ML20008F173
| ML20008F173 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 02/27/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20008F164 | List: |
| References | |
| NUDOCS 8103120444 | |
| Download: ML20008F173 (44) | |
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ENCLOSURE 4 COMANCHE PEAK DRAFT SER 3.6.2 Determination of Break Locations and Dynamic Effects Associated wiTh'ThTfostiilated Rupture oFP'iping
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The review peformed under this section pertains to the applicant's program for protecting safety-related components and structures apinst the effects of postulated pipe breaks both inside and out-side containment.
The effect tha: breaks or cracks in high and i
moderate energy fluid systems would have 6n adjacent safety-related components or structures are required to be analyzed with respect j
to jet impingement, pipe whip, and environmental effects.
Several maans are normally used to assure the protection of these safety-related itens.
They include physical separation, enclosure within suitably designed structures, the use o'f pipe whip restraints, and the use of equipment shields.
Our review under Standard Review Plan Section 3.6.2, " Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping", was concerned with the locations chosen by the rplicant for postulating piping failures.
Ue also reviewed for the size and orientation of these postulated failures and how the applicant calculated the resultant pipe whip and jet impingement loads which might affect nearby safety-related components.
I.
REACTOR COOLANT SYSTEM (RCS) MAIN LOOP PIPING The applicant referenced Topical. Reports WCAP-8082-P-A and WCAP-8172-A,
" Pipe Breaks for the LOCA Analysis of tiie Westinghouse Primary Coolant Loop", as the basis for determining the postulated design basis pipe 8108120
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breaks locations in the RCS.
It is the staff's position that these reparts provide an acceptable basis for Westingnouse reactors and that the implementation of the criteria therein provides a level of protection equivalent to that resulting from the appli-cation of the criteria of Regulatory Guide 1.46.
However, the staff's position is that each application that references these Topical Reports must also include additiodal information to ensure that the plant under review is within the limits of WCAP-8082-P-A.
The additional information required in this regard that is not covered in the FSAR at this time is:
- 1) The completion of Table 3.68.3 showing stresses, fatigue cumulative u::;e factors and a comparison of the reference analysis seismic moments with the CPSES seismic moments.
- 2) A !.isting of the component displace.nents at support interfaces and at each design basis break location. This should include the effects of any physical structural restraints which demonstrate that the consequent pipe opening areas and discharge under such postulated pipe break events are within the limits specified in
!! CAP-8082 asfustification of the design basis for pressurization levels adopted in the design of enclosing compartments.
- 3) Sketches showing the locations of the structural barriers, restraint locations, and constrained directions.
Our evaluation has shown that the applicant's RCS dynamic analysis criteria, a
dynamic analysis model, and dynamic analysis models 'Jr jet thrust forces are consistant with Standard Review Plan Section 3.6.2 with the exception
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, that in the determination of the jet thrust at the point of rupture, 1
the fluid pressure and temperature conditions on which the analysis was based were that of.the 100 percent power steady state condition.
The SRP stipulates that the limiting upset condition should be used.
j II. SYSTEMS OTHER THAN RCS MAIN LOOP The criteria for defining break and crack locatic 4 and configurations, the analytical methods used to define the forcing functions, and the dynamic analysis methods to verify the ' integrity and operability of mechanical components, component supports and piping systems are adequate and in compliance with Section,3.6.2 of the Standard Review
- Plan.
There are two points, however, where additional information is required, or there appears to be a misinterpretation of the requirements.
- 1) The FSAR should be clarified to show that the requirements of 0.8 (S ^ $ ).or 0.8 (1.2Sh+S) are based on the sum of Equations h
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(9) and (10) of paragraph NC-3652 of the ASME B&PY Code,Section III and not Equation (9) and (10) individually.
2)
The section on moderate energy piping, in general, gives the areas where cracks are not postulated. The stress criteria justifying the exclusion of these areas and the : tress criteria for the location of postulated cracks are not given. This criteria should be provided in the FSAR.
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4.
Although the applicant's criteria and methods of dynamic analysis are acceptable with the exceptions noted above, the results of the analysis to demonstrate that essential systems, components, and i
supports will not be impared as a result of high energy pipe breaks A
are not discussed in the FSAR. The calculated stresses, type and locations of postulated breaks, and locations of pipe whip restraints f
are shown for some of the high energy lin'es but not for all of the lines of concern.
Information should be provided to demonstrate i
that pipe whip restraints meet accepted design criteria.
Subject to resolution of the above open issues, our findings are as follows.
The applicant has prepcsed criteria for determining the location, type and effects of postulated pipe breaks in high energy piping systems and postulated pipe cracks in moderate energy piping systems.
The applicant has used the effects resulting from these postulated pipe failures io evaluate the design of systeas, compcacnts, and structures necessary to safely shut the plant down and to mitigate the effects of these postulated piping failures. The applicant has stated that pipe whip restraints, jet impingement barriers, and other such devices will be used to mitigate the effects of these postulated piping failures. We have reviewed these criteria and have concluded d
that they provide for a spectrum of postulated pipe breaks and pipe cracks which includes the most likely locations for piping failures, and that the types of breaks and their effects ~are conservatively
.. 1 assumed.
He find that the methods used to design the pipe s. hip restraints provide adequate assurance that they will functica properly in the event of a postalated pfping failure. We further 1
conclude that the use of the applicant's proposed pipe failure criteria in designing the systems, components, and structures necessary to safely shut the plant down and to mitigate the i
consequences of these postulated piping fnilures provides reasonable assurance of their ability to perform their safety function following a failure in high or moderate energy piping systems. The applicant's criteria comply with Standard Review Plan Section 3.6.2 and satisfy the app-]icable portions of General Design Criterion 4.
3.7.3 Seismic Subsystem Analysis The review performed under Standard Review Plan Section 3.7.3 includes the applicant's dynamic analysis of all seismic Category I piping
- systems, in addition to operating transient loads, this analysis also considers abnormal loadings such as an earthquake. This FSAR section is divided into two subsectlcas. Subsection 3.7N.3 is concerned
. primarily with the NSSS and includes all the Westinghouse supplied equipment. Subsection 3.78.3.is concerned with B0P Systems.
At this time, the information in the FSAR is not adequate to verify that-all the requirements of Section 3.7.3 of the SRP for iefsmic Subsy, stem Analysis have been met. Section 3.7N.3, 3.78.3 and the respect 4e subsections which appear in the following paragraphs refer to the corresponding sections in the FSAR and are considered as open issues.
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The sections not discussed have been found to have sufficient information and are sufficiently complete to meet the requirements of applicable sections of the SRP.
3.7h.3. Seismic Subsystem Analysis NSSS Section 3.7N.3 tends to give a general ditucssion of seismic analysis i
methods but does not state what was actufily done. More !ctailed information is required to allow the staff to make a determination as to the adequacy of the seismic subshstem analysis performed on NSSS piping systems.
3.7N.3.1 Seismic Analysis Method The discussion on seismic analysis methods in this section does not describe the methods actually used for the subsystem analysis.
The follcuing information is required as it pertains to the subsystem analysis before our review can be completed.
(_1) The method used (time history response spectra, or equivalent static load).
(2) How the torsional, rocking and translation responses of the components and their supports were considered.
l (3)
The cethod for determining that an adequate number of degrees l
l of freedom were used in the dynamic codeling to determine the I
response of all Category I and applicable Non-Category I structures and plant equipment.
-(4) Justificction that a sufficient number of modes were con-sidered to assure participation of all significant modes l
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(5) The nethods used to handle the relative dis;'n...~ m of Category I supports.
(6)
Hcw significant effects such as piping ir.teractions, externally applied structural restraints, hydrodynamic loads and non-linear responses are accounted for.
(7)
If the equivalent static load method was used, justification must be provided that the system can be represented by a simple model and that the relative motion between surr:::t points is accounted for.
3.7N.3.3 frocedures Used for Analytical Modeling The criteria and procedure given for th'e modeling of the seismic systems and the criteria for determining whether a ccaponent is analyzed as part of a system or independently requires amplificatice arid inclusion of all infocation required by the SRP. Before our review can be completed on this section, the criteria and procedures actually used must be described. This should include the modeling procedures used and the criterit for decoupling as outlined in SRP Section 2.7.2, paragraph 111.3.
3.7N.3.4 Basis for Selection of Frequencies A discussion of the methods actually used in determining he fundamental frequencies is required in this FSAR Section. Also explain how the three ranges of equipment / support behavior (rigid, ficaible, resonant) delineated are handled in the analysis. A statement or statements is required as to how these matters are considered in the CPSES analysis.
j
-B-3.7N.3.5 Use of Equivalent Stat _ic Load Method of Analyses l
For the subsystems analyzed by the Equivalent Static Load "ethod i
justification must be provided that the system can be represented i
by a simple model and that the method produces conservative results.
In addition, the relative motion between all points of support should i
be considered. The criteria used in justifying the use of the i
Equivalent Static load Method is required'before the evaluation of this section can be completed.
3.7N.3.6 Three Components of Earthquake Motion The approach for combining the three components of earthquake motion is satisfactory when the respon,se spect'ra method of seismic analysis is used.
Discuss the approach utilized for" combining these components when the time history method of analysis is used.
3.7N.3.7 Conbination of Modal Responses The method for combining modal responses is essentially the same as that outlined in SRP.3.7.2 Section II-7 an>f the grou-ing method of 20.alatory j
Guide 1.92 Section 1.2.1 with the exception that a coupling factor iias been included when considering the response of closely spaced modes.
Justification'is required for including this factor. Regulatory Guide 1.92 permits this factor to be used in.the double sum method but not with the grouping method.
3.7N.3.8 Analytical Procedures for Piping The comments made on Section 3.7N.3.1 Seismic Analysis Methods are also applicable to this.section.
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_9 3.7N.3.9 Multiple Supported Equipment & Components With Distinct Inputs The criteria to be used in the analysis of multiple supported equip-ment and components meet the staff requirements as outlined :n NRC Standard Review ? tan 3.7.3 Section II-9 with the exception that a commitment be made to combine the support displacements in the most unfavorable combinations.
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3.7 N.3.14 Seismic Analysis of Reactor Internals The applicant references three staff ahproved topical reports concerning the seismic anlaysis. This is acceptable for typical fuel assemblies.
However, there is no discussion as to,the seismic analysis of the reactor internal structure, control ro,d assemblies, and very little discussion of the Control Rod Drive Mechanism.
Before our review can be completed, evidence is required from the applicant that the analysis complies with the requirements of Section 11.1 and II.6 of SRP 3.7.2 concerning seismic analysis methods and the three components of earthquake motion respectively.
In addition, typical diagrams of mathematical dynamic modeling of reactor internal structure and damping values and the justification for must he included.
3.78.3.1 Seismic Analysis Methods The applicant's methods of seismic analysis or test methods in lieu of i
analysis have been reviewed per the requirements of Section 3.7.3.1 of the SRP and fouad to be sufficiently complete and adequate with the following exceptions.
(1) Assurance is required that an adequate number of masses or degrees
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. of freedom are used in the dynamic modeling to determine the response of the systems.
(2) Assurance is required that a sufficient number of modes have been investigated to assure participation of all significant modes.
(3) Assurance that all significant effects such as piping inter-actions, externally applied structural restraints, hydro-dynamic loads, and non-linear responses nave been considerec.
1 3.78.3.2 Determination of Number of Earthquake Cycles The number of maximum stress cycles considered for the OBE as stated in the FSAR is adequate. There is not, however, any n.ention of the SSE.
3.7B.3.3 procedures Used for Analytical Modeling Comments made earlier in Section 3.79.3.1 are also applicable to this section.
3.7B.3.5 Use of Ecui alent Static load Method of Analysis i
The criteria for the use of the Equivalent Static Load Method of Analysis is accept '-le... it is applied to single degree of freedo:.1 systems and systems with s' natural frequency equal to or greater than 33HZ.
If systems other than the aforementioned are analyzed by this method, the criteric used must be so stated.
3.73.3.7 Combination'of Modal Responses Oescribe the technique used for handling modal responses of closely spaced modea including the method or. methods (grouping method, ten percent method or double sum method) along with the applicable equations that are used in complying with Regulatory Guide 1.92.
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4 3.7B.3.8 AnalyticalProceduresofPipingJystems There are some areas that appear to be in error or have been oisin-torped by the reviewer. This section does not meet all the r ouire-ments of Section 3.7.3.8 of the SRP. Additional informatica or clarification is required for the following.
(1) The statement is made that "for piping that does not leave a building, no seismic relative motion is accounted for at supports because the relative displacement of supports points is very insignificant within the 'same building". Justification for this statement is required,.especially for piping systems that are routed over several different floor elevations.
(2)
The justification for exc,luding c'xternal and DBA loading
'I conditions from the analysis.
(3)
It is stated that "A simplification in the anlaytical effort was achieved for a given model and spectra by limiting all pipe cress sections, regardless of weights to the same e
maximun deflection".
Provide justification for this technique.
(4) For a of'..n deflection the stresses in a piping systen will vary considerably due to pipe diameter, distance between supports, weight distribution, etc. How is this accounted for?
(5) Justification is required for the assumption that all floors in a single structure are assumed to move in phase.
(6)
The statement is made "No separate evaluation was made for the stress requirement of the Faulted Condition because it is covered by the stress requirement of the upset condition if only pressure gravity and earthquake loadings are considered."
The stress due to LOCA should also be included in the faulted conditio
1-(7) The acceptance criteria of SRP 3.7.2 Section !!.1 as it applies to piping analysis is not covered adequately; 1
especially in the areas of degrees of freedom used, number of modes investigated to insure participation of all signi-4 ficant modes, consideration of relative displacement among supports, and inclusion of other significant effects.
l 3.78.3.9 Multiple 5'4pported Equipment & Components with Distinct Inputs Before the evaluation of this section csn be completed, additional information is requireo to insure that the maximum support relative displacement are considered.
3.78.3.13 Interaction of other Piping with Seismic Category I Piping This section concerning the interaction of other piping with seismic l
Category I piping adequately defines how these piping systems are I
handled when they are a part of the same system. However, informat'1n is required as to how Non-Category I piping systems are anlyzed and/or isolated fren Category I piping when the systcas are in close proximity not so that a failure of the Non-Cate2ory I piping would damage the Catagory 3
I piping.
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l Upon resolution of the above open issues, we will report our findings in l
a supplement to the Safety Evaluation Report.
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o 13-3.9 MECHANICAL SYST_EfS AND COMPONENTS ine review serformed under Standard Review Plan Sections 3.9.1 through 3.9.6 pertains to the structural integrity and operability of various safety-< elated mechanical components in the plant. Our review is not limited to AS"E Code components and supports, but is extended to other components such as contral rod drive mechanisms, certain reactor i
internals, ventilation ducting, cable trays, and any safety-related piping designed to industry standards other than the ASME Code.
We review such issues.as load combinations,k allowable stresses, nothods of anlaysis, summary results, seismic qualification, preoperr.tional testing, ant. inservice testing of pumps, and valves. Our review must arrive at the conclusion that there is a'dequate assurance of a mechanical ccmponent performing its safety-rel'ated function under all postulated combinations of normal operating conditions, system operating transient, postulated pipe breaks, and seismic events.
3.9.1 SPECIAL TOPICS FOR MECHANICAL COMPONENTS The review r< rformed under Standard Review Plan Secitn 3.9.1 pertains to the design transients, computer progams, experimental stress analysis and elastia-plastic analysis methods that were used in the -
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analysis of sei: dc Category 1 ASME Code and non-Code items.
We have identified the following open issues in our review.
The issues are identified by sections.of the FSAR. Section 3.9N pertains to the reactor coolant system, its component and supports. Section 3.98 pertains to all other safety-related systems.
f FSAR SECTION 3.9N.1 NSSS COMPONENTS The applicant has provided a complete. list of transients to be used in
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-16 the design and fatigue analysis of all Code Class 1 and CS components and of components supports and reactor internals within the reactor coelant pressure boundary. The number of events postulated for each transient has been included and is acceptable.
The applicant has listed the computer program used i,n the dynamic and static anlaysis to determine the structural and functional inegrity of seismic Category 1 Code and non-Code items. Verification of one of these programs, WECAN, is described i,n topical report WCAP-8929, I
" Benchmark Problem Solutions Employed for Verification of WECAN Compuper P rogram". He are currently reviewing this topical report and will report on the results of our evaluation in a su'pplement to this SER. There is no verification or reference included in the FSAR for the WESAN program.
It is stated in 3.9N.l.4.2 of the FSAR (under transients) that the vertical there31 growth of the reactor pressure vessel nozzle center-lines is considered in the thermal analysis. The treatment of the horizontal movrant should be provided.
i There appears to be an error in the equation for d T (t) on Page 21 3.N-36.
The equation should be:
T (t)
/hT (t) =
T(0,t) - T A (t) 1 21 2
4 The stress criteria for Class 1 components ana component supports as a
stated in Section 3.9N.i.4.7 are not in agreement with the criteria of Appendix F of the Code. Justification should be presented to demonstrate that the criteria used for design is as conservative as that of Appendix F.
4
s Several of the items posed in the staff quesiton 112.25 concerning asymmetric loading have not been addressed as yet or are incomplete.
The applicant has stated they will be supplied later in its response to this question.
FSAR SECTION 3.9B.1 B0P COMPONENTS 1
The applicable transients, component operating conditions and number of events for any ASME Class 1 systems covered by this section should be listed and/or referenced. The table hn Section 3.98.1.4 should be completed to include the combined shear and tension allowables of the anchor bolt m3terial.
Subject to resolution of these open issues, our findint.s are as follows.
The methods of caslysis that the applicant has employed in the design of all seismic Category I ASME Code Class 1, 2, and 3 components, component supports, reactor internals, and other non-Code items are in conformance with Standard Review Plan Section 3.9.1, "Special Inpics for Mechanical Components", and satisfy the applicable portions of General Design Criteria 2, 4,14, and 15. The use of these criteria in defining the applicable transients, computer codes used in analyses, and analytical methods provides assurance that the stresses, strains, and displacements calculated for the above noted items are as accurate as the current state-of-the-art permits and are adequate for the design of these items.
3.9.2 Dynamic Testing and Analysis'of Systems, Components, and Equipment The review performed under Standard Review Plan Section 3.9.2 pertains
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s r to the criteria, testing procedures, and dynamic analyses employed by the applicant to assure the structural integrity and operability of piping systems, mechanical equipment, reactor internals and their
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supports under vibratory loadings. This review is divided into three parts, each of which is discussed briefly below.
o' 3.9.2.1 Piping Preoperational and Startup Testing _ Program Piping vibration, thermal expansion, and dynamic effects testing will l
be conducted during the Comanche Peak p ant's preoperational and i
startuR testing program. The purpose of these tests is to confirm that the piping, components, restraints, and supports have been designed to withstand the dynamic loadings and op'erational transient conditions that will be encountered during service as required by the ASME Section III Code and to confirm that no unacceptable restraint of normal thermal motion occurs. We have identified the following open issues in our review.
The issues are identified by sections of the FSAR.
FSAR SECTI'"! 3.9t!.7..l HSSS C0;tP0i!E!!TS Many of the items required by the Stand 1rd Review Plan (SRP) Section 3.9.2 are covered only briefly or not at all in this FSAR section.
4 The staff question 112.8 was not answered as it. pertains to the NSSS System.
In addition to answering question 112.8, the requirements of the SRP Acceptance Requirements II.l.a through f and items a through d of-the Review Procedures should be addressed before this FSAR Section can be consideret to be acceptable.
FSAR SECTION 3.9B.2.1 BOP RELATED ITEMS The staff require.; a commitment to test all high energy piping and all' 9
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. seismic Cate;;ry I =cderate ener;y piping. inclucin; s;;;;rts an:
restraints f:r tiermal expansien, steady state and ejr.5,i: response (Cuestien 112.12). The FSAR if st Of the systens to be included in these tests d:es not include all of these syste7s. There is nct sufficient infor atien available to ade;uately revien tte test ;r: gram.
Items 1.a thr: ugh f of the SRP Acceptance Criteria sh:ald te c:vered in detail.
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Subject to resciutien of these c;en issues, cur findf n;s.f11 te as f 11cos.
The vibratica, thermal expansica, and dynamic effe:ts test pr:gra: hicn will be conducted during startup and initial ;eraticn :n s;e:ified high and moderate energy piping, and all asscciated systers, restraints and supports is-an acceptable program.
The tests ;r: vide ade;uate assuran:e that the piping and piping restraints of the system eave teen designed to withstand vibrational dyna:fc effects due te valve closures, purp trips, and other Operating nodes asscciated with the design tasis ficw conditiens.
In additien, the tests provide assurance that ade;uate clearances and free cove:ent of snubbers exist for unrestrained thenaal cove ent Of piping and supports daring normal system heatup and coltewn Operaticns. The p12nned tests will develop leads similar to these ex;erienced during react:r ;er-ation.
This test pregram cc: plies with Standard Review Plan Sectica 3.9.2
-and constitutes an acceptable ~ basis fer fulfilling, in part, the re;uire-
=ents of General Design Criteria la and 15.
a 3.9.2.2 Presperational Flew-Induced Vibration Testing of React:r Internals Flow-induced vibration testing of reactor internals sh uld be cenducted
~ during the preeperational a.nd startup test pregram. The purpose of this 9
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.. test is to d2.nonstrate that floa-induced vibrations similar to tnose expected during operation will not cause unanticipated fic,.-incace:
vibrations of significant magnitude or structural damage.
Indian Point No. 2 is designated as the prototype of the CPSES reactor internals and the design differences are noted in the FSAR.
- dditionally, i
prototype data has been obtained frca the Trojan Plant anc will te obtained fro,1 the Sequoyah No. 1 Plant. Tr.e applicant states in the FSAR that the guidelines in Regulatory Guide 1.20, " Comprehensive '.'itraticn Assesscent Program for Reactor Internals During Precperaticnal and *nf tial Startup Testing" will be met by ccnducting the preoperational testing exanination for integrity per Sectica 1 of Regulatory Guide 1.20,
" Regulations for Reactor Internals Simil'ar to Prototype Oesign".
'Je find this program to be acceptable provided that the applicant sutaits a corrilation of the Coranche Peak vibration predictions with the dynamic analyses or test results frc the prototype raactors and, if the results from the Trojan and Sequoyah tests indicate the need for any corrective action, the staff will review the applicant's evaluation of the need ror cinilar corrective action en CPSES.
Subject to resolution of these open issues, cur findings are as follc<s.
The preoperational vibration program planned for the reactor internals provides an acceptable basis for verifying the design adequacy of these internals under test loading conditions ccaparable to those that will be experienced during operation. The combination of tests, predicti.e analy is,
.. and p9st-test inspection provide adequate assurance that the reactor internals will, during their service lifetime, withstand the flow-induced vibrations of reactor operation without loss of structural integri ty.
The integrity of the reactor internals in service is essential to assure the proper positioning of reactor fuel assemblies and unimpaired operation of the control rod assemblies to permit, safe reactor operation and shutdown. The conduct of the preoperational vibration tests is in conformance with the provisions of Regulatory Guide 1.20 and Standard Review Plan, Section 3.9.2 and satisfies the applicable I
requirements of General Design Criteria 1 and 4.
3.9.2.3 Dynamic Analysis of Reactor Coolant Systhm Dynamic system analyses should be performed to confirm the structural design adequacy and ability, with no loss of function, of the reactor internals and unbroken loops of the reactor coolant piping to withstand the loads from a loss-of-coolant accident (LOCA) in combination with the SSE.
Our review covers the methods of analysis, the considerations in defining the nathenatical models, and descriptions of the forcing function, the calculational scheme, the acceptance criteria, and the interpretation of analytical results.
The applicant has stated that stresses due to the SSE were combined in the most unfavorable manner with the blowdown stresses and the results indicate that the maximum deflections and stresses in the critical structures are below the established allowable limits. The blowdown analyses require further amplification and clarification. 'Specifically, the Staff will require that the applicant:
(1) justify decoupling of the horizontal and vertical
components of the blowdown analyses, (2) justify the use of resuits of linear analyses for the inherent nonlinear problem, (3) justify the gap and damping values used in the mathematical models and (4) present a discussion outlining the effects of system flow upon the mass and flexibility properties.
In general, this Section does not include in sufficient detail the infor-mation required in the SRP Acceptance Requirement Section 5 to allow for meaningful review.
Also the computer prpgram FORCE is referenced with no explanation or verification of the program.
The.re is no verification thet stresses under the combined loads are within allowable limits of the applicable code and that the deformationi are within the limits set to assure the ability of the reactor internal structures and piping to perform needed safety functions.
Subject to resolution of the above open issues, our findings are as follows.
The dynamic system analysis performed by the applicant provides an accepta basis for confirming the structural design adecuacy of the reactar internals and unbroken piping loops to withstand the combined dynamic loads of a 1
- postulated loss of coolant accident (LOCA) and the safe shutdown earthquake (SSE).
The analysis provides adequate assurance that the combined stresses and strains in the components of the reactor coolant system and reactor internals do not exceed the allowable stress and strain limits for the materials of construction, and that the resulting deflections or diplace-ments at any structural elements of the reactor internals will not distort the reactor-internals geometry to the extent that core cooling may be
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impai red.
The methods used for component analysis have been found to be compatible with those used for the systems ' analysis.
The proposed com-binations of component and system analyses are, therefore, acceptable.
The assurance of structural integrity of the reactor internals under combined LOCA and SSE conditions provides added confidence that the design will withstand a spectrum of lesser pipe breaks and seismic events.
Accomplishment of the dynamic system analysis constitutes an acceptable basis for complying with Standard Review Plan Section 3.9.2 and for satisfying the applicable requirements of General Design Criteria 2 and 4.
3.9.3 ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures Our review under Standard Review Plan Section 3.9.3 is concerned with th structural integrity and operability of pressure-retaining components, their suppor*s, and core support structures which are designed in accor-dance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, or earlier industry standards.
this review is divided into four parts, each of which is discussed briefly bolcw.
3.9.3.1 Loading Combinations, Design Transients and Stress. Limits For Sections 3.9N.3 and 3.98.3 of the FSAR to be acceptable, the following j
. issues need to be resolved:
'l.
The discussion of plant conditions in Section 3.98.3.1 i of the FSAR requires clarification. The applicant should clearly define the normal a
upset, emergency and faulted plant conditions which apply to this section and to Table 3.98-1 in the FSAR.
2..
The methods of combining responses to the various loads listed in the
j loading combinations in FSAR Tables 3.9N-2, 3.9N-4 and 3.99-1 are not 1
defined in the FSAR.
The responses to Questions 112.7 and 112.17 do not clearly delineate the applicant's position and further clarification is needed.
We will require a description of the methods used for the combina-1 tions of responses to all dynamic loads for all NSSS and 30P supplied ASME Class 1, 2, and 3 equip, ment, components and their supports.
Our position on this issue is outlined in ftUREG-0484, " Methodology for Combining Dynamic Responses", Revision 1 dated May,1980.
Verify that the position taken in response to Qubstions 112.7 and 112.17 regarding loading combinations is consistent with the requirement of NUREG-0484.
3.
The design criteria specified for the internal parts of NSSS and BOP components should be provided.
4.
We have not received a complete reply to our Question 112.25 pertaining to the response of certain reactor coolant system components and their supports to postulated asymmetric LOCA loads.
We have contracted with the Energy Technology Engineering Center to perform an indepen419t mlysis of a sample piping system in the CPSES plant.
This anlaysis will not only verify that the sample piping systea meets the applicable ASME Code requirements, but will also provide a check on the applicant's ability to correctly model and analyze its piping systems.
The results of the above evaluations will be presented in a future supplement to this report.
Subject to resolution of the above open issues, our findings are as follows.
The specified design and service combinations of loadings as applied to
4 ASME Code Class 1, 2, and 3 pressure retaining components in systems designed to meet seismic Category I standards are such as to provide assurance that, in the event of an earthquake affecting the site or other service loadings due to postulated events or system operating transients, the resulting combined stresses icposed on system components will not i
-exceed allowable stress and strain limits for the materials of construction.
Limiting the stresses under such loading tombinations provides a conservative basis for the design of system components to withstand the most adverse combination of loading events without f.oss of structural integrity.
The design and service load combinations and associated stress and deformation limits specified for ASME Code Class (, 2, and 3 components comply witn Standard Review Plan Section 3.9.3 and' satisfy the applicable portions of General Design Criter 'a 1, 2, and 4.
3.9.3.2 Pump and "alve Operability Assurance Program The information presented in Sections 3.9N.3 and 3.93.3 of the FSAR requires more detail on acceptance criteria used to assure pump and valve operability.
The applicaS1 acceptance criteria discussed in Standard Review Plan, Section 7
3.9.3, Paragraph II.2 should be addressed in the FSAR for both USSS and UCP supplied active pumps and valves.
Subject to resolution of the above open issues, our findings are as follows.
The component operability assurance program for ASME-Code Class 1, 2, and 3 active valves and pumps provides-adequate assuran;& of the capability of such active components (a) to withstand the f;; ese d($lgn and service
.. =
loads without loss of structural integrity, and (b) to perform necessary
" active" functions (e.g., valve closure or opening, pump operation) during postulated events and conditions expected when plant shutdown is required.
The specified component operability assurance test program complies with Standard Review Plan Section 3.9.3 and satisfies the applicable portic. t of General Design Criteria 1, 2, and 4.
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3.9.3.3 Design and Installation of Pressure Relief Devices The design and installation criteria aphlicable to the mounting of pressure relief devices (safety and relief valves) for the overpressure protection of ASME Class 1, 2, and 3 components are reviewed. To be acceptable, the following issues need to be resolved:
l.
Section 3.98.3.3 of the FSAR should include an adequate description of the calculation procedures, computer progams, etc. which were used in the parametric studies for closed discharge systems.
2.
Information should be provided in Section 3.9B.3.3 of the FSAR relating to the various design and service loading conditions and combinatiens thereof, and the corresponding stress criteria used in the design for the mounting of pressure relief valves.
3.
lne method of evaluating the structural response of the piping and support system stiffness in the dynamic analysis of these mountings should be discussed in Section 3.9B.3.3 of the FSAR.
3.9.3.4 Component Supports The review of information submitted by the applicant includes an evaluation of Code Class 1, 2, and 3 component supports. The review includes an assessment of the design and structural integrity of the supports and their
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.. effect on the operability of active components. The review addresses three types of supports:
plate and shell, linear, and component standard types, and their function. For Sections 3.9N.3.4 and 3.98.3.4 of the FSAR to be acceptable, the following issues need to be resolved:
1.
In the response to the staff's quest:sn 112.20, the applicant st,ated that NSSS supplied Class 2 and 3 standard compenent supports which were procured prior to July if 1974 were designed in accordance with the criteria in Section 3.9N.3.4 of the FSAR. Component supports procured after July 1,1974 will homply with the requirements of ASME Section III, Subsection NF, " Component Supports". Since the criteria in the FSAR may not be as conservative as that in Subsection NF, we require more information f^or those supports procured prior to July 1,1974. A discussion which demonstrates that those components designed to the FSAR criteria have an adequate margin of safety should be submitted in the FSAR or as a revised reponse to Question 112.20.
'In addition, the applicant should verify that the allowable stresses of MSS-SP-58, " Pipe Hangers and Supports" are used without the addition of a shape factor to account for bending stresses.
2.
The missing information on NSSS component supports leads and stresses in FSAR Table 3.9N-14 through 3.9N-19 should be submitted.
3.
Information similar to chat requested in Item 2 above should be
. submitted in Section 3.9B.3.4 for the supports in the Auxiliary Feedwater System and all Emergency Core Cooling Systems.
4.
We will require an acceptable response to our request for preservice-inspection and testing information on snubbers which is outlined in a letter from Robert L. 'Tedesco to R. J. Gary dated January 14, 1981.
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.. Subject to resolution of the above open issues, our findings are as follows.
The specified design and service loading combinations used for the design of ASME Code Class 1, 2, and 3 component supports in systems classified as seismic Category I provide assurance that, fr. the event of an carthquake or other service loadings due to postulated events or system operating I
transients, me resulting combined stressos imposed on system components will not exceed allowable stress and strain limits for the materials of construction.
Limiting the stresses unyer such loading combinations pro-vides a conservative basis for the design of support ccmponents to with-stand the nost adverse combination of, loading events without loss of structural integrity or supported component operability. The design and service load combinations and associated stress and deformation limits specified for ASME Code Class 1, 2, and 3 component supports comply with Standard Review Plan Section 3.9.3 and satisfy the applicable portions of General Design Criteria 1, 2, and 4.
3.9.4-Control Red Drive Systems Our review under Standard Review Plan Section 3.9.4 covered the design of the. hydraulic control rod drive system up to its interface with the control rods.
We reviewed the analyses and tests performed to assure the structural integrity and operability of this system during normal operation and under accident conditions. We also reviewed the life-cycle testing performed to demonstrate the reliability of the control rod drive system over its 40 year life.
The review indicates that additional information is required as to the
design criteria for the non-pressurized components. The thermal deflecton problem of dissimilar materials is covered but there is no information as to the allowable and actual deflections due to the various loading conditions.
Design margins for stress, deformation, and fatigue should be presented and should be shown to be equal to or greater than those of other plants of similar design having a period of successful operation.
l Subject to resolution of the above open issues, our. findings are as follows.
i The design criteria and the testing program conducted in verification of the mechanical operability and life cycle capabilities of the control rod drive system are in conformance; with Standard Review Plan Section 3.9.4.
The use of these criteria provide reasonable assurance that the system will function reliably when required, and form an acceptable basis for satisfying the mechanical reliability stipulations of General Design Criterion 27.
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3.9.5 Reactor Proccure Vessel Internals Our review under Standard Review Plan (SRP) Section 3.9.5 is ccacerned with the load combinations, allowable stress limits, and other criteria used in the design of the Comanche Peak reactor internals.
In addition to the issues discussed in Sections 3.9.2.2 of this Safety Evaluation Report, resolution of. the following issues is also required:
1.
There is no commitment to design and construct the core support structure per the requirements of subsection NG of Section III of the G
e -
. ASME Code except for the allowable stresses for the faulted condition.
(Reference Section 3.9N.5.4) 2.
There are no calculated stresses and deformation to evaluate. The table of deflections (Table 3.9N-il) set 1s to be incomplete and is not clear as to the location and direction of the deflections.
Subject to esolution of these issues, our findings are as follows.
The specified transients, design and sehvice loadings, and combinations of loadings as applied to the design of the Comanche Peak reactor internals provide reasonable assurance that in the event of an earthquake or of a system transient during normal plant opbration, the resulting deflections and associated stresses imposed on these reactor internals would not exceed allowable stresses and deformation limits for the materials of construction.
Limiting the stresses and deformations ~under such loading combinations provides an acceptable basis for the design of these reactor internals to withstand the most adverse loading events which have been postulated t7 recur during service lifetime without loss of structural integrity or impairment of function. The design procedures and criteria used by the applicant 'in the design of the Comanche Peak reactor internals comply with Standard Review Plan Section 3.9.5 and constitute an acceptable basis for satisfying the applicable requirements of General Design Criteria 1, 2, 4, and 10.
a 3.9.6 Inservice Testing of Pumps and Valves In Section 3.9.3 of this Safety Evaluation Report, we discussed the design -
and operability lof safety related pumps and valves in the Comanche Peak Steam Electric Station. The design of these pumps and valves is intended i*r-
. to demonstrate that they will be capable of performing their safety function (open, close, start, etc.) at any time during the plant life.
However, to provide added assurance of the reliability of these components, the applicant will periodically test all its safety related pumps and valves.
These tests are performed in general accordance with the rules of Section XI of the ASME Code. These tests verify that these pumps and valves operate -
i successfully when called upon. Additionally, periodic measurements are made of various parameters and compared to baseline measurements in order -
to detect long term degradation of the pump or valve performance. Our review under Standard Review Plan Section 3.9.6 covers the applicant's program for preservice and inservice t.esting of pumps and valves. We give particular attention to those areat of the test program for which the applicant requests relief from the requirements of Section XI of the ASME Code.
The information presented in Section 3.9.6 of the FSAR does not contain sufficient detail to demonstrate how the applicant intends to implement the inservice testing of pumps and valves requirements of ASME Section XI,
" Rules.for Inservice Inspection of Nuclear Pcwer Plant Components".
We will require a description of the applicant's proposed program on this subject. Guidelines on the type of information that we require is contained in Attachment 1 of this SER.
In addition, we will require an acceptable respons to our request for information on inservice inspection pressure isolation valver (Reference letter from Robert L. Tedesco to R. J. Gray, dated 11/17/80).
We will report on the resolution of these. issues in a supplement to this' SER.'
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i CONCLUDING REMARKS j
The above draft may be incomplete as follows. We are continuing to. review 1
your responses to Qll2.24, Item II.D.1 in the NRC Action Plan, seismic and system quality group classifications, and compliance with Section 50.55(a) 1 and the ASME Code Cases.
)
We may request additional. information or commit-ments on these matters.
We will report on the resolution of the issues Identified throughout this i
draft'in the SER or a Supplement to the SER.
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ATTAC4"ENT 1 NRC STAFF CCK"ENTS 0. INSERVICE PU' P AND VALVE TESTIN3 PR03 RAMS AN5 RELIEF REQUESTS The NRC staff, after reviewing a number of pucp and valve testing programs, has determined that further guidance might be helpful to illustrate the type and extent of information we feel is necessary to expedite the I
review of these prcgrams. We feel that the Licepsee can, by incorpcrating these guidelines into each program submittal, reduce considerably the staff's review time and time spent by the Licensee in responding to NRC staff requests for additional information.
The pump testing program should include all safety related* Class 1, 2, and 3 pumps which are installed in water cooled nuclear power plants
~
and which are provided with an emergency power source.
The valve testing program should include all the safety related valves in the following systems excluding valves used for operating convenience only, such as manual vent, drain, instrument, and test valves, and valves used for maintenance only.
PWR a.
High Pressure Injection System b.
Low Pressure Injection System c.
Accumulator Systems d.
Containment Spray System a
- Safety related - necessary to safely shut down the plant, mitigate the consequences of an accident and maintain the plant in a safe shutd' own condition.
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. Primary and Secondary System Safety and Pe'ief Yalves e.
f.
Auxiliary Feedwater Systems 9
Reactor Building Cooling System h.
Active Components in Service Water and Instrument Air Systems ~
which are required to support safety system functions.
- i. Containment Isolation Yalves required to change position to isolate containrent,
- j. Chemical & Volume Control System k.
Other key compcnents in Auxiliary Systems which are required to directly support plant shutdown or safety system function.
1.
Residual Heat Removal System m.
Reactor Coolant System BWR High Pressure Core Injection System a.
b.
Low Pressure Core Injection System Residual Heat Renoval System (Shutdown Cooling System) c.
d.
Emergency Coenwr System (Isolation Ccadenser Systm) e.
Low Pressure Core Spray System f.
Containment Spray System g.
Safety, Relief, and Safety / Relief Yalves h.
RCIC(ReactorCoreIsolationCooling) System 1.
Containment Cooling System
- j. Containment isolation valves required to change position to isolate centainnent.
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k.
Standby liquid control system (Boron System)~
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1.
Automatic Depressurization System (any pilot or control' valves, associated hydraulic or pneumatic systems, etc.)
Control Rod Drive Hydraulic System (" Scram" function) m.
other-key components in Auxiliary Systems which are required to directly n.
support plant shutdown or safety system function.
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Reactor Coolant System Inservice Puno.and Valve Testing Program I.
Information required for NRC Staff Review of the Pump and Yalve Testing Program A.
Three sets of P&ID's, which include all of the systems listed above, with the code class and system boundaries clearly marked.
The drawings should include all of the components present at the time of submittal and a legend of the P&ID symbols.
B.
Identification of the applicable ASME Code Edition and Addenda C.
The period for which the program is applicable.
D.
Identify the component code class.
E.
For Pump testing:
Identify 1.
Each pu.:p-required to be tested (name and number) 2.
The test parameters to be measured 3.
The test frequency.
. F.
For valve testing:
Identify 1.
Each valve in ASME Section XI Categories A & B that will be exercised every three months during normal plant operation (indicate whether partial or full stroke exercise, j
and for power operated valves. list the limiting value for stroketime.)
,!. Each valve in ASME Section XI Category A that will be leak tested during refueling outages (Indicate the leak test procedure you i.. tend to use) 3.
Each valve in ASME Section XI Categories C, D, and E that will be tested, the type of test and the test frequency.
For check valves, identify those that. will be exercised every 3 months and those that will only be exercised during cold shutdown or refueling outages.
II. Additional Information that will be Helpful in Speeding Up the Review Process A.
Include the valve location coordinates or other appropriate location information which will expedite our locating the valves on the P& ids.
B.
Provide' PalD drawings that are large and clear enough to be a
read easily.
C.
Identify valves tht are provided with an interlock to other components and a brief description of that function.
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5-Relief Requests from Section XI Recuirements The largest area of co.ncern for the NRC staff, in the re.iaw of an inservice valve and pump testing program, is in evaluating the basis for justifying relief from Section XI Requirements.
It has been our experience that many requests for relief, submitted in these programs, do not provide adequate descriptive and detailed technical inf6rmation. This explicit information is necessary to provide reasonable assurance that the burden imposed on the licensee in complying with the code requirements is not justified by the increased level of safety obtained.
Relief requests which are submitted wit'h a justification such as "Impractictl", " Inaccessible", or any other categorical basis, will require I
additional information, as illustrated in the enciesed examples, to allow m
staff to make an evaluation of that relief request. The intention of this guidance is to illustrate the content and extent of information required by the NRC staff, in the request for relief, to make a proper
~
evaluation and adequately document the basis for that relief in our safety evaluation report. The NRC staff feels that by receiving this information in the program submittal, subsequent requests for additional information and delays in completing our review can be' considerably reduced or eliminated.
I.
Information Required for NRC Review of Relief Recuests, A.
Identify component for which relief is requested:
a 1.
Name and number.as given in FSAR 2.
Function 3.
ASME Section III Code Class 4.
For valve testing, also specify the ASME Section XI valve category as defined in IWV-2000 4
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B.
Specifically identify the ASME Code requirement that has been determined to be impractical for. each component.
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C.
Provide information to support the detcraination that the requirement in (B) is impractical; i.e., state and explain the basis for requesting reifef.
i D.
Specify the inservice testing that will'be performed in lieu of the ASME Code Section XI requirements.
l E.
Provide the schedule for implementation of the procedure (s) in (D).
Examples to Illustrate Several Possible dreas Uhere Relief May Se II.
Granted and the Extent and Content of Information Necessary to Make An Evaluation A.
Accessibility: The regulation specif!cally grants relief frca the code requirement because of insufficient access pro-visions. liowever, a detailed discussion of actual physical arrange unc of the cecponent ;n q.s ' ion to illustrate the insufficiency of space for conductit,.1he required test is necessa ry.
Discuss in detail the physical arrangement of the component in question to demonstrate that there is not sufficient space to perform the _ code required inservice testing.
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What alternative surveillance ceans t:hich will provide an
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acceptable level of safety have you considered and why are these means not feasible?
B.
Environnental Conditions (e.g., High radiation level, High i ~
temperature, High humidity, etc.)
Although it is prudent to maintain occupation radiation exposure for inspection personnel as low as pE,acticable, the request for relief-from the code requirements cannot be granted solely on the basis of high radiation levels alone. A balanced judgment between the hardships and cocpensating increase in the level of safety should be carefully established.
If the hcalth and safety of the public dictates the necessity of inservice testing, alternative means or even decontamination of the plant if necessary should be provided or developed.
Provide additional information regarding the radiation levels at the required test location. What alternative testing techniques which will provide an acceptable level of assurance of the integrity of the component in question have you considered and why are these techniques determined to be impractical?
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C.
Instrumentation is not originally provided Provide information to justify that compliance with the code requirements would result in undue burden or hardships without a compensating increase in the level of plant safety. What alternative testing methods which will provide an acceptable level of safety have you considered and vehy are these methods a
determined to be impractical?
D.
Valve Cycling During Plant Operation yould Put the Plant in an Unsafe Condition The licensee should explain in detai.1 why exercising tests during plant operation could jeopardi'ze the plant safety.
E.
Valve Testing at Cold Shutdown or Refueling Intervals in Lieu of the 3 Month Required Interval The licensee should explain in detail why each valve cannot be exercised during normal operation. Also, for the valves where a refueling interval is indicated, explain in detail why each valve cannot be exercised during cold shutdown intervals.
III. - Accentance Criteria for Relief Request The Licensee must sucessfully demonstrate that:
1.
Compliance with the code requirements would result in hardships or unusual difficultie$ without a compensating increase in the level of safety and noncompliance will provide an acceptable level of quality and safety, or
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2.
Proposed alternatives to the code requirements or portions thereof will provide an acceptable level of qual.'cy and safety.
Standard Format A standard format, for the valve portion of the pump and valve testing I
program and relief requests, is included as an apt'achment to this Guidance.
The NRC staff believes that'this standard fomat will reduce the time spent
^
by both the staff in our review and by.the licensee in their preparation of the pump and valve gesting program and submittals.
The standard format includes examples of relief requests which are intended to illustrate the applicatien of the standard format and are not necessarily a specific
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plant relief. request.
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0-14 X'
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0-15 X
6 DE flA C
DT 717 3
C-15 X
16 CK SA CV X
CS 702C 3
C-15 X
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3 REL SA CV 834 3
D-ll X
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Q X
l 7228 3
B-11 X
9-11 X
3/4 REL-SA SRV 715 2
A-10 X
B-10 X
0-14 X
10 GA M0 C
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MT 30 sec.
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Legend for Valve Testing Example Format Q - Exercise valve (full stroke) for ope ~rability every (3) months LT - Valvec : t leak tested per Section XI Article IWV-3420 MT - Stroke time.neasurements are taken and compared to the stroke time limiting value per Section XI Article IWV 3410 CV - Exercise check valves to the position required to fulfill their function every (3) months SRV - Safety and relief valves are tested per Section XI Article IWV-3510 DT - Test category 0 valves per Section XI Article inn-3600 ET - Verify and record valve position before operations are perfor ed and after operations are completed, and verify that valve is -locked or sealed.
CS - Exercise valve for operability every cold shutdown RR - Exercise valve for operability every reactor refueling e
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Relief Reauest 3 asis System: Auxiliary Coolant System, Component Cooling 1.
Valve:
71 7 Category:
C Class:
3 i
Function:
Prevent backflew from,the reactor coolant pump cooling coils
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Impractical test requirement: Exercise valve for operability every three conths Basis for relief: To test thir valve would require interruption of cooling water to the reactor coolant pumps notor cooling cails. This action could result in damage to the reactor coolant pumps and thus place the plant in an unsafe mode of operation.
Alternative This valve will be exercised for operability Testing:
during cold shutdowns.
2.
Valve:
834
. Category:
3-E Class:
3 Function:
Isolate the primary water from the component cooling surge tank during plant opertion.
It is normally in the closed position, but routine operation of this valve will occur during refueling a
and cold shutdowns.
Impractical Test Exercise valve (full streke) for operability Requirement:
every three (3) months.
2-Basis for Relief:
This valve is not required to change position
'during plant operation to accomplish its safety function.
Exercising this valve will increase the possibility of surge tank line contamination.
Alternate Verify and record valve position before and Testing:
after each valve operation.
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3.
Val ve:
744B l
Category:
A Class:
2 Function:
Isolate the residu'al heat exchangers from the ccid leg R.C.S. backflow and accumiator backflow.
Test Requirements:
Seat leakage test Basis for This valve is located in a high radiation field Relief:
(2000 mr/hr) which would make the required seat leakage test hazardous to test personnel. We intend to teat leak trst tuo other valves (8753 and 876B) i;hich are in series with this valve and will also prevent backflow. We S al th.
by complying with the seat leakage requirements we will not achieve a compensatory increase in the level of safety.
Alternative No alternative seat leak testing is proposed.
Testing:
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