ML20039D999

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Nonproprietary Version of Amend 3 to Gessar II
ML20039D999
Person / Time
Site: 05000447
Issue date: 12/14/1981
From:
GENERAL ELECTRIC CO.
To:
Shared Package
ML20039D990 List:
References
NUDOCS 8201060418
Download: ML20039D999 (68)
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Text

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O GENER AL $ ELECTRIC NUCLEAR ENERGY DIVISIONS 22A7000 s n o. 29 REv 3 SHEET REVISION STATUS Sheet ilev Sheet Rev Sheet Rev xiii 3 6A.6-14 3 xiv 0 6A.6-15 3 xvii 3 6A.6-16 3 xviii 3 6A.6-17 3 xxxi 3 6A.6-18 3 xxxii 3 6A.7-1/6A.7-2 3 xxvii/xxviii 3 6A.8-1 3 6.3-43a 3 6A.8-2 3~

6.3-43b 3 6A.8-3 3 6.3-44a/6 3-44b 3 6A.8-4 3 6.3-47 0 6A.9-1 3 6.3-48 3 6A.9-2 3 App 6A Title Pg 3 6A.9-3 3 6A.1-1 3 6A.9-4 3 6A.1-2 3 6A.9-5 3 6A.2-1 3 6A.9-6 3 6A.2-2 3 6A.9-7/6A.9-8 3 6A.3-1 3 6A.9-9 3 6A.3-2 3 6A.9-10 3 6A.4-1 3 6A.9-ll 3 6A.4-2 3 6A.9-12 3 6A.5-1 3 6A.10-1/6A.10-2 3 6A.5-2 3 6A.10-3/6A.10-4 3 6A.5-3 3 6A.5-4 3 6A.5-5 3 .

6A.5-6 3 6A.5-7 3 6A.5-8 3 6A.6-1 3 6A.6-2 3 6A.6-3 3 6A.6-4 3 6A.6-5 3 6A.6-6 3 6A.6-7 3 6A.6-8 3 6A.6-9 3 6A.6-10 3 6A.6-ll 3 6A.6-12 3 6A.6-13 3

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01221 8201060410 011230 gDRADOCK05000 {

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GESSAR II '22A7000 238 NUCLEAR ISLAND Rev. 3 f

121481 CIIAPTI:R 6 CONTENTS (Continued)

Page 6.3.3.5 Use of Dual Function Components for ECCS 6.3-34 6.3.3.6. Limits on ECCS System Parameters 6.3-35 6.3.3.7 ECCS Analyses for LOCA 6.3-35'-

6.3.3.7.1 LOCA Analysis Procedures and Input Variables 6.3-35 6.3.3.7.2 Accident Description 6.3-37 6.3.3.7.3 Break Spectrum Calculations 6.3-39 6.3.3.7.4 Large Recirculation Line Break Calculations 6.3-39 6.3.3.7.5 Transition Recirculation Line Break Calculations 6.3-41 6.3.3.7.6 Small Recirculation Line Break Calculations 6.3-42 6.3.3.7.7 -Calculations for Other Break Locations 6.3-42 6.3.3.7.8 Improved Decay Heat Correlation 6.3-43a l6.3.3.8 LOCA' Analysis Conclusions 6.3-43a 16.3.4 Tests and Inspections 6.3-43b

- 6.3.4.1 ECCS Performance Tests 6.3-43b .

j 6.3.4.2 Reliability Tests and. Inspections 6.3-44a

. 6.3.4.2.1 HPCI Testing 6.3-45 6.3.4.2.2 ADS Testing 6.3-46 6.3.4.2.3 LPCS Testing 6.3-46 6.3.4.2.4 LPCI Testing 6.3-47 6.3.5 Instrumentation Requirements 6.3-47 6.3.6 References 6.3-48 6.4 HABITABILITY SYSTEMS 6.4-1 6.4.1 Design Basis 6.4-3 6.4.1.1 Safety Design Basis 6.4-4 6.4.1.2 Power Generation Design Bases 6.4-6 6.4.2 System Design 6.4-6 ,

6.4.2.1 _ _ Control _ Building Envelope 6.4-6 xiii

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, GESSAR II 22A7000 238 NUCLEAR ISLAND Rev. 0 033180 CHAPTER 6 CONTENTS (Continued)

Page 6.4.2.2 Ventilation System Design 6.4-8 6.4.2.2.1 Control Room Drawings 6.4-8 6.4.2.2.2 Release Points 6.4-8 6.4.2.3 Leaktightness 6.4-9 6.4.2.4 Interaction With other Zones and Pressure-Containing Equipment 6.4-10 6.4.2.5 Shielding Design 6.4-12 6.4.2.5.1 Design Basis 6.4-12 6.4.2.5.2 Source Terms 6.4-14 6.4.2.5.3 Results 6.4-14 6.4.3 System Operational Procedures 6.4-16 6.4.4 Design Evaluations 6.4-16 6.4.4.1 Radiological Protection 6.4-16 6.4.4.2 Smoke and Toxic Gas Protection 6.4-17 6.4.4.3 Life Support 6.4-18 6.4.5 Testing and Inspection 6.4-19 6.4.6 Instrumentation Requirements 6.4-20 6.4.7 Nuclear Island / BOP Interface 6.4-20 6.4.7.1 External Temperature 6.4-20 6.4.7.2 Meteorology (X/Q's) 6.4-21 6.4.7.3 Toxic Gases 6.4-21 6.5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 6.5-1 6.5.1 Standby Gas Treatment System (SGTS) 6.5-1 6.5.1.1 Design Bases 6.5-1 6.5.1.2 System Description 6.5-2 6.5.1.3 Design Evaluation 6.5-5 6.5.1.3.1 shield Building Annulus Pressure Response Analysis 6.5-14 6.5.1.3.2 Auxiliary Building ECCS and RWCU Pump Rooms Pressure Response Analysis 6.5-16 6.5.1.3.3 Fuel Building Pressure Response Analysis 6.5-17 6.5.1.4 Tests and Inspections 6.5-18 llh xiv

GESSAR ' II - 22A7000 238 NUCLEAR ISLAND Rev. 3 121481 CHAPTER ~6

. CONTENTS (Continued)-

'Page-

.6.6.3 Examination Techniques and Requirements 6.6-2 6.6.4 Inspection Intervals 6.6-2

- 6.6.5 Examination Categories and Requirements 6.6-2 6.6.6 Evaluation of Results 6.6-2 6.6.7 System Pressure Tests 6.6-2 6.7 MAIN STEAM POSITIVE LEAKAGE CONTROL SYSTEM (MSPLCS) ,

6.7-1 6.7.1 Design Bases 6.7-1 6.7.1.1 Safety Criteria 6.7.2 System Description 6.7-4

~ 6.7.2.1 General. Description 6.7-4 6.7.2.2 System Operation .6.7-5 6.7.2.3 Equipment 6.7 6.7.3 System Evaluation 6.7-7 6.7.3.1 Functional Protection Features - 6.7 - 6.7".3.2 Effects of Single Act'ive Failures- 6.7-8 6.7.3.3 Effects of Seismic ' induced Failures 6.7-8 6.7.3.4 Isolation Provisions 6.7-9 6.7.3.5 Leakage Protection Evaluation 6.7-9 1 . 6.7.3.6 Failure Mode and. Effects Analysis 6.7-10 6.7.3.7 Influence on Other. Safety Features 6.7 ,6.7.4 Instrumentation Requirements 6.7-10 6.7.5- Inspection and Testing '6.7-11 6.8 PNEUMATIC SUPPLY' SYSTEM 6.8-1

'...1 Design Bases 6.8-1

_ System Description 6.8-1 6.8.3 Safety Evaluation 6.8-4 6.8.4 Inspection and Testing Requirements '6.8-5 6.8.5 Instrument Application 6.8.6 xvii ).

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

T GESSAR II- 22A7000 239 NUCLEAR ISLAND R v. 3 121481 CHAPTER 6 CONTENTS (Continued)

Page 6A APPENDIX 6A - IMPROVED DECAY HEAT CORRELATION FOR LOCA ANALYSIS 6A.1-1 6A.1

SUMMARY

6A.1-1 6A.l.1 Methodology 6A.1-1 6A.1.2 Objective 6A.1-2 6A.2 APPLICATION 6A.2-1 6A.2.1 Exemption Request 6A.2-1 6A.2.2 Applicability 6A.2-1 6A.2.2.1 Operating Plant Analysis 6A.2-1 6A.2.2.1.1 Analysis for New Fuel Only 6A.2-1 6A.2.2.1.2 Analysis for All Fuel 6A.2-1 6A.2.2.1.3 Limiting Break Analysis 6A.2-1 6A.2.2.1.4 Complete Recalculation 6A.2-1 6A.2.2.2 Near Term Plant Analysis 6A.2-2 6A.2.2.2.1 Plants with Completed Analysis 6A.2-2 6A.2.2.2.2 Plants without Completed Analysis 6A.2-2 6A.3 INTRODUCTION 6A.3-1 6A.3.1 Licensing Philosophy for Evaluation Models 6A.3-1 6A.3.2 Other Appendix K Conservatisms 6A.3-2 6A.4 OBJECTIVE 6A.4-1 6A.4.1 Prevent MAPLHGR Derate 6A.4-1 6A.4.2 Efficient Fuel Utilization 6A.4-1 6A.4.3 Cost Effective Allocation of Resources 6A.4-2 6A.5 TECHNICAL BASIS 6A.5-1 6A.5.1 Basis for BWR Core Power Limits 6A.5-1 6A.5.1.1 Peak Pellet Linear Heat Generation Rate 6A.5-2 6A.5.1.2 Core Average Power Density 6A.5-3 6A.S.I.3 Core Average Specific Power 6A.5-5 6A.5.2 Technical Advantages of the 1979 Decay Heat Standard 6A.5-5 6A.5.2.1 Background 6A.5-5 6A.S.2.2 Decay Heat Power from Fission Products 6A.5-6 6A.5.2.3 Effect of Neutron Capture 6A.5-7 6A.5.2.4 Pulsed and Infinite Reactor Operation 6A.5-7 6A.6 METHOD OF ANALYSIS 6A.5-1 6A.6.1 Summary of Decay Heat Terms 6A.6-1 6A.6.2 Total Energy per Fission 6A.6-2 6A.6.3 Fission Heat Induced by Delayed Neutrons 6A.6-4 6A.6.4 Fission Product Decay Heat 6A.6-5 6A.6.5 Decay Heat Due to Actinide Decay 6A.6-7 6A.6.6 Decay of Activated Structure and Poison Materials 6A.6-9 6A.6.7 Application Considerations

  • 6A.6-ll 6A.6.8 Evaluation of Uncertainties
  • 6A.6-15 6A.7 CONCLUSION 6A.7-1 6A.8 REFERENCES 6A.8-1
  • Proprietary material provided under separate cover.

xviii --

GESSAR II 22A7000 238 NUCLEAR ISLAND Rev. 3 121481 CHAPTER 6 TABLES (Continued)

Table Title Page 6.3-4 MAPLHGR, Maximum Local Oxidation and Peak Cladding Temperature Versus Exposure 6.3-55 6.3-5 Summary of Results of LOCA Analysis 6.3-57 6.3-6 Key to Figures 6.3-58 6.4-1 Identification of Failure /Effegt in the Control Room HVAC System 6.4-23 6.4-2 Control Room Heating, Ventilating and Aircondition System Failure Analysis '6.4-24 6.5-1 Standby Gas Treatment System Component Description 6.5-51 6.5-2 Standby Gas Treatment System Failure Analysis 6.5-55

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6.5-3 Post-LOCA Secondary Containment Flow Rates (CFM) to SGTS Time (sec) 6.5-57 6.5-4 Compliance Status with Regulatory Guide 1.52 6.5-59 6.5-5 _ Allocated Leakages 6.5-63 l 6.5-6 Failure Analysis - 6.5-65

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6.7-1 Single-Failure Analysis of Main Steam Positive Leakage Control System 6.7-15 7

6.8-1 Pneumatic Supply System Services 6.8-9 6.8-2 Pneumatic Supply System Capacity Requirements (Per Division) 6.8-10

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6A-1 Current Design Values for Net Energy Release Per Fission from Principal Contributors 6A.9-1 6A-2 Net Energy Release Per Neutron Capture in Selected Isotopes- 6A.9-2 l 6A-3 Total Net MeV/ Fission - Two Lattices at i 40 Void 6A.9-3 6A-4 Effect of Void on Total Net MeV/ Fission Lattice "A" 6A.9-4 l

6A-5 Relative Fission Rate After LOCA 6A.9-5 6A-6 $ and G Factors Versus Burnup and Time After

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Shutdown for a Typical BWR Fuel Assembly 6A.9-6 _

GESSA'R II . 22A7000 238 NUCLEAR ISLAND Rev. 3 121481 CHAPTER 6 TABLES (Continued)

Table Title Page 6A-7 Energy Released Per Actinide Decay 6A.9-7 6A-8 Decay Power Ratio at Various Exposures Relative to Decay Power Following Scram From __

10 GWd/STU* 6A.9-9 6A-9 Decay Power Ratio - Void Dependence

  • 6A.9-10 6A-10 Decay Power Ratio - Enrichment Dependence
  • 6A.9-ll 6A-ll Decay Power as Percent of Shutdowri Power Level, for Various Operating Histories, at Several Times After Shutdown
  • 6A.9-12 -
  • Proprietary material supplied under separate cover.

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xxxti

GESSAR 11 22A7000 238 NUCLEAR ISLAND Rev. 3 0 121481 CHAPTER 6 I'LLUSTRATIONS (Continued).

f Dgure Title Page 6.5-14 Standby Gas Treatment System (SGTS) Distribu-tion and Flow Rates 6.5-83 6.7-la Main Steam Positive Leakage Control System P&I Flow Diagram 6.7-17 0.7-lb Main Steam Positive Leakage Control

6.8-1 Pneumatic Supply System P&I Flow Diagram 6.8-11 6A-1 GE LOCA Model 6A.10-1 6A-2 Decay Power _ Ratio versus Exposure Level at Shutdown for Various Times T (seconds)' -

after Scram

  • 6A.10-3 O
  • Proprietary material provided under separate cover. -

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xxvii/xxviii

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

,' G5 SEAR II 22A7000 238 NUCLEAR ISLAND Rev. 3 121481 r~'s

( ,) 6.3.3.7.7 Calculations for Other Break Locations (Continued)

An analysis was also done for the main steamline break outside the containment. Reactor water level and vessel pressure from SAFE /REFLOOD and peak cladding temperature and fuel rod convective heat transfer coefficients from REFLOOD are shown in Figures 6.3-68 through 6.3-71.

6.3.3.7.8 Improved Decay Heat Correlation Section I.A.4 of 10CFR50, Appendix K, requires use of the 1971 ANS Standards Subcommittee proposed decay heat standard for ECCS licensing evaluations. The current method for applying the 1971 standards in BWR LOCA calculations is outlined in GE's approved ECCS evaluation model (Reference 6.3-2). In 1979, the American National Standards Institute approved and the ANS published a much improved decay heat standard (Reference 6.3-3). A detailed techni-

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cal basis for an improved GE BWR decay heat correlation based on the 1979 standard is outlined in Appendix 6A. Use of the improved correlation in the currently approved GE LOCA models will provide increased ECCS criteria margins.

Application of the correlation described in Appendix 6A is optional. To use it in place of the current method, a utility must provide the NRC with a request for exemption from Section I.A.4 of 10CFR50, Appendix K. The utility must reference Appendix 6A as the technical justification for the exemption.

6.3.3.8 LOCA Analysis Conclusions Having shown compliance with the applicable acceptance criteria of Section 6.3.3.2, it is concluded that the ECCS will perform its function in an acceptable manner and meet all of the 10CFR50.46 acceptance criteria, given operation at or below the MAPLHGRs in Table 6.3-4.

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6. 3-43 a )

GESSAR II - 22A7000 238 NUCLEAR ISLAND R3v. 3 121481 6.3.4 Tests and Inspections lll 6.3.4.1 ECCS Performance Tests All systems of the ECCS are tected for their operational ECCS function during the pre-operational and/or startup test program.

Each component is tested for power source, range, direction of rotation, setpoint, limit switch setting, torque switch setting, etc. Each pump is tested for flow capacity for comparison with vendor data. (This test is also used to verify flow measuring capability). The flow tests involve the same suction and dis-charge source (i.e., suppression or condensate storage tank).

All logic elements are tested individually and then as a system j to verify complete system response to emergency signals including l l

the ability to valves to revert to the ECCS alignment from other )

I positions.

Finally, the entire system is tested for response time and flow capacity taking suction from its normal source and delivering flow into the reactor vessel. This last series of tests is performed with power supplied from both offsite power and onsite emergency power.

See Chapter 14 for a thorough discussion of pre-operational testing for these systems.

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-e GESSAR II- 22A7000 238 NUCLEAR ISLAND Rev. 3 121481

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6.3.4.2 Reliability' Tests and Inspections The average reliability of a standby (nonoperating) safety system ,

is a function of the duration of theLinterval between periodic functional tests. LThe factors' considered in determining the periodic test interval of the.ECCS are: J(1) the desired system availability (average reliability); (2) the number of-redundant functional-system success paths;L(3) the failure rates of the individual components in the system; and (4) - the schedule of periodic tests (simultaneous versus uniformly staggered versus randomly staggered). For the ECCS, the'above factors were used-to determine safe test intervals utilizing the methods described in' Reference 6.3-1. ,

All of ' the active components of the HPCS System, ADS, LPCS and LPCI Systems are designed so that they may be tested during normal plant operation. Full flow test capability is provided by a-test-

) line back to the suction-source. The full flow test is used-to verify the capacity of each ECCS pump loop while the plant remains undisturbed in the power generation mode.- In addition, each individual valve may be tested during normal plant operation.

Input jacks are 'provided such that by racking out the injection valve breaker, each ECCS loop can be tested for response time.

6.3-44a/6.3-44b

GESSAR II . 22A7000 238 NUCLEAR -ISLAND Rev. 0' 033180 6.344.2.4 LPCI. Testing ^

Each LPCI' loop' can' be tested during reactor operation. -The test conditions are tabulated in Figures 6.3-4a, b and c.. During plarat operation, this test _does not inject cold water into the reactor because the injection line check' valve is held closed by vessel pressure, which is higher than the. pump pressure. The -

injection line-portion-is tested with reactor water when the reactor is shut down and when a closed system loop is created.

This prevents unnecessary thermal stresses.

To test an LPCI pump at rated flow, the_ test line valve to the suppression pool is opened, the pump suction valve from the suppression pool is opened (this valve is normally'open) and:the pwnps are started using the remote / manual switches in the control room. Correct operation is determined by observing the instruments in the control room.

If an initiation signal occurs during the test, the LPCI System returns to the operating mode. The valves in the test bypass lines are closed automatically to assure that the LPCI pump discharge is correctly routed to the vessel.

6.3.5 Instrumentation Requirements Design details including redundancy and logic of the ECCS instrumentation are discussed in Section 7.3.

All-instrumentation required for automatic and manual initiation of the HPCS, LPCS, LPCI and ADS is discussed in Subsection 7.3.1 and is designed to meet the requirements of IEEE-279 and other.

applicable regulatory requirements. The HPCS, LPCS, LPCI and ADS can be manually initiated from the control room.

6.3-47

GESSAR II

  • 22A7000 238 NUCLEAR ISLAND R v. 3 121481 6.3.5 Instrumentation Requirements (Continued) gg The HPCS, LPCS and LPCI are automatically initiated on low reactor water level or high drywell pressure. (See Table 6.3-8 for speci-fic initiation levels for each system) The ADS is automatically actuated by sensed variables for reactor vessel low water level and drywell high pressure plus indication that at least one LPCI or LPCS pump is operating. The HPCS, LPCS and LPCI automatically return from system flow test modes to the emergency core cooling mode of operation following receipt of an automatic initiation signal. The LPCS and LPCI system injection into the RPV begin when reactor pressure decreases to system discharge shutoff pressure.

HPCS injection begins as soon as the HPCS pump is up to speed and the injection valve is open, since the HPCS is capable of injecting water into the RPV over a pr'ssure range from 1177 psid* to 200 psid 3 ,

O 6.3.6 References 6.3-1 H.M. Hirsch, " Methods for Calculating Safe Test Intervals and Allowable Repair Times for Engineered Safeguard Systems", January 1973 (NEGO-10739).

6.3-2 " General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50, Appendix",

(NEDO-20566), submitted August 1974, and " General Electric Refill Reflood Calculation" (Supplement to Safe Code Description), transmitted to USNRC by letter, G. L. Gyorey to Victor Stello, Jr., dated December 20, 1974.

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6.3-3 " Decay Heat Power in Light Water Reactors", ANSI /ANS 5.1-1979, Approved by American National Standards Institute, August 29, 1979. _

3 psid - differential pressure betuaen RPV and pump suction source.

O 6.3-48 ]

GESSAR II 22A7000

-238 NUCLEAR ISLAND Rev. 3 121481 O

f APPENDIX 6A IMPROVED DECAY HEAT CORRELATION FOR LOCA ANALYSIS i

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GESSAR II 22A7000 238 NUCLEAR ISLAND Rev. 3 121481

() APPENDIX 6A IMPROVED DECAY HEAT. CORRELATION FOR LOCA ANALYSIS 6A.1

SUMMARY

This appendix outlines the technical basis for an improved corre-lation for determining decay heat generation in a General Electric boiling water reactor (BWR) for loss-of-coolant accident (LOCA) analysis. Currently,Section I.A.4 of 10CFR50 Appendix K requires the use of the 1971 ANS Standards Subcommittee 5 proposed decay heat standard 1'for all emergency core cooling system (ECCS) licensing evaluations. In 1979 the American National Standards Institute approved and the ANS published a revised decay heat standard.2 Both industry and the NRC acknowledge that the 1979 i

standard is superior to the 1971 standard because it deals in greater detail with the. physics involved and has a considerably-

.better data base.

6A.l.1 Methodology This analysis method not only includes fission product decay heat correlation but also contains descriptions of other major con-t tributors to post LOCA heat generation. 'These include:

1. Decay of actinides
2. Decay of activation products
3. Fission heat due to delayed neutrons Other important phenomena addressed in this technical basis
include
l. Use of correlations based on actual fuel histories rather than a hypothetical infinite irradiation case.

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GESSAR II- 22A7000 238 NUCLEAR ISLAND Rav. 3 121481 6A.l.1 Methodology (Continued)

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2. Use of exposure dependent energy / fission values.
3. Use of the correlation in a "best estimate" mode as well as with a 2a uncertainty factor.

6A.1.2 Objective The objective of submitting this correlation is to provide a more accurate assessment of the decay heat during LOCA. This will result in a more realistic, yet conservative, calculation of the PCT during postulated LOCA's. In addition this will demonstrate additional PCT margin which could provide the following major benefits for BWR owners:

1. Prevention of MAPLHGR (maximum average planar linear heat generation rate) derate.
2. More efficient utilization of fuel.

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3. Cost effective allocation of resources.

Both near term and operating plants could benefit from the use of this correlation when used in conjunction with the currently approved GE LOCA models. Utilities can request an exemption from the 1971 decay heat standard in 10CFR50 Appendix F and reference this appendix as the technical basis for using the more correct 1979 standard.

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G5SSAR II 22A7000 238 NUCLEAR ISLAND Rev. 3 121481 O

() 6A.2 APPLICATION 6A.2.1 Exemption Request In order to utilize the correlation-outlined in this appendix, utilities must provide the NRC with a request for exemption from Section I.A.4. of 10CFR50 Appendix K. They must reference this document as the technical justification for.the exemption.

6A.2.2 Applicability This technical basis is applicable to both near term and operating plants.

6A.2.2.1 Operating Plant Analysis 6A.2.2.1.1 Analysis for New Fuel Only O The correlation may be used in the heat-up model for. calculations on new reload fuel only.

6A.2.2.1.2 Analysis for All Fuel The correlation may be used in the heat-up model for calculations on all the fuel.

6A.2.2.1.3 Limiting Break Analysis The limiting break can be recalculated using the LOCA system model with the new decay heat correlation.

6A.2.2.1.4 Complete Recalculation A complete LOCA analysis for all break sizes incorporating the new decay heat correlation can be performed.

6A.2-1

GESSAR II' 22A7000 238 NUCLEAR ISLAND Rev. 3 121481 6A.2.2.2 Near Term Plant Analysis h

6A.2.2.2.1 Plants with Completed Analysis For plants with completed analysis the available options using the new correlation are the same as listed above for operating plants.

6A.2.2.2.2 Plants without Completed Analysis A complete LOCA analysis for all break sizes incorporating the new decay heat standard can be performed.

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6A.2-2

GESSAR II 22A7000 238 NUCLEAR ISLAND Rev. 3 121481 6A.3 INTRODUCTION 6A.3.1 Licensing Philosophy for Evaluation Models In the past, the General Electric Loss-of-Coolant Accident.(LOCA)

Evaluation Models (EM) have been developed as an interconnected series of analytical modules and empirical correlations. To a large degree, each of these modules is-evaluated independent of.

its interaction with the remainder of the analytical system. Most often, the result of the independent evaluation of these individual modules has been that the mathematical representation must bias each module in such a manner so as to maximize the calculated peak cladding temperature. As shown in Figure 6A-1, there are in. excess of 20 independently identified modules in the total analytical system. If each of these are independently biased to represent a 90% probable upper bound on calculated peak cladding temperature, the resulting calculated peak cladding temperature will not repre-

- sent a physically real outcome.

w This previous philosophy of biasing the mathematical' formulation of'each model module in a " conservative" manner results in a mis-representation of not only-the ultimate evnsequences of a poten-tial accident, but for many events this process also misrepresents the sequence and timing of events which are expected to happen and there'are is of no value in assessing the impact of potential design options or for being used as a basis for preparing operator guidelines. In fact, with the extreme to which the current phi-losphy has been implemented, using the current evaluation models for these objectives may actually result in degradation of real plant safety.

1 General Electric strongly advocates the use of " expected value" mathematical representation of each module for future versions of the evaluation model. Model uncertainties would be accommodated by the use of a realistic adder on the calculated output of the

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V 6A.3-1

e GESSAR II' 22A7000 238 NUCLEAR ISLAND Rav. 3 121481 6A.3.1 Licensing Philosophy for Evaluation Models (Continued) lh Evaluation Model. With this approach, the Evaluation Model could be used to:

e Ouantify the real expected safety margin for operating

,v .its e Provide a technical basis for the preparation of operation guidelines to be followed in the event an accident were to occur e Provide a technical basis for evaluating alternate equipment designs or modes of plant operation 6A.3.2 Other Appendix K Conservatisms This appendix does not address any conservative correlations cur-rently in Appendix K other than the decay heat model (i.e., Moody discharge model and Baker-Just 4metal / water reaction correla-tion). The purpose of this exemption submittal is to obtain ECCS margin without a real compromise of reactor safety and with a minimum of both NRC and industry resource expenditure. The decay heat correlation is an external rodule which is used with the General Electric evaluation model codes as an input table. The utrrent GE LOCA codes do not have to be internally modified in ordec to use the later 1979 decay heat standard. Only the decay heat input parameters need to be changed.

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GISSAR II -22A7000 238 NUCLEAR ISLAND Rav. 3 121481

- (Q 6A.4 OBJECTIVE

%.J The primary objective of'this General Electric correlation based on the 1979 ANSI /ANS-5.1 decay heat standard2 is to provide a more accurate assessment of the decay heat during LOCA resulting in

~ increased ECCS margin for both operating and requisition plants.

The,use of this decay heat correlation, with a realistic uncer-tainty, in the current GE LOCA evaluation model will result in anywhere from 200*F to 400*F peak cladding temperature (PCT) margin. There are three major benefits available for both oper-ating and requisition plants due to this increased PCT margin.

6A.4.1 Prevent MAPLHGR Derates Any plant which has less than 54 maximum average planar linear heat generation rate . (MAPLHGR) operating margin available is vul-nerable to derate from unexpected events. The increased PCT

,, margin available from this analysis will be sufficient to prevent

([) any MAPLHGR related derates.

6A.4.2 Efficient Fuel Utilization For all plants the increased PCT margin provides for better utilization of fuel due to the increased operating range available with a more realistic LOCA limit. ECCS operating limits based on local power levels require BWR's to intentionally flatten power dis-tribution. However, a BWR inherently contains sources of hetero-geneities (i.e., water gaps,_ voids, etc.) resulting in a natural tendency toward varying spatial power distribution. Arbitrarily flattening of power distribution to meet a local limit may have a detrimental effect on neutron economy resulting in inefficient utilization of fuel thereby increasing utility fuel costs. As a rule of thumb, relaxation of a local limit by 10% can result in a fuel cost savings of about 3%. A 3% fuel cost savings for a 1000 MWe BWR equates to about 3 to 6 million dollars per year.

O 6A.4-1

22A7000 GESSAR II ' R v. 3 238 NUCLEAR ISLAND 121481 6A.4.3 Cost Effective Allocation of Resources g An additional advantage of increased margin is to allow cost effective allocation of industry resources. Current evaluation models result in tabulations very near the licensing limits.

Different phenomena can always be postulated in the models which result in an increase in calculated PCT over the limit. This causes an immediate allocation of utility and GE resources to fully investigate the phenomena to prevent derating the plant in question. With a more realistic margin between the licensing limit and the calculated model PCT, the postulated phenomena could be addressed in a cost effective manner.

O O

6A.4-2

'r

~

GESSAR II- 22A7000 i

238 NUCLEAR ISLAND Rev. 3 121481

(-~)

q, 6A.5 TECHNICAL BASIS 6A.S.1 Basis for BWR Core Power Limits The decay heat model described in this submittal will be applied to-the analysis of-BWR's whose core power limits do not exceed the following values:

o Peak pellet heat generation rate 14.4 kW/ft (for 8x8 fuel) e Core average power density 57 kW/l e Core average specific power 28 kW/kg Uranium

.These limits represent a near term workable plan and no need 1:s seen to extend them for General Electric supplied BWR's operating in the United States. If a future occasion arises which requires the

+

extension of one of these limits, the requested increase would be

{}

accompanied by detailed submittal covering its justification.

The sole basis for these specific values is that they envelop the

. values for all currently operating GE supplied BWR's as well as those under design for fature operation. The purpose of specifying these limits in this submittal is to commit to the Regulatory Agency that GE will not use any operating margin obtained as a result of this submittal as an opportunity to increase the' total core power output beyond these specified values. The licens-ing analysis for these plants, performed with LOCA models 5 cur-rently approved by the USNRC, has demonstrated that plant rated power is achievable within all constraints of 10CFR50 Appendix K.

The objective of this submittal is not to upgrade the core power 6A.5-1

GESSAR II 22A7000 238 NUCLEAR ISLAND Rev. 3 121481 6A.5.1 Basis for BWR Core Power Limits (Continued) level, rather it is to obtain licensing recognition of a portion of the existing BWR safety margin for two primary objectives:

o Provide some margin between the licensing calculations and the licensing limit such that future potential modeling or plant performance issues can be addressed

. in a planned systematic manner rather than under the threat of imminent operating plant derate.

e Provide some operating flexibility in terms of local power distribution such that the utility owner may increase the efficiency of the uranium ut.11zation, thereby reducing the resulting power generation costs.

A discussion of each of these parameters is presented below.

6A.5.1.1 Peak Pellet Linear Heat Generation Rate O

The fuel rod linear heat generation rate is measured in terms of kilowatts of thermal power per foot of fuel rod length. This parameter has been chosen as a key performance indicator because the fuel pellet peak and average temperatures are directly pro-portional to t: Tis parameter. Therefore, important factors which may impact fuel thermal mechanical performance (such as stored energy, ther.m.al expansion, margin to incipient fuel melting, etc.)

may be inferred from this single parameter.

The instrumentation and data processing required for on-line moni-toring of the peak pellet value of parameter are relatively straightforward in BWR's through the use of the in-core neutron detectors and the analytical models encoded into the process com-puter. Controlling this parameter specifies a firm limit on the total power spatial peaking factor at high power operation.

6A.5-2

GESSAR II 22A7000 238 NUCLEAR ISLAND Rev. 3 121481

,y

() 6A.5.1.1 Peak Pellet Linear Heat Generetion Rate (Continued)

One of the key checkpoints on this parameter is that the point of incipient centerline melting of the UO fuel ccurs at approximately 2

22 kW/ft. The operating limit for the 7x7 fuel assemblies remaining in the operating plants is 17.5 kW/ft for BWR/2-3 and 18.5 kW/ft for BWR/4. This fuel type has provided an operating data base of approximately 750,000 fuel rods over a 10-year operating period at linear heat generation rates substantially above the committed 8x8 limit of 14.4 kW/ft. The conclusions of this experience were that there are no detrimental consequences of steady state operation within the ranges of linear heat generation rates tested. However, the plant data revealed that certain operational fuel duty situa-tions could result in cladding perforation due to pellet-cladding interaction (PCI) with a frequency that increased with increasing peak values of linear heat generation rate. Based on this extensive data base, limiting the peak pellet linear heat generation rate to

(~N 14.4 kW/ft will provide a greater margin with respect to PCI and s/

even greater margin with respect to fuel melting than the large data base provided by 7x7 fuel operation.

6A.5.1.2 Core Average Power Density The core average power density is measured in terms of kilowatts per liter. This parameter serves as an indicator of both the thermal power being releasel to the cooling system and the neutron flux density level experienced by ccre structural material. The total core volume is fixed by the plant design; therefore, this parameter is controlled through the continuously monitored value of total core thermal power.

The 57 kW/l core power density translates to an average bundle power of slightly less than 5 MW. With the channeled modular construction of the BWR, there is no hydra;'lic coupling between

,_ fuel assemblies in the core region. Theref(re, the thermal

(_) hydraulic consequences of power density can be completely 6A.5-3

GESSAR II 22A7000 238 NUCLEAR ISLAND Rev. 3 121481 6A.S.l.2 Core Average Power Density (Continued) determined by full scale testing of a single channeled fuel assembly. Extensive full scale tests have been conducted in the 6

ATLAS test facility on GE supplied fuel. The range of test parameters now covered are e Pressure 75 to 1600 psia (rated N1015)

~

6 e Mass Flux -0.1 x 10 to 1.50 x 10 6 lb/hr-ft 2 (rated %1.0 x 106) e Inlet Subcooling 0 to 100 Btu /lb (rated %20) e Bundle Power 0 to 12 MW The bundle powers tested represent power density ranging from 0 to sl30 kW/1. Therefore, all thermal hydraulic test conditions exceed ggg the range of values expected to occur during BWR plant operation.

No adverse thermal hydraulic phenomena were observed up to the power level at which the onset of transition boiling occurred. The conditions under which this phenomenon occura have been accurately described in the General Electric Critical Quality Boiling Lettgth (GEXL) correlation. Proper consideration of this phenomenon is assured by the specification of bundle minimum critical oower ratio (MCPR) during the licensing process and by monitoring this parameter during plant operation to assure cc;pliance.

The value of 57 kW/l for core average power density is well below the values tested in full scale fuel bundle tests for General Electric supplied fuel. Furthermore, the transition boiling phenomenon is specified and controlled in BWR operations indepen-dently of this parameter limiting core average power density.

O 67.5-4

GESSAR II 22A7000 238 NUCLEAR ISLAND Rev. 3 121481 6A.5.1.2 . Core Average Power Density (Continued)

There is no technical basis to indicate that variations in this value would result in any perceptible degradation of plant safety.

The particular value of 57 kW/l was selected since it envelops the GE supplied BWR plants currently operating and being designed.

6A.5.1.3 Core Average Specific Power The core average specific power is measured in kilowatts per kilogram of uranium loaded. This parameter is controlled through the weight of fuel loaded and the total core thermal power which.

is a continually monitored parameter during plant operation.

Specific power is an indicator of the rate at which fuel burnup (mwd /MT) occurs and of the energy generation rate relative to the energy storage capacity of the fuel.

Substantial data exist for values of specific power as high as

( )-

75 kW/kg. This data base has resulted in no identified adverse phenomena related to this parameter. However, the rate at which the fuel temperature would increase, if the fuel surface heat transfer coefficient were severely reduced (such as in the case of a film boiling event) , is directly proportional to the operating specific power. Therefore, low specific powers provide additional insurance in the event of degradation of the surface heat transfer coefficient. The commited value of 28 kW/kgU of core average specific power is well below the values for which experimental data confirm that no threshold for new adverse phenomena exists.

6A.5.2 Technical Advantages of the 1979 Decay Heat Standard 6A.S.2.1 Background In October 1971, the ANS Standards Subcommittee 5 proposed the 3 adoption of a standard entitled " Decay Energy Release Rates Follow-k) ing shutdown for Uranium-Fueled Thermal Reactors." It has since 6A.5-5

GESSAR II 22A7000 238 NUCLEAR ISLAND Rev. 3 121481 6A.S.2.1 Background (Continued) remained in the status of a proposed ANS standard; however, the Atomic Energy Commission, now Nuclear Regulatory Commission (NRC),

has used the proposed standard in the regulatory process. This proposed ANS standard was based on the curve recommended by K. Shure for infinite irradiation of uranium and for cooling times from 0 to 10 9 seconds. The approach was simplistic in that a single curve was chosen to represent the decay heat power of

" uranium-fueled thermal reactors." Many phenomena that make the decay heat unique to each case were ignored and assumed to be in-cluded within the appropriately large uncertainties that were adopted.

In August 1979, the American National Standards Institute approved and the ANS published the new revised ANSI /ANS-5.1-1979 standard entitled "American National Standard for Decay Heat Power in Light Water Reactors." This new revised standard has significant technical advantages over the original version.

lll 6A.S.2.2 Decay Heat Power from Fission Products In this new revised standard, values are provided for decay heat power from fission products from fissioning of the major fission-able nuclides present in LWR's, i.e., U 235 and Pu'39 thermal and 238 U fast, and methods are prescribed for evaluating the total fission product decay heat power from the data given for these specific fuel nuclides. The original standard gr.ve one standard curve for " uranium-fueled reactors."

These new values for the major fissionable isotopes are based upon a statistical evaluation of new experimental data and summation calculations. The experimental data is from Friesenhahn, et al.,

at IRT and Dickens, et al., at ORNL using radiation detection methods and from Yernell, et al.,l1 at LASL and Schrock, et al.,1 at the University of California, Berkeley, using calorimetry. The 6A.5-6

GESSAR II 22A7000 238 NUCLEAR ISLAND- Rev. 3-121481 (O,j 6A.5.2.2 Decay Heat Power from Fission Products (Continued).

summation calculations.were performed using the ENDF/B-IV data file and demonstrated that idestical results;were obtained using any of the well known codes, ORIGEN, CINDER, and RIBD. 6 Schmittroth and Schenter 17 developed and applied the generalized least-square method to obtain the best value and uncertainty of decay heat power from the experimental data and summation calcula-tion results. This method is the basis of the decay heat power data

^

chosen for the new revised standard data for 235U and 239P u thermal fission for cooling times less than 10 5 seconds. All othe;- data in the new revised standard are from summation calculations only and have been assigned a slightly inflated uncertainty. It is explicitly noted that in the new revised standard, the uncertainty is expressed in a statistical sense as one standard deviation in a normal distribution.. The uncertainty specified in the original

< standard was ambiguous.

)

6A.5.2.3 Effect of Neutron Capture The effect of neutron capture in fission producta during reactor operation is accounted for in the new revised standard. Shay,18,19 i England, Shure,21 and Tasaka have evaluated the effect of neutron capture from summation calculations and have shown that the effect is on the order of only a few percent even for the longest practical reactor operating times in LWR's for cooling times less than 10 4 seconds. 'Spinrad 23 has completed a parametric study of j the influence and formulated an empirical representation for short times. England, et al.20 have si. 3wn that the correction for long cooling times can be large and is dependent upon a larger number of reactor parameters.

6A.5.2.4 Pulsed and Infinite Reactor Operation The new revised standard presents decay heat for two irradiation

() conditions: a fission pulse and infinite reactor operation. In 6A.5-7

GESSAR II 22A7000 238 NUCLEAR ISLAND Rev. 3 121481 6A.S.2.4 Pulsed and Infinite Reactor Operation (Continued) lh the absenca of neutron capture in fission products, these two cases are related to one another by an exact mathematical expression. This makes it possible to obtain the decay heat and itu uncertainty, in the absence of neutron capture in fission prod-ucts, for any finite or infinite and/or variable or constant power history correctly accounting for exposure dependent fuel compositions.

O O

6A.5-8

GESSAR II- 22A7000 238 NUCLEAR ISLAND Rev. 3.

121481 6A.6 ' METHOD OF ANALYSIS This section describes the methods used to calculate the decay heat soarce in a reactor during a loss-of-coolant accident.

The sources of heat are identified and the methods used to calcu-late them are listed. Sample numerical results are also presented to illustrate the sensitivity cf decay heat values to fuel burnup and reactor operating history.

6A.6.1 Summarh of Decay Heat Terms There are four potential nuclear energy sources in a reactor following shutdown: fission heat induced by delayed neutrons, decay heat from fission products, decay heat from actinides and decay heat from irradiated structural materials. The-decay heat is proportional to the total fission rate just prior to shutdown.

This fission rate is most conveniently calculated as the power

(~3 c'ivided by the total energy released per fission, and the decay

("/

heat is given by P

t " C(T N( +Df(t,T) + A(t,T) +A s(t,T)} .(6A-1) where t = time after shutdown (sec)

T = irradiation time (sec). This time is usually defined as the total time a particular fuel-bundle is in an operating reactor.

Q(T) = total energy released per fission (MeV/ fission).

This quantity includes fission energy from all fissioning nuclides as well as energy released from

! n, Y reactions. The calculation of Q is discussed

/~' in Subsection 6A.6.2.

d 6A.6-1 i

GESSAR II 22A7000 238 NUCLEAR ISLAND R r.v . 3 121481 6A.6.1 Summary of Decay Heat Terms (Continued) lg Df(t,T) = Decay heat (MeV/ fission) from fission products.

This term is calculated in accordance with ANS Standard 5.1 (1979). The calculation is discussed in detail in Subsection 6A.6.4.

A(t,T) = Decay heat generated by actinide decay (MeV/

fission). The evaluation of this term is discussed in Subsection 6A.6.5.

As(t,T) = Decay heat generated by activated structural and poison materials (MeV/ fission). The evaluation of this term is discussed in Subsection 6A.6.6.

Pss = Steady state power just prior to shutdown.

DN(t) = Fission heat induced by delayed neutrons (MeV/

fission). This term is discussed in Subsec-tion 6A.6.3.

The ratio R(t) is defined by the following equation:

R(t) =

Q T) {D N

+

f

' ^ ' +As (t,T) } (6A-2)

It is the ratio of decay heat at time t to the heat just prior to shutdown and will be used as the basis for comparison in determining the relative magnitude of the decay heat terms as a function of irradiation time and power history.

6A.6.2 Total Energy per Fission The total energy release per fission is determined in the General Electric Company design calculation by assigning a net energy release to each fission, and adding the energy produced by the non-fission neutron capture effect. The net energy release per 6A.6-2

GESSAR II 22A7000 238 NUCLEAR ISLAND Rev. 3 121481

<m

( ,) 6A.6.2 Total Energy per Fission (Continued) fission is the sum of the fission fragment and neutron kinetic energies, the prompt and delayed gamma energies, and the (delayed) beta energies. The energy of the neutrinos are not included. The kinetic energy of the neutron which produced the fission must be subtracted from the kinetic energy of the fission-produced neu-trons in order to get the net energy. Table 6A-1 shows the values utilized in the most current lattice design code for the net energy per fission in the principal fission isotopes. They are 4

taken from the ENDF/B-2 data base.

There are approximately 1.5 neutrons captured in various non-fission reactions for every neutron which causes a fission. These capture reactions produce more energy, usually from n,y reactions.

In addition, certain reactions create unstable products which eventually produce more energy as they decay to a stable isotope.

,r^]

\

An example is the formation of U237 by the absorption of a neutron in U236. There is a prompt release of 5.12 Mev in gamma energy followed by a 8- emission, to Np237.

The present design code determines the non-fissile capture in each material in the fuel bundle, and calculates the capture energy produced by utilizing the following equation:

3 90 9m,j #c m,j m ij O(MeV)/ fission = $"1 3 *"125 (6A-3)

"fm,j #j "m j=1 m=1 where j = flux group

,_ m = material identification

()

6A.6-3

GESSAR II 22A7000 238 NUCLEAR ISLAND Rav. 3 121481 6A.6.2 Total Energy per Fission (Continued) h q = net energy produced per neutron captured in material m in group j Nm = atom density of material m The summation occurs over the whole lattice. The energy assigned per non-fission capture for each material is shown in Table 6A-2.

It includes the 8- and y energy that is released by the decay of short-lived daughters, such as U237, U239, Np239, and Gdl59.

Table 6A-3 shows the net MeV/ fission for two BWR lattices at various exposures. The lower enrichment design has u flatter reactivity shape at low burnups. Therefore, the number of non-fission capture per fission increases with exposure over the whole range of expo-sure. The low enrichment bundle also has a greater portion of the fissions occurring in plutonium than does the higher enrichment bundle.

Table 6A-4 illustrates the effect of voids on the total net MeV/ fission for one lattice design. There is slightly less net energy per fission at low voids than at high voids. However, the effect is small and decreases as the discharge exposure level is approached.

6A.6.3 Fission Heat Induced by Delayed Neutrons When a reactor shuts down either during normal or accident conditions, the power level does not drop immediately to zero.

Instead it decays away with time due to fissions caused by delayed neutrons. The decay rate is determined by the physical properties of the reactor fuel and by the magnitude of the negative reac-tivity insertion. During a LOCA there are two main sources of negative reactivity: void feedback which begins immediately and O

6A.6-4

GESSAR II 22A7000 238 NUCLEAR ISLAND Rev. 3 121481

(~1 6A.6.3 Fission Heat Induced by Delayed Neutrons (Continued) v!

control rod insertion which is delayed for a short period due to instrument response and the inertia of the control rods.

The current method for determining the decay of the fission heat is a computer code which is a one-dimensional coupled nuclear and thermal-hydraulic model. It uses a single neutron energy group and makes the prompt jump approximation. The thermal-hydraulic model is identical to the model in the SCAT code. Further details about the model and its application may be found in pages I-10 to I-35 of General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50, Appendix K.

The ratio R(t) =Dy(t)/Q is tabulated in Table 6A-5. The values listed in Table 6A-5 are conservative for Design Basis Accident (DBA) LOCA events because they were calculated assumir;g a slow blowdown rate which results in a smaller void feedback term, and henca a slower decrease in neutron flux following the break.

(/

6A.6.4 Fission Product Decay Heat The fission product decay heat evaluation is based on the ANS standard 5.1 published in 1979. The procedures used in evaluating the total decay heat follow as closely as possible those outlined in the standard. Fission products are created from fissions occurring in three principal nuclides, U235, U238 and Pu239.

There are fissions which occur in other nuclides but we shall assume, as directed in the standard, that all nuclides other than Pu239 and U238 have the same fission product characteristics as U235. The fission product decay term is given by 3 T (t'

Dg(t,T) = G(t,T) Pss Q ' # i(t') fg(t+T-t') dt' 1=1 o (6A-4) o 6A.6-5 t

GESSAR II 22A7000 238 NUCLEAR ISLAND R;v. 3 121431 6A.6.4 Fission Product Decay Heat (Continued) where T is the total irradiation time and Pss and Q(T) are the power density and energy per fission immediately before shutdown.

The quantity rf(t') is the fraction of total fissions in isotope i at time t. This quantity is calculated by the GE BWR bundle design nuclear code as a function of fuel bundle irradiation time or bundle exposurc. Fuel exposure is defined as T

E (T) = P(t') dt' (6A-5) f o where o f is the initial weight of heavy metal (fissile and fertile) material. The quantity r U235 is calculated as

  1. (t)U235 " 1 - U238#(t) - #(t)

Pu239 (6A-6)

The quantity fg(t) is defined as the decay heat power t seconds after a fission pulse from fissionable nuclide i (MeV/sec/ fission) .

This definition is identical to that given in the ANS standard.

The exponential fit form is used to evaluate fi(t).

23

-A"iD fi(t) = ani e (6A-7) n=1 The coefficients ani and Ani are given in t?e ANS 5.1 standard.2 The function G(T,t) accounts for neutron capture in the fission products and is given in the standard as 0.4 G (T, t) = 1. + t (3.24 x 10-6 + 5.23 x 10-10 t) T 9 (6 A-8 )

The quantity Q, defined as the fissions per initial fissile atom, has been calculated for a typical BWR bundle, and is tabulated as a function of exposure in Table 6A-6.

6A.6-6

GESSAR II 22A7000 238 NUCLEAR ISLAND Rev. 3

'121481 (q~,s j 6A.6.5 Decay Heat Due to Actinide Decay Actinides are heavy elements produced from neutron capture in uranium and plutonium isotopes. Two isotopes produce the vast majority of actinide decay energy in light water reactors. These isotopes are uranium 239 and neptunium 239. The next four nuclides, U237, Am244, Np238, and Pu243, represent less than 8% of the total actinide contribution for times less than 10,000 seconds after sautdown from a high burnup core. The actinide decay energy ,

- is given by 6

A(t) =

}[ A f Ng(t) Qg/(I f$)

1=1 (6A-9) where N

i = atom density of actinide nuclide i

( If c = fission rate immediately before shutdown

Q1 = energy released from decay.of nuclide i It is conservative to assume that at shutdown, the concentration of each actinide is at its equilibrium value. The decay is then given.by t

e' i Nf(t) =Nf(o) ,

where Ai is the decay constant for nuclide i. The above expression is true for all nuclides except Np239 which is formed from the l decay of U239. In this case

+

Ny(t) =Ny(o) yfy,j 1 e it, i e

~A j t (6A-ll) j i 6A.6-7 i

y 4 - -w.-- w -- -,_._,e,, 3, .,ry,,.m._ s4,-1 - -.+ , _ , _ _- . . . - -

GESSAR II 22A7000 238 NUCLEAR ISLAND j R v. 3 121481 where A 3 = decay constant of U239. The equilibrium concentration is calculatcd assuming no captures in the radioactive nuclide.

h i

I &

^

Ni(o) =

(6A-12) i P

where I,g ? is the number of captures in the parent nuclide (nuclide with mass number 1 less than the actinide). The expression for A(t) becomes P

6 Q [L A(t) =

4 Q.

1 gg(t) (6A-13) i=1 (E f /

whcre gg(t) =e -A i t (6A-14)

O for U239, U237, Am244, Np238 and Pu243 and gi(t) = e ~it_ Ai e

-A i t (6A-15)

A -A g A -A g for Np239.

TheratioI[/E g is obtained from bundle analysis results as a function of bundle exposure. The quantity Q i, the energy release per actinide decay, is tabulated in Table 6A-7.

O 6A.6-8

GESSAR II 22A7000 238 NUCLEAR ISLAND R;v. 3 121481

.~

(,), 6A.6.6 Decay of Activated Structure and Poison Materials The principal structural material in the reactor is zirconium.

Two isotopes of zirconium have daughters which have long half-lives. These are Zr94 and Zr96. The reactions are as follows:

B- B-Zr94 + n + Zr95 + Nb95 + Mo95 (6A-16) 8- B-Zr96 + n + Zr97 + Nb97 + Mo97 (6A-17)

The half-life of Nb95 is 3.5 days while the Zr95 has a 64 day half-life. The comparable values for the 97 chain are 73.6 minutes and 16.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Therefore, the Nb isotopes decay at the same effective rate as the Zr parents.

The total energy (6- and y) from the decay of Zr95 and Nb95 is 1.65 MeV. The Zr97 and Nb97 produces 2.65 MeV. A typical lattice,

(^)

~

at high exposure and 40% void, has a neutron balance which shows that 0.0084 MeV/ fission is due to the decay of the unstable Zr isotopes and their daughters. This is between 1 and 2% of the energy produced by the U239, Np239 decay chain during the first 3

10 seconds after a shutdown.

Other structural materials in and around the reactor are the steels in the control blades, the shroud, the bottom support plate, etc. The captures in the control blades produce less than 0.0003 MeV/ fission in decay heat. The decay heat produced by neutron captures outside of the core is of a similar m tanitude.

Both are excluded from the decay heat model.

Two Gd isotopes have long-lived daughters. These are Gdl58 and Gdl60. The reactions are as follows:

B-Gdl58 + n + Gdl59 + Tbl59 (18.6 hr) (6A-18) g_ g_

(_) Gdl60 + n + Gdl61 + Tbl61 + Dyl61 (3.7 min, 6.92 day)

(6A-19) 6A.6-9

GESSAR II 22A7000 238 NUCLEAR ISLAND R :.v . 3 121481 6A.6.6 Decay of Activated Struc_ture and Poison Materials (Continued)

The S- and y energy produced the Gdl59 decay is about* 0.354 MeV/ decay. The 161 chain yields 0.931 + 0.247 = 1.178 MeV.*

The number of captures in these isotopes was determined at discharge exposure levels, where this value per fission event is maximized, in a lattice with seven 4 wt % Gd O r ds. The Gdl58 23 absorbs 0.408 x 10 -2 neutrons / fission, while the Gdl60 captures 0.0306 x 10- neutrons per fission. The total energy obtained from these daughters is 0.18 x 10 -2 MeV/ fission, or about one-fifth of the Zr chain at shutdown. The seven 4 wt % Gd 0 design 2 3 is a rather heavily poisoned lattice. Most of the energy is from the 159 chain. Fuel assemblies with less Gd O will have less 23 Gd159 and Gdl61 produce 6.

O

  • These are one-third of the averaged maximum 6- decay energies, plus the remaining y energies.

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GESSAR II 22A7000 238 NUCLEAR ISLAND Rev. 3 121481 6A.6.8 Evaluation of incertainties (Continued)

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l O

6A.6-16

' GESSAR II 22A7000 238 NUCLEAR ISLAND Rev. 3' 121481 6A.6.8 Evaluation of Uncertainties (Continued) i, l

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GESSAR II 22A7000 238 NUCLEAR ISLAND Rev. 3 121481 6A.6.8 Evaluation of Uncertainties (Continued)

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6A.6-18

GESSAR II 22A7000 238 NUCLEAR' ISLAND Rev. 3 121481

() 6A.7 CONCLUSION i Use of.the improved LOCA analysis decay heat methods outlined in this appendix will' provide from 200*F_to 400*F peak cladding tem-perature (PCT) margin in current ECCS-calculations. To be allowed  ;

to use this-analysis method a utility must. apply for and be l granted an exemption by the.NRC to Section I.A.4 of:10CFR50 Appendix K. Details regarding the application of this evaluation correlation must be addressed on a plant specific basis.

4 C:)

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4 GESSAR II 22A7000 238-NUCLEAR' ISLAND Rev. 3 121481 6A.8 REFERENCES

(

1. - ' Decay Energy Release Rates Following Shutdown of Uranium Fueled Thermal Reactors,". Proposed American Nuclear Society Standard, Approved by ANS-5 subcommittee, October 1971.
2. " Decay Heat Power in Light Water Reactor," ANSI /ANS 5.1-1979, Approved by American National Standards Institute, August 29, 1979.
3. F. J. Moody, " Maximum Flow Rate of a Single Component, Two Phase Mixture," Journal.of Heat Transfer, Trans American Society of Mechanical Engine 6rs,.87, No. 1, February 1965.
4. L. Baker and L. C. Just, " Studies of Metal Water Reactions

}

at High Temperatures, III. Experimental and Theoretical Studies at the Zirconium-Water Reaction," ANL-6548, page 7, ,

, May 1962.

t O.

i'- 5. " General Electric Company Analytical Model for Loss of Coolant. Anslysis in Accordance with 10CFR50, Appendix K,"

General Electric Company, November 1973 (NEDE-20566-P)

(GE Company Proprietary).

6. " General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application," General Electric Company, Novemh7r 1973 (NEDO-10958).
7. " General Electric BWR Thermal Analysis Basis (GETAB) Data, Correlation and Design Application," General Electric Company, January 1977 (NEDE-10958-PA).

8.- K. Shure, " Fission-Product Decay Energy," in USAEC Report WAPD-BT-24, pp. 1-17, December 1961.

6A.8-1 w3 -,y,~ a y. ,m ....,4~n n .- , --' - m m =

GESSAR II 22A7000 238 NUCLEAR ISLAND Rev. 3 121481 6A.8 REFERENCES (Continued) llh

9. S. J. Friesenhahn, N. A. Lurie, V. C. Rogers, and N. Vagelatos, "235 U Fission Product Decay Heat from 1 to 10 seconds," Electric Power Research Institute, February 1976 (EPRI NP-180).
10. J. K. Dickens, J. F. Emery, T. A. Love, J. W. McConnell, K. J. Northcutt, R. W. Peelle, and H. Weaver, " Fission-Product Energy Releass for Times Following Thermal-Neutron Fission of 235 U Between 2 and 14,000 Seconds," Oak Ridge National Laboratory, 1977 (ORNL/NUREG-14 ) .
11. J. L. Yarnell and P. L. Bendt, " Decay Heat from Products of U Thermal Fission by Fast Response Boil-Off Calorimetry,"

Los Alamos, 1977 (LA-NUREG-6713) .

12. V. E. Schrock, L. M. Grossman, S. G. Prussin, K. C. Sockalingam, F. M. Nuh, C-K Fan, N. Z. Cho, and S. J. Oh, Final Report on Contract RP 230, "A Calorimetric Measurement of Decay Heat from U Fission Products from 5

10 to 10 Seconds," Project 230 Final Renort, February 1978, Vol. 1, Electric Power Research Institute (EPRI NP-616).

13. " Fission-Product Decay Library of the Evaluated Nuclear Data File," Version IV (ENDF-B-IV), available from and maintained by the National Nuclear Data Center (NNDC) at the Brookhaven National Laboratory, Upton, New York.
14. M. J. Bell, "ORIGEN - The ORNL Isotope Generation and Depletion Code," Oak Ridge National Laboratory, May 1973 (ORNL-4628).
15. T. R. England, R. Wilczynski, and N. L. Whittemore,

" CINDER-7: An Interim Report for Users," Los Alamos, April 1975 (LA-5885-MS). lll 6A.8-2

e GESSAR II 22A7000 238 NUCLEAR ISLAND Rev.-3 121481

() A

6.8 REFERENCES

(Continued)

16. D. R. Marr, "A User's Manual-for Computer Code RIBD-II, A
Fission Product Inventory Code," Hanford Engineering Develop-ment Laboratory, January 1975 (HEDL-TME-75-26).
17. F. Schmittroth and R. E. Schenter,_" Application of Least-Squares Method to Decay Heat Evaluation," Nuc. Sci & Eng.,

69, 389-397 (1979).

4

18. M. R. Shay, " Summation Evaluation of Reactor Afterheat

-Including the Effect of Neutron Capture in Fission Products,":

I Thesis, Oregon State University, Corvallis, Oregon (August 4

1976).

19. M. A. Bjerke, J. S. Holm, M. R. Shay, and B. I. Spinrad, "A Review of Short-Term Fission Product Decay Power," Nuclear Safety,-Vol. 18, 596 (1977).

O

20. T. .R. England, M. G. Stamatelatos,'R. E. Schenter, and F. Schmittroth, " Fission-Product Source Terms for Reactor Apolications," Proceedings of the American Nuclear Society Topical Conference cut Thermal Reactor Safety, Vol. 2, 240,_

August 1977 (CONF 77078).

" 5

21. K. Shure, 0 Fission Product Decay Energy - 1972 Re-Evaluation," Uestinghouse Atomic. Products Division,

, October 1972 (WAPD-TM-1119).

22. K. Tasaka, "Effect of Neutron Capture Transformations on the Decay Power of Fission Products," Nuc. Sci. & Eng., 6 2, -

167-174 (1977).

23. B. I. Spinrad and A. Tripathi, "Modeling the Effect of Fission Product Capture on Reactor Decay Power," Nuc. Sci. &-

() Eng., 66, 140-141 (1978).

6A.8-3

GESSAR II 22A7000 238 NUCLEAR ISLAND R:;v. 3 121481 A

6.8 REFERENCES

(Continued) gg

24. " Evaluated Nuclear Data File," Version V (ENDF/B-V) available from, and maintained by, the National Nuclear Data Center (NNDC) at the Brookhaven National Laboratory, Upton, New York.
25. B. C. Slifer and J. E. Hench, " Loss-of-Coolant Accident and Emergency Core Cooling Models for General Electric Boiling Water Reactors," General Electric Company, April 1971, (NEDO-10329), Appendix C.
26. C. L. Martin, " Nuclear Basis for ECCS (Appendix K) Calcula-tiens, " General Electric Company, November 1977 (NEDO-23729).

O O

6A.8-4

F GESSAR II 22A7000 238 NUCLEAR ISLAND R v. 3 121481 f

1 ),, Table 6A-1 CURRENT DESIGN VALUES FOR. NET

  • ENERGY RELEASE PER FISSION FROM PRINCIPAL CONTRIBUTORS Nuclide Basis - MeV/ Fission **

U235- 194.02 1 0.12-U238 195.01 t 0.36 Pu239. 200.05 0.30 Pu241 202.22 1 0.34 U234 190.3 1.1 U236 192.8 t 1.1 Pu240 197.4 1.0 Pu242 200.6 1 1.0 U233 191.29 i 1.1 Th232 185.85. 0.58

  • Kinetic energy, prompt plus delayed gamma energy, and delayed-data energy, less the fission-indacing neutrons' energy. No capture gamma energy. Neutrino energy not included. ENDFB-234 values. -
    • The lattice design code does not include the uncertainty term.

s I

d 6A.9-1

GESSAR II

^

R 238 NUCLEAR ISLAND jg 3 Table 6A-2 NET ENERGY RELEASE PER NEUTRON CAPTURE h

IN SELECTED ISOTOPES Design Basis Isotope Intermediate Decay (MeV/ Capture)

U235 6.541 U236 (8-)U237 5.12+0.35=5.47 U238 (8-)U239, (8-)Np239 4.80+0.46=0.42=5.68 U234 5.312 Pu238 5.65 Pu239 6.53 Pu240 5.24 Pu241 6.30 P242 (8-)Pu243 5.04+0.18=5.22 Th232 (8-)Th233c(8-)Pa233 4.78+0.449+0.449=5.678 U233 6.83 Np237 (8-)Np238 5.48+0.46=5.94 Am243 (8-)AM244 5.34+0.48=5.82 Gd154 6.439 Gdl55 8.53 Gdl56 6.364 Gdl57 7.93 Gdl58 (8-)Gd159 5.936+0.354=6.290 Gd160 ( 8-) Gdl61, ( 8-) Tbl61 5.631+0.951+0.247=6.829 HO2 2.224 B10 2.788 Zirc-2 thermal 7.705*

Zirc-2 epithermal 8.185*

2 O 4.142 1

All captures in H, except fast neutrons which are 90%O2, 10%H.

2 Thermal and epithermal. Fast captures are n,a, with little net energy release (assumed zero).

  • Based on isotopics of pure zirconium. Two isotopes have complex decay chains. g_

Zr94 + n + Zr95 +g_N b 95

  • M 95 llh i

Zr96 + n + Zrg7 + Nbg7 + Mo97 6A.9-2

GESSAR.II.

l 3 238 NUCLEAR ISLAND 121481

) Table 6A-3

-TOTAL-NET MeV/ FISSION - TWO LATTICES AT 40 VOID Lattice "A" = 3.03 enrichment 7 4 wt % Gd 02 3 rods Lattice "B" ='2.575 enrichment 2 7 wt % Gd203 rods MeV/ fission-

' MWD /STU Lattice A Lattice B 0 -201.82 201.08 0.2 202.38 201.67 1 202.56 202.10 3 202.71 202.89 5' 202.66 203.49 7 202.50 203.95 10 202.88 204.45 15 204.37 205.70

) 20 205.85 207.35 25 207.30 208.97

'30 208.75 210.48 35 210.12 -211.83 40 211.36 212.98 O

f6A.9-3

22A7000 GESSAR II R:v. 3 238 NUCLEAR ISLAND 121481 Table 6A-4 EFFECT OF VOID ON TOTAL NET MeV/ FISSION LATTICE "A" MeV/ fission mwd /STU Void =0 __40 70 0 201.23 201.82 202.45 5 202.10 202.66 203.25 10 202.15 202.88 203.67 15 203.72 204.37 205.06 20 205.30 205.85 206.44 25 206.95 207.30 207.74 30 208.69 208.75 208.99 35 210.37 210.12 210.11 40 211.92 211.36 211.12 O

O 6A.9-4

GESSAR II 22A7000 238 NUCLEAR ISLAND Rev. 3 121481

. Table 6A-5 RELATIVE FISSION RATE AFTER LOCA-Time After" Shutdown Relative Fission Rate t (sec) R (t)

-0.0 1.000 0.1 0.982 0.2 0.920.

0.4 0.724 0.6 0.559 0.8 0.457 1.0 0.293.

2.0 0.101 3.0 ~1 0.378 x 10 4.0 ~1 0.191 x 10 5.0 -1 0.109 x 10 6.0 0.834 x 10 7.0 0.744 x 10 8.0 0.600 x 10

-2 9.0 0.460 x 10 10.0 -2 0.330 x 10 20.0 0.130' x 10 -2

-3 30.0 0.900 x 10 40.0 -3 0.500 x 10 50.0 -3 0.400 x 10

-3 60.0 0.290 x.10 O

6A.9 -

GESSAR II 22A7000 238 NUCLEAR ISLAND R:~.v . 3 121481 Table 6A-6 h

$ AND G FACTORS VERSUS BURNUP AND TIME AFTER SHUTDOWN FOR A TYPICAL BWR FUEL ASSEMBLY

  • G factor - 1.0 Burnup Gid/STU $ t=10 see t=10 see t=10 sec

-6 -6 -6 0.2 0.00926 6.64x10 7.70x10 17.33x10

-5 -5 1 0.0463 6.33x10~ 7.34x10 16.51x10

-4 -4 ~4 3 0.1389 2.94x10 3.41x10 7.67x10

~4 -4 -4 5 0.2315 6.02x10 6.98x10 15.71x10 10 0.463 1.59x10~ 1.84x10~ 4.14x10~

20 0.926 4.19x10~ 4.86x10~ 10.93x10~

-3 -3 30 1.389 7.39x10~ 8.57x10 10.29x10

-3 -3 40 1.852 ll.06x10~ 12.83x10 28.86x10

  • 2.5% enriched bundle operating @ 50 kw/ liter O

O 6A.9-6

i GESSAR II 22A7000

, 238 NUCLEAR ISLAND Rev. 3

, 121481 Table 6A-7 ENERGY RELEASED PER ACTINIDE DECAY Actinide Avg. Q-Mev/ disintegration t1/2 ,

U239 0.474.1 23.5 minutes Np239 0.419 1 -23.5 days U237 0.358 2 6.75 days Np238 2 0.884 2.12 days i

Pu243 0.223 2 4.956 hours0.0111 days <br />0.266 hours <br />0.00158 weeks <br />3.63758e-4 months <br />.

Am244 0.482, 1.170 2 26m, 10. lh (3) 9 fANSI/ANS-3.1-1979

,l/3 averaged max. S- energy + gamma energy

~%70% decays with 26 min half-life, 30% with 10.1 hr.

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GESSAR II 22A7000

.238 NUCLEAR ISLAND Rev. 3 121481

'O Table 6A-8 l' . DECAY POWER RATIO AT VARIOUS EXPOSURES RELATIVE TO

DECAY POWER RATIO FOLLOWING SCRAM FROM 10 GWd/STU k

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GESSAR II 22A7000 238 EUCLEAR ISLAND Rev. 3 121481 Table 6A-9 h DECAY POWER RATIO - VOID DEPENDENCE GE PROPRIETARY - provided under separate cover O

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, GESSAR II 22A7000 -

j _- 238. NUCLEAR ISLAND. Rev.--3 ,

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GESSAR II 22A7000 238 NUCLEAR ISLAND Rev. 3 121481 Table 6A-ll O

DECAY POWER AS PERCENT OF SHUTDOWN POWER LEVEL, FOR VARIOUS OPERATING HISTORIES, AT SEVERAL TIMES AFTER SHUTDOWN GE PROPRIETARY - provided under separate cover 1

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l l Power Peaking Factors l l Initial Gap Conductance l Ififif Ififif PLANT INPUTS

- GEOME TRY >

RADIATION VIEW FACTORS h

~

FISSION GAS RE LE ASE I CRITICAL FLOW CORRELATION j PERFORMATION STR AIN VS TEMP h VOID FR ACTION CORREL ATION YlELD STRESS VS TEMP h y y

UTP CCFL CORRE LATION METAL WATER RE ACTION CORRE LATION h o UTP CCFL BREAKDOWN CORRELATION hO OM CORE SPR AY HE AT TRANSFER COEFFICIENT I PE AK CLADDING t-* m e

gM TE MPE RATURE >

N SEO CCFL CORRE LATION E m LOCA PRESSURE VS TIME W W 38 SYSTEM = HEAT UP SEO CCFL BRE AKDOWN CORRELATION MODEL TIME OF NODE UNCOVERY MODEL g g O  % mH TIME OF RATED SPRAY FLOW REACTION t'*

y l CORE EXIT BYPAS* CCFL TIME OF NODE REFLOOD m

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SYPASS LEAKAGE FLOW CORRELATION HE AT TRANSFER COEF VS TIME PRESSURE LOSS COEFFICIENTS h TO TIME OF RATED CORE SPRAY l VAPORIZATION CORRELATION HE AT TR ANSFER COEFFICIENT

- CRITICAL POWE R CORRE LATION

- REWET CORRELATION

- NUCLEATE BOILING A

- TRANSITION BOILING M

- FLOW FILM SOILING

- POOL FILM BOILING "y g# p M

- ST E AM F LOW H <; a

b. O m o Figure 6A-1. GE LOCA Model Hwo

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j. GESSAR II .22A7000 238 NUCLEAR ISLAND -Rev.-3 121481 LO I

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.. after Scram r

6A.10-3/6A.10-4 '

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