ML18026A219
| ML18026A219 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 08/28/1980 |
| From: | Curtis N PENNSYLVANIA POWER & LIGHT CO. |
| To: | Youngblood B Office of Nuclear Reactor Regulation |
| References | |
| PLA-536, NUDOCS 8009080364 | |
| Download: ML18026A219 (147) | |
Text
O'PIT,IL TWO NORTH NINTH STREET, ALLENTOWN, PA.
18101 PHONEs (215) 821-5151 August 28, 1980, Mr. B. J. Youngblood, Chief Licensing Branch No.
1 Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC 20555 Docket Nos. 50-387 50-388 SUSQUEHANNA STEAM EIECTRIC STATION
RESPONSE
TO DRAFT SER DATED 4/21/80 AND MEETINGS ON JULY 14, 15, 16, 1980 ER 100450 FIIE 841-2 PLA-536 s
Dear Mr. Youngblood:
The purpose of this letter is to document the agreements reached and the actions t'o be taken as a result of the meeting held wi.th the NRC staff to resolve open items in FSAR Sections 3.6 through 3.10 inclusive on July 14 through 16, 1980.
1.
SER Pa e
2 Para ra h 3:
. Break opening times should be assumed to be one millisecond.
Break opening times greater than one millisecond must be justified by experimental data or analytical theory.
A ta'ble of break opening times or assurance that times of one millisecond or less were used is necessary for completeness.
RESPONSE
A break opening time of.one millisecond or less is assumed in the analysis for pipe break effects.
ACTION:
Revise FSAR Subsection 3.6.2.2.2.1 as shown in Attachment A to this letter.
2.
SER Pa e 2 Para ra h 4:
a In order for us to complete our review, the applicant should provide isometric sketches which show the locations of postulated breaks in high energy piping within the drywell.
The applicant'as not yet
///
responded to our Question 110.26 which requested some of this
>~~L/,)o'nformation.
We consider this an open.issue.,
~ Qg <<kn R., 61~.A~ <'~-g ao o gosh'6(
PENNSYLVANIA POWER 8
L IGH7 COMPANY
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H H l l
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Mr. B. J. Youngblood, Chief August 28, 1980 Page 2
RESPONSE
See response to Question 110.26 (Attachment B).
Also there are no "no break zones" for CS and RHR supply and return (Figures 3.6-5 through 3.6-8).
ACTION:
Revise the FSAR as shown in Attachment B.
3.
SER Pa e 2 Para ra h 5:
Another open issue exists in the applicant's analysis for the effects of pipe breaks in non-seismic Category I piping.
The applicant has assumed that this piping will fail under the effects of the SSE.
Further assurances must be provided to show that breaks or cracks in these fluid systems have been assumed in the worst case locations.
RESPONSE
For non-seismic Category I piping, only one pipe break at a time is postulated to occur concurrent with the SSE.
In non-seismic Category I high energy piping, break or cracks are assumed at the worst locations in evaluating the effects of pipe failure.
In non-seismic Category I moderate energy piping, cracks are assumed at the worst locations in evaluating the effects of cracks.
The break and crack locations for non-seismic Category I piping are consistent with Standard Review Plan 3.6.2 and Branch Technical Po'sition MEB 3-1, as described in FSAR Subsection 3.6.2.1.1.
ACTION:
Revise FSAR Subsection 3.6.2 as shown in Attachment C.
4.
SER Pa e
3 Para ra h 1:
Pipe whip and get impin ener lines havin res gement need only be considered in those high gy g
ervoirs with sufficient capacity to develop a
jet stream.
Our review cannot be completed until information showing how reservoirs having sufficient capacity to develop a jet stream are identified and a list of these systems are provided.
Mr. B. J. Youngblood, Chief August 28, 1980 Page 3
RESPONSE
Some of the high energy fluid system piping do not have any flow during plant normal,and upset operating conditions.
These lines have either a check valve or a normally closed valve in the system.
Only that portion of the piping between the RPV and the check valve of the normally closed valve is considered to be high energy system.
For a postulated pipe break in the high energy portion of the system, that portion of the piping towards the normally closed valve, is considered not to have fluid energy reservoir with sufficient capacity to cause pipe whip.
Table 3.6-13 lists these systems and the reservoirs with and without sufficient capacity to develop a jet stream.
High Energy Piping Fluid Reser-voir Breaking Location ACTION:
Assumed Reservoir h This side is considered not to be reservoir with sufficient capacity to cause pipe whip Revise FSAR Section 3.6 as shown in Attachment D.
5.
.SER Pa e
3 Para ra h 2:
The applicant has stated that.
a design basis for the Susquehanna SES is that a postulated pipe break inside, containment (up to and including a rupture of the recirculation piping), in conjunction with the SSE, plus a single failure will not prevent the plant from obtaining cold shutdown with no loss of containment integrity and no dose levels above the limits in 10CFR100.
The applicant further states that piping not designed to seismic Category I standards is assumed to fail under the effects of the SSE.
In the FSAR Section 3.6 the applicant, states that pipe breaks inside and outside of containment are not postulated to occur simultaneously.
This statement appears to be in conflict with the stated design basis.
This inconsistency must be resolved.
I
~ 4 r
it 0
Mr. B. J. Youngblood, Chief August 28, 1980 Page 4
RESPONSE
BTP-APCSB 3-1 paragraph B.3.a., titled Anal sis and Effects of Postulated Pi in Failures states that "...each...break in high-energy fluid system piping or leakage crack in moderate-energy fluid system piping should be considered separately as a single postulat'ed initial event occurring during normal plant conditions."
For purposes of piping failure analysis only one 'initial event is postulated during normal plant conditions.
That is, for any single postulated seismic Category I or II pipe break, the plant is assumed to be i.n a normal plant condition (no SSE) with only that 'single postulated pipe break.
For conservatism and defense-in-depth, Susquehanna SES has committed to the design basis stated in Subsection 3.6.1.1 of the FSAR that for inside the containment pipe breaks would be postulated in conjunction with an SSE.
But once again for this particular case, only a single pipe break (the LOCA) was considered in the evaluation of the piping failure.
It should be, pointed out that the combination of a LOCA and an SSE is used as a loading combination for the design of systems, components and structures required to bring about a safe shutdown.
But except as noted above, this combination is not used to analyze the effects of postulated piping breaks.
In all cases the design basis includes the requirement to be able to bring the reactor to a safe shutdown and maintain it in a safe shutdown condition.
For additional -information on the failure of non-seismic Category I (Category II) piping, see the response to SER Page 2, Paragraph 5.
ACTION:
Revise Subsection 3.6.1.1 and Question 110.18 as shown in Attachment E.
6.
SER Pa e 3 Para ra h 3:
The applicant has installed 'pipe whip restraints to prevent ruptured pipes from whipping into and damaging nearby safety-related equipment.
These pipe whip restraints are designed to withstand the resultant loads and remain intact to assure protection of nearby safety-related structures, systems and components.
The only area requiring further clarification is the level of plastic strain allowed in the restraint under loads due to the whipping pipe.
. Mr. B. J. Youngblood, Chief August 28, 1980 Page 5
RESPONSE
Subsection '3.6.2.2.1 has been modified to provide clarification (Attachment F).
ACTION:
Revise FSAR Subsection 3.6.2.2.1 as shown in Attachment F.
7.
SER Pa e 4 Para ra h 1:
In those piping systems in which breaks are not postulated, an augmented inservice inspection program that commits to 100$ volumetric examination of all welds in these systems is required.
The proposed augmented inspection program does not meet those requirements for ASME Class 1 piping or for weld-o-lets, half couplings, and socket welds.
Further discussions are required before the augmented inspection program can be accepted.
Additionally, credit for examination of these welds cannot be applied to the general examination of welds in the plant.
RESPONSE
See revised Subsection 6.6.8 (Attachment G).
ACTION:
Revise FSAR Subsection 6.6.8 as shown in Attachment G.
8.
SER Pa e 6 Para ra h 1:
If different excitations were present at different anchor and support points, the response spectrum analysis was performed using the response spectrum at or above the center of mass of the piping system.
This tends toward conservative results since the response spectrum increases with building height.
Relative displacement between anchor points was determined from the dynamic analysis of the associated structure.
CLARIFICATION OF THE STATEMENT:
Add the following to the above referenced paragraph:
Alternately, the multiple excitation analysis methods may be used where separate acceleration time histories or response spectrum were applied to each piping system support points.
Mr. B. J. Youngblood, Chief August 28, l980 Page 6
ACTION:
The NRC staff will add statement to SER.
9.
SER Pa e
7 Para ra h 1:
Standard Review Plan Section 3.7.3, "Seismic Subsystem Analysis",
requires 5 OBE's with a minimum of 10 cycles each to be utilized in fatigue evaluation.
This requirement has not been met.
The applicant must justify this deviation from Standa'rd Review Plan 3.7.3 or commit to meet our requirements.
RESPONSE
One OBE intensity earthquake with 10 peak stress cycles is postulated during the life of the plant for fatigue evaluation.
This position was approved as a licensing basis on other plants based on the following justification.
A BVR dynamic model subjected to 3 different recorded time histories and modal responses truncated to study the response of three different frequency bandwidths (0-10 Hz, 10-20 Hz and 20-50 Hz) was analyzed.
This study showed that during a 40 year life, the probability of one OBE with 50/ of the SSE intensity is extremely remote.
It takes 20 quarter-SSE's, which are more realistic to produce the same level of stress of one OBE.
Therefore, to cover the combined effects of these earthquakes and the cumulative effects of even lesser earthquakes, one OBE intensity earthquake is postulated for fatigue evaluation.
In addition, the number of stress cycles between one-half peak stress and full peak stress is less than 4g.
Therefore, the assumption of 10 peak stress cycles provides an added margin of conservatism.
ACTI0N:
No action, is required since one OBE with 10 cycles is acceptable for the Susquehanna SES fatigue evaluation.
10.
SER Pa e
7 Para ra h 2:
Regulatory Guide 1.92, "Combining Modal Responses and Spatial Components in Seismic Response Analysis," outlines the procedures for combining modal responses.
Specifically, modes having frequencies falling within 10/ of each other are defined as closely spaced modes and must be combined by the absolute sum method.
Our review of CESAR Section 3.7.3 cannot be completed until assurance is provided that this criteria has been met or that an equivalent level of safety has been achieved.
It should also be noted that Bechtel topical report BP-TOP-l, Revision 2, has not been accepted.
Revision 3 of this
Mr. B. J. Youngblood, Chief August 28, 1980 Page 7
report has been accepted and, therefore, the FSAR should be clarified to reflect the differences between the two versions.
RESPONSE
(NON-NSSS):
Compliance with Regulatory Guide 1.92 "Combining Modal Responses and Spatial Components in Seismic Response Analysis",
was not a design basis requirement for the construction permit for Susquehanna SES.
However, the criteria employed in the seismic analysis of the piping system is conservative and provides an equivalent level of safety as that with the criteria currently acceptable to NRC staff.
The criteria employed in the seismic analysis of the piping systems for Susquehanna SES is described below:
l.
Damping values of 1/2g for OBE and lg for SSE are used.
Regulatory Guide 1.61 permits use of higher damping values (lg and 2/ for OBE, and 2g and 3g for SSE).
2.
Seismic analysis is based on the envelope of response spectra at different anchors or supports in piping system.
(This is more conservative than using the response spectrum at or above the center of mars of the piping system.
The approach of using the response spectra at or above the center of mass of the piping system has been acceptable to NRC staff =for the plants of the vintage as Susquehanna, as stated previously in this SER.)
3.
Piping system response to three spatial components of motion are combined by the SRSS method (as, required by Regulatory Guide 1 92 ~ Section C, 2).
(For a plant of the vintage as Susquehanna, combination of the largest horizontal responses and verti'cal response, using absolute summation method has been acceptable to NRC staff, as stated previously in this SER.)
BP-TOP-l, Revision 3, which was approved by the NRC, is complied with in the analysis of piping systems with the following exception:
In seismic analysis the modal responses are combined by SRSS and lower damping values are used.
Th'is was previously stated in the FSAR in response to Question 110.56 which was submitted in Revision 12, September 1979.
ACTION:
Copies of the slides used in the presentation were requested and are supplied as Attachment H.
Revise FSAR Subsection 3.7b.3 as shown in Attachment I.
Mr. B. J. Youngblood, Chief August 28, 1980 Page 8
RESPONSE
(NSSS):
The NSSS seismic design of Susquehanna SES was established prior to the issuance of Regulatory Guide 1.92 and, therefore, the consideration of closely spaced modes in the response spectrum method of seismic analysis
'as described in this regulatory guide was not a requirement for the issuance of the Susquehanna SES construction permit.
Therefore for the NSSS equipment where response spectrum method of seismic analysis was used, the closely spaced modal responses were combined by the Square Root of the Sum of the Squares method (SRSS).
However for Susquehanna seismic analysis of the NSSS ASME Safety Class 1 piping the double sum method of Reference 2 of Regulatory Guide 1.92 approved by NRC on GESSAR 251 docket is used to combine the closely spaced modal responses.
ACTION:
PP&L will provide the following information as requested by the NRC staff:
1.
The modal shapes and natural frequencies of the piping systems and equipment.
2.
The portion of the loadings which is due to seismic.
3.
Verify that when the double sum method is used that it is ch'ecked by the SRSS method as described in GESSAR 251.
Demonstrate the equivalent level of safety achieved by the double sum method.
ll.
SER Pa e
8 Para ra h 1:
In FSAR Section 3.7a.2.3.2 the applicant has stated that "the two rotational coordinates about each node point are excluded because the moment contribution of rotary inertia from surrounding nodes".
We require clarification of this statement.
RESPONSE
The statement in FSAR Subsection 3.7a.2.3.2 has been clarified (see Attachment J).
ACTION:
Revise FSAR Subsection 3.7a.2.3.2 as shown in Attachment J.
Mr. B. J. Youngblood, Chief Augus't 28, 1980 Page 9
12.
SER Pa e 10 Para ra h 4:
For the control rod drives and their housing, the list of design transients and their associated number of cycles is acceptable with the exception of the number of OBE's assumed.
Standard Review Plan 3.7.3, "Seismic Subsystem Analysis", requires 5 OBE's with a minimum of 10 cycles each to be utilized in fatigue evaluations.
Only 1 OBE has been assumed by the applicant in the analysis of the control rod drives'ESPONSE:
The above concern is addressed in the response to SER, Page 7,
Paragraph l.
ACTION:
None 13'ER Pa e ll Para ra h 1:
The above, criteria also applies to CRD housings, in-core housings, hydraulic control units, core supports, other reactor.internals, reactor vessel support skirt, shroud support, shroud plate, MSIV's, SRV's, recirculation pumps, and recirculation gate valves.
It is not clear whether the applicant has considered the OBE in the fatigue evaluation of these items'hese apparent exceptions conflict with FSAR Table 3.7a-4 and a statement in Section 3.7a.3.2 that "the OBE is an upset condition" and, therefore, must be included in fatigue evaluations according to ASME Section III.
The applicant must provide clarification of the consideration of OBE loads for NSSS and BOP scope Class 1 components to resolve these apparent conflicts in the FSAR.
RESPONSE
(NSSS):
The above concern has been addressed in the response to SER Page 7,
Paragraph 1 and NRC Question 110.37 (submitted in Revision 12, September 1979), which requested an amendment to the various sections of the FSAR.
In addition, it should be noted that in the normal duty fatigue analysis calculations of the MSIV's, SRV's, and recirculation gate valves, significant factors of safety on the code permissible cumulative usage factor (Q < 1) and'fatigue cycles (Na
>2000) were obtained.
Although dynamic'-,cycles associated with OBE were not considered a design basis at the time these valves were ordered, in light of the above indicated safety margins it is concluded that the
Mr. B. J. Youngblood, Chief August 28, 1980 Page 10 consideration of the additional cycles associated with OBE will not jeopardize the design fatigue life of this equipment.
ACTION:
None
RESPONSE
(NON-NSSS):
For the non-NSSS ASME Section III, Class 1 components, the OBE loads are considered in fatigue evaluation.
ACTION:
None 14.
SER Pa e ll Para ra h 2; In FSAR Section 3.9.1.1.2 it is stated that for th'e CRD housing a
scram with no buffer is considered a normal/upset condition with 1 cycle.
For the CRD the same event has 10 cycles.
The 'applicant must resolve this inconsistency.
RESPONSE
The statement in Section 3.9.1.1.2 for the CRD housing is a
typographical error and should be 10 cycles instead of 1 cycle.
ACTION:
Revise FSAR Section 3.9.1.1.2 as shown in Attachment K.
15.
SER Pa e 13 Para ra h 3:
Our.review has uncovered a few open issues which require resolution.
For NSSS piping, the applicant's acceptance criteria for piping vibration is'hat the stress due to pressure and the measured steady state or transient vibration should be less than Service Limit B.
For common ferritic steels at 600~F, this stress limit would be about 30 ksc.
We believe it more appropriate to have a limit on stress due to vibration alone (neglecting pressure) which is relate'd to the material endurance limit in some fashion.
RESPONSE
Acceptance criteria are divided into two categories, i.e., Level 1 and Xevel 2. If Level 1 criteria are violated, the test must be placed on hold until'orrective action is taken.
If Ievel 2 criteria are
Mr. B. J. Youngblood, Chief August 28, 1980 Page ll violated, the test can continue, but the measurements must be evaluated to verify that continued test operation will not result in exceeding piping fatigue requirements.
For steady state vibration the piping peak stress (zero to peak) due to vibration only (neglecting pressure) will not exceed 10,000 psi for Level 1 criteria and 5,,000 psi for Level 2 criteria.
These limi'ts are below the piping material fatigue endurance limits as defined for 10 cycles in Appendix I of ASME Section III.
The material fatigue endurance limits for carbon steel and stainless steel are 13,000 psi and 25,000 psi, respectively.
For operating transient vibration the piping bending stress (zero to peak) due to operating transient only will not exceed 1.2 S
or pipe m
support loads will not exceed. the Service Ievel D ratings for Level 1 criteria.
The 1.2 S limit insures that the total primary stress including stress and'ead weight will not exceed 1.8 S< the code Service Level B limit.
Level 2 criteria are based on pape stresses and support loads not to exceed design basis predictions.
Design basis criteria require that operating transient stresses and loads not to exceed any of the Service Ievel B limits including primary stress 1'imits, fatigue usage factor limits, and allowable loads on snubbers.
ACTION:
None 16.
SER Pa e.14 Para ra h 1:
The other open issue in this review area is the scope of the test program for BOP piping.
Further detailed discussions will be required to resolve this issue.
RESPONSE
The scope of the test program for BOP piping systems is shown in FSAR Table 3.9-33.
Also see revised Subsection 3.9.2 (Attachment L).
ACTION:
Revise FSAR Subsection 3.9.2 as shown in Attachment L.
17.
SER Pa e 14 Para ra h 2:
Also, we will require the applicant to provide a brief summary of the results of this test program upon its completion.
Mr. B. J. Youngblood, Chief August 28, 1980 Page 12
RESPONSE
PAL will provide a brief summary of the piping test program after completion of test.
ACTION:
Provide a summary of piping test program.
18.
SER Pa e 15 Para ra h 2:
The applicant has committed to test the reactor int'ernals in accordance with the provisions of Regulatory Guide 1.20, "Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Start-Up Testing", Revision 2, for non-prototype Category I plants.
Test procedures will require operation of the recirculation system at rated flow with internals installed (except for fuel).
Test duration will assure that a minimum of 10 cycles of vibration will be experienced by the critical components during two-loop and s$.ngle-loop operation of the recirculation system.
At the completion of the flow tests, the vessel head will be removed and the internals will be inspected for evidence of vibration, wear, and loose parts.
The inspection will cov'er all components which were examined on the prototype design including the shroud, shroud head, core support structures, jet pumps, control rod drive, in-core guide
- tubes, and lower plenum.
The test results will be compared with the analytical results.
CXARIFICATION OF PARAGRAPH:
Revise the above stated paragraph as follows:
"The applicant has committed to test the reactor internals in accordance with the provisions of Regulatory Guide 1.20, "Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Start-Up Testing", Revision 2, for non" prototype Category I plants.
Test procedures will require operation of the recirculation system at rated flow with reactor internals important to safety installed.
Blade guides, incore instruments, neutron, sources and fuel will not be installed.
Control rods will be either not installed or fully withdrawn and prevented from inserting.
Test duration will assure that a minimum of 10 cycles of vibration will be experienced by the critical components during two-loop and single-loop operation of the recirculation system.
At the completion of the flow test, the vessel head will be removed and the internals will be inspected for evidence of vibration, wear and 1oose parts.
The inspection will cover all components which were examined on the prototype design including the shroud, shroud head, dryer, core
Mr. B. J. Youngblood, Chief August 28; 1980 Page 13 support structures, jet pumps, peripheral control rod drive guide tubes, peripheral incore guide tubes, and lower plenum."
Also see revised, Subsection 3.9.2.4 (Attachment M).
ACTION:
NRC staff will revise the paragraph as stated above.
PAL will revise CESAR Subsection 3.9.2.4 as shown in Attachment M.
19.
SER -Pa e 16 Para ra h 3:
We will require the applicant to provide a brief summary of the results of this test program upon its completion.
RESPONSE
PPSL will provide the NRC staff with a brief summary of the Reactor Internals Vibration test results upon test completion.
ACTION:
Provide a brief test summary.
20.
SER Pa e 17 Para ra h 1:
The applicant has analyzed its reactor 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.
We cannot complete our review in this area until the applicant submits the information requested in Question 110.32.
RESPONSE
The information re'quested in Question 110.32 will be provided at the completion of the New Loads Adequacy Evaluation program (4Q1980).
ACTION:
Provide information requested in Question 110.32.
21.
SER Pa e 17 Para ra h 2:
We have also requested in Question 110.41 that the applicant provide response time histories't one key location for each of the following internal components:
Mr. -B. J. Youngblood, Chief
~
~
~
August 28, 1980 Page Z4 jet pump shroud wall shroud head control rod instrument guide tube core plate This location should be the one having either the maximum stress combination or the most critical deflection combination, whichever governs the design.
Separate responses are required for the loads associated with the SSE and the most severe pipe break event.
The method of combining=the dynamic responses should be outlined and justified.
Our review of this information is incomplete.
RESPONSE
The response to Question 110.41 was provided in FSAR Revision 14 dated
- February, 1980 and addressed all the concerns raised by the NRC staff.
ACTION:
No action is required since the response to Question 110.41 is acceptable to the NRC staff.
22.
SER Pa e 19 Para ra hs 2 and 3:
The first area of review is the subject of load combinations and allowable stresses.
Our required loading combinations for the Susquehanna plant have been included in Question 110.42.
With one exception the applicant has provided a commitment that all ASME Class 1 y 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 the referenced loading combinations.
This one exception is related to our position that for load cases 1
and 2, as identified in Question 110.42, all ASME Code Service Level B requirements are to be met, including fatigue usage factor requirements, and should take into account all SRV discharge load effects (initial actuation and continuous suppression pool vibratory) taken for the number of cycles consistent with the 40 year design life of the plant.
On the other hand, the applicant states that its ASME Class 1 piping fatigue calculations have considered OBE and SRV loads separately, not in a combined fashion as we require.
Mr. B. J. Youngblood, Chief August 28, 1980 Page 15 Justification for Classif in the Loadin Combination P + DW + (FV
+ OBE ) 1/2 as Emer enc Condition It is conceivable that an Operating Basis Earthquake (OBE) could cause a plant trip with consequential turbine stop valve closure (FV). It is recognized that the OBE and turbine stop valve closure events could occur simultane'ously.
However, the total number of stress cycles with the amplitude equal to the maximum OBE combined with the maximum FV stress amplitudes during plant life are limited and classification of the loading combination as Emergency Condition is appropriate.
OBE and FV are separately combined with pressure and dead weight, and are considered as upset conditions.
.On the Main Steam system there is a.known transient flow induced unbalanced load imposed on the system due to the fast valve closure,.
The time duration for loads induced during the transient flow is of the order of milliseconds.
The results of time history dynamic analysis, using'onservative damping ratios, 'indicates that approximately 3 to 5 cycles (but only one cycle with the maximum amplitude) occur.before the vibration ceases.
Since the OBE is postulated to occur 5 times during the plant life, this means that 15 to 25 stress cycles could occur during the entire plant life with only about 5 stress cycles with combined maximum stress due to OBE and FV.
Further, the time histories of the two transients are unrelated.
The OBE is a random excitation while the FVC is a predictable dynamic behavior.
The probability of the maximum stresses occurring at the same location in the system, in phase and at precisely the same instant of time during the 100-200 millisecond duration of the fast valve closure is very, very low.
To consider the combination of the loads produced by these 'two events for comparison to an upset stress limit (which is approximately 1/2 the material yield stress) is not the intent of the ASME Code.
RESPONSE
See response to FSAR Question 110.42 (Attachment N).
ACTION:
1.
Revise FSAR as shown in Attachment N.
2.
PAL is to provide justification for Footnote 6 on Table 110.42-l.
Mr. B. J. Youngbl'ood, 'Chief August'8, 1980 Page 16 23.
SER Pa e 20 Para ra h 1:
We consider the subject of seismic sloshing loads in the Susquehanna suppression pool to be an open issue because the applicant has not responded to our Question 110.4.
RESPONSE
The issue of seismic sloshing loads has been,addres'sed in the response to Question 021.73.
Question 110.4 will be revised to reference Question 021.73.
ACTION:
Revise Question 110.4 to reference Question 021.73.
'24.
SER Pa e 20 Para ra h 2:
Another open issue related to load combinations is the applicant's method for combining peak responses to multiple dynamic loads.
The applicant has used the "square root of the sum of the squares" method (SRSS) for all dynamic loads.
Our position, as outlined in NUREG-
- 0484, "Methodology for Combining Dynamic Responses",
is that the SRSS method is acceptable for combining peak dynamic responses due to IOCA and SSE.
For other dynamic loads we are currently 'preparing a generic position which should be available in the near future.
RESPONSE
For all mechanical
- systems, components and supports, the dynamic responses to dynamic loads such as LOCA, and SRV and OBE or SSE are combined using the "square root of the sum of the square" (SRSS) method.
We understand that the NRC Topical Report Evaluation of "Technical Basis for the Use of the Square Root of the Sum of Squares (SRSS)
Method for Combining Dynamic I,oads for Mark II Plants" (letter from Roger J. Mattson to H. Chau, Chairman of MKII Owners Group dated June 25, 1980) and Revision 1 of NUREG-0484 accept the SRSS method for the Mark II load combinations.
ACTION:
None 25.
SER Pa e 20 Para ra h 3:
The applicant has not yet responded to Question 110.53 concerning the use of ASME Code Cases.
This is an open issue.
Mr. B. J. Youngblood, Chief August 28, 1980 Page 17
RESPONSE
PAL will provide a statement, that no unapproved code cases were used or a list of the unapproved code cases which were used.
ACTION:
Provide the response to Question 110.53.
26.
SER Pa e 21 Para ra h 2:
Our review has resulted in two open issues concerning the applicant's operability assurance program for active values.
The applicant's response to Question 110.22 stated that the 20 inch Lunkenheimer recirculation loop gate valves were qualified by similarity to a 4 inch Anchor/Darling gate valve.
Our review has not concluded that these values are, indeed, similar.
We require the applicant to either justify this similarity or qualify the 20 inch valve prototypically.
RESPONSE
The similarity of the 4 inch Anchor Darling gate valve to the 28 inch Lunkenheimer valve was on the basis of similar major parts, i.e. body,
The 4 inch valve was tested to ensure conservatism in design.
A combination of analysis and testing to assure operability of the discharge gate valve is described below:
In the recirculation loop, only recirculation discharge gate valves are required to be operable for the safe shutdown of the plant in the case of a recirculation line suction nozzle break.
Analysis results of the valve body, bonnet and yoke under the limiting loading conditions indicate that the deformations do.not exceed the elastic limit of the materials.
Hence, this assures that the components will return to their original position after the loads are removed.
Since these discharge valves are required to operate after the IOCA induced loads are not present, the operability of the valve is assured.
In addition, the representative Susquehanna SES motor operators designated by the vendor as one SMB family have been qualified for operability under expected environments and loading conditions.
Therefore the above analysis and testing adequately assures that the recirculation 1'oop discharge gate valve will be operable when required to operate.
ACTION:
1.
Revise the response to Question 110.22 to delete reference to the 4 inch valve.
Mr. B. J. Youngblood, Chief August 28, 1980 Page 18 2.
Add a statement to the response that during the loading there is no binding that prevents valve operation when the load is released.
27.
SER Pa e 21 Para ra h 3:
The applicant has stated in the FSAR that as long as a valve and its internal parts remain elastic, valve deformations will remain so small that valve operability will not be impaired.
We do not necessarily agree.
It is our position that even elastic deformations must be checked to provide assurance of the valves not binding.
Similarly, elastic deformation in active pumps should also be checked.
RESPONSE
The following additional information pertaining to operability assuran'ce for the active valves and pumps in GE scope of supply is provided on a component-by-component basis.
Main Steam Safet Relief Valves (SRVs)
The operability of the SRV valves without binding will be assured by
.in-process inspection of critical dimensions and clearances to verify similarity to that specimen which has undergone full seismic and full load qualification.
Testing of each Susquehanna SRV will be conducted with steam flow for setpoint verification.
Main Steam Isolation Valves (MSIVs)
The only part which could possibly bind inside the valve is the poppet.
The poppet, travels along guides with generous clearance of approximately 0.020-0.040 inch on the diameter and has a self-aligning cone for a conical seat.
Gross distortions and unusually abnormal loading in the valve body would be required for binding of the poppet.
Imposed loads result, in a maximum stress ratio of 0.34 which is well within the elastic limits and binding on the MSIV is not possible.
Recirculation Gate Valves The recirculation discharge valves are designed to maintain their pressure and structural integrity during a IOCA event (defined as the recirculation suction nozzle break) and be capable of closing after the break.
After this break event, no IOCA induced stresses remain on the valve, therefore, any elastic deformation occurring in the valve due to LOCA induced stress will not be present when the valve 'is required to function.
Thus'binding will not occur and valve operability is assured.
Mr. B. J. Youngblood, Chief August 28, 1980 Page 19 Standb Li uid Control E losive Valve Ci H
The operability of the standby liquid control explosive valve is not based on an elastic deformation criterion, instead, the operability is demonstrated by testing.
RHR and Core S ra Pum s
The RHR and core spray pump and motor assemblies are anal'yzed for static and dynamic loads to demonstrate structural integrity and operability.
This is completed using a beam and lumped mass finite element model of the assembly with nodes located at critical points where stress and deflections are evaluated to check clearances.
The critical locations where.deformation is evaluated to assure operability are listed below:
o 'elative horizontal displacement between impeller and bowl.
o Relative horizontal displacement between shaft and mechanical seal.
o Relative vertical displacement between shaft and mechanical seal.
o Relative vertical displacement between first stage impeller and bowl.
o Relative horizontal displacement between motor rotor and stator.
Standby Liquid Control Pump (SLC), Reactor Core Isolation Cooling Pump (RCIC)
Hi h Pressure Coolant In ection Pum (HPCI)
Elastic deformation in pumps comes from two sources, the first being internal pressure and the second being applied loads including dynamic loads.
For all pumps the pressure distortion acceptability is demonstrated as part of the pump acceptance testing.
The HPCI, RCIC and SLC pumps are extremely rigid and the operabilities are not affected by applied loading.
The alignment between the pump and driver is analyzed for acceptability during all -loading conditions to assure operability.
ACTION:
PPSL will provide the actual versus the allowable deformations for the RHR and Core Spray Pumps, Standby Liquid Control Pump, RCIC Pump and HPCI Pump.
j
Mr. B. J. Youngblood, Chief August 28, 1980 Page 20 28.
SER Pa e 24 Para ra h 1:
We have reviewed the applicant's design criteria pertaining to buckling of component supports and the design of bolts used in component supports.
With respect to buckling, the applicant has not yet responded to our Question 110.43.
RESPONSE
The response to Question 110.43 was provided in CESAR Revision 15, dated April 1980.
The response to that question has been revised (see ).
ACTION:
Revise Question 110.43 as shown in Attachment 0.
29.
SER Pa e 26 Para ra h 2:
The only open issue concerning the applicant's design criteria is the statement in the CESAR that "Deformations are not a limiting factor in the analysis of the CRD's components since stresses are in the elastic region."
Further information must be provided to show that the elastic deformations can not affect the performance of the CRD system.
RESPONSE
The General Electric position on BWR/4-5 plants is that the drive is capable of performing its safety function, i.e.
scram, following a seismic event.
By definition, elastic deformation of parts would mean that interfacing components would be back to their original state when the loads were removed.
After the seismic event the drives would scram normally even if deformation had been great enough to completely prevent motion during the seismic event.
Testing has been done to determine CRD ability to scram with deflections equivalent to those calculated for the seismic conditions.
These tests have imposed fixed fuel channel bows and effects on scram time were measured.
Testing has also been run with dynamic deflection of the CRD housing at its natural frequency.
Additional analysis is expected to be completed the 4th quarter of 1980.
This analysis is underway to evaluate the effect of seismic and other accident loading on the entire driveline.
ACTION:
None
Mr. B. J. Youngblood, Chief August 28, 1980 Page 21 30.
SER Pa e 29 Para ra h 2:
The applicant has not yet submitted its program for the preservice and inservice testing of pumps and valves, as requested by Question 110.47, therefore we have not yet completed our review.
RESPONSE
The information requested in Question 110.47 will be provided 6 months prior to fuel load.
ACTION:
Provide requested information.
After the above actions are taken, we assume that the items and concerns are resolved.
Very truly yours, N.
M. Curtis Vice President-Engineering and Construction-Nucl'ear CTC/cak CTC 59:1
Attachment A'SES-PSAB Page 1 of 2 breaks.
These fiqures indicate the breaks for vhich dynamic analysis vas performed and the. type of the break assumed.
d 3.6.2.2.2 Analytic Methods to Define Blovdown Forcing Functions and
Response
Models for Recirculation Piping System 3.6.2.2. 2.1 Analytical Methods to Define Blowdovn Poxcing
~unCtiddnS The rupture of a pressurized pipe causes the flov characteristics of the system to change, creating reaction forces vhich can dynamically excite the piping system.
The reaction forces are a function of time and space and de pend upon fluid state within the pipe prior to ruptuxe, break flov area, frictional losses, plant system characteristics, piping
- system, and other factors.
The methods used to calculate the reaction forces for recirculation piping system are presented in the follovinq sections.
d*dodd 'd'd blowdovn forcing functions includes:
ddd d
d ~
d d
pipe severance and separation amounting to at least a
one-diameter lateral displacement of the ruptured piping sections unless physically limited by piping restraints, structural membex's, or piping stiffness as t~~ demonstrated by the inelastic pipe vhip analy si s (Subsection
- 3. 6. 2. 2. 2. 2)
.'2)
The 'dynamic force of the jet dischar'ge at the break location s~CW~ based on the effective cross-sectional flov area of the pipe and on a calculated fluid~ pxessuxe as modified by an analytically or experimentally determined thrust coefficient.
Limit'ed pipe displacement at the break location, line restrictions, flov limiters, positive pump-controlled flow, and the absence of energy reservoirs caybe~
taken into account, as applicable, in the reduction of get discharge.
zJ'sed for the initial pulse, c
1 s<-bee.~
I I
3 6-20
Attachment A
SSES-PSAR Page 2 of 2 Blovdown forcing functions are determined by either of tvo methods qiven in
{1) and (2) belov:
(1) 'he predicted blovdovn forces on pipes fed by a
pressure vessel can be described by transient and steadystate forcing functions.
'The forcing functions used are based on methods described in Reference 3.6-6.
These ~ sis ply descpihed as follows:
a e The transient forcinq functions at points along the pipe, result, from the propagation of waves (wave thrust) along the pipe, and from the reaction force due to the momentum of the fluid leavinq the end of the pipe (blowdovn thrust)-
b.
The vaves cause various sections of the pipe to be loaded with time-dependent forces.
It is assumed that the pipe is one-dimensiona1, in that there is no attenuation or reflection of the pressure waves at bends,
- elbovs, and the like.
Pollowing the rupture, a decompression wa ve is assumed to travel from the break at a
speed equal to the local speed of sound vithin the fluid.
Rave reflect'ions wi11 occur at the.
break
.end, changes in direction of piping, and the pressure vessel until a steady flow condition is established.
Vessel and free space conditions are used as boundary conditions.
The blovdovn thrust causes a reaction force perpendicular to the pipe break.
Ce d
The initial blovdovn force on the pipe is taken as the sum of the vave and blowdovn thrusts and is equal to the vessel pressure
{P6 ) times the break area (A).
After the initial decompression period (i.e. ~ the time it takes for a vave to reach the first change in direction), the force is assumed to drop off to the value of the blovdovn thrust (i.e., 0.7 P0A).
Time histories of transient
- pressure, flov rate, and other thermodynamic properties of the fluid can be used to calculate the blovdown force on the pipe using the following equation:
F
= [(P-Pa)
+
] A pu2 2g
- 3. 6-21
2UF. STION 1 10 26:
Attachment B
Page 1 of 26 Sections 3.6.2.5.2.
1 (page 3.6-31) and 3.6. 2.5.,2.'3
{page 3.6-32) are indicated as "Later."
provide a schedule for their inclusion in the FSAR.
NESPONSE:
npkpi-u~xy Se~
icuis~J S'
s ecV-I~~a J.G. / ~-~
~. < ~ ~
(
Rev.
>4, 2/80 I(
110 2 6-1.
4
Attachment B
Page 2 of 26 2)
The safety/relief valves protect the main steamlines from abnormal pressure.
The safety/relief valves, besides protecting the steamlines against over-pressurization, precipitate the initiation of the LPCI mode of the BHR system for smaller pipeline breaks by rapidly depressurizing the reactor vessel.
3)
Separate main steam loops supply high pressure steam to run the turbine driven pumps of RCIC and HPCI systems.
Should steam power not be available to drive the reactor feedwater pumps during shutdown, part of the residual steam will be used to drive the turbine in the RCIC system which returns makeup water to the reactor from the condensate storage tanks, the suppression pool, or from the RHR heat exchanger during the steam condensing mode.
In addition, the following design features are incorporated into the design to ensure isolation valve operability and the leaktiqht inteqrity of the containment:
a)
The piping between the containment isolation valves is designed to meet "no break" criteria stress limits of Subsection 3.6. 2. 1. 1 b)
Homent limiting restraints are placed upstream of inside containment isolation valves and downstream of outside containment isolation valves.
c)
Plate barriers are provided to protect the main steam isolation valve operator from other high energy fluid system pipe breaks.
gi2~eB eak Locations and pi~en~hi nestcaints 4,
The postulated p pe break locations and the type of the break are determined b sed on the criteria given in Subsection 3.6.2.
Figure shows the locations of postulated pipe breaks and pipe whip restraints.
The main steamlines are restrained inside the primary containment to prevent the main steam pipe whip.
The main steamlines in the turbine building are separated from essential systems and components.
I Verific~aion of the Safe Shutdown of the Plant 1)
The routing of main steam piping, locations of pipe whip restraints, and the protective measures described in Table 3. 6-2 ensure that the emergency core cooling systems are not adversely affected by a postulated pipe break in the main steam lines.
3 6-5
Attachment B
Page 3 of 26 a) b)
ht each location of potential high stess~ such as pipe fittings (elbovs,
- tees, reducers, etc),,'valves,
- flanges, and velded attachments l
ht each location vhere the folloving stress and fatigue limits are not met:
For jtSHE Code,Section III, Class 1 Piping under normal and upset conditions, 1)
The primary plus secondary stress intensity range, Sn, as calculated by equation (10) of Paragraph NB-
- 3653, does not exceed 2.4 Sm.
2)
The stress intensity range, Sn, as calculated by equation
{10) of Paragraph NB-3653 exceeds 2.4 Sm, but is less than 3.0 Sm, and the cumulative fatigue usage factor is less than 0. 10.
3)
The stress intensity range, Sn, calculated by equation (10) exceeds 3.0 Sm, -'but the 'ranges of stresses computed by equations (12) and (13) of subparagraph NB-3653 are less than 2.4 Sm and the fatigue usage factor is less than 0.10.
For ASIDE Code,Section III, Class 2 and '3 Piping 4)
In the event that at least tvo intermediate pipe break locations cannot be determined by the above stress and fatigue usage criteria, a
minimum of tvo locations of highest stress as calculated by equation (10) in Paragraph NB-3653, and vhich are separated by a change of direction'><in the pipe ran,
~a're selected.
j,(.:
For-hSHZ Code,Section III, Class 2-a'nd 3 piping:
5)
The maximum range of stress, as calculated by the sum of equations (9) and (10) in Paragraph NC-3652, for normal and upset plant conditions, does not exceed 0.8 (1.2 Sg
+ S<)
6)
If tvo intermediate break locations cannot be determined-hy the above stress and fatigue usage
- criteria, a
minimum of tvo locations of highest
- stress, as calculated by the sum of equations (9) and
{10) in Paragraph NC'-3652, and vhich are separated by a change in direction of the pipe run, are selected For piping not designed to seismic Category I standards:
I 3 6-12 REV 1.", 7/,9
Attachment B
Page 4 of 26
~
~
~
~
~
response analysis is provided on piqure 3.6-12.
Zn this analy'sis pipe restraint rebound effects are also considered.
~
~
~
~
~
~
~
~
~
The criteria for the dynamic analyses're as follovs:
X 1)
An analysis of the piping sys'em is performed for e'ach longitudinal and circumferentiaL postulated rupture at the break locations determined in accordance vith the criteria of Subsection 3.6.2. 1.
2)
The Loading condition of a piping system prior to postulated rupture in terms of internal pressure, temperature, and stress state is that condition associated vith reactor operating at 100 percent pover 3)
For a circumferential rupture, pipe vhip dynamic analyses are performed only for that end (or ends) of the pipe or branch that is connected to a contained fluid energy reservoir having sufficient capacity to develop a get stream.
4)
Dynamic analytic methods used for calculating the pipinq and piping/restraint system response to the pipe break forces adequately account for the effects of.
a)
TransLational masses (and rotational masses for ma)or components) and stiffness properties of the piping system, restraint system, major components, and support valls')
Transient forcing function(s) acting on the pipinq system c)
ELastic and inelastic deformation of piping and/or restraint d)
The design clearance betveen the pipe and the restraint.
5)
An aLLovable desiqn strain limit of 0.5 ultimate uniform strain of the materials of the zestraints is used.
6)
A 10 percent increase of minimum specified design yield strenqth (Sy) is used to account for strain rate effects in inelastic nonlinear analyses.
0
'Fs g4Zeg l~ct.'fotag of Apl Roe eedwater an 3 ~ 6 'I fg 3 6 g
y,~ g<
PtPc
~rc'ale toeQ:~ ~
P.Pc brc+
'fglelc s
'1 ~ 6 6 ta
'x ~ 6~ l3
~~~+
+c 5
<hiked(g 1$ Sc RtlAlysI $
gaackw ~Cf Q~
HP<l RCI C cc K5 ~PILY RNg
~eA ~
'.6-'19
<Asar+
4yg g
~
~
Attachment B
Page 5 pf 26
~
+~~4.~.
~~ee.
These figures~indicate the breaks for vhich dynamic analysis vas performed and the type of the break'ssumed 3.4.2.2.2 Analytic Hethods to Define Blovdovn Forcing Functions and
Response
Models for Becirculation Pipin S st m
3.6.2.2.2.1 Analytical Methods to Define Blovdovn Forcing l
'"'he rupture of a pressurized pipe causes the flov characteristics of the system to change, creating reaction forces vhich can dynamically excite the piping system.
The reaction forces are a function of time and space and depend upon fluid state vithin the pipe prior to rupture, break flov area, frictional losses, plant system characteristics, piping
- system, and other factors.
The methods used to calculate the reaction forces for recirculation piping system are presented in the folloving sections The criteria that should be used for calculation of fluid blovdovn forcing functions includes:
r Circumferential breaks should be assumed to resuult in pipe severance and separation amounting to at least a
one-diameter lateral displacement of the ruptured piping sections unless physically limited by piping restraints, structural
- members, or piping stiffness as xay be demonstrated by the inelastic, pipe vhip analysis (Subsection 3.6.2.2.2.2).
j ( ii (2) 'he dynamic force of the get discharge at the break location should be based on the effective cross-sectional flov area of the pipe and on a calculated fluid pressure as modified by an analytically or experimentally determined thrust coefficient Limited pipe displacement at the break location, line restrictions, flov limiters, positive pump-controlled fLov,( and the absence of energy reservoirs may be taken into account, as applicable, in the reduction of get discharge.
(3)
A ris'e time not exceeding one millisecond should be used for the initial pulse, unless lonqer crack propagation times or rupture opening times can be substantiated by experimental data or analytical theory i
3'-20
Attachment B
SSES-FS AR Page 6 of 26 linear approximation vas made by NSC for the restraint load-
".'deflection curve supplied by GE.
This approximation is demonstrated in Figure 3 6-15.
The effect of this approximation is to give lover energy absorption of a given restraint deflection.
Typically, this yields higher re st ra int de flections and lover restraint to structure loads than the GE analysis.
The deflection limit used by NSC is the design deflection at one-half of the ultimate uniform strain for the GE restraint design.
The restraint properties used for both analyses are provided in Table 3 6-4 A comparison of the NSC analysis vith the PDA analysis, as presented in Table 3.6-5 and Figure 3 6-16, shovs that PDA predicts higher loads in 15 of the 18 restraints analyzed.
This is due to the NSC model includinq energy absorbing effects in secondary pipe elements and structural members.
Ho ve ver, PDA predicts hiqher restraint deflections in 50% of the restraints.
The hiqher deflections predicted by NSC for the lover loads are caused by the linear approximation used for the force-deflection curve rather than by differences in computer techniques.
This comparison demonstrates that the simplified modelinq system used in PDA is adequate for pipe rupture loading, restraint performance and pipe movement predictions vithin the meaninqful design requirements for these lov probability postulated accidents.
3.6.2.3 Dynamic Analysis Nethods To.Verify Integrity and'O erabili~t g~ Q~o~Pjpjng~theg than gecirculgtion giping+ystem Pipe vhip restraints and compensatinq struts are used to control pipe vhipping during a postulated" rupture of the pipe.
Barriers are used to protect components against get impinqeaent e
Compensating struts are mechanical snubbers used to perform the follovinq functions:
a.
Permit unrestrained thermal motion of the pipe.
Restrain pipe motion under seisaic and other dynamic
- loads, and c.
Resist sustained loads resulting from a pipe break.
The pipe vhip restraints used to protect the mechanical components are designed either as a part of the normal restraint Rev. 12, 9/79
- 3. 6-25
Attachment 'B,
,SSES-ZSAR Page 7 of 26 3.6.2.5 lfaterial To Be Submitted for Operating License Reviev
$.6.~.~ fog ~Pi
~in Otheg tta~necirenlation piping goatee The folloving paragraphs indicate hov the criteria for protection against dynamic effects associated vith postulated piping features are iaplemented 2)
The criteria qiven in Subsection 3.6.2.1 have been adhered to in locating the pipe failure locations and type of the failure.
These locations are shovn on Figures 3.6-'I to ~~Is 3.6.B egypt <ct ) f g4 Steak ws+<o'<t t5
%%own Protective devices such as pipe vhip restraints and the barriers are used.
The in-service inspection requirepents are implemented as discussed in Section 6.6.
3)
Analytical methods to analyze the effects of pipe break are discussed in Subsections 3.6.2 2 and 3.6.2.3.
Summary of the results are shovn on ~
Ta.thea n. i, - c Mro+
- n. t- <t.
0)
All safety related systems and components have been protected from the efEects of pipe vhip and their design intended function vill not be impaired to an u naccep ta ble level.
/
3.6.2.5.2 Implementation of Criteria for Pipe Break and Crack Location and Orientation for Recirculation 3.6.2.5.2.1 Postulated Pipe Breaks in Recirculation Piping System Inside Containment The criteria for selection of postulated pipe breaks in the recirculation piping system, i'aside containment, is provided in Subsection 3.6.2.1.2.
Break locations on recirculation piping have not yet been finalized and all pertinent data relatinq to this vill be supplied at a later date, pending issuance of piPing stress analyses reports of the respective projects 3 6-3)
Attachment B
TABLE 3.6-6 Page 8 of 26
SUMMARY
OF STRESS IN HIGH ENERGY ASME CLASS l PIPING MAIN STEAM LINE INSIDE CONTAINMENT I
Table 3.6-6 will be supplied the fourth quarter of 1980
Attachment B
.TABLE 3-6-7 Page 9 of 26
SUMMARY
OF STRESSES IN HIGH ENERGY ASME CLASS 1 PIPING
. @I FEEDWATQR,,ZiXQE".INSIDE.
CONTAINMENT.;".'ODE STRESS (KSI)
CUMULATIVE USAGE PIPE BREAK (EQ.
10)
FACTOR STRESS LIMIT (KSI)
REMARKS 2.4 SM 125 (Tapered transition joint) 122A (Butt weld) 122B
('utt weld) 118A (Butt weld) 118B (Butt weld) gg (Tapered transition
)oint) 64.049 46.923 46.870 46.490 46.473 68.311
.5809
.1235
.1227
.1179
.1176
.5939 42.480
- 42. 480.
42.480 42.480 42.480 42.480 A
82A (Butt weld) 82B (Butt weld)
- 85A, (Butt weld) 46.846 46.880
.1234 42.480
.1193 42.480
.1238, 42."480'-
85B
4 0.
(Tapered transition g oint.)
tt weld) 46.884 81.514 62.889 I
47.182 I
=.1194
.. 4594
.5458
.1232 42.480 42.480 42.480 42.480 A
Attachment B
,TABLE 3.6-7 SHEET 2
OF 2
~
0 o< 26 I
NODE
SUMMARY
OF STRESSES IN HIGH ENERGY ASME CLASS 1 PIPING FEEDWATgR-":GXNE-'NSIDE'ONTAXBRBHKl....
"--'TRESS (KSI)
CUMULATIVE USAGE PIPE BREAK (EQ.
10)
FACTOR STRESS LIMIT (KSI)
REMARKS 2.4 SM f42B (Butt weld) 45A (Butt weld) 45B (Butt weld) 35 (Tee) 25 (Tapered transition joint) 20 (Tapered.
transition g oint) 12 (Tapered transition joint) 10 (Tapered transition'oint) 47.383 47..561 47.539 84.20'4.033 70.568 71.423 71.537
.1243
.1261
.1258
.7874
.4919
.3436
.3823
'. 3355
'42.480 42.480 42.480 42.480 42.480 42.480 42.480 42.480 A
OTES: A-B CD Terminal End break Breaks determined by "Minimum Break, Locations" Criteria Breaks determined by stress reguirem6nt See Figure 3.6-2 for Node locations
Attachment B
TABLE 3'.6-8 Page ll of 26
SUMMARY
OF STRESSES IN HIGH ENERGY ASME CLASS 1 PIPING
//j'<X STEAM SUPPLY LINE INSIDE CONTAINMENT NODE STRESS (KSI)
CUMULATIVE, USAGE PIPE BREAK (EQ.
10)
FACTOR STRESS LIMIT (KSI)
REMARKS 2.4 SM 405 (Tapered transition joint) 409 (Elbows 42.113
- 53. 971
.5512
.0352 42.096 42.096 423 (Elbow) 50.896
.0315 42.096 433 (Elbow beginning) 36.404
.0011 42.096 NOTES:
A Terminal end break B Breaks determined by C Breajcs determined by D 'See.~Figure 3.6-3 for I
"Minimum Break Locations" Criteria Stress requirement'ode locations E
Attachment B
PaSe 12 of 26 4
TABLE 3.6-9
SUMMARY
OF STRESSES IN HIGH ENERGY ASME CLASS 1 PIPING RCIC STEAM SUPPLY LINE. INSIDE CONTAINMENT NODE STRESS (KSI)
CUMULATIVE USAGE PIPE BREAK (EQ.
10)
FACTOR STRESS LIMIT (KSI)
REMARKS 2.4 SM 646 (Tapered transition joint) 647 (Elbow) 29.837 31.523
.1424
.0004 42.096 42.096 658 668 (Elbow beginning)
~
~
30.040 21.621
.0029
.0000 42.096 42.096 NOTES:
A B
C D
Terminal End Break Breaks determined by "Minimum Break Locations" Criteria Breaks determined by stress requirement See Figure 3.6-4 for node locations
Attachment B
TABLE 3.6-10 Page 13 of 26
SUMMARY
OF STRESSES IN HIGH ENERGY ASME CLASS 1 PIPING CORE SPRAY LINE INSIDE CONTAINMENT NODE STRESS (Ksi)
E.10 CUMULATIVE USAGE PIPE BREAK STRESS FACTOR LIMIT Ksi 2.4 Sm REMARKS 10 (Butt Weld) 15 (Butt Weld) 20 A (Butt.Meld) 126.217 126.540 106.167
.5805
.3878
.6427 40.2 40.2 40.2 A
A 20 B
'08.467 (Elbow Beginning)
.3476 40.2 25
'62.'595
,.0054 40.2 NOTES:
A.
Terminal End Break B.
Breaks determined by "Minimum Break Locations" Criteria C.
Breaks determined by stress requirement D.
See Figure 3.6-5 for Node Locations
Attachment B
TABLE 3.6-11 Page 14 of 26
SUMMARY
OF STRESSES IN.HIGH ENERGY ASME CLASS 1 PIPING RHR SUPPLY LINES INSIDE CONTAINMENT NODE STRESS (Ksi)
E.10 CUMULATIVE USAGE PIPE BREAK STRESS FACTOR LIMIT Ksi '.4 Sm REMARKS 55 A (Butt Meld) 35.547
.00 40.200 A
55 B
79.274
{Elbow Beginning)
.2033 40.200 55 C
{Elbow End) 78.234
.1926 40.200 58 67.653 (Tapered Transition Joint)
.2355 40.200 62 64!442 (Tapered Transition,',C Joint),f
.1372
- 40. 200 K>
64 63.664 (Tapered Transition,
Joint)
. 1192 40.200 NOTES:
A.
Terminal End Break B.
Breaks determind by "Minimum Breaks Locations" Criteria C.
Breaks determined by stress requirement D.
See Figur'e 3.6-6 for Node Locations
Attachment B
TABLE 3.6-12 Page 15 of 26
SUMMARY
OF STRESSES IN HIGH ENERGEY ASME CLASS 1 PIPING RHR RETURN LINE INSIDE CONTAINMENT NODE STRESS (Ksi)
E
.10 CUMULATIVE USAGE PIPE BREAK FACTOR STRESS LIMIT Ksi 2.4 Sm)
REMARKS 125 (Tapered Transition Joint) 78.749
.8698 40.200 A
132 (Tapered Transition Joint) 50.803
.0052 40.200 145 (Tapered Transition Joint) 53.744
.0157 40.200 NOTES:
A.
Terminal End Break'.
Breaks determined by "Minimum Break Locations" Criteria C.
Breaks dhterminM by stress requirement D.
See Figure 3.6-7 for Node Locations
c ITI I
ITT C >
Z CO
+ +> m Z
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<Om h) O 37.
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~
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- 5. ata flti iCCAK QCSTRANT Ia. ~ ~ RvaatCC TYtC CliTRAI4f
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COtIPC4SATI4O CTRL y A ~ ARaOC IO. DRCAR'T SRV ARC CRCA<
CI SwaltOl\\TS MTN Tti WAC scoat AATC or Tci iclu(H SIRC i
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- purposes, a longitudinal break area shall be assumed to be the eguivalent of one circumferential pipe area unless analytical methods representinq test results can conservatively reduce forces based on a mechanistic approach.
- normally closed valves PV052A 6 PV052B.
- 3. 6.'1.1 (first paragraph on page 3.6-2) states that "A design basi for Susquehanna SES is that a postulated pipe break inside containment
- purposes, a longitudinal break area shall be assumed to be the equivalent of one cizcuaf erential pipe area unless analytical aethods
- 3. 62
- 3. 6>>2.
- 3. 6-2 Alternatively, nonlinear time history dynamic analyses are pezforaed The forcinq function used in piping dynamic analysis is obtained using Reference 3 6-0 A typical forcing function used foz the dynaaic 3~ 6-18
- I c)
- 3. 6-'(9
- Loads, Structural Des~in'f Nuclear Plant Facilities, Vol 1, (1973).
- 1970, Reactor Technology TID-9500.
- 3. 6'-7
- 3. 6-8 3.6-9 Relap 4
- Systems, ANCR-NUREG-1335, issued in September, 1976.
- Campbell, Ca.
- 5. Et
- 1975, as modified by Appendix III o Minter 1975
- addenda, and A-2232 of the Summer 1976 addend
- geometry, thus, t y are not implemented:
- however, the extent of examination of ASME Section XI will be supplemented to comply with the augmented inspection program requirements outlined above.
- SPECTRA, SRSS OBE 1l2%
- 1. FEEOKATER
- 2. RQfCU
- 3. HEAD SPRAY
- 1. RI-IA HX SEBUICE QUATER SYSTEM
- 2. ESW RETUBM LIMES
- 3. Elf SUPPLY UMES
- 4. RHR PUMP OltSCHABGE
- ALL,
- . EQUIVALENT'LEVELOF SAFETY'AS THE".'.
- 3. 7b-15
- Forces, Portland Cement Association (1974).
- Hence, the modeling of intermediate length members becomes unnecessary.
- pumps, in-core guide tubes and housinqs,
- sparger, and their'upply headers.
- hence, the couplinq effects are negligible.
- 3. 7a-8
- compared, dependinq on the specific safety class of the equipment, to Industrial Codes, ASIDE, ANSI or Industrial Standards, AISC, allowables.
- 3. 6.
- son, vxsua o servatzon x.n >cat tha bration is significant, measurements will be made hand he ibroqraph.
- 3. 9. 2-. la~
- loads, caused by mai'n stop valve closure or relief valve blow for example,,aze analyzed for the time dependent forces.
- Rev, 12, 9/79 3 9~2
- qeneral, the specific causes of the steady state vibration is not known beforehand; therefore, desiqn enqineers vith stress analysis experience and familiarity with the sub ject pipinq system vill visually observe the lines during all siqnificant
- page y of 2
- rr we' p~+~
- hence, each structural member (between tvo mass points) can only have an axial load.
- 3. 9. 5.2.
- 2) ought to be considered as an Emergency condition.
- OBE,
- 2) the number of combined stress cycles expected over the plant lifetime, and
- 3) the intent of the ASME code.
- inquiry, GE agreed to meet Upset limits (Service, Level B require-ments) without fatigue analysis.
- chugging, condensation oscillation, pool swell, drag loads, annulus pressurization, etc.
- Desiqn, Normal, Upset, and Emerqency Faulted The pipinq shall conform to the requirements of Section III, paraqraphs NC-3600 and ND-3600.
- Company, Atomic Power Equipment Department, APED-5460, July 1968.
- Document, General Electric Company, NEDE-20566.
- November, 1977.
- 3. 9-8 "Functional capability criteria for essential Mark II piping",
- . 'hn.,analysis of. reactor pressure vessel support skirt buckling for faulted. condi'tions shows that the support skirt has the capability to meet ASilE Code Section III, Paragraph P-1370 (c) faulted condition limits of 0.67, times the critical bucklinq strength of the, support at temperature assuminq that the critical bucklinq stress~limit corresponds to the material yield stress at temperature.
- pipe, and the compressive ef fects on the support skirt due to the thermal and pressure expansion of~ the reactor vessel.
RESPONSE
BTP APCSB 3-1 paragraph B.3.a., titled Anal sis and Effects of Postualated
~Pi i Fil t
tt'"...
l...l ki
Ãgh-gyf1 6
l piping or leakage crack in moderate-energy fluid system piping should be
~
considered separately as a single postulated initial event occurring during normal plant conditions."
For purposes of piping failure analysis only one 'initial event is postulated during normal plant conditions.
That is, for any single postulated seismic Category I or II pipe break, the plant Is assumed to be in a normal plant condition,(no SSE) with only that single
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postulated pipe break.
For conservatism and defense-in-depth, SSES has committed to the design basis stated in Subsection 3.6.1.1 of the FSAR that for inside the containment pipe breaks would be postulated in conjunction with an SSE.
But once again for this particular case, only a single pipe break (the LOCA) was considered in the evaluation of the piping failure.
It should be pointed out that the combination of a LOCA and an SSE is used as a loadin combination for the design of systems, components and structures required to bring about a safe shutdown.
But except as noted above, this combination is not used to analyze the effects of postulated piping breaks.
In all cases the design basis includes the requirement to be able to bring the reactor to a safe shutdown and maintain it in a safe shutdown condition without taking credit for operation of non-seismi'c Category I equipment.
110 18-1
'SES-FSAR Attachment E
Page 2 of 2 A design basis for Susquehanna SES is that a postulated pipebreak inside
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the containment (up to and including a rupture of the recirculation piping),
in conjunction with the SSE, plus a single fail'ure will not prevent the plant from accomplishing the above.
For outside the containment, the single failure is qualified per NRC Branch Technical Position APC'SB BTP 3-1, paragraph B.3.6.3.
No credit is taken for non-seismic system in plant shutdown following an SSE.
Components which are required to operate for a safe shutdown of the plant are protected from the below listed effects of postulated pipe failures, unless it can be demonstrated that the function of the safety equipm'ent is not impaired.
Pi e WhiP Pipe whip is assumed to be one consequence of a quillotine failure of a high energy pipe.
Cracks in moderate energy systems do not cause pipe whip.
Pipe whip is an unrestrained pipe movement of either end of the ruptured pipe in any direction about a plastic hinge formed at the nearest pipe whip restraint.
A whipping pipe is assumed to rupture an impacted pipe of smaller nominal pipe size, and of the same nominal size with smaller wall thickness.
A whipping pipe is assumed to have sufficient energy to cause the failure of impacted electrical cable ways and instrumentation unless the equipment can be shown to be sufficiently strengthened or protected.
High energy piping is located away from the essential safety 'system wherever practical.
Otherwise, pipe whip restraints are located on the piping to prevent pipe whip.
Jet Impingement
'Jet impingement loads (due to pipe failures) on equipment and safety systems are considered.
Protection against the jet impingement is provided either by separation or by the addition, of! barriers and enclosures.
Environmental ~:l' Pipe failures in high" and moderate energy lines will release fluid which can increase temperature,i pressure, and humidity in the vicinity of the pipe failure and also in remote areas that communicate with the local atmosphere.
Safety equip-ment required after the pipe failure may be exposed to abnormal conditions which can degrade the capability of the equipment to perform its function.
I Safety related equipment is qualified to meet the postulated environmental conditions.
3.6-2
At:tachmenC F
Page 1 of 7 diffezence,'etveen the circu a ferential and longitudinal stresses is deterained, then both types of breaks shall be considered.
d ht intermediate locations chosen to satisfy the miniaua break location criteria, only circumferential breaks shall be postulated.
For desiqn
, representing test results can conservatively reduce forces based on a m'echanistic approach.
f6)
Poz both longitudinal and circuaferential breaks, after assessing the contribution of upstream pipinq flexibilities, pipe vhippinq is assuaed to occur in the plane defined by the piping geometry and confiquration for circumferential breaks and out-of-plane for longitudinal bzeaks, and to cause pipe aoveaent in the direction of the get reaction.
P (7)
For a circuaferential break, the dynaaic force of the get discharge at the break location vi11 be based upon the effective cross-seCtional floe area of the pipe and on a calculated fluid pressure as aodified by an analytically or experiaentally deterained'hrust coefficient Justifiable line restrictions, flov liaitezs, and the absence of energy reservoirs shall be used, as applicable, in the reduction of the get discharge 3.6.2.2 Analytica1 models To Define Porcing Functions and es onse Models
~ ~
. 1 For Pi in Othe than Recirculation Pi~i'ystem Analyses are performed for the pipe failure postulated in Subsection 3.6 2
0 Analysis of get thrust forces vhich result in the event of a pipe rupture are described in Section 2.2 of Reference 3 6-2 Plaid get iapinqement forces are discussed in Section 2
3 of Reference
Zmpulsi ve loadinq and iapact combined with impulsive loading are described. in Sections 3.2 and 3.3 respectively of Reference
Zvw<0 A Attachment F
SSES-PSAR Page 2 of 7
g,6 )~ nn d, t=i'g 3 6" l2 4- ~
!. !~ !
The criteria for the dynamic analyses are as follovs:
1)
An analysis of the piping system is performed for each longitudinal and circumferential postulated rupture at the break locations determined in accordance vith the criteria of Subsection 3.6 2.1.
2)
The loading condition of a piping system prior to postulated rupture in terms of'nternal pressure, temperature, and stress state is that-condition associated vith reactor operating at 100 percent po vere 3)
For a circumferential rupture, pipe vhip dynamic analyses are performed only for that end (or ends) of the pipe 'or branch that is connected to a contained fluid energy reservoir having sufficient capa'city to develop a get stream.
4)
Dynamic analytic methods used for calculating the pipinq and pipinq/restraint system response to the pipe break forces adequately account for the effects of:
!')
Translational masses (and rotational masses for ma)or components) and stiffness properties of the piping system, restraint system, major components, and support vali's j
b)
Transient forcing function(s) acting on the pipinq system
Elastic and inelastic deformation of piping and/or restraint d)
The design'learance betveen the pipe and the restraint.
I Piqures of main A 10 percent increase of minimum specified design yield strength (Sy) is used to account for strain rate effects in inelastic nonlinear analyses.
3 6-5 through 3.6-7 shov the summary of the analyses steam, feed vater, and HPCZ lin'es for postulated pipe j',
Attachment F
Page 3 of 7 Insert 'A' typical piping system model used in the dynamic analysis is provided'on Figure 3.6..ll(a).
Protection against the pipe'hip is accomplished by restraining the motion of the pipe after pipe break.
The pipe whip restraints are designed with energy absorbing components, ie., crushable honeycomb, in the direction of the pipe whip.
Crushable honeycomb limits the reaction load in the whip restraint in most cases to about 80% of the design yield load for the restraint and absorbs the energy to greatly reduce the tendency of the pipe to rebound after impact.*
When the required energy absorption is too great to be entirely accomplished by the honeycomb, the plastic deformation capability of the whip restraint itself is taken into account.
The structural steel whip restraint is permitted to have plastic deformation that results in ductility ratio no greater than 20.~*
For structural steel subjected to shock and impact loading, ductility ratio of 20 is an acceptable practice (Reference 3.6-9).
Reference 3.6-8 was used in determining the response of the piping system under pipe break loads.
Energy absorption capacity of the honeycomb associated with crushing up to
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60% of its original height is used in the design calculations.
The load deflection curve in this region is relatively flat.
~* Ductility ratio is defined as plastic strain (deformation) divided by the strain (deformation, at yield strength of the material.
SSES-FSAR 3
6.4 REFERENCES
Attachment F
Page 4 of 7
,'/
F.J.
Moo'dy, Fluid Reactions and Impingement
3 6-2 3 6-3 3 6-4 "Design for Pipe Break Effects<<,
BN-TOP-2A, Bechtel Power Corporation.
GE Spec.
No.
22A2625 "System Criteria and Applications for Protection Against the Dynamic Effects of Pipe.
Brea k<<
Relap 3 -
A Computer Program for Reactor Blowdown Analysis XN-1321, issued June
3 6-5 3 6-6 GE Report NEDE-10813 "PDA - Pipe Dynamic Analysis Program for Pipe Rupture Movement<<.
(Proprietary Piling)
Nuclear Services Corporation Report No.
GEN-02-02
<<Final Report Pipe Rupture Analysis of Recirculation System for 1969 Standard Plant Design".
A computor program for transient Thermal Hydraulic analysis of Nuclear Reactors and Related
I PIPRUP A computer program for pipe Rupture Analysis, developed by.nuclear services corporation (1977),
'/
"Design of Structures for Missile Impact",
BC-TOP-9A, Revision 2, Bechtel Power Corporation, September 1974.
ll, 7/79 3 6-34,
'I
~
Attachment F
Page 5 of 7 clearance honeycomb APE C t.l
'3
'0 pipe in Amid.
conditian 8'ipe in 1xrt cnxlditicn pipe in cold condition a)
. Pipe whip restraints with honeycomb (EAC) and clearance all around the pipe
(.,y b)
Bumpers SUSQUEHANNA STEAM ELECTRlC STATlON UN(TS 1 AND 2 FlNALSAFETY ANALYSISREPORT TYP1CAL P1PE NHiP RESTRAlNTS FlGURE 3.6-10.
0
N
~ ~
F'R-9 S'<
PR 35-tl~
hlAF'shc7oR r/ozzLE D)O 8
ID AC W M W
0 w I
~78 Ol
'x Oa C'l 0
Z mz I
ol ZZ m CZ cn m
+am M O CO m
o 0Z
+p 8NAK Locsrtel
~ BLoNDA!dyi]gdsr
, Rs'Wcrrea Adcu~C Fug AuALyszs'uRPo5F'aul y'
Attachment F
Page 7 Qf 7 CO Pi P~ ~ 121,230,LBF 6 t ~.0024 sec 50 I
O O
COa F..~
73,440 LBF 9 t= 0.0 sec F ~
=77i000 LBF ste-~
0
.04
~ 08
.12
.16
.20
..24
.28
.32 TIME SECONDS HPCI INSIDE CONTAINMEHT PIPE BREAK FORCING FUNCTIONS REACTOR SIDE
(
SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 FINALSAFETY ANALYSISREPORT FORClNG FUNCTIONS MODE!.
AUX:IATEDWITH PIPE WHIP DYNAMIC ANALYSIS FIGURE f.e-12 ~
Attachment G
SSES-FSAR
'Page 1 of 2 6 6 7 SYSTEM PRESSURE TESTS System pressure tests will meet the requirements, of IQC-5000 IQD-5000 of Section XI.
and "
6 6 8
hUG5ENTED IHSERVICE INSPECTION TO PROTECT AGAINST POSTULATED PIPING FAILURES The prese vice/in-service inspection program complie with the AS5E BGPV de,Section XI, 1974 edition, includin addenda through Summ
, using 50% of the reference 1
el as criteria for investigat ng reflectors.
The following augm ted inspection program pplies to piping between the containm nt isolation valves r which no breaks aze
.postulated:
1)
For ASIDE III, Class 1
iping t "equirement of the above-referenced Code applie as is.
2)
Fo'r ASIDE III, Class 2 pip@ g the requirement of the above-referenced Code applies as
, except that 100'%f the circumferential welds wi in he containment isolation boundary receives 100%
oluxet c examination during each inspection interval.
hese weld are included in the 50% of the total.n'umber of elds for exa 'nation category C-G.
The following branch-o-main zun welds wo ld not reveal
'meaninfgul volumetr'c examination results e to fitting
a)
Meld let couplings b)
Ha X'rrzc~Y R he augmented inserrice inspection program to provide assuranc a
inst postulated piping failures of high energy systems bet een con ainment isolation valves vill be reviewed and implement as descr ed below.
There a e no guard pipes used to enclose h gh energy pipin on the Susquehanna SES c)
Socket welds None f t'e piping between
+he containment isolation val s for whi no breaks are postulated involves a longitudinal wel
.'EV ll, 7/79
r Attachment G
Insert A
Page 2 of 2 AUGMENTED INSERVICE INSPECTION TO PROTECT AGAINST POSTULATED PIPING FAILURES The augmented inservice inspection program to provide assurance against postulated piping failures of high energy systems between containment isolation'alves will be-reviewed and implemented as described below.
There are no guard pipes used to enclose high energy piping on the Susquehanna SES.
The following augmented inspection program applies to piping between the containment isolation valves for which no breaks are postulated:
FOR ASME III, Class 1 and 2 piping, the requirements of the applicable Code applies as is, with the exception that the extent of examination will be augmented such that 100X of the circumferential welds within the contain-ment isolation boundary will receive 100/ volumetric examination during each inspection interval.
Volumetric examinat'ion of branch connections containing weldolets, half-couplings, and socket welds would not be meaningful due to the geometry of the branch connection and the small pipe sizes involved.
Full coverage of
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the weld and required volume cannot be obtained.
Therefore, surface examination will be performed on all branch to main run welds and all socket welds up to the first isolation valve on the branch line.
All butt welds included in the branch piping up to the first isolation valve will receive full volumetric examination.
The inspection program will be performed, completely in accordance with ASME Section XI requirements,
R.G.1.92 WAS MOT A DESIGN CRASIS REQUIREMENT FOR THE CONSTRUCTION PERMIT FOR SUSQUEHANNA PLANT THE CRITERIA EMPLOYED OM SUSQUEHAMMAIS CONSERVATIVE AMD PROVIDES AM,EQUIVALENTLEVEL-OF SAFETY AS THE CURRENT NRC CRITERIA-
COMPARISO OF THE CRITERIA SSES CRITERIA CURRENT NRC CRITERIA MODALRESPONSE COMBINATION DAMPING SPATIAL COMPONENT RESPONSES
RESPONSE
SSE 1%
SRSS Q ENVELOPE OF
RESPONSE
SPECTRA AT DIFFERENT SuPI ORTS @
R.G. 1.92, C1;2.
OBE 1% B 2%
SSE 2% &3%
SASS ENVELOPE OF
RESPONSE
SPECTRA AT DIFFERENT SUPPORTS k, FOR SSES COMBINATIONOF TIRE LARGEST HORIZONTAL RESPONSE AND VERTICALRESPONSE USING ABSOLUTE SUM HAS BEEN ACCEPTED
@ RESPONSE SPECTRA AT OR ABOVE THE CENTER OF MASS OF THE PIPING SYSTEM HAS BEEN ACCEPTED FOR SSES
JUBTIFICATIOM A STUDY WAS CONDUCTED TG DEMONSTRATE THATTHE CRITERIA EMPLOYED OM SSES PROVIDES AM EGUIVALEMYLEVEL GF SAFETY AS THATWITH CURRENT MRC CRITERIIA STUDV CONSISTED GF:
A. FOUR PIPING SYSTEMS INSIDE CGMTAIMMEMT
~
. 4. RHA SPRAY HEADER B. FOUR.PllPIMG SYSTEMS OUTSIDE CGMTAIMIttIEMTFOR GBE AND SSE.
THE EIGHT PII IhlG FOR TWO CASES:
CASE A STEMS WERE ANALYZE CLOSELY SPACED MODALRESPONSES ARE COMBINED BY SRSS DAMPINGVALUES OF 1/2% FOR OBE AND 1% FOR SSE RESPONSES OF THREE SPATIAL COMPONENTS OF MOTION ARE COMBINED BY THE SRSS METHOD CASE 8 CLOSELY SPACED MODALRESPONSES ARE COMBINED BY GROUPING METHOD, REGULATORYGUIDE 1.S2, PARA C1.2.1 DAMPINGVALVESOF 1% FOR OBE AND 2% FOR SSE RESPONSES OF THREE SPATIALCOMPONENTS OF MOTION ARE COMBINED BY THE SASS METHOD (FOR PIPING OVER 12" DAMPINGVALUES OF 2% FOR OBE 3% FOR SSE ARE ACCEPTABLE)
CHARACTERISTICS F THE PIPING SYSTEMS STUDIED ARE SUMMARIZEDIN THE TABLE BELOMf SYSTEM NO. OF NO. OF CLOSELY NO. OF FIRST PIPE MODES SPACED RESTRAINTS MODE SIZE NO.-OF BELOW MODES AND NO. OF FREQ.
INCH D.D.O.F.
33 HZ GROUPS ANCHORS VALVES HZ FEEDWATER 12 & OVER 231 RWCU HEAD SPRAY UNDER 6 646 17 6
300 17 RHR.,
SPRAY HEADER OVER 12 132 60 34 0
10A o"
7.1 2
169 10 10.3 RHR SERVICE.j VIATER 68 20 147-10 7
14 "
6.3
'ETURN LIMES 2 TO '14 262 21 30 8A ESVV SUPPLY LIMES RHR PUMP DISCHARGE 2 TO 14 231 17 6 TO 24 396 24 10 30 9.6 9.6 0
C THE RESULTS OF CASE A AMD CASE 8 WERE COMPARED FGR EACH PIPING SYSTEM STUDIED. THE FOLLOWING MAXIMUM RESPONSES WERE COMPARED:
FORCES AMD MGMENTS IM EACH DYNAMICRESTRAINT/ANCHOR THE RESULTANT VALVEACCELERATIGMS
l UDY CASE A RESPONSE EXCEEDS CASE 8
RESPONSE
FOR SYSIEH
,FEEDHATER RI",CU HEAD SPRAY RHR SPRAY lkEADER, RHR SERVICE WATER ES>l RETURN LINE ESI'l SUPPLY 'LIflE RflR PUNP DISCHA(GE e
PIPE
~'KK ALL ALL ALL ALL ALL
='LL ALL.
ALL ALL ALL ALL ALL ALL
'LL ALL ALL ALL ALL
'ALL ALL.
ALL-Ql ALL'LL'~1' NO.
OF NO, OF 8DI/O"'ML V$~LLVS I
. i Qi:
FOR Tl{O VALVES THE CASE a RESPOllSE EXCEEDS CASE A RESPOHSE FOR SSE OtlLY, THIS RESPONSE IS tlOT CONTROLLIt>G
~
~
SEISMIC CRITERIA USED IM PIPING
'NALYSISOM SUSQUEHANNA PROJECT IS CONSERVATIVE AMD PROVIDES AN
. CURRENT NRC CRITERIA.
Attachment I SS ZS-FS AR Page 1 of 5
-3 7b.3 1.2 Piscina Svstems x'ngcI-X '
.*.3 3.7b. 3 1.3 Class IE Cable Trays The cable trays are seismically qualified by, the capacity evaluation method which consists of the followinq:
a)
Calculation of the fundamental frequency of the cable tray based on the tray properties obtained from static tests b)
Seismic load computation based upon the tray frequency and the design spectra
")
Calculation of the tray allowable capacity d)
Evaluation of the tray. capacity by interaction formulae 3.7b.3.1.4 Supports for Seismic Category I HYAC Ducts and Cable Trays Tpe supports for HVAC ducts are analyzed by the response spectrum method.
The supp'orts for cable trays are based on response spectrum meth'od and/or experimental data acquired. from, actual seismic testing performed on cable tray and support systems.
c)
. Section 7
1 8 of the Design Assessment Report (DAR) describes the methods used for dynamic analysis of the raceway system.
3 7b 3
2 Determination of Number of Earthquake Cycles Equipment that", is qualified by analysis is designed to remain elastic during, the earthquake.
Any 'fatigue effects in tested equipment are 'accounted for by the duration of the test.
c>>
I\\
3 7b-13 Rev.. 16, 7/80 In general, tPe design of the equipment is not fatigue controlled because the eQuipment is elastic and the number. of cycles in an ea rthqua ke is lo w.
Attachment I Page 2 of 5 Insert C
-TOP-l, Rev.
3
( Ref. 3.7b-6
) describes the, methods used for seismic analysis of piping systems.
Reference 3.7b-6 is followed on Susquehanna SES with the following exceptions":
In seismic analysis the modal responses are combined by SRSS and lower damping values than specified in Reference 3 'b-6 are used.
See Subsection 3.7b.3.7.
Attachment I Page 3 of 5 The.criteria used for.combining the, results of horizontal and vertical seismic responses for piping systems aze:described in ction 5. 1 of Reference 3.'7b-6
,/I 3
7b 3
7 Combination of Nodal Bes2onses The modal responses of equipment are. combined by the square root
.of the sum of the squares method.
The absol ute values of tvo closely spaced modes are added first befor'e combining vith the other modes by the sguare root of the sum of the squares method Tvo consecutive modes are defined as closely spaced vhen their frequencies differ from each other by,lO percent'or less.
procedures given in Regulatory Guide 1.92 foz combining modal responses, vhen closely-spaced modes are present, are not complied vith in the seismic response spectra analysis for piping.
All modal responses are combined, by square root of sum of squazes (SRSS) in the response spectra method of modal analysis for seismic loading (OBE and SSE)
Seismic response spectra used in the piping analysis corresponds to conservative damping values of 1//2g for OBE and 1% for SSE.
k
'he procedures used in evaluating the piping system: for hydrbdynamic loads (SRV. and LOCA} by response spectra method is in compliance vith Regulatory Guide 1.92.
The modal responses in this case are combined in accordance,vith.section 5
2 of BP-TOP-Hev
)-,- vhich has been accepted by the HRC staff, per'he-ter dated September 29, 1976, from Karl Kniel Chief Light tez Reactors Branch Ho 2, Division of Project management to
~ ~~ L. Lex ~ Bechte1 Pover Corporation Serac~
/,
The criteria used; for piping systems are described in Sections 5
1 and 5 2 of Reference 3 7b-6 3 7b.3 8
Analytical Procedures for Pi in
/
The design criteria and the analytical procedures applicable to piping systems are Ias described in Section 2.0 of Reference 3 7b-6 The methods used to consider differential piping support movements at different support points are as described in Section 0 of Reference 3 7b-6
~,
Attachment I Page 4 Qf 5 REFERENCES N.C. Tsai ~ <<Spectrum Compatible Motions for Design Purposes",
Journal of Enqineering Mechanics Division, ASCE, Vol. 98; No.
EM2, Proc, Paper 8807 (April 1972),
pp '45-356 N.M. Nevmark, "Design Criteria for Nucleh.r Reactors
= Subject to Earthquake Hazards<<,
Proc IAKA Panel on Aseismic Des~i n and Testing of Nuclear Facilities, Japan
. Earthquake Enqineerinq Promotion Society, Tokyo, Japan
.(1967).
"Seismic Analyses of Structures and Equipment for Nuclear Power Plants<<,
BC-TOP-4A, Rev 3, Bechtel Pover Corporation, San Francisco, California (November 1974).
Uniform Building Code (UBC), by International Conference of Building Officials, Mhittier, California, 1970 Edition.
A.T. Derecho, D. M. Schultz, and M. Fintel, Analysis and Des~in of Small Reinforced Concrete Buildings for Earthguake
8 "Seismic Analysis of Piping Systems",
BP-TOP-1, Rev g, Bechtel Power Corporation, San Francisco, California (Ja nua ry 197g).
3.7b-23
Attachment I SSES-PSAR Page 5 of 5 TABLE. 3.9-15(Page 20 of 21) tBECHTEL CRITERIA)Q 1/2 SSE Cycles (Operating Basis Earthquake)
Condition Upset Expected number of equivalent 1/2 SSE in life of pipe system Average duration of strong motion vibration 1/2 SSE Average number of maximum seismic load. cycles of pipe system for each 1/2 S SE Total lifetime number of maximum seismic load cycles of piping system 5
15 sec lo 50 SSE Cycles (Design Basis Earthquake)
Condition-Paulted
~
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~
~
~
Expected number of equivalent SSE in life of pipe system Average duration of strong motion vibration SSE
,j/l'5 sec I
Rev.
12, 9/79
-Attachment J
Page i of I q rouped into the two representative lengths shown.
These I.engths represent the lonqest and shortest 'housings in order to adeguately represent the full ranqe of frequency response of the
~
~
~
housinqs.
The hiqh fundamental natural freguencies of the CRD housinqs results in very small seismic loads.
Furthermore, the small frequency differences between the various housings due to
.the. length differences result in neqligible differences in dynamic response.
Not included in the mathematical model are light components such as jet
This is done to reduce the complexity of the dynamic model. If the seismic responses of these components are needed, they can be determined after the system response has been found.
The presence of a fluid and other structural components
{e.g.,
fuel within the RPV) introduces a dynamic couplinq effect.
Dynamic effects of water enclosed by the RPY are accounted for by introduction of a hydrodynamic mass matrix, which will serve to link the acceleration terms of the equations of motion of points at the same elevation in concentric cylind'ers with a fluid entrapped in the annulus.
The details of "the hydrodynamic, mass derivation are qiven in Reference 3.7a-1.
The seismic model of the RPV and internals has two horizontal coordinates for each mass point considered in the analysis.
The remaininq translational coordinate (vertical) is excluded because the vertical frequencies of RPV and internals are well above the significant horizontal frequencies.
Futhermore, all support structures, buildinq and containment.walls have a
common centerline, and
A separate vertical 'analysis is performed.
Dynamic loads due to vertical motion are added to or subtracted.from the static weight of components,
'whichever. is the more conservative.
The two rotational coordinates about each node point are excluded because the ~~~ con tribution of rotary inertia Since a11. deflections are assumed to be within the elastic range, the riqidity of some components may be accounted for by equivalent linear sprinqs.
5 Y7cp it/I l/Q.
The shroud support plate is loaded in its own plane during a
seismic event and hence is extremely stiff and therefore may be modeled as a rigid link in the translational direction.
The shroud support leqs and the local flexibilities of the vessel and shroud contribute to the rotational flexibilities and are modeled as an equivalent torsional spring.
Attachment K
Page 1 of 1
Transient Category Cycles 8.
Operating BasisEarthquake+
9.
Safe Shutdown Earthquake+<
10.
Scram with inoperative buffer 11.
Scram with stuck control blade normal/upset'aulted normal/upset normal/upset 10 1
10 All ASME Class 1 components of the CRD have been analyzed accordinq to ASME Section III Boiler and Pressure
.Vessel Code.
The capability of the CRD ~ s to withstand the emergency and faulted conditions is verified by test rather'- than analysis.
3 9.1.1. 2 CHD Housi~cC and ?score Housing Transients Cycles Category The number of transients, their cycles, and classification as considered in the design and fatigue analysis of the CRD housing and incore housing are as follows:
I Transient 120 130 10 80 normal/upset normal/upset normal/upset normal/upset normal/upset no rma 1/upset emergency/faulted l.
Normal startup 6 shutdown 2.
Vessel pressure tests 3
Vessel overpressure tests 4.. Interruption of feedwater f1'ow 5.
Scram 200 6
OBE* (Operating Basis 10 Earthquake) 7.
SSE+~
(Safe,.
Shutdown 1
Earthqua k'e ';
CED Housing Only 8.
Stuck Rod Scram normal/upset 9.
Scram no Buffer normal/upse t
/
The freguency of occurrence at this transient would indicate emergency category.
However, 'for conservatism, this OBE condition was analyzed as an upset condition.
The single event is assumed to consist of 10 cycles.
SSE is a faulted condition; however, in the stress analysis report it &as treated as emergency with lower stress limits.
r Rev. 12, 9/79 3 9-2i
A'ZTACHNENT L Page l of ll
'I qenerate loadinqs.
This analysis utilizes appropriate seismic floor response spectra and combines. loads at frequencies up to 33 HZ in three directions.
Imposed stresses are qenerated and combined for normal ~ upset, and faulted conditions.
Stresses are
3.9.1. 0. 12 Seismic C~a:e~oy I Items Othe~than HSSS For statically applied loads, the stress allowables of Appendix "F of iASNE Section III, Minter 1972 were used for code components.
For noncode components, allowables were based on tests or accepted standards consistent with those in Appendix F of the code.
Dynamic loads for components loaded in the elastic range were calculated using dynamic load factors, time history analysis, or any other method that assumes elastic behavior of the component.
The limits of the elastic ranqe are defined in Pazagraph 1323 of Appendix F for the code components.
The local yielding due to stress concentration is assumed not to affect the validity of the assumptions of elastic behavior.
The.stzess allowables of Appendix F for elastically analyzed components were used for code components.
Foz noncode components, allowables were based on tests or accepted material standards consistent with those in Appendix F for elastically analyzed components.
The methods used in evaluating the pipe break effects aze discussed in 'ection
~3.9.
DYNANEC TgSTENG AND ANALYSES 3.9.2.la Preoperational Vibration and Dynamic Effects Testing on Pipit The test program is divided into three phases:
preoperational vibration, startup vibration, and operational'transients.
3 9.g,la.l Prestoerational Vibration Testing The purpose of the pzeoperational vibration test phase is to verify that operatinq vibrations in the recirculation aaaMhH9 xm~xr pipinq are aiMku acceptable Rim~.
This phase of the test uses visual observation+
3 9-27
ATTACHMENT L Page 2 Of ]g a)
Recirculation
/
s minimum urxnq s ea y s ate opera
Visual observations, manual an mote.
measurement i,ll be made during the following st state conditions:
b)
Reci'rculation pumps of rated flov; c)
Recirculatio mps at 75A of d flov; d)
Reci ation pumps at 100% of rated,'
I e
RHR suction pipinq at 100% of rated;flov in the cooling mode.
tdow 3 9.2.la.2 Small attach~ed Pi in'ua o servatxon o
eac o
e a
(a) throuq ecial attention vill be give attached piping an ment connect' nsure that they are not in resonance vith t ation pump motors or zlov ind'uced vibrations.
peratin ion acceptance criteria ar appropriate corrective a
'll be t
S tartup Vibration The purpos~e yf this phase of the program is to verify that the main steam recirculation piping vibration are,'within acceptable limits.
Because of limited access due to/'high radiation levels, no visual observation is made during this phase of the test.
Remote "measurements shall be made. durinq the'ollovinq steady state conditions:
(a)
Mairi steam flov at 25'5 of rated; (b)
Main s'team.flov at 50% of rated; I
(c)
Main steam flov at 75% of rated; (d)
Main steam flow at'00% of rated; (e) ov in the shutdovn (f)
RHR suc x.nq mode.
CIC turbine steam line at 100
/
at 100%
o 3 9-28
TTACFBKNT L 3
9 2 la. 4 02enating Tcansient Loads page 3 of 11 The purpose of the operatinq transient test phase is to verify that pipe stresses are within Code Limits.,The amplitude of
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~
displa'cements and number of cycles per transient of the main steam and recirculation pipinq will be measured and the displacements compared with acceptance criteria.
The deflections are'orrelated with the calculated defi'ections to assure that the stresses remain within Code limits.
Remote vibration and deflection measurements shall be taken during the folloving transients:
a)
Recirculation pump starts; b)
Recirculation pump trip at 100% of rated flow; c) d)
Turbine stop valve closure at 100% power; REPRESS BH747 /VH (g)
Manual discharge of ~ash S/R valve~at 1,000 psig and at planned transient tests that result in S/R valve discharge.
~92 14.5 Test valnaLion and acc~etance Criteria The piping response to test conditions shall be considered acceptable if the o'rqanization responsible for the stress report revieys the test results and determines that the tests verify that the pipinq responded in a manner consistent with the
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~
~
~
~
~
~
predictions of the stress report and/or that the tests verify'hat pipinq stresses are within Code limits.
To insure test data integrity and test safety, criteria have been established to facilitate assessment of the test while it is in progress.
These criteria,'esignated Level 1 and 2, are described in the followinq paraqraphs.
~92.1aa5.1 Level 1 CLitenia Ef in the course of the tests, measurements indicate that the piping is respondinq in a mannner that would make test termination 'prudent, the test shall be terminated.
Level 1
criteria establishes bounds on movement that, if exceeded, make a test hold or termination mandatory.
The limits on movement are based on maximum allowable Code stress limits.
3 9-29
ATTACHMENT L Page 4 of 11 SSES-PSAR seismic event by dynamic analysis only Piping systems having
'ignificant anticipated transients
In addition, piping steady state vibration and dynamic transient tests vill be performed as summarized belov, to ensuze that a)
Excessive steady state vibration is not present in the piping that would zesult in piping stresses and restraint loads above the allovables b)
The piping is adequately restrained'" to withstand the dynamic transient loads.
Cognizant design personnel familiar with the systems to be tested Mill develop the test plans, and evaluate the test results.
Also the cognizant design personnel vill vitness the test.
The data acquired fzom the tests vill be compared vith the expected results to determine the acceptability of the total system res ponse.
A list of all piping systems in the BOP is provided in Table 3.9-33.
ASFiE Section III Class 1, 2 and 3 piping systems, high energy piping systems, moderate energy piping systems, seismic Category I and seismic Category II systems aze identified in the Table.
%he Table also identifies the tests to be performecd foz each system.
Piping thermal expansion tests are performed for the safety-related piping systems with normal operating temperature exceeding 300 oP.
Safety related piping, systems with normal operating temperature less than 300 oP do not have significant thermal expansion to varzant thermal expansicnytests.
Engineering ze'vieM of all seismic Category I "piping systems including the'ir supports, restraints or snubbers is performed after completion of construction and prior""z,o fuel load to ensure that no restraint of normal thermal movement occurs due to interferences and obstructions and that the suppozt and restraints are, in accordance vith the design intent.
1'or the systems receiving thezmal expansion tests, the pipe movements are monitored to ensure that no restraint of normal thermal movement occurs at locations other than at the designed restraint locations.
The thermal expansion test program verifies that t'e free thermal expansion of piping systems take place at the snubbers by monitoring thd'hermal movement.
Performance of the snubbezs designed for transient loads such as due to Hain Stop Valve Closure or Hain Steam Relief Valve Discharge are verified by measuring the load in the snubber during the dynamic transient tests.
The snubbers are qualified by dynamic testing for cyclic loading as described in Subsection 3.9.3.4.1
Ill CP
ATTACHMENT L.
SSES-FSAR Page 5 of 11 Acceptance criteria for thermal expansion tests and dynamic transient tests is that the measured pipe displacements or restraint loads shall be below the calculated or design values.
y',r Acceptance criteria for the ste'ady state vibration tests is:
Either The maximum measured amplitude of the piping vibration shall not induce a stress in the pipe more than the endurance limit of the material.
By limiting the maximum stress in the pipe'ue to steady state vibration below one-half of the endurance limit (allowable stress corresponding to 10 cycles in Appendix I of ASME Section III), the steady state vibration induced stress will not contribute to the reduction of fatigue life of piping.
Or Acceptance criteria are divided into two categories, i.e., Level 1 and Level 2.
If Level 1 criteria are violated, the test must be placed on hold. If Level 2
criteria are violated, the test can continue, but the measurements must be evaluated to verify that continued test operation will not result in exceeding piping fatigue requirements.
For steady state vibration the piping peak stress zero to peak due to vibration only (neglecting pressure) will not exceed 10,000 psi for Level 1 criteria and 5,000 psi for Level 2 criteria.
These limits,.are below the piping material fatigue endurance limits as defined foi'0 cycles in Appendix I of ASME Section III.
The Table 3.9-33 provides cross reference between the FSAR Section 3.9
~
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~
~
and'the appropriate.teat;description in FSAR Chapter"14'."
3.9.2.1b.l Piping Dynamic Transient Tests During the preoperational and/or startup testing the following piping dynamic transient tests will be performed for the indicated modes of operation.
a)
Main steam piping outside the containment for main steam turbine stop valve closure at 50 percent, 75 percent and 100 percent power b)
Main steam bypass piping to the anchor near the bypass valves for the turbine stop valve closure c)
Selected main steam safety/relief valve discharge piping for the main steam relief valve opening.
The selected SRV discharge piping brackets all the SRV discharge piping.
d)
HPCI turbine steam supply piping for HPCI turbine trip.
3.9-32a
I II
SSES-FSAR" age o
11 From past experience, the dynamic transients in other piping systems are not significant.
1 Dynamic transient analysis of the subject lines except for'e) above has been performed to determine the response of the piping system and the restraint loads.
During the test the displacement of the pipe, loads in the snubbers and restraints and pressure at representative locations will be measured.
Acceptance criteria for this test are that the measured loads in the snubbers and restraints shall be below the design values of the snubbers and restraints.
In the case (e) the acceptance criteria is that the measured response shall be less than the acceptable response determined by analysis.
m
. 2 1b 2
~Pi ing Ste~ad State Yibnatinn Testing Page 7 of 11 The piping sy tern associated with the following components operation vill be observed for steady state vibration during preoperational test programs or pover ascension:
a)
RHR pump e)'
ar'~
S7 e>r~
b)
HPCI pump and turbine I
c)
RCIC pump and turbine d)
Core spray pump 4)
Fee Jwe he~
cJ )
/Pean 4r 'Ale4>>
C /LQ>> c/
Prom experience on other nuclear power plants, the steady state vibration in other piping systems is not critical.
Hovever abnormal vibrations of other systems during system valkdovn on
'initial startup or pover escalation will be noted and instrumented if necessary to determine the acceptability of such vibration.
Steady state vibration in the subject piping systems is primarily induced by the flov in the pipe and the equipment motion.
In
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~
mohes of system opera ion and classy y each line as acceptable if the vibration is not siqnificant, or questionable if vibration is siqnificant.
The lines
- vith questionable steady state vibration vill be monitored by suitable instrumentation to determine the system response.
I rTTOn i+r tiT~cess f &le j:<<>
d~,.p4
< ~i'r'a4 le r
2 sr ru~ en 4a r'ran The type of the instrumentation, if necessary, vill be determined by the desiqn enqineer so that the aiaximum amplitude and frequency response of the pipinq system can be determined.
The instrumentation vill not screen out the significant frequencies.
For lines vith questionable steady state vibration, the acceptance criterion i pipinq.
This is applicable to nuclear and non-nuclear
+-S 4/Sc'gcCed in" Su j~<~<<<ST g,er, Z'g, Qhen required, additional restraints vill be provided to reduce the steady state vibration and to keep the stresses belov the acceptance criteria le vels.
3 9-33
W0 TABLE 3.9-33 BOP-PIPING SYSTEMS POWER ASCENSION TESTING
~Pa e
2'i in S stem Code(s)/
S.C/H.E M.E (1)
Thermal Dynamic Expansion Transient Tem
>200 F Test (2)
Test (3)
Steady State Vibration Test (4)
Remarks ocess Sam lin lorination ressed Air strument Gas ed Pump Turbine earn keup Water lve Stem Leakoff cid Injection 831.1 SC IX M.E a831.1, SC II
'ME 831.1, SC II ME ASME III-2,3 931.1, SC I SC II ME 831.1, SC XI H.E 931.1, SC II, kK 931.1, SC II, HE 831.1, SC IX ME No No No No Yes.
No Yes No N A,~
N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A a
N/A N/A N/A N/A N/A N/A N/A
'/A N/A dro en Stora e
iesel Engine Auxiliaries 931.1, SC II lK ASME XII-3 831.1, SC I 6 II ME No N/A N/A N/A Yes Yes g)
N/A N/A
(
S e e Rema r k s ).-
Emergency Diesel Exhaust Has T >300~F and Thermal Ex ansion Test Performed.
SIV Leakage Control eactor Recirc otor / Generator igh Pressure oolant Injection ev.
12 9r, ASME III-122 931.1, SC I SC II HE 931.l, SC II ME ASME III-1,2 831.1, SC I SC II, H.E, 1K Yes No Yes N/A Yes
~
a N/A N/A Yes N/A N/A Yes HPCI Turbine Stop Valve Closure Transients for Steam Supply, Steady Stat Vibration for Steam Suppl p~+p. /Qp wgqim J
TABLE 3.9-33 BOP PIPING SYSTEMS POMER ASCENSXON TESTING gA'cWi~
p rr/4lise 4p'r p'p
~~PM&.~+ viP~HA 7 c ~F
~Pa e 3 Pi in S stem actor Core Isolation olin Code(s)/
S.C/H.E M.E (1)
ASME XII"1,2 B31.1, SC I SC XI HE ME Yes Yes N/A Thermal Dynamic Expansion Transient Tem
>200 F Test (2)
Test (3
Stead
.State Vibrati Test
(
Yes Remarks teady State Vibration for R
/team Supply and Turbine Rxhaust
~C actor Mater ean up ASME III-1,2 B31.1; SC I SC IX, HE, ME Yes Yes N/A Yes Steady State Vibration for 18KQIjge Inside Containment sidual He'at moval Leanup Filter
.'mineralizer ASME IIX-1,2, 3, B31:1, SC IRII, IK,
. ME ASME III-2 B31.1, SC II ME Yes Yes N/A
( S e e Remarks)
N/A N/A N/A Major ity of the Sys tern has Normal Operating Temperature less than 300~F Thermal Expansion Tests are done for SCI Systems MAA T >300'F.
Steady State Vibration for Inside Containment Piping and RIB Pump Dischar e.
control Rod Drive ASME III"2 B31.1, SC I SC II ME No N/A N/A N/A tandby Liquid Control ASME XII-1,2 B31.1, SC I 6 IX, HE, ME No N/A N/A (See Remarks)
N/A Only a small portion of the Line near RPV has Tem erature
>200~F nre Spray ASME III-1,2 B31.1,.
SC I 6 II HE ME Yes N!A Yes Steady State Vibration For Core Spray Pump Dischar e
-=0 z
0 Cg TABLE 3.9-33 BOP PIPING SYSTEMS POWER ASCENSION TESTING P~ae 4
Pi in S stem Code(s)/
S.C/H.E M.E (1)
Thermal Dynamic Expansion Transient Tem
>200 F Test (2)
Test (3
Steady State Vibration Test (4)
Remarks uel Pool Cooling, leanup Zc Demineralizer ASME III"3 B31.1, SC I SC -I.I ME No J
N/A '~
N/A N/A ontainment tmos heric Control olid Radwaste and adwaste Solidification ff-Gas Recombiner bient Temperature harcoal Off-Gas reatment hilled Water ASME III-2 B31.1, SC I 6 II ME B31.1 SC II ME ASME III"3 831.1, SC I SC II H.E 831.1, SC II ME ASME III-3 B31.1, SC I SC II
>iE No No Yes No No N/A N/A jU/'/5 N/A N/A
'N/A N/A N A N/A N/A N/A N A N/A Rev.
12 3
ATTACHMENT L Page 11 of 11 TABLE 3. 9-33
~Pa e5 NOTES (1)
Code(s):
ASME lII Boiler and Pressure Vessel Code, -1, -2, or
-3,Denotes Nuclear Class 1,
2 or 3 Piping.
S.C I or II Denotes Seismic Category I or II H.E:
Denotes High Energy Piping System i.e. Pressure
>275 PSI or Temperature
>200 During Normal Plant Operation M.E:
Denotes Moderate Energy Piping System (2)
Thermal Expansion Tests for the indicated systems corresponds to test, descriptions.-'R-,
Chapter 14.
sT-3'R (3)
Dynamic Transient Tests for the indicated systems corresponds to test descriptions~
Chapter 14.
s7=ER (4)
Steady Stage libration Tests for the indicated systems corresponds to test description~-,
Chapter 14.
5~ /0 (5)
N/A - Denotes Not Applicable and it means test is not performed for the below reasons:
I A)
For Thermal Expansion Rests:
The"system is not, safety-related.or the normal operating temperature is less than 300 F.
B)
For Dynamic Transient Test:
The system is not safety-related or does not experience any significant transients.
C)
For steady State Vibration Tests:
The system is not safety-related or no significant vibration expected.
(-
Eav. 12, a9/79
f2)
Attachment, M
, I
'Data from previous plant vibration measurements is assembled and examined to identify predominant vibzation response modes of major'components En'general, zespon-e modes are similar but response amplitudes vary among BMBs of differinq size and design.
l33 Parameters are identified vhich are expected to influence vibration response amplitudes among the several reference plants.
These include hydraulic parameters such as telocity and steam flov rates, and structural parameters such as natural frequency aad siqaificaat dimensions
{4)
Correlation functions of the variable p'arameters are devel oped vhich, multiplied by res ponse'mplitudes, tend to minimize the statistical variability. between plants correlation function is obtained for each major component and response mode.
I (5)
Predicted vibration amplitudes for components of the prototype plant are obtained from these cozzelation functions, based on applicable values of the parameters for the prototype plaat.
'Zhe preditect a mplitude for each dominant response mode is stated in terms of 'a range, taking into account the deqree of statistical variability in each of the correlations.
The predicted mode and frequency are obtained fzom the dynamic analyses of paraqzaph 1 above.
The dynamic modal analysis also forms the basis for interpretation of the prototype plant preoperational and initial startur. test results (Subsection 3.9.2.0).
NOdal stresses are caiculat;eQ and relationships are obtained between sensor zesponse amplitudes and peak component 3 9 2 4 greope'rational Plow-Induced Vibration Testing of Reactor Internals Susquehanna reactor internals-vill be tested in accordaace with provisions of Regulatory Guide 1.20, Revision 2, for Non-prototype Cateqory I plants.
The test procedure vill require +
operation of.the recirculation system at rated flow with r<>~
11~, fll gy eve ence of vibration, wear, oz loose parts The test duration
~~to k w.
sePe+]y 3 9-46 r'g @pre'~err+5 nc </4~4 s~~~
> >per J
~~J
'Pic'/
~~g~/rty
~~ J pp (~'Jr eM 8 cv I
Attachment M
Page 2 of 2 I
vill be sufficient to subject critical components to at least 10~
cycles of vibration during tvo-loop and single-loop operation of the recirculation system.
At the completion of the flow test, the vessel head and shroud head vill be removed, the vessel vill t
~
~
be detained and major components vill be inspected on a selected basis.
The inspection vill cover all. components vhich vere examined on the p-ototype desiqn, includinq the shroud, shroud head, ~ core support structures, ~ jet pu m ps, peripheral control rod drive ~ inshore guide tubes.
Access vill he provided to the reactor love 1
um ui Q v 85 N~Vt + ~~/j Reactor internals for Susquehanna
.are substantially
=he same as the internals design confiqu"ations vhich have been tested in prototype Bi R/4 plants.
Results of the prototype tests are presented in a Licensinq Topical Report (RZF. 3. 9-7).
This rep'ort also contain additional information on the confirmatory inspection program.
3 9 2.5 Dynamic System Analysis of the Reactor Internals Under Faulted Conditions
. In order to assure that no siqnificant dynamic amplification of load occurs as a
esult of the oscillatory nature of the blovdovn forces (Figure 3.9-5),
a comparison vill be made of the periods of the applied forces and the natural periods oi the core support structures beinq acted upon by the applied forces.
These periods vill be determined from a comprehensive dynamic model, of the BPV and internals vith 27 degrees of freedom.
Only motion in the rertica$ d'irection will he considered here;
Besides the'eal masses of the BPV and core support structures, account vill be made for the vater inside. the BPV.
Typical curves of the variation of pressures during a steam line break are shown in Figure 3.9-5.
The accident analysis method is described in Subsection
The time varyinq pressures are applied to the dynamic model of the reactor internals described above Ercept for the nature and locations of the forcing functions and the dynamic model, the dynamic analysis method is identical to that described for seismic analysis and is detailed in subsection 3.7.2.l.
The loads and load combinations acting upon the jet pumps are listed in Subsection 3.9 3 l.
Attachment N
Page I of 9 OESTXOW 110 42:
PSAR'ections 3.9.3, 3.9.4 and 3. 9.5 reference several tables (3.9-2, 3.9-6, 3.9-17, 3.9-27, etc.) that describe the various loading combinationSconsidered in the design of ASME Class 1,
2, and 3 components, component supports, core support structures, control rod drive components, and other reactor internals.
Re have had discussions with the Mark II Owner's Group concerning the load combinations appropriate for the design of 3WR Mark II plants Our position vith respect
.to load combinations has been documented as Attachment A to Enclosure 5 of the HRC Mark IZ Generic Acceptance Criteria for Lead Plants.. This staff position is repeated here as Enclosure 110-2.
These loading combinations are applicable to the Susquehanna plant.
Therefore, provide 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'or vill be analyzed or othervise qualified in accordance vith Enclosure 1 10-2, as modified by the follovin'g tvo clarifications:
{a) Por load cases 1 and 2 all ASME Code Service Level B
requirements are to be met, including fatigue usage factor requirements, and should take into account all SRV discharge
~ load effects (initial actuation and continuous suppression pool vibratory) taken for the number of cycles consistent vith the 40 yr. design life of the plant.
(b) Por load case 10, SRV~ should be assumed to be one SRV.
/lion For load cases 1 and 2 as identified in Question 110.42, enclosure 110-2,all ASHE Code Service Level B requirements,'ncluding fatigue consideration for Class 1 components, are met, for piping in non-NSSS's scope.
For load case 10, SRVz (one SRV) is not considered in com-bination with DBA induced loads.
However, the loads re-sulting from condensation oscillation and chugging are con-sidered in combination with the effects of SRVADS.
FSAR Tables 3.9-17, 3.9-18, 3.9-23 and 3.9-27 are revised to reflect the loading combinations'nd acceptance criteria that are used for ASME Code Class 1,
2 and 3 components and their supports.
p
Attachment N
I Page 2 Of 9 I
/
oad Case 1 combinations meet the'ited staff position.
General Electric believes that the loading combination OBE + SRV ll All (Load Case
The classification of this low probability combination of loads as Emergency (Service Level C requirements) is consistent with 1) the encounter frequency of the
However, in response to continued regulatory staff
The considerati'ons for not conducting the fatigue analysis involve the same technical )ustifications enumerated above.
For load case 10, SRV (one SRV) is not considered in combination with DBA induced loads.
TABL
.42 1
<v)
ACCEPTANCE CRITERIA FOR NSSS PIPING
& E UIPMENT Load Case N
~S SRV OBE SSE IBA DBA Acce tance Cri'teria 3
4 X
X X
X X
X(2) x(')
X{2)
B(l) ".
B(6)
D(3) cz)
D(3)
(3) 7 NOTES.:
X X
g cv)
(1)
For load case 1, all ASME Code Service Level B requirements are to be met, including fatigue usage.
All SRV discharge load effects vill be combined with mechanistically associated loads and taken for the number of cycles consistent with the 40 yr. design life of the plant.
(2)
Loading due to DBA/SBA/IBAis determined from rated steady-state conditions.
(3)
Piping functional caPability will be assurect for essential piping Per Fnclosure 110-2 o
NEDE~21o85 (4)
Not used.
(5)
IBA and DBA includes all associated loads such as annulus pressurization, pool swell, chugging, etc.
(6)
For load case 2, all ASME.Code service requirements are to be met, excluding fatigue usage.
t g)
For specific load combinations r'efer to -Table 3,"-.-. 2, 1771/cak
45
attachment N
TABLE 3.9-17 Page 4 of 9 DESIGN LOADING COMBINATIONS FOR ASME CODE CLASS 1, 2, AND 3 COMPONENTS CONDITION DESIGN LOADING COMBINATIONS Design Normal Upset (a)
(b)
(c)
PD PD+
DW PO
+ SRV x )
PO
+ DW +
(RVC
+ OBE )
I 2
PO +
PO +
+ SBA )
2 2
+
28 (b)
PO+ DW+ (FV
+ OBE )
Faulted (a)
(b)
(c)
PO +
+
2 2
+ SRV
+
2+
2 PO +
+ DBA )
IBA )
IBA )
Design Pressure Operating Pressure Dead Weight Operating Basis Earthquake (inertia portion)
Safe Shutdown Earthquake (inertia portion)
Loads due to Safety Relief Valve Blow Axisymmetric or Asymmetric Loads due to Automatic Depressurization SRV Blow-Axisymmetric Small Break Accident Intermediate Break Accident Design Basis Accident Transient response of the piping system associated with fast valve closure.
Transients associated with valve closure time less than 5 seconds are considered PD PO DW OBE SSE SRVx SRVADS SBA IBA DBA FV Note::
(1)
As required by the appropriate subsection, i.e.
NB, NC, or ND, of ASME
~
~
~
~
~
~
~
~
~
~
~
~
~
~
~
~
~
Section IIIDivision I, other loads, such as thermal transient, thermal gradients, and anchor point displacement portion of the OBE or SRV may require consideration in addition'o those primary stress-producing loads listed.
j ) ri Legend:
Attachment N
TABLE 3.9-'17 Page 5 of 9 Page 2 of 2
~ r+
l RVC
Transient response of the piping system associated with relief valve opening in a,closed system RVO
Sustained load or response of the piping system associated with relief valve opening in an open system or last segment of the closed system with steady state load.
SBA, lBA', and DBA include all event induced loads, as applicable, such as
Attachment N"
Page 6 oZ 9
TABLE 3 9-1S DESIGN CRITERIA FOR ASIDE CODE CLASS 1 PIPING CONDITION STRESS LIMITS(t l(<)
Design Normal and Upset Emergency Faulted HB-3221 and NB-3652 NB-3223 and NB-3654 HB-3220 and HB-3'655 NB-3225 and NB-3656 As specified by ASIDE Section III, 1971 through Minter 1972 Addenda (2)
Functional capability of essential piping will be assured per Rodabaoeh Criteria, Ref. 3.9-8.
Attachment N
TABLP-Page 7 Of 9 DESIGN CBITERI'A FOB ASIDE CODE CLASS 2'ND 3 PIPING CONDITION
'I STRESS LINZTSCxvkL)
The piping shall conform tc the requirements of ASIDE Code Case 1606.
As specified by ASIDE Section III, 1971 through 1972 Minter Addenda (2)
Functional capability of essential piping will be assured per Rodabaugh Criteria, Ref. 3.9-8.
i
ac men Page 8 of 9 TABLE 3,9-27 DESIGN LOADING COMBINATIONS FOR SUPPORTS FOR ASME CODE CLASS 1, 2, AND 3 COMPONENTS (NSSS AND NON-NSSS)
CONDITION Hydrostatic Test Normal and Upset DESIGN LOADING COMBINATIONS (a).
HTDW (a)
DW + TH +
(OBE
+ SRV
)
X 3
(b)
DW + TH +
(RVC
+ OBE )
1 (c)
DW + TH + FV ALLOWABLE'TRES S~
0.8 Sy.
S1, (d)
DW + TH + OBE + RVO Emergency Faulted
'a)
DW + TH +
(OBE2 (b)
+ SRV
+ SBA )
2 2h
+FV).
+
+ IBA )
2 2b
+ SRV
+ IBA )
AD)
+ DBA )
P 1.8 S1 0.9 Sy
~
~
~
~
~
~
~
~
Section III, Subsection NF.
(They are not,commercially available to the require" ments of ANSI B 31.1.)
Notes:
(1)
Loads due to OBE SSE SRVxp SRVADS SBA, IBA and DBA include both inertia portion and anchor motion portion when response spectra metho'8 is used.
The loads from inertia portion and anchor motion are combined by the method of Square Root of the Sum of Squares (SRSS).
(2)
The allowable stress shall be limited to two-thirds of the critical buckling'tress.
LEGEND:
HTDW TH Sy Piping dead weight due to hydrostatic test Reaction at the support due to thermal expansion of the pipe Yield Stress h
See Table 3.9-17 Allowable stress per ANSI B31.1 l
for additional nomenclature.
1 S',
Attachment N
SSES-FSAR Page 9 of 9 Preservice and in-service valve tests will be done in accordance with IWV-3000 of ASME Section XI.
The testing schedule for valves is discussed in Chapter 16.
3.9.7
~
REFERENCES 3.9-1 3.9-2 "Design and Performance of G.E.
BWR Jet Pumps," General Electric
Moen, H.H., "Testing of Improved Jet Pumps for the BWR/6 Nuclear System," General Electric Company, Atomic Power Equipment Department, NED0-10602, June 1972.
3.9-3 General Electric Company, "Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50', Appendix K," Proprietary
3.9-4 "BWR Fuel Channel Mechanical Design and Deflection," NEDE-21354-P, September 1976.
3;9-5 3.9-6 3.9-7 "BWR/6 Fuel Assembly Evaluation of Combined Safe Shutdown Earthquake (SSE) and Loss-of-Coolant Accident (LOCA) Loadings,"
NEDE-21175-P, November 1976.-
Seismic Analysis of Piping Systems, BP-TOP-l, Bechtel Power Corporation, San Francisco, California, Rev 2, January 1975.
"Assessment of Reactor Internals Vibration in BWR/4 and BWR/5 Plants",
NEDE-24057-P (Class III) and NEDO-24057 (Class I),
NED0-21985, 78 NED174 (Class I) September 1978.
3,9-.95
Attachment 0
"I l
QUESTION II0.43:
/
~
~
I Provide the bases for the all.owable buckling loads, including the bucklinq a liowa ble stress limit, under faulted conditions for all HSSS and BOP ASNE Class 1 component supports.
I'or the reactor vessel support skirt, provide a comparison of the calculated bucklinq loads against the critical buckling loads of the skirt under the most limitinq faulted loadinq condition.
Describe the analytical techniques used in determining both the calculated buckl'ing loads under faulted conditions and the critical bucklinq loads of the Susquehanna support skirt.
Provide the most limitinq load combination considered in the bucklinq analyses for the.reactor vessel. support skirt.
RESPONSE
Commercially available ruts and snubbe have been specified in R
accordance with ve rs catalogue data.
Miie
'pin to. pin T
dimen:-i:ons exce the standard lengths and buck is to concern
~
~
the capaci of these items has been reduced in acco nce with A
.the ca oque information.
Calculations and tests suhst iating
.th data are on file at the vendors engineering office..
~ I
~
Pez GE design specificaiton the permissible compressiVe load on the reactor vessel support skirt.cylinder modeled as plate anQ "helltype comoonent support was limited to 90 percent of the load which produces yield -tress, divided by the safety factor for the condition beinq evaluated.
The effects of fabrication and.
operational eccentricity was included..
The safety factcr for faulted*conditio'ns was 1.125.
t
The faulted condition analyzed included the compressive loads due to the design bases maximuin earthquake, the overtu4in'ing moments and shears due to the jet reaction load result'i'n'q from a'evered
LThe expected maximum earthquake loads for the I
I Rev.. 15; 4/80 110 43-1 Str ural elements used in the support of AStlE Class 1
componen an 7 piping are beinq evaluated for bucklinq 'g a
stress czit.
a that limits the allowable axial corn ssion and':
bendinq stres"e or linear type supports to 2/
f the critical buckling stress.
T acceptance criteria f plate and shell type support limits the owable memh e stress intensity or I
membrane plus bendinq stress 'en y
(compression only) to L/2,N
'f the'critical bucklinq tres
~ S-.
E'
S
Attachment 0
'INSERT A Page 2 of 3 Structural elements used -in the support of ASME Class 1
compone'nts and piping are evaluated for buc'kling using a
stress criteria that limits the allowable stresses for supports to 2/3 of the.critical buckling stress.
Commerically available struts and snubbers have been specified in accordance with vendor catalogue data.
The capacity of these items, stated in the catalogue, relative to buckling is limited by the maximum permissible pin to pin dimension.
Calculations and t&ts substantiating these data are on file at the vendors engineering office.
1
Attachment 0
SSES-FS'AR Page 3 of 3 Susquehanna 1
6 2 reactor vessel support kirts are less than 50%
"of the maximum design bases loads used in the. bucklinq analysis 6escr ibed; th refore, the expected faulted loads are ~ell below the critical bucklinq limits of paragraph P-1370 fc) for this
~
reactor vessel support skirt.
The expecte'd earthquake loads for this reactor vere determined'sinq the seismic dynamic analysis methods described in Section 3.7 of the Susquehanna 162 Safety Analysis Report.
/(
Rev. 15, 4/80I
~
1 110 03-2