ML20199G980
| ML20199G980 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 01/15/1998 |
| From: | Hammer M NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20199G986 | List: |
| References | |
| REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR GL-87-02, GL-87-2, TAC-M69460, NUDOCS 9802040387 | |
| Download: ML20199G980 (18) | |
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4 Northern States Power Company s ** Monticello Nuclear Generating Plant
/ 2807 West County Road 75 Monticello. MN $5362 January 15,1998 U S Nuclear Regulatory Commission GL 87-02 Attn: Document Control Des' USI A-46 Washington, DC 20555 MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 Response to Request for Additional Information on the Resolution of Unresolved Safety issue A-46 (TAC NO. M69460)
By letters dated October 8,1997 and November 12,1997, the NRC requested additional information on Resolution of Unresolved Safety issue (USI) A-46, Seismic Evaluation Report (TAC NO. M69460). This letter provides Monticello's response to the NRC's request for additionalinformation.
In our December 17,15:i7 letter to the NRC on this same subject, NSP requested additional time to respond to questio .s 6,7, and 8 after the industry position perta'ning to these questions had been determined. Based on a m.nuary 12,1998 conversation with the Monticello NRC Project Manager, Mr. T. J. Kim, Monticello will submit responses to questions 7 and 8 by February 15,1998. Response to question 6 is enclosed.
This submittal contains no new NRC commitments, nor does it modify any prior commitments.
Please contact Sam Shirey, Sr Licensing Engineer, at (612) 295-1449 if you require additional information related to this request.
$ML4& I Michael F. Hammer j Plant Manager Monticello Nuclear Generating Plant c: Regional Administrator-Ill, NRC NRR Project Manager, NRC j Resident inspector, NRC State of Minnesota Attn: Kris Sanda .,sa J Silberg I , 6 b U *g Attachments: Affidavit to the US Nuclear Regulatory Commission.
Exhibit A: Response to NRC Request for Additional Information mwe se mcenstwouovunse ooc
'l((lfll 9802040387 980115 '
PDR ADOCK 05000263 u
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0 UNITED STATES NUCLEAR REGULATORY COMMISSION 1
NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 Response to Request for Additional Information on the Resolution of Unresolved Safety Issue A-46 (TAC NO. M69460) i Northern States Power Company, a Minnesota corporation, hereby provides the information requested by the USNRC in letters dated October 8,1997 and November 12,1997, titled
" Request for Additional Information on the Resolution of Unresolved Safety issue A-46, Seirt.,ic Evaluation Report. (TAC No. M69460)"
This letter contains no restricted or other defense information.
NORTHERN STATES POWER COMPANY By OA4'l4MhD Michael F. Hammer Plant Manager Monticello Nuclear Generating Plant 8
- On this 16 day ofTawuq (MB before me a notary public in and for said County, personally appeared Michael F. Hammer, Plant Manager, Monticello Nuclear Generating Plant, and being first duly sworn acknowledged that he is authorized to execute this document on behalf of Northem States Power Company, that he knows the contents thereof, and that to the best of his knowledge, information, and belief the statements made in it are true and that it is not interposed for delay.
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I RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION L
' PERTAINING TO USl A-46, SEISMIC EVALUATION REPORT
- MONTICELLO NUCLEAR GENERATING STATION-
' January 15,1998:
Exhibit A t
The following questions are from a letter oent by the NRC to Northern States Power Co., ;
"MONTICELLO NUCLEAR GENERATING PLANT- REQUEST FOR ADDITIONAL :
INFORMATION RELATED TO UNRESOLVED SAFETY ISSUE (USI) A-46, SEISMIC EVALUATION REPORT (TAC NO. M69460)" dMed October 8,1997.-
L : In Reference A, the response to Question 2 statixi that damping independent power spectra density (PSD) curves were generated from the available floor response spectra (FRS) to produce response spectrum curves for otner damping values. Provide the algorithm used in
- generat!ng the " damping independent PSD" from the original 0.5% of critical damping FRS and the threshold criteria used for the error function. Also provide key examples of the newly generated spectra and their corresponding PSD functions.
Response
" The algorithm and criteria used in generating damping independent PSD from the original 0.5%
floor response spectra (FRS) are provided in Attachment 1. Figures 1 to 4 provide key examples of generated spectra.
- 2. Question'10 in Reference B specifically mquested the rationale (supported by engi leering analysis) for concluding that no large rel: se motion between the tank and the pump house will take place during a safe shutdown euthquake. Provide an applicable engineering calculation supporting the conclusion stated in Reference A las needed, mfer to Standard Review Plan Section 3.7.3). ,
Response
The original evaluation was based on the engineeringjudgment that the relative seismic
' displacement between the tank and the Pump house would be acceptably low. An engineering 4
' calculation has subsequently been created which supports this original judgment.- This -
- . engineering calculadon has been provided in Attachment 2.
The engineering calculation evaluated the relative displacement between tank T-44 and the Diesel Fuel Pump House. Seismic displacements were determined for these phenomena
- o Seismic soil waves 1
r Response to NRC Request for Additional Information - USI A-46 January 15,1998 Exhibit A
- The respoase of tank T-44
- The response of the Pump 11ouse Both T-44 and the Pump House are stiff structures founded on soil. Foundation flexibility was inodeled by the use of soil springs. The component displacements were combined in an squve root of the sum of the squares (SRSS) manner to arrive at a relative displacement of less than 0.03" for the safe shetdown earthquake. Piping between T-44 and the Pump House was evaluated for an enforced dieplacement of 0.03". The resulting stresses in the piping were found to be low,less than 300 psi in bending and 35 psiin torsion. The pipe material yield stress is 30,000 psi. These results confirm the original engineering judgment that relative seismic displacement between the tank and the Pump house would be acceptably low.
- 3. In Reference A, the msponse to Question 11 stated that "these amas were all rooms within a much larger building floor, it is estimated that these areas represent less than 2% of the area under review...." Please indicate how many rooms in the inaccessible areas consisted of the 2%
of the ama under review. Also, provide the basis for how the "CEVUE" pictures were used for all of the inaccessible rooms to determine that the cable trays and conduits in these rooms are all well supported and show no visible degradation.
Runens The following areas are the areas that were inaccessible due to radiological concerns during the cable tray and conduit walkdown.
Radwaste Tank Room, RBA18 figure F-6 Steam Jet Air Ejector Room, TBA5 figum F-9 Radwaste Building 947' floor, RW-A3 Tank room & Hopper room Radwaste Building 962' floor, RW-A4 Centrifuge & Pump room TIP Drive room area, RBA9 figure F-6 The Radwaste Tank Room, the Tank room & Hopper room on 947' of the Radwaste Building, the Centrifuge and Pump room on 962' of the Radwaste building, and the Steam Jet Air Ejector Room in the Turbine Building wem determined from conduit and tray routing drawings to contain only tray or conduit that served equipment in the area.
There is not any equipment on the Safe Shutdown List in these areas. These areas were judged acceptable based on the walkdown experience of the acceptability of the conduit investigated in the other loc . and because the tray in these areas does not contain any cable required for SSEi equipment.
A walkdown of the cable tray and conduit was not done due to ra 'iotascal exposum concerns in the TIP Drive Room. The TIP Drive Room was the only area that utilized CEVUE for the evaluation that the conduit and tray v'ere acceptable. CEVUE is a PC 2
, Response to NRC Request for AdditionalInformation - USI A 46 January 15,1998 Exhibit A based video system that allows a user to " walk" through the plant. The system allows users to view areas of the plant that have radiation concems thus maintaining the ALARA concept. The CEVUE pictures show the cable tray and conduit in the TIP Drive room such that an estimate of the support spacing could be made and a determination if the support type should be included in the Limited Analytical Review.
There were no signs of degradation found during the inspection of the Reactor Building conduits and cable trays, thus it was concluded that Tip Drive Room (located in the Reactor Building) which is a clean dry area would also b we no signs of degradation.
- 4. NSP letter to the NRC dated April 29,1997, which povided an update on the status of NSP's response to NRC Generic Letter 87-02, Supplement 1, indicated that the resolution of outliert listed in Table 8-1 of Report 91C2687.A46,"USNRC l>S' A-46 Resolution, Seismic Evaluation Report, Monticello Nuclear Generating Plant" was completed. Provide a summary of the specific analysis or resolution methods adopted for the outlier resolution.
Also in reference to the response to Question 14 in reference A regarding the unresolved outliers in Table 8-1, explain the safety implications for not resolving these outliers, in accordance with Item 17 in Section 9.1 of the Generic Implementation Procedure (GIP-2).
Elaborate on the NSP's decision to defer the resolution of identified outliers and the evaluation in support of the conclusion that the licensing basis for the plant will not be affected by the decision.
Ecnome The following is a summary of the resolution of the outliers that were listed in Table 8-1 of Report 91C2687.A46, "USNRC USI A-46 Resolution. Seismic Evaluation Report, Monticello Nuclear Generating Plant" All Control Room Overhead lights were supported so as to prevent a seismic items interaction.
C-129B The steel angle support member was upgraded to remove the concern of the weld quality of the joint. The vertical duct support was upgraded to preclude a potential seismic interaction.
C-19, C-32, C-41 The duct was modified such that it is not in direct contact with the panels and any motion of the duct will not cause the duct to impact the panels.
C-56 A conduk rupport was added to prevent any potential swinging that could pull on the cables
-C-253B The cart has been removed DIO The missing anchor has been installed.
D312 The cart has been removed D313 The panel has been top braced and the anchorage upgraded.
3 w
, Response to NRC Request for AdditionalInformation - USI A-46 January 15,1998 Exhibit A D3A,D3B Thimissing spacers have been replaced and the end batten has been tightened against the cell.
E-200A, E-2008 An evaluation was done that found that the capacity of the cinch anchors are adequate for the loads.
K 10A The tape rack and cabinet have been removed P 109A, P-109B The pumps are in the Intake Structure. Per Monticello's P-109C, P-109D commitment, the original GIP evaluation was performed P-111 A, P-111 B using Reactor Building floor spectra. For outlier resolution,
. floor spectra were calculated for the Intake Structure.
Because the Intake Structure is squat and stiff,in structure demand for the pumps was found to be lower than that used in the original GIP evaluation. The pumps were evaluated with the new demand and found to be acceptable.
T-79A,B,C,D,E,F A tightness check was performed on all of the clamps to T-80A,B,C,D,E,F verify they had adequate torque.
RV-1746, An additional support was added to the duct to assure it V-AC-4 would not become a seismic interaction hazani Y10,Y20,Y30 The existing anchorage was evaluated and determined to be adequate.
~Y01,Y22 The transformers were inspected and the coils were found to be restrained.
Y72,Y82 The coil hold down bolts were tightened to assure that there the coils could not impact the side of the transformer.
DPII-Al, RB A17 The conduit or raceway supports were modified such that RB- A19, TB-A7 they would meet the GIP criteria.
CB-A2 The raceway was modified such that they would meet the GIP criteria.
The GIP methodology involves an initial screening process, if safe shutdown equipment should fail to pass the initial screening process, i.e. it is classified as an outlier, more detailed methods for verifying its seismic adequacy may be used as described in Section 5 of Part II of the GIP. The outliers listed in Table 8-1 did not pass the initial screening process. However, the GIP outlier resolution process which included analysis, modifications, and replacement was implemented such that all of these outliers were verified as seismically adequate.
The safety implications for not resolving the identified outliers would be a potential failure of the equipment during or after a seismic event which could potentially prevent a safe shutdown.
The GIP Section 2.3.1, Backfitting, states the following on page 11:
"When a licensee concludes that no further action is necessary for an equipment condition that fails to meet the GIP initial screening or outlier resolution guidelines, but the condition is not deficient against the plant's current license or design basis, the Staff must comply with backfitting requirements pursuant to 10 C.F.R. 50.109 before the licensee 4
Response to NRC Request for AdditionalInformation USI A-46 January 15,1998 Exhibit A can be required to take any further action. The licensee must notify the Staff of the condition in the summary report (see Paragraph 2.2.8 of Part I) and provide an explanation of the safety implications of not modifying the outliers and equipment found to be deficient against the A-46 criteria."
NSP made a decision to resolve all of the identified outliers and themfore did not provide an explanation of the safety implications of not resolving the outliers.
- 5. In Reference A, the response to Question 15 implied that the peer reviewers selected a sampling of safe shutdown equipment list (SSEL) items that are located outside of the containment, the mactor water cleanup room, and main steam tunnel for the purpose of both document review and a walkdown. It is not readily obvious how the peer reviewers reached a conclusion that the remaining set of the SSEL components am acceptable based on their mview/walkdown of a subset of the SSEL items. Discuss any review work performed by the peer reviewers for the SSEL items located within the above noted areas to support your conclusion about the adequacy of the peer review.
Response
The scope of the peer review was in concordance with Section 2.2.7 of Part I of the GIP, which states:
"This evaluation will provide an assessment of the walkdown and analyses by audit and sampling This is intended to be a one- or two- day overview to detemiine if gross errors have been made. This overview effort should be subsubstantially less than the original walkdown and analysis. The individual (s) performing this evaluation will meet the qualification requirements pmscribed for a Seismic Capability Engineer (as described herein) and will document the review in letter form."
The SSER No.2 (Reference C) section 1.2.2, Interpmtation and Guidelines, page 7 states:
"For a meaningful third-party audit (Section 2.2.7 of Part I), the NRC expects that the auditor (s) should have broad engineering experience and have completed the SQUG developed training course on seismic adequacy verification of equipment in operating .
nuclear power plants. This is because the third-party audit will involve substantially less time and effort than the original wa!kdown and analyses. Thus, the auditor (s) should have sufficent qualification and experience to be able to assess the adequacy of the entire plant-specific implementation program during the limited time of the audit."
F The Screening Evaluation Work Sheets (SEWS) and all other mferenced documentation were available to the peer reviewers. The Monticello peer review included an overview of seismic l
l l
Respoue to NRC Request for Additional Information - USI A-46 January 15,1998 Exhibit A evaluation results, a general walkdown of the plant, and detailed examination of a sample of ;he SEWS. The peer reviewers reviewed the SSEL to determine the equipment classes that would be inaccessible at the time of their review. They then selected to review SEWS of equipment thai were in accessible areas. These SEWS encompassed the classes of equipment in the areas that were inaccessible. The review also included caveat compliance, anchorage, and interaction issues that are common to all equipment. The peer review was performed by senior engineers highly qualified to make this assessment. Dr. R.P. Kennedy was a member of the Senior Seismic Review and Advisory Panel (SSRAP). SSRAP reviewed the validity of the GIP methodology for Seismic Qualification Utility Group in cooperation with the US Nuclear Regulatory Commission. lie is a member of the National Academy of Engineering and has worked extensively in the field of safety related nuclear stmetures. He has been a member of ACI and ASCE eMe committees related to design of safety related nuclear structures. Dr. J. Stevenson has also worked extensively in the field of safety related nuclear structures. He has served on committees of the ASCE, ASME, ANS, ACI and AISC charged with the development of standards devoted to design of nuclear plant facilities. Dr. Stevenson's relevant experience includes seismic walkdowns of many nuclear power stations and over 25 years as a structural-mechanical engineer with particular application to earthquake design and analysis. Both Dr.
Stevenson and Dr. R.P. Kennedy have completed completed the SQUG developed training course on seismic adequacy verification of equiptr.ent in operating nuclear power plants.
The SEWS that were reviewed by the peer rev'. ewers encompassed equipment classes that wem in areas that were inaccessible dur'ag the peer review. The sampling of SEWS that were reviewed by the peer reviewers were believed adequate to assure the peer review covered items in the inaccessible areas. The qualifications of the two peer reviewers make them highly qualified to more than meet the requirements of the Seismic Capability Engineers thus adding assurance that their review was adequate.
- 6. In Reference A, the response to Question 21(a) indicated that the n: actor building response spectra (which is also used for buildings other than the reactor building and the emergency filtration train (EFT) building) and the EFT response spectra have in-structure response spectra for elevations within 40 feet that have amplitudes higher than 1.5 times the Seismic Qualification Utilities Group (SQUG) bounding spectrum. Please indicate whether there are any SSEL components within these buildings having a natural frequency ofless than 8 Hz. For such components, use of the floor response spectra provided in Attachment 2 to Reference A rather than the 1.5 times the bounding spectra may be necessary.
Response
There are SSEL components that were evaluated as having a natural frequency below 8 Hertz.
The GIP reonires that these items be evaluated using Floor Response Spectra (FRS) for demand.
All items that were less than 8 Hz were evaluated demand using the applicable FRS as required by the GIP.
6
.- = . . -. - . - - - -
( . Response to NRC Request for Additional Information - USI A 46 ,
January 15,1998 Exhibit A !
- 7. In Reference A, the response to Question 21(b) indicated that for Monticello, application of Generic Implementation Procedure (GIP) method A has been completed in accordance with the requirement thet the safe shutdown earthquake (SSE)is defined at the ground surface. As requested in Reference B, please provide a technicaljustification for not using the in structure response spectra provided in Attachmen' ? fo Reference A.
Raipons Response to this requested information will be provided at a later date as stated in the cover letter.
- 8. The NRC staff has concerns about the way the issue regarding the A-46 cable trays conduit raceway was being disposed of by licensees. The sta'f had issued requests for additional information (RAl) to several licensees on the issue. SQUG responded instead of the licensees because SQUG considered the RAI to be generic in natum. The staff issued a subsequent RAI to SQUG as a follow up to SQUG's response. However, the staff found that the correspondse veitl. SQUG did not achieve tL intended results in that it did not address the technical concerns of the staff. Therefore, the staffis directing the following question to each affected licensee.
The GIP procedun: recommended performing a limited analytic evaluation for selected raceway. and cable trays. The procedure further recommended that when a certain cable tray system can be judged to be ductih and if the vertical load capacity of the anchorage can be established by a load check using three times the dead weight, no further evaluation is needed to demonstrate lateral resistance to vibration from carthquakes.
The GIP procedure eliminates horizontal force evaluations by invoking ductility. However, the staff believes that "non-ductile cable trays" would eventually become ductile by inelastic deformation, buckling or failure of the non-ductile cable tray supports and members. The GIP procedure is a departure from conventional methods of engineering evaluation and the GIP does not provide an adequate bases for dealing with those cable trays that initially are judged to be non-ductile but are eventually called ductile by postulating failure of the lateral supports.
If this procedure was followed for eliminating a cable trays for further assessment at Monticello, then all the cable trays could conceivably be screened out from A-46 evaluation.
The following questions are to elicit information that would support the staffs safety evaluation of cable trays at Menticello.
- a. - Define ductility in engineering terms as used at Monticello foithe USI A-46 evah'ation.
Clarify how this definition is applied to actual system configurations at Monticello plant consistently for the purpose of analytical evaluation.
7
Response to NRC Request for Additional Information - USI A-46 January 15,1998 Exhibit A
- b. Provide the total number of raceways that weie selected for worst-case analytical calculations and were classified as ductile in your A-46 evaluation and .L: which you did not perfomi a horizontal load evaluation. Indicate the approxim' : percentage of such raceways as compared with the population selected for analytical review. Discuss how the ductility concept is used in your walkdown procedures,
- c. Describe the typical configurations of your ductile raceways (dimension, member size, supports, etc.).
- d. Justify the position that ductile raceways need not be evaluated for horizontal load.
When a reference is provided, state the page number and paragraph. The reference should be self contained, and not refer to another reference.
- c. In the evaluation of the cable trays and raceways,if the ductility of the attachments is assumed in one horizontal direction, does it necessarily follow that the same system is ductile in the perpendicular direction? Ifit is not ductile in the perpendicular direction, how was the seismic adequacy of the attachments evaluated?
- f. Discuss any raceways and cable trays including supports in your plant that are outside of the experience data. Explain what criteria are used for establishing their safety adequacy and specify your plan for resolution of outliers that did not meet the acceptance criteria.
Provide examples of the configurations of such raceways and cable trays including supports. Also, indicate the percentage of raceway and cable trays examined during the walkdowns of the safe shut down path, How are they going to be evaluated and disposed?
- g. Submit the evaluation and analysis for four of the representative sample raceway (one single non-ductile, one single ductile, one multiple non-ductile, and one multiple ductile raceway), including the configurations (dimension, member size, supports, etc.).
Responsc Response to this requested infom1ation will be pmvided at a later date as stated in the cover letter.
- 9. In Reference B, the staff had specifically requested the analytical calculations for the resolution of outliers. The NSP's response (Reference A) did not provide the analytical calculations.
8
m Response to NRC Request for Additional Information - USI A-46 ;
January 15,1998 !
Exhibit A
, ' Rather, the NSP requested the NRC to review the material at the Monticello site. The staffis -,
requesting the NSP to provide the analytical calculations for relief valves RV-4236 and RV-4673, Tanks T-45A and T-458. After a veview of the calculations, the staff will determine the - ;
necessity for an audit at Monticello. .
Response 3 The requested analytical calculations are included in Attachment 2.
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Response to NRC Request for AdditionalInformation - USI A' 46 l=
January 15,1998 :
. Exhibit A '
BrJerences: ,
A. Letter fmm William J. Hill to NRC titled," Response to Request for Additional Infonnation on the Resolution of Unirsolved Safety Issue' A 46 (TAC No.-M69460)," dated April 29,'1997;
- B. letter from NRC to Northern State Power Company titled, " Request for Additional -
1Information on the Resolution of the Unresolved Safety issue A-46 (TAC No. M69460),"
dated January 29,1997
- C. - Supplement No.1 to Generic Letccr (GL) 87 02 That transmits Supplemental Safety; Evaluation Report No,' 2 (SSER No. 2) on SQUG Generic Implementation Pmcedure, Revision 2, as Corrected on Fcbruary 14,1992 (GIP 2) h'-
4 10
c Attachment 1 -
Calculation of RS for Other Dampings Y
c w
Calculation of RS for Other Dampings Step 1 - Calculate PSD From RS
- An acceleration power spectral density function (pSD) is developed from a given response acceleration spectrum (RS). He equivalent PSD is extracted using methodology developed by Vanmarcke (1,2) and Kaul (3).
The relation between a response spectrum and the associated PSD is derived from random process theories. The response spectrum at a certain frequency is the maximum re:ponse of a single degree of freedom (SDOF) system.
. He distribution of the maxima can be computed knowing the energy content (or PSD) of the random process and the probability of exceedance. The relationship between the PSD and the RS is established by assigning a probability of exceedance (r) to the RS. His exceedance value is the probability that the value of the RS will be exceeded during a seismic event of duration T. Treating the seismic event as a Gaussian process, the relationship between the RS (a measure af peak response) and the PSD (a statistical measure of motion) can then be derived. Given a target acceleration response spectrum, tne initial approx!mation PSD is determined fre n the response spectrum by f " ('9 I) r(m)o = 2(,(S(m,())2 t[ -
21n x In(1-r) xm u <mT n where F(m)o is the initial approximation PSD; T is the effective earthquake duration; r is the probability of exceedance, and S(m,() is the target acceleration response spectrum (frequency e and damping ratio (). The effective damping term is (, = (+ 2/mT .
The values of t and T are set to 0.5 and 15 seconds, respectively. The actual values of these parameters are not critical to RS generation since they are in turn used in the reverse process of PSD to RS conversion.
To ensure proper representation of the PSD from the response spectrum, the solution is iterated and a new response spectrum is determined. The PSD is updated by F(m)m = F(m)l(8(*'E)- S(e,(),j where S(m,{),is the acceleration RS determined from F(m); using eqs. 6 threugh 8. The process is continued until the calculated response spectrum closely matches the target spectrum. The convergence error c is calculated by the following m 2 (eg, 3)
RSUM = [(S(a3,($
i.,
= _
2 (eg,4)
RSUMD = [(S(4.()-S(4,()<)
/=t
(cq 5) c (RSUMDQ 1
E E
where a4,ja'I to nrs, are frequency points at which the response spectrum is defined.
The calculations are continued until the maximum number ofiterations specified by the user is reached. Five _'
iterations were perfonned and a convergence of c < 5% was achieved for all cases, i Step 2- Check Envelope Criteria
- The final PSD from Step I is checked by comparing the final RS to the target RS using the enveloping criteria of the Standard Review Plan, Section 3.7.1 (reference 4). If needed, the criteria of the Standard Review Plan are achieved by scaling up the results of Step 1. .
Step 3 - Calculate RS From PSD at Desired Damping The following formula is used to calculate an acceleration response spectrum from an acceleration PSD - ;
r gn y sta (eq. 6) 7
-. S(mo,() = < -2moln
~# E In(1-r) -
T \ms>
< s.
- where (eq. 7)
- m. = m<mo) = ,m"llle(s)l' P(m)dc)n=0,2,4..,
and where mo' + 4 mo' 4' m' (*9 8)
=
lHo(m)la (mo -m2 ):,4 g,2 42g2 2
. P(m) is the PSD obtained from Steps I and 2. S(mo,() is the RS at the desired damping ratio 4 References
- 1, ' EJI. Vanmarcke and D.A. Gasparini. " Simulated Earthquake Motions Compatible with Prescribed Response Spectra." MIT, Department of Civil Engineering, Publication No. R76-4, Order No. 527, January 1976. ;
2, . "SIMQKE: A Program for Artificial Motion Generation." MIT, Department of Civil Engineering, Users-Manol and Documentation, November,1976.
. 3. - = M1 Kaul. " Stochastic Characterization of Earthquakes Through Their Response Spectrum." Earthauake Engineerine and Structural Dynamics, Vol. 6,1978, pp. 497 509.
4? . NUREG 0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, US Nuclear Regulatory Commiss, ion, July 1981 2
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- 9 Monticello' Nuclear Generating Plant Acceleration Response Spectre ReactorSuWing Elevetlon 936 Horleontal 8
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.._ g o1 1 10 100 Frequency (Het l Figure 1 Monticello Nucleet Generating Plant Acceleration Power Sps . vel Donelty Reactor BuWing Elevation 935 Hortrontal 100E .
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Menscoto Nuclear Generating Plant Accelerotion Response Spectre
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Figure 3 Monticello Nuclear Generating Plant Acseierstion Power Spectrel Denelty
- Reactor Building Elevation 943 Hortsontal t out42 100E43
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Figure 4:
4-4
i Attachment 2 Calculations
-