ML17333A565
| ML17333A565 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/18/1996 |
| From: | John Hickman NRC (Affiliation Not Assigned) |
| To: | Fitzpatrick E INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| References | |
| NUDOCS 9609200290 | |
| Download: ML17333A565 (24) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 18, 1996 5o-3I5 Mr. E.
E. Fitzpatrick, Vice President Indiana Michigan Power Nuclear Generation Group 500 Circle Drive
- Buchanan, MI 49107
SUBJECT:
REVIEW OF PRELIMINARY ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF AN EVENT AT D.C.
COOK
Dear Mr. Fitzpatrick:
Enclosed for your review and comment is a copy of the preliminary Accident Sequence Precursor (ASP) analysis of an'perational event which occurred at D.C.
Cook on September 12,
- 1995, (Enclosure 1),
and was reported in Licensee Event Report (LER) No. 315/95-011.
This analysis was prepared by our contractor at the Oak Ridge National Laboratory (ORNL).
The results of this preliminary analysis indicate that this event may be a precursor for 1995.
In assessing operational
- events, an effort was made to make the ASP models, as realistic as possible regarding the specific features and response of a given plant to various accident sequence initiators.
We realize that licensees may have additional systems and emergency procedures, or other features at their plants that might affect the analysis.
Therefore, we are providing you an opportunity to review and comment on the technical adequacy of the preliminary ASP analysis, including the depiction of plant equipment and equipment capabilities.
Upon receipt and evaluation of your comments, we will revise the conditional core damage probability calculations where necessary to consider the specific information you have provided.
The object of the review process is to provide as realistic an analysis of the significance of the event as possible.
In order for us to incorporate your comments, perform any required reanalysis, and prepare the final report of our analysis of this event in a timely manner, you are requested to complete your review and to provide any comments within 30 days of receipt of this letter.
We have streamlined the ASP Program with the objective of significantly improving the time after an event in which the final precursor analysis of the event is made publicly available.
As soon as our final analysis of the event has been completed, we will provide for your information the final precursor analysis of the event and the resolution of your comments.
In previous years, licensees have had to wait until publication of the Annual Precursor Report (in some cases, up to 23 months after an event) for the final precursor analysis of an event and the resolution of their comments.
9609200290 9609%8 PDR ADQCK 050003%5 P
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v September 18, 1996 We have also enclosed several items to facilitate your review.
Enclosure 2
contains specific guidance'for,performing the requested review, identifies the criteria which we will 'apply to determine whether any credit should be given in the analysis for the use of licensee-identified additional equipment or specific actions in recovering from the event, and describes the specific information that you should provide to support such a claim.
Enclosure 3 is a
copy of LER No. 315/95-011, which documented the event.
Please contact me at (301) 415-3017 if you have any questions regarding this request.
This request is covered by the existing OHB clearance number (3150-0104) for NRC staff followup review of events documented in LERs.
Your response to this request is voluntary and does not constitute a licensing requirement.
Sincerely, i/rigina1 signed by:
t t
John B. Hickman Project Manager Project Directorate III-1 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation Docket No. 50-315
Enclosures:
1.
Preliminary ASP Analysis 2.
Review Guidance 3.
LER No. 315/95-011 cc w/encl:
See next page DISTRIBUTION:
IDocket Fi 1 e RIngram PUBLIC PD 3-1 Reading RZimmerman JRoe EAdensam (e)
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E. Fitzpatrick Indiana Michigan Power Company CC:
Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, Illinois 60532-4351 Attorney General Department of Attorney General 525 West Ottawa Street Lansing, Michigan 48913 Township Supervisor Lake Township Hall P.O.
Box 818 Bridgman, Michigan 49106 Al Blind, Site Vice President Donald C. Cook Nuclear Plant 1 Cook Place Bridgman, Michigan 49106 U.S. Nuclear Regulatory Commission Resident Inspector's Office 7700 Red Arrow Highway Stevensville, Michigan 49127 Gerald Charnoff, Esquire
- Shaw, Pittman, Potts and Trowbridge 2300 N Street, N.
W.
Washington, DC 20037 Mayor, City of Bridgman Post Office Box 366 Bridgman, Michigan 49106 Special Assistant to the Governor Room 1 - State Capitol Lansing, Michigan 48909 Drinking Water and Radiological Protection Division Michigan Department of Environmental equality 3423 N. Hartin Luther King Jr Blvd P. 0.
Box 30630 CPH Hailroom
- Lansing, Michigan 48909-8130 Donald C.
Cook Nuclear Plant Mr. Steve J.
Brewer Indiana Michigan Power Nuclear Generation Group 500 Circle Drive
- Buchanan, Michigan 49107 Aught 1996
LER No. 315/95-011 LER No. 315/95-011 Event
Description:
One Safety Injection pump unavailable forsix months Date ofEvent:
September 12, 1995 Plant:
D. C. Cook, Unit 1 Event Summary During a surveillance test with the unit shutdown in Mode 6, personnel determined that the West Centrifugal Charging Pump (CCP) had been inoperable for about six months. This was the result of an incorrect breaker relay calibration performed six months earlier. The unavailability ofthe West CCP primarily affects the units'esponse to steam generator tube rupture (SGTR) and small-break loss ofcoolant accident (SLOCA) events.
The estimated increase in core damage probability for this event is 7.7 x 10~, over a nominal value for the same period of2.9 x 10'.
Event Description On September 12, 1995, the plant was shutdown in Mode 6 when personnel started the West CCP in order to perform the "ECCS Full Flow Test" surveillance.
The West CCP provides injection flow on the receipt of a Safety Injection (SI) signal. Afler operating at full flow for seven minutes, the pump tripped on motor overcurrent.
Personnel determined that the pump tripped because the setting of the 1-51-TA8 time overcurrent relay was incorrectly set. Itwas determined that this relay was last calibrated on March 15, 1995 (180 days prior to the full flow test).
This effectively rendered the West CCP inoperable for the preceding six months.
During the review of this event, the Instrumentation and Control (f&,C) technicians involved in calibrating the relays demonstrated how they typically dctcrmine thc relay pickup current.
Because their technique was incorrect, the relays were miscalibrated.
Both INC technicians involved with the relay calibration were Enclosure 1
LER No. 315/95-011 trained and qualified within the D. C. Cook Nuclear Plant relay training program.
However, it was determined that a significant amount of time elapsed between the end of the training program and the time that the 1-51-TAS time overcurrent relay associated with the West CCP breaker was incorrectly calibrated.
Additional Event-Related Information During normal plant operation, both charging pumps (East and West) are configured for their charging function.
One charging pump is suQicient to supply full charging flow and reactor coolant pump seal injection during normal leakage and normal letdown conditions.
A third positive displacement charging pump is available, but is not normally used.
On receipt of a valid SI signal, the CCPs operate in the high pressure injection mode.
D. C. Cook also has a separate SI system.
This system, with two pumps operating in parallel, operates in an intermediate pressure injection mode. The two SI pumps deliver flow&om the Refueling Water Storage Tank (RWST) at a maximum injection pressure of approximately 1100 psig.
The residual heat removal (RHR) pumps can be aligned for recirculation from the containment sump to the suction of either the SI pumps or the CCPs.
The licensee indicated that the East CCP had been inoperablc for less than 18 h during the six-month period that the West CCP was unavailable.
Additionally, the emergency diesel generator (EDG) supporting the East CCP was unavailable for less than 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> during the six-month period that the West CCP was unavailable.
Modeling Assumptioas The CCPs were subject to common cause failure during this six-month period based on incorrect maintenance practices.
Because the success criteria in the IRRAS model assumes both CCPs are required for success of the CCP portion ofthe high pressure injection function in response to either a SLOCA or a SGTR, no further changes werc required to model the increased potential for common cause failure. Success ofone ofthe two SI pumps also ensures success of the HPI function in the IRRAS model, independent of the success of the
LER No. 315/95-011 CCPs.
This is not as stringent as the assumption of the plant Individual Plant Examination that one oftwo CCPs and one oftwo SI pumps are required in response to a SLOCA.
The IRRAS response to a SGTR was modified. Previously, a loss ofthe high prcssure injection function lead directly to core damage.
The possibility of lowering RCS pressure below the Steam Generator safety valve setpoint within 30 minutes was allowed following the loss of high pressure injection capability.
Based on the operator burden under a short time constraint, a failure probability of 0.1 was assigned to the new basic event, PCS-XHE-DEPRES.
Loss of offsite power (LOOP) sequences were prominent in Case 2 with only one EDG available.
LOOP probabilities of short-term and long-term offsite power recovery, and the probability of a reactor coolant pump (RCP) seal LOCA following a postulated station blackout were developed based on data distributions contained in NUREG 1032, Evaluation ofStation Blackout Accidents at Nuclear Power Plants.
The RCP seal LOCA models were developed as part ofthe NUREG 1150 PRA efforts. Both ofthese are described in Revised LOOP Recovery and PWR Seal LOC'AModels, ORNL/NRC/LTR-89/11, August 1989.
Analysis Results Determining the overall increase in the core damage probability involved determining the increase in the core damage probability for three different cases and then summing these cases together.
The three cases are:
Case 1
the increase in the core damage probability due to the long term unavailability of the West CCP (4252 h).
Case 2 the increase in the core damage probability due to the unavailability of the opposite train EDG that was periodically unavailable while the West CCP was unavailable (50 h).
Case 3 the increase in the core damage probability due to the time that both CCPs were simultaneously unavailable because ofvarious maintenance activities (18 h).
Combining the probability estimates for the three cases results in an overall increase in the core damage probability for the 180 day period of 7.7 x 10
. Most of the increase (56%) is driven by the long term
at
LER No. 315/95-011 unavailability ofthe West CCP (Case I). An additional 44% ofthe increiise in core damage probability was added by Case 2. The dominant core damage sequence, highlighted as sequence number 6 on the event tree in Figure 1, contributes approximately 44% to the combined increase in the core damage probability estimate for all three modeled cases.
Sequence number 6 involves:
~
a SLOCA,
~
the successful trip ofthe reactor,
~
the successful operation ofthe AuxiliaryFeedwater (AFW) system, and
~
the failure ofthe High Pressure Injection (HPI) system to provide suKcient cooling Qow.
The next most dominant sequence involved a LOOP and contributed approximately 13% to the combined increase in the core damage probability estimate for all three modeled cases.
The nominal core damage probability over a six-month period estimated using the Accident Sequence Precursor (ASP) models for D. C. Cook is approximately 2.9 x 10~.
The failed West CCP increased this probability by 28% to 3.7 x 10'.
The latter value (3.7 x 10 ') is the conditional core damage probability (CCDP) for the six-month period in which the West CCP was inoperable.
For most ASP analyses of conditions (equipment failures over a period of time during which postulated initiating events could have occurred), sequences and cut sets associated with the observed failure dominate the CCDP (i.e., the probability of core damage over the unavailability period, y'ven the observed failures).
The increase in core damage probability because of the failures is, therefore, essentially the same as the CCDP, and the CCDP can be considered a reasonable measure ofthe significance of the observed failures.
However, for this event, sequences unrelatc;.'o thc failure ofthe West CCP dominated the CCDP estimate.
The increase in core damage probability given the West CCP inoperability, 7.7 x 10, is, therefore, a better measure ofthe significance ofthe failure ofthe West CCP.
Definitions and probabilities for selected basic events are shown in Table 1. The conditional probabilities associated with the highest probability sequences for the condition assessment are shown in Table 2.
The sequence logic associated with the sequences listed in Table 2 are given in Table 3.
Table 4 dcscribcs the
LER No. 315/95411 system names associated with the dominant sequences for the condition assessment.
Minimal cut sets associated with the dominant sequences for the condition assessment are shown in Table 5.
Acronyms AFW ASP CCDP CCP ECCS EDG HPI HPR IRRAS IR,C LOCA LOOP MOV PRA RCP RHR RWST SGTR SI SLOCA Auxiliary Feedwater'ccident Sequence Precursor Conditional Core Damage Probability Centrifugal Charging Pump Emergency Core Cooling System Emergency Diesel Generator High Pressure Injection High Pressure Recirculation Integrated Reliability and Risk Analysis System Instrumentation and Control Losswf-Coolant Accident Loss ofOIIsite Power Motor-Operated Valve Probabilistic Risk Assessment Reactor Coolant Pump Residual Heat Removal Refueling Water Storage Tank Steam Generator Tube Rupture Safety Injection Small-Break Loss'-Coolant Accident References
LER No. 315/95-011 LER 315/95/011, Rev 0, "West Centrifugal Charging Pump Inoperable Due to InaMity to Meet Design Basis Requirements for Six Months as a Result ofPersonnel Error During Relay Calibration,"
November 20, 1995.
Indiana Michigan Power Company, Donald C. Cook Nuclear Plant Individual Plant Examfnatlon Summary Report.
Indiana Michigan Power Company, Donald C CookNuclear Plant Final Sajetyr1nalysis Report.
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lt4LOCA SEACTOS TED IIAII fEEOVTATES STSTISI fttD ANO STEED COOISIO fSS SECTSIOAAY COOIJIO AECOVEAEO SCS COOlDCHW USSIO 7 SVa SESSSIAL IIEAT SEIIOVAL SEQ ¹ END STATE OK OK CD OK CD CD OK OK 9
10 11 12 13 CD OK CD CD OK 14 15 16 17 18 19 20 21 CD OK CD OK CD CD CD COOK1, ASP PWR 8 SMALLLOCAEVENT TREE
LER No. 315/95-011 Table 1. Definitions and probabilities for selected basic events for LER No. 315/95-011 Event name CVC-MDP-FC-IA HPI-MD'-ALL HPI-MDP-FC-IA HPI-MDP-FC-IB HPI-MOV4C-SUC HPI-MOV~RWST HPI-XHE-NOREC HPR-XHE-NOREC PCS-XHF DEPRES RHR-MD'-ALL RHR-MDP-FC-IA RHR-MDP-FC-IB RHR-MOVE-SU C1 RHR-MOVE-SUC2 RHR-MOV~RWST RHR-XHE-NOREC Description Failure ofCharging Pump A HPI Pump Common Cause Failures HPI Motor-Driven Pump A Fails HPI Motor-Driven Pump B Fails HPI Serial Component Failures Failure to Isolate RWST From HPI System Operator Fails to Rccovcr thc HPI System Operator Fails to Recover the HPR System Failure to dcprcssurizc thc RCS within30 minutes RHR Pump Common Cause Failures RHR Motor-Driven Pump IA Fails RHR Motor-Driven Pump IB Fails Failure ofRHR Hot Lcg Suction MOVA Failure ofRHR Hot Lcg Suction MOVB Failure to Isolate the RWST During RHR Operator Fails to Recover thc RHR System Base probability 9.0 E404 7.8 FA04 3.9 FA03 3.9 E403 1.4 FA04 3.0 E403 8.4 E401 1.0 E+000 1.0 E401 4.5 E404 4.1 E403 4.1 FA03 3.0 FA03 3.0 E403 3.0 E403 1.0 E+000 Current probability 1.0 E+000 7.8 E404 3.9 E403 3.9FA03 1.4 E404 3.0 E403 8.4 E401 1.0 E401 4.5 E404 4.1 E403 4.1 E403 3.0 E403 3.0 M$3 3.0 FA03 Modified for this event No No No No No No No No No No No No No No No
LERNo. 315/95-011 Table 2. Sequence conditional probabilities for LERNo. 315/9&411 Event tree name SLOCA SGTR SLOCA TRANS SLOCA Sequence name 06 08 03 08 05 3.3 E-06 5.4 E-07 2.2 E-06 8.9 E-08 9.6 E-08 2.9 E-08 4.8 E-09 2.0 E-06 7.9 E-10 3.2 E-08 Change to CCDP (Importance) 5.4 E-007 1.6 E4)07 8.8 E-008 63 E-008 Percent Contribution to Importance 77.2 12.5 3.7 2.0 1.4 Subtotal Case 1
(shown)'ubtotal Case 2b Subtotal Case 3'otal (all sequences) 6.3 E-06 3.6 E-06 2.7 E-08 9.9 E-06 2.0 E-06 2.4 E-07 8.9 E-09 2.2 E-06 43 E-006 3.4 E-006 1.8 E-008 7.7 E-006
'ase 1 rcprcscnts thc increase m thc core damage probability due to thc long term unavailabiTity ofthe West CCP (4252 h).
Case 2 represents the increase in the core damage probability due to thc unavailability ofthc opposite train EDG that was periodically unavailable while the West CCP was unavailablc (50 h).
Case 3 rcprcscnts the increase in the core damage probability due to the time that both CCPs were simultaneously unavailable bccausc ofvarious maintenance activities
(lg h).
LER No. 315/95-011 Table 3. Sequence logic for dominant sequences forLER No. 315/95-011 (Case 1 only)
Event tree name SLOCA SGTR SLOCA TRANS SLOCA Sequence name 06 08 03 08 05 Logic
/RT, /AFW, HPI
/RT, /AFW, HPI, RCS-SG-H
/RT, /AFW, /HPI,
/COOLDOWN, RHR, HPR
/RT, /AFW, PORV, PORV-RES, HPI
/RT, /AFW, /HPI, COOLDOWN, HPR Table 4. System names for LER No. 315/95-011 (Case 1 only)
System name COOLDOWN HPI HPR RCS-SG-H RT Logic No or InsuQicient AFW Flow RCS Cooldown to RHR Pressure Using Turbine-Bypass Valves, etc.
No or Insufficient Flow from HPI System No or InsuQicient HPR Flow No or InsuQicient Flow from RHR System Failure to depressurize the RCS below the SG safety valve setpoint without HPI Reactor Fails to Trip During Transient 10
LER No. 315/95-011 Table 5. Conditional cut sets for higher probability sequences for LER No. 315/95-011 Cut set No.
Percent Contribution CCDP importance)'LOCA Sequence 06 83.1 14.9 1.6 SGTR Sequence 08 83.1 14.9 1.6 SLOCA Sequence 03 86.4 3.2 1.7 1.7 1.7 TRANS Sequence08 SLOCA Sequence 05 Subtotal Case 1~
(shown above)
Subtotal Case 2'ubtotal Case 3'.3 E-006 2.8 E-006 5.0 E-007 5.4 E-008 5.4 E-007 4.5 E-007 8.0 E-008 8.6 E-009 1.6 E-007 1.5 E-007 6.2 E-009 3.6 E-009 3.6 E-009 3.6 E-009 8.8 E-008 6.3 E-008 4.3 E-006 3.4 E-006 1.8 E-008
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HPI-MDPCF-ALL,HPI-XHF NOREC HPI-MOV4C-SUC, HPI-XHE-NOREC HPI-MDP-FC-IA.HPI-MDP-FC-IB, HPI-XHE-NOREC 4K': '(i%:@%Ãk5@><NxN>C'4.:0>+SVpA:;P4xAAÃ<98>hN~':~PM?MNV+><"I<">N5%~44 HPI-MDPCF-ALL,HPI-XHE-NOREC, PCS-XHE-DEPRES HPI-MOY4C-SUC, HPI-XHE-NOREC, PCS-XHE-DEPRES HPI-MDP-FC-IA,HPI-MDP-FC-IB, HPI-XHF NOREC.
PCS-XHE-DEP RES RHR-MDPZF-ALL,RHR-XHE NOREC, HPR-XHE-NOREC RHR-MDP-FC-IA, RHR-MDP-FC-IB, RHR-XHF NOREC, HPR-XHE-NOREC HPI-MOV4O-RWST, RHR-MORC-SUC2, RHR-XHE-NOREC, HPR-XHF NOREC HPI-MOV4O.RWST, RHR-MOV4O-RWST, RHR-XHF NOREC, HPR-XHE-NOREG HPI-MOV4ORWST, RHR-MOV4C-SUCI, RHR-XHF NOREC, HPR-XHE-NOREC i;4~@jjkQj+%~!:-'.".'3(+YA4%4M%~%QIt">Ã*4iop4'j PA%.
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LER No. 315/95-011 Table 5. Conditional cut sets for higher probabiTity sequences for LER No. 315/95-011 Cut set No.
Percent Contribution CCDP importance)'otal (all sequences) 7.7 E4)06 The change in conditional probability (importance) is dctcrmincd by calculating thc conditional probability for the period in which the condition existed and given the condition, and subtmcting the conditional probability for the same period but with plant equipmcnt assumed to be operating nominally.
Thc conditional probability for each cut sct within a scqucnce is dctcrmincd by multiplying thc probability that the portion ofthe sequence that makes the precursor visible (c.g., the system with a failure is demanded) willoccur during the duration ofthe event by thc probabilities of the remaining basic events in the minimal cut set. This can be approximated by 1 - e.
where p is dctcrmincd by multiplying the expected number of initiators that occur during thc duration of the event by the probabilities of the basic events in that minimal cut sct. Thc cxpectcd number of irutiators is given by At, whcrc A is the frcqucncy ofthe initiating cvcnt (given on a pcr hour basis), and t is the duration time ofthe event. This approximation is conservative for precursors made visible by the initiating event.
The frequencies of interest for this event are: ~
5.3 x 10 /h,~
3.8 x 10~/h, Asset, 1.0 x 10 /h, and~
1.6 x 10~/h.
Case 1 represents the increase in the core damage probability due to thc long term unavailability ofthc West CCP (4252 h).
Case 2 rcprcscnts the increase in the core damage probability due to the unavailability ofthc opposite train EDG that was pcriodicaHy unavailable while the West CCP was unavailable (50 h).
Case 3 rcprcscnts the increase in the core damage probability due to thc time that both CCps were simultaneously unavailablc because ofvarious maintenance activities (18 h).
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GUIDANCE FOR LICENSEE REVIEW OF PRELIMINARY ASP ANALYSIS,
Background
The preliminary precursor analysis of an operational event that occurred at your plant has been provided for your review.
This analysis was performed as a part of the NRC's Accident Sequence Precursor (ASP)
Program.
The ASP Program uses probabilistic risk assessment techniques to provide estimates of operating event significance in terms of the potential for core damage.
The types of events evaluated include actual initiating events, such as a loss of off-site power (LOOP) or loss-of-coolant accident (LOCA), degradation of plant conditions, and safety equipment failures or unavailabilities that could increase the probability of core damage from postulated accident sequences.
This preliminary analysis was conducted using the information contained in the plant-specific final safety analysis report (FSAR), individual plant examination (IPE),
and the licensee event report (LER) for this event.
Nodeling Techniques The models used for the analysis of 1995 and 1996 events were developed by the Idaho National Engineering Laboratory (INEL).
The models were developed using the Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) software.
The models are based on linked fault trees.
Four types of initiating events are considered:
(1) transients, (2) loss-of-coolant accidents (LOCAs), (3) losses of offsite power (LOOPs),
and (4) steam generator tube ruptures (PWR only).
Fault trees were developed for each top event on the event trees to a supercomponent level of detail.
The only support system currently modeled is the electric power system.
The models may be modified to include additional detail for the systems/
components of interest for a particular event.
This may include additional equipment or mitigation strategies as outlined in the FSAR or IPE.
Probabilities are modified to reflect the particular circumstances of the event being analyzed.
Guidance for Peer Review Comments regarding the analysis should address:
Does the "Event Description" section accurately describe the event as it occurred?
Does the "Additional Event-Related Information" section provide. accurate additional information concerning the configuration of the plant and the operation of and procedures associated with relevant systems?
Does the "Modeling Assumptions" section accu} ately describe the modeling done for the event?
Is the modeling of the event appropriate for the events that occurred or that had the potential to occur under the event conditions?
This also includes assumptions regarding the likelihood of equipment recovery.
Enclosure 2
1
~ ~
i Appendix H of Reference I provides examples of comments and responses for previous ASP analyses.
Criteria for Evaluating Comments Modifications to the event analysis may be made based on the comments that you provide.
Specific documentation will be required to consider modifications to the event analysis.
References should be made to portions of the LER, AIT, or other event documentation concerning the sequence of events.
System and component capabilities should be supported by references to the
Comments related to operator response times and capabilities should reference plant procedures, the FSAR, the IPE, or applicable operator response models.
Assumptions used in determining failure probabilities should be clearly stated.
Criteria for Evaluating Additional Recovery Measures Additional systems, equipment, or specific recovery actions may be considered for incorporation into the analysis.
However, to assess the viability and effectiveness of the equipment and methods, the appropriate documentation must be included in your response.
This includes:
normal or emergency operating procedures.
piping and instrumentation diagrams (P&IDs),
electrical one-line diagrams, results of thermal-hydraulic
- analyses, and operator training (both procedures and simulator),
etc.
- Systems, equipment, or specific recovery actions that were not in place at the time of the event will not be considered.
Also, the documentation should address the impact (both positive and negative) of the use of the specific recovery measure on:
the sequence of events, the timing of events, the probability of operator error in using the system or equipment, and other systems/processes already modeled in the analysis (including operator actions).
For example, Plant A (a PWR) experiences a reactor trip, and during the subsequent recovery, it is discovered that one train of the auxiliary feedwater (AFW) system is unavailable.
Absent any further information regrading this event, the ASP Program would analyze it as a reactor trip with one train of AFW unavailable.
The AFW modeling would be patterned after information gathered either from the plant FSAR or the IPE.
However, if information is received about the use of an additional system (such as a standby steam generator feedwater system) in recovering from this event, the transient would be modeled as a reactor trip with one train of AFW unavailable, but this Unavailability would be Revision or practices at the time the event occurred.
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mitigated by the use of the standby feedwater system.
The mitigation effect for the standby feedwater system would be credited in the analysis provided that the following material was available:
standby feedwater system characteristics are documented in the FSAR or accounted for in the
- IPE, procedures for using the system during recovery existed at the time of the event, the plant operators had been trained in the use of the system prior to the event, a clear diagram of the system is available (either in the
previous analyses have indicated that there would be sufficient time available to implement the procedure successfully under the circumstances of the event under analysis, the effects of using the standby feedwater system on the operation and recovery of systems or procedures that are already included in the event modeling.
In this case, use of the standby feedwater system may reduce the likelihood of recovering failed AFM equipment or initiating feed-and-bleed due to time and personnel constraints.
Haterials Provided for Review The following materials have been provided in the package to facilitate your review of the preliminary analysis of the operational event.
~
The specific
- LER, augmented inspection team (AIT) report, or other pertinent reports.
~
A summary of the calculation results.
An event tree with the dominant sequence(s) highlighted.
Four tables in the analysis indicate:
(1) a summary of the relevant basic events, including modifications to the probabilities to reflect the circumstances of the event, (2) the dominant core damage sequences, (3) the system names for the systems cited in the dominant core damage sequences, and (4) cut sets for the dominant core damage sequences.
Schedule Please refer to the transmittal letter for schedules and procedures for submitting your comments.
References L. N. Vanden Heuvel et al., Precursors to Potential Severe Core Damage Accidents:
- 1994, A Status
- Report, USNRC Report NUREG/CR-4674 (ORNL/NOAC-232)
Volumes 21 and 22, Martin Harietta Energy Systems, Inc.,
Oak Ridge National Laboratory and Science Applications International Corp.,
. December 1995.