ML17291A946
| ML17291A946 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 08/07/1995 |
| From: | Clifford J NRC (Affiliation Not Assigned) |
| To: | Parrish J WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| References | |
| TAC-M74489, NUDOCS 9508140056 | |
| Download: ML17291A946 (21) | |
Text
August."7, 1995:
Mr. J.
V. Parrish (Mail Drop 1023)
Vice President Nuclear Operations 3000 George Washington Way Washington Public Power Supply System P.O.
Box 968
- Richland, Washington 99352-0968
SUBJECT:
RE(VEST FOR ADDITIONAL INFORMATION RELATED TO INDIVIDUALPLANT EXAMINATION FOR WASHINGTON PUBLIC POWER SUPPLY SYSTEM (WPPSS)
NUCLEAR PROJECT NO.
2 (WNP-2)
Dear Mr. Parrish:
During the staff's continuing review of the Washington Public Power Supply System (WPPSS)
Individual Plant Examination (IPE) submittal and its associated documentation, we have identified the need for additional information to complete our review.
The request for additional information (RAI) is detailed in the enclosure.
To assist the NRC staff in meeting its target review schedule, we request that you respond to the RAI in writing within 60 days of receipt of this letter.
The requirements affect nine or fewer respondents and, therefore, is not subject to the Office of Management and Budget review under P.L.96-511.
If you have any questions, please contact me at 301-415-1352.
Sincerely, ORIGINAL SIGNED BY'ames W. Clifford, Senior Project Manager Project Directorate IV-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation Docket No. 50-397
Enclosure:
RAI cc w/ enclosure:
See next page DOCUMENT NAME: WNP74489.RAI DISTRIBUTION:
Docket File PUBLIC JRoe EAdensam WBateman JClifford EPeyton
TWFN PDIV-2 Reading
- OGC, 015B18
- KPerkins, WCFO
- JDyer, RIV
- HWong, RIV KThomas OFC PDIV-2 NAME
@eeyon DATE 8 Q 95 PDI -2 JCl>fford 8
95 OFFICIAL RECORD COPY
'P50814005b
'P50807W PDR ADQCK 05000397, PDR~
IIIFC RILE CKItIR CSPV
August."7, 1995 Hr. J.
V. Parrish (Hail Drop 1023)
Vice President Nuclear Operations 3000 George Washington Way Washington Public Power Supply System P.O.
Box 968
- Richland, Washington 99352-0968
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION RELATED TO INDIVIDUALPLANT EXAMINATION FOR WASHINGTON PUBLIC POWER SUPPLY SYSTEM (WPPSS)
NUCLEAR PROJECT NO.
2 (WNP-2)
Dear Hr. Parrish:
During the staff's continuing review of the Washington Public Power Supply System (WPPSS)
Individual Plant Examination (IPE) submittal and its associated documentation, we have identified the need for additional information to complete our review.
The request for additional information (RAI) is detailed in the enclosure.
To assist the NRC staff in meeting its target review schedule, we request that you respond to the RAI in writing within 60 days of receipt of this letter.
The requirements affect nine or fewer respondents and, therefore, is not subject to the Office of Management and Budget review under P.L.96-511.
If you have any questions, please contact me at 301-415-1352.
Sincerely, ORIGINAL SIGNED BY:
James W. Clifford, Senior Project Manager Project Directorate IV-2 Division of Reactor Prdjects III/IV Office of Nuclear Reactor Regulation Docket No. 50-397
Enclosure:
RAI cc w/ enclosure:
See next page DOCUMENT NAME: WNP74489.RAI DISTRIBUTION:
Docket File PUBLIC JRoe EAdensam WBateman JClifford EPeyton
TWFN PDIV-2 Reading
- OGC, 015B18
- KPerkins, WCFO
- JDyer, RIV
- HWong, RIV KThomas OFC NAME DATE PDIV-2 ey on 8H 95 PDI -2 JCllfford 8/
95 OFFICIAL RECORD COPY
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&4001 August 7, 1995 Hr. J.
V. Parrish (Hail Drop 1023)
Vice President Nuclear Operations 3000 George Washington Way Washington Public Power Supply System P.O.
Box 968
- Richland, Washington 99352-0968
SUBJECT:
RE(VEST FOR ADDITIONAL INFORMATION RELATED TO INDIVIDUALPLANT EXAHINATION FOR WASHINGTON PUBLIC POWER SUPPLY SYSTEH (WPPSS)
NUCLEAR PROJECT NO.
2 (WNP-2)
(TAC NO. H74489)
Dear Hr. Parrish:
During the staff's continuing review of the Washington Public Power Supply System (WPPSS)
Individual Plant Examination (IPE) submittal and its associated documentation, we have identified the need for additional information to complete our review.
The request for additional information (RAI) is detailed in the enclosure.
To assist the NRC staff in meeting its target review schedule, we request that you respond to the RAI in writing within 60 days of receipt of this letter.
The requirements affect nine or fewer respondents and, therefore, is not subject to the Office of Hanagement and Budget review under P.L.96-511.
If you have any questions, please contact me at 301-415-1352.
Sincerely, Docket No. 50-397
Enclosure:
RAI cc w/encl:
See next page J
mes W. Cli ford, Senior Project Hanager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Hr. J.
V. Parrish cc w/encl:
Hr. J.
H. Swailes WNP-2 Plant General Hanager Washington Public Power Supply System P. 0.
Box 968
- Richland, Washington 99352-0968 Chief Counsel (Hail Drop 396)
Washington Public Power Supply System
>>P.O.
Box 968
- Richland, Washington 99352-0968 Hr. Frederick S. Adair, Chairman Energy Facility Site Evaluation Council P. 0.
Box 43172 Olympia, Washington 98504-3172 Hr. D. A. Swank (Hail Drop PE20)
MNP-2 Licensing Manager Washington Public Power Supply System P.O.
Box 968
- Richland, Mashington 99352-0968 Hr. Paul R.
Bemis (Hail Drop PE20)
Director, Regulatory and Industry Affairs Washington Public Power Supply System P.O.
Box 968
- Richland, Washington 99352 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Harris Tower
& Pavilion 611 Ryan Plaza Drive, Suite 400 Arlington, Texas 76011-8064 Chairman Benton County Board of Commissioners P.O.
Box 69
- Prosser, Washington 99350-0190 Hr. R.
C. Barr, Senior Resident Inspector U.S. Nuclear Regulatory Commission P.O.
Box 69
- Richland, Washington 99352-0968 H. H. Philips, Jr.,
Esq.
Winston 8 Strawn 1400 L Street, N.W.
Mashington, DC 20005-3502
Enclosure J
REQUEST FOR'ADDITIONAL INFORMATION ON THE WASHINGTON NUCLEAR PLANT UNIT 2 INDIVIDUALPLANT EXAMINATION SUBMITTAL 1.
Please address the following topics related to initiating events.
a)
The submittal states that plant specific initiating events were identified using an IDCOR technical report, with two additional initiating events considered.
Discuss the process used to identify the two additional initiating events and discuss how the process ensures that a plant-specific identification of initiating events was performed.
b) Provide the basis for not considering the following events as initiating events:
loss of HVAC for electrical switchgear or for the control room, loss of division I of DC power, loss of all DC power, loss of an AC bus, and loss of reactor building component cooling water.
c) Discuss why loss of standby service water was considered as an initiating event since it does not satisfy the definition of an initiating event used in the IPE (event causes reactor trip).
2.
The impact of loss of containment cooling on the ability to provide core cooling is not clear from the submittal.
The event trees indicate that loss of containment heat removal and failure of containment venting lead to core damage;
- however, the discussions of the event trees indicate that with loss of containment heat removal and failure of containment
- venting, high pressure core cooling with feedwater,
- HPCS, and RCIC remains available until the containment fails, and that if the containment fails in the drywell these core cooling systems remain available after containment failure.
3..
a) Please address this apparent discrepancy between the event tree structure and the descriptions of the event trees.
b) Provide the basis for assuming that HPCS and RCIC remain available until the containment fails by overpressure; specifically address the trip of RCIC on high containment pressure and loss of RCIC and HPCS on high supression pool temperature (e.g., failure of pump seals or pump/motor bearing).
The IPE models a small LOCA as a break of diameter between I and 4
inches.
a) Provide the basis for assuming that RCIC alone can mitigate a water line LOCA of 4 inches in diameter.
b) Provide the basis for assuming that adequate steam pressure is available during the entire mission time for RCIC to mitigate a
4 inch steam line LOCA.
c) Dimus~hy RCIC was not credited for mitigating a stuck open SRV given that RCIC was credited for mitigating a break in either a steam or a water line up to 4 inches in diameter.
The IPE credits natural circulation for cooling of the RCIC pump room during station blackout, if operator action is taken to open doors to the RCIC pump room.
Please address a) Indicate weather there are procedures directing operators to open the door to the RCIC pump.room, and if there are procedures directing them to use fans to ensure pump room cooling; if not, please provide basis for credit of opening doors and use of fans.
Please
- also, discuss how fans can be used during station blackout events, and whether or not these fans are needed to assure adequate pump room cooling.
- Also, discuss how much time is available to open the doors before RCIC overheats, and the basis used to determine this time.
b)
Discuss the indications available to operators that RCIC pump room overheating may be occurring.
Please address the following topics associated with the quantification of component failures.
a)
The NUREG-1335 guidelines requested that the rationale be provided if plant-specific experience is not used for. important components such as diesel generators.
Discuss why diesel generator failures were not quantified with plant-specific data.
Also, summarize the process used to select components for quantification with plant-specific data.
b) The generic data in the submittal are failure frequencies; without times between testing and system mission times, we cannot review the actual failure probabilities used.
Please provide the actual, event failure probabilities used in the analysis.
c) The plant-specific data in the submittal are failure frequencies and no distinction between standby and run failures is provided.
Please provide the actual event failure probabilities used in the analysis.,
d) The submittal lists beta factors for the following pumps: safety injection,
- RHR, and containment spray.
Clarify which pumps are the "safety injection pumps."
Clarify which pumps are the "containment spray pumps" and discuss whether they are not the same pumps as the RHR pumps.
If the containment spray and RHR pumps are the
- same, provide the rationale for the use of different beta factors for these items of equipment.
e) Discuss the basis for using only an 18 month period for the collection of plant-specific data used in your IPE.
It is not clear in the submittal if plant changes due to the Station Blackout rule were credited in the analysis.
Please
a) Identify whether plant changes (e.g.,
procedures for load shedding, alternate AC power) made in response to the blackout rule were credited in the IPE and what are the specific plant changes that were credited.
b) If available, identify the total impact of these plant changes to the total plant core damage frequency and to the station blackout CDF (i.e.,
reduction in total plant CDF and station blackout CDF).
c) If available, identify the impact of each individual plant change to the total plant core damage frequency and to the station blackout CDF (i.e., reduction in total plant CDF and station blackout CDF).
d) Identify any other changes to the plant that have been implemented or planned to be implemented that are separate from those in response to the station blackout rule, that reduce the station blackout CDF.
e) Identify whether the changes in ¹4 are implemented or planned.
f) Identify whether credit was taken for the changes in ¹4 in the IPE.
g) If available, identify the impact of the changes in ¹4 to the station blackout CDF.
The status of the potential plant improvements to reduce the likelihood of core damage and/or improve containment performance discussed in the submittal is not clear.
Please clarify the submittal information by
'roviding the following:
a) The specific improvements that have been implemented, are being
- planned, or are under evaluation.
b) The status of each improvement, i.e., whether the improvement has actually been implemented, is planned (with scheduled implementation date),
or is under evaluation.
c) The improvements that were credited (if any) in the reported CDF.
d) If available, the reduction to the CDF or the conditional containment failure probability that would be realized from each plant improvement if the improvement was to be credited in the reported CDF (or containment failure probability), or the increase in the CDF or the conditional containment failure probability if the credited improvement was to be removed from the reported CDF (or containment failure probability).
e) The basis for each improvement, i.e., whether it addressed a
vulnerability, was otherwise identified from the IPE review, was developed as part of other NRC rulemaking, such as, the Station Blackout Rule, etc.
The submittal states that use of the fire water system was not credited for core cooling due to the time required to implement injection with
this system.
However, the event tree for station blackout identifies the fire water system as available for core cooling.
a) Please clarify the treatment of injection with fire water as a means of core cooling in the IPE.
b) Discuss whether specific procedures are available to operators to perform the actions necessary to use the fire water system for core cooling; whether these procedures have ever been practiced; and whether all equipment needed for use of this system (e.g.,
spool pieces, special hose connections) are currently available at the plant.
b) The tables of system interdependencies indicate that fire water requires AC power for injection to the vessel.
Please discuss this dependency.
The table of success criteria credits one LPCI pump for mitigation of a large LOCA, but a footnote to the table states that "in the long term
(>2 hours) a different combination may be required."
Explain this footnote and how it was factored into the success criteria for mitigating a large LOCA.
Please clarify the credit taken, if any, for DC load shedding during station blackout to prolong battery life.
The IPE credits containment venting with the soft vent lineup.
Please provide a schematic and list of system dependencies for the containment venting system.
Please provide the following information related to the modeling of an interfacing systems LOCA in the IPE.
a) Discuss how the model evaluated the possibility of piping exposed to greater than design pressure not failing.
b) Discuss how the model addressed long term makeup to the vessel given success'ful depressurization.
Please address the following system dependencies as given in the tables of intersystem dependencies.
Please a) Explain why onsite AC power is not totally dependent on DC power (for field flashing and closing of circuit breakers).
b) Explain why AC and DC power have a partial/delayed dependence on room cooling instead of a complete dependence, and discuss whether or not any compensatory actions for cooling electrical components were credited in the IPE.
c) Discuss how control room HVAC was considered in the IPE during the mitigative phase of accident sequences.
14.
WNP has.mperienced reactor power oscillations during its operational history.
Please provide the basis for not including power oscillation events in your IPE.
Specifically take into account the power oscillation event that has occurred at WNP-2 and discuss the potential for more severe. events..-'..
15.
The IPE has used guidance from NUMARC 91-04 to identify vulnerabilities.
The submittal states that'o vulnerabilities were found.
- However, NUMARC 91-04 indicates.that events representing greater than 50X of the CDF should be further;assess'ed.
Please explain how LOSP-initiated events were not considered vulnerabilities since as group they contribute 67X 'of.'the CDF:.;.-,
16.
The submittal's discus'sion on..the resolution of USI A-45 is based on a
"narrow" DHR definition;.th'at being loss of containment heat removal.
The IPE model,. however,.'addressed DHR for both its core and containment heat removal functions.'.:.: Ii order to resolve USI A-45, the Generic Letter 88-20 '(and the 'guidance provided in NUREG-1335) requested the licensees to examine DHR for its capability during both core cooling and containment heat removal phases and for all accidents except large
Please expand the submittal.'s discussion on. the resolution of USI A-45 (i.e.,
DHR redundancy, diversity, etc.)..to include the DHR core cooling capability at the WNP-2 plant 17.
The submittal does,'not clearly,, discuss the process that was used to identify 'and select're'-initiator human events involving miscalibration of instrumentation,'nd.'those associated with failure to restore properly to,se'rvice after-test'and maintenance.
The process used to identify and 'select these 'types'f human events may include the review of procedures, and di'scussions w'ith.'appropriate plant personnel on interpretation and implementation of the plant's calibration, test,'nd maintenance procedures.
a)
Please provide a description of the process that was used to identify these kinds of human events.
Please provide examples illustrating this process.
b) It is not clear why certain events listed in Table 3.3.3-1, Pre-Accident Human Errors; were included in the scope of the analysis of failures that can lead to core damage conditions.
These include the following events: '.'......
CIAHUMNBUTAK3XX,'EOfails,to replace specific bottle when required;
- CIAHUMNTK20AX3XX, EO fails to replace bottle (CIA-TK-20A) when required; CH-HUMNTK20AX3XX, Personnel fail to reorder liquid nitrogen when required; RCIHUMN-PIX3LL, Human error during pump oil change.
II It
Please=review these similarly designated pre-initiator human events contai'ne'd in Table 3.3.3-1 and briefly describe how these events can contribute to the frequency of core damage.
18.
It is not clear from the submittal how error probabilities values were assigned for pre-initiator human events.
It is understood from the WNP-2 submittal that the method was based on the ASEP HRA approach described in NUREG/CHEWIER-4772.
In that method, a basic human error probability (HEP) of 0.03 is assigned for each critical action, followed by application of one or more recovery factors if appropriate.
These recovery factors can reduce the basic HEP by a factor of 10 for each applicable factor.
These factors include:
the performance of a functional test following the action that would reveal any errors in the task execution, or separate checking by another individual.
By way of illustration, please explain how these recovery factors were
- assigned, and the overall error probability calculated, for the following events listed in Table 3.3.3-1, Pre-Accident Human Errors, of the WNP-2 submittal:
CIAHUMNBUTAK3XX, EO fails to replace specific bottle when required; CIAHUMNV-20J3XX, TM error on CIA-V-20; CN-HUMNTK-IX3XX, Personnel fail to reorder liquid nitrogen when required; HPSHUMNPVOPEX3LL, Operator error, pump
& valve operability test; and RCIHUMN-PIX3LL, Human error during pump oil change.
19.
The submittal provides no description of the plant-specific experience of pre-initiator human events.
Please provide a comparison of any operating experience of pre-initiator human events (e.g.,
data from LERs or other plant records) with the failure probabilities calculated in the HRA modeling.
In other words, please describe to what degree do the results of the modeling of pre-initiator human events represent the WNP-2 plant experience.
Include in your discussion a description of your experience associated with the effectiveness of checking actions and functional testing discussed in guestion 18.
20.
Failure to identify and evaluate different types of dependencies between pre-accident human events that could potentially exist can result in failure to recognize vulnerabilities associated with the design, operation, maintenance or surveillance of the plant.
In addressing dependencies, whether miscalibration or failure to restore, the process utilized should consider plant conditions, human engineering, performance by same crew at same time, adequacy of training, adequacy of procedures, and interviews with training, operations and various crews.
Please illustrate this discussion with at least two examples where dependencies were explicitly analyzed.
21.
The WNP-2 submittal does not clearly describe the method used to identify and select post-initiator human events for analysis.
The method utilized should confirm that the plant emergency procedures, design, training programs, and operational practices and policies were
Il
examined-and understood to identify potential severe accident sequences.
Please provide a description of the process that was used for identifying and selecting the post-initiator human events evaluated.
22.
The submittal does not clearly indicate whether a screening process was utilized to help differentiate the more important post-initiator human events.
If a screening process was used, please provide the screening value(s) used and the basis for the value(s); that is, provide the rationale for how the selected screening value did not eliminate (or truncate) important human events.
Also, provide the list of errors initially considered and those screened.
23.
It is not clear 'from the submittal if plant-specific performance shaping factors were used in the assessment of post-initiator human events.
This plant-specific information could include the size of crew, availability of procedures, quality of the human-system interface, etc.
Please describe the types of plant-specific performance shaping factors considered and their values, and discuss by way of examples how these performance shaping factors were used in the assessment of post-initiator human events.
If plant-specific performance shaping factors were not considered in the
- analysis, please explain why the analysis can be considered to be representative of the actual WNP-2 plant design and operating practices.
24.
The time available for operators to perform actions is often considered in HRA analyses to be an important influence in assessing the probabilities of'post-initiator human events and the'submittal describes that "time available" was used as one input parameter in the WNP-2
'tudy.
However, the'submittal does not discuss the basis for estimating these timescales and the procedures used to ensure that the timescales were truly representative of the scenarios for which human event probabilities were being calculated.
Please illustrate this process with the following examples of post-initiator human events taken from Table 3.3.3-5.
RHRHUMNSP-COOLL SLCHUMN20MINUTES VENTFAIL.
25.
Section 3.3.3.1 of'the WNP-2 submittal states that "Basic events representing human interactions are defined in system fault trees, but not in event trees."
While representing post-initiator human events in the system fault trees can be appropriate, that form of representation requires two important considerations.
These are:
(I) incorporating changes in the sequence quantification process for the quantification of such human events for different accident sequences involving such factors as different timescales available for performing the actions, and (2) ensuring that dependencies between multiple human events in accident sequences are modeled and calculated correctly.
The submittal does not discuss how these considerations were accounted for in the WNP-2 study.
Please
- describe, using the analysis of human actions in
26.
27.
28.
29.
ATWS sceiamo as examples, how each of these considerations were incorporated in the WNP-2 analyses.
In the analysis of post-initiator human events in Table 3.3.3-5, event RCIHUHNRCOLH3LL is identified as a human error in following procedure (PPN 5.6. I) to provide alternate room cooling to the RCIC room.
Table 3.4.2-1 identifies this event as opening the RCIC pump room door and starting cooling fans, whereas Section 3.4. 1.3 describes this action as only opening the room door within 30 minutes, and the definition of event UZ on page 3. 1-45 states that this action involves allowing only natural-circulation cooling of the room.
Please clarify the specific scope of this event, particularly as the action as it is performed under station blackout (SBO) conditions.
(Also see guestion 4.)
In the discussion of internal flooding (Section 3. 1.2.4.7 of WNP-2 submittal),
the role of operators in responding to flooding events is discussed.
For example, in the discussion of Category FLD14 floods, operator actions within a timescale of 8-9 minutes are discussed; in the sequence involving a TSW pipe break in area R504, operator actions are described as "important."
Other categories of internal flooding also describe operator responses.
However, the discussion of human reliability modeling in Section 3.3.3 does not discuss the modeling of such actions, and they are not listed in the tables of human actions associated with that section.
Please provide a description of the procedure used for modeling these actions, and provide two examples of '
the human error probability calculations to illustrate this procedure.
The submittal does not clearly describe the method used to identify and select recovery type actions for analysis.
The method utilized should confirm that the plant emergency procedures, design,. and operational practices and policies were examined and understood to identify potential severe accident sequences.
Please provide a description of the process that was used for identifying and selecting the recovery type actions evaluated.
Section 4.4 of the WNP-2 submittal describes the quantification of potential recovery actions, particularly those important for the analysis of Level 2 issues.
While it is recognized that the use of a failure probability of O.l for human actions associated with the recovery of operable equipment is often conservative, such a value may not be considered conservative in those sequences where human errors have already caused the sequence.
Please describe the process by which the analysis of recovery actions took account of prior post-initiator human events in the sequences, and provide illustrative examples of that process.
Please include in that description the actions to reclose or otherwise isolate containment penetration failures described on page 4.5-29 of the WNP-2 submittal, taking account of any environmental limits (e.g.
radiation or heat limits in access).
1
)l
30.
Section-$.-3-, Review Comments and Resolution, of the WNP-2 Submittal identi'fi'es an intention to apply limits to the number of human actions that can be applied in accident sequences to prevent truncation of accident sequences.
This is a concern that reflects the limited analysis of dependencies between human events in accident sequences (see guestion 25 above).
The submittal identifies this concern as an area to be evaluated as "part of the configuration control of the IPE."
Please describe any work performed since the submittal in response to this commitment by WNP-2.
31.
Section 6.2, IPE Recommended Improvements, includes a discussion of the use of the ADS inhibit switch.
The discussion related to the use of this switch in non-ATWS scenarios is not understood.
Please expand your discussion of this issue, describing how such use will result in a lower core damage frequency.
Include in your discussion why use of this switch in non-ATWS scenarios will not result in an increase in core damage frequency by creating a new accident scenario involving failure to initiate ADS when needed, following earlier inhibiting of ADS by operators.
32.
In Generic Letter 88-20, Supplement No. 3, NRC requests that licensees consider insights from the CPI Program, specifically for Hark II containments:
additional containment heat removal capability (e.g.,
hardened vent or improved reliability of suppression pool cooling).
Also, Supplement No.
3 requests Hark II licensee consideration of insights directed at the Hark I BWRs and delineated in Generic Letter 88-20, Supplement No. I:
(a) alternate water supply for drywell spray/vessel injection; (b) enhanced RPV depressurization reliability; and (c) improved EOP and training.
Throughout the IPE submittal, comments are made about these issues.
For example, it is stated that (1) improvement in the reliability of the RHR for containment heat removal has a significant beneficial impact on the Level 2 results (page 6.0-5);
(2) the advantages and disadvantages of drywell sprays and the importance of procedures to optimize benefits (page 6.0-5);
(3)
SBO depressurization and related coping time that can reduce the CDF by up to 34X (page 6.0-4);
and (4) the ineffectiveness of a hardened vent in reducing offsite consequences (page 4.9-2).
Large releases dominate the results of the IPE.
(About 55X of the CDF is associated with large releases (about lE-5/yr).)
CPIs should address reduction of these large releases.
Please provide a more cohesive and comprehensive picture of the Supply System's approach to CPI including the prioritization of any action items, the schedule for actions, and the risk-significance of actions taken.
33.
In Section 4.7 the CET quantification is discussed.
In previous sections the CETs are described along with the associated node-specific fault trees and the plant models and methods for physical processes.
The topic areas are well explained,
- however, the link between the mechanistic analysis using the HAAP code and the actual quantification
V of the..GETS is, not clear.
By selecting MAAP runs for several of the more important accident sequences, please a)
Show how the MAAP pressure and temperature or MCCI values were used to quantify the.CETs..
b) Provide temperature and pressure histories for both the drywell and the wetwell.
c) Explain'hat was 'the process of determining whether a
temperature/pressure combination was sufficient for containment failure, taking into account'he: uncertainties associated with containment failure and the MAAP analyse's themselves.
~
~
34.
On page 4.3-8'of the APE submittal, it is stated that the containment failure pressures provided are conservative because a plastic strain failure critepioa of I-2X.was used where, in reality, the maximum principal strairt at'upture would be more like 16X. If a "16X" criterion were 'used, would there be any changes in the containment failure characterization,=
other than a higher containment failure
- pressure, e.g.; failure locations and/or displacement versus force-type loading of the biological,shieldf State the reasons that there would (or would not) be any significant changes.
35.
36.
On page 4.4-2, under "Bins and Plant Damage States,"
recovery actions are discussed and,it is stated that the possibility of collateral plant damage'when the'.s'upport systems become available) could make hardware less reliable than'xpected.
Because of this, conservative values for failure to function for some key systems is used.
Please give specific examples of "col.lateral plant damage" and explain how such damage would affect the accident progression
'and'radionuclide releases.
It is not clear how."success" for the "Short term station blackout" for the Top Event "High pressure injection before VF" was determined.
On page 4.5-3, it's stated that,. if the high-pressure injection systems are recovered prior to vessel failure, the sequence can be terminated if the in-vessel debris is in a eoolable configuration.
Please state the success criterion used for eoolable configuration and its basis.
Further, it is stated on page 4.5-6 that:
e
. successful restoration of injection during the additional time which 'ma'y be available is unlikely to have much effect.
Even if injection's restored there is still a significant probability that the core cannot be. cooled so injection has no effect on vessel protection.,
The node of the event tree (Power recovered prior to VF) is resolved simply with a split fraction which segregates the sequences by the relative likelihood that power is or is not restored in time to prevent vessel failure."
The above suggests that not much credit should be given for successful injection prior to vessel failure, which would then prevent vessel 10
V
"4l failurg..
However, the CET shows in Figure 4.7-1 a power recovery probabil.ity of 55X with a subsequent high pressure injection successful probability of 80X, thus yielding an intact vessel conditional probability of 44X for this station blackout PDS.
Please discuss the apparent inconsistency between the quoted discussion and the CET quantification.
37.
On page 4.9-3, it is stated that ".
as the fragility of the
[containment] shell increases
. the fraction of very early containment failures
. decreased from 8.7X to 2.2X,
[emphasis added].
Should the word increases actually be "decreases"7
- Further, the use of the descriptor "very early" indicates that the containment fails prior to core damage (see page 4.6-6).
Would sole use of the word "early" be more accurate2 38.
On page 4.9-4, it is concluded from the sensitivity study on debris coolability that the current assumption on coolability is conservative.
It is not clear why you draw that conclusion.
When the split fraction for cooled/not cooled was changed from 43/57 to 90/10 it was found that the conditional probability of an intact containment remains unchanged and that there was a small increase in scrubbed fission products (certainly expected).
These results indicate that the 43/57 value (and even the 90/10 value) may not be conservative.
Please explain the basis for your conclusion.
39.
The treatment of direct bypass of the containment is not clearly stated.
On page 4.6-2, containment bypass is discussed in the context of source-term grouping.
It is stated that "Even though the Level-1 study contains no containment bypass sequences above the cutoff frequency, for the sake of completeness it is considered in the grouping process."
The groups STG-15 and STG-16 are assigned to containment
- bypass, and no contribution to releases from these groups is shown (note Tables 4.8-2 and 4.8-3).
- However, although it is not labeled as
- such, the "Large LOCA Outside Containment With Failure To Isolate (AO-S14, see pages 3.4-6 and 3.4-8)," with a 0.9X contribution to TCDF, appears to be a
containment bypass event.
Therefore, it appears that containment bypass contributes about 0.9X to the core damage frequency.
Please explain this apparent discrepancy.
What are the STG-15 and/or -16 frequencies resulting from the bypass (AO Sum) events'lso, on page 1.0-7, it is stated that a sensitivity analysis was performed for containment bypass but it does not appear in the Section 4.9, "Sensitivity of Back-end Results."
Please provide the results of this sensitivity analysis.
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