ML20205L245
| ML20205L245 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 03/25/1986 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19306D588 | List:
|
| References | |
| TASK-2.C.2, TASK-TM NUDOCS 8604080122 | |
| Download: ML20205L245 (11) | |
Text
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- UNITED STATES j
- 1 NUCLEAR REGULATORY COMMISSION
- i wasmNGTON, D, C. 20555 e
k p# MAa 2 51986 MEMORANDUM FOR: All NRR Employees FROM: Harold R. Denton, Director Office of Nuclear Reactor Regulation
SUBJECT:
NRROFFICELETTERNO.47(REY.1)DISSEMINATIONOF-PLANT SPECIFIC PRA RESULTS This Office Letter, which supersedes Office Letter No. 47 dated September 13, 1984, updates the procedures that must be followed for the timely dissemination of information from plant specific probabilistic risk assessments (PRAs). The information for dissemination includes the identification of potential safety problems that may require prompt regulatory action or sumaries of . insights from PRAs that can be factored into other ongoing programs or longer term follow-up actions. In the case of potential safety problems that may require prompt regulatory action, it is the responsibility of each member of the staff to bring such issues to management's attention so that action can be initiated in a timely manner. ~ Completion"of ongoing actijf ties in the development of the Severe Accident
- ' Policy,SafetyGoals,andInltegratedSafetyAssessmentProgrammayhavean impact on h'ow we process PRA;results in the future. However, in the interim, the Enclosure describes the procedure that ensures the timely dissemination of information fror. plant specific PRAs. This procedure is directed at 1) plant specific probabilistic assessments sponsored by RES that are not integrated into an ongoing regulatory activity such as resolution of Unresolved Safety Issues and 2) probabilistic assessments performed by licensees /appifcants and '
submitted to NRR. The procedure is to be implemented for all PRA reports received after September 1, 1984. The Director of RES has agreed to the enclosed procedure. Harol R. nton, irector Office of Nuclear Reactor Regulation
Enclosure:
Procedure For Disseminating Plant-Specific PRA Results cc: J. Sniezek s 4 e4 o?A L22 1 f J
l1 . ENCLOSURE PROCEDURE FOR DISSEMINATING PLANT-SPECIFIC PRA RESULTS
- 1. During the conduct of plant-specific probabilistic assessments sponsored by RES, if information is developed that indicates a high risk situation in the subject plant that might warrant prompt regulatory action, such as
.- the issuance of an order, IE Bulletin, etc., RES will promptly notify the Director, Division of Safety Review and Oversight (DSRO), NRR, by ,
i memorandum, and the Director, Division of Emergency Preparedness and Engineering Response, IE. The memorandum will contain as a min'imum: J Ic '
- a. A brief description of the perceived situation with simple I
supporting diagrams attached if appropriate.
- b. An explanation of why the situation is thought to require prompt l
regulatory action. i' L c. Identification of any items of poss'ible non-compliance that were l. noted incidental to the performance of the PRA. ?$ i The requirement for this memorandum does not preclude immediate informal discussions with DSR0 personnel on situations deemed high priority by RES
- - management. A meeting to brief interested NRR and IE personnel should be V
scheduled as soon as the evidence of unusually high risk surfaces. i~ j
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. - 7 The Safety Program Evaluation Branch (SPEB/DSR0) will be responsible for transmitting the memorandum over the signature of the Director, DSR0, to the appropriate NRR directors for consideration of potential board notification and/or other imediate action for the subject plant.
Potential board notification will be pro' essed c in.accordance with NRR Office Letter No. 19. Other imediate plant-specific actions will be developed and coordinated by the appropriate licensing division (PWR-A, PWR-B, BWR) in NRR as they would for Experience Memoranda processed under NRR Office letter No. 24.
- 2. Upon completion of plant-specific probabilistic assessments sponsored by RES, lucid, easily unde,rstood sumaries of the salient feat'ures of the plant safety or risk profiles will be delivered promptly to SPEB. The transmitted memorandum will' clearly state the RES position on the content. The sumary will contain dominant core-melt sequences, offsite consequences (ifavailable),significantinsights(plant-specificor generic including non-compliance issues), and any known modifications being considered by the owner affecting plant risk. RES is responsible for providing an assessment of each distinct generic issue that RES believes warrants further study and submitting it with the memorandum.
SPEB will be responsible for transmitting the sumary to the appropriate NRR divisions for consideration of plant-specific follow-on effort or development of new generic issues. The responsible NRR divisions will fill out the form appearing at the back of NRR Office Letter No. '40 for l each generic issue they believe warrants further study. Potential board l k
+ 3-notifications will be processed by the appropriate licensing division in accordance with NRR Office Letter No. 19. Within 20 work days following receipt of the summary, the cognizant licensing division will develop an action plan and schedule for plant-specific follow-up activities consistent with NRR Office Letter No. 24 and ensure that any resulting actions are processed in accordanu with NRC Manual Chapter 0514. Generic issues that may stem from the plant specific PRAs will-be prioritized for further staff consideration by the Safety Program Evaluation Branch (SPEB/DSRO) in accordance with NRR Office Letter No.
- 40. The Summaries should also be distributed to AE00, I&E, RRAB/DSRO, all licensing divisions and the Operating Reactor Assessment Staff /NRR for information. In 1!he longer term, a Research Information Letter t, .
should be prepared that, *among other things, flags, for particular attention insights of particular interest to specialists throughout those programatic offices dealing with reactor safety.
- 3. For.probabilistic risk assessments or analyses, of any scope, submitted to the staff by a utility, the recipient licensing division will transmit the study to the Branch Chief, Reliability and Risk Assessment Branch, DSRO within ten work days and 1,dentify a need for a. modest review.
4 The Reliability and Risk Assessment Branch (RRAB/DSRO) will perform an overview of the study within 60 work days to identify potentially significant safety issues (plant-specific or generic including non-compliance issues) and provide appropriate sumaries of the results e- - -
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4-(based on available documentation) to other NRC units. The overview will be coordinated with technical branches under the Assistant Director for
~ Technical Review in the appropriate licensing division and will sumarize the significant results of the study based on the available documentation; it will not evaluate the credibility of the study or resolve questions that may arise in the initial reading. Specifically RRAB will:
a) Issue an Experience Memorandum in accordance with NRR Office Letter No. 24 within 10 work days over the signature of the Director, DSRO, to the Director of the respective , licensing division if there is reason to suspect that prompt. regulatory action may be warranted
; and/or board notification in accordance with NRR Office Letter No. 19. This memorabdum will characterize the safety significance of the issue and explain the need for immediate action.
Inform the Director, Division of Emergency Preparedness and u Engineering Response, IE, within 10 work days over the signature of the' Director, DSRO, if there is a reason to believe that a prnmpt generic comunication in the form of an IE Bulletin or Information l: Notice may be necessary. For insights and information not requiring imediate action, b) Provide the Safety Program Evaluation Branch (SPEB/DSRO) any i l c r - _ . . , _ , - - - , ,. ,. _-- .- .%, .,_y.. -.-. . . , . , . , _ . . - - - - . . , _ _
y L, significant generic safety issues that may be identified in the inital. reading of the report.- This information will be processed in accordance with NRR Office Letter No. 40. c) Provide the project director for the subject plant (s) an Information
- - Memorandum that sumarizes the dominant accident sequences, core-melt likelihoods, offsite consequences (if available), and potentially significant safety issues. -RRAB will also make a determination about the potential usefulness of extending the overview to a modest review of the PRA (if not already initiated) based on consideration of the Policy and Planning Guidance on the Use of PRAs and on other regulatory activities such as Generic Is' sues studies, resolution of Unresolved Safety Issues, Severe AccidentProgram,a'ndTMIActionItem,II.C.2(continuationof IREP). This modest review would include flagging for particular attention insights of particular interest to specialists throughout the programatic offices dealing with reactor safety. Within 20 work days of receipt of the Information Memorandum, the cognizant licensing division will develop an action plan and schedule for plant-specific follow-on activities consistent with NRC Office letter No. 24 and ensure that any resulting actions are processed in accordance with NRC Manual Chapter 0514. This activity will include coordinating support from the appropriate technical groups and informing the subject utility.
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d) For information purposes, the overview of the PRA results will be transmitted to the appropriate Region. I&E, AE00, ORAS/NRR, other licensing divisions NRR, and RES. I 9
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ENCLOSURE TO ANSWER NO. 8 MEETINGS AND TELEPHONE CONVERSATIONS
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c I MEETINGS AND TELEPHONE CONVERSATIONS 9/?/86 Telephone conversation between V. Stello and W. Derrickson 10/14/86 Telephone conversation between V. Stello and W. Derrickson 10/78/86 Public meeting between Ohio Governor and V. Stello on emergency planning 10/29/86 Conference call between .NRC (Ragchi, Long, Lyon) and PSNH (Maidrand) 10/30/86 Meeting with BNL (Bari, Hofmayer, Pratt) in Novak's office. NRC participants included Novak, Rossi, Noonan, Long 11/3/86 Telephone conversation between V. Stello and R. Sweeney , 11/4/E6 Telephone conversation between V. Stello snd W. Derrickson 11/12/86 Public meeting between NRC and PSNH in Bethesda to discuss issues involved in BNL review of Seabrook EPZ study 11/17 R. Vollmer travel to Seabrook
-18/86 11/19/86 Telephone conversation between V. Stello and W. Derrickson ,
11/?1/86 Telephone conversation' between R. Vollmer and V. Noonan on responses to Markey cuestions l 11/26/86 Telephone conversation between R. Vollmer and T. Novak concerning l Markey Ticket #2346 12/4/86 Meeting of V. Stello, R. Vollmer and T. Novak on Seabrook emergency i planning end Congressional mail 12/4/86 Telephone conversation between V. Stello and W. Derrickson 12/12/86 Meeting in T. Novak's office to discuss staff actions following receipt of RNL draft report (some personal notes provided) l 12/16/86 Meeting of W. Darrickson and J. DeVincentis with V. Stello, R. Vollmer, and Christenbury 12/1f/86 Public meetir.g between PSNH and regional staff at Region I Office in King of Prussia, PA to discuss pending submission of petition to ASLB
- and pending organizational changes at NHY 12/24/86 Telephone conversation between V. Stello and W. Derrickson 1/7or8/87 Meeting with R. Vollmer. V. Noonan and D. Fioravante 1/8/87 Telephone conversation between D. Fioravante and T. Romano concerning conflict of interest Markey/NRR 07/03/87
1 1/8/87 Meeting of T. Novak, J. Scinto, E. Reis, E. Rossi, R. Perlis, D. Matthews, V. Nerses, S. Turk,'E. Jordan, and L. Soffer to discuss how to handle the the PSNH motion to the ASLB for a waiver of the 10 mile EPZ requirement 1/7-9/87 Meetings at BNL for S. Newberry, S. Long and W. Lyon to discuss comments with BNL investigators 1/12/87 Meeting in Rethesda for NRC staff members to discuss draft PP!L report with BNL personnel (Mere to participants provided) 1/12/87 Meeting in F. Kantor's office attended by D. Perrotti, P. Perlis, L. Soffer and D. Matthews 1/12 or Telephone conversation between NRC (S. Long and S. Newberry) and PSNH 13 (J. Moody and D. Maidrand) to discuss agenda for public meetino on 1/14/87
, 1/14/87 Public meeting with NRC, BNL and PSNH in Rethesda for PSNH to make commnents and discuss draft BNL report (Meetino summary provided) 1/16/87 Meeting J. Scinto, J. Murray, V. Stello, H. Denton, R. Vollmer, T. Novak, E. Jordan, V. Noonan (and others) to discuss staff . schedule 1/27/87 Meetings J. Scinto with R. Vollmer; J. Scinto with T. Novak and V. Noonan; J. Scinto with V. Stello to discuss staff response to ASLB i
l-l: I Markey/NRR 02/03/87
u _ . m__ ..a . _. . , ._ .__ . . _z _ . ___ ._ ._ EXTERNAL COMMUNICATION DOCUFENTS THAT ARE ON THE DOCKETA i October 8, 1986 Meeting Notice - Forthcoming Meeting with Public Service Company of New Hampshire to discuss certain technical aspects of the Seabrook Risk Management and Emergency - Planning Study. t October 15, 1986 - Seabrook Station Training Center. Seabrook, NH October 16 and 17, 1986 - Brookhaven National Laboratory, i Upton, NY i October 24, 1986 Meeting Notice - Forthcoming Meeting with Public Service Company of New Hampshire to Discuss Certain Technical Aspects of the Seabrook Risk Management and Emergency Planning Study - Bethesda, Md. November 7, 1986 Letter from Public Service Company of Now Hampshire concerning Response to Request for Additional Information 7 (PAIs). . November 17, 1986 Letter for Public Service Company of New Hampshire con-cerning Response to Request for Additional Information (RAfs). November 21, 1986 Meeting summary and Simulator Tour Held on October 15, 1986. l November 21, 1986 Letter from Public Service Company of New Hampshire .: concerning Emergency Planning Sensitivity Study. November 24, 19P6 Letter from Public Service Company of New Hampshire con-
; cerning Seabrook Station Probabilistic Safety Assesment (SSPSA) Update.
November 25, 1986 Summary of Meeting Held on November 12, 1986 to Discuss
- Seabrook Emergency Planning Sensitivity Study.
November 26, 1986 Summary of Meeting held on October 16 and 17, 1986 to ! Discuss Seabrook Emergency Planning Sensitivity Study. i December 12, 1986 Letter to Chairman Zech from S. A. Mack concerning allegations at Seabrook. December 23, 1986 Memorandum And Order issued by the Atomic Safety and Licensing Board. December 31, 1986 Meeting Notice for Forthcoming Seabrook Unit 2 Construction Permit Meeting to be held on January 6,1987 in Bethesda, Md. Markey/NRR 02/03/87 _ _ _ . . . _ _._ _ _.-._ _ .._ _ .._._ _ . ~ . _ ____ _ _ .__._ _ . _ _ ..-. _ _ _ -_ _
_ . _ _ . .z_ ... a _ _ ._.. . ._ _ _ _ ._,_ _,_,_ _ _ _ _ _ __ January 6 -1987 Meeting Notice for Forthcoming Meeting with Public Service Company of New Hampshire to Discuss Draft
' Brookhaven National Laboratory Report on Seabrook
' to be held on January 14, 1987 in Bethesda, Md.
January 7, 1987 Memorandum and Order (Supplement to Board Order of December 23,1986) issued by the Atomic Safety and Licensing Board. January 9, 1987 Order issued by the Commission. January 9, 1987 Memorandum and Order (Modifying Board Order of December
- 4, 1986) issued by the Atomic Safety and Licensing Board.
January 9, 1987 Letter to Public Service Company of New Hampshire concerninc
- summary of public meeting held in Region I office on December 16, 1986.
January 15, 1987 Peeting Summary for meeting held on January 14, 1987 with Public Service Company of New Hampshire to discuss the Brookhaven National Laboratory draft report on the Seabrook Energency Planning Sensitivity Study. i January 20, 1987 Letter from Public Service Company of New Harrpshire transmitting coments on the Brookhaven Nationat Laboratory draft report on the Seabrook Emergency Planning Sensitivity Study. ji 5 Parkey/NRR 07/03/87
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POLICY ISSUE (Infortnation) N ' For: The Comissioners 3 , William J. Dircks 2 : From: , Executive Director for Operations ,j - y
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Subject:
PERF0riANCE u OF PROBASILISTIr. RISX ASSESSMENT OR OTHOt TYPEC OF SPECIAL ANALYSES AT HIGH POPULATI0ti DENSITY SITES U6 provide the Comission with an initial list of sites
Purpose:
orioritized by population density; ano to infom the Comis-
~siori of the nature, extent, and priority of probabilistic risk analyses (PRA) or ather typet of special analyses that the staff is perfonning or planning for~all operatir.g reactors and applications for reactor licensees. This subject relates '
to the Statement of Interim Policy on Nuclear Power plant Acci-dent Considerations affimed on May 15,1980, and published in - the Federal Renister on June 13,1979 (45 FR 40101). It also - relates to the development of a transitien policy bet.: ten the 4
-old and new siting criteria requested oy the staff recuire-ments memorandum dated June 30,1980 (Enclosure 3). t:o action is requested of the Coemission at this time. Complete respense-to ite.n 2 of tne June 20 memorandam will be developed af ter review and disc.ussion of this paper and the Indian Point and Zion reviews witt the ACES.
Backorcund: During the Ccmission's discussion on April 16, 1980, conc'erhing SECY-80-131, " Accident Considerations Under NEPA*, there were 81 ~ expressions of ir. crest in plants in early stages of construction r j
>- that might be car.didates for special risk studies. This was in A addition to an err.andad treatment of accident considerations for J
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cases undergoing !! censing review and resulted in language being added to'the Sta s.ent of Interim Policy Sat the staff .shculd -
" Consider the. likelihood that substantive changes in plant design features which may compensate further for adverse site features-may be more easily incorporated in plants where construction has not yet progressed very far." ]
Contact:
M. L. Ernst, NRR 492-8016 S I 01 i n 3 C5"
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s-a . l i hl l s . I a . f. j During the discussion of SECY-80-131.on April 16, 1980 Limerick was identified for a special PRA study, following the Indian Point / Zion special risk studiss (NUREG-0660, TM! Action Plan, Task II.B.6). By letter cated May 6,1980, the staff requested
.. the Philadelphia Electric Company to perform within 120 days a t
j preliminary risk assessment of the Limerick facility. However.
- the results of this st'Jdy fil not be available from the licensee 3 until at least March. Consumers Power Co. has initiated a risk f study at Big Rock Point to support the utility's desire to not l backfit certain of the TM1 requirements because of a relatively
; low power level and low. site population density. In addition, i Crystal River was evaluated by the NRC under Phase I of the IREP
- (Interim Reliability Evaluation Program); and reviews of Calvert i Cliffs 1. Arkansas 1. Millstone 1, and Browns rerry 1 were
) recently initiated as Phase II of IREP.
! One purpose of the IREP studies is to develop a common methodol-ogy for the evaluation of all operating plants under NREP,(the-
; National Reliability Evaluation Program). The industry is also j interested in developing a comon methodology for the apolication of PRA, as is evidenced by the EPRI-sponsored review of Sequoyah i and the NSAC-sponse.d review of Oconee. Currently, arrangements
; have been made among t.'.a NRC, AIF, IEEE, and the ANS for the
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cooperative development of a common PRA. methodology that would
- draw upon all of the above experience. If this development is
.I successful, the results likely will be used for the NREP in lieu of tne methodology being developed under IREP.
In addition to IREP/NREP, other programs are currently underway or soon will be underway which will directly or indirectly impact the assessments of the safety of operating plants. These include the SEP program; implementation of the Bingham amendment (Sec-
', tion 110 of the NRC FY-80 Authorization Act (Public Law 96-295));
continued examination of the Indian Point petition, includng ! t special accident mitigation features; degraded core rulemaking; evaluation of shutdown heat removal capabilities; and a number of L other TMI related subjects and possible rulemakings. r j At an appropriate time a coordinated, disciplined approach to the p review of operating plants must be adopted that considers all of j k these programs in an integrated manner and achieves an effective l i and efficient utilization of the considerable resources required i by these activities. This point shculd be weighed in any consid-
. eration to do additional interim studies.
Discussion: In the development of an appropriate course of action the staff addressed two basic questions: , N l G 48
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- 1. If additional studies should be performed at sites having high population densities or other possible adverse charac-teristics, which sites should receive priority consideration?
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l 2. At such tites, what type cf additional studies would be
' appropriate?
o To arrive at decisions regarding these questions, the staff con- { sidered several criteria as described in detail in Enclosure 1.
., Basically, however, the staff's considerations centered upon the t likelihood that such interim studies would result in sufficient.
- near-tem improvement in the safety of operating reactors to warrant the cocmitment of priority resources.
$ Site Identification - The first question involves the identifica-tion of plants which might be candidates for any special studies.
i y The only adverse site feature considered by the staff in this ' / analysis was population density, since little infomation was readily available on some potential risk factors, such as liouid pathways and evacuation effectiveness. ' Criteria for external hazards are being considered in siting policy, to be followed by rulemaking; and site-specific meteorology likely would not sub-stantially affect the results of the analyses. There are many ways :. hat population density might be considered,
. such as number of people nearby, total population out to 10, 20, 30, or 50 miles, or some process that weights nearby people higher (because of less opportunity for atmospheric dispersion) but still
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considers exposures to persons more remote frem the site. Trial application of these alternatives by the staff indicated that the general results of the analyses (i.e., sites identified as higher risk) are not sensitive to the chosen analysis method to any significant extent. - I The method employing the Site Population Factor (SPF) was selected, since it was considered by the staff to provide a fairly reasonable numerical representation of radiological risk, assuming a given
- airborne release of radiological materials. The SPF methodology.-
weights population by proximity to the site in accordance with the typical average atmospheric dispersion factors at various distances from a sitet i.e., si:e specific meteorology is not used. The resultant SPF values were weighted by plant power level j (themal MW). A discussion of this process and the results of the screening for all reactor sites are provided in Enclosure 2. In general terms, most of the reactor sites in the United States have a power 1 ciel weighted SPF value that is within a factor of four of the median weighted SPF for all reactors, which is a value of 206 out to a
! radius of 30 miles (equivalent to a unifom population density of i
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, i 5 -less than 100 people per square mile out to 30 miles). However, 8 sites have "Above Average" power level weighted SPF values, i.e., values four to eight times greater than the median value; and three sites -- Indian Point, Limerick, and Zion -- have "Substantially Above Average" values (10-15 times higher than the median).
i-Special studies are already being performed on Indian Point, Limerick, and Zion. The staff believes that the 8 "Above Average" L reactor sites are the only additional ones that might be construed l to have " adverse" population densities, and of these, only Bailly and Millstone 3 are in 'he early stages of construction -- Bailly
!- being about it complete, and Millstone 3 about one-third complete.
The staff is aware that the criteria used to separate the reactors into groups are somewhat arbitrary. For example, Pilgrim 2 (which is awaiting a CP) is not listed in the "Above Average" group, since it has a power level weighted SPF value only somewhat greater than 3 times higher than the median. However, Bailly (which is in the "Above Average" group) has a power level weighted SPF vaTiie only 20t higher than Pilgrim 2. . r Generally speaking, the risk represented by population density is linearly proportior.al to the population. Therefore, unless there is some other associated factor, such as evacuation efficiency,
. L-i that is nonlinear with increasing population, determi-stion of a
; point at which an increasing population suddenly becomes an
" adverse" site feature is necessarily arbitrary. Nevertheless, the staff has concluded that the categories chosen are reasonable, since: ,
- 1. Choice of a weighted SPF smaller than a factor of four above
- that of the median site as a demarcation would result in very little difference in risk between the median site and the
" adverse" sites.
- 2. The " median" site, upcn which the analyses wue based, has almost a factor of two less population than the " average"
; site, thus the chosen demarcation point is only about a
- factor of two higher than the average site.
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- 3. The errors in performance of risk analyses are of the same
} order of magnitude or perhaps greater than the factor of four; p
therefore, the significance of the conclusions of any compre-h, hensive risk analyses perfomed at these " adverse" sites would be somewhat lost in the uncertainties of the analyses, if a lower point of demarcation were chosen. l Ed W -.
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- 4. All sites (except Haddam Neck) that have a power level I
I weighted SPF within a factor of four of the median site
' also have less than an average cf 500 persons per square mile at a 30-mile radius, and Haddam Neck has less than 500 persons per square mile at all distances up to and including 20 miles.1 .To illustrate the lack of a clear i demarcation line for identifying sites with " adverse"
> population characteristics, three of the 11 sites that
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have a weighted SPF greater than a factor of four above the mediar. (Seabrook, Three Mile Island, and Waterford)
! have a population density less than 500 persons per square j mile at any distance up to at least 50 miles.1 h The staff believes that the method chosen to identify sites with
} potentially adverse population characteristics is reasonable.
However, one must not conclude from this analysis that the sites { so identified actually are the ones that contribute predominately to the overall risk from nuclear power reactors. Factors other than population density have a major influence on risk. No 3, consideration was given to plant design in thie paper; and varia-c tions in plant design have been demonstrated to have a majcr 5 - influence on risk -- potentially much greater than population i differences. This fact has been spotlighted in the recent Indian Point and Crystal River risk studies. Ty;:es of' Ris*= Studin - The second question involves the type of I, special risk stuales that might be worthwhile to conduct, if it . is determined that special studies are warranted because of : adverse site characteristics. Pcssible kinds of studies are: f, i
- 1. Complete risk studies.using WASH-1400 methocology. l l 2. Truncated WASH-1400-type evaluations that analyze only selected potential dominant accident sequences, based on judgment supported by experience.
; 3. IREP-type analyses, which are principally targeted at core melt sequences (no in-depth review of containment or of releaseconsequences).
4 CRAC-type analyses, which are aimed at consequences only l. (i.e., assume given releases and pathways, but use regional l. or site-specific meteorology).
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- 5. Plant-specific analyses of speciai mitigation features (such 3 l
as filtered vented containment), which muld require a i l reasonable knowledge of accident sequences and potential i release pathways to provide some judgment as to overall risk l } reduction. f h
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t l All of the above possibilities involve PRA to some degree, since ! PRA (at least as the term is connonly used) can involve varying . degrees and sophistications of probabilistic analyses of accident : sequences, mitigation, and/or exposure of people. A discussion i i of some of the considerations involved in conducting PRA h
! found in Enclosure 1.
i The staff considered the above types of analyses and has the following technical conclusic,n.- E
- 1. CRAC-type of analyses using regional or site-specific t meteorology and denography are a useful way of estimating .
the risk of accidents, provided that accident sequences and i release pathways can reasonably be assumed. If such cannut :
} be assumed, then these analyses still give interesting -
- e. relative information to permit comparisons of one site to i another, assuming the same reactor on each site. While such 5-
- relative analyses might appear to be useful, .they would not F yield good plant / site-specific risk profiles; and the use of i
- SPF values (which do not use site-specific meteorology) likely [
. would display the relative comparisons of sites sufficiently 2 i accurately to detannine whether population represents an !
i " adverse feature" without using the sophistication of the -
; CRAC. code. However, running CRAC calculations would perhaps [;
> improve the public's understanding of site-specific . radio-I logical risks at a nominal expenditure of manpower. [
- 2. Design-specific risk assessments (IREP, or using WASH-1400 I.
methodology) are very useful ways to estimate risk; however, j { they are strongly dependent on a full knowledge of final & plant design, operating experience, and operating procedures. f Perfonnance of such analyses without full knowledge cf the R above would be so subject to error that they likely would ! not warrant the expenditure of resources. Also, the results [ of such analyses could be misleading, and the studies likely h would have to be repeated at some later date when such c'etailed E L. , inforration becomes available. [ i The IF.EP analyses provide a good core-melt modeling of plants, ! which would give a useful perspective on plant safety. Wnile contairinent and other release mitigation features also are [= 4 t important to safety and should not be neglected, such full f risk studies would be much more demanding on resources than [ 6 the IREP-type studies. E h
- 3. Plant-specific anal g vented cor.tainment)yses are usefulofand mitigation are beingfeatures (such as filtered done for Indian p Point and Zion. These studies will be extended to other plants
- k under a program to be developed after the results of the Z/IP L
s 9 4 M
-7 f studies are available. While knowledge of accident sequences ; and release pathways are necessary to evaluate fully the ; safety benefits of such fe:tures, such knowledge can be some- } what more deteministic in nature than would be required i for the more probabilistic-criented risk studies. Considera-i tion of requiring the use of suci features on reactors would i
likely be dependent on the specific: of reactor design, acci-dent sequences, and release pathways. Whether the site has I an " adverse" population density would probably be of somewhat lesser consideration in any technical decision to implement ~ f' such a requirement. f Current and Fu'.ure Staff Actions - The course of action for per-p formance of risk studies currently being followed by the staff i for sites with " adverse" population features as well as for all
; other plants is sumarized below:
- 1. For all future enviremental statements, and for those currently in preparation, there will be an expanded discussion of accident risk similar to that provided in the statement for the Sumer c plant.
t 2. CRAC studies for all reactor sites are underway by the staff i in connection with the technical basis for rulemaking on i Part 100. The results of these studics, which use recional or site-specific meteorology, will be compared to SPF data, whi:h use a single, typical or average meteorology.
- 3. Using population density (SPF calculations) weighted by power level as a way of identifying sites with " adverse" population features, Indian Point, Limerick, and Zion are identified in the "Substantially Above Average" category. Risk studies using WASH-1400 methodology are being perfomed by the ifcensees -
for all of these plants, as are studies of special mit.3ation features for Indian Point and Zion. The results of these
, studies will be reviewed by the staff. No other actions are being taken at this time; but special mitigation features will ba studied for Limerick in the future, subsequent to completion of the Z/IP studics, as part of the overall long-range program.
. 4. Using population density (SPF calculations) weighted by power-3 level as a crude way of identifying additional sites with
, hpotentially adverse features, eight sites (Bailly, Beaver e
,hg* '. Valley, Femi, Millstone, Seabrook, Shoreham, Three MileIsland, and Waterfor category. Of these, Bailly and Millstone Unit 3 are the only
; ones in the early stages of construction.
l (
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For Millstone the staff intends to: identify and advise the licensee of special design considerations to be cognizant of in the final design of the plants based on the result 1 of the PRA studies performed on a risk study under NREP (plantsReliability the National with similar systems;' perform Evaluation Program) on a relatively high priority basis;-and perform - appropriate long-range studies of special mitigation features. dependent on the results of the Indian Point and Zion stud.ies. The NREP reviews Itkely would be performed prior to the issuance of the OL, provided that the methodology for such reviews is developed in a timely manner. j The special problems associated with Bailly will be addressed within the context of: outstanding petitions filed under 10 CFR a 2.206; and the staff reconsendations on applying NTCP requirements (NUREG 0718) to prevent CP holders for which onsite construction has been minimal. }
- 5. For the remaining six sites in the "Above Average" category, the licensees will be advist' of special design considera-e tions, and the NREP and app- wiate studies of spec.ial mitiga-tion features will be perfon4d in a manner sim#.1ar to Bailly and Millstone Unit 3. However. any OL issuance decisions likely would not be contingent on the performance of such studies, particularly for the NREP review.
. 6. For all other plants, activities similar to those identified in S., above, will appropriatily be undertaken. However, the details of tne NREr and special mitigation features pro-grams have not been developed, and the priority given to such reviews will be based on a number of factors other than population density, such as licensing status, age of the plant,
, and the initial results of the program to identify deviations
- from the regulations of particular significance (the "81ngham" amendment).
- 7. Emergency response studies are currently underway to determine the effectiveness of evacuation, sheltering, and other protec- j tive measures at all reactor sites. These studies are related
- to requirenents associated with the new mergency planning rule q and for the rulemaking on Part 100. Sensitivity studies (or h thtse efforts will provide the relationships to assess the influence of energency response on risks from high population j
H sites. f i Backfit Determinations - If, as a result of any of the studies identified above, accitional measures appear to be required to j reduce risk, the applicant will be requested to propose piant . d. i %hai, would be provided would be a fairly complete description of the potentially dominant accident sequences, the safety systems , used to mitigate the accidents, and identification of the inter-4 dependencies between major safety systems thanselves and their support systems. These functional interdependencies are critical paths whose failure could result in significant accidents. i
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modifications, including analyses to demonstrate the degree of safety improvement. In the near tenn, calculational refinements to " fine tune" the PRA analyses will be discouraged, since the basic intent'(until development of a comon PRA methodology for the NREP) should be to compare systems and no.t to try to demon-strate a new level of absolute risk for nuclear power plants. The time frame for implementing any needed proposed modifications would depend on the magnitude of the deviations from the currently
" accepted" risks and input from parallel NRC actions on Class 9 events, severe core damage protection, siting, emergency prepared-ness, ATWS, the systematic evaluation program, and' establishment of safety goals and criteria. If a plant already satisfies our current design criteria with respect to redundancy and diversity and the TMI requirements, experience demonstrates that the estimated the range of probab{lity of severe 10- to 10-5/RY, core damage in which case wewill likelythat believe be in any action to require modifications should await the consideration of other reviews and studies so that any backfits would be appro-priately coordinated with other possible requirements. Any plants that are found to have a higher probability for severe core damage, or which substantially exceed other currently
" accepted" or "nomal" levels of risk, will be reexamined against our current deteministic criteria and required to correct the
~
deficiencies or otherwise reduce the risk in a reasonably expeditious manner. sd W j Willhd J. Dircks l Executive 91 rector 1 for Operations 1
Enclosures:
- 1. Otscussion of Decisien j Criteria .
l 2.' Prioritization of Sites ) with Regard to Popula-tion Density
; 3. Memo from S. Chfik to W. Ofrcks dtd 6/30/80, "SECY-80-153, Advance Notice of Proposed Rulemaking on Reactor Si ting" DISTRIBUTION Comissioners Comission Staff Offices Exec Otr for Operations ACRS ASLBP ASLAF Secretariat t _
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Enclosure 1 Discussion of Decision Criteria Various NRC staff proposals recently have recocrnended a number of approaches for determining candidate sites for possible future risk studies. Selection criteria have included Class 9 considerations, high population densities, unique emergency preparedness cor siderations, and/or power level considerations. Enclosure 2 provides a proposal for a priority listing for any such near-tem studies. However, an even more basic question is whether any more ad hoc risk , studies need or should be performed, or whether future plant reviews utilizing
- PRA should await the development of the National Reliability Evaluation Program (NREP), utilizing a more disciplined and agreed-upon methodology. This enclo-sure focuses on the considerations that affected staff decisions regarding the performance of any new near-tem PRA studies at high population density sites.
The decision criteria'used by the staff were: , Confonnance with regulations
/ Improve NRC decision making on licensing actions 7.Enhancepublicunderstanding Effective use of pRA resources l 1. Confomance with Reculations - The first criterion is that the program cust .
I conform to NRC's regulations; recognizing, however, tr.at policy and regula-tions can be changed. The NRC's regulations address tre use cf PRA in tr,e Comission's statement of interim policy and guidance (not a " regulation". per se) concerning its position on the disclosure of accidene risks under { NEPA.1 i / A 14,5 FR 40101, Junes13, 1980.
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4 This interim policy states that events or accident sequences that lead to releases shall include, but not be Ifmited to, those that can reasonably be expected to occur. Such analyses shall incicJe in-plant and external causes of accidents that can result in inadequate cooling of reactor fuel . and melting of the reactor core. The enviromental cor. sequences of releases (resulting from such accidents) :vhose probability of occurrence has been estimated shall also be discussed in probabilistic terms. Health and safety risks to individuals and population groups shall he discussed in a manner that fairly reflects the current state of knowledge regarding sue.h risks, including associated uncertainties. Among other things, the Statement of Interim Policy states: .
"It is the intent 'of the Connission in issuing shis Statement of Interim Policy that the staff will initiate treatments of accident
- considerations in accordance with the foregoing guidance, in its ongoing NEPA reviews, f.e., for any proceeding at'a licensing stage where a Final Enviremental Impact Statement has not yet been issued. These new treatments, which will take into account significant site- and plant-specific features, will result in more detailed discussions of accident risks than in previous environmental statements, particularly for those related to.conven- i tional light water plants at land-basert sites. It is expected that these revised treatments will lead to conclusions regarding the environmental risks of accidents similar to those that would be reacheo by a continuation of current practices."
"However, it is alsn t'he intent of the Ccenission that the staff -
take steps to identify additional cases that might warrant early
' consideration of either additional features or other actions which-would prevent or mitigate the consequences of serious accidents.
Cases for such consideration are those for which a Final Environ-mental Statement has already been issued at the Construction Permit stage but for which the Operating License review stage has not yet been reached. In carrying out this directive, the staff should consider relevant site features, including population density, associated with accident risk in comparison to such 'eatures at presently operating plants. Staff should also consider the likeli-hood that substantive changes in plant design features which may compensate further for adverse site features may be more easily incorporated in plants where construction has not yet progressed very tar." 1 o. i hw
. - . . , ..e ~
..u r ^. M.i,- K s* *, Q.% 2,ptw*fxff$f Conclusions that may be drawn from the above statements r,egardirg the present regulations are:
. The Interim Policy Statement is silerit en the parformance of PRA for operating reactors.
. For reactors cu.eently under construction but not in the Ot. review stage, PRA studies might, be wareanted. Such studics, if perfomed, should consider site features compared to suen features at presently ,
operating reactors, and snould consider substantive changet in plant design features which might compensate further for (any) adverse site features.
. The Interim Policy Statement does not address the type of studies that should be perfomed. ,
l
- 2. Imorove t!RC Decision Making - The second criterio.1 relates to irnproving
- the ability of the itRC to make reasonable piant/ site decisim regarding i
l substantive changes in plant design features. The first aspect of this criterion considered by the staff concerns verifiability, and the seccnd deals with the treament of uncertainty and ...ter aspects of perfaming i PRA which affect the evaluation of the results. As will be seen in subsequent discussions, tne verification process would be substantially easier, if the FRA methodology used for these risk studies were reasonably consistent wit the techniques used in the Reactor Safety S tudy. This would enhance verifiability by pemitting a resonable compar-
.on of the resultant consecuence curves to ot'tain a relative risk compa'r.
Hon to typical reactor / site combinations. Obviously, modification of the I
4 methodology (models, data, assumptions) can change the results for both the subject plant and the reference plant. However, the staff would be prir.:fpally interested in changes in specific plant features that would provide substantive safety improvements (i.e., lower risk) at sites with adverse characteristics, and not " improvements" based on refinements in analyses. If the analyses are not reasonably constrained to the WASH-1400 data base
, and techniques, these studies (which should basically be comparative in nature) would begin to take on the character of absolute risk studies, in addition to making verifiability more difficult. The net result of such studies wuld be the e;:penditure of additional NRC resources to verify the analyses, plus increased controversy over the assumptions, data base, and ,
methodology at the exp'ense of a clear focus on design improvements (com-pared witn a " typical" reactor) to compensate for an' adverse site charac-teristic. As a matter' of interest, this is the na'ture of the Indian Point, s Ziori, ar.d t.imerick studies being performed by the industry. A great deal of eUo.t is going into phenomenological studies and the refinement of data bases and assumptions with an eye toward demonstrating that the overall risk from nuclear power plants is substantially lower than the estimates in WASH-1400. p The final aspect of the second criterion relates to the ability of the NRC l i to utilize uch PRA as a tool to assist decision making. The use of risk studies ta armine whether additional design features should be required 1 does nue .an that the staff is alsandoning established requirements. For 9 3 i
I instance, risk analyses performed for power plants at low population sites should not be used to pemit single train safety systems, even though the studies might show that the risks would be below those of the reference plant. Thus, at the present time the results of any such studies would I more likely be used to add new requirements, not delete present ones. t [ Since only high population density sites presently are being considered. the implicit concern is one of population exposure as well as individual
! exposure. Therefore, the results of these studies would likely be pre-h sented as complementary cumulative distribution functions (CC0F) which show the probability of exceeding consequences (early fatalities, latent cancers, property damage) of a given magnitude as a result of radioactive 6
releases. The use of severe core damage probabilities alone likely would be considered insufficient for this assessment, since such analyses would not include, for example, population densities and mitigation measures
- ~
such as containment. Thus, core melt calculations alone would not pemit a direct evaluation of risk to the population. In the absence of any previously sanctioned acceptance criteria for societal risk, it is very likely that the results of these risk studies would be com-pared with the CCDF's presented in the Reactor Safety Study (WASH-1400) for reference light water reactors to decide whether further action is required to reduce risk. However, such cmparative analyses using truncated risk studies would have several weaknesses, as discussed below: a. Comoleteness - Because of resource constraints, the studies envisioned Itkely would not treat external events, sabotage, fire, etc., and
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. s.
i likely would not include a full scope of detailed fault tree analyses; l 1.e., they likely would be characterized as truncated studies. These omissions might not be serious, since an absolute risk determination l would not be the objective -- only a comparative systems evaluation. i However, the limited analyses would introduce some uncertainty, since .i J l such studies would not include a complete risk modeling of all plant systems, subsystems, and components. This weakness should be limited by careful attention to inclusion of all the major unique plant differ-ences between the subject plant and the reference plant, and peer judgment when the studies are reviewed.
- b. Uncertainty - The results would represent probabilities at about the 50 percent confidence level (ignoring a certain lack of statistical ,
rigor in the analyses and certain bounding assumptions that are made) and would not address the potential that the uncertainties in the analyses could be significantly different between the plants compared. l This could distort the results compared with what might be achieved at higher confidence levels. Dr. Okrent (Science, April 25,1980)has suggested that the risk be assessed at a high level of confidence to ensure better treatment of the data. Mcwever, because of the likely j. h limited nature of any new studies, the substantial effort required to L f address propagating errors and to carry uncertainties forward, the l H effect of phenomenological assumptions, and the fact that many of the data are point estimate values or judgments, it is not believed that l detailed uncertainty analyses would be justified. The problems asso-ciated with uncertainty could be alleviated to scme degree by careful w - -ev-,--- -.,.-+ - , _ _._.O ,
mummummum
. .*e attention to the performance of sensitivity studies and by using the.
same methodology, assumptions, ard data base as is used in WASH-1400, with significant deviations clearly defined and justified. '
- c. Dependence on Final Design - Systems level PRA is not useful unless it
} ts perfomed on the as-built or final design of the plant. While generic analyses such as WASH-1400 (which uses two plant: as representa-tive prototypes) give a useful perspective regarding a particular design, small differences from the reference design in an as-built plant can have substantial impact on,the dominant accident sequences. For this reason, the performance of systems PRA on plants in the CP review stage or in the early stages of construction would have significantly fewer
' benefits than PRA performed during the OL stage or for an operating reactor. In a similar vein, a substantial contribution to risk origi-nates frcm th'e specific operating and maintenance procedures that are, in effect at the plant.' Since these will not be final until shortly before operation, the maximum benefit of performing a PRA would not be achieved until the plant is near operational.
l d. Realism - Past experience has shown that a considerable amount of judgment is used in the risk studies. These include human error, J 1, treatment of comon cause failures, physical phenomena related to contaircent failure, and equignent operability under adverse condi-p tions. In these areas, the staff's evaluations would likely tend to be conservative, particularly for these truncated studies. Such L L conservatism could result in the requirement of more marginal safety L improvements.
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- 3. Enhancement of Public Understanding - Whether or not any of the possible . !
programs to perform additional PRA studies would substantially enhance - public understanding is debatable. However, one certainly would not want , to embark on a program.that wuld not be understandable to the public, unless such a program likely wuld result in the identification and imple-mentation of significant safety improvements. On the one hand, the public is aware of the PRA tool, especially since pubitcation of the various TMI reports and the recent considerations regarding the interim acceptability of Indian Point. In this regard, it wuld be difficult for the public to understand any decision not to utilize PRA as an aid to making decisions in controversial areas, such as those associated with sites of high population density. On the other hand, there are such substantial uncertainties in the perfor-mance of PRA that any result (regardless of how reali, tic or hcw conserva-tive) will be very controversial. Therefore, in the face of such contro-versy the public will be unsure as to the merits of the results, which w uld be a disservice to public understanding. Therefore, it wuld appear that to optimize enhancement of public under. standing, the scope and nature of any study that is performed should be such as to minimize the effects of uncertainty in the results.
- 4. Effective Use of PRA Resources - As noted in the Background set, tion of this Comission Paper, there are a substantial number of PRA's underway at the present tiine by both the NRC and the industry. These analyses
;"~" _ _ _ . _
~ - - . . . . . . . _..._._.o_,_____ _ _ _
l 9 t are being perfomed principally to evaluate a the accept bil design and siting of specific plants ity of systems methodology. , and to develop a standardized PRA There are not many trained people in the field of PRA
-- indeed, half of NRR's professionals currently assigned are in to PRA
' .ed full-time in IREP, and the others are essentially fully e-comitt and Indian Point rev'fews. to the Limerick, Zion, fact, the utilities involved in the IREP prograTh t
of this program to train their people in PRA. m are taking full opport Therefore, we are essentially at the point where th new studies would begin to seriously dilute ongoing PRAE the development of a common methodology efforts, including I preclude the initiation of a new PRA effortWhile this fact shou effort would.likely have a safety payoff that would wa, one m dilution of the longer-tem program, rrant the further i
. Additionally, one must be careful that the usef l played.
u ness of PRA is not over-There is a danger that o, if ad ,h,cc studies are req i u red by the NRC but are not carefully controlled and monitored quently be questioned and perhaps found to be seri, the results ously flawed. If this were to happen on a few controversial cases , stantial black eye and might be perceived y useful as a to be not ver, th decisicnal tool. be dismissed out-of-hand for future use, only e of a premature and 1 l \
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- i incautious use. Again, the potential safety payoff of any early use of PRA in controversial cases should be sufficiently great to warrant taking the risk of perhaps perfonning seriously flawed studies.
l . The scope of all of these studies vary considerably, as can be seen from Attachnent 1. The IREP studies are the least manpower intensive, but they are limited essentially to the evaluation of severe core damage accident i i sequences. More detailed studies that include accident consequences and external events (such as seismic) require 5 to 10 times more manpower than the Phase !! 1 REP. The performance of the so-called short tenn Zion and Indian Point studies (not listed in Attachnent 1) required less manpower
- than the IREP studies, since they did not modal the entire plant but only focused on unique differences in plant design compared with the Surry Analysis in k' ASH-1400. '
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9
PROHARILISTIC RISK ASSESSIIENT . r
. STillllES CURRENTLY llNI)EinfAY '*
I g MSSS/ Cit 4TAINilENT PARTIES 1.EVEI, OF Pl. ANT i
'IT PE lHV01.VEI) EST1HATEI)
I . EFrol'T (HYl _COMPl.ETIoll DATE SCOPE OF EFFORT 4 Ocunce B&W/llry P -p 3 g NSAC 15-20 Ailg 1981 X X
) 8 Utilities ,
X Conseiltants Sestiinyals W/lce Condesiser EPHI 12-16 Utility Planse 1 - X X X i Dec 1980 { connnitants Phase 11 - l Dec 1981 } I.linerick , CE/flark II ? Utility 7-8 Sept 1980 X X HSSS Venulor - Connial tan t n - Zion /leidinnIPt W/ Dry lit t 11t t es " 30 Oct 1980 llSSS Vendor (3 innits X X X Consiit taint s at 2 alt.en) Crystal River R&W/ pry HRC-RES 6-7 (liil L lsil INEl' Sept 1980 X
+
58 nely) thissial t nesi a Calvert Cliffa I CElitry HRC-RES Arlinnaan i It&W/Isry 3-5 Jiine 1981 X tilllutune I Coinisitanta (per pinut) CE/iLerk I Browns Ferry 1 CE/flark I
,(INEP Follow-sep .
Stesdies) r)
- P -
Aci:Ident Prot.ala!!It les - 3 C - Si Accident Conseelnences *
.4 A -
Plant Availatallity E - Externial Eventa i N _ l
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Enclosure 2 prioritization of Sites with Regard to Population Density
- 1. Introduction In comparing and evaluating the population around nuclear power reactor sites, the staff has long recognized that the population characteristics of a site, that is, its density and distribution, are a relatively crude measure of the consequences associated with the accidental release of ra~dioactivity. The residual risk from an accident would depend not only upon the population den-sity of the site, but also upon many other factors, such as reactor design, onsite and offsite management and technical suppc. t resources, external hazards, liquid pathway considerations, meteorological conditions at the time of the accident; and effectiveness and nature of public protective actions taken.
In addition, the risk is not uniform for all merabers of the population regard-less of distance from the site, but would be higher for those persons relatively close to the site, and would generally decrease with distance away frem the ' site. An analysis has been carried out to obtain a first-order prioritization of sites based upon population density and distribution. The discussion that follows outilnes the rationale and methodology used and gives the results of this analysis. l l 2. Methodology In carrying out this analysis, the following assumption and cathodology were used: i
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2-(a) All sites where a reactor was either in operation, under construction, or where a construction permit was presently under active review were evaluated. This involved a total of 93 sites. . e (b) The population data used were taken from NUREG-0348, based on the 1970 census. The population data for the Fenni' site as reported in NUREG-0348 are in error and were corrected for this analysis by a special computer run of the 1970 census tare. (c) Although it is well-known that individuals closer to the reactor are at a higher level of risk, given an accident, than those more renotely located, the precise quantification of the variation of risk with distance is still somewhat uncertain; For the purpose of thi,s analysis, the distance weighting given by the $tte Population Factors (SpF), as given in WASH-1235, were used. Further, population beyond 30 miles was. neglected, beause the consequences at distances within 30 miles were considered to domir.ste any considerations of overall societal impact, and beyond 30 miles the potential population exposure differences from site to site become less snarp. Preliminary analyses carried out by the staff have indicated that
, somewhat differing weighting schemes, or the factoring in of copulation -
out to 50 miles, does not change the resulting prioritization 'of sites to a significant degree. (d) The power level of the largest reactor at the site was multiplied by the SpF value to account, in a firct-order way, for the variation of reactor fission product inventory from site to site. Only one reactor at a site was considered, even where multiple reactors exist or are contemplated, u i I
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3-because the probability of an accident involving more than one reactor simultaneously was considered negligible. Although it can be argued that the population around a 4 reactor site is at a higher level of, risk than those around a single reactor site, the prioritization of sites is intended to give a measure of the relative consecuences, given that an-accident has occurred. The number of reactors at a site presumably effects only the probability of an accident. Also, it could be argued th6t a multi-reactor site would have some attributes that would reduce risk, compared to a single-reactor site, because of greater management and technical resources that can be applied to reducing either the likeli-hood or. consequences of an accident. Using the above methodology, the reactor power level times the SPF value was calculated and tabulated for ' each of the 93 sites considered. The results are discussed below.
- 3. Results The reactor power level times SPF (P x SPF) was calculated for each of the 93 sites. The resulting values ranged from a high value of 2983 to a low value of 6. The median value is 205; and the median site has a population of 1.ess than 100 persons per square mile, which is almost a factor of two less than the population of the average site. The sites are not Ifsted in numerical order, since this would imply a greater degree of precision than is warranted by the uncertainties in the analysis. Also, as pointed out previously, the residual risk at a particular site cannot be measured in terms of consequences alone, since plant design and other factors are important contributors to risk. Therefore, we decided to place each site I
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{ I l into one of five groups or categories. The variation within a given group was selected to be sufficiently small so that each site within that group is considered to have about the same ranking. In selecting the groups we decided to use the median value and factor of two varia-
' tion about the median to demarcate the " average" group boundaries. The other groups were chosen as indicated below.
Group No. Title tE I Below Average PXSPF less than one-half the median value (PXSPF<100)
!! Average PXSPF.between one-half and
- twice the median value *
.. (PXSPFfrom100to400) 111 Slightly Above Average
.PXSPF between twice and four' times the median value (PXSPFfrom400to800)
IV Above Average PXSPF between four and eight times the median (PXSPF from 800 to 1600) V Substantially Above PXSPF greater than eight times Average the median (PXSPF > 1600) lif thin each group the sites have been listed in alphabetical order, as shown in the following tables. l l Group V - Substantially Above Averace
- 1. Indian Point
- 2. Limerick
- 3. Zion 6
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Group IV - Above Average 1
- 1. Bailly S. Seabrook
- 2. Beaver Valley 6. Storehar *
- 3. Femi 7.
- 4. Millstone Three Mile Island *
- 8. Veterford Gmup III - Slightly Above A erage
- 1. Byron
- 2. 11. Peacn Bottom Catawba
- 3. Cook
- 12. Perkins
- 4.
- 13. Pilgrim Cherokee 14. Perry
- 5. Erie 15. Salem
- 6. Forked River 16. Sequoyah
- 7. Haddam Neck 17. Susquehanna
- 8. Hope Creek
- 9. 10. Rancho Seco McGuire 19. Turkey Point
- 10. Midland 20. ifmer Group II - Average
- 1. Arkansas 21. Palisades
- 2. Bellefonte 22. Phipps Bend
- 3. Black Fox ;
4 Braidwood
- 23. Prairie Island
- 24. Quad Cities l S. Browns Ferry 25. River Bend l 6. Calvert Cliffs 26. Robinson
- 7. Clinton 8.
- 27. San Onofre Brunswick 28. Shearon Harris
- 9. Davis-Besse 29. Sumer
- 10. Duane Arnold 30. Surry
- 11. Fort Calhoun 31. St. Lucie
- 12. Fit 1 patrick 32. Skagit
- 13. Ginna 33. Trojan
- 14. Hartsville 34. Vogtle
- 15. LaSalle 35. 'datts Car
- 15. Maine Yankee 36. ~47PSS 3/5
- 17. Marble Hill 37. Yemont Yankee
- 18. Nine Mile Point 38. .Monticello
- 19. Oconee 39 Yellow Creek 20 Oyster Creek 1Ba111y and Millstone Unit 3 are the only plants in Group IV that are in the early stages of construction.
Table 1 Some Risk Insights vs. Design ,, ,
* . (
PCM Estimated P >1 Fatal /Pcm Probability
- Estimated Probability of of Severe core Damage Experiencing 1 or more Acute Fatalities *** . l.
Type of ,.Given Severe Core Damage ; Reactor Name Design per Reactor-Year l Surry 3-Loop PWR (Subatmospheric) $6x10 5 '
+1/10 i ~
+1/3 "CRAC" !
- BWR #4 (MARK I) +3x10 5 Peach Botton
+4x10 5 ' +1/25 e I.P. Site Sequoyah 4-LoopPWit(IceCondenser)
Indian Point 4-Loop PWR (Dry)
+3x10 5
- - To +1/100 f Oconee 2-Loop PWR (Dry) +2x10 4 (+1/4) '{~~~
~2x10 4
(~1/4) Estimated Calvert Cliffs 2-Loop PWR (Dry) .
+3x10 4
- +1/3 ,), ,
Crystal River-3 2-Loop PWR (Dry) 4-Loop PWR (Dry) +4x10.s 1/10 to 1/100 '! Biblis** -
"These are core melt estimates which derive from variou PRA studies and study groups and reflect median or point values. Large uncertainties can exist around such estimates.
**PRA results from fred. Republic of Germany
*** Acute fatalities used as risk Indicator here because of largest variability exhibited due to designs and siting.
- These factors hold for approximately
- 100 or less acute fatalities. !
9
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1 i Group I - Below iverage
- 1. Allens Creek 2, Big Rock Point 13. Kewaunee
- 3. Callaway 14. Lacrosse 4 15. North Anna
- Comanche Peak
- 5. Cooper 16. Palo Verde
- 6. Crystal River 17. Pebble Springs
- 7. Diablo Canyon 18. Point Beach
- 8. Dresden 19. South Texas
. 9. Farley 20. WPPSS 2
- 10. Ft. St. Vrain 21. WPPSS 1/4
- 11. Grand Gulf 22. Wolf Creek
- 12. Hatch 23. Yankee Rowe s
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~ ,0:,te:%i FOR:
.. William J. Dircks, Acting -
*..n t it.llo Executive Director for 0 rations-
~
f, . .ller po;; Samuel J. Chilk, Secretag ,
- l2 t
s.30ECT*-
~~ SECY-80-153 - ADVA';CE !!0T'is#; 0F FR070 SED RULCMI';G M CALE : CAR ITE!4)
REACTOR SITittG (C0tiSEt E
! In association with the advance notice of' proposed rule .aking on rc..ctor siting, and in recognition of the action of the Congress in the1 FY .'!O t.uthorization Conference, it is necessary for the !!RC to day 3 pp a transition
"- policy to move frem the old to the new siting criteria and a cethod of fnensina , . .. -
~
en existing sites. Therefore, the staff is requested to do the folfowinor, * ,..t .. (1)(a) For plants with applications for CP filed before October 1 '- 1979, but that do not have an LWA or a CP or for LWA plants
- with little construction conpleted, the staff is directed to determine for the purpose of this transitionary effort whicts of these would not be expected to meet the recoranendations of the Siting Policy Task Force, fiUREG-0625, as modified by OPE and This con:parison, ACRS co.ments as discussed with the. Coninission.
together with the other aspects of the safety evaluation, will be ,! used to determine if any additional measures in plant design or . i I operation should be reconinended. (, . . .~ ' .. y J. - (b) The staff discussion of siting matters should be placed in the staff Safety Evaluation Reoort or.should be in the form of an addendum to the SER. "The Coninission will'then review these
! matters when the LWA or CP. cases reach us.
(2)(a) For plants with CP applications filed before Octobsr 1,1979 and:
- f. with an OL, or:. .
~- T n . .. . . :,J:4:l,.. ;-;..*2. ' *"
~
. 11. , wit.hout an.0L bu,t with a CP or; .
. iii. without a CP but with an LWA and more than a little
' construct,$4n-completad- _.
. . .t - - .u. ,.y ,... .... .
. the staff is requested to continue to develop a priority list of those Plants in the highest population areas for additional safety evaluatinn.
As the proposed siting criteria .are developed, the staff should if.tennine l how far to extend the list based on those plants which might fail sone significant aspect. ~cf,t.he new cdteria. a - - ,, ,
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(b) Following cor.pletion of those analyses currently un<ferway, i.e., . Indian Point. Zion, etc., the staff should continue safety reviews - for the next plants on the list and reco a:cnd e.at if.any additional '
- measures are indicated. \
t , (c) These staff reviews should be in the form of SECY papers that e are presented first to the ACP.S and then, after appropriate 'y* , ,.: -- revision. to the Co. .ission. Based on the early reviews. . .. the Comuission r'.ay either truneett or require extensien of the
- original candidate list.
1 cc: . Chairman Ahearne . Comissioner Gilinsky . Comissioner Kennedy . . , _ Comissioner Hendrie . -~ - Comissioner Bradford -. ..,. - ..
; Comission Sta o ces ..f . a ,.. .
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'N .'.... '- ' .d* September 5,1920 -
MEMOR/iNCUM FOR: William J. Dircks, Acting Executive Director for Operations FROM: Harold R. Denton, Director
- Office of Nuclear Reactor Regulation
SUBJECT:
SECY-80.lS3 ADVANCE NOTICE OF PROPOSED RULD! AXING ON REACTOR SITING
. a This is in response to item la of the memorandum frem S. J. Chilk l
to W. J. Dircks, dated June 30, 1980. As directed by the Comission, the Staff will include the requested site evaluations in the safety evaluation reports for the applicable CP and ML applications.
- , V ..
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. .. .-. Harold R. Denton, Director
. .. .. Office of Nuclear Reactor Regulation
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4 ENCLOSURE TO ANSWER NO. 8 EXTERNAL COMMUNICATION DOCUMENTS i l
^ '
D ib *v b - D
%, UNITED STATES
[/
,, )
i g NUCLEAR REGULATORY COMMISSION . 1
; j WASHINGTON, D. C. 20665 l
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8 Ocr ses l I Docket No.: 50-443 MEMORANDUM FOR: Vincent S. Noonan, Director PWR Project Directorate No. 5 Division of PWR I.icensing-A FROM: Elizabeth I.. Doolittle, Project Manager PWR Project Directorate No. 5 Division of PWR L.icensing-A
SUBJECT:
FORTHCOMING MEETING WITH PU8l.IC SERVICE COMPANY OF NEW HAMPSHIRE TO DISCUSS CERTAIN TECHNICAt. ASPECTS OF THE SEABROOK RISK MANAGEMENT AND EMERGENCY Pl.ANNING STUDY DATES & TIMES: Wednesday, October 15, 1986 thru Friday, October 17, 1986 Wednesday: 10:00 a.m. - 5:00 p.m. Thursday and Friday: 8:00 a.m. - 5:00 p.m. l.0 CATION: .Wednesday, October 15, 1986
- Seabrook Station Training Center Seabrook, NH ,,
Thursday and Friday, October 16 and 17, 1986
- Brookhaven National I.aboratory Upton, NY PURPOSE: To discuss several topics involved in the staff's review of the Seabrook Risk Management and Emergency Planning Study: Topics include plant response, structural strength calculations, and steam generator tube integrity.
ATTENDANCE: Members of the public who wish to attend the meeting on October 15, 1986, must contact William Sanchez at (603) l- 474-9574, Ext. 2487 by noon on October 14, 1986. l g - .= 9PR '
o
-J. 1. - (-, I'_[
2_ Members of the public who wish to attend the meeting on October 16 and 17, 1986, must contact Dr. Charles Hofmayer at (5161282-2317 by noon on October 15, 1986. PARTICIPANTS: NRC- BNL PSNH S. Long R. Youngblood J. Moody G. Bagchi C. Hofmayer and other D. Hickman D. Wesley (SMA) representatives of the applicant Elizabek1 L. Doolittle, Project Manager PWR Project Directorate No. 5 Division of PWR licensing-A e 9 e G i
- ee r w yM-N "
P'- rw 'W 19'- M w-T -f y er
, .t.
- a. % '
- r. (
- b. Mr; Robert J. Harrison ,
!/ Public Service Company of New Hampsh_ ire Seabrook Nuclear Power Station . I cc-i Thomas Dignan Esq. E. Tupper Kinder Esq. .; John A. Ritscher, Esq. G. Dana Bisbee, Esq.
j Ropes and Gray Assistant Attorney General j 225 Franklin Street Boston, Massachusetts 02110 Office of Attorney General 3- 708 State Hosue Annex s Concord, New Hampshire 03301 Mr. Bruce B. Beckley, Project Manager
. Public Service Company of New Hampshire Resident Inspector . Post Office Box 330 Seabrook Nuclear Power Station Manchester, New Hampshire 03105 c/o US Nuclear Jtegulatory Commission
, Post Office Box 700 Dr. Mauray Tye, President Seabrook, New Hampshire 03874 Sun Valley Association 209 Sumer Street 'Mr. John DeVincentis, Director Haverhill, Massachusetts 01839 Engineering and L.icensing Yankee Atomic Electric Company Robert A. Backus, Esq. 1671 Worchester Road O'Neil, Backus and Spielman Framingham, Massachusetts 01701 J 116 Lowell Street Manchester, New Hampshire 03105 Mr. A. M. Ebner, Project Manager
! United Engineers & Constructors
- Wil.liam S. Jordan, III 30 South 17th Street Diane Curran .ost
" Office Box 8223 1
Hannon, Weiss & Jordan "hiladelphia, Pennsylvania 19101
! 20001 S Street, NW
- i Suite 430 ..
Washington, D.C. 20009 Mr. Philip Ahrens, Esq. Assistant Attorney General State House, Station #6 Augusta, Maine 04333 Jo Ann Shotwell, Esq. . Office of the Assistant Attorney General
- . Environmental Protection Division Mr. Warren Hall One Ashburton Place i Public Service Company of Boston, Massachusetts' 02108 i New Hampshire r Post Office Box 330 D. Pierre G. Cameron, Jr. , Esq.
I Seabrook, New Hampshire 03874 General Counsel
! Public Service Company of New Hampshire ! Se r.oast Anti-Pollution League Post Office Box 330 Ms. Jane Doughty Manchester, New Hampshire 03105 i 5 Market Street j Portsmouth, New Hampshire 03801 Regional Administrat6r, Region I
- } ,
U.S. Nuclear Regulatory Comission Mr. Diana P. Randall 631 Park Avenue - 70 Collins Street King of Prussia, Pengsy.lvania 19406 Seabrook, New Hampshire 03874 i Richard Hampe, Esq. New Hampshire Civil Defense Agency 107 Pleasant Street Concord, New Hampshire 03301 _ . __ _ . . . . , _ . . . , . . . ._ _ _ -. . _ _ -, _ _- ._ ._. _ _ . ~_ -- _ .
F -. .
. .b h '
Public Service Company of Seabrook Nuclear Power Station
. New Hampshire cc:
Mr. Calvin A. Canney, City Manager Mr. Alfred V. Sargent. City Fall Chainnan
.126 Daniel Street Board of Selectmen Portsmouth, New Hamnshire 03801 Town of Salisbury, MA 01950 Ms. letty Pett Senator Gordon J. Fumphrey Town of Brentwood ATTN: Tom Burack .
RFD Dalton Road U.S. Senate Brentwood, New Hampshire 03833 Washington, D.C. 20510 Ms. Roberta C. Pevear Mr. Owen B. Durgin, Chairman Town of Hampton Falls, New Hampshire Durham Board of Selectmen Drinkwater Road Town of Durham Fampton Falls, New Hampshire 03844 Durham, New Hampshire 03824 Ms. Sandra Gavutis Charles Cross, Esq. Town of Kensington, New Hampshire Shatnes,~Mardrigan and RDF 1 McEaschern . East Kingston, New Hampshire 03827 25 Maplewood Avenue Post Office Box 366 Portsmouth, New Hampshire 03801 Chairman, Board of Selectmen RFD 2 South Hampton, New Hampshire 03827 Mr. Guy Chichester, Chaiman Rye Nuclear Intervention Mr. Angie Machiros, Chairman Comittee Board of Selectmen c/o Rye Town Ha.ll for the Town of Newbury 10 Central-Road Newbury, Massachusetts 01950 R,e, New Hampshire 03870 Ms. Cashman, Chairman Jane Spector Board of Selectmen Federal Energy Regulatory Town of Amesbury Comission - Town Fall 825 North Capital Street, NE Amesbury, Massachusetts .01913 Room 8105 Washington, D. C. 20426 Honorable Peter J. Matthews Mayor, City of Newburyport Mr. R. Sweeney Office of the Mayor New Hampshire Yankee Division City Hall Public Service of New Hampshire Newburyport, Massachusetts 01950 Company - 7910 Woodmont Avenue Mr. Donald E. Chick, Town Manager Bethesda, Maryland 20814 Town of Exeter . 10 Front Street Mr. William B. Derrickson Exeter, New Hamp:, hire 03823 Senior Vice President Public Service Company of New Hampshire I - .- Post Office Box 700, Route 1 Seabrook, New Fampshire 03874
. ~ . - 1 L ,f 8,, UNITED STATES .
Y a NUCLEAR REGULATORY COMMISSION l 5 -) WASHINGTON, D. C. 20555 ( l
%, %. . */ 007 n igg Y
~
Docket No.: 50-443 , MEMORANDUM FOR: Vincent S. Noonan, Director PWR Project Directorate No. 5 Division of PWR Licensing-A FROM: Elizabeth L. Doolittle, Project Manager PWR Project Directorate No. 5 ; Division of PWR Licensing-A [
SUBJECT:
FORTHCOMING MEETINGS WITH PUBLIC SERVICE COMPANY !
.. OF NEW HAMPSHIRE TO DISCUSS CERTAIN TECHNICAL i ASPECTS OF THE SEABROOK RISK MANAGEMENT AND EMERGENCY PLANNING STUDY DATES & TIMES: Thursday, November 6, 1986 10:00 a.m. - 5:00 p.m. - FNII Wednesday, November 12, 1986 10:00 a.m. - 5:00 p.m.- fil4 Wednesday, November 19,1986 10:00 a.m. - 5:00 p.m. - ryste PURPOSE: To discuss technical issue involved in the staff's review of the Seabrook Risk Management and Emergency Planning !
- Study.
- PARTICIPANTS
- NRC PSNH S. Long, el d D. Maidrand, e_t t al CONTACT: ' Members of the public may contact E. Doolittle at (301) l 492-8379 for additional information.
Yfb Elizabe L. Doolittle, Project Manager (. PWR Project Directorate No. 5 L Division of PWR Licensing-A i
- - $LOJ1% '6d) 5?4 -
W-
$- . . - - , , , - + - -= ,. m, -, , - - , , ,f--- -,, - -. ,,,e. ,. . -.--m -r y - ,-.--, - , , - - , ----- , - - - -
~
SEABROOK STATION
- Enginasring Office IP T M ,
November 7, 1986 Pub 6c Service of New Hai,f.4 ire SBN- 1227 T.F. B7.1.2 NEW HAMPSHIRE YANKEE DIVISION United States Nuclear Regulatory Commission Washington, DC 20555 Attention: Mr. Steven M. Long , Project Manager PWR Project Directorate No. 5 Division of PWR Licensing - A
References:
(a) Facility Operating Licensa NPF-56, Construction Permit CPPR-136, Docket Nos. 50-443 and 50/444 (b) USNRC Letter, dated October 8,1986, " Request for Additional Information for Seabrook Station, Units 1 and 2, Emergency Planning Sensitivity Study", S. M. Long to R. J. Harrison (c) USNRC Letter, dated October 23, 1986, " Request
, for Additional Information for Seabrook Station, Units 1 and 2, Emergency Planning Sensitivity
) Study", S. M. Long to R. J. Harrison (d) P,SNH hetter (SBN-1225), dated October 31, 1986,
" Response to Request for Additional Information (RAIs)", J. DeVincentis to S. M. Long Subj ect : Response to Request for Additional Information (RAIs)
Dear Sir:
Enclosed herewith are additional responses to the Requests for Additional Information forwarded in Ref erences (b) and (c). Previous responses were submitted in Referenct (d). Attachment A identifies responses that are in-cluded in this transmittal. The responses are provided in Attachment B. An additional submittal addressing the remainder of the RAIs will be forthcoming in the near future. Very truly yours,
/s C John DeVincentis i Director of Engineering Attachment cc: Atomic Safety and Licensing Board Service List Director, Office of Inspection and Enforcement United States Nuclear Regulatory Commission Washington, DC 20555
, rt,1 Apl+3
.2 6,4 t= %
Secbrock Station Construction Field Office . P.O. Box 700 Secbrock. NH 03874
. . . _ . ~ . . . _ _ . . - _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ . . . _ . _ _ . _ . . . . . _ .
~
Dicn3 Currca, Esquire Pater J. Math ws , Mayor
'Harmon & Weiss .
City Hn11
-2001 S. Street, N.W. Newburyport, MA 01950 i
Suite 430 W :hington, D.C. 20009 Judith H. Mizner , Silvergat e, Gertner, Bake r , ' Sh rwin E. Turk, Esquire Fine, Good & Mizner ! Office of the Executive Legal Director 88 Broad St. U. S. Nuclear Regulatory Commission Boston, MA 02110 T nth Floor W:shington,lDC 20555 Calvin A. Canney
, City Manager -
Rob rt A. Backus , Esquire City Hall 116 Lowell Street 126 Daniel Street P. O. Box 516 Portsmouth, NH 03801 Manchester, NH 03105 Stephen E. Merrill, Esquire i Philip Ahrens, Esquire Attorney General Assistant Attorney General George Dana Bisbee, Esquire ! D:pSrtment of the Attorney General Assistant Attorney General Seccchouse Station #6 25 Capitol Street Auguata, ME 04333 Concord, NH 03301-6397 4 Srs. Sandra Gavutis Mr. J. P. Nadeau Chairman, Board of Selectmen Selectman's Office - RFD 1 - Box 1154 10 Central Road i Kensington, NH 03827 Rye, NH 03870 Carol S. Sneider, Esquire Mr. Angie Machiros Assistant Attorney General Chai rman of the Board of' Selectmen Department of the Attorney Genural - Town of Newbury, a Newbury,'MA 01950-On3 Ashburton Place,19th Floor - 5:ston, MA 02108 . - Mr. William S. Lord S:ancor Gordon J. Humphrey Board of Selectmen U. S. Senate Town Hall - Friend Street Wcchington, DC 20510 Amesbury, MA 01913 ( ATIN: Tom Burack) . k Richced A. Hampe, Esquire Senator Gordon J. Humphrey Hampa and McNicholas 1 Pillsbury Street 35 Pleasant Stre'et Concord, NH 03301 Con: sed, NH 03301 ( ATTN: Herb Boynton) , Thoacs F. Powers, III H. Joseph Flynn, Esquire l Town Manager Office of General Counsel l Tcwn of Exeter Federal Emergency Management Agency ! 10 Front Street 500 C Street, FW Ex3 tor, NH 03833 Washington, DC 20472 Brantwood Board of Selectmen Paul McEachern, Esquire RFD Delcon Road Matthew T. Brock, Esquire i Brentwood, NH 03833 Shaines & McEachern 25 Maplewood Avenue Grry W. Holmes, Esquire P. O. Box 360 H21 mas & Elis Portsmouth, NH 03801 47 Winnacun' net Road Hampton, NH 03842-Mr. Ed Thomas Robert Carigg FEMA Region I Town Office l 442 John W. McCormack PO & Courthouse Atlantic Avenue , BistCn. MA 02109 North Hampton, NH 03862
. - - . _ . , . - . . - - . . . . - - . . ~ . . . - - - . - - - ~ - -- - -- - - - - - ~ ~ - - - - -
ATTACHMENT A Responses to these RAIs were forwarded by SBN-1225, Reference (d): 1~ 12 24 44 61 2 13 25 45 62 3 '14 26 46 63 . 4 15 28 49 64 5 16 33 50 67
, 6 17 34 51 68 .
I 7 19 35 53 69 8 20 40 55 70 s 9 21 41' 57 71 10 22 42 59 72 11 23 43 60 73 Responses are included in this transmittal for the following RAIs: 18 48 75 . 27 52 - 30 54 . . 32 56 36 58 37 65 39 66 47 74 Responses to the following RAIs will be forthcoming in an additional submittal: . l 29 31 !. 38-h - l-
.a r-e,-- en+ m-e- .
am m e. 4 -o -a w ee n- m, + ..m , me- 6,w.- e4 9 ATTACH. MENT B RESPONSES TO REQUESTS FOR INFOPy_ATION e l I
_ . . ... . . _ . _ . ... _ _ _ _ . s. .. . . , , . . . _ . . . __ , . _ _ _ . . ,_ .
- j
( a I I
RAI 18
Provide the basis for concluding that the sight glasses in the hatches r will not fail under high containment temperature and pressure conditions.
't e .
ANSWER 18 The sight glass in the personnel hatch was tested by its supplier, Owen Corning Co., under the following conditions: Pressure = 150 psig l Temperature = 5500F i [ In addition the pressure was cycled from 0 psig to 150 psig ten times h at a constant temperature of 5500F. l The Owens Corning data sheet is attached. l l We are current 1'y pursuing discussion with Corning Glass to determine if any testing has been done above these values. Corning indicates that a conservative allowable working stress for the
#7740 tempered glass is 300 psi. At this working allowable, pressures in excess of 200 psi can be accomodated without glass failure.
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4 l 1 TEST SAMPLE: TEMPERED SIGHT GLASS, CORNING PYREX NO. 7740. 7 INCHES DIAMETER BY 3/4 INCHES THICK. W. J. WOOLLEY TESTING: , PERFORMANCE TEST: PRESSURIZED TO 150 PSIG FOR 3 MINUTES. , MECHANICAL CYCLING TEST: PRESSURIZED FROM 0 TO 150 PSIC, MAINTAINED FOR 10 SECONDS, THEN TO O PSIG, A TOTAL OF 100 CYCLES. RADIATION EXPOSURE: 60 MEGARADS GAMMA. (NCrrE 1) POST RADIATION PERFORMANCE TEST: PRESSURIZED TO 150 PSIG FOR 3 MINUTES 5 TDtPERATURE : AMBIENT CORNING TECHNICAL SPECIFICATIONS: MAXIMLN WORKING PRESSURE: 150 PSIG MAXIMLH WORKING TEMPERATURE: ,290 0 C, 5540 F ' NOPMAL SERVICE TEMPERATURE: 2600C, 5000F HEAT SHOCK TEST: ROCM TDtPERATURE M 270 0 C,' 5180 F
REFERENCES:
- 1. CORNING GLASS BULLETIN 15-20, 1971
- 2. CORNING GLASS CUSTOMER PRODUCT SPECIFICATION, JANUARY 15, 1973
- 3. AMERICAN ENVIRON >2NTS COMPANY QUALIFICATION TEST REPORT FOR PERSONNEL AIRLOCK VIEWPORT SIGHT GLASS REPORT NO. STR-62883-2
{ 1 NCYTE: . 4
- 1. THE SIGHT GLASS CHANCED COLOR FROM THE GREEN TINTED CLEAR TO A DEEP REDISH PURPLE, DUE TO THE GAMMA RADIATION EXPOSURE. VISIBILITY THROUGH THE SIGHT GLASS WAS REDUCED.
I L
l R I 27 Discuss the ef fect on risk of hydrogen deflagation/ detonation in the RHR vault. RESPONSE 27 Transient hydrogen burn analyses were not performed for the RRR vaults. During a V-sequence hydrogen is expected to be released into the RHR vaults via t,he RRR pump seals. Steam condensation in the water pool may increase the concentration of hydrogen released at the top of the water pool to a flammable condition. If a hydrogen burn were to. occur above the pool in the RHR vault it would almost certainly be a continuous burn at the pool surf ace because of the continuous release through the RHR pump seal. prior to vessel melt through. A postulated global hydrogen burn in the RHR vault could cause a pressure induced failure of the RER vault boundary. The RHR vault is part of the enclosure building area ventilation vault boundary and the pressure capacity of this boundary was analyzed in Section 6 of Appendix H.1 in the SSPSA. The weak points in the RRR , vault pressure boundary are readily apparent. Fi rs t , the double wing fire doors at grade elevation (+20 feet) communicate into a corridor which communicates to the outside through another large double door. These doors are located approximately 50 feet above the water ' pool surface. The second weak point, is the RHR vault ventilation system which communicates with the eiclosure building area. The weak points in the pressure boundary for the enciosare building were identified as (1) the metal siding in the fire walls between the ch' rging a pump cubicle and the primary auxiliary building, (2) the enclosure building ventilation ducting in the primary auxiliary building and (3) the HEPA filters in the enclosure building ventilat?1n exhaust. Table 11.3-4 in the SSPSA shows that both the ventilation duct work and the HEPA filters are expected to fail at a pressure below 2 psid and the metal siding in the charging pump cubicle is expected to f ail at a similar low pressure. The RER vault structure on the other hand is designed for an internal pressure of 3 psid which according to the analysis in SSPSA Appendix H.1. Section 6 has an expected pressure capacity of at least 6 psid. thus the expected failure locations due to overpressure in RNR vault are readily identified. The analysis of the V-sequence in the RNEPS takes credit for source term mitigation, by the deposition of fission products on the structure surf aces above the pool. . The f ailure mode involving the ventilation duct work or the HEPA filters both would increase the surfaces available for radionuclide deposition before release. The vault failure mode involving the two sets of fire doors at grade level would result in radionuclide releases to the environment by a release path that is essentially identical to that modeled in the RMEPS study. In the WASH-1400 sensitivity study (PLG-0465) credit was only taken for radionuclide scrubbing in the RHR vault pool, but not for any deposition above the pool. The only mechanism which could increase the consequences of a V-sequence in the sensitivity study as a result of a hydrogen burn would be the structural f ailure of the vault walls at such a low level that the water pool is lost down to the RHR pump level. This would require _ . _ , . _ _ _ + , ---.-n- . , _- _ _ _ - -
failure of the concrete vault walls at an elevation where the walls are 4' feet thick, twice the thickness of the walls at higher elevations. As described above the pressure capacity of the two foot thick concrete walls is at least 6 psid. Failure at these low elevations will thus be prevented by failure first of the metal siding, fire doors, ventilation ducts, or HEPA filters, and secondly by the thinner concrete walls above elevation (-)41. Loss of the water pool will thus not occdr and a postulated global hydrogen burn in the RHR vault would not increase the consequences of a V-sequence. '
+ 4 e
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- 4 f
1 l l
RAI 30
The S7W release category isotopic di.stribution reflects a decontamination factor (DF) of 1000, for all isotopes except noble gases, because the release point is ' submerged in the RHR vault. WASH-1400 source term methodology credited BWR releases-with-a DF of 100 when they occured through a subcooled suppression pool, but set the DF to I when the pool was at saturation temperature. a. ' Discuss the degree of subcooling that would be expected in the RCS water that pools in the RHR vault following blowdown through the RHR system.
- b. Justify the use of a DF-1000 in light of the WASH-1400 methodology and the degree of subcooling expected in the RHR vault water.
m RESPONSE 30 The subcooling of the water in the RHR vault is a function of the size of the break, the injection rate, and heat losses to structures in the reactor buildi ng . Considering the wide variations of accident scenarios that could be postulated, it is reasonable to assume that the bulk subcooling of the pool would be at least 100C due to concrete heat sinks. In addition, the . static head of the water in the vault also represents a significant subcool-ing, since a 10 meter deep pool saturated at the surf ace would have a subcooling of 100C at.the failure location. Due to these subcoolings (10 to 200C), a considerable amount of deposition of the aerosol fission products woul'd be created by the initial bubble collapse. Analyses that were performed to calculate the degrse 'of subcooling are described below. The S7W release category is the expected result of an interfacing systems LOCA, or V-sequence initiating event. The f ailure of the RHR isolation valves which initiates the event allows the pressurized primary water mass to blow down into the RHR pump vaults through the RER pump seals. The highly pressurized reactor coolant entering through the RHR break will flash upon contact with the low pressure atmosphere in the RHR pump vaults. The steam generated by this flashing process will enter into the RHR pump vaults and the saturated water generated will f all to the floor of the RHR equipment vault and begin to cool through heat transfer to the concrete. The transient will begin with highly pressurized water exiting through the break, then will continue with a short period of steam flow through the break as the rapid depressurization in the primary system causes some flashing in the remaining reactor coolant. Safety injection will
- be actuated by the decrease in RCS pressure and will reflood the core with water from the RWST within several minutes af ter the steam flow through the break begins. W a ter flow will then continue through the break driven by the pressure of the RCS which is maintained at or below the RHR relief valve setpoint pressure until vessel failure occurs.
' The RCS pressure is maintained initially by the safety injection pumps j until they are flooded by the rising level in the auxiliary building, and then driven by the charging pumps until the RWST is emptied. Once i the RWST is emptied, the water level will drop in the core until the hot leg is once more uncovered, and the breakflow will then consist d y -
-s y,i--,p + m am=-e- y-i-e< -e-,yy m 9g,y-,g wy.g .---y , . - - . ---%v.w -- gp-nw y % r, y,yy,- r9- y y-ge g-umr og+- evee gu--eme -1pw-sm----wwwwmvw*-g----+------vr.- .ww w vi
-. ~ . - -. __ _ _ . _ . . ._ _ _
solely of steam flow for the remainder of the transient. As the core melts and vessel failure occurs, the pressure driving the steam flow will drop off and the steam flow will also decrease. The original analysis consisted of a hand calculation using the model specified above which has since been verified with a confirmatory computer run. The steam flow from the flashing at the break was conservatively ignored until the break is covered, since condensation of the steam on the walls and in the air would add inventory to the auxiliary building pool and would cool the water in the pool. Af ter the break is covered by the rising water in the RHR vault , the water exiting the break was assumed to mix with the water in the pool. When steam is exiting the break, it is assumed to bubble through the pool without much condensation because if its initial flow velocity. l
-Heat transfer was only assumed between the water and the surrounding concrete floor and walls, and no heat tranfer was considered between the water and metal equipment or the RHR vault atmosphere, which would act to subcool the water in the vault even further. These assumptions were used to verify the results and conclusions of the original analysis. The results for the RHR vault bulk water temperature versus time for the duration of the transient are presented in graphical form in Figure I with some results in tabular form in Figure 2. Please note that this figure does not predict the return to saturated conditions near the end of the transient run time of 24 hours noted in the Emergency Planning Study near the top of page 4-36. This return to saturated conditions is due to the steam break flow race dropping sufficiently 4 af ter core melt and vessel failure to allow the steam to condense as it flows into the RHR vault water pool.. This condition was'modeled in the initial hand calculations, but not in t,he' computer results presented here.
The results presented in F'igure 1 show that the RER vault water pool remains subcooled throughout the period of maximum fission product release through the break (10 to 18 hours into the transient). Combined with the depth of the RHR water pool, the subcooled conditions of the pool should provide sufficient justification for the use of a DF of 1000 for this release category. The decontamination f actor is determined by subcooling of the pool, the particle size, and the steam mass fraction being lost from the break location. Typical representations show a decontamination factor as a function of these elements and the decontamination itself experiences a minimum with the lowest DF being at a particle diamter of about .25 L microns and an equilibrium steam mass f raction in the existing gas flow. In the attached memo, this is shown to have a DF of about 3,000 with a water depth in the vault of 10 meters. Considering that (1) the water would have subcooling both due to structural heat sinks and the static head of the water pool and (2) that the typical aerosol distribution would have particles in the range of a few microns, the average DF of 1000 used in the study is a conservative assessment for the complex processes involved in aerosol deposition. For typical V-sequence scenarios with flow down a long pipe, the aerosol particles would experience deposition on the pipe wall and re-entrainment by the high velocity gas stream before being discharged into the water pool. In this case, the particles which are entrained by the gas stream would be a few tens of i
,Sk
' cb
'.i-l microns to 100 microns in diameter. Particles of.this size would have a very high decontamination f actor, i.e. > 104 . This is an additional conservatism not accounted for in this analysis.
The above discussion of decontamination f actors represents our best state of knowledge about the unique conditions associated with RHR pump vault flooded release paths. In the sensitivity study, it was determined
. that WASM-1400 methodology did not have an accurate source term for this scenario. The closest match was the use of DF of 100 for subcooled BWR suppression pool releases. It was also known that the IDCOR program assessed BWR suppression pool scrubbing at a DF of 1000 (IDCOR 23.1 for Grand Gulf) and subsequently, based on additional experimental data, a BWR DF of 500 to 700 was supported (IDCOR 85.2). Therefore, the IDCOR work suggests that WASH-1400 may have been conservative for BWR s,uppression pool scrubbing by a f actor of 5 to 10. In view of the greater depth of water in the Seabrook RHR pump vault, the assumption of a DF = 1000 seemed to be appropriate for performing the sensitivity study. It is clear that the results of the sensitivity study are insensitive to assumptions regarding RHR decontamination f actor. If a DF of 100 had been used, the numerical results for early health risk safety goal and dose vs.
distance would not have changed appreciably. Even if a DF of I had been used, the 200 REM dose vs. distance would not have changed because of the low frequency assessed for S7. I e 9 e 9 r . - . .
\
FIGURE 1 ANALYSIS FOR SEABROOK V SEQUENCE SUBCOOLED WATER SCRUBBING TEMPERATURE OF WATER IN LOWER VAULT REGION 220.
'""%========
1 9 e E 180. 5 a 3 160. a 5 .
~
4 3 140. o u . 3 120. E 2
$ 100.
' 0. 20000. 40000. -
60000. 80000. TIME (SEC) l C
--- - - . . . . . - , ,--,,._n..,,, ,_ , , . , , , , , _ . , . , , . _ , . ._a., , _ _ . , , , _ _ _ _ , ,
. . . . . . - - . . . . - - -. .. . - - . . . - - . . k FIGURE'2 Tabular Results of Seabrook Auxiliary Building.' Water Analysis Time (sec) Water Temperature (OF) 192 .
156.3 268 18 0 .8 613 162.1 1281 200 .4 2418 205.1 6087 20 8 . 5 10553 209.5 20213 . 208 .9 29997 - - 208.1 40033 206 9 59140 205.0 86401 203 1
r 7...- E Fauske & Associates,Inc. DATE: October 31, 1986 TO: R. E. Henry FROM: P.G.EllisonIfk
SUBJECT:
Minimum Expected DF in a Pool 10 m Deep with Tyoical V-Secuence Conditions The decontamination of a gas stream of aerosol is given by
- kZ OF = e
(]) where Z is the pool height and k is given by ' 2h 8h Tg ,, s qod k=s1 2 l+ l 8 (2) (120V'+oVd Du [ 3 V0(3rud) The slip correction facto'rS is. given by
~
s = 1 + Kn (2.514 + 0.8e-0.55/Kn) Equation (2) has a minimum around a particle diameter d of 0.25 um. The minimum DF in a pool, assuming that k is not a function of height can be estimated using the minimum particle diameter, equilibrium steam fraction and neglecting entrance effects. For the typical V-sequence conditions given below: 3 o = 5000 kgm/m 0 = 5 x 10-3 , d = .25 x 10-6 , T gas
= 350*K u = 1.7 x 10-5 kgm/m-sec X = 1.38 x 10-23 joule /*K o = 0.5 kgm/m 3 V = 0.25 m/see s = 1.947 Kn = 0.353 Z = 10 m 16WQ70 West 83rd Street
- Burr Ridge, Illinois 60521 * (312) 323 8750 i
I TO: R. E. Henry
- 2- October 31, 1986 and from Equation (2)
- k = 0.023 + 0.358 + 0.441
= 0.822 or DF = e0.822 10 DF s 3000 Thus a minimum DF of the order of 1000 is expected for these conditions.
PGE:lak 9 h O ' S
- e 4
. _ . , _ - . _. . _ . _ . ~ . _ . . . . . .. _ . . - . - - ._ _ _
i . .
RAI 32
In your prediction of large deformation behavior of the containment, full bond was assumed between reinforcement and concrete between two adajacent ' vertical cracks;' assess the effect on containment behavior including penetration
. capability, if no bond stress is assumed between the reinforcing steel yield point and ultimate strength of steel. Based on our discussions in the meeting, it is our understanding that you will perform this assessment' assuming no bond stress. .
RESPONSE 32 As was agreed in the review meeting at Brookhaven National Laboratory on October 16 and 17,1986 we have recalculated the containment pressure versus deflection curv,e (Figure 4-2 in Appendix H.1 of the SSPSA) under the assumption of no bond stress, and we have interpreted the difference in the two curves in terms of changes in the containment failure pressure and in the' containment fa'ilure time. The attached' Figure shows both the original pressure displacement curve and the no bond strength curve. - The difference in failure pressure is seen to be 15 psi at th'e median failure pressure of 196 psig,10 psi at the 1 percent deformation strain and 7 psi at the type A containment failure pressure of 166 psig. The table below compares these pressure changes and the resulting changes in containment failure times. It is concluded that:
- 1. The small changes in the time of containment failure will not have any noticable impact on the source terms or on the consequences for either the RMEPS study (PLG-0432) or the sensitivity study (PLG-0465).
- 2. None of the conclusions in the above referenced studies would be affected.
- 3. The effects of assuming no bond stress are well enveloped by the results of the sensitivity study for containment failure which was documented in response to RAI 20. In that analysis it was arbitrarily assumed that containment f ailure occurs at a deformation strain of 1 percent.
S
- - _ _ . m___..___.. _ _ _ _ _ . - _ . . . _ , . _ .
, _ . , ~ --' * . _ . _ _ . . ----- .. . . . - - - .
- 4. e.;
.{ (.
z, ( CONTAINMENT FAILURE PRESSURE - HOURS) 'PSIG) AND i FOR TWO ASSUMPTIONS ON CONCRETE REBAR CASE CONTAINMENT FAILURE PRESSURE (PSIG) TIME (ROURS) f _ BOND STRESS YES , NO YES NO WET, MEDIAN WET, TYPE 'A" 196 18 1 166 54 i 3 159 30 38 36 i DRY, MEDIAN DRY, TYPE "A" 172 162 89 157 76 151 62 h l 56 l l e L I i! l 1 l l l l I i
31 ',- Figure 1 ~ , Comparison of Pressure Displacement Curve i with and without Bond Stress ' ' i . [* 250 y y inck4ng effect of bond stress -
/. (original curve) p # ~
p # : T / # - 5 150 assuming no bonding of , l g % bars ] O
~
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.E !
5C . O O 50 l !, r 1 i 0 ,
- O 10 20 30 40
- Radial Displacement (inches) t
. - . . . . - . ... -. ..= -.~ . -
RAI 36
Discuss the results of recent EPRI tests to address the potential for strain concentration in the liner at crack locations. 4 RESPONSE 36 . SMA is presently reviewing. the EPRI test reports to determine the applicability of any test ' results to the assumptions made in the ultimate capacity analysis. The NRC will be notified if the reviews identify any information that has a significant' negative impact on the SSPSA or the RMEPS conclusions. O e e 4 e
- , e - - , - - - - - -, --- -
--.-4 n--, .n-v-mwn-, , , , 4 4 q w--m q ..,,r. , ~ - - , , --- o q g - -+ ,
RAI 37
Demonstrate that your calculations fully account for the differences in stress-strain behavior between the reinforcing steel and the lower plate with regards to strain compatibility. RESPONSE 37 As discussed in the October 16th and 17th meeting at BNL we have investigated the availability of a stress-strain curve for the liner material in the plastic region. This curve could be utilized to demonstrate that liner strain at ultimate capacity is on the flit part_of the plastic region. We have not located a pertinent plastic stress-sta'in curve for normalized A516 Gr 60, however we have located extra liner plate material on site which is from the same " heat" as installed liner plate. Utilizing actual . liner material, we will perform necessary testing to develop the relevant plastic stress-stain curve. This testing and final curves will be completed in December, 1986. I l I
RAI 39
Only selected penetrations were analyzed in the calculations;-compile a list of all containment penetrations, categorize according to behavior and demonstrate that each penetration is adequately covered by the analyses that have been performed. t RESPONSE 39 Containment piping penetrations and their qualification methods are summarized in the attached table.
?
t
- - . , - - - - . , , , - , , . y, ,, , -- , , , - .r---n, -,,, -- - , , . ,
I Panetration Closure . Penetration Number Penetration Specifically Qualification Type Analyzed Method ' I. Flued Head X-1 to X-8, X-9 to X-15, . X-8 Report pages H.1-44 to i-X-63 to X-66 (18 inch, och 100 Carbon Steel) H.1-50 II. Flat PlateClosure X-25, X-26, X-27 X-26 Pages H.1-39 to H.1-44 : Thick Wall - Large (4 inch, och 160, stainless) { Bore Piping ' III. Flat PlatsClosure X-16 thru X-24 X-23 Pages H.1-39 to H.1-43 Thin Wall - Large X-28 thru X-34 (12 inch, sch 40, carbon Steel) , Bore and Small Bore X-39, 41, 42, 50, 60, - Piping 61, 67 , IV. Flat Plate Closure X-35 thru X-38 X-71 Page H.!-37, H.1-39 Thin Wall Piping X-40, X-4 3, X-47, X-48, Multiple Penetration X-49, X-50, X-52, X-57, X-71 thru X-76 V V. Fuel Transfer Tube X-62 X-62 , Page H.1-50 to H.1-55 s t O f 6
we: l
~
RAI 47
Address the risk from creep failure of the steam generator (S/G) tubes due to exposure to high temperatures during core melt sequences in 4 which the reactor coolant system (RCS) remains at high pressure and the secondary sides of the S/Gs are dry. Your discussion should reflect the recent experiments and modeling ef forts that show 3-dimensional t_ convective flows which transfer heat from the overheating core to other places within the RCS particularly into the upper plenum and from the upper plenum along the hot legs into the S/Gs and through the U-tubes. Also include the influence of pressure driven flows resulting from "reactor bumping" coolant the RCPs, pump (RCP) seal LOCAs, PORV/ safety valve actuations, etc. Localized heating effects due to -redistribution of fission products in the RCS should be included. a. What is the total probability of occurrence for the high RCS pressure core melt sequence with dry S/Gs?
- b. What ~
is the estimated conditional probability that the S/G tubes will fail due to overheating before the pressure is relieved by failure of the-RCS elsewhere?
- c. What is the effect of preexisting S/G tube leakage (within technical specifications) on the heating rate and temperature required for failure of the leakingi. cube (s).
'd.
What release category would creep failures of .the S/G tubes result in? RESPONSE 47 The risk from creep failure 'of steam generator (SG) tubes is very small for the followi,ng reasons: 4 o The frequency of high pressure core melt with dry steam generators is very small.
.i o Given the postulated occurance of a high pressure core melt with dry steam generators, creep rupture of the SG tubes is not a credible failure mode.
o A large number of tubes must fail to produce an early large
. containment bypass.
A simple sequence diagram in Figure I shows that three failures must occur for containment bypass to occur as a result of a high pressure melt sequence. The three failures are (1) failure to recover water to the SG, (2) f ailure to depressurize primary system and (3) creep f ailure of SG tubes. Success of any one of these three ensures success (no SG tube crecp failure). d
. . - - . . , . , , p. .~~..,e-. ,_..,.--,4 w - , . , , - , - , _ . . , , , ,. ,,,,, ,, , ,,-_ ,,,,y.-., _... , -
l l The potential for creep failure of the steam generator tubes for core melt sequences in which the steam generators are dry was evaluated using the best tools available including the MAAP 3.0 computer code developed in the IDCOR program. As a result of the technical issue resolution process between IDCOR and the NRC, this code includes models for the natural circulation flows between an overheated core, the upper plenum, the hot legs and the steam generator tubes. The analytical models used in the MAAP PWR program have been benchmarked with the recent published experiments carried out at Westinghouse under EPRI sponsorship. In addition, these analyses also include the influences of PORV/ safety L valve actuations, as well as models for fission product release, transport, deposition and localized heating. Several variations of station blackout sequences were considered: all assumed f ailure of main and emergency feedwater such that the secondary J sides of the steam generators remained dry. The sequences studied include: (1) no RCP seal LOCA or manual PORV actuations (2) a 50 gpm per pump seal LOCA with no manual RCS depressurization (3) manual RCS depressurization when the core outlet temperature equals 12000F, and (4) several uncertainty analyses on key physical parameters in the system , models. These analyses indicated that the steam generator tubes would not be subject to a sufficient temperature increase to result in creep
- rupture before failure of the reactor coolant system. In particular, since the secondary side temperature is limited to about 7000F due to thermal inertia of the SG shell and heat losses to the containment, the SG tube temperatures only achieve temperatures of about 11000F.. This is far less than that required for creep rupture. Based on information provided in Appendix B of .the attached draf t report, creep rupture of the SG tubes would not be' expected. These conclusions are valid for all of the sequence variations considered as well as all of the uncertainties considered in the analysis which include variations in the core eutectic formation and slumping behavior, flows through the steam generator y tubes, etc.
During station blackout seq 2ences, the Seabrook operating procedures require that the operators monitor the f unctional restoration guidelines. These guidelines would first lead the operator to consider restoration of P; reactor coolant system inventory control which would be met with the g discovery that no electric power is available for this postulated y sequence. Subsequently, the restoration guidelines would lead the operator to consider procedures for restoration of a RCS heat sink. 1, The procedure for loss of heat sink calls for manual depressurization
- of the primary system using the PORVs. While this procedure is not specifically implemented for station blackouts and is only monitored, l!
-' if new evidence subsequently turns up to support the view that creep
. rupture is a credible failure mode, the station blackout procedures R could be changed to require some appropriate method of RCS depressurization.
j The preferred method would be to blow down the secondary side of the l steam generators using the SG PORVs. Such action would permit introduction of steam generator cooling via the firewater pumps (diesel driven). An alternative method would be to depressurize the primary system using the pressurizer PORVs. With such an action, the pressure dif ference imposed on the entire primary system pressure boundary is dramatically reduced, thereby eliminating any potential for creep rupture f ailures. l l
~ , ,--.,---,--,,e- - , . ,, ,+,--,e--e, - , , - , - , , , - , - , , , - , 1-, - , , , , ,,r ,--g- ,eme.,-_,,,__,e -n-nwm,,.m---o,ee. - - --e,e-e-ee---
- - -. -_~ - . _ . __
The action of " bumping" the RCPs cannot be accomplished without the. return of of f site power. If one postulates of fsite power being returned, the operator would first restore ECCS systems. This action alone will recover core cooling and make bumping the RCPs unnessary.- In summary, extensive MAAP 3.0 analyses, which includes models for the pro-cesses in question, were carried out in response to this question. These analyses show that the steam generator tubes would not be threatened by the creep rupture mechanism due to either the natural circulation flows between the core, upper plenum, hot legs and steam generators or the pressure driven 11ows resulting from RCP seal LOCAs or manual RCS depressurization. Also, the calculations show that localized heating ef fects due to fission product deposition are not sufficient to increase the temperature of-the steam generator tubes to a level where creep rupture would be anticipated. Consequently, if any such mechanism were anticipated, it would occur elsewhere in the primary system long before the steam generator tubes would be threatened. The mean annual frequency of early high pressure core melt sequences of interest is approximately 4.5 x 10-5 events per reactor-year. 3 PDS Hean Annual Reference Frequencies 3D 1.5 x 10-5 RMEPS Page 3-58 3FP r 8.5 x 10-6 RMEPS Page 3-60 4A RMEPS Page 3-61 4C ' .1.4 1.7 xx10-7 10-5 SSPS A Page 13.1-23 4D 2.8 x 10-6 SSPSA Page 13.1-26 4E 2.2 x 10-I1 SSPSA Page 13.1-28 4 FP 1.2 x 10-7 SSPS A Page 13.1-28 8A 3.9 x 10-6 RMEPS Page 3-66; sequences 8A-27 through 30 and 8A-34 4.5 x 10-5 through 37 There are many specific accident sequences that comprise the above plant damage states. These sequences broadly include transient and loss of of fsite power sequences with failure of main and emergency feedwater and failure or inability to feed and bleed and transient without scram sequences. The above results for PDS 8A consist of 8 sequences involving station blackout and emergency feedwater f ailure in which ef forts to recover containment heat removal are successful . In the containment event tree analysis, all the "A" states are assigned a high chance of no containment f ailure, the "C" and "D" states 'a high chance of long term overpressurization, the "FP" states a high chance of small bypass and the "E" states a high chance of large bypass. Hence, in terms of relative consequences, the " A, C and D" states would experience the greatest increase in consequences if a SG tube f ailure were assumed to result from creep failure during high pressure melt sequences. 1
- ,- _ , ,- , - .., -,- _,... ,, ,___,.._-___,_,_._m.__,- .-_+__m ._..-.,_-.__..._,,_.--.,_.m__,_ .m.--4_. - , , ._
- __ _. . - . ~ . . . . . . . _ _ __ _ _ . . _ . _ _ . _ . ._
Early high pressure sequences are of interest because sequences with emergency feedwater f ailure (i.e. dry S/Gs) are mo. deled as early melts j (i.e. plant damage states 3 and 4). The one exception applies to 8 j sequences in RMEPS where containment heat removal recovery was considered. These particular recovered sequences were assigned to PDS 8A. PDS 3F and 4F are not of interest because a large containment bypass (release category S6) already exists. PDS 3FP and 4FP are not of interest i because a release path (S2) already exists. The above frequency of early high pressure core melt is conservative for three reasons.
- 1. There are still potential recovery actions to ensure wet S/Gs not yet i considered. For example, the dominant sequence (3D-1 on RMEPS page 3-58) involves a loss of main feedwater transient and f ailure of the solid
, state protection system (mean annual frequency = 8.3 x 10-6). Solid state protection system failure is assumed to result in no auto initiation of safety equipment such as emergency feedwater. Emergency feedwater and other safety equipment can be started manually from the Control Room but these actions were conservatively neglected in the PSA.
- 2. There are early high pressure melt sequences with wet S/Gs in some of these plant states (i.e. emergency feedwater available and 4 operating). For example, sequence 3D-4, a non-recovered station blackout and the operators don't depressurize S/Gs. Other such sequences include 4A-2 and 3, and 3A-29. (Exclusion of these sequences would not signif-icantly reduce the total estimated above).
- 3. No credit is taken for , operator actions to manually depressurize which reduces pressure.(i.e.",,not' high pressure core melt). Also, there is
! some chance the PORVs may fail to close.
I The analyses indicate that the steam generator tube temperatures and the hot
- legs are both well below the levels where creep rupture would occur. In addition, it is likely that the operator would be instructed to manually depressurize the RCS when the core outlet temperature exceeded 12000F.
This action would decrease the primary system pressure such that the stresses on the steam gemator tubes and the hot legs would both be l reduced to levels less than that observed under normal operation. Consequently, the estimated conditional f requency that the plant conditions for the hypothesized creep rupture of steam generator tubes would occur is dependent upon the likelihood that the operator would f ail to manually depressurize the primary system. With modified procedures and adequate training, the freguancy of operator failure to depressurize could be reduced below 10-4 to 10-3 per demand. This would lower the f requency of creep failure potential conditions to the 10-7 to 10-8 range. Given the occurrence of these conditions, the probability of creep rupture is a matter of our stata of knowledge about the laws of physics that govern heat transfer to the SG tubes as opposed to a statement about the relative frequency of a random process (which we make about the likelihood of achieving the necessary plant conditions in terms of
' frequency"). Due to cooling of the steam generator tubes by the
) secondary side steam, the SG tube temperatures are well below (approximately 200-3000F) that required for creep rupture. Due to the very strong j dependence of creep rupture on temperature, this consideration means that creep rupture would not occur, even if high pressures are maintained. .l .
Despite the uncertainties in our models and data, we are quite confident that errors as large as several hundred degress on the low side of the correct valve are not credible. To express our confidence in the models and data, we assign a 99% chance that failure of SG tubes will not occur before reactor vessel melt through or piping nozzle failure. g The ef fect of a pre-existing steam generator tube leakage (within technical L specification - 1 gpa per steam generator), would be a small flow rate compared to that created by the 50-gallon per minute flow assumed for the t pump seal LOCA case. Since the major influence of an additional flow would be to provide higher temperatures in the steam generators, this is bracketed by the results of the pump seal LOCA case which showed no major influence on the steam generator tube temperatures. 4 Given the assessment of e.he primary system response, potential for creep rupture of the steam generator tubes is extremely unlikely. Consequently, no specific calculations were carried out for the actual release categories since it is not risk significant. If specific questions were to arise regarding such releases, these would be analyzed with the MAAP 3.0 PWR code. If one postulates creep rupture f ailure of steam generator tubes, the pressure inside the previouly dried out and isolated stream generator secondary side would increase until the steam generator PORVs setpoint is. reached at which time these valves would lif t and modulate until restor vessel melt through occurs and the RCS depressurizes into the containment. There are no existing release categocies/ source terms analyzed for Seabrook or other - PWR plants that adequa'tely represent this scenario. 'During the periods of SG PORV opening, there wo,uld be a high leak rate bypass condition directly from the RCS to outside the containment. However, after vessel melt through, the leak rate out this path would be low corresponding to any low pressure leakage through the reclosed PORV. This leak path could be enhanced if the SG safety valves also lif t and fail to reseat properly; however, it is believed unlikely that the safety valve setpoint would be reached. If a source term were developed for this scenario, it would probably resemble S2 with the addition of an carly puff to cover the relief valve opened period of the release. Depending on the sequence, there may or may not be a long term overpressurization component to the source term. l' I 1 i
?
Figure.1 CREEP FRILURE' 0F STERM GENER ATOR TUBES 4 From SSPSR High Pressure Core Melt tulth Dry Steam Generators 1 P Recover IDater to YES Steam Generator , NO
- 1 P h
Manual RCS YES m S/G Depressurization " Success
- ~
d k
. NO 1 P Steam Generator YES Tube Integrity Maintained NO II Containment
- Bypass i
... .== ww 7" LU y
l g j ?
'33r M j SEABROOK STEAM GENERATOR INTEGRITY AN ALYSIS
. ~hlartin G. Plys Marc A. Kenton -
' Robert E. Henry Fauske & Associates, Inc.
Burr Ridge, Illinois Robert Lutz - Peter Kirby Westinghouse Electric Corporation Pittsburgh, Pennsylvania November,1986
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TABLE OF CONTENTS Page
1.0 INTRODUCTION
. . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.1 Background . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.2 Method . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 2.0 ACCIDENT SEQUENCES STUDIED . . . . . . . . . . . . . . . . . . . 2-1 2.1 Sequences Without Operator Actions . . . . . . . . . . . . 2-1 2.2 Sequence With Operator Action . . . . . . . . . . . . . . . 2-1 2.3 Uncertainty Analyses . . . . . . . . . . . . . . . . . . . 2-1 3.0 SEABROOK SPECIFIC INFORMATION . . . . . . . . . . . . . . . . . 3-1 3.1 MAAP 3.0 Parameter File . . . . . . . . . . . . . . . . . . 3-1 3.2 Operator Actions . . . . . . . . ... . . . . - . . . . . 3-1 4.0 RESULTS ............................4-1 4.1 No Ope ra to r Ac ti o n s . . . . . . . . . . . . . . . . . . . . 4- 1 4.2 Operator Actions . . ...................4-5 4.3 Influence of Uncertainties . . . . . . . . . . . . . . . . 4-13
5.0 CONCLUSION
S . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 ' 3 j
6.0 REFERENCES
. . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 APPENDIX A: MAAP 3.0 Seabrook Parameter File . . . . . . . . . . . . . A-1 i APPENDIX B: Steam Generator Tube Integrity Analysis . . . . . . . . . B-1 B.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . B-1 B.2 Tube Degradation . . . . . . . . . . . . . . . . . . . . . . B-1
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- - - . - - - _ . .- - - - - - + ---
I 11 - TABLE OF CONTENTS (Continued) i l Page B.2.1 Tube Properties . . . . . . . . . . . . . . . . . . . B-2 B.2.2 Thinning / Cracking Type Defects . . . . . . . . . . . . B-2 B.2.3 Tube Denting . . . . . . . . . . . . . . . . . . . . . B-5 8.3 Cree p Ru ptu re . . . . . . . . . . . . . . . . . . . . . . . . B-9 B.4 Summa ry . . . . . . . . . . . . . . . . . . . . . . . . . . . B-12 B.5 References . . . . . . . . . . . . . . . . . . . . . . . . . B-14' APPENDIX C: Estimation of Steam Generator Tube Wall Temperatures . . . C-1 e e a e t 4 2 . e 6 0 4 b
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- iii - .
LIST OF FIGURES Figure No. Page 1-1 Hot leg and steam generator natural circulation flow . . 1-2 4-1 Base case primary system pressure . . . . . . . . . . . . 4-3 4-2 Base case primary system structure temperatures . . . . . 4-4 4-3 Base case and seal LOCA case primary system Cs! re-tantion . . . .- . . . . . . . . . . . . . . . . . . . . 4-6 4-4 Base case and PORY case primary system pressure com-parison . . . . . . . . . . . . . . . . . . . . . . . . 4-8 4-5 Base case and PORV case steam generator inlet plenum gas temperature comparison . . . . . . . . . . . . . . . 4-9 4-6~ Sase case and PORY case steam generator average tube temperature comparison . . . . . . . . . . . . . . . . . 4-10 4-7 Base case and PORY case in-vessel hydrogen production compa ri son . . . . . . . . . . . . . . . . . . . . . . . . 4- 11 4-8 Base case and PORV case primary system Qsl retention . . 4-12 4-9 Base case and high eutectic temperature case hottest core node temperature comparison . . . . . . . . . . . . 4-14 4-10 Base case and high eutectic temperature case steam generator inlet plenum gas temperature comparison . . . . 4-15 4-11 Base case and natural circulation uncertainty cases steam generator inlet plenum h comparison . . . . . . . . . gas. temperature . . . . . . . . . . . . . . 4-17 4-12 Base case and natural circulation uncertainty cases steam generator tube temperature comparisons . . . . . . 4-18 , 4-13 Base case and steam generator PORV case primary sys-l tem pressure comparison . . . . . . . . . . . . . . . . . 4-21 4-14 Base case and steam generator PORV case water level comparison . . . . . . . . . pressurizer . . . . . . . . . 4-22 4-15 Base case and steam generator PORY case steam gen-erator inlet plenum gas temperature comparison . . . . . 4-23 4-16 No blockage case care-upper plenum flow . . . . . . . . . 4-24
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, LIST OF FIGURES Figure No. Page 1
4 17 No blockage case upper-plenum gas temperature . . . . . - . 4-25 4-18 No blockage case steam generator plenum gas temperature . . . . . . . . . . . . . . . . . . . . . . 4-26 B-1 High temperature tensile properties of annealed l (1600*F/1 hr.) hot-rolled plate (B-1) . . . . . . . . . . B-3 B-2a Burst data for 0.875 x 0.050 uniform thinnin mens - defect length variation (B-1) . . . .g .speci- . . . . . . B-4 B-2b Burst data for 0.875 x 0.050 uniform thinning speci- , mens - defect depth variation (B-1) . . . . . . . . . . . B-4 B-3a Burst data for 0.875 x 0.050 EDM slot specimens - defect depth variation (B-1 ) . . . . . . . . . . . . . . . B-6 B-3b Burst data for 0.875 x 0.050 EDM slot specimens - i defect length variation (B-1)
. . . . . . . . . . . . . . B-6 B-4 Burst pressure data of 0.875 x 0.050 uniform thinning specimens with and without denting (B-1) . . . . . . . . . B-7 B-5 Creep and creep-rupture comparisons (B-4) . . . . . . . . B-10 B-6 Master creep rupture curve for 316 stainless steel, taken from Ref. (B-7) . . . . . . . . . . . . . . . . . . B-13 C-1 Natural circulation flows on the inside and outside of a tube carrying fluid from the inlet to the out-let plenum . . . . . . . . . . . . . . . . . . . . . . . . C-2 l
p
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-v-LIST OF TABLES Table No.
Pace 4-1 Blackout Base Case Key Event Times, Without and With Seal LOCA . . . . . . . . . . . . . . . . . . . . . . . . 4-2 4-2 Operator Action Figures of Merit and Event Times
. . . . 4-7 4-3 Key Event Time Comparison for Stea Open Case . . . . . . . . . . . . m Generator PORY
............ 4-19 B-1 Comparison of Burst Pressures of Elliptically Wasted
.875 00 x .050 Wall Tubin (B-1) . . . . . . . . . .g With and Without Denting
................ B-8
. B-2 Creep Rupture Data for Inconel-600. Hot Rolled, and Annealed at 1600*F ................... B-ll g e 8
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1-1 .
1.0 INTRODUCTION
1.1 Background
A frequently-studied' postulated severe LWR accident is the station blackout sequence (TMLB). In this sequence, all off-site and on-site AC power are assumed to be lost. When analyses of this sequence are performed, high reactor vessel upper planum temperatures are computed (1). This is caused by two factors, the heating of the upper plenum due to core-upper plenum natural circulation, and that caused by volatile fission products (chiefly iodine isotopes) which are released from the core and deposit in the upper plenum. Experiments performed at Westinghouse (2) in a one-seventh scale test facility have shown that these high upper plenum temperatures will cause natural circulation to initiate between the upper plenum and steam genera-tors. This process, which 'is, shown schematically in Figure 1-1, would result in an increase in: steam generator inlet plenum temperatures. Conse-quantly, concern has been expressed that temperature-induced failures in the tubes or in the tube-to-tubesheet welds could result in a discharge of fission products to the secondary sides of the steam generators. Continued blowdown of the primary system to the secondary system through such failures could lift the steam generator 'afety s valves, bypassing the containment. The principle purpose of this study was to assess the likelihood of these
, failures in the Seabrook plant, given a station blackout sequence.
I 1.2 Method Several variations on station blackout sequences were simulated using the Modular Accident Analysis Program (MAAP) version 3.0. This code is an integrated thermal-hydraulic and fission-product analysis code for severe [ accidents developed in the IOCOR program. Earlier version have been used by l utilities, NSSS vendors, and consultants for approximately 4 years. The l code has been verified and has been extensively benchmarked against experi-mental data, actual plant transients, and detailed code calculations (3_). i i _ . _ _ . . . . ~ , . , - . , _ , . _ _ - _ , _ . . _ _ . __,_..,_.__,,_,_,.,m_,__._,,.,_--.,,,m_m.,m,.,,w,-.--_,_.,.-.-
1-2 l
- Ngo 7M (TOTAL FLOW) [
'OUT" TUBE <=
"BACK' TUBE :
e TSEC.. = J \hl r C'HL
/ T" CO TUP WHL H C
%S I
Figure 1-1 Hot leg and steam generator natural circulation flow.
. - . . , .. -,_.- - - - -. ..,..---- .-.....- ,-.---.~...-, - - --- - - _ -.-.- ._. - - _ - - - - - - - . - .
1-0 . MAAP is particularly suited for this study since it contains fully integra+ed models for:
- a. Core overheating, oxidation and melting.
- b. Natural circulation between the core and upper plenum, between the upper plenum and inlet steam generator plena, and between the inlet and outlet plena of the steam generators. Calculations using the natural circulation models have compared well to the Westinghouse one-seventh scale tests (1,2_).
- c. Fission product release, transport, deposition, and revapori-zation.
4 In addition, MAAP allows arbitrary operator actions to be applied so that' the efficacy of the likely operator responses can be studied. W
- l l
l l l l l
, .. _ ___ . . _ _ . i -
~ ~ ~ ~
L 2-1 . 2.0 ACCIDENTSE00ENCESSTUDIED In all the sequences presented below, it is assumed that off-site and on-site AC power remains lost indefinitely. In such a case, the emergency response procedures (ERPs) (4_) instruct the operators to initiate auxiliary feedwater (AFW) using the turbine-driven pump. As long as sufficient AFW flow can be maintained, decay heat would .be removed and the accident would not progress further. For the cases studied here, it has been further assumed that all AFW has been lost. .This reflects a major conservatism in the analysis. 2.1 Seouences Without Operator Actions Two sequences were studied in which no operator actions were credited. In the base case, no additional failures were assumed. In the second case, loss of cooling to the main coolant pump seals was assumed to result in a leak area corresponding to a flowrate of 50 gal / min of water per pump.
' This leak area is the same as Ithat assumed in the IOCOR program (5_). '
2.2 Sequence With Operator Action The Seabrook ERPs (6) specify that when AC power is available and all other measures have been attempted, if core thermocouple temperatures exceed 1200*F. the pri try system should be depressurized using the pressurizer PORVs. This enables the accumulators to be used to inject coolant into the core. While this action is not invoked for blackout sequences, it was considered that this action might well be recomended by the Technical Support Center. Accordingly, this action was simulated in one case. 2.3 Uncertainty Analyses Several variations on the base case were studied to investigate the impact of uncertainties in phenomenological parameters and accident pro-gression. In these cases, parameters were varied in such a way as to increase the potential for high steam generator temperatures:
l L 1 2-2 I
- a. A high (3000'K) core melting temperature was assumed compared to the nominal value (2500*K). This delays the onset of core geome-try degradation and thus enhances the potential for the core to heat the rest of the primary system.
- b. Low values of axial and cross-flow friction factors were assumed in the core modelling. This also tends to maximize convection of heat out of the core.
- c. Lower values of steam generator natural circulation flow WSG (see Figure 1-1) relative to hot leg natural circulation flow W were HL assumed to minimize cooling of the steam generator inlet plenum by flows from the outlet plenum. This was accomplished by choosing lower limit values of the number of steam generator tubes partic-ipating in the flow from the inlet to the outlet plena, guided by observationsintheWestinghcuseexperiments(2).
- d. A run was made where'1't was assumed that the steam generator PORVs ,
stick open. Depres.suMzing the secondary side tends to decouple the tubes from the shell after the steam generator dries out. This could potentially increase tube temperatures.
- e. A run was made in which it was not assumed that coolant channel blockage occurs in a core node when melting begins in that node.
MAAP ordinarily malies such an assumption to represent the rapid reduction in flow area that would occur after melting comences. In this run, complete blockage of the flow area in a node was only assumed to occur if the node completely filled with molten ma-terial. By allowing continued core oxidation and core-upper plenum flow, this assumption maximizes the potential for the core , to heat the rest of the primary system.
, -. _._ . - - - - --- - -~ ~ ~ ~
3-1 , s 3.0 SEABROOK SPECIFIC INFORMATION 3.1 MAAP 3.0 Parameter File Since MAAP 3.0 was used for this analysis, it was necessary to develop a Seabrook-specific parameter file. The major part of this effort was completed by using the MAAP 2.0 Seabrook parameter file which had been developed in support of the Emergency Planning Zone study (7_). i Additional changes were made to the Seabrook MAAP 2.0 deck to reflect the' model revisions in MAAP 3.0. The major additions and changes were in the areas of reactor ves:..I her.t sinks, peaking factors, fuel rod ballooning data, model parameters, and auxiliary building data. The new MAAP 3.0 data for this analysis was derived from Seabrook plant drawings, data deleted or modified from MAAP 2.0, the Westinghouse IMP data base, and other generic and Seabrook speciffe documents., A copy of the Seabrook MAAP 3.0 parameter file is included in the Appendix. . E' 3.2 Operator Actions In order to get a complete picture of steam generator tube response during a station blackout transient, operator actions were also considered. The potential operator actions were derived from ,the Seabrook ERPs (4,6). j ~ ' In the event of a station blackout, the operators are instructed to 1 initiate auxiliary feedwater (AFW) operation using the turbine-driven pump. ] This action was not modelled in the MAAP 3.0 runs, since it would result in li the removal of decay heat and would effectively prevent core uncovering from occurring. In sequences with AC power available, the symptom-oriented procedures (j,) call for depressurizing the primary system using the PORVs if core temperatures exceed 1200*F and all other means for cooling the core have i l l
__ _ _ . -. - .. . ~ . . - - - - - - ~ ~ ^ - ~ ~ ~~ lf 3-2 been attempted. While this action is not invoked in the procedures if AC power is not available, it was considered that such an action might well be recomended by the Technical Support Center in a blackout in order to obtain flow from the accumulators. For this reason, this action was simulated in one of the comparison cases. t 4 S O 9 4 9 9 h
4-1 , l 4.0 RESULTS 4.1 No Operator Actions Two cases were run with no operator action, the base case and the base case with a pump seal LOCA at 45 minutes after initiation. Since the results of these sequences are quite similar, the base case.will be de-scribed here and only differences will be noted. Major events are listed in Table 4-1. After the blackout is initiated, the steam generator inventory begins to be depleted. As the water level in the steam generators de-c eases, heatup and expansion of the primary coolant leads to the pressur-izer " going solid", i.e., completely full of water. The quench tank rupture disk then breaks after discharge from the pressurizer overpressurizes the tank. The steam generators dry out completely soon afterwards. 6 The seal LOCA case has slightly different timing until steam generator dryout because of the loss of primary system inventory through the failed seals. Thus, in this cade. the pressurizer goes solid later because there is less~ water which can expand. Steam generator dryout, on the other ha'nd, is determined simply by integrated decay power and the available secondary side inventory and differs little from the base case. Loss of primary coolant leads to eventual core uncovery, meltdown, and vessel failure. The primary system remains at high pressure until vessel failure as dictated by the pressurizer relief valve setpoints (Figure 4-1). Strong natural circulation occurs between the core and upper plenum after the core uncovers; this in turn sets up circulation between the upper plenum and the steam generators. Fission products leave the core during the heatup and can migrate through the upper plenum to the hot leg. The circulation and fission product transport are affected by temperature differences, an'd feed back to influence region temperatures (Figure 4-2). After vessel failure, the primary system blows down, the pump bowls clear, and the accumulators empty. Most core debris and water in the lower plenum at that time are entrained into the lower compartment, which is o O
__ . _ , . . .. _ . m_-._ -- - - - - 4-2 Table 4-1 BLACK 0UT BASE CASE KEY EVENT TIMES. WITHOUT AND WITH SEAL LOCA Event Base Case Seal LOCA
. Initiation 0.0 0.0 Seal LOCA -
2714. Pressurizer Solid 4829, 5920. Quench Tank Disk Rupture 5486. 5944. Steam Generator Dryout 5527. 5517. Steam in Pressurizer 6319. 6269. l Core Uncovery 7280. 7124. Vessel Failure 11648. 11054. Accumulator Deplet.ed ' 11700. 11102_.
"End of Simulation '
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_ . . _ _ _ _ . . _ - - - - - - - ~~~~ " . . _ _ _ . 4-5 connected to the reactor cavity. Airborne fission products are swept from the primary system at that time. With the pump bowls clear, natural circu-lation can occur throughout the primary system, and some long-term revapor-ization of fission products is possible. In the case. of the seal LOCA,
~ slightly more Cs! leaves the primary system due to leakage through the pump seals and the lower system pressure (Figure 4-3).
4.2 Operator Actions Operator action to depressurize the primary system by opening the pressurizer relief valves was assumed to occur when the maximum core temper-ature reached 1200'F, as discussed earlier. This resulted in complete blowdown of the primary system before bottom head failure. Thus, the time during which steam generator tubing was exposed to both high pressures and high temperatures was significantly reouced by operator action. Key events and figures of merit are surrenarized in Table 4-2 and com-
, , pared to the base case., It, .can be'seen that not only is the peak steam generator inlet plenum gas temperature lower after the operator action, but the primary system pressure is lower as well.
The action occurs at 8000 seconds, about 15 minutes after core uncovery, and leads to accumulator dis-
. charge well before bottom head fail ~ure.
This discharge occurs over a period of several thousand seconds, as seen by the gradual decrease in primary systempressure(Figure 4-4). The steam generator inlet plenum gas tempera-ture (Figure 4-5) is high only after depressuri =ation, and the tubes them-selves are relatively cool (Figure 4-6). While, as shown in Figure 4-6, tube temperatures continue to increase after vessel failure, simulations j continued for a longer time than those presented here show that the tempera-j j tures do not increase much beyond 700'K due to heat losses to containment. Since the differential pressures are much lower after vessel failure, these -f moderate temperatures are not limiting. The accumulator discharge causes only slightly more hydrogen production (Figure 4-7) through availability of steam. However, the large flows caused by steaming of accumulator water ? lush fission products from the primary system into the containment (Figure 4-8).
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4-7 ' Table 4-2 OPERATOR ACTION FIGURES OF MERIT AND EVENT TIMES Base PORV Seal Case 9 1200*F LOCA
- 1. Feak SG inlet plenum gas tem- 858. 682. 865.
perature while primary system isathighpressure(*K)
- 2. Primary system pressure at \
17.1 1.83 14.9 bottom head failure (MPa)
- 3. Peak SG inlet plenum gas 858. 843. 865.
temperature (*K) 4
- 4. Primary system pressure at 17.1 4.23 16.9 time of #3 above (MPa) '
- 5. Time of peak T #3 (sec) 9230. 9290. 9000,
- 6. PORYopen(sec);
8005. --
- 7. Time accumulators ' depleted (sec) 11701. - 10643. 11102.
- 8. Vessel failure (sec) 11648. 12601. '11054.
- 9. Approximate peak SG tube wall 750 low 760 temperature (see Appendix C)
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4-13 4.3 Influence of Uncertainties Five sensitivity cases are presentC aere which show the influence of high core melting temperatures, enhanced core to upper plenum natural circula' tion, reduced steam generator tube circulation, blowdown of the steam generator secondary sides, and neglect of core blockage phenomena. Each of those factors individually can increase hot leg and steam generator plenum temperatures relative to nominal cases. A high UO 2 -Zr eutectic temperature of 3000*K (versus the nominal 2500*K) was used for uncertainty analysis of the base case presented above. In the base case, vessel failure occurred at about 11,600 seconds (Table 4-2), while in this case vessel failure was delayed until 12,300 seconds. This is because more time is required to reach the higher eutectic tempera-ture and cause melting. Other thermal-hydraulic behavior and event t'i Ifng is similar to the base case, with exceptions relating to higher core temper-atures. I The. driving ' potential for natural convection can be compared between this and the base. case by considering the hottest core node tempera-h ture (Figure 4-9), which s' ows the effect of the input eutectic temperature . Only a small impact on the ' steam generator inlet plenum gas temperature is caused by , this parameter change (Figure 4-10), with' a slight delay in reaching the peak values. Thus, considering an uncertainty of 500*K in the core melting temperature, a difference of only tens of degrees in steam generator temperatures results. This is because the heatup rates in the core are very much more rapid than the heatup of the steam generator plena once temperatures are high enough to cause rapid Zircaloy oxidation. Natural circulation between the core and upper plenum was enhanced by lowering the axial and cross-flow care friction factors in one sequence . Circulation between the steam generator tubes and the inlet plenum was reduced by lowering the fraction of tubes carrying flow out of the plenum in another case. Each case acts to increase the steam generator inlet plenum temperature over the base case value. I changes on sequence timing: There is a small effect of these base case, 12100 seconds vessel failure occurs at 11600 seconds in the seconds in the high core circulation case, and 11200 in the low steam generator circulation case. In the high core
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4-16 circulation case, more heat transfer occurs to the steam generators and vessel failure is slightly delayed. In the low steam generator circulation case, less heat transfer from the plenum to the tubes occurs, and vessel failure is slightly hastened. The steam generator inlet plenum gas tempera-ture (Figure 4-11) is significantly higher (about 150*K) only for the steam generator circulation case ("SG NC"), while it is only slightly higher for the core circulation case (" CORE NC"). Tube temperatures (Figure 4-12) follow the same trend. It should be noted that the low steam generator circulation case, which assumed that only 10 percent of the tubes carried flow in the "out" direction, is in all likelihood unrealistically severe . More typical extremal values of the flow split observed in the Westinghous experiments (about 20 percent of the tubes carrying out flow) result in onl about a 900*K peak inlet plenum gas temperature. A blackout case with stuck open steam generator relief valves on all the units was run to minimize the heat transfer capability of the tubes, and thus increase the steam generator inlet plenum temperature. In tne base case, steam on the secondary side serves as an efficient heat sink for the small recirculation flow 'through the primary side, and also couples the tube mass with the steam generator shell heat sink. When~ the secondary side is blown down, and this heat capacitance is impaired, the tube outlet tempera ture is higher and thus the mixed mean steam generator inlet plenum temper ture is higher. ' The early behavior of this sequence differs dramatically from that of the base case due to enhanced cooling by the secondary side early in the transient. Comparing events (Table 4-3), the steam generators go dry much earlier in the steam generator PORV case, and the primary system cools down l enough that the pressurizer is drained due to contraction of the coolant. Later, of course, with the heat sink lost, primary system fluid reexpands and the pressurizer goes solid. Core uncovery and vessel failure occur slightly earlier in the steam generator PORY case because the overall integrated heat removal is lower with the stear, generators blown down.
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4-19 Table 4-3 KEY EVENT TIME COMPARISON FOR STEAM GENERATOR PORY OPEN CASE Event Base Case Shp,PORV Initiation 0.0 0.0 SG PORV Open - 0.0 SG Dryout 5527. 2738. Pressurizer Drained -- 350. Pressurizer Refilling -- 2000. Pressurizer Solid 4829. 4205. Quench Tank Disk Rupture 5486. 4937. Pressurizer Has Steam 6319. 59'20. Pressurizer Empti. 7000. 7400, Core Uncovery - 7280. 6626. Vessel Failure 11648. 11045. 1 u I 1 o 9 _ , , , , , , ------'# ~'"'^~'
_ _ _ _ . - -- _ . _ . _ . . . . . - - ~^ ^ ~ ^ ' 4-20 Primary system pressure (Figure 4-13) shows the cooling achieved while the secondary side blows down, resulting in a 10 MPa transient over 4000 seconds. The pressurizer water level (Figure 4-14) illustrates the con-traction of primary system coolant during the cooldown, and later reexpan-sion. Since the core uncovers earlier in the steam generator PORV open case, high gas temperatures in the steam generator inlet plenum are shifted to slightly earlier times (Figure 4-15). The initial high peak, prior to degradation of core geometry, is the same in each case both in magnitude and rise behavior. Thereafter and before vessel failure, the inlet plenum gas
~
temperature is about 50'X higher for the steam generator PORV case. How-ever, it is still below 800*K for most of the high temperature period. The last uncertainty case assumed that nodal core flow channel blockage does not occur at the onset of melting in the node. MAAP normally assumes such blockages to represent the rapid reduction in flow area that would
. occur after melting begins.
In this case, complete blockage was credited only when the geometry would allow no flow, i.e., when the node was com-plately full of refrozen eut.ectic. As shown in Figur_e 4-16, a rapid reduc-tion in core-upper plenum flow still occurs after the beginning of accel-erated oxidation. While the flow continues at relatively low values, heat removal from the upper plenum due to the hot leg flows about equals heat convected from the core and that due to fission product heating, resulting in a stabilization in upper plenum gas temperature (Figure 4-17). Steam generator inlet plenum gas temperatures (Figure 4-18) also stabilize. The peak sustained plenum gas temperature, which occurs just before vessel failure (12,000 secs) is about 1060'K. Thus, even when blockage in the core is essentially neglected, only relatively moderate increases in plenum gas temperature are seen over the base case. O _ ___-._____ - - - - - - - - - ' ' ~ ' ~ ~
e.
; , aanSE SEABROOK A!M NRV TMLB COMPARISON '
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5.0 CONCLUSION
S i i Various blackout sequences have been analyzed for Seabrook which co l a range of potential operator actions and phenomenological uncertainties. An important result of the base case is that for the period when the pr system is at high pressure, the peak steam generator inlet plenum gas temperature is only s 850*K, and average tube ' temperatures are well below 700*K. h Operator action to depressurize the system can successfully reduce I' the system pressure prior to the ' time high temperatures are reached in the steam generator. When uncertainties in circulation between the core and upper plenum or in the core melting temperature is considered, negligible increases in the steam generator temperatures are observed. Similarly, only slight increases in temperature were obtained in a case where the seco sides of the steam generators were depressurized. For an unmitigated case, 2 considering a large variation in the steam generator inter-plenum circula-tion indicated that gas; temperatures can briefly reach about 1000'K in the steam generator plenum. ~ 'Sim.ilar temperatures (s 1060*K) are achieved w the blockage of core node. coolant channels due to melting is neglected . Both of these~latter cases are considered unrealistically extreme. MAAP does not contain a detailed model for the change in tube wall temperature as one leaves the inlet plenum and moves toward the outlet
.computed).
plenum of the steam generator (i.e., only average wall temperatures are However, by using the average secondary side gas temperature and the temperature of the primary side gas entering the tubes, the wall tempe atures can be estimated by knowing the value of the heat transfer coefff-cients on the primary and secondary sides. As shown in Appendix C, these heat transfer coefficients are approximately equal. By equating the heat j ' flux convected to the tubes by the primary side gas to that conv from the tubes by the secondary side gas, the analysis in Appendix C leads j to the conclusion that the peak tube wall temperature will be the average h the inlet gas temperature and the average secondary side temperature . In the worst case discussed above, this results in a peak tube wall temperature of approximately 850*K and is only 750*K in the best estimate case. As
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6.0 REFERENCES
' (1) Fauske IDCOR Report & Associates. 85-2 (July,1985). Inc., " Technical Support for Issue Resolution", (2,) W. A. Stewart, et al., " Experiments on Natural Circulation Flows in Steam Generators Ouring. Severe Accidents", Proc. Inter. ANS/ ENS Topical (1986). Meeting on Thermal Reactor Safety. San Ofego, California (3_) R. E. Henry, " Benchmarking of Severe Accident Codes: How Should It be on Thermal Reactor Safety San Diego, California (1986).D (4_) New
" LossHampshire of All AC Power", Yankee (May Station 16,1986).Emergency Operating Procedure ECA-0.0, (5)
Comonwealth Edison Company. " Zion Station Integrated Containment Analysis". IDCOR Report 23.12 (1985). (6) New Hampshire Yankee Station Emergency Operating Procedure FR-C.1,
" Response to Inadequate Core Cooling", (May 16,1986). -
(7) Pickard, Lowe l and Emergency ,'Plantiing Study", PLG-0432 (December,1985).an l ,
A-1 l I l I l APPENDIX A __MAAP 3.0 Seabrook Parameter File O O e O S e 4
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e l i l l E t I 1
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st s** ssasssastamasassstastassassanss**** 3:SENERAL INFORMATION: 83 8!I.FORTHEGEOMETRICALRELATIONSHIPBElWEENVARIOUSCONTAINMEN as SYSTEM MANUAL N0DEse SEE THE REDIUN SUBROUTINE WRITE-UPS IN VUL 2 0F USER'S at s ss FOR LARGE: DRY CONTAINMENTS: SPECIFY ZERO VULUME FOR ICE-CONDENSER AND
** PARAMETERS TO BE IGNOREDUPPER PLENUM-fMIS CAUSES ALL OTHtR UPPER PLENUM AND ICE CONDENSER
** NOTE OUTER WALLS IN COMPARTMEN1S A AND D SEPARATE THE CONTA!hMEN1 FROM 832.
** THE ENVINUNMENTI THE UUIER WALL IN CUMPI B SEPARATES CUMPTS !$ AND Di at THE OUTER WALLS IN COMPTS I AND U (ICE CONDENSENS ONLY) ANE NUT MUDELED
** SINCE THESE WALLS ARE INSULAIEDI THE UUTER WALL IN CUMPT C.!S ASSUMtD
** TO THEBE INNER ADIABATIC FACE) ON ITS FAR SIDE (IE NU HEAT LOST FkOM THE SIDE 0PPOSITE as TO REPRESENT A FREE STANDING STEEL CONTAINMENT WITH A SHIELD NJILDING, as ss TREAf THE SHIELD BUILDING.AS THE WALL AND THE CUNTMT PRESSURE BOUNDARY ss AS A ' LINER'l ENTER.THE ' GAP' DISTANCE BE1 WEEN THk TWO WHERE CALLED FOR '
383.
** INTERNAL OR INTERIOR WALLS ARE WALLS TOTALLY CONTAINED IN A COMPT st PROPERTIES (THERMAL CONDUC ETC.) UF INTERIGR WALLS IN A AND 8 ARE sa ASSUMED THE SAME AND ARE ENTENED IN THE LOWER COMPT SECTION
** IF (AS IS USUALLY IHE CASE) YUU MUST LUMP WALLS OF SEVERAL THICKNESSES TOGETHER, YOU SHOULD LUMP DNLY RELATIVELY THILK WALLS (EG GkEATER THAN st
- ABUUT I FOOT OR .3 METER IN THICKNESS) AND ENTER IHE THICA> FESS s OF THE TNINNEST WALL CREDITLD
** .\
**4.
as DECK REFEks TO THE FLOOR (AND VERTICAL WALLS IN ICE CONDENSER PLAN)S)
** IMAT SEPtRATES THE UPPER CUMPARTMENT FRUM THE COMPARfMENTS LOWER IN THE CONTAINMENT at 3:$. TWO WAYS TO HANDLE CONTAINMEN1 FAILUkEI
** A. MECHANISITIC MUDEL:
3:
*: ENTER 0 FOR THE FAILURE PRESSURE (ACOMPT NO. 34) ALSO SUPPLY: .
3: (I) CONCRETE: SUPPLY ALL THE MAIERIAL DATA: CuMCRETE PARAMS 13 42. ETC.
** (2) FREE STANDING STEEL SHELL: ENTEN THE WALL THICKNESS IN THE IJPPER AND ANNULAN CUMPARIMENTS IN fME ' LINER
- THICKHESS ENTRIES as
** AND SUPPLY ONLY THE LINER MATERIAL PRUPERTIESI THE NUMBER OF
*s TENDONSeAND AMUUNT UF REBAR SHUULD BE SET TO 4ERO IN THIS CASE B. SIMPLE MODEL:
st SUPPLY ACOMPT NO. 34 AND 3/I FAILURE AREA ENTERED AS MUDEL PARAMATER st NO.23NEED MUT SUPPLY INE OTHER PARAMtTERS s
**4. SEDIMENTATION AREA' IS THE TOTAL UPWARD-FACING AkEA IN A GIVEN COMPARTMENT UPON WHICH FISSIUN PRODUCT AEROSULS CAN SETTLEI THIS g SHOULD INCLUDE (WHERE APPROPRIATE), FLOORfe CABLE TRAYSe EUUIPMENI ETC
**7. AS DESCRIBED IN THE
- CON 1ROL SECTIONe TW. AUXILIARY BUILDING MODELS ARE as
** ACTIVATED BY SUPPLYING A NUNZERO NO. OF AUX. NuotS TO BE MODELLED.
THE MODEL CAN BE RUN SIMULTANEOUS WITH A RUN OF THE CON 1MT AND FkIMARY
e -- ~~~ '" ' A-3 , i u s'
+
st SYSTEM MODCLS
** GRe BY SUPPLYING A NUNZERO INPUT FILE NO.e THE '
** AUX EAALIEft MODELSMAAP,RUN.ONLY CAN BE RUN USINU AN INPUT FILE OF T/H DATA FROM AN as s - 1 s*8; FISSION PRODUCT REMOVAL BY INERTIAL IMPACTION IS MODELLED ONLY IN 21-
*t UNE CUNTAINMENT COMPARTMENT. IN LARGE DRY'S SUCH PARAME1ERS
** SHOULD CHARACTERIZE GRATES WHICH ARE ASSUMED TO BE IN THE ANNULAR sa ~ IMPACTION AREA 0F ALL THE GRATES:ConPARTMENT. IF MORE fHAN
** OF THE GRATE ELEVATIOMS AND THE MAXIMUM FLOW AREA AT ANT st '
as IN ICE CONDENSER PLAN 1Ss
** THESE PARAMETERS NEVEN THOUGH LOCATED
** ; t
( tu FLOW AREAS AND STNAP WIDTHS IN THE ICE , s . >
,u BOX-SE ttf. THE UNITS FOR PARAMETER JMP'J13 ARE SPECIFIED BY EITHEN ' ' A aS1 (MET
, ** OR 882 (BRITISH) UNIT 3 CARD. ALL PARAMETERS FULLOWING SUCH A CARD ,
** ARE ASSUMED TO HAVE THESE UNITS UNTIL THf. NEXT UNITS CARD IS INLLUDE
** THUS A PARAMETER FILE CAM HAVE SECTIONS WITH DIFPERENT UNITSe DESIRED. IF
** TNC LAST UNITS CARD IM A PARAMETER FILE CONTROLS THE UNITS 1
s ** UF UTHER PROGPAn INPUTS IN TAPE S (EG START AND FINAL TIMES ETC.) I, AND
** THE UNITS TO K OUTPUT IN THE OU1PUT FILE AND PLOT FILES. >
** MkTRIC UNITS ARE M-KG-SEC-DEGREE MELVIN-PASCALS-Ma*2/SECsETC. I at BRITISH UefITS EXAMPLES: ARE FEEI-LBM-HOURS-DEGREE F-PSI-GPM '
3 st IN METRIC UNITSe FLOWRATES SPECIFIED TO Bk VOLUML1RIC SHOULD K ' .
** Ma*3/SECl OfHER FLOWNATES IE ALL THOSC NOT EXPLICI(LY STATED TO BE s
** VOLUMETNIC SHOULD BE KG/SECI HcADS SHOULD BE IN Mi PRESSURES IN PA) 4
> sa IN ENULISH THE UNITS ARE RESPECTIVELY GPMrLBM/HRerTe PSIA-NOTE TO MAAP/ SWR USENS-GPM IS USED IN MAAP/PWR INSTEAD'GF FTS*J/HR s
O.. ** IN THE
- TIMING SECTION ALWATS IN SECONDSfME UNLY EXCEP ' - '
**10.!M LARGEe DRY CDNTAINME01"S ' FANS' REFEkS TO FAN ta
** ANNULAR i COULERS ENILH TAK FUCTION FROM THE COMPT. AS SPECIFIED BELOW. UPPER COMPT AND DISCHARUE TO EITHER 1
THE LOWER OR sa PLANTS AME USED TO CHAPACfERIZE (HE tAIR PETURN s ,
, j FAi!LTHE SAME INPU
, i c
$33325tstsstSittststStsSSSSSSSttsssssssssssssstsstsststartstisssistSt$ts sa :h r "
83R ' s
** , (
COMPT 1a s* * * 'A* SUPPER COMPARTMENT (OR s t a t ts s t *** ***** stats x******2'** ** * * * * * *t t** *Is ** *5353 3 3 s 33333353833333388833833855333333383tS28888333333333333333333333358355833 01 2.13804 FREE VULUME : 02 1415. 03 141. AREA 0F REFUELING POOL 04 3&37. HEIUHT OF CONTAINMFMr SPWAY HEAD ABUVE BOT' 10M OF COMPARTMtNT 05 15394 FLOW AREA FROM UPPER COMPARIMEN) INTO ANNULAR COMPT CHARACT
** M CS ISTIC CROS*-SEC AREA 0F COMPT FOR BURN TiffE
** ANEA THE BURN f!ME IS THE SQUAkE ROOT OF THIS IVIDED BY TNC BURN VELOCI:Y 04 00 !
tt CURS HEIGHT IN REFUELING POOL 70.N. LOW OVERFLOW--N0knALLY
** 0 UMLESS CLASSICAL ICETUU CONDENSER ASSUME REFUELING PCUL DRAINS ARE BLOCKED (A SEQUENCE),
07 72122. THEN MAKE IT LARGE 08 .0357 SURFACE AREA 0F UUTER WALLS IN UPPER COMPAR(MENT Of 0.2S LINER THICKNE'S$ ON OUTEN WALL -
** CUTER WALL LINER u.m VSISTANCE--SEE N0lE IN
- LOWER COMPT FOR HOW TO MODEL FAEE STA.@IMU STEEL CONTMTS WITM A SHIELD ,
I ( fg , O _____.__________________.__________._.____.m_ . _ _ _ _
- . . . . _ . - - - '-= ~ ~ ~ ~ ~ ~~ '
,. A-4
** WALL 10 4.1 11 0.92 OUTkR WALL TOTAL THICKNESS as 12 0 157 WITH A LI fMERMAL CUNDUCTIVITY e THIS R OF UUTER WALL (FOR CONCNE1
- SPECIFIC 13 144 AT OF uuT ntS WALL.T0 THE CONCRETE PANT) 14 0 DENSITY OF OUTER WALL as 15 .11I14. ENTER A 1 IF THE QUIEN WALL IS SOLID WITH NO(WALLS SHIELD BUILDINO)e 0 FOR STEEL (IE A SIE CONCRETE WITH OR W 16 0 HALF AREA MODELED AS 1-D SLABS) 17 0 LINER THICKNESS UN INFERIOR WALLe IF ANT 0F IN1ERNAL WALLS 18 4. LINER GAP RESISTANCE IN INTERION WALL 19 I;990. THICKNESS DECK AREA OF INlERNAL WALLS 20 0 21 0 LINtR THICKNESS ON DECK o *2 4.2 LINER GAP RESISTANCE ON DECK
!3 0.92 DECK THICKNESS H 24 0.157 THERMAL CONDUCTIVITY OF KCK 25 144. SPECIFIC HEAT OF DECK 26 0 DENSITY OF DECK 27 3.43&D4 ENTER A MASS1 IF THE DECK IS SOLID STEEL:
METAL E0PT 0 FOR CONCRETE 28 29 - 2.0603D5
- 0. . EOPT HEAT TkANSFER AT.~A J
30 0. l AVERAGE DISTANCE OF THESE FROM THE CEILIN s* SOURCES SETHEIR RESPECTIVE ssCOMPARTMEdf-IF IN THE NO IGNITERS IGNORE LOWERS COMPARTMENTS 31-33 ANNULARE OR U WHICH CAN THEN PROPA 31 0. 32 0. i 1 u 33 34 at 0.
.90 NO.
DISTANCE OFFROM IGNITERTHE TOP OF$/IGN A TO THE SOURCES DECK ' IN D WHICH 35 .35 FRACTION REFUELING POUL OF'UFFEN COMPT SPRAY (VS. CONTINUING ON DIRECTLY WATER INT THAT RUNS- NTU T g g FRACTION OF WATER DRAINING 0010F REFUELING FOOL 1HA1 LO TO LOWER COMPT (REMAINING FRACTIUM RUNS INTO THE
** INPUTS FOR SIMPLE (FAILURE PRESSURE SUPPLIED) 34 RECALCULATED) 37 0 187. HUDELS FOR CUNTAIMtNT FAILURE--SEE GENER as FAILURE YNTER A 1 PRESSURE IF CONfMT FAILS OFINCONTAINMENT UPPER COMPT 10OR FOR0 TO USE DETAILE
*****stssas*****1:3*SsNEW283333333334f:
38 44.4 FAILURE IN THE ANNULAR CO 39 1.67D-4
** CONTAINMENT RADIUS FOR STkESS CALCULATIONS 40 41 0.D0 NCRMAL LEAKAGE IS ASSUMED TO COME FRO MASS OF WA1ER IN NEUIRUN SHIELD BAOS--WHEN BAGS RUPIU p~
10263. THEY DROP THEIR CONTENTS INIO REFUELING POOL
**THE REST OF THESE ARE NEW-SEDIMENTATION AREA FOR FISSION FRODUC
**** THESE ARE ONLY REQUIRED IF TH -
satsassssssssssssagssagEWs********E 42 425. DETAILED CONTAINMENT FAILUNE MODEL Ib USED. 43 .04590 NUMBER GIVEN IN ITtMOF 43 TkNDONS IN HOOP D1RECTION IN THE LENGTH OF 44
** .0512 VOLUME RUNNING IN INE VOLUME OF REBAR P OF REBAR HOOP DIRECTION FER UNIT AREA 0F QUTER WALL (EOUIV T 45 .1969 RUNNING IN l'HE I R CTIONNIT AREA 0F 001ER WALL (EGUIV THICKNESS) 46 164. DIAMETER OF HOOP TENDONS j ** ! TNAT PART OF THE WALL REPRESEN1EU IN D
. _ . . . . . . . u. -- _ , . . . . . - -. - -- - ~ ~ ~
A-5 f' ) as 47 16.4 (EG APPROX IHAT ABOVE THE OPERATING DECK) 48 .584 HkIGHT Or* INTENNAL WALLS 83 DISPLACEMENT IN AXIAL DIRECTION WHICH IS SUFFICIENT TO TEA THE CONTMT WALL (EG A1 A PENETRATION) 49 .984 ' 8x SAME AS 48 FOR 1HE RADIAL DIRECTION ss assssssssssssssssssssssssssssssssssssssssssssssssssssssssssss 3 LOWER COMPARTMENT (OR 'B' COMPT a sssssssss 01 02- 3375.
- 45. sssssssssssssss s s ssssss)ssssss ssss DISfANCE FRUM FLOOR TO TOP OF 8 CuMPARTMENT as AREA THE FLOUR 0F CORIUM
,ENIERED BELOW) POOLI THIS KUST BE LESS lHAN THE AREA Of 03 2.162 04 1840 ' HEIGHT OF CURS ON FLOOR (OVER WHICH WATER OVERFLOWS TO C)
OS. 2.523E5 CHARAC. FREE VOLUMECROSS-SEC AREA 0F LOWER COMPT FOR BURN TIME CALCS 04 40.
** VERf! CAL DISTANCE FNOM THE CAUITY BYPASS FLOW AkEA ss (EG AREA ARQUND VEShEL NUZZLES BUT SEE DEFINITION IN CAVITY SECTION BELOW) ss
-07 25. GF fME IUNNtl FLOW AREA TO THE CENTEN OF THE CAVITY END
- 08 0.00 DISTANCE FNOM THE FLOOR OF A TO THE OPENING FROM B IN10 D 3 FOR CASES WHERE IHE QUTER DOUNDARY OF CONfMT IS A as STEEL SHELL SEFER4TED FROM A CONCRETE SHIELD WALLe as ENTER DISTANCE BEiWEEN THE TWO AND (REAT THE STEEL ss SHELL ENIER 0 GIMERWISE AS A LINEN (ACOMPT AND DCOMPT GU1ER WALLS) -
- 0 UTER 09 WALL OF B DIVIDES IT FROM COMPT D 1125.0 AREA 0F QUTER WALL 10 00 OUTtR WALL LINER THICKNESS
- 11 00 GAP RESISTANCE OF QUTER WALL LINER 12 4.
THICKNESS OF 0 UTER WALL 13 0.92 14 0.157 THERMAL' CONDUCTIVITY OF OUTER WALL 15 144. SPECIFIC HEAT OF OU1EM WALL 16 0 DENSITY OF CUIER WALL - I I ENTER 1 IF THE QUIER WALL IS SOLID STEEL: 0 FOR CONCREIE l ssWALL 17 16290. FOR RADIATION CALCULATIONSssNOTE THAf CORIUM IN & 18 0 HALF SURFACE AREA 0F INTERIOR WAi' 19 0 INTERIOR WALL LINER THICKNESS 20 4.0 GAP RESISTANCE OF BUILDING INTERIOR WALL LINER THICKNESS OF IN1ENIOR WALLS 21 0.92 22 0 157 IHtRMAL CONDUCTIVIfY OF INTERIOR WALLS 23 144 SPECIFIC HEAT OF INIERIUR WALLS 24 6751. OtNSITY OF INTERIOR WALLS 25 O. AREA 0F FLOOR (USE WATEM POOL AREA IF LES$1 26 0. FLOGR LINER fHICKNESS 27 4.0 GAP RESISTANCE OF FLOOR LINEN 28 (HICKNESS OF FLOOR 0.92 THkRMAL CONDUCTIVITY OF FLOOR 29 0.137 i 30 144 SPECIFIC HEAT OF FLOOR DENSITY OF FLOOR 31 1.26E5 33 MASS REGIONIOF EQUIPMENT-THIS REFERS TO EUPT IN1ERNAL 70 THIS sa as THE PRIMARY SYSTEM MASS SHOULD NOT BE INCLUDED SINCE IT l' 32 21422. NAS A SPECIFIC IREATMkNT tLSEWHtRE HEAT TRANSFEM AREA 0F EDPT as00ANTITY 33 8.8 43 IS USED FOR ALL EXTERNAL WALLS st. HEAT TRANSFEM C0 EFFICIENT TO BE USED ON THE QUIER SURFACE 2334 NOT USED OF lHE CONTAINMENr UUIER WALLS (EG IN A AND D) 1
,---ye---s e- ., .,v - ,.y _,, y -- , .-r -
. , , , .,r. .- , -- . -mm..-.m. ,_e- . . . . - - . - = , , . . e -- - e
_ ,_ _- . . - - ~ - - - - - ~~ ~ i A-6 i ! 35 ' 0.D0 as FRACTIONAL AREA AVAILABLE FOR REVERSE FLOW ON 3-1 FLOWPATH as COMPARtD TO THE FORWARD DIRECTION (EG UuE TO ICE CONDENSER DOOR (S) ss SHUTTING)--THIS NO. MUST ss 82 NONZER0 LARGE, DRY CONIMTS AND POSITIVE IN ICE CONDLNSER PLANTS--IGNORED IN 36 1.D0 ss FRACTIONAL AREA AVAILABLE FOR REVERSE FLOW ON A-D FLOWPATH ss (E0 AIR ENTER TTURN 1 IF NO DAMPER FAN FLOW DAMPERS IN ICE CONDENSLRS) 37 384. FLOW AREA FROM B INTO D 38 1092. FLOW AREA FkOM B TO A 39 0. 40 0. NUMBtR OF IGNITERS / IGNITION SOURCES IN B 41 27. AVG DISTANCE OF THESE FROM THk CEILING OF B 42 4500.. MtIUNT OF FLOOR OF B ABOVE FLQQR QF C SEDIMENTATION AREA ss t
' ssssssssssssssssssssssssssssssssssssssssssssssssstssssssssssssssssssssst sCAVITY (CCOMPT) ssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssss strHE CAVITY INCLUDES ALL THE VOLUME 8ELUW THE REATOR N0ZILES INSIDE ssTHE ss BIOLOGICAL SHIELD AND ALL THE VOL QUT TO WHERE 1HE TUNNEL SLOPES UP
** NOTE THAT THE CAVITY HAS TWO FLOWPATHS-
- TUNNEL
- REFERS TO A WATER ssAND PERNAFS COR!un FLOW FATH THAT ENfERS NEAR THE BASE OF tHE CAVIfYi ss' BYPASS' REFERS TO A FLOWPATH HIGHER IN THE CAVITYI THIS COULD BE THk ssANEn ARQUND THE RV N0ZZLES, OR IN fHE CASE OF SOME PLANTSs BLOWOUT PANELS s*
ss HIGHER IN THE CAVITY--THE BYPASS AREA IS ASSUMLD TO EMPTY INTO B ' stIN SOME PLANTS WATER CAN FLOW DOWN FROM THE REFUELING FOOL TO THE C ssTHE RV ANNULUS--AT FRESENT GAS IS NOT EXCHANGE ss - - asIN MANY SE00kNCES, NAT. CIRC. IS SET UP WHEREBY COLD GAS ENTERS THE ssTHROUGH 01 7.74 THE BYPASS AREAstCAVITY IHROUGH THE IUNNELe IS HEA1LO BY as BYPASS (NON-TUNNEL) CUMPARTMENTSI FLOW AREA COUPLING CAVITY TO LOWER /UFFEk ss THIS SHUULD PT THE LIMITING FLOW AREA, EG ss THE AREA AROUND THE N0ZZLES AS THEY FENETRATE 1HE BIOLOGICAL 02 476.4 SHIELD OR THE ANNULAR FLOW AREA BETWEEN THE RV AND THE SHIELD 03 233.0 AREA 0F CAVITY POOL--THIS INCLUDES KEYWAY EIC WHERE APFLIC 04 17. CHARAC. CROSS-SEC AREA 0F COMPT FOR BURN TIME CALCULATION 05 158 2 HEIGHT OF VLSSEL ABOVE BOTTOM OF CAVITY FUNNEL CRUSS-SECTNL AREA
- 06. 253 ss LARGEST CHARAC CROSS-SECTML AREA 1 HAT CORIUM MUST as TRAVERSED UN ifs WAY TO THE OPENING WHERE IT MAY BE ss EN1RAINtB OR FLOODED TO COMPTS A OR B--IN FLANTS WITH b01 TOM HEAD PENT 1RATIONSe ss fHIE WILL TYPICALLY BE THE
** 'KLYWAY' AREA (THIS IS USED TO CALCULATE THE MINIMUM 07 16935 VELOCIIY WHICH CAN ENIRAIN IHE CORIUM AND WATtR)
CAVITY FFEL VOLUML 08 22. ss HEIGHT OF TOP OF TUNNEL ABOVE CAVITY FLOOR (MtASURED AT 09 2670. CAVITY END OF THE TUNNEL IF IT SLOPES) 10 0.0 AREA 0F CAVITY CUTEN WALLS LINER THICKNESS 11 0.0 LINER GAP RESISTANCE 12 10.
** THICKNESS OF WALL (UR DtPTH TO BE MUDtLLED FOR HEAT 13 0.92 TRANSFER IF IT IS VERY DEEP) 14 fMERMAL CONDUCTIVITY OF WALL - 0.157 SPECIFIC HLAT OF WALL 15 144 - DENSITY OF WALL 16 0.
NUMBER OF IGNITION SOURCES IN C
f . .. ~ g s_.
. _~ - w -.- . - --~-==*-*" ~~u - * - * * - - -
- A-7 .
17 0. 18 794. AVG DISTANCE OF THESE FROM IHE CEILING SEDIMEN)ATION AREA 19 43 1
** MINIMUM THROUGH TUNNEL FLOW AREA WHICH CONNECTS CAVITY TO LOWEN LOMPT 35
- CONCRETE AND CONTAIMMENI SHELL ESI .
asFIRST 12 OUANTITIES ARE USED FOR ALL CONCRE1E DECOMPOSITION CALCS.
**UNLESS OTHERWISE STAIEDs CONCRLTE PRUPERTIES ARE FOR ' PURE' N
02 Ah CIFIC HEAT OF CONCRE1E (UP TO MELT POINT) 1503. sassasNEW: DEFINITIONS OF FOLLOWING HAVE CHANGEDatst*** MELTING
**ALL lHE CONCREIE MASS FRACS SHOULD ADD UP IO ROUGHLY 0.9 TO 1.1 as(THE DIFFERENCE BETWEEN THE SUM AND 1 IS DUE TU N0Y ACCOUNIING FOR
**SMALL 03 .029 PERCENTAGES UF MELATIVELY INtRT MAIERIALS: EG AL203 AND MGO) 04 .0 MASS FRACTION OF CONCRETE THAT IS FREE WATER
**! TEM $0 SHOULD MASS BE SMALL FRACTION OF CONCHE1E tHAT IS CHEMICALLY BOUND WATER OS .01S FOR ' BASALTIC
- TYPE CONCREfES 04 2.74ES MASS FRACTION OF CONCHETE IMAT IS C02
** ENERGY ABSORBED IN ENDOTHERMIC CHEMICAL REACTIONS DURING CONCHETE DECOMPOSITION 07 S.SE5 08 LATENT HEAT OF MELTING 1 8E-2 09 S.4E-2 MASS FRACTION OF CONCRETE THAT IS NA20 SAME FOR K20 10 0.SS SAME FOR SIO2 11 0.30 SAME FOR CAO + OTHER L asasstaasstatsas s* NEWS **************ESS 12 183.
** VOLATILE CONCRETE COMPONENTS .
** REBAR DENSITT ==DCSRCN (MASS OF REBAR PEN UNIT VOLUME OF sa REINF0F.CED CONCRETE) = KG $1EtL'/ Ma*3 S1EE!. t CONCRETE
** RELATED TO -(KG STEEL / KG CONChkTE)==R BY
** DCSRCN RSDCN0/(1 + R10CN0/DCS) WHERE DCN0 IS THE -
** VIRGIN CONCRETE DENSITY ==2300 KU/M833 AND DCS== VIRGIN
** SLEEL otNSITY==8000 KG/Ma*3 - CONSIDER THOSE VALUES HARD-WIRED BECAUSE THEY WILL BE USED BY MAAP INTERNALLY.
** REMAINDER OF THE QUANTITIES AME USED IN IHE CONTAINMENT FAILURE MO
**AND
**NorES SECTION) NEED NOT BE SUPPLIEE IF THE ' SIMPLE' MODEL IS USED (SEE GENERAL 83 NOTE: FOR FREE-STANDING STEEL CONTAINMEN1Se YOU NEED SUPPLY ONLY THE
- ' LINEN' PROPERTIES (WHICH ARE TAKEN TO DESCRIBE lHE STEEL SHELL)
**AND THE STEEL THICKNESS (STEEL THICKNESS IS INPUT AS
*** LINER'
**SEE GENERAL THICKNESS NOTES SECTIONIN THE UPPER AND ANNULAR COMPARTMENT SECTIONS)--
13 3.E11 14 ELASTIC YOUNGS MODULUS FOR TENDONS 15 1.99E11 ELASTIC YOUNGS MODULUS FOR REMAR 3.9/E9 16 1.4E9 PLASTIC YOUNGS MODULUS F08t TENDONS 17 9.7E8 PLASTIC YOUNGS MODULUS FOR REBAR 18 PRESTRESS ON HOOP TENDONS 1.01E9 PRESTRESS ON AXIAL TENDOMS 19 1.53E9 20 TENDON YIkLD STRESS 21 4 137E8 REBAR YIELD STRESS 1.4SE9 TENDON ULTIMATE STRESS 22 6.2E8 23 REBAR ULTIMATE STRESS
~ 24 1.99E11 ELASTIC YOUNGS MODULUS FOR LINER 1.4E9 23 PLASTIC YOUNGS MODULUS FOR LINER 26' 4 137E8 LINER YIELD STRESS 6 2EB LINER FAILukE STRISS ss 35
, .. _.. . . .. - .-~ ..
A-8 SBR sssssssssss***ssssssssss**ssssssssssssssssssasts***?sssssss**ssatassass CONTROL CARDS sassssssssssssssssssssssssssssssss****sassssssstast***sssas*** as ssssa s:USE UF THE FAST SfEAM IABLES.IS NOT RECOMMENDED AND SAVES LIITLE TIME , 01 1 ENTEN A 0 TO USE FAST STEAM TABLES IN PRI SYS WHEN PUSSIBLE 02 1 ENTER A 0 TO USE FAST STEAM TABLES IN CONIMT WHEN POSSIBLE 03 1 INTEGRATION METHOD: RUNGE-KUTTA ORDER (1 3R 2)I 1 IS as RECOMMENDED 04 29 UNIT NUMBER (' TAPE' No. IN CDC JARGON) as f0 WRIIE PESTART FILES FOR MAIN PROGRAM FROM THIS RUN 06 30 UNIT NUMBER TO WRITL RESTART FILES FOR HEATUP FNOM THIS RUN ss07 NOT USED 08 10 UNIT NUMBER TO Put FRI SYSTEM OUTPUf ON 09 10 UNIT NUMBER TO PUT CONTAINMENT GUTPUT ON (MCST USERS PUT as IN THE SAME NO. WHICH APPENUS THk TWO FILES) 10 31 UNIT NUMBER FOR THE FIRST PLOT FILE (UfMERS SEQUENTIAL) 11 39 UNIT NUMBER FOR SCENARIO FILE ssNEXT 3 QUANTITIES CUNTROL IHE PLOT POINT STORAGE FREQUENCY (SEE V0L 1 0F sauSER'S MANUAL) 12 250 NON-SPIKE NUMBER OF POINTS (AVERAGE BEHAVIOR) S10 RED 13 15 NUMBER OF POINTS STORED DURING A SPIKE (TO RESOLVE FAST sa TRANSIENTS) 14 800 . MAXIMUM NUMBER OF FLOT FOINTS ALLOWED PER PLOT FILE
**SE. ESF LINEUP MENU IN SUBROUTINE ENGSAF WRI1E-UP IN ssVOL 2 0F USER'S MANUAL FOR NEXT TWO ENTRIES 15 2 ESF PUMP LINEUP IN RECIRC (1 FOR ZION, 2 FOR SEQUOYAH) 16 1 ESF PUMP /ACCUM DISCHARGE SE1UP (1 FOR ALL TO COLD LEGS) 17 0 EN1ER A 1 FUR B AND W PLANTS, 0 OfHERWISE (NOTE THAT MAAP as WAS WRITTEN FOR B AND W PLANTS WHOSE OTSG LOWER TUWESHEETS ss LIE BELUW THE LEVEL OF fHE PRIMARY SYSTLM NUZ4LES--NOT
. as .
YET TESTED FOR"THE HIGH TUBESHkEr CONFIGURATIONS) 18 13 FILE NUMBER TO WRI1E AUX DATA ON FOR LAIER STAND-ALONE as AUX RUNS (OR O NOT TO WRITE DAlA) ass **sasssssssssssEWsssssssssas*sssssssi - 19 0 FILE No. TO READ AUX DATA FROM (IF THIS NUMBER IS NONZER0e ONLY
** THE AUX BUILDING MODELS ARE RUNE INE INPUT T/H DATA FROM THE
- CONTAINMENT HAVING BEEN RECORDED FROM A PREVIOUS RUN) 20 4 NUMBER OF NODES IN INE AUX BUILDING (MAX =53IF Oe INE AUX BLDNG ss MODELS ARE NOT RUN, Bul A FILE MAY STILL BE CREATED FOR ss SUBSEQUENT STAND-ALONE AUX BUILDING ANALYSES BY SUFPLYING
*: A NONZERO No. FOR ITEM 18) c su
{ asssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssss l
- CORE t -
susssastas*** ssassssssssstrassssassssssss** ss*ssssssssssss**ssassas*** 01 .031167 FUEL PIN UUTER DIAMETER 02 46920 INTIAL ZIRCALLOY MASS 03 50952. NUMBER OF FUEL PINS 04 222739. TOTAL U02 MASS ssITEM 5 MUST BE ABOVE THE ELEVATION SUFPLIED FOR 1HE TOP GF THE RV HEAD
- SIN fME PRIMARY SYSTEM SECTION 05 10.132 ELEVATION OF BOTTOM OF ACTIVE FUEL ABOVE BOTTOM OF VESSEL 06 22.132 ELEVATION OF TOP OF ACTIVE FUEL AB0VE 80TTOM OF VESSEL 07 8766. TIME OF IRRADIATION 08 1.16E10 FULL POWER ssTHE CORE NODALIZATION ADMITS UP TO 70 N0DESI IN ADDITION: NO MORE THAN as20 RUWS MAY BE USED AND NO MORE THAN 7 RINGS OR COLUMNS y s*WHATEVER NODALIZATION IS UfED, INSERT FEAKING F ACTORS IN10 AFFROFRIATE L
L l. l
-e M YNy wr7-+y- --' - - - - = ' *N-w- wwm- ww--., w -g----n+ww-,r-w- WeWW-*-"iMW9-%-gaw a-G y.g g 4 9 e w9, @ @e$ E. 3 ' G r--g- e dr- - --m-- y--- mw--- ,m-----wg wr-y vy7i- P-M-, 7-eyevevii-
, ,_ _ ... .. - - . - - - - -- - - ~~
A-9
**tNTRY NUMBERS (EG SECOND RING I FROM INS DE RADIAL PEAKING FACTOR IS
. *sALWAYS ITEM 32 NO MATTER HOW MANY AXIAL NODES)
**f0P N0DE IS UNFUELED (FISSION GAS PLENUM ETC) AND MUST HAVE ZERO
** FACTOR 09 7 NUMBER OF RINGS 10 10 NLMBER OF RUWS
** TOP ROW IS STRUCTURE (UPPEN PLENA, ETC)e 50 PEAK =ZENO.
** 9-ROM VALUES HAVE BEEN 08fAINtd BT AVERAGING 10-ROW VALUES.THE F
' ** THIS MEANS THAT ROW 5 HAS A SLIGNILT LOWER FEAK THAN OTHENWISE-
** MkEVALUATE 11 0 498 AXIALTHESE PEAKING FACTOR PEAKING FACTORS IF DESIRtDs BUT DIFFERtNCES AR BOTiun 12 0./34 AXIAL PEAKING FACTOR 13 1 124 AXIAL PEAAING FACTOR 14 1 402 AXIAL PEAKING FACTOR 15 1 480 AXIAL PEARING FACTOR 16 1.402 4
17 AXIAL PEAKING FACTOR
' 1 124 AXIAL PEAKING FACTOR
' 18 0.~/34 19 AXIAL PEAKING FACTOR 0 498 AXIAL PEAKING FACTOR
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20 0.000 AXIAL PEAAING FACTOR TOP R 31 ENTRIES 1.09 21-30RADIALAXIAL PEAKING PEAKING FACTOR FACTO INSIDE S NOT USED IN 1HIS NODALIZATION 32 1.11 33 RADIAL PEAKING FACTOR 1.10 RADIAL PEAKING FACTOR 34 1.115 35 RADIAL PEAKING FACTOR 1.096 RADIAL PIAdING FACTOR 36 1.01 RAUIAL PEAKING FACTOR 37 0.75 RADIAL PEAKING FACTOR DU1 SIDE 38 0.047 AREA OR VOLUME FRACTIONS 39 0.062 INSIDE 40 AREA OR VOLUME FRACTIONS 0.145 AREA OR 70LUME FRACTIONS 41 0.124 AREA OR VOLUME FRACTIONS 42 0.207 AREA 02 VOLUME FRACTIONS
- 43 0.166 AREA OR VOLUME FRACTIONS 44 0.249 AREA OR VOLUME FRACTIONS l' OUTSIDE
**FOLLOWING 45 32000 QUANIITIES CONTROL ANSI DECAT HEAT CALCULATION
** FUEL EXPOSURE AT SCRAM (ALhdTS IN 4e .39 MEGAWATT-DAYS / METRIC TON NO MATTER WHAT UNITS CELECTED)
** FUEL ' ALPHA
- AT SHUTDOWN (FISSILE ISOTOPE CAPrukES/ FISSION) 47 .032 48 .442 INITIAL kNRICHMtNT OF FUEL IN ATOM FRACTION
** CONVERSION RATIO (PRODUCTION RATE OF U-239/ ABSORPTION RAT P 49 .487 IN FISSILE ISOTOPES) AT SHUTDOWN i ** FRACTION OF FISSION POWER MADE DUL TO F1SSIONS IN U-235 AND PU-241 AT SHUTDOWN 50 443 SAME AS 49 FOR PU-239 L $1 .069 52 6.5E-4 SAME AS 49 FOR U-238 (FAST FISSIONS)
FRACTIONAL IR02 MASS (COMPARED TO IN MASS) AT TIME O sass
$3 ****sts**ALL
.01344 THE REMAINDER IN iMIS SECTION ARE NEWE***
54 FUEL PELLET kADIUS 3 185
** CORE FLOW AFEA IN THE BYPASS AREA BETWEEN THE CUNE BAF/LE P
** AND THE CUME BARNEL (ENSURE fMIS IS CONSISfENT WITH FRI STSTEM CORE FLOW AREA PARAMETER NO. 5) .
** PARAMETERS asMOSTLT FROM TMI REPORTS) SS-60 ARE USED FOR CALCULATING BALLOCNING (DATA SHOWN
' 55 1.87bE-3 CLAD THICKNESS 54 GAS VOLUME PER FUEL PIN 57 450. 58 AS-BUILT ROOM TEMP FUEL PIN FILL GAS FRESSURE I CURE SUPPORT PLATE MASS--THIS PLAIE IS MELTED BY INE DEBRIS l
. , - . . , , . - . - , , - r,r,. , . . , , , , , , , .- , , . . . . _ _ ,- .,,,,-,,-..,w-,n,.rw,. ,.-.,.n,. - - . . , - - _ . . - , - - . , , . . - . . - , -
~. . .- - u .- - - - - - - - -- --- -~ - '-
L L A-10
-as AS IT LEAVES THE ORIGINAL CORE BOUNDARY 59 FRACTION OF THE TOTAL FUEL PIN GAS VOLUME WHICH IS ss CONTAINED IN THE LOWER GAS PLENUM OF THE PIN 60 SAME IN THE UPPER GAS PLENUM 61 RADIUS OF CORE BAFFLE (2sPISTHIS ENIRY SHOULD bE THE st CIRCUMrERENCE OF IHE BAFFLE) 62 0. ' FLOW AREA PER ROW' IN CORE BAFFLE (IMPORTAN1 ONLY IF ss IN-VESSEL NAIURAL CIRCULATION RETURN LEG IS IN BAFFLE-CORE
** BARREL ANNULUS--SEE *MODEL)--THIS REPRESENTS THE APPROXIMATE ss FLOW AREA AVAILABLE AS IHE FLOW IURNS SIDEWAYS AND PENETRATES ss THE CORE 63 0. FOR TMI-TYPE GEOMEIRIES THE FLOW AREA THROUGH EACH CORE ss FURMtR PLATE IN AXIAL DIRECTION 64 0. FOR TMI-TYPE CORESeNUMbER OF CORE FORMLR PLAihS IN THE ss BAFFLE-CORE ANNULUS ss ss s*
ssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssas \ s!CE CONDENSER ('I' COMPARTMENT) sssssssssssssssssssssssssssssansssssssssssssssssssssssssssssssssssssssst 01 0.E0 TOTAL VOLUME INCLUDING THE ICE
$302 EXIT GAS TEMPERATURE--THIS IS THE TLMPENATUNE OF GAS LEAVING THE 33 ICE BOX (SEE WRIBE-UP FUR SUBROUTINE HTICE IN VOL 2 0F USER'S MAN) 3x03 INITIAL TEMFERATURE OF THE ICE ss04 SPECIFIC VOLUME OF ICE--NOTE THE TOTAL VOLUME MINUS.lHE ICE MASS ss TIMES THE SPEC VOL SHOULD BE THE FREE VOLUML s*05 FLOOR AREA 0F WATER SUMP IN 80TTOM OF ICE CONDENSER 8806 HEIGH 1 QF SUMP (IE CUMB OVER WHICH WATER DRAINS INTO B)
*s07 VERTICAL HEIGHT OF ICE BOX 3s08 FLOW AREA BETWEEN LOWLR COMPARTMENT AND THE ICE CONDENSEN
*s09 SEDIMENTATION AREA ss ss ssssssssssssssssssssssssssssssssssssssttssst*stsssssssssssssssssssssssss sUPLENUM (UPPER PLENUM UF ICE CUNDENSER- 'u' COMPARTMENT) sssssssssssssssssssstatsstatss*****ssssssssst****ssassasssssssss**ssssas 01 0. VULUME--ENIER 0 VULUME FOR LARGES DRY CUNTAINMENTS
**02 CHARACTERISTIC CROSS-SEC AREA 0F THE COMPT FOR BURNS ss03 HEIGHT OF UFPER PLENUM s*04 LIMITING FLOW AREA WHICH COUPLES THE ICE CONDhMSER TO THE UPPER as COMPARTMENf--IE USE THE LESStR OF THE UPPER PLEN TO UPPER COMPT 3 FLOW AREA CR THAT CDUPLING THE UFFER PLEN TO THE ICE CONU.
*s05 NUMBER OF IGNITERS IN U 8804 AVERA3E DISTANCE OF IGNITERS BELOW THE CEILING OF U
*s07 AVG DISTANCE FROM THE TOP OF UP PLEN TO IHE PORTION OF THE ss CEILING OF THE UPPEN COMPT WHICH IS JUST OVER THE EXIT 001 0F UI as - THIS IS USED TO CALCULAlE LOCAL BURNING IN THE UPPER COMPT 3 INf!ATED BY FLAME PROPAGATION OUT GF U s*06 SEDIMENIAf!ON AREA IN U 85
. *s statss***stsssssssssssssssssssssssssssssssssssssssssssssssssssssssssssss sANNULAR COMPARTMENT (*D' COMPARIMtNT)
*Essssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssss ts!N LARGE DRY CONTAINMENTS:
stTHIS COMFARTMENT REFRESENTS THE VOLUME BEfWEEN THk CRANE WALL (IF ANY) ss4ND THE CONTMT WALLS AND BEIWEEN fHE DECK AND IHE LOWER COMPT FLCOR-- s*IF NO CLEAR DISTINCTION, ARBITRARILY DIVIDE THE SFACE BELOW THE UFFER
c-- - - - -
- A-Il .
** COMPT AND USE LANGE FLOW ANEAS TO KtEP THE GAS WELL MIXtD--AT PRESENTe
**CORIUM IS ASSUMED NOT TO GE1 INf0 THIS COMPARTMEN1
**IN ICE 01 2.968E5 CONDENSERS:
FREE VOLUMETHIS VOLUME REFLECTS THE ' DEAD-END' COMPARTMLNTS 02 4888. AREA 0F WATER P00L 03 0. DISTANCE THE >LOOR OF D IS A80VE THE FLOOR OF B 81 06
!!m. MahMimNa5Sa "' """ ' " '""" ""' c c*^ " o" 0.03125 WALL LINER THICKNESS 07 0.28 GAP RESISTANCE OF WALL LINER 08 45 THICKNESS OF WALL 09 0.92 IHtRMAL CONDUCTIVIlY OF WALL 10 0.157 SPECIFIC HLAT OF WALL
' 11 144. DENSIlY OF WALL 12 0.
** ENTER A 1 IF THE OUTER WALL (CONfMT OuiEN BOUNDARY)
IS MADE OF STELL 13 0. 14 0. HEIGHf 0F CUN8 SEPERATING D ANU B MEASURED FNOM B'S FLOOR 15 0. NUMBER OF IGNITENS OR IGNITION SOURCES IN D 16 9776. AVG DISTANCE OF THESE FROM THk CEILING SEDIMENTATION AREA 1' **sst****s33sALL THk REMAINUER IN THIS SECTION ARE NEW8333
**1HE NtXT THREE PARAntTEMS ARE USED TO DEFINE THE EFFICIENCY OF
**IWERTIAL IMPACTION
**IN LARGE: DRT'S THESE PARAMETERS SHOULD CHARACTERIZE
** GRATES WHICH ARE ASSUMtD TO BE IN IHE ANNULAR COMPARTMENT as
- SIN ICE CONDENSER PLANTS, THESE PARAMtTERS (EVEN THOUGH LOCATLD
**IN THE ANNULAR COMPARTMENT DATA SECTIGN) SHOULD REFLECT IMPACTION AND
*sFLOW 17 AREAS AND STRAF WIDTHS IN THk ICE BOX--SEE EG FCSTMA 733.2 18 0.010 IMPACTION AREA ( AREA 0F BANS IN GRAIES fHAT INTERCEPT FLOW)
WIDTH OF GRATE PARS 19 4154.8
** FLOW AREA THROUGH GRATES (DEFINES FLOW VELOCITY OF AEX0SOLS TRAVtRSING: GRATES) ~
- NOTE: IF MORE THAN ONE LEVLL OF GRATES EXISTS, USE THE TOTAL IMPACTION AREA
**0F ALL INE GRAIES: AND THE MAXIMUM FLOW ANEA AT ANY OF THE GRA1E as ELEVATIONS
*SI 8:
a:USEDDETAILED CONTAINMLNr FAILURE MODEL INPUIS--IGNORE IF SIMrLE MODEL 20 130. st NUM8ER 0F TENDONS IN HOOP DIRECTION IN THL FART OF THE WALL WHOSE AREA IS GIV6N IN ITEM 5 ABOVE 21 216. 22 NUMBER OF TENDONS WHICH RUN IN THE AXIAL (VERTICAL)DIRECTION
** .01399 VOLUME OF RtBAR PER UNIT AREA 0F QUTER WALL (EQUIV THICKNESS)
RUNNING IN THE HOOP DIRECTION 23 .06 DIAMtTtR OF HOOP TENDONS 24 .06E0 25 DIAMETER OF THE AXIAL TENDONS
.0156
*E VOLUME IN RUNNING OFTHE REBAR AXIALPERDIRECTION UNIT AREA 0F OUTtR WALL (EQUIV 1HICKNESS) 26 .3
** DISPLACEMENT IN AXIAL DIRECTION WHICH IS SUFFICIENT TO TEAR THE CONTMT WALL (EG AT A FENElRATION) 27 .3 SAME AS 26 FOR THE RADIAL DIRECTION SFR
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- ENGINEERED SAFEGUARDS stat *******************************
asasstassassanssasssssassas***3:****
**IN BRITISH UNITSe s**ss***ssas******* ass *************
**FLOWWATES SPECIFIED TO BE VOLUMETRIC SHOULD BE Ms33/SECl CTHtR FLOWRATES
**IE ALL THOSE NOT EXPLICITLY STATkD TO BE VOLUMETRIC
. - . . - . . .. -~ -. . - - . . . . - . _ . .
A-12
**SHOULD BE EG/SECI HEADS SHOULD BE IN Mi FRESSURES IN FAI IN ENGLISH THE
** UNITS ARE RESPECTIVELY GPMsLBM/HRerie PSIA--
**NOTE TO MAAP/BWR USERS--GFM IS USED IN MAAP/FWR INSTEAD OF Fl**3/HR
**IN
** CONDENSER PLANTS)
THE FOLLOWINGe' FANS' REFER TO FAN COOLERS--(AIR RETUNN FANS IN
**FOR BETTER ACCURACYe YOU MAY ELECT TO INPUI ' SYSTEM' FUMP HEAD CURVES WHICH
** INCLUDE fHE EPFECTS UF FMICTION IN IHE IHLtT AND OUTLET PIPING (WHICH IS
** IGNORED IN MAAP)I IF YOU DO S0e BE SURE THE ASSUMPTIONS ON STATIC HEAD
**WHICH ARE USED IN l' HEIR CALCULATION ARE CONSISIENT WIlH THE PUMP ELEVATIONS
**ETC. WHICH ARE INPUI BELOW--THIS IS GENERALLY A FACTOR ONLY IN CRITICAL
** APPLICATIONS SUCH AS FLED AND BLELD WHERE 1HE CHARGING PUMP FLOW IS
** BARELY (OR NOT) ADE00 ATE TO MATCH DECAY HEAT 01 0.833 ACCUMULATUR PIPE DIAMETER 02 195. FRESSURE SETFOINT FOR LPI 03 1336. PRESSURE SLTPOINT FOR HPI 04 615. INITIAL FRESSURE OF ACCUMULATORS 05 86.
** 1EMPERATURE OF REFULLING WATER STORAGE TANK (RWST)--IE THE TANK FROM WHICH THE CHARGING, HPI, LPle AND SPRAYS
** DRAW THEIR WATER DURING 1HE INJECTION PHASE 06 100. 1EMPERATURE OF ACCUMULATORS 07 2.91E6 INITIAL MASS IN RWST
- 08 5.28E4 INITIAL MASS PER COLD LEG ACCUMULATOR 09 1319.5 AREA UF 8ASE OF RWST 10 18 3 LENGTH OF AN ACCUMULATOR PIPE 11 45.1 PRESSURE SETPOINT OF BLDG SPRAYS 12 300. PRESSURE SETFOINT OF BLDG FANS
- 13 0 NUMBER OF OPERATING FAN COOLERS OR FANS
- Id 0.0 VOLUMETRIC FLOW THROUGH ONE FAN C00 lex OR FAN 1h
** 2.165E-3 NUMINAL PfAMETER OF CONTAINMtNT SPRAY DROPLETS AS THEY LEAVE THE SPRAY HEADER 16 1J50. VOLUME OF ONE COLD LIG ACCUMULATOR 17 4 NUMBER OF OPERATIONAL COLD LEG ACLUMULATORS 18 1 NUMBtR OF OPERATIONAL HPI PUMPS 19 1 NUMBER OF OPERATIONAL LPI FUMFS 20 5 21 3600. NUMBER OF ENTRIES USED IN HPI PUMP-HD CURVE TABLE (5 MAX)
HIGHEST HEAD IN TABLE (UNITS ARE MEIERS) 22 3200. NEXT HIGHEST HEAD IN HPI PUMP-HEAD CURVE TABLE 23 2900. 24 1650. NtXT HIGHEST HEAD IN HPI FUMP-HLAD CURVE TABLE NEXT HIGHEST HEt ,IN HPI PUMF-HEAD CURVE TABLE - 25 00 LOWEST HEAD IN HPI FUMP-HEAD CUkVE TABLE 26 0.0 VOLUMETRIC FLOWRATE CORESFONDING TO FIRST ENTRY IN
** THE PRESSURE TABLE 27 325. NEXT VOL. FLUWRATE 28 425. NEX1 VOL. FLOWRATE 29 650. NtXT VOL. FLOWRATE 30 450. NEX1 VOL. FLOWRATE .
31 5 NUMBER OF ENTRIES USED IN LPI TABLE 32 470. HIGHEST HEAD IN LPI TABLE 33 425. NEXT HEAD 34 390. NEXT HEAD 35 325. MtXT HEAD 36 0.0 NEXT HEAD 37 0.0 FIRST VOLUMETRIC FLOWRATE IN TABLE 38 2000. NEXT VOL. FLOWRATE 39 3000. NEXT VOL. FLOWRATE 40 4500. NEXi VOL. FLOWRATE 41 4500. NtXT VOL. FLOWRATE 42 2687. CHARGING PUMP FRESSURE SEff0IN1
. _ . - . - - - - - ~ ~- -
A-13 43 1.0 NUMBER OF WORKING CHARGING PUMPS 44 5 NUMutR OF EMIRIES IN CHARGING PUMP HEAD CUhVE TABLE 45 6000. FIRST HEAD 46 5800. MEXI HEAD 47 4800. NEXT HEAD 48 2000. NEXf HkAD 49 0.0 NEXT HEAD 50 00 FIRST VOL. FLOWNA1E 51 150. EXT VOL. FLOWRATE 52 300. NEXT VOL. FLOW ATE 53 b50. NEXT VOL. FLOWRATE 54 550. NEXT VOL. FLOWNATE 55 160. AREA 0F BASE OF CONIMT SUMP
$6 8 DEPTH OF CONTM1 SUMP
*sN0 fee IF DESIRED YOU CAN SUPPLY ONE NUMBER--IF DO SO GIVE IT A LARGE asHEADe TEM A CONSTANT FLOW MODEL WILL BE USED 57 1 NUMBER OF USED EN1 RIES IN SPRAY PUMP HEAD CURVES (5 MAX) 58 1000 FIRST ENTRY IN SPRAY PUMP HEAD TABLE a HEADS 59-62 NOT USED IN THIS SCHEME 63 1.650-1 FIRST VOLUR TRIC FLOW ENIRY IN SPRAY PUMP TABLE
** VOLUMETRIC FLOW VALUES 64-67 NOT USkD IN fHIS SCM ME 88 FOR NPSH TABLES: TR SAN FLOWS AS WERE GIVEN FOR HEAD CUNVES AkE '*
as ASSUMtD TO CORRESPOND T0 IHE NPSH HEADS GIVEN 68 28. NPSN (UNITS OF LENGTH) REQ'D FOR CHARGING PUMP ta AT FIRST FLOW IN TABLE 69 28. NEXT NFSH ENTRY FOR CHARGING PUMPS
- 70 28. MEXT NPSH kNTRY FOR CHARGING PUMPS 71 28. .
NEXT NPSH ENTRY FOR CHARGING PUMPS 72 28. NEXT NPSH ENTRY F3R CHARGING PUMPS 73 13.5 FIRST NFSH ENIRY FOR LPI 74 13.5 NEXT EN1RY FOR LPI 13.5 75 NEXf ENTRY FOR LFI 76 13.5 NtXT ENTRY FOR LPI - 77 13.5 NEXT ENTRY FOR LPI 78 25. FIRST NPSH ENTRY FOR HPI 79 25. NEXr ENIRY FOR HPI 80 25. NEXT kNTRY FOR HPI 81 25. NEXl-ENTRY FOR HPI S2 25. NkXT kNTRY FOR HPI 83 19. . FIRST NPSH EN1RY FOR SFRAY PUMPS 84 3.05 NEXT EN1RY FUR SPRAY PUMPS 85 3.05 NEXT ENTRY FOR SFRAY PUMPS 84 3.05 R XT ENTRY FOR SPRAY PUMPS 87 3.05 NEX1 ENTRY FOR SPRAY PUMFS 88 1 NUM8tR OF OPERATING SPRAY PUMPS FOR UPFtR COMFARTMENT 89 0 NUMBER OF OPERATING SPRAY PUMPS FON LOWEx COMFAkTMtNI 90 77.5 HEIuMT OF BOTTOM 0F RWST ABOVE INE ENG SAFE PUMFS 91 22.3 HEIGHT OF 80TTOM OF CONTAIN SUMP AbOVE THE ENG SAFE FUMFS 92 38. ELEVATION OF IHE RV INJECTIUM N0ZiLES ABOVE IHE SI FUMPS 93 7577. FLOW THWOUGH ONt SPRAY PUMP WHEN ITEM 94 IS MEASUNEu 94 40. DIFFERENTIAL PRESSURE ACROSS 1HE SFRAY N01ZLES 95 0.0 MASS F_(0WNATE OF EX1ERNAL RWST REFLACEMENT WATEke IF ANY 96 .00278 TIM MLAY FM HPI (IE TIME StTWEEN fHE ACTUATION AND WHEN
** ACTUAL OPERATION BEGINS) 97 .002778 TIME DELAY FOR LPI 98 .002778 TIME DELAY FOR CHARGING PUMPS 99 .00833 TIME DELAY FOR UPPER COMPAR1 MENT SFRAYS 100- .00833 TIME DELAY FOR LOWER COMPARTMENT SPRAYS 101 5.0 TIME DELAY FOR FAN CCOLERS 102 NUM8ER OF TUBES IN A FAN CDOLER
_ _ . , - . . - = ~ - - - - - ~~ A-14 t 103 OUTSIDE AREA 0F ALL fuBES IN A FAN COOLER 104 AREA QF ALL FINS IN A FAN COOLER 10S FAN CO3LtR FIN EFFICIENCY 106 FAN COOLER INSIDE FOULING FACTOR 107 FAN CUOLER FIN DIAMETER 108 FAN COOLER TUBE THICKNtSS 109 FAN C00LtR fuBE THERMAL CONDUCTIVITY 110 MINIMUM FLOW AREA THROUGH FAN COOLEN 111 FAN COOLER 1UBE ID 112 5 NUMsER OF NODES USED TO MODEL FAN COOLER (5 MAX) 113 INLti COOLING WATLR (EMP TO FAN COOLER--NOTE LHIS IS sa ALSO USED AS THE COOLING WATEN TEMP FOR ALL OTHtR
** SAFEGUARJS HEAT EXCHANGERS 114 INLET C0 CLING WATER FLOW TO A FAN COOLEN 115 1 NUMBER OF LFI PUMFS USED FOR RHR SPRAYS WHEN VALVE OPEN 116 0 ENTER A 1 IF FANS / COOLERS DISCHARGE TO 310 TO D as
**ESF HX'S
** CALCULATIONS CONTROLLED BY HEAT EXCHANGER TYPE
*sHEAT EXCHANGER TYPE:
** -1 SET GUTLET TEMP OF HX TO RWST TEMPERATURE
** O IS NO HX--0UTLET TEMP IS CONTMT SUMP TEMP
** 1 STRAIGHT TUBE HX ta 2 U-lube HX ss
**IMPORTANT NOTE:
**FOR HX TYPES 1 AND 2 EITHER SUPPLY ALL GEOMLIRIC FARAMLIERS
**0R IHE NTU (NUMBtR OF TRANSFER UNITS) PER HX--ALL KNOWN USERS DO
**THE LATTER--NTUS ARE AVAILABLE BY CONSULTING NAntFLATE DATA ANU- .
**USING GRAPHS IN: FOR EXAMPLE: HOLMAN HEAT TRANSPER
*s4LL FARAMETERS ARE ON A PER HX BASIS ff7 2.D0 TYPE OF HX'FOR SFRAY -
113 0.00 NUMBER OF 10BES IN SFRAY HXS
- 119 0.D0 NUMBER OF SHELL SIDE BAFFLES IN SFRAY HXS 1*0 0.00 SPRAY HX 1UBE ID 131 0.00 SFRAY HX TUDE THICKNESS 122 0.D0 IUBE TO IU8E SEPARATION IN SPRAY HX 123 0.D0 SHtLL LENGTH IN SPRAY HX 124 0.D0 IHtRMAL CONDUCTIVITY OF SPRAY HX TUBES 125 0.D0 126 0.D0 LARGEST FER? DISTANCE FROM SHELL TO BAFFLE (' BAFFLE CUf')
127 2.35E6 SHELL TO TUBE CLEARANCE AT GUTSIDE OF SPRAY HX IUBE BDL SFRAY HX COOLING WATER MASS FLOWNA1E
**128 NOT USED 129 2 TYPE OF HX FOR RHN 130 0.D0 NUMatR OF TUBES IN RHR HXS 131 0.D0 NUMBER OF BAFFLES IN RHN HXS 132 0.D0 TUBE ID IN RHR HXS 133 0.D0 TUBE THICKNtSS IN RHR HXS 134 0.D0 (UBE TO IUBE SEPERATION IN RHR HXS 135 0.D0 SHELL LENGTH IN RHN HXS 134 0.00 IUBE IHtRMAL CONDUCTIVIlY IN RHR HXS 137 0.00 BAFFLE CUf DISTANCE IN RHN HXS (SEL 125) 138 0.D0 SHELL 70 IU8E CLkARANCE AT OUTSIDE OF RHR HX IUBE BUNDLE 139 2.475E6 RHR HX COOLING WA1EN MASS FLOWRATE 140 0.996 SFRAY HX NTU 141 1 416 RHR HX NTU 142 0.00 SHELL ID OF SFRAY RECIRC HX 143 0.D0 SHELL ID OF RHR RECIRC HX
**tNTER XERO VOLUME FOR ITEM 148 IF NO UHI SYSTEM
- A-15 5:144 INITIAL MASS IN THk UNI WATER ACCUMULATOR ss145 LtNG1H OF THE UHI PIPE TO IHE RV 3:146 DIAMETER OF THE UHI FIPE
**147 INTIAL PRESSURE OF 1HE UHI ACCUMULATOR 148 0.00 TOTAL (WATER + GAS) VOLUME IN THE UNI ACCUMULATORS
**149 FAILURE DIFFERENTIAL PRESSURE OF 1HE UHI FIPE RUPTURE DISK asTHE ' CAVITY INJECTION SYSTEM' IS (RARELY) USED TO SIMULATE A asPRO/OSED DEDICA1ED ESF WHICH MtRELY DUMPS WAIER INTO IHE CAVITY 150 0.00 TOTAL MASS IN THE CAVITY INJtCTION SYS1EM TANK 151 0.00 MASS FLOWRAIE OF IHE CAU INJ SYS1EM WHEN ACTIVATED as USER HAS THE OPTION TO THROTTLE ESF SYSTEMS AT LESS THAN ss IHEIR FULL FLOW GIVtN THE CONDITIONS hXISTING--TO DO THIS, as ENIER FOR THE APPROPRIATE SYSTEM (ANU FOR THE AFW IN THE STM ss GENERATOR SECTION) A TOTAL FLOWRATE DESIREDI IHE CODE WILL USE
** THE MINIMUM OF THIS FLOW AND THAT CALCULATtG FkOM THE HEAD CUKVES sa AND THE NO. OF OPERATIONAL PUMPSIIF OPERATOR ISN'T 1HROITLING, as ENfER A LARGE NO.11F HE CHANGES THE DEGREE OF THNOTTLINGe ENfEN 88 PARAMETER CHANGES USING INIERVENTION NO. 1000 IN CONTROL CARDS 152 7.936E9 THROTTLED FLOW FOR LPI SYSTEM (TOTAL) 153 7.936E9 SAME FOR HPI 154 7.936E9 SAME FOR CHARGING FUMFS 155 7.936E9 SAME FOR UPPER COMPT NORY . SPRAYS 156 7.936E9 SAnt FOR UPPER COMPT RHR SPRAYS (WHEN AC1IVA1EU) 157 7.936E9 SAME FUR LOWER COMPT SPRAYS as as ssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssss sINITIAL CONDITIONS
* *stsstasstasssssssssssssssssssssssssssssas**stsssssstas**sssssssssstass:
01 591.8 NOMINAL FULL POWER PRIMARY SYSTEM WATER 1EMPERATURE 02 2250. NOMINAL FULL POWER FRIMARY SYSTEM FRESSURE OL 28.05 FRESSURIZER WATER LEVEL (ABOVE BOTTOM OF PZR HEAD) 04 14.7 CONIAINMENT BUILDING FRESSUKE 05 120 LUWER CONTAINMtNT BUILDING COMPARTMkNTS-( ALL BUT S: UFFER COMPT ANU ICE CONuCNSEN) TEMPERATUNE 06 0. ILE CONDENSER GAS IEMPERAluRE WHERE APPLICABLE l 07 1. LOWER CONIAINMtNT BUILDING COMPARTMENIS REL. HUMIDITY (0-08 0 INITIAL ICE MASS 09 99253. INITIAL MASS OF WATER ON SECONDARY SIDE OF EACH S/G
** VALUE TAKEN FROM MODtl F SG T1H DATA FOR MILLSTONE l 10 120. INITIAL TEMPERA 1UNE OF CONIAINMENT CONCRETE ANU
, ss METAL STRUCruRES i I 11 944. INITIAL PRESSURE ON SEC SIDE OF S/G'S
*s0FPER COMPT CONDITIONS COULD BE DIFFERENT IN ICE CONDENSERS 12 1 00 UFFER COMPARTMENT REL HUMIDITY (0-1) i 13 100. UPPtR COMPARTMENT 1EMPERATURE i 14 591 8 INITIAL FNIMARY SYSTEM WATER TEMrERATURE FOR TNIS RUN 15 2250 INITIAL PRIMARY SYSIEM PRESSURE FOR 1HIS FUN 16 00 AMOUNT OF SUPENHEAT AT EXIT N0Z OF AN OTSGI IGN0kED FOR ss U-TUBE STEAM UENERATORS sa as ssssssss** stssassssssssssssssasts**stssatastssstasasssssssssstassssss**
- PRIMARY SYSIEM
**ssssssstas**ssassassessassasts**ss****sattssssssssss****s** sass *** sass ssUNLESS OTHERWISE N01ED ALL ELEVATIONS IN THIS SECTION SHOULD BE
** REFERENCED TO THE LOWEST FOINI 0F THE INSIDE OF THE RV HEAU
*sWHEN A PARAMETER SUCH AS THE VOLUME OF fHE DOWNCOMER IS CALLED FORE asTHE ACTUAL DOWNCOMER VOLUML SHOULDe CF COURSE BE USED EVEN THOUGH THE saMAAF NODALIZATION LUMPS UTHER VOLUMES WI1H THE DOWNCOMER VOLUME (IHE l
. , - - . . _ - - -- - - - - - - - - - '-- " ~ '~ ~
d l A-16 ' .* SLUMPING IS DONE INTERNALLY IN THE CODE) 01 4 NUMBER UT COLD LEGS 02 2.42 INNtR DIAMETER OF A HOT LEG PIPE satsstaatsst**** s**NtWasssassassassssss: 03 - INSIDE RADIUS OF THE CYLINURICAL PANT OF THE REACTOR VEbSEL assssssssssssss**sssNEWassssssssssssssss: 04 VOLU.16 WHICH IS INSIDE THE CORE BARREL AND L1LS BETWELN ss THE BUITOM 0F IHE CORE AND 1HE LINE WHICH DENOTES IHE TOP as . OF THE RV HEAD (IE THE BOTTOM OF THE RV CYLINURICAL SECTION) 05 51.325 FLOW AREA 0F CORE PLUS CUME BYPASS AREA i 06 84.01 , VOLUME OF HORIZONIAL RUN OF PIPE IN CNL COLD LEG FNUM sa THE REACIGA VESSEL UUT TO 1HE MAIN COOLANT PUMP 07 .0623 RADIUS OF VESSEL PENtlRATION--IF NO VESSEL PENE1 RATION sa ttG SUME CE PLANTS) USE 1HE ASSUMED INITIAL RADIUS OF sa FAILUME WHEN THE RV HEAD FAILS Duk TO CORIUM ATTACK ANU
*s 3UPPLY 1 FOR 1HE NO. OF FAILED PENETRATIONS IN *MODLL 08 1.2E7 ENERGY INPUT FROM ONt PRIMARY SYSTEM PUMr (WHtN RUNNING) 09 0.0 IUTAL MAKEUP FLUW TO 1HE PRIMARY SYS1EM--UNDER NORMAL ss OPERATION SHOULD EQUAL LEfDOWN FLOW BELOWITHIS IS USED sa MAINLY IN fME TMI SCENARIO AND MOST USERS WILL INPUT ZERol as ss THIS WATER IS NOT SUDTRACTED FROM THE RWST ANU CONTINULS (IF POWEN IS AVAILABLE) UNTIL MANUALLY SHUT OFF 10 540. TEMPERATURE OF MAAEur WATENs IF ANie GIVtM IN 09 11 2 2917 INNtR DIAMEIER OF A COLD LEG PIPE
- 12 28.35 ELEVATION OF THE N0ZZLE WHICH ATTACHkS THE SUNGE LINE -
as TU THE HOT LEG--fMIS MUST BE GREAfkR 1HAN IIEM 47 ssNOTE: If IS HELPFUL IN LOCAS (ESP SMALL BREAKS) TO AVOID
** PUTTING INE BNLAK ELEVAi!0N IN THE VICINI1Y OF THE SURGE LINEl asART!FICIALLY INCREASING THE s* ELEVATION OF THE SURGE LINE 0.5-1 METER OR SU ABOVE THE BREAK IS SUGGESTED asFOR1HERe IT IS HELPruC'f0 AVOID PUITING BREAKS NEAR 1HE ELtVATION OF IHE ssTUDESHLET IN U-TUDE TYPE $/G. PRIMARY SYSTEMS--BOTH OF THESE MEASUNES ssHELP AVOID WATER SLOSHING INTO AND OUT OF NODES (WHICH CRANKS INE TIME STEP
! ssDOWN) ANU WILL GREATLY DECREASE RUNNING TIME Af NEGLIGIBLE LUSS OF ACCUNACY l sssssssssssssssssas*NEWas*******s**stass: I 13 6 ENIER BROKEN LOOP BREAK LOCATION KEY (NODE NO.)! ss 3--BROKEN HOT LEG NODE i as 4--BR0ktN HOT LEG 'TUDE' N0DE (B ANU W ONLY) as ss 6--8R0 KEN INlERMEDIATE LEG NODE (StfWEEN PUMP AND COLD SIDE OF
$/G) .
i as 7--BROKEN COLD LEG NODE (HORIZ PART OF COLD LEG) ! as - 8--D0kNCOMER NODE (IE DOWNCOMER PLUS LOWER HEAD) 14 0. BROKEN LOOP BREAK AkEA (Ffss2) IS 27.347 BROKEN LOOP BREAK ELEVATION--SEE NOTES ABOVE 16 VOLUnt IN A COOLANT LOOP (BOTH COLD LPGS FON FLANIS WITH st sa iWO COLD LEGS PER HOT LEG) WHICH IS UNutR A HORI20NTAL 17 LINE MAWN THNOUGH THE BOTTOM OF A COLD LEG N0Z4LE ss MAX VA UME OF WATkR IN ONE COLD LEG WHICH WILL STILL ALLOW GAS TRANSFER TO OCCUR PA3T THE LOWEST FART OF The CULD LEG 18 TOTAL VOLUME OF ONE COLD LEG 19 TOTAL VOLunt OF ONE HOT LEG 20 TOTAL PLUID VOLUME OF fME RX VES$tLe IE IHE VOLUME NOT as INCLUDING THE CORE ITSELF OR INftRNAL STRUCTUNES P
**21 SAS FLOWHA1E OF REACTOR HIGH POINT VENT (S)eIF ANie At s* NOMINAL SYSTEM PRESSURE s
' DOWNCOMER IS M00tLLED AS LNDING AT IHE POINT WHERE IHE LOWER HEAD as0F THE RV MEETS THE CYLINURICAL SECTION--NOTE THE C0kE 5ARREL IS
*s4LSO ASSUMtD TO STOP AT IHIS F0!NT 22 TOTAL VOLUME OF DOW4 COMER 23 PORTION OF DOWNCOMER VOLUME kHICH IS BELOW IHE i
9 4
_ _ . _ .. _ _ m:,-._ _ m _. . A-17
** ELEVATION OF THE BOTTOM OF THE COLD LEG N0ZZLES 24 3
< l*
- kNTkR A 3 FOR PZR TO BE IN BROKEN LOOPI 9 TO BE IN UNBROKEN
** LOOP FOR U-TUDE GEONETRIESI USE 4 AND 10 RE5FECTIVELY Fuk 25 4 8 AND W PLANTS (NODE NO. OF PRIMARY SYSTEM SURGE LINE N0Z)
NUMBER OF HOT LESS 26 0 10 VOID FRACTION AT WHICH REACTOR COOLANr PUMPS TNIP OR Fall
** SCRAM SEIPOINTS: IF A OIVEN 1 RIP DOES NOT EXISTS INPUT A VALUE WHICH THE
** CODE WILL NtVER CROSS -
27 1900. LOW PRESSURIZER PRESSURE TRIP POINT 28 12400. HIGH PRESSURIZER PRESSURE TRIP POINT - 29 -381.7 HIGH LOOP DELTA-T SCRAM SETPOINI MINUS 459
** SO THAT MAAP GETS 78 F AFTER CONVERSION 30 -100. LOW PRESSURIZER LEVEL TRIP----THEME IS NOME HENE 31 48 2 HIGH PRESSUAIZER LEVtL TRIP 32 5.556E-4 REACTOR TRIF DELAY TIMt 33 35.83 LOW S/G WAikR LkVEL SCRAM SETPOINT 34 5 NunsER OF POINfS IN MAIM C00LAMI PUMP COAST-DOWN CONVE
** (5 MAX) 35 3.55E7 FIRST MASS FLOWWATE IN MCP COAST-DOWN CUNVk(MUST BE THE ONE Fulf FLOW UNDtR NOMINAL CONDITIONS) 34 3.23E7 SECOND FLOWNATE 37 2.48E7 NEXT FLOWRATE
, 38 1.77E7 NEXT FLOWRATE 39 1.10E7 NEXT PLOWRATE 3 40 0.D0 FIRST TIMt IN COAST-DOWN CURVE--MUST BE 0 41 2.778E-4 NEXT TIME la b*.5kb TTIMk 44 5 833E-3 NEXT TIME .- - 45 33.13 ELEVATION OF SOTTOM'0F S/G TUsESHkET A80VL BOTTOM OF'RV
** (IGNORED IN 8 AND W PLANTS) 44 .50 THICKNESS OF RV MkAD 47 26.20 ELEVATION OF INE BASE OF INE COOLANT LO'OP N0ZZLES
** (DISTANCE FROM BOTTOM OF N0ZZLES TO BOTTOM OF RV 4 AD) 48 64 VkRTICAL DISTANCE FROM LOWEST POINT OF A COLD LEG TO THE
** ELEVATION OF THk BASE OF THE COLD LEG N0ZZLt ON THE RV 49 VULUME OF fHE HORIZONIAL RUN OF A HOT LEG PIPE 50 0.0 TOTAL LETDOWN FLOW--SEE NOTE NEAR MAKhur FLOW EN1RY ABOVE 51 35.5 NORMAL DIFFERtNTIAL PRESSUNE FROM COME INLET TO HOT LkG
** SIDE OF OUTLET N0ZZLES WHEN MAIN C00LANf FUMPS ARE ON
*************ALL THE RtMAINDER IN 1HIS SECTION ARE NkWS***
**MOST USERS WILL USE THE 'UN8ROKkN' LUOP BkEAK UNLY FOR PUMP SEAL LOCAS l **IN iMLB SEQUENCES 3 IT CAN ALSO BE USED FOR SPECIAL PURPOSES (EG LOFT FP/2
** SIMULATION) asTHIS BREAKe ALONG WITH THE BROKEN LOOP BREAK IS CONTROLLED BT EVtNr CODE
**2091 ONE CAN TURN THE BMEANS ON AND UPF SEPERAILLY SY USIN3 A FARAMETER
** CHANGE-TYPE INTERVENTION (CODE 1000---SEE V0L 1 0F USER'S MANUAL)
, 52 12 LOCATION KEY FOR UNBMOREN LOOP BREAKe IF ANY n ss ? --UN W QKEN HOT LIG NODE
** 10--UNBROKtM HOT LEG '1UBC' NODE (3 AND W QNLY) st 12--UN8ROKEN IN1ENMLDIATE LEG NODE--
** MTE BREAK IN UNBROKEN LOOP COLD LEG OR C ** DMGMLR NOT ALLOWtD AT THE PRESENT 53 0. AREA 0F UNBROKEN LEG BRtAK--PUT IN ZERO IF NONE
**54 35 ELEVATION OF UN8ROKEN LOOP 3REAK (SEE NUTES FERTAININO
** TO BREAK ELEVATION ABOVE
**THE ' DOME' REFERS TO THE REGION A80VL THt UFFEN PLENUM
**lHE ' DOME PLATE' IS THE PERFONAIED PLATE THAT DIVIDES IHE UPPER PLENUM
' **FROM 55 THE DOML--SEE DRAWINGS IN THE PRISYS SECTION OF THE USEM'S MANUAL ELkVATION OF THE RV DOME PLATE I
_ 1 - .m . . _ . ._ _s _ _ . . _ _ _ _ . _ . _ . _ . _ _ _ _ . _ . . A-18 56 ELEVATION OF THE INSIDE OF THE RV HEAD 57 ELEVATION OF IHE RV FLANGE (CLOSURE STUDS) as (NOTE THAT THIS FLEVATION IS 58 59 U D Af4 T M X Ob MASS OF fME CORE BARREL BELOW THE ELEVATION OF THE TOP OF ss THE CORE (' LOWER CORE BARREL') FROM IMF DATA + MAAP2 FILE N P E RE - 2ZOb NO 5 61 MASS OF UPPER PLENUM INTERNA 62 MASS OF THE RV DOMt PLATE - LS MAAP2 --CALC ZIUM MAAP2 ZION CALC NOTES NOTES 63 MASS OF 1HE WALL FORMING 1HE EXTERIOR OF lHE DOME (IE as INCLUDES THE RV CLOSUME HEAD) FNOM MAAP2 FILE 44 TOTAL MASS OF ONE HOT LEG + HOT INLET PLENUM WALL OF THE S/G+ ss THE TUl($HkET MASS ASSOC WITH THE INLET PLENUM 65 TOTAL MASS OF ONE COLD LEG + COLD QUTLET PLENUM OF INE S/G
** PLUS THE TUsESHtET MASS ASSOCIAltD WITH THE QUILEf PLENUM
** NOTE: FOR PLANTS WI1H 1WO COLD LEGS PER CUILET PLEMUMe
** ADD ONLY HALF THE OUTLET FLENUM MASS--THt OlHtR HALF IS
** IHEN ASSOCIATtD Wil'H IHE UlHER COLD LEG IN 1 HAT LOOP 66 MASS OF THE RV WALL (BELOW THE RV FLANGEi THE DUMt WALL as ENTERED ABOVE STARTS AT IHE PLANGE) FROM MAAP2 FILE 67 WATER LINE AREA IN THE UFPER ILENUM (ABOVE THE Cuke ANU SELOW ss ThE DOME PLATE) -- ESTIMATE WIlH D=12Fre 1/2 WATER 68 HYDRAULIC DIAMETER IN THE UPPER PLENUM - INP
~
69 TOTAL HEAT TRANSFtR AREA 0F IHE UPPER PLENUM INTERNALS 70 CONVECTIVE (NON-RADIA1IVE) HLAT LOSSES UNDER NOM CONDITIONS sa FNOM STEAM GENERATORSe PRES $URIIERe AND NEST OF PRIM. STS. Es NOTER DETAILED CALCULATIONS INDICATE THAT UNDER NORMAL
** OPtRATIONe IHE PRIMARY SYSTEM HEAT LOSS IS DUE VIRTUALLY ss ENTIRELY TO UNINSULATED FARTS OF THt SYSTEM (LOSS THNUUGH
*n INSULATION IS NEGLIGBLE)I 1Hyl iHIS NUMBER ss SHOULD BE APPROXIMATELY THE TOTAL NOMINAL PRIMAM
** SYSTEM HEAT LUSS (SEE IDCOR NEPORT 85-2 FOR DISCUSSION) 71 NO. OF PLATES IN PRIMARY SYSTEM REFLECTIVE INSULATION OR:
as tiNTEM 0 FUR CALCIUM SILICATE BULK INSULATION OR ss ENTER -1 FOR ROCK WOOL INSULATION--IF YOU HAVE A
** DIFFERENT BYPE OF INSULATION YOU SHOULD CONSIDER MODIFYING
** FUNCTION THCBUL WHICH SUPPLIES THE THtRMAL CONDUCTIVITY 7" TOTAL 1HICKNESS OF INSULATION 7I' ELEVATION OF THE BASE OF THE CTLINDRICAL PANT OF THE RV 74 VOLUME OF IHE LOWER HEAD OF THE RV 75 TOTAL HEAT TRANSFER AREA 0F LUWtR CORE BARREL / THERMAL
** SHILLDS (IE 1 HAT PORTION BELOW 1HE TOP OF fME CUME) 76 TOTAL MtAT TRANSFER AREA 0F UFFER COME BARREL sa as as assassssssssss** sssssas**** sts******* ss***ss*************************
- PRESSURIZER sasssssssssssssssssssssssssssssssssssssas s***ss*************sss********
b03 2235. NkhbkR -SECTIONAL AREA PRESSURIZtR HEATER PRESSURE SETPOINT 04 2325. FRESSURIZER SPRAY FRESSUkE SEfPOINf 05 7.8 WATER LEVEL BELOW WHICH PZR HEATERS TRIP 04 4 14E6 PRESSUKIZER NtATER TOTAL OUTFUT--IN MAAP THE HEATEki as ARE EIINER ALL ON OR ALL OFF 07 3.47E3 SPRAY SYSTEM FLOW RATE 08 4.2E5 FLOW RA1E OF SAFE 1T VALVE AT ITS SETPOINT
__. __ _ ~ ._ .
, .. ..._._ _ _ __ _ .m __ _ , _ . _ , ,
A-19 - 09 2500. LOWkST SETPOINI 0F A SAFETY VALVE (OFENING PRESSURE) 10 2500. HIGHEST SEIPOINT OF A SAFElf VALVE (OPtNING PRESSURE) 11 0.93 DIAMkTER OF THE SURGE LINE 12 44.25 ELkVATION OF SPRAY HEAD ABOVE BOTTOM OF PZR 13 64. LENGTH OF THE SURGE LINE 14 3 NUMBER OF SAFELY VALVES 15 3.281E-3 NOMINAL PZR SPRAY DROPLET 16 2350. LOWEST SET POINT OF PORY (OPENING PRESSURE) 17 2350. HIGHEST SET POINT OF PORV (OPENING PREShuME) 18 2 NUMBER OF PORVS 19 2.1E5 NOMINAL FLOWNATE OF A PORV AT ITS SETPOINI 20 1 654t3 EMPlY MASS OF PZR STEEL 21 0 ENTER A 1 IF THE SURGE LINT HAS A LOOP SEAL (EG TMIls ss 1HIS PREVENTS COUNTER-CURRkNT DRAINING OF PMESSURIZER
** THMOUGH SUNGE LINT WHEN THE PRIMANY C00LANI LOOP SIDE
** IS VOIDED (SEE WRIIEUP FOR SUSHOUTINE DRAIN) 22 38.5 SEDIMENIATION ARLA assassass***s4LL l'HE RLMAINDER IN IMIS SECTION ARE NEWassa 3 PRESSUNIZER RELIEFS ARE ASSUMED TU CLOSE AT PRESSUKE PSET-FDEAD WHExE PSET IS 1HE OPENING PNESSURE DEFINED ABOVE AND PDEAD IS GIVEN BELOW 23 100 DEADBAND ON PRESSUNIZER SAFETY VALVES
. 24 100 DEAD 8AND ON FRESSURIZER PORVS th
***sas:s******sas:s*****s ssassas**ssssssssssssssas*******************st sSTEAM OtNERATOR (VALUES REFER TO ONE UNIT) sassassassassassassassassassassas***sssssssssssssastsas***ssassassssssas 01 S900. TOTAL SECONDARY SIDE FREE VOLUMEe EG QUT TO THE MSIV'S
'02 DOWNCOMER CR0hS-SECTIONAL FLOW AREA -
03 TUFC SUNDLE (SECONDARY SIDE) PLOW AREA 04 0.D0 8 AND W ONLY--ELEVATION OF AUX FLED SPRAY HLAD A80VL st 1HE TOP GF 1HE LOWER IUBESHEET 05 1.158E6 INITIAL MASS IN CONpCNSATE STORAGE TANK--OR A LARGE sa NU. IF NO LIMIT ON AFW SUPPLY 04 2-PHASE WATER LEVLL IN TUWE PUNDLE AT THE SEC SIDE
! Hh! US D bAJUST L V0f kkAC N D S R PU N
- IN THE IUSE BUNDLE SO AS TO APPROXIMATELY MAKE UP FOR sa SIMPLIFICATIONS IN THE MAAP MODEL I THE CDRRECTION
** SHOULD MOST IMPACT LOSS UF FELD SE00LNCES 07 440. MAIN FEEDWATER TEMPERATURE 08 1199.7 LOWEST SETPOINT OF SECONDARY SAFE 1Y VALVES N11 9.19E5 k Ob Lhkh hG NOMINAL FLOWMATE OF A SAFETY VALVE AT THE SETPOINI i
12 1135. SETPOINT OF SEC RELIEF VLV (ASSUMtD SAME FOR ALL RtLIEFS) a:IF NO ' RELIEF VALVES'--SUFFLY A SET POINI PRESSURE HIGNEN THAN THL
** SAFETIES AND USE THE RELIEFS AS MANUALLY CONTHOLLED STEAM DUMPS 13 1 NUM8ER OF RELIEF VALVES FER S/G 14 4.0E5 NOMINAL FLOWRATE OF A RELIEF VALVE IS 3 97E6 MAX FEEDWATER FLOWNATE PER S/G sassassstaat: DEFINITION CHANGED FOR OTSG'S******ssats:
as!NCLUDE THAT PORTION OF THE TUBE VOLUME WHILH IS NOT CCOLED (IE IS INSIDE aslHE 1U8ESHEET(S)) IN IIEM 14 16 318.8 TOTAL (BOTH PLENA FOR 0T50'S) PRIMARY HtAO(h) VOLUMt-- ssMAIN STEAM ISOLATION VALVE (MSIV) CLOSURE: MAIN FEEDWATER SHUT 0FFe
**AND AUX FEEDWATER ACTUATION ARE ASSuntD TO OCCUR AT REACTOR SCRAM j s UNLESS DEFEAlED WIlH APPROPRIAIE tVtNT CUDES i
17 .0028 TIMt DELAY FOR ACTIVATION OF AUX FEED AFTER SCRAM j 18 .00138 TIME MEQUIRED FOR MSIVS TO LINEARLY RAMP FROM OPEN f l
___m , . - _ .m _._ c-- -- - -- A-20 l sa TO CLOSED 19 944 1 TOTAL PRIMARY SIDE VOLUME OF ONE STEAM GENERATOR 20 1.78E5 MAXIMUM AUX FEED FLOWRATE PER S/G 21 100. AUX FEED TEMPERATURE 22 5426 NUM8ER OF l'UBES IN A STEM GENERATOR 23 0.0033 THICKNESS OF STEM GENEMATOR TUsES 24 0.05047 ID OF STEAM GENERATOR TUBES 25 9 35 THERMAL COMUUCTIVITY OF STEAM GENkRATOR TUBES 26 7.93AE9 ss IHROTfLED FLOW PER STEAM GENERATOR FOR AFW SYSTEM OR LARGE ! ** NUMBER IF FLOW NOT THNOTTLED (SEE DISCUSSION AFIER ENGIN. SAFEGUARDS ITEM 151)
'7 1. FRACTIONAL AREA USED FOR STEAM DUMPS IN BROKEN LOOP S/G 38 1. FRACTIONAL AREA USED FOR STEAM DUMPS IN UNBKN LOOPS S/GS
- STEAM GENERATOR WATER LEVkL CONTROL (SbMLC) SYSTEM PARAMETER $;
29 34.8 DOWNCOMER PROGRAM WTR LkVEL FOR STEAM GENERATOR WATER t as LEVEL CONTROL SYSTEM IN PR0KLM LOOP S/G 30 34.8 DOWNCMR PROG WTR LVL FOR SGWLC SYSTEM IN UNBAN LOOP S/GS 31 STEAM GENEMATOR TUDESHEET DIMtTER ssFOR 34.8 ACCIDENT SIMULATION IT WAS NECESSARY 70 INCORPORATE A SANG-BANG asMODE OF S/G WATER LEVEL CONINGL--IE OPERATOR CONINGLS THE WATER LEVEt.
- IN AJ4.8CILLATORT WAY WIfMIN A DEAD 8ANDI MUST USENS WILL NOT WISH
**TO 32 USE 0.0 THIS MODE AND SHOULD LEAVL THE NEXi THNEk ENIRIES EDUAL TO O 8-LOOP S0WLC DEADBAND 33 0.0 A-LOOP SGWLC DEADSAND (NONZER0 VALuk ACTUATES as BATCH FEED MODE) 34 0.0 FOR BANG-BANG MODEe-THE MINIMUM AFW FLOWNATE PER S/G 70
** BE USED ON THE-DECNEASING CYCLE assssssssass ALL THE REMAINDER IN THIS SECTION AkE NkWstas 35 0 MAIN STEAM LINE SREAKS CAN BE SIMULATED; ENTER 0 FOR NO
** MAIN STEAM ~ LINT BREAKI 1 DIRECTS STEAM FMUM WOKEN as LOCP $/G TO CONTMil 2 DIRECTS STEAM FROM ALL S/GS TO as
[h / AM N) A$ NO [3 THIS PARAMETER 36 TOTAL HEIGHT OF S/G SHELL AB0VE TUBESHEET 37 ss MASS OF S/G SHELL--DON'T INCLUDE MASSES ASSOCIATED WITH PRIMARY l as HtADS OR TURESHEETS WHICH ARE LUMPED WIIM IHE ASSOCIATED COLD i AND HOT LEG MASSES IN THE PRIMARY SYSTEM SECTION I 38 12 as NUMSER OF PLATES IN REFLECTIVE INSULATION ON S/G SHkLLS OR
** CODE INPUTINu!CATING NO. 71) QTHER INSULATION TYPE (SEE FR1 MARY SYSTLM f I
as as sasasssssssssss****ssas****ss******************s*******sstastassassssses
- TIMING DATA sassassassassssas**ssssss**ss****sas******s3**s******sts***s**ssssas****
3301NOTUSgD ON MAX TIME STEP (ALWAYS INPUI IN SECONDS) 04 .005 MINIMUM TIME STEP (ALWAYS INPUT IN SECONDS) .
** TIME SELECTION ALGORITHMS ARE EXyLAINED IN THE WNITE-UPS FOR SUSROUf!NtS
*s!NTORT (T/M MODELS) AND INTGFP (FISSION PRODUCT MODELS) 05 .05 RELATIVE MASS CHANGE USED TO SELECT TIME STEP 04 .02 sa MINIMUM INTER-NODE FISSION PROD MASS TRANSPER CONSIDERED WHEN 07 P!CKIN6 TIME STEP IN FISSION PRODUCT MUDELS
.02 RELATIVE GAS TEMPtRA1URE CHANGE USED TO SELECT TIME STEP 08 .1 3: REL MASS CHANGE FOR FISSION PRODUCTS USED TO SELECT TIME j as STkP IN FISSION PRODUCT ROUTINES
. . . __ = _ _ - - . - . _ - .
_~. ~ . - - . . ~- - . . . . - - . . . - - - . - . - A-21 sssssssssss**ssssssssssssssas** ssssss**ssass*** ssssas*****sssas****ss 400ENCH TAMK ('OT* COMPT) asssssssssssssssssssssss***sassassassssssssssssss****ssas***stasassss s* 01 1800. TOTAL VOLUME INCLUDING THAT OF 1HE INITIAL WATER MASS 02 ~83495. INITIAL WATER MASS . 03 105.7 FAILURE DIFFERENTIAL PRESSURE OF RUPTURE DISK 04 12 5 HEIGH 1 0F RUPTUNE DISK ABOVt BCOMPT FLOON ss05 SEDIMENTATIQN AREA--NOTE AS SOON AS RUPluRE DISK FAILS: HOLD-UP ss IN THE GAS srACE OF THE QUENCH TANK IS NOT MODELLED ss ss sS1 assassssssssssssssssssssss**ss**ssass***ssssssssssssssssssssas**sassass: 3MODkL PARAMETERS sassssssssssssssssss************ssas**ssassssssssssssss*********s***ssas L
*sSEE DISCUSSION IN VOL 1 0F USER'S MANUAL FOR ALLOWABLE LIMITS ON
*MODEL PARAntTER VALUES AND THE DIFFERENI SENSITIVITY ANALYSIS MODkS ss as' SCALE FACTORS' MULTIPLY MODEL PREDICTIONS OF FLOWNATES ETC.-
!! * !!ssk as bEYO N bVE BEEN CHANGEDasts 01 .005 CORIUM FRICTION COETFICIENT FOR VESSEL ABLATION HEAT
*s TRANSFER (REYNOLD'S ANALOGY) CALCS 02 .002 LEAK-BEFORE-BREAK CONIMT LEAKAGE AREA (IF THE CONTMT STRAIN as MODEL IS NOT USEDs THIS IS THk AREA USED WNkN THL CUNIMI PRESS as EXCEEDS THE UskR-SUPPLILD FAILURE SIRES $)
03 60. TIME TO FAIL Vt'SSEL P(N. WELDS AF1ER CONinCT WITH CM 04 0.01 INE CONIMT FAILystt ARtA USED IF THE CONIMT FAILS DUE TO OVER-3: FRESSUNIZATION IN A GROS $ MANNkR (IE SCFORE THE STNAIN
** CRIIERION IS REACHED AT PENETRATIONS)) USkD.ONLY IF DETAILED
** CONIAIMMENT FAILUME MODEL ACTIVATED OS 2.0 MULTIPLIER OF NORMAL CLAD SURFACE AREA USED IN OXIDATION
** CALCS TO'.ACCOUNf FOR STEAM INGRLSS AFTER *
, as CLAD RUPlUNE (MUST BE Bt1WLEN 1 AND 2) 06 983.D0 CRITICAL FLAME TEMP AT ZER0 STEAM MOLE FRACTION
*s USED IF NO IGNITION SOURCESI THIS IS MULTIPLIED BY THE 3: WESTINGHOUSE FLAME TEMPERATURE MULTIPLIER CURRELATION sa FOR NONZERO STEAM MOLE FRACTIONS l
'. 07 1. SCALE FACTOR FOR FISSION PRUDUCT ANU INERT AERO RtLEASE RATES
** FROM CORE (SHOULD USE A N0. LESS IHAN OR EQUAL TO 1) 08 300.0 NON-RADIATIVE FILM BOIL. Mr. TRANS COEFF FROM CM TO POOL
, 09 850. NAT. CIRC. (MCP'S OFF) S/G PRIMARY SIDE FILM RESISTANCE i ** WHEN 2- OR 1- PHASE NATUNAL CIRCULATION IS OCCURING
** IN lHE COOLANT LOOPS-NOTE IMAT COOLANT VELOC11Y AND as VOID FRACTION DISTRIBUlION ARE NOT COMPUlLD UNDER THESE COND.
10 .5 FRACTION OF S/0 TUBES CARRYING '0UT' FLOWS IN 1HE HOT LEG ss NATUNAL CIRC MODEL (SEE SUWROU11Nt HLNC WHITE-UP)I IF YOU as WISH TO FORCE THE FLOW OFFe USE 0 (1HIS REQUIRES BYPASSING 3: PARAMATER CHkCKING SY USING THt: SENSITIVITY ANAL OPTION
** IBATCH=2)I PARAMETER DOES NOT AFFECT B AND W GEOMETRY UNLESS ss 0 IS INPUI SINCE OTSG TU145 DUN'T PARTILIPA1E IN FLOW p 11 .1 8 AND W ONLY1 FRACTION OF S/G TU8ES STRUCK BY AeW i 12 1.D3 MT. TRANSFER COEFF BETWLEN MOLTLN CURIUM AND A FROZEN CRUSTI ss USED IN DECOMP AND IN CALCULATIONS WITHIN A MOLTEN POOL IN '
3: THE CORE 13 0 ENTER A 0 FOR ENTRAINMENT FROM C TO Bf 1 FOR C TO A--LATTER !
** uENERALLY USED ONLY IF CAVITY HAS NO INSTRUMENT IUNNEL i 14 0.D0 IF 13 IS NONZERO: FRACTION OF THE ENfRAINED MASS WHICH )
as STRIKES IHE MISSLE SHIELD (BEFORE SIGNIFICANILY INTERACTING ' s WITH THE UPPER COMPARTMtNI GAS) l i
_ . _ _ _ . . _ . _ _ 1 . --- u A-22 15 1 00 DRAG COEFFICIENT OF RISING PLUME DURING BURNS IN UPPER
** COMPT- LARGER VALUES RESULT IN A SLOWER AND FATTER PLUML
**- ANS THUS INCREASE 1HE EFFICIENCY OF THE IGNITERf kh k '. b OMPT 18 1.00 SAME FOR D COMPT 19 1.00 SAME FOR U COMPT 20 1 33 CHURN-TUNSULENI CRITICAL VtLOCITY COEFFICIENr 21 3.7 DROPLkT FLOW CRITICAL YkLOC11Y C0 EFFICIENT 22 1. SPARGED POOL VOID FRACTION COEFFICIENr
$ h b1 E RffMtN UNS T OF
.5 .9 EMISSIVI1Y OF WATER 26 .85 EMISSIVITY OF WALLS 27 .85 EMIS$1VI1Y OF EQUIPMENT 28 .85 EMISSIVITY OF CORIUM SUNFACE 29 .6 EMIES!VITY OF GAS 30 .3 CORE MfDR0 DYNAMIC LIMIT KUTATELADtE NO. FOR RtFLOGDING M1'
** AND XIDATION CALCULATIONS 31 .33D0 NUM R TO MULTIFLY AulATELADZE CRITERION BT TO REPRESEN1
** DIFF CULTY (GT 1.00) UR EASE (LT 1.00) FOR DESRIS TO GET
** OUT OF CAVITY 32 3.0 P'LOODING CRITICAL VELOCI1Y COEFFICIENT 33 .14 FLAT PLATE CHF CRITICAL VtLOCITY COEFFICIENT 34 1. NUMBER OF VESSEL PENEfRATIONS 'lHAT FAIL 35 .75 DISCHARGE. COEFFICIENT FOR PRIMANY SYSTEM PREAK(S)
*** SCALE FACTORS
- MULTIPLT MODLL PREDICTIONS--lHE DEST-ESTIMATE VALUE
**!$ USUALLY 1 -
36 1.D0 SCALE FACTOR FOR BUNN VELOCITY CORRELATION 37 1.00 SCALE FACTOR FOR HEAT TRANSFER COEFFICIENTS TO PASSIVE
** HEAT SINKS - -
38 2.5D0 GAMMA SHAPE FACTOR (TO ACCOUNI FOR NON-SFHERICAL SHAPES IN
** 1HE COAGULATION EQUATION) UstD FOR AERos0LS 39 1.D0 CHI SHAPE FACTOR (TO ACCOUNr FOR NON-SFHtkICAL SHAPES IN
** STOKES LAW) USED FOR AEROSOLS 40 3.D4 RATIO 0F AIRBORNt AEROSOL MASS TO THg MASS WHICH WOULD LEAVt
** YOU IN STEADY-STATE WI1H 1HE CURRENT SOURCE STRtNGlHilHIS IS
** USED TO CONrM0L THE SELECTION OF DECAY VS STEADT-STATE Atx0 SOL
** SETTLING CORRELATIONS 41 10.D0 DECONIAMINATION FACTOR ASSOCIATED WITH THE FASSAGE THkOUGH 1
** MtTER (REFkRENCE LtNG1H USED FOR E11HER SET OF UNITS) 0F uATERI
** ASSUME DF IS LINEAR FUNCTION OF DtFTH FOR OTHtR DLPTHS 42 .02 CAPTURE EFFICIENCY OF CONTMT SPRAY FOR AER050LS--lHIS IS
** THE FRACTION OF THE TOTAL VOLUMt SWtPT fY FALLING D>.0FS WHICH
** IS CLEANSED OF AEROSOLS 43 1.D0 ABSOLUTE VALUE OF THE DtSIRED MULTIPLIER OF CbI AND
** CSON VAPOR PNESSURE--ENitR A NEGATIVE NUMBER TO SELECT
** JAMAF CSON FUNCTIONI FOS FOR SANDIA CURELLATION (3EST-EST) 44 .100 FRACTION OF CLAD OXI11IZED WHICH CAUSES CORE TO COLLAPSE ON
** REFLOOD (GIVES SMALLER KU FOR HkAI TWANSFER THAN INIACf
** MODEL) AND CAUSES CORE GEOMEIRY TO CHANGE 45 0.D0 FOR S AND W UNITS ONt.Ye FRACTION OF FtRFtCT CUNUENSA1 ION
** OF STEAM EN1ERING DOWNCOMER INROUGH PLAFFER VALVES
*S!
46 2100. TEMPERATUNE AT WHICH CLAD FAILS IF IT HASN'T ALREADY RUFTUh.EDI
** IHIS HALIS PUR1HER BALLOONING AND ALLOWS FISS FROD RELEASE 47 2.SE5 LATENI HEAT OF y-IR-ZR02 EUIECTIC 48 .25 VUID FRACTION OF A COLLAFSED CORE
. 1
A-23 49 .3D-4 SEED RADIUS ASSUMED FOR NfGROSC0FIC AEROSOL GROWiH CALC 50 -2 EN1ER A 2 FOR FISSION FRODUCT xELEASE TO BE COMPUTED at BY THE IDCOR/kFRI STEAM OXIDATION MODELi 1 FOR
** NUREG-0772 MODELI NEGATIVE NOS. ACTIVATE IHE SAME MODEL
** AS POSITIVE NUMsERS BUT ALSU TUNN ON A BLOCKAGE MCI'EL
** WHICH REDUCES INE RELEASE OF NONVOLATILE FISSION PRODS 88 WHEN THE NODE IS BLOCKED FOR GAS TKANSFOR1 -
51 0 ENTER A 1 IF TELLURIUM IS RELEASED IN-VESStLI O IF IT
** IS ASSUMED TO BE TOTALLY BOUNu UV WITH 2IRCALLOY
** (o IS *EST-EST) kk ss h.
2ALCSI R AL > ISCANBEESf!MATEDBYF=2.*DP*kN0/G3*2WNtRLgALLUSED FOR UPPER P
** VALUES ANE FOR NORMAL OPERATION WI1H MCP'S ON) DP=COME
** PRESSUNE DROF, RH0= DENSITY OF PRIMARY SYSTEM C00LANte Ga 4
** CURE AVERAGE MASS FLOW PtR UNIT AREA (IN BRIT UNITS.
** INCLUDE G0 AND OTHER NECESSARY CONVtRSIONS TO MAnt F sa DIMENSIONLESS)--USE GT 100 TO ARTIFICIALLY STOP PLOW
** (REC'JIRES USING THE SENSTIVITY OPTION IBATCHa2) 54 0. INStRT 0 IF IN-VESSEL NAluRAL CIRCULATION PLOW REIURN LEG sa IS IN DUTER FUEL ASSEMWLIES (USUAL CASE)i1NSERT 1 IF RETUKN
** IS D0bN ' BYPASS' (IE BAFFLE-CURE BARREL ANMULUS)--1HIS UQULD
< ** BE EXFECTED ONLY IF THERE WAS A LOT OF FLOW AhEA IN THE
** B1 PASS (EG PERHAPS B AND W PLANTS) 55 .1 A VOID FRACTION: BELOW WHICH A CORE N0DE IS AhSUMED BLOCKED
** FOR GAS PLOW OR OXIDATION 34 10. NO. OF SAMPLES AVLRAGED OVtN IN NC MCDEL (SEE UhER'S MANUAL) 57 .25 CROSS-FLOW PRICTION COEF IN NC MODEL (LIBERA 1UME SAYS .25 .45) 58 .05 FRACTION OF XENON INVENTORY IN THL'FtLLET-CLAD GAP DUE TO at
** LONG-TERM OPERATION (OFTEN CALLED IHE ' GAP RELEASE's 1HIS
** IS USED IN CACCULATING THE FKEShuRE INSIDE THE FUEL PIN FOR 84LLOONTM6 CALCS--NUMEG 0772 SAYS OBSERVED VALUES ARE 0-0.25) 59 .33 VOID FRACTION IN FRIMARY SYSTEM ABOVE WHICH THE FHASES
** SEPARATE AND 1WO-PHASE NAluRAL CIRCULATION STUFS .
60 1040.
** TEMPERATUNE OF H2 JEl ENIERING NON-INtRTED COMPARTMkNI WHICH as IS SUPFICIENT TO CAUSE A LOCAL BURN--FRUM HEDL-lME 78-50
************************a**********************************************s
- FISSION PRODUCTS
**********************************************************************sa tt
********************NEWa****************:
** FISSION PRODUCT GROUPING SCHEME:
83 GROUP 1: NOBLE GASSES AND 'INtRT' (NON-RADIUACTIVt) AEnOSCLS
** GROUP 2: USI
** GROUP 3: TELLURIUM (TAKEN TO BE ELEMEd(AL TE)
** GROUP 4: STRONTIUM (TAREN TO BE SRO WI1H BARIUM LUMPED IN AS BA01
** GROUP 5: MOLYBDENUM (TAAEN TO BE M003)
** GROUP 6; CSOM
** STRUCTURAL MATERIAL GROUPING SCHEMt
**USED IN CORE NODES (THACKED IN CONIAINMtNT AS LUMrED GROUP 1 AEROSOLS)
** GROUP 1: CD asGROUP 2: IN .
** GROUP 3: AG ssGROUP 4: SN
** GROUP 5: MN sa sSS$33333838 384*SESNEWas*S$$$$t33****S*:
4 6
.- ..,,.-w... - _ . _ - ,_,,.,,,...w.. _ _ . - ,_.__,_._,-,.w,,,m%.,_.w,.,-e,-y,,,,,ym.,,--.,
m,-- ,.w,-,,,,,me-...- _ _ - ,
. . . _ . . . ~ . -- -- - - -- --
A-24 01 .0428 FRACTION OF FISSION PRODUCT POWER IN GROUP 1 02 .222 SAME FOR GROUP 2 03 .04A7 SAME FOR OROUP 3 h5 04
'.bb
.0415
! b$
SAnf FOR GROUP 6 07 433. INITJAL MASS OF FIS$!ON FR0 DUCTS IN GROUP 1 (NOBLES ONLY) 08 23.49 INITIAL MASS IN GROUP 2 09 40.92 GROUP 3 10 181.5~ GROUP 4 (TOTAL KG OF SR ANu BA EXFRESSED AS OXIDES) 11 552.4 group 5 (LXPRESS AS KG OF TRI-0XIDE)
!$ .' 8 A MASS OF CD IN CORE (STRUC MATERIAL GROUr 1) 14 305.4 INITIAL MASS OF IN IN CORE 15 g8484 ENITIAL MASS OF A6 IN CORE if4:18 A L' NOT USED 3fil2 11 er a f H al ss asssssssssssssssssssssssssssssssssssssssssaanssusas********** 34assassess
- AUXILIARY BLUILDING ssssssssssssssss*********ssassastsaassas****sssssssstas***sssssssassassas asCAN M0 L A MAXIMUM OF 5 SERIALLY-CONNECTED NODES asTHESE R CEIVE FLOW FROM THk CONIAINMENI FAILUME ANue IN V-SEQUkMCES:
ssFROM T PRIMARY SYSTEM SREAK(S) ssYOU NEED SUPPLY INFORMATION ONLY FOR TH( NO. OF NODES YOU SELECTED IN~ s*THE SCONfNOL SECTION--NCIE 1HAfe FOR EXAMPLE 1HE MAES OF WATER ss!NITIALLY IN N0DE 1 ALWAf5 GOES IN INFUf NO. 4 NO MATTER HOW MANT' ssN0 DES.YOU ANE USING - - 01 421.4 VOLR3(!) VOLUME OF NODE 1 02 711.8 N0DE 2 81 1118.i kN ! 8:05 0. NODE 5 asMASS OF WATER CAN SE USED TO REPRESENre FOR EXAMrLE: REFUgLING FOOLS N) Sh0FWATERINNODE1 07 0.0 NODE 2 08 0.0 NODE 3 Of 0.0 N0DE 4 ss10 0 0 N0DE 5 1{ 0.0 AWATR9(!) SUNFACE AREA 0F WATER F00Le IF ANY IN NODE 1
- 1. 0.0 NODE 2 0.b 3:15 0.0 k
NODE 5 844T PRESENT ONLT ONt EXTERIOR WALL FER NODE IS MODELED IN THL AUX CODEI ss!E INE WALL HAS AUX CONDITIONS ON ONE SIDE AND IHE ENVIRONMENT ON 1HE ss0THER-EITNER A STEEL OR CONCRETE WALL CAN BL MODELED BY INVUITING THE asAPPROPRIATE MATERIAL PROPERTIES le 211.8 ANSRS(I) ONE-SIDED WALL ARLA FOR NODE 1 f8 19 710.2 e NODE 4 i as20 0. NODE 5 g .g X g8g!) THICKWtSS FOR NODE 1 WALL 23 .4572 NODE 3 24 .4572 N0DE 4 ss25 0. NODE 5 l
. .. . _ . _ _ _ . . - . . . _ . ___ _ . . - _ _ _ . . . - - . ~ . - - - .__ .- . l 1 l A-25 26 1 59 KHSRB(!) THtRMAL CONUUCTIVITY OF WALL IN NODE 1 27 1.59 N0DE 2 28 1 59 NODE 3 29 1.59 NODE 4 as30 0. NODE 5 31 656.7- CFHSRB(I) SPECIFIC HLAT OF WALL IN NODE 1 32 656.7 NODE 2 33 656.e NODE 3 34 654.7 NODE 4
**35 0. NODE 5 36 2.65 ZHSR8(I) HEIGHI 0F WALL (FOR NC CALC 3) FOR NODE 1 37 4.48 NODE 2 38 14 61 NODE 3 39 8.69 NODE 4 as40 0. NODE 5 41 2308.5 DHSRB(!) DENSITY OF WALL IN NODE 1 42 2308.5 NODE 2 43 2308.5 NODE 3 44 2308.5 NODE 4
**45 O. NODE 5 s*THE VENTILATION (OR 'SGTS') SYSTEM IS MODELED Bf SUFFLTING
- **A FORCED 007 FLOW AND/UM A F0HCED IN FLOW 84THIS FLOW IS ON UNf!L THE FIRE DAnrER SETFu1Ni(SEE ILLOW) IS
**REACHtD IN A COMPARIMkNT--lHIS SHUTS FLOW DOWN IN IHAT C0hPT
** NOTE fME AC F0WER EVtNr CODE DOES NOT AFFECT THE SbTS FLOWI TO w
** SHUT THE FLOW OFF, SUFPLY O'S BELOW OR A LOW FIRE DAnFER ftMP 46 0.0 WVOR8(I) FORCED VOLudtiRIC (Ma*3/fEC OR GFM)
** VENTI 47 0.0 NODE . (ATION FLuW OUT OF NODE 1 48 0.0 NODE 3 49 22.26 NODE 4
**50 0.0 NODE 3 .. ..- .
51 17.96 WVIR9(I) FORCED. VOLUMETRIC VtNIILAll0H FLOW INf0 NODE 1 52 0.0 NODE 2 53 4.30 NODE 3 . - 54 0.0 NODE 4 a*55 0.0 NCDE 5 56 200 6 ASEDRB(I) AEROSOL SETTLING AREA FOR NODE 1 57 200.4 NODE 2 58 200.6 NODE 3 59 200.6 NODE 4
**60 4900. NODE 5
**AEROSQL IMPACTION DATA
**SEE DISCUSSION OF IMPACTION FARAMETERS IN
- ANNULAR SECT!bN AbOVL
**IF IMPACTION IS MODELkD IN A NODE, thE InFACTION AxEA, DIAnETER (EG ORATE
** THICKNESS), ANu FLOW AAEA MUST ALL St GIVth 41 0. AIMPR5(I) IMPACTION AREA FOR NODE 1 6' O. NODE 2 4$ 0. NODE 3 64 0. NODE 4
**65 0. NODE 5 66 0. XDIMMB(I) IMFACTION DIAMtTER FOR NODE 1 67 0.0 NODE 2 68 0.0 NODE 3 69 0.0 NODE 4 4870 0.0 NODE 5 71 0. AGRARB(I) GRATE FLOW AREA FOR NCDE 1 72 0.0 NOUE 2 73 0.0 NODE 3 74 0.0 NODE 4
I w A-26
$$75 0.0 N0DE 3
- SPRAYS (EG FIRE SPRAYS)--THESE ARE TURNED ON AND OFF USING EVEN: ,
** CODE 240 MANUALLY--NO AUTOMATIC INITIATION 76 0.0 WSPR8(I) SPRAY MASS FLOW RATE FOR NODE 1 77 0.0 NODE 2 78 0.0 NODE 3 79 0.0 NODE 4 8880 0 0 N0DE 5 81 0.0 XHSPR8(I) SFRAY FALL HEIGHf FOR NCDE 1 82 0.0 NODE 2 83 -0 0 NODE 3 84 0.0 NODE 4 3*85 0.0 ADDE 5 as!NITAL CONDITION DATA 84 305. INITIAL TEMPTRATURE OF AUX BUILDING 87 305. AUX SLDG SPWAY WATER TEMP 88 1.D-3 AUX SLDG SPRAY DROF DIAntTER 89 .5 INIIIAL Rtl HUMID!lY OF AUX BUILDING COMPTS 90 311. ENVIRONMENI TEMr 9 1 05 ENVIRONMENT / Aux SLDO PRESSURE 9.} 355. FIRE DAMPER ACTIVATION TEMr (fur IN 0 TO Shui DUWN SGTS (SEE A80VE)I PUT IN A VERY HIGH NO. IF NO FIRE DAMPERS)
**IF DESIRE TASULAR OUlPUI ANU OFERATOR INiERVtNIIONS IN SNITISH UNITS *
** INSERT A 18R HERE (EG FOR A PARAMElER DUMP IN SNITISHe OPERATOR
**INTERVENf!ONS: ANU TABULAR OurFul)
*SI e
e
...,--.-,c.. , --- .,.. , . - - - , , - - - - , - - . . - - ,,-,-..,. - , - - . , . , , , . - - . . , - . .
-- - - - - - - ' ~
E B-1 APPENDIX 8 Steam Generator Tube Integrity Analysis 8.1 Introduction In the previous sections, the analyses show that the steam generr*ar tubes can be subjected to temperature and pressure conditions beyond the design basis for certain severe accident sequences. After core uncovery, natural circulation in the primary system can bring hot steam and hydrogen into the steam generator inlet plenum and the tubes. Heatup of the tubes is retarded by heat transfer to the secondary side, so gas temperatures will exceed the tube temperatures. At this time, primary system pressure could t
.. f be at the pressurizer safety or relief setpoints, while the secondary side could be at the steam generator safety or relief points. The combination of high pressures and potektially high tube wall temperatures thus raises the +
possibility of tube failure and containment bypass. The purpose of this section is to provide information related to the strength of the steam generator tubes at these conditions, from which conclusions related to their integrity can be drawn. B.2 Tube Degradation l-During the course of operation, tubes may experience some degradation. l and the effect of such degradation must be considered in evaluation of tube integrity. On the basis of operating experience, tube degradation may be separated into two distinct categories: 1) denting and 2) thinning and/or ! cracking. The effect of these two types of defects on tube integrity are ,
- discussed in the following sections, in light of the predicted severe l accident loadings.
i Numerous studies have reported on the various rrodes of tube degradation , and the burst strength of steam generator tubing. The results discussed , below are based on test data reported in Reference (B-1). All of the tests l were conducted in a simulated steam generator environment at 600 degrees F ) +
, -- - - , e --e,,.,~,.,,,----.-m.,,,,n-en---m- g-,,.n-n,--n-,,.,.,,,,,w-,-,,,--,_---n,--m--,-
. . . ~ . . . _ _ . _
B-2 for 0.875 inch diameter and 0.05 inch wall thickness Inconel-600 mill-annealed tubing. S.2.1 Tube Properties The following sections sumarize results of tests conducted with steam generator tubes in research facilities to determine the burst strength of new and degraded tube at a temperature of 600 degrees F whereas the pre-dicted wall temperature for severe accidents may be substantially higher. In this section, a method of relating these data to the accident case is presented. Theories on ductile failure indicate that the flow stress is approximately proportional to the sum of the material yield and ultimate strength (B-2). The strength properties of mill-annealed Inconel-600, at elevated temperatures, taken from Reference (B-3), are shown in Figure B-1 and are sumarized below for a few selected temperatures. Temperature Yiel'd, 'Illtimate Flow Stress - Flow Stress (F) (ksi) . (ksi) (ksi) Ratio 600 43 94 137 - 1000 41 82 123 0.90 1350 30 47 77 0.56 The correlation of material properties shows that the ernected burst strengths at 1000 degrees F would be E3% of those reported in the tests; the expected burst strengths at 1350 degrees F would be 56% of those reported in the tests. The reported data at 600 degrees F can therefore be used and adjusted for the predicted higher severe accident temperatures. 8.2.2 Thinning / Cracking Type Defects In the case of thinning the defects were produced by machining uniform-
- ly on the tube outside diameter (00). Combinations of three different penetration depths and four defect lengths were t,ested, in addition to the undefected virgin specimens. Results of these tests are shown in Figures B-2a and B-2b. As indicated by these results for a tube with a 1.5 inch h' " - - -
8-3 100 % , g 3 g g g y w - g sory % 90 "
"8th j ,
l
- r. a0 -
4 e O . 70 - [ a w i . l 60 = ' - k$0 - b'bh, f' et) -
- m. m .. ----, - . - . .-
C - . _... _ _ . 7 %..' L i 40 " **Elo* n" ga
- I -
,. tio n ;
; I l
- 30 -
.. - - -- g - - -- - - -
s , [ 20 - l - t 10 - O l f I I I l l l 0 200 400 600 800 1000 1200 I400 1600 1800 Temperature. 'F Figure B-1 High temperature tensile properties of annealed (1600*F/1 hr.) hot-rolled plate (8-1). 1
n . _ - - . _ . - _ , . - - B-4
- l 10 =
! I ell -
_.....I g m e
% o j 7 -
8 0 0 w
- g. -
. e e "e -
t g4 - E 9 3 3 , 0 UNQEFECTED 8 o 25 30'. ' 2 - O
, SS 7g,,g. 60'.,.
1 - o ' ' ' ' ' ' . o o as o so o 7s oc as i so OEFECT LENGTM tlNCMESI Figure'B-2a Burst data for 0.875 x 0.050 uniform thinning specimens ' defect length variation (8-1). to - . -.
?
sll - . ; h ~ q e, 7 *
- S 1
w
- g. -
g * % l
.s -
- , g l'
E g4 - s l2 - o unoEFEcTEo e 3/14* TirNNED t!NGTM y a 2 - 0 3,8* TMi%*st o 6 E *s CTM \ I e 3/4" THeNNED LE* Gf M L g , a 1 t/2" TMINNEO LENGTM o
- o 10 20 30 40 SO 60 70 80 90 100 -
max mum DECm ACAfacN t'.WALu Figure B-2b Burst data for 0.875 x 0.050 uniform thinning specimens - defect depth variation (B-1). l
B-5 - long defect. 50 percent of nominal wall thickness is required to withstand the postulated event loadings when the degradation in flow stress at high temperatures is considered. That is, 5 ksi
- 0.56 = 2.8 ksi which exceeds any possible aP across the tubes, given the rather high 1350*F temperature.
Considering a aP of 1250 psi present in the reactor, a good margin of safety exists. For shorter and/or nonunifam defects, the allowable degradation " would be somewhat larger than 50 percent. To study the effect of tight cracks the defects were simulated mechan-I ically by EDM slots. Slot lengths of 0.25 inch 0.5 inch, and 1.5 inch and slot depths in the range of 25-30 percent, 55-60 percent, and 85-90 percent of nominal wall were tested. The results of these tests are shown in Figures B-3a and 8-3b. Again, from these results, it is seen that a tube
, with a 1.5 inch long and 50 percent through-wall crack would be ab'le to
. withstand the postulated event loading without rupturing.
4 B.2.3 Tube Denting' . Because of the inherent mechantsm of denting, namely the corrosion of carbon steel tube support plates, the dented region of tubes is confined
- within the support plate thickness. The denting was therefore simulated by hot-swagging a carbon steel ring onto the tube. The ring thickness was 0.75 inch, simulating the thickness of the support plate. The nominal denting
! depth of 0.04 and 0.05 inch.
To investigate the effects of tube thinning near a dent, tests were also perfomed on specimens with denting plus uniform thinning and denting plus elliptical wastage defects.
- Results of burst pressure tests with various denting and thinning L
configurations are sumarized in Figure B-4 and Table B-1. These results l clearly indicate that denting has no degradation effect on the burst strength of a tube with either the nominal or wasted wall. A similar conclusion has been derived regarding the strength of a dented tube with superimposedcracks(B-5). Thus, the observations made previously regarding l _. -~. _ __ ___. _ __ _ .. _ _ _ _ ___ _
- - - - - - - - - - - - - - - - --:~~
B-6 7 A C ell - 8 - o* Ey . .a E w - 5* a
@ e -
3s - 4 4 o
- i 3 -
0 UNDEFECTED 3 - o 1/4* SLOT , a 1/2'* SLOT g ,
- 11/2 SLOT 0 ' ' ' ' ' ' ' '
O 10 20 Jo 40 S0 s'e 70 80 90 500 MAXIMUM Os CR AO ATION l'. WALL) Figure B-3a . Burst data for 0.875 x 0.050 EDM slot
- specimens - defect depth variation (B-1). .
m oO g_____j e - V 3 l f7 - ! % l* o e ,. .I 4s V s i s . e uwospectso e ts.aos
, , e es. sos e es. sos
- 1 ,
,9 $ t t t ' t e e as o so on i oe i as i no Of f ECT Lih4TM tehCMf Si Figure B-3b Burst data for 0.875 x 0.050 EDM slot l specimens - defect length variation (B-1).
- - + + - - ---r. ---,,,--------w,--- - - - . - - - _m-m.--er--.-,--S-..-.e-. ,%-%- -.-,-e, - , - , - -.,--,.--w,- e,-.e,.,w , - . - -,-
.-~ .-
B-7 10 - W
'o a . :
e -
.g ,
3 8*
- 1 .
a E w ge - . e e
=
g' 4
< e s $ -
e
=
3 3
, 0 UNGFFECito #
4 02330**. e 4 55 609 8 - 1 e 7s. sos 51 1 - s3 2 3 .- 0 J_2 1 50 4 0.25 0.60 0 75 1.00 1.25
- OSPECT LENGTH (INCHES) l Figure 8-4 Burst pressure data of 0.875 x 0.050 uniform thinning specimens with and without denting (B-1).
l l l t l t
-- ...---. -.. .. .. . . . . .. - ~ .- . - - .
B-8~ Table B-1 COMPARISON OF BURST PRESSURES OF ELLIPTICALLY WASTED
. 875 OD x .050 WALL TUBING WITH AND WITHOUT DENTING (B-1)
Radius of Wrap Burst Pressures (psi) Cutter Angle Depth (in.) (degrees) (% Wall) Without Denting With Denting 24 0 25-30 8155, 8150 7905, 8150 24 45 25-30 7615, 7940 9200, 8060 24 135 25-30 5200, 7780 7800, 7770 12 0 55-60 5550, 5635 5520, 5605
~
12 45 '55-60 5680, 584,0 5650, 5675 i 12 135 55-60 5610, 5455 5320, 5230 l l 4
-----,----,,,_m,-e,__,,,- +,-v.-w,,-,,g-o--r~~,e -n a - g r wv.-gy, ge,---r,nn,vn,, gem--
I B-9 i i I I the required. wall thickness to withstand the event loads are unaffected by the presence of denting. t 8.3 Creep Rupture l The potential for creep rupture must be evaluated given the possible
- combination of high temperatures and pressures. Pressure differences of i 1230 to 1300 psi can exist across the tubes when the primary and secondary sides are both at the safety setpoint pressures. A pressure difference of 1215 psi can exist if both sides are at the respective PORV setpoints.
Given the tube dimensions, this translates to nominal hoop stresses between f 9330 and 9980 psi. Creep rupture times for various tube temperatures can be estimated given these stresses. Creep data for Inconel is presented in Figure 8-5 and sumarized below. At . temperatures of Il50*F and 1350'F, corresponding roughly to 900'X and i 1000*K respectively, rupture times are seen to be on the order of a thousand i. hours or greater for the stresses considered. , Rupture Stress (ksi) Temperature 1000 hr. 10,000 hr. 1
'F. ( *K) Ruoture Rupture f 1150 (894) = 15 = 12 1350(1006) 9 7 Additional creep rupture data for Inconel-600 from Ref. ( 8-8,) is l sumarized in Table 8-2. Data are provided here for shorter rupture times, corresponding to higher temperatures or stresses. The 1000 hour rupture data are consistent with that of Figure B-5. According to this table, a j temperature of 1500*F (1089'K) is required for a 10 hour period to rupture
- tubes with a 10 ksi stress. Thus, for the accidents considered, creep
~
I rupture is not a credible phenomena. l r n~~ wn- ,,o,,,w-,w,, -m m-~,.,---m-~~,-,,.. wwwm . - wwymn-m,m-=,~ ---~e---g-w-- ~ ~ ~ - , - m
. . . . . - . . Y. . ... _ . . . . . ~ . . . - N - . . . N. . -_. . N . . - . . .
B-10 s. 40 310 l ICCO.nr - 410 310 ' io.CCC.nr -
\\ i f-3C9 A '""
VN
*8'*
480 i ,3 i to I 430 / I /g . 8 ( i ig .
?
g - 304i 2 \ \ \ '\ \' ' ' g \ \\\\ ,'-incosay I \ .\\\ & - i i
;4 IN ' W- ,
\ \N@' ace'ay
,-le c ar ei l
i Incenety 43g,/ Type 444j
/ !
/ ,
I
\ .
4 4 stainless Tyge 444 I l i 4 l steiniess g l
'800 1000 1200 1400 16C0 18C0 8C0 6CQQ 1200 1400 i4CQ iSCO Testing temsereture. F Testing temperature, F ,
40 -
.. ... 01% Crees A- % ser I000 ar type 314 steiniess '
f {10 E I l l l
, , gyg ,g , , ,
i i \ \lh l %! \l i I I
. < 4 ~.e.e m o eo w w e-i 1 410 '* Color i
\ 3CI#
/ 310 444 g- ;
700 000 900 10C0 110 0 #200 1300 14CQ ISCO 14C0 . 7CC !C: Testing temeererwee, F i Figure B-5 Creep and creep-rupture comparisons (B-4). I i
.._ ~ . . _ _ . _ _
l B-11 l
. Table B-2 CREEP RUPTURE DATA FOR INCONEL-600, HOT ROLLED, AND ANNEALED AT 1600'F Rupture Stress (ksi)
Temperature 10 hr. 100 hr. 1000 hr.
'F. (*K) Rupture Rupture- Rupture 1100 (866) 48 35 22 1200 (922) 33 22 14 1300 (977) -
22 14 8.6 1400(1033) , 13 8.6 5.8 1500(1089) ;-
- 10 . 6.5 3.9 l e
e j
. . _ . . ~ . -
i B-12 , Prediction of the failure time given a temperature and stress, for failure times below 1000 hours, is possible by correlating a parameter combining time and temperature with the stress. The Larson-Miller parameter is such a parameter [Ref. (B-6)]. For 316 stainless steel, it has the fann LMP = T*(20 + log rt )
- 10-3, where T is in temperature in degrees Rankine,
- i. t ris the rupture time in hours, and log represents the base 10 logarithm.
A master rupture curve for 316 stainless steel is shown in Figure B-6, as
- presented in Ref. (Bd). The Inconel data from Figure B-5 maps onto the
- same curve as the ASME N-47 data for stainless steel, and from Figure B-5 it
- can be seen that these two materials exhibit very similar creep rupture behavior. Thus, for example, assuming stress of 10 ksi, a Larson-Miller parameter value between 40 and 41.5 is obtained. Given a temperature of I 1500*F = 1960*R = 1090*K, a rupture time between 2.5 and 14.9 hours is obtained. Checking the validity of this extrapolation, at 10 ksi and thus the same Larson-Miller parameter range, for tr = 1000 hours a temperature i between 1280'F and 1340'F is predicted. This agrees quite well with the value 1325'F from Figure By and'an interpolated value of 1275'F from Table l B-2. Therefore, the extrapolttion approach is seen to.be valid.,and temper-atures in excess of 1500*F would be required to fail tubes by creep rupture, since the time at elevated temperatures is well under _an hour. -
B.4 Sumary Based on the material properties of Inconel-600 at elevated tempera-tures and the results of the test program on burst strength of steam genera-tor tubes, it can be shown that the combined pressures and temperatures at which the tubes can continue to maintain integrity are within the predicted conditions for severe accidents described elsewhere in this report. The
- results are applicable to conditions of tube degradation up to 50 percent uniform through-wall wastage as well as for crack profiles which represent i the limits of operation.
Creep rupture is not considered to be a failure mechanism for steam generator tubes at the pressure, temperature and time conditions predicted for severe accidents. This conclusion is based on three independent data t I f
! i l
t j Master Rupture Curve for.316 Stainless Steel : I - I i 10 0 -- i i i i 1 -i t i . .
} -
_ ,x
. , 'y.v ASME Code Case N-47 ;
' ~
f g Lasson-MsNet data / ., l
- g for 18Cs-8Hi S.S
- a f l e -
. . '.,*.o -p f
l I ~
.. ' . h .,y
' x a 10
, , . yoo g g,oo .r)
.a~< !
) X 755 K (900 *F) ,' x _ l C 0 - 811 K (1000 *I) -[ f., l j
~
O - 922 K (1200 'F) " i A - 1033 K (1400 *F) i m y 1089 K (1500 *F)
- m. .
i :: ! u) . j_ . . ..i _______.I_.__ l________. _1- ... I __L - - _ . l; j 15 20 25 30 35 40 45 50 l l Larson-Miller parameter [T(20+1ogtr ) x 10-3] l' Figure B-6 Master creep rupture curve for 316 i stainless steel, taken from Ref. l (B-7). i i t i
B-14 sources, which indicate that temperatures in excess of 1500*F must be sustained for long periods for creep rupture. , 8.5 References - (B-1) M. Vagins, et al., " Steam Generator Tube Integrity Program - Phase ! Report". NUREG/CR-0718, PNL-2937 (September, .1979). (B-2) R. J. Eiber, et al., " Investigation of the Initiation and Extent of Ductile Pipe Rupture (Final Report)", BMi-1908 (June,1971). (B-3) " Source Book on Industrial Alloy and Engineering Data", American
. Society for Metals.
(B-4,) " Metals Handbook", 8th Edition American Society for Metals. , (B-5) D. L. Harrod and D. A. Kaminski, " Burst Strength of Dented and Cracked 1600 Steam Generator Tubes", Westinghouse R&D Memo 79-102-DEFIN-MI(January,1977). (B-6) F. R. Larson and J. Miller, "A Time-Temperature Relationship for Rupture and Creep' Stress", Transactions of the ASME, pp. 765-775 (July,1952). . - (Bd) V. Shah, Presentation to IDCOR/NRC Exchange Meeting on Ofrect Con-tainmentHeating(April 23,1986), (B-8) "Inconel-600". Technical Bulletin of the International Nickel Company, Inc.,1969, cited in Nuclear Systems Materials Handbook, Volume 1. l 1 f
C-1
. APPENDIX C Estimation of Steam Generator Tube Wall Temperatures MAAP 3 does not contain a detailed model for natural circulation on the gggdLrl side of the steam generators. An overall heat transfer coeffi-cient is calculated in the code which is used to represent the loss of heat from the tubes to the secondary side gas, but this does not consider the detailed behavior in the vicinity of the hottest tube surfaces. This area Ifes just above the tubesheet on the inlet plenum side on those tubes which carry fluid from the inlet plenum out to the outlet plenum ("out" tubes).
The purpose of this appendix is to detemine the relative magnitude of the heat transfer coefficients on the primary side of the inlet part of the out tubes to that on the secondary side. If, as will be shown to be the case. l these heat transfer coefficients are comparable, the tubes will achieve a
.. temperature about half-way between the inlet primary side gas temperature and the average secondary side
- gas temperature.
The procedure to be used is to calculate the natural circulation flows
; on the primary and secondary sides and to use these flows to calculate the heat transfer coefficients. Since both flows are driven by the high tem-peratures which exist near the inlets of the out tubes compared to the cold I
temperatures existing elsewhere, scaling arguments can be effectively used
; to compare the flows and heat transfer coefficients. '
f. As the primary side flow W, (see Figure C-1) enters the out tubes, it [ rapidly cools off from a temperature T, to a temperature which is essential-J ly the mixed average secondary side temperature 7,. If we assume a roughly j Ifnear drop in temperature occurs over some length t, the magnitude of the flow is given by equating the total frictional pressure drop required to j move fluid a distance 2L on the primary side to the difference in hydro-static head created due to the high temperatures in the zone of length t at i the inlet of the out tubest i L i i
C-2 Ts N L l. l U Ws ' a A O , 3 .. .. . o n o Tp . 1 inlet outlet 4 plenum plenum l
/
i i Figure C-1 Natural circulation flows on the inside and outside of a tube carrying fluid from the inlet to the outlet plenum. f
1 C-3 .
+ =o g8 1 (C-1)
'p P (a (1-a)2) where, op = average density of fluid on the primary sides of the tubes, .
Op = hydraulic diameter on primary side (tube diameter). Ag = total tube flow area, a = fraction of tubes carrying out flow; (1 - a) carry the return flow back from the outlet plenum, f = friction factor. . 8 = coefficient of volumetric expansion. g = acceleration o.f gravity.
. t, The temperature decrease in;the fluid AT, is given by aT, Wpcp =0 5G (C2) where, cp= specific heat, 0$g = total heat loss to secondary side.
Substituting 2 2 t 0, Ag W,3eo p 9, (C-3) 7 (1 - a)2 On the secondary side, we have by the same reasoning
C-4
" 2ft r=os sa AT, F-t (C-4) 2a z ,z 's s s
where. . AT, W, pc =Qg 3 (C-5) or I 2 W 3,o 98 Qgm$ 4 (C-6) 2f ("p where A, is the flow area. associated with the inlet half of the tube bundle. For simplicity, we have i'gnored the typically small differences in volu-metric expansion coefficient' specific heat, and friction factor between the primary and secondary sides. If we divide Equation (C-3) by Equation (C-6) we obtain (C-7) [ 1 \I/3 (' * (i a)'/ . For typical values of a s .2, the last term is very close to 1. The heat transfer coefficients can now be calculated by using the Dittus Boelter correlation for turbulent flow in passages. Since super-. heated steam has a Prandt1 number of 1, and if we neglect small differences in thermal conductivity and dynamic viscosity between the primary and secondary sides we have:
C-5' g.8 hs (C-8)
,,D.z or N 2' h ( (C-9)
For Seabrook: 2 Ag = 1.1 m 2 A, a 2.95 m Op = .0155 m , D, = .038 m , L = 17 m t s 1-2 m (based on the rate at which the primary fluid loses its heat to the tube wall in the MAAP calculations). At the time of peak inlet plenum temperature, the primary side pressure is about twice the secondary side value and has an absolute temperature which is approximately 1.3 times as high so that h>h=1.5 s Thus [b )(1.2)(1.2)(.5)(1)(1.3) s
% .9
.a . . . . . . . - - - . - . . . -.-. -
C-6 or
- h, g h, It should be noted that this conclusion is a relatively weak function of the assumptions (e.g., on the value of s), given the small values of the expo-nents in Equation (C-9).
As noted previously, if the heat transfer coefficients are equal, in steady-state the tube temperature will be half-way between the primary and secondary side gas temperatures. Further, by substituting in Equation
; (C-7), we obtain a
W, s 5 W, so that there will be a relatively small variation AT, in secondary side gas , temperature ' compared to AT,1 thus, the average secondary side temperature
- can conveniently be used for, estimating the tube temperature.
1 U
RAI 48
Host of the work pertinent to severe accidents has addressed plant behavior at full power, on the assumption that this represents the major contribution to risk. Also WASH-1400 assumed containment failure was probable following a core melt, making containment bypass sequences relatively less important. Therefore:
- a. Please address the possibility of accidents inside the containment building while in Modes 2-6 (Startup Hot Standby, Hot Shutdown, Cold Shutdown, and Ref ueling) insof ar as these accidents could impact upon risk. In particular, consider the ef fect of reduced safety equipment availability and containment integrity requirement permitted by technical specifications while shutdown or refueling.
- b. Event V and steam generator tube rupture provide a direct path from the RCS to the environment during severe accidents. Please describe the Seabrook work which identifies any other direct path,
- c. Please provide further information and/or specific references pertinent to release of radioactive material located outside of the containment building (e.g. spent fuel pool, radwaste systems) insof ar as the magnitudes are large enough to impact upon the issue under consideration here.
2 RESPONSE 48 We would like to note that while it is true that WASH-1400 assumed a high probability of containment failure, it is not true that bypass sequences were relatively less important. Most of the containment failure probability in WASH-1400 for PWRS was assigned to the base mat penetration and leakage failure modes and these modes did not result in high contributions to early fatality risk. The V-sequence in fact dominated category PWR-2 which, in turn, dominated early fatality risk. Hence, the V-sequence type of bypass was very important in the WASH-1400 results.
- a. Hodes 2-6 The Seabrook Station technical specifications generally require an equivalent level of safety equipment to be available in modes 1, 2, and 3, consistent with the actual mode of operation, for the reactivity control systems, instrumentation, coolant circulation, and ECCS subsystems. The technical specifications require that only one train of ECCS subsystems be available for mode 4 and that the accumulators be isolated. The saf ety injection pumps are not required to be operable in Mode 4. Technical specifications for containment integrity, containment cooling and support systems such as of f-site power, on-site power, component cooling and service water pumps are generally the same for modes 1, 2, 3, and 4. Considering
. . ~ . . . _ , - . . . . . - - - .. . . . . . . - . . . - . . - . . .
that the plant is in modes 2, 3, and 4 for only a small portion of time and the technical specifications generally require an equivalent level of safety equipment to be available, the SSPSA and RHEPS analyses account for the mode 1, 2 and 3 events. The ef f ect of reduced saf ety equipment and reduced containment integrity requirements permitted by the technical specifications while shutdown or refueling are addressed in the response to request for additional information #21.
- b. Direct Paths For completeness, all sequences in the SSPSA were considered with regard to their potential for early release. As shown in Figure 1, there are four ways of having an early large release (Release categories $1, $6 and $7). Each is described below:
Containment Svosse There are two types of containment bypass initiating events in the SSPSA. The interfacing LOCA (V) is quantified in SSPSA Section 6.6 and an enhanced new model is provided in the RMEPS. As shown in Table 1, the V sequences contribute approximately 12% to early release frequency (13% of which is in . release category $1 and 10.8% of which is in release category $7). A steam generator tube 'rppture (SGTR) . initiating event can result in an early bypass releasa if the secondary. side is not isolated and,the ability to cool the core is lost. The SGTR nodel has not changed since the $$PSA (SSPSA Sections 5 3.11 and 5 4.4). SGTR is an insignificant contributor to early release f requency for two reasons. First, the frequency of core seit is low ( Aprox . 10-6 / yea r) . This low frequency stems from a high degree of plant equipment availability for core cooling and a very long time available for the operator to tereinste break flow and maintain basic safety functions. Second, the conditional f requency that the secondary side is open and the operators don't isolate paths to the environment is extremely low. Therefore, SGTR core sett sequences with bypass are insignificant contributors to early release frequency. I I guternal svents, guternal initiating events have the potential of f ailing containment and starting an initiating event. Two such events (aircraft crash and turbine missiles) are in the $$PSA model but neither provide a significant contribution j to early release f requency or early health risk. It should be noted that other external events were considered. For example, the seismic capacity of the Containment ($$PSA, Section 9 2) was found to be greater than 23 and excluded from the detailed event tree quantification.
m.._. _ , - - , - - - - -
,._ .. - - - . _ _- ~.
Aircraf t crashes into the containment' that can potentially penetrate the Section c:ntainment 9.3).- arefrequency The essa assumedoftocorecausemelta and
,large LOCA early initiating release event is 1.03 x 10-($5PSA,10/ year '
(SSPSA, . Section 13.2). This analysis has not changed since the SSPSA. As shown in Table 1 air crashes ( APC) make an insignificant contribution to early release frequency. Terbine Nissiles that penetrate containment are also assumed to cause a large LOCA (SSPSA, Section 9.9). Core melt frequency is assessed in the plant model (88PSA, Section 3.3.7 and 3.4.3) with a sean core melt frequency equal to 3.23 a 10-10/ year ($5PSA, Section 13.2) 12 conclusion, enternal events tapacting containment are insignificant contributors to early release frequency as shown in Table 1 and are minor contributors to early h:alth risk in RMEPS. To the extent that WASH-1400 (and NUREG-0396) risk curves d3 not explicitly contain risk from external events, the RMEPS results are conserv-ctive in comparison. , Containment Isolation Failure Cantainment isolation systes failure and f ailure to recover during a core melt would result in a potential eatly release. The systems analysis without recovery is documented in SSPSA Section 0.13. The systems analysis model was updated to include operator recovery that was included at the sequence level in the SSPSA. This update.is provided in the Risk-Based Evaluation of Technical Specifications ' For Seabrook Station, PLG-0431 (Section 4.5). Presently, the update and the criginal SSPSA analysis are being integrated. The top contributor to early rolesse frequency and risk is f ailure of containesnt isolation caused by seismic . Initiating events that fail support systema such as the SSPS. As shown in Table 1 containment isolation (CIS) f ailure dominates early release f requency (491, all of which is 56). Again, RMEPS risk is more cosplete WASH-1400 and (NUREG-0396) since risk free earthquakes were not explicitly included in WASH-1400. Containment Structural Pa11ure cratainment structural f ailures due to core melt phenomena is assessed in SSPSA Section 11. This assessment includes an analysis of the ultimate strength of the containment, f ailure modes, and f ailure times as well as .ef fects of hydrogen burns, 'j. and steam esplosions. This analysis has not changed since the SSPSA. BNL reviewed
, the $5PSA severe accident phenomena, containment response, and radiological source teres and generally concurred (NUREG/CR-4540) that early f ailure of Seabrook cintainment is very unlikely. As shown in Table 1, steam explosion makes an insignificant contribution to early release frequency.
i l
- c. Other Sources As discussed in Section 5.2.2 of the SSPSA (PLG-0300), Initiating Event Identification, if a significant radiological release occurs at a nuclear power plant , "...it must originate either in a damaged core or in a non-core source of radioactivity such as the spent f uel storage pool or the gaseous, liquid, and soild waste f acilities. Past experience and analysis (Reference 5 2-1) have clearly shown that releases from the core are by f ar the only significant source of risk at a nuclear power plant
...It is generally recognised that sources of radioactivity at the plant (other than the reactor core) and the possible mechanissa for their release are such as to provide a negligible risk of public health impact." We also note that contributions from class 3-8 accidents, modes 2-6 events, external events and many conson cause f ailures were not included in WASH-1400 or NUREG-0396.
Reference 5.2-1 is NUREG/CR-0603, "A Risk Assessment of a Pressurized Water Reactors for Class 3-8 Accidents," R.E. Hall, et al, October 1979. T. e t 4
j. F Table 1 Distribution of Large Early Release Frequency ' Release Mean Fraction of Sequence Mean I of Fraction of Category Frequency
- Total Type
- Frequency
- RC, Total
$1 5.8 x 10-9 .016 V (pipe break) 4.6 x 10-9 79 .013 Sta Exp1 5.9 x 10-10 10 .002 TMLL 5.2 x 10-10 9 .001
.' .' APC '
1.0 x 10-10 2 .000 56 3.2 x 10-7 .889 Earthquakes /CIS 3.2 x 10-7 100 .889 57 3.9 x 10-8 .108 V 3.9 x 10-8 100 .108 l (RNR pump seal) l 3.6 x 10-7
- events per reactor year 0 V- interf acing LOCA Ste. Emp1- Reactor vessel steam explosion TMLL- Turbine missile resulting in large LOCA AFC- aircraf t crash into primary containment (resulting in large LOCA) ,
CIS- containment isolation systes f ailure RC- retsase category . t t l [ L__ __--________m __._.m.-_m_ _ _ _ _ . . _m __ __ _ _ _ . _ _ _ _ __. _ _ - - _
PRINCIPAL CONTRIBUTORS TO EARLY RELEASE FREQUENCY HTIATING EVENTS WITH CONTAINMENT BYPASS
- INTERFACING LOCAs * - STEAM GENERATOR TUDE RUPTURE EXTERNAL EVENTS WITH -
POTENTIAL CONTAWMENT DAMAGE FOR
- AIRCRAFT CRASH > EARLY
- TURBINE MISSLE RELEASE I
' LOSS OF CONTAINMENT . .
STRUCTURAL INTEGRITY ALL OTHER INITIATING - EVENTS , m
' CONTAINMENT ISOLATION .
FAILURE ' O
.. . , ~ .a. . . - ~ . - . - . , ..
{ l
. t i
RAI 52
Page 3-7 contains a discussion of vault behavior in response to RHR systes breaks. The emphasis is upon loss of equipment due to flooding. What consideration has been given to breaks which are small enough that the vault-is not flooded, but there is a significant thereal energy release that say impact equipment operation? Please include consideration that enough energy may be released to activate the fusible links in the ventilation system, thereby terminating ventilation and indirectly causing f ailure to pumps due to overheating of pump motors, and that this could occur at a time earlier than eight occur due to fleoding. RESPONSE 52 On subsequent pages in this chapter (e.g. pp 3-22 thru 3-28 and referenced tables) it described how environmental damage to pumps in the RHR vault due to causes other than submergence was assessed. Calculations are
- presented that show that the leak area of the RHR pump seals must be less than .09 in2 to enable the RHR vault sump pumps to keep up with the rate of flooding. In the event trees of Figure 3-4 and 3-5, leak areas of 0 to .9 inZ are covered by sequences in which the top event L1 is successf ul
(, sequences No. 4-42 in the VI tree and 4-34 in the VS tree, respectively).
- In addition to submergence by'RHR vault flooding, consideration was given to other f ailure modes to itHR vault pumps such as the thereal and moisture effects of the stese environment. The f ault tree in Figure 3-9 together with the quantifications in Tables 3-11, 3-12 and 3-13 of PLG-0432 were used to assess the probabilities of environmental failure of the RRR, C5g, and 51 pump, respectively. The result of these assessment is summarised in Table 1. As seen in this table, there is a high probability aspigned to environmental f ailure probability of each type of pump even for the range of 0 .09 in2 in which flooding of motors may not occur. While
.. these assessments are highly subjective, they demonstrate adequate consid-erstion of these failure modes and are believed to be conservative. Note that
. for larger leak rates there is some chance of non-submergence in the injection path (VI) event tree because of the possibility of operator action to ISOLATE the leaking check valves.
In spite of the conservative assessment of environmental damage of RHR vault pumps, it is clear that the conclusions of the sensitivity study are insensitive to these assessment. The total f requency of non-core seit sequences reguiting from success of event L1 is on the order of 2 to 3 x 10- /per reactor-year. Even if no credit were taken for any unflooded pumps in this vault, the results would be unaffected because the charging pumps located outside the vault would still be available for core cooling. Even if it is further assumed that the charging pumps would also f ail during these sequences, there would be no 1spect on the conclusions of the sensitivity study because of the low frequency currently assigned to the non-selt sequences in which the pumps in the RHR vault do not flood.
The conservative assessments of environmental damage in Table 1 accounts for the direct ef fects of water jets, steam, humidity and thermal damage. These ef fects were assessed to overshadow the rather indirect mechanism of fusible link actuation and overheating due to lack of ventilation. It is not clear whether the fusable links would actuate or not. Even if they did, the incremental thermal stresses acting on the pumps would be small in relation to the direct environmental stresses. Note that sensitivities were assessed on the impact of fusable links on source terms in Section 4 of RMEPS. . 9 9 9 9 9 . 4 6 e
+
9
. TABLE 1. Assessment of Environmental Damage to Pumps in RHR Pump Vault RHR Pump Seal Environmental Failure Proability of RHR Vault Pumps Leak Area RHR Pumps CBS Pumps SI Pumps (in 2)
Il 0 .09 .55 .1 .1
.09-1.05 .85 .44 .33 1.05-2.6 1.0 .75 .64
>2.6 1.0 -
1.0 1.0 3.5, 0 .09 .56 .11 .11
.09-1.05 1.0 1.0 1.0 1 05-2.6 1.0 1.0 1.0
>2.6 1.0 10 1.0 L
~ ' ~ ~ --
RAI $4 The authors concluded on page 3-9 that presence of water in the reactor i + cavity will decrease (significantly?) the revaporization of fission products from RCS and perhaps RRR surf aces. We anticipate that a significant quantity of heat producing radioisotopes will remain in the wreckage of the reactor vessel, and this may be ef fective in heating what-ever gases or vapor are flowing toward the break. Has this been investi-gated? RESPONSE 54 Water present in the reactor cavity will essentially eliminate heating of the reactor vessel.by the debris. Furthermore, volatile fission products trapped in the primary system and RHR piping will be cooled by the flow of steam f rom the cavity pool to.,the RRR pump vaults. Calculations indicate that this flow could cool somewhat more than half of the volatile inventory, were it trapped in the RHR line~alone. ' Larger inventories can be cooled as the decay heat drops in lohg duration sequences or if credit was taken for direct heat losses from the primary system to the containment. Somewhat smaller inventories could be cooled if the fission products were concentrated at one location, although natural processes would tend to distribute the fission products. l
. . . . - - ~. . ~ . .
. . .w .- . - . . . . . - . ~ . . . . . -
.~
RAI 56
There have been a number of indication (prior to and including page 3-11) that containment spray may be actuated due to RRR relief valve release into containment. What is the justification for this conclusion? Include the effect of containment heat sinks and containment cooler operation in the response. - RESPONSE 56 In the early stage of the V-sequence analysis, it was recognized that there could be some potential for containment spray (CBS) actuation since the RRR relief valves would be discharging to containment (through the pressurizer relief tank); in the interest of completeness, this possibility was included in the modeling. As discussed on page 3-11 (top), automatic CBS actuation occurs.on a P-signal which is generated by a high containment pressure of 18 psig (33 psia). The V-sequence behavior was analyzed using the MAAP computer program as discussed at ,the bottom of page 3-11 and throughout Section 4. . MAAP predicted .2,ontiinment pressure is plotted in Figure 4-13. MAAP does account for such affects as heat sinks and containment air coolers, though for this particular case the coolers were assumed to be not operating. In actuality, 5 of. the 6 coolers would typically be running and delay-CBS actuation signal even further. l It should be noted that CBS pump operation could have a negative affect on the V-sequence since it directs Refueling Water Storage Tank (RWST) inventory to the containment. That is, away from the RCS where it could contribute to core cooling and away from the vault where it could contribute to fission product scrubbing. Thus neglecting the containment coolers is conservative. u The discussion at the bottom of page 3-11 provides a somewhat simplified event timing discussion for a particular size of RHR leakage into the vault area. As discussed throughout section 3,-the V-sequence analysis includes event trees modeling of the V-sequence over the complete range of RER system overpressure failure modes and leakage rates. The MAAP model simply
- l. included both the possibility of CBS pump actuation and the possibility of p CBS pump flood-out. In the MAAP model, if actuated, the CBS pumps operate as long as they are not submerged. (see Table 3-7).
As discussed in response number 52, the event trees include pump failure probabilities as a function of RHR leakage for all pumps in the vault area. e
,w-- -
-- , - er - a sw-- - - -t:9we - y---e,--wyw-s-- w-meww v e --mg---- n-+-erv-- -r----w
l
RAI 58
The last paragraph on page 3-11 contains a number of timing of event-state-ments. Please provide justification of each. Plots of plant behavior showi'ng suitable parameters and indicating the event points are sufficient for most. Operator response information, in addition to RCC parameter information, is necessary to substantiate the statement that RCPs will be tripped within about 21 seconds of break initiation. RESPONSE 58 The last paragraph of page 3-11 discusses the timing of certain events as predicted by the MAAP program for one particular V-sequence event involving the maximum expected RHR pump seal leak area of 1.3 squared inches for each pump. This particular analysis and its results are discussed further
. th roughout section 4 of PLG-0432. Table 4-7 provides further sequence event timing information. Plots of several plant parameters versus time (as pre-dicted by MAAP) are provided in Section 4 including the following: s
- Primary System Pressure (Figure 4-11)
Core Water Temperature (Figure 9-12) , Vault Water Level (Figure 4-20)
- Seal *LOCA Flow Rate Into Vault (Figure 4-19)
Containment Pressyre (Fi'gure 4-13) Enhanced plots of the information provided there are-included here as Figures 1 and 2; they show RCS pressure vs. time with various set points superimposed and subcooling vs. time with Reactor Coolant Pump Trip criteria superimposed. CBS pump timing in Paragraph 3-11 is based on an approximate hand calculation for one particular break size. Further CBS pump operation is discussed in responses 52 and 56. Page 3-11 notes that MAAP predicts the RCS to be solid within 30 seconds; this particular event, which is of no significance to the V-sequence, is probably inaccurately predicted by MAAP because the RHR relief valve discharge to the Pressurizer Relief Tank was actually modeled via the PORV as discussed in Seccion 4.4.3 Operator Response Information: 1Due Response to Reactor Trip or Safety Injection (E-O') would be the first procedure utilized by the operators. I L The E-0 procedure specifies on the foldout page (see attachment) that the Reactor Coolant Pumps be tripped whenever the following criteria are met. TRIP ALL RCPs IF ANY CONDITIONS LISTED BELOW OCOUR: I
- CCPs or SI pumps - at least one running
- and -
RCS subcooling - less than 30*F Phase B Containment Isolation (loss of PCCW)
-. -. - , _ - _ _ = - - __ ,=. _,
These instructions are valid throughout E-0. For further information on this subject, see included excerpt from Westinghouse ERP Users Guide on RCP Trip Criteria, section 2.3.2, Evaluation of Alternate RCP Trip Criteria. The trip of the RCPS is a function of the present ERP sets, which is trained upon in both licensed operator training and requalification training. b 9 4 l l l -
RCS Pressure vs Time for Worst Case RHR Interf ace LOCA > 1 4
?
I 24QQ - 2200 K~~'~~~~< * ~ * * " ' ' ~ ' ~ ~ ~ ~ ~ <~ ~ ' " ' ' ~ ' " " ~ " ~ ~~~~ '""""'~*"*"""'"*~- 2000 2-.___..___ ._____________r,___.r,_..-~ ______-____ ___y___. ~ ~~ ~ Legend ' 1 R - i C 1800 {1 .
.o- RCS Pressure -
S 1600 _l: ~~ Low Pzr Press Alarm P 1400[: - -
-- Rx Trip Setpoint !
r :- ( 1200 - Si Setpoint i e, ::g s s 1000{ -
" RHR Hi Press. Alarm ' ~ i u 800 4- =- - --. - RHR Suction Relief Selpt.
r - e 600: _ ........ . . . . . . . , ....... ,.............. ....... ,........... .
.......I...w................I.......
'b - n- nl.
NOTE: Data provided by 1
._ m m , y ---- o O 400 ::.
N' - - - Westinghouse Electric 200 -~ -: : : : :::: :::: :::: :::: :::: . :::: :::: ::H :::: :::: :::: :::: , 0 5 10 15 20 25 30 35 40 45 50 55 60 65 t g Time in Flinutes
}
!l
Calculated Subcooling vs T ime Io Worst Case RHR Interface LOC A 60 o 1 i 5 55 =------ - -- - - - - - - - - - - - - - - -- - - - - - - - - ---- --- u 50 . b - 4 5 :. . c . o 40 - . . . . - .-- -. --- - .-- . - .-. - - - .-.:.. ... ..-. o 35 f -- . - - . . ..- - - .. .- -- . - - . . . - - - -- ... .. . . . .- . . - - - i : Legend 30;- ~~"" " " " . " " " < ~ ~ ~ ~ ~ ~ ~ ~ " " " - " - - - " " " " " - " ' " " " " " " ""- i I "" RCP Trip Criterea ! n 25 = - --- ---- --- -- - - --- - - - - --- --- --- - - - - - - 9 go :- A -o- Subcooling (calculated)
,n_ f. .% \
o g 15 7
-- - - - - --- - - - --- --- Ng/- - - - - --- - - --
n 10 j : -- - - - - - - - - Ag' - - - - -- - - - - - - - - - - --- - - - - - - - --- ---- NOTE: RCS Pressure and Core Water Temperature supplied D 5l -\ T by Westinghouse, Subcooling e 0o .. .,
....i ... ,...., ....i ....i .... ,... 1 ....e .... ,....i . . . . , ... 1 calculated by Seabrook.
g
-5 g o. - - ,
~
7 -10 ~ 0 5 10 15 20 25 30 35 40 45 50 55 60 65 Time in Ilinutes
._. _.. . . . ~ . . _ _ _._ _ . _ . _ . . - _ . _ . . . _ _ . . -
3 t (E-0) OPERATOR ACTION
SUMMARY
fur LA SSRIES PROCEDURES
- 1. RCP TRIP CRITERIA )
l Trip all RCPs if ANY conditions listed below occur. e CCPs or SI pumps - AT LEAST ONE RUNNING
-AND-RCS Subcooling - LESS THAN 30*F l
e Phase B containment isolation (loss of PCCW)
- 2. ECCS ACTUATION. CRITERIA Actuate SI and go to E-0, REACTOR TRIP OR SAFETY INJECTION , Step 1, if EITHER condition listed below occurs:
e RCS subcooling --LESS ,THAN ,30*F
.~
-OR- . ,
o Pressurizer level - CANNOT BE MAINTAINED GREATER THAN 5% [(35)% FOR ADVERSE CONTAINMENT]
- 3. EFW SbPPLY Commense CST askeup as soon as possible to avoid low inventory problems.
i 4 RED PATH
SUMMARY
- ATTACHMENT F
- 5. KEY CAUTIONS e If offsite power. is lost af ter SI reset, manual action any be required to restart safeguard equipment.
- RCS pressure should be monitored. If RCS pressure drops below 200 PSIG, RER pumps must be manually restarted to supply water to RCS.
9
, -n-- , , , . , - - , -
, _ . . _ ~ '
._,,_,l.--, -- .--,-e-.------
e The appropriate instrument uncertainties should be acded to the RCS pressure value established above. For normal containment conditions, the normal instrument uncertainties should be used, whereas with adverse containment conditions, the instrument uncertainties associated with post-accident containment conditions should be used. The instrument uncertainties should be determined for both the RCS pressure measurement and the steam generator pressure measuremes+, and the values should be combined in an appropriate manner to obtain the total uncertainty. The resulting two pressures are the indicated RCS pressure setpoints at which the operator should trip the RCPs, depending upon the steam generator pressure and the containment conditions. To facilitate the use of this parameter, a curve or table can be used which shows the RCS pressure setpoint for RCP trip as a function of steam generator pressure for normal and for adverse containment conditions. , The setpoint for this parameter could also be expressed as a RCS/ steam generator pressure differentili. With this method, the RCS/ steam generator pressure differential setpoint for RCP trip would be equal to the pressure difference from the steam generator pressure measurement location to the RCS pressure measurement location established above plus the combined RCS and steam generator pressure measurement uncertainties. 2.3.2 Evaluation of Alternate RCp Trio parameters )
. Analyses have been performed to evaluate the effectiveness of the tnree l alternate RCP trip parameters for small break LOCAs, SGTRs and non-LOCAs. For each of the accidents, a design basis accident was defined and analyses were performed for representative Westinghouse plants. The details of the accident t analyses are presented in Reference 5.
The. objective of the small break LOCA analysis was to demonstrate that the alternate parameters would provide an indication of the need for RCP trip prior to the time when trip is,actually required. For acceptability per NRC letters 83-1Cc and 10d, the time available from reaching a setpoint that indicates the need for RCP trip to the time RCP trip is actually recuired l l
.. . . ~ . . . . --
i should be at least 2 minutes. To provide a conservative minimum time period to trip the RCPs for each of the alternate RCP trip parameters, the small break LOCA analysis was based on Appendix K assumptions and the Westinghouse Small Break Evaluation Model ' fFLASH /LOCTA-IV Codes. The use of best estimate assumptions and mocals would result in longer time perices than tnose octained with the Appendix K assumptions and models. The results of the small break LOCA analysis demonstrate that the tnree alternate RCP trip parameters (RCS pressure, RCS subcooling and RCS/ steam generator AP) are essentially equivalent in providing an indication to the operator to trip the RCPs during a small break LOCA transient. The results also show that each of the parameters will provide the indication for RCP trip sufficiently early such that more than 2 minutes are available for operator action between the time the RCP trip setpoint is reacned and the time when ., trip is required. This was demonstrated for each of the RCP trip parameters , without adding any instrument uncertlinty in determining the RCP trip set-
- points. Thus, each of the alternate RCP trip paranieters will satisfactorily 1
indicate the need for RCP tri,p, for a small break LOCA with the instrument ' uncertainties based on e ther normal or adverse containment conditions.
- Because each of the alternate RCP trip parameters are adequate to quickly provide an indication of the need for trip during a small break LOCA, the choice of which one to implement at a given plant may therefore be basec upon the discrimination capability for SGTRs and non-LOCAs and other plant soecific instrumentation considerations.
For the SGTR and non-LOCA events, design basis accidents were cefined and analyses were performed.to determine the behavior of the alternate RCP trip parameters. The design basis SGTR was defined as a double ended ructure of one s' team generator tube on the outlet sice of the steam generator. The non-LOCA analyses were performed for credible steamline and feedline breaks since it was determined that these accidents result in the most limiting L transients among the non-LOCAs considered. The design basis steamline break was defined as an unisolable break of approximately 4-1/2 inches in diameter in one steamline, whien is equivalent to one steam generator PORV failing l RCP TRIP HP/d-Rev.1 0069V:1 18
... . .- . . . .. . - .= __.
)
RAI 65 - Relative water levels in the RHR vaults and the RCS are mentioned on pages 3-35 and 3-36. What are the water volumes in these regions as a function of elevation? (Of particular interest is the level at the top of the core and at the elevation of the hot leg connections to the RHR.) RESPONSE 65 . The volume of the RCS (Reactor Vessel and attached Cold Leg and Hot Leg piping) is 3,332 cubic feet. The center line of the reactor coolant piping lies at (-) 9 feet. The reactor vessel volume at the top of the core is 2,362 cubic feet. The top of the active fuel lies at -14' 2". The connec-
, tion of the RRR suction nozzle to the loop piping is shewn in the attached figure. The equipment elevations, flood elevtions and flood volumes are i shown in the attached figure for the concainment building spray pumps, the CBS vault sump pump, the RHR pump and the safety injection pump.
l l
FIGURE 1 RHR CONNECTIONS TO REACTOR COOLANT LOOP PIPING I tt I g . -, .
. .s
- l. . h..__ . _
%@9 12'Rh310UAL HEAT -
Q- b ti RESIDJAL HLAT REMOVAL .
%MOVAL
, / .
=
LOOP 4 LOOP 1 i t
=
_ _ - - . - , - - . ,._.-.y -[,. ^~'5__.
_;g. l -
- u
- e. se
. .t
- 6 Safetyinsection Pumo j .
j 8
- Motor Center hne Elevanon t-) 4T
/ Flood Elevat on 47.6 FT I'1' ik@ 2 i
]
# Flood volume 23634 FT3 III Contamment Busiemg Sorey Pumo
/ \
Motor Center une Elevation H $8 $* Flood Elevate 58.4 FT , . / RHRPumo FloodVoluene 4783 FT3 ,,,,, 04acnarge Elevation 14 56' 6-Seal Flood Erevanon 55.0 FT N \,8 , I, I i I.~~ Seal Flooc Voeu me 10729 FT3 qw
. n 6r o .
m/ 9, i _ - I
!s ,,
I!
- l. a r.
h l Casvaun
/
Surno Pumo , Flood Elevenon 59.S FT ricos voeume 2e98 FT3 Figure 2 RHR Equipment Vault Volumes and Elevations D e
.-w-- , - , - .
.._ . _. ____ _. _ . _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ . _ _ _ _ .-~._
I I l l l 1
RAI 66
! What is the justification for the statement on page 3-36 that the water
- level in the vaults will be approximately the same as that in the RCS?
(We do not. agree because of the potential that pressure in the vaults and containment are not the same,. and water temperature in the two locations may differ). l t a RESPONSE 66 , Page 3-36 (and preceeding) discusses potential operator actions to recover ; the V-sequence and prevent core melt. It states that if the LOCA can not be isolated, the water level in the vault will rise until the levels
; in the vault and in the RCS are approximately equal. This is a general-
. statement made only in evaluating possible operator action; the exact levels are not .important to o'perator. action.. As 'the question states, pressures '
and temperatures may dif f.er in the vault and in the RCS/ containment; however, , ! since these areas are communicating through the RHR leak,- the dif f erences would not be significant. Also, the dif ferences might tend to force additional water into the vault area (since RCS/ containment pressure would be greater than vault pressure) providing additional fission I product scrubbing. It should be noted that pressure / temperature differences of interest here are between the vault and the RCS not between the vault and ! containment. Other than RRR relief valve discharges, there is no communication between the containment and the RCS or vault area (at least until vessel ! melt-through) unless a'RRR relief valve should stick open. , i t i i I L i sw-m-w~r ,--.n- _ _ _ + _ _ _ _ _ _ _ _ _ _ _ _ _ _ ,
., _ . . . _ . _ . _ _ _ . . _ . ., . . _ _ m._ _ . . _ - _ . . -
RAI 74
a A tacit assumption appears to be incorporated into References 1 and 2 that check valves are always closed. .In reality, many check valves require a (substantial) reverse flow to force then to close, and they additionally of ten require a significant reverse. pressure to keep them closed. Is this the case for any of the valves of interest here? If so, please discuss the ,P implications. If not, what is the justification f or the conclusion? 4 ? J RESPONSE 74 1 It was not merely an assumption that resulted in the omission of sequences involving check valve failures to close but rather the result of a considered assessment. First of all, our knowledge of the design l. characteristics and tech spec surveillance requirements that govern the
- RHR system during normal power operation give us high confidence that. ,
'the check valves in the RER-and interf acing systems are initially closed.
The check valves of interest here, those that separate the RHR system from the CVCS, SI system, CBS, RWST and containment sump are all designed to close without the need to apply reverse pressure.' Hence it is necessary to postulate that. these valves f ailed to close the last time they were
. opened and that the failures remained undetected until the time of the i postulated "V-sequence". Secondly, even if one assumes, conservatively that all these check valves are initially open, our knowledge of check valve failure rates for f ailures to close, which have taken into account the San Onofre Unit 1 experience, gives us high confidence that any potential for check valve f ailed open bypass pathways is greatly over
!- shadowed by the probability already assumed for RRR piping f ailure given
- RHR pressurization. In the discussion attached we presented our assessment ji-of check valve failure rates in light of San Onofre Unit 1 experience and q
our conservative bounding assessment of sequences involving assumed failed i, open check valves. In view of these assessments, the conclusions of 4, the sensitivity study are generally insensitive to assumptions regarding ll check valve performance. I{ L -
CHECK VALVE FAILURE RATES USED IN THE SEABROOK EPZ STUDY , i The following provides further documentation for the estimated failure rates of the check valves in the pressure boundary of the ECCS and RCS, as modeled in the Seabrook EPZ study (Reference 1). The failure modes of -
, concern are disc rupture or gross leakage of a check valve that is
- initially seated.and tested to verify its position and failure of check ,
i valves to reseat on demand. s.
- 1. DISC RUPTURE / GROSS LEAKAGE 1.1 DATA COLLECTION The source of event descriptions for this analysis was Nuclear Power Experience (Reference 2). In this search, which was done manually based on the NPE key word index, a total of 610 check valve failure events were identified (Attachment A is an expanded list of these failure events, based on the NPE-automated retrieval system--a total of 692 events). The initial list was then reviewed to identify leakage events (external and internal) in all systems of both PWRs and BWRs. A total of 163 events i were identified. These events are marked "L" in Attachment B. This i review also provided further evidence that a large number of check valve s liakage events should not be considered for the failure mode of interest
, in the V-sequence analysis, either in terms of mode or cause of failure. It was. then decided to limit the data base to those events involving i check valve leakage in.the ECCS and RCS/ECCS system boundary of PWRs. l ? These were judged to be .the closest category to the initially seated and tested check valves modeled in the analysis. No di,sc rupture events were ! identified in these events, and the maximum observed leak rate was 200 gpm. The majority of ' events involved very small leaks. Those 4 considered to be more significant are listed in Table 1. Even for the i
- cases listed in Table 1, the exact leak rates were not always provided in
1 the event reports. Consequently, leak rates were estimated, based on other available evidence: the rate of boron concentration change, pressure reduction, and similarity to other occurrences for which the , leak rates were known. , Also considered was an event that occurred in San Onofre Unit 1 in J November 1985. The event involved failure of several main feedwater pump i discharge check valves to reseat on demand resulting in overpressurization of the main feedwater system. A summary of the event, ; as presentad in Nuclear Power Experience (Reference 2), is provided as Attachment C. , As can be seen from the event description, four of the five failed check 4 valves failed to reseat when the main feedwater pumps tripped. These failures obviously do not apply to the disc rupture / gross leakage mode of failure considered for the ECCS/RCS check valves. However, as it is described later, they are included in the estimated frequency of failure , to reseat on demand. The fifth valve (feedwater regulating valve bypass j line check valve) failed because of water hammer resulting from i ! l l \ l i 1 i
, 1423P091686 i
_.__---.m---.-m_ __.e _,_,__,,,,,___.m._ .g,-_.m.,-_,,, - , - - , , - -
. .. m. . _. . _ . .__.~. . - . _ . . . . _ __ ._ _
TABLE 1. CHECK VALVE LEAKAGE EVENT,0ATA BASE Sheet 1 of 2 upE p1 ant - ** *** Reference Event Description Range (date) (spm)- V11.A.126 Zion 2 A leak rate of 'O 25 sps was detected from y 0.25 (October 1975) the *A* accumulator check valve '
- wrong size gasket installed. -
~
V11.A.32 Turkey Point 4 One of the three check valves in the high-head y 0.33 (May 1973) safety injection Ifnes to the RCS cold legs c'eveloped 1/3 gpa leakage with 180 psf of water pressure applied. Two other chect valves showed only slight leakage - failure of sof t D seats. , V11.A.175 San Onofre 1 A tilting disc check valve located in the LPI y<5 (May 1978) system as the first valve inside containment. l, failed to close with gravity - valve installed in a vertical rather than a horizontal pipeline. V11.A.114 Surry 1 Check Valves 1-51-128; 130 leaked causing boron y < 10 . (July 1976) dilution in the "B" accumulator. y < 10 i i V11.A.182 Calvert C11ffs 2 The outlet check valves associated with the y < 10 (September 1978) safety injection tanks 218 and 228 leaked y < 10 L reducing the boron concentration from 1,724 t and 1,731 ppe to 1,652 and 1,594 ppe in 1-month period,respectively. p V11.A.306 McGuire 1 Of scharge check valves associated with the cold y < 10 i . (Aprf) 1981) leg indection accumulator A leaked.- cause a y < 10 { ianspecified. b V11.A.343 Point teach Check valve 1-853C. serving as the first-off y < 10 (October 1981) check valve from the RC3 for the low head safety injection. V11.A.291 Surry 2 Check valve associated with the safety y < 20 ! (January 1981) injection accumulator "C" leaked, resulting j in accumulator boron dilution - cause unknown. V11.A.63 Gfnna Accumulator "A" check valve leaked leading to y < 20 , (September 1974) boron dilution (from about 2,550 down to I 1,617 ppe) - cause unknown. ,
,' .V11.A.85 Surry 1 Check valve associated with the IC accumulator y < 20
( August 1975) failed to seat, resulting in increase in lJ J accumulator level - cause unspectfled. ] developed 6 gpa leakage. y V11. A.105 Robinson 2 "B" safety injection accumulator check y < 20 b January 1976 valve developed leakage - cause unspecified. 9, Y.A.122 Zion 1 Olscharge check valve on the accumulator 10 y < 20 {' (June 1976) developed back leakage - cause unspecified.
- V.A.407 McGuire 1 Cold leg injection accumulator check valve 20 < y < $0 g -
(May 1983) leaked, resulting in low accumulator baron ], concentration - cause unspecified. ! V.A.452 St. Lucie 2 The $1T outlet check valve oeveloped excessive 20 < y < $0 (December 1984) leakage - foreign material caused ball galling leading to joint binding. 2 1422P091686 2
TABLE 1(continued) . Sheet 2 of 2 ea ate NPE Plant Event Description Range Reference (date) (gpe)- V.A.456 Calvert Cliffs 2 $1T check valve developed excessive leakage - 20 < y < 50 (January 1985) ethylene propylene 0-ring material degradation. V.A.437 Farley 2 Loop 3 cold leg safety injection check valve 50 < y < 100 (September 1983) developed excessive leakage - incomplete contact between disc and seat. V.A.273 Davis Besse 1 Gross back leakage through core flood check 20 < y < 50 (October 1980) valve - cause unspecified.
~
V11.A.384 Calvert Cliffs 1 j!T outlet check valve leaked at the rate of y ~200 (July 1982) 200 gpa ring deteriorated. l l l l 1422P091644 3
@ he- .--.s D- .m 9 e a-the failure of other valves. This failure also does not apply to the failure model of interest here.
In the process of reviewing the available data, a recent review of eight BWR events (Reference 3) was also con.sidered. These events, listed in Table 2, involved testable isolation check valves in the pressure boundary and could be considered as precursors to an interfacing LOCA.- These events were judged to be inapplicable for this study because the valves involved are different from those considered here both in terms of design and operation. The reasons for inapplicability of each of the events are listed in Table 2. In summary, the BWR check valves have air operators, whereas the PWR ECCS/RCS check valves are enclosed and cannot be operated from outside. The latter group is verified seated, either continuously (for the upstream valve) or during startup (for the downstreamvalve). Thus, the same mechanisms that cause the eight BWR check valves to be open and undetected do not apply to the PWR ECCS/RCS check valves. 1.2 SUCCESS (EXPOSURE) DATA To estimate the total , check valve hours, the information provided in NUREG/CR-1363 on the number of valves in the ECCS and RCS in various PWRs was used. The details are pr The total number of check valve hours is 1.0 x 10gvided in Table 3. 1.3 FAILURE RATE ESTIMATE The various leakage events were grouped into five leak ranges, as shown in Table 4. For.each group? 'a frequency per hour was estimated using the exposure time discussed'above. Table 4 also provides the corresponding cumulative frequency points that are also shown in Figure 1. The curve fit on a log-log scale was done using an IMSL code, which uses the least square method. The parameters of the line obtained from this method correspond to the Bayesian most probable values based on a uniform prior distribution. The equation of the line is
-y = a x + b where x is the logarithm of the leak rate (gpm) and y is the logarithm of the frequency of exceedance per hour.
Using the data of Table 4, the following values for a and b were derived: e Parameter a . Mean = 0.0976 95th Percentile = 1.0127 Sth Percentile = 0.6915 4 l 1423P091686 r
i i TABLE 2. SlMtARY OF OPERATING EVENTS ! i 4 Event * * ' '*" ' Percent Systee
""* '*"'" '" Status Cause Reason far Inappifcability
! Vermont Yankee 12/12/75 99 LPCI/RUR Open Unknown PWR ECCS/RCS check valves are tested LER 75-24 and verified seated f altfally. They - can not be lef t open undetected. . I Cooper 01/21/77 97 HPCI Open Loose Part PWR ECCS/RCS check valves are tested I LER 77-04 I Obstruction and verffled seated. Any inttfal
- leakage or failure to be in the seated ;
! position will be discovered before the ! , plant goes to power. 1 LaSalle-1 10/05/82 20 HPCS Open- 'Drfed Lubricant and PWR ECCS/RCS check valves do not have LER 82-115 Insufficient Preload air operators. They can not, therefore, in Air Operator- be opened enternally. Opened typass LIne .
- LaSalle-1 06/17/83 48 HPCS Open Theriaal Sinding; Check valve failed to close due to disk i LER 83-066/03L Opened typass thermal binding. The PWR ECCS/RCS check i i
- Line valves are required to hold 49afast RC$r l-pressure after being verified seated Initially. These valves are closed and i stay closed. They are not cycled; j therefore, the failure modes are
- different.
4 LaSalle-1 09/14/83 0 LPCI Open Maintenance Errors PWR ECCS/RCS check valves ar'e tested i LER 83-)05/01T ; i and verified seated before the plant '
. goes to operation. .
j PfIgrie 09/29/83 98 HPCI Open Rusted L1nkage on LER 83-48 Air Operator t j Hatch-2 10/28/83 90 LPCI Open Mafetenance Errors PWR ECCS/RCS check valves do act have 2 LER 83-112/03L on Air Operator air operators and will not open due ] to a stellar maintenance error. > Browns Ferry-1 : 08/14/84 100 LPCS Open Maintenance Errors PWR ECCS/RCS check valves do not have LER 83-032 on Afr Operator air operators and will not open due i to a sjeflar maintenance error. , i ), i .i 1 1422P091686 }
I. . .. . . - . - - - . ..- . TABLE 3. CHECK VALVE EXPOSURE DATA
! N(suber of Start of Nuster cf Check Valves Total Number of Plant Name Code Casumercial Operation Years in ECCS Check Valve Ho,urs Arkansas Nuclear One 1 ARI Deceeber 1974 '10 20 1.75+6 Crystal F.tver 3 CR3 March 1977 7.75' 23 1.56+6 Davis-8 esse 1 081 November 1977 7.08 . 26 1.61 + 6 Oconee 1 OE1 July 1973 11.42 20 2.00+6 . Oconee 2 OE2 March 1974 10.25 21 1.89+6 Oconee 3 OE3 December 1974 10 21 1.84+6 Rancho Seco R$1 April 1975 9.67 30 2.54+ 6 Three Mile Island 1 Til September 1974 10.25 19 1.71+6 Three Mile Island 2 T12 December 1978 6 19 9.99+5 Arkansas Nuclear One 2 AR2 March 1980 4.75 30 1.25+6 Calvert Cliffs 1 CCI Me 1975 9.58 45 3.78+6 Calvert Cittfs 2 CC2 A 11 1977 7.67 45 3.02+6 Fort Calhoun FCI September 1973 10.25 45 4.04+6 Millstone 2 MI2 December 1975 9 47 3.71+6 Maine Yankee MY1 December 1972 12 49 5.15+6 Palisades PA1 December 1971 13 21 2.39+6 SL1 December 1976 8 30 2.10+ 6 it.Lucie1 eaver Valley 1 BV1 April 1977 7.67 36 2.42+6 D. C. Cook 1 DC1 August 1975 9.33 34 2.78+6
- 0. C. Cook 2 OC2 July 1978 6.42 34 1.91+6
- Haddam Neck HN1 1968 14 27 3.31+ 6 Indian Point 2 IP2 Januar574 July 1 10.42 36 3.29+6 Indian Point 3 IP3 August 1976 8.33 45 3.28+6 Joseph M. Farley 1 JF1 December 1977 7 33 2.02+6 Kewaunee KE1 June 1974 10.5 19 1.75+6 North Anna 1 NA1 J;ne 1978 6.5 3 2.05+6 Prairie Island 1 PRI Deces6er 1973 11 23 2.22+6 Prairie Island 2 PR2 December 1374 10 . 23 2.01+6 Point Beach 1 PT1 Decenter'1970 14 21. 2.58+ 6 Point 8each 2 PT2 October 1972 12.17 21 2.24+6 R. E. Ginna 1 RG1 March 1970 14 21 2.58+6 H. 8. Robinson 2 R02 March 1971 ' 13.75 '25 3.01+6 Sales 1 SA1 June 1977 7.5 32 2.10+6 San Onofre 1 Sol January 1968 14 18 2.21+6 Surry 1 Sul December 1972 12 25 2.63+6 Surry 2 SU2 May 1973 11.58 25 2.54+6 Trojan TR1 May 1976 8.58 22 1.65+6 Turkey Point 3 TU3 December 1972 12 34 3.57+6 Turkey Point 4 TU4 September 1973 11.25 34 3.35+6 Yankee Rowe YR1 June 1961 14 17 2.08+6 Zion 1 ZIl December 1973 11 50 4.82+6
! Zion 2 ZIZ September 1974 10.25 50 4.49+6 i Yotal 1.08+8 il ll NOTE: Exponential notation is indicated in abbreviated fons; i.e., 1.75+6 = 1.75 x 106 , lr l l i: 1 l \ 1422P091686 l l
4 TABLE 4. STATISTICAL DATA ON CHECK VALVE LEAXAGE EVENTS IN PWR, ECCS, AND RCS SYSTEMS T h Frequency Leak Rate Number of Frequency of (gpm) Events of Occurrence (perhour) Exceedance ~' 5 3 2.94-8 - 2.06-7 10 7 6.86-8 1.77-7 ' 20
- 5 4.90-8 1.08-7 50 4 3.92-8 5.90 100 1 9.80-9 1.96-8 ,
200 1 9.80-9 9.80-9 NOTE: Exponential notation is indicatgd in abbreviated form; i.e., 2.94-S = 2.94 x 10-o. 1 e i 7 1 1422P091686 i
.- +. . ._ -, _ , ,- _ . _ _ - , . . - . , _
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. . . . . . . . . . . . - - - ~
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- 6 4 4 6 6 6 44864l 66Ilig 6 4 4 4 4 4 44 l 4 4 6 66 44L s = \ % ~
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s LEGENo; \ \ 5 - -= - - STATISTICAL sOUNOS g . 4 . AT 90% CONFl0 ENCE \ ,,
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g -- T - 5 - 4 - s - 2 1 a 10*" ' ''' ' ' ' ! ' I ' ' ' t 0 too ' ' ' ' ' ' ',ooo i io coo CHECK VALVE LEAK RATE (CPM) FIGURE 1. FREQUENCY OF CHECK VALVE LEAKAGE EVENTS 8
. _ _ . . . _ _ _ . . . _ . ~ . . . _ . . - . - . .
e Parameter b Mean = 13.6943 95th Percentile = 14.2862 , 5th Percentile = 13.1024 , Based on the above values, the "best-fit" line is -
-y = 13.6943 + 0.0976 x '-
s. with the following bounds: 95th Percentile = -y = 14'.2862 + 1.0127 x 5th Percentile = -y = 13.1024 + 0.6915 x These lines are shown in Figure 1 as the "best-fit line" and " statistical bounds at 90% confidence." To account for uncertainty in the assessment of the leak rates, classification of data, estimation of exposure data, and the applicability of the data to the check valve and failure mode of concern in this analysis, the statistical bounds were further stretched by increasing the range factor of the frequency at 150 gpm from 3.7 to 10 . and increasing the range factor of the frequency at other. points proportionally (to RF = 14 at 1,800 gpm). The resulting new bounds are also shown in Figure 1.
- 2. FAILURE TO RESEAT ON DEMAND i , .
To estimate the frequency of check valve failure to reseat on' demand, two'
- types of data were used
- -(1) estimates from severa'l generic sources of i failure data, and (2) experiential data from eight U.S. nuclear power plants based on plant-specific PRAs performed by PLG.
- ' Since the majority of data sources provided information on check valve failure on demand without specifying failure to open and failure to close modes separately, the distribution developed here is based on failure on
- demand data. Review of check valve failure events from several plants
, indicate that the distribution is a good (and perhaps even conservative) estimate of the failure to reseat frequency. < An additional piece of information provided by the San Onofre event of November 1985 (Attachment C) was also incorporated into the estimate of j check valve failure on demand frequency. Four of the five check valve
- ~ failures (failures involving pump discharge check valves) apply to this mode of failure. NPE was reviewed for the period January 1, 1971, -
i through June 30, 1986, to see if there have been other check valve failures in the San Onofre main feedwater system. None were found. The corresponding success data (number of demands) were developad by
- assuming an average of 10 system-wide demands per year, a population of i eight check valves, and 15.5 years of operation from January 1,1971,
! through June 30, 1986. This resulted in an estimated 1,240 check valve ? - i
~
l 1423P091686 9
demands. The corresponding failure frequency estimate (counting four of the five check valve failures) is so *'f gg = 3.2 x 10~3 per demand. A This value was used together with the generic estimates as well as - plant-specific data from other plants in a Bayesian updating process described in Reference 4 to develop the failure on demand frequency distribution. '. 1 'The following summarizes the data used. e Generic Estimates Source Estimate Assigned Range Factor
- WASH-1400 1.00 x 10-4 5 NUREG-1363 1.10 x 10-4 3 EPRI-81 7.00 x 10-5 10 e Data from Nuclear Power Plants Number of Plant Events Number of Demands l Oconee . .- 3 6,855
.. Zion '. 0 6,970
- Indian Point 2 , 0 . 1,440
, Indian Point 3 0 1,550 Beznau (2 Units), 7 28,978 Pilgrim 0 2,394 TMI-1 12 8,716 e San Onofre Unit 1 Main Feedwater System Check Valves Estimate = 3.23 x 10-5 Assigned Range Factor = 5 (A moderate range factor is used to represent higher degrees of uncertainty than indicated by the estimatad four events in 1,240 demands. )
*The assigned range factor (ratio of the 95th to the 50th percentile of lognormal) represents our uncertainty of the accuracy of the estimate.
See Reference 4 for the details of the methodology. 10 1423P091686
The resulting distribution is shown in Figure 2. Some key
. characteristics are:
**" Median 95th Perce tile p,pg j]. _
1 5.46-4 1.18-5 1.58-4 1.63-3 NOTE: Exponential notation is indicated in abbreviated form; i.e., 5.46-4 = 5.46 x 10-4
- 3. REFERENCES
- 1. Fleming, K. N., A. Torri, K. Woodard, and R. K. Decemer, "Seabrook Station Risk Management and Emergency Planning Study," prepared for Public Service Company of New Hampshire and Yankee Atomic Electric Company, confidential PLG-0432, December 1985.
- 2. Nuclear Power Experience, S. M. Stoller Corporation, updated monthly.
4
- 3. U.S. Nuclear Regulatory Commission, " Preliminary Case Study Report,
- Overpressurization of Emergency Cooling System in Boiling Water Reactors," February 1985.
~
1 -
- '- "!!!"60^i,;;"^.'- " "** ' "' '"' " "* '"' - .
l i I i i l r 1423P091686 11
- . - . _ . _ d __ L . _ _ u_ _m_ a s_
Z p WASH 1400 l-W p
""a'*'='
E,..... 3 . SAN ONOFRE 1 , b 2
. LUMPED DATA FROM S PLANTS (22 FAILURES IN 56.903 DEMANDS)
O s . b . f N m !I O !
. \
i WW G i W i 10 0 10-5 10-4 10'3 10-2 FAILURES PER DEMAND FIGURE 2. CHECK VALVE FAILURE ON DEMAND FREQUENCY DISTRIBUTION
\.
ATTACHMENT A s 9
- e 9 % 8 9
9 9
- e l .
[ 6 e
_- - - . . . - . . . ._ -~ - - . . . . . - - 1 PAGE 1 NPE CAECK VALVE LISTI46 - 693 AaTICLES TetAL Tctet att 73 ARTICLEISI TS DE SISPLAFED: , 3.04.s.0010 : VAeIOUS CONF 4GL SLADE, 94tVE PRetLEnl SWE
- CHECK VALVE F AILU4ES
- 19 h-1975 30ESSEN 1 - 1960-12 "b
} S.64.8.3034 : DISK .
i
; AND SEAT Sutf4CCS DIATV - CRS CWECK W4LVE LEAKAGE *
--- DAESDEN 2 - M4T F5 - SEfGELING INuttedu -
1 d 4.05.E.0002 : CNECK VALVE LEAK 4GE
--- DRE59EN 2 - SPAING 72 l I
a.05.E.0003 3 FW CNECK W4LVE LE4KAGE *
-*- DRESDEN 2 - SPaING FI S.05.E.0014 s FW CNECK V4LVE LE4KA3E - CNA1 gee 33 SILIC3m SEAL RING ,
i --- DAESSEN 3 - SPAING 1974 *- S.C6.E.0010 3 3-AIgGS ADDED 10 FW CNECK VALVE 5 -
--- sat $ DEN 3 - IST 1/2 of 17F3 . > i 5.C6.E.0022 3 CNECK VALWg mLUG LEAK
--- SYSTER CREEK - JULY F3 ~
4 8.te.E.0023 3 , SILICONE O-AINGS UN4CCEPT4GLE - V4LVES LE4KED J
--- satseEN 2 - n4a a nar Fs *
{ i ,
,- t.01.E.0026 : FW CNECK DISC DID NOT MATE WITN SEAT g,
--- MONTICELLO - SPAING F6 l 8.05.E.0029 . .
LE4KT G4SKET IN FW CNECK WALVE j O 1
--- 4RNOLD - JULY 74 (PodE4 ESCALAf!3N TESTING) .
f I 8.06.E.0034 : , LEAK 4GE SETWEEN CNECK VALWE SEAT RIN3 AND VALVE 90DF -
, gg
--- 3U48-CITIES 1 - Art 76 (REFUELING $ nut 93WN3 i
[t 3.05.E.0040 s FLAW IN FW CdECK WALVE j 2 ;
--- Gunt-CITIES 2 - IES FS - REFUELING $NWTDOWN ,
S.06.E.0042 : CNECK VALVE SEAL RING LEAKAGE ! { --- * :J 2U49-CETIES 2 - DEC F4 & 75 - REFUELI16 SNUTDOWN s.06.E.0043 : , SEAT LEAKAGE AND CAACKES SELLOWS Im fd CNECES J
--- PE ACM SOEIOM 2 - M4V & JUNE 75 - SNUTDOWN AND SSE P3WER l
' 8.06.E.0048 s E400E9 CONT 43L V4LVES L 14sAPU4 3 6 2 - 1970 t i s.0$.E.0058 3 FW C1ECK , [ l WALVE 0-RINGS C09*RESSED - EXCE55tWE LEAK 4GE ,
--- DAESSEE 5 - Ara & nay 75 - AffufLING $ NUT 90W1 i g 1
3.05.E.0059 : FW --* CNECK v4LWE 0-4!NG$ 9!$ ASSOCIATED Dnf SOEN 2 - N3V F4 - REFJELI4G 5 NUT 90dN O s.05.E.0075 : PIN---NOLE LEA ($ IN FW CHEC( V4LVE SEAL PLATE, CAACKED C01v3Luft PEACH SOff0M 3 - AUG F5 - ST44tigG UP k - mw r- -
~- II I
! e i a . e , 4393 I e 8*09*3*0088 2 35V3334 JM AV143 53V141V13 E3110nS 6 !, --- d3V3N 00110N Z - tnN3 25 - 290 WnE j f 8*0S*S*0002 8 WIS3* 1GggIN3 31313 SAS13bS 450913kS
--- unN1ssaut - sadr-rt
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--- snve-3III35 t - vdu 4$ :SNnlecrN 40s s3sn31 INS ( IC i
f*02***0020 8 i 53I3 tnuSIN3 3MMVnSI 3N333 VSSEN0134 IN30453311A I --- NON113311c - rVN lE - SNnaecnN l 3( ' G*OJ'V*00tL 8 3N333 AV1AE 610nIkS 513VN* ECG 3ONIWC1$ NCA 3N3333G SND100PN A1114414133 - rR1A lE - )40n35 353V1V!!0N 1381!N!( - I 8*02*V*00C2 8 3NNVnSI 1!N3 3N333 AV1A3 31Vdd3W GIS3ONN33 'l --- tsnNSnI33 z - VdnI1 s - )4051 3uIII3V,13413snIN9t ts uCnla g 5*OJ'V*0019 8 33NVnSI 3N333 AvlAI A1Vdd3%.8VS rVhN34 *
--- SWONSnI33 t - 031 25 - ftI d0n35 f
' E*OJ*V*0090 8 33I3 4 tnes!NI SIS 3NV593 3N333 AV1AB A1Vdd33 4133 31nG SUCEIN g
--- EWnN$rt33 2 - J39 25 - MCS SIVhtlA f*CJ'V*CD?! 8 .
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--- 300d35 etS3MVt93
- 031 25AV1A3$ NVe eI51A SEVAS VNG NISSthS AVS13NIWS
- SMnat0PN g 9*OJ*3*0089 8 33V13 0N AV1A3 83VAS .
--- k0NAI33110 - tdWINS 21 M }
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--- k0NAI33110 - rVN 45 - SNnas0nN .
)
i
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^ ~ " ~ ~ - -
e PAGE 3 i i 1 3.0F.9.3074 : ANA AND NPCI VALVE PACKING LEAKS I
--- 3RUNSWICC 2 - SEPT F5 - 523 PedER ,
- 3.0F.D.00t! VALVE LEAKAGE, LPCI LD0P 3VERPRESSURIZED - NEAT EXCN445ER EASKET FAILED
--- WERMONT FANKEE - DEC F5 -.99I PedER SJ 4.0F.D.0104 100u$ SPRAT WALVE OPERATOR MOUNTINS PLATE WELD FAILED, CR4CKED
- V3EE ARM - I15UFFICIENT JELD PENEIRATION, UNDERSIIED NOWNTINE 90LTS
--- sROWNs FERAT 1 - NAT s SEPT F4 - COLD sNUTD0dM C3
, 3.0F.E.0007
' 5 TEAM VALVE DISC PIN FAILURE = AUPTURE DISCS RUPTURE,- TE1P
$dITCN DAM 4GE
--- MONTICELLO - JULV F2 *
)
4 - i 4.0F.E.0008 : STEAN VALVE LEAKAGE i MONTICELLO - JULT 72 "* 5.07.E.0011 : LOOSE TURRINE RUST - CNEtt VALVE LEA (ED TORUS WATER BACK TOWARDS 1PCI
--- PILGRIM - JULV F2 (P0 DER ESCALATION TESTING) '
3.0F.E.0018 : VACUJM I4 TURBINE EMNAUST.- JATIR NAMMER 2l
- --- WERMONT TANKEE - 1971 (PREOPER4T!0NAL TESTING) j D.07.f.0025 : J.
1 WATER NAMMER OF TURSINE EENAUST - INST 4LLED CONDENSI46'8PLAGER i
--- BROWNS FERAT 1 - OCT F2 (PREOP TESTINE) -
i B.07.E.0048 : - G GLAND CONDENSER GASKET FAILED - P338IOLE INJECT 134 VALVE LEAKAGE
--- SROWNS FERAT 2
- NOV F4
- 42X POWER i 3.07.E.0053 : SURF 4CE FLAW 5 IN VALVE DISC AND SEAT o.
h
--- PEACH BOTTON 2 - FEs F5 - SHUT 90dM # ,
i 8.07.E.0054 : SENT MINGE PIN AND SAD DISC 34 TURSI1E EMMAUST CNECK VALVE
--- MONTICELLO - JAN F5 - $NUTDOWN Gi r
3.0T.E.0058 : MISSING WELD IN SWING CHECK, INSUFFICIENT TACK WELDS IN STOP l4 CNEtt, VALVE DI55 A55EMBLED WITH C30L4NT, ERE ATER TN A1212 h DEGREES F
--- ARNOLD - APR F5 ~
l* 8.0F.E.0061 : STEA1 VALVE LEAKED
--- GUAD-CITIES 2 - DEC F6 - REFUEL!16 $NUTDOWN s #
- 5 9.0F.E.00FF : SCRATCHED SE4 TING $URfACES - TURSINE EXNAU$T VALit LE4K46E d
{ --- MONTICELLO - SEPT F5 ) ! 8.0F.E.0130 : MPCI TUROINE ENNAUST CNECK VALVES LEAKED - WORN VALVE $ EAT $ " i J j ,
----avaD-CITIES 1 2 - DEC F6 8 SEP FS - REFUELING --- 3UAD CITIES JAN 76 - REFUELING i
i s.0F.F.300F : VARI 3US PR00LEMS WITN VALVE LIMITOROUE OPERATORS e l ,
--- DRE5 DEN 2 8 3 - DEC 10 - JAN F1 i
1 5.08.C.0055 : DIRT IN 5ERVICE WATER CHECK VALVE O i ,
--- OYSTER CREEK 1 - NAT F5 - 390 MNE
! dB i
l 8
' ,,,, g' . .
>+
0.08.C.30A4 : Caost THREADED PIPE S Loost VALVE 30Lis - LEAKS 1
--- BRUNSWICC 2
- JUL F5 - Ft POWER 8.07.F.0004 : CNECC VALVE MAD MIS $ING DISK ANS P3P*ET *
.);
i{ --- SUAS-CITIES 1 - JUL F4 - $NUTt3WN gj
~
4: s.09.f.0007 : INSTRUMENT CNECK VALVES NAD dEAK $PRINGS, PITTED POPPET 3 ANS BAD SEAT .,
*-- SUAS-CITIES 2 - APR F5 - COLD $NJ'T90W4 O I
j 9.09.E.3147 * : DAMASED SEAT ON INSTRUMENT N2 VALVE
--- PEACM 30TTOM 2 - MAY F5 - 1005 P3WER (} .
i 3 11.A.0057 STUCK BIE$tL FUEL SALL CNECK VALVE 3l --- VERMONT rANKEE - JUL F4 i ,' j O.11.4.0103 : $ TUCK DIEEEL AIR CNECK WALVE lI 4 --- 3RUNSWICK 2 - DEC F5 - 360 MWE 'j 3.16.9.001F : SECONDARY C01TAIM1ENT VIOLATIONS * ] --- SAESSEN 2 8 3 - OCT FI - MAR F2 - ,!; 9.14.8.003F : CNECK WALVES LEAKED
--- DRESDEN 3 - JAN F3 i
t 8.16.0.0056 : CONTAINMENT CNECK VALVE MISALIGNMENT .
--- MONTICELLO - NOW F3
- t 3.16.e.0059 : .
LEAKING VACUJR BREAKER PENETRATIONS
- INADVERTENT RELEASE l' --- PEACN 30TTOM 2 - DEC F3 (POWER ESCALATION TESTIME) ,
gg O.14.8.3072 : VACUUM SREAKER LEAKAGE
--- 3VSTER CAEEE 1 - APR F4 - 442 90E .-
() 3.16.8.3081 : DEFORMATION 3F CNECK VALVE RUSSER SEAL * { --- PEACM SOTTOM 3 - MAR F5 - 1005 P3WER gg l S.14.t.0088 s REVERSE FLOW CNECES REMOVED FROM SETS - M3 TOR OVEnt0AD
--- COOPER - 94Y F5 - COLD SNUTe0W4 ,,
8.16.8.3115 : CNECK VALVE PISTON FOUND NUNE UP
--- PEACH 80TT3M 3 - OCT F5 - 555 PedER t
5.15.4.0007 : ININ WALLED VALVES - SWR $ IN SENERAL - 19F0 ((P.15.333 j t.15.A.0057 : MITR0EEN ACCJ4ULATOR VALVE LEAKED s
--- BRUNSWICC 2 - DEC F5 - SSI POWER '
I 3.16.C.0001 .3 i GUESTION ON SUITASILITV 0F CERTAIN PIPING AND VALVES
--- STSTER CREEK 1 - JUL 59 t .3 0.14.C.0036 : PROCCDURAL PROBLEM - LPCI VALVE MOTOR SUR4E9 OUT DURIuS TEST.
--- VERMONT FANKEE - M AY F3 e.15.C.0099 : PROCEDURAL LEFT OPEN PROBLEM - EXCESS FLOU CNECK VALVE'S STPA58 VALWE8
--- PEACM 30TTOM 2 s 3 - NOV F4 - 103 s 50 POWER C'
............. ALL DONE. PRESS (RETURN > (EV TO 63 T3 MEN 3 ................ 1
... - . . . . - ,m _ - . . -
g)]
.. . - -~,-a. - - . 't C
past- l' TKEAE 4RE 123 A4TICLEt$3 TO DE DISPLAVES: 19-4U4-1934 13:37:37 BW4 CNECK V4LVE5a 19F4 - 1980 0.04.3.3047 : Ces RETueN LINE ISOL4 TION VALVE N43 WOGN SEATa RJST ON PISieN *
--- 3RESSE9 3 - OCT F4 - aEFdELING SNuTD0d4 ,
B.04.3.0114 : CONTROL ROD SID NOT LATCN - DIRECTIONAL V4LVE FAILED
-- .An0WN5 FEsaf 2 - JuN 80 - 408 POWER 8.04 0.011F : WEST SDV NEADERS DID NOT taAIN AFTER M4NU4L SCR44
--- 3 RESTEN 5 - JUL 00 - 5NutDOWN ,
i 4.04.8.3145 : CONTa0L 800 BRIVE Sv5 TEM 1ALFuMCTIONS .
--- anuMSWICs 1 - Aus s0 - sJeCatrIC4L esgAaTUP) --- sauNsu!CK 2 O
!~ -
!' IES 81 - 1.53 P3WEA --- STSTER CREEK -N3V 30 - 435 P0dE4 O.05.C.0257 : TNSE4DED FAILED LOf(ING DEVICES 3N VALVE 8e PUMP 8a VALVE OPE 44T345
--- SWa*5 IN GENEAAL - M44 83 ((P.089.293)) - .*
i 9.05.C.3261 : 4D5 4It SUPPLT ACCunuLAT34 CNECK WALWE5 LEAKED , 1
--- NATCM 2 tra 53 - COLD $NuiseWN -
J 3.05.C.3264 : 4e54 Att suPPLT ACCuiut4T3R CNECK WALVES LEAKED '
--- COOPER
- APs 40 - attutLI E l
3.06.E.0054 : 'aust 04 rv C4ECK WALVE SE4TS kI --- PEACN SOTTOM 2 - ! TAR Fe - SNUTDeJN J !~ j s.06.E.0055 : FW CNECK V ALWE LE AKAsE - EuCEssIVE elsC-T3-SEAT CLEAs4NCE 1
--- DnE5aEN 2 - 1An 76 - nEtutLINs seufsodN ,
g lt
- 8.04.E.0040 4 IW CNECK VALWE5 LEAKED - WITON 0-AIN58 DETERIORATED .'
--- 3uAS-CITIES 1 - JAN FS - REfutLIga SNUTDOWN g B.04.E.0064 : DIAT IN FW CNECK VALVE 5a IMPROPER DISC-TS-SEAT CLEAA4NCE,
/
I EECESSIVE LE4KAGE - TESTING METN005 JEat INADEeueTE ' g i --- DRE5 DEN 3 - SEPT & OCT F6 - AEfutLING SNuT90WN .
~
j; 8.04.E.0045 : FWGASEET 54MPLE P43SES LODGES IN'NPCI & FW CNECK VALVEla FL4NSE SLOWN # II --- T&C00 PEA - JAN FF - APPRONIMATELT 74I POWERJ SAUNSWICK I - FEB
- SNuit0w1 4.05.E.3076 : W0aN FW VALVE SEAT / DISC ASSE18LT PINSa SEFORMED SEAT AI168
--- satsoEN 2 - OC; FF - atruft!Ns sMuTeodN .
s.C6.E.0086 : FWK4LREICNECK VALWES SEAT 5 M4:NINEDa SE4T SEAL 0-RINGS CN45GED TO l auAS-CITIES 2 - SEPT F4 - affufLING , O.06.E.0093 : W3mN SEAT / DISC A$$EntLT PINE IN FW CNECK5 ' l l enESDEN s - Man Fs - afrufLINs ! 8.04.E.009F : WORN SEATS Age AINGS IN FJ CNECK VALVES { --- MILLSTONE 1 - NAE F8 - affutLING U l l 8.04.E.010F : DIaT ON FW C4ECK VALVE SE4T O
--- SUA8-CITIES f - JAN FP - REFUELIVE I
O.05.E.0109 : WORN SuSWINGS - FW CNECK WALWE3 LEAKED 9 2
I
, PAGE 6-
-** BSUNSWICC 1 - AP4 79
- REFUELtg4 0.04.E.0810 I W004 FW CNEtt VALVE SU5NI465
--- 3eUNSWICC 2 - NAV 79 - AEFUELING e.04.E.3112 : FW CNECK VALWES LEntED - SEATS REPAI4ED, SEAL REPLACED
--- NILLETONE 1 - JUNE 79 - SEFUELINE +
8.04.E.0114 : FW CNECK VALVE KING PIN C3vte LEAE - NIEN CONTAINNENT TEMP
--- sauM5MICs 2 - Aus 79 - 94: POWEn S.06.E.012F FW CNECK WALWE LEAK - U044 SEAT # DISC ASSE99LY PINS
, DRESDEN 3 - FES 80 - SEFUELIES 8.06.E.0131 : FW LEAEED INte T0aus UNDETECTED - CNECK VALVE 804 NET l'EAL FAILED
--- WEANONT TANEEE
- JUNE 80 - 355 P3WER i S.06.E.3134 : FW ---CNECK VALWE SEAT / DISC ASSENGLV PI4S WORN #
DRE$ DEN 3 - ins 80 - REFUELING , 8.04.E.0141 : Wean SEAT, DIST IN FW CNECK VALVES *
--- SUAD-CITIES 1 - SEPT 80 - AEFUELINE $NUIDOWN
~
3.0F.A.0058 : BENT DISC WASMER, INP40 PEG SEATINE It ACIC EMMAUST VALVE
--- 10Nf! CELLO - 3CT FF - REFUELIN3 SNUID3WN 8.0F.A.0059 800GN SEATIN3 SU4 FACES IN RCIC ETEAN VALVES "'
--- COOPER - OCT FF - REF3ELING $NUIDOWN '
O.0F.4.3042 aCIC TuaSINE STEA1 EMMAUST CNECK VALWE LEAKED - FLAPPER 340EE '
--- 3UAD-CITIES 2 - SEPT F4 - AEFUELING S.0F.A.00F1 20064 AND CRACKED EMMAUST VALVE SEAT 5 ~
C00 pen - APs Fa - mEFUELINs
/
3.0F.A.00F5 : DInf --- 04 TUse!NE EuMAUST VALVE SEATE b MONTICELLO - OCT 78 - aEFUELI13 , 3.0F.A.3074 : RCIC VALVE PIM $ NEARED
--- Ba0WN1 FEnev 1 - DEC FS - REFWELING 5.0F.A.0080 : DISTf WALVE INTERNALS
--- 3 MAD-CETIES 1 - JAN FF - REFUELINE i
O.0F.A.3083 : Lo0SE VALVE DISC SLOCKED ACIC TueBINE EMMAUST LINE - BLSW4' AUPTuat DISC ,
, --- NATCM 2 - JUNE 79 - GE P3WER e.0F.A.0045 : DEFECTIVE ToteUE SWITCN, ROU3W SEATS - STEAN SUPPLY AND EENAUST
, VALVES LEAKED
--- COOPER
- APR 79 - AEFUELING . 0 #
S.0F.A.0094 : aCIC TueDINE FAILED TO START - SOLT LODGED IN STEAN EENAUST CNECc Watvt g
--- SRUNSWICC 1 - NOW F9 - SMUTDod1 a.0F.A.30as : aCIC ---
raIP - CNECc WALvf L3CK sa0cEg, VIen4TED CLOSED E NATCM 1 - DEC 79 - s0E P3WER 8.0F.a.0096 : aCIC saatN P3T LEvtL vntvE PACKIN5 agriURED - N64 MAL JEAR
3 O. o , , , 4 0 o e o o
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l
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\
k. 5 I . l
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l , - .= ** = .s .
=. o.
D
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. - . . s
. .= .
= = = = .
- . . = .
. . s . s. . . =
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, - .s = . . . = .
. w = .= .= .s
. = = =
g
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w
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.. . . - . .a . . . . a
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=
.= . .= . .
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.=
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e.
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- .=
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a -. . = = . 3
. .. == = . . . w- .a .
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= - ~ s.
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4 j . 3.
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o
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om . .
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=
h
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4
=
s .. .~ s. s. ~ .
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= -
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=.
o.
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s s o
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--- MaTCN 2 - nov 80 - 993 P3WEa 9.0F.D.8236 : ANa ---
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--- DaE5 DEN 2 - M47 SG - SNUTDowN 9.0F.D.3259 : CNECs VALVE stuCE - PL ANT NODIFIC4TI3N INSTALLED '
--- BauM5WIC4 2 - NOV 80 - 6tt PedER S.07.E.0076 : MPCI STEAM C1ECK VALVE E4SKET FAILED
--- auAD-CITIES 1 - JAN F6 - REFufLI46 SNJTDOWN +
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- 1
--- DRE5 DEN 2 - MAR F6 - $NUTDOWN I 0.0F.E.3090 : SINDING IN M*CI TuneINE EEN4WST V4LVES - auPTuat SISC SLtdN #
--- Ba0dNS FEsaf 3 - Aus F6 - 3E P3utt ," ~
S.0F.E.0005 : LOOst aETAINING Nuts, seedEN aETAINIts Stuts DISLODE'ED DISC 15 MPCI luatI4E EIN4WST CNEC( W4LVES k i
--- SauMSWICC 1 8 2 - OCT Fe - COLD SNuTD3uN '
4.0F.E.0113 : NPCT Tuastut EuM4Wsf VALVE DISC Stut FAILED, RISSING P4 aft 8
--- COOPEa - APa FF - COLD $NUTD0d1 .
f' 0.07.E.3114 : NPCI flou $$CILLATIONS AT LOJ SPEED - CNECK VALVE DISC WI9EE reaCTuaED -
- 1 i
--- saumsNICC 1 - NAT FF - C3LD $NuTDouN e-g, '
8.07.E.3121 : DIaTV EsN4ust VALVE SEnis
--- MONTICELLO - SEP FF - afrutLING suuTDOWN '
II 5.0F.E.3122 a3uGN aut $CSATCNit VALVE SEATS
--- COOPER - OCT FF - aEFJELING SNUTDeum .#
i - gg l 4.07.E.3136 : STICEING NPCI CNECK VALVE 5, WATEa N4MMEa - $Nuesta Sa4FT B40EEN ' !- --- saumsw!CC 2 - nan F8 - smuTDodt f- '; j 3.0F.E.0139 s MECN4NICAL 31TEaFERENCE IN CNECK V4LVE ;
--- Ba0NNS Feast 2 - Man F8 - aEFufLING j
B.0F.E.0152 : ELAND 3EAL C4ECK VALVE LE4KES \
~~~ PEACM 00ff0M 2 - OCT F8 - 51E 70dE4
- 0.0F.E.3155 : DIai 04 TunSINE EENAu5T V4LVE SEATS
{ l
--- MONTICELLO - 3CT F8 - affufLINE J 4
8.0F.E.0161 : CNECC --- VALVE 345KETED $ EAT FAILED SuaD-CITIES I - J44 FF - aEFufLI16 g 9.07.E.0209 j 10aus SUCTIO1 VALVES FAILED Det TO STEAM LEAK FASA A CNECC V4LVE NaTCN 2 - mer so - 99E P0wfa 08 l 0.0F.E.3213 : FOREIGN NaTEtIAL SETdEEN CNECK VALVE D0DV & SEAT l --- snowns Fraar 2 - sEP a0 - attufLING sNuT90wn '
- f3 l s.Or.r.003F : uut0i eETwEEs IssL4TIoM eL0s CNECc V4LVE aND sTE4n LI4E t.0v saussa was casss-rMafaaED g3 ,
PAGE ~9
--- saESSE1 3 - N3V F6.- SE Peuta .
4.0F.F.0050 : TILTING DISC CNECE V4LVE INSTALLE9 It Wa016 ATTITutt
--- ALL Swas - JUL FS (IP.0F.A.1FS)) ,
0.0F.F.3048 : PEseLE IN FIAE PUNP CNECE VALVE
--- #ITIP4TnICE - Aus 79 - Cate Suufstum #
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--- ORUNSWICC 2 - JUL 79 - SSS P3 DER C '
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--- SLOW T3 CLOSE - SWITCN NEEDED 49JUSTMENT ; i
' PItseIN - OCT 79 - s38 i I S.0F.F.00F1 : { VARI 3us VALVES LEAKED - DISTY SEATS *REVE1TES CLSSURE .
--- PEACM SOTTON 3 - NOV 79
- REFUELING .
i l S.0F.F.0072 :
- j. Vaa!3U5
--- 44TCMCONTAINMENT 1 - AP4, N4V, DEC ISOL4TIO1 VALWES*
F9 - REfJELINS LEAE AND 973e4TESP3dE4UNACCEP.TleLE
- NATCN 2 - M4T 79 - SuuTD3dm l y 3.0F.F.00F8 : VARI 3US PSIM4RT CONTAINNE1T ISOLATION VALVE LE4ES* .
--- auAs-CITIES 2 - Nov-eEC F9 - atrJELINs FEs as ji 3.07.f.0001 : .' VaaI3MS C01T4INNENT ISOLATIS1 VALVES FAILED LLRT %
li --- 104TICELLO - FES & N4R SS - AEFUELING
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4.07.7.3083 : VARI 3us VALVES LEAKED IN EXCESS OF TECN SPECS
*-* MATCN 2 - N44 8 4PR 83 - COLD Suutt0W1 '
j B.0F.F.0089 :
- j PRInant CONT 4INNENT PENETeATION LE4ES EXCEEste TECM S*ECS
.'. --- COOPER - NAV 80 - affutLING
- j. 9.0F.F.3131 : VAaI305 VALVE LE4ES
--- 1ILLST04E 1 - OCT a0 - afrufLI1s -
f i s.cs.C.00s4 : , 4, {. EuCESsIvE AaN PL4T sErvEEN Stav!CE WATEa CNECE WALVE eISC aus sISC
--- VERN0NT FANKEE - AuG F6 - 931 POWER 2
! 8.09.C.3121 : LEAKI+46 DRAIN VALVE, C0CEED SPRI16 31 AECISC PUNP SE4L PRES $dEE I VALVE 4 1
--- COOPEn - SCT FF - afrutLINs SWJTeeuN #
f s.03.C.3171 : C0ansete SERWICE WATER VAULT DAAIN V4LVE SPAINGS *
. --- auAD-CITIES 1 - FEa 77 - 4ErutLI16 #
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--- stuuSw!C4 2 - AUS FF - 338 Poeta ~#
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--- NATCM 1 - JAN 80 - 945 P3WER .>
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--- auAn-CITIES 2 - eEC Fa - afrutLIss l-l l
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