ML20054F709
| ML20054F709 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 06/14/1982 |
| From: | Longenecker J ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT |
| To: | Check P Office of Nuclear Reactor Regulation |
| References | |
| HQ:S:82:046, HQ:S:82:46, NUDOCS 8206170239 | |
| Download: ML20054F709 (100) | |
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Department of Energy F Washington, D.C. 20545 f Docket flo. 50-537 F HQ:S:82:046 JUN 1 4 l'I2 -
P Mr. Paul S. Check, Director CRBR Program Office -
Office of Nuclear Reactor Regulation -
U.S. fluclear Regulatory Commission Washington, D.C. 20555 _
Dear Mr. Check:
RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION
Reference:
Letter, P. S. Check to J. R. Longenecker, "CRBRP Request for E Additional Information," dated February 19 and 26,- -
March 11,15, 23, and 25; and April 9,1982
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This letter formally responds to your request for additional information -
contained in the reference letter.
Enclosed are responses to Questions CS 210.4, 8, and 10; CS 220.3, 5,10, , [
15, 25, 30, and 35; CS 250.1 and 3; CS 410.2, 3, and 19, CS 421.1, 5, 8, 9, F 14,17,18,19, 20, and 23; CS 491.18; and CS 490.11 and 35; which will L -
also be incorporated into the PSAR Amendment 69; scheduled for submittal later in June.
Sincerely, E J n R. Longene h r Acting Director, Office of the -
Clinch River Breeder Reactor Plant Project F Office of Nuclear Energy E-Enclosures ;
Y cc: Service List L Standard Distribution L Y[
Licensing Distribution 8206170239 820614 E
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Page 3 (82-0287)[8/22]#35 ouestion cs210 1 The ASE Section 111 Code does not require a f atigue evaluation for Class 2/3 piping design. However, for operating at elevated temperature, creep f atigue ef f ects may become severe in Class 2/3 piping. Describe the method.used f or creep f atigue evaluation f or Class 2/3 piping, or justif y why such -
consideration is not needed. Furthermore, verify whether rules in Code Case N-253 are met.
Resoonse The considerations given to creep-f atigue ef f ects in elevated temperature Class 2/3 piping can be clarlfled by discussion of the structural criteria available to evaluate these designs. In some elevated-temperature Class 2/3 piping lines, Code Case N-253 is planned to be adopted as an acceptable
c ta f nr demonstrating protection against unacceptable creep-f atigue dmnage. The project has developed a modification to RDT Standard F9-4T for application to ASME Code Section lil, Class 2 and 3 components and piping (Attachment 1). These project-developed criteria, (termed "F9-4TMOD" in the remainder of this response) is the generally applied approach for evaluating creep-fatigue. Both of these approaches will be discussed separately following an identification of where the two approaches are planned to be used. Where possible, an estimate will be provided of the potential for a design satisfying the F9-4TM)D rules to also satisfy Code Case N-253.
Bef ore proceeding with discussions of the two approaches, the first sentence of the NRC question needs to be put into perspective. It is acknowledge that ASME Section lli design rules for low tanperature Class 2 and 3 ploing (NC/ND-3600) do not require an exollett, evele-by-evele f atigue evaluation.
However, satisf action of stress range limits f or thermal expansion loeds (Equation 10) and combined pressure, weight, and thermal expansion stress The limits (Equation 11) uses the allowable expansion stress range value, S3.
determination of Sg is dependent on the cyclic service conditions by introduction of a stress range reduction value, f. The value of f was developed using f atigue test data and accounts for all temperature cycles l
during life. Since temperature cycling is the principle cyclic load for low-temperature piping, It is concluded f atigue ls addressed in the low-temperature Class 2 and 3 piping rules. The only shortcoming of the low-temperature rule approach might be the absence of " thermal gradient" stress in the stress range evaluations. As will be seen, these gradient stresses are picked up In most of the elevated tanperature evaluations.
For organizational clarity, four operating regimes for the Class 2/3 piping are def ined as f ollows:
Regime L PIolng - Piping that will never experience elevated temperatures 3nder Normal, Upset, or Emergency operating conditions, i.e., It's entire lif e, except possibly for Faulted events, is spent at temperatures f or which stress values exist in ASME B&PV Code, Section Ill, Appendix 1.
Regimes E1. E2. E3 Ploing - Piping that will at some time during Normal, Upset, or Emergency conditions experience elevated tanperaf'ures.
QCS210.4-1 Amend. 69 I May 1982
PM (82-0257)[8/223d35 Ragtme El Pining - Piping that will not be subjected to significant cyclic duty, i.e., it can meet certain exemption criterla f rom cyclic damage evaluations.
Ragtme D Pining - Piping that rit t not be subjected to significant elevated temperature service, i.e., it can meet certain material time-temperature limits that preclude significant creep ef fects from occurring.
Regime E3 PIntng - Piping that will be subjected to signif Icant elevetsd temperature service.
The majority of the Class 2/3 piping f alls Into Regime L. The low-temperature (NC/ND-3600) cyclic evaluation methods have been discussed, they address f atigue and will not be discussed f urther since the NRC question is basically concerned with Regimes E1, E2, and E3.
A list of elevated temperature Class 2/3 lines are:
- 1) The Steam Generator System (SGS) watsr/ steam piping which is a Class 3 piping system is anticipated to be in Regime El and thus exempted form creep-f atigue evaluations. The exemp+1on criteria contained in the piping specif Ication are Identical to those contained in Appendix A of Code Case N-253. Should it prove necessary to perf orm a creep-f at' gue avaluntion, the eriterIa of Code Case N-253 would be const dered for Impl ementation.
- 2) The Intermediate Heat Transport System (IHTS) hot leg and super heater dump lines downstream of the second valves are Class 3 piping lines which are expected to be in Regime E2. They have been optionally upgraded to Class 2. There exists gne emergency design event with fluid temperatures in excess of 800 F. It is expected this event will be exempt f rom elevated temperature creep and f atigue evaluation. If elevated temperature creep and f atigue evaluations become necessary, they will be perf ormed in accordance with F9-4TK)D.
- 3) The SGS leak detection module piping between the sodium isolation valve and the LDS modules is instrument piping optionally upgraded to Class 3 and expected to be in Regime F2. Design requirements are not finalized yet but it Is anticipated the Code Case N-253 rules will be invoked.
- 4) The Steam Generator Auxillary Heat Removal System (SGAHRS) piping between the superheater and the vent control valve is Class 3 piping expected to be in Regime E3. It will be designed to Code Case N-253.
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) 5) The SGS leak detection module piping between the IHTS and sodlum I isolation valve is Class 2 piping which has been optionally upgraded I ta Class 1 and will be designed to Code Case 1592 creep-f atigue criteria.
- 6) The IHTS main and auxiliary piping upstream from the second valve is Class 2 and 3 piping which has been optionally upgraded to Class 1 and wIII be designed to Code case 1592 creep-f attgue eriterIa.
QCS210.4-2 Amend. 69
P go 5 (82-0287)[8/22]i35
- 7) The impurity Monitoring and Analysis system (IMAS) and inert Gas Receiving and Processing System (lGRPS) contain several lines in Regime E2 which will be designed to F9-4TM)D. These lines include:
a) Primary sodium sampling loop ,
b) Sodlum sampling line to and f ran the Intermediate Sodlum Characterization System c) Equalization line between the primary sodium overflow vessel and the reactor vessel d) Primary pump equalization lines to the reactor equalization line e) Equalization and overflow vent iInei, to the vapor condensor f) Primary pressure relief line from the equalizer line to the overflow vessel g) The overflow vessel vent line.
- 8) The entire Intermediate Sodium Processing System (ISPS) except for dump lines is in Regime E2 and is being designed to F9-4TMOD.
- 9) The IGRPS Inlet lines to the vapor condensors, drain lines f rom the vapor condensors, and drain lines f ran the Intermediate Heat Exchanger are in Regime E3 and are being designed to F9-4TM)D.
- 10) The Auxillary Liquid Metal System ( ALMS) dis::harge piping f rom the primary sodium cold traps in the overflow makeup circuit will be designed per the Regima E3 requirements of F9-4TM)D. Additionally, this piping will have insignificant thermal transient stresses and is in Regime E2, per the time-temperature Iimits of both F9-4TMOD and Code Case N-253, thus the criteria provides essentid ly the same protection as the Regime E2 coverage of Code Case N-253.
l 11) Low-stressed Type 304SS portion of the ALMS Inlet piping to the primary sodium cold traps which are in Regime E3 will be designed to l the Code Case 1592 elastic method creep-f atigue criteria.
l High-stressed portions utilize Type 316SS end are evaluated per the Regime E3 per the time-temperature limits of both F9-4TMOD and Code Case N-253, which with the Insignllicant thermal transient stresses In this line, provides essentially the same protection as the Regime E2 coverage of Code Case N-253.
Discussions will now center of the various evaluative methods for creep-f atigue used in the Regime E1, E2, and E3 piping. Discussions will not l
be provided of Code Case 1592 methods since the NRC question is concerned with Class 2/3 methods and Code Case 1592 is a Class 1 methodology, which specif ically addresses creep-f atigue.
Regime E1 (Insignificant Cvelle Service) Evaluative Method 1
i The Code Case N-253 Appendix A creep-f atigue exemption criteria can be I summarized as f ollows:
l l Creep-f atigue evaluations ara not required when the maximum peak strain range during service is below the permitted value f or 10 cycles f ran the Code Case 1592 continuous cycling f atigue curve and, less than 25 "signif Icant" loed cycles exist during Level A or B service. A detailed QCS210.4-3 Anend. 69 May 1982
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P ge 6 (82-0287)[8/223#35 def inition of "signif icant" is provided to assure consistency in rule appl ication. Although Code Case N-253 limits this exemption to -3200 vessels, a similar 25-cycle exemption exists In F9-4TM00. A justif ication f or the F9-4TM)D 25-cyele exemption erIterIa for any component Is also provided in F9-4TM)D and with only minor modification can be used to justify use of the Code Case N-253 exemption f or piping. ?
Ragtme E2 (fnsfgnificant Creen Effects) Evaluative Methods RDT F9-4TMOD The Class 1 elevated temperature Code rules (Code Case 1592) recognize under certain combinations of stress level, time, and temperature, that creep ef f ects are minimal, and these ef f ects on cyclic lif e can be addressed by simple modif ications to the low-temperature design methodology. The Class 1 time-temperature limit (T-1325) which def ines this regime was adopted by the CRBR Project as Appendix E to F9-4TM)D. Satisf action of this time-temperature limit allows the piping designed to modify and utilize the low-temperature Class 1 design methodology (bB-3600), which includes an explic!t, cycle-by-cycle f atigue evaluation (Equation 11) plus consideration of thermal gradient stresses, not normally included in Class 2/3 evaluations. The modif ication also includes use of the Code Case 1592 " hold-time" f atigue curves (Figure T-1430) which has built into them the worst-case ef fects of creep on cycl ic damage. Theref ore, potential creep-fatigue is directly accounted f or in this F9-4TMOD design route, even though creep ef fects are restricted to minimal levels.
Egge Case N-253 The Code Case N-253 design route includes a time-temperature limit similar to F9-4TMOD and Code Case 1592; however, if met, it permits a modif ied Class 2/3 (NC/ND-3600) methodology to be used. Modif Ications do not require thermal gradient stresses to be considered nor explicit, case-by-case f atigue evaluations to be perf ormed. The logic f or this approach is that the time-temperature exclusion criteria means that creep ef fects are not significant and that piping in this category is really an extension of Regime L.
Regime E3 (Potentially Significant Creen Effects) Evaluative Methods RDT F9-4TMOD Two evaluation approaches are permitted in F9-4TMOD:
- 1) Meet the Code Case 1592 creep-f atigue criteria either using elastic analysis methods (requiring elestic analysis method strain limits also be met) or using Inelastic analysis methods (RDT F9-5T). This directly addresses creep-f atigue damage in a manner identical to that of a Class 1 component without upgrading the design to Class 1.
QCS210.4-4 Amend. 69 May 1982
Page 7 (82-0287)[8/22]d35
- 2) Perf orm a low-temperature Class 2 (NC-3600) evaluation using Section Vill Division I allowables (per Code Case 1481), but reduce the S g value of NC-3611.2 by the largest value of th CT; + C3Ea, (me.T. - %TL ) -
which occurs at each location during the specif led Normal, Upset, and Emergency operating conditions. This approach is thus similar to that used by B31.1 at elevated temperature, but it also includes an accounting of the potential deleterious ef fects of thermal gradients in addition to restrained thermal ex;,ansion.
Code Case N-253 provides the following approach. The design rules f or iow-temperature piplng (NC/ND-3600) must be met wIth the folIowing modifications:
- 1) The stress reduction f actor, f, Is determined using Appendix B to Code Case N-253, where f atigue f actors are determined as a f unction of material and temperature.
- 2) After excluding 25 cycles from evaluation, a ilmit is placed on acceptable thermal expansion plus thermal gradient stress range.
- 3) A modification is made to the NC-3600 Equation 11 allowable on thermal expansion range (to be the lessor of the current limit and a new limit l
based on material yleid strength).
Satisfaction of Code Case 4-2 % Rules l
As noted previously some portions of the CRBR Class 2/3 piping are or will be evaluated directly in accordance with the rules of Code Case N-253. Other portions of the Class 2/3 piping are or will be evaluated using RDT F9-4TK)D or optionally upgraded to Class 1.
In summary, a methodology has been developed and described in the above in order to provide eroop-f att gue avaluatton for ASME Class 2/3 plping. This methodology is structured to the specific piping in question. This pre-empts the question of across the board compilance to Code Case N-253.
QCS210.4-5 Amend. 69 May 1982
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Mac/drl Q, cs z/41 A CRBRP MODIFICATION TO RDT STANDARD F9-4 j FOR APPLICATION TO ASME CODE SECTION III, CLASS 2 and 3 COMPONENTS AND PIPING May 1975 4
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t O. FORWARD In the title, and in subsections 0.1, 0.1.1, 0.2 and 0.5.4, change the phrase
" Code Cases 1592,1593,1594,1595 and 1596" to read,
' " Code Cases 1481,1592,1593,1594,1595, and 1596" -
In 0.1 Scope, change to read, ". . . Classes 1, 2 and 3 nuclear components. . ".
In Subsection 0.5.2, change to read: Code Case 1592-2 ~
Code Case 1595-1 In Subsection 0.5.2, add the following: '
" Code Case 1481 Elevated Temperature Design of Section III Class 2 and 3 Components".
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" Code Case 160,6-1 Stress Criteria,Section III Classes 2, 3 Piping Subject to Upset, Emergency, and Faulted Operating Conditions."
" Code Case ,1607-1 Stress Criteria for Section III Class 2 and 3 Vessels Designed to NC/ND-3300 Excluding the NC-3200 Alternate".
" Code Case 1635-1 Stress Criteria for Section III, Classes 2 and 3, Valves subject to Upset, Emergency and Faulted Operating Conditions".
In subsection 0.5.4, change the phrase "to ' Code Case 1592-O' (or 1593-0,1594-0,1595-0,1596-1)"
to read.
"to ' Code Case 1592-2' (or 1593-0,1594-0,1595-1,1596-1,1481-0)"
l 5.0 SUPPLEMENTS TO CODE. CASE 1481 For RDT F9-4 applications, any reference in Code Case 1481 to the rules of sub-sections NC or ND shall mean the rules as supplemented by RDT E15-2 NC and RDT E15-2 ND, respectively. Where this supplement references Code Cases 1592 it '
means RDT F9-4.
the requirements of Case 1592 as supplemented by the applicable portion of l
Where this supplement references NB-3000 it means the requirements of Article NB-3000 of Section III 's supplemented a by RDT E15-2B.
The rules of Code Case 1481 shall be supplemented by the following additional requirements.
(2) Delete this paraoraph in its entirety.
Add the following additional paragraphs:
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(5) Applicability '
When this standard is invoked for Class 2 and 3 Components, it is. applicable only to vessels, piping, and valves. Electro-magnetic pumps may be treated as either piping or vessels. ~
These rules apply only to Class 2 and 3 components constructed fron Types 304 and 316 austenitic stainless steel, Ni-Cr-Fe Alloy 800H, and 2-1/4 Cr-1 Mo -
ferritic steel pemitted in Code Case 1592. For components constructed from other materials, specific desian rules shall be. provided in the Design Specifi-cation. When the temperature of integral component supports exceeds the upper temperature limit of Section III for the component, the component boundary shall be extended to include the elevated temperature portion of the support. These rules are not applicable to components containing pad type nozzles or non-integral attachments. Socket welds shall not be used for Class 2 or 3 Components which -
contain sodium or radioactive fluids with the sole exception of instrumentation b mel) lines.
(6) Basic Requirements The supplemental Class 2- and 3 requirements of this standard shall be applied only to components which meet the requirements of Section III via paragraph (1) of Code Case 1481. For Upset, Emergency, and Faulted Operating Conditions, Code Cases 1607,1635, and ,1606 shall apply for vessels, valves, and piping respectively.
~ The requirements of Code Cases 1607, 1635, and 1606 shall be satisfied using the 5 values from Section VIII, Division 1, per paragraph (1) of Code Case'1481.
The requirements of Appendix B of this standard shall be satisfied in addition to the external pressure limits of NC/ND-3133. Prior service experience or experimental demonstrations may be used in lieu of analysis. When a buckling evaluation of an identical component for equivalent or more severe service to Code Cases 1331-5, 1331-6, .1331-7,1331-8, or 1592 is available, and when a supplemental stress report (certified by a Registered Professional Engineer) is pr.epared which reconciles the previous evaluation with the current specification, the requirements of Appendix B are satisfied. The evaluation to the particular Code Case must be complete and shall demonstrate satisfaction of all the Code Case buckling requirements.
I Only the Normal Operating Conditions need be considered when detemining the most severe condition of coincident pressure and temperature per NC-3112.1 and ND-3112.1, The owner shall supply the modifications to the design limits which are appropriate for the ma.terials of construction in the specified service environment.
(7) ' Suppl'emental Reouirements:
Exemption from Ratchetting and Creep-Faticue Analysis Ratchetting and creep-fetigue evaluations of the specified Nomal Upset, and Emergency Operating Conditions are required unless one of the following require-ments is satisfied:
a) The number of significant load cycles is less than 25. Appendix A of this standard defines the tem "significant". This procedure may be used only where the peak strain ra~nge is less than the limit given in Appendix A of this standard.
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b) Prior service of equivalent severity is used as the basis for demonstrating structural integrity. Appendix C of this standard describes the " Prior Serv ^ Report" which shall be prepared when this procedure is utilized.
c) Testing of identical components under equivalent or more severe conditions is utilized as the basis for demonstrating structural integrity. Appendix D of -
this standard describes the Experimental Test Report which shall be prepared when this procedure is utilized.
d) When an evaluation of an identical component for equivalent or more severe service to Code Cases 1331-5,1331-6,1331-7,1331-8, or 1592 is available, and when a Supplemental Stress Report (certified by a Registered Professional Engineer) is prepared which reconciles the previous evaluation with the current specification, these requirements are satisfied." The evaluation to the particular (1331 or 1592) Code Case must be complete and shall demonstrate satisfaction of all of the Code Case ratchetting and creep-fatigue requirements.
c) for Class 3 components only, the owner may, via an explicit statement in the Design Specif.ication, exempt the component from ratchetting and creep-fatigue evaluation.
(8) Suppl ~emental Requirements:
Ratchetting and Creep-Fatigue Analysis
- When ratchetting and creep-fatigue evaluations are not exempted by (7) above, one of the following requirements shall be satisfied:
a) An evaluation which demonstrates compliance with all of the requirements of T-1300 and T-1400 of Code Case 1592 as supplemented by RDT Standard F9-4.
The alternate procedures in RDT F9-5 may be used. Note that the use of an elastic creep-fatigue evaluation procedure of T-1430 of Code Case 1592 is pennited only when the limits of T-1320 are satisfied. In NT-1325 Test No.
4(a) the extrapolation may be performed using the maximum slope at the current metal temperature.
i b) The temperature-time limits of Appendix E of this standard and the appropriate l requirement below:
l l Vessels: All of the requirements of NC-3200 with the elevated tempera-t ture S mvalues obtained from Appendix I-14 of. Code Case 1592 and with the design fatigue curves extended to elevated temperatures via Appendix G of this standard. If NC-3219.2 i
is utilized, the modification of Appendix F of this standard shall be employed. NC-3219.3 shall not be used.
Valves: The exemptions of Conditions A or B of. NC-3219.2 as modified by Appendix F of this standard shall be satisfied, or a detailed fatigue analysis shall be made in accordance with the rules of Apnendices XIII and XIV where the Sm values at elevated l
l temperatures shall be obtained from Appendix I-14 of Code Case 1592 and the design fatigue curves at elevated temperatures shall be obtained via Appendix G of standard, or the limits of NB-3512.2 (a) or (c) using, at elevated temperatures, the S values from Division 1,Section VIII per paragraph (1), the pressure-temperature ratings of ANSI B16.5, and the fatigue limits of Appendix G of this standard.
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_4 Piping: The limits of NB-3600 using, at elevated temperatures, the h values from Code Case 1592 and the design -fatigue curves of Appendix G of this standard.
c) For piping, the limits o'f NC-3600 when the basic allowable stress values at elevated temperatures are obtained from Division 1,Section VIII per paragraph (1) and the SA value.of NC-3611.2 (c) is reduced by the largest value of 1/2[2($v)lAT) l+CE 3 ab ! "aT , - a bb T !)
which occurs at each location during the specified Normal, Upset, and Emergency Operating Conditions (the terms are defined in NB-3653.2). *
(9) Design Reports Design reports shall be prepared and submitted to the owner which document all of the evaluations performed in addition to the minimum required for stamping by Sub-sections NC and ND and Code Case 1481. These reports shall follow the ASME Code guidelines for Stress Reports. The reports shall be certified by a Registered Professional Engineer as meeting the requirements of the Design Specification.
The fabricator numbers shall of cycles provide greater thanthe 10gustification
. for the fatigue limits employed for
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APPENDIX A Significant l'oad Cycles A load cycle is "significant" when any of the following is true: -
a) The . variation in the primary stress is greater than 25% of the' maximum allow- '
able primary stress as defined by Code Case 1481.
b) The secondary stress range is greater than 50% of the limiting value (2Sy at Section III temperatures or T-1300, Code Case 1592 at elevated temperatures).
c) The estimated peak stress (K" = 2.5 when local st,ructural discontinuities exist, unless o amplitude at 10gherwise justified) is greater than twice the allowable stress cycles from Figure I-9 of Section III (after the environmental effect correction is applied to the design fatigue curve.)
This procedure is invalid when the maximum estimated peak strain range from any one. cycle exceeds the maximum allowable value for 10 cycles as obtained from the design fatigue curve (Figure T-1430 of Code Case 1592) at the maximum metal temperature of the cycle. When the maximum metal temperature is below the value for which Code Case 1592 provides design fatigue curves, the allowable strain range may be obtained from Figure I-9 of Section III (Subsection NA) by dividing th) allowable stress amplitude, Sa. 'IUr one half of the Figure I-9 Young's Modulus.
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APPENDIX B Bucklino Limits -
1.0 General Requirements shall be maintained.Where buckling is a potential failure mode, the following m 1.1 specific geometrical configurations under specific loading These Section permittedIIIbylimits includetolerances.
fabrication the effects of initial geometrical imperfections the effects of creep due to long-term loadings at elevated temperatur the effects of the other loads or geometries. These rules provide additional limits which are applicable to general configurations and loading conditions of themay that cause buckling or instabi1Hy due to time-independent creep behavior material.
and Operating conditions.These additional limits are applicable to all specified Design 1.2 For the limits specified here, distinction is made between load-controlled buckling and strain-centrolled buckling. Load-controlled buckling is characterized by continued application of an applied load in the post-buckling regime le catastrophic f.ailure as exemplified by collapse of a tube under external pressure.
Strain-controlled buckling is characterized by the immediate reduction of strain-induced deformations.
resulting ioad upon initiation of buckling, and by the self-limiting nature of the Even though it is self-limiting, strain-controlled buckling must be avoided to guard against failure by fatigue, excessive strain, loss of functionality trolled instability. due to excessive deformation, interaction with load-con-t 1.3 For conditions under which strain-controlled and load-controlled buckling may interact, the Load Factors applicable to load-controlled buckling shall be against buckling in the interactive mode.used for the combination of loa 1.4 For conditions where significant elastic followup may occur, the Load Factors controlledapplicable buckling.to load-controlled buckling shall also be used for strain-1.5 For load controlled buckling the effects of initial geometrical imperfections and tolerances shall be considered in the time-independent calculations of-paragraph 2.1 of this appendix; and the effects of geometrical imperfections and tolerances, whether initially present or induced by service, shall be considered in the time-dependent. calculations of paragraph 2.2 of this appendix.
1.6 For purely' strain-controlled buckling the effects of geometrical imperfections and tolerances whether initially present or induced by service, need not be considered in calculation of the instability strain. However, if significant 9eometrical imperfections are present initially, enhancement due to creep may cause excessive deformation or straih.
the application of deformation and strain limits.These effects shall be considered in l
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s 1.7 The expected minimum stress-straih curve for the material at the specified temperatures shall be used. The expected minimum value may be obtained by normalizing the appropriate average hot tensile curve of Fig. T-1800 of Code Case 1592 to the minimum yield strength given in Fig.1-14.5 of. Code Case 1592.
1.8 The limits of both 2.1 and 2.2 shall be satisfied for the specified Design and Operating conditions. - -
2.0 Buckling Limits 2.1 Time-Independent Buckling For load-controlled buckling, the Load Factor, and for strain-controlled buckling, the Strain Factor; shall equal or exceed the values given in Table B-1 for the specitied Design and Operating conditions to protect against time-independent (instantaneous) buckling. .
TABLE B-1 TIME-INDEPENDENT BUCKLING LIMITS Load Factor (j) Strain Factor (l) (3)
. Design Conditions 3.0 52) 1.67 Operating Cond.itions ,
Normal 3.0 ' l.67 Upset 3.0 1.67 Emergency 2.5 1.4 Faulted 1.5 1.1 Testing I4) 2.25 1.67 (1) Load (Strain) , Load (strain) which would ,
Design or expected Factor. cause instant instability load (strain).
at the design or actual operating temperature.
(2) Changes in configuration induced by service need not be considered in calculating the buckling load. ,
.(3) For themally induced strain-controlled buckling, the Strain Factor is applied to loads induced by thermal strain. To determine the buckling strain, it may be necessary to artificially induce high strains concurrent with the use of realistic stiffness properties. The use of an " adjusted" thermal expansion coefficient is one technique for enhancing the applied strains without affecting the asscciated stiffness characteristics.
(4) These factors apply to hydrostatic, pneumatic, and leak tests. Other types of tests shall be classified according to-3113.7 of Code Case
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2.2 Time-Dependent Buckling To protect against load-controlled time-dependent creep buckling, it shall be demonstrated that instability will not occur during the specified lifetime for a load history obtained by multiplying the specified Operating condition loads by the factors given in Table B-2. A design factor is not required for purely strain-controlled buckling because strain-controlled loads are reduced concurrently with resistence of the structure 'to buckling when creep is significant. The time ,
temperature limits of Appendix E of this standard may be used to detennine whether time-dependent buckling need be considered.
TABLE B-2 TIME-DEPENDENT LOAD-CONTROLLED BUCKLING FACTORS Operating Conditions Normal 1.5 Upset 1.5 Emergency 1.5 Faulted 1.25 D
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APPENDIX C Prior Serv' ice. Experience A Prior Service Experience report may be prepared to demonstrate that the prior service experience of identical components was at least as severe -
as that specified in the Design Specification. In this manner, successful prior experience can be used in lieu of ratchetting and creep-fatique analysis for cases where the prior service is at least as severe as that currently specified. The Prior Service Experience Report shall provide a comparison of the proposed design and the design with which the successful service experience was accummulated to demon-strate that they are identical. The report shall also demonstrate that the prior service was at least as severe as that currently specified. Among the factors which are to be used to demonstrate equal load severity shal] be the maximum metal temperature and the total duration above the temperature limits of Section III. The Frior Service Experience report shall be certified to be completed and correct by a Registered Professional Engineer. The Prior Service Report shall be attached to the Design Report.' -
APPENDIX D Experimental Demonstration of Integrity Experimental tests of components may be used to demonstrate their structural integrity in lieu of a ratchetting and creep-fatigue analysis.
The experimental tests shall be performed using test conditions which are at least as severe as those identified in the Design Specification. The number of tests and degree of test severity- (beyond that defined in the Desig.n Specifi-l cation) shall be sufficient to demonstrate satisfactory component service. An l Experimental Test report shall be prepared which relates the test conditions to
! the specified service conditions and demonstrates that the test results are a l good and proper simulation of the specified service conditions. The report l
shall contain thermo-hydraulic calculations, where applicable, as well as structural calculations. The Experimental Test Report shall be certified by i
a Registered Professional Engineer to be correct and complete.
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-f0-APPENDIX E Temperature-Time Limits 1.0 When these limits are satisfied, creep effects do not need to'be accounted for in the evaluation of the primary plus secondary and primary plus secondary plus peak stress intensities.. Their use, however, requires that the fatigue limits be modified. Faulted Operating conditions (when specified) need not be -
considered in these evaluations. ,
The fatigue damage sum, when used, shall be limited to 0.9 instead of 1.0.
2.0 This limit is satisfied when both -
y(gi ) 1 0.1 id and f(cj)1 0.2%
where ti is the total duration of the metal temperature Tj, during the specified design lifetime and tid is the minimum time to rupture at a stress level of 1.5 times the minimum, yield strength at that temperature (1.5 SylD). The minimum stress to' rupture charts and' figures of Code Case 1592 are a valid data source.
If extrapolation to greater load durations is necessary, the extrapolation shall be performed using the greatest slope for that material on the plot of the loga-rithm of the minimum stress to ru ture vs the logarithm of the time to rupture for that metal temperature. The tj shall equal the specified design lifetime.
The ci value is the thermal creep strain which is accumulated due to the imposition of a uniaxial stress of 1.25 times the minimum yield strength (1.25 Smin) at the associated metal temperature, Ti, during the time durati.on', ti . Creefhardening from one time period (tj) shall not be accounted for in other time periods. The isochronous stress-strain curves of Appendix T of Code Case 1592 may be used to evaluate this limit.
3.0 Figures E-1 through E-4 may be used to evaluate these limits. The specific curves are based on Code Case 1592 Spin and SR values plus the thermal creep -
equations of the Nuclear Systems Materials Handbook.
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APPENDIX F Modifications to NC-3219.2 Fatique Exemption Rules ~
When be the procedure of NC-3219.2 is employed, the following modifications shall incorporated. -
Condition A
- 1) The S mvalues for elevated temperatures shall be obtained from Code
~
Case 1592.
2)
In NC-3219.2(c), the absolute value of aT) shall 'be a' dded to the change in metal temperature between two adjacent points, AT, before detemining the cycle factor.
The tem AT' is the maximum equivalent linear through the wall temperature difference (seeNB-3653.2(b)). All specified Nomal, Upset and Emergency conditions shall be considered. '
Condition B 6
- 1) The reduction in the allowable Sa value beyond 10 cycles shall be con,sidered.
- 2) The allowable Sa value shall be' reduced, as show'n below, to accoun+. for the maximum through the thickness thermal gradient during all specified Nomal, Upset and Emergency Operating conditions.
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Sal-Se-1/2[2ki,)l'^T113 APPENDIX G ElevatedTemperatureDeskonFatiqueCurves
- When this Appendix is invoked, the design fatigue curves of Figures T-1430 of Code Case 1592 shall be used, where Sa is one half of the product of the total strain range, et, and the Young's Modulus, E, at that temperature.
- 11. all cases the design fatigue curve shall be modified by the appropriate environmental effect corrections prior to use.
In all cases the ap amplitude beyond 10gropriate cycles shallreduction, if any, in the allowabi'e stress (or strain) be employed.
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A Basis for the Twenty-five Cycle txclusion This note provides an evaluation of the potential for failure of components which meet the twenty-five cycle creep-fatigue exclusion.
The minimum number of cycles of Figures T-1430 of Code Case 1592 s ten.
Assume that twenty-five cycles at a strain range equal to the strain range of Figure T-1430 at ten cycles are specified. The mean failure fatigue curve ,
is at least a factor of twenty on life above the design curve. Thus a strain range which is associated with ten cycles on the design fatigue curve is also associated with (10 x 20) 200 cycles on' the mean failure fatigue curve. It is judged that the minimum failure fatigue curve lies not more than a factor of four on life below the mean failure fatigue curve. Thus, a strain range which is associated with ten cycles on the design fatigue curve is also associated with (10 x 20 + 4) 50 cycles on the mean failure fatigue curve. Thus, twenty- ,
five cycles failure fatigue of that strain range is still : factor of two below the lower bound curve.
The effect of temperature is already accounted for by the temperature effects already built-into the Figure T-1430 design fatigue curves. The effects of slow straining rates and effects of hold periods at maximum strain are built into the Figure T-1430 design fatigue curves.
Thus, all of th'e factors thought to be important in determining fatigue life have been considered. A safety factor of tv;o on life exists between the allow-able number of significant cycles and the loser bound failure fatigue curve.
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JUSTIFICATION FOR USE OF STRESS RUPTURE EXTRAPOLkTION PROC Introduction In Appendix E of the CRBRP Modification to RDT Standard F9-4 (for bse with /
Class 2'and 3 equipment) a time-temperature threshold is' provided. When the time-temperature combination is below this limit, it is judged that low tempera-ture fatigue and ratcheting rules may be used. Amon~g the criteria used to define this threshold is the locus of points at which the time duration divided by the
" allowable time duration" at 1.5 times the minimum yield strength is equal to 0.1.
The " allowable time duration" is defined as the minimum time to rupture.
..__ :. 132 provides the minimum time to rupture vs. ctress and temperature (Sf") over a range of stress and temperature values. The yield strength varies relatively slowly with temp _erature while the minimum stress to rupture value is very temperature sensitive.
Thus, at low temperatures (400 F austenitic steels, 4700 F low alloy steels) 1.5 times the minimum yield strength may be smaller than the minimum stress to cause rupture, even at 1.0 x 105 to 3'x 105 hours0.00122 days <br />0.0292 hours <br />1.736111e-4 weeks <br />3.99525e-5 months <br />.
Thus, extra'polation of the stress-rupture curves to time above 3 x 10 5 $3
, necessary.
Objective The purpose of this note is to document the reasons for employing a stress-rupture extrapolation technique which is less conservative than that in Code Case 1592.
' Austenitic stainless steels, among others, can exhibit a slope discontinuity in their stress-rupture behavior. At long lifetimes (or high stress levels, or high temperatures) the linear relationship between the logarithm of the stress vs. the logarithm of the lifetime may decrease abruptly. In the practical use of stress-rupture data there are many cases where low stress applications are encountered. The designer is faced with the task of determining the damage due to these low stress it.. s. A conservative method for extrapolating the stress-rupture curves was needed.
If the curves were extrapolated linearly based on the maximum slope at the temperature of use, the results might be unconservative if a slope discontinuity existed just beyond the maximum lifetime of the curve.
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If, however, the curve was extrapolated at a slope equal to the highest value for all temperatures, there is great assurance that the result is conservative.
The second method, extrapolation at the maximum slope for all temperatures is believed to be conservative for all temperatures up to a few hundred degrees below the maximum for the curve. For types 304 and 316 austenitic stainless i steel, the maximum metal temperature for which Code Case 1592-3 provides minimum stress-to-rupture values is 1500*F. Since the LMFBR is not expected to operate with metal temperatures above 1200 F, the use of the 1500*F slope to extrapolate 1200*F stress rupture data is expected to be safe. .
In the case of time-temperature limits (below which low-temperature ratchetting and fatigue procedures are valid) the use of the stress-rupture extrapolation using the slope at the maximum temperature of the table is unrealistic. At low stress and temperatures, the stress-rupture slope discontinuity is not expected to occur at all. In 304 S/S the initial stress-time dependency of failure is .not even observed until lifetimes of greater than 104 hours0.0012 days <br />0.0289 hours <br />1.719577e-4 weeks <br />3.9572e-5 months <br />.- The existence of a slope discontinuity nearly is not expected. Even if a slope discontinuity were to occur just beyond the Code Case's time limit (3 x 105 hrs) l its slope is not expected to be as high as that for 1500*F.
1 The use of the maximum temperature slope extrapolation method, when applied to the Appendix E threshold, results in a severe and unwarrented design limitation I
at low service temperatures, which would have an unnecessary cost impact on the CRBRP Project. The Appendix E threshold is shown below:
Allowable Load Duration, Hours .
Temperature 304 US !
316 S/S I
'*F Max. Temp This Temp ' Hax. Temp This Temp j 800 F 200,000- 2,000,000 ,
900,000 large 850*F 80,000 80,000 400.000 l'arge 900 F 27,000 27,000 140,000 1,200,000 950 F 4,200 4,200 43,000 68,000 l
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- Summary The stress-rupture extrapolation procedure of the Code Cow is inappropriate for this use. The only place where it is more r'estrictive than current metal temperature linear extrapolation procedure is at low temperatures. (800-900'F). .
The reasoning behind the maximum temperature slope extrapolation procedure is not applicable at low temperatures. The result of using the Code procedure would be an unnecessary and expensive reliance on creep analysis for very low temperature applications.
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Paga 12 (82-0287)[8/22]435 Question OCS 210.8 Provide rules and acceptance criteria used in the design of mechanical component supports, especially those supports for piping at elevated tenperature service. -
Resocrse Camponent supports as defined by ASME Section III, Subsection NP include structural elenents, which carry the weight of cmponents or provide them with structural stability, or both. Se term includes hangers, supports, braces, snubbers and other devices which are designed to transmit loads frca:: the j pressure retaining barrier of the caponent to the load carrying building structure during any of the specified operating conditions. %erefore, l emponent supports intended to conform to the requirenents for Class 1 c-*rnc&im as set forth in ASME Section III, Subsection NB, are Soverned by Subsection NF, as supp1mented by RDT Standards. %ese hnnets provide the !
rules for the materials, design, fabrication, examination and testing of l cmponent supports. j QCS210.8-1 Amend. 69 May 1982
Page14(82-0287)[8/22]i35 g estion C D10 1D .
Describe the basis of " leek without breek" critoria usedsince in Section 3.6 of operating CRBRP-PSAR for the primary and secondary sodium loops.
temperature in the hot legs are significantly higher than the cold legs and recondary loops, explain why the same pipe leakage criteria is applicable.
Response
The appIIcability of the concept of leak bef ore break to the CRBRP hot leg piping in the PHTS cells of the Reactor Containment Building is discussed in the Project's topical report on this subject. This report " Clinch River Breeder Reactor Plant Integrity of Primary and Interined! ate Heat Transport System Piping in Contalrunent", WARD-0-0185, was issued in September 1977 and
!:: ref erenced in Section i.6 of the CRBRP PSAR. Copies of this report have been recently provided to the CRBRP Program Of fice at NRC.
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QCS210.10-1 Amend. 69 May 1982
Foge - 4 (82-0184)[8,223 - #36 -
Ouestion CS770.3 (3.5.4.5)
On page 3.5-13b ductility ratlos for concrete anc steel are listed. S m e of these ratios are dif f erent fra those specified in Appendix A to SRP Section 3.5.3. Conformance to SRP Section 3,.5.3 ductility ratios is requested unless i justification f or deviation is provided. g s Resoonse: . s \ s
't PSAR page 3.5-13b Is revised to incorporate fuctility rdtio's used In it.e , 'L design of concrete and steel structure <..
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knend. 69
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. s rage 1 [8,22]i39 {
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Where A = ductil!ty: ratio defined as the ratio of the deflection of a l structure at f ailure to its deflection at yleid.
T = natural period of vibration (seconds)
The Ductility Ratlos, which will be used in determining the overall structural s response due to missile impact load f or various structural components, are given as follows:
DUCTILITY RATIOS fp )
V Poinforced Concrete Members
?!aximum value of As for reinforced concrete members will be os stated in ACI 349, except stringent requirements speci fied in R.G.1.142 will be used.
Structural Steel Members
- a. For tension due to flexure
/Ad 1 10.0
- b. For columns with slenderness ratio (1/r) equal to or less than 20
/ Ed 1 1.3 Where i = ef fective length of the member r = the least radius of gyration For columns with slenderness ratio greater than 20
)Md 1 1.0
- c. For members subjected to tension N
/d10.5ftp Where eu = Ultimate strai n e = Yield strain 3.5-13b Amend. 69 May 1982 82-0184 _
Nge 2 $8,223939 Case 2. Missile Not Penetrattna the Structural El- nts Williamson and ALvy (Ref. 7) derived a formula for determining a static concentrated load (q equivalent to the force Induced by a missile striking a structurl)i element without penetration. the farmula is: .-
7y*(2[-,)m3 1+ [I+( % Y [L -
When the last term under the radical and its square root are large when compared to unity, the above equation reduces to
_. ?k%Y 8/ ~ T )lspl-/
Where:
m = Mass of missile (Ib-sec.2ff9,)
3.5.5 Misslie Barrier Features t Structures housing saf ety-related systems and components and located above the ground w11I be destgned to restst both the externally and Internally generated missiles. The steel Containment structure will be designed with suf ficient thickness of steel plate to prevent perf ormation and with stif feners proportioned to maintain the overall stability. Preliminary evaluations show i that a containment vessel shell thickness of 1-1/8" wil be capable of accommodating all the potential tornado missiles. Reinf orced concrete, l
Seismic Category I structures, which contain vital equipment will be designed with suf ficient strength and ductility to prevent penetration and to absorb impact f rom the postulated missiles.
3.5-13c Amend. 69 May 1982
Page - 10 (82-0184)[8,22] - #36 I
Ouestion CR770.5 j The major seismic Category I structures of the (RBR plant are supported on a !
mmmon basemat f ounded on competent rock with an embedmont of 100 f t of back fill. Under such a condition, it appears most appropriate to consider the structures as f lxed at the f oundation.
The embedment of f act can be accounted for by considering the soll-structure interaction between the lateral earth pressure and the structure in contact.
The seismic Input motion should be applied at the foundation level. The applicant has considered an analysis in which there is soll (rock) structure interaction at foundation level as well as on the lateral side with the seismic Input motion applied at the finished grade level. In staff's opinion such an analysis does not represent the realistic condition and the complexity of the analysis as used by the applicant precludes a prior assessment of the auequoq ut the method f or staf f review. As a resolution of staf f's concern it is required that seismic Category I structures, systems and components be designed to seismic ef fects obtained by enveloping the results of applicant's and the fixed base approach as stated above or equivalent.
Response
The major Category I structures of CRBRP with the exception of the Diesel Generator Building, are supported on a common basemat founded on rock with an average shear wave velocity of 4000 f t/sec. The material ebove the elevation of the f oundation mat (the embedment material) consists of sound and weathered rock, lean concrete fili and compacted Class A backfIl1.
The input motions were applled at the foundation level and not at grade level (Section 3.7.1.1 of the PSAR).
l The justification for using lumped springs and dashpots in lieu of " fixed" base for rock-structure Interaction is given below.
- 1) In the seismic analysis of the CRBRP Nuclear Island, the actual stif f ness of itie foundation material was evaluated in terms of equivalent springs and dampers and on this basis the seismic analysis was perf ormed.
f 2) In the CRBRP seismic analysis it was considered that using a fixed based l
analysis was unwarranted, since due to the large size and stif f ness of the l structure that is comprised of all the Nuclear Island buildings, some Interaction was expected between the foundation rock and the structure.
This was confInned by the calculated responses at the foundation mat which dif f ered f rcrn the "f ree f iel d" responses (Figure Q220.5-1)
Additional analysis, using a dif f erent analytical approach (the Computer Program FLUSH) showed that the f ree-f iel d and in-structure sei smic l responses at the f oundation level dif fered, confirming that rock (soll) structure Interaction will occur (Figure Q220.5-2).
l l QCS220.5-1 l
Amend. 69 l May 1982
Page - 11 (82-0184)[8,22] - #36
- 3) To calculate the f oundation springs, because of the Irregular layering of l
the site, and variation of foundation properties, a static finite element method was used. ( Section 3.7.1.6 of the PS AR).
- 4) Calculations using the theoretical half-space equations were also perf ormed (Response to NRC Question 130.53). The results of the elastic half-space calculations verified, within reasonable limits, the values obtained f rom the static finite element calculations.
- 5) To account f or uncertainity in foundation material properties three sets of spring stif f nesses were calculated: f or upper bound, average and lower bound of material properties. Analyses showed that responses with average f oundation material properties were enveloped by those f or the upper and lover bound. Two complete analyses were then perf ormed using the upper end lower bound material properiles and the responres were enveloped.
- 6) The radletion danping was calculated based on the hal f-space equations f or equivalent material properties deducted f rom the spring constants obtained by the finite element calculations. This approach was validated by the f act that the results f rom the finite element and hal f-space calculations were in good agreement.
SRP 3.7.2 of NUREG-0B00 def ines varicus acceptable methods for modeling As described above, the and analyzing soll structure Interaction ef fects.
CRBRP seismic design conf orms to acceptable methods ? hat are repre-sentative of the site geologic conditions and the Nuclear Island str uctures. Appropriate conservatism has been included in the model to account f or a range of foundation material properties that were developed f rom a rigorous subsurf ace investigation. The SRP def ines "a fixed base assumption" as an acceptable basis f or modeling if the structures are supported on rock. Such an assumption is not considered reasonable for CRBRP since analysis has confirmed that interaction will occur and a more realistic and conservative representation of the rock and embedment conditions has been accounted f or in the seismic design. The average shear wave velocity of 4000 f ps is not characteristic of a hard rock material that would be consistent with a " fixed" base assumption.
The requirement of a fixed base approach for CRBRP is inappropriate and arbitrary. The analysis of a "f ixed" base model, theref ore, will not be considered f or CRBRP.
o The largest historical earthquake In the tectonic province was assumed to occur in the CRBRP site.
o SSE maximum ground acceleration was increased f rom 0.18g to 0.25g.
l QCS220.5-2 Amend. 69 May 1982
Page - 12 (82-0184)[8,223 - #36 o Design Response Spectra consist of wide band envelope spectra based on statistical studies of many past earthquake records, o Artificial acceleration time-histories used in the seismic analysis envelope and for most f requencies are above the Design Response Spectra. -
o Floor response spectra are envelopes of two Independent analyses using lower and upper bound of soll-rock properties, o Floor response Spectra were widened at peaks 'and smoothed.
An Independent f Inite element analysis using the Fl.USH progran was perf ormed to compare floor response spectra with the CRBRP design spectra. The CRBRP spectra in escance envelopes the calculated spectra and spectra generated f rom the CRBRP and finite element analyses are very similar (Figure Q220.5-3).
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l QCS220.5-3 knend. 69 May 1982
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l Figure o 220.5 - 3 Q 220 5-6
Page - 18 (82-0184)[8,22] - #36
_Ouestion CS770.10 (3.7.1.6) i in last paragraph on page 3.7-4, you stated that a fixed base approach would
- be justifled. However, In order to account for soll-structure Interaction l
ef f ects, you made a number of simpilfying assumptions and also conducted a t
scoping study to take Into account variations in spring constants and damping l values. Indicate if you have taken the fixed base condition into i consideration In your scoping study since this is more represen5tive of the
- actual condition.
Resnonse:
See the response to Question CS220.5.
I i
l QCS220.10-1 Amend. 69 May 1982
P2ge - 13 (82-0184) [8,22] #38 Oues+1on CR220.B On page 3.7-8, It is stated that you perf orm a non-linear history analysis.
Provide more detalls of such an analysis. .
Resnonse The nonlinear seismic analysis of reacter systems was perf ormed on the Primary Control Rod System (PCRS) since this system was determined to contain signifIcant nonlinearltles. Tne description of this analysis is given in Section 3.7.3.15.3 of the PSAR.
l l
QCS220.15 Amend. 69 May 1982
Page 20 [8,22]#39 mathematical representation of the system or components. A sufficient number of masses with their appropriate degrees of f reedom are used in the model to adequately describe the behavior of the structural system, and to insure an accurate determination of the dynamic response. Significant non-linearities, such as gaps or clearances between PCRS components, are included in the [
mathematici model. In this case, a nonlinear time history analysis'is j performed, which considers the Impact forces generated at the gap locations. 1 Non-symetrical features of geometry, mass, and stif f ness, are modeled to I include their torsional offacts in the analysis. A description'of a !
preliminary reactor system linear model and a preliminary PCRS non-linear model is given in Section 3.7.3.15.
The methods of response spectra analysis and time history analysis are described in a number of publications. A description of these analyses techniques is provided in Appendix 3.7-A.
The system cr ccmponent is analyzed with the seismic Input (floor response spectra or time histories) derived at the particular points of support on the structure. AlI signifIcant modes of the mathematical model are incleded in the analysis. The significant, dynamic response modes are those prehninant modes which contribute to the total, combines modal response of the system.
Other modes, whose inclusion in the square root of the sum of the squares modal summation have snegligible ef fect on the total response would not necessarily be used. With this procedure the number of modes included will be such that inclusion of additional modes will not result in more than a 10%
increase in responses. Where the response spectrum method is used, the individual modsl responses are combined by the square root of the sum of the squares, except for closely spaced modes (frequencies less than about 10%
apart) where the modal responses are combined by the absolute sum. The analysis is performed independently in each of the two horizontal directions, and the vertical direction. Similar ef fects obtained f or each of the three directions are combined by the square root of the sum of the squares. This is consistent with Regulatory Guide 1.92.
A simplified analysis based on a single mass model or an equivalent static load method may be used when it can be demonstrated that the simptifled analysis provides adequate conservatism. For the simplified analysis, the equivalent static force, F , is distributed proportional to the mass of the compormni, and is calculat$d by the following equation:
F3 = 1.5 W A s where W is the total weight of the component, and A is the maximum peak acceleration of the response spectra, which apply ai the points of support of the component. Components whose fundamental frequencies are greater than 33 Hz in any direction, are assumed to be rigid in that direction and may be designed for at least the maximum acceleration at their supports.
All systems and components under the jurisdiction of the ASME Section til Nuclear Power Plant Components Code will be designed to accommedste seismic loadings in combination with other loadings without producing total combined stresses in excess of those allowed by the Code. For elevated temperatures, appIIcable ASE-Ill Code Cases and RDT Standards w!!! also apply. Stresses resultleg f rom loar' combinations which include 3.7-8 Amend. 68 May 1982
Pago - 32 (82-0184) [8,223 #38 Duestion CU70.25 l
On page 3.8-1, it is stated that ASE Section Ill Division 1,1974 Edition I with Addenda through winter 1974 and ASE Section ill Division 2,1975 Edition will be used for the design of the steel containment and the steel lined concrete containment foundation mat respectively. Indicate what will'be the ef fect on the design of the latest editions of the ASE Section lli Division 1 and 2 including Code Case N-284 (1980) are used.
Resoonse The PSAR design was perf ormed to the requirements of the 1974 Code edition specified In the design specifications. The specific criteria related to buckling are described in the PSAR Appendix 3.8-A. The Intent of these criteria is similar to the Code Case N-284 criteria, in that these address buckling modes, provide capacity reduction f actors and f actors of saf ety, and s i . . oi inieraction equations for buckling. A significant reanalysis would be required to demonstrate that the Containment Vessel meets the requirements of the new Code and Code Case N-284, however, the applicant has compared the PSAR to the 1980 ASE Code, and has evaluated the significance of the changes.
Several of these changes are considered to be of suf ficient significance to require additional study. This comparison wilI be provided by July 15, 1982, and the additional study of the significant changes will be provided by August 30, 1982.
The applicant believes that the Intent of N-284 and the 1980 ASE Code has been implemented by the PSAR and the PSAR Appendix 3.8-A, the vessel design is adequate and saf e, and that no analysis to the 1980 Code or to Code Case N-284 1s necessary.
I l
(
l QCS220.25-1 Amend. 68 May 1982
Pag 2 - 37 (82-0184) [8,22] #38 Ouestion CS 220.30
%e ultimate capacity of the steel contalment should be addressed.
Response
%e ultimate capacity of the steel contalment is addressed in the h=mt "CRBRP-3, Hypothetical Core Distruptive Accident Considerations in CRBRP, Volume 2, Assessmet of Wermal Margin Beyond the Design Base." A tabulation of the allowable pressures for different temperatures of the material is given in Table 3-10 of the above document. %ese pressures were calculated with primary membrane stress limits for Service Limit Level D given in NURIG-0800 (SRP 3.8.2), with yield and ultimate strength values fra the Nuclear Systems Materials Hancbook (TID-26666) . Se above allowable pressures are concervative since the thickness of the cylindrical portion of the contalment has increased subsequent to the calculations c12e to other design considerations.
OCS 220.30-1 Amend. 68 May 1982
Page - 50 (82-0184) [8,22] #38 -
Ouestion es??0 11 a) Code Case N-284 (1980) should be ref erenced and applied as applicable.
b) The abscissa on Figures 3.8 A-1, 3.8 A-4, and 3.8 A-6 should be labeled R/t and not R/1.
c) Both quadratic and IIncar interaction curves are used. Most authors recommend using linear interaction curves. Are the nonlinear interaction curves conservative?
d) The R/t range f or the containment shell is in a borderline region where either elastic or plastic buckling could occur. For f abricated shells of this type, Imperf ections can greatly influence the elastic buckling loads.
Also, plastic buckling can be !nf!uenced by large residual stresses that can be present. For these reasons the f actors of saf ety given in Table 3.8 A-2 seem to be low. Please justify these f actors.
e) How will buckling be evaluated f or dynamic loads?
Resoonse:
a) See answer to Question 220.25. Code Case N-284 addresses containment shell buckling. This non-mandatory Code Case Is a recent document which was not in existence at the time when the Containment Vessel design was initiated. The Project recognized the need f or specif ic buckling criterla and an appropriate criterion was developed which is described in PSAR Appendix 3.8- A. While the quantitative ef f ect on the design of using Code Case N-284 has not been determined, the intent of the buckling criteria used by the Project is similar to the Code Case, in that the criteria address buckling modes, provide capacity reduction f actors and f actors of saf ety, and similar Interaction equations for buckling.
b) The typographical errors on Figures 3.8 A-1, 3.8A-4 and 3.8 A-6 have been corrected, c) The selection of quadratic or linear Interaction curves is dictated by loading combination. For example, interaction curves f or loading combinations involving torsion are quadratic because modes shapes are dissimilar. Quadratic interaction curves are used in Code Case N-284. In the design buckling criteria, for each loading combination the shape of the nonlinear Interaction curve is identical to that given in a widely i
l QCS220.35-1 Amend. 68
"- *^^'
P2ge - 51 (82-0184) [8,223 f38 i
used shell design document, " Structural Analysis of Shells" by E. J. Baker, L. Kovalesky and F. L. Rish, McGraw-Hill,1972. Therefore, it is concluded that each Interaction curve is conservative, d) The f actors of saf ety given in Table 3.8A-2 are based on the requirements that the ef f ects of Initial Imperf ections and plasticity are adequately considered in the calculation of critical loads. These ef fects are accounted f or by the coef f icients C, K, and H in Figures 3.8 A-1 to 3.8 A-10. The ef f ects of residual stresses on predicted critical loads are negligible f or ring-stif fened cyI!nders subjected to axial compression (the most severe loading condition).
e) Buckling is evaluated f or dynamic loeds by comparing the peak dynamic stresses with the static critical (allowable) stresses.
l l
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QCS220.35-2 Amend. 68 May 1982
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PQ9e 1 WB2-0362 (8,22) 80 Ouestion CE750.1 in Section 3.2.2, "Saf ety Classif ications," of the PSAR, the applicant stated compi lance with the quality requirements of paragraph 50.55a, " Codes and Standarcs," of 10 CFR Part 50, except items (g), " Inservice inspection ,
F5quirements," (1), " Fracture Toughness Requirements," and (j), power reactors I for which a hearing r,otice on the application for a construction permit was l published on or bef ore December 31, 1970. l We concur with the applicant that the CRBRo is not subject to the conditions of paragraph 50.55a (g), (I), and (J) of 20 CFR Part 50. The conditions of pa agraph 50.55a are applicable to boiling or pressurized water-cooled plants.
However, pursuant to paragraph 50.55, " Conditions of construction Permits, "xs recuire conformance to the Ins 3rvice insoection and Fracture Toughness reautrements of the CRBR Principal Design Criteria stated in the PSAR.
Deviation f rcrn the CRBR Design Criteria should be Identiflad and justified.
The Inservice inspection and Fracture Toughness surveillance requirements to be specif led f or the CRBRP must provide an acceptable level of quality and saf ety in design and construction under operating, maintenance, testing and postul ated accident conditions.
Resoonse
- 1. Inservice insoection - The Inservice inspection plan for the Clinch River Breeder Reactor Plant is described in Appendix G of the PSAR. Any exceptions to the applicable Design Criteria are identifled and justifled in Appendix G. Also, see the response to Question CS250.2.
II. Material Surveillance
- 1. Reactor Coolant Boundary: CRBRP Criterion 30 Provisions f or material surveillance for the reactor coolant boundary are incorporated in the design of the reactor vessel Internals and ere documented as folIows:
a) The response to General Design Criterion 30 states that appropriate survel11ance sampies wIII be placed inside the reactor vessel thus providing means for monitoring and evaluating
! potential material degradations.
b) A CRBRP Material Surveillance Program for the Reactor Coolant Boundary has been draf ted, and provisions of this program have been incorporated in appi! cable specifications and other design documents. This program report is being f InalIzed and a summary l
will be included in the FSAR.
! QCS250.1-1
' Amend. 69 May 1982 l
I _ _ _ _ . .
W
- 2. Intermediate Coolant Boundary and Steam Generator System: CRBRP Criterion 33 The Intermediate Coolant Boundary and Steam Generator System surveillance program is being developed. A summary of this -
program will be included in the FSNI. -
QCS250.1-2 Ame'nd. 69 May 1982
Paga - 1 [82-0343] #70 Ouestion CR250 3 Identif y the components and supports in the reactor cociant system.and connecting systems (including the steam generator) which have been constructed, stating the purchase date and the Code, Standseds, and criteria to which they were f abricated. Descr!be the procedures used for their storage. Indicate the dif f erence in the purchase requirements and the Codes, Standards and criteria in ef f ect at the present time. The use of the components should be justified on the basis that they will provide en equivalent degree of system integrity and saf ety as if f abricated to the requirements of the current Codes, Standards, and criterie.
Resoonse The f ollowing components have been procured and f abricated:
Comoonent Contract Date Reactor Vessel 4/18/75 Closure Head 11/14/75 Guard Vessels 2/ 26/76 Core Support Structure 3/08/76 Intermediate Heat Exchangers 4/ 23/ 75 Primary Check Valves 11/75 Primary Control Rod Drive Mechanisms 5/30/75 Listing of the applicable Code editions, Code Cases, and RDT Standards are f ound in the PSAR as f ollows:
Comoonents Location l
Reactor Vessel, Closure Head, Tabl e 5.2-1 and Guard Vessels Core Support Structure Section 4.2.2.3 Primary CRDM Section 4.2.3.1.5 PHTS Components Section 5.3.1.2 j
IHTS Components Section 5.4.1.1 and 5.4.1.2 The procedures used f or storage of the identif ied components and , supports implement the intent and guidance of ANSI N45.2.2 as endorsed by Regulatory Guide 1.38.
Each item important to the saf e and reliable operation of the nuclear power plant is assigned by the designer, a classification level as defined in ANSI N45.2.2. The Project has established and maintains storage f acilities which provide the levels of storage as def ined in ANSI N45.2.2.
QCS250.3-1 Amend. 69 May 1982
P;ge - 2 [82-0343] #70 The Designer establishes packaging, handling, and storage requirements to be sett sf led f or each component which are based on the importance or complexity of the component, the classif Ication of an available storage f acility, and the component characterisites which must be protected or preserved (i.e., exterior surface, Interior surf ace, contamination, temperature variations, moisture, etc.). The packaging, handling, and storage requirements include: -
o Receipt Inspection erIterIa, o Unpacking and re-packing criteria, o Storage maintenance (purge, dessicant, rotation, lubrication, etc.)
criteria, and o Storage maintenance verification criteria.
The procedures control both housekeeping of storage f acilities per ANSI N45.2.2 and component maintenance per the packaging, handling, and storage requirements f rom the designer.
The current applicability of Codes and Standards used in early procurements of CRBRP components was addressed orally as part of the May 6-7, 1982, presenta-tion to the NRC on High Temperature Design. The documentation requested will be provided in Ref erence QCS210.1-1.
l l
l I
QCS250.3-2 Amend. 69 May 1982
PJim - 2 82-0287 [8,22] #37 l
1 l
Ouestion CS410.2 (9.1.d1 n
It is our position that overhead cranes whose f ailure could damage spent f uel or essential equipment be designed such that In the event of the SSE they can '
retain control of and hold their load. The bridge and trolley should be designed to remain in place on their respective runways with their The bridge wheels should remain on prevented f rom leaving the tracks during the SSE.
the runway with brakes applied, and the trolley should remain on the crane g girders with brakes appa led. The crane should be designed and constructed in accordance with Regulatory Position 2 of Regulatory Guide 1.29, " Seismic Design Classification." The maximum critical load plus operational and seismically induced pendulum and swinging load ef f ects on the crane should be cor:sidered In the design of the trolley, and they should be added to the trolley weight f or the design of the bridge.
For the polar crane, cask handling crane, and other cranes whose f ailure could damage spent f uel or essential equipment, demonstrate that you meet this position.
Resoonse The f ollowing responses apply to the RCB Polar Crane, and the RSB Bridge Crane (cask handling crane) and the SGB Gantry Crane:
(1) For critical loads, overhead cranes are designed such that the SSE will not result in dropping or losing control of the heaviest critical load.
The cranes are designed as Selsmic Category 1, single f ailure proof, redundant cranes to meet NUREG-0554.
(2) All cranes are provided with selsnic restraints such that the bridge and trolley will remain In place on their respective runways during the SSE.
The bridge will renaln on the runway with brakes applied and the trolley will renaln on the crane girders with brakes applied.
(3) The subject cranes are des!gned in accordance with Regulatory Position 2 of Regulatory Guide 1.29, " Seismic Design Classification."
load (4) Operationally and seismically induced pendulum (i.e. swinging) ef f ects have been incorporated in the crane designs by adding these loads
' to the maximum critical load and dead load.
QCS410.2-1 Amend. 69 l
May 1982
pag 3 1 W82-0357 (8,22) 29 Ouestion CS410.3 Provide the results of an analysis which demonstrates that spent fuel and essential equipnent will not be damaged by a heavy load drop &e to a handling system malfuncticn. Include consideration of cask handling crane failure resulting in a cask drop, polar crane failure resulting in drcpping the heaviest load handled by this crane over essential equipnent including the reactor vessel, cnd other pertinent potential handling syste malfunctions.
% e analysis should satisfy the guidelines of NUR E-0612, " Control of Heavy Ioads at Nuclear Poder Plants," Section 5 and Appendix A with due consideration for the differences between CRBR and IRR design.
Response
%e CRBRP approach to control of handling heavy loads is consistent with the guidelines of NURID-0612. %e approach takes advantage of a fundamental difference betwem CRBRP and IER designs, i.e., reactor refueling and spent fuel storage fuel handling are done "through-the-head" rather than "open-head" or "open-pool."
Selected heavy loads use the single-failure-proof cranes in the RSB and RCB.
In addition to the use of single-failure-proof cranes and associated lifting devices with the following features or proce& res are used, o Identification of safe load paths, o Administrative controls over handling of heavy loads over or near the reactor, Ex-Vessel Storage tank of Fuel Handling Cell, o Polar crane interlocks to prevent handling of heavy loads and polar crane operations over or near the reactor & ring reactor cperation, and o Design and analysis of the reactor and fuel storage facilities for the impact of the heaviest anticipated load carried by the single-failure-proof main hooks of the RSB and RG cranes (at the maximum lowering speed). %is is over and above the requirenents imposed by NURED-0612.
Other heavy loads use the auxiliary non-single-failure proof cranes and cssociated lifting devices. For the EVST, FHC and Reactor, the effects of a load drop of the mwh= weight fra the highest administratively allowed working height are analyzed. Due to a polar crane interlock, and identification of safe load paths, the drop fra the auxiliary hook is I analyzed only for reactor shutdown conditions. In these three locations, the postulated load drops do not affect criticality, damage the integrity of the liquid coolant boundary, or affect shutdown equipnent.
We results fra impact 1 cad analyses for the reactor head and fuel handling facilities are sumarized below.
QCS410.3-1 Amend. 69 May 1982
LPP4M7/
page 2 W82-0357 (8,22) 29 IMPACT IIRD ANALYSIS SLD91ARY Pertinent Location PSAR Sections Tamet rmdino Conseguences Ex-Vessel Storage 9.1.2.1.1 Load lowered by a single Nodamahe Tank failure proof crane Load drop from non-single Stresses less failure proof crane than ASME Code allowables for no gas leakage Fuel Handling Cell 9.1.4.10.1 Load lowered by a single No damage failure proof crane Load drop fr a non-single Damage factor failure proof crane of 4 for rein-forced concrete.
No structural failure. Radio-logical conse-quences of pos-sible leakage of FHC atmosphere enveloped by reactor cover gas leak (see PSAR 15.5.2.4).
1 Reactor 15.5.2.5.1 Load lowered by a single Possible damage failure proof crane to head mounted cmponents.
Radiological consequences of I cover gas re-l lease are enve-l loped by a l discussion in l
PSAR Section 15.5.2.4 due to nature of crushed CRms.
'Ihis event is not as severe as the reactor cover gas release event because the seal failure would result in a more gradual leakage of cover gas.
OCS410.3-2 Amend. 69 May 1982
LPRM67/
oogs 3 W82-0357 (8,22) 29
%e proce&res and identification of non-critical loads to be lifted over the reactor closure head & ring reactor shutdown using the non-single failure-prcof-crane are currer.tly under developnent.
h e Spent Fuel Shipping Cask (SFSC) is required to have single-failure-proof lifting devices and can only be handled by the single failure proof Reactor Service Building (RSB) overhead crane main book. In addition, a safe load pcth has been identified for the SFSC that does not pass over or near the EVSr, nic, or safety-related equignent.
% c radiological impact of the drop of the SFSC fr a the single-failure-proof crane onto the SFSC handling shaf t is provided in Section 15.7.3.2.
%e effect of heavy loads dropped on safe shutdown equignent in the RCB has baen analyzed. Any heavy load dropped fr a the auxiliary book can only affect one enannel of the shutdown systen. Since the Reactor Shutdown Systen is designed to IEEE 379, (Application of Single Failure Criterion to Nuclear Protection Systens), as described in PSAR Section 7.2.2, failure of that one channel, by damage to sensors, cabling, or transmitters, will not preclude operation of sufficient equipnent to achieve a safe shutdown.
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QCS410.3-3 Amend. 69 May 1982
p gt- 1 W82-0298 (8,22) 40 Ouestion CS410.19 (9.7.3)
The normal and emergency chilled water systems provide cooling for plant HVAC systems. HVAC units serving areas containing sodium or NaK are provided with drains to carry away chilled water leakage to prevent moisture carry-over in the HVAC ducting. Leak detectors are provided in the drains to defect chilled water system coil f ailure. Activation of the detectorJustify resultsthe In au'tomatic use of closure of the chilled water coil isolation valves.
non-saf ety related normal chilled water system piping and valves in HVAC units serving areas contelning sodium and NaK.
Response
With the exception of the SGB loop cells, the HVAC units provided wiih normal chilled water and serving areas containing sodium and NaK are located outside
" u Jium and NaK cells. These cells do not require safety-related cooling.
Accordingly, their associated HVAC units are classified as non-saf ety related.
For the SGB loop cells. Three barriers between the sodlum and water tre provided as follows:
a) Chliled water piping walls b) HVAC equipment walls which serve as spray shields c) Sodium piping walls The safety classification of these barriers is currently being evaluated to determine if they provide adequate protection against a sodium / water reaction.
The results of this evaluation will be provided in a future ammendment.
l QCS410.19-1 Amend. 69 May 1992
~~~~~~___ ~' ' ' - - .-
Page 1 [8,22]025
~
Question M1.01 , g ;, ,
t . . , ';
During meetings with the applicant and Westinc. house, several discussions have J t
been held concerning the fact that the primary and secondary shutooun systens '
l do not each, indivi& ally, ceply with Sectiori1.7.3 of IEEE-279 m Control !
and Protection System Interaction. he applicant should document forf s ss inclusion in the PSAR the justification of the adequacy of the proposed design for cm plying with Section 4.7.3 of IEES-279. %is justification should ,
l include a discussion of the syst s adequa q with respect to control and \
protection system interaction & ring periodic testing of a protection syste channel or when a protection system channel is cut of service for maintenance.
If the justification includes the use of a median selector for control ,
signals, plans to periodically test the median selector & ring plant operatPn should also be discussed. ,
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%e two PPS reactor shutdown systems, conside' red together, always meet Section 4.7.3 of IEED-279.
We provision of two diverse and independent shutdowa systems provides the plant with an unusually high degree of pr:Aection against cormnon mode failure ;
incidents.
Under normal operating conditions, and in the great majority of abnormal' conditions, each syst s separately tully meets the requirements of IEEE-279. ,
However, in a limited number of situaticas, it is possible that & ring testing .
or maintenance of a channel which supplies signals to both protection and i control, a single sensor failure could initiate plant control actions requiring protective action and simultareously prevent proper protective channel response. Applying the criteria of Section 4.7.3 of IEEE 279, which requires the assumption of a second rande failure for these channels, results in the assumption that the protective function in one of the systes would be disabled. In these situations, the re&ndant functicns in the alternate systs always provide protection as required by Section 4.7.3 of IEES-279. _
We likelihood of such potentially disabling situations is limited by the use in the control system of a median select arrangement which precludes rerponse '
to abnormally high or low signals. However, when one protection channel is placed in the trip condition, a simultaneous sensor failurt ,in & low direction could potentially result in disabling the protection systs at the same time that a faulted control system calls for a power increase.
l With regard to validation of the median selector, most failure 1 of the median selector are self-annunciating (i.e., abnormal plant response is detected by the operator or subsystems within both Shutdown Systens leading to a plant trip). Othcr median selector failures (eg., output not following the true t
median input) occurring together with a sensor failure do not cause a plant i
response. In any case, median selector failures do not invalidate the I performance of the related protection system. Nevertheless, functional testing of the median selector circuits is performed annually as part of the scheduled maintenance.
OCs421.01-1 Amend. 68 May 1982
j, Page 2 [8,22]G25 1
In the majority of situations, further built-in control features (such as built-in overall tanperature loops or rod movanent blocks related to alternate flux measurenetts) will prevent any such overpower conditions fra develqing.
Where this is not the case, as previously noted, protection is provided by the second independent protection systen.
Section 4.7 of IEEE 279 deals with contrc1/ protection systen interaction. .
. There is no control / protection systen interaction for the SCRS.
s.
l QCS421.01-2 l
l Amend. 68 l May 1982 ,
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Ouestion CM71 1 Section 7.6 of the PSAR states that the Radiation Monitoring System contains However, Chapter saf ety related components which are discussed in Chapter 11.
11 does not discuss these saf ety related components. 0; rect the PSAR to identify the saf ety related compoents of the Radiation Monitoring System.
Responsa A PSAR mondment w!!I be submitted by July 1982 which wll1 Identify safety related components of the Radiation Monitoring System. Safety-related process and of fIvent monitors ' w11I be deseribed in SectIon 11.4.2.2.8 and IdontIiled High range containment area monitors which are saf ety in Table 11.4-1.
related will be Identified in Section 12.1.4.1.
QCS421.21-5-1 Amend. 69
P:ge 10 [8,223425 Ouestion 421 03 Document the design provisions for conducting response time tests of BOP and NSSS protection systems in accordance with R.G.1.118. Identif y saf ety-related systems that do not have provlsions for response ting testing.
Discuss the techniques to be used to periodically measure saf ety-related sensor time responses.
IMsponse:
The CRBRP Protection systems are being designed to comply with R.G.1.118.
The NSSS protective systems (RSS, CIS and SGMRS Initiation) use an overlap testing technique to verif y that system relponse times are within acceptable
- limits. The f ollowing features are provided for response time testing:
A. For the neutron detectors, detector-cable capacitance checks are made with the detectors and cables in the as-Installed configuration to identify increases In capacitance which could af f ect time response of the channel.
B. For the reactor vessel level Instruments, the sensors are checked in the as-Installed conf iguration by Inserting a step increase In the primary coil excitation current at a steady state sodlum temperature and level and monitoring the increase In secondary coil output voltage due to the Increase in magnetic f leid strength.
C. For channel response time testing, not including sensors, a test signal source is connected to the channel which can simulate the sensor input over the entire range. Measurements of channel Input changes and channel output changes are recorded. These tests include Istrument channel Where equipment and the comparator trip outputs to the logic trains.
protective f unctions use two or more variables, channel time response is determined f or each input to the f unction one variable at a time. The remaining variables are adjusted to a conservative value within the normal operati ng range.
D. The Primary RSS and Steam Generator Auxillary Heat Removal System (SGMRS)
Initiating logic time response is checked during f unctional testing of the logic. Pulses are Inserted into the redundant instrument channel comparators to simulate the eight possible combinations of trip and reset.
For test Inputs with two or more redundant channels tripped, the log!c output is checked f or a trip condition. This trip output must be detected within the propagation delay limits of the logic train or it is flagged by the tester as a f ailure.
E. The Primary RSS scram breaker time response is checked by inserting trip signals to the logic train and monitoring the time to current interruption on the output side of the breakers. Each breaker is tested separately to f
l assure compilance with response time requirements.
( The Secondary RSS logic is tested by Inserting a trip input to the F.
comparator and monitoring the time required until the current to the solenold screm valve In Interrupted.
QCS421.08-1 Amend. 68 May 1982
Utge W[GT,WIEJW
- G. Time response testing of the Secondary RSS scram solenoid valves is i included with testing of the Secondary rods.
H. The CIS logic and breakers are tested in a similar f ashion to D and E.
- l. Response time measurement of pressure or dif forential pressure I,s performed by applying a pulso wIth a spectfIc ramp rate and comparing the output of a tested transducer with the output of a f ast acting ref erence transducer. The output of both transducers is recorded on a strip chart for subsequerd computation of the response time.
J. Response time measurement of thermocouples cr RTDs is f acilitated with the use of a loop current step response analyzer. tie loop current step response analyzer sends a pulse of current to the thermocouple or RTD.
The analyzer receives time varying voltage data f rom the temperature caneing device, digitizes and stores test data. The analyzer then computes the sensor time constar.t cr.d d! splays this on a panel meter.
K. A sweep generator shall be provided for measuring pump speed signal conditioning time response. The sweep generator shall have an auxiliary voltage output that is proportional to f requency output. The sweep generator shall supply an input to the pump speed signal conditioning modules and the proportional voltage output will be supplied to an oscilloscope. The output of the pump speed signal conditioning modules will also be supplied to the oscillograph.
L. Response time measurement of the sodim flow signal conditioning modules shall be accomplished utilizing a millivolt generator that provides a ramp output. The methodology used is analagous to the methodology described in j the previous paragraph.
l M. A sensor response time monitor will be used to determine if time response degradation has occurred in PPS channels. The device samples the normal fluctuations around the average valve of the sensor output and determines the time response characteristics of the sensor based on the number of times the output signal crosses its average valve in a fixed time interval. This device may be used on RTDs, thermocouples, pressure sensors, level sensors, and flow sensors as a method of determining time l response degradation of PPS channe,I s. Use of this methodology is f ast and of fIclent but cannot be used for initial time response measurements. Only l
time response degradation can be measured.
N. A test and calibration signal source is permanently installed on eacn PPS Shutdown Panel. The signal source Is used to inject a step test signal for mmparater response time measurement. The step test signal is Inserted into the channel to cause a PPS comparator to trip The output of the comparator is compared to the output of the test signal and time response characteristics are calculated.
i i Q421.08-2 Amend. 68 May 1982 l
nhnxn _ _
Page 12 CB,22]I25 O. SGB flooding is detected with temperatures and humidity sensors. When flooding is detected the f eedwater Isolation valves shut to prevent damage to SGNiRS components. A method of verifying humidity sensor time response characteristics is to be determined.
P. Radiation monitors are provided with a built-In check source. 'The check source is normally shleided f rom the detector, and the shlald is solenoid operated. Response time can be measured by electrically actuating the solenoid operated shleid, and obsreving the monitor output, taking due account f or solenold response time.
Q.
Gas detectors can be tested similarly to pressure Instruments by injecting known compositirn of gas through the test valves and observing the instrument outrut.
- n. r i v. >ensors w,1ose operation is based on the hot wire technique are not tested in situ. They would be removed f rom the pipe or duct and placed in a test fixture In which flow can be suddenly terminated, and the flow sensor output observed.
S. Sound powered phone Jacks, test cabling, and test points are incorporated into the design to f acilitate the testing methods descr! bed above. Most of the testing hardware is portable but some permanently mounted equipment is provided to minimize the ef f ort requ! red to measure the time response of the PPS comparaters, i
1 QCS421.08-3 Amend. 68 Mav 1QA7 l
LtR@E9 Page 13 [8,22]#25 005TICri CS 421.09 Identify where instrment sensors or transmitters supplying information to more than one protection channel, to both a protection channel and control channel, or to more than one control channel, are located in a coman -
instrument line or connected to a cannon instrumet tap. Se intent of this its is to v( ify that a single failure in a omnon instrment line or tap (such as break or blockage) cannot defeat required protection system re&ndancy.
RESPONSE
Instr mentation sensors or transmitters located in instr ment lines or connected to instr ment taps do not supply more than ont protection channel or control channel. %erefore, the required protection action will not be JALJ by ; blockage or breakage of an instrment line or instrument tap.
However, there are instrumet sensors and transmitters located in instrument lines or connected to instrummt taps which provide signals to both a protection cnannel and a control channel. In all cases, both the protection and control function has a three channel input re&ndancy. % ese re&ndant channels use separate instr ment lines and taps. For example, 0
the superheater steam flow Venturi provides three separate taps located 120 apart for the redundant sensors. Sus, a single failure resulting fra a blockage or a breakage in an instrment line or tap will not defeat the required protective action.
l l QCS421.09-1 l Amend. 69 i May 1982
82-0260 Pag 314 [8,22]025 Resoonse to NRC Ouestion 171.09 In general the instrumentation shall meet the criteria of IEEFe279-1971 and RDr Standard C16-1T-1%9, also the single fallure criteria of IEEE 379-1972 and Regulatory Guide 1.53. .-
A. Pertaining to the Plant Protection Systen (PPS) channels, each reactor shutdown systen (RSS) shall use three redandant channels of instrumentation to measure plant permeters.
B. Accident monitoring equipnet redandance is per WARD-D-0307.
C. For Class 1E equipnet other than A&B, redindancy will be accmplished by one of the following criteria.
- 1. Redindant process loops.
- 2. Redandant instrtunent channels.
Sc- redundancy requirenents of the instrumentation applicable to this question are satisfied as follows:
o Wree redindant channels for each loop are provided for the following PPS channels:
- 1. Feedwater flow
- 2. Feedwater tenperature
- 3. Superheater steam flow
- 4. Superheater steam tenperature,
- 5. Superheater stem pressure
- 6. Reactor prod 2 cts vent flow (east evaporator, west evaporator, superheater, IHTS Na expansion tank.
'No auxiliary feedwater flowmeter transducers are provided for each AEW loop to insure isolation capability with loss of one 1-E power division (each transducer pair is powered frm different 1-E power sources).
OCS421.09-2 Amend. 68 Mav 1QR?
Page - 1 [82-0362] 484 Question CSd21.14 Discuss the CRBR design pertaining to bypassed and inoperable status indication. As a minian, provide information to describe:
- 1. Means to be used for compliance with the reca mendations of R.G. 1 47.
- 2. %e design philosophy to be used in the selection of equipnent/systens to be monitored.
- 3. How the design of the bypass and inoperable status indication systens will emply with positions B1 through B6 of ICSB Branch Technical Position No.
I 21.
%e design philosophy should describe as a minimm the criteria to be unployed in the display of inter-relationships and dependencies on equipnent/systens
! and should insure that bypassing or deliberately induced inoperability of any auxiliary or support system will autmatically indicate all safety systens l affected.
l Response ne following responds to Itens 1, 2 and 3:
t i 1. A discussion regarding c m pliance with R.G. 1.47 is provided in new PSAR j section 7.5.12.
- 2. We safety functons and systems to be monitored by Inoperable Status Monitoring Systen (ISMS) are based upon Engineered Safety Features of Chapter 6 of the PSAR and are shown in the Table 7.5-4 of PSAR Section 7.5. %ere are 2 active safety systems which are not included in ISMS:
the Reactor Shutdown System, which provides separate indication, and the contaimmt Isolation Systen, for which no bypasses or deliberately induced inoperable states have been identified. (Reactor Shutdown Syst s bypasses are discussed in Section 7.2.1.1 of the PSAR.) %e design of ISMS will enploy the following steps to ensure that the system l inter-relationships and dependencies on auxiliary systens are properly identified.
1
) o For each systen and subsystem in the attached table, the states of the l caponents which lead to systs inoperability will be identified.
o %e Maintenance Outline Proce&res and Operating Outline Proce&res will be reviewed to identify any maintenance testing, or surveillance activities which would cause any active caponent to be bypassed or
' inoperable, o %e results of the first two itens will be cmbined to define the caponents to be monitored and to develop the logic for identifying inoperable systems per Regulatory Guide 1.47.
QCS421.14-1 Amend. 69 May 1982
^
82-0362 Page - 2 [82-0362] 084 o In addition, the auxiliary and support systems required for operation of all active cmponents in the safety systems will be identified and the above 3 steps repeated for these identified auxiliary and support systems. Inoperability of these auxiliary and support systems would be indicated by auxiliary system inoperability indications and an indication of the incperability of the affected emponents.
- 3. We design of the GBRP bypass and inoperable status indication systes is intended to emply with positions Bl through B6 of ICSB Branch Technical Position No. 21 as follows:
BTPl. ["he bypass indicators should be arranged to enable the operator to determine the status of each safety syst e and determine whether continued reactor operation is permissable."]
Bypass indication for safety systems is to be ccabined on a single ISMS indicator panel with separate indications for each of the following subsystems: (Ref. PSAR Sec. 7.5.12) o Decay Heat Removal Syst m -
o Fuel Storage Heat Renoval Systen o Control Rom Habitability o Annulus Filtration o Reactor Service Building (RSB) Filtration nese dedicated indicators are activated whenever a system is determined bypassed or inoperative.
In addition, the ISMS is supported by Plant Annunciator System (PAS) and the Plant Ihta Handling and Display Systs (PDH&DS), and changes in the safety Eysten status are transmitted to the PAS for audible and visual annunciation to the operator. PDH&DS cathode ray tube (CRI) displays may be used to provide the operator information about safety systens.
1 BTP2. ["When a protective function of a shared systen can be bypassed, indication of that bypass condition should be provided in the control rom of each affected unit."]
CRBRP shares no safety systen with other units.
l BTP3. ["Means by which the operator can cancel erroneous bypass indications, if provided, should be justified by denonstrating that the postulated cases of erroneous indications cannot be eliminated by another practical design."]
Activation of bypass indication is provided by a caputer l program which is not accessible to the plant operator.
l Cancellation of bypass indication is normally only possible l through removal of the condition which caused the bypass indication (e.g., reclosing of a critical valve or breaker).
If the condition is erroneous, the cause of the error (e.g., a short-circuited wire) nust be determined and corrected in order to cancel the bypass indication.
QCS421.14-2 Amend. 69 Mm, 1000
[ LTFdLtf
- Nge - 3 [82-0362] #84 l
l BTP4. ["Unless the indication systen is designed in conformance with criteria established for safety systens, it should not be used to perform functions that are essential to safety.
l Adninistrative proce&res should not require innodiate operator l
action based solely on the bypass indications."] ,.
I
( %e CRBRP bypass and inoperable status indication systen is not I used to perform functions essential to safety. G BRP operating proce&res will not require insnediate action in response to l
bypass indications.
BTPS. ["The indication systen should be designed and installed in a manner which precludes the possibility of adverse effects on plant safety systens. Failure or bypass of a protective function should not be a credible consequence of failures occurring in the indicatica e:;uipt.ent, and the bypass indication should not re&ce the required independence between re&ndant safety systens."]
he ISMS equipnent shall be isolated from the associated safety related equipnet so as to preclude any abnormal or normal action of the ISMS from preventing the performance of a safety function. It is intended that all electrical input connections to ISMS frcun safety related equipnent are electrically isolated at the safety related equipnent.
BTP6. ["he indication system should include a capability of assuring its operable status & ring normal plant operation to the extent that the indicating and annunciating function can be verified."]
he ISMS will provide a means of verifying that the indicator lights are functional.
QCS421.14-3 Amend. 69 May 1982
Page 1 (82-0362) [8,22] #79 7.5.12 Inocerable StaH= Monitorina svaten 7.5.12.1 Design Description
%e Inoperable Status Monitoring System (ISMS) provides for an autmatic indication at the system level of the bypassed or deliberately induced inoperability of selected safety systems. Se indication of inoperative and bypassed status will be provided in conformance with Regulatory Guide 1.47,
" Bypassed and Inoperative Status Indication for Nuclear Power Plant Safety Systens". %e Reactor Shutdown System (RSS) includes indication of bypassed status and is not part of ISMS. See Section 7.2 for the RSS details.
Specific capabilities of the ISMS include:
a *~itering of safety systen variables and alerting the unit operator ualy and audibly of a bypassed or deliberately induced inoperable i safety system.
o A manual capability to activate the safety systen indicators.
No direct safety or protection function is performed by the ISMS. Se list of status information and monitoring provided by ISMS is provided in Table 7.5-4.
W e data acquisition and processing c a ponents of the ISMS are located in the control rom area behind the Main Control Panel (MCP). S e unit operator interface centrols and indicators are located so they can be seen fra the MCP. W e ISMS data acquisition system obtains the sensor status information processes the sensor data, and provides the status indication to the unit operator. n e processed data is transmitted to the Plant Data Handling &
Display Systen (PDH&DS) for presentation via cathode ray tube (CRT) displays in support of ISMS.
7.5.12.2 Desian Analysis l he ISMS is empletely designed to conform to the requirements of Regalatory Guide 1.47, " Bypassed and Indication for Nuclear Power Plant Safety Systems".
% e ISMS in c m bination with the bypass status indication portions of the Reactor Shutdown Systen provides the complete systen level coverage of safety system indication and manual activation of indicators required by Regulatory Guide 1.47. g he safety systens provide the isolation devices for associated safety-related equipnent so as to preclude any action of the ISMS frm preventing the performance of a safety function.
7.5-33d Amend. 69 May 1982
82-0362 Page 2 (82-0362) [8,22] 079 Table 7.5-4 Rafety Functions and Primary Svstama Monitored by 19E SAFETY SYSTDIS/ SUBSYSTEMS FUNCTI W Mm rIORED MONTIORED Decay Heat PACC newva.L Aux FW/ Vent SGAHRS Steam Generator Heat Transport PH'IS Heat Transport DHRS Direct Heat Removal System Fuel Storage Forced Circulation EVST Cooling Heat Ranoval Natural Circulation Control Room Control Rocra Filtration H&V Habitability l
l Annulus RCB/ Annulus H&V Filtration l
l RSB Filtration RSB H&V l
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7.5-42 Amend. 69 May 1982
page 1 W82-0263 (8,22) 33 ouestion m.17 2e information supplied for renote shutdown (PSAR Section 7.4.3) frun outside the control rom is insufficient. %erefore, provide further discussion to describe the capability of achieving hot or cold shutdown from outside.the control rom. As a minimum, provide the following information: -
a) A table listing the controls and display instrumentation required for hot and cold shutdown fra outside the control rom. Identify the train assignments for the safety related equipnet.
b) Design basis for selection of instrunentation and control equipnent on the hot shutdown panel.
c) Location of transfer switches and the renote control station.
d) Description of transfer switches and the remote control station.
e) Description of isolation, separation and transfer / override provisions.
his should include the design basis for preveting electrical interaction between the control rom and remote shutdown equipnent.
f) Description of control rom annunciation of remote control or overridden status of devices under local control.
g) Description of cmpliance with the staff's Remote Shutdown Panel position.
Response
W e response to this question is provided in the amended text for Section 7.4.3.
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l QCS421.17-1 l
Amend. 69 May 1982
Pag 2 - 1 [8,22] 87 7.4.3 pnnte Shutdown Systm 7.4.3.1 Design Description 7.4.3.1.1 Functim f he Rmote Shutdown Syste provides the means by which (1) aafe shutdown conditions of the reactor plant can be established and maintained fr a locations outside of the Control Room in the event that the Cmtrol Rom must be vacated; (2) hot shutdown conditions can be achieved and maintained; and, (3) if desired, the plant can be cooled to and maintained at the refueling temperature.
7.4.3.1.2 Design Basis h e R mote Shutdown system is designed to use equipment located outside of the Control Rom to place the reactor and plant into a safe shutdown condition under the following conditions:
(a) W e evacuation of the Control Room is not coincident with any other abnormal plant condition with the cme exception that loss of offsite power may occur.
(b) No severe natural phenmena such as earthquake, tornadoes, hurricanes, floods, tsunami and seiches (fra 10CFR50, Appendix A, Criterion 2) occur coincidently with the excavation of the Control Rom.
l (c) % e plant remains in an orderly shutdown status from the initiation of I the evacuation of the Control Room to the time that comand of the shutdown is re-established outside of the Control Rom.
(d) We rmote shutdown operations will be ccananded from one location and will use plant systes operated in their local mode to effect the shutdown and decay heat removal.
(e) Plant instrumentation and control syst es required for remote shutdown operations will have transfer switches located at the local panels to permit the plant operating personnel to select to operate frm the local panels while isolating the remote controls or, conversely, to operate fr m the control room while isolating the local controls. % e transfer of control of a plant system fr m the remote to the local mode is annunciated in the control rom.
(f) Cmmunicatims between the Rmote Shutdown Monitoring Panel (RSMP), the cm mand location for remote shutdown operations, and the SGAHRS panels and other local panels during rmote shutdown operations will be by the Maintenance Canunication Jacking (MCJ) system utilizing a sound-powered telephone.
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7.4-8a Amend. 69 l May 1982
m . _ _ - -.
Page - 2 [8,22] 87 7.4.3.1.3 Ranote Shutdown Operations Se RSMP will be located in Cell 271 of the 836'-0" level of the SGB. ' 2e RSMP will have indications (see 7.4.3.1.4) fra which an operator can assess the progress of the shutdown, and it will be the location from which that operator will connand the operation of the plant systems being operated in their local mode to effect shutdown.
Se Division 1,11 and 111 SGAHRS (Section 7.4.1) local panels will be located in Cells 272A, B and C respectively, in close proximity to the RSMP, on the 836'-0" level of the SGB-1B. Se SGAHRS, cperated in its local mode, will be used to control the renoval of heat fra the reactor plant to achieve and stabilize the plant at the desired plant tenperature (hot shutdown or refueling tarperature). Se local SGAHRS panels will have all of the controls ;
and indications necessary to cepletely enntrol the systen. All signals fra the Control Rom to the SGAHRS panels are buffered to prevent faults occuring in the Control Room fr a propogating back to the SGAHRS panels. All SGAHRS caponent controls can be transferred to local at the local SGAHRS panels.
Placing the transfer switches in " local" overrides all control functions in the control Rom.
We Division 1,11 and 111 OSIS local panels are located in SGB Cells 272A, B and C with the SGAHRS panels, and will be operated in the local mode when required to control heat removal fr a the plant in conjunction with the operation of SGAHRS. Isolation of OSIS panel controls fr a the Control Room is incorporated in the design. Steam drum drain and superheater outlet isolation valve controls can be transferred to local at the local OSIS panels.
Whenever any SGAHRS cmponent control transfer switch is placed in the " local" position an alarm is initiated in the Control Room to alert the Control Room operator. W e same statement is true for the steam drum drain controls and superheat outlet isolation valve controls on the OSIS panels.
If offsite power is lost coincident with having to achieve a safe shutdown condition in the reactor plant frm outside of the Control Rom, the diesel generators will start and function in accordance with the design provided by the Building Electrical Power Systen. Any operator actions required in conjunction with operating and loading the diesel generators will be done in the local operating mode at the DG local panels.
In the event that the Control Rom must be vacated, reactor scram and SGABRS operation will be initiated manually. Se operating personnel will move to the 836'-0" level of the SGB where the SGAHRS in the local mode will effect heat removal and stabilization of the plant tenperatures. Operation of the SGAHRS in the local mode will effect heat renoval and stabilization of the plant temperatures. Se plant shutdown will be directed by the operator at the RSMP who will also assign operating personnel not continuously occupied in l operating SGAHRS to oversee or operate other systens as required.
l Moyenent of personnel within the plant and access to building cells and local panels will be controlled by the facilities and procecbres of the Industrial Security Systen.
7.4-8b Amend. 69 May 1982
g -
Itge - 3 [8,22] 87 7.4.3.1.4 nmir=mit Desian
%e RSMP is the mly piece of equipnent provided by the Renote Shutdown Systen. It will be a vertical sided cabinet amaa=hly containing meters and a phone jack panel. Se meters will receive buffered signals fra the .-
initiating systens and, thus, do not require transfer switches to isolate then fran the Control Rom. Se phone jack panel will permit the operator at the ;
R91P to conmunicate with the five NSSS or Nuclear Island buildings by means of any of the three MCJ circuits provided in each of the buildings. In addition, connunications among the buildings can be established through the phone jack panel on the RSMP.
%e indications provided on the RSMP are as follows:
"-- ~h primary heat transport systen loope, 1 - Pump outlet sodium tenperature indication (3 total) 1 - Reactor inlet soditun tenperature irdication (3 total) 1 - Sodium pm p shaft speed indication (3 total) o For each intermediate heat transport systen loop, 1 - IHX outlet sodium tenperature indication (3 total) 1 - IHX inlet soditan tenperature indication (3 total) 1 - Sodium pump shaft speed indication (3 total) o For each superheated stecm loop, 1 - Tenperature indication (3 total) 1 - Steam flow indication (3 total) o One reactor vessel sodium level meter (long probe) o For each Diesel Generator (3 total) 1 - Wattmeter 1 - Frequency meter 1 - Varmeter 1 - Voltmeter with phase selector switch 1 - Ammeter with phase selector switch In addition to the foregoing indications, other indications used during remote shutdown operations that are not on the RSMP will be available as follows:
o SGAIES Controls and indicators used for the operation of each SGAHRS division are located on the three separate SGAHRS panels in cells 272A, B, and C. Each SGAHRS division is separate and redindant fran the other divisions. See the response to question 421.04 for additional information about SGAHRS division assignments.
7.4.8c Amend. 69 May 1982
U'RAUJ Page - 4 [8,22] 87
'Ibe following controls, indicators and alarms are on each SGAHRS panel.*
Controllers Auxiliary Feedwater Flow -
AEW Steam Wrbine Stean Inlet Pressure PAOC Inlet Louver Position PAOC Fan Blade Position Steam Drtan Level Steam Drtzn Vent Superheater vent Analog Indicators Protected Water Storage Tank Level
. . . . .:'.d t'ater Storage Tank Tenperature Auxiliary Feedwater Flow Auxiliary Feedwater Planp Discharge Pressure steam Driven Wrbine Steam Inlet Pressure Steam Driven Wrbine Speed FAOC Outlet Air 'lirnperature TACC Outlet Water Flow and Temperature PACC Inlet Louver Position PAOC Fan Blade Pitch Position Steam Drum Pressure and Water Level Annunciators Protected Water Storage Tank Level PWSr Tenperature AEW Supply Tenperature Steam Driven Wrbine Speed Driven Wrbine Steam Inlet Pressure Steam Driven Wrbine Bearing and Lube Oil Tenperature High Motor Bearing Tenperatures SGAHRS Initiation o Diesel speed and fuel oil indications will be available at the diesel generator local control panels in the Diesel Generator Building Cells 511 and 512.
- Each indicator, alarm and controller is repeated on each of the SGAHRS panels except for those associated with the AEW pumps. Panels A and B have the controls, alarms and indicators for motor driven AEW pumps A and B; Panel B has those associated with the steam driven AEW pump.
7.4-8d Amend. 69 May 1982
eu,w>s Page - 5 [8,22] 87 7.4.3.2 Desian Analysis
%e Remote Shutdown Systen provides the RSMP fran which an operator can assess the progress of the plant shutdown and cannand the local operation of the plant systems (primarily SGAHRS) to effect the shutdown. It should be noted that the PACC subsysten of SGAHRS is autanatically initiated by all reactor trips, and it remains in operation for the cbration of the plant shutdown or as lang as the reactor generates significant decay heat.
%e Renote Shutdown Systen inposes no special requirements on the plant systens, but takes advantage of the following systen design features:
o %e ability to operate in both local and renote modes with isolation fran and annu m ation in the Control Room when operating in the local mode.
a w reamrkncy diversity, separation, isolation and reliability of the safety grade systens.
o %e design and location of safety grade systems equipnent that minimize the probability and effect of fires and explosions on the ability of the systems to perform their safety function.
o %e rechndant safety grade SGAHRS provides the capability to achieve and maintain hot shutdown and, if desired, to cool the plant to and maintain the plant at refueling conditions.
o When transferring SGAHRS to the local mode, the operator manually starts SGAHRS. Once started, SGAHRS autanatically controls those parameters used i
to remove decay heat.
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7.4-8e Amend. 69 May 1982
Pags - 4 [82-0362] 084 Question CSA21.18 Prwide documentation that verifies that control prwided for safe shutdown fra outside the control rom will include the capability for reset of any engineered safety features equipnent having a high likelihood of being -
autmatically initiated & ring the normal transient occurring follwinci a manual reactor trip. For example, the auxiliary fee &ater syste may be in this category.
Response
he Auxiliary Fee &ater (AEW) and Protected Air Cooled Cmdenser (PACC) are subsystems of the Steam Generator Auxiliary Heat Renwal Systen (SGAHRS). Se AEW Subsystem is not initiated during the normal transient occurring follwing a manual reactor trip.
We PAOC subsystem is autmatically initiated by all reactor trips, and it remains in operation for the & ration of the plant shutdown or as long as the reactor generates significant decay heat. Se PACC has the capability of being reset at the local panels. E en, the operator can manually start and stop the PACC units. Once started the PAOC units will aut matically control stern drum pressure the same as in the main control rom.
OCS421.18-1 Anend. 69 Mau 10 A9
P;gs - 5 [82-0362] 984 Ouestion CSD1.19 A nunt>er of concerns have been expressed regarding the adequacy of safety systens in mitigation of the kinds of control systen failures that could i actually occur at nuclear plants, as opposed to those analyzed in PSAR Chapter -
! 15 safety analyses. Although the Gapter 15 analyses are based on j
conservative assumptions regarding failures of single control systens, I
systenatic reviews have not been reported to denonstrate that multiple control systen failures beyond the Gapter 15 analyses could not occur because of l single events. 7 nong the types of events that could initiate such multiple failures, the most significant are in our judgernent those resulting fran failure or malfunction of power supplies or sensors ccanon to two or more control systens.
l l
To provide assurance that the design basis evet analyses adequately bound
' multiple control system failures you are requested to provide the following informaticm l 1) Identify those control systens whose failure or malfunction could seriously impact plant safety.
- 2) Indicate which, if any, of the control systens identified in (1) receive power frcra ccanon power sources. he power sources considered should l
l include all power sources whose failure or malfunction could lead to failure or malfunction of fnore than one control systen and should extend to the effects of cascading power losses che to the failure of higher level distribution panels and load centers.
- 3) Indicate which, if any, of the control systens identified in (1) receive input signals frcrn ccanon sensors, costrnon hydraulic headers, or carnon impulse lines.
%e PSAR should verify that the design criteria for the control systens will be such that simultaneous malfunctions of control systers which could result fran failure of a power source, sensor, or sensor impulse line supplying power or signals to more than one control systen will be bounded by the analysis of anticipated operational occurrences in Chapter 15 of the Final Safety Analysis Report.
Response
1 he design criteria for the Plant Protection Systen prohibits Werefore, there control systen are no control malfuncticos frca endangering plant safety.
system failures or malfunctions that seriously impact plant safety Failure because of in the protection provided by the Plant Protection Systen (PPS) .
following control systens could, however, cause a reactor scram to occur:
Supervisory Control, Reactor Control, PHTS and IHTS Sodium Flow WeControl, PH'IS and IH'IS Pump Speed Ccntrol, Drum Level Control and Turbine Control.
Chapter 15 analysis envelopes the failure of multiple control systens che to loss of power since:
OCS421.19-1 Amend. 69 Mw 100
82-0362 Pagg - 6 [82-0362] 084
- 1) For loss of offsite power, the PPS trips the control rods upon loss of power to the sodim pumps. Action of the control syst e is irrelevant.
- 2) Primary rod control has reindant Ki cets powered fra non-UPS normal A and B sources. Ioss of A or B does not affect rod motion. For loss of A and B, a PPS trip occurs &e to steadfee&ater mismatch resulting from a turbine / generator trip.
- 3) Failure of electrical power (non-UPS normal A) to the Supervisory Control and Reactor Control Systers will not result in primary control rod withdrawal. W e control rod rate circuit will produce a zero rod rate signal with zero power available. %e worse that can happen on the loss of norr-UPS normal electrical power is a re&ction in coolant flow which is enveloped in the Chapter 15 analysis.
- 4) For Supervisory Control, Reactor Control, PHIS Sodium Flow control and IHTS Sodium Flow Control, the design provides for controllers in different cabinets each with reindant power supplies to eliminate power supply failures affecting several controllers.
Superheater exit steam flow sensors are shared by the Supervisory Control and Drum Level Control Systems, but median select circuits are used to prev et single sensor failures form causing an abnormal condition and resulting reactor scram. Loss of power to the median select circuits will result in a lowering of the steam drum level and a re&ction in reactor power. %e Plant Protection Systs will trip the reactor on a " low steam drum level" trip. Se loss of power to the median select causes the superheater exit steam flow signal to go to zero indicating zero steam flow. %is causes the steam drum level control systs to close the feedwater control valves resulting in a decrease in the steam drum level. It also causes the supervisory control systs to decrease reactor power in order to keep reactor power qual to plant thermal power as indicated by the superheater exit steam flow signal.
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QCs421.19-2 Amend. 69 l Mav l oR?
Page - 9 [82-0362J #85 Due511Da.CSA2L20 As a result of the Loose Parts Monitoring Briefing held on February 24, 1982, the staf f requires a f ormal submittal for the following: )
(1) An analysis f or all loose objects that can occur in the primary, intermediate, and the steam systems and their ef f ect on saf ~e ty.
(2) An arialysis f or the potential ef f ects of crud on saf ety.
(3) An analysis f or a system to detect f ailures through a noise diagnostic progran.
(4) Criteria f or a system that will satisfy Regulatory Guide 1.33 (i.e.,
CRBR needs to develop their own threshold analysis).
(5) The design concepts being co...!dered with a demonstration of feasibility.
EcspDDSD The CRBRP Project has developed and is implementing an action plan that will The action plan f ully evaluate the need f or CRBRP Loose Parts Monitoring.
will be conducted in two phases as follows:
o Phase I - Development of the basis for CRBRP Component Degradation Moni tor i ng o Phase II - Identification of the requirements and implementation needed to support the basis !n Phase 1.
Phase 1 of the action plan will establish data needed to a_sure that the 079RP that the CRBRP provides a level of saf ety comparable to LWR's and identify general approaches to obtain the needed data.
Phase II of the action plan will (a) Identify the specific monitoring requirements and design changes (If any) needed to support monitoring requirements; (b) establish the operational and limiting criteria for (a); (c) determine the specific methods for Implementing (a) and (b); and (d) develop a plan f or research, development and test, if needed, to demonstrate the practicability of (c) above.
Phase I of the action plan is scheduled f or completion by September 15, 1982 with the Phase il ef f ort scheduled f or completion by February 28, 1983.
Reports on the outcome of these actions will be available af ter the above da te.
l QCS421.20 Amend. 69 May 1982 l
l .. ....
P ge 2 (82-03630 [8,223f 86
_0uestion CS421.23 in the CRBR PSAR Section 7.6, several instrumentation and control systems are IIsted as being required f or saf ety which have not been included in the
' f ollowing discussion in Section 7.6. It is apparent f rom our review of this section that these are systems which have been omitted and also, have' not been com pl eted. The staff requires additional Inf ormation to complete out review of Section 7.6.
Etsponse The inf ormation related to Instrumentation and control for the following saf ety related systems is being developed and will be provided in a July 1982 amendment of PS AR Section 7.6.
5-marcency Plant Service Water System;
- 2. Emergency Chilled Water System;
- 3. Recirculation Gas Cooling System;
- 4. Nuclear Island Heating, Ventilating and Air Conditioning System.
I 1
QCS421.23-1 Amend. 69 May 1982 l
l
Pag 2 - 20 [8,22] #32 Ouestion CS491.18 (4.3.3)
Cite references where your proce&res (methods, codes, models and data) have been clearly empared with some other laboratory's proce&res for the calcula- l tion of Doppler coefficients, sodium void coefficients, control rod worths, ;
power and flux distributions, material worths, burnup, bowing reactivi'ty 1 coefficients, power coefficients, taperature defects, startup ooefficients, etc. Some of the fundammtal neutronics parmeters of the CRBR are Doppler i coefficients, sodim void coefficients, control rod worths, power and flux distribution, burnups, and bowing reactivity coefficients. Identify the particular safety consideration that you feel is the most impacted, limited, or made uncertain by each of the above parmeters. Ren, identify the most uncertain link in the calculational chain for each of the parmeters.
ResNmse Were are two major areas where published CRBRP fast reactor analysis results can be cm pared with independent results fr a other laboratories. % e first is the Zero Power Plutonium Reactor (ZPPR) Cooperative Analysis Progra with participation by Westinghouse, Argonne National Laboratory and General Electric. In the ZPPR analyses, cross section data and calculational methods are benchmarked against measured integral parameters (criticality, control rod worth characteristics, fission rates, sodium void worth, small-sample material worths, and others) in a zero-power full-scale mockup of the CRBRP core. %e Westinghouse calculations can be cmpared with these from ANL and GE, as well as with the measured parameters. %e results from these caparisons are used to assess bias factors and uncertainties in CRBRP nuclear performance characteristics calculated with the same methods and cross section data, a sumary of which is contained in Sections 4.3.3.3 through 4.3.3.9 of the CRBRP
! PSAR. % e following references contain det 'Is of the analysis of the ZPPR-7 experiments:
Westinghouse: CRBRP-ARD-0237, "ZPPR-7 and 8F Cooperative Analysis Program: Critical Experiments," R. V. Rittenberger and J. A. Lake, March,1979, (Availability: USDOD-TIC).
ANL: Epclear Technoloav 44, " Physics Studies of a Heterogeneous Liquid-Metal Fast Breeder Reactor," M. J. Lineberry, et. al., pp. 21-43, June, 1979.
GE: CRBRIH3ER-00025, " Analysis of the ZPPR-7 Critical Experiments," A. K. Hartman and J. T. Hitchcock, July, 1977, (Availability: USDOS-TIC).
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QCS491.18-1 Amend. 69 May 1982
www Page - 21 [8,22] 032 W e second area where CRBRP calculational methods can be c apared with those fr a independent laboratories is the Large Core Code Evaluation Working Group (IIrDC) . IOCDC is a cooperative effort, supported by USDOE, and including participants fr a all the major fast reactor analysis organizations
- where the results of fast reactor nuclear performance characteristics, calculated with various neutronics codes, are c apared. Rese characteristics include'k peak-to-average fission rate, breeding ratio, control rod worth, burnup ,ff, reactivity, neutron balance, and sodim void worth. Se following references summarize the results of the analysis of the cmpleted benchnark problens:
DOF/IIC-1027427, "he Large Core Code Evaluation Working Group Benchnark Analysis of a Emogeneous Fast Reactor," Septaber, 1981.
DOF/IIC-2005709 and 2005710: "h e Large Core Code Evaluation Working Group Benchnark Analysis of a Heterogeneous Fast Reactor," January,1982.
All of the neutronic parameters listed in the question, except sodim void coefficients, factor into a variety of design basis events. %ey are treated at length, including uncertainties, in Chapter 15 of the PSAR. Se sodi m void coefficients are significant for the HC R. Bis parameter, with uncertainty variations, is covered in depth in GBRP-3, Vol.1 (Ref. 491.10-1) and CRBRMIER-00523 (Ref. 491.18-2) which are on the docket.
Ref. 491.18-1 CRBRP-3, Vol.1, " Energetics and Structural Margin Beyond the Design Base," Dept. of Diergy, CRBRP Project Office, March 1982.
Ref. 491.18-2 CRBRM;ER-00523, "An Assessment of HOR Diergetics in the CRBRP Heterogeneous Reactor Core," General Dectric Co., Dec.1981.
- Participants in IICDG include Atomics International, Argonne National Laboratory, Cmbustion Engineering, General nectric, Hanford Engineering Developnent Laboratory, Los Alamos Scientific Laboratory, Massachusetts Institute of Technology, Oak Ridge National Laboratory, and Westinghouse Mvanced Reactors Division.
QCS491.18-2 Amend. 69 May 1982
Page - 21 [8,22] #62 Ouestion CM90.11 r The basis for construction of 99% conf idence bands f or the CDF f uel evaluation model was criticized by the partial draf t saf ety evaluation report (SER)
The question involved has prepared in 1977 on pages 4.2-44 through 4.2-46.Please discuss the relative merits of the not been resolved.
Ref erence (58) to Section 4.2 and the method suggested in the partial SER.
Please also perf orm the evaluation of the two methods suggested on page 4.2-46 and provide the result. All page numbers ref er to the partial draf t SER.
Resoonse The Project is unaware that any SER for CRBRP has ever been issued by the NRC.
The following response is provided on the assumption that the basis for construction of the 995 confidence bands for the CDF f uel evaluation models is a concern to the WRC.
In the COF procedure, as outlined in Ref erence (58) of PSAR Section 4.2, the properties of the cladding are described byFor linear regression design equations level analyses, thewhich were formulated using experimental data.
uncertainties in the cladding properties are treated via 99% confidence bands about the regression equation. .
The partial draf t SER raiseo Shree issues related to the conf idence bands and their mode of application In the CDF procedure. In the following response, each issue is addressed separately.
Glossarv A = Intercept of regression equation, B = slope of regression equation (i.e., regression coef ficient),
N = number of data points, S = standard error of estimate, E
T = student's T-stati stic, Vg = variance in the Intercept, V = variance in the slope, B
V = variance about the Independent variable in the design environment, D
V(7) = variance in the expected true value, X = mean of Independent variable f ran test data, X = value of Independent variable used in calculation, X = possible random verlate about 2,
= I-th measured value of the Independent variable from test data, X,
Y = mean of dependent variable f rom test dste, QCS490.11-1 knend. 69 May 1982
Page - 23 [8,22J #62 k = Dest estimate of the true value of the property, Y = possible random variate of the true value of the property, Y
D
= Possible true value of the property in the design environment,-
7, = possible measured value of the property in the source test ,'
environment, Y, = l-th measured value of the property from test data, CA = probabliIty (confIdonce) level,
= standard deviation aoout X, 73
= standard deviation aoout Y, 9
V = degrees of freedom.
A. Valldity of the Ecuation Used to Comoute the Confidence Bands in the CDF technique, as described in PSAR Ref erence (58), the variances used to calculate the confidence bands about the regresslor. equations are computed according to vCh = S$ ((VN) + (A-E /[(N-0#I] f(A.1)
The validity of the aDove equation has been questioned and the suggestion made by NF<C that the proper variance is V = Sf [L l 4 (VN) + ( ~IO [(N-1) 6 3 (A.2)
Howe /er, Equation ( A.2) is meant to deal exclusively with predicting possible measured values of the property in the source test. Whereas, Equation (A.1)
! defines the vart ence in the true value of the property.
Clearly, the Intent of 1he design procedure is to deal with expected values of the property in the design environment. Thus, the CDF procedure employs l
Equation (A.1) and adds to this the anticipated variances (i.e.,
uncertaintles) In the Independent variables as encountered in the design l
environment, l
To highlight the f undamental dif ferences between the two equations and their application, they are derived in parallel in the discourse which follows.
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QCS490.11-2 Amend. 69 May 1982
9tge - NJLCld,7EGRAZ5 Given a test which yloids N measured values of a property, Y, and the cssociative values the Independent verlable, X, then, by definition:
et
- ( i p gg,3) u 2 y e (l[(N-D [ h 9 (A 4) sol H
(I[N i 3 gg,5)
- 2 l
8 h ~
~
- (A 6) t l
If Y is linearly dependent on X then a linear regression equation can be f ormulated so that ,
9 = ~A t B X (A.7) l describes a line (i.e., a regression !!ne) yhich is a best estimate of the mean (true) value or the property when X = X.
The standard error of estimate for the N measured points is given by S[ = [lANs( tal Ni-%) c3,33 where is the value of the regression equation evaluated at the test's I-th valuo ot X. The standard error of estimate is somewhat camparable to the l standard deviation vis-a-vis the regression line; It accounts for all random variations in the test data such as random measurement errors, verlebility in the pre-conditioning of the test specimens, variability in the test's environment and the related variability in the Independent variable.
The variance in the intercept of the regression equation is given by A (A.9)
QCS490.11-3 Amend. 69 May (982 on_ntoi
Phge - 25 [8,22J #62 cnd the variance in the slope Is given by v, = sl/y" d-xsf = s;/cof-o oh ( A.10)
T.i ,
The variance in k about the regression line at X=X is obtained by summing the variances in the slope and intercept, i.e.,
! Vc9) = % + Vs ( A-R )* (A.11)
. Note that Y(Y) is the variance In the estimate of the property's true value as c.etermined f rom the test and is not the variance in the test data.
Assume for a moment that alI colIateral varIablIities need not be considered.
In this case, a single possible value of the true property at X=I would be given by Y=f + RV( )u ( A.12)
In other words, tchere R is understood to be a suitable multiplicative f actor. e Equation ( A.12) defincs the variability of the expected true value of the property if X were known precisely and if there were no extrinsic random errors.
i l in reality, the overall variability in the expected value of the property
' depends on the collateral variabilities that are operating under any given set I of conditions; these collateral variances must, of course, be added to Equation ( A.12) .
If one were Interested in estimating a possible measured value of the property in the source test
- Itself then the appropriate collateral verlance must include the random measurement errors, the variability in pre-conditioning, etc. This, of course, is described directly by the standard error of estimate given by Equation (A.8). It follows theref or3, that in the source test, a possible measured value or the property at X=X is given by
'A source test is defined as the test used to obtain the N measured data points one which is laentical in every way.
l QCS490.11-4 Amend. 69 l May 1982 l
P:ge - 26 L8,22.1962
+ R S, s h + RV ( ) .
Expanding Equation (A.13) yleids
( = A + Bk + RSg () + (14d + ($t-lif(((N-D c)1 (A.14)
Note that the variance employed in Equation (A.14) is exactly the form suggested by NRc Lt.e., see Equation (A.1)] and that (A.14) is relevant only in oomputing possible measured values in the source test.
If the property in question is to be used in a design calculation, then there is na Interest in the collateral variance associated with the source test's random errors, indeed, for a design calculation, the appropriate collateral vcriance(s) must reflect the uncertainties in the design environment as well es the specific nature of the design problem.
Typically, the collateral variances associated with the design environment are treated via an assigned uncertainty in the Independent variable. Thus, let Y D be the assigned variance about x in the design environment. Then, a possible true va!ue of the property in the design environment is given by
~ + Yg Yo = Y + R i VM^ Vg + B R2 Vp
( A.15)
Expanding Equation (A.15) yleids Y
% = A + a [f( + Rz [*] + R Se { ( 1/n) +
($ 4- RzYD *- X )(((N-1)Crj } E ( A.16) f.e.,
Let I be a possible random variate of the independent parameter, l
I
^ %.
(
X=x&RVpt
~
( A.17)
Then, it follows from Equation (A.16) that
~ s Yo = A + BX + R Sg { (IM + (~K-Xf(((q.f)a ]7 l/2.
y i x ]
l
' (A.18)
QCS490.11-5 Amend. 69 May 1982
P ge - 27 L8,22J #62 Equation ( A.18) is a direct expression of the typical application of a regression equation to a design problem. Specifically, the design level estimate of the true property is computed using the regression equation and its variance, both evaluatd at the design level value for the independent variable. -
Equation (A.18) represents the procedure employed by the CDF technique th compute cladding properties; note that the variance used in (A.18) is exactly that specified in PSAR Ref erence (58).
B. The Distribution of Data Points ReIatIve to the ConfIdance Points About a Rearession Eauntion.
in Section IV of PSAR Reference (58), a series of Figures compare the regression equations to their source data, included in these figures are the 99% confidence bands about the regression line. In all cases, a large f raction of the data f all outside the confidence bands.
The question has been frequently asked as to why 99% of the data are tot snclosed with the confidence bands and condequently, questions the bands' vaildity.
As conventionally defined, confidence Iimits address the probabliIty that the true value of the property (i.e., the meen) lies within a certain range.
Thus, there is no requirement that the confidence limits encompass the data.
On the other hand, the limits which are expected to encompass a specified f raction of the data are f requently ref erred to as tolerance limits; these limits may be used to acdress the probability that a measured value (i.e., a dato point) will f all within e certain range.
Conventionally defined confidence p,andt about a regression equation can be described by Equation (A.18) with X = X and lR 3 l taken to be a proper value of the student's T-statistic. In other words, Y = A t B$ + TS,-
IM+ BEN d 3 (B.1) l l
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l QCS490.11-6
' Amend. 69 May 1982
Pcge - 28 L8,22J #62
)
i.e.,
Similarly, Equation (A.14) would describe the tolerance bands,
( = A+Bk 2 TSg fI + O/d + d 'N(((N-D ef] (8.2) .
Equation lies with the (B.1) addresjed range Y * \Y-T tjeg)obability
- Equation that the true h value of the property b bttitr that dala points wlil felI with the range k(B.2) addresses t e pro a1 l$-fl .
1%-Yl ; thus, one would anticipate that data points may fall outside the conf idence bands.
C. Conservativeness of the CDF Design Procedure.
As mentioned earlier, for design level analyses, the uncertainties in the cleading's properties are treated In the 99% confidence bands about their regression equations; these are taken to combine in the most unf avorable way possible. Depending on the property in question the most unf avorable level may be the upper or lower confidence band.
In Section V of PSAR Ref erence (58), the design procedure is examined and verified using a number of Independent sources of data. In these verification examples, the design proceaure successf ully bracketed, or would have precluded, all or the f ailure data from (a) multistage stress-rupture tests, (b) creep / tensile and creep / burst tests and (c) FCTT tests.
The veracity of the soove verifications has been questioned *.
Specifically, it has been suggested that if the confidence bands do not encompass all of the data from .e respective source tests, then the limits computed in the verificati mi examples could not have bracketed the f ailures as shown.
In view of t'ie discussions in Parts A and B, it should now be clear that, within the capabliltles of the models, the design procedure should bracket a vast majority of the f ailures. Specifically, it was shown that the confidence bands describe the limits of certainly about the true values of the properties.
'All of the inf ormation and data necessary for the duplication of the verification analyses are contained within PSAR Ref erence (58).
I i
QCS490.11-7 Amend. 69 May 1982
P ge - 29 L8,22J #62 Thus, the design level CDF, computed with the worst combination of 995 confidence bands should schlese unity prior to any CDF computed with random combinations of the properties as might be encountered in nature.
It was with this point in mind 15at the following study was undertaken. .
The objective of the study was to compare the results of a typical design level fuel pin analysis with that from a Monte Carlo analysis. In the Monte Carlo treatment, the values f or the various properties are randomly chosen f rom simulated experimontal populations and then combined at random. This is done in a large number of FURFAN computations. Thus, within the capabisities of the models, the Monte Carlo treatment yields a sample of CDF values crown f rom a population having variability comparable to that which would be obtained in nature, in this regard, the Monte Carlo distributions in the CDF cay be taken :ss estimates of the conditional probability of achieving givon CDF values.
The operating parameters assumed for this analysis are summarized in Tabie QCS490.11-1. These parameters correspond to coeditions with 2cFplant f actors prevailing; note that these parameters remain the same within each Monte Carle trial.
The results of the typical deterministic design level analysis for the subject pin is given in Figure QCS490.11-1. As shown, by combining all properties at their worst confidence levels a CDF = 1.0 is achieved af ter 540 EFPD. In comparison, a combination of properties at their nominal levels (with the 2 tr f actors still prevelling) yleids a CDF of 0.905 af ter 825 EFPD.
The Monte Carlo analysis involved 100 trials each using a set of properties unich were randomly chosen from their respective distributions and randomly combinea. rigure QCS490.11-2 gives the resultant distributions of CDF values at 540 EFPD, the time the design limit is achieved. Note that at this time, the entire population of possible CDF values is below unity with 995 of the values less than 0.8.
Figure QCS490.11-3 shows the cumulative distributions in the CDF's at various times during trhe 825 EFPD operating period. As such, this figure represents QCS490.11-8 Amend. 69 May 1982
P;ge - 30 LB,22j #62 cstimates or the time dependent, cumulative conditional probabilities of achieving CDF values less than some specific value, i.e., P(CDF S X).
Figure QCS490.11-4 compares the cululative conditional probability estimate for CDF S 1.0 to the design level and nominal CDF's. As shown, at the 540 EFPD limit P(CDF S 1)
In other words, at the steady state limit (i.e., 540 EFPD) there is almost a 100% probability that the (X)F is less than unity; also, ef ter 825 tFPD there is as 54% probability that CDF 11.0.
l l
QCS490.11-9 Amend. 69 May 1982 0 9_ AT 99
P;ge - 31 La,22) #62 TABLE QCS490.11-1 stme4ARY OF PARAMETERS tlRFD IN ANALYSIS inside Radius of Cladding: 0.100 inch .
Outside Radius of Cladding: 0.114 Inch -
Axial Location: X/L = 1.0 2c'PIant and 20' Hot Spot Factors TIE TEMPERATURE ( F) PRESSURE (DAYS) 1.D. 0.D. (PSI) 0 1308 1284 180 275 1236 1211 500 275 1259 1234 540 559 1195 1170 1150 550 1214 1187 1150 825 1152 1128 1750 QCS490.11-10 Amend. 69 May 1982 1
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4 l \s -
f o
I hN ~ d e
d a N
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a*
3 8
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N l h l mfk \
i l==W N\ -
I i leE
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s E' i
E Ns= \
g g x N s
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\ *
\ E
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- 8 g- s
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=s T i I E I -- E f f l f l
' ;. : : : : : : : : : ; =
wousannavnva nuv,nnna !
7I90-1 QCS490.111I l
l l
l l
8.30
- FRE00ECCY NISTOSRAM s *25 --
E man = ws STc.sEV.=t.152 g 0.20 ~
MALLEST VALUE = 0.071 g LARGEST VALUE = Etat
,, g,t g -
8.10 50.0s - -
l i i i 0.00 0.0 0.1 0.2 0.3 0.4 0.5 0.0 0.7 0.0 0.0 1.0 CDF 1.0 0.9 X -
vi 0.8
- u. CUMUL ATIVE FREQUENCY o
y 0.7 -
E a 0.6 -
o l 3:
5 0.5 -
?
E 0.4 -
o
! 0.3 13 E 0.2 -
0.1
! I I I I I I I I I t 0.0 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 X
l Figure QCS490.11-2. Distribution of Possible CDF Values at X/L = 1.0 After 540 EFPD of Steady State Operation 7I90-2 QCS490.l l-12
1.00
/
0.ee / ,s
/
' ,s.-
a
/
tal FFro I
/
- 8.10 j ene 0.50
- 0 I I
!8.50 / =
g i / m g
=
0.40 7 7
- E j I
/
,e 8.30 l ,
I 0.20 l
/
0.10 l
1 1 I I I I l
l l l O.00 2.0 0.8 1.0 1.2 1.4 1.6 1.8 0.0 0.2 0.4 0.6 X
l Figure QCS490.113. Cumulative Frequency Distributions in Possible CDF Values at X/L = 1.0 After Various Periods of Steady State Operation 7190-3 QCS490.11 13 v
i l
I*N PICOF S1) 0.00 0.00 0.70 w CDF AT MAXIMUM g 0.60 -
UNCERTAINTY M
$ O.50 E ,
2 m
$ O.40 0.30 0.20 0.10 CDF WITH NOMIN AL PROPERTIES V
i l l \
0.00 0 100 200 300 400 500 600 700 000 TIME (EFPD)
Figure QCS490.11 4. Comparison of the Conditional Probability that CDFS 1.0 and the Corresponding CDF Values at the Design and Nominal Levels 7190-4 QCS490.11 14
Page 22 (82-0321) [6,223 #63 Ouestion CSd90 33 Many computer codes were used by the CRBRP designers to perf orm theSome thermal of and hydraulle analyses presented in section 4.4 of the CRBR PSAR.
these codes are proprietary and some were developed by the CRBRP or.its contractors and are not widely evallable. To evaluate the applicability of these codes to the thermal and hydraulic analyses presented in section 4.4, .
substantially more Inf ormation is needed than is presented in section 4.4, in Theref ore, please 1 Appendix A, and in the ref erences cited in Appendix A.
provide code manuals and/or detailed descriptions along with code listings for the f ollowing codes.
- a. CATFISH
- b. CORINTH r mTEC
- d. CRSSA
- e. DEMD
- f. FATHOM-360
- g. FATHOM-360S
- h. FLODISC
- l. FORE-2M J. NICER
- k. OCTOPUS
- 1. TRITON Resconse Since last submission of the CRBRP PSAR, two of the codes in the above list have been superseded. The (MRINTH code has been replaced by the DOE national program code (UBRA-WC. Also, the FL90lSC code has been replaced by COBRA-WC In the f low for analysis of very low flow (natural circulation) conditions.
range f rom 100% to ~ 10% flow the FLODISC code has been replaced by CATFISH.
The CRAB-Il code which analyzes the primary control assembly steady state hydraulics, scram dynamics and flotation behavior should be added to the list.
Descriptions of the major features, models appilcations and typical results of the above codes have been reported in the open literature; a list of these papers / articles / reports is attached.
An extensive validation ef f ort is ongoing f or all of the above codes.
Manuals and validation reports f or all the above codes will be provided prior to FSAR submittal. Appendix A will be modif ied to reflect the above changes.
QCS490.35-1 Amend. 69 May 1982
P$9e23(82-0321)[8,22]#63 AVAILARfE OPEN LITERATURE PUBtICATlONS CATFlSH
- 1) M.D. Carelli and J.M. Willis, "An Analytical Method to Accurately Predict LMFBR Core Flow Distribution", Trans. Amer. Nucl. Soc., 32, pp. 575-576, 1979.
- 2) M.D. CarellI and J.M. WilIis, " Analytical ModelIng of Core Hydraulics and Flow Management in Breeder Reactors", Proceedings of the XVill Congress of the International Association for Hydraulic Research, Cagliari (Italy),
September 10-14, 1979.
PtnTrf'
- 1) E.H. Novendstern, " Mixing Model for Wire Wrap Fuel Assemblies", Trans.
Amer. Nucl. Soc. ,15., pp. 866-867, 1972.
- 2) E.H. Novendstern, " Turbulent Flow Pressure Drop Model for Fuel Rod Assemblies Utilizing a Hellcal Wire Wrap Spacer system", Nuct. eng.
Des ign, 22, pp. 19-27, 1972.
- 3) Y.S. Tang, M.R. Yeung and M.D. Carelli, "A Core Design Subchannel Ana'ysis Code Calibration and Validation", to be presented at the ANS Annual Meeting, Los Angeles, June 1982.
- 4) F.C. Engel, R. A. Markley and B. Minushkin, " Buoyancy Ef f ects on Sodium Coolant Temperature Profiles Measured in an Electrically Heated Mockup of a 61-Rod Breeder Reacter Blanket Assembly", ASE-78-WA-HT-25.
- 5) F.C. Engel, R. A. Markley and B. Minushkin, " Heat Transfer Test Data of a 61-Rod Electrically Heated LMFBR Blanket Assembly Mockup and Their Use f or Subchannel Code Calibration", in Fluid flow and Heat Transfer Over Rod or Tube Bundles, pp. 223-229, American Society of Mechanical Engineers, New Yor k, 1979.
- 6) F.C. Engel, R. A. Markley and B. Minushkin, "The Ef fect of Heat input l Patterns on Temperature Distribution in LMFBR Blanket Assembiles",
ANS/ASE International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Saratoga Springs, NY, October 1980, RJREG/CP-0014, Vol. 3.
DEMO
- 1) W.H. Alliston, "LMFBR Demonstraton Plant Simulation Model (DEM3)",
CRBRP-ARD-0005, February 1978.
FATHOM 360 AND FATHOM-360S
- 1) M.D. Chuang, M.D. Careli I, C.W. Bach and J. S. K!l I imayer, "Three-Dimensional Thermal-Hydraulic Analysis of Wire Wrapped Rods in Liguld Metal Fast Breeder Reactor Core Assemblies", Nuclear Science and Ina, fL4, pp. 244-257,1977.
QCS490.35-2 Amend. 69 May 1982
! Page 24 (82-0321) [8,22] #63 l
- 2) M.C. Chuang, M.D. Careill and M.R. Young, " Distributed Parameter Analyses of the Thermal-Hydraulic Behavior of Wire Wrapped Rods in LWBR Cores",
paper submitted to the 2nd international Topical Meeting on Nuclear Reactor Thermalhydraulics, Santa Barbara, January 1983. .
QCS490.35-2a knend. 69 May 1982
Page 25 (82-0321) [8,22] #63 NfCER
- 1) M.D. Carelli, C. W. Bach and R. A. Markley, " Analytical Techniques f or Thermalhydraulles Design of LMFBR Assembiles", Trans. Amer. Nu c I . Soc.,
12, pp. 423-424,1973. .
EIte.US
- 1) M.D. Carel I 1, A. J. FrIendl and, C.W. Bach and R. A. Markt ey, "An Optimized Method f or Orlf Icing LMFBR Cores", Trans. Amer. Nuef. Soc., 26, pp.
437-438, 1977.
- 2) M.D. Carelli and C.W. Bach, "Orf ficing interchangeable LMFBR Cores",
Trans. Amer. Nucl. Soc. ,1.4, pp. 268-270, 1980.
TT'ITON
- 1) M.D. Carelli and C.W. Bach, " Thermal-Hydraulle Analyses f or CRBRP Core Restrai nt Design", Trans. Amer. Nucl. Soc. , 21, pp. 393-395, 1975.
- 2) F.C. Engel, B. Minushkin and R. A. Markley, " Comparisons of Design Code Predictions with LMFBR Blanket Heat Transf er Test Results", American Nuclear Society and the European Nuclear Society November 1980 International Conf erence, Washington, D.C.
CRAB and CRAB-1i
- 1) M.D. CarellI, C.W. Bach and R. A. Markiey, Hydraulic and Scran Dynamics Analysis of LMFBR Control Rod Assemblies", Trans. Amer. Nucl. Soc., 16, 1, pp. 216-219,1973.
- 2) M.D. Careill, H.W. Brandt, C.W. Bach and H.D. Kulikowski, "LMFBR Control Rods Scran Dynamics", Trans. Amer. Nucl . Soc. ,18, pp. 278-27 9, 1974.
- 3) M.D. Carei l l, L. A. Baker, J.M. Wil l is, F.C. Engel and D.Y. Nee, " CRAB-l l:
A Computer Program to Predict Hydraulles and Scram Dynamics of LMFBR Control Assemblies and its Validation", to be presented at the ANS Topical
( meeting on Reactor Physics and Core Thermal-Hydraulics, Kiamesha Lake, NY, Septanber 1982.
COBRA-WC
- 1) T.L. Geerge, K.L. Basehore, C.L. Wheeler, W. A. Prather and R.E. Masterson, "CGBRA-WC: A Version of COBRA for Simple-Phase Multi-Assembly Thermal-Hydraulic Transient Analysis", PNL-3259, July 1980.
- 2) E.U. Khan, W. A. Prather, T.L. George, J.M. Bates, "A Val idation Study of the COBRA-WC Computer Program f or LMFBR Core Thermal-Hydraulle Analysis",
l PNL-4128, December 1981.
QCS490.35-3 Amend. 69 May 1982
Vtge 2FIM'F&RNTlDJi& ORA 9 FORE-2M
- 1) J.N. Fox, B.E. Law ler and H. R. Butz, "F9RE-1 I : A Computational Progren f or the Analysis of Steady State and Transient Reactor Perf ormance",
GEAP-5273, September 1%6. .
- 2) J.V. Mil l er and R.D. Cof f lei d, " FORE-2M: Modifled Version of ttie FORE-Il Computer Progran f or the Analysis of LMFBR Transients", CRBRP-ARD-0142 (avalIabt e f rom US/ DOE Technical Inf ormation Center), November 1976.
QCS490.35-3 a Amend. 69 May 1982
.. . _ . .