ML20197H350
| ML20197H350 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 12/24/1997 |
| From: | Holden J FLORIDA POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 3F1297-47, TAC-M98991, NUDOCS 9712310255 | |
| Download: ML20197H350 (47) | |
Text
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Florida PGwer como Dess W303 December 24,1997 3F1297-47 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Subject:
Technical Specification Change Request Notice 210, Request for Additional Information (TAC No. M98991)
References:
- 1. FPC letter dated June 14,1997 (3F0697-10) " Technical Specification Change Request Notice 210" 2 NRC letter dated December 19,1997 (3N1297-12) " Summary of Meeting with the Florida Power Corporation"
Dear Sir:
In Reference 1, Florida Power Corporation (FPC) submitted Technical Specification Change Request Notice (TSCRN) 210, which proposes amendments to Operating License No. DPR-72 for Crystal River Unit 3 (CR-3). TSCRN 210 is necessary to address design and licensing basis changes primarily involving plant systems used to mitigate the consequences of certain small break loss of coolan' accide.nts (SBLOCA). In Reference 2, the NRC requested additional information to complete their review of TSCRN 210.
The requested information is provided in the following attachments. This additional information does not alter FPC's. previous conclusions or the No Significant Hazards Consideration provided in Reference 1. _
Attachment A - List of Commitments i The attachment provides the list of commitments made in this submittal. : ,
' CRYSTAL RIVER ENERGY COMPLEX 15760 W. Power Lirw Street Crystal River, Florida 34428-6708 a (352) 7&5-6486 A Florida Progress Company 9712310255 971224
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Anehment B - Additional Informatig)
-The; attachment provides additional information requested.by Reference 2 'and certain~ t
, clarifications to the meeting summary. .
Attachment C ; Timed Actions and Simulator Validations
.The attechment provides a table listing the time taken by the operator to complete each of--
the 17 operator actions associated-with TSCRN 210'during various simulator scenarios.
' Accompanying the attached table is a description of each referenced scenario. .
Attachment D - Operator Actions with Prior NRC Review ;
The attachment provides a table listing whether each of the 17 operator actions associated
- with TSCRN 210 have received prior NRC review and appropriate references.
-If you have any questions concerning this submittal, please contact Mr. David Kunsemiller, Manager Nuclear Licensing at (352) 563-4566.
Sincerely, l
M/4-John J. Holden Director
- Site Nuclear _ Operations i JJII/ mal; cc: Regional Administrator, Region 11
! - Senior Resident Inspector
- . NRR Project Manager -
b Attachments:
fA; Additional Information -
- . B.' List of Commitments C. Timed Actions and Simulator Validations D. . Operator Actions with Prior NRC Review l
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FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 ATTACHMENT A LIST OF COMMITMENTS 4
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- ATTACHMENT A:
LIST OF COMMITMENTS The following table identifies those actions committed to by Florida Power Corporation in this document. Any other actions discussed in the submittal represent inteixled or planned actions by Florida Power Corporation. They are described to the NRC for the NRC's information aint are not regulatory commitments. Please notify. the Manager, Nuclear Licensing of any questions regarding this document or any associated regulatory commitm(.nts.
ID Number. Commitment - Due Date 3F1297-471 FPC_ will update the FSAR to tellect the Prior to entering Mode 2.
use of manual llPI actuation. The Technical Specification Hases will also be updated to reflect the use of manual llPI actuation in accordance with CR-3 Technical Specification 5.6.2.17, i
4 FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 ATTACIIMENT B ADDITIONAL INFORMATION 1
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ATTACHMENT B ADDITIONAL INFORMATION In letter dated December '19,1997 (Reference 2), the NRC summarized a meeting between the
[ - NRC staff and representatives of FPC regarding TSCRN 210 (Reference 1). The NRC meeting summary also identifies additional information the NRC staff requests to completc their review for Operator Actions, Risk Analysis, and General Design Criteria (GDC) 20.
> FPC's response to each of the requests, as well as clarifications to the information contained in the NRC meeting summary, follows.
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.- U.S. Nuclear Regulatory Commission Attachment B 3F1297-47 Page 2 CLARIFICATIONS TO NRC AtEETING SUh1 MARY CImARIFICATION 1 The discussion of the Limiting Single Failures states, in part:
"A loss oiSUP-2 would result in both EDGs [ Emergency Diesel Generator]
being available anu EFP-1 [ Motor-Driven Emergency Feedwater Pump] being available, however, due to the EDG capacity, the EFP-1 (power supplied by the A train EDG) would need to be secured prior to starting the LPI pump "
While the statement is essentially correct, it should be noted that the load management strategies proposed by TSCRN 210, for this single failure, would preserve EFP-1 operation as presented in TSCRN 210, Attachment P, " Safety Assessment," pages 5 and 6.
CLARIFICATION 2 The discussion of the operator actions states, in part [ emphasis added]:
"FPC stated, during its presentation, that a total of 17 operator actions are included in TSCRN 210,15 are required to be executed for any one of the 3 scenarios."
The complete list of operator actions identified to mitigate the consequences of the SBLOCA scenarios proposed by TSCRN 210 are identified in Tables 3A and 3B of FPC letter dated September 25,1997 (3F0997-30). Although 15 operator actions are identified in these tables as being associated with the single failure of Loss of Battery "B" (LOBB), at most,14 may be required te mitigate the accident. Operator Action # 16 is to establish Reactor Coolant System (RCS) cooldown using the Turbine Bypass Valves (TBV) or Atmospheric Dump Valves (ADV) and is listed on the tables as associated with the LOBB single failure. However, Operator Action #16 is not required to mitigate the consequences of a SBLOCA. Please refer to FPC letter dated November 15,1997 (3F1197-40), Attachment B, page 8, which states:
"RCS cooldown using the ADVs or TBVs is specified by the EOPs [ Emergency Operating Procedure] in accordance with the generic technical guidelines, llowever, this cooldown is associated with post-LOCA plant recovery and is not required to satisfy 10CFR50.46 requirements." -
Further, many of these operator actions will be eliminated after Cycle 11. In TSCRN 210 and in a letter dated September 17,1997 (3F0997-27), FPC has committed to reduce the number of operator actions by implementing certain plant hardware modifications prior to the beginning of Cycle 12. As discussed in the December 15,1997 meeting, these modifications have the potential to eliminate 7 operator actions (Operator Actions 3, 4, 5, 9,11,15, and 17).
. U.S. Nuclear Regulatory Commission Attachment B 3F1297-47 Page 3 One commitment is to complete the permanent modifications to resolve the CR-3 EDG capacity limitations. Presently, the two primary options under consideration are to (1) modify the existing EDGs, further increasing their capacity, or (2) install a diesel-driven emergency feedwater pump. Included with either of these options is the removal of the automatic Emergency Feedwater Initiation and Control System (EFIC) trip of the motor-driven feedwater pump. Such modifications will increase the capacity margins of the EDG beyond the margins proposed by TSCRN 210 and eliminate the need for the operator actions proposed by TSCRN 210 for load management.
The second comniitment is to install the cross-ties and passive now control valves in the liigh Pressure Injection (IIPI) system. The installation of passive flow control valves will serve primarily to minimize the need for operator action in the first 20 minutes. Properly designed cross-tie piping on the llPI discharge piping will improve llPI flow delivery, thus reducing peak clad temperature, as well as reducing operator burden.
CLARIFICATION 3 The discussion of the operator actions states, in part [ emphasis added):
FPC stated, during its presentation, that a total of 17 operator actions are included in TSCRN 210,15 are required to be executed for any one of the 3 scenarios."
As discussed in TSCRN 210, FPC has been able to reduce the number of operator actions required m the first 20 minutes of these SBLOCA scenarios relative to the previous requirements. Not all of the operator actions listed in this table would be required for all SBLOCAs. As indicated in the table, some of these actions would only be required to be implemented under certain low probability scenarios. Additionally, some of these actions are considered to be " defense in depth" and are not considered in design basis mitigation analyses.
At most,14 operator actions may be required for the LOBB limiting single failure. For the remaining limiting scenarios, Loss of Battery "A" (LOBA) and Loss of EFP-2 would require, at most,12 operator actions.
. U.S. Nuclear Regulatory Commission Attachment B 3F1297-47 Page 4 CI ARIFICATION 4 The discussion of operator actions states, in part:
"llowever, only three of the 11 (operator actions already in the EOPs] are included in the licensing basis."
Attachment D is a listing of the operator actions used in the mitigation of SBLOCAs, as discussed in FPC letter dated September 25,1997 (3F0997-30). For each of these operator actions, Attachment D identifies whether there has been prior NRC review. The table indicates that ten operator actions (Operator Actions 1, 2, 3, 5, 6, 7, 8,12 13, and 16) have received prior NRC review. Clarifications regarding the context of the prior NRC review for four of these operator actions is as follows:
. Operator Action 2 involves a single action to initiate llPI and RBIC. This operator action combines two previous operator actions (isolate normal makeup and isolate letdown) into one action and was developed as part of TSCRN 210 in order to reduce the number of operator actions required in the first 20 minutes of the SBLOCA scenarios. As discussed in Attachment D hereto, the NRC has reviewed the operator action to initiate llPI, but has not speciFcally addressed that the single operator action would also manually initiate RBIC.
- Operator Action 5 involves ensuring adequate HPI flow by isolating a broken injection line using the new IIPI criteria. As discussed in Attachment D hereto, the NRC has reviewed operator actions for IIPI now balancing in order to maximam IIPI now to the core. While the balancing criteria did change to isolation criteria, the purpose md intent of this operator action to maximize HPI flow to the core did not change.
. Operator Action 7 involves ensuring Control Complex Ventilation is mnning in emergency mode. As discussed in Attachment D hereto, the NRC has reviewed the operator action to manually initiate the Control Complex Ventilation but did not specifically review the time required to complete the action.
- Operator Action 12 involves verifying the Control Complex Chiller is running.
As discussed in Attachment D hereto, the CR-3 FSAR, Table 8-1, " Emergency Diesel Generator "A" Auto & Manually Connected Loads," has listed the Control Complex Chiller as a manually connected load.
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,- U.S. Nuclear Regulatory Commission Attachment B 3F1297-47 Page 5 NRC REQUEST. OPERATOR ACTIONS 1 FPC expand the Simulator Validation table (Attachment E, Page i) submitted in their September 25,1997, supplement to TSCRN 210, to provide completion times for all 17 of the operator actions, in addition, for operator actions that e.rceed 20 minutes, include the maximum times allowed for operators to perfonn each action without exceeding any safety limits.
FPCRINPONSE Procedure Validations Emergency Operating Procedures (EOPs) and Enclosures' were validated throughout the development process to ensure their completeness and accuracy. The attached table (See
- Attachment C) of timed actions for the 17 operator actions associated with TSCRN 210 reflects the procedure validation effort. Accompanying the attached table is a description of each referenced scenario. The table presents the validations most closely related to the TSCRN 210 solution sets. Additional validations of EOP-03, Inadequate Subcooling Margin, not depicted on the table were performed to demonstrate the ability of the procedure and the referenced EOP enclosures to mitigate ioss of subcooling margin (LSCM) symptoms. These included a scenario involving a Station Blackout (SBO), two validations of a large once through steam generator (OTSG) tube rupture and LOCA cooldown, a makeup system letdown line rupture, a loss of subcooling margin with no high pressure injection (HPI), and a Mode 4 LOCA.
Between the time period of July 30, 1997 and the end of October 1997, the procedure validations shown on the attached table were performed using three separate operating crews who were in training. The individuals involved in the validations participated as a functional crew. No mixing of crew members was used for any of the validations performed through October. While operating crews were in the process of taking their final exams, those validations during November 1997 were performed by licensed operators who were not part of an assigned crew.
The maximum times allowed for operators to perform each action are established by the supporting analyses. The Framatome Technologies Incorporated (FTI) reevaluation of HPI EOP-03, inadequate Subcooling Margin EOP-08, LOCA Cooldown EOP-13, Rules EOP 14, Enclosures - Enclosu.e 7, EFP-2 Management EOP-14, Enclosures - Enclosure 11. EDG A Load Management EOP-14. Enclosures - Er. closure 17, Control Complex Emergency Ventilation ,
I EOP-14. Enclosures Enclosure 18, Control Complex Chiller Startup EOP-14, Enclosures - Enclosure 19, ECCS Suction Transfer l
e U.S. Nuclear Regulatory Commission Attachment B 3F1297-47 Page 6 requirements during small break LOCAs, FTI document 51-12d5866-00, assumes the first six operator actions will be accomplished during the first 20 minutes of the LSCM. These are maximum assumed times. Although these actions, as shown on the attached table, can easily be accomplished in the required timeframe, the analysis does not take credit for the actions until the 10 and 20 minute points, consistent with the NRC staff position contained its the NRC letter to FPC dated September 26, 1978 which defines allowabl9 operator actions for which credit may be taken following a Condition 111 event (small LOCA). These actions are defined as " simple" (characterized as pushing of a button or turning a switch) which can be assumed after at least 10 minutes, or " complex" which cannot be assumed in less than 20 minutes.
The maximum times associated with the post-twenty minute operator actioris can be best explained by first understanding the linear nature of FPC's emergency operating procedures.
d As explained in FPC letter dated December 11,1997 (3F1297-38), EOPs are written in a step sequence fashion. Each step must be perfonned in the specified sequence and be completed prior to progression to the next step. The only deviations from this are the carryover steps and
" concurrently perform" steps. The first " concurrently perform" step in EOP-03 is a contingency action found in Step 3.9 (ensure OTSG level is trending towards the inadequate subcooling margin, "lSCM" setpoint) to address the " defense in depth" elements of TSCRN 210 to establish an alternate source of feedwater if emergency feedwater is unavailable. For the TSCRN 210 solution sets, emergency feedwater is not challenged until the control complex chiller is loaded to the emergency diesel generator (EDG); therefore, the action required by Step 3.9 to depress the ISCM push buttons will be performed prior to progressing to the next step.
The next " concurrently perfonn" step is Step 3.12 to start the control complex ventilation on the emergency mode. This step must be completed within 30 minutes. The next step, to transfer the emergency core cooling system ('CS) pump suction from the horated water storage tank (BWST) to the reactor building (RB) sump if the BWST level is <20 feet, can be performed prior to completion of Step 3.11. Ilowever, based on the break sizes for small break LOCAs, the BWST switchover level is not expected to be reached until well after (>2 hours) the next time critical action contained in Step 3.17, which is starting the control complex chiller.
Prior to the " concurrently perfonn" step for starting the control complex chiller, EOP-03 Step 3.16 is a "perfonn" step to manage EDG "A" load. This step must be accomplished prior to beginning the step to start the control complex chiller. The balance of the required TSCRN 210 operator actions are in EOP-08 and are a function of break size, cooldown rates, or other contingencies based on system parameters. The only " concurrently perform" steps in this sequence not already previously performed in EOP-03, is a step to maintain a temperature log for heatup and cooldown of the RCS and pressurizer.
e_ ; U.S. Nuclear Regulatory Commission . Attachment B-.
3F1297-47 Page 7
- NRC HQUEST. OPERATOR ACTIONS 2 FPC discuss how the average times identified in Attachment B (Table of EOP Step Changes) of-their December 11,1997, RAI response were derived.
FPC RESPONSE The attached table (Attachment C) lists one validation performed on December 8,- 1997, with no validation number. This simulator mn was performed during the NRC EOP Inspection 97 :
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12, to demonstrate for the inspectors, the ability to mitigate a .small break LOCA with a LCBB. As noted in FPC's letter dated December 11, 1997 ' (3F1297-38), this scenario incorporated a radio / telephone link between the simulator located in the offsite training center and the CR-3 plant where field actions were performed. This special setup contributed to
. several delays in completion of steps. In actual conditions, communications would occur using the PAX phone system and radios, or face-to-face. The average recorded times contained in the listing of operator actions included in FPC's letter dated December 11,1997 (3F129738) is a simple average of the times depicted in the attached table, including those identified for the -
December 8,1997, scenario.
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, U.S. Nuclear Regulatory Commission Attachment 11 3F1297 47 Page 8 SRCREQLEST, OPERATOR ACTIONS 3 FPCprovidc information that demonstrates that all crews can accomplish each of the operator t.ctions required, for all three scenarios, in the maximum time allowed operators to perform each at tion without e.tcreding safety limits.
The NRC meeting summary also noted thefollowing:
Subsequent to the meeting, FPC ir(formed the staff that all crews were trainedfor each of the
' three scenarios, but none of the crews were evaluated (i.e., simulate: tested) for a LOBB or failure of EFP 2. Crews were evaluatedfor a loss of EDG A, which FPC state:is very similar to a LGBA. FPC stated that it would venfv this inforinatfort and provide it in writing to the staff. 1 1
EPC.RESEOSSE Operator TIaining EOP training on the TSCRN 210 solution set scenarios was developed and implemented in accordance with standard industry practices for performance based training using the systematic approach to tiaining. The process began with a thorough review of the three solution set scenarios, EOP 02, EOP-08, and associated rules and enclosures. The program developers performed an assessment of the kanwled; ., skills and abilities required to perform the procedures and had them reviewed by subject matter expert; in the EOP development group. The developers then compared the identined knowledge, skills and abilities against the current qualincations of the intended target population to determine the required scope of training, it should be noted that the target population for this training is a group of highly trained licensed control room operators. Therefore, the primary focus of the EOP training program wTs to reinforce when and why the specified actions would have la be taken. To accomplish this, the training was broken into three phases.
- All operating crews (including backup licensed operators) were fully trained on each of the dirce solution set scenarios. Each solution set was discussed in detail including initiating events, assumptions, mitigation strategies, and additional " defense in depth" options that may k available.
- Following the solution set discussion, all operating crews (including backup licensed operators) received comprehensive classroom training on the applicable EOPs, rules, and enclosures. The training focused on how the actions specified in the EOP would mitigate each of the solution set scenarios and highlighted those actions which were identined as being time critical to the mitigation stantegy.
. U.S. Nuclect Regulatory Commission Attechment 11 3F1297-47 Page 9 e Next, each operating crew (including backup licensed operators) next received eight hours of simulator instruction on EOP-03, EOP-08, and associated mies and enclosures. This instruction consisted of the LOBil and emergency feedwater pump EFP-2 failure solution sets. The simulator training focused on (1) exercising the skills and abilities necessary to implement the requirements of the procedures under conditions similar to those that would I exist during a real event, (2) ensuring the licensed operators could perform the procedure l in the ptoper sequence, making appropriate branches and transitions. Time critical steps J were carefully n'onitored for timely completion as part of the success criteria. The ;
scenarios were delivered as unannounced casualty exercises with instructor intervention as necessary to ar.:.wer student questions or provide additional instruction.
Opxrator Evaluations Each operating crew (including backup licensed operatots) participated in an unannounced casualty training effectiveness evaluation exercise at the end of their training week. The evaluation scenario challenged the crew's ability to perform the important mitigation strategies and time critical actions of EOP-03 and EOP-08 in response to a LOOP, failure of the "A" emergency diesel generator (EDG), and an llPI line break LOCA. This scenario is essentially the same as the TSCRN 210 solution set involving a small break LOCA, LOOP, and loss of llattery "A." Crew performance was measured a,ainst the requirements of ECP-03 and EOP-08, again, with time critical steps monitored for success The exercise involved no instructor intervention during the scenario. The performance of all crews was satisfactory. Each crew participated in a detailed post exercise critique where individual pe-formance challenges were identified and remediated.
Condusion Operater training and crew evaluations indicate that the SilLOCA solution sets are understood by CR-3's operators. Any additional evaluations performed of these scenarios would have the potentia' to overly bias the operators toward them and lead to an increased potential for misdiagnosis. FPC strongly feels that creating an event-bias through training is not consistent with the industry-wide premise for training on symptom-based procedures such as emergency operating procedures. FPC is satisfied that the systematic approach to training employed dming the recent operator training has provided sufficient assurance that all crews can accomplish each of the operator actions required, for all three single failure scenarios.
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, U.S. Nu: lear Regulatory Commission Attachment B 3F1297-47 Page 10 NRC REQUEST. _ RISK ANALYSLU }
A detailed discussion on FWP-7 affects the results of FPC's risk analysis (providirg quantitative as well as qualitative description), including the assumed unavailability / failure probability of FWP 7.
FPC REMEONSE .
In the CR 3 trobabilistic Safety Assessment (PSA), there are thre: different core damage sequence types for small break LOCA. They are: 1) small break LOCA, failure of IIP 1, 2) sniall-break LOCA, successful IIPI, failure of IIPR (high pressure recirculation), and 3) small break LOCA, successful llPI, failure of all feedwatcr (main, emergency, and auxiliary),
failure to open the Pilot Operated Relief Valve (PORV). The first two sequence types are i straightforward and involve failure to make up the inventory being lost out the break, short.
term and long term. The third sequence addresses those sinall-break LOCAs which do not relieve enough water out the break to remove all of the decay heat. These breaks r: quire secondary cooling, or, given a loss of secondary cooling, opening of the PORV. During a LOOP, secondary cooling can be provided by emergency or auxiliary feedwater (FWP-7).
Procedural guidance exists to implement each of these means of secondary cooling, as well as the opening of the PORV if secondary cooling is unsuccessful. i i
FWP 7 Fallure/ Unavailability Data ,
FWP-7 Event Probability Source FWP-7 maintenance 9.3x10 CR-3 plant specific data - ,
unavailability FWP-7 failure to start 1.33x10 ' CR-3 plant specific data 4
FWP-7 fails to run 3.48x10 CR 3 plant-specific data in TSCRN 210, Attachment A, Commitment 5. FPC stated:
"FPC also will have available Auxiliary Feedwater Pump 7 (FWP-7) which will be powered by a dedicated diesel generator mstalled during the current curage. :
The use, maintenance, and testing of FWP-7 will be controlled by plant procedures that will be approved prior :0 CR-3 restart to ensure that availability and reliability is appropriately addressed commensurate with its importance.
The i WP 7 dedicated diesel generator, MTDG 1, is installed and tested. Procedures to ensure the availability and reliability are approved and issued.
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U.S. Nuclear llegulatory Commission Attachment il 3F1297-47 Page 11 Specifically, the requirements of NOD-31, " Equipment Reliability improvement Program,"
include-l FWP-7, its flow path and diesel generator are listed as " Equipment Needing Special Consideration for Timely Actions." l If FWP-7, its flow path or diesel generator are out of service, then I i
e Enter equipment in the Out of Service log and Nuclear Shift Supervisor log.
- Gererate necessary workpackages within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for troubleshooting ,
and/or repair.
- Complete repairs in 30 days.
- Approval of Directors of Nuclear Plant Operations and Nuclear Regulatory Affairs is required to extend the out of service time beyond 30 days.
Surveillance for FWP 7 and its diesel generator are in SP-348A, " Auxiliary Feedwater Pump (FWP 7) Testing and h1TDG-1 Surveillance Test," which includes:
- hionthly valve strokes e Quarterly pump tests e Quarterly tests of diesel generator -start and load of FWP-7 Preven'ative maintenance of the diesel generator is part of Pht 275, Enclosure 5. "hiTDG 1 Diesel Generator hiaintenance Schedule," which includes:
- 6 months - replace lube oil, fuel filters and air filters; drain any water and sediment from fuel tank; inspect / replace cooling system hoses and drive belts e 12 months 6 month surveillance plus inspect air box drain tubes and clean breather screens
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. U.S. Nuclear Regulatory Commission Attachment B 3F1297 47 Page 12 NACAEQUEST. El%K ANALYKis 2 A discussion on how the estimated risk at the plant uvuld change of FWP-7 is not credited in i the risk analys.i. In other nords, what uvuld be the estimated overall CDF if FWP-7 uns not I creditedfor SBLOCA/ LOOP sequences?
-i VPC RESPONSL f l
if no credit for FWP-7 is taken, as well as no credit for operator actior., the baseline core !
4 4 damage frequency for all internal events would incre..a from 7.$lx10 per year to 1.0$x10 l per year. This increase is within the "non-risk-significant" region of the EPRI PSA ,
. Applications Guide. ,
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, U.S. Nuclect Regulatory Commission Attcchment B l 3F1297-47 Page 13 i P
NRCREQl1EST RASK ANALYSIS 3 ;
Just(Ocation and references for the single failure probabilities presented in the risk analysis ,
table provided in the meeting, and to expand the table to include single failures that are less l limiting but more likely than the LOBB, LGBA, and EFP-2, (i.e., failure of EDG). .
EEC BWNPONSE
SUMMARY
FPC has prepared a table of the operator actions and the corresponding sequence frequencies due to the ove.all single: failure probabilities (See Table 1) and assuming that the operator fails
- to perform the required task 100% of the time. The overall single failure probabilities include the less limiting single failures. The percent increase for each of these sequence frequencies 4
over the baseline core damage frequency of 7.19x10 , as discussed in FPC letter dated 2
December 3,1997 (3F1297 27), is within the "non-risk-significant" iegion of the EPRI PSA Applications Guide.
FPC has re-evaluated the sequence frequencies using more realistic human error probabilities and equipment failure probabilities as presented in Table 2. This analysis demonstrates the changes in procedures and the new load management strategies proposed by TSCRN 210 do not have an appreciable effect on the low risk of core damage frequency at CR-3.
DISCUSSION The three limiting :; ingle failures addressed by TSCRN 210 are: 1) loss of ES Bus A or loss of "A" battery,2) loss of ES Hus H or "B" battery, and 3) loss of EFP-2.
The dominant contributor to the failure probability for loss of ES Bus A or loss of "A" battery, given a LOOP, is the "A" diesel generator. It contributes over 97% of the single failure probability. The failure modes are failure to start, failure to run after a successful start, and maintenance. The failure rates for fail to start and fall tr ..m are from plant-specific data. The diesel maintenance unavailability is also taken from plant-specific records.
Diesel generator fails to start: 6.06x10
Diesel generator fails to run: 3.08x10-2 Diesel generator in maintenance: 6.80x 10
Consequently, the overall single failure probability for LOBA or LOBB considering the above failure rates is 4.49x10-2 ,
As with the loss of ES Bus A or loss of "A" battery, the dominant contributor to the failure
= probability for loss of ES Bus H or loss of "B" battery, given a LOOP, is the "B" diesel generator. The percent contribution and the data sources are the same as for the "A" diesel.
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, U.S. Nuclear Regulctory Commission Attachment B l i
3F1297-47 Page 14 For loss of EFP 2, the dominant contributor is EFP-2 itself. It contributes 98% of the single failure probability. The failure modes are failure to start, failure to run after a successful start, and maintenance. The failure rate for fail to start is from plant-specific data. The failure rate for failure to run is taken from generic imiustry data. The failure rate is an average of the failure rates fot standby turbine-driven pumps taken from NUREG/CR-1205, Seabrook Probabilistic Safety Study, the Oconee PRA, the Zion Probabilistic Safety Study, and the Indian Point Probabilistic Safety Study. The EFP-2 maintenance unavailability is calculated from plant specific records.
l EFP-2 fails to start: 1.58x10 l EFP 2 fails to run: 3.05x10~2 ,
' 4 EFP-2 in maintenance: 9.60x10 Consequently, the overall single failure probability for ne loss of EFP-2 considering the above 2
failure rates is 5.7x10 .
.The sequence frequency for each new operator action specified by TSCRN 210 is shown by the following risk analysis table updated to reflect revised limiting single failure probabilities discussed above and given a human error probability of 1.0. The percent increase for each of these sequence frequencies over the baseline core damage frequency of 7.19x10*, as discussed j in FPC letter dated December 3,1997 (3F1297-27), is within the "non-risk-significant" region ;
d of the EPRI PSA Applications Guide.
Table 1 Operator SBLOCA Overall Single Failure iluman Error Sequence Frequency Given Action LOOP Probability- Probability No Operator Action 9 2.24E-05 LOBB 1.0 1.0lE-06 4.49E-02 10 2.24E-05 LOBB 1.0 1.01 E-M 4.49E-02
- 11 2.24E-05 Loss of EFP-2 1.0 1.28E-06 5.70E-02 14 2.24E-05 LOBA or LOBB 1.0 1.01 E-06 4.49E-02 15 2.24 E-05 Loss of EFP-2 1.0 1.28E-06 5.70E-02 A human error probability of 1.0 assumes that the operator fails to perform the required task '
100% of the time. The human error probability of 1.0 was originally used by FPC to demonstrate the small significance the new operator actions had on the total CR-3 core damage frequency. More realistic human error probabilities have been developed using the techniques of NUREG/CR-1278, "llandbook of Iluman Reliability Analysis with E:nphasis on Nuclear Power Plant Applications." The realistic human error probabilities reflect that each of the operator actions is addressed by explicit procedure guidance, has been included in operator
, U.S. Nuclear Regul: tory Conunission Attachment 11 3F1297 47 - Page 15 training, and can be consistently completed with the required timeframes as shown by recent i
simulator exercises.
'lhe equipment failures associated with the operator actions (e.g., such a !!FW crosstie valve, r712, for Operator Action #9) were determined using fault trec modeling based on CR 3
.lant specific failure data. Operator Action #15 has an equipment failure of 0.00 since no a'iditional equipment is required to complete this action.
The changes in the sequence frequencies has been recalculated using these more realistic human error probabilitics and equipment failures as shown below. It should be noted that no credit has been taken for additional mitigative equipment such as llPl/PORV cooling or
-I M P 7.
Table 2 {
Operator SilLOCA Overall Single - lluman lirror liquipment Fa lure Sequence
- Action - 1,00P Failure - (11) (liF) Probability Frequency Give n Probability _ Probability II+I!F Probabilitics ~
9 2.2411-05 1. 0 1111 1.6011-02 2.0911-02 3.721!-08 4.491!-02 10 2.2411-05 1, 0 1111 2.60li-03 1.4111-02 1.6911-08 4.4911-02 11 2.2411-05 less of liFP-2 2.30li-03 5.1011-05 3.01!!4)9 5.70l!4)2 14 2.2415-05 1, Olla or 1.01111 4.7011-03 1.671102 2.1511 08 4.4911-02 15 2.2411-05 less of I!FP-2 1.0011-03 0.00 1.28154)9 5.7011-02 in summary, FPC considers that this analysis demonstrates the changes in procedures and the
, new load management strategies proposed by TSCRN 210 do not have an appreziable effect on the low risk of core damage frequency at CR 3.
i.
U.S. Nuclear Regulatc,ry Commi::slon Attachment B 3F1297-47 Page 16 NRC REQUEST. RISK ANALYSIS 4 Justification for the frequen'y assumed in FPC December 3,1997, RAI response for a SBLOCA occurring in the llPI (7E-7/ year).
FPC RESPONSE The frequency for a small break LOCA in a llPI line was taken from EPRI TR-102266, " Pipe f Failure Study Update." The failure rate for PWR safety injection and recirculation piping is 1.42x10* per section per hour. For four llPI lines, this translates to a frequency of:
/shlECA. HPi tu, = (1,42x10* per section per hour)(4 lines)(1 section)(8760 hrs /yr)
= 5.0x10-5 per year i
With the additional requirement of an accompanying LOOP, the frequency becomes:
/suwc4 oscisi=> toop = (5.0x10'5 per year)(1.4x10-2) !
= 7.0x10 per year i
4 f
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. U.S. Nucle:r Reguletory Conunission Attachment 11 3F1297-47 Page 17 NBC REQUEST. GENERAL DESIGN CRITERIA (GDC)R 1he discussion of operator actions led the staf to request that FPC address Crystal River design criterion 14 Core Protection Systems" (similar to GDC 20, " Protection Sprem Functions"), which is described in the Final Safety Analysis Report (FSAR) ai;d requires the high pressure inJcction to be initiated automatically. The analysis used to support Amendment 210, for some of the small-break loss-of coolant accidents, crrdits manual
<- Initiation ofIIPI by the operators, rather than automatic initiation. The licensee indicated that they would correct their licensing documentation (FSAR and Technical Specyications flases) to clanfy the credit given to manual initiation ofIIPl.
EPC.RESPOSSE SLIMMARY FPC has completed a review of the CR-3 design and licensing basis and concluded that the NRC has reviewed and approved the use of manual llPI actuation at CR 3. FPC will update the FSAR to reflect the use of manual llPI actuation. That update will be included in Revision 24, which will be issued prior to entry into Mode 2 as stated in FPC letter dated Nove.nber 19,1997 (3Fil97-47). The Technical Specification liases will also be updated to reflect the use of manual llPI actuation in accordance with CR 3 Technical Specification 5.6.2.17 and provided prior to entry into Mode 2.
TSCRN 210 Attachment 11 " Safety Assessment," of TSCHN 210 provides a topically-oriented safety assessment of the proposed SitLOCA solution sets. Included is a discussion of the systems resp (mse to the initiating event of a SilLOCA concurrent with a loss of offsite power (LOOP) and the limiting single failures. Page 3 of the Safe!> Assessment states:
"An ES [11ngineered Safeguards) actuation occurs when RCS [ Reactor Coolant System) pressure reaches 1500 psig. Operators manually initiate llPI if a loss of subcooling margin occurs before the 1500 psig automatic actuation."
Attachment D, "Framatome Document 171'I 51-1266138-01," of TSCRN 210 provides the 171'l safety analyses and evaluations of accident mitigation challenges to the SilLOCA solution sets, Page 1 of this 171'l document describes the initial event sequences and states:
"Subcooling margin will be lost anJ, if not automaticahy actuated on low RCS pressure, the operators will manually imtiate high pressure injection (llPI) flow,"
, U.S. Nuclear Regulatory Commission Attachment 11 3F1297-47 Page 18 Attachment F, "Supponing Information," provided, in part, a description of the operator actions required for the implementation of TSCRN 210. Table 3A of Attachment F, Operator Action #2, states:
"If Subcooling Margin is lost, and ES has not actuated, initiate llPI ..."
During a meeting between representatives of FPC and the NRC on December 15,1997, FPC explained that the CR-3 accident analysis ti.kes credit for the manual actuation of IIPI for certain very small break LOCAs that are not expected to obtain an ES signal within 20 minutes. The NRC noted ti.at reliance on manual IIPI actuation was inconsistent with the CR-3 FSAR.
EXISTING FSAR AND TECIINICAL SPECIFICATION HASES Section 1.4.14 of the CR-3 FSAR states (emphasis added):
1.4.14 CRITERION 14 - CORE PROTECTION SYSTEMS (Category B)
Core protection systems, together with associated equipment, shall be designed to act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits.
DISCUSSION The reactor design meets this criterion by reactor trip provisions and Engineered Safeguards (ES). The RPS is designed to limit reactor power which might result from unexpected reactivity changes, and provides an automatic reactor trip to prevent exceeding acceptable fuel damage limits. In_.a Loss-of-Coolant Accident (LOCat the Engineered Safeeuards Actuation SystenL(ESAS) automatically actuates the liigh-Pressure Iniection (IIPI) and Low. Pressure Inlection (LPI) Systems. The Core Flooding Tanks (CFTs) are self actuating. Certain long-term operations in the Emergency Core Cooling Systems (ECCS) which do not require immediate actuation, such as remote switching of the LPI pumps to the recirculation mode and sampling of the recirculated coolan' are performed manually, Sections 7.1.2 and 7.1.3.
The CR 3 Technical Specification Bases also indicates that ilPI is initiated automatically, without mentioning the need for manual IIPI actuation. Most significantly, the Applicable Safety Analysis of thses 3.3.6, "ESAS Manual Initiation," states:
"The ESAS manual initiation Function is a backup to automatic initiation and allows the operator to initiate ES Systems operation whenever plant conditions dictate. The manual initiation Function is not assumed or credited in any accident or safety analysis."
, U.S. Nuclear Regulatory Commission Attachment B 3F1297-47 Page 19 PREVIOUS NRC REVIEW AND APPROVAL i
in letters dated August 24, 1979 (300879-25) and November 14, 1979 Fl179-25), FPC :
l provided its responses to Short Term Actions Items 2 and 5 of I.E. Bulletin 79-05C, " Nuclear Incident at Three Mile Island - Supplement." Short Term Action item 2 required, in part, a ,
LOCA analysis for a range of small break LOCAs and a range of time lapses between the reactor trip and Reactor Coolant Pump (RCP) trip. Short Term Action item 5_ required analysis and the development of guidelines and procedures related to inadequate core cooling ;
and a dennition of the conditions under which a restart of the RCPs should be attempted. In j both FPC letters, the plant behavior during a SBLOCA is described. The analysis provided discusses various sizes of SHLOCAs and under different conditions of plant responses.
Ilowever, the analyses describes incidents in which automatic ES actuation would not occur and the need for operator actions to initiate IIPl. Further, the November 1979 letter provides operating guidelines for SDLOCAs. The list of symptoms and indications of SBLOCA include a statement that ES actuation may not occur for all SBLOCAs. The list of immediate actions i
include:
2.5 Monitor system pressure and temperature. If saturated conditions occur, initiate IIPI.
2.6 If ESFAS [ Engineered Safeguard Features Actuation System] has been bypassed due to heatup or cooldown, initiate safety injection.
In letter dated June 3,1981, the NRC stated that FPC's responses to IE Bulletin 79-05C would '
be reviewed as part of Task Action Plan item II.K.3.5. In letter dated March 15,1989, the NRC stated:
" Florida Power Corporation has adopted and implemented the BWOG [ Babcock
& Wilcox Owners Group] methodology, and our review of the Crystal River Unit 3 (CR-3) plant specific information has progressed to the degree that we
[the NRC] can conclude the BWOG methodology has signincantly improved reactor safety, and that there are no major safety-significar.t concerns regarding the plant-specific information.
"Therefore this issue is considered satisfactorily resolved for CR-3 and our effort on the subject TAC [ TAC No. 49668] is concluded." -
CONCLUSION The NRC has specincally reviewed and approved the use of manual IIPI actuation at CR-3. ,
Further, the statements made in the FSAR reDect a description of the principle safety features associated with a large break LOCA occurring in hie 1. The description also dx3 not i reflect the manual actuation required when the automatic actuation features are bypassed -
, U.S. Nucle:r Regulatory Commission Attachment B 3F1297-47 Page 20 duriag heatup and cooldown, nor does it reflect the requirements for manual operator actions for a SHLOCA occurring in Mcxle 4.
The CR 3 ISAR and Technical Specification Bases do not reflect the current licensing and design basis of CR 3 for manual IIPI actustion and will be updated.
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FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 l
l ATTACIIMENT C TIMED ACTIONS AND SIMULATOR VALIDATIONS l
U.S. Nuclear Regulatory Commission Attaclanem C :
- 3F1297 Page 1 ;
Time Deperdent Steps Based on Loss of SCM Operator Timed Actxms (Min:sec) Durmg Simulator Vclidations Regaired ;
Acta Date Validation 7/30/97 8/5/97 8/6/97 8/11/97 8/12/W 8/13/97 ._ .1/97 9/15/97 9/1997 11/7/97- 11/12/97 12/8.V7 Perfonned Shift Cunaaws 2RO 1 RO 2RO 2 P.O 3RO 2RO 2RO 3RO 1RO IRO 1RO 2RO ISRO ISRO 2 SRO ISRO 2 SRO ISRO ISRO ISRO 1SRO 1SRO ISRO 2 SRO Scenarios Scenario Scenarm Scenarm Scenano Scenano Scenarm Scenano Scenano Scenano Scenarm Scenarm LOBB 1 2 71 74 69 75 78 88 97 88 Ill NRC' Trip RCPs < 2 min < 2 min <2mm <2 min <2 min <2 min N < 2 min OJ:25 Uth23 00 2 1AU Initiate llPI/RBIC <2 min <2 min <2 min <2 mm <2 rein <2mm v <2 min 0006 00:42 0199 4dU l Ensure llPI Valves 3:10 4 22 2:05 210 N i 4d)1 3U5 02:57 02:56 800
! OPen i
- Isolate high flow 14:30 14:15 N/A N/A N/A 5:00 N/A 13:13 N/A 07:12 07:34 15d)0 I line Isolate RCP Seal ~ 16:00 i 10:45 6:00 6:00 6:00 i 15:08 5:30 05.37 08:51 17:00 l Injection l Raise OTSG tevel i N v N 1532 ~9 min 09:05 0992 1710 to 90%
Ensure CC Vent in 23:13 V 18:45 15:40 i 1810 i 32:37 1335 22:44 22AU 2310 ;
Emergency Mode j NOTEI ;
l Transfer ECCS 39:00 ~36dj0 ~3810 -4210 i ;
Suction to Sump !
Crosstic EFP-2 to ~ 32:00 i N/A i 481M 3237 4610 "A" Train
, i Close EFW Block ~32dXs i N/A N 48dM 32:57 4110 Valves - I t
. -. . ~ . . . . - __
k U.S. Nuclear Regulatory Commission Attachmem C 3F1297-47 Page 2 Operator Timed Actxxis (Mm sec) Durmg Simulator Vali.lanons Regured Action Dase Validanon 7/30/97 8/5 M 8/& 97 8/11/97 8/12/97 8/13 5 8/21 M . 9/15/97 9/19/97 - 11/7/97 11/12/97- 12aL97
- Performed RWP/SWF in PTL -32d>0 N N/A i 48iM N/A 29:49 EFP-1 Trip Defeat Ensure CC Chiller 41:00 59:15 50XX) i N Y V 51:13 47:12 42:32 70dK)
NME isolate RB Sump i i V 45:37 Cycle EFP-2 V EFP-1 Trip Defeat N/A 29:14
<50C # interlock RCS Cooldown V $9%)3 i using TBVs/ADVs - j Periodically This is a contin;;ency action not exercised because skills for isolating affected IIPI line previously exercised in EOP-3 (see 4th operator acton abme)
Evaluate llPI line twak criteria i Means step performed but time not captured Empty box means step not covered by scenano because the sceruno ended at the transition to cooldomm (EOP-8).
- NOTE I - Action delayed due to damper indicanon problem from battery failure - resolution by modification (MAR 97-07-0542).
NOTE 2 - Fourth and fifth operator actions are reversed from the OA numbers shown in previous tables to indicate actual procedure sequence.
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, U.S. Nuclear Regulatory Commission - Attachment C 3F1297-47 Page 3 l
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Scenark_L i 100% Power, MCC-3AB is aiigned to the B Bus. RCS leak develops in the B2 IIPI nozzic. ,
Leak starts at 50 ppm and increases to 200 gpm. Rx trip on low pressure or manual operator i action. When Rx trips, a LOOP occurs including the 22KV backfeed to the air compressors. B j EGDG fails to start due to loss of B Battery. EOP-02 is entered and immediate actions - j performed. EOP-03 is entered on loss of SCM. The failed IIPI line is isolated. EFW is cross-
- tied to allow shutdown of EFP-1. The "A" chiller is started. Scenarin ends when EOP-08 is
- entered. ,
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4 4
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. I U.S. Nuclear Regulatory Conunission Attcchment C 3F129747 Page 4 I
Sco==4 2 100% Power MCC-3AB is aligned to the B Bus. RCS leak develops in the B2 IIPI noule.
i leak starts at 50 gpm and increases to 200 gpm. Rx trip on low pressure or manual operator action. When Rx trips, a LOOP occurs including the 22KV backfeed to the air compressors. B EGDG falls to start due to loss of B Battery. E0P-02 is entered and immediate actions perfornwd. E0P-03 is entered on loss of SCM. The failed ilPI line is isolated. EFW is cross-tied to allow shutdown of EFP-1. The "A" chiller is started. Scenario ends when E0P-08 is entered.
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, U.S. Nuclear Regulatory Commission Attachment C l
. 3P129747 Page5 ;
i Se:---Le 171 hr Sopwe I1=> nunoure l Sinudator Seng/hogrammed Falures: l
- RTP l
- EOL ,
- No equipment OOS i e Pressurizer surge line ruptures (full sheer). l
- Rx is tripped by RPS on low pressure. l
- Access to the AB is los-; due to radiation levels. The PPO was not allowe:1 back in to l perform post event actions.
Scenado rs nonees:
i e SCM will be lost and EOP43 will be entered.
- All EIM will be lost due to the LOOP and the failure of EFP-2, but EIM is not required I for LHLOCAs.
- Time dependent actions for starting CREYs and the Chiller will be balanced with ECCS transfer to the sump.
l Frnectedprocedure usace.' l
- EGP42 inunediate actions will be pe fonned. !
- EOP43 will be entered due to a loss of SCM.
- EOP-13 Rules will be used as appropriate.
. . . =_ - _ __ ~ . -- .. -_
i
- U.S. Nuclear Regulatory Commission Attachment C
.3F1297 47 Page 6 Scenarks 176 Inst of SCM wkh no HPl i
%-t-sar srauniprogr--a pastures; }
e 100% RTP I e EOL :
- MUP 1 A R/T'd to Mechanics for PM and is ready to release clearance l e Small RCS leak develops (50 gpm) in Per steam space (.0067) ;
e Ink increases to 300 gpm (.(402) ,
e' Rx tripped by manual operator action or RPS l
MUP-1C fails to start l
- When Subcooling Margin Monitors are selected to incores, MUP-1B trips on motor !
overload. l
. Transition to no llPI branch in EOP-03 is made. .
- Scenario terminates when LPI flow is re-establist:ed. !
Scenado CA-II nges l e EOP-02 immediate actions, e EOP-03 Loss of SCM with no llPI branch e EOP-13 Rules 1,2,3 arul 4
. EOP-14 Enclosure 17 (CREVs) d B
1 a J
- . .. , - , . n.-c -. L ... ,, .,-,L ...
U.S. Nuclear Regulatory Commission Attachment C 3P1297-47 Page 7 Scenvia # 69 LOCA Cooldown with loss of A burry and Offsite Power Sinudator Setup /Programnted Failures:
e Small leak occurs in cold leg (200 GPM). l I
- Rx is tripped by opc <itor or RPS.
e Concurrent with Rx trip a loss of all off-site power c: curs.
- A-Ilattery failure occurs (single failure) l
- The leak in the cold leg increases to 0.006 of a full cold leg break. !
e Access to the AB is lost due to radiation levels. The PPO was not allowed back in to perform post event actions.
- When EOP-08 is entered, the cold leg break will increase in size to 0.1 of a full break.
ScenadtLChQlkBECE e Loss of the A train will require transfer of ES status lights, transfer of ES MCC 3AB and i closure of MUV-18 and 27 to allow determination of IIPI flows.
- Iess of A-Battery will also result in the use of the ES test switches to bypass ES to defeat the 6S 480V hxkout. This will permit the starting of CREVs and the chiller.
. When EOP 8 is entered, the cold leg break will increase significantly. Rapid depressurization of the RCS will occur. Building spray actuation may occur.
- Wher LPI flow reaches 1400 GPM, a transition to the large break branch of EOP-08 will occur, e The ES t .: switches will prevent a reactuation requiring manual LPI actuation. The reactuation of ES will require restarting CREVs and the Chiller, e Transfer if ECC5 ouction transfer to the sump will occur en the B train only, e The scenario will end when EOP-08 is exited and the TSC will be contacted.
- Boron precipitation control will not be lerformed due to only one train of LPI and faability to open drop line.
Expectedpmcedure usage:
- EOP-02 immediate actions will be performed.
- EOP-03 will be entered due to a loss of SCM.
- EOP-14 Enclosures 2,17,18, and 19 will be used.
-_, - , _ , m - -- ,, ,#
U.S. Nuclear Regulatory Commission Attachment C !
3F1297-47 Page 8 Scenario 175 Inss of SCM with tw EFW and Degrad-d HPl S[musanar Serun/Proerarnmed Failures:
- 100% RTP
- EOL
- FWP 7 Rff'd for motor rebuild
- Units 1 aix! 2 steam unavailable i
- A break occurs in the llPI line downstream of MUV-26 (App 200 gpm). When the Rx is tripped, a loss of offsite power occurs and A EDG fails to start. l
- MUP IC is degraded to a maximum of 40% output. EFP-2 fails to start. ;
- Operators perform the immediate actions of EOP-02 then transition to EOP 93. ES MCC 3AH is transferred to B side and MUV-18 and 27 are closed. Operator isolates broken llPI line to increase flow to the core but degraded MUP-lC is not adequate to remove core heat. :
With EFP-2 unavailable, operators will transition to EOP-04 and attempt to regain OTSGs.
After PORV is opened, ASV-50 is reset and OTSG cooling is established. Transition to EOP-08 is made. The scenario ends when MSIVs are isolated and operators are managing steam for EFP 2.
Scenario Challenges:
1
- Flowpath through EOP-03 wita degraded llPI
- Transition to EOP-04 when lack of heat transfer is recognized
- Transition from IIPl/PORV cooling to OTSG coeling
- Control of cooldown arxl steam management for EFP-2 in EOP-08 >
1
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U.S. Nuclear Regulatory Commission Attachment C
- 3F1297-47 Page 9 Scenario #78 Isss of B Bauery with Cold Leg SBLOCA Si=ulaint Selun/ Programmed Fauurig
- 100% RTP
- EOC Life
- No equipment OOS
- Small leak occurs in cold leg (200 gpm)
- Rx is tripped by operator or RPS
-
- Concurrently with Rx trip LOOP occurs
- B Battery failure occurs (single failure)
- 1.cak increases to 0.003 of a full cold leg break less of the B train will require transfer of ES status lights, transfer of ES MCC 3AB and 4 closure of MUV-18 and 27 to allow determination of IIPI flows. ES will be bypassed along with the ES 480V lockout. CREVs will be started per Enclosure 17. Because the A EDG is running, Enclosure 11 must be performed to cross-tie EFP-2 to EFP-1 control valves. EFP-1 will be shut down to make room on the EDG to start the CC Chiller per Enclosure 18.
The scenario will end when the TSC is contacted and EOP-08 is exited.
Fryettedprocedure usage:
- EOP-02 immediate actions
- e EOP-03 will be entered due to loss of SCM e F.OP-08 will be entered after completion of EOP-03
. EOP 13 rules as appropriate
. EOP f A Enclosures 2,11,17, and 18
i U.S. Nuclear Regulatory Commission Attachment C 3F1297-47 Page 10 .;
i Scenada 188 SBLOCA resuldng in loss of SCM and LOCA Cooldown ,
i l
Simulninr Setup /lYogmmmed Failures:
. Mode 1100% FP
. SBLOCA on *B2" IIPI line
- Ioss of offsite power '
- Iess of B" Battery :
e less of Berm air compressors -
Initially, only one RO and one SRO will be present in the Control Room. Second RO will not ,
be allowed to enter until 2 minutes after loss of SCM. Full Emergency Plan response is ;
required for NSS participation. STA will not be available until 10 minutes into the event.
Time critical actions will be timed. .
A SHLOCA develops on the "B2" IIPI line. When the reactor trips, a loss of offsite power B Battery and Beim air compressors occur. EOP-02 immediate actions are performed with transition to EOP-03 due to loss of SCM. Power is transferred to MUV-25 and 26. The high flow line is isolated and EFP-2 is cross connected to supply the A EFW train. Adequate SCM will tot be regained and transition to EOP-08 will occur. Scenario ends when hold point in EOP-08 is reached for RCS pressure < 400 psi.
Scenario Challengts: '
e Flow path from EOP-02 to 03 to 08 e EFP 2 management ,
o EDG-1 A load management 1 3
w-- , v-,_-. - , -- , . , _ . , . , , -
, U.S. Nuclear Regulatory Commission Attachment C 3F1297-47 Page 11 Set ===io 197 Cold lig Break (Fuu Shgar)
.ct-~lanar Segun/Progr==med Failures:
e 100% FP, EOL e Nothing OOS e Cold leg break (full shear) e DilV-42 fails to open during sump swapover Break is inserted. Rx trip by RPS. AB access is lost due to radiation levels which won't allow PPO in to perform post trip actions. EOP-02 immediate actions will be perfonned with transition to EOP4)3 due to loss of SCM. EOP-08 will be transitioned to from EOP-03 to perform sump swapover.
Strantic Challenges.*
DilP-1 A, ilSP 1 A and A Train MUP must be tripped during sump swapover due to failure of DilV 42 s
U.S. Nuclear Regul: tory Commission Attachment C 3F1297-47 Page 12 Sta== ia illi KnfDCA & EFP-2 failure
.ct-uinant brun/1%er---d Failurtu
. 100% power, MOL e No equipment DOS e RCS leak occurs on B2 IIPI line (Full double ended shear) e Loss of offsite power occurs concurrent with Rx trip e EFP-2 trips on overspeed (single failure)
Expectedprocedure ucare;
-* EOP 13
- - EOP-14 . Enc. 2,11,17, and 18 Scenario Challenges:
Broken llPI line will be isolated by EOP-03. EDG-1 A load management will be performed per Enc.11 to shutdown RWP-2A, SWP-1 A, and defeat the 500 psig trip of EFP-1. CREVS and CC Chiller will tv: started on the B side uilng Rule 5.
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FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 ATTACHMENT D OPERATOR ACTIONS WITII PRIOR NRC REVIEW
Attachment D, U.S. Nuclear Regulatory Commission Page1 .
3F1297-47 Prior NRC R;;;.ese OA Operator Action Time Basis Review 1 Trrp ail runnog RCPs < 2 mnutes Requk ed for loss Yes NRC letter to FPC dated 5/29/S6 (Gawc Letter 66-05) refes to B&W of subcoohng Owners Group (BWOG) studies which conduded that comphance with to margin based on CFR 50 46 is achewed if operator aden to tnp RCPs is taken withm 2,.
voiding condmon enutes. NRC to FPC letter dated 3/15/89 (3NO389-11) stated FPC har of reactor coolant adopted and Wplar,w-aed the previously approved (GL 86-05) BWOG methodologies for RCP tnp and dosed NUREG 0737 ftem II.K.3.5.
2 If Subcoohng Margrn (SCM) is lost 10 minutes Required for loss Partial NRC lettes to FPC dated 75/79 (SER for Order dated 5/16/79 based on and ES . has not actuated, initiate of subcochng TMI-2 Acodent) recognizes CR-3 revision to Emergency Procedure EP- ,
manual HPI and Reactor Building margin (precedes (Manual 106, which defines opera;or action in response to a spectrum of break isolaton and Coohng (RBIC) automatic initiat on of sizes. States EP-106 was pdged to provide adequate guidance to the l inrtiaton) RBIC has operators to cope with small break LOCA? EP-106 (currently EOP-03. I not been " Loss of Subcooling Margin ~) contained guidance to inrtsate HPI and reviewed) ensure adequate HPl flow. In letters dated 8/24/79 and 11/14/79 FPC
-is:Aates letdown (USO6) provided its responses to Short Term Actons Items 2 and 5 of I.E.
Bulletm 79-05C. In toth FPC letters, the plant behavior dunng a
- initiate? HP1 flow SBLOCA is desenbed The analysis provided desenbes meidents in which automatic ES actuation would not occur and the need for operator
- isoletes normal makeup (USO6) actons to initiate HPl. The 11/79 letter provides operatmg guadelmes for (contingency actons are provided SBLOCAs. The list of symptoms and indicatons of SBLOCA include a in OA 4 wrthin 20 minutes if power statement that ES actuaten may not occur for all SBLOCAs in letter is not available) dated June 3.1981, the NRC stated that FPC's responses to IE Bulletc 79-05C would be reviewed as part of Task Acton Plan item IIK3.5. NRC to FPC letter dated 3/15/89 (3NO389-11) stated FPC has adopted and
- isolates RCP seal control bleed off vanes implemented the prevtously approved (GL 86-05) BWOG methodologes f r RCP trip and closed NUREG 0737 Item Il K.3.5.
- actuates EFIC
-initiates Emergency RB cooling 3 Ensure all four HPI injection valves 10 minutes Required only for Yes NRC letter to FPC dated 5/29/79
- Permanent Soluton to SBLOCA issue' are open loss of 1 train of recognizes operator action to tum associated transfer switch to open Class 1E power affected HPI valves by 10 mmutes.
- switch power supply for affected injection valves by manipulating switches in control room
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t Attachment D, U.S. Nuclear Regulator, Commission Page 2 3F129M7 .
Reference OA Operator Action Tsme Basis Prior NRC Rev'ew isolate RCP seal Mjechon (USQ6) 20 minutes Required to No FPC lettsr to NRO dated 2/28/79, answers a prevous queston of 4
maximize HPI whether or not it was necessary to isolate any flow paths in the ma6eup flove to reactor system after a LOCA. FPC tefers to RCP seal 6ecten 1 erd normal makeup and refers to a Gnbert AsssMes report that concludes Jguate HPi flow is achieved without these hetes isolated. NRC letter to hcensees with B&W designed systems (Genenc Letter 86-05) dated (As a contingency action, sf pr is ST29/86 states the coohng water sources supporting the RCP with the lost 1: MilV-27 (normal mabup) aM potential of being isolated are seal injectc), seat bleedeff Osp06(,it MUV i8 (RCP sealinjection) transfer coc!ing water to seat line coolers. and wmponent coohng water to RCP to an energized bus and close valves) motors and oil coolers. The need to isolate RCP Seal injechon was discovered in 1995 to be necessary due to discovery that operators rehed on non-Reg Guide 1.97 instrumentation to measure this flow when determning HPl pump runout flow hmits (see LER 95-026). Seal injection isolation was also determined necessary during Refuel 10 in 1996 upo.i discovery that worst case instrument error may result in inadequate HP1 flow (see LER 96-006). FPC letter dated July 7,1997 (NOV 96-07) discusses the additional need for closure of RCP sea!
controlled bleed-off (CBO) valves after 90 seconds if seal injecten has I
' riot been restored. See OA-2.
w ,
.V +E s. Ca. a L .~
,. U.S. Nuclear Regulatory Commission Attachment kh 3F1297-47 Page 3 .
OA Operator Action _ Time - Basis Prior NRC Reference Review 5 Ensure adequate HPi flow (USQ6) '20 minutes Reqwred on!y for Partaaf FPC letter to NRC dated 10/27/o' 9 states HPI must be soccessfu#y
{ isolate a broken injection line using break in HPiline balanced to support SBLOCA mitigation as desenbed in vanous B&W new isolation enteria) (BalaM#g topical reports accepted by NRC. Subsequent FPC letter dated 10'31/89 replaced states that tr@gation strategy employed from the late 19717s through the by reviews done en response to NUREG 0737 relied on balancmg HPl flow isolation) for breaks in HPI injection lines. These letters relate to LER 89-037, issued in November 1989 reporting a design basis condition in which instrumentation used for balancmg HPI flow was inadequate NRC letter dated 12/20/89 confirmed verbal concurrence to resume power operabon with the HFI instrumentation problems One w,-4aion was operator acton for HPI flow balancing. NRC letter dated 2/17/95 from Gary Holahan '.o Ed Jacks (BWOG Operator Support Committee) states staff .
has completed its reWw of BWOG response to NUREG 0737 P m I.C.1 regarding EOP Guidelires and is finalizing an SER on the topic.
Balancing HPl flows was a part of the ATOGITBD guidelmes incorporated into FPC procedures. FPC issued LER 96-007 on 3/15/96 to report another design basis condition involving l ' HPI flow instrumentation The flow deficences described therein were addressed by revised SBLOCA analyses provided by Framatome Technologies in April 1996 which required isolation of the affected HPt line versus balancing. Most recent FTl analyses have provided new isolation criteria (50 gpm vs. 75 gpm).
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U.S. Nuclear Regulatory Commission Attachment Dt-Page 4 . ,
" Operator Action Time Basis Prior NRC Refereiwe l OA Review 6 Er.sure adequate EFW flow (USO6) 20 minutes Ratse OTSG Yes B&W (Taylor) letter *o NRC (Baer) dated 5/1/78 ru,Gs i@cid iePG1 levels to ISCM 10104, '8&Ws ECCS Evaluabon 'Jodel," which notes operator acton is setpoint (90'ya necessary dunng earty stages of the accident to trutigate consequences and meet 10 CFR 50 46. Auxiliary feedwater is assumed to be available.
(EriC was enthated in OA2; therefore, NRC letter to FPC dated 7Ex79 provides a SER for actions taken in ensuring EPV flow is a confim.ation response to Commtssion Order dated 5/16779. The SER states that a step only) geac review of B&W anatyses endt!ed
- Eval < 2ation of Transient Behaver and Small RCS Breaks in the 177 Fuel Atsembty Plant" resulted in a principle finding that reconfirms SBLCCA analyses demonstrate a combination of heat removal by the steam generator and This step manually raises OTSG the HPI system combined with operator action to ensure adequate core levels to the inadequate Subcoohnt, coohng These results are apphcab!e to CR 3 cons:dering the abihty to Margin. ISCM level manua!!y start the redundant EFW pumps and HPl pumps from the control room. assuming failure of automatic EFW actuaton. NRC letter to FPC dated 8/30M5 provides a SER for NUREG 0737 Item I!X320.
"SBLOCA Me* hods? Sect:on 111.5 a of the SEF. "tates De tmag of operator actior, to raise the secondary system water level to 95% was l
found not to be entical?
The safety evaluaton cu wnpenying issuance of Amendment 104 (NRC 7 Ensure Control Complex Ventiation is 30 minutes To assure control Pertial runnirv)in emergency mooe room operator (time rett letter to FPC dated 2/19/88) states "The staff reviewed these scenanos dose is not not [SBLOCA among them], including equipment operaton and exceeded and tc, previously provide cc tol reviewed) systems ficws, and found them appropriate and complex coolnig.
Required to be therefore acx:eptable. The SER recogued that manually adding control complex !ans to the EDG can be done after other load icw.irpucd accomplished within 30 minutes acboru have been bken. CR-3's FSAR Secton 9.7.2.1 desenbes the normal and emergency modes of operaton of the Plant Ventilabon Systems. Subsecten g.1 desenbes the emagm-(y reorculaton mode of g
operaton initiated by radiaton monitor RM-A5 which trips the normal duty supply unds and states "the emergency fans AHF-18A and AHe-188 must be manuatry started from the control rty C The results of engineenng calculation H97-0001 indecate the control cor plex ventdation can be oct of sennce for 30 minutes coircident with the control complex chilled water system being out of service for 80 minutes without impainng the abehty of these systems to maintain the control complex temperature witrdn the ermronmental quahfication fmts specified.
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iU.S.jNuctedr Regulatory Commission ? Attachment Bh .
~ 13F1297-47? .. Page5 .;
OA Operator Acthm Twne Beeis Prior NRC Reference
- Revsow ,
3
- 8. If at any tune 9WST is < 20 ft, transfer Funcbon of To ensure Yes improved Tecroical Specificaton (ITS) BASES 3.52 stabs. "When Sie, ECCS pump suchon to RB sump Inventory sufficent source BWST has been nearly empbed, the.suchon for the LPI pumps'is manuany transferred to the reactcr building emergency sump? ' FSAR of borateo water -
forinfection by Sechon 6.1.2.1.2 states that *When the BWST inchested level reaches 15 HPl/LPl fL, the operater will Me schon to open the LPI System suchon volves Depenchng on from the reactor busidmg emergency sump, pemisitng reorculebon of Wie .3 break size, spilled reactor coolant,and inpecton weler from the reactor buBding J achon may be . sumpf EOP-03, Step 3.13 starts the transfer prowss at 20 fL so Wiet Wie 1 required between transfer can be accomplished by 15 ft 25 mmutes and '
1-1/2 hours '
9 If "B* DC power is lost, crosstie EFP-2 . EDG-A Load : EFP-1 can only No ,
to A train (EFV-12) - ManagemerW provide flow for a ;i (Prior to specific time ,
manuaNy period, then loading Chaler) EFP-2 must be AND aligned to ensure sufficent margin is maintained on the 'A' EDG for - -
Secure EFP-1 i later adding of Control Complex :
ChiNer - i 10 - Put EFIC in manual pemiessive . ' EDG-A Load Requeed to No .
Management prevent cycling ..
(Prior to of thelimited '
manuaily duty motors on AND loading ChiNer) the EFW block ralves Close EFW block valves (deenergized after c6sure) - 1
....mm ..,m... ____....___m____2._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -_____m.. 2 ___ . ___.,_ m,o # e , , . . r-r 3 , iv.,%.,_,g,, ,,m. , , . ,, ,,yi,,3 r. 3 ,.-,.,,g.g.. .gg,-
Attachment D, U.S. Nuclear Regulatory Commission Page 6 .
3F1297-47 Time Basis Prior NRC Reference CA Operator Action Review 11 Manage EDG load in order to extend EDG-A Load Defense in No EFP-1 operation by - Management Depth action for (Prior to postulated singte
. Shutdown SWP-1A & RWP-2A manua!!y failure of theloss after verifying redundant pumps loading Chiler) of EFP-2. These are operati.g and placing actions extend switches in Pull-to-Lock to the trme EFP-1 is prevent reactuation of pumps available for (EDG loading) CTSG cooling Place EFP-1 Trip Defeat Switch in defeat position to prevent automatic trip of EFP-1 on RCS pressure of 500 psig 80 Minutes Required within Partial Pnor to Revisen 23. She CR-3 FSAR. Table 8-1, ~ EMERGENCY DIESEL 12 Venfy Control Complex Chsller is running 80 minutes to (time req't GENERATOR A* AUTO & MANUALLY CONNECTED LOADS.*lisW the Cnsure centrol not Cc, trol Complex Chiller as a manuaffy connected load.
complex previously instrumentation reviewed) remains within analyzed temperature j ranges for instrument accuracy l
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o U.S. Nuclear Regulatory Commiosion Attachmeru Dt Page 7 .
' Prior NRC Reference I OA Operator Action Time Basis Review l
t Isolate the RB sump by placing RB Sequence Required to Yes Required by IE Buttetin 79-05A and appicved by NRC in letter dated 13 i
Action Prior to isolate unneeded June 3,1981. (3N0681-34). Required by IE E917etin 79-05A to isolate stanp pumps in Pull-to-Lock, closing RB sump pump discharge valves, and Cooldown penetration flow systems utilized to trans'er radioactive liquid and gases from the paths. These containment. These penetrations go to the waste gas header and the closing waste gas header isciation penetrations go Miscellaneous Waste Storage tank. Isolating the RB sump penetrateon valves to the waste gas will maintain inyt.atory in the containment for possible ECCS pump header and the suction for long term recirculation Miscellaneous Waste Storage tank. Isolating the RB sump penetration will maintain inventory in the containment for i possible ECCS pump suction for long term recirculation 14 If only EFP-2 is supplying feedwater Function of To maintain EFP- f>a Maintaining steam generator pressure in order to continue the operation to the OTSG, the RCS cooldown will Cooldown 2 as an avakble of the steam-driven emergency feedwater pump has been reviewed by be stopped prior to reaching an EFP-2 Rate scurce of the NRC for conditions involving inadequate core cooling. Refer to FPC feedwater and letter dated November 14,1979 (3F1179-25) and NRC letters dated operational limit. Manage operation of EFP-2 t:y closing ASV-5 and ASV- operate the June 3.1981 and March 15.1989.
204 on low OTSG pressure (Cycle pump within EFW) and restart EFP-2 when analyzed pressure increases. regions. Use of FWP-7 orovides addi' ca d resources (Mitigation strategy includes operation available to of diesel backed FWP-7 as a Defense operators during in Depth action. Isolation valves a LOOP.
located in the intermediate building may be opened to allow use of FWP-7)
a U.S. Nuclear Reguictory Commission Attachment D, 3F1297-47 Page 8 .
OA Operator Action Time Basis ~ Prior NRC Reference Review 15 If EFP-2 is not operating when in a Function of If EFP-2 is not No LOOP condition with inadequate Cooldown available, steps subcooling, limit coo!down prx to Rate must t e taken to EFP-1/LPl. ensure EFP-1 operates as ing as needed 16 Establish RCS Cocidown using TBVs Cooldown initiates RCS Yes ITS bases B3.7.4 states in part that the TBVs provide a method for or ADVs Rates are cooldovm to cooling the plant to Decay Heat Removal (DHR) System ef fJy CundItions provided in achieve end via the main condenser. Fo!!owing an accident, this is done in Table 3 of point of event conjunction with the Emergency Feedwater (EFW) System. providing EOP-08 (start of decay flow from the EFW tank (EFT-2). In the event of a LOOP. the heat) Atmosphe-ic Dump Valves (ADVs) would be relied upon to perform the secondary side heat removal function. Analysis performed during initial licmsing demon $ated that the ADVs can be used to cool down the plant and still mee 0 CFR 100 limits, although the offsite dose would be significantly higher than tnose associated with a TBV-based coc:down.
The ADVs are air-operated valves equipped with pneumatic controllers to permit control of the cooldown rate.
17 Periodically re-evaluate HPl line break Contingency Required for No criteria on RCS repressurization specific HPlline pinch areas to ensure a broken line will be isolated if isolation criteria is not met early in the event while in ECP-3
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