ML20011A431
| ML20011A431 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 10/10/1973 |
| From: | GULF UNITED NUCLEAR FUELS CORP. |
| To: | |
| Shared Package | |
| ML20011A420 | List: |
| References | |
| SS-1126, NUDOCS 8110130342 | |
| Download: ML20011A431 (46) | |
Text
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SS-1126
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SUPPLEMENTAL INFORMATION ON THE LACROSSE BOILING WATER PEACTOR EMERGENCY CORE COOLING SYSTEM r
October 10, 1973 Work Performed on Gulf United Project 3431 for Dairyland Power Cooperative
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GULF UNITED NUCLEAR FUELS CORPORATION
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Elmsford, New York i
8110130342 010929 PDR ADOCK 050004o9 P
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TABLE OF CONTENTS 1.
INTRODUCTION 1-1 2.
ECCS TEST RESULTS AND OPERATING EXPERIENCE...
2-1 2.1 Introduction........................
2-1 2.2 Technical Specification Requirements for ECCS 2-1 2.3 Dascription of LACBWR ECCS.............,
2-1 2.4 Operating History and Test Results for ECCS......
2-2' 2.5 Events Causing Ataomatic Operating of ECCS, Description of Resulting Thermal Transients C
and Their Effect on LACBWR Systems and Com pon e nt s........................
2-8 3.
MODIFICATIONS fO IMPROVE THE LACBWR ECCS....
3-1 3.1 In;roduction............ -
3-1 3.2 Modificatiens Completed to Date or in Progress 3-1 3.3 Modifications Proposed to Comply with AEC Interim Criteria...................
3-1 4.
CLAD TEMPERATURES DURING A LOSS-OF-COOLANT ACCIDENT OF THE BIG ROCK POINT REACTOR 4-1 4-1 4.1 Introduction........................
4.2 Analysis Based on LACBWR Model............
4-3 4.2.1 Fuci Rod Heatup Analysis 4-3 4.2.2 Fuel Assembly Heatup Ana.ysis 4-5 4.2.2.1 Description of Model 4-5 4.2.2.2 Initial Temperatures 4-9 4.2.2.3 Effects of Core Spray...
4 4-9 4.2.2.4 Radiation Interchange Factors 4-9 4.2.2.5 Metal-Water Reaction..........
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4.2.3 Clad Temperature 4-15 f.
4.3 Analysis with Revised Model 4-15 4.3.1 Description of Model 4-15 4.3.2 Clad Temperature 4-18 5.
REFERENCES 5-1 APPENDIX A - LACBWR FACILITY CHANGE NO. 73-11....
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- f TABLES 2.1 Summary of Operating History for LACBWR ECCS 2-3 2.2 Summary of Results of Semiannual Tests for LACBWR ECCS High Pressure Core Spray System 2-7 2.3 Summary of Results of Annual Tests for LACBWR ECCS High Pressure Core Spray System............
2-8 2.4 Summary of Results of Semiannual Tests for LACBWR ECCS Alternate (Low Pressure) Core Spray System......
2-9 2.5 Summary of Results of Annual Tests for LACBWR ECCS Alternate (Low Pressure) Core Spray System 2-10
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2.6 Summary of Stress Analyses on LACBWR Vessel and Fuel Due to Thermal Transient of May 15,1970......
2-12 3.1 Summary of Modifications to Improve LACBWR ECCS Completed to Date or in Progress 3-2 3.2 Summary of Proposed Modifications to Improve' LACBWR ECCS 3-4 4.1 Big Rock Point Loss-of-Coolant Calculations 4-4 4.2 Cladding and Fuel Temperatures During Rod 4-7 Heatup Period 4.3 Big Rock Point High Power Fuel Assembly Rod Data 4-10 4.4 Input Parameters, Fuel Rod Assembly Heatup Analysis, Design Basis Accident -Big Rock Point Reactor........
4-11 4.5 Radiation Interchange Factors with Dry Shroud.........
4-12 4.6 Radiation Interchange Factors with Wet Shroud.........
4-13 4.7 Material Properties Fuel TIGE11 Code 4-16 t
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FIGURES 4.1 Center Rod Cladding Temperatures During LOCA, Big Rock Point Plant...................... 4-2 4.2 lieat Transfer CoefficientsforDesign Basis Accident..... 4-6 4.3 Big Rock Point Fuel Assembly -Type F............ 4-8 4.4 Revised TIGER Model for Fuel Assembly Heatup....... 4-17 l
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- 1. INTRODUCTION As a basis for issuing a full-term operating license for the Lacrosse Boiling Wate'r Reactor (LACBWR), the AEC in a letter has requested t
additionalinformation on certain plant design features, engineered safety systems and management and administrative procedures.
This report gives supplementalinformation on the adequacy of the Emergency Core Cooling System (ECCS) requested under Item 2.2 of the 1
AEC letter which states:
" Discuss the reliability of the emergency core cooling system based on test results and operating experience. Describe the events when the ECCS was initiated automatically and discuss the consequences of these transients. Describe the thermal transients, conditions of the core and effects on reactor vessel of core spray operation incidents.
List the modifications to improve the emergency condenser and core spray reliability and capability. Provide analyses or references to show that the requirements of AEC Interim Acceptance Criteria for ECCS are satisfied."
An analysis to show that the present ECCS meets the AEC Criteria 2
was submitted in May 1972 with supplementalinformation on densification effects and specific replies to questions by the AEC supplied in April and 3
May 1973 . These analyses showed that the Lacrosse ECCS would meet
- e the requirements of the AEC Interim Acceptance Criteria.
The present report gives the additionalinformation requested under 1
Item 2.2 of the AEC letter concerning operating experience and test results with the existing ECCS and information on modifications, both planned and already executed, to improve the system. Also included in this submittal
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g is a calculation of the cladding temperature of the Big Rock Point Reactor during a loss-of-ccolant accident (LOCA) and a comparison with clad temperatures calculated by the reactor manufacturer (General Electric).5 Results show that up to clad temperatures where the metal-water reaction becomes significant, there is good agreement between the GE and Gulf United temperature calculations. The Big Rock Point calculations shown here have been performed at the request of the AEC to serve as a validation of the analytical tools used by Gulf United in the LACBWR loss-of-coolant analyses.
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- 2. ECCS TEST RESULTS AND OPERATING EXPERIENCE
2.1 INTRODUCTION
The reliability of the LACBWR ECCS is evaluated based on its operating history and test results. The thermal transients and stresses in the core and vessel resulting from partialloss of cooling water in the core during the incident of May 15,1970 is reviewed briefly and a summary of the important temperatures and stresres presented. References to relevent studies and reports are included. This incident represents the most severe transient which required ECCS action in the course of the LACBWR operation to date.
2.2 TECHNICAL SPECIFICATION REQUIREMENTS FOR ECCS The functional requirements for the LACBWR ECCS are described in some detailin the LACBWR Technical Specifications, Sections 2.4.6, 2.4.7, and 4.2.2.14 through 4.2.2.18.
Similarly, the tests to be performed on the ECCS and their frequency are described in Sections 5.2.7 and 5.2.8 of the LACBWR Technical Specifi-cations
2.3 DESCRIPTION
OF LACBWR ECCS The ECCS consists of two separate subsystems: (1) the high pressure emergency core spray system and (2) the low pressure alternate core spray (core deluge) system. These systems automatically supply cooling water to the reactor vessel to minimize fuel damage due to a loss-of-coolant incident.
Detailed descriptions o. these systems with related flow diagrams are given in Gulf United Report SS-942.2 4
2-1
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- t 2.4 OPERATING HISTORY AND TEST RESULTS FOR ECCS 4
The operating history for the LACBWR ECCS is summerized here in two compilations of reported data. Incident reports and repair and maintenance accounts are given in Table 2.1 and the results of all semi-annual and annual tests required by the Technical Specifications are given in Tables 2.2 through 2.5.
i The test procedures for the annual tests required by the LACBWR Technical Specifications for the high pressure corp spray system are contained in the LACBWR Operating Manual, Volume II, Sections 8.5 and 8.6.
l The semiannual test on the high pressure core spray system valves is performed in conjunction with the semiannual test of the boron injection system as described m Sections 9.4.3 through 9.4.5 of the LACBWR Operating Manual, Volume II.
The test procedures for the sen'iannual and annual tests required by l
the Technical Specifications for the alternate core spray system are contained in the LACBWR Operating Manual, Volume II, Section 17.5.
Results of te*'s on this system are given in Tables 2.4 and 2.5.
With the exceptan of the single instance reported in the LACBWR
.c Monthly Opelating Report DPC-851-38 fo: December 1971, when a fitting was damaged during disassembly of the core spray assembly all semi-i annual and annual tests conducted to date have satisfied Technical Specifi-cation requirements.
Review of the events listed in Table 2.1 indicates that both the High Pressure Core Spray System and the Alternate Core Spray System have always operated satisfactorily when required. Certain auxiliary systems, however, such as the dieselfire pumps which supply river watcr when the high pressure service water pressure drops below a set point of 60 psig and the speed control systems for reactor feed pumps 1A and 1B appear to require modific9 tion to improve their reliability. In the case of the speed control systems particularly, there appears to be a controlinsuf-ficiency which causes reactor water level to~ wander to scram set point levels when the feed pump control is cwitched from automatic to manual operationorirom one pump to the other. Over the years, this action has resulted in at least six unplanned scrams, during several of which the ECCS was activated. This undesired operating characteristic has been reviewed and corrective measures initiated under Facility Change No. 73-11, included as Appendix A. The replacement of the scoop tube posilloner with
!E a Beck rotary actuator with integral position indicator and control system l
should prodace a more linear and wider range of controller span to obtain the desired feed flow,
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TABLE 2.1 -
SUMMARY
OF OPERATING HISTORY FOR LACBWR ECCS Date Report No.
Description of incident Corrective Measures Required O a on j
9/67 ACNP-67-532 Not required Modifications recommended by Alternate (Low Pressure) Core Spray System was instalk d i
10/67 ACNP-67-534 AEC-D R L 11/67 12/67 ACNP-68500 Not required Maintenance Replaced valves and 5, eats for 1 A Core Spr ny Pump.
6/68 ACNP-68508 Not required Maintenance Soleno'd valve 38-25-001 was disassembled, cleaned and tested.
i 12/G9 DPC 851-2 Not required Maintenance Flainless steel core spray oc ute safe ends found sensitized.
i 1/70 DPC-851-3 Not required Maintenance Removal and replacement planned.
1/15/70 DPC-70-1 Not required Reactor not in operation. While washing Defective control relay 20CR replaced and Battery Bank A circulating intake water screens, Diesel charged to normal voltage.
4 Fire Pump 1B did not start when high
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pressure service water pressure dropped below 60 psig set point.
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1/70 DPC-851-3 Not reqHred Precautionary erodification Relief valve set at 150 esig was installed in Alternate (Low i
Pressure) Core Spray line to protect line from normal reactor pressure. Tested in May 1970. (See DPC 051-11) 4 j
2/70 DPC-851-4 Not required Maintenance Replacement of core spray nozzle safe ends completed thh i
month. (See DPC-851-2 above).
2/2/70 DPC-70-2 Not required Reactor not in operation. During monthly Pressure sensing lines were insulated and relocated away from test, Diesel Fuel Pump 1 A failed to start outside wall to prevent future freezing.
when hijh prem.re service water pressure dropped below 60 psig set point.
7 4/70 DPC-851-10 Not required Maintenance Replaced stainless steel core spray flange bolts because they co were overstressed. Higher strength bolts were installed.
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TABLE 2.1 -
SUMMARY
OF OPERATING HISTORY FOR LACBWR ECCS (Continued) i i
i ECCS j
Date Report No.
Operation Description of incident Corrective Measures Required i
5/15/70 DPC-70-3 Automatic Malfunction of Turbine Main Steam Loose cover screws on hydraulic system directional valve and Manual Bypass. Valve system caused uncon-which caused malfutat.an of Turbine Steam Bypass valve were I
trollable oscillation of Reactor Water t;ghtened and safety wired to prevent repetition of troubla.
f Level necen!tating manuni reactor Hydraulic system was overhauled. Reliei valves were recali-scram.
brated; damaged oil reservoir sight level caps and hydraulic oil i
replaced. A high pressure control room alarm was installed on hydraulic system. Reactor feedwater pump controls were i
adjusted to operate at lower discharge pressures. MSIV oper-ation changed to require manual reopen af ter automatic j,
closure.
7/14/70 DPC-70-9 Automatic During routine switchover from 18 to Feedwater pump se:oop tube settings were correted.
i 1 A reactor feedwater pump, reactor water level dropped below -12 inches, i
scrammed the reactor and activated ECCS.
l 2/5/71 DPC-71-1 Automatic Plant electrical power failure with Relocated lubricating oil pressure switch for diesel generator
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reactor in hot standby. Emergency engine for faster response and therefore ouicker startirg of j
shutdown condenser brought into diesel generator. Decreased cut in pcassure setting for the j
e.arvice to remove decay heat. ECCS pressure switch and provided immersion heater to keep oil was activated automatically when fluid during standby:
reactor water level reache.
- 12 inches.
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3/24/71 DPC-71-7 Manual 69-kw LAC 8WR tie line breakers were Af ter isolating faulted line for repair, power was restons.
j tripped due to fire in yard transformer A number of indirectly related modifications were made.
j resulting in load rejection. Reactor j
scrammed with loss of plant power.
i Shutdown condenser and ECCS were l
controlled manually for orderly i
shutdown and recovery.
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t TABLE 2.1 -
SUMMARY
OF OPERATING HISTORY FOR LACBWR ECCS (Continued) 4 I
l ECCS Date Report No.
Operation Descriptir of incident Corrective Measures Required 8/12/71 DPC-71-16 Not required 1 A Diesel Fire Pump starting motor fail-18 Diesel Fire Pump iw.s temporarily cut into service while ure due to welded starting relay contacts.
new relay and st:;rter motor were installed on 1 A. Service Reactor power was reduced during re-then restored to normal.
quired repairs.
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12/7!
DPC-851-38 Nc,t required S.S.1/4. inch NPT fitting for ECCS was Damaged fitting was replaced with new one.
damaged during disassembly of distri-bution header. Reactor shut down for i
maintenance.
6/72 DPC-851-44 Not required Maintenance Dye-penetrant inspections made or welds in high pressure core spray system. All tests were satisfactory.
l 8/15!72 DPC-72-8 Automatic Malfunction of Reactor Feed Pump 1 A Scoop tube position can be checked only 'v. ally at the pump.
and 18 Speed Control System resulted A remote indicator will be provided in the c,ntrol room to I
in unplanned reactor sciam. Reactor provide a continuous check on servo amplifier motor in feed water level dropped below -12 inches pump control syr~cm. Completion expected in fall 1974.
activating the ECCS. Closure of the MSIV placed the Emergency Shutdown Condenser in operation, further droppism water level.
11/2/72 DPC-72-13 No Instability in reactor water level with investigation revealed no equipment failure or malfunction.
Reactor Feed Pump under normal Technical review and/or inspectic
- of the fluid coupling control caused reactor scram due in system and the feed pump speed reatrol system was per-high water level. Similar to incident formed and modifications noted under DPC-72-8 above. Will I
reported in DPC-72-8.
be completed in fall 1974.
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A TABLE 2.1 -
SUMMARY
OF OPERATING HISTORY FOR LACBWR ECCS (Continued)
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ECCS Description f incident Cecrective Measures Required Date Report No.
Operation 4/4/73 Abnormal d.mual Test During annual surveillance test of ECCS, All twject pipe nipples and couplings were replaced with a new I
Occurrence lower than normal differential pressure single piece fitting designed to reduce threaded areas which are Report 73-2 measurement revealed several cracked subject to stress coreosion.
pipe nipples in the ECCS distribution i
I system.
5/23/73 Facility Change No Replace existing scoop tube positioner with a Beck rotary i
73-11 g
activator with integral inCMator and control system. fiarr. ve l
present servo positioner and indicator. This change should l
produce a more linear and wider range of controller span in order to obtain desired feed flow control. Completion expected fall 1974.
7/t3/73 DPC-73-14 No Similar to DPC-72-13 See entry above under DPC-72-13 8/23/73 DPC-73-17 No Similar to DPC-72-13 See entry above under DPC-72-13 8/23/73 DPC-73-18 No Similar to DPC-72-13 See entsy above under DPC-72-13 to b
TABLE 2.2 -
SUMMARY
OF RESULTS OF SEAHANNUAL TESTS FOR LACB;VR ECCS HIGH PRESSURE CORE SPRAY SYSTEM Corrective Measures Date Report No.
Test Results Required 11/2/67 ACNP-67535 Satisfactory None 5/6/68 ACNP-68507 Satisfa tory None 10/23/68 ACNP-68515 Satisfactory Diesel Pump Speed Governor settings corrected 3/6/69 ACNP-69503 Satisfactory None 4/3/70 DPC-851-10 Satisfactory None 5/3/70 EPC-851-11 Satisfactory None 12/70 DPC-851-19 Satisfactory None 8/25/71 DPC-851-33 Satisfactory None 12/12/71 DPC-851-38 Satisfactory See Table 2.3 6/8/72 DPC-851-44 Satisfactory None 1/16/73 DPC-851-54 Satisfactory None 6/18/73 DPC-851-61 Satisfactory None 2-7
r TABLE 2.3 -
SUMMARY
OF RESULTS OF ANNUAL TESTS FOR LACBWR ECCS HIGH PRESSURE CORE SPRAY SYSTEM Corrective Measures -
Date Report No.
Test Results Required 11/20/67 ACNP-67535 Satisfactory None 11/13/68 ACNP-60616 Satisfa;: tory None 4/27/70 DPC-851-10 Satisfactory None 12/15/71 DPC-851-38 Satisfactory Fittin; damaged during
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disassembly of Core Spray Assembly was replaced 6/8/72 DPC 851-44 Satisfactory None 4/4/73 DPC 351-59 Satisfactory None 2-8
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TABLE 2.4 -
SUMMARY
OF RESULTS OF SEMIANNUAL
)
TESTS FOR LACBWR ECCS ALTERNATE (LOW PRESSURE) CORE SPRAY SYSTEM Corrective Measures Date Report No.
Test Results Required 11/67 ACNP-67535 Satisfactory Mane 5/20/68 ACNP-67507 Satisfactory None 10/7/68 ACNP-68515 Satisfactory None 3/6/69 ACNP-69503 Satisfactory None 9/23/69 ACNP-69510 Satisfactory Hone 3/30/70 DPC-851-8 S& sfactory Nor.=
4/70 DPC-851-10 Satisfactory Hone 12/5/70 DPC-851-19 Satisfactory None 12/9/71 DPC-851-38 Satisfactory None l
6/12/72 DPC-851-44 Satisfactory None 2/2/73 DPC-851-55 Satisf& tory None 6/17/73 DPC-851-61 Satisfactory None l
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l TABLE 2.5 -
SUMMARY
OF RESULTS OF ANNUAL TESTS FOR LACBWR ECCS ALTERNATE (LOW PRESSURE) CORE SPRAY SYSTEM Corrective Measures Date Report No.
Test Results Required 11/13/67 ACNP-67535 Satisfactory None 10/7/68 ACNP-68515 Satisfactory Diesel Pump Speed
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Governor settings corrected 4/3/70 DPC-851-10 Satisfactory None 12/9/71 DPC-851-38 Satisfactory None 6/12/72 DE C-851-44 Satisfactory None 4/17/73 DPC-851 -59 Satisfactory None
=4 2-10
From the abcVe record it can be concluded that in eve /y instance when either the High Pressure Core Spray System or the Alternate Core Spray System has been called upon to operate either automatically or manually, the systems have responded satisfactorily to sustain reactor water level.
2.5 EVENTS CAUSING AUTOMATIC OPERATION OF ECCS, DESCRIPTION OF RESULTING THERMAL TRANSIENTS AND THEIR EFFECT ON LACBWR SYSTEMS AND CO2.12ONENTS
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Of all of the incidents listed in Table 2.1 during which ECCS was
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initiated automatically, the one causing the largest thermal transients in the vessel, core, etc., c ucurred on May 15,1970 as a result of a mal-function of the main steam bypass valve during reactor operation at 60% power. After mari.lal reactor scram the reactor water level dropped 2-1/2 feet below the top of the core, and reactor pressur ' decreased from 1280 psig to 200 psig in approximately 7 minutes. Details of this incident-are covered in LACBWR Incident Report No. DPC-70-3.
(q In respor se to requests from AEC-RDT and the Chicago Operations Of' ice following this inc! dent, analyses were performed to evaluate the effects of the in-cident on the LACBWR fuel and vessel These analyses are reported in i
8 detailin Gulf United Report SS-588 and further supported in Gulf United Report SS-591.7 The results of these analyses are' summarized in Table 2.6. The ss calculated stresses for the vessel are within the requirements of Section III, p
CME Boiler and Pressure Vessel Code, Nuclear Vessels, and the calcu-
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A lated maximum temperatures Teached by the fuel and its cladding are 838* F and 828* F, respectively, which are less than the average fuel tem-perature expected in the core (930* F) during full power operation.
1 Based on observed water level fluctuations during the transient, these I
temperatures were calculated for a core which stayed covered for 10 minutes l
after reactor scram during which time the heat was transferred from the fuel rods to saturated coolant with a heat transfer coefficient based on bulk boiling. This was followed by a heatup period of 12 minutes during which 2-1/2 feet of the core was uncovered, and heat transfer was to saturated steam. Afterwards, the normal water level in the core was restored.
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1 TABLE 2.6 -
SUMMARY
OF STRESS ANALYSES ON LACBWR VESSEL AND FUEL
- DUE TO THERMAL TRANSIENT OF MAY 15, 1970 Max Max Max Allowable Appucable Location Temp, F Stress, psi Stresc, psi ASME Code i
i Junction of Closure Head Spherical lj Shell to Flange Region 547 30,782 80,100 Section III J
At Vessel Lamination 547 5,606 No requirement None i
Fuel 638 None Fuel Cladding 828 5,051 Yield strength 19,000 None i
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- As given in Gulf United Report SS-588 i
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- 3. MODIFICATIONS TO IMPROVE THE LACBWR ECCS 3.1 INTRODUCTIGN Since the LACBWR inception, a number of modific.Vions to improve its ECCS have been made or proposed. These are listed in the succeeding sections with a statement of their status.
3.2 MODIFICATIONS COMPLETED TO DATE OR IN PRCGRESS Modifications to the LACBWR ECCS which have been completed to date or in progress are listed in Table 3.1. It may be noted that except for the addition of the alternate core spray system in 1967, all of the modifications which have been made or which are in process are 'n ancill-ary systems which can influence the reliability of the ECCS.
3.3 MODTF1 CATIONS PROPOSED TO COMPLY WI'IH AEC INTERIM CRITERIA In response to an AEC request for reevaluation of the LACBWR ECCS for compliance with the AEC Interim Acceptance Criteria for ECCS, DPC 2
presented analyses which showed that the existing ECCS system was capable of providing adequate cooling for all postulated locs-of-coolant accident situations and which did indeed comply with the interim criteria for ECCS systems adapted by the AEC in June 1971.
However, as a result of the failure mode and effects analysis performed to the specifications of the single failure criterion as defined in IEEE Standard 279, it was concluded that there was a possiblity that single failures of active components or circuit faults could inhibit the function of the ECCS A
(See Section 8 of Reference 2). In order to eliminate th'.s possibility, DPC proposed to make the modifications listed in Table 3.2.
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TABLE 3.1 -
SUMMARY
OF MODIFICATIONS TO IMPROVE LACBWR ECCS COMPLETED TO DATE OR IN PROGRESS i
Date Description of Mdfkation Status u ent Sept.1967 ACNP-67532 Alternate Core Spray System was installed uport
' Completed late 1967 Oct. 1967 ACNP-67534 recommendation by AEC-DRL.
Nov.1967 ACNP-67535 Jan. 1970 DPC-851-3 Check and relief valves were iristalled in Alternate Core Completed May 1970 Spray Syste.n to protect lines from normal reactor operating pressure.
Feb.1970 DPC-70-2 Pressure sensing lines in high pressure service water system Completed Feb.1970 were insulated and relocated away from outside wall to prevent future freezing.
May 1970 DPC-70-3 A high pressure alarm for the hydraulic system which Completed May 197C operates the Turbine Main Eram Bypass Valve was installed in the Control Room. The MSI valve operation was t.nanged to require manual reopen after automatic closure.
Feb.1971 DPC-71-1 Relocate lubricating oil pressure switch for diesel generator Completed Feb.1970 to obtain faster response and therefore quicker starting at diesel generator. Install immersion heater to keep oil fluid during standby in cold weather.
March 1971 DPC-71-7 The following recommended modifications were listed in IR DPC-71-7:
- 1. Cross connect line should be installed between the high
- 1. This modification has been pressure service water system and the low pressure ser-investigated and will not bc vice water system to attw cross feed under certain
- made, emergency conditions. The cross connect valve should c.o k
be manually operated and normally locked closed.
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TABLE 3.1 -
SUMMARY
OF MODIFICATIONS TO IMPROVE LACBWR ECCS COMPLETED TO DATE OR IN PROGRESS (Continued)
Reference I
Date Document Description af Modification Status
- 2. A reliable diesel fuel storage tank level indicator
- 2. Completed March 1971 independent of plant power will be installed.
- 3. The manual control system for the shte:down
- 3. Completed June 1971 condenser will be modified to provide faster response time.
- 4. The feasibility of providing power to the seal
- 4. Under consideration injectico pumps from the Essential Bus will be invesdgated. This is needed to supply continuous seal water flow to the CRD and FCP seals.
- 5. Backup emergency power will be provided to the
- 5. Completed approximately counting room and to other locatior.; so that May1J72 scalers and portable air monitors can be operated during loss of normal power supplies.
Aug.1971 DPC-71-6 Starting motot relays for Diesel Fire Pumps were Completed August 1971 replaced with heavy duty contact type.
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Aug.1972 DPC-72-8 Replace the existieg scoop tube posi<ioner with a Facility Change No. 73-11, Beck rotary actuator with integral position indicator May 23,1973. Completion of and control system to produce a more linear and installation expected by Oct.1974.
wider range of control of feed flow.
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TABLE 3.2 - SUMMAR'l OF PROPOSED MODIMCATIONS TO IMPROVE LACBWR ECCS Reference Date Document Description of Modification Status June 5,1972 LAC-1144 with
- 1. Install a redundant emergency diesel
- 1. Detailed study completed attached reference generator and assoc *ned switchgear.
Sept.1973. Equipment document GU-SS-942
$pecifications have been developed. Following review
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by DPC, bids will be solicited.
- 2. Modify reactor wa;er level safety
- 2. Under consideration.
channel No.1 to eliminate the HPCS pump automatic trip func-tion by a high water level signal.
- 3. Provide separate power supplies
- 3. 12SV A.C. Essential Bus feeds
' for each reactor water level separate power supplies for I
sensing system.
each sensing system. No further changes planned.
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- 4. Eliminate the single HPCS pump 4 Under consideration.
l selector switch.
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- 4. CLAD TEMPERATURES DURING A LOSS-OF-COOLANT ACCIDENT OF THE BIG ROCK POINT REACTOR
4.1 INTRODUCTION
As part of an evaluation of the adequacy of the LACBWR ECCS, a number of analyses of the consequences of a loss-of-coolant accident have been performed.2,3,4 In these analyses, the blowdown following a s
pipe break was modeled by means of the RELAP 3 Code and the rod heatup by means of the ARGUS and TIGER codes.810 2
In order to verify the rod heatup calculations, the AEC has requested
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that the clad inmperatures following a LOCA be calculated for the Type "F" fuel assemblies of the Big Rock Point Plant with the same methods used in the LACBWR los:s-of-coolant analysis and results be compared with those submitted by General Electric to the AEC in Docket 50155-112.5 l
The GE calculations were performed by means of the MOXY code.!!
Gulf United performed the Big Rock Point clad temperature calculations in a n.anner similar to the LACBWR loss-of-coolant analysis and obtained significantly higher clad temperatures than GE, as shown by curve A in Fig. 4.1 To come closer to the GE results, another calculation was performed with a revised calculational model which comes closer to the GE model and whose details are explained in Section 4.3. As shown by curve B of Fig. 4.1, with the revised model, the clad temperatures, although still somewhat higher, were closer to the GE results, at least up to the time where the metal-water reaction became significant. Gulf United calculates a peak cladding tem-perature of 3145 F versus 2700* F for GE.
O 4-1
., _. ~ _... _,.. _.. _.. _ _, _,.., _ _,.. - _ _ _ _ _ _ _ _ _.. _ _. _ _
o l
3500' n
I i
e I
i I
l l
g g
g 3000
~
Gulf Uniterf Revised Model Curve B g
/
I
/
2500
~
/
y
/
Gulf United e'
./
j Original Moctet,
/
Curve A N
j 200D
/
>g GE Rewit:5 j
/
ha j
1500
(
1000 f
I 500 2
4 6 8 10 00 '00 200 400 600 gon gang Time After Break,see M
4.1 - Center Rod Cladding Temperatures During LOCA' S
Big Rock Point Plant
(
4-2
o f
It is therefore concluded that with both the original and. Me revised i
model, Gulf United calculates higher clad temperatures for s Lig Rock l
Point LOCA than GE. It should also be pointed out that the Gulf United calculations for Big Rock Point are based on a shroud wetting time of 92 seconds quoted by GE in the AEC submittal.3 However, according to Gulf United's methods of~ calculating shroud wetting time, presented in the LACBWR loss-of-coolant analysis,2 the shroud wetting time for Big Rock Point would be ~210 seconds rather than % seconds. This represents additional conservatism in the LACBWR analysie.
t 4.2 ANALYSIS BASED ON LACBWR MODEL As in th original LACBWR loss-of-coolant analysis, the core heatup phase was Cvided into two parts. The first part, fuel rod heatup, considers the initial period of the accident when blowdown occurs and the heat transfer coefficient drops from its steady-state value to zero due to the cescation of nucleate boiling. During this period each fuel rod can be analyzed inde-pendently of its neighbors.
The second part of the analpis, fuel assembly heatup, considers the period of the accident during which the fuelis uncovered. Radiative heat transfer between fuel rods, the shroud, and the control rods is of major importance. Analysis is on a fuel assem1'ly basis, taking into account the various radiative heat transfer interactions that occur htween the assembly components, as well as heat transfer to the core spray after initiation of the ECCS.
Basic parameters for the Big Rock Point calculations are given in Table -i.1.
l 1.2.1 Fuel Rod Heatup Analysis 1
As in the LACBWR analysis, the temperatures of the fuel mds of the hottest fuel assembly during the period immediately following the bre::.k till the core uncovers (8 seconds for Big Rock Point) were calcu-lated by means of the ARGUS code.'
Temperatures were calculated for three fuel rods covering the range of local peaking factors, 0.83 to 1.19, encountered in the hot assembly.
Heat transfer *vas considered in the radial direction only. As in the LACBWR cnalysis, the fuel was divided into 11 radial nodes and the cladding into three nc.les. The variation of the surface film coefficient with time was i
4-3 1
.=
\\
s TABLE 4.1 - BIG ROCK POINT LOSS-OF-COOLANT CALCULATIONS Average core power before LOCA, kw/ft 6.18
' Radial peaking factor 1.45 Axial peaking factor 1.51 2
Gap coefficient, Btu /hr-ft _oF 1000 Rod array 9x9 Clad OD, in.
0.5625 Clad ID, in.
0.4825 Pellet OD, in.
0.4715 Heated length, in.
68 Shroud inside dimension, in.
6.5 x 6.5 Shroud thickness, in.
0.10 Time when cora uncovers, sec 8
Time when spray starts, see 16 Time when shroud wets, sec 92 9
0 4-4
.,m,
a 3
taken from Docket 50155-112, Fig. 2 which is reproduced here as Fig. 4.2.
g The coolant was assumed to be at the saturation temperature of 550.5* F.
Initial fuel and cladding temperatures at time zero were based on steady state temperature distributions at full power obtained by means of the ARGUS code. These initial temperaturer., and the temperatures at t = 8 seconds are given in Table 4.2.
4.2.2 Fuel Assembly Heatup Analysis Following the rod heatup period, there is a period from 8 seconds to 16 seconds during which heat transfer to and from the rods is by ra-diaticn only. At 16 seconds, the spray comes on providing additional heat transfer by convection. At 92 seconds, the shroud wets, further improving the heat transfer.
As in the LACBWH analysis, heat transfer calculations were per-formed using the TIGER code except that a metal-water reaction model similar to that used in the MOXY code"was incorporated to account for the additional heat generation due to the Zr-H O reaction. The analysis 2
was performed for the highest power assembly. Heat transfer between fuel assemblies was assumed negligible.
4.2.2.1 Description of Model The model configuration is shown in Fig. 4.3. It is a two dimensional.
representation of the 9 x 9 Type F fuel assemblies of Big Rock Point. It was assumed that axial conduction and radiation could be neglected. The temperature distribution and heat generation in the fuel assembly were assumed to be symmetrical. Thus, one nuJs only to consider one-eighth of the rod bundle. The division of the rod bundle into rod groups numbered from 1 to 15 is shown in Fig. 4.3.
l Each fuel rod was divided into one fuel node and one cladding node.
The shroud and control rods were represented by one internal node each.
Each cladding node was providd with surfa.ce nodes from which convection and radiation connections were made to other surface nodes. The spray was represented by a constant temperattire boundary node with convection heat transfer between this node and the rod surfaces and the shroud. Ra-diation between rods, between each rod and the shroud, and between the shroud and the control rod was also accounted fe-.
s 4-5
O f
105 4
104 I
l Wet Rods
/ etted Channel W
and Rods
'i., 103 l
IE I
B as
.s E
I r.
o
(
l e!! 102 0
1 Channel i
101 Rods (Dry)
~
VI d'
I 100 10 100 1000 Time After Accident, sec 4
Fig. 4.2 -Heat TransferCoefficients for Design Basis Accident i
f TABLE 4.2 - CLADDING AND FUEL TEMPERATURES DURING ROD HEATUP PERIOD Local Rod Peaking Cladding Temperature _
Fuel Temperature Factor Outside, 'F Inside,1 Surface. "F Centarline, "F Initial Rod Temperatures (t = 0) 0.83 564.0 681.0 992.0 2800.0 1.03 566.8 712.2 1098.1 3575.0 1.19 569.3 737.2 1183.0 4120.0 Rod Temperatures at t = 8 seconds 0.83 1093.3 1102.1 1167.6 1840.0 1.03 1257.6 1270.0 1362.1 2407.9 1.19 1404.6 1419.9 1533.4 2815.5 e
4 4
e 4-7
J f
saeouo eeseOOOd O-ee00000 00 ee0000 000 eOOOO OOOOwDOOOO 00000 0000 000000000
, 000000000 sO O O O O O O O Os Fig. 4.3 -Big Rock Point Fuel Assembly -Type F 4-8
w a
- e__,
(.
g The decay heat curve used in the analysis is the one developed for the LACBWR analysis.2 The heat generation rate in each rod was taken to be proportional to the peaking factor of the rods, given in Table 4.3.
The peaking factors were taken from Docket 50155-112.5 4.2.2.2 Initial Temperatures The initial temperatures for the 15 rod groups were obtained from the ARGUS runs at a time of 8 seconds.by averaging the fuel temperaturcs and the clad temperatures to obtain single fuel and clad temperatures for each rod. The initial fuel and clad temperatures for the 15 rod groups are given in Table 4.3.
i 4.2.2.3 Effects of Core Spray The varbtion of rod and shroud heat transfer coefficients and rod and shroud emissivities duri':; the various stages of the heatup period are given in Table 4.4. The heat transfer coefficients between the indi-vidual rod groups and the spray were taken from 71g. 4.2 and are tabu-lated in Table 4.3.
(
4.2.2.4 Radiation Interchange Factors
?r}..
es a The radiation interchznge factors fof the Big Rock Point fuel as-g.,
sembly were caRulated according to th'e methods described in the origi-8 12 nal LACDWR analysis,2 with emissivities specified by GE for Zircaloy clad fuel. These emissivities are listesi in Table 4.4. The calculated interchange factors for dry and wet shrouds are given in Table 4.5 and 4.6, respectively. The rod numbering system is as shown in Fig.' 4.3.
4.2.2.5 Metal-Water Reaction The reaction between zirconium and steam is expressed by the chemical equation:
Zr + 2H O - ZrO + 2H 2
2 2
Following the procedure employed in the MOXY code," it was assumed that an unlimited amount of steam was available. The reaction rate was therefore assumed to follow the parabolic rate. law proposed by Baker and Just:
)t 4-9 i
L._,__-,___.__...,..._____.___________,.__._,_._...._.____,__._____._..__._.__
O I
TABLE 4.3 - BIG ROCK POINT HIGH POWER FUEL ASSEMBLY ROD DATA 4
i.
l Heat i
No. of Local Heat Transfer Initial Temperatures Generation Rod Rods in Peaking Coefficient, (t = 8 seconds)
- Rate, i
Group Assembly Factor Blu/hr-ft - F Clad, *F Fuel, *F Blu/hr-in!
2 1
I i
1 0.83 1.5 1098.
1504.
18,318.
I 2
4 0.96 1.7 1210.
1748.
21,187.
3 4
1.03 1.7 1273.
1885.
22,732.
1
.4 4
0.84 1.7 1105.
1522.
18,538.
D 8
0.91 1.7 1167.
1653.
20,083.
6 4
1.04 1.7 1280.
1897.
22,952.
7 4
1.06 1.7 1298.
1934.
23,394.
i 8
8 1.12 1.7 1350.
2047.
24,718.
S 9
8 0.97 1.7 1220.
1707.
21,408.
10 4
1.10 1.7 1333.
20' d.
24,277.
l 11 4
-1.08 2.0 1317.
1970.
23,835.
j 12 8
1.19 2.0 1413.
2174.
26,263.
13 8
1.08 2.0 1317.
1970.
23,835.
l 14 8
1.19 2.0 1413.
2174.
26,263.
+
550.6 550.6 0.0 l
15 4
0.0 I
I
- From 16 to 92 seconds h " 3.2, after 92 seconds, h = 1002 i
bo j
i
1-g' L
TABLE 4.4 -INPUT PARAMETERS, FUEL ROD ASSEMBLY HEATUP ANALYSIS, DESIGN BASIS ACCIDENT -
BIG ROCK POINT REACTOR Prior to Spray On, Spray On, Spray Dry Shroud Wet Shroud Time, seconds 8 to 16 16 to 92
> 92 Rod heat transfer 2
coefficient, Btu /hr-ft _oF 0
1.5 to 3.2 1.5 to 1000 Shroud heat transfer C
2 coefficient, Btu /hr-ft - F 0
20 1000 Rod emmisivity 0.67 0.67 0.67 Shroud emmisivity 0.67 0.67.
0.96 Spray temperature, F 316 316 4
1 9
4-11
TABLE 4.5 - RADIATION INTERCHANGE FACTORS 3,
}
WITH DRY SHROUD l
tg./
i #
I 4
(g 7
t s
- 3 lp Il Il y
4 y
. i S$ 2:01__.2JL5I ?CIL_ dS 35 t C O.__.112 hC 1_ J.510 _0 L._,1 ts t.91_a5 E t p:ftg., t se 9 -01 pe
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.3532-02
.1914-02
.3395-03
.2917-03
.3442-02 1
7115-01
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.2755-01
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11 L7-01__. 195 2:0 2__,2_7.15-Q2.__. 54 2 3 :02_. 312 9 -02_ _. s t it r o.L. 42 57-0 3__. 4 515t- 02 4113-01
.1544.C0.3748-01
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.1091-01. E 125- 01
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510 L-01.. 13 0 -0.J __.1tD o rQ L.1.3.s 3 -01 1415 -0L _.11 s2 _01.. 15 31.-0 2._ _,1.151-01.
5 J
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.1509-01
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6-a $50-02_.1113-OL_.225270 L. 132L-0L.
10.3S.1h2133-01 113A-0.t _.1_en t3e
.1993-01.3774-02 7115-01
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.7354-02
.3145-02
.4129-01 7
.1995-02.1379-01
.1152-01
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.5052-01
.7258-01
.3043-01
>C
.a 411-41, A E N -02.__,.4 9 31 ?o t. _., S 9 51.-C 1 5 e Zi-C.1_., t n e at_._, a 2 2.3.-0 2_.,4.2.12.-g L 8
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.5527-01.7199-01
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2 75.8-03. 1932.-g2, _35 30 02 33 2.5-C2,_,225 7-01.
4 S,8 3.0L_.3J7 4 -02
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19.1,8 + 00__.
IL 4539-03
.2714-02
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.1s64-0i 9130-01. 7319-01 1223-02
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1.
4-12
l
+
i
~
l TABLE 4.6 - RADIATION INTERCHANGE FACTORS
[',
WITH WET SHROUD 7
a' I
u 6
1 b
b
~,
IC iI IL i
q "b,
- .i g
e t
4
.3867-01
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. tis t -02
.5423-03
.tE25-03
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. 6942-02
.71333 1
.1211*n0
.1544 00
.94E9-01
.1225*00
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.3195-01
.2752-01 7
.1110-01
.1992-02
.2558-02
.5132-02
.292t-02
.1075-02
.2705-03.5028-02,
.4333-01
.1544.C0
.3797-01
.1005+C0
.16570
.5243-01
.22'4T-T1
.2 315-C1 2
.2417-01
.1811-02
.1951-02
.524% 02 4571-02
.1871-02
.4914-01
.? t i t -0 2
.7917-02
.5458-01
.1005*00.253?-C1
.1570+C0
.1606-01
.7313-01
.1013e00
.2517-01
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.3095-02
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. 7C O S-02
.2!43-02
.E425-03
. 427-01 4
. -.1911-nt.
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. 3 t w -0 2
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1r!!S-qL. 112J-ql. 2.1f 9.9 t 1910-02
(
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.3594-01 5
10 AI3E_-f13._ 2555-AZ 1353-02....s09t-0Z_. 2325.41stif 2.QZ s70.EG-01. a35SL-01 C
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(.
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12
~.3504-02
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.2704*00 i
11
.tSC 2 2 -0 4 __m.5 3 7 4 -O f _. S t s e.-Q L. 12 f r -02.._,5 617 -02_.
1ct s: n1_. 2150-0 2_
17.46 :01 C
.54E4-01
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.5754-02
.7753-01'.5492-01
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. 2 97 8-01_,8 721 -O L 2 3 95.-f 2... 5 3 01.-C2.,.18 4 5 -C1_.at *C d? 00...,1371-01._,83021.00 IQ
.3108-03.1529-02
.2003-02
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.1951*00
.19 9 5 +00
.1643*00
.1502NO g
1 4-13
- - -. -.-.~ -
I l.0615' I-41200I 0
I dr dl
- k Rs-r j eXp T
j where r
= radius of the reacting metal-oxide interface (in.)
Rs = initial rod radius (in.)
T
= temperature at the reacting metal-oxide interface (*R) t
= time (sec)
Integrating this equation over a time step and assuming the expression on the right side of the equation is constant over the time step gives:
l-I 2 = 0.123 exp '( 41200 2
DRPI - DRP
' At T
j where DRPI = depth the reactica has penetrated at the end of the time step.
DRP = depth the reaction has penetrated at the beginning of the time step.
At
= time step size.
2 2
By defining the variable AP = DRPI - DRP and rearranging the equation, the mass of zirconium reacted during the time step is given by:
MZR = #p
[2Rs (DRPI-DRP)-AP]
zr Assuming a heat of reaction of 2800 Btu per pound of zirconium '
reacted gives the followin'i for the cladding heat generation:
MWR * (2800) (MZR) 9 A
At n
where An = surface sea of the node 4-14
...w a.
i
(
Time above equations were pregrammed into th'e TIGER code to j
evaluate the amount of heat generated in the cladding. A temperature of 1200* F was assumed as the threshold temperature above which metal-water reaction occurs.
1 4.2.3 Clad Temperature The clad temperature for the central -od of the hcitest fuel assembly of Big Rock Point, calculated by means of the original LACBWR methods, is shown as Curve A in Fig. 4.1 as a function of time. As the temperature was substantially above the one calculated by,GE, the calculation was terminated at t = 92 seconds. New calculations were made with a revised model described in the Section 4.3.
4.3 ANALYSIS WITH ftEVISED MODEL 4.3.1 Description of Model As the rod heatup model described in the previous section resulted' in much higher cladding temperatures for Big Rock Point than those st-culated by GE, it was decided to revise the model to obtain better agree-C ment. Two changes were made. To calculate fuel rod heatup during ths
,first 8 seconds, the TIGER code was usec in place of the ARGUS code.
The particular version, FUEL TIGER, provides temperature-dependent fuel and clad properties, as shown in Table 4.7, which are expected to improve the accuracy of the fuel and clad temperatures at the end of the 8 second period. Secondly, la the TIGER model representing the 15 rods of the hottest fuel assembly the fuel region of each fuel rod was repre-sented by 4 nodes, as shown in Fig. 4.4, rather than by a sinctie node, as in the original model. This corresponds closer to the model used by GE in their LOCA calculations.
In performing the calculations, the initial temperatures for the 15-rod groups were obtained from the 10-fuel node FUEL TIGER runs at a time of 8 seconds. The temperature profiles from these runs were plotted and the tempe*catures for a four fuel node model were obtained for input into the M rod TIGER fuel assembly model. The heat transfer coefficients, radir. tion interchange factors and metal-water heating rates were the same as described in Section 4.2.
i 4-15
I
)
1 k
' i TABLE 4.7 - MATERIAL PROPERTIES FUEL TIGER CODE Mt.TERIAL:
UO q
2
?
Heat Conductivity:i' (T in *F, k in Stu/hr-ft oF) 4 23M 4 = 446 + 0.555T + 0.00735e 1.044 x 104 T 0 < T < 3000* F 6
.k = 1.104 + 0.00735e 1.044 x 10-3 T 3000 < T< 5070o F Heat Capacity:52 (T in aF,C in Stu/lb *F) p p = 0.0726 + 3.33 x 10-8 T - 1467 0.555 T+ 255 C
80< T< 2196= F Cp = 0.0304 + 1.967 x 10-3 T 2196 < T < 5730* F MATERIAL:
Zitcalo; 4 Heat Conductivity:'2 (T in F, k in Stu/hr-ft *F) k = 2.35991 x 10-7 T2 + 4.22414 x 10-3 T+ 6.36497 200 < T < 1900* F 2
k = 2.37524 x 10-s T - 1.70923 x 10-4 T+ 6.78158 1900 < T < 3360 F Heat Capacity: 2 (T in 'F, C ic Stu/lb *F) p p = 0.0753 + 9.44 x 10 6 T - 890 0.555T+ 255 80 < T< 1746' F C
i k
t 4-16 s
--,,,.---...e..-.--..
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T Fuel,4 Radial Nodes
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- i**
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1000 Stu/hr ft.op h.-
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Cladding,1 Node Fig. 4.4 -Revised TIGER Model for Fuel Assembly Heatup
(
4-17
)
r 4.3.2 Clad Temperature The center rod clad temperature calculated with the revised model is shown in Fig. 4.1 as Curve B. Agreement with the GE results is. fairly good till the temperabres are affected by the heat due to the metal-water reaction. While the GE clad temperature reaches a peakat 2700* F, the clad temperature calculatedby Gulf Unitedpeaks at 3145' F. This discrepancy is probably due to a difference in the method of accounting for the metal-water reaction. The metal-water reaction is not important in checidng out the basic LACBWR calculaticnal methods, as Big Rock Point has Zircaloy clad fuel while the LACBWR fuel is clad in stainless steel, where the metal-water reaction is of less significance.
e e
E 4-18
9 r
d' e
- 5. REFERENCES 1.
Letter of Sept. 26,1972 from D. J. Skovholt, Directorate of Licensing, AEC, to J. P. Madgett, Dairyland Power Cooperative.
2.
LAC-1144 - Letter, Madgett, OPC to Block, AEC-DL, dated June 5,1972 with Attachment SS-942. Technical Evaluation - Adequacy of Lacrosse Bolling Water Reactor Emergency Core CocMn3 System - dated May 31,1972.
3.
Review of Densification Effects in Lacrosse Boiling Water Reactor.
Gulf United Report SS-1085 (May 15,1973).
4.
Response to Questions by AEC/DL With Regard to Gulf United Report SS-942, Technical Evaluation, Adequacy of Lacrosse Boiling Water Emergency Core Cooling System, Gulf United Report SS-1075 (April 1973).
5.
Big Rock Point Nuclear Plant, Design Basis Accident, Loss-of-Caolant Analysis for NFS Demonstration Fuel Assemblies, Docket bO155-112 l
(May 18,1972).
6.
Reactor Vessel Stresses, Fuel Temperature and Cladding Stres9 l
Calculations Following Main Steam By-Pass Valve Malfunction, Gulf United Report SS-588 (June 1970).
l 7.
ibid., Addenda to SS-588, Gulf United Report SS-591, (June 1970).
8.
Rettig, W. H. et at: RELAP 3, A Computer Prcgram for Reacter Blowdown Analysis, IN-1321 (June 1970).
9.
Schoeberle, D. F., Heestand, J., and Miller, L. B.: A Method of Calculating Transient Temperatures in a Multiregion, Axisymetric, Cylindrical Configuration. The ARGUS Program,1089/RE 248, written in FORTRAN H, ANL-6654 (Nov.1963).
i 5-1
e
.g-3
,p 10.
Briggs, D. L.: TIGER, Temperatures from Internal Generat!cn Rates, KAPL-M-EC-29, UC-32 (Feb.1,1963).
11.
Evans, D. RJ MOXY: A Digital Computer Code for Core Heat Transfer Analysis, IN-1392; (August 1970).
12.
Silfer, B. C. and Hen'ch, J. E.:
Loss-of-Coolant Accident and Emergency
- Core Cooling Models for General Electric Boiling Water Reactors, NEDO-10329 (April 1971).
13.
Hammer, R. R.: Zircaloy-4, Uranium Oxide and Materials Formed by Their Interaction, a Literature Review With Exploration of Physical Properties to High Temperatures, IN-1093 (TID-4500) (Sept.1967).
14.
SEQUOYAH Nuclear Plant Preliminiary Safety Analysis Report, TVA, Vol.1, 3.2.2-2.
1
'f e
a 5-2
Q6 - e e
4 E
4 APPE. DDC A - LACBWR FACILITY CHANGE NO. 73-11 N
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I A-1
1,,tbVisU"3/20/73
.f
- IACE*R FACILI?( CGNE Faci.'ity Change No.
73-11 l
Date Originated 5-23-73
' Originator W.
Nowicki Subject Systen Rea P b um
. System Fateria[si I.
P..
DPC-72-8, 'ol. III, Ch. 12 Beck Instruction Reference.
k Manual fe
._marv Actuators Reference Drawings:
Attached, 41-503-754
/
Refere.Te Code /s:
A-C 41-650 Electrical. Installation (Acolicable sections)
Descriction of Change:
1 I
Replace the existing scoop tube positionerwith a Beck rotary actuator with integral position indicator and control system.
The present servo positioner and indication will be removed.
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reason for Change:
(
1 To produce a more linear and wider rarge of controller span in order to obtain desired feed fic.i.
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