ML20199L573
| ML20199L573 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear, 07201007  |
| Issue date: | 10/31/1977 |
| From: | Arlotti M, Eicheldinger C, Skaritka J WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20199L511 | List: |
| References | |
| WCAP-8846-A, NUDOCS 9901280059 | |
| Download: ML20199L573 (88) | |
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TECHNICAL LIBRARY ..
PORTLAND GENERAL ELECTRIC COMPAH1
" Westinghouse Non-Proprietary Class III 4
V$ $E!,' ~. l~': '.,7 Do vur ,
HYBRID B C ABSORBER CONTROL 4
ROD EVALUATION REPORT Edited by J. Skaritka October 1977 Work Perfonned Under DGRF 60201 Approved: [
- 54. art 6tti,lanager
,4uel Licensing and Coordination
( Nuclear' Fuel Division I
Approved: .
d C. Eicheldinger, Manager F N
Nuclear Safety PWR Systems Division, Westinghouse Electric Corporation I Nuclear Energy Systems P. O. Box 355 Pittsburgh, Pennsylvania 15230
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Mr. C. Eicneldinger, tianager j Nucledr Safety Department 1 Westinghouse Electric Corporation P. U Box 365 Pittsburgh, Pennsylvania 10230
Dear Mr. Eicneldinger:
SudJECT:
[- EVALUATION OF WCAP-6846 (NON-PROvdlEfARY) 4 "HYdRIO B C ABSORuER CONTROL Ruu EVALVATION xEPORT" SEvitiidER 1970 Ine statt nas completed its review of Westinghouse Electric Corooration f topical report WCAP-oo*u ano the acuitional informatton transmitted oy letters is dated clarch 31, 1977 and ilay 26, 1977. A summary of our evaluation enclosed.
WCAP-db46 cescribes the Westinghouse hyDrid boron carnice full-length rod cluster control assemoly for three-loop and four-loop reactors using 12-foot long fuel assemolies with 17 x 17 fuel roa arrays. The report provides tne design description, design Dases, platerials evaluation, Cesign evaluation, prototype test results, surveillance program, and accident analyses.
As a result of our review, we have concluded that the hybrid boron csrDide absorcer control roa design is acceptable. Tne surveillance program is acceptable proviaed the visual exatnination of selected 0-bank roos at the end of tne third cycle is supplemented oy several control roo reactivity worth checks during the first core fuel cycle and rod wortn measurements of all rod banks at refueling outages. The accioent analyses cemonstrate that the design is likely to provide acceptable results of transients and accicents; nowever, specific acciaent analyses are required for our evaluation on a case by case basis for plants using t,he nyorld Doron carbide rod design.
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L In accordance with estaolished procedures, we request thai; within three months of receiving this letter, you issue a revised version of WCAP-6646 to include this acceptance letter and tne additional information provioed for our review.
( The revised version of this report may be referencea in license applications to acce is support tne p,M bl g con (:usion tnat tne hybrid boron carDice control roa design
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Mr. C. Eiche1dinger, Manager SEP O 21977 b
We do not intend to repeat our review of WCAP-6846 when it appears as t
a reference in a particular license application except to assure that the information plant involved. presented in the report is applicable to the particular If you have any questions concerning our evaluation of WCAP-dd46, please contact us.
Sincerely, 9 id s
Jo in F. Stolz, Chief L ht Water Reactors Branch No.1 I ivision of Project rianagement l l-
Enclosure:
i Topical Report Evaluation i
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, ENCLOSURE g 1
i REVIEW OF TOPICAL REPORT WCAP-8846 Report Identification: WCAP-8846 (Non-Proprietary)
Report
Title:
" Hybrid 84C Absorber Control Rod Evaluation Report" Report Date: September 1976 Originating Organization: Westinghouse Reviewed by: Core Performance Branch and Reactor Systems Branch, DSS Summary of Topical Report The topical report describes the Westinghouse hybrid boron carbide (B4 C) full-length rod cluster control assembly (RCCA) for three- and four-loop 12-foot 17x17 fuel assembly cores. The report covers tne following areas: (1) Design Description, (2) Design Bases. (3) Materials Evaluation, (4) Design Evaluation-Mechanical, Nuclear, Thermal, Hydraulic, Prototypical Performance Tests and Surveillance. and (5) Accident Evaluation.
l l Summary of the Regulatory Evaluation 1
1 We have reviewed the subject topical report including the l
Vestinghouse responses to our questions and the listed references.
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! Reaulatory Position i
We find the hybrid B 4C absorber control rod design acceptable i
and approve the topical report for reference in licensing actions.
l The extent of our review and basis for our conclusions are given in the enclosure. A routine surveillance program will, however, be required of all licensees wh'o use these rods because of the water- l l soluble nature of irradiated 5 4C.
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ENCLOSURE - WCAP 8846 EVALUATION a
l Materials and Mechanical-Thermal D e s i r, n Evaluation l
1 The full length hybrid absorber rod proposed by West-inghouse is composed of B4 C pellets and Ag-In-Cd cylinders encapsulated in 304 stainless steel cladding. Within the cl a d d,in g , the silver alloy rods occupy the lower 20 to 40 inches with the B4 C peliets stacked above to give a total absorber column of 142 inches. The column is positioned in the lower part of the absorber rod by an Inconel 718 coil spring.
The external dimensions of the hybrid rod are identical to those for current 17x17 assembly rods composed wholly of A g'-I n- Cd in 304 SS cladding.
Westinghouse experience with stainless steel cladding is ,
both extensive and nearly fault free. Three early PWRs, Indian Point Unit 1, San Onofre Unit 1 and Haddam Neck, were designed and operated with fuel rods which had 304ss cladding. l The fuel performance of these cores over the past 8 year period has shown an approximate failure of 10 to 15 fuel rods out of I a total of %100,000 rods. In addition, Westinghouse has fabri-l cated and operated approximately 20,000 304ss clad control rods of the Ag-In-Cd design. A number of these (s1600) have operated fo.r ten years with no apparent difficulties or failures.
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. l The only difference between current.Westinghou,so control rods and the proposed hybrid control rod is the use of 84 C pellets.
Thus, our review of the proposed design a Mressed primarily this new material. Boron carbide (B4 C) has been in use in other reactor Its designs for control and shielding purposes for many years.
unirradiated and irradiated physical and chemical properties have been extensively studied for both thermal and fast reactor applica-that, for tion. The major differences between B 4C and Ag-In-Cd are 5 C, (a) the melting point is higher by a factor of 3 (b) the 4
irradiation swelling is higher by a factor of 4 (c) the coefficient of thermal expansion is lower by a factor of 5, (d) the thermal con-ductivity is lower by a factor of 10 (e) the solubility in the f a reactor coolant is higher and (f), unlike with the silver alloy, gaseous product (helium) is generated by the neutron reaction and a portion of this gas is released to the control rod plenum.
Westinghouse discusses all of these material properties for 5 4C and appropriately accommodates them into their proposed hybrid design.
A standard 1-D heat transfer calculation conservatively predicts a maximum temperature for 54 C of less than 1200*F which provides a large margin to melt (4400'F) and for contact with As-In-Cd f inves-(m.p. = 1454*F). Compatibility of 8 C4 and Ag-In-Cd has been to be acceptab.1e at these tempera-tigated by Westinghouse and found tures., Should unexpected compatibility problems arise, they would The higher be detected in the surveillance program described below.
irradiation swelling rate is accommodated in the design by increasing f
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the pellet-to-cladding gap. The lower thermal expansion for the absorber, on the other hand, reduces the potential for pellet /
cladding interaction, and, in this case, the cladding actually expands away from the absorber pellets. Hard contact between the pellets and the cladding is, therefore, not expected during the lifetime of the control rod. The chemical compatibility between BC 4 and 304ss has been demonstrated to be acceptable based on the successful operating experience in GE BWRs where B4 C granules are constantly in (loose) contact with the 304ss cladding. For the gaseous product (helium), the predicted release fraction of 30% is acceptably conservative based on the referenced studies. The end-of-life internal pressure due to helium release for the bounding or worst-case irradiation conditions (Mode B) is predicted to be less than the coolant system pressure. Thus, the dimensional para-meters for the cladding during plant operation will not be affected
{ by the thermal conductivity, the irradiation swelling, materials compatibility, or the helium release of the B C.4 Although the probability of cladding failure appears to be quite low, we have considered the behavior of the hybrid control rod in the postulated case of cladding perforation or failure.
Slightly irradiated B C 4 has demonstrated a higher solubility rate in water than for the unirradiated material. This fact was illus-trated recently when a manufacturing error led to the failure of the Zircaloy cladding of many A1 23 0 -B 4C burnable poison rods in St. Lucie' Unit 1. Boron from the dissolution of slightly irradiated B C4 was h
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redistributed within and partially lost from the fa,iled rods.
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The total boron loss was less than 10% during the approximate 2-month
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period the failed rods were in water. The resulting change in core nuclear characterir. tics was detected by in-core instrumentation in the St. Lucie plant, and corrective action was taken before safety margins were significantly affected. Many of the Westing-house hybrid absorber rods, however, will be in safety banks that
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are held in reserve for emergency situations; they are not normally used and therefore their reactivity worth is not normally observed.
Although the Westinghouse design seems to account for everything possible to preclude failure, it seems unwise to allow safety rods containing soluble poison to remain in the reactor coolant for years
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without checking their integrity or reactivity worth. Consequently, as a routine matter, licensees who use B4 C control rods will be asked to submit plans for surveillance designed to assure that the
[ reactivity invested in the control roda is not being lost through some common mode failure mechanism. An acceptable program might include several rod reactivity checks during the first core cycle and worth measurements of all the rod banks at refueling outages thereafter.
In addition, for these initial plants using the new hybrid absorber rods, Westinghouse has agreed to perform a visual surveil-lance program. Selected D bank rods will be examined at the end of ,
their third cycle in this program. The underwater television or ,
binocular scans of these rods should establish any abnormalities,
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such as cladding defects, crud buildup, wear or diametral growth.
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. Based on the review of 5 4 C material properties (unirradiated
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and irradiated), the mechanical and thermal design considerations and the details of the in-reactor surveillance program, we find the hybrid boron carbide (B4 c) full-length absorber rod design acceptable.
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Tuclear Design Evaluation Westinghouse has established design criteria for the
,( design of hybrid control rods which assure that the reactivity worth of the rods is at lesst as great throughout life'as the presently used Ag-In-Cd rods, but not more than 15% greater than Ag-In-Cd rods. These criteria assure preservation of minimum shutdown margins, design basis radial peaking factors, maximum controlled and uncontrolled (ejected) reactivity in-sertion races.
6 The topical report indicates that the hybrid rods are
( calculated to have a worth exceeding that of Ag-In-Cd r.ods.
In response to our questions, Westinghouse has satisfactorily provided the experimental basis to support these calculational results. In addition, startup test requirements in first cycle and reload cores will provide further verification that the
[ hybrid control rods have adequate reactivity worth. We routinely keep abreast of these test results. We, therefore, conclude the nuclear design of the hybrid cuntrol rods is acceptable , .
for use in reactors.
Accident Analysis Control rod drop tines for hybrid control rods exhibit an increased sensitivity to flow rate due to the reduced B C absorber 4
weight. For the hybrid control rod design, Westinghouse proposes .
to use a scram time appropriate to flow rate assumed in analyses
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for specific accidents which are sensitive to variations in control r' .g.
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, rod drop time. Traditionally, the slowest scram time cor- ;
I responding to the mechanical design flow rate has been used in all accident analyses for the Ag-In-Cd control rod design.
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The report presents results of transients and accidents for l
a typical four-loop plant showing required hybrid control rod drop l
time for acceptable consequences are greater than expected l
l (technical specification) drop times. This enables us to make a generic finding of acceptability of the hybrid control rods in the area of accident analysis. However, specific analyses '
will continue to be presented and evaluated on a case by case
( basis for reactors using the hybrid control rod design.
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,~ l TECHNICAt LIBRARY PORTLAND GENERAL ELECTRIC COMPANY j Westinghouse Non-Proprietary Class III e
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HYBRID B C ABSORBER CONTROL 4
R00 EVALUATION REPORT 1 i
l Edited by i J. Skaritka
- j October 1977 l Work Performed Under DGRF 60201 l
1 Approved:
- 54. art 6tti
[f ,~Panager
,'7uel Licensing and Coordination
( Nuclear Fuel Division Approved: .
C. Eicheldinger Manager r N
Nuclear Safety PWR Systems Division, i
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Westinghouse Electric Corporation Nuclear Energy Systems P. O. Box 355 Pittsburgh, Pennsylvania 15230 l
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TABLE OF CONTENTS Iltle Page Table of Contents i fit List of Tables List of Figures 19 1-1
1.0 INTRODUCTION
AND
SUMMARY
1-1 1.1 Introduction 1.2 Sumary 1-1 2-1 2.0 DESIGN DESCRIPTION 3-1 3.0 DESIGN BASES 4-1 4.0 MATERIALS EVALUATION Irradiation of Boron Carbide 4-1 4.1 4-1 4.1.1 Helium Generation and Release 4.1.2 Density Changes 4-1 4.1.3 Mechanical Integrity of B 4C Pellets 4-2 4.1.4 - Tritium Generation and Release 4-2 4.2 Physical and Chemical Properties of B 4C 4-2 4.2.1 Mechanical Properties 4-2 4.2.2 Porosity of Sintered Pellets 4-3 4.2.3 Chemical Compatibility 4-3 4.2.4 B C Thermal Conductivity 4-4 4
4-4 4.2.5 Corrosion in Water 4-5 4.3 References 5-1 5.0 DESIGN EVALUATION 5-1 5.1 Mechanical Evaluation 5.1.1 Thennal Expansion and B4C Swelling 5-1 5-1 5.1.2 Rod Internal Pressure 5.1.3 Spring Characteristics 5-1 5.1.4 Control Rod Drive Mechanism Simulated Stepping Test 5-2 5-2 5.2 Nuclear Evaluation 5-3 5.2.1 Fluence Limitations i
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I l _ TABLE OF CONTENTS (continued)
Iitig Pggg 5.2.2 Load Follow Considerations 5-5 5.2.2.1 Allowed Depletion 5-5 5.2.2.2 Depletion in Withdrawn Banks 5-5 5.2.2.3 Depletion in Banks D and C 5-6 5.2.3 Helium and Tritium Generation 5-8 5.2.4 Reactivity Worth 5-10 5.3 Thermal and Hydraulic Evaluation 5-11
- 5. 3.1 Maximum Absorber Temperature 5-11 5.3.2 Dypass Flow 5-12 5.4 0 Loop Performance of Prototypical 17x17 Hybrid B C4RCCA 5-12
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5.4.1 Test Conditions 5-12 !
5.4.2 Determination of Control Rod Drop Time 5-13 5.4.3 Control Rod Stall Tests 5-13 5.5 Surveillance Program 5-13 5.6 References 5-14 )
6.0 ACCIDENT EVALUATION 6-1 6.1 General 6-1 6.2 Incidents with Minimal Consequences Due to Variations in 6-2 Control Rod Drop Time 6.3 Loss of Flow 6-2 6.4 Underfrequency 6-4 6.5 Rod Ejection 6-5 l 6.6 Verification of In Plant Rod Drop Time 6-6 6.7 References 6-7 APPENDIX A - AEC CORRESPONDENCE A.1 Letter from J. F. Stolz to C. Eicheldinger, March 10,1977 Request for Additional Information A.2 Letter from C. Eicheldinger to J. F. Stolz, March 31, 1977 Response l to March 10, 1977 Letter A.3 Letter from C. Eicheldinger to J. F. Stolz, Septenber 26, 1977, j l
Aeplacement Response to Q-232.2 (WCAP-8846)
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l LIST OF TABLES ;
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2-1 Typical Parameters; Hybrid Full Length B4 C Rod Cluster Control 2-2 Assen61y 5-1 17x17 Hybrid 84C RCCA Drop Time versus now Rate; D Loop Test . 5-15 Results Reference Plant Characteristics: Loss of Row /Underfrequency 6-8 L
6-1 Initia1 ' Conditions Parameters Used and Results for the Ejected Rod Cluster 6-9 ;
6-2 control Assen61y Analysis (Mechanical Design Row Initial l Conditions) i i
Parameters Used and Pesults for the Ejected Rod Cluster 6-10 l
6-3 Control Assen61y Analysis (Thennal Design Row Initial l
Conditions) l l
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LIST OF FIGURES
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2-1 Hybrid Full Length Absorber Rod 2-3
( 4-1 Themal Conductivity of Unirradiated Boron Carbide, 70% to 4-7 i 75% Theoretical Density Pellets 4-2 Irradiated Boron Carbide Themal Conductivity 4-8
{ 6-1 Hybrid Rod Position Versus Time 6-11 r 6-2 Trip Reactivity Insertion Versus Rod Position 6-12
[ 6-3 Complete Loss of Flowp Nuclear Power Versus Time 6-13 6-4 Complete Loss of Flow; Heat Flux Versus Time 6-14
.( 6-5 Complete Loss of Flow; Core Flow Versus Time 6-15 6-6 Complete Loss of Flow; DNBR Versus Time 6-16 6-7 Underfrequency; Nuclear Power Versus Time 6-17
{ 6-8 Underfrequency; Heat Flux Versus Time 6-18 6-9 Underfrequency; Core Flow Versus Time 6 19 6-10 Underfrequency; DNBR Versus Time 6-20
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1.0 INTRODUCTION
AND SUt94ARY
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l.1 Introduction
( This report evaluates the planned introduction of a Westinghouse hybrid bcron carbide (B4 C) full-length rod cluster control assembly (RCCA) into three and four loop 12' 17x17 fuel assemblies cores.
The hybrid RCCA consists of 24 hybrid absorber rods, each containing f silver-indium-cadmium (Ag-In-Cd) absorber and 84 C absorber pellets stacked on top of the Ag-In-Cd. Except for the substitution of B C pellets for Ag-In-Cd absorber length, the hybrid RCCA hardware
( 4 remains unchanged from the 100% Ag-In-Cd full-length RCCA presently being used. The part-length RCCA remains unchanged with 100% Ag-In-Cd absorber rods. The hybrid RCCA is designed primarily in response to the projected limited future availability of silver and indium.
In order to satisfy control rod drop time requirements, the hybrid RCCA plus its associated drive red is specified to weigh a minimum
[ of 211 pounds for the present fuel assenbly design. Using the drive rods currently available. 40" of Ag-In-Cd rod length is required to satisfy the design minimum weight. When available,
{ a heavier weight drive rod design can be substituted and will permit decreased Ag-In-Cd lengths to as little as 20".
1.2 Sun,ima,ry
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Based on the materials, design and safety evaluations in this report, it is concluded that the hybrid B C4RCCA satisfies the performance
{ and safety requirements for three and four loop 12',17x17 fuel assembly cores. These requirements are stated in this report (Section
- 3) and Safety Analysis Reports which address reactors using either the 100% Ag-In-Cd RCCAs or the hybrid B C4 RCCAs.
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l 1.2 Summary (continued) l Section 4 evaluates the physical, chemical and irradiation properties of boron carbide. Conservative values of these properties are utilized in evaluating the design and perfonnance requirements of the hybrid RCCA and the chemical compatibility of boron carbide.
Section 5 (Design Evaluation) considers the mechanical, nuclear j and thennal hydraulic performance of the hybrid RCCAs during its planned usage under reactor conditions. During the most severe load follow conditions planned, the design bases (Section 3) are I satisfied, and the hybrid rod cladding integrity is maintained for the design lifetime. Acceptable amounts of 84C pellet swelling and helium generation art satisfied by limited 8 C 4 radiation exposure.
l Test results are presented to confirm the adequacy of the hybrid
! RCCA drop (scram) times.
l Section 6 (Accident Evaluation) shows that the present SAR safety limits are satisfied for the hybrid RCCA design with at least a l
211 pound drive line weight. The most limiting accidents of loss of flow, under frequency and rod ejection receive detailed evaluations.
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]I 2.0 DESIGN DESCRIPTION _
, A full length hybrid absorber rod is shown in Figure 2-1. These rods are pemanently attached to the RCCA spider assemblies and are raised and lowered within fuel assembly guide thimbles by RCCA motion, as required to control core reactivity.
The approximately 142" of absorber material used in the hybrid rod is composed of 4B C and Ag-In-Cd, with4B C pellets stacked in the top portion and Ag-In-Cd solid rod placed in the lower portion of the rod. The absorber material is stacked into a 10% cold worked 304 stainless steel tubing. The length of the Ag-In-Cd absorber rod is fixed to assure I at least a 211 pound minimum drive line weight (RCCA plus drive rod).
With drive rods in current use, 40" of Ag-In-Cd rod length is required.
A future heavier drive rod weight would pemit decreased Ag-In-Cd length to as low as 20".
I Sufficient diametral and end clearance is provided to accommodate relative thermal expansions and irradiation induced swelling of the absorber materials. The absorber materials am positioned in the lower part of the absorber rod by an Inconel 718 coil spring. This spring is designed to provide a holddown force on the absorber stack so that the absorber pellet stack will not move and generate gaps during transport, handling and operation.
Except for the 84 C pellets, the hybrid absorber rod is fabricated in I the same manner as the 100% Ag-In-Cd absorber rod presently used. Design and fabrication details on the remaining portions of the hybrid RCCA are the same as its associated 100% Ag-In-Cd RCCA which is given in Section 4.2.3.2.1 of RESAR 3S.
Typical parameters of the hybrid RCCA are given in ' sable 2.1.
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Table 2-1 TYPICAL PARAMETERS HYBRID FULL LENGTH B C4 R0D CLUSTER CONTROL ASSEMBLY 12',17 x 17 Three 809_Fgut,Lggg,P{ ants l 1
Absorber, Material B4 C +'Ag-In-Cd Length of Absorber Material, inches 142 (total stack) l Minimum Weight RCCA + Drive Rod, pounds 211 Length of Ag-In-Cd Rod, inches 20 to 40 (drive rod l weight-depend.ent, 40 with present drive rod j design) l B C Pellet Diameter, inches 0.334 l 4
B4 C Pellet Density, pounds / inches 3 0.0637 (nominal 70%
of theoretical)
Composition of Ag-In-Cd 80% Ag,15% In, 5% Cd Diameter of Ag-In-Cd, inches 0.341 3
Density of Ag-In-Cd, pounds / inches 0.367 Claddigg, Material 304 Stainless Steel, Cold Werked Cladding Thickness, inches 0,0185 Number of Full Length RCCAs 53 (four loop) 48(threeloop)
Number of Bank D Full Length RCCAs 9(fourloop) 8 (three loop)
Number of Absorber Rods per RCCA 24 2-2 l
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52 A50$RBER MATERIAL y q
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3.0 DESIGN BASES The following are considered design conditions under Article NB-3000 of the ASME Boiler and Pressure Vessel Code,Section III. The control rod clad, cold worked Type 304 stainless steel tubing, is the only non code material used in the control rod assembly. The stress intensity limit Sm for this material is defined at 2/3 of the 0.2% offset yield stress.
- 1. The external pressure equal to the Reactor Coolant System operating pressure with appropriate allowance for overpressum transients.
- 2. The wear allowance equivalent to 1,000 reactor trips.
- 3. Bending of the rod due to a misalignment in the guide tube.
- 4. Forces imposed on the rods during rod drop.
- 5. Loads caused by accelerations imposed by the control rod drive mechanisms.
- 6. Radiation exposure for maximum core life.
A hybrid RCCA design life of fifteen years minimum was asstaned for purposes of evaluation.
In addition, the following design criteria have been established to assure that perfonnance and safety requirements will be satisfied.
Criterie.1 Throughout the design life the reactivity worth of a full length RCCA containing 84 C shall be at least as great as the reactivity worth of the analogous RCCA containing all Ag-In-Cd.
Basis The excess over the design basis minimum shutdown margin and the controlled reactivity insertion rates during load follow, startup, and shutdown must not be degraded by the B4 C substitution.
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3.0 DESIGN BASIS _ (continued)
Critgr!gg,@
Control rod drop time shall be sufficient to assure that accident limits are not exceeded.
Basis Reactivity must be controlled rapidly enough under all accident or abnomai conditions in order to prevent reactor safety limits from being exceeded.
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3.0 DESIGN BASES (continued)
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Throughout the design life the reactivity worth of a full length RCCA containing B4C shall not be more than 15% greater than for the analogous RCCA containing all Ag-In-Cd.
Basis The Design basis radial peaking factors and maximum controlled reactivity insertion rate must not be exceeded by the 4B C substitution. Higher worth increases these values. The ejected rod accident also becomes more severe with higher rod worths.
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The standard load following maneuvers using Ag-In-Cd rods (i.e., constant axial offset control) should be available when 84 C is substituted.
Basis Substitution of B C 4
for Ag-In-Cd must not degrade plant perfonnance.
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The maximum temperature of the absorber materials shall not exceed its melting temperature (1454*F for Ag-In-Cd and 4400*F for B4 C).*
Basis l
l Melting of the material would lead to loss of dimensional control, filling of the clad-absorber gap and possible clad-absorber interaction.
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- The melting point basis is detennined by the nominal material melting point minus uncertainty.
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4.0 MATERIALS EVALUATION 4.1 Irradiatign,gf,Borgn,Carbi,de 4.1.1 Helium Generation and Release The generation of helium in B4C occurs due to the B-10 (n.
a) Li-7 reaction, and helium generation can be equated to B-10 depletion. The radial distribution of helium generation is peaked strongly toward the surface of the pellet.
For the nominal 70% T.D. 84 C pellets the design uses a helium gas release of 30%. This represents an upper bound gas release for temperatures less than 1200'F and for all burnups conside red. The gas release information was obtained from References 2, 4, 5, 7, 8, 9,11,12,13 and 14. The hottest B C pellet has a maximum temperature of less than 1200'F 4
(see Section 5.3.1). .
4.1.2 Density Changes For 70% T.D. B 4 C pellets irradiated under thermal reactor based on the references conditions, below, is 0.15% a conservative AV/V per 10 2 atoms B /cmC swelling, depletion up to 3
approximately 95'x 10 20 depletions /cm . At higher burnups, microcracking causes a significant increase in the swelling rate and the recommended swelling is % aV/V = .468 D - 30.2 where D is in depletions x 10 20 Data on irradiation induced swelling of B 4C were obtained from References 1, 2, 3, 4, 6, 7, 9,10,11 and 13.
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4.1.3 Mechanical Integrity of B C Pellets 4
B4 C pellets for the Westinghouse design are expected to 20 develop macrocracks up to burnups of 7 x 10 depletion /
cm3 . The causes of macrocracking are the stresses introduced by both irradiation damage and thermal expansion. Macrocracking of the pellets does not adversely af fect the B4 C stack distribution.
Information on the pellet mechanical integrity were obtained from References 7, 8, 9 and ll.
4.1.4 Tritium Generation and Release l
The tritium production in B4 C occurs initially from the B-U (n, a) a. T reaction. Cross sections are negligible below 1 Mev, and the epithermal flux weighted average is 12 millibarns. Additional contribution to tritium production occurs from the Li-7 (n, n') a, T reaction which has a 2.81 Mev threshold and 9.6 millibarn cross section. The amount of teitium generated during the rod life is evaluated in Section 5.2.3.
Limited data for release rates of tritium from 84C is available.
Since tritium does diffuse more rapidly than helium, the conservative design assumes 100% release rates of tritium from B C into the pellet clad gap and a 100% release rate 4
through the clad into the coolant (Reference 16).
4.2 Phys i ca l ,and ,Ch emi c a l ,P rrjge rti es ,o f,84g 4.2.1 Mechanical Properties _
The B C pellets are fabricated to exceed a minimum axial 4
compressive strength of 25,000 psi at room temperature.
This minimizes pellet powdering.
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_ _ _ - - . . _ - -. -. . - - . . _ . - . . _ _ - . . _ _ _ - -- ~~
I l
4.2.2 Porosity of Sintered Pellets _ l The total porosity of the B 4C pellet is maintained within 30% + 2%. Measurments from each of four potential pellet suppliers show more than 98% being open porosity with the remainder closed porosity. l l
1 Of the total void space within a rod, 92% is represented I by the open porosity in the 70% B 4C pellets and only 6%
by the plenum and annular gap. Therefore, it is important to verify that the voids in the pellets are actually open to store generated helium. Results of gas displacement measurements confirm that from 98.1% to 99.4% of available void volume of pellets is interconnected open porosity.
4.2.3 Chemical Compatibility _
A negligible interaction rate is predicted between 8 4C pellets in contact with the 304 stainless steel clad. References 6 and 12 provide equations to detennine the interaction rate in mils of SS/ year. For the hottest B 4C pellet under the most severe conditions (full power, load follow, beginning I
of life), a reaction rate of less than 0.1 mils / year is calculated. Based on a maximum of 12 Effective Full Power Years in Bank D under load follow conditions, the total chemical interaction is expected to be less than 1.5 mils of SS.
I No interaction between the Ag-In-Cd rod end and contacting B C pellet end is expected. This is based on the Reference 4
10 observations of no interaction between Ag and B4 C in contact near the melting point of Ag (about 1760*F).
l 4-3 4
I _ _
[
[
4.2.4 B C Thennal Conductivity .
The thennal conductivity for B4 C pellets is a function of density, temperatum and bumup. Out of pile thennal conductivity measured values (References 17 and 18) for 70% T.D. 8 4C are approximated by a linear fit for temperatures ranging
[
from O'F to 1900"F, as shown in Figure 4-1. The observed decrease in thennal conductivity with buniup saturates at
{ approximately 5 x 10 20 captures per cm . The irradiated 3
values (Reference 18), relative to their out of pile values, for a ser'es of temperatures are shown in Figure 4-2.
4.2.5 Corrosion in Water Although unlikely, it is conceivable that an occasional hybrid rod will have its cladding barrier breached, even though the rods are designed to minimize this occurrence.
In pile corrosion data9indicates that B C 4 under load follow operating conditions could have an irradiation enhanced
( corrosion rate up to about twenty mils per month per surface (worst case data plus conservative margins). Therefore, it is conservatively assumed that B4 C pellet disintegration
{ could occur within one year after a postulated cladding breach. Postulated 8 C washout through a cladding breach 4
of 'a few isolated hybrid absorber rods would not cause a violation of the design bases (Section 3).
Except for the hybrid D bank control rods, B4 C pellets are essentially in an out of pile environment. Out of pile data shows B C corrosion rates more than an order of magnitude 4
less than the worst case assumed in pile information. During /
verification testing in Westinghouse D loops, two hybrid rods had 0.031" diameter holes drilled through the cladding
( to simulate defects. There was no degradation of the pellets directly opposite the holes after about 880 test hours in flowing 585*F water.
[ ,
4-4
4.3 References
- 1. A Jostons, et. al. " Defect Structure of Neutron Irradiated Boron
. Carbide", J. of Nuclear Materials 14 (1973/1974), page 136 through page 150.
~
- 2. F. J. Homan, " Performance Modeling of Neutron Absorbers", Nuclear Technology, Volume 16, October 1972.
- 3. G. L. Copeland, et. al. , " Irradiation Behavior of Boron Carbide",
CONF-720420-1.
~
- 4. J. A. 3asmajian, et. al. , " Irradiation Effects in Boron Carbide Pellets Irradiated in Fast Flux Spectra". Nuclear Technology, Volume
] 5.
16, October 1972.
C. R. Mefford and H. E. Williamsom, "The Performance of Boron Containing l Control Rods in Water Cooled Power Reactors", Nuclear Applications, Volume 4, June 1969. 1 R. E. Dahl, "Special Topic Presentation to Thirty-Second High Temperature I 6.
Fuels Committee Meeting Control Materials Development Program". HEDL-SA-206, May 12, 1971.
]
1
)
- 7. A. L. Pitner and G. E. Rasscher, " Irradiation of Boron Carbide Pellets !
' and Powers in Hanford Thermal Reactors", WHAN-FR-24, December 1970 I E. W. Hoyt, D. L. ZimerNan, " Radiation Effects in Borides Part I ~
8.
I", GEAP 3743, February 13, 1962.
)
- 9. R. G. Gray and L. R. Lynam, " Irradiation Behavior of Bulk B C and
~
B C + sic Burnable Poison Plates", WAPD-261, October 1963. 4 4
L 10. E. W. Hoyt, "Information on the Use of Boron Carbide as a Nuclear Control and Poison Material", GEAP 3680, March 15,1961.
- 11. T. W. Evans, "The Effects of Irradiation on Boron Carbide", BNWL-679, February 1968.
- 12. R. E. Dahl, " Boron Carbide Development for FFTF Control Elements".
HEDL-SA-565, April 1973.
- 13. B. Weidenbaum, et. al., " Properties of Some High Temperature Control Materials". Chapter 14 of unknown United States Atomic Energy Comission at(04-3)-189, P. A. Number 4 work.
1 I - 14. G. L. Copeland, et. al. , " Evaluation of Fast Reactor Irradiated Boron Carbide Powders", ORNL-TM-3729.
- 15. W. K. Anderson, J. S. Theilacker, " Neutron Absorber Materials for j
Reactor Control",1962.
- 16. J. H. Austin, et. al. , " Surface Effects on the Diffusion of Tritium in 304 Stainless Steel and Zircaloy 2", J. of Nuclear Materials 48(1973).
l, 4-5 I
4.3 References (continued) ,
- 17. H. W. Deem, C. F. Lucks, " Thermal Conductivity of Boron Carbide i from 100*C to 800*C", BMI-713 December 10, 1951.
- 18. D. E. Mahagin, et. al. , " Boron Carbide Thennal Conductivity". HEDL-TME 73-78. UC-79 b. September 1973.
I I
I I .
I I
I I
I I
I
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[
4-6 T.
u
10 ,-
9- -
- 8. .
7--
y
- TC 6--
0 3*
EC e 5- -
e ce
- 8k E
4- ~
~
ge 3- -
3 w
2-1 1 : ; ; '
0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 1400 1500 1600 1700 1800 1900 Temperature (*F) i Figure 4-1 Thennal Conductivity of Unirradiated Boron Carbide, 70% to 75% Theoretical Density Pellets V
s i i i 6
1.0 n
a \
~
> \
\ ;
b \ U i E \ l a * \ O >__ _ _ T gn=870 i C (15?8"F)
& ,E N
" W--- T gg-750 i C (1382*F)
$ ~7vg _
% ~ ~ _ _ _ _ _A _ _ _ _ _ . Ti g g 600 C (Ill2*F) g e
N OTIRR=500 C ( 932*F)
I .
'a e i.
- x 0 10 15 20 O 5 20 Captures /cm3 ) ;
Burnup (10 t
t I
'igure 4-2 Jrradiated Boron Carbide Thermal Conductivity i
_- ~. -_- . _ _ _ -
i 1
5.0 DESIGN EVALUATION _
i l
5.1 Mchani, cal _ Evaluation 5.1.1 Themal Expansion and B4 C Swelling l 4
i The rod accomodates relative thermal expansions and irradiation induced swelling in both the radial and axial directions.
A radial gap is provided at the Ag-In-Cd/ clad and B4 C pellet /
clad interfaces for the worst case conditions of reactor operating temperature and maximum irradiation exposure of the B C pellets. The B 4C swelling rates of Section 4.1 4
and the burnups of Section 5.2.2.1 are used. Similarly, relative i
themal expansions and irradiation induced swelling is considered in the axial direction. An axial gap is maintained at the top portion of the rod. Sufficient gap is provided to prevent ;
the hold down spring from being fully compressed, i
1 5.1.2 Rod Internal Pressure _
Sufficient void volume is provided within the rod for helium and tritium gas generation. The available volume consists of the open porosity of the B 4C pellets, the absorber material /
' clad gap, and the void volume at the top portion which houses the spring. Whil'e considering a 30% and 100% release rate l
for helium and tritium respectively, the clad stresses from the resultant internal pressure are within the clad design bases of Section 3.0.
5.1.3 Spring Characteristics The absorber materials are held down by an Inconel 718 coil spring. The spring deflection accomodates thermal expansion l
i of the various rod components and irradiation induced growth of the B C pellet stack in the axial direction. The spring 4
hold down force exceeds the separation force on the absorber materials from the CRDM stepping loads.
! 5-1 l
l l ._.
5.1.4 Control Rod Drive Mechanism Simulated Stepping Test The simulated CRDM stepping test objective was to verify the mechanical integrity of the B4C peHets. Spec h ns,
. which were fabricated with Ag-In-Cd rod ends, were tested with an equivalent CRDM step loading. The rod specimens were radiographed at various test intervals to assess the pellet condition as testing commenced. At 300,000 steps no cracks or dimensional changes were observed from the radiographs. All specimens were tested to an accumulated 625,000 CRDM steps, the equivalent of fifteen years reactor operation.
Subsequent radiography of the specimens indicated the existence of cracks / chips in several pellets in one specimen. All rod specimens were then cut open at the bottom end plug for examination. One specimen with an intentional zero spring load on the pellet stack revealed two cracked pellets, some chipped pieces and some powdering effect on the pellets.
The remainder of tae 8 C pellets in the specimen were intact.
4 No cracks, chips, powdering or dimensional changes were found on the other specimens due to the equivalent CRDM loading on pellet stacks with a nominal spring hold down force.
5.2 Nycl, ear, Evaluation It is shown in the following sections that the parameters used in the design of the hybrid B C4 RCCA meet all the nuclear design criteria given in Section 3,0 The initial reactivity worth satisfies Criteria 1 and '2.
i 5-2
5.2 Nuclear, Evaluation (continued)
The use of 100% Ag-In-Cd in part length (P/L) RCCAs, Ag-In-Cd reduced rod lengths in full length (F/L) RCCAs, and residency limitations of Bank D F/L RCCAs while operating in load follow modes satisfies Criteria 3 as well as the lifetime reactivity requirement of Criterion
- 1. The hybrid RCCA duty is based on a fif teen year calendar life, I
which is equivalent to twelve effective full power years based
~
l on an assumed 80% capacity factor. Planiop~erationisassumed to be either base loaded or load following using Constant Axial Offset Control (CAOC). RCCA overlap of 100 steps and power swings between 100% and 50% are assumed.
5.2.1 Fluence Limitations Satisfactory nuclear perfonnance for fifteen years requires limitation of the neutron fluence seen by the B C.4 A period of fif teen years exposure to the typical local core flux environment of P/L RCCAs and F/L Bank D RCCAs would cause 6
violation of:
l l a. Criterion 1 due to excessive loss of reactivity from B-10 depletion,
- b. The cladding stress design bases due to excessive pellet swelling and excessive helium generation.
Excessive pellet radial swelling occurs first. Thus local l
fluence in 8 C is the limiting nuclear design parameter.
4 The value of local fluence in F/L hybrid control rods can be limited to two ways. The axial location of the B C can 4
be restricted to regions above the top of the fuel, where the flux is low. This is done by providing an appropriate f rod length of Ag-In-Cd in the bottom end of the rods. If control rods were in their fully withdrawn position except during reactor shutdown, the required length of Ag-In-Cd rods would be about 5". Actually all control rods are never 5-3 l
l-l
5.2.1 Fluence Limitations (continued) fully withdrawn. Even in base load operation Bank D is normally inserted about fifteen to thirty steps (the bite position). In CAOC Mode A load follow Bank D may be deeply inserted and Bank C partially inserted, depending on rod bank worth and the power swing involved. The Bank C md ends are 128 steps above the Bank D rod end in the one-hundred step overlap mode. In CAOC Model B load follow the F/L rod insertions are even greater. In proposed Mode B operation the P/L rods could be in the core as much as 60% of the time each month of operation, and under these conditions fif teen years for P/L rod insertion can only be obtained using 100% Ag-In-Cd, i.e. , 36". The chosen F/L hybrid absorber rod design satisfies the fluence limitations by limiting
- the residence time of rods in Bank D.
Four loop plants have 53 F/L RCCAs with nine RCCAs in Bank D. The largest number of Bank D rod sets which can be obtained by rotation is five, and acceptable fluence exposure is obtained by limiting the Bank D residency to three years calender time (assumed to be 2.4 years Effective Full Power Years). Three Icop plants have 48 F/L RCCAs with eight RCCAs in Bank D, and the design for three year residency I
could allow as much as eighteen years of operating by rotation in sets of eight RCCAs.
The loss in reactivity which occurs during exposure of the control rods could cause asymmetric core-wise power distributions l
in the rodded planes if all rods in the same bank had not seen the sane duty. Ideally the rods should be rotated by banks to preserve symmetry; however, the maximum local reactivity loss due to exposure in this design is only about 7% (see Section 5.2.4). As an extreme case F/L rods depleted to 90% of their original worth were placed in Bank D on one side of the core and undepleted rods were used on the other side. The calculated core tilt in the rodded plane f
5-4
was less than 2%, and no corrective action would be required by present technical specifications. Therefore, there is no nuclear requirement for maintaining rods with the same burnup history in the same bank.
5.2.2 Load Follow Considerations 5.2.2.1 Allowed,hel,etion The specifications of maximum B4 C pellet 0.D. and 304 stainless steel clad minimum I.D. provide the minimum diametral pellet clad gap at room temperature.
The increase in clad I.D. at power is greater than the increase in pellet 0.D., and the gap at cold conditions is the minimum. The allowed fractional swelling for the designed no pellet-clad contact corresponds to an isotropic volume expansion of 6.851. As stated in Section 4.1.2. the maximum B C swelling is 0.15% AV/V per 10 20 atoms /cm depletion 3
4 up to 95 x 1020 depletions per cm . Thus the allowed 3
average local depletion is 45.7 x 10 20 atoms /cm .
3 The maximum initial number density of B-10 atoms based on the upper limit of the specifications is 160 x 1020 atoms /cm . The most conservative 3
fractional depletion is a local radial average of 45.7/160 = 0.286.
5.2.2.2 h21gtign,in,Mit M rgg Bggs Studies show that under the most extreme exposures caused by continuous load follow operations, either i 100% Mode A or the anticipated monthly mix for Mode B (60% B. 40% A), the minimum length of Ag-In-Cd rods required to protect.B 4C from the high fluxes is about 17" (see Section 5.2.2.3).
5-5
5.2.2.2 De21 etion,in,With,drawr}, Banks (continued)
The minimum length of Ag-In-Cd rod is set at 20",
based on mechanical considerations. Since fully l ~
withdrawn rods place the bottom of the absorter at least 1" above the top of the fuel, the location of B4C in fully withdrawn rods will begin at least 21" above the top of the fuel. The neutron fluence at that point is about three orders of magnitude l
smaller than the limiting value for fifteen year residence, and all effects of neutron fluence in j
fully withdrawn rods are negligible compared with f any duty due to rod insertion. Thus, only rod j exposures while in Bank D or Bank C need consideration, l
since all other banks are fully withdrawn at operating power levels.
s.2.2.3 Ptelet19a_1a BenM_9_and C The duty of B 4C in the rods depends on how the core is actually operated and on the length of l
Ag-In-Cd rods. Cores supplied with 40" rod lengths and operate base loaded (Bank 0 nominally at about two-hundred steps) could tolerate B4C exposure f in Bank D for fifteen years, because the bottom j of the B 4C is about 24" above the top of the fuel.
j
' ^
Even with 20" Ag-In-Cd rod lengths. Bank D insertions ;
' at about 203 steps could be tolerated; however, l
the design must also withstand the most extreme l
exposures inposed by continuous load follow.
In Mode A, the maximum flux in F/L rods occurs at the bottom end of the rod and decreases monotonically above that point. In Mode B with deep Bank D insertions which occur at reduced powers, the P/L rods can i
push the peak flux far above the bottom end of the Bank D rods. Calculations were perforned to determine what load follow condit,fons lead to the 4
s largest fluxes in the F/L rods.
I 5-6 l
l __ _ _ . . _ _ _
5.2.2.3 Degletion,in, Banks,0,a,nd,,C (continued)
A 12' Cycle 1 core, having insertion Ifmits which
. are deeper than expected for hybrid rodded plants, was used for the study. The largest flux occurmd at the end of the cycle at 100% power with F/L rods at their insertion limit (70% inserted).
The axial offset of about + 3% at end of life contributes to the larger flux in the F/L rodded plane. The peak flux occurred at the bottom end of the hybrid rod and decreased at points above. The exposure of B 4C under these conditions can be reduced to any desired value by increasing Ag-In-Cd rod length.
However, at 50% power with P/L rods positioned to produce the largest flux in Bank D (at their insertion limit of 90%), the peak occurred 70" above the bottom of the F/L rodded plane and was about half the value occurring at the bottom end of the Ag-In-Cd rod for the full power case. The axial ' offset was + 14%, which is only slightly above current operating bands. The peak flux in Bank C at this sane 50% power condition decreases from a maximum at the bottom end of the rods.
At 50% power Bank D was inserted 61% (at 84 steps),
which is the maximum insertion expected under normal operation because of insertion limits. With P/L rods at 90% insertion, the maximum flux in Bank D rods occurs beyond any length of Ag-In-Cd considered in this design. The peak reaches but does not exceed the value which limits Bank D to a three-year residency.
During load following maneuvers, the peak location noves, and no B 4 C pellet will see the peak continuously.
5-7 1
l l
I f
5.2.2.3 De21 etion,1,n, Banks,0,and,C (continued) f The use of 17" or more of Ag-In-Cd tips will provide a three-year Bank D residence capability with some
[ margin, but the actual operating history must be examined to calculate the margin to be expected.
At 50% power, the Bank C location, corresponding to Bank D at 84 steps, is at 212 steps. With P/L rods at 90% insertion, the peak flux occurs at the rod end. Even at this extreme condition, the fluxes at 20" above the rod end produce negligible exposures, and no nuclear effects due to Bank C residency need to be considered.
5.2.3 Helium and Tritium Generation f
The generation of helium in B 4C is dominated completely by the B-10 (n, a) Li-7 reaction, and helium generation can be equated to B-10 depletion. Although the radial distribution of helium generation is peaked strongly toward f the surface of the pellet,' only the radial average is needed for calculating the total helium generation in a rod. The same procedures for calculating B-10 depletion described in
{ Section 5.2.2 were used to estimate helium generation. The additional information needed is the axially integrated fluence, which depends on the actual mode of plant operation.
As a conservative case, assuming Bank D insertion of 22" into the fuel exists for 50% of the time at 100% power during Bank D residency of three years and an insertion of 88" applies for the remaining 50% at 50% power using Mode B load follow, the
[
total helium generation using the maximum tolerances 'bn B-10 content in a rod with an assumed 20" Ag-In-Cd rod is about 2.0 x 1023 atoms or 0.32 moles. If operation is base load and a 22" insertion applies for full power operation 100% of f the time, the helium generation is only 0.04 moles. Bank C and shutdown banks add negligible contributions. Only a fraction of the helium generated is released from the
[.
5-8 o
[
5.2.3 Helium and Tritium Generation (continued) f lattice and contributes to the internal pressure on the clad (seeSection4.1.1).
The increase in axial length of the pellet stack due to unconstrained swelling in a B4 C length of 122"(142-20)
[
can be estimated from the average depletion. The calculated depletion of 2.0 X 10 23 atoms per rod produces an average
[ depletion of 11.3 x 10 20 atoms /cm3 when weighted over the length of the stack. This will give a A L/L of 0.56% or 0.69" increased stack length.
f The tritium production in B4 C occurs initially from the B-10 (n, a) a, T reaction. As Li-7 is generated in the lattice from the B-10 (n, a) Li-7 reaction, an additional contribution to tritium production occurs from the Li-7 ,
(n n') a, T reaction. Other reactions leading to tritium production occur, but the yields are negligible, primarily because the neutron energy tnresholds are so high. The
{ B-10 reaction has no physical threshold, but cross sections are negligible below 1.0 Mev. The flux weighted value of o,j is 11.9 mil 11 barns. The Li-7 reaction has a threshold
{
of 2.81 Mev and a,j of 9.6 millibarns. At the neutron energies involved in these reactions, there is virtually no local flux variation due to material inhomogenetties.
The initial rate of tritium production decreases very f slowly as the B-10 depletion is replaced by Li-7 because the Li-7 cross section is only about 20% smaller than the B-10 cross section. Even at the point of maximum allowed
{
B-10 depletion (29%) the rate of tritium production is only about 6% less than in undepleted 04C.
d l If tritium were confined to the rod, it would build up to
! equilibrium with its decay to He-3 (T1/2 = 12.33 years)
) and its further production from the He-3 (n, p) T reaction.
I k
5-9
(
{
5.2.3 Helium,and Tritium _@neration (continued)
In this design, it is assumed that tritium is released to
[ the coolant as it is generated (see Section 4.1.4). Using the conservative assumption of 22" and 88" Control Bank D insertions for each 50% (full power and half power in
[ Mode B follow) of a three-year residency, each hybrid rod of 122" of B4 C length gives a tritium production rate of two curies per year. In a four-loop plant with 9 x 24 Bank 0 rods, the plant load is 430 curies per year'.
I Assuming a 100% base load operation for three years (22" insertion), tritium yearly production results in 0.16 curies per rod and 35 curies for the plant.
[
5.2.4 Reactivity Worth
{
The initial reactivity worth of hybrid rods in this design is calculated to be greater than the worth of the corresponding 100% Ag-In-Cd rods. In a typical four loop plant at beginning of life, HZP, the calculated worth of 53 F/L rods containing all B 4C at the maximum initial B-10 content allowed by tolerances is 12% greater than the present Ag-In-Cd rods, and 10% greater
( at the minimum initial B-10 content.
The depletion of Ag-In-Cd and hybrid rods was calculated using the neutron flux typical of values near the peak in the rodded
[ assemblies. Over the range of fluence considered, the fractional loss in local B4C reactivity with exposure is nearly identical to the loss of Ag-In-Cd reactivity. This occurs because the
{
rapid depletion of Cd-ll3 in Ag-In-Cd causes its reactivity loss to keep pace with the effect of B-10 depletion until the Cd-113 is nearly depleted. At larger fluences the 8 4C reactivity loss rate continues to increase while the .
- f. reactivity loss rate in Ag-In-Cd decreases because of the slower depletion rates in Ag and In isotopes.
l.
r 5-10 l'
5.2.4 Reactivity Worth (continued)
The three-year Bank D residency and 20" Ag-In-Cd rod ends used in the analyses limits the maximum local reactivity loss in B C4 to about 5%. The maximum loss at the bottom end of the Ag-In-Cd amounts to about 7%. The average loss in the inserted length is much less, and the effect on reactivity control due to depletion of F/L hybrid rods during fifteen years is negligible.
53 Ihermal_and_Wdraulic.gvaluation l 5.3.1 Maximum Absorber Temperature The maximum centerline temperature, under nonnal reactor operating conditions, of the 0.334" diameter 84 C pellet with a gas environment of air is calculated to be less than ;
1200*F. This value is well below the melting temperature of 4400*F for 8 4C and 1454*F for Ag-In-Cd even conservatively assuming the B 4 C hot spot occurs at the interface contact with Ag-In-Cd. The maximum flux locally in the Ag-In-Cd ;
rod has acceptable material temperatures, as is the case of the F/L 100% Ag-In-Cd control rods.
The B C temperature calculations are based on the following:
4
- 1. The maximum heating rate in the hottest B4 C pellet.
- 2. Heat is assumed to be generated uniformly in the B 4C pellet. The actual heating distribution, where approximately l
60% of the heat is generated in the outer one-fifth of the B4C pellet, results in lower centerline tenperatures.
! 3. Control rod surface temperature was assumed to be 660*F
! and t.he worst dimensional stackup was considered.
5-11
e 5.3.1 Maximum Absorber Temperature (continued)
- 4. Thermal conductivity of B4C decreases under irradiation and the amount of reduction is a function of the irradiation 1 temperature. The thermal conductivity was appmpriately reduced in the analysis to account for the irradiation effects.
5.3.2 Bypass Flow Since the clad outer diameter of the B4 C control rod is I the same as the clad outer diameter of the Ag-In-Cd control rod, there is no change in the total thimble bypass flow.
5.4 DS992. Test,Perfgrmance,0f,Prototygi, cal,17x1,7,1jgb,ri,d,B 4 9,R([$,$s,seyy The purpose of the 0-loop functional test was to confirm the control rod drop time adequacy of a prototypical 12' 17x17 hybrid B4C control rod' assembly. The functional adequacy of the drive line components which drive, guide and protect the RCCA assembly have already been verified and reported in Reference 1. The geometry of the RCCA is identical to the 100% Ag-In-Cd RCCA except for the B 4C and clad interface.
5.4.1 Test Conditions The hybrid control rod was tested, in the 0-loop test facility, f for a total of 880 hours0.0102 days <br />0.244 hours <br />0.00146 weeks <br />3.3484e-4 months <br /> (at a temperature above SCYF).
During the tests the rod was subjected to 245 control rod drops and accumulated 57,319 steps. The purpose of this l
testing was to demonstrate'the insensitivity of the control rod drop time to component changes (i.e. , guide tube and drive rods) and to establish the stall characteristics of I the hybrid 84 C control rod (drive line).
5-12
. - , - . . . - _ . - . _ - - - . - - - - - - - . _ . _ ~
j 5.4.1 Test Conditions (continued) l The tests were conducted at 585'F i 5'F, 2000 psi i 100 j psi with assembly flow rates. that varied from 0 gpm to 3445 l gpm which is 166% of thermal design flow. The control rod drop time was recorded over all phases of testing. The control rod drop time and the variation in control rod drop l time was within acceptable limits for core safety (see Section l 6), and the adequacy of the 17x1712' hybrid control rod was confimed for use in Westinghouse Pressurized Water l Reacto rs.
5.4.2 Determination of Control Rod Drop Time The control rod drop time was detennined by monitoring both l the digital rod p'osition indicator electrical output signal and the electrical signals from the velocity coils mounted
, on the fuel assembly (refer to Reference 1 for a pictorial description). These signals were recorded on an oscillograph with a 100 Hz reference time code. The control rod drop
- time (Table 5-1) was found to be within the design limits ,
'over the range of plant flow rates.
l 5.4.3 Control Rod Stall Tests At the completion of the first phase of testing, control rod stall tests were conducted. Control rod drop times were recorded at 127,139,150,162 and 166% thennal design flow (TDF) of 2070 gpm per assembly. Under the test conditions, ;
in all cases of flow up to 167% TDF, the control rods did not f stall .
55 hrvtillenct_Pggggm t l
l For those initial plants utilizing the hybrid RCCAs, selected hybrid !
rods will be visually examined. Visual examinations will look i for abnormalities such as clad defects, enJd buildup and wear. !
l \
5-13
_e ,_
- ,, ,yc
5.6 References
- 1. Cooper, F. W., Jr., "17x17 Driveline Components Test - Phase IB, !! and II D Loop Drop and Deflection", WCAP 8446 (Proprietary) and WCAP 8449 (non Proprietary), December 1974
{-
4 f -
p L .
I-5-14
{a:
Ssu w '
I TABLE 5-1 17 X 17 HYBRID B C RCCA DROP TIME VERSUS FLOW RATE 4
D LOOP TEST RESULTS C0')LANT CONDITIONS - 585'F, 2000 PSI A_
I D Lcop Rod Drop Plant Loop Design,Cgnfitign [lgw, gate,;,gd Iige,;,Seggnps [lgwggtg,;,R$
Thennal Design Flow 2070 1.87 94,400 Best Estimate Flow 2208 1.97 100,700 Mechanical Design Flow 2305 2.04 105,100 I
I I
I 5-15 4
[
[
6.0 ACCIDENT EVALUATION 6.1 General f Control rod drop times for hybrid 8 C4 RCCAs exhibit an increased sensitivity to flow rate due to the reduced B C4 absorber weight. The
( substitution of B C4 neutron absorber material for Ag-In-Cd results in no changes to the nuclear and thermal hydraulic parameters previously assumed in accident analysis, which was based on 100% F/L Ag-In-Cd control rods. Although B4 C has a slightly greater neutron absorber worth than Ag-In-Cd, this benefit has not been considered in evaluating accident control rod drop time requirements.
I for transients which are sensitive to variations in control rod drop time, scram time requirements have been evaluated on a case-by-case basis. Minimum and maximum flow rates, based on the best estimate of expected plant flow rate, have been used for accident analysis. Mechanical design flow (MDF) is the maximum flow rate employed in SAR accident analysis of limiting accidents, and it is based on the best estimate flow rate increased by an uncertainty allowance. This flow rate is used to calculate forces on reactor vessel internals and control rod drop time. Thermal design flow (TDF) corresponds to the best estimate flow rate minus an uncertainty allowance. This flow rate is the initial flow rate assumed in accident analyses for evaluation of core thermal conditions. ' Traditionally, the scram t'ime corresponding to the mechanical design flow rate has been used in accident analysis for the 100% Ag-In-Cd RCCA design.
Limiting thermal conditions for accidents occur at thermal design flow. The combination of the slowest scram time and the limiting thermal condition for the two extremes of flow rate have been a conservative simplifying assumption. Due to the increased sensitivity of scram time to flow rate associated with hybrid RCCAs, the appropriate
{ flow rate for establishing control rod drop time requirements has been evaluated for specific accident analyses. Based on this evaluation, accident analysis rod drop requirements have been specified in terms of allowable scram time for the flow rate assumed in the
[ analysis. ,
6-1 r
[
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1 The analyses presented were performed for a typical four-loop plant with a 17x17,12 f t. core. The effect of control rod drop time depen-dency on flow rate will be less severe for three loop plants, due to a lower average assembly flow rate for a given loop flow rate.
f Incidents with Minimal Consequences due to Variations in 6.2
. _ _Co.n.t.r.o.l. .R.o.d_ _D_r.o.p
.. T.i.m.e. . . . _ _ _ . . . . . . . . . . . _ _ _ _ . . . . . . . . . .
Slow transients are relatively insensitive to changes in control rod droo time and will therefore not be significantly affected by increased scram l time due to the use of hybrid RCCAs. Variations in control rod drop time will have a minimal affect on minimum DNBR for the rod withdrawal at power incident, since minimum DNBR for the transient occurs at relatively low reactivity insertion rates as shown in Section 15.2.2 f of Reference 1. No .aalysis was required for the rod withdrawal from subcritical transient, since previous analyses have shown that peak heat flux and fuel pellet average temperature associated with this transient remain well below the nominal full power values. Sensitivity studies performed for the locked rotor transient showed that peak reactor coolant system pressure and peak clad temperature for this transient are relatively insensitive to variations in control rod drop time; therefore, no analysis is presented for this transient.
6.3 Los,s ,of ,F,1,ow f ,
M_et, hod of Analysis l
Control rod drop time requirements for the complete loss of flow transient were evaluated using standard methods and computer codes as described in Section 15.3.4 of Reference 1. Reactivity insertion as a function of time following reactor trip based on 8 C4 RCCA scrat. characteristics f
calculated for various rod drop times was employed in this analysis.
Figures 61'and 6-2 show normalized rod position vs. time and reactivity l vs. rod position, respectively, used to calculate reactivity as a function of time following reactor trip. Reactor trip due to reactor
[* 6-2
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[
{ coolant pump bus undervoltage was assumed to result in control rod motion 1.5 seconds after the trip setpoint was reached. Although rod drop velocities are somewhat lower at higher flow rates, the loss of flow accident was analyzed at thermal design flow. During the 1.5 second delay time before rod motion is assumed to begin, the flow would coast down to approximately 90% of its initial value. For the analysis case initiated at thermal design flow, minimum DNBR occurred at a particular value of flow rate during the transient. For an initial flow higher than thermal design flow, this particular flow rate would occur later in the transient. For such transients, control rod drop would be initiated earlier in the transient with respect to reaching the parti-( celar flow rate, and a more negative reactivity would have been inserted by the control rods by the time this particular flow rate was reached.
This would give a reduced heat flux and thus a Df4BP. benefit would
{
resul t. Therefore, the analysis case initiating the loss of flow transient from thermal design flow was considered to be the most limiting Case.
Resul ts Table 6-1 shows reference plant characteristics assumed for the evaluation.
The transient analyzed was a loss of power to all reactor coolant pumps assuming all loops in operation. Partial loss of flow events were not
( ,, investigated, since previous analysis has shown that these transients are less limiting with respect to DNB than a complete loss of flow.
Figures 6-3 through 6-6 show nuclear power, heat flux, reactor coolant
{
flow, and DNBR as a function of time corresponding to a control rod drop time of 2.9 seconds.
Conclusion This analysis demonstrates that for a complete loss of forced reactor coolant flow, minimum DNBR is maintained above 1.30 providing that control rod entry to the dashpot occurs within 2.9 seconds following
(
p 6-3 e
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initiation of rod drop motion. Results of hybrid RCCA rod drop per-
{
formance tests presented in Table 5-1 show that the expected rod drop time at thermal design flow is well within the required scram time.
6.4 4q(gr.fr.qqqqqcy.
Method of Analysis Control rod drop time requirements for a grid frequency disturbance resulting in a 5 HZ/sec frequency decay rate assuming a 57 HZ under-frequency trip setpoint were evaluated using the methods and computer codes described in Reference 2. This transient is slightly more
[ limiting with respect to DNBR than a loss rf power to all reactor coolant pumps due to a faster flow coastdown. As for the complete loss of flow, scram time requirements for underfrequency events initiated at a flow f
rate greater than thermal design flow need not be considered. Plant characteristics used for this evaluation were identical to those assumed for analysis of the complete loss of flow transient, as shown in Table 6-1.
Results
(
This evaluation showed that a control rod drop time of 2.5 seconds is f required to assure that minimum DNBR is maintained above 1.30. For this 1
transient however, evaluation of scram characteristics of 95% of thermal design flow is conservative, since this analysis as well as previous f analyses has shown that flow is reduced by approximately 10% at the time of reactor trip. Nuclear power, heat flux, reactor coolant flow,
{ and DNBR as a function of time corresponding to a control rod drop' time of 2.5 seconds are shown in Figures 6-7 through 6-10.
Conclusion i For the limiting grid frequency decay transient, minimum DNBR remains above 1.30 providing that control rod entry to the dashpot evaluated
{ at 95% of thermal design flow does not exceed 2.5 seconds. Table 5-1 shows that the expected rod drop time for 95% of thermal design flow
' satisfies this requirement.
6-4 r
6.5 hi.Eieqtd.qq Me_thod of Analysis The rod ejection transient was analyzed using standard methods and computer codes as described in Section 15.4.6 of Reference 1, with the exceptions noted. This transient has traditionally been analyzed using thermal design flow to evaluate heat transfer, based on a rod drop time calculated at the mechanical design flow rate. The use of the most conservative combination of flow assumptions for evaluation of the themal transient and control rod drop time has previously been used to eliminate the need for transient analyses for a range of flow rates and corresponding scram times. Due to the increased sensitivity of control rod drop time to flow, however, md ejection scram time requirements corresponding to both themal design flow and mechanical design flow have been evaluated. Analysis has shown that rod ejection transients initiated at themal design flow and mechanical design flow and their corresponding scram times bound the consequences of transients initiated at intermediate flow rates. Analysis of the rod ejection transient for both TDF and fiDF, based on a scram time consistent with the flow rate assumed in the analysis, will therefore, be required to determine the consequences of rod ejection accidents using hybrid RCCAs.
~
Rod scram posi6 ion as a function of time has traditionally been modeled assuming that all control rods are completely withdrawn at the time of reactor trip.- Peaking factors and ejected rod worths employed in analyses, however, are nomally calculated assuming contml banks inserted to- their maximum insertion limit. In the analyses presented below, therefore, credit was taken for partial control rod insertion at the time of reactor trip. For calculation of scram reactivity associated with the full power cases, the transient was assumed to be initiated with the shutdown banks completely withdrawn, control banks A, B, and C at the top of the fuel, and Bank D 10% inserted.
h 6-5
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This assumption is extremely conservative with respect to initial con-ditions used to calculate peaking factors and ejected rod worths for this transient. The zero power cases were analyzed assuming a scram reactivity calculated for Bank D fully inserted, Bank C 10% inserted, Banks A and B at the top of the fuel, and the shutdown banks completely
- wi thdrawn.
Results Rod ejection parameters assumed in this evaluation are shown in Table I 6-2 and 6-3. The parameters used represent a conservative combination of values expected for a four-loop,17x17,12 ft. core. This analysis showed that for rod ejection transients initiated at mechanical design flow, a control rod drop time of 3.5 seconds is required to assure that critical parameters are within the limiting values presented in Section 15.4.6.1.2 of Reference 1. A scram time of 2.7' seconds is required to produce acceptable results for a rod ejection transient initiated I at thermal design flow. Results of both analyses are presented in Table 6-2 and 6-3.
Conclusion Table 5-1 shows that the expected scram times corresponding to thermal design flow and mechanical design flow are well below the required limits I of 2.7 and 3.5 seconds, respectively. It can be concluded, therefore, that the fuel and clad limits presented in Reference 1 are not exceeded due to the substitution of B C 4 neutron absorber material for Ag-In-Cd.
6.6 Yetificetion.of.lodlaut.BQd.Qton Iime I
Verification of control rod drop performance in plant tests is required I to demonstrate that RCCAs are within expected control rod drop time requirements. Expected control rod drop times based on experimental data, evaluation of uncertainties, and the variation of scram performance l for postulated conditions are utilized in accident analyses. The 6-6 W
l t .
predicted scram performance is presented in terms of the time after scram to reach the top of the dashpot as a function of reactor coolant l flow rate. The objective of the plant control rod drop test is to demonstrate
! adequate scram performance for normal plant operating conditions.
i l
A technical specification limit is placed on drop time to require a value of rod drop time in the plant which corresponds to predicted values from the test conditions. The limit is evaluated at the hot normal l plan! operating conditions. The effect of test condition drop time i verification up to the technical specification limit has been included l
in the analysis of accidents assuming abnormal plant conditions. For I
12' plants, the technical specification limit for B C4control rods is f
j 2.2 seconds for the test conditions described above.
6.7 References l
- 1. RESAR 35, Reference Safety Analysis Report, Westinghouse Nuclear j j Energy Systems, July 1975 and later amendments.
l t 1
( 2. Balwin, M. S. , Merrian, M. M. , Schenkel , H. S. and Vandewalle, j l D. J., "An Evaluation of Loss of Flow Accidents Caused by Power j l
System Frequency Transients in Westinghouse PWRs", WCAP 8424, i Revision 1 July 1975. l l
i l
t 4
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6-7 E
TABLE 6-1 REFERENCE PLANT CHARACTERISTICS Loss of Flow / Underfrequency Initial Conditions Power Rating, kt 3.425 Thermal Design Flow, gpm per loop 94,400 Nominal inlet temperature, 'F 558.1 Analysis inlet temperature. *F 564.6 l 2,250 Nominal pressure, psia -
Initial pressure, psia - 2,220 Initial flow, gpm per loop 94,400 l
I, 6- 8 I
- . ~ - ~ . - -
. - . _ ~ - . - -.-.- . . - . .. - .
i I
i TABLE 6-2
. i PARAMETERS USED AND RESULTS FOR Tile EJECTED R0D l t CLUSTER CONTROL ASSEMBEY ANALYSIS' (
(MECHANICAL DESIGN FLOW INITIAL CONDITIONS) _
Time in Life Beginning Beginntnco.
End End I Power Level, % 102 0 102 0 l
Ejected rod worth, 5 Delta K 0.16 0.626 0.195 1.038
- Delayed neutron fraction, % 0.55 0.55 0.44 0.44 Feedback reactivity weighting 1.60 2.31 1.60 5.00 Trio res.-tivity. % Delta K 5.0 2.0 4.0 2.0 F before rod ejection 2.52 -- 2.52 --
l q
F after rod ejection 5.60 12.60 6.50 24.01 q
4 2 4 2 Number of operational pumps Max, fuel pellet-average temp. 'F. 3772 3224 3923 3874 Max. fuel center temp, *F -
4900* 3991 4800* 4615 Max. clad average temp, F 1940 2161 2060 2697 Max. fuel stored energy, cal /gm 162 135 170 168 Time to dashpot, secs 3.5 3.5 '3.5 3.5 Flow,' gpm per loop 105,100 48,346 105.100 48,346
- r.105 melt l
i I
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6- 9
. ._ = . ._ _ _ . _ _ _
TABLE 6-3 l
1 i
PARAMETERS USED AND RESULTS FOR THE EJECTED i ROD CLUSTER CONTROL ASSEPELY (THERMAL DESIGN FLOW CONDITIONS)
Time in, Life Beginning. Be.g_i nning End End :
i Power level, % 102 0 102 0 Ejected rod worth, % Delta K 0.16 0.826 0.195 1.038 ,
Delayed neutron fraction, % 0.55 0.55 0.44 0.44 l Feedback reactivity weighting 1.60 2.31 1.60 5.00 i Trip reactivity, % Delta K 5.0 2.0 4.0 2.0 ;
.F before rod ejection 2.52 -- 2.52 --
q '
F after rod ejection 5.60 12.60 6.50 24.01 q
Number of operational pumps 4 2 4 2 Max. fuel pellet average temp, F 3632 3173 3759 3684 Max. fuel center temperature 'F 4823 3850 4800* 4305 l Max clad average temperature. *F 1929 2168 2004 2645 j Max fuel stored energy, cal /gm 155 134 162 159 i Time to dashpot, secs 2.70 2.70 2.70 2.70 Reactor coolant flow, gpm per loop 94,400 43,424 94,400 43,424 -
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Figure 6-3 Complete Loss of Flow; Nuclear Power Versus Time r 6-13 1'
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Figure 6-4 Complete Loss of Flow; Heat Flux Versus Time
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/ % UNITE 3 5TATES l . f\ s' , NUCLEAR REGUl.ATORY COMMISSION
! y nAsHtNGTON, D. C. 20606
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1 M I 01977 l
1 Mr. C. Eiche1dinger, Manager l Nuclear Safety Department Westinghouse Electric Corporation P. O. Box 355 Pittsburgh, Pennsylvania 15230 l
Dear Mr. Eiche1dinger:
L
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING WCAP-8846, i
" HYBRID B 4C ABSORBER CONTROL R00 EVALUATION REPORT" Additional infomation is needed to complete our review of Westinghouse l i
l Electric Corporation topical report WCAP-8846. The requested additional information is enclosed. !
This additional infonnation is needed by March 25, 1977 to meet our ;
review schedule. If you cannot meet this schedule, please inform us !
within ten days after receipt of this letter of the date you plan to
, submit your response. -
Sincerely, ,
m !
hn F. Stolz, Chi ight Water Reactors i Branch No.1 Division of Project Managenent
Enclosure:
Request for Additional Infonnation l
l A.1-1 t
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l ENCLOSURE C: QUEST FOR ADDITIONAL INFORMATION REGARDING WCAP-8846, " HYBRID B C ABSORBER CONTROL R00 EVALUATION, 4
DATED SEPTEMBER 1976 l
i A.1 -2
1 L
Reactor Fuels Section - Core Performance Branch 231.1 Provide the following control rod perforraance data:
{
a) Plastic strain design limit for cladding.
f b) End-of-life calculated plastic strain in the cladding due to B 4 C swelling. Is isotropic swelling shown in experiments to date? -
c) End-of-life internal pressure in psia due to helium release and calculated stress on cladding under normal operating con-ditions,
[
d) Calculated stress on cladding at end-of-life due to LOCA (small breaks) and normal plant depressurization.
f 231.2 In reference to maximum absorber pellet temperature (Section 5.3.1), it is implied that the radial gap is air-filled. Will the hybrid control rods be' fabricated with air-filled gaps?
f 231.3 Identify the materials performance code used to predict the B 4C behavior.
231.4 A visual surveillance program is described in Section 5.5.
f Provide a tentative frequency for visual examination of selected rods. Discuss how rod diametral growth will be examined.
231.5 Provide a discussion on the fabrication techniques and quality assurance procedures for the B4C pellets.
\.
A.1-3 I. .
2-f 231.6 The fo11ow2ng references (Section 4.3) appear to be in error:
a) No. 3 - Incorrect conference number.
b) No. 6 - Need report number or better reference notation.
c) No. 10 - CEAP notation appears inconsistant with year of issue.
d) No. 13 - Reference as stated is untraceable. Provide
{
reference or copy of reference if not readily available.
l 9
A.1-4 l.
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.* 3 Physics Section - Core Performance Branch 232.1 Please describe the data base you use to validate your nuclear calculations of hybrid BgC control rod properties.
232.2 Control rod worths have been overpredicted for several Westinghouse reloads. This can lead to problems in maintaining shutdown margins. What assurance can you provide that the likelihood of this occurring again does not increase with B C 4
/ rods, in view of the fact that calculational uncertainties may increase with the new material and lack of startup test data?
A complete answer will discuss this problem for the old.as well as the new control rod material.
9 A.1-5
~
Section B, Reactor Systems tranch 212.1 What is the effect of an off-center pellet (B4 C) on steady-state and transient cooling of the control rod?
212.2 Is there any potential chemical corrosion of fuel eladding in the event of B 4C " washout" from a control rod?
212.3 Provide details of control rod drop tests, including o in g each test.
212.4 Sect. 6.5, Page 6.5 ,
" Credit is taken for partial control rod insertion at time of reactor trip." Justify this statement.
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y,ts;!:ghcuse E!cc;;!c Cciporatiori PGwir Sysisms -
amas PimiughPersytvate15230
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, March 31, 1977 f* NS-CE-1395
.Mr. John F. Stolz, Chief Light Water Reactors Branch No.1 Division of Project Management
-[ U. S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, Maryland 20014
Dear Mr. Stolz:
As requested by your letter, dated March 10, 1977, the attachcant provides
[ additional information on WCAP-8846, " Hybrid B4C Absorber Control Rod Evaluation Report". This infonnation follows the format of the enclosure to your letter.
f Pursuant to the phone conservation between the NRC and my Staff on March 25,
. 1977, this information is being supplied one week later than stated in f your letter. Contrary to our phone conservation additional time will-t not be required to provide our response to Question 231.ld. It is hoped that upon receipt of this infonnation' your review can be completed without schedular impact. -
f.
Very truly yours.
A 6-
[. C. Eich dinger, anager i Nuclear Safety Department CJR:pj
( Attachmen
{
A.2-1
(* l
I RESP 0f4SES T0 f4RC 00EST1045 ON B C RCCA WCAP-8846 Juestion 231.1 , .
Provide the following control rod perfonnance data:
A. Plastic strain design limit for cladding.
B. End-of-life calculated plastic strain in the cladding due to B 4C swelling. Is isotropic swelling shown in experiments to date?
C. End-of-life internal pressure in psia due to helium release and calculated stress en cladding under normal operating conditions.
D. Calculated stress on cladding at end-of-life due to LOCA (small breaks) and normal plant depressurization.
Response 231.la
- The plastic strain design limit is 1".
Response 231.1b The end-of-life calculated plastic strain in the claddina due to B4C swelling is zero, since the design provides for a radial gap under worst-case normal operating conditions (see Section 5.1.1 of WCAP-8846). The design of no circumferential total pellet-clad contact under any normal plant operating conditien is satisfied during the hybrid RCCA 15 year design life, as' long as the design maximum B4 C local depletion of 45.7x10 20 atoms /cm 3 is satisfied (see Section 5.2.2.1 of WCAP).
For plants operating under current Technical Specification, the maximum allowable B4 C depletion can be satisfied with Control Bank D residency time well in excess of the 2.4 Effective Full Power Years limitation of the assured Mode B C*OC '
operation.
Microscopically, the information available (References 1, 2. 3 below) indicate anisotropic straining of the B C4 unit cell, which is rhombohedral with icosahedral units on the corners and a three atom chain along the diagonal. There is conflict between authors, and also between experimental data of indivi?.al authors, as to whether the unit cell expands or contracts.
Macroscopically, there nas not been any evidence of anisotropic swelling and isotropic swelling is assumed. This is to be expected in pel',ets made from powders sin:e individual power particles are not significant1 alligned. Uni form thermal egansion values also support the random particle ali;* ent argu ent.
A.2-2
2 References for Question 231.lb
- 1. G. L. Copeland, et. al. , "Ir-adiation Behavior of Boron Carbide", Nuclear Technology, Vol.16, October 1972
- 2. R. G. Gray and L. R. Lynam, Irradiation Behavior of Bulk B 4C and 3 C + sic Burnable Poison Plates ', WAF0-261, October 1963 4
- 3. C. W. Tucker Jr. and P. Senio, "X-ray Scattering by Neutron Irradiated Single Crystals of Boron Carbide I, Acta Cryst. 8,1955 Response 231.lc End-of-life internal pressure due to helium release is less than 2100 csia for a rod with a 20 inch Ag-In-Cd ends and a total residence time in Bank D c' 2.40 Effective Full Power Years under the assumed Mode B luad follow operat':n. The corresponding calculated stress is less than 5,000 psia with a maxirve aormal coolant pressure of 2300 psia.
The assumed Mode B case evaluated represents bounding or worst-case irradiation conditions for helium and tritit, production in possible, but not expected, future plant operations. Fcr p' arts operating under current Technical 5:ecifications, both the helium and tritium prodsction would be substantially less than the values quoted in WCAP-8846. For example, for hybrid B C RCCAs with 20 inch 4
Ag-In-Cd ends, a daily 12-3-6-3* Mode A load follow would allow a 15 calendar year Bank D residency time while satisfying the 45.7x10 20 atoms /cm 3 B C depletion limit, assuming an 80% plant cacacity factor. Gas production would be -educed by a factor greater than 10, and tritium release would result in only an additional 21 curies per year fc- a four loop plant.
Response 231.ld During the small break LOCA transient it is possible that the claddin; a yield and expand to the thirble tube '-:ernal diameter. The thirble wall wc.,'d act as a barrier against further claddi ; expansion.
During the Condition II plant de:ressurization transient, the pressure trop across the cladding is less than 1,350 :sia. Taking the end-of-life maximum :' adding temperature and the trans4 eat !v' um pressure drop, the claddira stress alculated is less than 24,000 psi, wr.ich se-isfies the design basis li-it g'ven i- Eection 3 9.I N ~E!f5...........................................................................
- 12-3-6-3 is ce rined as 100', power for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> ranp to 50t scwer and then a 6 .h qu r,,h,qlj ,,a nj ,Qs, ,a ,3, ,$,qu r,,r a, ,q ,b,a,c,k,, t,q ,1,q0,f ,p ge c.,,,Q,c,e j ,w,s gay ,hl hl,gg , ,;,qn.
A.2-3
i Question 231.2 In reference to maximum absorber pellet temperature (Section 5.3.1), it is implied l that the radial gap is air-filled. Will the hybrid control rods be fabricated with l air-filled gaps?
Response 231.2 It is planned to fabricate the hybrid control rods with air-filled gaps.
Question 231.3 Identify the materials perfomance code used to predict the B 4C behavior.
Response 231.3 A material perfonnance code was not used to predict B4C behavior. The B 4C perfonnance is predicted using the materials properties given in the WCAP, in particular the Materials Evaluation Chapter 4.0.
Question 231.4 A visual surveillance program is described in Section 5.5. Provide a tentative I frequency for visual examination of selected rods. Discuss how rod diametral growth will be examined.
Response 231.4 1
Since inserted rods experience the highest fluence, selected D bank rods of the initial plants having hybrid RCCAs should be visually examined at the end of their third cycle. The red diameter would be examined for potential abnormalities by T.V. scanning and/or binocular inspection.
Question 231.5 Provide a discussion on the fabrication techniques and quality assurance procedures for the B4C pellets.
Response 231.5 Existing fabrication requirements allow cold-pressed and sintered, extruded and sintered, or hot-pressed B 4C pellets to be made from B 4C powder. At present pellets are planned to be supplied in the cold-pressed with binder and then sintered state. Sintering temperatures range from about 1200-1300 C to approximately 2275 C, depending upon the vendor.
Consistent quality is obtained by the many requirements of the B4 C pellet specifications. Quantitative assurance of the quality is satisfied by measuring pellet:
A.2-4
4
- 1. Dimensions A. Diameter B. Length
- 2. Chipping
- 3. Chemical Composition A. Bor on and Boron plus Carbon B. Impurities, including soluble boron and boron oxides C. Moisture
- 4. Isotopic Composition
- 5. Geometri: Density
- 6. Axial Compressive Strength
- 7. Gas Evolution
- 8. Microstructure
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0 A.2-5
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Question 231.6 The following references (Section 4.3) appear to be in error:
a) No. 3 - Incorrect conference number b) No. 6 - Need report number or better reference notation c) No.10 - GEAP notation a : ears inconsistent with year of issue.
d) No.13 - Reference as stated is untraceable. Provide reference or copy of reference if not readily available.
Response 231.6 a) No. 3 - The correct CONF curber is 720420-1.
l b) No. 6 - The report number is HEDL-SA-206.
c) No. 10 - The correct GEA: e. umber is 3680.
d) No.13 - A copy of the re'erence is attached.
In addition, the No.12 refererce should be corrected to report number riEDL-SA-565, dated, April,1973.
Question 232.1 Please describe the data base you use to validate your nuclear calculations of hybrid B4 C control rod propertdes.
Response 232.1 The experimental data base usec to validate the nuclear characteristic: :f B 0 4
control rods consists of four : itical experiments conducted at the Westinghouse Reactor Evaluation Center (WRE;. in 1967 and one critical eceriment ccaducted by B&W in 1972. .
f The WREC critical experiments .ere carried out in 30 x 30 and 31 x 31 'sel rod arrays with a 0.600-inch pitc". wo types of B4 C rods were ,easured. C e set
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contained B4 C powder compacte: :: 69.0i T.D. in stainless steel tu:es .esing a 0.435-inch 0.D. and a 20 mil t ':kn;ss. The other type was conpacted to 63.81 T.D. in :tair.les:. steel tubes anng a 0.285-inch 0.D. and a 26.5 mil : .:ckness.
A.2-6
Response 2.321 (continued)
Measured axial bucklings were ootained by gamma scanning the fuel roos. The worths of the B 4C control rods werc measured in terms of the changes in the experimental axial bucklings.
The B&W experiment was carried out in the CX-10 facility. The results are reported in BAW-3647-24. The core contained a center 15 x 15 array in which 16 B4 C full length control rods were arranged to simulate an RCC pattern. The B4 C rods were made by filling 0.438-inch 0.D. aluminun tubes having a 35 mil
, wall thickness with material B C4powder compacted to 50.8% T.D. The worth of the B 4C control rods was measured in terms of the change in the critical boron concentration of the water moderator.
l A comparison of the measured and calculated control rod worths shows that the calculated values are in the range of 2.3 to 5.1" less than the measured values.
Ques tion 232.2 Control rod worths have been overpredicted for several Westinghouse reloads.
This can lead to problems in maintaining shutdown margins. What assurance can you provide that the likelihood of this occurring again does not increase with B4C rods, in view of the fact that calculational uncertainties may increase with the new material and lack of startup test data? A complete answer will discuss this problem for the old as well as the new control rod material.
Resconse 232.2 As stated in Response 232.1, the discrepancy between tne measured and calculated value is conservative in that in every case the measured wortn is higher than
{ that calculated. This consistent underestimate of rod worth in the critical experiments provides assurance that an overestimate of rod wortn in :ne plant is'unlikely.
An overprediction of Ag-In-Cd control rod worths has been observed in tre past
( in several Westinghouse reloeds. As a result a control rod worth prograr has been carried out by Westinghouse in cooperation with a "unber of utilities to collect accurate rod worth dat3, identify tne cause o' :ne overes*,i 3te of rod worth, and develop' an i . proved rod worth calculation ethod. End point ;oron samples at various rodded core conditions in both first and reload cores .iere collected Dy the participating utilities and sent to Westinghouse, where the boron A.2-7
Response 232.2 (continued) concentration under controiled conditions could be determined to within 1 ppm 4 (lo). While comparisons :etween the Westinghouse and plant boron titrations showed that measurement uncertainties have contributed to the c: served over-estimates.in control rod worth, the need for an improved calculational method was confirmed. Therefore, these rod worth data were then used to develop im-proved transport corrected two groups constants for the control rod cells for subsequent use in two group spatial diffusion codes. The use of these improved group constants has resulted in significantly better agreement between measured and calculated rod worths.
The good agreement demonstrated for rod worth comparisons using the improved calculation method provicis assurance that the nuclear characteristics of B4 C control rods are accurately detennined.
I Question 212.1 What is the effect of an off-center pellet (8 C) 4 on steady-state and transient cooling of the control red?
Response 212.1 The effect of an off-center B 4C pellet results in no increase in the maximum pellet temperature when c;mpared to the case of a concentrically located pellet.
This is proven by the analyses shown in the three references bel:w.
References for 212.1 1
- 1. Nijsing, R., "Tencerature and Heat Flux Distribution in uclear Fuel Element Rods," Nucl . Enar. and Desian, Vol . 4,1966, c. *-20.
- 2. Goldbert, M. D. , and Rust, J. H. , " Parametric Study d -ett Trans fer from LQ Fuel Pelle:s Located Eccentrically within tw :: 1: ding," ANS Transactions, Oc:::er.1974, p. 318-319.
- 3. Olson, C. I... and Soman, L. H., "Three Dimensicnal Efi::s in PWR Fuel Rods", ASPE Ssper, 75-WA/HT-78, ASME Winter 1975 "ee:ing.
1 A.2-8
Question 212.2
' 15 there any potential chemical corrosion of fuel cladding in the event of B4C "wasnout" from a control rod?
Response 212.2 There is no chemical corrosion of fuel cladding anticipated due to the unlikely event of B4 C washout from a control rod. In such an event, thE B C in the 4
coolant could degrade to B 23 0 and carbon or CO 2
. Neither the S 4 C particles, the boron oxide (already in the coolant as a deliberate reactivity control agent), nor the carbon and/or oxides would accelerate the expected minimal corrosion of the fuel cladding. -
Question 212.3 Provide details of control rod drop tests, including q in each test.
Response 212.3 The control rod drop tests are descri, bed in Section 5.4 of WCAP-8846 and the reference in Section 5.6. Referring to Table 5-1, the rod drop times and their measured standar: deviations for the flow conditions are:
Thennal Design Flow - 1.87 1 007 seconds Best Estimate Flew - 1.97 1 015 seconds Mechanical Design Flow - 2.04 1 020 seconds 110% Mechanical Cesign Flow - 2.27 1 030 seconds The hybrid RCCA drop ti ne characteristics considered in the Accident Evaluation c:tapter are:
Cesign Condi tions
- Rod Drop Time (Sec.) Flowrate 'e* Loop (GPM) 95% TDF 2.5 89,720 TDF 2.7 94, 0 3 MDF 3.3 105,1:?
110% MDF 5.3 115,600
- TDF = Thermal Design ;bw MDF = Mechanical Desig. Flow A.2-9
L Question 212.6- Section 6.5, Page 6.5
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' " Credit is taken for partial control rod insertion at time of reactor trip".
Justify this s tatement.
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Response 212._4
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Since the ejected RCCA is assumed to occur at the D Bank maximum insertion limit, it would be valid to assume that the remaining (unejected) 0 bank RCCA's are
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tripped from that insertion limit. The ejected rod worths and peaking factors were conservatively calculated at the full power insertion limit (about 40%
b in the core). However, it was assumed that the D Bank RCCA's are tripped from a 10% insertion position, which has a lower negative reactivity trip rate
[ than the 40% insertion position. Therefore, taking credit for the 10" partial control rod insertion at the time of scram is conservative and valid.
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For the zero power rod ejection cases in Section 6.5, maximum ejected rod worths and peaking factors were calculated with Bank 0 fully inserted and
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Bank C at its insertion limit of about 60%. For the same reasons given for the above full power rod ejection cases, Bank C is assumed to trip from the
[ 10% insertion position. At initiation of scram, Bank D is fully inserted and Banks A and B are at the top of the fuel. The report states that Banks A and
[ 8 are 10% inserted at scram initiation, and this error is corrected in the attached errata sheet *to this submittal.
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- The errata sheet of the March 31, 1977 transmittal letter is not included since the corrections have been made in the main text of WCAP-8846-A (October 1977).
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[ D r Westinghouse Electric Corporation Power Systems sum PittsburghPennsyNna15230
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September 26, 1977 NS-CE-1561 f -
Mr. John F. Stolz, Chief
, Light Water Reactors Branch No.1 Division of Project Management U. S. Nuclear Regulatory Comission Washington, D.C. 20555
{
Subject:
Replacement Response to Q.232.2 (WCAP-8846) f
Dear Mr. Stolz:
Enclosed are thirty-five (35) copies of Question and Response No. 232.2 related to the review of WCAP-8846, " Hybrid 84C Absorber Control Rod i Evaluation Report". This replaces the response to the same question transmitted by NS-CE-1448, dated May 26, 1977.
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Very truly yours, db
, C. Eicheldinger, Manager j Nuclear Safety Department CJR:pj Enclosure cc: M. S. Dunenfeld f
f L
A.3-1 r.
Q232.2 Control rod worths have been overpredicted for several Westinghouse reloads. This can lead to problems in maintaining shutdown margins.
What assurance can you provide that the likelihood of this occurring again does not increase with B 4C rods, in view of the fact that cal-culational uncertainties may increase with the new material and lack of startup test data? A complete answer will discuss this problem for the old as well as the new control rod material.
R232.2 The response below supplements the previously submitted response trans-mitted to the NRC (Mr. J. F. Stolz) on March 31, 1977 (N5-CE-1395).
Control Rod Worth Predictions In 1975 an examination of measured and predicted rod worths in first cores and reloads led to the conclusion that the Ag-In-Cd control rod group constants in use at that time produced about a 4% overprediction of individual control rod bank worths. Therefore, a control rod worth program was undertaken to improve the agreement between the calculations and the measurements. The program which required the cooperation of several utilities was completed in early 1977. Formal docurentation of the results and conclusions is in progress. A summary of the results is included in the following paragraphs.
The control rod worth program consisted of collecting accurate rod worth data, identifying the cause of the previous overestimate of rod worth, and developing an inproved rod worth calculation method.
The plant rod worth data consisted of boron endpoints. Since plant chemistry is not as accurate as can be obtained under controlled labvatory conditions, tne boron samples from ;he plants were shipped to Westinghouse for titrations. Isoto:':
analysis by mass spectrometer were also performed and showeo no significant doit-tions from nominal B-10 content. The critical experiments data consisted of egeri-ments conducted at the Westinghouse Reactor Evaluation Centar (WREC), the Battal'9 Nort%est Laboratories (BNWL), and at B&W.
l A. 3-2
2-The calculational method was improved by modifying the treatment of resonance absorption in the higher resolved resonance energy groups and the procedure for selecting the buffer thickness used in the transport theory calculations.
The resulting transport corrected two group constants were then used in spatial two group diffusion codes and compared to the plant and critical experiments data. The results of these comparisons, sumarized in Table 1, show excellent agreement between the measured and calculated rod worths.
A comparison of measured and predicted values for B 4 C critical experiments data is sumarized in Table 2. These comparisons show that all calculated values are lower than measured but are within 2c of the measurements.
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A. 3-3
TABLE 1 Comparison of Measured and Predicted Ag-In-Cd Control Rod Worth a
Measured Predicted Percent Worth Worth Discrepancy Core Rod Configuration ppm ' ppm (P-M)/M b
? Critical Experiment No. I 17 x 17 Rodlets 141.4+1 (140.3)ti 140.7 -0.5
$ Critical Experiment No. 2 0.400 inches 0.D. Rods 10113 103.0 +2.0 Plant A One Control Bank 11511 116.9 +1.7 Plant A One Control Bank 10811 107.3 -0.6 Plant B One Control Bank 8511 75.9 -10.7 Plant B One Control Bank 15511 150.2 -3.1 c
Plant B One Centrol Bank 4811 79.0 +64.6 Plant C Four Control Banks 42411 423.5 -0.1 Plant A Four Control Banks 42611 432.7 +1.6 Plant B Four Control Banks 42511 419.0 -1.4
- a. Error shown is lo.
- b. Critical Experiment Facility Chemistry.
- c. Bank produces a shift in power to the inside of the core where there is a large positive axial offset. This 3D effect accounts for the large overprediction.
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TABLE 2 Comparison of Measured and Predicted 8 C Control Rod Worths 4
Critical Percent Experiment Measured Predicted Discrepncy Rod Configuration Worth
- Worth (P-M)/M No. 1 0.395 inch 0.D. Rods 8.20+0.3'% 7.91% -3.5
, No. 2 0.232 inch 0.D. Rods 4.81+0.3% '
4.70% -2.3
{g No. 3 0.232 inch 0.D. Rods 6.57+0.3%
"' 6.21% -5.5 No. 4 0.232 inch 0.D. Rods 5.98+0.3% 5.71% -4.5 No. 5 0.438 inch 0.D. Rods 117+ 3 ppm
\
- 111 ppm -5.1
- Ermr shown is lo.
I ANO-109.002.218 Reference 15 l
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