ML20247M172
| ML20247M172 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 05/15/1998 |
| From: | Hammer M 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 9805260190 | |
| Download: ML20247M172 (13) | |
Text
, _ _ _ _ _ _ _ _ _ - - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
l . .,
Northern States Power Company M " ' "
Monticello Nuclear Generating Plant 2807 West County Road 75 Monticello, MN 55362 l
May 15,1998 j
U S Nuclear Regulatory Commission GL 87-02 Attn: Document Control Desk USl 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 Ssfety issue A-46 (TAC NO. M69460)
By letter dated February 27,1998, the NRC staff requested additional information on Resolution l of Unresolved Safety issue (USI) A-46, Seismic Evaluation Report (TAC NO. M69460). This letter provides the NSP response to this request.
l This submittal contains no new NRC commitments and does not modify any prior commitments.
Please contact Joel Beres at (612) 295-1436 if additional information is required.
Michael F. Hammer Plant Manager Monticello Nuclear Generating Plant c: Regional Administrator-lll, NRC NRR Project Manager, NRC Resident inspector, NRC State of Minnesota Attn: Kris Sanda J Silberg Attachments: Affidavit to the US Nuclear Regulatory Commission Attachment 1 Request For Additional Information Pertaining To USI A-46, Seismic Evaluation Report Monticello Nuclear Generating Plant Attachment 2 Revised Pages to Appendices C and D of NSP Submittal dated November 20,1995 9805260190 980515 \
PDR ADOCK 05000263 P PDR k i sness e acessasouowuS98 DOC y
\ .
UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 l Response to Request for Additional Information on the Resolution of Unresolved Safety Issue A-46 (TAC NO. M69460)
Northern States Power Company, a Minnesota corporation, hereby provides the information I requested by the USNRC in letter dated February 27,1998, titled "Monticello Nuclear Generating Plant - Request for Additional Information Related to Unresolved Safety issue (USI) A-46, Seismic Evaluation Report. (TAC No. M69460)."
l This letter contains no restricted or other defense information.
NORTHERN STATES POWER COMPANY l
By A4%MIA/
MichdelV. Hammer Plant Manager Monticello Nuclear Generating Plant On thisl3 day of h/ l$ 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 Northern 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. !
l k
h' I COLLEEN A.HANNON ,
l- .
NM . ~ .
Colleen A. Hannon.
Notary Public - Minnesota Sherbume County My Commission Expires January 31,2000 i
ATTACHMENT 1 Request For Additional Information Pertaining To USl A-46, Seismic Evaluation Report Monticello Nuclear Generating Plant l
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- 1. Describe the reviews that were performed to determine if any local operator actions required to safely shut down the reactor (i.e., implement the SSEL [ safe shutdown equipment list]) could be affected by potentially adverse env ronmental conditions (such as loss oflighting, excessive heat or humicuty, or in-plant barriers) resulting from the seismic event. Describe how staffing was evaluated and describe the reviess that were conducted ta ensure operators had adequate time and resources to respond to such events.
NSP Response:
In the process of responding to this letter, NSP reviewe.d the safe shutdown equipment that requires local operator action after the seismic event. NSP has now determined that the affected equipment is in fact seismically qualified, and consequently no local operator actions are necessary to safely shutdown the reactor.
The subject equipment includes the 125 VDC and 250 VDC divisional battery chargers. The original review of the seismic qualification of this eauipment did not locate sufficient information to demonstrate qualification of certain relays. Without this information, it was conservatively decided to assume that these relays would reposition during the seismic event. Local operator action was required to reset these relays. NSP has recently determined that the chargers are seismically qualified based on the generic equipment ruggedness spectrum contained in Appendix C of EPRI NP-5223-SL Rev.1, Generic Equipment Ruggedness of Power Plant Equipment. In addition, seismic test reports were located for the 250 VDC chargers.
By Attachment 1 of a submittat dated April 29,1997, NSP provided a response to a similar question posed by the staff (Question 18). Given the above, NSP requests that the staff consider our previous response to Question 18 superseded. By letter dated November 20,1995, NSP submitted a Safe Shutdown Equipment List as part ofits USl A-46 Resolution Report. The battery charger portion of this list contains data fields which include information on qualification and local operator actions. The affected pages have been revised and are provided as Attachment 2 to this letter.
- 2. Have you identified any control room structures that could impact the operator's ability to respond to the seismic event during your review? Such items might include but are not limited to: MCR [ main contml room] ceiling tilcs, nonbolted cabinets, and nonrestrained pieces of equipment (i.e., computer keyboards, monitors, stands, printers, etc.). Describe how each of these potential sources ofinteractions has been evaluated and describe the schedule forimplementation of the finalmsolutions.
NSP Response:
All control room structures and components were considered for potential seismic interaction and for the potential to impact the operator's ability to respond to a seismic event.
The method used for evaluating these potential sources of seismic spatial interaction is described in GIP-2, Part 11, Section 4.5 and Appendix D. After performing this
$n4/90 eb J M.lCENSE\SQUG\RMS98-8 DO 2
review, NSP concluded that all of the control room structures and components passed the GIP screening criteria except for the ceiling lights which could have adversely impacted the operator's ability to respond to an SSE. The ceiling lights have been modified and are now seismically restrained to prevent dropping during a seismic event.
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- 3. Describe what reviews were performed to determine if anylocal operator actions were required to reposition " bad actor relays." For any such activities describe how adverse environmental conditions (such as loss oflighting, excessive heat or I hur'dity, orin-plant barriers) resulting from the seismic event were analyzed and dispositioned. Describe how staffing was evaluated as well as the reviews that were conducted to ensure operators had adequate time and resources to respond to such l events.
l f NSP Response:
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The " bad actor" description of relays is synonymous with the " low ruggedness" term used in EPRI NP-7148, Procedure for Evaluating Nuclear Power Plant Relay Seismic Functionality. Appendix E of this document contains a list of low ruggedness relays.
l These relays have low seismic ruggedness or demonstrated sensitivity to high i- frequency vibration.
. A review of system logic diagrams was performed to determine the expected position or state of equipment affected by low ruggedness relay repositioning. A comparison .
l of the relay endpoint state to the required state was performed, and conflicts were evaluated. No local operator actions to reset low ruggedness relays were identified.
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- 4. Describe which of the operator actions associated with resetting SSEL equipment affected by postulated relay chatter are considered to be routine and consistent with the skill of the craft. If not considered skill of the craft, what training and operational aids were developed to ensure the operators willperform the actions required to reset affected equipment?
NSP Response:
l As described in the response to Question i herein, there are no local operator actions required for operators to safely shutdown the reactor following a seismic event. Certain annunciators, however, which do not involve safe shutdown of the reactor, are included within the SSEL by virtue of their interaction with safe shutdown circuits. NSP has determined that some local annunciators may alarm and cause a sealed-in control room annunciation. These annunciators are identified by an OA (Operator Action) field in the Relay Evaluation Report. If necessary, after the reactor is safely shutdown, the local annunciators could be acknowledged, and the annunciator panel reset. This action involves a standard alarm response procedure, and the action is routine and consistent with the skill of the craft.
1 5/14#9e job J1LICENSBSQUGMAl$e64 DOC 3
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- 5. If the alarms associated with " bad actor relays" are expected to annunciate during the seismic event, would the operators then have to respond to those annunciators and review the annunciator response procedures for potential action? How would those additional actions impact the operators' ability to implement the Normal, l Abnormal, and Emergency Operating Procedures required to place the reactor in a safe shutdown condition?
NSP Response:
As described in Section 3.5.3 of EPRI Report NP-7148,50 to 100 or more alarms are expected to annunciate following an earthquake which causes a turbine trip and reactor scram. In addition, there may be several earthquake-induced, spurious alarms resulting from such events as water sloshing in tanks, oil sloshing in 5 transformers, actuation of vibration protective instrumentation on rotating equipment, and contact chatter of relays.
When these alarms occur, the operator will clearly be aware that the plant has tripped. Plant procedures and operator training require that operators respond to the turbine trip and reactor scram. This is accomplished by confirming the reactor scram and turbine trip and checking important levels, temperatures, pressures, flows, and electrical switching resulting from associated power transfers. These confirmatory checks will take longer than a minute to complete during which time the operators will not be responding to alarms unrelated to the scram response procedures. The earthquake motion is assumed to last less than a minute, and the causes of the spurious alarms will have cleared during tus initial response period.
The NRC staff and SQUG representatives have discussed this topic in detail. The results of an August 3,1988 meeting are summarized in Section 3.5.3 of EPRI NP-7148. The following statement is included in page 3-12 of this section.
"Accordingly, there appear to be no reasonable bases or evidence which would suggest that spurious alarms resulting from an earthquake may lead to abnormal operator responses. Therefore, special operating procedures or relay evaluation actions to address potential spurious alarms are not considered warranted and relays affecting alarms need not be seismically adequate."
The NRC staff accepted the relay functionality review procedure summarized in GlP-2 and described in detail in EPRI NP-7148 by Supplemental Safety Evaluation Report No. 2 on GIP-2. Based on the above, NSP does not consider it necessary to perform any additional reviews of spurious alarms from low ruggedness relays or from other causes as a result of a seismic event.
Alarms associated with low ruggedness relays are not specifically identified to the operators. The operators are expected to respond to all alarmed annunciators.
Recommended operator responses to alarmed conditions, as identified in C.6 Annunciator Response Procedures-General Instructions, require verification by independent confirmatory indications if available. These independent confirmatory indications receive a signal from a different transmitter or source than that signal uwm , accusecouomaisos4 ooc 4
which alarmed the annunciator. If no independent confirmatory indications are available, the operating crew would determine the validity of the annunciator using other means. The annunciators associated with low ruggedness relays would not be valid alarms, and no additional actions would be necessary such that the ability to place the reactor in a safe shutdown condition would not be adversely affected.
- 6. To the extent that Normal, Abnormal, and Emergency Operating Procedures were modified to provide plant staff with additional guidance on mitigating the A-46 Seismic Event, describe what training was required and provided to the licensed operators, nonlicensed operators, and other plant staff required to respond to such events.
NSP Response:
Normal, abnormal, and emergency operating procedures were reviewed to assure that procedural guidance was adequate to assure safe shutdown for the seismic l event. No changes to these procedures were required, and no additional training was required.
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1 ATTACHMENT 2 I
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Revised Pages to Appendices C and D of NSP Submittal dated November 20,1995 1
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MNGP A-46 Fitri Report Rety Evaluation Report SOUG Relav Review Functional Screenina Results p, 34 Relav Designation Relav Tvoe Egngj w
Ruouedness Resolution (SAT)
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SSEL_LineMumber: 4015 Plant Sy_stemLCRM Component /Subystem: CV-3-32B I POS. f. C. NAMCO EA180 CV-3-32B No CA POS. l. O. NAMCO EA180 CV-3-32B No CA
_SSEL_LineMumberL4014 Plant Systeml_CRH Component /Subystem: CV-3-32_C POS. l. C. NAMCO EA180 CV-3-32C No CA l POS. l. O. NAMCO EA180 CV-3-32C No CA S_SEL_Une_ Number: 4016 Plant _SystemLCRH Qomponent/Subystem: CV-3-32D POS. l. C. NAMCO EA180 I l CV-3-32D No CA l POS. l. O. NAMCO EA180 CV-3-32D No CA l
SSEL_Line Number: 400.3 Plant _ System:_CRH Component /Subys_ tem:_CV-3-33A 1
POS.1. C. NAMCO EA180 CV-3-33A No CA POS. l. O. NAMCO EA180 No CA C(/-3-33A SSEL._LineEumber; 4004 Plant System: CRH_ Component /S_ubystem: CV-3-33B POS. l. C. NAMCO EA180 CV-3-33B No CA POS. l. O. NAMCO EA180 CV 3-338 No CA i i
SSEL LineEumberL.4011 Plant _ System: CRH Component /Suby_stemj_CV-3-33C POS. l. C. NAMCO EA180 CV-3-33C No CA >
POS. l. O. NAMCO EA180 CV-3-33C No CA SSE.L Line Number: 4012 Plant System: CRH Component /Subystem: CV-3-33D
! POS. I. C. NAMCO EA180 CV-3-33D No CA POS. l. O. NAMCO EA180 CV-3-33D No CA I
SSEL Line Number: 5002 Plant System: 125 Component /Suby_ stem: D10 Battery Charger Controls Exide US 130-3-50 D10 No GERS l SSEL_Line_NumberL5003 Plant _ System: 125 Component /S_ubystem:_D20 Battery Charger Controls Exide US 130-3-50 D20 No GERS l SSEL_Line_NumberL_500.5 Elant_SystemL125 Componerit/Subystem;_D40 Battery Charger Controls Exide US 130-3-50 D40 No GERS l
.SSEL_ Lina Nurnber: 6022 Plant _ System;_250 Component /SubystemLD.52 l Battery Charger Controls C&D Batteries ARR130HK150F D52 No GERS l l
.- MNGP A-46 Final Report Relay Evaluation Report SOUG Relav Review Functional Screenina Results p [ofg4 L9E.
Relav Designation Relav Tvoe Eitngj Ruaoedness Resolution (SAT)
SSEL_Line Number;_6023 Plant System;_250 Component /Subystem: D53 Battery Charger Controls C&D Batteries ARR130HK150F D53 No GERS l
.SSEL_Line Number;_6024 Plant SystemL250 Component /Subystem:_D.54 Battery Charger Controls C&D Batteries ARR130HK150F D54 No GERS l
_SSEL Line_ Number;__6 017 Plant Systam:_250 Component /SubystemLD70 Battery Charger Controls C&D Batteries ARR130HK150F D70 No GERS l SSEL Line Number: 6018 Plant System: 250 Component /SubystemLD80 Battery Charger Controls C&D Batteries ARR130HK150F D80 No GERS l
.SSEL_Line NumberL6019 Plant SystemL250 Component /Subystem:_D90 Battery Charger Controls C&D Batteries ARR130HK150F D90 No GERS l l
SSEL Line Number: 2154 Plant SystemLRHR Component /Subystem:_DPW-10_-130.8 K2 P&B MDR 163-1 C292 No GERS K3 P&B MDR 163-1 C292 No GERS MEL Line NumberL7_151C Plant System: DOL Component /Subystem: fP_M_(11 DG1 FPR EMD 8299025 C91 CR
_SSEL_Line. Number: 7157D Plant System: DOL Component /SubystemLEPM (12 DG)
FPR EMD 8299025 C92 CR SSEL_Line_ Number; 71.57A Plant _ System: DOL Component /Suby. stem: FTM-1_(j1 DG)
FTC1 8474707 C93 No CA SSEL Line Number: 7156 Plant System: DOL Component /Subystem: FTM 1 (12_D_G)
FTC1 8474707 C94 No CA
.SSEL Line Number: 7157B Plant System: DOL Component /Subystem: FTM-2_(1_1_DG)
FTC2 8474707 C93 No CA SSEL L!.ne_ Number 1_7157 Plant _ System: DOL Component /Subystem: FTM-2112 D.G)
FTC2 8474707 C94 No CA SSEL Line Number: 7045 Plant System: DGN Component /SubystemLG-3A l 14A-K11 A GE 12HFA151 A2F C32 No GERS 14A-K22A GE 12HFA151 A2F C32 No GERS L - _ _ _ _ _ _ _ _ _ _
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