ML20138G043
| ML20138G043 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 04/29/1997 |
| From: | Hill W NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR GL-87-02, GL-87-2, TAC-M69460, NUDOCS 9705060145 | |
| Download: ML20138G043 (39) | |
Text
44 Northem States Power Company l
April 29,1997 U S Nuclear Regulatory Commission GL 87-02 Attn: Document Control Desk 'JSI A-46 Washington, DC 20555 MONTICELLO NUCLEAR GENERATING PLANT l Docket No. 50-263 License No. DPR-22 i l
Response to Request for Additions!Information on the Resolution of Unrevolved Safety issue A-46 (TAC NO. M69460) ;
i By letter dated January 29,1997, tilted " Request for Addtional Information on the Resolution of l Unresolved Safety Issue A-46 (TAC NO. M69460)", the NRC requested additionalinformation j beyond Monticello's original November 20,1995 submittal on the same subject. This letter provides Monticello's response to the NRC's January 29,1997 request for additional information.
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 j information related to this request. )
. I i
} kbs William J Hill l Plant Manager Monticello Nuclear Generating Plant c: Regional Administrator-l!I, NRC NRR Project Manager, NRC Resident inspector, NRC State of Minnesota Attn: Kris Sanda J Silberg r.
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Attachments: Affidavit to the US Nuclear Regulatory Commission.
f I Attachment 1: Response to Request for AdditionalInformation.
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UNITED STATES NUCLEAR REGULATORY COMMISSION !
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. NORTHERN STATES POWER COMPANY )
MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 Response 'o Request for Additional information on the Resolution of UrJevolved Safety Issue A-46 (TAC NO. M69460) ;
, Northem States Power Company, a Minnesota corporation, herby provides the l Information requested by the USNRC in letter dated January 29,1997 titled " Request i for AddtionalInformation on the Resolution of Unresolved Safety issue A-46." l 1
This letter contains no restricted or other defense infonnation.
NORTHERN STATES POWER COMPANY By M5 4 \
William J Hill u !
Plant Manager !
Monticello Nuclear Generating Plant l l
On this@l ay of Atd 1947 before me a notary public in and for said County, personally appeared William J Hill, Plant Manager, Monticello Nuclear Generating Piant, and being first culy swom acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, that he knows the contents thereof, and that to the best of his knowledge, information, and belief the statements j made in it are true and that it is not interposed for delay. I Nwua
/ Samuel
-4 Yy . ... _ __ _____. _ -
I. Shirey SAMUEL 1. SHIREY Notary Public - Minnesota (% /alb umm euauc - MINNESOTA Sherbume County '
. . , uf comm rs. Jan. 31,20@ ,
My Commission Expires January 31,2000 '
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t s RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION - USl A-46 MONTICELLO NUCLEAR GENERATING PLANT
, April 29,1997 Attachment 1
Reference:
Letter from NRC to Northern States Power Company "Monticello Nuclear Generating Plant - Request for Additional Information on the Resolution of Unresolved Safety issue A-46 (TAC No. M69460)", January 29,1997.
I
, The following questions are from the aboved referenced letter.
l
- 1. With respect to Section 3.2, page 3-1, discuss the basis for defining the seismic demand for equipment housed in buildings other than the reactor and emergency filtration train (EFT) buildings as equivalent to the dernand based on the response spectra generated for the reactor building.
1
Response
The intake Structure and the Emergency Diesel Building are relatively short squat i structures that will exhibit low amplification compared to the Reactor Building which is a much taller structure. The use of the Reactor Building response spectra for corresponding floors for these buildings is conservative due to the higher amplification of the Reactor Building.
The Plant Control and Cable Spreading Structure and the Turbine Building are of similar reinforced concrete construction as the Reactor Building, it is reasoreble to expect the response of these buildings would be similar to that of the Reactor Building. The use of the Reactor Building response spectra for corresponding floors i on these buildings is believed to be adequate for the seismic design of equipment in both the Plant Control and Cable Spreading Structure and the Turbine Building :
On May 22,1992, the NRC issued Generic Letter 87-02 Supplement I, which consisted of the NRC 's review of the Generic Implementation Procedure (GIP).
The letter requested that Seismic Qualification Utilities Group (SOUG) members provide to the NRC, within 120 days, information conceming the in-structure response spectra that would be used for the resolution of USI A-46. On September 21,1992, NSP responded to this request (Reference 3) by providing the Reactor Building and EFT Building response spectra and by stating " seismic demand for equipment housed in buildings other than the Reactor Building and the EFT Building will be based on the response spectra generated with the Reactor Building moder.
On December 10,1992, the NRC sent its evaluation (Reference 4) of NSP's September 21,1992 response letter. This evaluation stated that as a result of the NRC's review of the floor response spectra that was submitted the "... staff is I
/
- Response to NRC R:qu::st for Additional information - USI A-46 April 29,1997 Attachment 1 satisfied that the licensee has produced " conservative, design" FRS for ensuring the safety of the plant". Based on this NRC letter NSP performed the USl A-46 evaluation using the floor response spectra when Method B in section 4.2 of the GIP was utilized.
- 2. With respect to Section 3.2.1, page 3-2, explain in some detail as to how the original floor response spectrum (FRS) data were used to calculate response spectra for the additional oscillator damping.
Response
Only original 0.5% oscillator damping floor response spectra (FRS) were available for major elevations of the Reactor Building. To produce response spectrum cuves for other damping values, damping independent power spectra density (PSD) curves were generated from ava.4able FRS. FRS for other oscillator dampings were then calculated from the PSD curves.
Each PSD was calculated by a successive approximation techniquo. An original estimate is obtained and then, by tracking an error function, the estimate is successively updated. The error function measures the difference between the target response spectrum and the trial response spectrum (target spectrum: original FRS, trial spectrum: the one from the current PSD). The iteration continues until a suitable fit of the target spectrum is obtained. For all cases invo!v J, the trial response spectrum very closely matches or envelopes the target asponse spectrum through the frequency range of interest or both.
Information related to Monticello FRS was also summarized in NSP's response to GL 87-02 (Reference 3).
- 3. Section 4," Screening Verification and Walkdown," of GIP-2 under item " Caveats" reads "In order to ... or the generic seismic testing GERS, the eauipment (underline added) should be similar (underline added) to the equipment in the earthquake experience equipment class or the generic seismic testing equipment class and..."
However, SecSon 4.1.2, " Caveat Compliance" of the Monticello report (page 4-2) states that "the eauipment characteristics (underline added) are aenerally similar (undorline added) to the earthquake experience equipment class." Explain the basis for this apparent deviation frcm the specific caveat wording of GIP-2 and discuss in detail how the change in the specific wording of GIP-2 caveats impacted the finallist of outliers.
Also provide a list of additional outliers which would have been identified if the wording in GlP-2 caveats was not changed.
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/ s i R:sponse to NRC Requ st for Additional Information - USI A-46
. April 29,1997 I Attachment 1 !
l i
Response
There was no intended significance in writing " generally similar" as opposed to "similar". The guidelines of reference 2 (the " GIP"), including all equipment caveats, were followed as written in that document. A synopsis of the GlP analysis methodology was provided in the Monticello report as an aid to future readers and did not substitute in any way for the GIP document.
i
- 4. Referring to Table 4-1," items Meeting Intent but Not Specific Wording of Caveats,"
each of the items listed in the table involve some degree of judgment or estimation by licensee personnel in concluding that the intent of applicable caveats was met. For i each item listed in the table, provide a discussion of the bases for the judgment or estimation. Also, provide the following additional information:
- a. For Control Cabinets C-19, C-289A, C-30, C-32 and C-33, provide a comparison to show that a through-bolt is equivalent to a cast in place bolt in meeting the intent of caveats.
Response
The through-bolts are different from cast-in-place anchors in that: (1) there is no bond between the anchor and the concrete, and (2) the head of the anchor is not embedded, it is on the far face of the slab. Per reference 6, the GIP cast-in-place bolt allowables are based on ACl-349 Appendix B principles for a smooth bar with mechanical anchorage.
Therefore, the principles and requirements of ACl-349 are relevant in evaluating whether there is any significant difference in behavior.
The Commentary on ACl-349 Appendix B states that, for a smooth bar with mechanical anchorage, " bond failure will occur such that the entire tensile load is trsosferred into the concrete by the anchor head."
Consistent with this statement, the formulation of ACI-349 Appendix B does not credit bond between the bar and the concrete. Therefore, it is reasonable to conclude that the absence of bond does not preclude the use of the ACI-349 Appenoix B principles for determining the strength of a through-bolt.
There is no requirement in ACI-349 Appendix B to maintain a minimum distance between an anchor head and the far face of the concrete. On the contrary, when the anchor head is located bamnd the far face reinforcement, a more liberal strength reductior. . tor may be used I
($=0.85 instead of 0.65, GIP allowables are baseu on a 0.65 factor).
3
Responsa to NRC Requ st for AdditionalInformation - USl A-46 April 29,1997 Attachment 1 When the anchor head is located beyond the far face reinforcement, ACl-349 Appendix B Commentary states that the anchor load will be equivalent to a concentrated load on a slab, and that the use of $=0.85 is consistent with the requirements for that type of load. Therefore, the placement of the anchor head on the far face of the slab does not preclude the use of the ACl-349 Appendix B principles for determining strength (actually increases code-based capacity).
- b. For Hydraulic Control Units CRD HCU E and W, provide the computation pertaining to the adequacy of fluid operated valves, the pneumatic controls (e.g., solenoid valves) and the overhead lines of the HCUs.
Response
The fluid operated valves for the Hydraulic Control Units are diaphragm operated valves on 1" lines. The valves were evaluated using the GIP Class 7 requirements (fluid operated valve class) and were found to be adequate. The rack and pneumatic controls were evaluated using GlP Class 18 requirements (instruments on racks class) and were found to be adequate. Per Appendix B of GIP, the experience database for Class 18 includes solenoid valves and pneumatic system components for control of pneumatically actuated valves. The small-bore insert /withdrawallines from the HCUs were checked for vulnerability to seismic proximity interaction with the building work platform in the proximity of the HCUs.
Because the work platforms are well braced, the interaction concern was dismissed (consistent with the guidelines of Appendix D of GlP). The Class 7 and Class 18 Screening and Evaluation Work Sheets (SEWS) have been completed and are available for review at the Monticello site.
- c. For ECCS Area Drain Pumps, P-88A, B, C and D, discuss in detail how these vertical centrifugal drain pumps meet the intent of Class 6 caveats.
Response
The Bounding Spectrum caveats as applied to pumps P-88A through P-88D are discussed below.
VP/BS Caveat 1- Earthauake Experience Eauipment Class: The equipment description for this class excludes submersible pumps since they cannot be inspected. The Monticello pump is not submerged, nor is it being used in that fashion. This pump is being used as a single stage, vertical pump. Thus, it is evaluated as a vertical pump.
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R:sponse to NRC R:qu:st for Additional Information - USl A-46 April 29,1997 Attachment 1 VP/BS Caveat 2 - Cantilever Impeller Shaft Less Than 20 Feet Lona: Not applicable; tuis pump is an inline vertical pump with no impeller shaft.
VP/BS Caveat 3 - Check of Lona Unsupported Pipina: There was no instance of long unsupported attached piping.
VP/BS Caveat 4 - Sufficient Slack and Flexibility of Attached Lines:
Attached lines had sufficient flexibility. The pump is rigidly anchored and the seismic displacements would be very small.
VP/BS Caveat 5- Adecuate Anchorace: The pump is small and
, anchored with three 5/8" anchors. The anchorage was evaluated and was found to have a high margin of safety.
VP/BS Caveat 6 - Potential Chatter of Essential Relays Evaluated: Not i applicable; this pump does not house essential relays.
VP/BS Caveat 7 - No other concerns: There were no other concerns. l l
A!! anchorage caveat requirements were met and the pumps were found to be free of any spacialinteraction effects.
- d. For relief valves RV-4236 and RV-4673 (each of which is mounted off a %"line),
i provide the calculation to show that the seismic stress is less than the allowable stress for the attached piping to ensure the seismic adequacy of the valve support.
1 Response j i
Per GIP Appendix B, the intent of the FOV/BS Caveat 4 is to assure that !
pipe stresses associated with eccentric valve loads are not excessive, in l this case, although the line was less than 1", the valve was clearly smaller than that allowed for a 1"line. The assessment of the 3/4" line deviation was performed by estimating the potential increase in stress relative to that considered acceptable under the GIP for a 1" line. The assessment is documented in the Screening Evaluation Work Sheets (SEWS) for RV- i 4236 and RV-4673 which are available for review at the Monticello site, j
- e. For Conduit in Area RB-A30 and Dampers Controllers TC-8089C and TC-8089L in i the diesel room, discuss in detail the tug test performed by the SRT, and provide the basis for concluding the acceptability of the clamp supports and anchorage.
Response j I
l 5
i <
Response to NRC Requ st for AdditionalInformation - USl A 46 :
April 29,1997 Attachment 1 i
in Area RB-A30, a small number of very lightly loaded beam clamps, that are supporting an 1" diameter conduit, were found to be oriented such that tney rely on friction for vertical loads. The support spacing of the clamps is 3 to 6 feet, indicating a dead load of no more than 15 pounds per clamp.
The SCE performed a tug test that verified that the clamps were well preloaded and the system was rigid. The clamps would be required to resist a small total load during a seismic event due to the small supported weight and rigid response of the conduit. The SCE concluded that the tug test exceeded the credible total load during an Safe Shutdown Earthquake i
(SSE), and therefore the vertical friction concem with the beam clamp ;
orientation was found to be acceptable. 1 TC-8089C and TC-8089L are each anchored to a concrete wall with #8 screws using unknown inserts. The items in question are very small, with !
each estimated to weigh less than 5 pounds. The items were tug tested by the Seismic Capability Engineers (SCE) to an estimated 20 pounds and the anchorage was judged to be acceptable. The SCE judgments were documented in the SEWS. l 1
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- 5. In reference to Section 4.1.3, tha report states that anchorages were rigorously I analyzed using hand calculation and ANCHOR software package. Provide samples of l the anchorages engineering calculations. Also discuss a few cases of anchorage verification based on results of tug tests conducted and provide a description of the l tests and the engineering justification for such an approach.
Response !
Sample ANCHOR software evaluations for equipment Y-82 and P-11 are included in Attachment 1.
The engineering justification used for a tug test on a small component is that the applied load is larger than any credible seismic demand, or, the small component is clearly ruggedly anchored and the tug test verifies no gross installation defects. One example case of a tug test anchorage verification is TC-8089 which was discussed in Responso 4e. Another example is panel N3346A, which is a small wall-mounted panel housing a hand switch. It is 9 inches wide by 22 inches high by 6 inches deep. The item was pull tested to greater than 75 pounds (conservative estimate) with no looseness detected and thus was judged acceptable for anchorage (documented in a SEWS).
An example of a small, ruggedly anchored item is panel C-253D. The item is a wall-meunted panel,20 inches wide,30 inches high, and 7 inches deep. It is anchored to a reinforced concrete wall with four 3/8" anchor bolts. The 6
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Responsa to NRC Requnst for Additional Information - USI A-46
- April 29,1997 Attachment 1 4
- item was pull tested to greater than 50 pounds per side and no looseness was detected. The anchorage was judged acceptab!o because the anchorage was clearly rugged since there were no gross installation defects j and the capacity obviously far exceeds demand). The SCE judgments were 4
documented in the SEWS.
s 4
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. 6. With respect to Section 4.3.2," Comments About Anchorage," the last sentence of the second paragraph on page 4-7 reads: " Wall mounted equipment was not subject to a tightness check as allowed by the GIP because..." Identify the specific provision of the
- GIP-2 which allows such an exemption from performing the needed tightness check.
Response
The tightness check is used to detect gross installations errors which would
- leave the anchor loose in the hole. Per Section 4.4.1 of the GlP, anchors loaded in tension due to dead weight need not be tightness checked since
] the anchor set is effectively proof-tested by the dead weight loading. Wall mounted panels often presented a case of inaccessible anchors, because wall anchors were hidden behind the panel, and only disassembly of the panel would gain access to the anchors. In that case, the SCE relied on the i
dead load proof-testing for upper anchors and tug testing at the bottom and
,: sides to verify no gross installation defects. In all cases, wall mounted panels l were small enough such that the tug test was a meaningful exercise in
! verifying no gross installation defects.
i 7. With respect to Section 4.3.2, third-paragraph, discuss the extent of inspection implemented by a so-called " random ' spot' embedment check," and elaborate on the validity of the conclusion drawn from such a spot check with no identified installation problems.
J
- Response i
Original construction expansion anchors used Phillips shell anchor bolts. As stated in GIP Appendix C, appropriate installation is assured if the shell of these anchors does not protrude above the surface of the concrete. For most
. cases, it is not possible to see the installed shell because it is covered by the
. anchored equipment. Per the GIP Section C.2.4, rather than disassembling 4
many anchors, a spot check is acceptable, so long as no installation
! problems are observed. If installation problems are observed then the
. inspection should be expanded accordingly. The SCE was able to inspect
)
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I o .
l 1
Responsa to NRC Requ:st for Additional Information - USI A-46 l
April 29,1997 i Attachment 1 many anchors and no protrusions were encountered. In addition, during the Seismic Safety Evaluation of the Anchorage of Safety-Related Electrical Equipment program (Reference 10), a sampling of anchors under grout pads were completely disassembled, and no protrusions were encountered. I Therefore, based on the GlP, the level of inspection was valid.
- 8. Provide information regarding the seismic adequacy verification implemented under IE Bulletin 80-11 for a list of masonry block walls which were identified by the SRT to possess a 11/1 implication and a potential of collapsing on items listed in the SSEL.
Response
IE Bulletin 80-11, Masonry Wall Design, required licensees to identify masonry walls and the intended function associated with such walls. The licensees were also required to present reevaluation criteria for the masonry walls with anMysis to justify !
those criteria. NSP complied with this Bulletin and received a Safety Evaluation I from the NRC. This safety evaluation, " Safety Evaluation by the Office of Nuclear Reactor Regulation on Masonry Wall Design (Generic issue B-59) at the Monticello I Nuclear Generating Plant Docket No. 50-263", found that NSP had fully ;
implemented the requirements of NRC IE Bulletin 80-11. ;
When the SRT had determined that a masonry block wall had the potential of collapsing on a piece of equipment on the SSEL it was verified that the wall was evaluated as part of the 80-11 program and found to be seismically adequate. The wall was noted on the SEWS and referenced to as a wall that was previously evaluated in the NRC IE Bulletin 80-11 program.
- 9. Section 7.1.2 of GlP-2 specifies that a SOUG licensee perform a four-step engineering evaluation for verifying the seismic adequacy of tanks and heat exchangers according to guidelines provided in the section. However, under Section 5.1 of the licensee's report, it is stated on page 5-1, that "the Seismic Capability Engineers performed the evaluation such that they meet the intent (underline added) of these guidelines..."
Clarify whether the GIP-2 guidelines were always met, or identify any deviations from the guidelines where only the intent of the GIP-2 was satisfied.
Response
Section 7.1.2 of GIP-2 lists four checks to perform for the equipment class.
Two of these checks apply only to large, flat-bottom vertical tanks. There were no large, flat-bottom vertical tanks on the SSEL. The other two checks, 8
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Responsa to NRC Rcquast for Additional Information - USI A-46 April 29,1997
- Attachment 1 the anchor bolt and embedment strength; and the anchorage connection strength between the anchor bolts and the shell of the tank or heat
- exchanger, were applied as required by the GlP for all tanks on the SSEL.
Therefore the guidelines in the GIP-2 were met.
10.With respect to the Diesel Oil Storage Tank (T-44) listed in Table 5-1, provide the i
rationale (supported by engineering analysis) for concluding that no large relative motion between the tank and the Pump House will take place during the SSE.
. Response i
Since the tank, pipe and the pump (housed in the Pump House) are below .
j ground level, the significant concem is relative motion between the tank ar:d '
1 the Pump House that could break or crimp the supply line. However, since 4 both the tank and the Pump House are founded on soil and are very close to each other, large relative motions are not expected for the SSE. In addition, the bends in the piping will accommodate relative motion. 1 a
11.With respect to Section 6.1, discuss in greater detail the basis for judghig that conduit and cable tray supports in some inaccessible areas are acceptable and provide examples which led to such a conclusion.
Response
in limited instances, high radiation levels in segregated areas prevented access for cable and conduit raceway walkdown. These areas were all rooms within a much larger building floor, it is estimated that these areas represent less than 2% of the area under review. Information about the type of conduit and cable tray supports was obtained from pictures, drawings, and from plant personnel with knowledge of these room. As an example, for Area RB-A9, pictures of the area were viewed via NSP's "CEVUE" facility and it was judged to be acceptable (the area contains conduit only, no SSEL j equipment, and the conduit was seen to be well supported). CEVUE is a computer program containing images of inaccessible areas. With CEVUE, the users can conduct a virtual" tour" of the area interactively, )
12.Regarding Section 6.5, indicate why there were no rod hanger supports chosen for the I Limited Analytical Review (LAR). Also, discuss the basis for selecting the cases shown !
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u -
- Respons3 to NRC R
- qu st for Additional Information - USI A-46 April 29,1997 Attachment 1 on Table 6-2 and the rationale for judging that selection of 12 LAR cases is sufficiently adequate to represent the entire population of raceway and conduit supports.
Response
The vast majority of cable and conduit raceway hangers were of light steel
' frame construction, the "Unistrut" brand framing system is predominately used. The only raceway or conduit rod hangers identified were in the Non-1E j Electrical Room (part of area TB-A8) As documented on the Plant Area i Summary Sheet, these hangers were lightly loaded, there were no short
- hangers, rods were not vulnerable to fatigue loading, and there were no instances of rods threaded into ceiling-embedded expansion anchors.
Because the hangers were lightly loaded, had good dead load support, and
- had lateral ductility, it was determined that they passed the GIP screening j
. criteria. -
The basis for the selection of LAR cases was Section 8.2.4 of the GIP. The l Seismic Capability Engineers sought out representative samples of the major
, different types of raceway support configurations, and the most heavily loaded raceway support for each configuration. The rationale forjudging that 12 LAR cases are sufficient is Section 8.2.4 of the GIP. As a guideline, Section 8.2.4 specifies a minimum of 10 LAR cases. The sample size should vary with the diversity and complexity of the design and construction. At M nticello, the raceway support population was relatively uniform with the large majority are either cantilever bracket or a rigid wall bracket.
13.Regarding Section 7, describe the approach which will be used to obtain the more realistic in-structure demand for the outlier pumps in the intake structure.
. Response
- More realistic in-structure demand will be obtained by creating a dynamic model of the intake Structure. For USl A-46 resolution, in-structure demand
, for the intake Structure was based on Reactor Building floor response spectra (see also responses to Request 1 and Request 14).
14.In Table 8-1, only proposed outlier resolutions were included. Provide results of evaluations, tests, calculations and equipment modifications and replacements used to
- resolve outliers, as required by Section 9 of the GIP-2. Also, provide the justification to 4 ensure that the proposed schedule for resolving all the identified outliers or open items 10
f
- j Responsa to NRC Request for Additional Information - USl A-46 April 29,1997
- Attachment 1 prior to the end of 1998 refueling outage, does not lead to a potential safety significant l scenario. i
Response
j Reference 1 documents impleraentation of the GIP. The GIP provides no
- specific procedure for outlier resolution, and the task of outlier resolution is being performed indepeadently from implementation of the GIP. Per Section 9 of the GIP, a description of equipment outliers, a list of unresolved outliers, i and a proposed schedule for outlier resolution should be inc!uded in the j' submittal to the NRC. This information was submitted with the proposed i
schedule in reference 11. All identified outliers are treated as unresolved
- until they are resolved by documenting the follow-on engineering evaluations.
i 3
The outliers that are identified in Reference 1 were classified as outliers because l they did not meet the screening criteria in GlP-2. The resolution of the outliers will
- include additional analysis, investigation or a modification to the equipment all of i which would allow the equipment to meet the screening criteria. The method of r resolution used is based on which is the easiest appropriate method for the equipment to meet the screening criteria. The GIP-2 screening criteria is not part of the licensing basis at Monticello, therefore not meeting the GIP-2 screening criteria does not, of itself, give rise to an operability or safety concern. The SRT determined there was no reason to believed that the equipment that was classified as an outlier, as a result of the GIP-2 screening process, did not meet the plant licensing or design basis. l
- 15. Referring to the Monticello Nuclear Generating Station A-46 Seismic Third Party Audit Report, the peer reviewers did not perform the walkdown of inaccessible areas due to radiological concerns incluaing the Primary Containment, the Reactor Water Cleanup Room and Main Steam Tunnels. Discuss the general approach taken by the SRT in dealing with items listed in the SSEL that are located in inaccessible areas, and the rationale for concluding that the items are properly verified for their seismic adequacy.
Response
The approach of the SCE was to walk down all items on safe shutdown equipment list (SSEL). A walkdown was performed for all equipment on the SSEL and documented in a SEWS. Access to equipment often revolved around changes in plant status e.g., Primary Containment items were walked down during a refueling outage. The peer reviewers selected a sampling of SSEL items for review and performed a walkdown on those items. The peer reviewers walkdown was done during plant operation which made the 11
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- Response to NRC R quest for Additional Information - US; h-46 April 29,1997 1 Attachment 1 i
4 Primary Containment, the Reactor Water Cleanup Room and Main Steam Tunnel inaccessible due to radiological and inert gas conditions. However, the . sample of SSEL items that were selected by the peer reviewer provided !
assurance for all the SSEL components. l I
- 16. Describe any corrective measures taken to address the peer reviewers comments as I described in Appendix E of Attachment 1, for example, (1) Cabinet C-253A is about %" l
, away from cabinet Y-25 and may be an impact hazard, and (2) Cabinet C-27 contains a !
flexible RPIS Translation Electronics rack with circuit boards that could pop out under i seismic loads. The findings in these examples were not addressed in Table 7-1, Equipment Outliers. l
Response
(1) The SCE evaluated the potential for relative displacement between l C253A and Y-25. Y-25 is a small panel rigidly mounted to a building column. l C253A is a single section welded steel control panel. The SCE concluded that the relative displacement between C-253A and Y-25 during an SSE l would be well below 1/4". These conclusions were documented in the SEWS for C-253A. I l
(2) Cabinet C27 was listed and walked down based on a preliminary version I of the SSEL. Cabnet C27 is not part of the final SSEL. !
l 17.Regarding Appendix C of Attachment 2,"USl A-46 Resolution, Relay Evaluation Report, Monticello Nuclear Generating Plant," of the submittal, provide a discussion and !
specific examples of relays whose malfunction (i.e. chatter) is acceptable. These relays !
were identified as " Chatter Acceptable (CA)" in the appendix. !
Response
Evaluation of consequences of relay malfunction was performed following the methods contained in GIP Section 6, " RELAY FUNCTIONALITY REVIEW",
and EPRI NP-7148, " Procedure for Evaluating Nuclear Power Plant Relay Seismic Functionality". ,
A simplified failure modes and effects analysis was used to examine the l
consequences of relay malfunction (i.e. chatter). This step screened from l further consideration those relays, or complete circuits of relays, whose l 12
t '
R:sponsa to NRC Request for AdditionalInformation - USI A-46 April 29,1997
, Attachment 1 malfunction would not prevent system / component functioning or cause other unacceptable conditions.
Malfunction of specific relays and their associated circuits is considered '
acceptable for shutdown after an earthquake if:
. The function provided by the system and associated relays is not needed during the period of strong shaking and relay malfunction does ;
not make essential functions unavailable when needed after the strong i shaking.
, e Ralay malfunction does not prevent a desired function from occurring.
. Relay malfunction does not cause a spurious, unacceptable event.
Examples:
SSEL Line Number: 8066 Plant System: 4kV .
Component / Subsystem: 152-308 Relay Designation: 152-301/b Relay Type: GE-AMH-4.76-250 These contacts provide a trip signal to breaker 152-308 through a 5 second time delay relay when both breakers 152-301 and 152-302 are open, thereby initiating the transfer of the essential busses to 1 AR or the EDGs. This transfer is also initiated by the degraded / loss of voltage relays. Chatter in these contacts can not cause an undesired transfer or prevent a desired transfer, therefore, chatter is acceptable in this circu!t.
SSEL Line Nurnber: 12000 Plant System: APR Component / Subsystem: ADS Logic -
Ptelay Designation: 10A-K85A Relay Type: GE 12HGA11 A52F ADS function is not required. ADS logic is reviewed for possible malfunction (i.e. inadvertent opening of ADS valves). Since ADS logic includes a 107 second time delay (2E-KSA), relay chatter in the initiation logic (other than 2E-K5A) is acceptable.
SSEL Line Numoer: 20000 Plant System: ANN Component / Subsystem: Alarm Logic on all control circuit drawings Spurious operation of alarms, caused by relay chatter and other spurious events, such as water sloshing, may likely occur during the strong earthquake motion but will be corrected by normal operator actions before their effect can be of consequence. Accordingly, chatter of relay contacts and other contacts 13
Response to NRC Request for Additional Information - USl A-46 April 29,1997 Attachment 1 feeding the a! arm system is acceptable and the relays in these systems need not be seismically qualified. Reference EPRI NP-7148 SL, Appendix C, Section C.3.3.
SSEL Line Number: ~7157A Plant System: DOL Component /Subsystern: FTM-1 (11 OG) 4 Relay Designation: FTC1 Relay Type: 8474707 Contactor FTC1 controls the operation of the fuel transfer pump which
, transfers fuel between the day tank and the base tank. The base tank contains sufficient fuel volume for more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of diesel operation.
Therefore, momentary interruption of the transfer pump during the 30 seconds of strong shaking would have no effect on diesel operation and chatter in this contact is acceptable.
SSEL Line Number: 8003 Plant System: 480 Component / Subsystem: LC-103 (52-305)
Relay Designation: 83-305 Relay Type: GE HFAS4H Chatter or change of state of this relay could prevent auto-start of the Service Water pump. Service water is not required and this relay cannot trip a running pump or prevent the load-shed of the pump. The desired function for breaker 52-305 is to trip during required load shed conditions. Therefore, chatter of this relay is acceptable.
SSEL Line Number: 12000 Piant System: APR Component / Subsystem: Low-Low Set SCRAM Permissive - Division l Relay Designation: 5A-K13A Relay Type: GE CR305D102 Contacts provide a SCRAM signal permissive to the Low-Low Set logic.
Chatter in these contacts can not cause a Main Steam Safety Relief Valve (SRV) to open. Chatter in these contacts could cause an open SRV to close before the close setpoint is reached. Low-Low set logic would then prevent the reopening of this SRV for 10 seconds. However, this would require multiple coincident opening of these contacts affecting both divisions of Low-Low set. Only three of the eight Main Steam SRVs have the Low-Low Set logic. Furthem1 ore, the pressure relief function of all SRVs will still be avaihble. Therefore, chatter in this contact is acceptable.
SSEL Line Number: 1091 Plant System: RHR Component / Subsystem: MO-2006 Relay Designation: 10A-K58A Relay Type: GE 12HFA151 A2F 14
i l Response to NRC Request for Additional Information - USl A-46 April 29,1997
- Anachment i l
l' Normal valve MO-2006 position is closed. Relay 10A-K58A provides a permiscive to manually open the valve but cannot open the valve without the
)
j i corresponding control switch in the open pocition. Valve MO-2006 does not '
need to open until after the seismic event. Relay also blocks the auto close
- signal. However a close signal tvonid seal in until the valve is closed and
- chatter in this relay can nol provent the valve from closing. Therefore, chatter in this relay is acceptable.
! SSEL Line Number: 1121 Plant System: RHR l l Component / Subsystem: MO-2014 Relay Designation: 10A-K66A (NC, DE Contact)
Relay Type: GE 12HGA11 A52F
- l I
j i he normally closed, de-energized (NC, DE) contact of relay 10A-K66A j provides a permissive for closing valve MO-2014. This contact cannot cause i automatic operation of the valve withcut a close signal present. A close signal seals in and chatter in this contact would not prevent valve closure.
1 SSEL Line Number: 11000 Plant System: MST Component /Gubsystem: MSIV Control Logic Relay Designation: 16A-K1 A Relay Type: GE 12HFA151 A9F
, Main Steam isolation Valve (MSIV) isciation is initiated by a 1-out-of-2-twice logic from relays 16A-K7A, B, C, & D. Once the 1-out-of-2-twice logic is d
initiated, the MSIV isolation logic seals in until tha initiating signal clears, all four valve control switches are brought to CLOSE, and the Main Steam Isolation is reset. The desired action is for the MSIVs to close. Chatter in the initiation logic could cause the valves to close but cannot reopen them. Thus, j chatter in the initiation lod ic places the MSIVs in a closed position. !
L i
- 18. Appendix C of Attachment 2," Operator Actions," specifies relay / component combinations which are resolved by operator actions. Clarify how these operator actions were verified and validated to ensure that under the postulated conditions of a design-basis earthquake they could be adequately executed. What field and control room simulator scenarios were developed to verify and validate that these operator actions could be accomplished in the time frame required to facilitate safe shutdown?
How were potentially harsh environmental conditions (e.g. blackout, high temperature and high pressure) factomd into these analyses?
Response
15
Response to NRC Request for AdditionalInformation - USl A-46 April 29,1997 Attachment 1 ,
Personnel from the plant Operations Department conducted a review of the components which require operator actions as identified within Appendix C of Attachment 2. This review ericornpassed identifying the component, the specific operator action required for that component, and the time required to perform the specified action. Adequate time was shown to be available for the operator to perform the requirea actions. Plant Operating, Abnormal, and Annunciator Response procedures were reviewed to determine that adequate guidance to perform said actions are available.
No field or control room scenaries were utilized. The review completed by the Operations Department venfied that adequate time is available to complete the required actions.
J The components of concem are e4her located witnin the Piant Control and Cable Spreading Structure or the Emergency Filtration Building. Neither of these assas are identified in the USAR APPENDIX I," Postulated Pipe Failures Outside Containment", Revision 14, as having harsh environments. If a blackout occurs flashlights, will be used to carry out the operator actions.
With these potential harsh environment adequate time is available to perform the required operator action.
d 19.In addition to the outliers addressed in Tables 5-1 and 7-1 nith respect to use of Clinch anchors, confirm that RHR Heat Exchangers E-200A and E-200B were evaluated and found ciructurally adeqt: ate in accordance with the rules and procedures given in Section 7 of the GIP.
Resoonse The detailed analysic procedures provided in Section 7 of the gip are for the following types of equipment: (1) large, flat bottom, cylindrical, vert' cal tanks and (2) horizontal, cylindrical tanks and heat exchangers on saddles. E-200A and E-200B are waist supported, vertica! cylindrical heat exchangers and are not specifically covered by either of those procedures. Per Section 7.2 of the GlP, an existing analysis which verifies seismic adequacy may be used in lieu of the GIP provided the other analysis addresses the same type of loading as the GIP.
In reference 9, the heat exchangers were subject to a detailed seismic analysis for SSE loading. The analycis empicyed detailed finite e!ement modeling of the heat exchangers. As a result of the analysis, the support of cach exchanger was upgraded. The analysis included detailed checks of 16
1 l
i Response to NRC Request for Additional Information - USl A-46 l
- .~
April 29,1997 '
Attachment 1 !
l trunnica support members, connections, supporting beams and anchorage
- for the upgraded configuration. The SCE reviewed the existing analysis and j
. concluded that the existing analysis comprehensively addressed the same i l type of loading as the GIP, with the stated exception regarding Cinch i anchors. l j 20. Prev lde computations pertaining to the seismic adequacy of the Standby Diesel
- Generator Day Tanks (T-45A and T-458). Include the calculations of the seismic
{ adequacy of their saddle supports and anchorages. i R_espons.e
- The detailed evaluation of the seismic adequacy of T-45A and T-458 is documented in calculation 91C2687-C-016, Rev. O, which is available for review at the Monticello site.
I i
?
l f
- 21. Referring to the in-structure response spectra provided in your 120-day-response to the NRC's request in Supplement No.1 to Generic 1.etter 87-02, dated May 22,1992.
provide the following information.
I a. Identify structure (s) which have in-structure response spectra (5% critical
- damping) for elevations within 40-feet above the effective grade, which are
- higher in arnplitude than 1.5 times the SOUG Bounding Spectrem.
- b. With respect to the comparison of equipment seismic capacity and seismic
- demand, indicate which method in Table 4-1 of GIP-2 was used to evaluate the seismic adequacy for equipment installed on the corresponding floors in the j structure (s) identified in item (a) above. If you have elected to use method A in i
Table 4-1 of the gip-2, provide a technical justification for not using the in-structure response spectra provided in your 120-day-reuponse. It appears that some A-46 licensees are making an incorrect comparison between their plant's safe shutdown earthquake (SSE) ground motion response spectrum and the SQUG Bounding Spectrum. The SSE ground motion response spectrum for most nuclear power plants is defined at the plant foundation level. The SQUG Bounding Spectrum is defined at the free field ground surface. For plants
,i located at deep soil or rock sites, there may not be a significant difference between the ground motion amplitudes at the foundation level and those at the ground surface. However, for sites where a structure is founded on shallow soil, the amplification of the ground rnotion from the foundation level to the ground surface may be significant, i 17 i
i * -
8 Response to NRC Request for AdditionalInformation - USl A 46 l i April 29,1997 )
j Attachment 1 1
l
- c. For the structure (s) Identified in item (a) above, provide the in structure response I
, spectra designated according to the height above the effective grade if the in- i j structure response spectra identified in the 120-day-response to Supplement No. l
- i. 1 to Generic Letter 87-02 was not used, provide the response spectra that were !
j actually used to verify the seismic adequacy of equipment within the structures l identified in item (a) above. Also, provide a comparison of these spectra to 1.5 !
times the Bounding Spectrum.
l l
RSsponse !
- - l
- a. The Reacter Building response spectra (which is also used for buildings l other than the Reactor Building and the EFT Building) and the EFT '
s response spectra have in-structure response spectra for elevation
- within 40 feet that have amplitudes higher than 1.5 times the Bounding i Spectrum.
N \
- b. For Monticello, application cf GlP method A for this site has been !
completed in accordanc9 with the requirement that the SSE is defined l l at the ground surface. This definition of the situ SSE is consistent with
- the Monticello Updated Safety Analysis Report.
i c. Attachment 2 includes the in-structure response spectra compared with
! the 1.5 times the Bounding Spectrum. These in-structure response i spectra is the same as provided in the 120 day response letter
! (reference 3).
l 1
k b I i
i 1
18 m .
Response to NRC Request for AdditionalInformation - USI A-46
, April 29,1997 Attachment 1 References
- 1. Monticello Nuclear Generatina Plant Verification of Seismic Adeguacy of.
M_echanical and Electrical Eau?oment, Unresolved Safety Issue A-46 (SOUG_).,
Northem States Power Company, November 1995.
- 2. Generic implementation Procedure (GIP). for Seismic Verification of Nuclear _
i Plant Equipment, Revision 2, Corrected,2/14/92, Seismic Qualification Utility Group.
- 3. NSP Response to GL 87-02, Supplement 1 on SQUG Resolution of USI A-46 Monticello Nuclear Plant, Thomas M. Parker (NSP) to USNRC, dated September 21,1992.
- 4. USNRC Letter " Evaluation of Monticello Nuclear Plant 120-day Response to
- Supplement No.1 to Generic Letter 87-02 (TAC No. M69460)", A. H. Hsia (USNRC) to T. M. Parker (NSP), dated December 10,1992.
- 5. Northem States Power Company,"Montice!!o Nuclear Generating Plant Updated Safety Analysis Report".
- 6. EPRI Repcrt NP-5228-SL," Seismic Verification of Nuclear Plant Equipment Anchorage (Revision 1)." Electric Power Research Institute, Palo Alto, CA, prepared by URS/ John A. Blume & Associates, Engineers, June 1991.
- 7. " Code Requirements for Nuclear Safety Related Concrete Structures", ACI 349-90, American Concrete Institute.
- 8. " Commentary on Code Requirements for Nuclear Safety Related Concrete Structures", ACI 349R-90, American Concrete institute.
- 9. Bechtel Calculation, "RHR Heat Exchangers E-200A (10-2A) and E-200B (10-2B) Support Evaluation", Project Monticello, Job No. 10040-054 Calculation.
No. 054-C1, Rev. O,9/27/84.
- 10. Seismic Safety Evaluation of the Anchorage of Safety Related Equipment at the Monticello Nuciaar Generating Station, prepared by URS/ John A. Blume &
Associates,1982.
- 11. Monticello Nuclear Generating Plant Docket No, SC-263, License No, DPR-22,
- Response to Supplement 1 to Generic Letter 87-02, Submittal of USI A-46
- Seismic Evaluation Report (TAC M6946C), November 20,1995.
l t
19
e 3
ATTACHMENT 1 Calculations Note the enclosed samples of anchorage calculations done with the ANCHOR program are part of an existing SEWS for Y82 (pages 4 through 8 of 8) and P-11 (pages A2 through A6 of A6).
4 4
i i
i
)
i i
1 l
1 j
~
gip Rev 2, Corrected,2/14/92 l Monticelio NGP Status: No
. SCREENING EVALUATION WORK SHEET (SEWS) Sheet 4 of 8 ID : YS2 ( Rev. 0 ) l Class : 4. Transformers ~
Description : DIV 2120 VDC TRANSFORMER Y83 !
~"
l Room, Row / Col: MAIN Building : EFT l Floor El. : 960.00 Manufacturer, Model, Etc.
ANCHOR Report HiuihIWeal Response Spectrum: ssp Frequency : User - 8.00 ,
Percent Damping : User - 5.00 /
l SpectralValues : _
Direction Acceleration (g's)
North - South 0.450 /
East - West 0.450 /
l Vertical 0.150 /
Angle (N-S Direction makes with the X Axis): 0.00 Combination Criteria: SRSS
, Weichts ;
Number of Weights:1 Weight X Y Z No ,
l 300.00./ 9.000 / 7.500 / 19.000 /
1 l
Forces : ,
Number of Extemal Forces: 0 l
I fAoments; Number of Extemal Moments : 0 a[lgwaMes ;
i 6mbec i Number of Anchor types :1 Tension Shear L;ttimate inter inter Saf Ultimate l Shear Coeff Coeff Fact Manufact Product Tencion N o.
1 Dia 3/8 Hi!!i Kwik-Soit 1460.00 1420.00 1.00 0.30 1.00 /
_,N) (
acosu1e:
Ultimate Stress : 4000.00 psi. /
Reduction Factor: 0.85 Wild; Allowable Stress : 30600 psi.
6
~
" Gir' Rev Corrected,2/14/92 Monticello NGP Status: No SCREENING EVALUATION WORK SHEET (SEWS) Sheet 5 of 8 _
1D : Y82 ( Rev 0 ) l Class : 4. Transformers Description : DIV 2120 VDC TRANSFORMER Y83 l Poor El. : 960.00 l Room, Row / Col: MAIN l
Building : EFT Manufacturer, Model, Etc. :
Surfacena Number of Surfaces : 1 Surface Orientation _
Direction Direction Direction Comp Comp Comp No Nx 4 Ny ,
Nz 0 000 0.000 1.000
- 11
- ' Anchor Pattem for Surface #1 4
l t
-X ll Legend for Anchor Pattems i
! Anchor Bolts: ]
Concrete Unes : ..
Concrete Points : M l l
Weld unes: [R i
GD.pmetry; i A_Dehor :
Number of Anchors : 4 f I
- ~
Monticello NGP GIP Rev 2 Corrected,2/14/92 Status: No SCREENING EVALUATION WORK SHEET (SEWS)
Sheet 6 cf 8 ID : Y82 ( Rev. 0 ) l Class : 4. Transformers Description : DIV'E 120 VDC TRANSFORMER Y83 Building : EFT l Floor EL : 950.00 ! Room, Row / Col: MAIN Manufacturer. Model. Etc. -
Anch X Y Z Surf No. Id Coord Coord Coord Id 1 1 0.000 0.000 0.000 1 2 1 18.000 ,0.000 0.000 1 3 1 0.000 15.000 0.000 1 4 ,
1 18.000 15.000 0.000 1 C.QD.CELLLIGLU
- of elements perline : 4 Number of Concrete Lines : 2 Stati End End End Sf Line Start Start Z-Coord Y-Coord Z-Coord Id Width No X-Cocrd Y-Coord X-Coord_
0.000 0.000 15.000 0.000 1 2.000 1 0.000 0.000 0.000 18.000 15.000 0.000 1 2.000 2 18.000 0.000
't Concrete Points :
Number of Concrete Points : 0 WeldL101u
- of elements perline : 4 Number of Weld Lines : 0 Determination of Reduction Factors ;
Reduction Factorinput for Anchor # 1
. Adequately Insta!!ed : Yes Embedment Length : ( 1.63 in. Min Reqd. to achieve full capacity) := 1.63 in.
Gap at Thieaded Anchor: 0.00in.
Crack Size : 0.000 in. - Cracks Affect <= 50% Bolts Essential Relays in Cabinet: No Adequate Equipment Base Strength and Structural Load Path : Yes Embedment Steel and Pads Adequately Installed : Yes Reduction Factor input for Anchor # 2 Adequately Installed : Yes Embedment length : ( 1.63 in. Min Reqd. to achie /e full capacity) := 1.63 in.
Gap at Threaded Anchor : 0.00 in.
Crack Si2.e : 0.000in. - Cracks Affect <= 50% Bolts Essential Relays in Cabinet : No Adequate Equipment Base Strength and Structural Load Path : Yes Embedment Steef and Pads Adequatefy Installed : Yes Reduction Factor input for Anchor # 3
Monticello NC# ~ ~ GIP Rev 2. Corrected,2/WB2
- Status: No
- SCREENING EVALUATION WORK SHEET (SEWS)
Sheet 7 of 8 _
ID : Y82 ( Rev. 0 ) l Class : 4. Transformers _ _ _
Description : DlV 2120 VDC TRANSFORMER Y83 Building : EFT l Floor El. : 960.00 l Room, Row / Col: MAIN Manufacturer, Model, Etc.
- Adequately Installed : Yes Embedment length : ( 1.63 in. Min Reqd. to achieve full capacity) := 1.63 in.
Gap at Threaded Anchor: 0.00 in.
Crack Size : 0.000 in. - Cracks Affect <= 50% Bolts Essential Relays in Cabinet: No Adequale Equipment Base Strength and Structural Load Pa+h : Yes Embedment Steel and Pads Adequately Installed : Yes Reduction Factor input for Anchor # 4 Adequately Installed : Yes l Embedment Length : ( 1.63 in. Min Reqd. to achieve full capacity) := 1.63 in.
i Gap atThreaded Anchor: 0.00 in. .
I Crack Size : 0.000 in. - Cracks Affect <= 50% Bolts Essential Relays in Cabinet: No Adequate Equiprnent Base Strength and Structural Load Path : Yes !
Embedment Steel and Pads Adequately Installed : Yes Reduction Factors Data Current: Yes Pattrl Anc Pall / l RM Valtr RT RN Rt. RG Rs RE RF Rc RR RP RB l No M Vall 1.00 1.00 1.00 1.00 1.00 X 1.00 1.00 1.00 1.00 1.00 1.00 )
1 1 1460.00 N/A 1.00 1.00 X 1.00 1.00 1.00 1.00 1.00 1.00 l 1420.00 N/A 1.00 1.00 1.00 1.00 1.00 X 1.00 1.00 1.00 1.00 1.00 1.00 1 2 1 1460.00 N/A 1.00 1.00 1.00 1.00 1.00 X 1.00 1.00 _ 1.00 1.00 1.00 1.00 1.00 1420.00 N/A 1.00 1.00 1.00 1.00 X 1.00 1.00 1.00 1.00 1.00 1.00 )
3 1 1460.00 N/A 1.00 1.00 1.00 '
1.00 1.00 X 1.00 1.00 1.00 1.00 1.00 1.00 1420.00 N/A 1.00 1.00 1.00 4 1 1450.00 N/A , 1.00 1.00 1.00 1.00 1.00 X 1.00 1.00 1.00 1.00 1.00 1.00 )
1.00 1.00 X 1.00 1.00 1.00 ' 1.00 1.00 1.00 1420.00 N/A 1.00 1.00 1.00 Legend:
N/A = Not Applicable Pall = Allowable Pull without Reduced inspection Vall = Allowable Shear without Reduced Inspection Pa!!r = Allowable Pull with Reduced Inspection Vallr = Allowable Shear with Reduced Inspection
= Outlier X = Reduction Factor Not Used l RT = Reduction Factor for Type of Anchorage j RN = Reduction Factor for Insta!!ation Adequacy RL = Reduction Factor for Embedment RG = Reduction Factor for Gap at Anchors RS = Reduction Factor for Spacing RE = Reduction Factor for Edge Distance RF = Reduction Factor for Concrete Strength RC = Reduction Factor for Concrete Cracks RR = Reduction Factor for Essential Relays RP r: Reduction Factor for Base Stiffness and Prying Action
MonticeIlo NGP EIPRET250Yearzam Status: No SCREENING EVALUATION WORK SHEET (SEWS) Sheet 8 of 8 ID : Y82 ( Rev. 0 ) l Class : 4. Transformers Description : DlV 2120 VDC TRANSFORfAER Y83 Building : EFT l Floor El. : 960.00 l Room. Row / Col: fAAIN fAanufacturer Model, Etc. :
RB = Reduction Factor for Base Strength and Load Path RM = Reduction Factor for Embed. Steel and Pads Anaivsfr Results :
Analysis Performed : Yes Type of Analysis : Regular Spectral Accelerations (G's)
N-S E-W Vertical Safety Factor No 0.450 0.180 0.060 20.244 1
-0.450 -0.180 -0.060 22.679 2
-0.450 0.180 0.0S0 20.244 3
0.450 -0.180 -0.0S0 22.679 4
0.450 -0.180 0.060 20.244 5
-0.450 0.180 -0.060 22.679 6
0.450 0.180 -0.060 22.679 7
-0.450 0.180 0.060 20.244 8
0.180 0.450 0.060 14.992 9
-0.180 -0.450 -0.060 16.223 10 0.180 -0.450 0.060 14.992 11
-0.180 0.450 -0.0S0 16.223 12 0.450 0.0S0 14.992 13 -0.180 0.180 -0.450 -0.060 16.223 14 0.180 0.450 -0.0S0 16.223
_15 0.060 14.992 16 -0.180 0.450 0.160 0.180 0.150 30.105 17
-0.180 -0.180 0.150 46.794 18 0.180 0.180 -0.150 46.794 19
-0.180 -0.180 0.150 30.105 20 0.180 0.150 30.105 21 -0.180
-0.180 -0.150 46.794 22 0.180 0.180 -0.180 0.150 30.105 23
-0.180 0.180 -0.150 46.794 l 24
\
i Minimum Safety Factor: 14.992 / l l
The anchorage can withstand 14.992 times greater selsmic demand J 1
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I y-l i ST A5 Earthauake :
Response Spectrum : User Frequen::y : User - 0.00 Percent Damping : User- 0.00
. SpectralValues :
l Acceleration (g's) l Direction North - South 0.23 _
East - West 0.23 0.15 l
Vertical l
Angle (N-S Direction makes with the X /wis): 0.00
' l Combination Criteria : SRSS l I
4 Weichts :
Number of Weights :1 l Weight X Y Z No 1 242.00 14.500 6.000 9.750 Forces :
Number of Extemal Forces : 0 Moments :
Number of Extemat Momerits : 0 Affowabies ;
&nchor:
Number of Anchor types :1 Tension Shear Ultimate Ultimate inter inter Saf Tension Shear Coeff Coeff Fact N o. Dia Manufact Product 1870.00 1.00 0.30 1.00 1 3/8 Other J-Bolt (90 3740.00 deg) 4 Concrett Ultimate Stress : 3000.00 psi.
Reduction Factor: 0.85
- kVeld
Allowablo Stress : 30600 pst.
Ettrfaces :
Number of Surfaces : 1 Surface Orientation a
p-s i c + s.) $
l Direction Direction Direction Comp Comp Coma No Nx NV Nz 1 0.000 0.000 1.000
- Anchor Pattem for Surface # 1
- . -, .. ,~ - . ,:
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_ my, i, _ _ h'y h SC:k5i ':~5;~b=M
- -- u%JrL3[ -..runy . ., _ m .Z.k..w Legend for An: hor Pattems 4
ggctibilBotth ,
yys =._. -------m.,
Ppf5((o[cht551Asd (I 7 .' h3 L - -
~22 NII!
- _ L .
. ' f$
ms g_ m - - = = - .y y -
Geometrv :
- An; hor; Number of Anchors : 4 i
Anch X Y Z Sud N o. Id Coord Coord Coord Id 1 1 12.000 1.500 0.000 1
- 2. 1 27.500 1.500 0.000 1
~3 1 27.500 10.500 0.000 1 4 1 12.000 10.500 0.000 1 Concrete Lines :
- of e!ements perline : 1
]
4
. >-l) 5 f tJ 5 Number of Concrete Lines : 4 Start Start End End End Sf Line Start X-Coord Y-Coord Z-Coord X-Coord Y-Coord Z-Coord fd Width No 1.030 0.000 29.000 1.000 0.033 1 2.000 1 0.000 0.000 0.000 28.000 12.000 0.033 1 2.033 2 28.000 11.000 0.003 0.000 11.000 0.000 1 2.000 3 29.000 12.000 0.000 1.000 0.000 0.000 1 2.000 4 1.000 Concrete Points :
Number of Concrete Points : 0
. Weld Unes :
- of elements per line : 4 Nurnber of Weld Lines : 0 Deformination of Reduction Factors :
a Reduction Factorinput for Anchor # 1 Embedment Length : (20.50 in. Min Reqd. to achieve full capacity) :=14.87 in.
Gap atThreaded Anchor: 0.00 in.
Reduction Factorinput for Anchor # 2 Embedment Length : (20.50 in. Min Reqd. to achieve full capacity) :=14.87 in.
Gap at Threaded Anchor: 0.00 in.
Reduction Factorinput for Anchor # 3 Embedment Length : (20.50 in. Min Reqd. to achieve full capacity) :=14.87 in.
Gap at Threaded Anchor : 0.00 in.
Reduction Factorinput for Anchor # 4 Embedment length : (20.50 in. Min Reqd. to achieve full capacity) :=14.87 in.
Gap at Threaded Anchor : 0.00 in.
Reduction Factors Data Current: Yes Anc f'alv Pallr/
RP RB RM Valir RT RN RL RG RS RE RF Rc RR No ld Vall X 0.76 1.00 1.00 X 0.93 X X X X i 1 2640.05 N/A X X 1.00 1.00 X 0.93 X X X X X 1731.28 N/A X X 1.00 1.00 1.00 X 0.93 X X X X X 2 1 2640.05 N/A X X 0.76 1.00 1.00 X 0.93 X X X X X 1731.28 N/A X X 1.00 1.00 1.00 X 0.93 X X X X X 3 1 2640.05 N/A X X 0.76 1.00 X 0.93 X X X X X 1731.28 N/A X X 1.00 1.00 1.00 1.00 X 0.93 X X X X X 4 1 2640.05 N/A X X 0.76 1.00 1.00 X 0.93 X X X X X 1731.28 N/A X X 1.00 Leoend
- N/A = Not Applicable Pall = Allowable Pullwithout Reduced Inspection Vall = Allowable Shear without Reduced inspection I
l
p ti Spw 5 Palir = Allowable Pullwith Reduced inspection Valir - = Allowabie Shear with Reduced inspection
= Outlier X = Reduction Factor Not Used RT = Reduction Factor for Type of Anchorage RN = Reduction Factor for installation Adequacy RL = Reduction Factor for Embedment RG = Reduction Factor for Gap at Anchors RS = Reduction Factor for Spacing RE = Reduction Factor for Edge Distance RF = Reduction Factor for Concrete Strength RC = Reduction Factor for Concrete Cracks RR = Reduction Factor for Essential Relays
_RP = Reduction Factor for Base Stiffness and Prying Action RB = Reduction Factor for Base Strength and Load Path RM = Reduction Factor for Embed. Steel and Pads I
Analysic Reg _qlts_;
)
Analysis Performed : Yes a
j Type of Analysis : Regular
' Spectral Accelerations (G's)
E-W Vertical Safety Factor i No N-S 0.092 0.0S0 74.013
- 1 0.230
-0.092 -0.060 91.678 i 2 -0.230 0.092 0.060 85.371 l3 -0230
-0.092 -0.060 86.383 j 4 0.2.30
-0.092 0.060 74.013
- 5 0.230 0.092 -0.060 91.678 j 6 -0.230 0.092 -0.060 86.383 7 0.230 0.092 0.0S0 85.371 8 -0.230 0.230 0.0S0 58.823 9 0.092
-0.230 -0.060 64.719 10 -0.092
-0.230 0.060 58.823 11 0.092
] -0.060 64.719 2 12 0.092 0.230 0.230 0.060 58.892 13 -0.092
-0.230 -0.060 63.756 14 0.092 0.230 -0.0S0 63.75S
_15 0.092
-0.230 0.060 58.892 16 -0.092 j f 0.092 0.150 91.649 17 0.092 !
-0.092 -0.150 170.807 18 -0.092 l 0.092 -0.150 162.194 19 0.092
-0.092 0.150 113.906 20 -0.092 0.092 0.150 113.906 21 -0.092
-0.092 -0.150 162.194 22 0.092 j
-0.092 0.150 91.849 23 0.092 l 0.092 1 50
-0.__
170.807 24 -0.092 l !
I Minimum Safety Factor: 58.823
j ATTACHMENT 2 l in-structure Re' pense Spectra ;
l l
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Monticello NGP Reactor Building 5% in-Structure Response Spectrum Elevation 935 Horizontal Direction Envelope 1.20E+00 -
1.00E+00 - - - -
1.5 x BS ,
8.00E - - - - - - - - - - - - - -- -- - - - - - - - - - - - - - - - - - - - - - - - - -
S C
2 3 6.00E --- -- - - -- - - - - - - - - - - - - - - - -
2 8
o '
4 4.00E -- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - --- - - - - - - - - -
i 2.00E '
0.00E+00 . . . . . .
O 5 10 15 20 25 30 35 Frequency (Hz) ,
_ _ _ _ _ - _ _ _ _ _ _ ._.---__.-_.-_-__.__________-_.-_-.___.___-_____-.__-____.-_.___-___.._m.-___.______.___.m-_..___._-_--_m_-__-..m. -
. _ _ . . _ ._ . . . _ _ _ . . _ _ _ . _ . _ . _ . . . . m . ,__
s Monticello NGP Reactor Building 5% In-Structure Response Spectrum Elevation 963 Horizontal Direction Envelope 1.40E+00 -
1.20E+00 -
i 1.5 x BS 1.00E+00 -
3 8.00E - - - - - - - - - - - - - - -
8
=
8 2
a>
8 6.00E ---- - - - - - - - - - - - - - - - - - - - - -- ---
4.00E -- -- - - -- - ----
2.00E ,
0.00E+00 , , , , , .
0 5 10 15 20 25 30 35 i
Frequency (Hz) '
I i
O
's Monticello NGP Reactor Building 5% in-Structure Response Spectrum Elevation 986 Horizontal Direction Envelope 1.60E+00 i
1.40E+00 - - - - - --- -- -- - - - - - - - - - - - - - - - - - - -
1.20E+00 - ---
-1.5' x BS i 1.00E+00 -
E c
2 3 8.00E 2 8
M i 6.00E -- - - - - - - - - - - - - - - - - - --- -- - - - - - - - - - - - - - -- - - - -
i 4.00E -- - - - - - - - - - - - - - - - - - - - - -
i 2.00E -- - - - - - - - - - - - -- --- - - - - - - - - - - - - - - - - - - - -
i 1
0.00E+00 . . . . . .
0 5 10 15 20 25 30 35 Frequency (Hz)
.. . . - . , . ~ . . . . .-. . . . - - . = . .. .- ... . ~ .. . . _..- .. _. .-
l ,
t +
Monticello NGP EFT Building 5% In-Structure Response Spectrum Elevation 932 Horizontal Direction Envelope 1.60E+00 -
1 1.40E+00 - -- -- -- - - - - - - - - - - - - - - - - - - - - - - - - - ~ - - - - - - - - - --
1.20E+00 - -- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
1.00E+00 - - - - - - - - - - - --- - - - - - . ..1.5 x B S .___ _ . . _ _ _ _. _ _ _ _
2 !
c >
.9 3 8.00E -- - - - - - - - - - -
.2 8
o 6.00E - - - - - - - - - - - - - - - - - - - - - - - - - -
4.00E -- - --- - - - - - -
2.00E -- --- - - --- - - - - - - - - - - - - - - - - - - - -
0.00E+00 , , , , , ,
0 5 10 15 20 25 30 35 Frequency (Hz)
. e
_ _ . . . _ _ . _ _ _ _ _ _ _ _ . _ . . _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ . _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _.m - -
Monticello NGP EFT Building 5% In-Structure Response Spectrum Elevation 944 Horizontal Direction Envelope ,
2.00E+00 -
1.80E+00 - - - - ~ - - - - - - - - - -- --- -- ---- - - - - --
1.60E+00 - - -- -- - - - - - - - - - - - - - - - - -
1.40E+00 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
1.5 x BS g 1.20E+00 - - - - - - - - - - - - - - - - - - - - - - -- - - - - -
'c
.e 3 1.00E+00 - - - - -
.2 8
u
< 8.00E - -- -
6.00E 4.00E - - -
2.00E - -- --- - - - - - - - - - - - - -
0.00E+00 , , , , , ,
0 5 10 15 20 25 30 35 Frequency (Hz)
i
. i J
Monticello NGP EFT Building 5% in-Structure Response Spectrum Elevation 960 Horizontal Direction Envelope ;
i 1
2.50E+00 -
t 2.00E+00 - ---- --
1.5 x BS g 1.50E+00 - --- - - - - - - - - --- - - - - - - - - - - - - - - - - - --- --
8
=
2 '
.92 0
o 4 1.00E+00 - - - - - - - - - - - - - - - - -
5.00E - --- - - - - - - - - -
0.00E+00 , , , , , ,
0 5 10 15 20 25 30 35 Frequency (Hz) e l
b
_.__ .. . m. _.___. _ _ . _ . . _ ..-____ _ _ _ . _ _ _ _ . . . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . - ,,e-. . -, _