ML20063P470
| ML20063P470 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 10/06/1982 |
| From: | Devincentis J PUBLIC SERVICE CO. OF NEW HAMPSHIRE, YANKEE ATOMIC ELECTRIC CO. |
| To: | Kerrigan J Office of Nuclear Reactor Regulation |
| References | |
| SBN-339, NUDOCS 8210130436 | |
| Download: ML20063P470 (126) | |
Text
PUBLIC SERVICE ssAanOOK STATION sa in dne om c.
Companyof New HempeNro 1671 Wore. Wor Rood Frominghom, Massachusetts 01701 (617) - 872-8100 October 6, 1982 SBN-339 T.F. B 7.1.2 United States Nuclear Regulatory Commlssion Wa shington, D. C. 20555 Attention:
Ms. Janis B. Kerrigan, Acting Chief Licensing Branch 3 Division of Licensing Ref erences:
(a) Construction Permits CPPR-135 and CPPR-136, Docke t Nos. 50-443 and 50-444 (b) USNRC Letter, dated March 1,1982, " Request for Additional Information" F. J. Miraglia to W. C. Tallman (c) USNRC Letter, dated March 5,1982, " Request for Additional Information" F. J. Miraglia to W. C. Tallman (d)
PSNH Letter, dated April 21,1982, " Response to 210 Series RAIs; (Mechanical Engineering Branch)" J. DeVincentLs to F. J. Miraglia (c) PSNH Letter, dated April 30, 1982, " Response to 210 Series RAIs; (Mechanical Engineering Branch)" J. DeVincentis to F. J. Miraglia (f) PSNH Letter, dated May 10,1982, " Response to RAI 210.56; (Mechantcal Engineering Branch)" J. DeVincentis to F. J. MLeaglia Subject :
Revised Responses to 210 Series RAIs; (Mechanical Engineering Branch)
Dear Ms. Kerrigan:
l The referenced PSNil Letters [ References (d)-(f)] provided responses or l
commitments to respond to the 210 Series Requests for Additional Information l
(RAIs) which were forwarded in References (b) and (c), specifically RAI 210.3 through RAI 210.69.
O Meetings were conducted with the NRC Mechanical Engineering Branch on May 11-13, 1982, at the offices of United Engineers and Constructors in Phlladelphia, PA during which the above RAI responses were discussed.
Based on this meeting, many of the original responses have been revised for clarification, addLttonal analysis or commitments, etc.
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United Stctac Nacicar Reguictcry Cosmicaion Octobar 6,1982 Attention:- Ms. Janis B. Kerrigan Page 2 -
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'Je have enclosed responses to RAIs 210.3 through 210.69. Original responses which have been revised have been sarked " Revised", and include bars in the right margin to indicate the location of the chanSed information.
Responses which were presented for the first time-at the May 11-13, 1982 meeting have been marked "New Response at 5/11/82 Meg".
The revised FSAR Sections which correspond to the enclosed RAI responses will be incorporated into Amendment. 47.
Very truly yours,
. YANKEE ATOMIC ELECTRIC COMPANY John DeVincentis Project Manager h
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1 SB 1 & 2 FSAR RAI 210.3 (3.2.1. Table 3.2-2. Sheet 4)
Justify the non-seismic classification of the containment recirculating filter system. Show that its failure will not impair either the fans or ductwork.
RESPONSE
The air cleaning or filter unit is not safety-related and is not listed as Seismic Category 1.
The unit is seismically anchored to restrain movement and is structurally identical to other EST air cleaning units (containment enclosure and fuel storage building). Therefore, the unit casing will not fail during a seismic event.
such Internal components of the air cleaning unit may f ail structurally, but a failure will be contained within the unit and will not impair operation of the safety-related f ans, dampers and ductwork.
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Explain note 9 as it applies to the reactor coolant pump flywheel.
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RESPONSE
Note 9 does not apply directly to the RCP flywheel. The flywheel is designed to maintain structural integrity during an SSE and under overspeed I
conditions that could result from a LOCA. A detailed discussion of the RCP f
flywheel is contained in FSAR Chapter 5.
Sheet I of FSAR Table 3.2-2 $ A.Ms t
W revised to delete the reference to Note 9 for the RCP flywheel.
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Describe methods used to confirm the structural integrity of non-seismic category I components whose failure or collapse could result in loss of,
inction of seismic Category I equipment.
RESPONSEt E
1.
For non-seismic Category I components, except piping, attachments to l
structural members (i.e., anchoring devices) are analysed to demonstrate their ability to withstand applicable seismic loading. Each component's fundamental frequency in each of an X, Y and Z direction is determined, and a conservative 1.5 times the corresponding accelerations from the applicable Amplified Response Spectra (ARS) curves are the applied seismic loadings. Equivalent static analysis methods (Subsection 3.7.3.1) are then used to determine anchorage loadings and stresses. 1.cadings from each ear hquake direction are applied individually, and the square-root-of-the-sum-of-the-squares (SRSS) method is used for the determination of final results.
2.
For those non-seismic Category I components, including piping,which are l
not seismically supported, the failure modes offacts analysis (FMEA) performed assures that they are isolated by their location to prevent.
tp any potential impact on seismic Category I components, should there be
.h any failure or collapse of the non-seismic Category I componsats or NA piping.
Y 3.
Also see response to RAI 220.19.
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4.
Isometric drawings of pipe breaks and associated pipe whip restraints mentgedPf % gy lines inside containment will be provided in FSAR Am for high ener NSSS Consistent with Regulatory Guide 1.29. Westinghouse seismically qualifies any component whose failure could adversely effect a safety system. A specific Nj /
example of such a component is the fuel handling machine located in the spent gg(
l fuel pit. When seismic qualification of such components is required, they are qualifi~ed in accordance with the methods described in Sections 3.7(N)
[ggg, st scope which has been identified by Westinghouse as having a potential adverse Ilg and 3.10(N) of the Seabrook FSAR. In summary, any component in Westinghouse impact during a seinsic event on other safety-related equipment is seismically P
qualified.
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O SB 1 & 2 FSAR RAI 210.6 (3.2.1. Table 3.2-2. Sheets 20 and 21)
Why are the computer room system components and the primary auxiliary building dampers and ductwork not considered seismic Category 17 i
RESPONSE
The computer room air conditioning components are not required for safe plant shutdown. In the event of a f ailure of the equipment, provisions have been made to directly connect the computer room supply duct to the control room air conditioning supply air system. The latter system has sufficient cipacity to f urnish cooling to the computer room. The computer room ductwork is seismically supported, non-safety-related. Details of this system are contained in Section 9.4.1 and on Figure 9.4-1.
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With the exception of that equipment associated with the FCCW pump area, the primary auxiliary building ventilation system is not safety-related. Where this system extends over or near safety-related equipment, the ductwork and cesponents are seismically supported. Refer to Section 9.4.3 for further information and details.
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Sg1&2 FSAR RAI 210.7 (3.2.2. Table 3 2-2)
It is the staff's position that certain systems important not identified in gegulatory Guide 1.26 should be classified Quality Group C, or its Among these systems ares diesel fuel oil storage and transfer equivalent.
system, diesel engine cooling water system, diesel engine lubrication system, diesel engine starting system, and diesel engine combustion air intake and exhaust system. Justify the absence of a quality group classification of portions of those systems listed below:
Diesel Generator Fuel Oil Storage and Transfer System A.
1.
Remaining On-Engine Equipment and Piping B.
Diesel Generator Cooling Water System 1.
Auxiliary Coolant Fusp 2.
Remaining On-Engine Equipment and Piping C.
Diesel Generator Starting System s
1.
Air Compressor 2.
Remaining On-Engine Equipment and Piping D.
Diesel Generator Lubrication System 1.
Auxiliary Lube Oil Fump 2.
Remaining on-Engine Equipment and Piping E.
Diesel Generator Combustion Air Intake and Emb ust System 1.
Piping 2.
Air; Intake Filter 3.
Exhaust silencer
RESPONSE
All of the. piping and equipment associated with the diesel engine is 1
designed to seismic Category I requirements and is consistent with standards of Quality Group C or D of Regulatory Guide 1.26.
The quality standards l
used for specific components are considered in complia' ice with Regulatory Guide 1.26, which states that systems such as diesel engine and auxiliary support systems should be designed to standards comensurate with the safety f unction to be perf ormed. For the specific components identified, the -
following consents are noted:
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On-engine equipment and piping is considered integral with the engine, and is designed to manufactuer's standards, which is consistent with the engine itself.
3.
1.
The engine-driven pumps (c'oolant and lube oil) are the primary (and only) pumps required for emergency starting and i
operation of the diesel generator. A failure of these pumps-is considered an engine failure. The auxiliary of f-skid pumps are not excpected to function under emergency I
conditions, but could be used administratively for back-up or mai,ntenance purposes.
2.
See Response A above.
C.
1.
The air compressors function is to meintain air reciever pressure between starts, but is not required for emergency starting and operation of the diesel generator. The air receiver pressure is monitored to provide ample warning for corrective actions or air compressor problems.
2.
See Response A above.
D.
1.
See Response 3.1 above.
2.
See Response A above.
E.
There are no moving or rotating parts associated with the air intake filters, exhaust sinlencers, or interconnecting piping.
All of these items are seismically supported within the diesel building.
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Explain your rationale for classifying the shell side of the reactor coolant posp thermal barrier heat exchanger as ASME Code class 3 although the tube side is Code Class 1.
RESPONSE
Definition of the RCP thermal barrier relative to a tube side and shell side is not totally accurate. The tubes are located inside the pump casing and reactor coolant flows around the tubing. There is no shell side in the strict sense of a heat exchanger. Table 3.2-2 has been revised to correctly describe l
the thermal barrier and its classification.
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Justify the absence of a quality group classification f or the entire computer room air conditioning system.
RESPONSE
The computer room air conditioning system is not required for safe plant shutdown, therefore, it has no ANS safety class. Refer to our answer to RAI 210.6 which is directly related to this question.
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Control room complex emergency cleanup filter system fans and filter unit have been given ANS safety classification Non-Nuclear Safety, and the ductwork no safety classification at all. This system is considered important to safety. Provide justification for your classification.
RESPONSE
The control room complex make-up air system from the redundant intakes (remotely spaced on opposite sides of the two units) to the control room complex emergency cleanup filter unit is ANS Safety Class 3.
The placement of these air intakes essentially eliminates the possibility of having both inlets abaultaneously exposed to accidently released activity. This design considerably reduces the likelihood of the emergency filter even being needed. We consider the remote air intake design to be our primary protection against the possibility of control room contamination. The filter system is, therefore, considered only as a backup and for recirculating cleanup.
Tha duc'twork from the air cleaning unit to the associated fans and dempers, and finally to the air conditioning suppy air duct, will be upgraded to ANS Safety Class 3, Seismic Category 13 TSAR Table 3.2-2, Sheet 20, will be revised to show this change.
While The emergency cleaning unit f ans are redundant, with Class IE motors.
the filter units and associated fans are considered ANS Safety Class NNS, as they "can influence safe, normal operation" as defined in TSAR Subsection 3.2.2.1, they are not deemed essential to control room habitability.
As a retter of note, the overall design of the control room ventilation system was considered by the NRC in their August 1974 Saf ety Evaluation on Seabrook Station to meet the guidelines of General Design Report criterion 19 with respect to potential radiation doses to control room personnel.
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FSAR RAI 210.11 ( 3.2.2.2, Table 3.2-2 )
The following ventilation systems that serve the control room er engineered safety feature rooms have portions of their systems lacking a quality group classification. Assign an appropriate quality group classification or its equivalent or justify the noncitssification:
1.
Control room complex ventilation system ductwork.
2.
Fuel storage building ventilation system, ventilation fans, ductwork.
RESPONSE
1.
The control room air conditioning system ductwork located within the mechanical equipment room will be classified as ANS Safety
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The remaining ductwork, has no safety classification, but is seismically supported. Local failure of this ductwork will have no adverse effect on the safety-related components, equipment, or systems located in the control room complex. Table 3.2-2, Sheet 20, "' ' L. ?
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agree with the statements above.
2.
As explained in the answer to RAI 410.36, Table 3.2-2, Sheet 7 is (Note 12) to state that the ductwork from the
-4 downstream side of the air cleaning units to the fan intakes and the discharge of the fans to the building boundaries is Safety
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This is further clarified in Secticn 6.5.1.
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TSAR RAI 210.12 (3.2
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Exphin the NNS ANS safety classification of the entire liquid and solid waste systems, f
RESPONSE
The liquid and solid waste systems are classified as NNS since these systems perform no safety function and are 'ot required for safe shutdown of the Table 3.2-3 lists the quality standards applicable to these reacto,r.
systems per Regulatory Guide 1.143. These quality standards correspond to the NNS classification.
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SB 1 & 2 TSAR RAI 210.13 (3.6(B).2.1, Page 3.6(B)-6)
Confirm that the " elastica 11y calculated basis" for loadings of operating plant conditions plus an operating basis earthquake is the maximum stress as calculated by equation 9 in Paragraph NB-3652 of the ASME Code,Section III.
RESPONSE
This criterion refers to Class 1 piping in the Seabrook plant design.
Class 1 piping is all located inside containment, and therefore, the requirements of Regulatory Guide 1.46 must be met in postulating the locations of piping breaks in Class 1 piping.
The formula used in determining primary plus secondary stress intensities is equation 10 of NB-3652 of the 1971 ASME Code,Section III, with addenda up to and including Winter,1972. This formula considers the primary plus secondary stress intensity range due to internal pressure and the range of moment loading due to thermal expansion, anchor movement, earthquake ef f ects and other mechanical loads.
Equation 9, which is not required to be used for postulated break location, considers only the primary stress intensity due to internal pressure and t
moment loading due to earthquake, deadweight and other sustained design mechanical loads.
The elastica 11y calculated basis referred to in FSAR Paragraph 3.6(B).2.1.a.1.(b) describes the criteria used to calculate the maximum stress range under the applicable load combinations, which is based upon the assumption that stresses are directly proportional to strains.
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Provide drawings of all postulated pipe breaks, showing the type of break, structural barriers, restraint locations, and constrained directions in each restraint. Also provide a table showing calculated stress intensities, cumulative usage factors, and primary plus secondary stress ranges for each postulated break.
RESPONSE
Piping drawings are not available showing structural barriers. Agameenen p.avingsofpostulatedpipebreaklocationsandassociatedpipewhiprestraints I
for high energy lines inside containment es=Wur included in FSAR Amendment es09' b6VSbeen kf Stress intensities were calculated only for Class 1 lines, using generic stress data from Westinghouse and adding UE&C-derived seismic and thermal stresses to provide very conservative values. Cumulative usage factors were assumed to exceed 0.1 at every fitting. In order to avoid performing an enormous number of calculations, we have postulated breaks at every fitting weld for Class 1 lines.
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Specify where pipe whip restraints or anchors have required welding to the outer surface of the pipe. Provide details of the stress analysis performed as in the case of a riser clamp lug.
RESPONSE
Lug attachments welded to Class 2 and 3 pipes are qualified by a procedure whose methodology is equivalent to, but more conservative than, that presented in Code Case N-318.
Local stress levels in the pipe resulting from applied lug loads are obtained by multiplying the nominal stress in the lug at the lug / pipe interface by the appropriate B or C index (as defined in Code Case N-318) for each individual loading condition. The local stresses are superimposed upon the general pipe stress as determined from program ADLPIPE to establish the total stress level in the pipe for that loading condition.
Loading conditions required to be considered for Plant Normal, Plant Upset, Plant Emergency, and Plant Taulted Operating Condition are defined (per appropriate FSAR section), and total stress in the pipe is obtained from
- suussing the stresses for each individual loading condition that must be con-sidered.
Local stress levels determined using It indices are added to the general stress levels from ADLPIPE and this sum is compared against allowable limits to demonstrate structural integrity. For the pipe wall, local stress leve'Is determined using C indices are added to the general stress levels from ADLPIPE, and this sua is compared against the allowable range of ctress (S *S )-
h a Finally, weld stress is evaluated considering the absolute sua from all loads, independent of the operating condition, and compared against allowable stress from Table NF329.1-1, Subsection NF, ASME III.
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RAI 210.16 (3.6(B).2.1, Page 3.6(B)-g Inservice inspection of break-exclusion piping must include 100% volumetric examination of all pipe welds. Augment your inservice inspection description j
to include this requirement at intervals shown in IWA-2400 ASME Code,Section XI.
RESPONSE
All high energy pipe penetrations are evaluated for postulated pipe breaks.
Breaks are not postulated at the main steam and feedwater penetrations.
Augmented inservice inspection, including 100% volumetric examination of these penetrations, will be performed as defined in FSAR Section 6.6.
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Justify the 90% of yield stress criteria in plastic restraint design. Pro-vide examples of analysis of such a design.
RESPONSE
Critik-Except for those components (i.e., M pads, U-bolts) of the pipe rupture restraint (PRR) which are identified as elasto-plastic elements, the other components of the PRR structure are designed to remain elastic. The stress limit set for design of the elastic components is taken as 90% of the minimum yield strength of the material.
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SB 1 li 2 FSAR RAI 210.18 (3.6(B).2.3, Page 3.6(B)-15)
Provide a ref erence or further ju'stific4 tin for c5*.i<< of a maximum fiber strain of 50% of ultimate strain as an adequancy requirement for the load carrying capacity of piping.
ESPONSE:
We refer to ANSI /ANS 58.2 (draf t) November,1978 Section 6.3.2.a.l.for the use of this vel ae es defining an upper-bound design limit.
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What strain-rate and strain-hardening effects you have included in plastic system analysis?
RESPONSE
The strain-hardening effect of the material has been considered in the design of 11-bolt type pipe rupture restraints by utilizing an elastic-plastic bilinear stress strain curve which is approximated from either a test data stress strain curve or the minimum code yield and ultimate strength values. In the case of crash pad type pipe rupture restraints, the load-deformation characteristic of the pad is defined by the manufacturer's testing data which is basically an elastic perfect plastic curve.
At present, the specific effect of strain-rate has not been considered. It is our criteria to use a ten percent increase in yield and ultimate strength values to account for this effect.
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SB 1 & 2 FSAR RAI 210.20 ( 3.6(B).2.5, Pagt 3.6(B)-18)
In order to complete our revies', we must examine Appendix 3B, "Line Designation Tabulation". Frovide a copy of this appendix.
RESPONSE
Appendix 3B was incorporated into the FSAR as part of Amendment 44.
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In the primary loop, what size breaks are postulated for the design of pipe whip restraints? What size breaks are postulated in the primary loop for determination of compartment pressurization and asymmetric loads? If breaks for either case are less than size, provide justification.
RESPONSE
For the circumferential breaks postulated in the RCS, all break locations were assumed to have full double-ended breaks with the exception of the breaks postulated at the reactor vessel inlet and outlet nozzles. At these locations, the break opening area was assumed to be 144 square inches. The limited break opening area is based on the physical constraint provided by the pipe whip restraints located at the reactor vessel nozzles. Actual break opening areas based on the restraint design at these locations were calculated to be 80 square inches at the reactor vessel inlet nozzle and 30 square inches at the reactor vessel outlet nozzle. The calculated break opening areas are l
well below the 144 square inch break opening area assumed in the analyses.
As p art of the normal design interface, UEEC has provided Westinghouse with stiffness values for the RCS pipe whip restraints. These stiffness values were tien incorporated in the RCS structural evaluation to determine appropriate loading conditions on the restraints. This loading information was then transmit?ed to UE&C to verify the adequacy of their restraint design.
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SB1&2 FSAR ft VIsd RAI 210.22 (3.6(N).2.3, Page 3.6(N)-7)
Provide a copy of test results of pipe-to-pipe impact. Also provide test results that show whipping or bending of a stainless steel pipe does not cause the section to become a missile.
RESPONSE
Westinghouse has performed pipe whip tests demonstrating that a pipe will not break pipes of equal or greater size of the same material. These tests are documented in WCAP 7503, Supplement 1, and were submitted to the staff on the Trojan docket in response to a similar question. These test results provide justification for the Westinghouse position on this subject. It.
should be noted that the Westinghouse position is consistent witn the staff positions identified in Regulatory Guide 1.46 and Standard Review Plan 3.6.
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RAI 210.23 (3.6(N).2.5, Page 3.6(N)-ll)
Review of this section shows that you have vaed a cumulative usage factor of O.2 for postulated pipe rupture criteria. Aranch Technical Position EB 3-1 specifies a cumulative usage factor of lest than 0.1.
Provide a consnitment to meet this criteria.
RESPONSE
The number of break locations censidered in the reactor coolant system using a cumulative usage factor of 0.2 is adequate. Additionally, the NRC has determined that the 0.2 criteria in WCAP-8082 (Reference 1 to FSAR Section 3.6(N) which is applicable to the Seabrook plant is acceptable. It should be noted that this item was discussed with the EB during previous review meetings on ehme eh
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In addition to showing postulated break locations, they must be identified as either circumferential or longitudinal. Structural barriers, if any, restrain location and constrained directions must also be included in order
- to complete our review.
RESPONSE
Break opening areas, are discussed in detail in the response to RAI 210.21.
Additionally, the location of pipe whip restraints in the RCS was also dis-L cussed. f- '4a?^"--..___'
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What criteria is used to determine the number of degrees of freedom in your dynamic analysis?
RESPONSE
P!2t Each lumped mass will have specified for it those degrees of freedom which represent the possible and/or predominant directions of motions. In some circumstances, individual masses need to be lumped for short, stiff members which exhibit rigid range behavior. An example model of a typical cable tray assembly is shown in Figure 3.7(B)-32.
NSSS For flexible equipment, Westinghouse utilizes many degrees of freedom (e.g.,
200 for steam generators) in their dynamic analysis models. It should be noted that Westinghouse assures that a sufficient number of modes is con-sidered in the analysis, consistent with SRP 3.7.2.
Westinghouse has also gY provided test results at a previous EB review meeting which cupport their r
modeling techniques (e.g., for the Reactor Coolant System, NRC Docket 50-206, pg,{
April 24, 1977, " San Onofre Nuclear Generating Station Seismic Re-evalua;. ion P
and Modificati,ons"), and additional data on tanks, valves, and typical. piping b
g \\*
systems.
11 For auxiliary mechanical equipment with natural frequencies below 33 Hz,
[{,),
test results were presented at a previous MEB review meeting to support the number of modes considered in the analyses.
The above information, which was presented to the NRC at previous MEB review meetings, is applicable to the Seabrook plant.
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SB 1 & 2 FSAR RAI 210.26 (3.7.3.1, Pane 3.7(B)-11)
Demonstrate that the equivalent static load method analysis you have used accounce for relative motion between all parts of support.
RESPONSE
E I
When significant relative motions among the parts of any supporting system are encountered, their effects are determined statically and superimposed with other analytical results associated with any particular dynamic event.
NSSS g,k k
The equivalent static load method has not been used on any Westinghouse minine 8
' gh [1 and piping supports.
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RAI 210.27 (3.9(B).l.1,'Page 3.9(B)-1)
Are there any reactor coolant pressure boundary, ASME Code Class 1 or CS components in BOP 7 If so, provide or reference an appropriate design transient list.
RESPONSE
E Yes; see Subsection 3.9(N).1.1 for design transient list applicable to BOP Class I components.
Class 1 lines in the BOP scops are identified below:
Line No.
Line Size P&ID 91-1 1"
9763-T-805002 91-2 1"
9763-F-805002 328-6 2"
9763-F 805003 328-7 1h" 9763-F-805003 328-10 3/4" 9763-F-805003,
329-4 2"
9763-T-805004 329-5 1h" 9763-F-805004 329-8 3 /4" 9763-F-805004
.g 330-4 2"
9763-F-805005 N
330-5 1"
9763-F-805005 h
330-6 3/4" 9763-F-805005 b
331-4 2"
9763-F-805006 331-5 1h" 9763-l'-805006 331-8 3/4" 9763-F 805006 80-1 6"
9763-F-805007 80-2 3"
9763-F-805007 80-6 3"
9763-F-805007 74-1 6"
9763-F-805007
[
75-1 6"
9763-T-805007 76-1 6"
9763-F-805007 NSSS k[
Westinghouse has responsibility for Class 1 component core support structures and specific Class 1 piping. UE&C has responsibility for pressurizer safety 4t>
relief line, the reactor coolant system drain line and Class 1 reactor coolant g
pump seal piping.
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SB 1 & 2 FSAR RAI 210.28 (3.9(B).1.2, Fage 3.9(B)-1)
NUREC-0800 requires that computer programs in analyses of seismic Category I Code and non-Code items have the following information provided to demonstrate their applicability and validity:
The author, source, dated version and facility.
a.
b.
A description and the extent and limitation of its application.
Solutions to a series of test problems which shall be demonstrated to c.
be substantially similar to solutions obtained from any one of sources 1 through 4 and source 5:
1.
Hand calculations.
2.
Analytical results published in the literature.
3.
Acceptable experimental tests.
4 By an MEB acceptable similar program.
5.
The benchmark problems prescribed'in Report NUREG/CR-1677, " Piping T.enchmark Problems".
Demonstrate compliance with these requirements and provide sumary comparisons for the computer programs used in seismic Category I analyses.
RESPONSE
E The above information is documeraed and available for review for all structural analysis computer programs used by UE6C. The verification package for the b
in-house version of ADLPIPE has been supplemented by Problem #4 from NUREG/CR-1677.
4 Sumary comparisons show excellent agreement. (A copy of the supplemental D
verification is provided by a separate transmittal ) Revisions to FSAR Sub-seccion 3.9(B).1.2 which describes the verification methods used for each structural computer program have been provided in Amendment 45.
[(M The computer codes used by W,for Class 1 analyses are described in FSAR Sub-section 3.9(N).1.2 and in References 1 (WCAP-8252) and 2 (WCAP-8929) to Section 3.9(N). WCAP-8252 has been approved by the NRC, and WCAP-8929 is N
currently undergoing review. This information has been sufficient to address N
this concern during previous MEB review meetings.
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RAI 210.29 (3.9(B).1.1., Pare 3.9(B)-6)
Where is AISC criteria used in evaluation of faulted conditions? Justify its use.
RESPONSE
The AISC criteria were used in the support designs of the following mechanical components:
Containment Spray Pumps Primary Component Cooling Water Puers Spent Fuel Pool Pumps Primary Component Cooling Water Head Tank Cation Bed Demineralizer Tank Mixed Bed Demineralizer Tank Containment Spray Heat Exchanger Spent Fuel Pool Cooling Heat Exchanger Eastgency Feed Pumps (See the response to RAI 210.39 for justification).
were che eted
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RAI 210.30 (3.9(B).1.4. Page 3.9(B)-7)
This section does not address the criteria used to assure the functional capability of essential systems when they are subjected to loads in excess of those for which Service Limit B limits are specified. By essential 4
systems are meant those ASME Class 1, 2 and 3 and any other piping systems which are necessary to shut down the plant following, or to mitigate the consequences of, an accident. Provide such criteria.
Resp 0NSE:
Design service limits, as defined in NCA2142, are not tified in the i
design specification for piping.
For Seabrook, design condition's are s ied to be normal, upset, emergency and faulted, as defined in the Code, 1971 edition, with addenda up to and including Winter, 1972 e criteria used to assure the functional capability of essent systems to shut down the plant safely following, and to mitigate the sequences of an accident, are described in FSAR Subsectio (B).l.4.b.1 (a), (b) and (c).
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SB 1 & 2 FSAR cy/ke RAI 210.31 (3.9(B).2.1, Page 3.9(B)-8)
What are the acceptance limits for steady state and transient vibration?
The program must include a list of different flow modes and a list of selected locations for visual inspection and measurements.
RESPONSE
The preoperational and startup vibration test prc,-am is under development and is scheduled to be available in October, 1982.
Also, see revised FSAR paragraphs 3.9(B).2.1.a.1 and 3.9(B).2.1.a.5 (Amendment leo.
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What Code-allowable stress limits are used for acceptability of motion due to dynamic effects?
RESPONSE
See revised FSAR paragraph 3.9(B).2.1.c (t.nendmentJe6).
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What piping systems are not designed according to ASME Section III7 What design criteria was used for these systems?
RESPONSE
All piping systems desig-Any non-safety related system is designated NHS.
nated as NNS are designed to ANSI-B31.1 requirements. Item 4 of-Bege 3.9(B).3.1.b.M
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is being revised in Amendment /a6 to delete " Category I".
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RAI 210.34 (3.9(B).3.3, Pare 3.9(B)-22)
Regulatory Guide 1.67 does not address closed systems or systems with a water slug. How was 1.67 used for the installation and design of pressure relief devices?
RESPONSE
l Although Regulatory Guide 1.67 does not address closed systems, Subsection 3.9(B).3.3 of the FSAR addresses the evaluation of safety and relief valves for closed and open systems.
The only relief valves presently having a water seal which could introduce a water slug are the pressurizer relief valves. However, an alternate piping layout has been developed and will be implemented to eliminate the water slug concern. Paragraph 3.9(B).3.3a eMMe revised as followsl*^ ~"---
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"The three safety valves are mounted on the pressurizer nozzles with the short inlet pipe and elbow necessary to position the valves vertically. The total length of pipe, elbow and weld aeck flange is approximately 24 inches and is as short as possible to minimize the pressure drop on the inlet side of the valve.
The two power operated relief valves have inlet piping shaped to form a water seal below each valve seat to reduce the problem of steam and hydrogen leak-age through the valve seats. When the valves open, water from the seals is discharged ahead of the steam as the valve disc lifts. The dynamic effects from the flow of water and steam are included in the design analysis."
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Was Regulatory Guide 1.67 used to determine the spacing of the safety valves on the main steam lines?
RESPONSE
Spacing of the safety valves on the main steam lines is in compliance with Regulatory Guide 1.67 and referenced Code Case 1569.
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Provide a scheoule for completion of dynamic analyses results.
RESPONSE
The information now indicated as "later" on FSAR Tables 3.9(B)-19 and 3.9(B)-20 vill be provided 1,
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RAI 210.37 (3.9(B).3.4), Page 3.9(B)-26)
Provide your interpretation of jurisdictional boundaries as they pertain to NF supports. Justify your position.
RESPONSE
BOP NF requirements as shown in MF-1000 for plates, welding and bolting is as lh$gg'dt Jurisdictional boundaries of supports designed and fabricated to Subsection follows:
1.
Plates A.
Support plates that are embedded in concrete with integral embedded anchors (studs) do not fall under NF jurisdication, whether or not they protrude from the surface of concrete.
B.
Loose or adjustable base plates which only support compressive loads do not fall under NF jurisdiction.
C.
Loose plates that are welded to component supports do fall under NF jurisdiction. (Surface mounted plates)
- 2.
Welding A.
The weld used to attach NT supports to building steel, supplementary steel or intervening members is considered to fall within the jurisdiction of NF.
j 3.
Bolting t.
A.
Embedded custom-designed anchor bolts are designed and purchased to A15C requirements and the additional materials, Certification and NDE Examination Requirements of ASME Subsection NT.
B.
Standard expansion anchors which are manufactured and stocked as catalogue items such as hilti-kwik bolts, fall under the p
jurisdication of Subsection NF.
E NSSS lf The Class I component supports supplied by Westinghouse for Seabrook are consistent with the requirements of Subsection NF of the ASME Code. Westinghouse 6
supplies Class 1 supports from the base plate or concrete to the component.
9 Therefore, the jurisdictional boundary for Westinghouse supplied Class 1 gff'ly supports is well defined, Class 2 and 3 jurisdictional boundaries were also discussed. Design criteria fd for Class 2 and 3 component supports are described in FSAR Subsection 3.9(B).3.4 i
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SB 1 & 2 T'iAR Class 2 and 3 cc onent supports are generally attached to the component and
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the base plate.
The BOP designer is responsible for anchoring the component to the support-ing structure. Therefore, as in the case of Class 1, the jurisdictional i
boundary for Class 2 and 3 supports is well defined.
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RAI 210.38 (3.9(B).3.4. Page 3.9(B)-26)
Provide an example of the analysis performed on ASME Code Class 1, 2 and 3 valve supports.
RESPONSE
The only valve supports that were analyzed as valve supports were the pressurizer safety and relief valve supports. These are ASME Class I supports, and the stress report.... : ;-- haan ~ P * -
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The design criteria used for mechanical equipment supports needs clarification.
Subsection NF, ASME Code,Section III is applicable to these supports. Justify the use of AISC allowable stresses to demonstrate that your design criteria satisfy the requirements of Subsection NF.
RESPONSE
E.p The supports of certain mechanical equipment purchased circa 1974 were designed in accordance with the requi,rements defined in the AISC Manus 1 of Steel Con-struction. In addition, the following criteria were included in the support designs:
1)
Material properties used in conjunction with the support design were obtained from the tables for material strength values in the ASME III, Subsection NA, Appendix I.
2)
The allowable bolt stresses were derived from the AISC Specification, without use of one-third increase factor for Normal and Upset Conditions.
For the faulted condition the AISC allowable of 0.6 Fy was multiplied by the strength factors noted in SRPs 3.8.3 and 3.8.4.
3)
The loading considered in the design of the supports and anchor bolts are the same as those imposed on the components. More specifically, the appropriate loads are applied to the components and the resulting reactions are used to design the supports.
4)
For the faulted condition, tensile and bending stresses were limited to 90% of the material yield strength and shear stresses were limited to 60% of the material yield strength which compare f avorably with the p*)
I-limits defined by ASME III, Subsection NF, for faulted conditions.
5)
Buckling evaluations were performed in accordance with the AISC criteria S
without use of increase factor for faulted conditions.
6)
The highest value of KL/R is less than 20 for all mechanical components (excluding piping systems).
The following tabulation shows the stre'ss limits used for various bolt materials:
1)
Stress Limits for Anchor Bolts for Equipment Allowable Tensile Allowable Shear l
Bolt Material Stress
. Stress ASTM A193 Grade B7
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Fy = 105 ksi Ft = 0.6 Fy 0.5 Fu Fv = 0.4 Fy l
Fu = 125 kai
= 62.5 ksi
= 42 kai l
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FSAR ASTM A540 Crede B23 Class 4 Up to 3"9 Fy = 120 ksi.
Ft = 0.6 Fy 0.5 Fu Fv = 0.4 Fy Fu = 135 ksi
= 67.5 kai
= 48 ksi ASTM A354 Crede BD i
For 1/4" to 2-1/2"9 Fy = 130'ksi Ft = 0.6 Fy 0.5 Fu Fv = 0.4 Fy Fu = 150 kai
= 75-ksi
= 52 kai 2)
Bigh strength bolts for equipment on structural steel and for steel-to-steel connections ASTM A325 1/2" to 1"9 1-1/8" to 1-1/2"9 Fy = 92 kai Fy = 81 kai Fu = 120 kai Fu = 105 ksi ASTM A490 1/2" to 1-l'/2"9 Fy = 130 kai Fu = 150 kai AllaklowabletensionandshearvaluesareinaccordancewithManualofSteel Cons'truction - AISC.
For the faulted condition, the strength factors of 1.6 or 1.7 as noted in SRP 3.8.3 and 3.8.4 were applied to the above.
}LSES The Class 1 component supports supplied by Westinghouse for Seabrook are consistent with the requirements of Subsection NF of the ASME code. Westing-house supplies Class 1 supports from the base plate or concrete to the component.
q, Therefore, the jurisdictional boundary for Westinghouse supplied Class 1 E
supports is well defined.
W e.
b Class 2 and 3 jurisdictional boundaries were also discussed. Design criteria for Class 2 and 3 component supports are described in FSAR Subsection 3.9(N).3.4 k
Class 2 and-3 component supports are generally attached to the component and 3
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s the base plate.
The BOP designer is responsible for anchoring the component to the supporting structure. Therefore, as in the case of Class 1, the jurisdictional boundary for Class 2 and 3 supports is well defined.
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SB 1 & 2 FSAR RAI 210.40 (3.9(B).3.4, Page 3.9(B)-26)
Provide design criteria for any snubbers.
RESPONSE
E A revised FSAR Subsection 3.9(B).3.4 which incorporates snubber design criteria has been provided in FSAR Amendment 45.
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i NSSS The only snubbers supplied by Westinghouse are located at the steam generator upper support points. These snubbers are analyzed in accordance with the D
criteria described in Section 3.9(N).1 for Class 1 component supports.
3 ll Additional information for these snubbers is provided in FSAR Section 5.4.14.
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Attachment A I
NRC STAFF COMMENTS ON INSERVICE PUMP AND VALVE TESTING PROGRAMS AND RELIEF REOUESTS The NRC staf f, af ter reviewing a number of pump and valve testing programs,.
has determined that further guidance might be helpful to illustrate the type and extent of information we feel is necessary to expedite the review of these programs. We feel that the Licensee can, by incorporating these guidelines into each program submittal, reduce considerably the staff's review time and time spent by the Licensee in responding to NRC staff requests for additional information.
The pump testing program should include all safety related* Class 1, 2 and 3 pumps which are installed in water cooled nuclear power plants and which are provided with an emergency power source.
The valve testing program should include all the safety related valves in the following systems excluding valves used for operating conven*ence only, such as manual vent, drain, instrument and test valves, and valves used for maintenance only.
PWR f
a.
High Pressure Injection System.
h.
Low Pressure Injection System.
l c.
Accumulator Systems.
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d.
Containment Spray system.
l Primary and Secondary System Safety and Relief Valves.
e.
f.
Auxiliary Feedwater Systems.
f g.
Reactor Building Cooling System.
(
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Active Components in Service Water and Instrument Air Systems which are required to support safety system functions.
l i.
Containment Isolation Valves required to change position to isolate l
containment.
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- Safety related - necessary to safely shut down the olant and mitir, ate the f
consequences of an accident.
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210.41 As required by 10 CFR 50.55 a(g), we request that you submic your (3.9(B).6) preservice and initial 120 month inservice testing program for.-
pumps and valves. Attachment A provides a suggested format for this submittal and a discussion of information we require to justif y any relief requests.
The scope of pumps and valves operability testing, including
RESPONSE
adherence to ASME Boiler and Pressure Vessel Code,Section XI.and 10 CFR 10 CFR 50.55 a(g), is provided in FSAR Section 3.9(B).6.
50.55 a(g) does not specif y a submittal.date for a preservice or inservice testing program for pumps and valves; however, the pump and valve test program will be submitted, with the Inservice Inspection Program, within six months of the anticipated date for commercial operation. This submittal date is consistent with previous NRC staff guidance.
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Chemical & Volume Control System.
k.
Other key components in Auxiliary Systems which are required to directly support plant shutdown or safety system function.
1.
Residual Heat Removal System.
m.
BWR High Pressure Core Injection System.
a.
b.
Low Pressure Core Injection System.
Residual Heat Removal System (Shutdown Cooling System),
c.
d.
Emergency Condenser System (Isolation Condenser System).
e.
Low Pressure Core Spray System.
f.
Containment Spray System.
g.
Safety, Relief and Safety / Relief Valves.
h.
RCIC (Reactor Core Isolation Cooling) System.
i.
Containment Cooling System.
j.
Containment isolation valves required to chanFe position to isolate containment.
k.
Standby liquid control system (Boron system) 1.
Automatic Depressurization System (any pilot or control valves, associate hydraulic or pneumatic systems, etc.)
Control Rod Drive Hydraulic System (" Scram" function) i m.
Other key components in Auxiliary Systems which are required to direct l
n.
l support plant shutdown or safety system function.
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Inservice Pume and Valve Testing Program I.
Inforuation required for NRC Staf f Review of the Pump and Valve Testing Program 1
A.
Three sets of P&ID's, which include all of the systems listed i
above, with the code class and system boundaries clearly marked.
The drawings should include all of the components present at the time of subesittal and a legend of the P&ID symbols.
B.
Identification of the applicable ASME Code Edition and Addenda.
C.
The period for which the program is applicable.
D.
Identify the component code class.
E.
For pump testing: Identify 1.
Each pump required to be tested (name and number) 2.
The test perameters to be measured 3.
The test frequency F.
For valve testing: Identify 1.
Each valve in ASME Section XI Categories A & B that will be exercised every three months during normal plant operation (indicate whether partial or full stroke exercise, and for power operated valves list the limiting value for stroke time.1 i
2.
Each valve in ASME Section XI Category A that_ vill he leak tested during refueling outages (Indicate the leak test procedure you intend to use) 3.
Each valve in ASME Section XI Categories C, D and E that will be tested, the type of test and the test frequency. For check valves, identify those that will be exercised every 3 months and those that will only be exercised during cold shutdown or refueling outages.
II.
Additional Information That Will Be Helpful in Sneeding Up the Review Process A.
Include the valve location coordinates or other #propriate location information which will expedite our locating the valves on the P&ID's.
B.
Provide PAID drawings that are large and clear enough to be read easily.
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Identify valves that are provided with an interlock to other components and a brief description of that futa tion.
Relief Requests from Section XI Requirements The largest area of concern for the NRC staff, in the review of an inservice valve and pump testing program, is in evaluating the basis for justifying relief from Section XI Requirements. It has'been our experience that many requests for. relief, submitted in these programs, do not provide adequate descriptive and detailed technical information. This explicit information is necessary to provide reasonable assurance that the burden imposed on the licensee in complying with the code requirements is not justified by the increased level of safety obtained.
Relief requests which are submitted with a justification such as
" Impractical", " Inaccessible", or any other categorical basis, will require additional information, as illustrated in the enclosed examples, to allow our staff to make an evaluation of that relief request. The intention of this guidance is to illustrate the content and extent of information required by the NRC staff, in the request for relief, to make a proper evaluation and adequately document the basis for that relief in our safety evaluation report. The NPC staff feels that by receiving this information in the program submittal, subsequent requests for additional information and delays in completing our review can be considerably reduced or eliminated.
I.
Information Required for NRC Review of Relief Requests l
A.
Identify component for which relief is requested:
1.
Name and number as given in ySAR 2.
Function 3.
ASME Section III Code Class 4.
For valve testing, also specify the ASME Section XI valve category as defined in IWV-2000 B.
Specifically identify the ASME Code requirement that has been determined to be impractical for each component.
C.
Provide information to support the determination that the requirement in (B) is impractieml; i.e.,
state and explain the basis for requesting relief.
D.
Specify the inservice testing that will be performed in lieu of the ASme Code Section XI requirements.
E.
Provide the schedule for implementation of the procedurefs) in (D).
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i II. Examnles to Illustrate Several Possible Areas Where Relief Mav Re Granted and the Extent and Content of Information Necesserv to Make An Evaluation A.
Accessibility: The regulation specifically grants relief from the code requirement because of insufficient access provisions.
. However, a detailed discussion of actual physical arrangement of the component in question to illustrate the insufficiency of space for conducting the required test is necessary.
Discuss in detail the physical arrangement of the component in question to demonstrate that there is not sufficient space to perform the code required inservice testing.
i What alternative surveillance means which will provide an L
acceptable level of safety have you considered and why are the,se means not feasible?
B.
Environmental Conditions (e.g., High radiation level, High temperature, High humidity, etc.)
Although it is prudent to maintain occupation radiation exposure for inspection personnel as low as practicable, the request for relief from the code requirements cannot be granted solely on the basis of high radiation levels alone. A balanced judgment between the hardships and compensating increas'e in the level of safety should be carefully established. If the health and safety of the public dictates the necessity of inservice testing, alternative 2
means or even decontamination of the plant if necessary should be provided or developed.
Provide additional information regarding the radiation levels at the required test location. What alternative testing techniques which will provide an acceptable level of assurance of the integrity of the component in question have you considered and whv are these techniques determined to be impractical?
C.
Instrumentation is not originally provided i
d Provide information to justify that compliance w th the co e i
requirements would result in undue burden or hardships without a compensating increase in the level of plant safety. What alternative testing methods which will provide an acceptable level of safety have you considered and why are these methods determined to be impractical?
D.
Valve Cycling During Plant Operation could put the Plant in an Unsafe. Condition The licensee should explain in detail why exercisine tests during plant operation could jeoperdize the plant safety.
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Valve Testing at Cold Shutdown or Refueling Intervals in lieu of the 3 Month Required Interval The liensee should explain in detail why each valve cannot be exercised during normal operation. Also, for the valves where a refueling interval is indicated explain in dedail why each valve cannot be exercised during cold shutdown intervals.
III. Accentance Criteria for Relief Reouest The Licensee must successfully demonstrate that:
1.
Compliance with the code requirements would result in hardships or unusual dif ficulties without a compensating increase in the level of safety and noncompliance will provide an acceptable level of quality and safety, or 2.
Proposed alternatives to the code requirements or portions therof will provide an acceptable level of quality and safety.
Standard Format 5
A standard format, for the valve portion of the pump and. valve' testing program and relief requests, is included as an attachment to this Guidance.
The NRC staf f believes.that this standard format will reduce tha time spent by both the staff in our review and by the licensee in their preparation of the pump and valve testint program and submittals. The standard format includes examples of relief requests which are intended to illustrate the application of the standard format and are not necessarily a specific plan relief request.
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. Ca tr: gory S ir.e Valve Act.intor Normal Require-Relief
- TestirP, usesI for I,'
Class Coordinates ABCnE (inchen) Type Type Position ments Requests
- Alternettve relief hasis) i i
3 D-14 X
4 CA H
I,0 ET I
1 n-15 X
6 DE NA C
DT 3
C-15 X
16 CK SA CV X
CS CV 3
C-15 x
16 CK SA 3
E-14 X
3 REl.
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'l R-11 X
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3/4 REl.
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< e NT 30 sec.
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!,,J Legend fo-Valve Testing Examnle Format Q
- Exercise valve (full stroke) for operability every (3) months.
LT - Valves are leak tested per Section XI Article IW-3420.
NT - Stroke time measurements are taken and compared to the stroke time limiting value per Section XI Article IW 3410.
CV - Exercise check valves to the sosition required to fulfill their function every (3) months.
SRV - Sefety and relief valves are tested per Section XI Article IW-3510.
DT - Test Category D valves per Section XI Article IW-3600.
ET - Verify and record valve position before operations are performed and after operations are completed, and verify that valve is locked or sealed.
CS - Exercise valve for operability every cold shutdown.
RR - Exercise valve for operability every reactor refueling.
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e Relief Recuest Basis System: Auxiliary Coolant System, Component Cooling 1.
Valve:
717 Category:
C 3
Class:
Function:
Prevent backflow from the reactor coolant pump cooling coils.
Impractical Test Requirement:
Exercise valve for operability every three months.
Basis for relief:
To test this valve would require interruntion
' f cooling water to the reactor olant pumps motor cooling coils. This action could result in damage to the reactor coolant pumps and thus place the plant in an unsafe mode of operation.
Alternative Testing:
This valve will be exercised for operability during cold shutdowns.
2.
Valve:
834 Category:
B-E Class:
3 Function:
Isolate the primary water from the component cooling surge tank during plant operation. It is normally in the closed position, but routine j
operation of this valve will occur during durint refueling and cold shutdowns.
i Impractical Test' Requirement:
Exercise valve (full stroke) for operability every three (3) months.
l Basis for Relief:
This valve is not required to chance position l
during plant operation to accomplish its safety i
function. Exercising this valve will increase l
the possibility of surge tank line contamination.
I s
l y Alternate Testine:
Verifv and record valve oosition before and af ter each valve operation.
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3.
Valve:
7t+4B Category:
A Class:
2
' Function:
Isolate the residual heat exchangers from the leg R.C.S. backflow and accumulator backflow.
Test Requirements:
Seat leakage test.
Basis for Relief:
This valve is located in a high radiation field (2000 ar/hr) which would make the required seat leakage test hazardous to test personnel. We intend to seat leak test two other valves (875B and 8765) which are in series with this valve and will also prevent backflow. We feel that by comp 1ving wiht the seat leakage requirements we will not achieve a compensatory increase in the level of safety.
Alternative Testing:
No alternative seat leak testing is proposed.
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s' SB 1 & 2 FSAR RAI 210.42 (3.9(N).l.2, Page 3.9(N)-20)
Provide references 1 and 2 for our review.
RESPONSE
Reference 1 (WCAP-8252) has been reviewed and approved by the NRC. Reference 2 (WCAP-8929) has been submitted to the NRC and is currently being reviewed by the NRC and Oak Ridge National Labs.
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RAI 210.43 (3.9(N).l.4, Pete 3.9(N)-33)
How is the critical buckling strength for component supports determined?
RESPONSE
Westinghouse performs buckling analysis in accordance with 'the requirements i
of the ASME Code,Section III, Appendix F, and meets the 2/3 critical buckling criteria. Subsection L 'ini...- of the FSAR..
-, revise to delete the exceptiontoAppendixFwhich'[currentlystates'e; dhat 90% of critical buckling vill be met.
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lands 1 W *=4 t.: 0.67 times the eritisal If, as. a t of uses. d=*=tt-d evales the supports the member ive axial leede ehema to safely esamed 0.67 z'
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tanding eeubi=efg== and allamable stresses for III class 1 '
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med. supports are sives in Tables 3.9(s)-I and 3.9(s)-3..
m e yaelted.==w <,4 = a=.r= e4===,.the effects of the safe abstdoom
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(loca) are sembiand,
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.amatiff==e4-for this mathed. et leas'emmbiascians is, eestained is
- asfaresses (4 and (5).."
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l 9(E).1 Dyannis Testiar and Assitois '
3 9(E).1.1....
_- --' vibratise and Dynamic Efforts Testian esi pisiar.'
.A.peesperatismal. piping vibrational amt dyameias offsets testias progran
.. sill be **=+ for thm seester aoelmac laep/ supports system during
_ - ++-' toeting. The purpose ei t'asse toees wiIl be to -84= that the eyecan baa beam; adegastel, designed as.1 supported' for.vibraties, as-requires by en ef== III of tha'AssE Code, p.aragraph E3 3622.3. Ihm tests sill. Instade :===e-emelaec pony starts set tripe. If vibraties is.
==y= =di *isk,. freer vissaL eheervation, appears to be 'ensessive,
=4 *hme+-
- 1) an. Immew====ead-tese progener ser the piping, will be condested aos tho' system reemelysed to dessestrate thee tha ebeerved. levels will met.
samme. AatE. Code strese aos fuisse limits to.be essended, 2) the sauce of ths.exesseive vibratise will be,alf=i==*=d, oc 3) ther support system will be
==dified ter redeem: the vibrat on. pare 4==t=e attantion will be provided se i
theos lasatisme eheuer the vihrstiam la==y==e=d ts he the meet severe for the par *4
'- er==l==e *esadities being stadiad.
It ahestd be meted-thee the layout, sise, ets., of the rematar emelaat loop and merge line pipias used in the sesbrook plaats is very similar to that employed. Ese Wese4 =gh==== planer ase is operatics. The operating +--g-- ' n :
that has been obtained from these plante indicates that the reester seelanc
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SB 1 & 2 FSAR RAI 210.44 (3.9(N).2.5, Panes 3.9(N)-38 to 44)
Previous analysis for other nuclear plants have shovu'that certain reactor system components and their supports may be subjected to previously under-estimated asymmetric loads under the conditions that result from the postulation of ruptures of the reactor coolant piping at various loentions.
The applicant has described the design of the reactor internals for blowdown loads only. The applicant should also provide information on asymmetric loads. It is, therefore, necessary to reassess the capability of these
- reactor system components to assure that the calculated dynamic asymmetric loads resulting from these postulated pipe ruptures will be within the bounds
~
necessary to provide high assurance that the reactor can be brought safely to a cold shutdown condition. The reactor system components that require reassessment shall include:
a.
b.
Core supports and other reactor internals.
c.
Control rod drives.
d.
ECCS piping that is attached to the ' primary coolant piping.
e.
Primary coolant piping.
f.
Reactor vessel supports.
The following information should be included in the FSAR about the effects of ' postulated asymmetric LOCA loads on the above mentioned reactor system components and the various cavity structures.
1.
Provide arrangement drawings of the reactor vessel support systems in sufficient detail to show the geometry of all principal elements and materials of construction.
2.
If a plant-specific analysis will not be submitted for your plant, pro-vide supporting information to demonstrate that the generic plant analysis under consideration adequately bounds the postulated accidents at your facility. Include a comparison of the geometric, structural, mechanical, and thermal-hydraulic similarities between your facility and the case analyzed. Discuss the effects of any differences.
3.
Consider all postulated breaks in the reactor coolant piping system, including the following locations:
a.
Steam line nozzles to piping terminal ends.
b.
Feedwater nozzles to piping terminal ends.
c.
Recirculation inlet and outlet nozzles to recirculation piping terminal ends.
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Provide ar. assessment of the effects of asyuusetric pressure dif ferentials*
on the systems and components listed above in combination with all external loadings including safe shutdown carthquake loads and other faulted condition loads for the postulated breaks described above. This assess-ment may utilize the following mechanistic effects as applicables a.
Limited displacement break areas.
b.
Fluid-structure interaction.
c.
Actual time-dependent forcing function.
d.
Reactor support stiffness.
e.
Break opening times.
5.
If the results of the assessment on item 3 above indicate loads leading to inelastic action of these syste as or displacement exceeding previous design limits, provide an evaluation of the inelastic behavior (including strain hardening) of,the material used in the system design and the effect of the load transmitted to the backup structures to which these systems are attached.
6.
For all analyses performed, included the method of analysis, the structural and hydrsulic computer codes employed, drawings of the models employed and comparisons of the calculated-to-allowable stresses and strains or deflections with a basis for the allowable values.
j E
7.
Demonstrate that safety-related components will retain their structural integrity when subjected to the combined loads resulting from the loss-of-coolant accident and the safe shutdown earthquake.
l 8.
Demonstrate the functional capability of any essential piping when sub-l jected to the combined loads resulting from the loss-of-coolant accident and the safe shutdown earthquake.
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RESPONSE
l Westinghouse, in its analyses of reactor system components and their supports, has considered asymmetric LOCA loadings in the Seabrook plant design and E
analysis. The analysis methods used by Westinghouse are consistent with l
NUREG-0609. This question is addressed in Seabrook FSAR Subsections 3.9(N).2.5, 3.9(N).3 and the revised / updated Subsections 3.9(N).1.2, 3.9(N).1.4b, 3.9(N).1.4c, i
3.9(N).1.4d, 3.9(N).1.4e, 3.9(N).1.5, 3.9(N).1.6, 3.9(N).1.7, 3.9(N).4.2,
/Of/c4 Jr4 ffsvik l'
3.9(N).4.3 and 3.9(N).4.4 d * ' 211 L.
y-II;J 12::
in h en) utnt. +7.
Blowdown jet forces at the location of rupture (reaction forces), transient differential pressures in the annular region between the component and the vall, and transient differential pressures across the core barrel within the l'
reactor vessel.
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in the fatigue usage factor caused by these tests is easily i
covered by the conservative number (200) of primary side leakage tests that are considered for design.
l 2.
Secondary Side Hydrostatic Test The secondary side of the steam generator is pressurized to 1.25 design pressure with a minimum water temperature of 1200F, coincident with the primary side at 0 psig.
For design purposes, it is assumed that the steam generator will experience 10 cycles of this test.
~
These tests may be performed either prior to plant startup, or subsequently following shutdown for major repairs or both.
3.9(N).1.2 Comouter Programs Used in Analyses The following computer programs have been used in dynamic and stat!c analyses to determine mechanical loads, st'resses, and deformations of seismic Category I components and equipment. These are described and verified in References (1) and (2).
a.
WSTDYN-7
~
Static and dynamic analysis of redundant piping systems.
b.
FIXFM Time history response of three-dirnensional structures.
2 c.
WSDYN-2 Fiping system stress analysis from time history displacement data.
d.
STERUST Hydraulic loads on loop components from blowdown information.
e.
WSAN Reactor coolant loop equipment suppore structures analysis and l
evalua'. ion.
f.
WCAN l
_ Finite element structural analysia.
~-
Dynamic transient response analysis of reactor vessel and internals.
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3.9(N).1.3 Erperimental Stress Analysis No experimental stress analysis methods are used for seismic Category I systems or components. Ho e ver, Westinghouse askes extensive use of measured results from prototype plants and various scale model tests, as discussed in Subsection 3.9(N).2.
3.9(N).1.4 Cor riderations for the Evaluation of the Faulted Condition a.
Loading conditione The structural stress analyses performed on the reactor coolant system consider the loadings specified as shown in Table 3.9(N)-2.
These loads result from thermal expansion, pressure, weight,
+
Operating Basis Earthquake (OBE), Safe Shutdown Earthquake (SEE),
design basis loss of coolant accident, and plant operational thermal and pressure transients.
i b.
Analysis of the teactor Coolant Loop and Supports IMsEET) j The loads used.in the analysis of the esaccor coolant 1-,,*,'-.
are described in detail below.
f a[
1.
Pre ssure Pressure loading is' identified as either membrane design pressure or general operating pressure, depending upon its application. The membrane design pressure.is used in connection with the longitudinal pressure stress and minimum well thickness calculations in accordance with the ASME Code.
. The term operating pressure is used in connection with determination of the system deflections and support forces.
The' steady-state operating hydraulic forces based on the
~
system initial pressure are applied as general operating
~
pressure loads to the reactor coolant loop model at changes l
in direction or flow area.
2.
Weirh t A weight analysis is performed' to meet Code requirements by l
applying a 1.0 g load desnward on the complete piping system.
The piping is assigned a distributed mess or weight as a function of its properties. This method provides a distributed loading-to the piping system as a function of the weight of the pipe and contained fluid during normal operating conditions ~.
3.
Seismic
~
The forcing functions for the reactor coolant loop. piping seismic analyses are three orthogonal components of 3.9(N)-21 e
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a2/0.W INSERT A (Modification to 3.9(N).1.4b)
The reactor coolant. loop piping is evaluated in accordance with the Criteria of ASME III,'NB-3650 and Appendix F.
The leads included in the evaluation result from the SSE, dead weight, pressure, and LOCA loadings (loop hydraulic forces, asymmetric sub-compartment pressurization forces, and reactor vessel motion).
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The component upper and lower lateral supports are inactive during plant heacup, cooldown and normal plant operating conditions.
)
However, these restraints become active when the plant is at power and under the rapid motions of the reactor coolant loop l
components that occur from the dynamic loadings and are represented by stiffness matrices and/or individual tension or compression l
spring members in the dynamic model. The analyses are performed i
at the full power condition.
Me total response is obtained directly by d8, rect time integration of the equations of motion. The results of the time history analysis are forces and displacements. The time hist:,ry displacement response is then used in computing support loads and in performing the reactor coolant loop piping stress evaluation.
3.
Loss of Coolant Accident The mathematical model used in the static analyses is modified for the loss of coolant accidant analyses to represent the severance of the reactor coolant loop piping at the postulated break location.- Modifications include addition of the mass characteristic of the piping and equipment. To obtain the proper dynamic solution, two masses, each containing six dynamic degrees of freedom and located one on each side of the break, are included.in the mathematical'model. The natural frequencies and eigenvectors are determined fron this broken loop model.
The time-history hydraulic forces at the node points are combined 2
to obtain the forces and moments acting at the corresponding structural 16mped-mass node points.
The dynamic structural solution for the full power loss of 1
coolant accident and steam line break is obtained by using a i
modified predictor-corrector-integration technique and normal mode theory.
When elements of the system can be represented as single acting members (tension or compression members), they are considered as nonlinear elements, which are represented mathematically by the combination of a gap, a spring, and a viscous damper.
The force in this non-linear element' is treated as an externally applied force in the overall normal mode solution. Multiple l
[
non-linear elements can be applied at th6 same node, if necessary.,
The time-history solution is performed in subprogram FIXFM3 The input to this subprogram consists of the natural frequencies, normal modes, applied forces and nonlinear elements. The natural frequencies and normal modes for the modified reactor l
j coolant loop dynamic model are determined with the WESTDYN-7
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program. To properly si nulate the release of the strain energy I
l 3.9(N)-25 4
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FSAR 9
+
in the pipe, the internal forces in the syste the postulated break location due to the initial steady-sutt
.,Jr.ulic forces, thermal forces, and weight forces are determined. The release of the strain energy is accounted for by spriying the negative of these internal forces as a step function leadhr. The initial conditions are equal to zern because the -tution is only for the transient problem (the dynamic respx e of the syseen from the static equilibrium position). The time history displacement solution of all dynamic degrees of freedom is obtained using subprogram FIXndand employing 4 percett critical,
damping.
The loss of coolant accident displacements of the reactor vessel are applied in time history form as input *o the dynamic analysis of the reactor coolant loop. The loss o-oolant accident analysis of the reactor vessel includes a.1 the forces acting on the vessel including internals reactions, cavity pressure loads, and loop mechanical loads. The reactor vessel analysis is described in Subsection 3.9(N).1.4f.
.IN3fA'7)
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.The time-history displacement response of the loop is used in computing support loads and in performing stress evaluation of the reactor coolant loop piping.
supp rt Los are omput Dy Acap ing e sup rt '
p i ffne a er d h di 1 e y et e t e u ort oi Th pport ads eu in ee luatt of he p,.pp e-2 v-L The time-history displacements of the FIXFl6 subprogram are d
l used'as input to WESDYN-2' to determine the internal forces, i
deflections,. and stresses at each and of the piping elements.
For this calculation the displacements are treated as imposed l
deflections on the reactor coolant loop masses. Sitr.sesuJar l
Alghie AMurion areMnfahVpiping"s&cese sadOtiegy I-4.
Transients Operating transients in a nuclear power plant cause thermal and/or pressure fluctuations in the reactor coolant fluid.
The thermal transients cause time-varying temperature distributions across the pipe wall. These temperature distributions resulting in pipe well stresses may be further-subdivided in accordance with the Code into three parts, a uniform, s. linear, and a non-linear portion. The uniform portion results in general expansion loads. The linear portion causes a-bending moment across the well and the non-linear portion causes a skin stress.
3.9(N)-26 i
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INSERT B (Modification to 3.9(N).la4c) l The resultant asymmetric external pressure loads on the RCP and stesu generator resulting from a postulated pipe rupture and pressure build up in the loop compartments, are applied to the uane integrated RCL/ supports i
system model used to compute loadings on components, component supports and RCL piping as discussed above. The response of the entire system is obtained for the various external pressure loading cases from which the internal member forces and piping stresses are calculated for each pipe i
break case considered.
The equipment support, leads and piping stresses resulting from the external pressure loading are added to the support loads and piping stresses calculated using the loop LOCA hydraulic forces and RPV motion.
The break locations considered for swbcompartment pressurization are those postulated for the RCL LOCA analysis, as discussed in Section 3.6N and WCAP-8172 (Ref 1 of Section 3.6N). The asysmetric subcompartment, pressure loads are provided to Westinghouse by United Engineers & Constructees.
TheanalysistodetethnetheseloadsisdiscussedinSection6.2.
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For all possible load set combinations, the primary plus-secondary and peak stress intensities, fatigue reduction factors and cumulative usage factors are calculated. The WESTDYN-7 program is used to perform this analysis in accordance with the ASME Boiler and Pressure Vessel Code,Section III, Subsection NB-3650. Since it is impossible to predict the order or occurrence of the transients over a forty year life, it is assumed that 1
the transients can occur in,any sequence. This is a very l
conservative assumption.
The combination of load sets yielding the highest alternating stress intensity range is used to calculate the incremental
]
usage factor. The next.most severe combination is then determined and the incremental usage factor calculated. This procedure is repeated until all combinations having allowable cycles j
< 100 are formed. The total cumulative usage factor at a point is the summation of the incremental usage factors.
d.
Primary t'monent supports Models and Methods The static and dynamic structural analyses employ the matrix method and normal mode theory for the solution of lumped parameter, multimass structural models.. The equipment support structure models are dual purpose since-they are required tot 1) quantitatively represent the elastic restraints which the supports impose upon the loop,.
and 2) evaluate the individual mapport member stresses due to the forces imposed.upon the sup'> orts by the loop.
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A description of the supports is found in Sinbsection 5.4.14.
Detailed models er's developed using beam elements and plate elements, where applicable. (/NSEE1'C)
The respective cougouter programs are used with these models to obtain support stiffness matrices and member influence coefficients for the steam generator, reactor coolant pump, pressuriser and reactor vessel supports. Unit force along and unit moment about each coordinate axis are applied to the models at the equipment vertical centerline joint. Stiffness analyses are performed for each unit. Ioad for each model.
Joint displacements for applied unit loads are formulated into flexibility matrices. These are inverted to obtain support stiffness matrices which were included in the reactor coolant loop model.
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3.9(N)-29
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INSERT C (Modification to 3.9(N).1.4d)
The reactor vessel supports are modeled using the WECAN computer program.
Structure geometry, topology and member properties are used in modeling.
Steam generator and reactor coolant pump supports are modeled as linear or non-linear springs.
For each operating condition, the loads (obtained from the RCL analysis) acting on the support structures are appropriately combined. Reactor coolant loop normal and upset conditions thermal expansion loads are treated as pr'. mary loading for the primary component supports. The adequacy of each member of the steam generator supports, reactor coolant pump supports, and piping restraints is verified by solving the ASME III subsection NF stress and interaction equations. The adequacy of the RPV Support Structure is verfied using the WECAN t!omputer program and comparing the resultant stresses to the criteria 3 s s in ASME III subsection NF.
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FSAR Loads acting on the " supports obtained from the reactor coolant loop analysis, support structure member properties, and influence into the WESAN coefficients at each and of each member are input program.
For each support case used, the following is performedt Combine the various types. of support plane loads to obtain 1.
operating condition loads (Normal,' Upset, Leergency or Faulted).
Multiply member influence coefficients by operating condition 2.
loads to obtain all member internal ferees and moments.
3.
Solve appropriate stress or interactica equations for the specified operating condition. Maximum normal stress, shear stress, and combiced l'oad interaction equation values are printed as a ratio of maxisua actual values divided by limiting values. ASME Boiler and Pressure Vessel Code Section III, Subsection NF, stress and interaction equations are used with limits for the operating condition specified.
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.o
'o e.
Analysis of Primary Components Equipment which sert,es as part of the pressure boundary in the reactor coolant loop include the steam generators, the reactor coolant pumps, the pressurizer, and the reactor vessel. This equipment is ANS Safety Clasa 1 and the pressure boundary meets the requirements of the ASMr. Boiler and Pressure vessel Code,Section III, Subsection This equipment is evaluated for the loading combinations outlined NB in Table 3.9(N)-2.
The equipment is analyzed for: 1) the normal 3
loads of deadweight, pressure and thermal, 2) mechanical transients
/
of OBE, SSE, and pipe rupturesdaad 3) pressure and te eratuq transients outlined in Subsection J.9(N).1.1. b ? S UF _.j(.
~
~
i:
4..
The resdTes 'of the reactor coolant loop analysis-are used to determine the loads acting on the equipment nozzles and the support / component interface locations. These loeds are supplied for all loading conditions on an "unbrella" load basis. That is, on the basis of previous plant analysis, a set of loads are determined which should The "unbrella" be larger than those seen in any single plant analysis.
loads represent a conservative means of allowing detailed component analysis prior to the completion of the system analysis. Upon completion of the system analysis, conformance is demonstrated between the actual plant loads and the loads used in the analyses e
e 3.9(N)-30 4
d 2/O YY' r
)
INSERT D (Modification to 3.9(N).1.4e) including the effects of sayumetric subcompartment pressurization (for i
vessel nozzle breaks) i l
4 1
0 e
t 4
i e
e e
O e
e w%=,
.-...s.
,,e,
--,,,m,
-,,--,,,c 3-9 t
J/0.f SB 1 & 2 FMR of the components. Any deviations where the actual load is larger than the "uabrella" load will'be handled'by individualized analysis.
Seismic analyses are performed individually for the reactor coolant pump, the pressuriser, and the steam generator. Detailed and complex j
4 dynamic models are used for the dynamic analyses. The response spectra corresponding to the building elevation at the highest component / building attachment elevation is used for the component analysis. Seismic analyses for the steam generator and pressuriser are performed using 2 percent damping for the OBE and 4 percent damping for the SSE. The analysis of the reactor coolant pump for determination of loads on the motor, main flange, and pump internals I
is performed using the damping for bolted steel structures, that is, 4 percent for the OBE and 7 percent for the SSE (2 percent for OBE and 4 percent for SSE is used in the system analysis). This damping is applicable to the reactor coolant. pump since the main flange, motor stand,.and motor are all bolted assemblies (see Section 5.4).
The tsaccor pressure vessel is qualified by static stress analysis based on loads that have been derived from dynamic analysis.
I The pressure boundary portions of Class 1 valves in the RCS are designed and analysed according to the requirements of NB-3500 of ASME III.
Valves in sample lines connected to the RCS ara not considered to be ANS Safety Class-1 nor-ASME Class 1.
This is because the nozzles where the lines connect to the primary syrtes piping are orificed to a 3/8 inch hole This hole restricts'tLa flow such that loss.
through a severance of one of these lines can be made up by normal charging flev.
-_ %vw vessel support LOCA Loads
\\
LOCA anal is which is perfo or the reac or sel suppo e nayumetric pr ure distribu s on the i ernals los includes a
and e vesse exterior walls A detailed amic el of t e react asel and i carnal is prepared c
nelu s the tiffne of the reacto ve el support d the tach p ing dra lic forces ar d veloped re t the re tor vessel n zle; these in the ' ternals rg*the force are character d
ime dependen forcing fu e 'ons on (e#
the seel and core b In the deriva~ on of t se reing (or hydroela ic) i eracti in g [$f*
fu tions, the flu ru r
th downconer etwee barrel and e
esel is e en,
F 3, 8 $g Lato account ab sk at th ssel nozzle 1 o allows an asyneetri essure istributio d a subseque force ou the
. tide o e vessel a calculated o time his o basis for these I
F as ric loads. As a result of th ipe b 'ak, loop sechanicalg-
.i r q;- e.
a.
....t
(
\\
_ f _-
~~
1
.e V.
3.9(N)-31 N
i o
I
.~ -
(.---------
l0 YY S b 1 $ t.
A
)
FsMt.
o Thefpr' s e b unda artio s of C s 1 valv i the esi ed and naly a o ding to he equir nts f
-3500 the A
Code,. ec III.
se v ves ar entifi Sec on
.9, 3.2 Valv ple1 es onn ted t - he RCS e et nsider to be Safe lass nor AS Class This s beca e the ozz wh e t
11 con to the rim ystem ping a or i to a 3
/
nch ol e.
i hole stric the f s h at lo hrough s
f of the liner e ' ha
'; ku J-= 1 rh=q a
~
=da
- ance j
~
5 13.9N.1./;g Dynamic Analysis of Reactor Pressure Vessel for Postulated Loss of Coolant Accident 1.
Introduction This section presents the method of computing the reactor pressure vessel response to s postulated loss' of coolant acc.ident (LOCA). The structural analysis considers simultaneous application of the time-history loads on the reactor vessel resulting fran the reactor coolant loop mechanical loads, internal hydraulic pressure transients, I
and reactor cavity pressurization (for postulated breaks in the reactor coolant pipe at the vessel nozzles). The vessel is restrained by reactor vessel sup. port pads ard shoes beneath four of the reactor vessel nozzles and the reactor coolant loops with the primary supports of the steam generators and the reactor coolant pumps.
Pipe displacement restraints installed in the primary shield wall limit em g,
M =.
m 3.1C & 3l V
W
g f4 I (1
!$2 /O-o Rgt the break opening area of the vessel nozzle pipe' breaks to less than 144 square inches. This break area was determined to be an upper bound by using worst case vessel and pipe relative motions based on similar plant analyses. Detailed studies have shown that pipe breaks at the I
hot or cold leg reactor vessel nozzles, even with a limited break area, would give the highest reactor vessel support loads and the highest vessel displacements, primarily due to the influence of reactor cavity j
pressurization. By considering these breaks, the most severe reactor vessel support loads are detemined. For completeness, an additional break outside the shield wall, for which there is no cavity pressurization, was also analyzed, specifically, the pump outlet nozzle
}
pipe break.
2.
Interface Infomation
-~.
Asynnetric reactor _cayity_pyessurization loads were provided to Westinghouse!
b) dmied. En3 meets An L Constew. tors.
~
~~
^
All other input infomation was developed within Westinghouse. 'This
^
information includes: reactor internals properties, loop mechanical i
loads and loop stiffness, internal hydraulic pressure transients, and
{
reactor support stiffnesses. These inputs allowed formulation of the
}
mathematical models and perfomance of the analyses, as will be described.
1 3.
Loading Conditions Following a postulated pipe rupture at the reactor vessel nozzle, the reactor vessel is excited by time-history forces. As previously mentioned, these forces are the combined effect of three phenomena:
(1) reactor coolant loop mechanical loads, (2) reactor cavity l
pressurization forces and (3) reactor internal hydraulic forces.
The reactor coolant loop mechanical forces are derived fran the elastic
?
._W T._-.
ow-
--r
-e e
-w--re-e-
-..we.-..v,--e,ea-. _ - -
+-
- ^
"^
^
~ " - -
/OiY
/
analysis of the loop piping for the postulated break. This analysis is described in Section 3.9N.1.4C ;. The loop mechanical forces which are released at the broken nozzle are applied to the vessel in the RPV blowdown analysis.
Reactor cavity pressurization forces arise for the pipe breaks at the vessel nozzles from the steam and water which is released into the reactor cavity through the annulus around the broken pipe. The reactor cavity is pressurized asymmetrically with higher pressure on the side of the broken pipe resulting in horizontal forces applied to the reactor vessel. Smaller vertical forces arising from pressure on the bottom of the vessel and ti.e vessel flanges are also applied to the reactor vessel. The cavity pressure analysis is described in Section 6.2.
The internals reaction forces develop from asymmetric pressure
' distributions infide the. reactor vessel.. For a vessel inlet nozzle break and pump outlet nozzle break, the depressurization wave path is through the broken loop inlet nozzle and into the region between the core barrel and reactor vessel. This region is called the downcomer annulus.. The initial waves propagate up, down and around the downcomer annulus and up through the fuel. In the case of an RPV outlet nozzle break the wave passes through the RPV outlet nozzle and directly into the upper internals region, depressurizes the core, and enters the downcomer annulus frcyn the bottom of the vessel. Thus, for an outlet l
nozzle break, tgdowncomer annulus is depressurized with much smaller
' differences in pressure horizontally across the core barrel than for the inlet break. For both the inlet and outlet nozzle breaks, the depressurization waves continue their propagation by reflection and l
translation through the reactor vessel fluid but the initial depressurization wave has the greatest effect on the loads.
l The reactor internals hydraulic pressure transients were calculated including the assumption that the structural motion is coupled with the
~.. _
3.101J-31 C
~
~
L
/a W
)
~
\\
pressure transients. This phenomena has been referred to as s
hydroelastic coupling or fluid-structure interaction. The hydraulic analysis considers the fluid-structure interaction of the core barrel by accounting for the deflections of constraining boundaries which are i
represented by masses and springs. The dynamic response of the core barrel in its beam bending mode responding to blowdown forces compensates for' internal pressure variation by increasing the volume of the more highly pressorized regions. The analytical methods used to i
develop the reactor internals hydraulics are described in WCAP-8708$].
4.
Reactor Vessel and Internals Modeling The reactor vessel is restrained by tw mechanisms: (1) the four attached reactor coolant loops with the steam generator and reactor coolant pump primary supports and (2) four reactor vessel supports, two beneath reactor vessel inlet nozzles and two beneath reactor vessel outlet nozzles. The reactor vessel supports are described in Section 5.4.14 and are shown in Figures 5.4-12, and 3.8-17.
The support shoe provides restraint in the horizontal directions and for downward recctor vessel mction.
2 The reactor vessel model consists of two non-linear elastic models
[
connected at a common node. One model represents the dynamic vertical characteristics of the vessel and its internals, and the other model represents the translational and rotational characteristics of the structure.1 These two models are combined in the DARI-WOSTAS code [1] to mpresent motion of the reactor vessel and its internals in the plane of the vessel centerline and the broken pipe centerline.
The model for horizontal motion is shown in Figure 3.9N-12. Each node has one translational and one rotational degree of freedom in the vertical plane containing the centerline of the nozzle attached to the broken pipe and the centerline of the vessel. A combination of beam-elements and concentrated masses are used to represent the components
~.
I WI S.A M
W
, W W,
m,
~
A_
\\
/0.</ /
(
)
including the vessel, core barrel, neutron panels, fuel assemblies, and upper support columns. Connectioris'between the various components are
'~
either pin-pin rigid links, translational impact springs with damping or rotational springs.
The model for vertical mction is shown in Figure 3.9N-13. Each mass node has one translational degree cf freedom. The structure is represented by concentrated masses, springs, dampers, gaps, and frictional elements. The model includes the core barrel, lower support columns, bottom nozzles, fuel rods, top nozzles, upper support structure, and reactor vessel.
The horizontal and vertical models are coupled at. the elevation of the primary nozzle centerlines. Node 1 of the horizontal model is coupled i
with node 2 of the vertical model at the reactor vessel nozzle elevation. This coupled node has external restraints characterized by
~
a 3 x 3 matrix which represents the reactor coolant loop stiffness characteristics, by linear horizontal springs which describe the tangential resistance of the supports, and by individual non-linear R
vertical stiffness elements which provide downward restaint only. The supports as represented in the horizontal and vertical models (Figures
~
3.9N-12 an'd 3.9N-13) are not indicative of the complexity of the
^
support system used in the analysis. The individual supports are located at the actual support pad locations and accurately represent the independent non-linear behavior of each support.
l
~
5.
Analytical Methods The time-history effects of the cavity pressurization loads, internals loads and loops mechanical loads are combined and applied simultaneously to the appropriate nodes of the mathematical model of the reactor vessel and internals. The analysis is performed by numerically integrating the differential equations of motion to obtain I
the transient response. The output of the analysis includes the
- 1. f pf)-lic.
t ww
~-
w M
g m
w
.. :..T.C..L. L.~
m-2/0.W displacements of the reactor vessel and the loads in the reactor vessel supports which are combined with other applicable faulted condition loads and subsequently used to calculate tne stresses in the supports.
Also, the reactor vessel displacements are applied as a time-history input to the dynamic reactor coolant loop blowdown analysis. The resulting loads and stresses in the piping components and supports include both loop blowdown loads and reactor vessel displacements.
Thus, the effect of vessel displacements upon loop response and the effect of loop blowdown upon vessel displacement's are both evaluated.
l 6.
Results of the Analysis As describad, the reactor vessel and internals were analyzed for three i
i postulated break locations. Table 3.9N-12 summarizes the displacements and rotations of and about a point representing the intersection of the centerline of the nozzle attached to the leg in which the break was postulated to occur and the vertical centerline of the reactor vessel.
t Positive vertical displacement is up and positive horizontal displacement is away from and along the centerline of the vessel nozzle l
in the loop in which the break was postulated to occur. These displacements were calculated using an assumed break opening area for the postulated pipe ruptures at the vessel nozzles of 144 in2 and a double-ended rupture at the pump outlet nozzle. These areas are estimated prior to performing the analysis. Following the reactor coolant system structural analysis, the relative motions of the broken ~
pipe ends are obtained from the reactor coolant loop blowdown analysis.
1he actual break opening area is then verified to be less than the estimated area used in the analysis and assures that the analysis is conservative.
sw The maximum loads induced in the vessel supportsYdue to the postulated pipe break.mpaigngiU5il33EiWiWWuet. These loads are per vessel support and are applied at the vessel nozzle pad. It is conservatively assumed that the maximun horizontal and vertical loads occur
..'[*
O me, 4m M
h
- 1/0 W i
l simultaneously and on the same support, even though the time-history results show that these loads occur neither simultaneously nor on the same support, The largest vertical loads are produced on the support opposito the broken nozzle. The largest horizontal loads are produced -
on the supports which are perpendicular to the broken nozzle horizontal centerline. Note that the peak loads are conservative values since the break opening area for the vessel inlet nozzle break (as obtained from the dynamic loop analysis) is actually less than the estimated 144 square inch area used to generate the applied loads. If additional analysis was performed using the lower break opening area, the loads would be considerably reduced.
4 0
O 9
b e
6 3.f(Aff-jf
--._n.
, -- e
---r
,,m
A_
~
J2/0.W)
/
SB 1 & 2 FSAR The loads from these three sources, the internals reactions, reactor cavity pressure loads, and the loop mechanical forces, are applied simultaneously in a nonlinear elastic dynamic time history analysis on the model of the vessel, supports and internals. The results of this analysis are the dynamic loads on the reactor vessel supports and vessel time history displacewints. The maximum loads are combined with other applicable loads, such as
- eismic and deadweight and applied statically to the vessel support structure. The maxissum stresses in the support are calculated and compared to faulted condition stress allowables given in Subsection 3.9(N).1.4g.
U 3.9(N) /.65 Stress Criteria for Class 1 Components and component Suoports J.-
- "-=-
l --
All Class 1 components and supports are designed and analyzed for the Design, Normal, Upset, and Emergency Conditions to the rules and requirements of the ASME Code Section III. The design analysis or test methods and associated stress or load allowable limits that will be used in evaluation of Faulted Conditions are those that are defined in Appendix F of the ASME Code with supplementary options outlined below 1.
Elastic System Analysis and Component Inelastic Analysis This is an acceptable method of evaluation for Faulted Conditions if the rules of F1323.1(a) are met for component
(
supports, within the scope of Subsection NF and if primary stress, limits for components are taken as greater of 0.70 Su or Sy + 1/3 (Su - S ) for membrane stress and greater of 0.70 y
Sut or Sy + 1/3 (Sue - Sy) for membrane-plus-bending stress, where material properties are taken at appropriate temperature.
If plastic component analysis is used with elastic system analysis or with plastic system analysis, the* deformations and displacements of the individual system members will be shown to be no larger than those which can be properly calculated by the analytical methods used for the system analysis.
2.
. Elas tic / Inelastic System Analysis and Component / Test 1.oad Method The test load method given in F-1370(d) is an acceptable method of qualifying components in lieu of satisfying the stress / load limits established for the component analysis.
If the component / test load method is used with elastic or plastic system analysis, the deformations and displacements of the individual component members taken from the test load 3.9(N)-32
.~
~
s,
c2/O.4 o
SS 1 & 2
(
FSAR method data at the loads resulting from the system analysis will be shown to be no larger than those which can be properly calculated by the analytical methods used for the system analysis.
3.
Component Support Buckling Allowable Load
~
~
In the design of component supports, members compressive axial loads are limited to 0.67 times the critical buckling strength.
If, as a result of more detailed evaluation of the supports the member compressive axial loads can be shown to safely exceed 0.67 times the critical buckling strength for the faulted condition, verification of the support functional adequacy will be documented and submitted to the NRC for review. The me2ber compressive axial loads will not exceed 0.67 times the critical buckling strength without NRC acceptance.
In no case will the compressive load exceed 0.9 times the critical buckling strength.
Loading combinations and allowable stresses for ASME III Class 1 components and supports are given in Tables 3.9(N)-2 and 3.9(N)-3.
For Faulted condition evaluations, the effects of the safe shutdown earthquake (SSE) and loss-of-coolant accident (LOCA) are combined using the square-root-of-the-sum-of-the-squares (SRSS) method.
Justification for this method of load combinations is contained in gp References (4) and (5).
3.9(N).2 Dynamic Testing and Analysis 3.9(N).2.1 Preoperational Vibration and D mic Effects Testing on Piping A preoperational piping vibrational and dynamics effects testing program will be conducted for the reactor coolant loop / supports system during preoperational testing. The purpose of these tests will be to confirm that the system has been adequately designed and supported for vibration, as required by Section III of the ASME Code, paragraph NB-3622.3. The casts will include reactor coolart pump starts end trips. If vibration is L
experienced, which, from visual observation, appears to be excessive, either:
- 1) an instrumented test program on the piping, will be conducted and the system reanalyzed to demonstrate that the observed levels will not cause ASME Code stress and fatigue limits to be exceeded, 2) the cause of the excessive vibration will be eliminated, or 3) the support system will be modified to raduce the vibration. Particular act.ention will be provided at those locations where the vibration is expected to be the most severe for I
the particular transient condition being studied.
It should be noted that the layout, size, etc., of the reactor coolant locp l
and surge line piping used in the Seabrook plants is very similar to that L
employed in Westinghouse plants now in operation. The operating experience l
that has been obtained from these plants indicates that the reactor coviant l :
3.9(N)-33 l
J.es G.
-w----
w yy-w.-
e.
.2/0.W INSERT E - Qiodification to 3.9(N).1.6 and 3.9(N).1.7)
For faulted conditions analysis of class 1 branch piping attached to the reactor coolant loop, Equation (9) of ASME III subsection NB-3652 is applied with a stress limit of 3.0 % This criterion provides sufficient assurance that the piping will not collapse or experience gross distortion such that the function of the system would be impaired. The basis for this position is described in Westinghouse response to NRC Question 110.34 on the RESAR-414 application (Docket no. STN 50-572),
which subsequently received a preliminary design approval (PDA) in Nov.,1978.
3.9(N).1.7 Analytical Methods for RCS Class 1 Branch Lines The analytical methods used to obtain the solution consist of the transfer matrix method and stiffness matrix formulation for the static structural analysis, the response spectrum method for seismic dynamic analysis, and dynamic structural analysis for the effect of a reactor coolant loop pipe break.
l The integrated Class 1 piping / supports system model is the basic system model used to compute loadings on components, component and piping supports, and piping. The system models include the stiffness and mass characteristics of the Class 1 piping components, the reactor coolant loop, and the stiffness of supports which affset the system response. The deflection solution of the entire system is obtained for the various loading cases from which tl.: internal member forces and piping stresses are ca:culated.
Static The Class 1 piping system models are constructed for the WESTDYN computer program, which nwaarically describes the physical system. A network model is made up of a number of sections, each having an overall transfer relationship formed foam its group of elements. Thelinearelastic[propertiesofthesectionareusedtodefine the characteristic stiffness matrix for the section. Using the transfer relationship
~
for a section, the loads required to suppress all deflections at the ends of the section arising from the thermal and boundary forces for the section are obta6ned.
I' After all the sections have been defined in this manner, the overall stiffness matrix i
and associated load vector to suppress the deflection of all the network points is determined. By inverting the stiffness e
ne
e.,m,
-,,m.
ye.e--e-wwe*
- W
- '*"--+-F-*""Fe
- -'-"-B'e-*--~
'*"Tr' me.-
mer----w-:wiw+ - - -gu-wwe'*m'a-e---eve-m'- - - - -
-s--wn-me*"+",*mm-es-y--
meme.--.=ew-+w,-v
%-++s,-d-w-aw ee---
ww-w-
b j { g, l0-(tel l$
matrix, the flexibility matr5 is detehned. The flexibility matrix is' multiplied by the negative of the lead vector to determine the network point deflections due to the thermal and boundary force ef.fects. Using the general transfer relationship, the deflections and internal forces are then determined at all node points in the system.
The support loads are also computed by mutiplying the stiffness matrix by the displacanent vector at the support point.
Seismic The models used in the static analyses are modified for use in the dynamic analyses by including the sass characteristics of the piping and equipment.
The llanping of the distributed mass of the piping systems is
~ ~ - - - - -
. accomplished by locating the to':a1 sass at points in the system dich
- - ~ ~ ~
will appropriately represent the response of the distributed system.
-l'-~"
. Effects of the prinary equipment motion.,that is, reactor vessel, steam
~ ~ ~ ~ ~ -
generator, reactor coclant pump, and pressurizer, on the Class 1 piping
' " ~ "
system are obtained by modeling the nars and the stiffness
~~~
characteristics of the primary equipment and loop piping in the overall
. system model.
~
The supports are represented by stiffness matric.es in the system model
[~
~~'
for the dynamic analysis. Shock suppressors which resist rapid motions
]'-
- are also included in the analysis. The solution for the seismic disturbance employs the response spectra method. This method employs
[,
- the lumped mass t.3chnique, linear elastic properties, and the principle of mod $ superposition.
The total response obtained from t,he seismic analysis consists of two parts: the inertia response of the piping system and the response from differential anchor motions. The strisses resulting from the anchor motions are considered to be secondary and, therefore, are included in the fatigue evaluation.
- 3. p.n o
+
.,----.,c--
l pd. tw g)
S 6.! $ L..._
. ?/0 Tl,
. b- ---
fI$
1 I
Loss of Coolant Accident 6
The mathematical models used in the seismic analyses of the Class 1
' - " ~
lines are also used for RCL pipe break effect analysis. To obtain the
' ~ ~ ~ -
proper dynamic solution both for lines attached to the unbroken loops and lines attac,hed to the broken loop, the time history deflections.
~
" ~ " - ~ ~ ~
from the analysis of the reactor coolant loop are applied at branch nozzle connections.~
- ~ - -
_.. Fatique A thermal transient heat transfer analysis is performed for each different piping component on all the Class 1 branch lines. The normal, upset, and test condition transients identified in Section 3./9.1.1 are considered in the fatigue. evaluation.
- ~ ~ - - -
T,- T, and a,T,, -a Tb b.are calculated on a The thennal quantities g
2 time history basis, using a one-dimensional finite difference hea't transfer computer program. Stresses due to these quantities were I
- '-~
calculated for each time increment using the methods of NB-3650 of ASME
~~ -' "
'l For each thermal transient, two loadsets are defined, representing the
~
maximta and minimum stress states for that tr,ansient.
~
As a result of the normal mode spectral technique employed in the
-~'
seismic analyds, the load components cannot be given signed values.
Eight load sets are used to represent all possible sign permutations of,_,,,,
~
the seismic mannents at each point, thus insuring the most conservative combinations of seismic loads are used in the stress evaluation.
The WESTDYN computer program is used to calculate the primary-plus-secondary and peak stress intensity ranges, fatigue reduction factors and cumulative usage factors for all possible load
- = = - -. = - -
...a...... N f N N--... - -.-... -
- --.~.- -.. - - -
l
.-_.~..-.-.
tcoa. t.ua E)
.2/o W v
$$ 11,t 2 T-fM ll set combinations..It is conservatively assuned that the transients can occur in any sequence, thus re:ulting in the most conservative and restrictive combinations of transients.
The combination of load sets yielding the highest alternating stress intensity range is detemined and the incremental usage factor calculated. Likewise, the next most severe combination is then detemined and the incremental usage factor calculated. This procedure 6
I' is. repeated util all combinations having allowable cylces <10 are
,"~~
formed. The total cumulative usage factor at a point is the sumation of the incremental usage factors.
~
[..
u l
1 l
l 1
l 1.
1 i
\\
l 1
1 l'Shll0
- w
.4
- E 9
---ee-e,..-e,
. _ _ _ = _.
y 1
s210.cty s
1 SB1&2 FSAR c.
Dynamic Analysis The cyclic stresses due to dynamic loads and deflections are com-bined with the stresses imposed by loads from component weights, hydraulic forces and thermal gradients for the determination of the total stresses on the CRDS.
d.
Control Rod Drive Mechanisms The control rod drive mechanism (CRDM) pressure housings are class I components designed to meet the stress requirements for normal operating conditions of Section III of the ASME Boiler and Pressure f
Both static and alternating stress intensities are Vessel Code.
considered. The stresses originating from the required design l
transients are included in the analysis.
i A dynamic seismic analysis is required on the CRDN's when a seismic disturbance has been postulated, to confirm the ability of the pressure housing to meet ASMg Code,Section III, allowable stresses
]ggy and to evaluate the effect of the seismic event on the drop time.
s
' Control Rod Drive Mechanism Operational Requiremeces Tha basic-operational rdquirements for the CIDM's are 5/8 inch step, h
1.
t
- 2...
147 inch travel,.
j 3..
360 pound maximum load,'
I Step in or out at 45 inches / minute (72 steps / minute),
ha i
5.
Electrical pouer interr,ption stall initiate release of drive -
rod assembly, 6..
Trip delay time of less than 150 milliseconds - free fall of drive rod assembly shall begin less than 150 milliseconds after power interruption, no matter what holding or stepping action is being executed with any load and coolant temperature of 100*F to 550*F, 7.
40 year design life with normal refurbishment.
- 3. 9(N).4. 3 Desian fonds. Stress Limits, and Allowable Deforestions Pressure Vessel a.
The pressure retaining components are analysed for loads corres-ponding to normal, upset,. emergency and faulted conditions.
The e
9 3.9(N)-59
-M
+_.e-we
f,,2/aw 1
(
INSERT T (Modification to 3.9(N).4.2) 1 The Control Rod Drive Mechanisms (CRDM's) nra evaluated for the effects of postulated reactor vessel inlet nozzle and outlet nozzle limited displacement breaks. A time history analysis of the CRDM's is performed for the vessel motion discussed in Section 3.9(N).1.5. A model of the CRDM's is formulated with gaps at the upper CRDM support modeled as nonlinear elements. The CRLd's are represented by beam elements with lumped masses. The translation and rotation of the vessel head is applied to this model. The resulting loads and stresses are compared to allowables to verify the adequacy of the system.
5 a
e I
l b
(2/dW l
c>
5
~
SB 1 & 2 rSAR t
(c) Coil Stack Assembly - Thermal Clearances The assembly clearances of the coil stack asstably over the latch housing were selected so that the assembly could be removed under all anticipated conditions of thermal expansion.
At 700F, the inside diameter of the coil stack is 7.308/7.298 inches, and.the outside diameter of the latch housing is 7.260/7.270 inchss.
Thermal exoansion of the mechanism, due to operating temperature of the CRDM, results in the minimum inside diameter of the coil stack being 7.310 inches at 2220F and the maximum latch housing outside diameter being 7.302 inches at 5320F.
Under the extreme tolerance conditions listed above, it is necessary to allow time for a 700F coil housing to heat during a-replacement operation.
To verify the acceptability of the above tolerances, four coil stack assemblies were removed from four hot control rod drive mechanisms, mounted on 11.035 inch centers on a 5500F test loop, allowed to cool, and then placed without incident.
(d) Coil Fit in Coil Housing CRDM and coll housing clearances are selected so that coil heat up results in a close to tight fit. This is done to. facilitate thermal transfer and coil cooling in
- pf
-(
a hot control rod drive mechanism.
l
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s 3.9(N).4.4 CRDS Performance Assurance Program Evaluation of Material's Adequacy a.
The ability of the pressure housing components to perform throughout the design lifetime, as defined in the equipenne specification, is confirmed by the stress analysis report required by-the ASME Code,Section III.
l Internal components subjected to wear will withstand a minimum of f
3,000,000 steps without refurbishment, as confirmed by life tests
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(Reference 12). Latch assembly inspection is recommended after 2.5 x 106 steps have been accumulated on a single control rod drive mechanism.
To confirm the mechanic'ai adequacy of the fuel assembly, the control rod drive mechanism and rod cluster control assembly, functional 1
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3.9(N).4.3.1 Evaluation of Control Rod Drive Machinsias and Supports The control rod drive mechanishes (CRDM's) and CRDM support structure are evaluated for the loading combinations outlined in Table 3.9(N)-2.
A detailed finite element model of the CRDMs and CRDM supports is constructed using the WECAN computer program with beam, pipe, and spring elements. For the LOCA analysis, nonlinearities in the structure are represented. These include RPI plate impact, tie rods, and lifting leg clavis/RPV head interface. The time history action-of the reactor vessel head, obtained from the RPV analysis is input to the dynamic model. Nazimum forces and moments in the CRDMs and support structure are then determined. For the seismic analysis, the structrual model is linearized and the floor response spectra corresponding to the CRDM tie rod elevation is applied to determine the maximum forces and mossents in the structure.
The bending asaants calculated for the CRDMs for tha various loading conditions are compared with mur h m allowable moments determined from a detailed finite element stress walutation of the CRDMs. Adequacy of the CRDM support structure is verified by comparing the calculated stresses to the criteria given in ASME III, Subsection NF.
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test programs have been conducted on a full-scale 12 foot control rod. The 12 foot prototype assembly was tested under simulated conditions of reactor temperature, pressure, and flow for approxi-steps and. 600 trips.ype mechanism secumulated about mately 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />.
The protot At the end of the test, the 3,000,000 control rod drive mechanism was still operating satisfactorily. A correlation was developed to predir,c the emplitude of flow-excited vibration of individual fuel rods and fuel assemblies.. Inspection of the drive line components did not reveal significant fretting.
These tests include verification that the trip time achieved by the CRDN meets the design requirement of 2.2 seconds from start of rod cluster control assembly motion to dashpot entry. This trip time requirement will be confirmed for each control rod drive mechanism prior to initial reactor operation and at periodic intervals after initial reactor operation, as required by the pro-posed Technical Specifications.
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,These tests have been reported in Reference (12).
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,_,_. '[here are no significant differences between the prototype contro rod drive mechanisms and the production units. Design materials,
~ tolerances and fabrication techniques are the same (see Section 4.5).
It. is expected that all control rod drive mechanisms will meet specified operating requirements for the duration of plant life, with normal refurbishment. Latch assembly inspection is s
recommended af ter 12.5 x 106 steps have been accumulated on a i
single CEDN.
If a rod cluster control assembly cannot be moved by its mechanism, adjustments in the boron concentration ensure that adequate shutdown margin would be achieved following a trip.
'Thus, inability to move one rod cluster control assembly can be t
f tolerated.. More than one inoperabic rod cluater control assembly could be tolerated, but would impose additional demands on the plant operator. Therefore, the immber, of inoperable rod cluster control assemblies has been limited to one as discuss 6d in the l
proposed Technical Specifications.
l In order to demonstrate proper operation of the control rod drive mechanism, and to ensure acceptable core power distributions i
< bring rod cluster control assembly partial-movement, checks are i
performed on the rod cluster control assemblies (refer to Technical Specifications). In addition, periodic drop tests of the rod cluster control assemblies are performed at each refueling shutdown, to demonstrate continued ability to meet trip time requirements.
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In addition, dynamic testing programs have been conducted by Westinghouse and Westinghouse Licensees to demonstrate that control rod scram time is not adversely affected by postulated seirmic events. Acceptable scram Performance is assured by also including the effects of the allowable displacements of the driveline components in the evaluation of the test results.
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gsse RAI 210.45 (3.9(N).2.5, Page 3.9(N)-41)
Your statement that the loading imposed by the SSE is generally small compared to blowdown loadings implies that in certain cases you have neglected loads due to an SSE.
If this is true, provide analysis details justifying your doing so.
RESPONSE
The loading imposed on the reactor internals by the Safe Shutdown Earthquake (SSE) for the Seabrook plant is small compared to the blowdown loading. As stated on FSAR Page 3.9(N)-43, "The stresses due to the Safe Shutdown Earth-quake (vertical and horizontal components) were combined in the most unfavor-able manner with the blowdown stresses in order to obtain the largest principal stress and deflection". The FSAR ui.11 be revised to clarify the subject statement.
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, the loeding imposed by the earthquake is4small compared to%he blowdown loading.
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a.
Vertical Excitation Model for Blowdown For the vertical excitation, the reactor internals are represdated by a multi-mass system connected with springs and dashpots simu-lating the elastic response and the viscous damping of the components.
Also incorporated in the multi-mass system is a representation of the motion of the fuel elements relative to the fuel assembly grids.
De fuel elements in the fuel assemblies are kept in position by friction forces originating from the preloaded fuel assembly grid fingers. Coulomb-type friction is assumed in the event that sliding between the rods and the grid fingers occurs.
In ordct to obtain
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an accurate simulation of the reactor internals response, the effects of internal damping, clearances between various internals, snubbing action caused by solid impact, Coulomb friction induced by fuel rod motion relative to the grids, and preloads in hold down springs have been incorporated in the analytical model. De modeling is conducted in such a way that uniform masses are lumped into easily identifiable discrete masses, while elastic elements are represented by springs.
De appropriate dynamic differential equations for the multi-mass I
model describing the aforementioned phencaena are formulated and the results obtained using a digital computer progras which computes the response of the multi-mass model when excited by a set of time-l dependent forcing functions. He appropriate forcing functions
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are applied simultaneously and independently to each of the masses in the system. De results from the program give the forces, displacements and deflections as functions of time for all the reactor internals components (lumped masses). Reactor internals response to both hot and cold leg pipe ruptures are analyzed.
b.
Transverse Excitation Model for Blowdown Various reactor intarnal components are subjected to transverse excitation during blowdown. Specifically, the barrel, guide tubes, and upper support columns are analyzed to determine their response to this excitation.
c.
Core Barrel For the hydraulic analysis of the pressure transients during hot leg blowdown, the maximum pressure drop across the barrel is a uniform radial compressive impulse.
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SB 1 & 2 FSAR yis RAI 210.46 (3. 9(N).4.3 Page 3. 9(N)-60) l The statement "The stress limits are established not only to assure that peak stresses will not reach unacceptable values, but also limit the amplitude of the oscillatory stress component in consideration of fatigue characteristics of the materials" needs clarification. What are these stress limits and from what source were they obtained?
RESPONSE
The stress limits associated with the design of the Control Rod Drive System are those defined in Section III of the ASME Code. The subject statement in the FSAR is only intended to reiterate the basic intent of the ASME Code.
For the Control Rod Drive Systes, ASME Code limits have been satisfied where required.
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Provide assurance that deformation limits are sufficient to guarantee control rod drive system integrity and functioning after a dynamic event such as an OBE.
RESPONSE
The control rod drive system (CRDS) integrity (deformation) during an OBE is assured by limiting the allowable stress levels for the pressure retaining components and reactor internals to those defined by Subsections NB and NG, respectively, of Section III of the ASME Code. Thus, after the OBE, the geometrical relationship between the various components of the CRDS is basically the same as the pre-OBE configuration. It is emphasized that both the stress and deformation limits are considered in the evaluation of the CRDS and reactor internals to ensure the integrity of the CRDS and insertability of the control rods.
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The statement "The stres: limits are established not only to assure that peak stresses will not reach unacceptable values, but also limit the amplitude of the oscillatory stress component in consideration of fatigue characteristics of the material" needs clarification. What are these stress limits and from what source were they obtained?
RESPONSE
The stress limits associated with design of the reactor internals are th 2f0.h defined in NG-3000 of Section III of the ASE Code. As identified in M
the above statement is intended to reiterate the basic intent of the ASE Code. Additionally, as identified in
(, the extent of compliance with the ASE Code will be included in an FSA change.
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Subsection NG, ASME Code Section III should be referenced as the design criteria for all design analyses, not just for the design basis accident.
RESPONSE
The design and fabrication of the Seabrook core support structures conform to the requirements of the Subsection NG of Section III of the ASME Code.
By contract, this plant preceeded the application of Subsection NG and, there-fore, these internals are not " Code Stamped" and no specific Code stress report is required The Seabrook plant reactor internals are identical in nature to the SNUPPS reactor internals which are stamped and documented to code requirements. The Seabrook FSARaund4=6e revised to reflect the above stated comparisen with ASME Code requirements.
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Pump overspeed Seismic loads (Operating Basis garthquake and Safe Shutdown Earth-o.
quake) p.
Blowdown forces (due to cold and hot leg break).
De main objective of the design analysis is to satisfy allowable stress limits, to assure an adequate design margin, and to establish deformation limits which are concerned primarily with the functioning of the components.
He stress limits are established not only to assure that peak stresses will not reach unacceptable values, but to also limit the amplitude of the oscil-1 story stress component in consideration of fatigue characteristics of the materials. Both low and high cycle fatigue stresses are considered when the allowable amplitude of oscillation is established. Dynanic analysis on the reactor internals is provided in Section 3.9(N).2.
As part of the evaluation of design loading conditions, extensive testing and inspections are performed from the initial selection of raw materials up to and including component installation and plant operation. Among these tests and inspections, are those performed during component fabrication, I
plant construction, startup and checkout, and during plant operation.
" 3.9(N).5.3 Design Ioading Categories
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he combination of design loadings fit into either the normal, ency or faulted conditions, as defined in the ASME Code,Section III ofV g
Loads and deflections imposed on components due to shoch and vibration are determined analytically and experimentally in both scaled models and operating reactors. De cyclic stresses due to these dynsmaic loads and deflections are combined with the stresses imposed by loads from component weights, hy-draulic forces and thermal gradients for the determination of the total stresses of the internals.
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he reactor internals are designed to withstand stresses originating from various operating conditions, as summarized in Table 3.9(N)-1.
De scope of the stress analysis problem is very large requiring many different techniques and methods, both static and dynamic. De analysis performed depends on the mode of operation under consideration.
l 3.9(N).5.4 Design Bases he design bases for the mechanical design of the reactor vessel internals
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INSERTAforResponsetoQuestionfl0.4 However, it should be noted that by contract the reactor internals for the Seabrook plant ;rcesed the applicability of subsection NG of the ASME Code.
Therefore, these iscarnals are not " Code Stamped" and no specific stress report is required. Nevertheless, these reactor internals are designed to meet the intent of subsection NG of the ASME Code.
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Verify that reactor internals are designed in accordance with Standard Review Plan 3.9.3 " Core Support Structures" or justify siternate design criteria.
RESPONSE
1.
Design and service loading conditions for core support structures are given in Subsection 3.9(N).5.2.
2.
The combination of design and service loadings fit into either the normal and upset, emergency or faulted conditions, as defined in Subsection NG of Section III of the ASE Code.
3.
The stress limits associated with the design of the core support structures are those defined in Subsection NG (NG-3000) of Section III of the ASE Code.
4 The deformation criteria for reactor internals components a*4 core support structures is established in regard to mechanical integritv.uch that adequate core cooling and core shutdown must be assured.
the deformation limits for reactor internals and core support structures are given in Table 3.9(N)-12.
5.
Dynamic system analysis of the reactor internals for faulted conditions is performed as disedssed in Subsection 3.9(N).2.5.
The above information demonstrates that the reactor internals design and analysis are consistent with Section 3.9.3 of the Standard Review Plan for Core Support Structures.
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SB 1 & 2 FSAR di M RAI 210.51 (3.9(N).5.4, Page 3.9(N)-71J What are the stresses associated with the maximum deflections in Table 3.9(N)-177 State the basis for these deflection limits. Justify the lack of safety margin for radial outward deflection.
RESPONSE
Stresses corresponding to the maximum deflections given in Table 3.9(N)-12 are not calculated because there deflections characterize certain limits (called no-loss-of-function) so that the operability of the reactor is not impeded (e.g., control rod insertability is assured). For example, the radial inward deflection of 8.2 inches describes the no-loss-of-function of upper barrel during hot leg break of 1,0CA.
This implies that during hot leg break, the radial inward deflection of the upper barrel be limited to 8.2 inches so as not to impair the operation of (deflect or bend) guide tube assemblies.
The 1.0 inch rzdial outward deflection of the upper barrel during cold leg break defines the limit in front of the unbroken nozzles so as not to impair the efficiency of the Emergency Core Cooling System.
For the upper package, the no-loss-of-function deflection 0.15 inch refers tc the vertical upward deformation of the guide tubes. This limit is set to preclude any axial compressive buckling loads on the guide tubes. Finally, 1.75 inch no-loss-of-function deflection defines the limit on the transverse displacement of the guide tubes so as not to impair the control rod insertion function.
Therefore, the basis for these. deflection limits is that an adequate core cooling and core shut down is assured. The fact that the calculated deflections I
are less than the allowable displacements, provides additional conservatisms on the design of reactor internals.
Consequently, the limitations established on the internals are concerned principally with the maximum allowable deflections in addition to a stress I
criteria to assure integrity of the components.
l Although the Seabrook reactor internals are not contractually required to l
meet the ASME Code as identified in 49, the code criteria for faulted I
conditions (Appendix F) is used as design basis for evaluating acceptability
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of calculated stresses, and the resulting stresses and deformations are below the established limits.
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RAI 210.52 (3.9(N).2.3. Table 3.9(N)-1)
Provide information on how the number of occurrences of steady state fluctu-ations were determined. How was the effect of transients listed in this table considered for BOP equipment?
RESPONSE
Page 3.9(N)-5 of the FSAR provides a description of the steady state fluctu-ations.
Initial Steadv State Fluctuations Initial steady state fluctuations result from conditions of small moderator and doppler coefficients and high rod worthe, Control rod cycling can occur causing temperature cycles of 1 30F.
The corresponding RCS pressure fluctu-ations are limited to 1 25 psi by pressurizer spray and backup heaters. Using maximum rod worth data from operating plants, the cycling rate was found to be about one step per minute, giving a continuous cycling period of about two minutes. Because of fuel burnup and an increasingly more negative moder-stor, coefficient, the cycling period increases and no cycling occurs beyond the first two fuel cycles (20 full power months). The total cumulative number of cycles is approximately 150,000.
Ra' dom Steadv State Fluctuations n
Random RCS temperature and pressure fluctuations were obtained from operating plant data. Maximum fluctuations of + 0.50F and + 6 psi were found to occur over a period of approximately six minutes. Assuming plant availability to 6 total fluctuations during be 85%, this would result in approximately 3x10 the 40 year design life of the plant.
The determination of the number of steady state fluctuations is therefore based on NSSS design and operating experience.
Steady state fluctuations (initial and random) have no effect on BOP equipment.
l However, the remainder of the RCS design transients listed in Table 3.9(N)-1 may impact BOP equipment. Determination of such impact, if any, falls within the scope of UE&C. Interface criteria are transmitted to the utility and/or UE&C in the form of internal Westinghouse design documents. This information may be used by the BOP designer on an as applicable basis.
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RAI 210. 53 ( 3.9.6.2)
There are several safety systems connected to the reactor coolant pressure boundary that have design pressure below the rated reactor coolant system (RCS) pressure. There are also some systems which are rated at full reactor pressure on the discharge side of pumps but have pump suction below RCS pres-In order to protect these systems from RCS pressure, two or more iso-sure.
lation valves are placed in series to fons the interface between the high pressure RCS and the low pressure systems. The leak tight integrity of these valves must be ensured by periodic leak testing to prevent exceeding the design pressure of the low pressure systems thus causing an inter-system LOCA.
Pressure isolation valves are required to be Category A or A0 per IWV-2000 and to meet the appropriate requirements of IWV-3420 of Section XI of the ASME Code except as discussed below.
Limiting Conditions for Operation (LCO) are required to be added to the tech-nical specifications which will require correction action; i.e.,
shutdown or system isolation when the final approved leakage limits are not met. Also j
surveillance requirements, which will state the acceptable leak rate testing f requency, shall be provided in the technical specifications.
1 Periodic leak testing of each pressure isolation valve is required to be performed at least once per each refueling outage, af ter valve maintenance prict to return to service and for systems rated at less than 50% of RCS design pressure each time the valve has moved from its fully closed position
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unless justification is given. The testing interval should average to be approximately one year. Leak testing should also be perf ormed af ter all disturbances to the valves are complete, prior to reaching power operation following a refueling outage, maintence, etc.
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The staff's present position on leak rate limiting conditions for operation must be equal to or less than 1 gallon per minute for each valve -(GPM) to ensure the integrity of the valve. Demonstrate the adequacy of the redundant i
l pressure isolation function and give an indication of valve degradation over l
a finite period of time. Significant increases over this limiting valve l
would be an indication of valve degradation from one test to another, l
I Leak rates higher than 1 GPM will be considered if the leak rate changes are below 1 GPM above the previous test leak rate or system design precludes me suring 1 GPM with sufficient accuracy. These items will be reviewed on a c.se by case basis.
The Class 1 to class 2 boundary will be considered the isolation point which j
must be protected by redundant isolation valves.
In cases where pressure isolation is provided by two valves, both will be independently leak tested. When three or more valves provide isolation, l
only two of. the valves need to be leak tested.
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SB 1 & 2 FSAR Provide a list of all pressure isolat' ion valves included in your testing program along with four sets of Piping and Instrument Diagrams which describe your reactor coolant system pressure isolation valves. Also discuss in detail how your leak testing program will conform to the above staff position.
RESPONSE
A valve test program is in preparation and, when completed, will address inservice testing of valves whose function is to perform pressure isolation between high pressure reactor coclant and low pressure systems. Specific inf ormation regarding valve testing criteria, frequency, and exceptions will be made available to the NRC Project Manager for review by auary 1,1983.
This date is consistent with the previously stated submittal date for information relative to the Inservice Inspectica Program.
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RAI 210.54 It is the staff's position that all essential safety-related instrumentation lines should be included in the vibration monitoring program during pre-operational or start-up testing. We require that either a visual or instrumented inspection (as appropriate) bt conducted to identify any excessive vibration that will result in fatigue failure.
Provide a list of all safety-related small bore piping and instrumentation lines that will be included in the initial test vibration monitoring program.
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RESPONSE
Subsection 3.9(B).2, Dynamic Testing and Analysis, is presently undergoing an extensive review to define scope and programatic requirements. An update of this section, in conjunction with applicable portions of Chapter 14, will be provided to the NRC by October, 1982.
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SB 1 & 2 FSAR RAI 210.55 Due to a long history of problems dealing with inoperable and incorrectly installed snubbers, and due to the potential safety significance of failed snubbers in safety related systems and components, it is requested that main-tenance records for snubbers be documented as follews:
Pre-service Examination A pre-service examination should be made on all snubbers listed in Tables 3.7-4a and 3 4b of Standard Technical Specifications 3/4.7.9.
This exam-ination should be made af ter snubber installation but not more than six months prior to initial system pre-operational testing, and should as a minimum verify the following:
1.
There are no visible signs of damage or impaired operability as a result of storage, handling, or installation.
2.
The snubber location, orientation, position setting, and configuration (attachments, extensions, etc.) are according to design drawings and specifications.
3.
Snubbers are not seized, frozen or jammed.
4.
Adequate swing clearance is provided to allow snubber movement.
5.
If applicable, fluid is to be recommended level and is not leaking from the snubber system.
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Structural connections such as pins, fasteners and other connecting hardware such as lock nuts, tabs, wire, cotter pins are installed cor-i rectly.
If the period between the initial pre-service examination and initial system-l pre-operational test exceeds six months due to unexpected situations, re-examination of items 1, 4, and 5 shall be performed. Snubbers which are installed incorrectly or otherwise f ail to meet the above requirements must be repaired or replaced and re-examined in accordance with the above criteria.
I Pre-Operational Testing During pre-operational testing, snubber thermal movements for systems whose operating temperature exceeds 250 F should ve verified as follows:
l During initial system heatup and cooldown, at specified temperature a.
intervals for any system which attains opperating temperature, verify the snubber expected thermal movement.
b.
For those systems which do not attain operating temperature, verify via observation and/or calculation that the snubb.er will accommodate the projected thermal movement.
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Verif y the snubber swing clearance at specified heatup and cooldown intervals. Any discrepancies or inconsistencies shall be evaluated for cause and corrected prict to proceeding to the ner:. specified interval.
The above described operability program for snubbers should be included and documented by the pre-service inspection and pre-operational test programs.
The pre-service inspection must be a prerequisite f or the pre-operational testing of snubber thermal motion. This test program should be specified in Chapter 1* of.the FSAR.
RESPONSE
The response to this RAI was included in our November 27,19, submittal in response to the Acceptance Review RAls; and was subsequently incorporated into FSAR Amendment 44 (Subsection 3.9(B).3.4(d)).
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SB 1 & 2 FSAR RA? 210.56 (3.2.1. Table 3.2-2, Sheet 3 of 31)
Explain the use of ASME VIII for the design of the pressurizer discharge piping.
RESPONSE
W3S Refer to revised Sheet 3 of Table 3.2-2, which 6 provided in Amend-ment 45.
This was a typographical error.
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SB 1 & 2 FSAR RAI 210.57 (3.6(B).2.1, Page 3.6(B)-7)
Explain in more detail how intermediate break points are determined in Class 1 BOP piping when stresses and usage factors are below 2.4 Se on 0.1, respectively.
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RESPONSE
BOP The criteria of Regulatory cuide 1.46 were followed for the intermediate l
- d ggS f break point determinations. FSAR Paragraph 3.6(B).2.la ir S:' : revised in i
Amendment 45.
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For Class 1 analysis outside the RCS, Westinghouse provides stress and usage pd factors to UE&C. UE&C then takes the analysis results and determines break N,Jp locations.
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l 210.58 It is the staff's position that if a structure separates a high energy line from an essential component, that separating structure should be designed to withstand the consequences of the pipe break in the high-energy line which prodoces the greatest effect at the structure irrespective of the fact that the pipe break criteria aight not require such a break location to be postulated. Provide assurance that the Seabrook plant meets the abovs requirement.
RESPONSE
In evaluating the offacts of high energy line breaks on essential components, it was found that protection is most of ten provided by separation distance due to the high energy lines being located far enough away from essential components that the danger of impact or jet ispingement did not exist. In the case of the electrical trays in the control building, which were separated by the building wall from the adjacent main steam and feedwater lines, a guard pipe was provided to prevent jet impingement from the main steam line, and an energy-absorbing bumper was provided to prevent impact f rom main steam or feedvater lines that could damage the wall.
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SB 1 & 2 FSAR RAI 210.59 (Appendix 3C, Page 4)
For circumferential breaks that are axially restrained, provide justification that the axial and lateral movement of the pipe will result in the fan jet impingement that is assumed The staff's position is that a fan jet can occur when the broken piping is physically eeetrained from significant separation (axial pipe movement equal to or less than 1/2 pipe diameter and lateral pipe movement less than pipe wall thickness).
RESPONSE
Seabrook BOP piping is not axially restrained so as to restrict pipe motion in such a fashion as to form a fan jet. Fan jet calculations were not used in the analysis.
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RAI 210.60 What is the maximum allowable tip deflection of a restrained whipping pipe?
Provide assurance that the deflections of the restrained whipping pipe will not affect the function of any safety-re'.ated components.
RESPONSE
BOP For BOP piping, pipe whip restraints are provided to maintain the motion of the ruptured pipe end within controlled limits. The limit of motion is the area within which no essential component can be affected by impact or jet impingement.
f}qyd NSSS Westinghouse designs pipe whip restraints for the RCS to eliminate eny post-fb46[#
ulated pipe whip concerns. Based on the Westinghouse restraint design, the 9g ffh 1 maximum allowable tip deflection of a restrained whipping pipe is generally limited to the thickness of the pipe., This criterion precludes the whipping p4 pipe affecting the function of any safety related component.
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SB 1 & 2 FSAR RAI 210.61 Provide a description on examples of the different types of pipe whip restraints and jet impingement barriers that are used in the plant.
RESPONSE
BOP For' BOP piping, pipe whip restraints are either bumpers (with or without energy-absorbing crushable pads), guides, U-bolt restraints, or a combination of these. Guides are structural steel frames surrounding the pipe. U-bolts A
are only used in tension. Jet impingement barriers consist of sleeves or lp,ggl guard pipes.
I f0 NSSE The NSSS supplier only provides pipe whip restraints for the RCS. These y
restraints are described in FSAR Chapter $.
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De L SB 1 & 2 FSAR Obg RAI 210.62 (3.6(N).2.3. Page 3.6(N)-9)
When calculating the dynamic effects of jet impingement, what values are assumed for Ko for the various initial fluid conditions?
RESPONSE
When performing jet impingement saalysis for the RCS, Westinghouse uses a Ko value of 1.3 for hot leg breaks and a K, value of 1.57 for cold leg breaks.
The:,e values are based on ANS standard 58.2.
Tne FSAR -.'.'. Lv revised to incorporate a reference to ANS 58.2.
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F 531&2 F3AR The shape of the target affects the amount of somente change in the jet, and thus affects the impingement force on the target.
The target shape factor is used to account for target shape which j
do not denect the now 90 degrees away from the jet axis.
The method used to compute the jet impingement load on a target is one of the following:
1.
The dynamic efface of jet impingement on the target structure is evaluated by applying a step load whose magnitude is given by Fj = E,F,AmB18 abere Fj; jet ispir.gement imd on target
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fraction of jet intercepted by target E.
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=J target shape factor Discharge now arena for limited flow area circumferential l
breaks are obtained from reactor coolant loop analyses j
performed to determine the asial and lateral displacements of the broken ends as a function of time. Aug is the
==wi== break now area occurring during the transient, and
.L'a calculated'as the total surface area through which the fluid most-pass to emerge from the broken pipe. Using geometrical formulations, this surface area is determined to be.e function of the pipe separation (axial and transverse) and. the dimensions of the pipe (inside and outside dia==cer).
If a. simplified scacic analysis is performed instead of a dynamic analysis, the above jet load (Fj) is multiplied by a dynamic load f~ actor. For an equivalent static analysis of the target structure, the jet impingement force is multiplied by a dynamic load factor of 2.0.
This factor assumes the target can be represented as essencially a one degree of freedom. system and the impinsement force is conservatively applied as a step load.
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m SB 1 & 2 FSAR RAI 210.63 (Table 3.6(N)-2)
Why are only seven break locations listed? There are eleven design breaks postulated. Provide the CliF and moments for the other four design break locations. In addition, p: ovide any other modes where the CUF exceeds 0.1.
Provide the CUF values spec'ific for Seabrook at each design break location and at any other locations where the CUF exceeds 0.1.
RESPONSE
Footnote a in Table 3.6(n)-2 addresses the other four pipe break locations postulated in the RCS..The fatigue analysis for the RCS is performed on a generic basis. The cuum:ulative usage factors do not change from plant to plant. For the Seabrook plant, a specific moment analysis is performed to verify that the specific plant moments are enveloped by the generic analysis performed by Westinghouse and described in WCAP-8082.
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It is the staff's position that the use of the overlap technique in piping subsystem analysis is not generally acceptable. Provide a discussion on the use of the overlap technique with respect to NUREG/CR-1980. Identify those piping subsystems that use the overlap technique.
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RESPONSE
The overlap method is used primarily for those cases in which branch pipias 4
cannot be decoupled from the main run piping. When this situation arises, a sufficient subsection of the branch piping is included in the main run model to ensure inclusion of the branch effects. Conversely, subsequent branch analyses include adequate sections of the run piping.
The criterion employed requires that each decoupled analytical model includes a region of overlap c'ommon to both subsystems which contain the necessary supports to minimize translations as well as bending moments transmitted across overlapping boundaries. In practice, the decoupling is accomplished by extending the model to include supports which provide restraining actions in at least two x-x, two y y and one z-z direction (s) (x-x, y y, and :-z are the local piping coordinates with z-z being parallel to the piping centerline).
Stresses and support loads for the run are those determined for the model elements representing it.
Branch pipe stresses are determined in a similar However, in either case, whether the run or the branch pipe is being manner.
analyzed, the pipe stresses in the overlapped or extension regions are checked against the appropriate allowables.
In those rare cases where the overlapping technique is used for analyzing large main run piping systems, the criterion discussed above for selecting supports and restraints in the overlap region which effectively decouple the subsystems is used. Here, results from overlapped subsystem models ere treated in a conservative manner. Pipe stresses in the overlapped region are deemed to be the maximum found in either of the subsystem models, whereas total reaction loads for supports in these regions are computed by taiting the i
absolute sum (for seismic loads) of the loads determined for each subsystem.
I Direct comparisons of the preceding techniques with the test cases presented in NUREC/CR-1980 are difficult because of the rather simplified nature of l
those test cases. Nevertheless, UE&C's methods are considered adequate.
The specific boundary selection criteria are accompanied by other conservatisms.
These conservatisms are inherent in the applied response spectra. In fact, response spectra envelopes are applied when applicable. In addition, UE&C's support envelopes are applied when applicable. In addition, UE&C's support design philosophy dictates adequately stiff supports relative to the pipe sizes for which they are designed. These measures, in conjunction with the above modeling techniques, ars considered to provide acceptable analytical methods and designs. Substantiating test cases comparing coupled and decoupled models using the techniques presented herein are being performed.
I The piping systems which hav: been analyzed using the preceding mechods are t
listed below l
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CS - Chemical and Volume Control.
CC - Primary Component Cooling Water CBS - Containment Spray 3
RH - Residual Heat Femoval 4
SW - Service Water I
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The FSAR states that the load combination's and stress limits of Table 3.9(B)-7 provide assurance that essential Class 2 and 3 piping systems will retain their functional integrity. The staf f does not agree that the use of 1.8 Sh stress limits will assure the piping functionality. We have provided the applicant with our acceptance criteria for functional capability (Reference NRC evaluation of General Electric Topical Report NEDO-21985 dated September 1978). Provide assurance that the Seabrook plant meets our acceptance criteria for piping functional capability.
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RESPONSE
1.
The allowable stress 1 Lait of 1.8 Sh for essential Class 2 and 3 piping systems is lower, in magnitude, than the functional capability limit of 1.5 S at temperature below 4000F for piping materials used.
y Therefore, functional capability is assured below 4000F.
l 2.
At temperatures above 4000F, the stress limit (for Service Level C for austenitic steels. The only) of 1.8 Sh is greater than 1.5 Sy actual percentage difference between the allowable stress limit (1.8 S )
h and the functional capability limit (1.5 Sy) is small (6-8% depending upon material). NUREG/CR-0261, which is the basis for the NRC position on functional capability, states: "The C-Limits of 1.5 S involve an engineering judgment in which we take advantage of......y
..... It is our judgment that, under C-Limits, straight pipe'will not be subject to excessive strains nor will it lose a significant part of its flow area."(1) Since the criteria used for piping analysis have significant i
built-in conservatisas (e.g., magnitude of loads, method of determining
" worst" stress condition, and the functional capability limit was l
chosen based to a large extent on judgment, we believe that the esall difference between limits does not impair functional capability in
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this situation.
In conclusion, we belive that the allowable limit of 1.8 Sh assures functional capability.
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(1)
Evaluetion of the Plastic Characteristics of Piping Products in Relation to ASKE Criteria, E. C. Rodabough, S. E. Moore. July 1978 NUREG/CR-0261, Page 59.
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i SB 1 & 2 FSAR RAI 210.66 Provide a more detailed description of the loads, load combinations, and stress Ibnits that are used in the design of ASME Code Class 1, 2 and 3 com-ponents and component supports.
RESPONSE :
BOP The plant loading conditions and load combinations are summarized in Table 3.9(B)-2 for ASME components and their supports. Stress ihmits applicable to the various components (excluding supports) for each loading condition are summarized in Tables 3.9(B)-3 through 3.9(B)-5.
Note (1) on each of these tables instructs the reader to refer to Table 3.9(B)-2 for definitions of the plant loading conditions.
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Supports are addressed in FSAR Subsection 3.9(B).1.4(a). All component support designs, except those listed in the response to RAI 210.29, satisfy the design requirements and stress allowables of ASME III, Subsection NF.
Those listed in the RAI 210.29 response are designed according to AISC criteria.
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For loading combinations and stress limits for Westinghouse ASME Code Class p)tV 1, 2, & 3 components and component supports, refer to tables in FSAR Section
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FSAR Subsection 3.9(N).3.4 will be revised to state compliance with the ASME y (h.
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Code Subsection NF for Class 2 and 3 components supports procured after 4
July 1, 1974.
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fp0W ArrACEMENT 76 / rEh1 44 i
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velve emereising and inspection to ass =" e the functional lj r
ability.of, the valva.
I The pressuriser safety valves are qualified by the following
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se 1) stress and deforestion analyses of critical items which any affact operability for faulted aa=M ion t
lasda, 2) La shop bydrostacia and seat leakage tests, and 3) periodic in-situ velve inspection. In addition to these tears,
- s. static land 'equivalast to that applied by the faulted condition la applied at the top of the bommet and the pressure is increased until the valve==eh==f== actuates. Soccessful actuation withis the design requirements of the valve assures,its overpree-sarisation safety sepabilities during a seismia. event.
Using these~asthods, active. valves are qualified' for operability enrias a faulted eyesc. These asthoda est11and above conser-
. votively simmiate the seismic event and assure that the active-
'.valver will perform. their safety-related fuestion when necessary.
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Pump
- Meter and valve operator Qualification
. Activa pump motore (laeloding vital pump apportenances) and active
- 7. valve motor operators are seismically qualified. La accordmaca with
.G.IEEE' standard 344-1977 "The seismic q=mf Mcation program for 7 this' electrical ' quipment. is further~ esocribed in Section. 3.10(N) e
}, and the Equipment qualification Data Packages referenced thereia.
3.9(N).3.3 Neustier of Pressere Ballsf Devices ~
EnfertoSubsestion3.9kB).3b.,
3.9(N).3.4
- t --.s (ASMg Code Class 2 and 3)
See Subsection 3.9(N).1'for,ASur' Cede Class 1 eenponese supporta.
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t-Supports for P __ - - es Procured 'After July 1.1974 Class 2.and 3 Pe supports are designed and analysed 'for' Design, Borasi, Upest, Emergency, and Famited eondiciosa to the rules med.r'equirementa of Subeestion NF of Section III of the ASME B&FF Code.(1974. Edition). The desigs analy?es or test methods and associated stress or load allowablo limits that are used in the evaluation of linear supports for Faulted conditions are those i
.. defleed in Appendix. F' of the ASME Code.. 7-_ 1,
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Class 2 and 3 supports are designed as follows:
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--- t Smooorts (a) Norest - The allouable stresses or load ratings of MB5-SP-54.are.used.
(b) Upset - por goet somditions, the allouable ' stresses or Load ratings are 20 percent higher 25am those specified for normal aseditions
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(c) Emergency - For emergency conditions, the allouable stresses or load ratings are 80 percent higher thaa those specified for morosi conditiosa. Sepports (red hangers and struts) are checked for elastic stability
, when-applicable, ande===h compressive load does not onceed critical buckling load as specified by the
~ '- '.spplicable codes sad. desiga se==dards.
(d) Faulted - The alloushie stresses or lead retings of MES-SP-58 are based on a factor of safety which is grestar than or equal to four, i.e., the allowbie stress is less than. or equal to ese-fourth the slaiman tessile striss of the material. The allouable stresses"for faulted conditions are thus less than or equal to 0.6 times, the minimum tensile stress of the asterial, i.e.,
2A times one-fourth the =Ich tensile stress of the
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asterial equals 0.6 times the minimum tessile stress.
This. low allowable stress (associated factor of safety equals 1.67) is. adequate to assure that active componentsare properly supported for faulted conditions.
f 2.
Linear Type Sueoorts (a) Normel. - The allouable stresses of A13C-69 part 1. are employed for aerust condition limits.
(b) Upset - Stress limits for upset canditions are 33 percent higher than those specified for normal conditions. This is consistest with paragraph 1.5.6 of AISC-69 part 1 tdsich permits one-third increase in allowable stresses for wind or seismic loads 1
(c) amersency - not applicable.
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SB 1 & 2 FSAR RAI 210.67 Provide the stress limits used for NF bolts.
RESPONSE
BOP 1.
SA-325 3olts:
1/2" to 1"9 bolts - Sy = 92 kai Su = 120 kai a.
Friction Type Joint Allowable slip resistance per ASME XVII - 2461.4 Ps = unTiKs Ps = 21 x As kips / bolt where:
Slip coeff. Ks = 0.25 Initici clamping force Ti = 0.7 x 120 x As = 84 As kips As = tensile stress area of the bolt, sq. in.
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Number of shear planes per bolt = m = 1 g
'Y Number of bolts = n = 1 l
b.
Bearing type joint:
1 Allowable tensile stress = Sta
= 60 kai 2
i Allowable shear stress
= 0.62Su = 24.8 kii l
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SA-307 Bolts 1/2" 9 to 1"$ Bolts - Sy = 36 kai Su = 58 kai Allowable tensile stress = Su
= 29 kai 2
Allowable shear stress
= 0.62Su = 12 ksi l
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SB 1 & 2 FSAR NSSS In the design of bolting for ASME Code Class 1 component supports, Westing-house uses the design criteria in Subsection NF, paragraph 3280. Stress limits a
for normal and upset conditions are thora defined in Appendix XVII, paragraph 2460 of the Code, and Code Case 1644.
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For emergency and faulted conditions, Westinghouse limits stresses te 0.9 Sy ipt [#
at temperature. For bolts supplied by Westinghouse, the ultimate strength is approximately 4600= ksi and the yield strength is 130 kai at temperature.
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SB1&2 FSAR RAI 210.68 Describe to what exterit high-strength NF bolts are used.
RESPONSE
SOP Of these high
, g*
Approximately 5% of the NT supports use high strength bolts.
g strength bolted component supports, approximately 90% of the supports use friction type joints.
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Nsss Westinghouse only provides bolts for Class I component supports. Additional information concerning the properties of these bolts are discussed in the 7
response to RAI 210.67.
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h' 38 1 & 2 FSAR RAI 210.69 The Seabrook plant incorpo;ates the Westinghouse Model F ateam generator.
Ws vill requir2 that the results of the analysis to determine the tube plug-ging criteria be presented to the staff when they become available.
hEMPONSE Westinghouse will provide a report prior to commercial operation describing the tube plugging criteria which will be employed for Model F steam generators.
Additionally, Westinghouse will review the current position in the Seabrook FSAR on Regulatory Guide 1.121 to confirm its applicability to the Seabrook plant.
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