ML20128D949
| ML20128D949 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Seabrook |
| Issue date: | 06/05/1984 |
| From: | Cleland L, Ariuska Garcia, Jerrica Johnson LAWRENCE LIVERMORE NATIONAL LABORATORY |
| To: | Davis S Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19292B772 | List:
|
| References | |
| CON-FIN-A-0801, CON-FIN-A-801, FOIA-84-624 NUDOCS 8505290253 | |
| Download: ML20128D949 (18) | |
Text
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PrejpctTitle:
Review of Seabrook Nuclear Power Plant FIN /189 No.
A0501 Prorfacilistic Risk Assessment B&R No. 2'0-19-40-41-5 Wang No. 0473x/6-il-84 1.0 0 EJECTIVE OF PROPOSED WORK l.1
Background
The Office of Nuclear Reactor Regulation is conducting a probabilistic risk assessment (PRA) review program in which PRAs performed and submittet to the NRC by license applicants and licencees receive comprehensive review and ev al u a't ion. The program is the responsibility of the Reliability ana Risk Assessment Branch (RRAS).
A PRA of. the Seabrook Nuclear Power Plant (Seabrook PRA) has been submitted to the NRC. by Public Service Compapy of New Hampshire, an operating license (OL) appl i c' ant. The review of this document, whose title is "Seabrook Station Probabilistic Safety Assessment," is being performed as one. project in the larger NRC program.
1.2 ' Objective The objective of, this_ pr.oject is to review those aspects of the Seabrook PRA leading to the estimates of the frequencies of each plant damage state to 4
determine the accuracy of these estimates.
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2.0
SUMMARY
OF PRIOR' EFFORTS LLNL has been involved in PRA methods development and applications for the NRC for a number of years dating back to our involvement in the Reactor Safety-Stuoy. Current activities' include a review of the Millstone 3 PRA for NRR/ DST; the oevelopment of a seismic risk assessment method 31ogy for RES/DET in our Seismic. Safety Margins Research Program (SSMRP), and a systems interaction methodology for NRR/ DST.
Relateo PRA activities are ongoing for DOE anc FEMA projects.
3.0 WORK TO BE PERFORMED AND EXPECTED RESULTS Perform a comprehensive review and evaluation of the qualitative and quantitative aspects of the risk assessment being submitted by the licensee for the Seabrook PWR power plant to determine if the estimates of the frecuencies of tne plant damage states reflect appropriate use of risk assessment metnoos, plant / site information, and reliability data.
.3 L
, Prc' ject
Title:
Review of Seabrook Nuclear Power Plant-FIN /189 No. 'A0801
-Prcoabilistic F.isk Assessment BLR No. 20-19-40-41-5 Wan; No. 0473x/6-il-54 The defensibility of the licensee's submittal of the frequencies of plant damage states and the associated urtertainty spread with respect to (1) use of state-of-the-art risk assessment methods, (2) thoroughness and comprehensiveness-cf analysis, (3) availability and appropriate use of data, and (.4) realism of modeling assumptions will be evaluated.
The impact of H
deficiencies of understated uncertainties in the licensee's analysis of sequences will be described and discussed.
The review will cover all aspects of the study up to the point of the calculation of the frequencies,of the plant damage states, including methodblogy, assumptions, data, information sources, models, plant understanding, completeness of the analysis, and any other' area where inconsistencies may;ar.ise which could affect the quantitative or qualitative
~
results. _ A.ltergatives.ideritified in the review willle, considered in appropriate combinations to oetermine the incremental thange resulting from the use of. alter.nativ.es.in.the dominant sequences corresponding to each plant dhmage state."In general', these alternatives will be evaluated by performing
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localized calculations'within the analysis.
For example, minor incons.isten-
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cies identified in the event trees will be reevaluated to determine how alternative approaches _would impact the sequence probabi.lities.
An approximation of the impact of identified major errors or inconsistencies will be made; however, extensive requantification of event tree accident sequences are not within the scope of this project.
Should such grave errors 'or inconsistencies be' identified, the LLNL project' manager wilf consult with the-NRC project manager to determine the appropriate action to be taken.
T1e schedule for completic,n'of the PRA review and delivery of a draft final
' report has been established at five mor>ths, with the final report due at eight months. The project start date has been established, as indicated on Page 1 (NRC Form 189).
Tne projecteo short schedule and relatively small budget for this review is predicatec on several critical assumptions, principally concerneo with access 4
'_j r'oject
Title:
P Review of Seabrook Nuclear Fower Plant FIN /189 No. A0801 Probabilistic Risk Assessment B&R No. 20-19-40-41-5 Wang No. 0473x/6-il-84 to information requirements and the overall scope of our review. We have specifically mace the following assumptions:
a.
Extensive accident sequence requantification requiring computer
-support is out of scope.
b.
' Review of any consequence analysis ' included in the PRA is out of scope.
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q..
Although identifying significant' dependencies is a critical part of the review, per'orming a complete dependency analysis for front-line and support systems (at the train-to-train level) is out of scope.
Insufficient' plant data will be ava'ilable to the reviewers to
~ perfo'rm such' ariinalysis.
~
LLNL will have contact with a designated individual (a senior level d.
representative) from the engineering staff of the plant or utility, and access to the plant site, exclusively coordinated (at least initially) by the designated individual and the NRC project manager.
Site visits and requests for information by subcontractors /cpnsultants will be coordinated by LLNL to coincide with similar LLNL visits and requests for information in order to minimize the buroen placed on the uti.lity.
,.. ~
e.
Our review will be based on the NRC material listed in Section 12 herein and any additional information supplied by the NRC or licensee to support this review.
The tasks to be performed are described below.
Task 1 Internal Events Evaluation (a) Review and evaluate the scope, assumptions, and systems analysis for internal events. Review for completene: S the initiating events in the,
Pro [ectTitle:
Review of Seabrook Nuclear Power Plant FIN /189 No.
A0801 Prebabilistic Risk Assessment B&R No. 20-19-40-41-5 Wang No. 0473x/6-il-84
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l PRA including both the following events and combinations of initiating events with subsequent failures:
1.
Scation blackout (loss of offsite and emergency AC power) e.
a)'
Reactor Coolant Pump (RCP) seal f ailure (af ter station -
blackout).
b)
Loss of DC af ter finite time, due to battery depletion (after station blackout), or immediate loss of DC because of prior undetected batte,ry failure (after loss of offsite power).
2.
Loss of DC power 3.
Loss of instrument and control power.
(
4.
Multiple instrument tube LOCA below core level.
5.
Overcooling events - (Pressurized thermal shock).
6.
' Steam ieneraior~t'ube f ailure, including su'bsequent stuck-open
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secondary side safety relief valve 7.
ATWS" '
r 8.
Stuck open S/R valve.
9.
Loss of main feedwater.
10.
Containment isolation f ailure.
l,
- 11. ' Turbine trip.
l 12.
Loss of componen.t cooling water.
13.
Loss of service water.
14.
Loss of ventilation in auxiliary buil. ding.
l 15.
Pipe breaks in auxiliary building.
16.
Reactor Coolant Pump seal f ailure (as contribution to small LOCA initiator).
17.
baron dilutio'n.
18.
Excess feedwater events, including their possible contribution to j
loss of main feedwater 19.
Loss of instrument and control air.
l Assess the completeness and treatment of transients /LOCA initiating events f'
wr.ich aisc degrade mitigating systems.
! r l
L
- {. 'PYcjectTitie
Review of Seabrook Nuclear Power Plant FIN /189 No,'
A0501 L
^' Prebabilistic Risk Assessment
'B&R No. 20-19-40-41-5 Wang No. 0473x/6-11-84
' ' (b) Develop a table of assumptions used in the analysis and make a finding on their validity..These assumptions include such considerations as pump operabilities with insufficient cooling or net positive suction pressure, L
operability of components in degraded environments, minimum components needed to sustain a function, and availability of equipment based on technical specifications.
(c) Review the event trees for completeness and validity, and identify all functional and support system interdependencies that were considered or s'hould be considered in t'ne evaluation.
This should include coupling with the initiating event.
Evaluate the significance of areas which are incomplete.. Review the. stated and implicit assumptions associated with success paths in the event: trees (e.g., feed 'and bleed).
Review the
, evedt tree"logici add assess the vaiidity of,_ ass.ignment of combinations
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of successes and f ailures 'of systems to.the plant damage states (system f
success criteria).
Consider system dependencies arising from interaction related to the physical phenomena; identify and evaluate any omissions.
Compare system success criteria to those used in other PRAs for similar systems under similar circumstances.
Assess the validity of the treatment of transient-induced loss of coolant accidents, and other such' transfers'from one type of event tree to another.
(d) Review the f ault trees to determine.their. accuracy, validity, and comp'leteness in quantifying the high risk sequences.
The subjects listed below are of particular interest, o
definition of the top eve ~nts in fault, trees, o
test and maintenance unavailabilities, o
consistency of fault trees with system success criteria, o
common cause interactions, o
human errors (cognitive, procedural, latent, dynamic) o design errors, o
consequential failures.(coupled to initiating event),
o failure rate data and uncertainties, o
completeness of f ault trees. u
,. P 5 ject
Title:
Review of Seabrook Nuclear Power Plant FIN /189 No.
A0801 Prebeilis:ic Risk Assessren B&R No. 20-19 40-41-5 Wang No. 0473x/6-ll-84 Review ano evaluate the depencency analysis performed in the PRA.
The PRA snoulc include an analysis, on a train-to-train basis, for the depencency of front-line systems on support systems, and the dependency cf support systems on other support systems.
If this analysis is not proviceo, or is incomplete, use available information to perform a limitec dependency cnalysis on a train-to-train basis - e.g., which trains of a given frent-line system depend on which trains of a support system. Display the results in the form of a dependency table or diagram.
Identify and ev,aluate the impact of any omissions in the PRA.
(e) Determine whether the limit of resolution in the fault tree analysis is acequate for the identification of common mode f ailures.
Determine the adequacy of'the identification and qualitative analysis of common mode f aiiures in'd systems interactions including the. treatment of common mode f ailures arising from hard-wired coupling, spatial / environmental coupling, coupled human errors, design errors and inadequate testing resulting in the f ailur'e to recognize that a system may respond successfully during a test but not during a true demand..
Develop a table of com.on cause considerations and compare quantification in this PRA with previous PRAs.
(f)
Assess the completeness, accuracy, and adequacy of the conditi'nal o
probabilities for some system's success /f ailure given.tnother system's previous success / failure.
(g)
Determine the sources of data used for Qu.antification of the f ault trees, and comoare them t'o data sources used in other PRAs.
Highlight data used in quantifying trees if the data deviate significantly from data previously used in other PRAs. Particular interest should be accorded recovery or repair rates and failure rates.
Examine the validity of the method of combining generic data with plant specific data.
P dject
Title:
Review of Seabrook !?uclear Power Plant Fili /189 tio. A050i Probabilistic Risk Assessment B&R No. 20-19-40-41-5 Wang rio. 0473x/6-11-S4 (h)
Identify all known omissions and deficiencies in the systems analysis including support system dependencies and estimate the impact where
- possible, include the technical basis for these estimates.
For those omissions and deficiencies for which evaluation is believed to be beyond the state-of-the-art, provide a list and the basis for this belief.
(i)
Incorporate fiRR technical review cornents as provided (success /f ailure criteria, environmental qualifications, etc.).
(f)
Review the following issues and discuss their applicability to Seabrook.
o how NUREG/CR-2497, " Precursors to Potential Severe Core Damage Accidents:
1969-1979, A status Report" relates to the plant
[
~'nd its risic analysis, a
o how the recent failure of scram breakers to operate satisfactorily at the Salem plant relates to Seabrook and its risk analysis, o
the conditional probability that " containment f ails to isolate on demand",(provide a basis for this value).
(k) Assess the validity of the treatment of human errors,,both those occurring af ter an accident (dynamic) and prior to an accident (latent, e.g., test and maintenance errors).
Assess the completeness ar.d treatment of cognitive errors.
Develop a human. error table of significant actions showing time frame for action, availability of procedures and HEP, and compare with HEP from other PRAs.
(1)
Assess the analytical modeling ano quantification of accident sequences.
(m)
Review the sequence frequency calculations corresponding to each plant damage state to ascertain accuracy, uncerta.inty, completeness, and acequacy.
Wnere the reviewers disagree with the PRA in a significant way
-g-t
Prc: ject
Title:
Review of Seabrook Nuclear Power Plant FIN /189 No. A0801
".PEcbabilistic Risk Assessment-SLR No. 20-19-40-41-5 Wang No. 0473x/6-11-84 f
(e.g.,. assumptions, f ailure rates, success criteria, or$lissions), the reviewers will reevaluate the sequence frequencies by performing localized calculations within the analysis to determine how alternative approaches.would impact the sequence probabilities, y,-
(n)' Compare the dominant sequences corresponding to each plant damage state to existing PWR PRAs, generic data, and operating experience to help assure that no significant sequences are omitted.
(0) Assess the uncertainty ana' lysis. Examine propagation and completeness in treatment of uncertainty, data uncertainty, and modeling sensitivity / uncertainty.
Task 2' E'xternaTEvent's E'viluation
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(a) For each type of external event considered, concur with or modify the f
estimated hazard curves, assess the completeness -of the sequences considered for each plant damage state, assess the treatment of common cause interactions and human errors, and ev'aluate the fragility or
^
f ailure estimates of'the components identified by the' appropriate event / fault trees.
(b) Review the event and f ault trees to.determ.ine their accuracy, validity, and completeness for external events.
Review and evaluate the implicit and explicit assumptions of the external events analysis.
(c)
Identify all know'n' omissions and deficiencies in the external event analysis, and estimate the impact where possible.
For those omissions and deficiencies for which evaluation is believed to be beyond the state-of-the-art, provide a list and the basis for this belief.
(d) Review the sequence frequency calculations corresponding to each plant da age state to ascertain acct. racy, uncertainty, comple:eness, and adeouacy. Wnere the reviewers disagree with the PRA in a significant way (e.5., assu otions, failure rates, success criteria, omissions), the
. 10-
Frpject
Title:
- R'e' view of Seatrook. Nuclear Poaer Plant flN/189 No.
A0801
.,,-Procabilistic Risk Assessment ELR No. 20-19-40-41-E-Wang ho. 0473x/6-11-S4 reviewers will reevaluate the sequence frequencies by performing
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localized calculations within the analysis to determine how alternative approaches would impact the sequence probabilities.
(e). Assess the uncertainty analysis.
Examine propagation and completeness in treatment of. uncertainty, data uncertainty, and modeling sensitivity / uncertainty.
=
Task 3 Draf t Final Report on Tasks 1 and 2 (a). Fbr each of the PRA areas (e.g., initiating events, initiating event frequencies, component failure rates, corrrnon cause failures, human reliability analysis, test and maintenance, event trees, f ault trees, success /f ailure criteria, external events, quantification, and -
uncertainties)-determine if the analysis reflectL. appropriate use of risk assessment methods, plant / site information, and ' reliability data; define
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-the basis for the fincing and describe what was considered in the review.
The finding will include reestimates, as appropriate, of the dominant sequence frequencies corresponding to each plant damage state, identification of areas that were not pursued and identification of grey areas.where sensitivity studies rnight be used to bound a central estimate.
(b)
Describe areas of incompleteness (all known omissions.a'6d deficiencies) determined in the review. Quan.tify, where possible, the potential impact of these areas.
Discuss the basis for quantification values.
(c)
Based on the review of sequence calculations, discuss the accuracy, uncertainty, and adequacy of the PRA author's sequence quantifications.
(d)
An approximate outline of the report is given here..
t
i Projet: Titie:
Reviea of Seabrook Nuclear Power Plant FIN /189 No.
A0801 Probabilistic Risk Assessment BLR NO. 20-19-40-41-5 Wang No. 04~3x/6-il-Ei 1.
Executive Sum. mary 2.
Introduction
2.1 Background
2.2 Scope 2.3 Assumptions 2.4 Summary of Results and Insights Presented in the PRA 3.
Internal Events Analysis 3.1 Initiating Evints 3.2 Event Trees LOCA Transient
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dther 3.3 Success Criteria 3.4-Systems Descriptions Fault Tree Models 3.5 Human Factors 3.6 Failure Data 3.7 Operating Experience Analysis 3.8 Analysis Codes 3.9 Accident Sequences 3.10 Dependencies 3.11 Quantification 4
External Event Analysis 4.1 Seismic 4.2 Fire 4.3 Industrial Accioents
?.4 Other o
Project
Title:
Review of Seabrook Nuclear Power Plant FIN /lS9'No. A0801 Procabilistic Risk Assesscent B&R No. 20-19-40-41-5 a-Wang No. 0473x/6-il-54 5.
Summary and Conclusisr.1 5.1 Dominant Sequences Corresponding to Each Plant Damage State 5.2 Important Problems and Omissions 5.3 Treatment of Uncertainties 5.4 Overall Evaluation of Seabrook Risk Assessment
_m
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6.
Appendices (as required)
Task 4 Final Report The final report will take into account pertinent comments on the draft final report by NRC and other interested parties.
It will be published as a NUREG/CR.
This task assum'es that the NRC'will respond within the 45 calendar days allowed for comments in the project schedule.
Delay beyond this time will necessarily delay, completion of thE final report on a~ day-for-day basis.
Task 5 Ouestions to Licensee Provide questions for forwarding to the licensee covering significant all aspects of the systems analysis.
(Responses already may have been received informally).
4 DESCRIPTION OF ANY FOLLOW-OS EFFORTS None.
5.
RELATIONSHIP TO OTHER PROJECTS LLNL has developed and applied probabilistic risk assessment technology on the Reactor Safety Study, the Seismic Safety Margins Research Program (SSMRP), and various systems interaction studies for NRR.
Information from the SSMRP will be used in the evaluation of the component fragility or f ailure estimates.
In addition, information developec as a result of w.ork done on the Seismic Hazaro Characterization of the Eastern U.S. will be used, as appropriate, in the,
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Project
Title:
Review of Seabrook Nuclear Power Plant FIN /lE9 No.
A0831 Prchabilistic Risk Assessment B&R No. 20-19 40-41-5 Wang No. 0473x/6-11-84 external event evaluation of the seismic hazard curve for the plant site.
Some infermation from a DOE program to assess natural hazards at a number of DOE sites may be used.
6.
REPORTING REOUIREMENTS 6.1 Technical Reports Upon completion of Task 3, LLNL will formally submit a draf t final report covering Tasks 1, 2, and 3 as defined in Task 3.
Upon completion of Task 4 LLNL will submit the final report which takes into account industry and NRC comment s'.
Upon completion of Task 5, LLNL will submit questions for forwarding to the licensee.
6.2 Business Letter R'eport A monthlf busin'ess letter will be submitted at the snd'of each month of effort to the Project Manager with copies to the Director, Division of Safety Technology, ATTN:
J. Halvorsen; A. Thadani, DST; F. Rowsome, DST; R. Frahm, DST; and L. Solander, NRR. These reports will identify the title of the project, the IIN, the Principal Investigator, the period of performance, and the reporting period and will contain 3 sections as follows:
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Project Status Section 1.
A listing of the efforts completed during the period; milestones reached, or if missed, an explanation provided.
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ii. Any problems or delays encountered or anticipated and recommendations for resolution.
1/
1/
If the recommended resolution involves a contract modification, i.e.,
change in work requirements, level of effort (costs), or period of performance, a separate letter shoulo be prepared and sucmittec to the Direcict, Division of Safety Technology, ATTN:
J. Halversen anc a copy previ ec te the NRR Project Manager an: L. Solander, ND:
e Project
Title:
Revie. of Seatrock Nuclear Power Plant FIN /159'No.
A0E01 Probabilistic Risk Assessment B&R No. 20-19-40-41-5 Wang No. 0473x/6-li-84 iii. A summary of progress to date (this may be expressed in terms of percentage completion for the project).
iv.
Plans for the next reporting period.
Financial Status Section i.
Provide the total cost (value) of the project as reflected in the proposal (NRC Form 189), the total amount of funds obligated to date, the balance of funds required to complete the work by fiscal year, and a summary of costs as shown in Attachment A.
ii.
Provide the estimated total amount of funds expended (costed) during the period prior.to the report' period and cumulative to date (see Attachment Fee Recovery Cost Status Section P
Pursuant to the prov'isions of NRC Regulations,10' CFR 170, provide the total amount of funds expenced (costed) during the period and cumulative to date and report them on a_ separate page as part of this report in the format shown in the second page of Attachment A:
7.
SUBCONTRACTS, CONSULTANTS AND/0R MATERIALS
,~
We anticipate the use of subcontractors and consultants to assist in this work.
The two subcontractors and one consultant listed below have already been specifically identified, others may'also be used.
a.
Subcontractors (1) Applied Risk Technology Corporation ( ARTECH); P. J. Amico (2)
Jack R Benjamin and Associates, Inc. (JBA); J. W. Reed, M. W. McCann, Jr.
b.
Consult 6nts l
(1)
F. R. Davis i
'x 15
Pro 3ect
Title:
Review cf Seao cok Nuclear Power Plant FIN /lS9 No.
A0801 Probabilistic Risk Assessrent BLR No. 20-19-40-41-5 Wang No. 0473x/6-11-54 S.
LIST NEW CAPITAL EQUIPMENT RE0UIRED Not Applicable.
9.
DESC::EE SPECIAL FACILITIES REQUIRED Not Applicable.
10.
CONFLICT OF INTEREST INFORMATION No known conflicts or apparent conflicts of interest.
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11.
MEETINGS AND TRAVEL t
c Contractor may attend a 2-day visit at an unspecified site with the licens.ee to ' discuss questions on his analysis and may attend six
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'l-daf~ meetings at NRC headquarters in Washiy[gton, D.C. for four-participants.
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o Four individuals may conduct two three-day visits to the Seabrook site to look for potential common cause vulnerabilities, review of control operations, and a general overview of the plant layout and
' potential idiosyncrascies.
12.
NRC FURNISHED MATERI ALS The NRC staff will furnish LLNL with copies of,the following documents:
i.
The Probab.ilistic Risk Assessment (PRA) ii.
Tne plant's Final Safety Analysis Report (FSAR) lii.
The cuestion and answer volumes of the FSAR, if applicable iv.
A set of P&lD's 'f or' the plant Other material may be requested as required.
13.
PROPRIETARY DATA In the even; any proprietary information is submitted by the NRC in connection with tnir ;rc.'e:t, it must be specifically identifiec.
The University agrees to exercise its best efforts tc protect sucn proprietary cata.
F u r t ner;ho r e,
g.
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- - PNeject1
Title:
' Review of1Seabrook Nuclear Power Plant FIN /IE9 No. A0801
~
- Proba'bilistic-RiskJ Assessment B&R No. 20-19-40-41-5
~
Wang No. 0473x/6-ll-54
.nsa.
- limitation ~ shal1 not-be. imposed on the use of any information and data previously delivered ~to the University or _ government without limitations or previously published in any form as to be generally available.
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14.
PATENT STATUS Not: Appl.icable.
1 15.
TASK "X" ON-CALL ASSISTANCE Not. Applicable.
l 16 ~.. SUMMARIZE DELIVERABLES At the completion. of the project a final report will.be submitted summarizing
' all project work and results.
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ATTACHMENT AD
- 4 MONTHLY PROJECT FINAN:IAL STATUS-1..
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- ( Mo nth ).'
I eviei of-the Seabrook Prebabilistic' Risk Assessmen:
FIN-A0801
. A'.
PROJECT COSTS :
To.:ai ?rojected Funds Obligated Balance of_ Funds-Projedt Cost to Date by Fiscal Year-FYS4 FY85 S-K-
5 K
5 K'
5 K
L. ~ COST ANALYSIS
- B Period Cumulative
~
-Direct Lab Staff _Ef. fort - FTE
. Direct-Salardes.
K 5
K Materials &~ Services
~~ ^
ADP. Suppo. r_t, Subcontracts' 1 Travel. Expenses Indirect Labor Costs'
--Other:(TID).
~
General L. Administrative.
-Total Expenses 5
K 5
K
. Liens Total. Costs'and tieas~.
E K
( %) of fiindin g available
. ticte: These figures are for cost analysis only and may differ slightly from f inal sil. ling _ figures.
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Y ATTACHME!.TA;(continued)'
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.C; FEE-RECOVERY COST STATUS FIN:
A0801 TITLE:
Review of the Probabilistic Risk Assessment for the Seabrook Nuclear Power Plant PERIOD:
(Month)
Docket
~
Costs Facility Name Number Perico Cumulative Seabrook 50-443 5
K 5
K Common Costs 2/
50-900 Total Expenses 5
K 5
K
-.~
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2/ - Common costs are.for those costs incurred for effort such as preparatory or. start-up efforts to. interpret and reach agreemsnt -onLmethodology, approach,
.I acceptance criteria, ~ regulatory position, or TER f.ormat; efforts associated
-with the " lead piant" concept that might be involved during the first one or
. 'two plant' reviews as estimated by the contractor; meetings / discussion involving the above efforts to provide orientation, background knowledge, or
. guidance at:the beginning or during performance of work; and any technical effort applied to a category of f acilities, e.g., reactor analyses on all BWR facilities.. In the case of the latter example, these costs,should be prorated equally to only those facilities to which the effort applie's.
Development of' methodologies, criteria, or technical positions are not
. fee-recoverable. and, therefore, should not be categorized.as common costs.
Management and related support costs chargeab-le directly to the effort are not.
to be reported as common costs and should, therefore, be charged in proportion to the'other direct costs associated with those f acilities during the
. -reporting period.
NOTE:
Common costs are to be accrued on a monthly basis and prorated equally J amona the' f acilities identified under~ the effort at the end of each fiscal
~
. year cr -at the ccepletion of the effort, whicnever occurs first.
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