, PARA II.4:
Intersystem leakage can occur either across closed valves or through heat exchanger tube leaks.
Intersystem leakage across closed isolation valves at the reactor o coolant pressure boundary is detected by pressure and temperature V sensors. Leakage to the standby liquid control system is indicated by pressure and tank level sensors. Leakage to the reactor water cleanup system and past the main steam isolation valves will be l detected by pressure sensors and by temperature and flow indication '
sensors, respectively.
The Essential Service Water System (ESWS) is the only fluid system which normally communicates with the environment outside the plant.
The RHR system and the CCW system are cooled by the ESWS. Any leakage from these systems to the ESWS would be detected by the radiation detectors in the ESWS.
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O GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSON l
C.2.4.35 SRP SECTION: 3.2.2 REVISION: 0 DATED: 11/24/75 CATEGOPY IV, ITEM B-23 ;
l TITLE: Main Steam Isolation Valve Leakage Control System (SRP Section 3.2.2 BTP RSB-3.2)
O BRIEF EXPLANATION OF THE SRP 3.2.2 - BTP-RSB-3.2 In the course of GESSAR review, the NRC have identified a sys-tematic basis for classification of such components that are (a) not classified as safety-related items but are located downstream of the isolation valves, (b) not specifically designed to seismic Category I standards, and (c) not housed in seismic Category I structures. Therefore mainsteam and feedwater system components of BWR/6 plants should be classified in accordance with BTP-RSB No. 3-1 or alternately, in accordance with Table 3-2.1.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO SRP 3.22:
The shutoff valves and the leakage control system equipment which j n
'O are interfaced with the main steam lines comply with SRP 3.2.2 and BTP 3.2-1 of GESSAR page 3.2-19.
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U C-142 113078
.GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION O C.2.4.36 SRP SECTION: 3.5.3 REVISION: 0 DATED: 12/1/76 )
CATEGORY IV, ITEM C-1 TITLE: . Ductility of Reinforced Concrete and Steel Structural Elements Subjected to Impactive or Impulsive Loads NRC GUIDELINES'PROVIDED IN ENCLOSURE E STATEMENT OF NRC LETTER, R. S. BOYD TO G. G. SHERWOOD, DATED OCTOBER 13, 1978,
SUBJECT:
EXTENSION REVIEW MATTER FOR PDA's:
O IstaooUcTIon In the evaluation of overall response of reinforced concrete structural
- elements (e.g., missile barriers, columns, slabs, etc.) subjected to impactive or impulsive loads, such as impacts due to missiles, assumption 3 on non-linear response (i.e.,-ductility ratios greater than unity) of the structural elements is generally acceptable provided that the safety functions of the structural elements and those of safety-related systems and components supported or protected by the elements are maintained.
The following summarizes specific SEB interim positions for review and acceptance of ductility ratios for reinforced concrete and steel structural elements subjected to impactive and impulsive loads. j SPECIFIC POSITIONS
- 1. REINFORCED CONCRETE MEMBERS 1.1 For beams, slabs, and walls where flexure control designs, the permissible ductility ratio ( p ) under impactive and impulsive loads should be taken as p= g5p . for p p ' >_ .005 p = 10 for p p ' < .005 where p and p' are the ratios of tensile and compressive reinforcing as defined in ACI-318-71 Code.
1.2 If use of a ductility ratio greater than 10 i.e., p>100 is required to demonstrate design adequacy of structural elements agains impactive O or. impulsive loads, e.g., missile impact, such a usage should be identified in the plant SAR. Information justifying the use of '
this relatively high ductility value shall be provided for SEB staff review.
I
- 1. 3 For beam columns, walls, and slabs carrying axial compression loads and subject to impulsive or impactive loads producing flexure, the ;
permissible ductility ratio in' flexure should be as follows:
(a) When compression controls the design, as defined by an interaction disgram, the permissible ductility ratio shall be 1.3.
1 C-143 113078
1 GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION i O tte= c-1 Page 2 (b) ,When-the compression loads'does not exceed 0.lfc Ag or one-third of that which would produce balanced conditions, whichever is smaller, the permissible ductility ratio can be as given in Section 1.1. j (c) The permissible ductility ratio shall vary linearly from 1.3 to that given in Section 1.1 for conditions between those O- spec 4ried 4a ce) eno ch). csee F49ure 1. )
1.4 For structural elements resisting axial compressive impulsive or impactive loads only, without flexure, the permissible axial ductility ratio shall be 1.3.
- 1. 5 For shear carried by concrete only p = 1.0 '
For shear carried by concrete and stirrups or bent bars y = 1.3 '
For shear carried entirely by stirrups O p = 3.0 2.0 STRUCTURAL STEELMEMBERS 2.1 Foglexurecompressionandshear p = 10.0
- 2. 2 For columns with slenderness ratio (1/r) equal or less than 20 p = 1.3 where 1 = effective length of the number r = the least radius of gyration O For columes with sienderness ret 4o oreeter then 20 p = 1.0 2.3 For members subjected to tension p=.5 E EY where cp = uniform ultimate strain of the material ,
cy = strain at yield of material O .
C-144 113078 ,
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Item C-Page 3
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- 0 2&
w 063 Yy=b -
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Figure 1. Proposed Ductility Ratio for Beam Columns ,
GESSAR STANDARD. REVIEW PLAN ASSESSMENT i
PDA EXTENSION !
O ' Item C-1 Page 4 EVALVATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REFERENCE SRP STATEMENT:
The requirement on ductility ratios is really a statement as to i acceptable levels for review and acceptance (of ductility ratios) for reinforced concrete and steel structural elements subjected to impactive and impulsive loads, not general dynamic loads.
O The values given in the SEB position are relatively low values originally chosen as typical of what ordinary structures may be capable of without special design considerations.
The GESSAR design stress limits are used rather than ductility ratio limits. The stress limits are considered to provide adequate protection.
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O C-146 113078
I Q GE35AR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION .
C.2.4.37 SRP SECTION: 3.7.1 REVISION: N/A DATED: August 1, 1976 (Category IV Item C-2) 1 TITLE: Response Spectra in Vertical Direction NRC GUIDELINES PROVIDED IN ENCLOSURE E STATEMENT OF NRC LETTER, i R. S. B0YD TO G. G. SHERWOOD, DATED OCTOBER 13, 1978, SUBJECT EXTENSION REVIEW MATTER FOR PDA's:
IV.C.2 Response Spectra in the Vertical Direction
~
(3.7.1)
Subsequent to the issuance of Regulatory Guide 1.60, the report " Statistical Studies of Vertical and Horizontal Earthquake Spectra" was issued in January 1976 by NRC as NUREG-0003. One of the important conclusions of this report is that the response spectrum for vertical motion can be taken as 2/3 the response spectrum for horizontal motion over the entire range of frequencies in the Western O United States. According to Regulatory Guide 1.60, the vertical response spectrum is equal to the horizontal response spectrum between 3.5 cps and 33 cps. For the Western United States only, consistent with the latest available data in NUREG-0003, the option of taking the vertical design response spectrum as 2/3 the horizontal response spectrum over the entire range of frequencies will be accepted. For other locations, the vertical response spectrum will be the same as that given in Regulatory Guide 1.60.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REFERENCE SRP STATEMENT:
O The vertical response spectra used in GESSAR is consistent with Reg. Guide 1.60 rather than NUREG-0003 because of GESSAR's applica-bility even outside the Western United States.
O C-147
GESSAR STANDARD REVIEW PLAN' ASSESSMENT PDA EXTENSION j O I C.2.4.38 SRP SECTION: 3.8.1 REVISION: N/A DATED: 4/1/76 3.8.2 CATEGORY IV, ITEM C-3 TITLE: BWR Mark III Containment Pool Dynamics NRC GUIDELINES PROVIDED IN ENCLOSURE E 0F NRC LETTER, R. S. B0YD TO G. G. SHERWOOD, DATED. OCTOBER 13, 1978,
SUBJECT:
EXTENSION REVIEW -
MATTER FOR PDA's: '
O 1. P00L SWELL-
- a. Bubble pressure, bulk swell and froti swell loads, drag pressure and other pool swell loads should be treated as abnormal pressure loads, Pa. Appropriate load combinations and load factors should be applied accordingly.
- b. The pool swell loads and accident pressure may be combined in accordance with their actual time histories of occurrence. i
- 2. SAFETY RELIEF VALVE (SRV) DISCHARGE
- a. The SRV loads should be treated as live loads in all load combinations 1.5P totheappropriat$wherealoadfactorof1.25shouldbeapplied SRV loads. I
- b. A single active failure causing one SRV discharge must be considered in combination with the Design Basis Accident (DBA).
- c. Appropriate multiple SRV discharge should be considered in combination with the Small Break Accident (SBA) and Inter-mediate Break Accident (IBA).
- d. Thermal loads due to SRV discharge should be treated as T for normal operation and aT f r accident conditions.
- e. The suppression pool liner should be designed in accordance with the ASME Boiler and Pressure Vessel Code, Division 1 Subsection NE to resist the SRV negative pressure, considering O strength, buckling and low cycle fatigue.
The pool swell and safety relief valve loadings are described in Appendix 3B entitled, " Load Definition and Structural Capability of Mark III Containment." The design capability of the Mark III to withstand pool swell and safety relief valve discharge are discussed below.
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GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION e-Q . Item C-3 Page 2 EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REFERENCE SRP i STATEMENT; i
! 1. Pool Swell The pool swell loadings are treated as abnormal pressure 4
loads and combinations are provided in Sections 4 through
, 10 of Appendix 38. .The pool swell loads and postulated 1
O accidemt press #re are co biaed for the coateia eat str ct"re in accordance with the sequencing shown in Appendix 38, figure 6.1. Figure 6.1 is the bar chart showing the
- loading conditions that the containment structure may .
- experience during the postulated design basis LOCA. The 3
GESSAR design has been demonstrated to satisfy the staff positions regarding pool swell.
j 2. Safety Relief Valve (SRV) Discharge
- a. The SRV loads are treated as dynamic loads with
- appropriate pressure time histories. No load factor 2
multiples are utilized since extremely conservative methods are used. Analysis indicates that for a 95-95 percent confidence limit, approximately one percent of the number of safety / relief valve actuations O may result in containment loads above-the design valve. Such a low probability that SRV's may exceed their specified limit is acceptable considering the conservatism of the method of prediction. The actual expected loads will not exceed the design value,
- b. This position is complied within the GESSAR design i and is discussed in Section 2.4 of Appendix 38. The i
opening of a single valve is not a direct result of
' the LOCA and is not an expected occurrence during the accident sequence. However, load chart figures
, show a single S/R valve actuation as an additional load to demonstrate additional capability (Ref.
Figure 6.1 of ICCR).
O c. Appropriate muitinle SRv discheroes ere considered with the IBA condition. One S/R valve actuation is considered with the SBA and it is assumed that the
, operator action shuts down the reactor via normal procedures (Ref. Section 2.3 of Appendix 3).
- d. This thermal load condition is being considered.
Final determination to be made at FSAR stage,
- e. The GESSAR design complies with this position.
O.
C-149 113078
Q GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION C.2.4.39 SRP SECTION: 3.8.4 REVISION: 0 DATED: 9/1/76 CATEGORY IV, ITEM C-4 TITLE: Air Blast Loads
' O NRC GUIDELINES PROVIDED IN ENCLOSURE E, STATEMENT OF NRC LETTER, R. S. B0YD TO G. G. SHERWOOD, DATED OCTOBER 13, 1978,
SUBJECT:
EXTENSION REVIEW MATTER FOR PDA's.
AIR BLAST LOADS (Pa, Ta, To as defined in ACI 359-740)
The following interim position on air blast loadings on Nuclear Power Plant Structures should be used as guidance in evaluating analyses.
- 1. An equivalent static pressure may be used for structural analysis purposes. The equivalent static pressure should be obtained from the air blast reflected pressure or the overpressure by multiplying these pressures by a factor of two. Any proposed use of a dynamic load factor less than two should be treated on a case by case e
basis. Whether the reflected pressure or the overpressure is to be used for individual structural elements depends on whether an O incident blast wave could strike the surface of the element.
- 2. No load factor need be specified for the air blast loads, and the load combination should be:
U=0+L+B where, U is the strength capacity of a section 0 is dead load L is live load B is air blast load.
- 3. Elastic analysis for air blast is required for concrete structures of new plants. For steel structural elements, and also for reinforced concrete elements in existing plants, some inelastic response may O be ner 4ttee witn ennropriete ii its oa ovcti14ty retioe.
- 4. Air blast generated ground shock and air blast wind pressure may be ignored. Air blast generated missiles may be important in situations where explosions are postulated to occur in vessels which may fragment.
- 5. Overturning and sliding stability should be assessed by multiplying the structure's full projected area by the equivalent static pressure and assuming only the blast side of the structure is loaded.
Justification for reducing the average equipment static pressure on O. curved surfaces should be considered on a case by case basis.
C-150 113078
$ 4 Q GESSAR STANDARD REVIEW PLAN ASSESSMENT ,
PDA EXTENSION .
I
- i SRP 3.8.4 Item, C-4 4
- 6. Internal supporting structures must also be analyzed for-the effects 4
O or eir est to deter iae their ebiiity to cerry 'oeds eP9 ied directly to exterior panels and slabs. Moreover, in vented structures, interior structures may require analysis even if they do not support exterior structures.
f
- 7. The equivalent static pressure should be considered as potentially acting both inward and outward.
r EVALUATION OF GESSAR (BWR-6/MK-III) WITH RESPECT TO REFERENCE SRP
! STATEMENT:
The air blast load is only applicable to sites that are in the vicinity of potentially explosive areas. The assignment of this criteria is the same as was stated for Regulatory Guide 1.91, i
Section C.2.2.4 of this Appendix. !
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C-151 113078 t
GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION C.2.4.40 SRP SECTION: 3.5.3 REVISION: 0 DATED: 11/24/75 (Category IV, Item C.5)
TITLE: Barrier Design Procedures 1
1 O. NRC GUIDELINES PROVIDED IN ENCLOSURE E STATEMENT OF NRC LETTER, R. S.
O B0YD TO G. G. SHERWOOD, DATED OCTOBER 13, 1978,
SUBJECT:
EXTENSION REVIEW MATTER FOR PDA's:
l IV.C.5 TORNADO MISSILE PROTECTION i As an interim measure the minimum concrete wall and roof thickness for tornado missile protection will be as follows:
Concrete Strength Wall Thickness Roof Thickness (psi) (inches) (inches)
Region I 3000 27 24 i 4000 24 21 1 5000 21 18 Region II 3000 24 21 4000 21 18 5000 19 16 Region III 3000 21 18 4000 18 16 5000 16 14 These thicknesses are for protection against local effects only.
Designers must establish independently the thickness requirements for overall structural response. Reinforcing steel should satisfy the provisions of Appendix C, ACI 349 (that is, .2% minimum, EWEF).
The regions are described in Regulatory Guide 1.76.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REFERENCE SRP STATEMENT:
The answers to the above concerns are as follows:
The procedures used to transform missile impacts into loads on structures and assess damage to barriers and targets considered local penetration, spalling, scabbing, and overall structural response per SRP 3.5.3. For local damage, formulas were used such n as the " Modified Petry Formula", " Standard Formula", and " Ballistic L)
C-152 113076
- - =_ - - ._ ,_ . -
Q GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION SRP 'SECTION: 3.5.3 - Page 2 Research Laboratories Formula" along with full scale missile test data from the EPRI NP-440 Report entitled " Full-Scale Tornado Missile Tests" conducted at Tonapah, Nevada (July 1977). For O overall structural response, conservation of momentum methods and other conventional analyses methods are used. Minimum concrete wall and roof thicknesses are 24 and 21 inches on an interim basis using 3000 psi concrete with a minimum of .2% reinforcing steel EWEF per ACI 349. STRIDE design meets new SRP interim criteria except for concrete strength of 4000 psi.
However, from quoted excerpts of the " Abstract" of the EPRI NP-440 report on Page V using 3000 psi concrete test panels: " Data from the 18 tests can be used directly for structural design or for validating design and analysis techniques." --- "The results show that a 1500 pound utility pole, 1-inch rod, and 3-inch pipe are ineffective for producing significant local damage even under the improbably severe tornado-missile inspect conditions represented by the tests." --- "Although 12-inch pipes produced craters in the face of the panels of impact tests with these 743 pound missiles showed that 18-inch thick walls are adequate for preventing back-face scabbing (secondary missiles) in the highest tornado' intensity region of the U.S., while 12-inch thick panels are adequate for other regions." --- " Contrary to the predictions of conventional structural design methods, which do not account for missile defor-mation, no overall permanent deflections of the panels were produced by any of the missiles."
Based on the above quotation and GESSAR, it can be stated that the GESSAR design meets SRP 3.5.3 and its supplement.
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C-153 113078
GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION C.2.4.41' SRP SECTION: 6.3 REVISION: N/A DATED: 10/13/78 e (Category IV, Item C-6)
TITLE: Passive ECCS Failures During Long-Term Cooling Following a LOCA O NRC GUIDELINES PROVIDED IN ENCLOSURE E STATEMENT OF NRC LETTER, R. S. BOYD TO G. G. SHERWOOD, DATED OCTOBER 13, 1978,
SUBJECT:
EXTENSION REVIEW MATTERS FOR PDA's:
Passive failures in the ECCS, having leak rates equal to or less than those from the sudden failure of a pump seal and which may occur during the long-term cooling period following a postulated LOCA, should be considered. To mitigate the effects of such leaks, a leak detection system having design features and bases as described below should be included in the plant design.
The leak detection system should include detectors and alarms which would alert the operator of passive ECCS leaks in sufficient time ;
so that appropriate diagnostic and corrective actions may be taken '
p on a timely basis. The diagnostic'and corrective actions would include the identification and isolation of the faulted ECCS line 1 before the performance of more than one subsystem is degraded. The i design bases of the leak detection system should include:
(1) Identification and justification of the maximum leak rate; (2) Maximum allowable time for operator action and justification therefore; (3) Demonstration that the leak detection system is sensitive enough to initiate an alarm on a timely basis, i.e., with sufficient lead time to allow the operator to identify and isolate the faulted line before the leak can create undesirable consequences such as flooding of redundant equipment. The minimum time to be considered is 30 minutes.
O (4) Demonstration that the leak detection system can identify the J
1 faulted ECCS train and that the leak can be isolated; and (5) Alarms that conform with the criteria specified for the control room alarms and a leak detection system that conforms with the requirements of IEEE-279, except _that the single failure criterion need not be imposed.
O C-154 113078.
I
-Q GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION SRP SECTION: 6.3 Page 2 EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO THE REFERENCE SRP STATEMENT:
O Each ECCS equipment room is provided with a sump equipped with a Class "IE", seismic Category I Level Instrument which will alarm in the main control room on sump high level.
Upon receipt of the sump high level alarm, the control room operator has sufficient time to take the necessary action to identify and isolate the source of leakage before the leak can cause unacceptable consequences.
The equipment provided meets the requirements of this SRP. Leakage rate limits if required will be included in the technical specifications as part of the FSAR.
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O-C-155 113078
-l l
Q GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSON C.2.4.42 SRP SECTION: 6.3 REVISION: 2 DATED: 9/1/77 CATEGORY IV, ITEM C-7 l TITLE: Control Room Position Indicatfore of Manual (Handwheel) Valves NRC GUIDELINES PROVIDED IN ENCLOSURE E STATEMENT OF NRC LETTER, R. S. ;
BOYD TO G. G. SHERWOOD, DATED OCTOBER 13, 1978,
SUBJECT:
EXTENSION O aeview MATrea roa eo^'s:
1 Regulatory Guide 1.47 requires automatic position indication of l each bypass or deliberately induced operable condition if the l following three conditions are met: ;
(1) The bypass or inoperable condition affects a system that is designed to perform an automatic safety function.
(2) The bypass or inoperable condition can reasonably be expected to occur more frequently than once per year.
(3) The bypass or inoperable condition is expected to occur when the system is normally required to operate.
Revision one of the Standard Review Plan in Section 6.3 requires O- conformance with Regulatory Guide 1.47 with the intent being that any manual (handwheel) valve which could jeopardize the operation of the ECCS, if inadvertently left in the wrong position, must have position indication in the control room. In the PDA extension reviews it is important to confirm that standard designs include this design feature. Most do but it was probably not specifically addressed in some of the first PDA reviews.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REFERENCE SRP STATEMENT: ;
The ECCS, RCIC, and Essential Service Water System comply with the j intent of the above NRC position. The designs of these systems are considered acceptable for the following reasons:
(1)' Position indication is provided in the control room for manual (handwheel actuated) valves in the systems' main flow paths which cannot be verified by administrative procedures as being in the correct position during normal plant power generation and which could jeopardize the systems' safety function if inadvertently left in the wrong position.
(2) Incorrect position of a manual valve in the systems' main flow path that is not' provided with control room position indication is detectable during routine systems' flow testing during normal plant power generation and by normal administrative rontrol procedures.
C-156 113076
V n
U Item C-7 Page 2 (3) The changing of the position of a manual valve in the systems' main flow' path is only required for maintenance purposes and is expected to occur on an average of no more than once per five years. This is much lower than the bypass frequency of "more than once per year" that Regulatory Guide 1.47 established as the criteria to be used to determine when automatic bypass indication is required.
(4) Regulatory Guide 1.47 states that automatic indication of an O
inoperable condition is required if "the bypass or inoperable condition is expected to occur when the system is normally required f to operate." The closure of manual valves is expected to occur only when the system has been taken out of service for maintenance and hence is not required to operate (occurs approximately once in i five years). Therefore, Regulatory Guide 1.47 does not require I position indication on ECCS or RCIC Systems main flow path manual valves.
(5) The position of manual valves in lines connected to the main flow l path which could degrade the system safety function if inadvertently left in the wrong position are normally detectable by local area flooding or below normal system pressure indication or excess discharge to the radwaste system, or low suppression pool water level alarm or excess leakage rate in the drywell and normal admini-O strative control.
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C-157 113078 i
GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION C.2.4.43 SRP SECTION: 15.1.5 REVISION: 0 DATED: 4/01/77 CATEGORY IV, ITEM C-8 TITLE: ' Long-Term Recovery from Steamline Break: Operator Action to Prevent Overpressurization This is a PWR item and is not applicable to the GESSAR design.
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O C-158 113078
h GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION C.2.4.44 SRP SECTIONS: 5.4.6 REVISION: DATED: 12/1/77 S.4.7 CATEGORY IV, ITEM C-9
6.3 TITLE
Pump Operability Requirements O NRC GUIDELINES PROVIDED IN ENCLOSURE E STATEMENT OF NRC LETTER, R. S. BOYD TO G. G. SHERWOOD, DATED OCTOBER 13, 1978,
SUBJECT:
EXTENSION REVIEW MATTER FOR PDA's:
Describe the tests performed to demonstrate that the pumps are l capable of operating for extended periods under post-LOCA condi- <
tions including debris. Discuss the damage to pump seals caused by debris over an extended period of operation.
Provide detailed diagrams of all water cooled seals and components in the pumps.
Provide a description of the composition of the pump shaft seals J and the shafts. Provide an evaluation of loss of shaft seals. I Discuss how debris and post-LOCA environmental conditions were factored into pump design specifications and design.
1 EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REFERENCE SRP STATEMENT:
Suction strainers are provided in the suppression pool on suction lines that provide water to ECCS pumps. The suction strainers are designed so that particles passed through them are too small to block or plug the system critical flow paths such as system pump seal flushing / cooling water supply lines / components.
The system main process water is used for pump seal flushing /-
cooling. The process water is provided from the pump discharge O nozzle and passed through a cyclone separator before being dis-charged to the pump seal. The cyclone separator is a hydraulic centrifuge device whicn is capable of removing fine particulate debris ~ from the seal flushing / cooling water that could damage the pump seal. The cyclone separator is a continuous self-cleaning device which insures that it will function properly during long term post-LOCA' operation.
Mechanical type pump seals typically consisting of a carbon base material operating against a hard metallic mating face are used in all ECCS pumps. The high reliability of these types of seals has pd C-159 113078
, Item C-9' -
Page 2 resu'lted in broad commercial application. These types of seals L have an average operating life of five years which is substantially greater than the postulated 100 days of post-LOCA ECCS pump operation.
Additionally, post-LOCA steam venting may suspend particulate material in the. suppression pool water due to violent suppression pool circulation. But in a short period of time after the LOCA,
'O f ows wil be more orderly and most of the potentially damaging particulate debris will have settled out of the suppression pool 1 water. Therefore, it is expected that the main process water would contain potentially damaging debris for a minimal time, post-LOCA.
The pump mechanical seal is an active component. Since only one single active failure needs to be considered following a LOCA and failure of the pump shaft seal has substantially less impact on post-LOCA conditions than other failed events, such as failure of a diesel, the consequences of seal failure are considered acceptable.
If a gross seal failure did occur, the system with the failed seal could be isolated by closing suction and discharge line valves.
Due to the dependability of the cyclone separator and hydraulic centrifuge and the fact that these are standard devices, General Electric does not see the need for a test to demonstrate that the
( pumps are capable of operating for extended periods under post-LOCA conditions including debris.
Detailed diagrams of water cooled seals and components in the pump are not included in GESSAR because each purchaser may have a different pump design and therefore a detailed pump design would be out of the GESSAR scope.
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C-160 113078
'GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION O C.2.4.45 SRP SECTION: 3.5.1 REVISION: 0 DATED: 3/28/78 CATEGORY IV, ITEM C-10 '
~
TITLE OF ASSESSMENT ISSUE: Gravity Missiles, Vessel Seal Ring Missiles Inside Containment NRC GUIDELINES PROVIDED IN ENCLOSURE E, STATEMENT OF NRC LETTER, R.' S. B0YD TO G. G. SHERWOOD, DATED OCTOBER 13, 1978,
SUBJECT:
EXTENSION REVIEW MATTER FOR PDA's:
O Safety related systems should be protected against loss of function <
due to internal missile impact. Pressurized components and rotating machinery are potential internal missile sources. These include retaining bolts, control rod drive assemblies, the vessel seal ring, valve bonnets, and valve stems. Metastably supported equipment could also fall upon impingement. Protection against such potential i missiles includes preferential orientation of potential missile sources, missile barriers, and physical separation of redundant safety systems and components.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REFERENCE SRP STATEMENT:
Section 3.5 of GESSAR identifies the potential missiles which could O be generated inside containment. GESSAR also includes discussions on missile characteristics and barrier design criteria.
The program to define potential drywell missiles and their effect on potential targets is currently being conducted for the Mark III, 238 Standard Plant drywell design. The effects of potential retaining
' bolts missiles have been considered. However, a bolt failure does ,
not result in a postulated break in the system where the failure occurred since there is a redundancy of bolts in pressure retaining components to effect the seal. A target impacted by a bolt missile is not expected to result in a postulated break exceeding the DBA.
Postulated control rod drive (CRD) assembly missiles will be stopped by the CR0 support structure. Since the CRD housings are located within the reactor pedestal, their postulated missiles are prevented from entering the drywell area.
Gravity missiles originating from the vessel seal ring have not been evaluated for the Mark IV Standard Plant drywell, because GESSAR (Section 3.5) does not list the vessel seal as a potential missile source. During plant operation the vessel seal is not pressurized, and the drywell head cavity is dry, therefore the seal structure carries its own weight only. A missile size and weight has to be. defined prior to any evaluation of potential gravity missiles.
O C-161 113078
GESSAR STANDARD REVIEW PLAN ASSESSMENT Item C-10 Page 2 PDA EXTENSION
(:)
Piping thermowells have been considered as postulated missiles.
For example, there are two piping thermowells on each of the suction ,
legs.of the recirculation systems. The thermowells are located so as to prevent unacceptable damage due to a postulated thermowell :
missile. One thermowell is oriented toward the weir wall, and will not impact any essential targets. The second thermowell is expected 3 to. impact the recirc pump motor, and thus be prevented from being
() ejected into the higher drywell elevation. The recirculation pump and motor are not classified as essential equipment, and the thermowell missile will not result in escalating the size and effects of the original break size at the exit location of the thermowell missile.
2 Design and manufacturing requirements, and ISI procedures are
- considered when determining if valve stem and bonnets will become missiles.
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C-162 113078
GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION.
i O C.2.4.46 SRP SECTION: 4.4 REVISION: 0 DATED: 1/01/77 CATEGORY IV, ITEM C-11 TITLE: Core Thermal Hydraulic Analysis NRC GUIDELINES PROVIDED IN ENCLOSURE E STATEMENT OF NRC LETTER, R. S. BOYD TO G. G. SHERWOOD, DATED OCTOBER 13, 1978,
SUBJECT:
EXTENSION REVIEW MATTER FOR PDA's:
O Ia eveiuetiaa the thermei-hydreuiic Performeace of the reector core the following additional areas should be addressed:
- 1. The effect of radial pressure gradients at the exit of open lattice cores. l
- 2. The effect or radial pressure gradients in the upper plenum.
l
- 3. The effect of fuel rod bowing.
In addition, a commitment to perform tests to verify the transient 1 analysis methods and codes is required. ,
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REFERENCE SRP STATEMENT:
- 1. The problem of radial pressure gradients at the exit of open lattice cores is not relevant to the GE BWR design because BWR fuel assemblies have a channel enclosing the rod bundle, i.e.,
it is not an open lattice core.
- 2. The effect of radial pressure gradients in the upper plenum is not a BWR issue becaue BWR fuel assemblies have channels, the core flow distribution and pressure drop analyses are based on an equal pressure drop, from lower plenum to upper plenum, for each bundle. The effect of radial gradients at upper plenum is not considered relevant to BWR design.
- 3. The effect of fuel rod bowing on BWR core performance has been discussed with NRC in 1977 (Reference 1). The highlights of O GE's response to NRC on BWR fuel rod bowing effect on thermal performance are as follows:
(a) Fuel rod bowing effect on bundle thermal performance (critical power) has been considered in General Electric's BWR thermal analysis basis (GETAB, Reference 2) in that numerous tests have been performed to assess the impact of rod bowing and it was concluded that rod bowing of the magnitude expected in GE BWR fuel has no impact on critical power performance.
O (b) A fun scele 8x8 u uS test with m it4no rods inter 4er (typical to GE BWR/6 fuel design) and bowed has been conducted (Reference 3). The four critical rods were C-163 113078
l GESSAR STANDARD REVIEW PLAN ASSESSMENT Item C-11 Page 2
)
PDA EXTENSION i
4 O
l bowed toward the limiting subchannel with resulting
, clearance between 0.053-0.061/ inch. Critical power
, performance was unaffected in the bowed assembly at BWR
- operating condition.
l Based on the comprehensive test results, it is apparent that even for the severe local geometry deviations tested, there is a negligible effect on critical power performance.
! REFERENCES
! 1. G. G. Sherwood (GE) to D. G. Eisenhut (NRC), "NRC Question on l Rod Bowing," March 29, 1971.
- 2. " General Electric BWR Thermal Analysis Basis (CETAB), Data, i Correlation and Design Application," NED0-10958A, January 1977. l i
- 3. R. B. Nixon, B. Matzner, R. T. Lahey, Jr. , "The Effect of Reduced Clearance and Rod Bow on Critical Power in Full Scale Stimulation of 8x8 BWR Fuel", ASME Paper 75-HT-69.
i In addition, the transient methods have been extensively verified.
A These model comparisons against the transient tests data have been V documented in NED0-10802, NED0-10802-1 and NED0-10802-2, which are GE topical reports. Additional tests data have been recorded at
. Peach Bottom-2 and KKM. These data are compared with the ODYN 2
plant transient model in the letter report, E. D. Fuller (GE) to D. F. Ross (NRC), " Transmittal of Draft ODYN Qualification Report,"
] MFN 014-78, January 13, 1978. These results show the adequacy of I the GE plant transient analysis tools. Further testing will of course be considered by General Electric Company from time to time to further qualify plant performance and analytical models as
- j. appropriate.
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O 113078 C-164 I
...---m _ .- - - - , _ -
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GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION C.2.4.47 SRP SECTION: 8.3 REVISION: 0 DATED: 1/01/78 CATEGORY IV ITEM C-12 TIlLE: . Degraded Grid Voltage Conditions NRC GUIDELINES PROVIDED IN ENCLOSURE E STATEMENT OF NRC LETTER, R. S. BOYO TO G. G. SHERWOOD, DATED OCTOBER 13, 1978,
SUBJECT:
EXTENSION REVIEW MATTER FOR PDA's:
G DEGRADED GRID VOLTAGE CONDITIONS As a result of the Millstone Unit Number 2 low grid voltage occurrence, the staff has developed additional requirements concerning (a) sustained degraded voltage conditions at the offsite power source, and (b) interaction of the offsite and onsite emergency power systems. These additional requirements are defined in the following staff positicn.
- 1. We require that a second level of voltage protection for the onsite power system be provided and that this second level of voltage protection shall satisfy the following requirements.
a) The selection of voltage and time set points shall be determined from an analysis of the voltage requirements of the safety-related loads at all onsite system distribution O levels; b) The. voltage protection shall include coincidence logic to preclude spurious trips of the offsite power scurce; c) The time delay selected shall be based on the following conditions:
(i) The allowable time delay, including margin, shall not exceed the maximum tie delay that is assumed in the FSAR accident analyses; (ii) The time delay shall minimize the effect of short duration disturbances from reducing the availability of the offsite power source (s); and O (iii) The allowable time duration of a degraded voltage condition at all distribution system levels shall not result in failure of safety systems or components; (iv) The voltage sensors shall automatically initiate the disconnection of offsite power sources whenever the voltage setpoint and time delay limits have been exceeded; O-C-165 113078
4 GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION Item C-12 O- Page 2 (v)- The voltage sensors shall be designed to satisfy the applicable requirements of IEEE Standard 279-1971,
" Criteria for Protection Systems for Nuclear Powr
-Generating Stations"; and
('vi)'The Technical Specifications shall include limiting condition for operation, surveillance requirements, 0': tr4P setPotate with 4a4=" ead =ex4=" 14 4ts. ead allowable values for.the second-level voltage protection sensors and associated time delay devices.
- 2. We require that the system design automatically prevent load shedding of the emergency buses once the onsite sources are supplying power to all sequenced loads on the emergency buses.
lhe design shall also include the capability of the load shedding feature to be automatically reinstated if the onsite source supply breakers are tripped. The automatic bypass and reinstatement feature shall be verified during the periodic testing identified in Position 3.
- 3. We require that the Technical Specifications include a test requirement to demonstrate the full functional operability and independence of.the onsite power sources at least once per 18 O. months _during shutdown. The Technical Specifications shall include a requirement for tests: (a) simulating loss of offsite power; (b) simulating loss of offsite power in conjunction with a safety injection actuation signal; and (c) simulating interruption and subsequent reconnection of onsite power sources to their respective buses.
- 4. The voltage levels at the safety-related rt.w should.be optimized for the full load and minimuin % conditions that are expected throughout the' anticipated i ,e of voltage variations of' the offsite power source by appropriate adjustment of the voltage tap settings of the intervening transformers.
We require that the adequacy of the design in this regard be verified by actual measurement, and by correlation of measured values with analysis results.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REFERENCE SRP STATEMENT:
This SRV is applicable to plant equipment that is outside the scope of GESSAR. The equipment and conditions prescribed will be provided by the purchaser.
O C-166 113078
Q GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION C.2.4.48 SRP.SECTION: 6. 2.1. 2 REVISION: 0 DATED: 6/1/76 CATEGORY IV, ITEM 0-13 TITLE OF ASSESSMENT ISSUE: Asymmetric Loads on Components Located Within Containment Subcompartments O NRC GUIDELINES PROVIDED IN ENCLOSURE E STATEMENT OF NRC LETTER, R. S. BOYD TO G. G. SHERWOOD, DATED OCTOBER. 13, 1978,
SUBJECT:
EXTENSION REVIEW MATTER FOR PDA's:
In the unlikely event of a pipe rupture inside a major component subcompartment, the initial blowdown transient would lead to pressure l loadings on both the structure and the enclosed component (s). The i staff's generic Category A Task Action Plan.A-2 is designed to develop generic resolutions for this matter. Our present schedule calls for completing A-2 for PWR's during the first quarter,1975.
Pending completion of A-2, the staff is implementing the followirg program:
1
applicants to commit to address the safety issue as part of O their application for an operating license.
- 2. For PWRs at the OL stage of review,-the staff requires case-by-case analyses, including implementation of any indicated corrective measures prior to the issuance of an operating license.
- 3. For BWRs, for which this issue is expected to be of lesser safety significance, the asymmetric loading conditions will ]e evaluated on a case-specific basis prior to the issuance of in operating license.
For those cases which analyses are required, we request the per-formance of a subcompartment, multi-node pressure response analysis of the pressure transient resulting from postulated hot-leg and O cold-les <nuen suct4oa end eisenerse) reecter cooleat system nine ruptures within the reactor cavity, pipe penetrations, and stean generator compartments. Provide similar analyses for the pres-surizer surge and spray lines, and other high energy lines locatec in the containment compartments that may be subject to pressurization.
Show how the results of these analyses are used in the design of structures and component supports.
Q.
H 3078 C-167
O Item C-13 Page 2 EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REFERENCE SRP STATEMENT:
The' attached diagram-details the subcompartments within containment and their blowdown venting configuration for postulated high energy line breaks O With respect to pressurization of the subcompartments, a double ended line break at the worst case location is postulated to occur.
The RELAP4 computer code is used to calculate the mass / energy release rate for the design base accident in each subcompartment.
'The code considers pipe friction and equipment loss coefficients to ensure realistic conservatism in the analysis. The reservoir pressure is conservatively assumed to remain constant during the break / isolation period.
The calculated mass / energy release rates were input to the SCAM 04 computer code for determining the subcompartment design pressures and/or the required through-wall blowdown vent areas. This multi-node code is used to calculate the environmental conditions in all adjacent subcompartments affected by the break, Calculated pres-sures are multiplied by a factor of 1.4 in accordance with Standard O Review Plan 6.2.1.2.
Due to the large through-wall blowdown vent openings in some loca-tions, assurance of the required equipment environmental conditions during plant operations became a concern. This was resolved by installing blowout' panels in specific locations. These panels have a rapid opening time and low differential' release pressure to minimize the initial pressure transient from the line break. The SCAM 04 computer code was used to model the opening characteristics of these panels.
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C-168 113078
Item C-13 Page 3 O' _
F/D F/D
' A' 'E' y 1r VALVE hlE$rPecM v*
O 4 REccAY R e M. >
CONTA I W dEkIT.
O +
- Rw/ C. U n F/D WEAT EscHANCEE DRAisi VALVE STEAM TUNMEL.
O a m ecc.a r a c:._
F /D = Fu_rER .bEM auER AL 2Eid.
, tvJc.o : RsAcroR W Attre. ct@w UP.
m = cm.wA1 EwWoor hELS.
O CONTAINMENT SUBCOMPARTMENTS - BLOWDOWN VENTING CONFIGURATION C-169 113078
se GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION O
C.2.4.49 SRP SECTION: 6.2.6 REV: N/A DATED: 9/1/77 CATEGORY IV, ITEM C-14 TITLE: Cuntainment Leak Testing Program I
l NRC GUIDELINES PROVIDED IN ENCLOSURE E STATEMENT OF NRC LETTER, R. S. BOYD TO G. G. SHERWOOD, DATED OCTOBEP 13, 1978,
SUBJECT:
O EXTENSION REVIEW MATTER FOR PDA'S:
To avoid difficulties experienced in this area in recent OL raviews, the staff has increased its scope of inquiry at the CP/PDA stsge of review. For this purpose, the following information with regard to the containment leak testing program should be supplied,
- a. Those systems that will remain fluid filled for the Type A test should be identified and justification given.
- b. Show the design provisions that will permit the personnel air-lock door seals and the entire air lock to be tested.
- c. For each penetration i.e., fluid systems piping, instrument, electrical, and equipment and personnel access penetrations, A identify the Type B and/or Type C local leak testing that
! V will be done.
- d. Verify that containment penetrations fitted with expansion bellows be tested at Pa. Identify any penetration fitted with expansion bellows that does not have the design capability for Type B testing and provide justification.
l EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REFERENCED SRP STATEMENT Dntainment leak testing for the GESSAR plant is discussed in detail in Chapter 6 of NED0-21424 "238 Nuclear Island Containment Bypass Leakage Sealing and Testing Methods."
O This top icei descriees the Prov4sions 4nciuded 4a the conteinment barrier for 10CFR50, Appendix J. , Type A, B. and C tests. .
- a. Type A test will be performed in a conventional manner and with the positive leakage control systems deactivated so that air seal in leakage will not mask the results. However, lines open to the containment which have seal systems (which will be deactivated) will be closed off, plugged, or otherwise sealed to prevent through-line leakage during Type A testing.
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GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION Item C-14 Page 2
- b. ,Each door of the personnel air lock will be equipped with shaped double inflatable seals. . The space between the douole seals on each door will be capable of being pressurized to 30 psig without the use of test clamps. A 1/4" pressure barrier (buffer) connection will be provided to monitor for leak tightness of the
, double inflatable seals. All shafts which penetrate door will have double packing and test connections accessible from outside O the orvweii persommei iock.
- c. Table 12 of NED0-21424 identifies the leak testing that will be done en Type "B" and "C" penetrations. ,
Following are the penetration tested with Type B tests:
-(1) Equipment hatch (2) Personnel locks: 1.5 SCFH per lock (2 locks)
, (3) Fuel Transfer Tube Blind Flange Seals ,
1 (4) Mechanical Penetration Bellows: I (43 bellows)
(5) Electrical Penetrations (34 penetrations)
Type C tests (items 6 through 9 of Table 12): These are the ,
individual 10CFR50 Appendix J tests of isolation valves which I do not have any form of bypass leakage seals. The assumed leakages are based on valve size, type, and expected usage, and are summarized below:
(1) RHR system relief valve E12-F030 (one 1-in. relief valve)
.(2) RWCU sample drain (two 1-in. gate valves in series)
O <3) orvweii bleed /veat velve <two 2-4n. sete veives 4e ser4es)
(4) Containment vacuum relief valves (one 24-in. check valve in series with one 24-in. butterfly valve in each line, 2 lines)
- d. The containment penetrations dotted with expansion bellows will be tested at about 20 psig. These penetrations will be provided with a test itting to allow pressurizing between the two lamina-tions of bellows.
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GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION O C.2.4.50 SRP SECTION: 6.2.1.4 REVISION: 0 DATED: 1/01/77 CATEGORY IV, ITEM C-15 TITLE: ' Containment Response Due to Main Steam Line Break and Failure of MSLIV to Close NRC GUIDELINES PROVIDED IN ENCLOSURE E STATEMENT OF NRC LETTER, R. S. B0YD TO G. G. SHERWOOD, DATED OCTOBER 13, 1978,
SUBJECT:
EXTENSION REVIEW MATTER FOR PDA's:
O- CONTAINMENT RESPONSE DUE TO MAIN STEAM LINE BREAK AND MSLIV FAILURE In recent OP and OL application reviews, the results of analyses for a postulated main steam line break accident (MSLB) for designs
[
utilizing pressurized water reactors with conventional containments show that the peak calculated containment temperature can exceed for a short time period the environmental qualification temperature-time envelope for safety related instruments and components. This matter was also discussed in Issue No. I of NUREG-0138 and Issue No. 25 of NUREG-0153. The significance of the matter is that it could result in a requirement of requalifying safety-related equipment to higher time-temperature envelopes.
O EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REFERENCE SRP STATEMENT:
The referenced statement from above states that this is applicable to " designs utilizing pressurized water reactors with conventional containments."
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GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION O
V C . 2. 4. 51 SRP SECTIONS: 3.6.1 REVISION: 0 DATED: 11/01/77 3.6.2 CATEGORY IV ITEM C-16 TITLE: Main Steam and Feedwater Pipe Failures NRC GUIDELINES PROVIDED IN ENCLOSURE E STATEMENT OF NRC LETTER, R. S. BOYD TO G. G. SHERWOOD, DATED OCTOBER 13, 1978,
SUBJECT:
EXTENSION REVIEW MATTER FOR PDA's O Provide an identification of the " break exclusion" region of the i
l main steam and feedwater lines. Compartments that contain break '
exclusion regions of main steam and feedwater lines and any safety related equipment in these compartments should be designed to l withstand the environmental effects (pressure, temperature, humidity ;
and flooding) of a crack with a break area equal to the cross sectional area of the break excluded pipe.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REFERENCE SRP STATEMENT:
" Break exclusion" regions for the main steam, feedwater and other lines is that section of piping between the primary containment p isolation valves. Specifically for design basis environmental i calculations, the regions between the:
(a) Main steam inboard and outboard isolation valves 821-F022 and B21-F028, respectively.
(b) Feedwater inboard and outboard check valves B21-F010 and 821-F032.
Structurally, the design base accident within the subcompartments in which " break exclusion" regions are located is a single, double ended, non mechanistic, main steam line break. All ESF equipment is designed to withstand the environmental effects due to temperature, pressure and humidity. The calculated design pressure of the steam line tunnel is multiplied by a factor of 1.4 in accordance with SHP 6.2.1.2.
O The design base flooding accident within the steam line tunnel is the 900,000 lbs. mass loss from a feedwater line break, detailed in GESSAR 15.1.42. The phase separation for flooding evaluations was conservatively neglected. Seismic Category 1, Class 1E temperature monitors and flood detection systems are utilized to detect and alert control room operators of potential piping failures. The steam line tunnel drainage system is such that the postulated flooding incident does not violate plant safety.
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PDA EXTENSION
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l C.2.4.52 SRP SECTIONS: 5.4.1 REVISION: N/A DATED: 1/1/76 9.2.2 CATEGORY IV, ITEM C-17 TITLE: ' Design Requirements for Cooling Water to Reactor Coolant Pumps NRC GUIDELINES PROVIDED IN ENCLOSURE E STATEMENI 0F NRC LETTER, R. S. B0YD TO G. G. SHERWOOD, DATED OCTOBER 13, 1976,
SUBJECT:
EXTENSION REVIEW MATTER FOR PDA'S:
O Demonstrate that the RCS pump seal injection flow will be automatically maintained for all transients and accidents or that there is sufficient t
time and information available to permit corrective action by an operator.
We have established the following criteria for that portion of the component cooling water (CCW) system which interfaces with the reactor coolant pumps to supply cooling water to pump seals and
, bearings during normal operation, anticipated transients, and accidents. j
! 1. A single active failure in the component cooling water system shall not result in fuel damage or a breech of the reactor coolant pressure boundary (RCPB) caused by an extended loss of cooling to one or more pumps. Single active failures include O operator error, spurious actuation of motor-operated valves, and loss of CCW pumps.
- 2. A pipe crack or other accident (unanticipated occurrence) shall not result in either a breech of the RCPB or excessive fuel damage when an extended loss of coo'ing to two or more RC '
pumps occurs. A single active failure shall be considered when evaluating the consequences of this accident. Moderate leakage cracks should be determined in accordance with Branch Technical Position APCSB 3-1.
In order to meet the criteria established above, an interface requirement should be imposed on the balance of the plant CCW system that provides cooling water to the pump seals and motor and pump bearings, so that the system will meet the following conditions:
O 1. That portion of the component cooling water system which supplies cooling water to the reactor coolant pumps and motors may be designed to non-seismic Category I requirements and Quality Group D if you demonstrate that your reactor coolant pumps will operate without CCW for at least 30 minutes without loss of function or the need for operator projective action.
In addition, safety grade instrumentation including alarms must be provided to detect the loss of component cooling water to the reactor control room. The entire instrumentation system, including audible and visual alarms, should meet the O requirements of IEEE Std 279-1971.
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GESSAR STANDARD REVIEW PLAN ASSESSMENT l PDA EXTENSION Q Item C-17 Page 2
- If it is not demonstrated that the reactor coolant pumps and motors will operate at least 30 minutes without loss of function or operator protective action, then the design of the CCW system must meet the following requirements
- )
- 1. Safety grade instrumentation consistent with the criteria for j the reactor protection system shall be provided to initiate I O automatic protection of the plant. For this case, the component cooling water supply to the seals and pump and motor bearings may i
be designed to non-seismic Category I requirements and Quality '
Group 0; or
- 2. The component cooling water supply to the pumps and motors shall be capable of withstanding a single active failure or a moderate energy line crack as defined in our Branch Technical Position APCSB 3-1 and be designed to seismic Category I, Quality Group D and ASME Section III, Class 3 requirements.
The RC pumps and motors are within the RESAR-414 scope of design; therefore, if you choose to demonstrate that an RC pump design can operate with loss of component cooling water for at least 30 minutes without loss of function or the need for operator action, the following must be provided:
- 1. A detailed description of the events following the loss of component cooling water to the RC pumps and an analysis demonstrating that no consequences important to safety may result from this event. Include a discussion of the effect that the loss of cooling water to the seal coolers has on the RC pump seals. Show that the loss of cooling water does not result in a LOCA due to seal failure.
1 2. A detailed analysis to show that loss of cooling water to the RC pumps and motors will not cause a loss of the flow coastdown characteristics or cause seizure of the pumps, asuming no administrative action is taken. The response should include a detailed description of the calculation procedure including:
O a. The equations used,
- b. The parameters used in the equations, such as the design parameters for the motor bearings, motor, pump and any other equipment entering into the calculations, and material property values for the oil and metal parts.
- c. A discussion of the effects of possible variations in part dimensions and material properties, such as bearing clearance-tolerances and misalignment.
O- e. A descrint4ee of the cooiins end luer 4cet4no systems <with appropriate figures) associated with the RC pump and motor and their design criteria and standards.
C-175 113078
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, GESSAR STANDARD REVIEW PLAN ASSESSMENT 4
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Q Page 3
)
- e. Information to verify the applicability of the equations and material properties chosen for the analysis (i.e.,
references should be listed, and if empirical relations are used, provide a comparison of their range of appli-cation to the range used in the analysis).
Should an analysis be provided to demonstrate that loss of component cooling water to the RC pumps and motor assembly is '
C acceptable, we will require certain modifications to the plant Technical Specifications and an RC pump test conducted under l
operating conditions and with component cooling water terminated 2
- for a specified period of time to verify the analysis.
EVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REFERENCE SRP STATEMENT:
The recirculation pump's integrity is maintained by the following design features, analysis, and tests.
- 1) There are two functionally independent pump seal cooling systems.
1 l 2) Low flow and high temperature alarms exist to protect the pump and motor's bearings, windings, and seals.
- 3) Reactor coolant leakage due to a gross pump seal failure has j been evaluated by analysis and supported by tests. Recirculation I
, pump seal leakage through severely degraded seals is not a '
safety concern.
- 4) Transient analysis is included in the FSAR's appendix for the recirculation pump seizure accident. The analysis identifies that the effects and consequences is not threatening to reactor safety.
Discussion a
Safety Review Plan 5.4.1 states the following design criteria for the recirculation pump cooling system.
O 1) A single active failure in the component cooling water system shall not result in fuel damage or a breech of the reactor coolant pressure boundary (RCPB) caused by an extended loss of cooling to one or more pumps. Single active failures include operator error, spurious actuation of motor operated valves, and loss of CCW pumps.
- O C-176 113078 e ., - . ,,n -- _ , , . , - - -,y , w--
t GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION Item C-17 Page 4
- 2) A pipe crack or other accident (unanticipated occurrence) shall not result in either a breech of the RCPB or excessive fuel damage when an extended loss of cooling to two or more RC pumps occurs. A single active failure shall be considered when evaluating the consequences of this accident. Moderate leakage cracks should be determined in accordance with Branch Technical Position APCSB 3-1."
Q INDEPENDENT SEAL COOLING SYSTEMS Seal cooling for the reactor coolant recirculation pumps is provided by the Reactor Building Closed Cooling Water (RBCCW) system. The seal purge system, which is part of the control rod drive (CRD) system, also provides cooling to the recirculation pump seals even though the primary purpose is recirculation pump seal purge. If the RBCCW system failed to provide cooling for any reason, the seal !
purge system would adequately cool tiie recirculation pump seals for l as long as required. These two systems are functionally independent and failure of one system will not lead to failure of the other system.
i The functional independence of the two systems is manifest from the fact that, although the two systems do provide a common function (seal cooling), the designs are based on diverse requirements.
5 Each a' these two systems has a completely independent piping system (the pumps, the piping, and even the water sources are completely independent). The CRD system (seal purge) uses the condensate storage system as its source of water, while the RBCCW system is a closed system which draws make up from the demineralized water system.
ALARMS The following is a sample of the alarms used to alert the operator of performance degradation, o low flow alarm - RBCCW to motor winding c high flow alarms (2) - seal leakage lines o temperature alarms - motor windings (3)
O motor beer 4ao o41 motor thrust bearing motor winding coolant If the operator ignores all the alarms, the pump will continue to run until a winding short occurs due to excessive winding tempera-ture. This would cause an immediate pump trip.
O C-177 113078
GESSAR STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION Item C-17 Page 5 GROSS SEAL FAILURE Analysis has shown (Ref. 2) that under worst case conditions maximum seal leakage is 75 G.P.M. (@ 1550 F). This fluid loss is well within the normal fluid fluctuations and can be compensated by
-normal vessel water level controls. Tests have shown that with loss of seal cooling the seals will maintain their integrity for at least 2 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. There has been no gross seal failure in G.E.'s operating plant experience. In all cases the operator had time to O monitor the seal's performance and plan an orderly shutdown to replace the seals.
RECIRCULATION PUMP SEIZURE The case of recirculation pump seizure represents an incredible event of an instantaneous stoppage of the pump motor shaft of one recirculation pump. This produces a very rapid decrease of core flow as a result of the large hydraulic resistance introduced by the stopped rotor. The operator must ascertain that the reactor scrams with the turbine trip resulting from reactor water swell.
The operator should then regain control of reactor water level through RCIC operation or by restart of the feedwater pump, and he must monitor reactor water level and pressure control after shutdown.
This transient has been modeled and its description is documented O in GeSSAR ChePter 15.
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O C-178 113078
{} STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION C.2.4.53 SRP SECTION: 10.4.7 DATE: 8/01/76 CATEGORY IV, ITEM C-18 TITLE: Design Guidelines for Water Hammers in Steam Generators
(] with Top Feeding Design (BTP ASB-10.2)
This item is applicable to the steam generator for a PWR and therefore was not assessed against the GESSAR design.
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P STANDARD REVIEW PLAN ASSESSMENT PDA EXTENSION C . 2. 4. 54 SRP SECTION: 3.11 REVISION: 0 DATED: 1/01/76 Category IV, Item C-19 TITLE: Environmental Control System For Safety Related Equipment O NRC GUIDELINES PROVIDED IN ENCLOSURE E STATEMENT OF NRC LETTER, R. S. B0YD TO G. G. SHERWOOD, DATED OCTOBER 13, 1978,
SUBJECT:
l EXTENSION REVIEW MATTER FOR PDA's:
IV.C.19 - ENVIRONMENTAL CONTROL SYSTEMS FOR SAFETY RELATED EQUIPMENT l Most plant areas that contain safety related equipment depend on the continuous operation of environmental control systems to maintain the environment in those areas within the range of environmental qualification of the safety related equipment installed in those areas. It appears that there are no requirements for maintaining these environmental control systems in operation while the plant is shutdown or in hot standby conditions. During periods when these '
environmental control systems are shutdown, the safety related equipment could be exposed to environmental conditions for which it has not been qualified. Therefore, the safety related equipment s should be qualified to the extreme environmental conditions that could occur when the control equipment is shutdown or these environ-mental control systems should operate continuously to maintain the environmental conditions within the qualification limits of the safety related equipment. In the second case, an environmental monitoring system that will alarm when the environmental conditions exceed those for which safety related equipment is qualified shall be provided. This environmental monitoring system shall (1) be of high quality, (2) be periodically tested and calibrated to verify its continued functioning, (3) be energized from continuous power sources, and (4) provide a continuous record of the environmental parameters during the time the environmental conditions exceed the normal limits.
CVALUATION OF GESSAR (BWR-6/MK III) WITH RESPECT TO REFERENCE SRP Q STATEMENT:
The same environmental control systems used for postulated accident conditions are used during normal plant operation, including plant shutdown and hot standby conditions. This restricts the safety related equipment from being exposed to environmental conditions beyond which they are justified.
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ite= c-19 O Page 2 Alarms are provided to alert the operator before environmental con- t
'ditions exceed equipment qualification points so that alternate l equipment may be activated. The environmental monitoring system is I (1) of high quality components, (2) capable of being tested and calibrated, (3) energized from a Class lE buss (4) contained in rooms where room temperatures are logged by the process computer to indicate'if normal parameters have been exceeded.
' Chapter 3.11.4 of GESSAR describes in detail the environmental control systems for rooms which contain safety related equipment.
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