ML20115D210
| ML20115D210 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 10/13/1992 |
| From: | Parker T NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR GL-88-20, NUDOCS 9210200288 | |
| Download: ML20115D210 (18) | |
Text
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i Northem States Power Company 414 Nicollet Mall Minneapolis, Minnesota 55401 1927 Telephone (612) 330 5500 October 13, 1992 Generic Letter 88-20 Supplement No. 4 9
U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING P1H1T Docket Nos. 50 282 Licenses No. Dt'R 42 50-306 DPR-60 Response to Schedular concerns associated with neneric Letter 88 20, Supplement No. 4 Individual Plant Examination of External Events In a letter dated July 29, 1992, you informed us that our schedule for the submittal of our response to Generic Letter 88 20, Supplement No. 4 was unacceptable. We have discussed this schedule with the hRC Staff on several occasions, including a meeting with the NRC Staff on September 29,1992.
At this meeting, we were unable to agree on an acceptable submittal date.
This letter documents our response to your request for our schedule to be changed Please contact us if there is a problem with this schedule for the Individual Plant Examination of External Events for Prairie Island.
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1m omas arker Manager t
Nuclear Support Services c:
Regional Administrator Region III, NRC i
Monticello Senior Resident Inspector, NRC o
Prairie Island Senior Resident Inspector, NRC NRR Project Manager, NRC J E Silberg John Butler, NUMARC Attachment 9210200288 921013
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i We have reviewed the IPEEE schedule and have not been able to modify the schedule to conform with the requested June 1995 date proposed in your July 29, 1992 letter.
The reasons for this are discussed below.
j The NSP PRA philosophy is to develop living PRAs which will assist the corporation in making sound decisions on plant modifications, meet the l
requirements of the NRC and develop in-house personnel with the knowledge-to use the PRAs.
We plan to maintain living PRAs for both the NSP nuclear plants.
1 In the process of doing this we will understand the important factors affecting 1
risk, allowing us to reduce risk as much as possible.
t We have 7 full time employees dedicated full' time to PRA work. Two of these positions exist at Monticello, eiree in our general office and two at the-I Prairie Island site. This is a significant commitment in a atmosphere of financial restrain with every additional staff person being reviewed very carefully.
In addition, substantfal. financial resources are required for i
consultant work.
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The Montic611o PRA has been submitted and the Prairie Island PRA will be submitted in March of 1994. We have found the PRA tool to be very useful.
The a
PRA staff has many requests to evaluate different modification configurations as well as evaluate outagn scenarios. We believe our PRA personnel are performing a uignificant safety function in evaluating modification and j
scheduling.
j The specific objectives of the NSP IPEEE program are'as follows:
1 Develop appreciation of severe accident behavior.
l Understand most likely accida..t sequences.
1 Cain a quantitative nd qualitative understanding of risk.
I Reduce risk where a propriate.
s Maximize use of NSP staff.
l Insure the IPEEE reflects the current plant.
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Support licensing and plant modification process, j
Coordinate with SQUG.
l To meet these objectives NSP plans to perform a seismic PRA, fire PRA, and screening for the other external events.
The overall schedule including both the IPE and IPEEE for both Monticello and Prairie Island is shorn-on the attached figure..The completion of the'IPE has-the.hignest priority-followed by the IPEEE with'aupporting work for the plant having lower priority.
We have reviewed the schedule for the Prairie Island IPEEE very carefully and I
will be able to revise the scheduled due date from March 1, 1997 to December j
15, 1996. We realize this f ar from the June' 1995 dato requested..The-fact that the Prairie Island PRA is not scheduled to be completed until March of i-1994 makes the Jane 1995 unattainable using.NSP personnel. Attachment 2 contains a more detailed schedule for the Prairie Island IPEEE.
The details of I
the schedule may be revised when the project is started. The reasons as to why the scheduled date is beyond the requested date of June 1995
- e es follows:
1.
In the description of the examination. process for che IPEEE, ~the NRC strongly requests each licensee to uce its staff to the-maximum extent practical in conducting the IPEEE.
NSP has schedulad inter.ial resources in a way to ensure maximum technology transfer to NSP personnel which also reduces the consulting costa, thereby meeting the IPE goal of using.
the util.;y engineers to the maximum extent possible.
The Prairie Island IPE is scheduled for completion by March 1, 1994.
1 Completion of the 'PEEE by June of 1995 is not pessible.without having i-
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US NRC october 13, 1992 Page 3 J
.most all the work done by consultants at a very large cost to NSP which 4
would not meet the intent of the IPEEE process to have utilities involved in the analysis.
The IPE and IPEEE have been scheduled in this mann c because NSP has two l
sites with one site having two units.
This requir significant additional resources to complete the IPE and IPEEE than that required for i
a single site or a single unit. To manage the addicional burden, the 1
work is first being completed at Monticello followed by Prairie Island.
Resources are being shared between the sites to assure consistency and to i
avoid duplicatien.
Thic will more effectively use the personnel and reduce cost, but Soes somewhat extend the schedule.
I 2.
The NRC has stated that the IPEEE must reflect the current plant design.
With changes and modifications being performed routinely to the plant,.
the PRA must be updated to ensure the IPEEE reflects the currant plant design.
Included in-the schedule is updating the PRA before the j
quantification of the IPEEE. This is needed because the IPE and IPEEE j
i procesa naturally spans several Saars.
The update will include a review 1
of initiating events, modifications to the plant, changes to procedures, changes in maintenance and other inputs to the PRA to determine if there to a significant change that could effect the conclusions of-the IPEEE results. Only those significant changes will be updated.
3.
In Gtneric Letter 88-20, the NRC suggests the benefits of a PRA include i
8upporting licensing and the plant modification process. Modifications and other changes to Prairie Island are being reviewed using the PRA i
methods. This includes a review of the new diesel generators being installed in the fall of 1992. The review of plant changes is an ongoing effort using the same personnel that are developing the IPE and IPEEE.
In addition, work has been done reviewing the outages at Prairie Island using the PRA models. This is a qualitative look at the Prairie Island outages to confirm the outage plan is reasonable.
The same personnel who perform the IPE and IPEEE work on this review.
In the September 29, 1992 meeting between NSP and NRC Staff personnel, several' ideas for schedular improvement were discussed. We have evaluated each:
)
1)
Do not update the IPEEE to current plant conditions.
i To ensure the IPEEE conclusions are correct the PRA must be updated to include significant changes which may effect the results of the i
analysis. To minimize the effort only significant changes that may effect the conclusions of the IPEEE will be included in.the update.
i It is anticipated that there will be changes that will require a j
requantification of the PRA because of the length of time since the PRA was last qu.ntified.
f 2) combine the walkdowns with USI A-46 pcogram work.
This is already included in the schedule. The IPEEE will be coordinated with the USI A-46 work.
3)
Schedule the consultant work earlier.
The only two taske done by the consultants which can be completed earlier are the response and fragility analyses. Those two tasks are not critical path and to schedule them earlier would not improve the schedule.
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October 13, 1992 i
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Have more work done by consultants.
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To have the consultants perform more of the effort would put a j
large budgetary burden on NSP and would result in noh being able to meet the intc;? of the severe accident issues of having the utilities performing the work.
It is very important to have the j
utility staff understand the insights and methods used.
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5)
Remove " slop" time from the schedule.
l The " slop" in the schedule was removed by moving the date from March 1, 1997 to December 45, 1996.
To move the schedule more would increase the probability of a request for extension during j
the project.
1 We believe the schedule for performing the Monticello and Prairie Island IPE j
and IPEEE is appropriate-to meet,the intent of the Generic Letter.
By applying internal resources for both conducting and using-the PRAs, NSP is being fully consistent with the intent of the safety issues including the IPEEE.
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'US WRC October 13, 1992 i
Page 5-i NSP IPE and IPEEE Schedule 1987 1986 1989 1990 1991 1992 1993 1994 1995 1996 1997
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12/87 12/88
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Mnt IPEM 3
4 9/89 3/92 Mnt IPE 4
4 4 4 4 4
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i 4/92 3/95 Mnt IPEEE 2 2 2
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-1 10/88 5/91 PI IPEM 1
2 2 2 2 8/90 3/94 l
1 3
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5 5 3 1/94 3/97 i-PI IPEEE 1
4 4 4
4 4 4 Shutdown Review 3
3 i
Update PRA 4
value under each project is the number of people working on the project.
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US NRC 4
October 13, 1992 Page 6 4
Prairie Island IPEEE Schedule The main objective is to perform an IPEEE for Prairie Island.
This includes documentation consistent with the current NRC guidelines for submittal.
In addition to addressing the Severe Accident Policy requirements, the models will be developed and documented in a manner that will support future use and application to regulatory, safety, and-operating issues.
Another objective is that of technology transfer. The project will be conducted primarily by NSP staff under the guidance of consultants who are experienced in PRA and external events technology. The goal of this objective is to train the NSP staff to conduct similar studies with little or no consultant guidance and thereby provide NSP with the capability to update and utilize their analysis for related internal decisions with minimum outside assistance. The only exception is the fragility analyses, because of the detailed expertise required and there is not a need to reproduce the analyses in the future.
The schedule is made up of five major parts: 1) Project Management, 2) Seismic,
- 3) Fire, 4) Other, 5) Documentation, and 6) Update PRA. 'The attached figure provides the schedule for the empletion of the Prairie Island IPEEE work.
1.0 Project Management This task depicts the functions associated with effective project management.
The key tasks include: Develop a detailed project plan that defines the objectives, scope, schedule, and cost for the tasks required.
Ensure the
- objectives of the project are being met as work progresses. Provide the day-to-day direction of the technical work on the project and ensure that the tasks are i
completed in a timely manner. Provide the day-to-day managemene of the resources allocated to accomplish the project. Ensure that the containment analysis and 4
i the accident sequence analyses are fully integrated and consistent. Ensure that the work being done by the analysis team is reviewed at approprie'- intervals in order to monitor the quality of the work being performed.
2.0 seismic
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Seismic is the largest part of the IPEEE and will be coordinated with the SQUG A-46 work, Thn, scope of work will include a seismic PRA analysis which has three major parts to its 1) Plant Information, 2) Quantification, and 3) Documentation.
2.1 Plant Information The purpose of this task is to gather and compile the information necessary to perform the seismic evaluation. completion of this task is comprised of i
four subtasket 2.1.a Corsponent Inforeatic This task includes gathet g.tructural information,-system information, and relevant analyses perforno previously for seismic use to rupport the seismic IPEEE. Existing documentation on building construction, equipment support drawings and suspended system (piping, cable tray, conduit, and HVAC) configurations, and supports will be gathered together and reviewed.
This will include types of materials, anchor types, foundations, supports, etc. which are available in construction and design basis documentation.
In addition, relevant sections pertaining to seismic issues from the FSAR need to be ec' lected. -
Aaalyses from the FSAR, Level 1 PRA work on
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US NRC
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October 13, 1992 Page 7 containment capability and flooding, Appendix R work on component locations and barriers, and A-46 tasks will b-reviewed here as to coroponent and structural response.
Equipment seismic qualificati-reports will be reviewed for components on the seismic PRA list.
2.1.2 Review Hazard Analysee seismic hazards performed by both the EPRI (EPRI NP-6395D) and LLNL (NUREG/CR-5250) will be reviewed to understand their basis and to determine hoo they are applicable to the Prairie Island site.
2.1.3 Walkdown This task will Snvolve planning the walkdown and documenting the findings for use in the IPEEE. Much of this work will be coordinated with the SQUG work.
The initial walkdown will be based upon recommendations and inputs from the EPRI seismic margins approach (EPRI NP-6041),
the Generic Implementation Procedura (GIP) developed for A-46 by the Seismic Qualification Utility Group (SQUG) and EPRI, as well as the IPEEE Generic Letter.
The walkdowns will be performed to satisfy requirements of the IPEEE and tha A-46 issues.
A-46 issues and IPEEE seismic issues overlap in some areas, but are different in scope in other areas.
A-46 is concerned only with one success path of equipment needed to bring the plant to safe shutdown, whereas the PRA approach for IPEEE is concerned with multiple success paths.
A-46 and the GIP have more specific guidance on the walkdown, the GIP calls for the use of a wrench test for the bolt tightness check of uxpansion anchor bolts, whereas no specific testing is called for by the IPEEE.
The GIP methods can be used for overlapping equipment to satisfy both the IPEEE and A-46 concerns, and sufficient information can be taken on the GIP forms to ensure fragilities.can be generated for the PRA.
The CIP must be used for equipment that is only within the scope of A-46, while the fragility methods defined in the IPEEE generic letter must be used for equipment only within the scope of the IPEEE.
Preparatory work for the walkdown includes:: 1) Preparing the Team.
The team should include plant operations pe.connel,-systems engineers, PRA analysts, and seismic capability engineers. 2: Engineering work before the walkdown which: includes: Familiarization with plant der:gn features'by reviewing P& ids, FSAR, electrical one-line diagrams, operating procedures and recovery actions, arrangement drawings, and dependency matrices.
Identifying important plant functions and systems which satisfy those design features.
Identifying potential " success paths" of systems and actions which result in plant shutdown.
And, identifying relays and contractors prone to chatter, this task is difficult and requires iterations between systems and seismic capability engineers.
- 3) Review the seismic capability which includes: Review seismic sections of FSAR, sample equi.pment qualification reports, selected equipment specifications, structural
- drawing, seismic
- analyses, available floor
- spectra, representative equipment seismic anchorage analyses and designs,-seismic qualification review team forms if avaul able, and review derwentation available on qualification of relays and contractors against re ihatter Training will be required for the PRA group for either the SQUG or IPEEE effort.
This will include industry training by SQUG and other training given by EQE or TENERA for the valkdown.
Confirmatory walkdowns will be conducted and documented as necessary following the preliminary assessment to verify the results of te sequence quantification and to support the recovery analysis.
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US NRC October 13, 1992 Page 8 2.1.4 System Analysis j
Included in this task is the identification of components and equipment
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that may be needed in response to seismic events.
IPEEE and A-46 issues will be addressed here.
i Equipment required for safe shutdown must be identified. Much of the work done in the previous tasks will La used to develop lists of important j
equipment and structures.
This task needs to be completed before the i
walkdown. Dif ferent lists are necessary for the IPEEE and the A-46 issues i
because of different requirements and scopes.
In performing the seismic evaluation for the IPEEE it is necessary to assess the impacts of the seismic event on not only active and passive components and systems, but upon the structures as well.
Seismic system 4
i interactions should receive particular attention because o2 the potential
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for common cause f ailure mechenisms attributable to the seismic event.
Active components are those which must change states in order to satisfy their associated safety function. Passive components need to continue to operate in the state that they are in.
Structurec of importance include masonry block walls, tanks, and other items whose-integrity and position are vital to the continued operation of compononts and systems. Secondary interactions include failure or collapse of local equipment such as ceiling tiles, lighting fixtures, etc. onto systems being credited in the analysis.
These interactions also include environmental concerne like temperature, pressure, and humidity..These will be identified through a combination of system familiarity, IPE results, and plant walkdown information.
1 The A-46 issue is concerned with the seismic qualification of equipment needed to bring tne plant to safe shutdown. Seismically induced LOCAs and j
high energy line breaks need not be considered, and therefore piping and tubing require examination only if they have the potential to cause seismic interaction with other necessary 9quipment, above-ground tanks sre.
an exception.
Therefore, this list of equipment should be smaller than and a subset of the-list developed for the IPEEE. Licensing concerns may dictate that an expanded A46 list be considered.
Part of this task will j
include FSAR changes, broader A46 listed, replacement pa es, etc.
m Failure Modes and Effects Analyses (FMEAs) will be performed to bcL.
3 assess and screen the impact of failures associated,with structural and 4
l equipment interactions. The FMEAs will be used to modify existing fault trees to reflect the impact of seismic events on plant systems.
The equipment will include that identified by the - IPE as important to safety.
Once again, passive and active equipment, structures, and secondary interactions will be considered.
Other components such as i
relays, contractorc and switches that are prone to chatter will be added to the list as well if judged or shown to be important. However,-careful planning and use of generic insights are necessary at this. stage to avoid large resource expenditures.
An assessment of relay chatter is required by the Generic Letter for the IPEEE and is also required for A-46.
This task will be coordinated with the A-46 effort.
1 FMEA will be performed to both assess'and screen the impact of failures associated with the incore instrumentation and seal table.
This will.
address Generic Issue 131.
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2.2 Seismic Quantification Thia task deale with calculating the impact of seismic events upon the risk to the public in terms of core damage frequency and containment performance.
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US NRC october 13, 1992 Page 9 2.2.1 Response Analysis This task will include the steps necessary to determine the response of the auxiliary / turbine building, reactor building, and containment to various magnitude seismic events. The building respor.se is required for the fragility assessment and is a function of elevation, ground acceleration, and the ground acceleration spectral shape.
Building response will require development of median centered response spectra to generate more accurate fragilities. The impact on plant operation will be assessed f rom the point of view of both potential for and consequences of liquefaction.
HUREG-1150 and EPRI NP-6041 contain procedures for assessing soil liquef action. The issue of incore instrumentation and seal table failure and the impact of various magnitude earthquakes.
2.2.2 Risk Models The IPE models will be reviewed and modified to reflect the impacte of seismic events.
The tasks completed for the IPE will essentially be i
repeated, with a specific focus on seismic events. Initiating events that could be caused by a seismic initiator such as Loop, LOCA, fire, flood, etc. will be identified.
Event trees will be reviewed and modification which includes new events in response to the impact of the seismic event.
The seismic f ailure modes will be added to the appropriate f ault trees to model potential f ailures of active and passive systems, components and structures, and secondary interactions. The fault tree cut sets will be formulated as !
lean algebra expressions, far systems, functions, and accident sequences as input into the sequence quantification.
Idettification of basic events in the Boolean expressions that represent similar components being subject to the same seismic motion and determining what should be coupled so as to not overlook common mode responses and failures to the seismic event.
2.2.3 Fragility Analysis The fragility of structures relative to the earthquake level identified in earlier tasks, as impacted by the structural analyses, is determined by this task.
Fragility curves and High Confidence of Low Probability of Failure (HCLPF) calculations will be performed.
The list of components for which these-calculations are necessary will be limited if possible through the use of importance rankings and seismic capacities of the components.
Those con.ponents identified as having inherent high seismic capacity and HCLPF will not require generation of fragility curves. Fragility curves for components relatively sensitive to seismic motion in the range of intarest will be generated.
2.2.4 sequence Quantification This task is performed to calculate the impact of a seismic event upon the plant and its systems.
The hazard curves will be convolved (combined) with the fragility information and the risk models to derive sequence frequencies. The Boolean expressions generated earlierifor the accident sequences will be combined with the hazards and fragility analyses. HCLPF-results for systems, sequences, and the plant will be generated.
Recovery actions will be considered at +.his time. The emphasic will be on-actions taken from the control room and on accessioility for actione that must be taken outside of the control room. Time frames developed from the sequence analysis performed for the internal events evaluation will be used.
2.2.5 Containment performance Similar to the "back-end" analysis performed for the IPE, an evaluation of containment performance is requircd for the IPEEE. Starting with tho IPE results, the containment strs tres, penetrations, isolation paths, and supporting. systems will be aluated to determine the impact of the postulated seismic events upon the ability to contain radionuclides.
Information gathered during the seismic walkdown will be of use here. The a
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US NRC october 13, 1992 Page 10 emphasis will be pl6.ad on the following three systems / functions: (i) penetrations, (ii) inviation, and (iii) containment heat removal / pressure
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.y containment failures that might result in high A
consequence seg aces or which may initiace accident sequences are also of concern.
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Containment penetrations are generally s o ically ri gga.
A rigorous q
tragility analysis is only needed at revia > ).svels greater than 0.3g.
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evaluatior u the back-up air system of the aquipment hatch and personnel locks that r.mploy inflatable seals will be performed for mil review b.
tevels.
The possibility of }ssing cooling to penetrations requiring oling will be considered as well,
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9f f '.vec in containment isolation systems are expected to be seismically vg 4 -
.gged (NUREG/Ch-4734). The walkdown should assure tha; they are similar dus o tust data on valves having high capacities, and that there a a no apatial intera:t'on issues.
Seismic failures of actuasion and control nystems are more likely to cause isolation system failures.
The air
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- ystem should be included if valves rely on a back-up air system.
Components of the conta1 ament heat removal /pressuro suppression systems
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which ue not included elsewhere and which are not known to have high t
capacities should be exa.nined (e..g.,
a fan cooler unit supported on ke isolator shims).
The walkdown will include these components and the.
'g anchorages.
Support systems and other system intoraction effects (e.g g
relay chatter) should be examined ar applicable. The potential for gro, s
structural failure will onsidered.
y 2.2.6 Sensitivity Studieu The results with the LLNL.uzard curvea 4111 N T. viewed to determine if I
there are insights that were not antablishe2 v' 9a -he EPRI hazard curves.
Assump _
e ut the potential for the sais.oic event causing a fire and also
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L.9 the ability to tight thn fire by causing damage to plant 4
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.he same time will be assessed. This is one of the six Sandia fire issues required to be evaluated in the IPEEE.
p sensiti. ity studies will be performed to address uncertainties in elements such as hazard estimates, system responses, and structural capabilitias.
She potential uncertainties associated with recovery will be Ldressed g
here as well core damage results with and without operator action will be 4
assessed la accordance with the generic letter. A sens).tivity study will be performed to addrcas the generic issue 131 on the seismic ef fect on the incore instrumentation and the seal table.
2.3 Documentation This task's objective is to complete a set of documentation that will (i) 2
seismic concerns, and (lii) be conPistent with the IPE documentation and the
" liv ng PRA" concept.
1-A ' detailed technice.1 and edit orial review will be completed prior to g
completing the docucentarian.
The review will be conducted by a team of experts. A separate revisa will be conducted by an in-house NSP raview team.
To facilitate the NSP review team's work, the NSP rev. ewers will receive T
trainLng on the methods and results of the seisc.ic evaluation.
- 3. FIR 5'.
I Fire is the second largest part of the IPEEE. The scope of work will include a fire PRA analysis whico has three major parts to it: 1) Data, 2) Quautification, and 3) Documentation.
3.1 Data The purpose of this tack is to gather and compile the ir ~armation necessary E-
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US NRC October 13, 1992 Page 11 to perform the fire evaluation.
Thit task includes gathering plant and system information including relevant analyses and procedures for use in supporting the fire IPEEE.
3.1.1 Plant Information Existing documentation on plant layout such as room configuration, and fire barrier locations is required.
Also necessary is information concerning the composition of the fire brigade, its training, and procedures relative tc fire brigade actions.
During this time, the analysis team will ensure that issues identifJed in NRC I&E notices concernir.g fire barrier qualification (see, the list of notices in Section 7 of the FIVE method) have been addressed.
This will satisfy the Fire Barrier Qualification issue from the Sandia fire scoping study.
The Seismic / Fire Interactions iss"e is addressed in the seismic evaluation.
Verification that there are remote control anc monitoring circuits independent of the control room for alternate shutdown purposes will be done to address the Sandia issue on Coatrol Systems Interactions.
If a site does not have independent remote circuits, Appendix R must be reviewed to verify that the safe shutdown circuits have been located physically independent of, or can be isol sted from, the control room for an exposure fire that causes a loss of control from the control room.
Analyses from the FSAR, Level 1 PRA, and f rom the Appendix R analysis will
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be included as relevant to the plant's response to postulated fire events.
,y 3.1.2 Identify Fire Areas
't Before the detailed fire evaluation can beg.tn, it is necessary to identify the areas within the plant where a fire might be an important concern.
The likelihood of a fire starting in a particular area, the equipment located thero, and other considerations must be f actored into this task.
The likelihood of a fire initiating in a particular area _s cansitv'e to the locatir, within the area where the fire is postulatad tc begin. This is particularly true if fuel loading conditions, the potential for fire spread, and other factors are considered.
Plant-specific information on fires and combustible loadings will be coupled with industry data.
The plant fire walkdown information will also be incorporated as appropriate From plant layout drawings and P& ids, arets adjacent to the area in which a fire is postulated to occur will be identified. The area boundaries can be defined by existing FSAR and Appendix R work, using approve fire barriers as the dividing line between areas.
Sepacation w' thin and between areas will be considered, as well as potential pathways such as ventilation systems, penetrations, grates, etc.
Equipment located in an area will be used as a means of defining whether or noc the area is an important fire area.
The amount and type of equipmcat locateo in an area can be used as a screening mechanism to minimize the number of areas which must be investigaued.
An area might not contain any equipment important to safety, but it may be adjacent to and connected freely with an area tb7t does contz' important equipment.
Therefore, areas such as those cannwt be discounted automatically.
The potential for cross-zone spreading of fires, and the possibility of transient fuels being located in an aren, are important considerations.
Detection anu suppression capabilities as a function of fire area will also be identified.
P& ids and other drawings and documentation such as Appendix R, should contain this information.
The types of detectors and suppression systems in fire reeas will be delineated.
Failure Modes and Effects Analyses (FMEAs) will be used to determine the impacts of postulated fire induced failures of equipment that might not be modeled in the fault trees. This information will be used to supplement the fault tree models.
US NRC '
October 13, 1992 Page 12 3.1.3 Walkdown This task will involve planning the walkdown and documenting the findings for use in the IPEEE.
The walkdown will be conducted to support the evaluation and documentation requirements of the IPEEE generic letter.
Preparatory work for the walkdown includes the f amiliarization with plant design features, P& ids, FSAR sapecially fire-related sections, electrical one-line diagrams, operating procedures and recovery actions, arrangement drawings, dependency matrices, and Appendix R.
Identifying it.4rtant plant functions, Pystems which satisfy those functions, plant fire zones, potential propagation paths for fire, smoke, and water, and possible s
access paths for fire brigade w W responding to fire alarms.
The walkdown has the following cbj=c ines: 1) A tw previously cevMoped plant system models for obvious fire impacts.
- 2) Select or verify potential success paths.
- 3) Verify assumptions on fire detection and senpression capabilities, or gather information to determine capacilitiee, order to assist in the fire evaluation.
- 4) Verify or develop assumptions regarding cross-zone fire spread, fire brigade accessibility, transient and permanent fuel loedings, etc.
- 5) Identify failure modes of concern. 6) Ensure that cable routing ueed in the analysis represents as-built information, and verify the treatment of any existing dependence between remote shutdown and contrcl room circuitry.
The walkdowns will be performed to satisfy requirements of the 1PEEE issues. Confirmatory walkdowns may be necessary following the preliminary assessment to verify or modify the results of the sequence quantification and to support the recovery analysis.
1.1.4 COMBRN Code The COMBRN code will be setup to evaluate tl e fire progrescion.
The COMER" code will have been check out during the Monticello effort.
3.2 Quantification The initiating evant frequencies for fires are combireJ with the probabilitier aeLociated with fire detection, fire greath, and fire suppression, as well as with random failure probabilities, 'co calculate the impact of fire upon the plsnt risk.
3.2.1 Initial 1areenin-
~
A fire analysis could rrquire a large effort.
In order to minimize the g
resources devoted to the fire analysis, an initial screening effort is conducted.
Only fires and fire areas that have a significan; impact on plant risk are investigated in detail.
Initiating events will be related to the svent trees. The fires that are postulated to occur in different arcas could result in a variety of plant trips, ranging from manually initiated plant shutdowns or scrams to conditions such as loss of feedwatfr, turbine trip, loss of off-site power, or even a LOCA.
An assessment will be made for each fire and area of concern as to what the bounding initiating event caused by a postulated fire would be.
The equipuent within a specified area that might be affected by a fire in that area is next related to the functions and systems contained within the IPE event trees.
In this way the analyst 4
begins to get an idea abcut the possible magnitude of impact the fire might have on the plant.
A cuantitative screen will be performed by using the event tree and f ault tree logic models and the quantification computer codes, and modifying the unavailabilities to reflect the t,stulated impact of the fire upon the equipment.
By initially assuming that all equipment in a given area is disabled by tre fire without crediting detection or suppression systems, and by quantitatively removing those equipment from the logic models by setting the f ailure probabilities to unit, the maximum impact of the fire
i US NRC October 13, 1992 Page 13 can be quantified. Those areas which produce high core damage frequencies when a fire is postulated in this manner are retained for further refined investigation.. The quantitative screen is first performed on each area individually, without considering the potential for fire spread.
Areas which produce h.tgh CDFs in this quantification are maintained. They will-also produce high CDFs if the fire spreads to adjacent areas. Areas with low CDFs will be investigated further to ensure that fire spread is not a factor.
Areas which have low UDFs from the first screen are now invcatigated further. If the fire has the possibility of spreading, Lhen the adjacent areas are also considered to be involved in the fire, and the equipment in those areas is conside.ed disabled.
A new CDF is calculated.
If it is low, then the area and its adjacent areas are dropped from further consideration. Otherwise, they are maintained.
This screening is. done in two steps.
It is first performed without considering the alternate shutdown panel. Low CDFs produced 4.n this uay indicate that the areas can bs dropped.
High CDFs continun to the next step.
The probability of being aole to use the alternate shutdown panel as a means - of. shutting down the plant is derived.
This value is a combination of equipment fallare -probabilities and human error probabilities. Effects of the postulated fire must be considered. Once the unavailability is derived, it car be combined with the effects of the fire under consideration.
A qualitative screening approach will also be applied.
This approach credits separation between equipment, including cable trales, and the layout of. the area and its potential for preventing the fire from impacting all equipment. The high CDF fire areas maintained are reviewed 1
qualitatively by fire experts to determine if any mitigating factors might exist which would lead the analyst to believe that the fire cannot impact sufficient equipment to cause a significant effect on plant riuk.. The results are documented thoroughly by the analyst, including asuumptions and rttionale, and the list of fire areas for. further investigation is
.)
reduced more, if possible.
3.2.2 Fire Frequency Thr. initiating event frequency - will be derived for those areas most impe.,rtant as determined by the initial screening. Combustible loadings in the areas under study will be determined or estimated.
The amount of fixed combustibles should be available from plant documents such as the I
Appendix R report. Transient combustibles such as paint cans, wood,.and.
trash are harder to pin down. Estimates must be made based on maintenance records, daily work traffic patterns through the areas, administrative controls, etc.
The 1.'e frequency is derived as a function of the area, the combustible loadinp, and the number and type of potential ignition sources.
Plant-specific data on fires may exist for some areas, but dependence on generic -
information is usue.lly accessary.
The frequency can be chrived from s.urces such er dasa
'.4 EPRI's FIVE method, FIVE, NRC Daily Plant Status imports, and shoreham and Seaurook PRA related studies. Where possible, a frequency should be estimat for both "large" and "small" fires within an area.
With the fire frequency assigned, the sequences will ba re-quantified, still without any consideration of detection and suppression.
If the combination of initiating event frequency and the conservative assumption that all equipment fails produces a low CDF, then this can be-documented and the area can be dropped from further consideration.
3.2.3 Fire Damage Now the f actors of detection, propagation, and suppression must be applied 1
US NRC October 13, 1992-Page 14 to the remaining fire areas and postulated fires.
This task is to establish time windows for propagation and detection. The items that raust be established for each area are:
- 1) Time from luitiation of fire to spread of fire to important locations or equipment within the area.
2) ilme from initiation of fire until detection by available automatic detection devices or by manual dotection, by observing the fire or by observing eftects of the fire upon equipment and inutrumentation.
These time frames are important beccuse they impact the probability of successfully suppressing the firo before suf ficient damage to equipment to cause the onset of core damage occurs. The time available to suppress the fire is dependent on the initiating event. and will be established as available from existing transient analyses performed for internal events.
Methods of estimating these times include use of available computer codes to model the fire scenario, specifically, the COMPBRN code or the use of look-up tables included in the FIVE method.
Bounding nlues can also be used, with sensitivities assessed by varying these times.
The fire modeling techniques incorporated into the FIVE method are derived from the same basic correlations used in the COMPBRN IIIE code. The look-up tables in the method shculd be compared to results from COMPBRN.
If the results appear to be satisfactory, then the look-up tables be used by themselves.
Otherwise, the use et COMPBRN may be required.
This will satisfy the Sandia fire risk scoping issue Improved Analytical codes.
Sensitivity analyses will be performed to test the impact of major assumptions upon the results.
Possible egndidates for these studies include:
- 1) Starting size and locnion of the -' fire.
- 2) Amount of transient combustibles in the fire. 3) Assumed per onnel traffic patterns through or near the area..
- 4) Effect of area ventilation (e.g.,
vents open/ vents closed). 5) Effects of multiple combustibles (i.e., original fire starts a different set of combustibles on fire).
3.2.4 Fire Suppression Suppression of the fire via man.nal or automatic systems must be taken inte account.
The effectiveness of various systems (e.g., water'va. COi vs.
halon; hoses vs. sprinklers vs. deluge pipes vs. hand-held extinguishers, etc.) varies considerably and is a function of the type and size of the fire.
1 Fault trees for suppression systems,. such as tne fire water system will be developed and quantified us'aa clant-specific data on fire equipment reliability where availablt.
.3d on surveillance frequencies of the actual' system con guration. Gueric aata may be necessary to supplement plant datn.
c If manual suppression is necessary, it in first necessary to determine if and wher. the fire is detected.
For areas with automatic detection devices, such as smoke, firme, or heat detectors, the fire evaluation anslyst needs to assign a reliability to the detection system.
The ef fectiveness of the systems in detecting a firs influences the tability of the fire brigade to extinguish the fire.
If plant-specific data >is available for detector reliabilities, it should be used.
Otherwise, reliance on generic. data is expected.
With the reliabilities of detection and suppressic. systems derived, the effectiveness of those systems will now ue calculY.ed. Detection system effectiveness, a function of the type-of detector and the type of. fire, will probably require reliar.ce on generic data. Some published PRAs have references which may be useful, as does the FIVE method.
Estimating automatic suppression system ef fectiveness is next oL the list of tasks to t womplish.
The size and type of the fire, and the type of suppression system respondir.g to the fire, impact the ef fectivtMens of the
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US NRC
- October 13, 1999 Page 15 t
suppression efforts.
The data in FIVE on fire suppression system unavailability refers to the probability of not extinguishing a fire upon dennd.
In other words, "raallability" is defined as-operation of the s). :en and successful suppression of the fire given that the system operates.
Halon and CO systems are not included in the FIVE data, and 2
will require estimation from other data sources.
If automatic suppression systems do not exist or do not work, the fire.
brigade must suppress the fire. The FIVE method contains a discussion on
}
hcw to determine the effectiveness of the fire brigado.
The relevanc elapsed time is the time from fire initiation until the fire is controlled, as opposed to the time f rom detection. The impact of heat and smoke on manual fire fighting effectiveness, the fire fighter preparation time, and the time needed to locate the fire in a smoke filled area will also be considered. If these topics are addressed, this will satisfy the Sandia fire scoping issue on Manual Fire Fighting Effectiveness.
3.2.5 FMEA On Suppression System Operation An FMEA will be used to supplement the existing fault tree models.ind results.
This FMEA will focus on areas for which detailed models do not exist or may be difficult,to complete.
During this time the Sandia fire scoping issue -on Total Environmental Equipment Survival-should be-addressed. This requiras-that an assessment of three topics be completed:
l
- 1. Potential for adverse effects on equipment caused by combustion i
products.
l
- 2. Spurieus or inadvertent actuation of fire suppression equipment. This conc ~cn la being investigated in Generic Safety Issue G7.
The fire-evaluation will include an assessment of the susceptibility of safe shutdcwn equipment to damage from the suppressant being used.
The plant walkdown will identify potential concerne regarding redundant trains of shutdown equipment being simultaneously damaged from inadvertent actuation of a. suppression system.
- 3. Operator effectivenens in smoke-filled environmente.
Appendix ' R evaluations on the ability of the plant to implement fire procedures and use the alternate shutdown panel should be of use-here.
3.2.6 Containment performance The IPEEE G3neric Letter does not really discuss containment performance in response to fire events. Past PRAs have indicated that the effects of fires upon containment have been small.
However, if the fire evaluation determines that containment isolation - or cont runment integrity may be impacted by the fire, a discussion of these events will;be included. At a minimum, the fire evaluation will ensure that no sequences dif ferent from those of the internal events e Talysit are identified during the fire analysis.
3.3 Documentation This task *.s objective is to complete a set of documentation that will satisfy the NRC's IPEEE and be useful for identifying and addressing fire concerns.
It will also be consistent with the IPE documentation and the "living PRA" j.
concept.
A detailed technical and editorial review will be completed prior to completlo.* of the documentation. The revies will be conducted.by e team of a
experts.
A separate review conducted by an in-house NSP review team.
To facilitate the NSP revies team' work, the NSP reviewers will receive training on the methods and results of the fire evaluation,
- 4. Other This is the evaluation of other external events not part of the seismic or fire analysis. This includes external flooding, high winds, tornadoes, transportation
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j US NRC
-l October 13, 1992 Page 16 accidents and nearby facility accidents.
This task will also ensure that no other plant unique external event hazard exists. This task in broken down into three parts;
- 1) Review Data and Licensing, 2)- Screen Hazard, and 3)
Documentation.
)
4.1 Review Data and Licensing 1
This task is the review of the currently available data and licensing requirements.
Review all current hazard data available for the plant to determine to what extent the hazards have been documented.
Review the current licensing bases for each external event.
Review any significant changes in the operating license with respect to 1) military or industrial facilities within 5 miles, 2) on-site storage or other activities involving hazardous materials, 3) Transportation, or 4) developments that coM d affect the original design flooding conditions.
4.2 Screen Hazards This is a screening of the hazards. Only thoso hazards which do not pass the -
previous screen are reviewed by the next screen.
Compars information obtain'ed for conformance to the current criteria and perform a confirmatory walkdown on the plant. The walkdown would concentrate on outdoor facilities that could be affected by high winds, on-site storage of hazardous materials, a n.1 off-site developments.
If the comparison indicates that the plant conforms to the current criteria and the walkdowns reveala no potential vulnerabilities not included in the originsi design basis ar; vais, it is judged that the contribution form that hazard to core damage frcquency is less that lE-6 per year and the IPEEE screening criterion j
is met.
Otherwise, one er more of the following steps will be taken to 3
further evaluate the situation.
If the original de m.t. basis does not meet the current requirements, then the original design ba m will be reviewed to demonstrate that the-frequency of 3
the original design bases is sufficiently low.
This is done by showing that the design bases is less then lE-5 per year and that the conditional core damage frequency is less then lE-1.
If the original design bases with the conditional core damage frequency is not sufficiently low then additional analysis will be required.
This analysis is intended to provide a conservative calculation showing that either the hazard would not result in core damage or the conditional core damage frequency is below the reporting criterion.
The level of detail is that level required to prove the point.
If the bounding analysis cannot be shown than 1dditional analysis will be nquired.
A PRA consists-of hazard frequency, plant systems, accident analysis and i
radioactive material release analysis. A core damage frequency J ess than lE-6 per year would screen the event from further consideration.
the level of detail is that level required to prove the core damage frequency is below lE-1 6 or the determination of recommendations to reduce the frequency below lE-6.
4.3 Documentation i
This task'.s objective is to complete a set of documentation that will satisfy the NRC's IPEEE an'* me useful to NSP for identifying and addressing concerns.
i It should also be e sistent with'the IPE documentation and the "living PRA" concept.
A. detailed technical and editorial review will be completed prior to com, stion of the documentation. The review will be conducted by.a team of experts.
)
5.0 Documentation This is all the final documentation and reviews. This task's objective is to put together all the documentation from the other main tasks into one consistent and complete a set of documentation that will satisfy the NRC's IPESE and be useful for identifying and addressing concerns. It should also be censistent with the
i f
US NRC*
october 13, 1992 Page 17 IPE documentation and the "living PRA" concept.
A detailed technical and editorial review will be completed prior to submitting the final IPEEE report.
The review will be conducted by a team of experts.
The final IPEEE report for submittal to the NRC which incorporated the comments from the final review will be developed.
6.0 Update PRA Ther6 will be several years since the PRA model was completed, and with changes and modifications being performed routinely, the PRA must be updated to ensure the IPEEE reflects the c. crent plant design. Only significant changes which will 4
effect the results of the IPEEE will be updated. The update includes reviewing the data, event trees,. fault trees,_quantification, and sensitletty studies to en9ure that the IPEEE reflects the current plant ' design, and updating the data, i
models and documentation as needed. It is anticipated that a new quantification will be required to moet these objectives.
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US NRC l
October 13, 1992
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Page 18 Prairie Island IPEEE Schedule u
I Task 1994 1995 1996 I
t
- 1. Project tianage.Sent 1/1 12/15
- 2. Geismic 2.1 Plant Information 2.1.1 Component Information 1/1 6/2
.2.1.2 Review Hazard Analyses 6/2-6/11 2.1.3 Walkdown 8/6 10/1 4/27 5/17 lj9--1/30 2.1.4 System Analysis
'6/11--8610/1 3/16
/
2.2 Quantilicatier j
7.241 Response Analysis 3/16 8/11 2.2.2 Risk Models.
8/11 1/6 2.2.3 Fragility Analysis 7/29 6/8 2.2.4 Sequence Quantification 1/30 4/17 2.2.5 Containment Performance 4/17 5/17 1.
2.2.6 Sensitivity Studies 5/17 7/21 2.3. Documentation 4/17 8/30
- 3. Fire 3.1 Data T.1.1 Plant Information 1/1-1/13 3.1.2 Identify Fire Areas 1/13 4/7-3.1.3 Walkdown l'4/7 4/23-3.1.4 COMBRN Code 1/13 3/10 3.2 Quantification-3.2.1 Initial Screening 4/23 6/21 3.2.2 Fire Frequency' 6/21 7/28 3.2.3 Fire Damage 7/28 8/30 3
3.2.4 Fire Suppression
.8/3C 11/9 3.2.5 FMEA Suppression 11/9 - 11/19 3.2 : Containment Performance 11/19 - 12/1 4
+
3.3 Documentation 12 /1 --- --
2/2
- 4. Other j..
4.1 Review Data and Licensing 1/1-1/27 4.2 Screen Hazards 1/27 2/24 4.3 Documentation 2/24--3/17 5 Documentation 7/21 12/15 6 Update PRA 2/2 9/7 4
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