ML19347E939
| ML19347E939 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Callaway  |
| Issue date: | 04/22/1981 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Byran J, Koester G KANSAS GAS & ELECTRIC CO., NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| NUDOCS 8105140335 | |
| Download: ML19347E939 (34) | |
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'S UNITED STATES 8i'
^n NUCLEAR REGULATORY COMMISSION f
WASHINGTON, D. C. 20555
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and STN 50-483 g
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3 N,_, \\-fd Union Electric Company Kansas Gas & Electric Comp
/b ATTN: Mr. J. K. Bryan ATTN: Mr. Glenn L. Koester Vice President Vice President - Nuclear P. O. Box 149 201 North Market Street St. Louis, Missouri 63166 Wichita, Kansas 67201
Dear Gentlemen:
Subject:
Callaway Plant, Unit 1 and Wolf Creek Generating Station, Unit 1 Meeting to Resolve Open Items in SNUPPS FSAR Sections 3.2, 3.6, 3.7 and 3.9 Our review of your applications for the Callaway Plant, Unit 1 and Wolf Creek Generating Station, Unit I has progressed to the point that we now have a draft Safety Evaluatia Report (SER) in the mechanical engineering area.
We propose to accelerate the remainder of our review of the mechanical engineering area on an experimental basis, as discussed below.
We request that your staff (1) review the enclosed draf t SER input, (2) prepare responses to resolve each of the open items and (3) meet with the NRC staff at a mutually agreeable time and place where your staff will present and discuss your responses and proposed solutions for each open item.
It would be helpful if you could sut,mit draft responses regarding each of the open items to us one week prior to the meeting.
The objective of this me'eting is to resolve each open item during the meeting without resorting to further written questions and answers. We propose to have appropriate NRC technical personnel attending the meeting, including the Branch Chief and we request that you be represented by an equivalent level of management so that decisions can be reached. Any necessary appeals would subsequently be handled at the NRC Assistant Division Director and Director level.
When your staff has had sufficient time to review the draft SER input and has prepared responses to resolve each open item you should develop an agenda for a meeting to close out the open items. You are further requested to recommend 1
a location and time for this meeting. Since you bear the burden to achieve resolution, we want to give you preference on the meeting agenda, time and place. For this particular meeting, we suggest that you consider holding the meeting in the Bechtel Gaithersburg facilities due to the availability of information needed to resolve the matters of concern.
8105140 M K.
When the proposed agenda and location have been received and reviewed, we will issue a meeting notice. As intervenors and interested members of the public may wish to attend the meeting, notice must be issued a minimum of two weeks before the meeting is held. Please select a site which will not preclude members of the public from attending.
If you have any questions regarding this matter, please call the Callaway Plant, Unit 1 or Wolf Creek Generating Station, Unit 1 Licensing Project Managers.
Sincerely,
, )(. ][.f. _ e t :.-
Robert L. Tedesco, Assistant Director for Licensing Division of Licensing
Enclosure:
As stated cc:
See next pcGe i
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i Mr. J. K. c-"an Vice Presidant - Nuclede lir. uienn L. Koester Vice Fresidcr. - Nuclear Union Electric Comoany P. 0. Scx 1;C Kansas Gas ano Eicctric Campany 201 North Marke: Street St. Lcuis, Missouri 63166
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ichi ta, Kansa! 6 ?dt" c:: C:r:Id Charr.ari, Esq.
Sn;... Pi ttnaa, Fc tts, Dr. Vern Starks Trowbridge & Maccen Route 1, Box S53 1800 M Street, N. W.
Ketchikan, Alaska 99901 Washington, O. C.
20036 Mr. William han:en Kansas City Pc.:er S L ;nt Comoany U. S. 'Lcle3- ;eculatory Cocaissicn "TTN:
Mr. :. T. ":Pnee
- esite-t Int;.ec _;rs
- ':s i ts cre5 c=rst - Procuction RR ='
1330 Baltimore Avenue Steednar., '4iss:uri 65077 ransas City, Missouri 6a101
':s. Treva Hearn. Assistan: ';eneral Caom.el Mr. Nichala: A. Detrick Missouri 3 colic Service Co m.ission Executive Girector, SNUPPS P. O. Box 360
- Choke Cherry F.oad Jeffert:n City, '<isscur'.
65102 Rockville, Maryland 20550 Mr. J. E. Birk Jay Silberg, Es:uire Assistant to the General Counsel Sha.., Pittnan, 3:t:s & Trowbridge 1800 M Street, N. W.
Union Electric Comoany Washingtcn, C. C.
20036 St. Louis, Missouri 63166 Mr. D. F. Scnnell Kansans for Sensible Energy P. O. Box 3192 Manager - Nuclear Engineering Union Electric Company Wichita, Kansas 67201 P. O. Box 149 Francis Blaufuse St. Louis, Missouri 63166 Westphalia, Kansas 66093 Ms. Mary Ellen Salava Route 1, Box 56 Mr. Tom Vandel Burlington, Kansas 66831 Resident Inspector / Wolf Creek NPS c/o USNRC Mr. L. F. Orbi P. O. Box 1407 Missouri - Kansas Section Emporia, Kansas 66801 American Nuclear Society 15114 Navaho Mr. Michael C. Keener Olathe, Kansas 66052 Wolf Creek Project Director State Corporation Comnission State of Kansas Ms. Wanda Christy 515 N. 1st Street Fcurth Floor, State Office Building Burlington, Kansas 66839 Topeka, Kansas 66612 Floyd Mathews, Esq.
Birch, Horton, Bittner & Monroe 1140 Connecticut Avenue, N. W.
Washington, D. C.
20036 P00ROR8M1
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S DRNiT SNUPPS SER l
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1 DRE 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS AND SYSTEMS 3.2.1 Seismic Classification Criterion 2 of the General Design Criteria requires that nuclear power plant structures, s stem, and ce"1ponents important to safety be designed to withstand the e'.ects of earthquakes without loss of capability to perfonn their safety runction. These plant features are those necessary to assure:
- 1) the integrity of the reactor coolant pressure boundary; 2) the capability to shutdown the reactor and maintain it in a safe condition; or 3) the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to 10 CFR, Part 100 guideline exposures.
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Structures, systems, and components important to safety that are required to be designed to withstand the effects of the safe shutdown earthquake and remain functional must be classified as seismic Category I items in accordance with Regulatory Guide 1.29, " Seismic Design Classification". All other structures, systems, and components that may be required for operation of the facility have been designed to other than seismic Category I require-ments. This includes those portions of seismic Category I systems such as vent, drain and test lines on the downstream side of isolation valves which are not required to perform a safety function.
The following discusses open issues found in our review of FSAR Section 3.2.1.
It concludes with our findings contingent upon resolution of all open issues.
The applicant states that nonsafety-related structures, systems and components that must be designed to retain structural integrity during and after an SSE, but do not have to function, are seismically analyzed.l nThe s.vice.L.
- applicant should make assurances that these items meet the ?d. ad limits.
It is also stated in the FSAR that the above mentioned items are not controlled by a 10 CFR 50 Appendix B Quality Assurance Program. The items should be included in the Quality Assurance Program / or in a proyom (avia;3 an ayiva t a lent o f q ual;9 a ssa ca.n ee.,
2 The applicant does not fully comply with Regulatory Guide 1.29.
The applicant should commit to comply with Regulatory Guide 1.29 or provide justification that an equivalent level of safety is ensured.
Based upon our review of FSAR Section 3.2.1 and subject to the satisfactory resolution of the open items, our findings will be as follows:
Structures, systerr", and components important to safety that have been designed to withstand the effects of a safe shutdown earthquake and remain functional are identified in an acceptable manner in Table 3.2-1 of the Final Safety Analysis Report. The basis for acceptance in our review has been conformance of the applicant's designs, design criteria and design bases for structures, systems, and components important to safety with the Comission's regulations as set forth in General Design Criterion 2 and in Regulatory Guide 1.29, " Seismic Design Classification", our technical positions and industry codes and standards.
We conclude that structures, systems, and components important to safety that are designed to withstand the effects of a safe shutdown earthquake and remain functional have been properly classified as seismic Category I i'. ems in conformance with the Commission's regulations, the applicable regulatory guides, and industry codes and standards and are acceptable. Design of these items in accordance with seismic Cttegory I requirements provides reasonable assurance that in the event of a safe shutdown earthquake, the plant will perform in a manner providing adequate safeguards to the health and safety of the pullic.
3.2.2 System Quality Group Classification Criterion 1 of the General Design Criteria requires that the nuclear power plant systems and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety function to be performed.
Fluia system pressure-retaining components important to safety must be designed, fabricated, erected and tested to quality standards commensurate with the importance of the safety function to be performed.
3 eur-Based upon e/ review of FSAR Section 3.2.2 and contingent upon satisfactory resolution of the open issues, our findings will be as follows:
The applicant has identified those fluid-containing components which are part of the reactor coolant pressure boundary and other fluid systems important to safety where reliance is placed on these systems: 1) to prevent or mitigate the consequences of accidents and malfunctions originating within the reactor coolant pressure boundary, 2) to permit shutdown of the reactor and maintain it in a safe shutdown condition, and 3) to contain radioactive material. These fluid systems have been classified in an acceptable manner in Table 3.2-1 of the Final Safety Analysis Report and on system piping and instrumentation diagrams in the Final Safety Analysis Report based on conformance with Regulatory Guide 1.26, " Quality Group Classification and Standards".
The applicant has applied Quality Groups A, B, C, and D in Regulatory Guide 1.26, " Quality Group Classifications and Standards", to the fluid system pressure-retaining components important to safety. These components that are classified Quality Group A, B, C, or D have been constructed to the codes and standards identified in Table 3.2-2 of the Final Safety Analysis Report.
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4 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING The review performed under this section pertains to the applicant's program for protecting safety-related components and structures against the effects of postulated pipe breaks both inside and outside containment.
The effect that breaks or cracks in high or moderate energy fluid systems would have on adjacent safety-related components or structures has been analyzed with respect to jet impingement, pipe whip, and environmental effects. Seversi means are used to assure the protection of these safety-related items. They include physical separation, enclosure within suitably designed structures, the use of pipe whip restraints, and the use of equipment shields.
3.6.2 Determination of Break Locations and Dynamic Effec s Associated With the Postulated Rupture of Piping The review performed under Standard Review Plan Section 3.6.2 pertains to the applicant's program for protecting safety-related co ponents and structures against the effects of postulated pipe breaks both inside and outside containment. The effect that breaks or cracks in high and moderate energy fluid systems would have on adjacent safety related components or structures have been analyzed with respect to pipe whip, jet impin@ ment and environmental effects. Several means are used to assure the protection of safety related systems cnd components. These include shysical separation, enclosure within suitably designed structures, pipe whip restraints and equipment shields.
Standard Review Plan 3.6.2 also sets forth certain criteria for the analgsis and subsequent in-service inspection of high energy piping $$
b r d $ W 5} 5 M thd. Breaks need not be postulated in those portions of piping that meet the requirements of the ASME Code,Section III, Sub-article NE-1120 and the additional design requirements outlined in Branch Technical Position MEB 3-1.
Additional in-service inspection is also required for those portions of piping.
5 The applicant has stated that the failure of seismic Category I and seismically supported nonseismic Category I piping was caused by some mechanism other than an earthquake and that nonseismic Category I equipment could be used to bring the plant to a safe shutdown. The applicant should provide assurance that neither the cause of failure of the seismic Category I piping nor the failed seismic Category I piping could fail the nonseismic equipment mentioned above. Assurance must also be made that only seismic Category I equipment will be used to bring the plant to a safe shutdown in the event of an SSE.
The applicant should provide assurances that in the event of a postulated pipe break, all potential targets are considered.
The applicant states that the calculation of thrust and jet impingement forces considers any line restrictions (e.g., flow limiter) between the pressure source and break location and the absence of energy reservoirs, as applicable.
The applicant should list all instances where line restrictions or the absence of reservoirs were used.
The applicant should provide assurances that breaks and leakage cracks in nonseismic Category I piping are postulated in worse case locations and failure of non-seismic Category I piping will not cause failures of seismic equipment.
The applicant's criteria for postulating breaks in Class 1 piping is not in compliance with Standard Review Plan 3.6.2.
The applicant must either comply with SRP 3.6.2 or provide justification for the acceptability of their criteria.
The applicant has stated that welded attachments to those portions of piping included in the "no break zone" was not allowed except where detailed stress analyses could be performed to demonstrate compliance with appropriate limits. The applicant should provide details of all such locations where welded attachments were used.
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6 Various sheets of Figure 3.6-1, Table 3.6-3 and Table 3.6-4 indicate that pipe break restraint locations, Class I analysis pipe break locations, effects analysis for high-energy pipe breaks located within containment are all under review. The applicant should provide a schedule for completing these reviews. Our acceptance of Section 3.6.2 is contingent upon final review of these updated figures and tables.
Based on our review of FSAR Section 3.6.2 and subject to the satisfactory resolution of the identified open items, our findings will be as follows:
The applicant has proposed 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 to evaluate the design of systems, components, 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 imp)tingement barriers, and other such devices will be used to mitigate the effects of these postula-ted piping failures.
We have reviewed these criteria and have concluded 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 assumed. We find that the methods used to design the pipe whip restraints provide adequate assurance that they will function properly in the event of a postulated piping failure. We further conclude that the use of the applicant's proposed pipe failure criteria in designing gely shut the plant the systems, components, and structures necessary s
down and to mitigate the consequences of these pct; Acted piping failures 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 applicable portions of General Design Criterion 4.
7 j
3.7.3 Seismic Subsystem Analyses (Seismic Category I Piping)
The review performed under Standard Review Plan Section 3.7.3 included the applicant's dynamic analysis of all seismic Category I piping systems.
In audition to normal operating loads, this analysis also considers abnormal loadings such as an earthquake.
For the dynamic analysis of seismic Category I piping each pipe line was idealized as a mathematical model consisting of lumped masses connected by elastic members. The stiffness matrix for the piping system was determined using the elastic properties of the Dipe. This includes the effects of torsional, bending, shear, and axial deformations as well as change in stiffness due to curved members. Next, the mode shapes and the undamped natural frequencies were obtained. The dynamic response of the system was calculated by using the response spectrum method of analysis.
When the piping system was anchored and supported at points with different excitations, the response spectrum analysis was performed using the envelope response spectrum of all attachment points. Alternately, the multiple excitation analyses methods may have been used where acceleration time-histories or response spectra were applied to all piping system attachment points.
The relative displacement between anchors was determined from the dynamic analysis of the structures. The results of the relative anchor point displacement were used for a static analysis to determine the additional stresses due to relative anchor point displacements.
The applicant's procedures for the dynamic analysis of Category I piping have been reviewed and found to be generally acceptable. However, the following discusses open issues which must be resolved.
It concludes with our findings which are contingent upon the resolution of all open issues.
of The applicant's methods fg,r gglance, plant seismic analysis of piping Clarification of the methods used in systems is discussed in BP-TOP-1g combining closely spaced modes is required.
If the methods used deviate from Regulatory Guide 1.92, justification for the acceptability of the methods should be provided by the applicant.
j The applicant's method of response spectrum enveloping requires clarification.
The damping values used by the applicant are consistent with Regulatory Guide 1.61, except in the case of the primary coolant loop system components and large piping (excluding reactor pressure vessel internals) within the a fou r-NSSS for which a damping value percent is used for the faulted conditions.
g The value of damping was obtained by testing as described in WCAP-7921-AR and ensures a level of safety consistent with Regulatory Guide 1.61.
The applicant's method of combining modal responses for liSSS components is not necessarily conservative with respect to the requirements of Regulatory Guide 1.92.
Justification for the acceptability of the method tsed should be provided.
Based on our review of FSAR Section 3.7.3 and subject to the satisfactory resolution of the identified open items, our findings will be as follows:
The scope of review of the seismic system and subsystem analysis for the StiUPPS Plants included the seismic analysis methods for all Category I systems and components.
It included review of procedures used for modeling and evaluating Category I systems and components.
The review included design criteria and procedures for evaluation of the interaction of non-Category I piping with Category I piping. The review also included criteria and seismic analysis procedures for reactor internals and Category I piping outside containment.
The system and subsystem analyses are performed by the applicant on an elastic basis. Modal response spectrum multidegree of freedom and time history methods form the bases for the analyses of all major Category I systems and components. When the modal response spectrum method is used, governing response parameters are combined by the square root of the sum of the squares rule. Appropriate methods are used for combining modes with closely spaced frequencies. The square root of the sum of the squares of the maximum co-directional responses is used in accounting for three components of the earthquake motion for both the time history and response spectrum methods.
- f. ';crtisi sci =ic cytt~" t'"*c := lysis i:-c: picycd fcr :l' systc= =d 4 =;c-am + =.
9 We conclude that the seismic system and subsystem analysis procedures and criteria proposed by the applicant provide an acceptable basis for the seismic design of systems and components.
J
10 3.9 MECHANICAL SYSTEMS AND COMPONENTS The review performed under Standard Review Plan Sections 3.9.1 through 3.9.6 pertains to the structural integrity and operability of various safety-related mechanical components in the plant. Our review is not limited to ASME Code components and supports, but is extended to other components,
such as control rod drive mechanisms, certain reactor internals, ftpf.g:/< MY tier 2;....
daga...g,ayg;w;,;yc, and any safety-related piping designed to industry c
standards other than the ASME Code. We review such issues as load ccmbinations, allowable stresses, methods of analysis, summary results, and pre-operational testing. Our review must arrive at the conclusion that there is adequate assurance of a mechanical component performing its safety-related function under all postulated combinations of normal operating conditions, system operating transients, postulated pipe breaks, and seismic events.
3.9.1 Special Topics for Mechanical Components The review performed under Standard Review Plan Section 3.9.1 pertains to the design transients, computer programs, experimental stress analyses and elastic-plastic analysis methods that were used in the analysis of seismic Category I ASME Code and non-Code items.
Additionally, we have contracted with Pacific Northwest Laboratories to perfonn an independent analysis of a sample piping system in the SNUPPS plants. This analysis will verify that the sample piping system meets the applicable ASME Code requirements. We will report the results of this independent piping analysis in a supplement to this Safety Evaluation Report.
Computer programs were used in the analysis of specific components.
A list of the computer programs used in the dynamic and static analyses to determine the structural and functional integrity of these components must be included in the FSAR along with a brief description of each program.
Design control measures, which are required by 10 CFR Part 50, Appendix B, require that verification of the computer programs also be included. The applicant has not provided verification for all of the listed computer programs.
11 y Amt Aaek in th R P V i&nnis aLra 1a am s 'cc.s add m M& &
g d uiy le ases Loc ^ s I, ost 4 fier y h, n y e-s a 3 f o II. + s a Fcrartfication is needed for design transients used in the analyses of mechanical components for the BOP scope.
There appears to be contradictory statements in the FSAR regarding the use of inelastic analyses in the design of Code and non-Code components for the faulted condition. The applicant should clarify this area.
Based upon our review of FSAR section 3.9.1 and contingent on the satisfactory resolution of the open items, our findings will be as follows.
The methods of analysis that the applicant has employed in the design of all Seismic Category I ASME Code Class 1, 2 and 3 corponents, component supports, reactor internals, and other non-Code items are in conformance with Standard Review Plan 3.9.1 and satisfy the applicable portions of General Design Criteria 2, 4,14 and 15.
The criteria used in defining the apolicable transients and the computer codes and analytical methods used in the analyses provide assurance that the calculations of stresses, strains, and displacements for the above noted items conform with the current state-of-the-art and are adequate for the design of these items.
3.9.2 Dynamic Testing and Analysis of Systems, Components, and Equionent The review performed under Standard Review Plan Section 3.9.2 pertains to the criteria, testing procedures, and dynamic analyses employed by the applicant to assure tre structural integrity and operability of piping systems, mechanical eg;ipment, reactor internals and their suoports under vibratory loadings. Seismic qualification of safety-related mechanical equipment will be reviewed by the Equipment Qualification Branch.
Piping vibration, thermal expansion, and dynanic effects testing will be conducted during a preoperational testing program. The purpose of these tests is to assure that the piping vibrations are within acceptable limits and that the piping system can expand thermally in a manner consistent with m r<.nr the design ic.;_n During the SNUPPS plants' preoperational and startup testing program, the applicant will test various piping systems for abnormal l
12 steady-state or transient vibration and for restraint of thermal growth.
This test program must comply with the ASME Code,Section III, paragraphs NB-3622, NC-3622, and ND-3622 which require that the designer be responsible by observation during startup or initial operation, for ensuring that the vibration of piping systems is within the acceptable levels.
In addition, pipe whip restraint initial clearances will be checked, as wiil snubber response. The test program should consist of a mixture of instrumented measurements and visual observation by qualified personnel. The applicant will be required to provide a summary of tne results of this test program upon its completion.
The applicant's discussion of the testing program in the FSAR needs clarification and amplification. More information is needed regarding the piping vibration, thermal expansion and dynamic effects testing program.
Details are requested on the acceptance criteria for steady-state and transient vibration.
In particular, how are the stresses associated with measured deflections calculated? The applicant should make a commitment to provide the NRC documentation of any corrective action resulting from the tests and conformation by additional testing that substantiates effectiveness of the corrective action, L
a p e ro r -ia+<. eacapP s The applicant should reference the-ppc H :t
- ticnc of ASME Section III in Table 3.9(N)-3.
Based upon or review of FSAR Section 3.9.2.1 and contingent upon the satisfactory resolution of the open items, our findings will be as follows:
The vibration, thermal expansion, and dynamic effects test program which will be conducted during startup and initial operation specified high and moderate energy piping, and all associated systems, restraints and supports is an acceptable program. The tests provide adequate assurance that the piping and piping restraints of the system have been designed to withstand vibrational dynamic effects due to valve closures, pump trips, and other operating modes associated with the design basis flow conditions.
In addition, the tests provide assurance that adequate clearnaces and free
13 movement of snubbe rs exist for unrestrained thermal movement of piping and supports during normal system heatup and cooldown operations. The planned tests will develop loads similar to those experienced during reactor operation. This test program complies with Standard Review Plan Section 3.9.2 and constitutes an acceptable basis for fulfilling the applicable require-ments of General Design Criteria 14 and 15.
Based upon our review of FSAR Section 3.9.2.3, 3.9.2.4, and 3.9.2.6 and subject to resolution of the above open issue, our findings are as follows:
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 comparable to those that will be experienced during operation. The combination of tests, predictive analysis, and post-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 integrity. 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 one applicable requirements of General Design Criteria 1 and 4.
The applicant has analyzed the reactor, its internals, and unbroken loops of the reactor coolant pressure boundary, including the supports, for the combined loads due to a simultaneous loss-of-coolant accident and safe shutdown earthquake.
I Based upon our review of the FSAR Section 3.9.2.5 and subject to resolu-tion of any open items, our findings are as follows:
The dynamic system analysis performed by the applicant provides an acceptable basis for confirming the structural design adequacy of the reactor, its internals, and unbroken piping loops to withstand the combined dynamic loads of postulated loss of coolant accident (LOCA) and the safe shutdown
14 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 displacements at any structural elements of the reactor internals will not distort the reactor internals geometry to the extent that core cooling may be impaired. The methods used for component analysis have been found to be compatible with those used for the system analysis. The proposed combinations of component and system analyses are, therefore, acceptable.
The assurance of structural integrity under LOCA and SSE conditions for the most adverse postulated loading event provides added confidence that the design will withstand a spectrum of lesser pipe breaks and seismic loading 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 Suoports and Core Suoport Structures Our review tnder Standard Review Plan Section 3.9.3 is concerned with the structural integrity 2nd cper;ii'ity of pressure-retaining components, their supperts. and core support structures which are designed in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, or earlier industry standards. Our review covers Sections 3.9.3.1, 3.9.3.3 and 3.9.3.4, each of which is discussed briefly below.
The first area of review, Section 3.9.3.1, is the subject cf load combinations and allowable stresses. 46 LL m.; caccatica,'Theapplicant has provided a commitment that all ASME Class 1, 2 and 3 components, component supports, core support structures, control rod drive components, and other reactor internals have been analyzed or qualified in accordance with appropriate loading combinations.
Thi; onc cxc;pti;n it that the eppli;;nt should ' :'u'a thnr-=1 rhoct ta the leadi"g C;..iinetien.
Subject to rc:clutic c# the :bcvc ^p;n.52uca, Our findings are as follows.
15 The specified design and service combinations of loadings as applied to 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 :ite or other service loadings due to postulated events or system operating transients, the resulting combined stresses imposed on system components will not exceed allowable stress and strain limits for the materials of construction.
Limiting the stresses under such loading combinations provides a conservative basis for the design of system components to withstand the most adverse combination of loading events without loss of structural integrity. The design and load combinations and associated stress and deformation limits specified for ASME Code Class 1, 2, and 3 components constitute an asceptable basis for design in satisfying applicable portions of General Design Criteria 1, 2, and 4.
The second area of -eview in this section, Section 3.9.3.3, concerns the criteria used by the applicant in designing its ASME Class 1, 2, and 3 safety and relief valves, their attached piping, and their supports. We have specifically reviewed the applicar,t's compliance with Regulatory Guide 1.67, "Installaticr. of Overpressure Protection Devices".
The applicant states that a dynamic load factor of 2.0 was used unless a lower value can be justified. A list of all instances when a dynamic load factor of less than 2.0 was used should be provided along with the needed justification.
E,asad upon our review of Section 3.9.3,3, our findings are as follows.
The criteria used in the design and installazion of ASME Class 1, 2, and 3 safety and relief valves provide adequate assurance that, under discharging conditions, the resulting stresses will not exceed allowable stress and strain limits for the materials of construction. Limiting the stresses under the loading combinations associated with the actuation of these pressure relief devices provides a conservative basis for the design and installation of
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16 the devices to withstand these loads without loss of structural integrity or impairment of the overpressure protection function. The criteria used for the design and installation of ASME Class 1, 2, and 3 overpressure relief devices constitute an acceptable basis for meeting the applicable require-ments of General Design Criteria 1, 2, 4,14, and 15 and are consistent with those specified in Regulatory Guide 1.67 and Standard Review Plan Section 3.9.3.
The third area of our review in this :ection, Section 3.9.3.4, was the criteria used by the applican' in the design of ASME Class 1, 2, and 3 component supports. All component tupports have been designed in accordance with I
Subsection NF of the ASME Code,Section III.
We have determined the applicant's compliance with two major issues a
addressed by Regulatory Guides 1.124, " Service Limits and Loading Combinations j
for Class.1 Linear-Type Component Supports", and 1.130, " Service t imits and Loading Combinations for Class 1 Plate-and-Shell-Type Component Supports".
These subjects are buckling of component supports and the cesign of bolts used in component supports.
Based upon our review of Section 3.9.3.4, our findings are as follows:
The specified design and service loading conbinations used for the design of ASME Code Class 1, 2, and 3 component supports in systems classified as I
seismic Category I provide assurance that, in the event of an earthquake or j
other' service loadings due to postulated events or system operating transients, the resulting combined stresses imposed on system components will not exceed allowable stress and strain limits for the materials of construction.
Limiting the stresses under such loading combinations provides a conservative:
basis for the design of support components to withstand the most adverse combination of loading events without loss of structural integrity or supp1rted 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 constitute an acceptable basis for utisfying applicable portions of General Design Criteria 1, 2, and 4.
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i 17 3.9.4 Control Rod Drive Systems Our review under Standard Review Plan Section 3.9.4 covers the design of the h,fr: 'i: control rod drive system up to its interface with the control rods. We reviewed the artlyses 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.
Based upon our review of FSAR Section 3.9.4, our findir.gs are as follows.
The design criteria and the testing progran 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 will form an acceptable basis for satisfying the mechanical reliability requirenents of General Design Criterion 27.
3.9.5 Reactor Pressure Vessel Internals Our review under Standard Review Plan Section 3.9.5 is concerned with the load combinations, allowable stress limits, and other criteria used in the cesign of the SNUPPS reactor internals. The applicant has stated that the reactor internals have been designed in accordance with Subsection NG, " Core Support Structures", of the ASME Code,Section III. The description of the configuration and general arrangement of the reactor internal structures, components, assemblies and systems has been reviewed and found to be quite conplete.
Based upon our review of FSAR Section 3.9.5 and contincent upon the satisfactory resolution of the open items, our findings will be as follows.
O 18 The specified transients, design and service loadings, and combination of loadings as applied to the design of the SNUPPS reactor internals provide reasonable assurance that in the event of an earthquake or of a system transient during normal plant operation, 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 to occur during the. 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 SNUPP5 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.
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19 3.9.6 There are several safety systems connected to the reactor coolant pressure boundary that have design pressure below the rated reactor cooland system (RCS) pressure. There are also soma systems which are rated at full reactor pressure on the discharge side of p* imps but have pump suction below RCS pressure.
In order to protect these systems from RCS pressure, two or more isolation valves are placed in series to fom the interface between the high pressure RCS and the low pressure systems. The ice.k tight integrity of these valves must be ensured by periodic leak testing to prevent exceeding the design pressure of the low pressure systems thus causing an inter-system LOCA.
Pressure isolation valves are required to be category A or AC per IW-2000 and to meet the appropriate requirements of IW-3420 of Section XI of the ASME Code except as discussed below.
Limiting Conditions for Operation (LCO) are required to be added to the technical specifications which will require corrective action; i.e., shutdown or system isolation when the final approved leakage limits are not met.
Also, surveillance requirements, which will state the acceptable leak rate testing frequency, shall be provided in the technical specifications.
Periodic leak testing of each pressure isolation valve is required to be performed at least once par each refueling outage, after valve maintenance prior to return to. service, and for systems rated at less than 50". of RCS design pressure each time the valve has moved from its fully closed position unless justification is given.
The testing interval should average to be approximately one year. Leak testing should also be performed after all disturbances to the valves are complete, prior to reaching power operation following a refueling outage, maintenance, etc.
The staff's present position on leak rate limiting conditions for operation must be equal to or less than 1 gallon per minute for each valve (GpM) to ensure the integrity of the valve, demonstrate the adequacy of the
20 redundant pressure isolation function and give an indication of val w degradation over a finite period of time.
Significant increases over this limiting valve would be an indication of valve degradation from one test to another.
Leak rates higher than 1 GPM will be considered if the leak rate changes are below 1 GPM above the previous test leak rate or system design precludes measuring 1 GPM with sufficient accuracy. These items will be reviewed on a case by case basis.
The Class 1 to Class 2 boundary will be considered the isolation point which must be protected by redundant isolation valves.
In cases where pressure isolation is provided by two valves, both will be independently leak tested. When three or more valves provide isolation, only so of the valves need to be leak tested.
Provide a list of all pressure isolation valves included in your testing program along with four sets of Piping and Instrument Diagrams which describe your reactor coolant system pressure isolation valves. Also discuss in detail how your leak testing program will conform to the above staff position.
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ATTACHMENT A DRAFT 00ESTI0f1S FOR SNUPPS SER
A-1 JMA(NRC SNUPPS/SER 4/1/81 QUESTIONS ON SNUPPS FSAR
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Page 3.2-2 "Nonsafety-ralated structures, systems, and components that must be designed to retain structural integrity during and after an SSE, but do not have to function, are seismically analyzed." Assurance should be made that the above items meet the faulted limits.
It is als; stated that these above items are not controlled by a 10 CFR 50 Appendix B Ouality Assurance Program. These items should be included in the Quality Assurance Programf on e g val.ot I v a.I of palig
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Section 3.6.1.1.h.2(b), Page 3.6-3 It is stated that it was assumed the failure of seismic Category I and seismically suoported nonseismic Category I piping was caused by some mechanism other than an earthquake and, therefore, that nonseismic Category I equipment could be used to bring the plant to a safe shutdown. What mechanisms are postulated for failure of seismic Category I and seismically supported nonseismic Category I piping? Assurance must be made that the failed seismic pipir.g does not damage the nonseismic Category I equipment mentioned above. Assurance must also be made that only seismic Category I equipment will be used to bring the plant to a safe shutdown in the event of an SSE.
Section 3.6.1.1.j, Page 3.6-4 It is stated that the pipe whip was assumed to occur in the plane defined by the piping geometry and to cause movement in the direction of the jet reaction. Assurances must be made that this criteria was used only in the design of pipe whip restraints and that failed piping was considered capable of swinging in any direction about a plastic hinge following a pipe rupture and all potential targets were considered.
A-2 Section 3.6.1.1.k All instances where line restrictions or the absence of energy reservoirs were used in the calculation of thrust and jet impingement forces should be listed.
Figure 3.6-1 Various sheets indicate that pipe break restraint locations, Class I analyais pipe break locations, and effects analysis for high-energy pipe breaks located within containment are all under review. We cannot complete our review until these reviews are completed.
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A statement should be made that breaks and leakage cracks in nonseismic Category I piping are postulated in worse case locations and that failure of non-seismic Category I piping will not cause failure of seismic equipment.
Table 3.6-3 What is the difference between Sheet 1 and Sheet 2, Sheet 3 and Sheet 4, Sheet 5 and Sheet 6, and Sheet 7 and Sheet 87 Table 3.6-3 Sheets 28, 32 and 36 indicate that the stress analysis is under review.
We cannot complete our review of E ct13n 3.6.2 until this information is furnished.
Table 3.6-4 Data in this table under effects analysis are listed as (under review).
We cannot complete our review of Section 3.6.2 until this information is furnished.
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A-3 Page 3.7(B)-11 Reference is made to FSAR Section 3.7(B).2.7 which references Sections 5.1 and 5.2 of BP-TOP-1 for the criteria used for combining modal responses for piping systems. The last sentence in Section 5.2 of BP-TOP-1 (page 14) includes the words "if they do occur in-phase" with regard to when the grouping method or the double sum method will be used for closely spaced modes.
Please indicate how closely spaced modes were determined to " occur in-phase" and give an example of when they were determined not to occur in phase.
Figures 3.7(B)-5 through 3.7(B)-8 The response spectra of the synthetic time-history does not envelope the corresponding design spectra for all frequencies. Please explain this apparent non-conservatism.
Figures 3.7(B)-9A and 3.7(B)-9D Please explain the significance and conservatism of these figures.
Page 3.7(?!)-14 Equation [3.7(ft)-29] is not necessarily conservative with respect to the requirements of Reg. Guide 1.92.
Provide justification for its accept-ability. Equation [3.7(fi)-30] is in error.
Section 3.9(B).l.1, Page 3.9(B)-1 Reference is made to Section 3.9(fi).l.1. Section 3.9(fi).1.1 discusses the transients considered ir the design of the reactor coolant system (RCS),
RCS component supports, and reactor internals. Are these the same transients used in the design of the BCP components?
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A-4 Section 3.9(B).l.2.1.1, Page 3.9(B)-1 This section references Appendix 3.9(B)A which states that ME-632 results were compared with the results of the previously approved Engineering Data System (EDS) computer programs. Where is a discussion of the verifica-tion of the EDS programs and when was it approved?
Section 3.9(B).1.3.2, Page 3.9(B)-3 It is indicated that inelastic methods are not used in the design of Code or non-Code components for the faulted condition. On Page 3.9(B)-4, Section 3.9(B).l.4.2 it is indicated that inelastic analyses were used.
Please clear up the discrepancy.
Section 3.9(B).2.1, Page 3.9(B)-4 More information is needed regarding the piping vibration, thermal expansion and dynamic effects testing programs. Please list those systems to be monitored for 1) transient induced vibration, 2) steady state vibratici.
and 3) thermal expansion. Also list the flow modes of operation to be included in the testing program. List those locations where visual inspection will be utilized and those locations where measurements will be taken and also the associated acceptance criteria. A commitment should be included that the NRC will be provided documentation of any corrective action resulting from the tests and confirmation by additional testing that substantiates effectiveness of the corrective action.
Section 3.9(B).2.1, Page 3.9(B)-5 The applicant should indicate whether the listed systems meet the SRP 3.9.2 requirements with respect to the scope of this program.
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A-5 Section 3.9(B).3.3.1.g. Page 3.9(B)-15 Please list all instances when a dynamic load factor of less than 2.0 was used and provide the needed justification.
Section 3.9(N).2.1, Page 3.9(N)-33 Please describe the acceptance limits that will be used for visual inspection of vibration.
How will the stresses associated with the vibration be calculated? What ASME Code stress and fatigue limits will be used?
What measJres will be taken to monitor the thermal movement of the primary loop during heat up to ensure that no restraint to thermal growth is encountered?
Section 3.9(N)-2.4, Page 3.9(N)-36 The FSAR should clearly state that the SNUPPS plants are classified as non-Prototype Category I in accordance with Reg. Guide 1.20.
Table 3.9(N)-3 (mpmphs The appropriate sastoons of ASME Section III should be referenced for the various components listed.
Section 3.6.2.1.1.9.2(b), Page 3.6-10 The pipe break criteria is not in compliance with SRP 3.6.2 in that the 3.0 S,value should be 2.4 S.
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Section 3.6.2.1.1.e, Page 3.6-13 Please provide details of all locations where welded attachments were made to portions of piping covered under this section.
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A-6 3.9.6 There are several safety systems connected to the reactor coolant pressure boundary that have design pressure below the rated reactor cooland system (RCS) pressure. There are also some systems which are rated at full reactor pressure on the discharge side of pumps but have pump suction below RCS pressure.
In order to protect these systems from RCS pressure, two or more isolation valves are placed in series to form the interface between the high pressure RCS and the low pressure systems. The leak tight integrity of these valves must be ensured by periodic leak testing to prevent exceeding the design pressure of the low pressure systems thus causing an inter-system LOCA.
Pressure isolation valves are required to be category A or AC per IWV-2000 and to meet the appropriate requirements of IWV-3420 of Section XI of tne ASME Code except as discussed below.
Limiting Conditions for Operation (LCO) are required to be added to the technical specifications which will require corrective action; i.e., shutdown or system isolation when the final approved leakage limits are not met.
Also, surveillance requirements, which will state the acceptable leak rate testing frequency, shall be provided in the technical specifications.
Periodic leak testing of each pressure isolation valve is required to be perforned at least once per each refueling outage, after valve maintenance prior to return to-service, and for systems rated at less than 50% of RCS design pressure each time the valve has moved from its fully closed position unless justification is given. The testing interval should average to be approximately one year. Leak testing should also be performed after all disturbances to the valves are complete, prior to reaching power operation following a refueling outage, maintenance, etc.
The staff's present position on leak rate limiting conditions for operation must be equal to or less than 1 gallon per minute for each valve (GPM) to ensure the integrity of the valve, demonstrate the adequacy of the l
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A-7 redundant pressure isolation function and give an indication of valve degradation over a finite period of time.
Significant increases over this limiting valve would be an indication of valve degradation from one test to another.
Leak rates higher than 1 GPM will be considered if the leak rate changes are below 1 GPM above the previous test leak rate or system design precludes measuring 1 GPM with sufficient accuracy. These items will be reviewed on a case by case basis.
The Class 1 to Class 2 boundary will be considered the isolation point which must be protected by redundant isolation valves.
In cases where pressure isolation is provided by two valves, both will be independently leak tested. When three or more valves provide isolation, only two of the valves need to be leak tested.
Provide a list of all pressure isolation valves included in your testing program along with four sets of Piping and Instrument Diagrams which describe your reactor coolant system pressure isolation valves. Also discuss in detail how your leak testing program will confonn to the above staff position.
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3 ALL APPLICANTS:
Sec#m 3, % 'A Due to a long history of problems dealing with inoperable and incorrectly installed snubbers, and due to the potential safety significance of failed snubbers in safety related systems and components, it is requested that maintenance records for snubbers be documented as follows:
Pre-service Examination A pre-service examination should be made on all snubbers listed in tables 3.7-4a and 3.7-4b of Standard Technical Specifications 3/4.7.9 This exami-l nation should be made after snubber installation but not more than six months l
prior to initial system pre-operational testing, and should as a m.mimum verify the following:
1 (1) There are no visible signs of damage or impaired operability as a result of storage, handling, or installation.
(2) The snubber location, orientation, position setting, and configuration (attachments, extensions, etc.) are according to design drawings and specifictions.
(3) Snubbers are not seized, frozen or je nmed.
(4) Adequate swing clearance is provided to allow snubber movement.
(5) If applicable, fluid is to the recomended level and is not leaking from the snubber system.
(6) Structural connections such as pins. fasteners and other connecting hardware such as lock nuts, tabs, wire, cotter pins are installed correctly.
If the period Letween the initial pre-service examination and initial system pre-operational test exceeds six months due to unexpected situations, re-examination of items 1,4, and 5 shall be performed. Sndbers which are installed incorrectly or otherwise fail to meet the above requirements must be repaired or replaced and re-examined in accordance with the above criteria.
Pre-Operational Testing During pre-operational testing, snubber thermal movements for systems whose operating temperature exceeds 250' F should be verified as follows:
(a ) During initial system heatup and cooldown, at specified temperature intervals for any system which attains operati'ig temperature, verify the snubber expected thermal movement.
(b) For those systems which do not attain operating temperature, verify via observation and/or calculation that the snubber will accommodate the projected thermal movement.
(c) Verify the snubber swing clearance at specified heatup and cooldown intervals. Any discrepencies or inconsistencies shall be evaluated for cause and corrected prior to proceeding to the next specified interval.
2-The above describ;d operabilit,y program for snubbers should be included and documented by the pre-service inspection and pre-operational test programs.
The pre-service inspection must be a prerequisite for the pre-operational testing of snubber thermal motion. This test program should be specified in Chapter 14 of the FSAR.
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