ML17272A374
| ML17272A374 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 03/22/1979 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | Strand N WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| References | |
| NUDOCS 7904160310 | |
| Download: ML17272A374 (48) | |
Text
MAR aS 1979 Docket No.:
50-39 Mr. Neil O. Strand washington Public Power Supply System 300 George tlashington May P.
O. Box 968 Richland, Washington 99352'ear Nr. Strand:
SUBJECT:
FIRST ROUND QUESTIONS QH THE Distribution Docket File Ll<R g4 Reading D. Lynch "NRC PDR Local PDR R.
Boyd D. Vassallo F. l<illiams S. Varga M. Service R. Mattson D. Ross J. Knight
, R. Tedesco R.
DeYoung V. Moore HHP-2 OL APPLICATIOH - MEB R. Vo11mer M. Ernst R. Denise ELD IE (3) bcc:
J.
In our review of your application for an operating license for the MHP-2 facility, we have identified a need for,additional information vAich we require to complete our review.
The specific requests are contained in the enclosure to this letter and are the seventh set of our round one questions; additional requests related to other portions of the l<NP-2 facility will be sent during this month.
In order to maintain our present
- schedule, we need a completely adequate response to all questions in the enclosure by June 13, 1979.
The attached set of round one questions represent the review effort of the Mechanical Engineering Branch.
Mhere we have been able to formulate statements of staff positions (RSP),
we have included these in our questions.
Please contact us if you require any discussion or clarification of the enclosed requests.
Sincerely, R.~EALLMggeQ Qgf Se Ae VQ+Qa Steven A. Varga, Chief Light llater Reactors Branch Ho. 4 Division of Project Management
Enclosure:
As stated
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Washington Public Power Supply System ccs:
Joseph B. Knotts,'r.; Esq.
Debevoise*8 Liberman 700 Shoreham Building 806 Fifteenth Street, N.
W.
Washington, D.
C.
20005 Richard Q. Quigley, Esq.
Washington Public Power Supply System 3000 George Washington Way P. 0.
Box 968
- Richland, Washington 99352 Nepom 5 Rose Suite 101 Kellogg Building 1935 S.
E. Washington Milwaukie, Oregon 97222 Ms. Susan M. Garrett 7325 S.
E. Steele Street
- Portland, Oregon 94206 Mr. Creg Darby 807 So. Fourth Avenue
- Pasco, Washington 99301
~
Nicholas Lewis, Chairman Energy Facility Site Evaluation Council 820 East Fifth Avenue Olympia, Washington 98504
,Mr. 0. K. Earle Licensing Engineer P. 0.
Box 968
- Richland, Washington 99352
ENCLOSURE STATEMENT OF REGUL'ATORY STAFF POSITIONS AND RE UEST FOR ADDITIONAL INFORMATION
.1<PPSS NUCLEAR PLANT NO.
2 DOCKET NO.
50-397
110.0 'ECHANICAL ENGINEERING BRANCH 110. 01 (3. 6. 1)
(3. 6. 2) 110.02
('3. 6. 1) 110.03
, (3.6.1)
You indicate in the FSAR that the fol,lowing portions will be provided at a later date:
Sections 3.6. 1.6 through 3.6. 1. 10, 3.6. 1.20, and 3.6.2.5.4.4.b and 3.6'.5 '.4.c.
Additionally, there are about 40 figures in Section 3.6 of the FSAR which are intended to be summaries of postulated pipe break locations.
- However, these figures have only a single entry; i.e., "later".
Indicate when the missing secti.ons and figures will be submitted.
'Provide in Section 3.6. l. 11.2. l of the
- FSAR, a definition of what is meant by the term "contiguous grid."
Indicate "clearly whether it includes the corner,grids (i.e., the grids which are diagonally adjacent).
Discuss the vertical extent of a contiguous grid.
Additionally, provide 'justification,for not assuming the simultaneous destruction of equipment in more than one contiguous grid.
Describe in Section 3.6. 1. 11.2 of the
- FSAR, how you evaluate the environmental effects of leakage cracks in high energy fluid systems postulated in accordance with the criteria contained in Sections 3.6.2. 1.3 and 3.6.2. 1.4.2.
110. 04.
(3.6.1)
Expand Section 3.6.1.11.3. l of the FSAR to:
(l) provide justification for not assuming the simultaneous malfunction of equipment in one or more contiguous grids; (2) describe your procedures to evaluate the effects of flo'oding which are discussed in Section 3.6.2. 1.4.2.c of the FSAR.
110. 05
"(3.6.2) 110. 06 RSP (3.6.2)
State in Section 3.6.2.1.1 of the FSAR, the criteria for postulating break locations in high energy piping not designed to seismic Category I criteria.
It is our position that a branch pipe connection to a main run of pipe need not be considered as a terminal end when all the following conditions are met: '1) the branch and main runs are of comparable size and degree of fixity (i.e., the nominal size of the branch is at least one half that of the main); (2) the intersection is not rigidly constrained by the building structure; and (3) the branch and main runs of pipe are modeled as a common piping system in the stress analysis of these pipes.
Revi.se Note (a) of Section
- 3. 6. 2. l. l. l. a to correspond with this definition of terminal ends.
110-1
110.07 (3.6.2) 110. 08 (3.6.2) 110. 09 (3. 6. 2)
Indicate in Nate (f) of Section 3.6.2.l.l.l.b(2)(c) of tPe FSAR, the range of plant operating conditions considered in your evaluation of Equation 13 of subsection NB-3653.6(b) of the ASME Code.
Indicate in Section
- 3. 6. 2. 1. 1. 3 of the
- FSAR, how the criteria for piping which is not designed to comply with the ASME Boiler and Pressure Vessel Code, differs from that for piping which is designed to this code.
Provide in Section 3.6.2.1.2.3 of the
- FSAR, a definition of the phrase "through-wall leakage crack" for which the tunnel structures are designed.
Me note that this section cross-references Section 3.6. 1.20 for further discussion.
- However, as noted in Item 110.01, this latter section is not in the FSAR.
110. 10 RSP (3.6.2) 110. 11 (3.6.2) 110. 12 (3.6.2) 110. 13 (3. 6. 2) 110. 14
'(3.6.2)
It is our position that the piping which is between the contain-ment isolation valves'nd for which no breaks are postulated, will receive a 100 pe'rcent volumetric examination of all welds, including the circumferential, the longitudinal, and the branch to main run welds, during each inspection interval.
(Refer to subsection IMA-2400 of the ASME Code. )
Accordingly, revise Section
- 3. 6. 2. 1. 2 of the FSAR to provide a commitment to such an augmented inservice inspection program.
Indicate in Section 3.6.2.1.3 of the FSAR, the criteria for postulating cracks in:
(1)
ASME Class 1 piping designated as moderate energy lines; and (2) piping not designed to seismic Category I criteria but which are designated as moderate energy lines.
Indicate in Sections 3.6.2.1.4.l.e(l) and (2) of the
- FSAR, how a consideration of the maximum stress range is used to exempt certain break orientations when the postulated break is due to a usage factor,in excess of 0. l.
Indicate in Section 3.6.2.1.4.1. f of the FSAR, where the postulated breaks are located with respect to the fittings.
Describe in Section 3.6.2.1.4.l.g of the FSAR, the "mechanistic approach" which you propose to justify longitudinal breaks having a break area less than the flow area of the pipe.
110-2
110.15 Oescribe in Section 3.6.2.1.5 of the FSAR, the criteria for (3.6.2) providing protection for safety-related structures, systems and components which might be subject to jet. impingement from postulated cracks.
110. 16 (3.6.2) 110. 17 (3.6.2) 110. 18 RSP Provide assurance in Section
- 3. 6. 2. 2. 3. 3 of the FSAR, that the mechanical strain in the energy absorbing, flexural support members does not exceed half of the ultimate uniform strain of the materials under the design loads using your design procedures.
Indicate in Sections 3.6 '.3.2 and 3.6.2.5 of the FSAR, whether:
(1) the environmental effects of postulated pipe breaks (i.e.,
- pressure, temperature, humidity, wetting of exposed equipment and flooding), have been considered in the design of the WNP-2 facility; and (2) these environmental effects are at least as severe as those associated with a postulated crack of the same size as the postulated break.
Expand Table 3.6-4 to include the control rod drive hydraulic piping system and the condensate piping 'system (i. e.,
the'iping which runs between the condensate pump discharge and the condensate demineralizers);
110. 19 RSP (3.e.2)
Previous analyses of other nuclear plants have shown that certain reactor system components and their supports may be subjected to asymmetric loads which are higher, on a conservative
- basis, than the loads estimated i'n the original analyses.
These asymmetric loads could result from postulated ruptures of the reactor coolant piping at various locations.
Accordingly, it is our position that you assess the capability of these reactor system components of the WNP-2 facility, including their supports, to provide assurance that the calculated
- dynamic, asymmetric loads resulting from these postulated pipe ruptures will be adequately conservative (ice., provide assurance that the reactor can be brought safely to a cold shutdown condition).
The reactor system components that require this reassessment are:
(1) the reactor pressure vessel; (2) the core supports and other reactor internals; (3) the control rod drives; (4) the emergency core cooling system (ECCS) piping which is attached to the primary coolant piping; (5) the primary coolant piping; and (6) the reactor vessel supports.
The effects of postulated asymmetric LOCA loads on these reactor system components and the various cavity structures should be submitted, including the following information:
110"3
Provide arrangement drawings of the reactor vesse],support systems in sufficient detail to show:
(1) the geometry of all principal elements; and (2) the materials of construction.
b.
If you choose to reference a generic analysis in your response to this item rather than submitting a unique analysis for the WNP-2 facility, demonstrate that the analysis of the referenced generic plant adequately bounds the postulated accidents in the WNP-2 facility.
Additionally, provide a comparison of the geometric, structural, mechanical and thermal-hydraulic similarities between the WNP-2 facility and the referenced analysis.
Discuss the effects of any differences between your facility and the generic plant.
c.
Consider all postulated breaks in the reactor coolant piping system, including the following locations:
(1) the steam line nozzles at the piping terminal ends; (2) the feedwater nozzles at the piping terminal ends; and (3) the recirculation inlet and outlet nozzles at the recirculation piping terminal ends.
Additionally, provide an assessment of the effects. of asymmetric pressure differentials on the systems and components listed above in combination with all external
- loadings, including the safe shutdown earthquake (SSE) loads and other faulted condition loads, for the postulated pipe breaks described in Item (c) above.
These pressure differentials are the blowdown jet forces at the location of the rupture (i. e.,
the reaction forces), transient differential pressures in the annular region between any affected component and its cavity wall, and transient differential pressures across the core barrel within the reactor pressure vessel.
This assessment may utilize the following mechanistic effects as applicable:
(1) the limited displacement break areas; (2) the effect of fluid-structure interaction; (3) the actual time-dependent forcing function; (4) the reactor support stiffness; and (5) the break opening times.
The following information should also be submitted:
t d.
If the assessment in Item (c) above indicates that loads could be imposed on the affected reactor system components which would exceed the elastic limit of the materials in these components and their supports or which could cause
e.
displacements that exceed previous design limits, provide:
(1) an evaluation of the inelastic behavior of the affected
- material, including a consideration of the strain hardening of the material; and (2) the consequent effect on the loads transmitted to the backup structures to which these systems are attached.
For all the analyses performed in responding to this
- request, indicate the method of analysis and provide:
(1) the structural and hydraulic computer codes employed; (2) schematic drawings of the models; (3) comparisons of the calculated
- stresses, strains and deflections with the allowable values;.
and (4) the basis for selecting the allowable stresses and strains.
Demonstrate that all safety-related active components will function properly when subjected to the loads resulting 110. 20 (3.9.1) 110. 21 (3. 9. 1) 110. 22 (3.9.2) from a postulated LOCA 1n comb>nation with the SSE loads.
g.
Demonstrate the functional capability of the safety-related piping when subjected to the loads resulting from a postulated LOCA in combination with the SSE loads.
Provide your basis for placing the Operating Basis Earthquake (OBE) in the "Emergency" category in Section
- 3. 9. l. l. 3 of the FSAR.
Since continued operation of the plant after the OBE is required by Section III(d) of Appendix A to 10 CFR 100, provide justification for the exclusion of fatigue considerations in the design of essential components (e. g., the control rod drives. )
Provide descriptions, including the testing procedures, of the experimental stress analysis referred to in Section 3.9.1.3.1 of the FSAR and which was conducted to verify the design adequacy of the piping seismic shock suppressors.
Supplement the preoperational piping vibration test program described in Section 3.9.2.1 of the FSAR with detailed informa-tion in the manner discussed in Section 3.9.2 of the Standard Review Plan (SRP),
NUREG-75/087.
In your response, emphasize the measures you wi 11 take to conduct visual inspections and measurements of vibration.
In addition to the piping discussed in Section 3.9.2.
1 (i.e., the recirculation piping and the RHR suction piping), include the following piping syst'ems in your response:
(1) all safety-related ASME Class 1,
2 and 3 piping systems; (2) other high energy piping systems inside seismic Category I structures; (3) those portions of high-energy systems whose failures could adversely affect the functioning 110"5
110.23 (3.9.2) of any safety-related structure, system or component; agd (4) the seismic Category I portions of moderate energy piping systems. located outside containment.
In Section 3.9.2.4 of the FSAR, you indicate that you will perform a non-prototypical preoperational flow test in the WNP-2 facility to determine whether there are any vibrational effects (i.e., wear or loose parts).
The basis you provide for this approach is that the internal design configurations of the WNP-2 facility are substantially similar to the prototype BMR/4 plants.
You further i'ndicate that the effects of the only design change made in the jet pumps would be verified by the vibrati'on. measurement and inspection programs in the Tokai-2 facility.
Accordingly, provide the Tokai-2 jet pump test data.
If such information cannot be provided in a timely manner or in the event that the Tokai-2 information is not
.acceptable, we will require that you classify the WNP-2 facility.
as prototypical in accordance with the guidance contained in Regulatory Guide 1.20, Revision 2, "Comprehensive Vibration
'ssessment Program for Reactor Internals During Preoperatinal and Initial Startup Testing,"
Hay 1976, and perform the following tests and analysis:
a.
Hake instrumented measurements at the jet pumps and at the shroud head during cold-flow, precritical and start-up tests; b.
Provide an analysis to verify that the BWR/4 data and the measurements made in Item (a) above provide assurance that the internals of a BWR/5 facility will not be adversely affected by flow-induced vibrations.
110. 24 (3.9.2)
In the first amendment of the General Electric topical report, NEDE-24057, "Assessment of, Reactor Internals Vibration in BWR/4 and BMR/5 Plants,"
(Reference 3.9-6 of the FSAR), it is implied in the third sentence of the response to request No. 3.a that feedwater spargers with a triple thermal sleeve and top-mounted discharge nozzles will be used in the WNP-2 facility.
The report also indicates that spargers of this type have performed satisfactorally with respect to vibrations during testing.
On this basis, the cited report indicates that no additional vibration measurements are planned.
Accordingly, provide detailed information in Section 3.9.2.3 of the FSAR regarding these sparger tests to demonstrate that these tests are applicable to the WNP-2 facility.
This i nforma-tion should include descriptions of the test setup, the testing procedures and 'the test results.
110-6
110. 25 (3 ~ 9. 3)
Section III of the ASME Boiler and Pressure Vessel Code defines the primary stress limit under design conditions for the combination of membrane plus bending stress rather than the stress limits for bending alone.
Accordingly, revise Section 3.9.3.
1 of the FSAR to include the membrane plus bending stress limits in the design of the pumps in the reactor core isolation cooling system (RCIC), the ECCS and the standby liquid control systems (SLCS).
110. 26 (SRSS) 110.27 RSP (3.9,3)
We have accepted for reactor coolant pressure boundary components, the use of the square-root-of-the-sum-of-the-squares (SRSS) methodology for combining the dynamic structural responses.due to LOCA and SSE loads.
Our acceptance of this approach is documented in NUREG-0484, "Methodology for Combining Dynamic Responses,"
September 1978.
At this time, we have not accepted the use of the SRSS methodology for combining responses from other combinations of dynamic loads and for other components and supports.
However, our review of additional applications of the SRSS methodology is continuing and we are concentrating on the proposed Kennedy-Newmark criteria.
Refer to Appendix I of the GE topical report, NEDO 24010-1, "SRSS Application Criteria As Applied to Mark II Load Combination Cases,"
Supplement 1, October 1978.
It-is anticipated that we can support our position and criteria for a more general applica-tion of the SRSS methodology.
Accordingly, provide a list of all components for which a combination of dynamic responses by the SRSS methodology is proposed, including a list of the dynamic loads which are combined.
This listing should specific-ally include such loads as the OBE inertia loads, the OBE anchor point movement loads, the safety/relief valve (SRV)
- loads, the turbine stop valve closure
- loads, the Mark II pool dynamic loads, the SSE loads, and the LOCA loads, including the annulus pressuri'zation loads.
Verify that the design of the WNP-2 facility complies with the resolution of the generic issues proposed by the Mark II
~
Owner's Group.
In particular, provide verification that you comply with our positions on the load cases, the structural acceptance criteria for piping and the demonstration of functional capability of the safety-related piping systems.
Appendix A
to Section 110 contains the load cases and the str'uctural acceptance criteria while Appendix B to Section 110 presents the criteria for demonstrating'unctional capability.
Note
~ 'he following two clarifications to Appendix 110-A:
110"7
a.
For load cases 1
and 2, all Service Level B requiJ;ementy of the ASME Boiler and Pressure Vessel Code are to be met, including the requirements regarding the fatigue usage factors for Class 1 systems.
The loads resulting from the initial actuation of the SRV's and the subsequent continuous suppression pool vibrations should be considered in your analysis for the number of cycles consistent with the 40-year design life of the WNP-2 facility.
b.
For load case 10, SRV should be assumed to be those loads resulting from the actuation of a single safety-relief valve.
110. 28 (3.9.3) 110. 29 (3.9.3) 110. 30 (3.9.3)
Provide information regarding the effects of seismic sloshing loads on'the safety-related piping and components'n accordance with the agreement between the NRC staff and the Mark II Owner's Group that this information will be provided in the applications of each individual Mark II facility.
Provide in Section 3.9.3.4 of the FSAR, the bases for the allowable buckling loads, including the allowable buckling
- stress, under faulted conditions for all ASME Class 1 component supports
.in the nuclear steam supply system (NSSS) and in the balance of plant (BOP).
Provide a comparison of the calculated loads in the reactor vessel support skirt with the critical buckling loads of the skirt under the most limiting faulted loading condition.
Oescribe your analytical techniques to determine both the calculated loads under faulted conditions and the critical buckling load of the WNP-2 support skirt.
Indicate the most limiting load combination considered in the buckling analyses of the reactor vessel support skirt.
Provide the following additional information in Section'.9.3.4 of the FSAR, regarding the operability ass'urance of snubbers:
a.
Identify and tabulate all mechanical and hydraulic snubbers installed on safety-related systems.
b.
Oescribe your methods and procedures for verifying operab-ilityy of those snubbers identified in your response to Item (a) above, during the startup test'rogram.
c.
If additional snubbers are installed after plant startup, commit to provide documentation for verifying:
(1) the operability of these additional snubbers; and (2) that these additional snubbers will not interfere with the normal operation of the WNP-2 facility.
110"8
110. 31 (3.9.3) 110. 32 (3. 9. 3) d.
Provide an inservice inspection and testing program for the snubbers',
including a discussion of the accessibility for:
(1) their maintenance; and (2) the repair and replacement of snubbers, if required.
Verify that the criteria for the design of pressure relief device stations conform with our positions in Regulatory Guide 1.67, "Installation of Overpressure Protection Devices,"
October 1973'f you adopt alternate criteria, demonstrate that your criteria provide a level of conservatism equivalent to that in the regulatory guide cited above.
In Section 3.9.3.2 of the FSAR, you indicate that active valves will be qualified for operability under seismic loading on a prototypical basis.
We agree. that a prototypical test can qualify a limited range of similar valves.
However, you do. not adequately describe the characteristics you consider in determining whether a valve is similar to the tested prototype valve and, therefore, can be qualified by analysis only.-
Accordingly, provide a discussion of how you establish the similarity of valves to a tested prototype.
This discussion should include, but not be limited to, those characteristics such as valve type,,size,
- geometry, pressure rating, stress level, manufacturer, actuator type and actuator load rating.
110. 33 (3. 6. 2)
(3. 9. 3) 110. 34 RSP (3.9.6)
For ASME Class 1,
2 and 3 components that could be exposed to either jet impingement loads or to pipe whip impact loads resulting from postulated pipe breaks in adjacent high energy piping, describe how you determine the stress levels in the targeted components.
In your response, include a discussion of the structural effects throughout the targeted system from the loads cited above (i.e., those loads associated with postulated pipe breaks) in combination with other applicable loads.
Provide assurance that the calculated stress levels are kept below the Service Level D limits of Section III of the ASME Code.
If applicable, more conservative limits on stress levels should be imposed for active components or where piping functional capability is required.
In accordance with 10 CFR 50.55a(g),
we require submittal of your program for inservice testing of ASME Class 1,
2 and 3
pumps and valves.
Our positions on this matter are presented in Section 3.9.6 of the SRP.
Appendix C to Section 110 provides a suggested'format for this submittal and includes a discussion of the information we require to justify any requests for relief from our positions on this matter.
110"9
110.35 (3.10)
(3. 9. 2)
It is not clear in the FSAR how the seismic analyses of seismic Category I electrical and mechanical equipment have taken into consideration the three components of the seismic accelerations.
Accordingly, describe how your analyses have considered the three spatial components of seismic excitation.
Regulatory Guide 1.92, Revision 1, "Combining Modal Responses and Spatial Components in Seismic
Response
Analysis," February 1976, provides methods acceptable to the staff for combining the responses to the three spatial components of seismic excitation.
110. 36 RSP (3. 9. 2)
(3. 10)
Hydrodynamic vibratory lo'adings in the suppression pool can be induced by the flow of a steam-water-air mixture into the suppression pool.
This flow may result either from actuation o'f the safety/relief valves or from a postulated pipe break.
In either
- case, the resultant vibrations in the suppression pool may affect structures, systems and components in other portions of the reactor building.
These induced vibratory loadings can be of various 'magnitudes and of various frequency content for the following load cases:
- SRV1, SRV
Accordingly, we require you to demonstrate that the electrical and mechanical equipment which is necessary to achieve and maintain a cold shutdown, are capable of performing their safety-related function under the most severe of the following combinations of seismic and vibratory loadings induced by the vibratory hydrodynamic loads in the suppression pool; a.
SRVx or SRVALL (whichever is controlling) + OBE b.
(whichever is controlling) + SSE c.
SRVADS + OB'k 4 IBA d.
SRVA S + SSE
+ IBA e.
~ DBA f.
SRVl + SSE
+ DBA Provide a commitment that all NSSS and BOP seismic Category I mechanical and electrical equipment will be qualified for the most severe combination of seismic and hydrodynamic vibratory loadings.
(Note that the applicants for operating licenses for similar facilities have stated that, in general, the load cases which include SRVR impose the most severe hydrodynamic
'ibrary loadings on safe) -related equipment.
However, this does not preclude the possibility that other hydrodynamic loads might be limiting for particular components at the WNP-2 facility.)
110. 37 (3. 10)
(3. 9. 2)
A review of the design adequacy of your safety-related elec-
- 'rical and mechanical equipment under seismic and hydro-dynamic loadings will be performed by our Seismic qualifica-tion Review Team (SgRT) during a site visit when this team will inspect and evaluate selected equipment.
(Note that Item 110.26 of this enclosure is concerned with the structural capability of safety-related components whereas this item is concerned with the functional operability of components.)
The SgRT effort will primarily be focused on the two topics listed below:
a.
A demonstration of the adequacy of the original single-axis, single-frequency tests or, alternatively, the analyses of equipment qualified in accordance with the criteria of IEEE Std. 344-1971.
b.
The qualification of equipment for the combined seismic and hydrodynamic vibratory, loadings.
The frequency of response due to this vibratory input may exceed 33 hertz and negate the original assumption of a component's rigidity in some cases.
Appendix D to Section 110 describes the SgRT and its procedures.
Section V.2.A of this appendix indicates information which we require for our review.
Several OL applicants with similar facilities have stated in their Design Assessment Closure Reports that equipment will be qualified by testing for the required respon'se spectra (RRS) representing the hydrodynamic and seismic loads combined by the SRSS methodology.
Similarly, when qualifi.ed by analysis, these applicants have indicated that the peak dynamic responses of equipment to the hydrodynamic and seismic loads will also be combined by the SRSS methodology.
For your information, we do not accept at this time, the combination of the hydrodynamic and seismic loadings using the SRSS methodology to obtain the RRS or the peak dynamic responses.
Accordingly, provide a compilation of the RRS listed below for each floor of the seismic Category I buildings:
c.
The RRS for the OBE or the SSE, whichever is controlling.
If the OBE is controlling, explain why.
d.
The controlling RRS due to hydrodynamic loads in the suppression pool.
110. 37 (3. io)
(3. 9. 2)
A review of the design adequacy of your safety-related
.el.ec-trical and mechanical equipment under seismic and hydro-dynamic loadings will be performed by our Seismic qualifica-tion Review Team (SgRT) during a site visit when this team will inspect and evaluate selected equipment.
(Note that Item 110.26 of this enclosure is concerned with the structural capability of safety-related components whereas this item is concerned with the functional operability of components;)
Ae~dingly
, provide-the-fo:Howing-additional information P~
7 a ~
A demonstration of the adequacy of the original single-axis, single-frequency tests or, alternatively, the analyses of equipment qualified in accordance with the criteria of IEEE Std. 344-1971.
b.
The qualification of equipment for the combined seismic and hydrodynamic vibratory loadings.
The frequency of response due to this vibratory input m'ay exceed 33 hertz and ne'gate the original assumption of a component's rigidity in some cases.,
Appendix D to Section 110 describes the S(RT and its procedures.
Section V.2.A of this appendix indicates information which we require for our review.
Several OL applicants with similar facilities have stated in their Design Assessment Closure Reports that equipment will be qualified by testing for the required response spectra (RRS) representing the hydrodynamic and seismic loads combined by the SRSS methodology.
Similar ly, whe'n qualified by analysis, these applicants have indicated that the peak dynamic responses of equipment to the hydrodynamic and seismic loads will also be combined by the SRSS methodology.
For your information, we do not accept at this time, the combination of the hydrodynamic and seismic loadings using the SRSS methodology to obtain the RRS or the peak dynamic responses.
Acco'rdingly, provide a
compilation of the RRS listed'below foi each floor of the seismic Category I buildings:
c.
- SSE, whichever is controlling.
If the OBE is controlling, explain why.
d.
The controlling RRS due to hydrodynamic loads in the suppression pool.
e.
Items (c) and (d) above combined using the SRSS methodology, f.
Items (c) and (d) above combined by absolute sum.
110-12
APPENDIX A TO SECTION 110 ACCEPTANCE CRITERIA I'OR MARK II PIPING SYSTEMS MECHANICAL ENGINEERING BRANCH DIVISION OF SYSTEMS SAFETY LOAD CASE SRVX SRYADS OBE SSE IBA(
)
DBA(
ACCEPTANCE CRITERIA 4
5 X
X X
x(')
X(2)
B
,(4)
,(4)
,<<4)
C(4)
(4) 10 X(2)-
B (4)
Use SBA or IBA whichever is governing.
Loading due to PBA/SBA/IBA is determined from rated steady state conditions.
(3) N - Normal load consists of pressure, dead weight, thermal h fluid reaction loads.
(.)Piping functional capability should be assured per Appendix ll0-2 or alternate means.
Service leuc)
(4) limits higher-than the level specified in this table may be used, provided piping functional capability is demonstrated.
- SBA, IBA and DBA shall include all event induced loads whichever are appl icable, Such as possible annulus (5) pressurization load, pool swell load, condensation oscillation load, chugqing load, etc.
APPENDIX B TO SECTION 110 P
INTERIM, TECHNICAL POSITION FUNCTIONAL CAPABILITY OF PASSIVE PIPING COMPONENTS MECHANICAL ENGINEERING BRANCH DIVISION OF SYSTEMS SAFETY I.
Introduction The functional capability of all piping components in essential ASME Class 1,
2 and 3 piping systems designed to Levels C or D service limits is required to be demonstrated.
Applicants may choose to use the criteria in Section II which require no further proof of functional capability.
Piping components within Section III require additional analytical or experimental proof that functional capability has been maintained.
The technical content of this position is based upon integrated experimental and analytical studies of piping system components performed at the Oak Ridge National Laboratory for the U.S. Nuclear Regulatory Commission.
The program of studies, the analytical and experimental
- results, discussions and recommendations have been
'ocumented in a report, "Evaluation of the Plastic Characteristics of Piping Products in Relation to ASME Code Criteria, ORNL/Sub-2913/8."
II.
Situations in which Functional Ca abilit is Assured without Further roof A.
Class 1 Piping Components:
Functional capability may be considered assured without further proof for any Class 1 piping component when the Level "A" or "B" or "C" limit,is used with Equation (9) of NB-3650 provided D /t < 50, where D.is the outside diameter and t is the wall Phickness o'f the IIiping component.
The Level "C" limit to be satisfied for the above verification procedure is 1.5 Sy.
A value of 8, not less than 0.5 may be used in Equation (9) for functional capability evaluation.
2.
For tees and branch connections, the Level "D" limit may be used with Equation (9) of NB-3650 without additional requirements for functional verification, provided D /t 50.
The Level "D" limit to be satisfied for the above verification procedure is 2.0 Sy.
t A value of B, not less than 0.5 may be used in Equation (9) for functional capability evaluation.
3.
Pd concurrent with Mi may be used in Equation (9).
B.
Class 2/3 Piping Components:
2.
Functional capability may be considered assured for Class,2/3 piping components for Levels "A" and "B" limits in'Equation (9) of NC-3652.
1 or ND-3652. 1 provided Do/t < 50.
For tees and branch connections, Level "C" limits may be used without additional requirements for functional verification.
However, for elbows or bends, the following addition'al requirements shall be met whenever Level "C" limits are specified:
'a)
Use (0. 8 Bq) instead of (0.75 i) but not less than l. 0.
(b)
Use (1.5 Sy) for the right-hand side of Equation (9).
In each of the above cases, D /t shall be'qual to or less than 50.
3.
Class 2/3 piping components may be evaluated as Class 1
piping components for verifying functional capability, provided the rules and limits as specified in item I'I.A.,
- above, are met.
4.
Pd concurrent with Mi may be used in Equation (9}.
III. Situations in which Functional Ca abi lit Re uires Additional Demonstration A.
Class 1 Piping Components; l.
Piping components other than tees'and branch connections, such as elbows, pipe bends and straight pipe, using Level "D" limits.
2.
Any piping components with D /T > 50.
0 B.
Class 2/3 Piping Components:
1.
Straight pipe when Level "C" limits are used.
2.
Elbows or pipe bends which cannot meet the requirements specified in item II.B.2, above, when Level "C" limits are specified.
3.
All piping components when Level "D" limits are used.
(NOTE:
The ORNL report.recommends against the use of Level "D" limits when functional capability must be maintained.)
4.
Any piping components with D /t > 50.
0 IV.
Definitions Functional Capability - Capability of piping components to deliver rated flow and retain dimensional stability when the design and service loads, and.their resulting stresses and strains, are at prescribed levels.
Piping Components - These items of a piping system, such as tees,
- elbows, bends, pipe and tubing and branch connections, constructed in accordance with the rules of Section III of the ASME Code.
Piping System -
A group of connected piping components and other associated Code components (i.e.,
- pumps, valves, vessels) performing jointly a specified plant function or, in the case of multifunctional
- systems, more than one function.
Essential Piping Systems - Piping systems which are necessary (a) for safe shutdown of the plant and to maintain the plant in a safe shutdown condition, or (b) to prevent or mitigate the consequences of an accident which could result in potential, offsite exposures exceeding the guidelines of 10 CFR Part 100.
. APPENDIX C 70 SECTION 110 hi!C Sin.--
CC.":;=iliS 0,( I)iS~R/iCE PUt'iP AND YSL>c, 7ESTI iG P~OuP~iS A'l0 REL IEr RE""ESTS The NRC sta'f, af:er reviewino a number o,
pump and valve.testing
- programs, has determined that further guidance might be helpiul to illustr>-~
~
VM the type an'x.ent of in, or.;.,ation we feei is recessary to expedite the revi w of these programs.
Ve feel that the License
- can, by incorporatinc these guidelines in.o each program submittal, reduc considerably the I
staff s review time and tim spen.
by the License in responding to HRC sta f req.ests
,or additional'n ormaticn.
The pump te:ting p, ogran should include all 'afe.y related* Class 1
~
and 3
p ;..ps v;nich are installed in v,a:er cooled ruclear pow r plants and v;hich a~e provided with an emergency power source.
The valve t sting program should in"lude all the sa ety related valves in the iollowin". systemis excludirg valves useC for operating convenien e
only, such as ianual vent, drain, instru-.ent and test valves, and valves used or maintenance only.
Pi'JR High Pressure Injection System b.
Low Pressure Inject'on System c.
Accu:-.,ul a.o.
Sys temis d.
Con'a',nri;en".
Spray System e'.
Primary and Secondary System Safety and Relief Yalves
. f.
Auxiliary fe d<<ater Systems h.
Reac.nr Building Cooling System Active Coo."onents in Service l/at r and Instrument Air Systems which are re" ired to suppor" safety system functions.
Con:ain;.en". [so!~tier Yalv s r quired to change position to i'soia e
contain;er...
Chemical '6 Yolu'e 'Control System Other key corn"oren'.s n Auxiliary Systems which are required to dir c-.]y suo<o".t plar, shu.."ov;.
oi safety system fu'iiction.
+Safety related
- necessary to safely shut Covn'the plan.
and riit'.jate the c."hseque;:ces of aii accident.
l.
Pos>dual i.=at R"....ov'. Sys:e.-.
le.. 'eac tot Ci o
< ailt Sys ie.,l b.
~ C.
e.
h.
n.
H'ich P. ess"re Ccr Injection Sys'.."m Low PI ssure Core In'ec. on S>s.e:-.
Resid.al H-.". Re~oval Svste;.i
(=hutdown Cool ina Syste.;,)
E',";sr." ncy Cond ns r Sys'.=-;-.i (Is"la:ion Cond n
r Syst v,)
LG'K Pr ssure Col e S~."ay Sys e.
Con a
i r.."el.t Spl ay Sys '.e.".1 q-r~~.,
P-s.'-=
-)C i:= <c~>>i "~l ~ er ii lv vCI v) i am i
~ Cl i G ~
vv
~
i~/s
~
i I
i4 RC IC (Reac.or Core Iso; a tiion Co 1 'i. g )
ys.e.
Co. tcir.:.-I Cccl inc Svs..-.i Contai,".'nt isola.ion valves r quir d to chanoe position to isola-.=
cor,tb
.".;.. n SI Gn.'.b" 1 icuid cc."trol sJ stei
('oron System)
Au-.c;..a-ic Depressuri:.at',on S stem (any pilot or control valves',
ass cia=a hy"~ aulic or pneum:at'c sys.e...s, e.c.)
C"ri.roi Rod Drive Hydrau'i'c Sys".e;.. ("Sera.<<" furc"ion) 0trer i:ey coaoonen:s ir Auxiliary Syster.s which are required to Cirec.lg'u"port plar,t shutd":m or safety yste.a function.
o.
Reac.or-Cco i.ant Systen Ins:, vice P..-.,'>>.".d Valve Tes ir-orr crag Infor-...a'on require"'for iiRC Staff Review c7 the Pu;ip and Yalve Tes
'.,"c Pro" ra' A.
Th.eo se'ts of P~ID's, which irclud all of the svs. ns lis-. "
p above wi"h:he code class and syste.",
boundaries clearlv rarked.
The dr awins shouid inclu all of the components presen at th=
i sub@ i
=. ard a
1 ece~<
oc
~a PliID s). 'l Is ~
B.
Iden.iiica".ion.ol 7, se applicable AS.".= Code =di:icn and Adcenda
" C.
D.
The period for which the pro",."as is applicabie.
Identi-y the coi-..ponent co e class,
E.
For Pu;.p testing; ICen.i y l.
Each pump recuired to be -"stad (nz".a and number) 2.
Th test para."...a.ers to be r:.
sur="
3.
The tes.
r c: ency For valve tes:."r.":
Id"nti;y l.
f-ch valve in AS.".f Sec".ior, XI Cate"vries A 6 8 that will 2.
3.
be exercised every.hr ronths Curing nonaal plant op raticn (indicat wh ther partial or full s.roke exercis and for "ower cp rat d valves list th li,",iiting value Tol stroke tel"..e, )
fa h valve in A'l'.E Sec:icn XI Cate"ory A tha. will be leak tested duriti~ refuelir".u'; ges (Irdicate "e leak:est procedure ycu inten" t" us=-)
Each vzl ve in AS)',E Se io" Xi Cate ories C,
D and E Pat will be tested, the type cf test ar.'l e test f 4)uency.
For check valv"=s, identify those
.hat will be exercised ever" 3 renths and those that will orly be exercis d.durirg cold shutco:;n or refuel ir>c outaaes.
II.
Additional Infer.ation That ~ill E'c Help-ul in Speeding Up the Pevie;v Process Inc 1 u e the val v 1 o i On coord inat 5 ol c
">el a Dpl opl iat lo a.icn inform;:a.i"." which wi';1 ex-edite our avoca-.i,ng the 8.
C.
valve.s on the PEIOs.
Provid PEIO drawings that are larg and clear enouch to b read easily.
Identi v valves ".h-'re provided with an irt=rlcck to other co
. Dc/'ts and a brie f Cescr i p-.ion o, tha ~ function, Rel >'ef geeues.s ircn 5ection XI 2ecuir<i-.ents The lar".est area o-,.co.". em I
ins rvice val "e ar C pu.-..p t s=ing jus.i-.y'n", rel '-:
rc'-,i 5 cti "n XI s:, f, in the review o, an fcr '",e ilk" pr ". a.-, is,in eva qua ino th basis for It has b:-en our. ex-'ri..-.ce Reoui, 2...ents.
~'> t I 'ny recues".
Qr re 1 e
sui
>. 1 t in t.ese pl ool a,.s Co n>o
>>i ovide a"equa".e descri"".':ve and C ta >1 d t8 hrical i>rior aticn, 7his expi
>"it information is necess'ry to provice reasona"le assurance that the burden i",',"os d on the'icensee in cc.".."1yin," with.h coCe r quir ".,ents is not iust fied by the ncr =se" 1
ve1 of safety c'a ired.
Relic> requests which ar su
-1t d wi h a justification such as "Inpractical", "Inaccessible",
or any other cate,orical basis, will requir additional infor...a.ion, as illus.rated in tf o enclosed exa,".,pies,
.o allcw our sta, f to rake an evaluation o
that relic, request.
The intention of-this cu >dance is.o V.lustra e the content and by the tiRC s.aff in tie r " es-
~or relief> to ext nt of information requi f.'.aLe a prep r eval u
>on an
=-"'.=':v Coc~ent '"~ "as'>s for th>a-. relief in our safa.y eva",u>a,"n r por..
Tho h.
C staff "
ls tha:
by r
eivino I
progr'" su".;:.'t.al, sJbsequent reques s
>or ad"i this info:".;ation in the tioral infcrration an" delays in co-...pic..in'ar review c~n be cor sid=rably reduced or elininate.".
Ir,for.-.ation Recuir d for N"-C I eview o; >:-'";ef P.o u.=t~
A.
I'tifycc.-..pcnent for which r lie. is request~'d:
I.
t(are arid nur..". r as aiven in FSAR 2.
Function 3.
4.
Fot valve tes.in.-.,
also s"ec> J
>h AS>"
tio> X'alve C.
cateoory as defin o in IWY-2000 ec'fically iden ifv the AS,.',r Code require>".. nt that has been d temined to be impract:cal for each corpcncnt..
Provido info....ation to su~port:he determination reou
~ re; ent in (2) 1 5 il"practic 1; i. e.,
sta.
and ezpl a in the b=s',s
-,or requestinc rel'.ef.
D.
Specify he inserv'.'ce testin>"
tn>a will be "e'r or.ad in lie'.
'of tt o AS>'.:-
Code Section XI requir."ants.
Provi"'e the sch>ed le fcr ir>"ienenta'.i."." o:he proceCur(.,'n (Dj.
-110 C-4
""-a--',as to illustrate ~~viral rossible Are>>s khere Relic flay B Cranted an t!,e Extent and Conten. of Infor,.at'.'on Necessary to Yiake An Evaluatirn A.
nccessibi i ity:
I'2
!~adulation speci fical ly grants r lief
.ro:; tne code requir=..: n. because of insi.f ic',ent access pro-visiic;is.
- However, a detailed disc ssicn o, actual physic=-1 arranc r, nt of the component in questicn to illustrat the insuf-, i"iency of s-ace for conduct ing the required test is ne essary.
Oiscush in detail tie physical arrange.".,ent of the component in cuesticn to de-:.,onstrate ihat -~ere is not su,-icient."ace to o-.rfor t,".e code requ',re" irservice :esting B,
ilhat alterna
.ive surveillance
.means whic! will provide an acceptable level of sa.eiy have you ccnsiC red and why are thesfe reans not;=->>sible?
Environmen:al ConCit ons (o.g.,
High radiaticn level, igh temper~",ture, fiigh hu;.,idi;y, tc.)
Althouch it is pr; crt to ra'.ntain occup tier, radia:icn ex"".s re for'inspection personnel as low as pract cab'.e, the r quest cr r lie from the code r quir="ments canno.
be cranted sclely cn th b>>sis cf high radia.ion level's alon A balanced Judc~n:
between the hards.'iips and cc.-,"ensating increase in the level of sa,eiy should be carefull.v establishe '.
If the health and sa.ty of the public dic.a.es
.h necessi".y of inservice testing alternative romans or even decontamiination oi tho plan.
I if necessary should be provided or develcpad.
fI Provide additicnal infor...ation regardinc the radia.
cn levels 4
at tI:e required test lcca.ion.
Nh'at alternative t sting t c..i;.""es w! ich will provic an acc ptable level of assuranc of.h I
~ ~
+
t ea i
i Ca~>>
'":nteority of the co~"."."en-. in c.esticn have 4'ciJ conside, e
why are these tochniques Ca'.ermine" to be imprac.ical?
5'1 C.
D.
Inst-..:.-an;ation is nc.
r'g'.ra;ly pr "viC "
Prov Ce in -"r.-..a "n to Ju v "i:a co.31 iance wi.h the co"e re-u'r='-.=n.
'..":;;" resu'.
in Jn" e burden or hard+" ips wi:hcu.
a co;.'cosa.
ng incr 5
in the lfevel o",
pl nt sa.ety.
t'hat a'i terra'.ive tes..",no;etho s>>iich will pr"vide an acce~:able level o
sa e y have.you
- o. 5 Car C anC wry are tiese f7=hods detef-..fin~d to be -'.~"rac:ical?
Valve Cycling Durin", Plan; Cperation Could Put the Plant in an Uns ;
Condition The licen:-ee sh""lo explain in Cetailf l;hy exer"ising t
.s Curin, plant of".erati n coulfd 'eooard ze the plarit safety.
Valve Testir.g a.
Colfd ShJ;"-'o: n cr P, fuelino Intervals
',n 1ieu o~ t"~ 3 "oo f'n~"i-~~ '."
="v='.
licensee shc.lC x."lain in c.tail why each valve c nn" be exercised C".rin,, ror;..al c"er"iicn.
- Also, or.he valves where a re-ueling in: r,al
'. 5 ind i "a:e~ exp I a
.'1 in d all >>"lv each valve caf;not be exer"iseC Cur no cold shutco>>n intervals.
III. Acceo-ance The C 1'="la io" <~ f1~
i ~""~s.
gars g << i(i c'
'f i gss
~
. 1
~ i co. )65 i
0 t
~
code re-uir ~. -.s woi ld resul:
-in 2.
hardships or unusJ.'-1 difficulties wi=hoJt a co;.ipensa".ing in"rease in the level o-. safety and coflco:pliance will provide an ~'ccep:a"le level oi cual ity and sa e.y, or Pro,".o'sed al terna:ives to trie co'equirer;.ents or portions thereof will pr".;i-'e Safety.
an ac e",tabl 1;el of quali:y and S'an~+sr 1 Av' A stari"-rC for the valve p."rtion o; the p ;..p and valve testir,"
proor=-;", an" rel f PO
~ 1 I
es 5)
~
J 5
iilC f u "
5 afi a.
<<i"ieseerit ZO thlS Guidanoe.
by both Sta ho f
~ aves
~
~
TT lfl ou e~Si >)r >
5 i
~ I>>
~ 'v r review anC fbv:h 1 ice.,
w i lf lf reCuce s
in their the t' sveni PP'paf a
l ofl,.
o=
he pu.-:ip a.".0 ',elve:es'.ing "ro"ra;, er.." sub;,i". als.
7he stanCarC for..".at inclu"es exa.-..".les o= re 1 i e i rendu' t s
'::?: ich are intenCed to i 1 1 us. ra te relief rendu st.
the stan 0
Tors 2nc a1 e no necesseri1 j "p. spe ifIc 3 la.
0
l I p
~ J Al it'c.a;.
Vial ve
!,'uiiP>>er I/l I/i
~d I/)
PJ
~
I E3 C)O L'al ve C0 te(i,ol )//
I/I Ol EJ I
O gJ IUl
~)0 r.
CJ I
~
4)
II/lA CL I/l
)J l)I III J
C.
n) 0 In 0)Dcr i))
C)
Ii I;)
C)
CV III n>
C.
1iii
) )
C7l wI/l I))
l BEY)nl(KS (Hot to be used for relief basis)
. 710 700
/17 702C 3
3 3
3 3
3 p-ln 0
iC)
C-15 C-l5 E-14 0-11 X
X X
6 16 16 3
cn CK CK IICL GL sn snl C
CY CV CV H
LO fT Iln C
DT X
CS X
fT 60 sec. 7220 722C 715 729 7440 3
3 2
2 2
X X
X Bin 3/)', I 3
10 ItEL IlfL rfL sn C(
)/0 Sl(V SllV SRV C
I LT 30 sec.
(.cool;d i o, Va1 ve jes,inn xa e polio+
Exercise valve (full stroke) for o"era"ili".y everv (3)
~g,on hs LT -
Yalv s ar i
ak tes.
d per S c.icn Xi 'r:icle I'ilY-.3'20 HI S I ro~e ti, -=
..e=-sure."-.,en. s a r taken and cca" a red to tha stroke tire 1
, liriitirg val e
p r Sec='."ri XI Article InV Sw10 CV -
Exercise c.",e.=k valves:o the posi:ion r quired to fulfilltheir fu..".ion every (3);,onths SRV-Dj-ET-Safety arC rel.ie valves ar tested per Sac ion XI Article I'tlV-3510 Test ca.egory D valves per Sectior, XI Art cle IV!-3600 Veri=y and r.cord valve position.b= or operations are performed ard el 0"'2 ions are cc;.,pl eted, and v rif r hat val ve is locked or sea'i ec.
CS Exercise valve for operability every cold s1U dc'ivn Exercis va "
Gl Gp rab'iity ev.ry reactor refu ling 110C-10
R~iief Req ~s'asis Auxiliary Coolant
- System, Componert Cooling Valve:
717 Category:
Class:
Function:
Imprac tica 1 Prevent backflow from the reactor coolant pump cooling coils test requirement:
Exercise valve for operability every three months Basis for relief; To test this valve would require interruption of Alterna" i ve Testing:
Valve:
Category; cooling water to the reactor coolant pumps motor cooling coils.
This ac ion could resul t in damage to the reactor coolant pumps and thus place the plant in an unsafe mode of operation.
This valve will be,ex rcised for operability during cold shutdowns.
834 8-E Class:
Function:
3 Isolate the primar'y water from the component cooling surge tank during plant operation.
It is normally in the closed position, but routine operation of this valve will occur during refueling an'd cold shutdowns.-
Impractical 'Test'xercise valve {full stroke) for operability Requirement:
every three (3) months.
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 link contamination.
'1(er'nate Verify and record valve position before and Testing:
and after each valve operation.
Valve:
Cate('cry:
Class; Func.icn:
7$ 8B Isolat the I ~sidval heat xch-n--"rs fr..-..:.';e ccl" le0 R.C,S, b'"
kT lc'"'nG a"
- u. Jl at'icr back 1
'>'( ~
7est Requirer~nts:Seat leakage test Basis for This valve, is located in a hich raci tion field R lief:
(2000 nr/hz) which would make:he r guirec sea-.
leakage (es hazardous to tes. perscnnel.
he in.er;" tn sea".
le>>'. te"t twc o.'.",er arid 8763) which are in series wi h and
>; ill al so pre Ye1" backf'low,
)ie Ya 1 ves (G7 th1 s Ya! v"-
c>>o 1 I
1 4 IC by co...plyinc wi h tho i.. will not a hieve a
the lev i of sa ety.
seat 12>>k?ce cc.-."en sa to ry r "uir -.en:s incr ?se in n 1 te rn?. t 1 ve Testino:
Iio alterna ive seat leak testing s ~roocse~
APPENDIX D TO SECTION 110 SEISMIC QUALIFICATION REVIEW TEAYi {SgRT)
Interim Procedures I.
SCOPE
'ORTtasks include both generic and site specific reviews.
Generic reviews cover equipment supplied by the NSSS and A/E common to more than one plant.
Specific plant reviews as delineated in the Standard Review. Plan Sections 3.9.2 and 3.10 will be supplemented by SgRT site visits and evaluation.
II.
OBJECTIVES SgRT is a group of NRC.staff members established to conduct reviews of the design adequacy of safety related mechanical components, instrumentation and control equipment, and their supporting structures for various vibratory loads.
S(}RT is charged with accomplishing the following three tasks.
l.
Determine the design adequacy of mechanical and electrical components and their supports for the required vibratory loading conditions which include:
(a)
Seismic
,(b) hydrodynamic (as applicable)
{c) explosive
{as applicable)
(d) other vibratory inputs from the operating environment (as applicable)
(e) appropriate combinations of the above events.
2.
Changes in seismic qualification criteria, such as the revision of IEEE Std. 344 and other IEEE Standards, and,the issuance of Regulatory Guides 1.100 and 1.89 require that the staff verify:
(a)
For c'.der plants having components qualified by previous criteria; that components have adequate margin to perform their intended design functions during and after a seismic event.
{b)
For new plant applications; that there has been uniformity and consistency in implementing the current criteria.
3.
In the case of plants which have design basis seismic ground motion levels and/or other required vibratory'oads increased, review to assure adequate design'margin exists at the revised levels.
I I I ~
GENERAL CRITERIA The bases used by the staff to determine the acceptability of equipment qualification will be IEEE Std. 344-1975 as supplemented by Regulatory Guides 1.100 and 1.92, and Standard Review Plan Sections 3.9.2 and 3.10.
IY.
GENERAL PROCEOURES SgRT will conduct generic and plant specific reviews:
1.
Generic reviews will be conducted of all NSSS vendors and most architect engineers (major equipment vendors and testing laboratories may be included if necessary) to assure proper interpretation and implementation of the current equipment qualification criteria applied 1100-2
to plants applying for construction permits and operating licenses.
2.
A plant specific equipment qualification review will be conducted of each plant now undergoing licensing review having compo. ants qualified to criteria different from current requirements.
A.
For components having multi-plant application (such as those within the scope of an NSSS vendor),
an equipment qualification review at specific sites will provide generic qualifications.
B.
For components which have only specific plant application (mostly those within the scope of the BOP supply),
an equipment qualification review at specific sites will provide site-specific qualifications.
3.
Equipment qualification review for plants with revised increased design basis seismic ground motion levels and/or other required vibratory loads will be conducted on a plant by plant basis.
V.
SPECIFIC PROCEDURES SgRT procedures provide for both generic discussion meetings and plant site visits.
1.
Generic Discussion Meeting:
To implement the generic review specified in IV.1 and IV.2.A, a generic discussion meeting will be held to discuss the following:
A.
Heeting Agenda Heeting Objectives by SgRT 110D-3
F.
SgRT concludes the meeting and specifies the follow-up'tems.
N 2.
Plant Site Reviews:
To implement plant specific equipment qualificacion reviews specified in IV.2 above, on-site inspection of equipment and supporting structures in question is required.
Site visits generally follow the following procedures'.
A.
Pre-visit information submission:
The applicant= (plant owner) receives initial i nformation concerning the intended visit, and should subsequently submit two summary equipment lists (one for NSSS supplied equipment and one for BOP supplied equipment).
These lists should include all safety related mechanical components, instrumentation, and control equipment, including valve actuators and other appurtenances of active pumps and valves.
In the lists, the following information should be specified for each item of equipment:
(1)
Method of qualification used:
(a)
Analysis or test (b) If by test, describe whether it was a sing'le or multi-frequency test and whether input was single axis or bi-axial.
(c)
If by analysis, describe whether static or dynamic, llOD-5
single or multiple-axis analysis was used.
Present natural frequency of equipment.
(2)
Indicate whether the equipment is required for:
(a) hot stand-by (b) cold shutdown (c) both (d) neither The scenario to be considered for this determination is:
(i)
SSE or OBE, with coincident (ii) loss of offsite power, and (iii) assumption of any single active failure.
(3)
Location of equipment, i.e., building, elevation.
(4)
Availability for inspection
( Is. the equipment already installed at the plant site?)
(5)
Provide a description of how cold shutdown is reached using the equipment in item (2) above.
B.
SgRT screens the above information and decides which items will be evaluated during our forthcoming site visit; The applicant 110D-6
will be informed of these items and will be expected to 'submit two weeks prior to the visit an equipment qualification summary as shown on pages 10-12 for each of the selected items.
C.
A brief meeting is held at the beginning of a site visit with the following agenda:
(1)
SgRT explains the objectives of the site visit and procedures to conduct equipment inspection.
(2)
Utility personnel or their designees present an overview of the seismic qualification program conducted.
(3)
The seismic qualification of certain specified items may be discussed as necessary.
(4)
SgRT specifies items that need to be inspected.
D.
SgRT conducts'nspection of specified items.
SgRT describes findings of the inspection.
F.
General discussion.
G.
SgRT concludes the visit and specifies needed information and the follow-up actions.
3.
After each visit SgRT will issue a trip report, which identifies findings, conclusions and follow-up -items.
Status reports may be issued as necessary.
The site review will include the issuance of
an Evaluation Report. for the specific plant.
Generic evaluations will be referenced to the NSSS vendor or A/E.
VI.
RESPONSIBILITIES OF NRC PARTICIPANTS:
A.
The Seismic Qualification Review Team consists of members of the Mechanical Engineering Branch (HEB), the Instrumentation and Control Systems Branch (ICSB),
and the Power Systems Branch (PSB).
One additional member from MEB will join the team when a review of a specific plant is going to be conducted.
This member will be the reviewer of the plant.
The Team Leader is responsible for scheduling actions, coordinating staff positions, and contacting appropriate authorities for work.
assignments to each member.
Ke reports to the HEB Branch Chief regarding the progress of SQRT performance.
He'ill set up necessary
't contacts for generic reviews and will contact project management for specific plant site visits.
He will specify the meeting objectives and concludes meetings.
The HEB members and Team Leader are responsible for reviewing L
assigned equipment qualifications in the area of responsibility of the Mechanical Engineering Branch, including the methods and procedures used in test and analysis.
Members representing the Power Systems Branch (PSB) and the Instrumentation h Control Systems Branch
( ICSB) are responsible for reviewing assigned equipment qualification in the area of responsibility of
.1100-8
0
their branch, including equipment signal interpretations for functional verification.
They serve as a liaison between SQRT and.ICSB and PSB.
All members shall present their opinion and professional judgement to the Team Leader in order to arrive at consistent and uniform SQRT positions.
B.
The MEB, PSB, and ICSB project reviewers will be advised of SQRT activities which relate to specific plants.
The MEB project reviewer is responsible for evaluating the impact of SQRT activity on the specific plant review and for taking appropriate action to include pertinent information in the plant safety evaluation.
The HEB project reviewer is expected'o partici pate in the site visit and attend pertinent generic meetings
'as necessary.
=
The DPH project manager, after being informed of the intended plant visit, is expected to contact the applicant and arrange for the visit.
The project manager serves as'a liaison between the SQRT and the applicant.
C.
Generic meetings will be arranged by the SQRT or via the DPM generic project manager if one is assigned.
D.
Representatives from IAE Regional Offices and other interested or'g'ariifational groups within NRC are welcome to attend either generic meetings or plant site visits as observers.
The SQRT should be informed of expected attendance at such meetings or site visits.
llOD-9
Oualification Sunmar of E ui ment I.
Plant Name:
1.
Utility:
2.
NSSS:
3.
~Te:
PWR BMR 1.
Scope:
[ ] HSSS 2.
Hodel Number:
3.
Vendor:
[ ] BDP l}uantity:
4.
If the component is a cabinet or panel, name and model No. of the devices included:
5.
Physical Description a.
Appearance b.
Dimensions c.
.Weight 6.
Location:
Building:
Elevation:
7.
Field Mounting Conditions L
3 Bolt (No., Size l
.f j 8.
Natural Frequencies in Each Direction (Side/Side, Front/Back, Vertical):
'/S:
F/B:
Y'.
a.
Functional
Description:
b.
Is the equipment required for [ ]
Hot Standby
[ ]
Cold Shutdown
[ ]
Both 10.
Pertinent Reference Design Specifications:
III.
Is E ui ment Available for Ins ection in the Pl'ant:
[ ]
Yes IY.
E ui ment uglification Method:
Test:
Analysi s:
Combination of Test and Analysis:
Test and/or Analysis by
[]
No name of Company or Laboratory 8 Report No.
1.
Loads considered 1.[ ]
Seismic only 2.[ ]
Hydrodynamic only 3.[ ]
Explosive onlt 4.[ ]
Other (Specify) 5.[ ]
Combination of 6.
Method of combini,ng RRS:
[ ] Absolute Sum
[ ] SRSS [ ]
2.
Required
Response
Spectra (attach the graphs):
3.
Required Acceleration in Each Direction:
S/S
=
F/B =
YI. If uglification b Test then Com lete:
[ ] random l.
[ ] Single Frequency
[ ] Multi-Frequency:
[ ] Sine beat
[]
2.
[ ] Single Axis
[ ] Multi-Axis 3.
No. of Qualification Tests:
Frequency Rang'e:
5.
TRS enveloping RRS using Multi-Frequency Test [ ] Yes (Plot TRS on RRS graphs)
[]No 6.
Input g-level Test at S/S
=
F/B
=
Y =
2.
Laboratory Mounting:
1.
[
3 go1t (go.,
Size
)
[ I geld (Length
)
[
1 8.
Functional operability verified:
[ ] Yes
[ ] No
[ ] Not Applicable 9.. Test Results including modifications made 10.
Other tests performed (such as fragility test, including resul.ts):
II YII. If uglification b Anal sis or b the Combination of Test and Anal sis then
~Com 1 etc:
l.
Description of Test including Results:
2.
Hethod of Analysis:
[ ] Static Analysis
[ ] Equivalent Static Analysis
[ ] Dynamic Analysis: [ ] Time-Hi story
[ ] Response Spectrum 3.
Hodel'ype:
[ ] 3D
[ ] 2D
[]1D
[ ] Finite Element
[ ] Beam 4.
[ ] Computer Codes:
Frequency Range and No. of modes considered:
[ ] Hand Calculations
[ ] Closed Form Solution 5.
= Hethod of Combining Dynamic Responses:
[ ] Absolute Sum
[ ] SRSS
[ ]Other:
specs y
6.
Damping:
Basis for the damping used:
7.
Support Considerations in the model:
8.
Critical Structural Elements:
Governing A.
Identification Location Res onse Combination Seismic Total Stress Stress Stress Allowable B.
Hax. Deflection Location Effect Upon Functional 0 er abi1 it 110D-12
4 1