ML19221B048
| ML19221B048 | |
| Person / Time | |
|---|---|
| Issue date: | 11/24/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-75-087, NUREG-75-087-03.9.4, NUREG-75-87, NUREG-75-87-3.9.4, SRP-03.09.04, SRP-3.09.04, NUDOCS 7907120311 | |
| Download: ML19221B048 (6) | |
Text
NU R E G-75/087 fg, Rg%
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U.S. NUCLEAR REGULATORY COMMISSION h['b STANDARD REVIEW PLAN g
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OFFICE OF NUCLEAR REACTOR REGULATION SECTION 3.9.4 CONTROL POD CRIVE SYSTEMS FEVIEW PESPGNSIBILITIES Primary - Mechanical Engineering Branch (PEB)
Secondary - Ptactor Systtms Branch (RSB)
Materials Engineering Branch (MTEB)
I.
AREAS OF PEVIEW Infarnation in the areas noted below is provided in the applicant's safety analysis report and is reviewed by the PER in accordance with this plan.
This information pertains to the reactor control rod drive system (CPDS), wnich is considered to extend to the coupling interface with the reactisity control elements in the reactor pressure vessel. For electro-nagretic systems, the review under this pla7 is limited to jult the control rcj drive rechanism (CPDM) portion of the CRDS. For hydraulic systems, the review covers the CPDM and also the hydraulic control unit, the condensate supply systen, and the scram discharge volume. For both types of systens, the CRDM housing should be treated as part of the reactor coolant pressure boundary (RCF B); the relevant mechanical engineer ing -formation may be presented in this section or by reference to the sections on tW FCPB.
If other types of CRDS are proposed nr if new features that are not specifically mentioned here are incorporated in CRDS of current types, infonaation should be supplied for the new systems or new features similar to that described below.
1.
lhe descriptive information, including design criteria, testing programs, drawings, and a survaary of the method of operation of the control rod drives, is reviewed to permit an evaluation of the adequacy of the system to perform its mechar.ical function p n]pe rl y.
2.
A review is performed cf information pertaining to design codes, standards, specifica-tions, and standard practices, as well as to General Design Criteria, Regulatory Guides, and branch positions tnat are applied in the design, fabrication, construction, and operation of the CRDS.
USNRC STANDARD REVIEW PLAN Standard re*w p
..e em prep.r.d f., a. eued.nce of the oence Of Naceeer meeeter neevietw ete+f rup.asibie for me revi.w Of appiicatione is conet,.ci.nd opereto nuc6 ear power plante These documente are mode eve 84eble to the pwblic es port of the Commioo6on's po#cy to infonn the nucieer industry and the 9
genered pubile of reguletery precedwees end poucle. Stenderd rowiew piene are not oubetttweee for regvisN 4. guidos er the Cor.mtesten a regwieelene end eemplience wfth them 60 not regulfed The Stenderd rewleg pien sectione tre heyed tg Rewfolen 2 of the Stende 4 Formet and Coetent of Cefety AneSyste Reporte for Nuclear power piente Not GFl se4* lone Of the $tenderd Fomet have e corresponding review plan.
pwednahed standard rewtow peone weM be rev* sed periodicelty. et appropriete. to accons modate sommente end to reMeet new informetten and emportance Cgen, ente end suggest6gne for lepreg gment wlM be ceneidered end showed be cent to the U $ Nuclear Regulatory Cornmaeolon.Ot'.ce of Nucteer Reector Nogulatlen. Weehengtem D C 2 Gile D HI 790moaan
The various criteria, described in general terrs above, shoulJ be supplied along with the nares of the apparatus to which they apply. Pressurized parts of the system are reviewed to determine the extent to which the applicant complies with the Class 1 requirements of Section III of the American Society of Pechanical Engineers (ASPE)
Boiler and Pressure Vessel Code (hereafter "the Code") for those portions which are not part of the reactor coolant pressure beundary, and with other specified parts of Section III, or other sections of the Code for pressurized portions which are not part of the reactor coolant pressure boundary. The PEB reviews the non-pressurized portions of the control rod drive system to determine the acceptability of design rargins for allowable values of stress, deformation, and fatigue used in the analyses. If an experirental testing pragram is used in lieu of andlysis, the program is reviewed to deterTaine wheth* it adequately covers the areas of concern in stress, deformation, and fatigue.
3.
Infor ation is reviewed which pertains to the applicable design loads and their appro-priate combinations, to t.e corresponding design stress limits, and to the corresponding allowable deformations. The deformations are of interest in the present context only in those instances where a failure of movement could be postulated due to excessive deformation and 3uch r:ovement would be necessary for a safety-related function.
If the applicant selects an experimental testing option in lieu of establishing a set of stress and defornation allowables, a detailed description of the testing program riust t o provided for reviow.
In the preliminary safety analysis report (PSAR), the load corbinations, design stress limits, and allowable def ormations criter'a should be provided for review.
In the final safety analysis report (FSAR), the actual design should be compared with the design criteria and limits to demonstrate th + the criteria and limits have not been exceeded.
Loadings imposed during normal plant operation and startup and shutdown transients in-clude but are not limited to pressure, deadweight, temperature ef fects, and anticipated operational occurrences. Loadings associated with specific seismic and other dynamic events are then combined with the above plant-type loads. Each set of con.bined loads has 3 selected stress or deformation limit. The selection of a specific limit is influenced by the probability or the postulated event occurring and the need to assure operation during and after the event.
4.
The portion of the SAR is reviewcd that describes plans for the conduct of an operability assurancs igram or t:.at references previous test programs or standard industry pro-cedures for similar apparatus. For xample, the life cycle '.est program for the CRDS is reviewed. The operability assurance program is reviewed La ascertain coverage of the follow'eq:
a.
Lite cycle test program.
b.
Proper service envircnment irposed during test.
3.9.4-2 i46 312
c.
Mechanism functional tests.
d.
Program results.
II.
ACCEPTANCE CRITERIA The acceptance criteria for the areas of review are the following:
1.
The descriptive information is determined to be suf ficient provided the ninimum require-ments for such information meet Section 3.9.4 of Reference 14 2.
Construction (as defined in NA-ll10 of Section III of the ASME Code, Reference 10) should meet the following codes and standards utilized by the nuclear industry which have been reviewed and found acceptable:
a.
Pressurizet Portiens of Ecuipment Classified as Quality Group A, B, C (Regulatory Guide 1.26 Section III of the ASME Code, Class 1, 2, or 3 as appropri3te (Ref.10).
b.
Pressurized Portions of Equipment Classi'ied as Quality Group D (Requlatory Guide 1.26)
(1)Section VIII, Division 1 of the ASME Code for vessels and purp casings (Ref.
10).
(2) Applicable to Piping Systems (American National Standards Institute, ANSI)E -
Bi6.5 Steel Pipe Flanges and Flanged Fittinas (Ref.16).
B16.9 Steel Butt Welding Fittings (Ref. 17).
B16.ll Steel Socket Welding fittings (Ref. 18).
B16.25 Butt Welding Ends (Ref.19).
B31.1 Piping (Ref. 20).
SP-25 Standards (Ref. 21).
SP-66 Valves (Ref. 22),
c.
Non-Pressurized Equipment (Non-ASME Code)
Design margins presented for allowable stress, defornation, and fatigue should be equal te or greater than those for other plants of sinilar design having a period of successf ul operation. Justification of any decreases shoJld be proVided.
3.
For the varicus plant operating conditions defined in NB 3113 of Section III of the ASME Code (Ref. 10), load combination sets are as given in Standard Peview Plan 3.9.3 (Ref. 15). The stress limits applicable to pressurized and non-pressurized portions of the control rad drive systems should be as given in Ref erence 15 for each loading set.
4.
The operability assurance program will be acceptable provided the observed performance as to wear, functioning times, latching, and overcoming a stuck rod meet system design requirerents.
III.
REVIEW DROCEDURES The reviewer will select _nd emphasize material from thE procedures described below as may be appropriate for a particular case.
1/ This list can be extended by a staff review and acceptance of other ANSI & MSS Standards in the piping systen area.
3.9.4-3 146 313
1.
The objectives of the review are to determine that design, fabrication, and construc-tion of the control rod drive mechanisms provide structural adequacy and that suitable life cycle testing programs have bcen utilized to prove operability under service conditions.
In the construction permit (CP) review, it should be determined that the design criterio utiliza proper load corbinations, stress and defortnation limits, and thai. Operability assurance is provided by reference to a previously accepted testing program or that a cocinitment is made to perform a testing program which includes the essential elements listed below. In the operating license (OL) review, the results of any testing program not previously reviewed should be evaluated.
2.
The design criteria presented should be evaluated for both the internal pressure-containing portions and other portions of the CRDS. These include the CRDM housing, hydraulic control unit, condensate supply system and scram discharge volume, and portions such as the cylinder, tube, piston, and collect assembly.
Of particular concern are any new and unique features which have not been used in the past. Pressure-containir.g components are checked to ensure that they meet the design requirerents of the coGas and criteria which have been accepted by the Reactor Systems Branch, and are identified in Standard Review Plan 3.2.2.
The review of the functional design of reactivity control systems, including control rod drive systems, is the responsibility of RSB (See SRP 4.5).
The loading corbinations for the various plant operating conditions are checked for consistency with Reference 15; given these loading combinations, the stress limits of the appropriate code should not be exceeded, cr the limits in Peference 15 should not be exceeded if not specified in the listed design code. Exceptions taken by the applicant to any of the accepted codes, standards, or AEC criteria must be identified and the basis clearly justified so that evaluation is possible. Engineering judgment, experience, comparisons with earlier cases and design margins, and cc.isultation with supervisors permit the reviewer to reach a decision on the acceptability of any exceptions posed by the applicant.
The choice at materials of construction for unpressurized equipment that is not goverrM by accepted codes or standards is reviewed by the MTEB.
3.
Loading combinations are defined as these loadings associated with plant operations which are expected to occur one or more tines during the lifetime of the plant dnd in-clude but ure not limited to loss of power to all recirculation pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all offsite power, combined with loadings caused by natural or accident events. The load combinations which are postulated to occur are specified for each of the plant cperating conditions as defined in Paragraph NB-3113 of the ASME Code (Ref.10). These load combinations are defined in Reference 15 and are compared by the reviewer with those provided by the applicant.
3.9.4-4
The desion stress lin its, including fatigue limits, and deforration limits as appro-priate to the cortponents of the contral rod drive rechanisn are compared by the review:r with those of specified codes, previously designed and successfully operating systers, or with the results of scale model and prototype testing programs.
4.
The control rod drive mechanisms of a new design or configuration should be subjected to a life cycle test program to deterrine the ability of the drives to function over the full range of temperatures, oressures, loadings, and mi3alianr.ent (xpected in service, The tests should include functional tests to determine times of rod in-scrtion and withdrawal, latching operation, scram operation and time, system valve operation and scram accumulator leakage for hydraulic CPDS, ability to overcome a stuck rod condi' ion, and wear.
Pod travel and nurber of trips expected during the mecnanism operational life should be duplicated in the tests.
The reviewer checks the elerents of the test program to be sure all recuired paran-eters '. ave beer included and finally reviews the test results to determine acceptability.
Excessive wear, malf unction of compcnents, operating times t'eyond determined limits, stran accurulator leakage, ett<, all w Mid be cause for retesting.
IV.
EV4 UATION FINDINGS The reviewer verifies that suf ficient information has been provided to satisfy thm require-rents of this review plan, and that his evaluation is sufficiently complete and adequate to support conclusion: 7f the following type, to be included in th" staff's safety eval-u3 tion report:
"The design criteria and the testieg program conducted in verification of the nechanical operability and life cycle capabilitics of the reactivity control system are in cc fonmance with established criteria, codes, standards, and specifications acceptable to thc "equlatory staff.
The use of these criteria provide reasonable ar,surarce thit the systen will function reliably when re-quired, and fom an acteptable basis for satisfying the rechanical reliability stipulations of General Dt;ign Criterion 27.'
V.
REFERENCES 1.
10 CFR Part 50, Appendix A, General Pesign Criterion 2, " Design Bases for Protect:an Against Natural Phenomena."
2.
10 CFR Part E0, Appendix A, General Design Cr terion 14, "Peactor Coolant Pressure i
Boundary."
3.
10 CFR Part 50, Apperdix A, Gene al Design Criterion 15, " Reactor Coolant System Design."
4.
10 CFR Part 50, Appendix A, General Design Criterion 20, " Protect:on System Functicns '
n 5.
10 CFR Part 50, Appendix A, General Design Criterion 26, "Reacti /ity Control Sys tera Redundancy and Capability.'
h} 3.9.4-5
6. 10 CFR Part 50, Appendix A, General Design Criterion 29, " Protection Against Anti ci p3 teri Operational Occurrences." 7. 10 CFR Part 50, Appendix A, General Design Criterion 30, " Quality of Peactor Coolant Pressure Boundary." 8. 10 CFR Part 50, Appendix A, General Design Criterion 31, " Fracture Prevention of Peat. tor Coolant Pressure Boundary." 9. 10 CFR Part 50, Appendix A, Ger.ra! Design Criterion 32, " Inspection of Peactor Coolant Pressure Boundary." 10. ASME Boiler and Pressure Vessel Code, Sections III and VIII, American Society of Mechanical Engineers, 11. Regulatory Guide 1.26, " Quality Group Classifications and Stanaards." 12. Regulatory Guide 1.29, " Seismic Design Classification." 13. Regulatory Guide 1.48, " Design Limits and Loading Combinacions for Seismic Category I Fluid System Components." 14. Regulatory Guide 1.70, " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," Revision 2. 15. Standard Review Plen 3.9.3, "ASME Code Class 1, 2, and 3 Corponents, Component ";pports, or<i Cora Support Structures." 15. ANSI B 16.5, " Steel Pipe Flanges and Flanged Fittings," Arerican National Standa d Institute. 17. ANSI B 16.9, " Wrought Steel Butt Welding Fittings," Arerican National Standard Institute. 18. ANSI B 16.11, " Steel Fittings Steel Welding and Threaded," American National ~tandard Institute. 19. ANSI B 16.25, " Butt Welding Ends - Pipe, Valves, Flanges, and Fittings," American National Standard Institute. 20. ANSI B 31.1, " Power Piping," American National Standard Institute. 21. MSS-SP-25, "Maiking for Valves, Fittings, Flanges, and Unwns," Manufacturers Standardization Society. 22. MSS-SP-66, " Pressure-Temperature Ratings for Steel Butt Welding End Valves," Manufacturers Standardization Society. 3.9.4-6 146 316}}