RBG-31174, Application for Amend to License NPF-47,revising License Condition 2.C.14,Attachment 5 Re Emergency Response Capabilities,Requiring Implementation of Reg Guide 1.97 Mods to Neutron Monitoring Sys
| ML20245L817 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 06/28/1989 |
| From: | Deddens J GULF STATES UTILITIES CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 RBG-31174, NUDOCS 8907060100 | |
| Download: ML20245L817 (19) | |
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, a b GULF; STATES UTELETIES COMPANY
' RIVER 8 TEND STATfDN POET OFFICE BOX 22D ST FRANCISVILLE LOUIS 1ANA 70776 AREA CODE 604 635 6094 - 346 8661 l
June 28,1989 RBG-31174
! File Nos. G9.5, G9.42 U.S. Nuclear Regulatory Commission Document Control Desk =
Washington, D.C. 20555 h
Gentlemen:
River Bend, Station - Unit 1 Docket No. 50-458'
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. Gulf States Utilities (GSU) Company hereby files an application to amend the
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River Bend Station - Unit 1 Facility Operating License NPF-47, pursuant to 10CFR50.90. This application is filed to revise' License Condition 2.C.14, Attachment 5, Emergency Response Capabilities. Item 3, which requires
-implementation of Regulatory Guide 1.97 modifications to the neutron 4: monitoring system. GSU previously filed' an amendment request to require implementation after the NRC completed its evaluation of the' Boiling Water
. Reactor Owners' Group'(BWROG) report (NEDO-31558) submitted in April 1988 in response to this issue.- As a result of the extended NRC review of the BWROG report and the currently required modification completion date of January 1.
1991,. insufficient time remains to incorporate NRC findings on the BWROG report into a system design change prior to the required completion date.
Therefore, 'GSU. requests an extension 'to allow completion of the final resolution to this issue' prior tc. its required implementation. The Attachment to' this letter includes the proposed revisions to NPF-47 and justifications for this proposed change.
GSU is prepared to promptly implement actions to resolve this issue upon completion of the NRC-BWROG efforts. Because no further GSU specific action remains prior to the generic resolution, GSU requests the previous commitment to -provide quarterly u, dates be revised to attain resolution after the NRC issues an SER on the BWRDG report. This submittal is GSU's second quarter report for 1989.
Your prompt attention to this application is appreciated.
Sincerely, fk-J. C. Deddens 4
Senior Vice President A. d River Bend Nuclear Group JEB/E4E/RJK/196 h Attachments p3 I
.. 8907060100 890628 '
i PDR ADOCK 05000458 i P FDC
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l cc: .U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000
' Arlington, TX 76011 1
Mr.. Walt Paulson, Project Manager U.S. Nuclear Regulatory Commission Washington, D.C. 20555-NRC Resident Inspector P. 0.-' Box'1051
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St. Francisville, LA- 70775 o Mr. William H. Spell, Administrator Nuclear Energy Division Louisiana Dept. of Environmental Quality y P. O. Box 14690 Baton Rouge, LA 70898
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l UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION STATE OF LOUISIANA )
PARISH OF WEST FELICIANA )
Docket No. 50-458 In the Matter of )
GULF GTATES. UTILITIES COMPANY )
(River Bend Station - Unit 1) j AFFIDAVIT J. C. Deddens, being duly sworn, states that he is a Senior Vice President of Gulf States Utilities Company; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the documents attached hereto; and that all such documents are true and correct to the best of his knowledge, information and belief.
t J. C. Q66 dens ~ hv Subscribed and sworn to before me, a Notary Public in and for the State and Parish above named, this 48b ' day of
()A14_L , 1939 . My Commission expires with Life.
8 0 AA &s . LULY Cla'udia F. Hurst Notary Public in and for West Feliciana Parish, Louisiana
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ATTACHMENT GULF STATES UTILITIES COMPANY f RIVER BEND STATION DOCKET 50-458/ LICENSE NO. NPF-47 REGULATORY GUIDE 1.97 - Neutron Flux Monitoring (89-002)
LICENSING DOCUMENT INVOLVED: Facility Operating License NPF-47 ITEM: License Condition 2.C.(14)
Emergency Response Capabilities Attachment 5. Item 3 REASON FOR REQUEST:
A _ change is being requested in accordance with 10CFR50.90 to revise the implementation date for modifications to the River Bend Station (RBS) neutron flux monitoring system (NMS) as currently required in License Condition 2.C(14) Attachment 5, Item 3 of Facility Operating License NPF-47. This proposed change revises the implementation date for NMS modifications from prior to January 1, 1991 to prior to restart from the next refueling outage starting after 18 months from the date of receipt of the NRC Staff safety evaluation report on the Boiling Water Reactor Owners' Group (BWROG) licensing topical report (NED0-31558). Approval of this proposed change will allow operation with the currently installed system until this issued is resolved. The proposed change is required to allow start-up following the third refueling outage currently scheduled to begin during September 1990. The BWROG submitted a report to the Staff on April 1, 1988 regarding implementation of Category I neutron flux monitors as addressed in Regulatory Guide (RG) 1.97. The NRC Staff's review and safety evaluation report (SER) was originally scheduled to be completed by the fourth quarter of 1988. As a result of the delay in resolving this issue, GSU may no longer be able to evaluate all of the available systems and install certain options without extending presently scheduled outages or be forced into an unscheduled outage.
GSU finds it desirable to await issuance of the NRC Staff's SER until installation of any revised NMS is pursued further at RBS. Installation of a modification without due consideration of the technical arguments forwarded by the BWROG would be inconsistent with the NRC policy encouraging common solutions. GSU believes the BWROG 1etter to be technically sound and likely to pursuade the Staff that alternatives to the RG l'.97 guidance are practicable.
In addition, RBS has found its current license condition requirements to be significantly in excess of the requirements of other plants similarly designed. RBS has initiated reviews of other design alternatives and followed industry activities in a good faith effort to fulfill its current license condition requirements. However, GSU believes that implementation of a revised NMS at this time could result in undue hardship, diversion of resources, and potentially unnecessary costs in excess of those contemplated when the license condition was issued.
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DESCRIPTION:
The intent of RG 1.97 is- to. ensure that all light-water-cooled nuclear-power plants are instrumented as necessary to measure certain prescribed.
variables and systems during and after an accident. A portion of RG 1.97 provides design and qualification guidance for NMS instrumentation.
Attachment . 5,. Item 3 to NPF-47 currently requires that GSU modify systems as required to meet RG 1.97 guidance before restart from the next refueling outage starting after 10 months-from the date of receipt of'the NRC Staff SER on the BWROG report'but no later than January 1, 1991. With the present . license condition requirements. and the . third refueling outage scheduled to begin' September 1990, GSU would be required' to issue the bid
. specification in May 1989. This restriction is based on the following
. constraints:
Six months-needed for: Vendor preparation of bids (2 months); bid review by GSU to ensure vendor compliance with specification requirements (3 months); and purchase order preparation and issuance (1 month)'. ' A minimum 10 month equipment lead time must be allowed to ensure' equipment receipt and inspection prior to the currently scheduled . third refueling outage starting date of September 1990. Two months will'be required for the installation and acceptance testing after the system is received on site to allow the subsequent startup.
.The. total period from initial specification release to completed installation of a system is therefore 18 months as identified in the proposed revision to the License.
The original 10 month period was based on schedule and regulatory conditions at that time. The assumptions which contributed to the .
-previous delay period were the NRC was expected shortly, possible l modifications were minimized to achieve compliance with the guidance in lieu of optimizing for plant operation and with resolution expected soon, a minimum number of vendors and options were considered.
GSU's efforts ftom NRC approval of the previous extension request to the present'have included: Submitting quarterly reports to the NRC as to the status of action with respect to procurement of a system meeting RG 1.97
. requirements; maintaining communication with both the BWROG and the NRC as to the status of the NRC's response to NED0-31558; and maintaining a bid specification for the purchase of a system meeting RG 1.97 requirements ready for issuance. On April 1, 1988, the BWROG provided to the NRC a licensing topical report (LTR), NE00-31558, to the NRC which described functional design criteria for the post accident NMS and provided appropriate justifications. In support of this position, GSU conducted a plant specific evaluation using the criteria provided in the LTR. This plant specific evaluation is included in this submittal as Enclosure II.
As discussed in the enclosure, the current RBS design meets all criteria provided in the LTR. Should complete approval be obtained, the complete specification could be cancelled and no additional procurement action be required. Denial of the alternate requirements regarding Range, Equipment Qualification, Seismic Qualification or QA requirements, which are primarily the result' of defining the critical scenario for NMS use as an anticipate transient without scram event, would result in significant revisions to the corresponding sections of the bid specification. Revision of 'the bid specification and obtaining the necessary reviews to reflect Page 2 of 5
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. s modified requirements is estimated to. require one month. An alternative L allowed by this license condition is to obtain NRC approval of an alternate design.
In addition lto the BWROG LTR 'and an RBS plant specific comparison, GSU has l- followed. industry development of equipment designed to meet stringent RG 1.97 requirements in-this area for several . years. .Several options have been identified. .However, concerns have been recognized regarding the ability-of these systems to comply with all criteria' of RG 1.97 or regarding installation and; operational' considerations. GSU is continuing to pursue resolution ~to these' concerns.to establish an acceptable alternate system installation. However, delivery constraints will require a purchase order to be placed between June and September, 1989, depending on the option, to ensure delivery and . completion of the final design for installation during the. third refueling outage.
To procure, design.and install a NMS prior to receiving the final NRC position on the BWROG LTR could result in undue hardship and unnecessary costs. The effect'of the NRC's response to the LTR BWROG LTR on the bid specification _is greatly dependent on the specific sections approved.
In conclusion,.'it is .GSU's position that based on alternate criteria 1 presented. in the BWROG letter dated April 1, 1988, the current RBS NMS
, ' meets the functional safety intent of RG 1.97. Continued operation.with the currently' installed NMS is acceptable based on the plant specific evaluation which shows that the existing NMS will continue to provide appropriate information to the operator to assure that the proper actions will, be taken to respond to events addressed by the Emergency Operating Procedures. GSU will continue to work with the BWROG and NRC St=ff to resolve this license condition issue.
SIGNIFICANT HAZARDS CONSIDERATIONS:
In accordance with the requirements of 10CFR 50.92, the following discussion is provided in support of the determination that no significant hazards are created or increased by the changes proposed in this amendment request.
- 1. - No-significant increase in the probability or the consequences of an accident previously evaluated results from this proposed change because:
There is no change in system design or operation. The license condit' ion currently requires upgrade of NMS during the third refueling outage. This proposed change will allow operation with the currently installed NMS which has been found to comply with all criteria proposed in the BWROG letter. This system is required to provide neutron flux indication and is not postulated to initiate any accidents. .e NMS is used .to verify reactor shutdown as part of the Emergency Operating Procedures (E0Ps). The use of neutron monitoring in the E0Ps is conservative in that, if it is not available, actions are specified which will lead to safe shutdown without the system. The requirements of RG 1.97 concerning neutron monitoring are additions to the existing system abilities. Therefore, delay in upgrade to RG 1.97 requirements will not significantly increase the probability of an accident and Page 3 of 5
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t would not lead to an increase in the consequences of an accident as defined in the safety analysis because of the conservative E0P actions.
- 2. This proposed change will not create the possibility of a new or different kind of accident than any previously evaluated because:
The current system has been evaluated using alternate criteria proposed in NED0-31558 and found acceptable for continued operation. This change does not involve any changes to design or operation. In addition, the neutron monitoring system is not postulated as the initiator of any accidents. Therefore, no new or different accidents are created.
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- 3. This proposed change does not involve a significant reduction in the margin of safety because:
i Design, function, and operation of the existing NMS remain the same.
There is no specified " margin of safety" associated with this system as used in RG 1.97 other than to assure reactor shutdown following a transient or accident. E0P actions are conservative with respect to the use of the NMS for verification that the reactor is shutdown. When not available during an accident or transient scenario, actions are specified which will lead to safe reactor shutdown. Because these actions lead to a safe plant condition (reactor shutdown), the margin of safety is not reduced. In addition, this request does not result in a reduction to the margin of safety as defined in the bases of the RBS Technical Specifications.
Because the present RBS design meets all criteria provided in the BWROG License Topical Report, NED0-31558, which was submitted to the NRC April 1, 1988, as supported by the plant-specific evaluation attached, extension of the implementation date for a NMS meeting RG 1.97 guidance is justified.
This proposed extension allows the NRC to complete their evaluation of the report, which provides an alternative design as allowed by the current license condition to comply with the RG 1.97 requirements. In addition, GSU will be able to better plan its resource utilization to address the NMS pursuant RG 1.97 after the Staff's SER is received.
REVISED LICENSE CONDITION: j I
The requested revision is provided in Enclosure I.
SCHEDULE FOR ATTAINING COMPLIANCE:
RBS is currently in compliance with the applicable section of the license condition. As discussed in this amendment request, the proposed change is requested promptly to allow continued operation without affecting the presently scheduled outages. If the proposed change is not granted, ,
delivery constraints on the required equipment regrire a bid specification to be released immediately to allow the maximum nur' er of options to be considered.
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4 4 NOTIFICATION OF STATE PERSONNEL:
A copy of this amendment application has been provided to the state of Louisiana, Department of Environmental Quality - Nuclear Energy Division.
ENVIRONMENTAL IMPACT APPRAISAL:
Gulf States Utilities Company (GSU) has reviewed the proposed license amendment against the criteria of 10CFR51.22 for environmental considerations. As shown above, the proposed changes do not involve a significant hazards consideration, nor increase the types and amounts of effluents that may be released offsite, nor significantly increase individual or cumulative occupational radiation exposures. Based on the foregoing, GSU concludes that the proposed changes meet the criteria given in 10CFR51.22(c)(9) for a categorical exclusion from the requirement for an Environmental Impact Statement.
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r ENCLOSURE'I Proposed License' Condition Change l
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ATTACHMENT 5 TD-hPF 47 EMERGEhCY R N CAPABILITIES l
GSU shall complete the following requirements of NUREG-0737 Supplement No. I on the schedule noted below:
- 1. Actions and schedules for correcting all human engineering discrepancies (HEDs) identified in the " Detailed Control Room I Design Review Summary Report" dated October 31, 1984 and Supplements dated May 14. June 12, 1985, and July 31, 1985, shall be implemented in accordance with the schedule conunitted to by GSU in the sununary report and supplements and accepted by the NRC staff in Section 18.1 of SSER 3.
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- 2. Prior to startup following the first refueling or outage )GSU shall implement modifications (installation upgrade for those. items listed below consistent with the guidance of Regulatory Guide 1.97, Revision 2.unless prior approval of an alternate design of these items is granted by the NRC staff. These items as listed in GSU's letter of June 24, 1985 are:
coolant level in the reactor; a)))
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dp suppression pool water level; drywell atmosphere temperature; primary system safety relief valve position; eJ standby liquid control system storage tank level; f) emergency ventilation damper position; and g) airborne radiohalogens and particulate.
- 3. GSU shall implement modifications (installations or upgrade) for neutron flux monitoring consistent with the guidance of Regulatory Guide 1.97, Revision 2 or the NRC Staff's Safety Evaluation Report !
of the BWR Owners Group Licensing Topical Report (NED0-31558, Position on NRC Regulatory Guide 1.97, Revision 3. Requirements forPost-AccidentNeutronMonitoringSystem). Modifications, if required, shall be completed ore restart from the next months from the date of I refueling outage starting afte i a Sa et luation Renorton3ED0-315_58.
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' ENCLOSURE II RIVER BEND STATION NEUTRON MONITORING J PLANT SPECIFIC DESIGN CONSIDERATIONS This evaluation provides the plant specific information relative to the existingneutronmonitoringsystem-(NMS)capabilitiesatRiverBendStation (RBS) as it applies' . to the alternative' design requirements -stated in NED0-31558,- " Position on _NRC Regulatory Guide 1.97, Revision 3, Requirements for Post Accident Neutron Monitoring System."
The topics of discussion in the following sections of this evaluation can be' directly ' correlated with subsections 5.2.1 through 5.3 of NED0-31558.
The individual NED0-31558 subsection headings and requirements are stated followed by a discussion of existing. capabilities as they apply.to RBS.
The basis for the alternative requirements are not restated as they can be obtained from NED0-31558.
The discussion provided under each subsection applies primarily to the average power' range monitoring (APRM) subsystem. . When appropriate, information will also be provided for other NMS subsystems to show the capability to provide: backup or confirmatory support function to the APRMs when at'the lower end of the operating range.
Because the position of NED0-31558 is based on operator actions stated in the Emergency Operating Procedures (E0Ps) and the utilization of NMS for these actions, a discussion of the applicable RBS E0P and comparison to the generic BWR Emergency Procedure Guidelines (EPGs) is included in the following section.
River Bend E0P Overview
- The River Bend E0P were developed from Revision 4 of the EPG with minor deviations resulting from plant unique design differences. Because core power (neutron flux) is the parameter of interest, discussion will be limited to the E0P which is concerned with the maintenance and control of this parameter. The E0P which deals directly with core power is E0P-0001,
" Emergency Procedure-RPV Control" and the associated flowcharts, E0P-1 "RPV Control" and E0P-1A " Anticipated Transients Without Scram" (ATWS).
Consistent with the intent of the EPGs, the RPV control flowchart provides the operator with direction to control reactor power under conditions where it can be determined that the reactor will remain subcritical under all conditions without Boron injection while the ATWS flowchart provides instructions. under conditions where Boron-injection may be required. The entry conditions for E0P-0001 are any condition requiring an automatic or manual reactor scram, or drywell or primary containment temperature above 212 degrees Fahrenheit or 185 degrees Fahrenheit, respectively. The scram condit Mns encompass the condition where the operator may not be able to determiae reactor power. The Bases document for the EPGs discusses the fact that loss of electrical power to the APRMs does not, in itself, I require'that reactor power is indeterminate. The ensuing discussion provided by the Bases document further supports the variables / methods used Page 1 of 9
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1 to determine reactor power that were described in NED0-31558 section 6.3.
The general guidance provided by E0P-0001 regarding control of reactor power is as follows:
If all control rods are not inserted to or beyond position 02 (maximum subcritical banked withdrawal position) the alternate rod injection (ARI) system is initiated. If reactor power is above 5% or indeterminate, recirculation pumps are tripped, all other methods to insert control rods implemented, and if required, Boron injection is initiated prior to the suppression pool reaching 110 degrees Fahrenheit (Boron injection initiation temperature). If at any time during the performance of E0P-0001, all control rods are inserted to or beyond position 02, terminate Boron injection (if previously initiated), perform the scram recovery procedure, and exit E0P-0001.
The injection of Boron into the RPV for the above listed action is a limiting suppression pool temperature of 110 degrees Fahrenheit (suppression pool temperature is a Category I variable as defined in RG.I.97). Action is conservatively taken at this temperature to ensure suppression pool heat capacity is 3dequate to provide pressure suppression during reactor shutdown. Once Boron has been injected, operator actions are those which will ensure that the hot shutdown Boron weight is injected and that preferential injection systems are utilized to promote the Boron effectiveness as a shutdown agent.
5.2.1 Range Alternate Requirement: 1 to 100% (RBS downscale alarm is 5%)
RG 1.97 Requirement: 10-6% to 100%
The operatigg range associagd with the APRM subsystem at River Bend is 2.8 X 10 nv to 2.8 X 10 nv or 1 to 100% core thermal power.
This range satisfies the alternate requirement stated above.
an operating range of 1 In gddition, X10 ny to 1.5the Rgnv X 10 IRM instrumentation or approximately 10hag % to at least 15% power.
5.2.2 Accuracy Alternate Requirement: +2% of Rated Power RG 1.97 Requirement: None stated The loop accuracy of the RBS APRM subsystem is +2% (for normal operations) based on G.E. setpoint methodology calculations. To maintain this degree of accuracy, the LPRM subsystem is calibrated every 1000 MWD /T using the TIP subsystem to compensate for sensitivity degradation due to depletion of the uranium coating of the detectors with increased exposure. In addition, relative sensitives are determined corresponding to the increased exposures .
on approximately a six month frequency. Whenever power is greater than 25%, each APRM channel is checked weekly against power as determined by a heat balance and the APRM channel is adjusted as required to produce a deviation of no more than 2%. Due to the l Page 2 of 9
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- exhaustive. measures. taken to . assure loop accuracy, the ~APRM lg subsystem meets'the alternate requirements as stated in NED0-31558.
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5.2.3 Response Characteristics
' Alternate Requirement: 5 Sec/10% Change I
RG41.97 Requirement: None Specified For the APRM: subsystem. this characteristic has been previously
. stated in NED0-31558.
5.2.4 Equipment Qualification Alternate Requirement: ' Operate in ATWS Environment RG 1.97 Requirement: RG 1.89 and 1.100 RBS Expected Environmental Conditions From An ATWS Event As discussed in NED0-31558, the bounding events for determination of design basis requirements for NMS as it applies to RG 1.97 are the lesser _ ATWS events in which partial control rod insertion occurs or the plant is not. isolated _from the main condenser.' The event selected to be bounding for this category of events is " Inadvertent
- SRV opening with partial scram failure". This event, therefore, establishes the. environmental conditions and function time requirements for the NMS as it applies to post accident event monitoring.
The above identified event has been analyzed in NED0-24222 assuming complete scram failure (including ARI failure) which would result in harsher (more conservative) environmental conditions than the partial scram failure scenario presented in NED0-31558. As the case of complete scram failure is bounding for the special case of partial scram failure, a site specific evaluation based on
- NED0-24222 was performed to determine the enveloping environmental conditions. The conservative environmental conditions determined by
~the evaluation is a peak suppression pool temperature of 177 degrees Fahrenheit'and peak containment pressure of 8.5 psig reached at' 67 minutes into the event, indicative that the event produces a gradual increase in both parameters during the event. If it is conservatively assum.ed that these same conditions then translate to the conditions in the drywell, this identifies the worst case conditions existing in the drywell during this event. No degradation of environmental conditions is expected to occur within areas of the Auxiliary and Fuel Buildings during this event. The NED0-24222 analysis of this event also assumes the unlikely failure of the ARI system currently installed at RBS. In cases where ARI is accomplished, maximum suppression pool temperature would be considerably.less than that determined assuming ARI failure.
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Other Environmental Conditions The -. analysis ' of . Large Break LOCA, Small Break LOCA and Control Rod Drop.
Accident presented in section 4.3.2 of NED0-31558 parallel RBS operator actions.- environmental impact and impact. of NMS failure. As stated in NEDO-31558,_ the LOCA events 'will. produce a harsher environment in containment and drywell than the ATWS events.
RBS Environmental Design Considerations The following information provided.for environmental qualification is based 4 upon review of the RBS environmental qualification files.
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'LPRM/APRM l
The. components _of the LPRM/APRM are currently qualified to 10CFR50.49 for normal, abnormal, and accident conditions. Specific qualification is contained within the RBS equipment qualification files which demonstrates operability for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> into a small high energy line break in the drywell i or containment. The bases for environmental qualification of the equipment considers testing of the detector assemblies to 608 degreas Fahrenheit for normal plant operations, and the fact that design basis events result in negligible changes in the environments of the detectors, which are mounted in dry tubes in the core. All other components (e.g. cable, penetrations) located in a harsh environment have been qualified as Class 1E components capable of operating.during and following a design basis event. The lesser environmental conditions postulated for an ATWS event are enveloped by the, existing qualification bases.
Intermediate Range Monitors (IRM)
The components within the IRM neutron monitoring subsystem located in a harsh environment (with the exceptions of the drives / motor modules) are presently qualified to 10CFR50.49 for normal, _ abnormal, and accident conditions. The only design basis event for which the IRM subsystem is required to be operable is a small high energy line break inside or outside of drywell. The IRM subsystem has been demonstrated to be environmentally qualified for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> into this event. The cable, connectors, and penetrations used in IRM' subsystem have been demonstrated to be qualified for a design basis event where conditions are postulated to consist of 330 3
' degrees Fahrenheit in a steam environment. The lesser environmental j conditions postulate for an ATWS event are envelop'ed by the existing !
qualification bases. l l
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' 5.2.5 Function Time Alternate Requirement: I hour-I
, RG 1'97 Requirement: None Specified
-Both .the APRM/LPRM and'the IRM subsystem (with the exception of the IRM drives / motor modules)l have been environmentally qualified for 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s: in. small break LOCA conditions. which envelope .the ATWS
~ conditions determined'for RBS. Thus,.as the equipment is qualified
_for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. in a harsher environment than that for which the function time requirement is based, the RBS NMS meets.the alternate requirement specified.
.5.2.6' Seismic Qualification Alternate. Requirement: Seismic Qualification Not. Required RG 1.97. Requirement: Seismically Qualify: Category I. Equipment As Important to- Safety Per RG-1.100 and !'
- Since the event- which has been determined.to set the design basis' requirements for the NMS is an ATWS event, seismic requirements for
-the 'NMS'should be consistent with the ATWS rule (10CFR50.62). This rule. specifies-ATWS environmental conditions which do not require seismic qualification.
However, the APRMs and certain portions of.the IRM subsystem are designed ~to- operate during the design basis earthquake. This capability. exceeds the alternate requirement of NED0-31558. Seismic qualification for all components (except the drive / motor. modules for IRMs) located. outside of~ the main control room is available from
, either the RBS equipment qualification files or General Electric files.-
-The IRM/APRM recorders are Bailey Model 771 series. This particular
- recorder is not seismically qualified; however, this model of recorder (Bailey. '771 series) has been previously qualified for use in other applications / systems.
Based on-the above, the IRM subsystem would meet the seismic qualification requirements of RG 1.97, except in the case of a seismic event that di.sabled the eight IRM drives and motors.
Therefore, for all. cases, except for motor and drive disablement, the IRMs would also be available to provide additional supporting information to the operator, following a seismic event, for j monitoring power at or around the 5% downscale alarm.
5.2.7 Redundancy & Separation I i
Alternate Requirement: Redundancy to Assure Reliability RG 1.97 Requirement: Redundant in Division Meeting RG 1.75 Page 5 of 9
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,. .c The APRM' subsystem consists' ofL eight independent.ch'annels-'each channel consisting of inputs from up to twenty-four LPRM ' detectors, and_the necessary signal conditioning equipment to provide an output signal .directly reflecting average. power in the _ core. . The eight
. channels are divided into four separate divisions with each division consisting of two APRM channels. Because of_ the. redundancy: in.
' detector; inputs, -the practices: of: power and. equipment separation.
- and.the total number.of channels,.the'APRM subsystem satisfies the alternate. redundancy and separation criteria. The methods used for identification of power cable, signal cable, and cable trays as safety: related . components and .the identification ~ scheme used to u- distinguish between redundant cable, cable. trays, and instrument panels is: in accordance with Regulatory Guide 1.75. The IRM subsystem is near identical-in design to the APRM subsystem with respect.to redundancy and separation.
5.2.8 Power Sources Alternate Requirement: 'Uninterruptible and Reliable Power Sources RG 1'.97 Requirement: Standby Power Source (RG 1.32)
The: four divisions _of the APRM subsystem are normally powered from the RPS bus. Backup power is. supplied by Class 1E divisional' power via manual control in the event normal RPS power supplies fail. The recordersLlocated on the operators control console are supplied power from a separate UPS power source with non-divisional battery backup. .This power source arrangement for -the AFRM subsystem satisfies the alternate requirement specified above.
The IRM divisional arrangement and recorder power supplies are the same as the APRM subsystem.(NOTE: APRM's and. IRM's; share-
-recorders). IRM drive motors and associated control logic circuits
.are not supplied with uninterruptible power.
5.2.9 Channel Availability-Alternate Requirement: Available Prior to Accident RG 1.97-Requirement: Available Prior to Accident As discussed in NE00-31558, the power range instrumentation is available and in service while the plant is operating; therefore, the existing design satisfies this requirement.
5.2.10 Quality Assurance Alternate Requirement: Limited QA Requirements on Generic Letter 85-06 (Reference 3)
RG 1.97 Requirement: Application of Specific Regulatory Guides
~The entire APRM subsystem is safety related with the exception of the APRM recorders located on the operators control console. The Page 6 of 9
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APRM, subsystem was: constructed;in accordance with 10CFR50 Appendix B. 'As. can bet expected,Jthe quality requirements associated with 0 Lnon-safety related components and for .this reason, the recorders j were not~ designed . procured, and installed to the same quality level 'J requirements asithose associated with the remainder of the APRM' l c* equipment., .Nonetheless. -the eighteen. criteria of Appendix B to 10CFR50'and the; guidance provided under NRC Generic Letter 85-06 for
'non-safety (related ATWS. equipment have been fully. satisfied by the procurement, design, installation,'and ongoing operational quality assurance; program :for the APRM. subsystem. ' Based on the above, the APRM subsystem satisfies-the. alternate requirement stated above.
7 4C The IRM subsystem shares the same safety class levels as does the ;
APRM . subsystem with the exception of the.IRM drives / motor modules;. {
however.usince'the alternate requirements above specify. compliance-with'. Generic Letter 85-06 and-all IRM equipment was installed to the '
requirements of.10CFR50 Appendix B, even though the' drives / motor modules- are non-safety class components, this requirement is satisfied by the IRM subsystem.
5.2.11 Display'and Recording Alternate Requirement: Continuous Recording
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RG 1.97 Requirement: Continuous Recording Every NMS channel has a: built-in neutron flux meter provided with it's . instrument drawer located on.the Control Room backpanels and
! continuous recording capability provided by strip chart recorders located' on' the operators control -
console. In addition, the
. individual LPRM detector readings showing local power information can. be displayed on' a digital meter on the operators control console with-the use of the.RCIS system, 5.2.12 Equipment-Identification Alternate Requirement: Identify in Accordance with CRDR RG 1.97 Requirement: Identify as Post-Accident Monitors The NMS recorders 'are all clearly marked and labeled by division',
and signal input. These recorders are located on operators control console along with the other plant parameters which are of primary significance to the operator. Located between the four APRM recorders are the APPM status indicators, clearly-identifying alarm levels. IRM channel status indication and annunciation is near identical to that of the APRMs. This instrumentation was reviewed 1 from a Human Factors standpoint for both useability and identification during' performance of the DCRDR effort. Based on the j above, the identification of the equipment satisfies the requirement j of NED0-31558.
5.2.13 Interfaces -
-Alternate Requirement: No Interference with RPS Trip Functions Page 7 of 9
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RG 1.97 Requirement: Isolators to be used for Alternate Functions At RBS, the non IE porticns of the'NMS are isolated and separated as required from the IE portions of the system. The NMS; therefore, satisfies the alternate requirement as stated above.
I 5.2.14 Service, Test, and Cal?.' oration Alternate Requirement: Establish In Plant Procedures
, RG 1.97 Requirement: Establish In Plant Procedures The NMS is tested and calibrated on the frequencies as specified in the RBS Technical Specifications. Channel checks are generally performed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and channel functionals performed weekly when the' particular instrumentation is required to be in service (depends on plant operational conditions). The IRMs (trips, alarms, recorders, power supplies, regulators, etc.) are calibrated every 18 months while these same functions on the APRMs are calibrated semi-annually. On a weekly basis (with core power 25%) each APRM is checked against core thermal power as indicated by heat balance and adjustments are made when the APRM output deviates by more than 2% from power as indicated by the heat balance. Every 1000 MWD /T, the LPRM detectors are calibrated using the TIP system. In addition, LPRM sensitivities are trended to determine expected detector lifetimes and, on a periodic basis, computer programs are ,
run to verify consistency between calculated core thermal power and '
NMS indicated core power.
Plant section procedures cover the above described items. The control of the frequency of performance of these procedures is performed in the same manner as all other Technical Specification surveillance procedures. Based on the above discussion, this requirement, as specified in NED0-31558, is satisfied.
1 5.2.15 Human Factors l Alternate Requirement: Incorporate HFE Principles RG 1.97 Requirement: Incorporate HFE Principles i The DCRDR effort has been performed for the instrumentation and controls located on the operators control console. Human factors engineering principles were incorporated into this review process; therefore, the NMS satisfies this criteria.
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p 5.2.16 Direct Measurement Alternate Requirement: Direct Measurement of Neutron Flux RG 1.97 Requirement: Direct Measurement of Neutron Flux The NMS utilizes fission detectors and, as such, directly monitors neutron flux in the core. Therefore, thiis criteria is satisfied.
5.3 Conclusion l In all cases, the APRM subsystem of the NMS meets or exceeds the alternate requirements established by NEDD-31558 and in many cases, ,
complies with RG 1.97 requirements. Because the only operator 1 actions that are predicated based on a known core power level are those actions taken as a result of core power being above or below the APRM downscale alarm value of 5%, the acceptance of a reduced monitoring range for RG 1.97 is considered justified. In the event that core power is indeterminate, the operator has actions delineated such that the requirement to monitor core power becomes unnecessary (although not undesirable) and only serves as an enhancement to the operator.
The APRM subsystem incorporates acceptable range, acceptable environmental and seismic survival and class 1E power capability with redundancy, channel accuracy and availability and multiple indications providing the operator with adequate means to determine reactor power during both normal operations and accident conditions where core power indication would be most useful. In addition, the IRM subsystem would probably be available during events analyzed by NED0-31558, and in its attachment, since the drive unit function time is approximately 3 minutes (operator action to insert following a scram), thus providing supporting monitoring capabilities to the APRMs for core power below 1 percent.
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