ML20203G249
| ML20203G249 | |
| Person / Time | |
|---|---|
| Issue date: | 06/02/1986 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2422, NUDOCS 8608010041 | |
| Download: ML20203G249 (47) | |
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> DATE ISSUED: 6/2/86 d
, s) Advisory Committee on Reactor Safeguards
- s. '
- Thermal Hydraulic Phenomena Subcommittee s
p Meeting Minutes
.g May 21, 1986 Washington, DC PURPOSE: The purpose of the meeting was to review selected portions of the NRC RES research Program that deals with thermal hydraulic issues for the Committee's Report to the Commission on the FY 1988 budget.
ATTENDEES: Principal meeting attendees included:
ACRS NRC D. Ward, Acting Chairman B. Sheron, NRR ,
J. Ebersole, Member L. Shotkin, RES
- v. Schreck, Censultant D. Bessett, RES C. Tien, Consultant C. Troutman, RES P. Boehnert, Staff F. Odar, RES B. Agrawal, RES G. Ree, RES Y. Chen, RES N. Zuber, RES J. Hopenfeld, RES J. Reyes, RES R. Jones, NRR L. Marsh, NRR N. Lauben, NRR T. King, NRR A complete list of meeting attendees is attached to the Office Copy of )
these Minutes.
8608010041 860602 ACRS DESIGNATED ORIGINAL PDR Certified By t
T/H Phenomena Meeting Minutes May 21, 1986 MEETING HIGHLIGHTS, AGREEMENTS AND REQUESTS
- 1. D. Ward, Acting Subcommittee Chairman convened the meeting at 8:30 a.m. J. Ebersole noted that the ACRS should work to see that the NRC research Program is looking at the big picture.
- 2. L. Shotkin (RES) discussed the FY 88 budget in closed session. He noted that this will be a " snapshot" of the budget as RES is reformatting to conform with the Agency Mission Area Code approach.
Figures 1 and 2 provide the details. Figure 1 shows the budget for each MAC that includes T/H funding along with the related programs in each MAC. Figure 2 shows the budget cross-cut by reactor type.
RES has proposed a request to Congress for B&W-related testing.
For FY 87 and 88, this amounts to $7.2M and $7.4M, respectively.
In response to Mr. Ebersole, Dr. Shotkin noted that B&W is now evolving into a service oriented company. Mr. Ward asked why the code related work is split between MACS. Dr. Shotkin said the logic for the split is difficult to defend, noting that the distribution of this work effort is still evolving.
Dr. Shotkin also noted that Industry support to the tune of
$10M/ year for FY 87-88 is being sought for follow-on testing to support B&W reactors. RES :,3id lacking industry support, there would be no high-pressure OTSG. If no Congressional support is found, there would no low-pressure OTSG and the MIST follow-on Program would be severely restricted. The Congressional request is devoted solely to B&W geometry related work.
Mr. Ward said he believes that the B&W Owners Group should strongly support the follol-on effort as he said they have a serious problem to resolve here.
s .
T/H Phenomena Meeting Minutes May 21, 1986 In response to Mr. Ebersole, Mr. Jones (NRR) said the question of increasing the size of the candy-cane vent would be investigated as part of the MIST follow-on effort. NRR also indicated that obtain-ing Industry support for the follow-on effort would be subject to the provisions of the Backfit Rule.
- 3. D. Solberg discussed the T/H Technical Integration Center (TIC) that has been established at INEL. The key objectives of the TIC are: (1) to assure that sufficient breadth of technically expert staff and experimental capability are available and can rapidly respond to priority safety issues as they arise in the future; (2) to provide expertise to assist NRC in defining potential future safety issues and defining new research programs for application of existing research results that are needed to resolve both current and future safety issues; and (3) to demonstrate the feasibility and value of expanding the scope of the TIC to include risk assess-ment and human reliability expertise. In response to Dr. Catton, RES indicated that NRC will exercise strong management control of the TIC. Mr. Solberg also indicated that in order to assure competitiveness and access to other expertise, some work will be contracted to othen Organizations (Universities, etc.). Mr. Reed expressed concern that using a TIC may result in delay in resolving critical issues such as assuring reliable DHR. Figure 3 shows a schematic of the TIC work structure. A new effort at the TIC is demonstration of the T/H codes' applicability to model the vendor designs and the relevant plants' operations (see below). In response to Dr. Catton, RES said the codes to be used for this effort are TRAC-P&B and RELAP-5.
During a lengthly question and answer session, Dr. Tien suggested that an oversight Group and/or a Review Counsel should be formed to guide the overall TIC effort. RES noted the suggestion. Mr. Reed expressed concern with the tendency of the TIC towards self-
T/H Phenomena Meeting Minutes May 21, 1986 perpetuation. Dr. Catton observed that the TIC seems to be tending to the problems of the past and not the current concerns. Mr.
Schrock asked how the other TICS will interact with the T/H TIC.
Dr. Shotkin said there are plans to set up two other TICS in plant m6intenance and plant aging. He also said RES wants to broaden the T/H TIC to include plant operational related matters. Mr. Ward, echoing Mr. Schrock's concern, noted that the TIC concept will te greatly changed if this TIC is to be broadened to include other areas of expertise. He would prefer the T/H TIC access, not include, other areas of expertise. RES expressed concern that if other expertise is not included, it may not be available at all.
Further discussion resulted in the concern expressed by the Subcom-mittee that the T/H TIC may not be focused tightly enough on T/H issues. It was suggested that such areas as human reliability, PRA etc., should be input to the TIC not be a part of the TIC. Dr.
Catton cited the RES effort on PTS as a good example of how the TIC should be run; i.e., a diverse work effort is " tied together" at the TIC. RES responded that they see their effort as assuring the T/H people are interacting with, for example, the PRA people as necessary.
In response to Mr. Ward, RES said the "NRC Contact" (L. Shotkin) would assure coordination between the TIC and other work efforts at, say, LANL.
The proposed code applicability study was outlined (Figures 4-5).
In effect, this is an Assessment Program directed at the modeling of each of the four vendors' reactor types. In response to Mr.
Ward as to how one judges when the code is adequate, RES said one can judge the code's adequacy by careful comparison against separate effects and integral tests.
T/H Phenomena Meeting Minutes May 21, 1986 Dr. Solberg reviewed the status of the experimental program (Figure 6). He noted two key points (1-2, Figure 6). The future experi-mental program at the TIC was outlined (Figure 7). This Program assumes no Industry support and is considered a baseline effort.
It does assume additional Congressional funding will be available.
In response to Mr. Ebersole, NRR said the concerns raised with the ICS and NNI failures at B&W plants will be addressed as part of NRR's on-going reevaluatiori of B&W plant design.
Turning to the proposed new integral test facility, RES said it is focusing on a 1/10 linear scale full-pressure-water type. Mr. Ward asked the Consultants if they concurred with this approach. The Consultants generally agreed that this approach is acceptable.
(Note: V. Schrock independently consulted with INEL on this matter putting him in conflict on this item.) Drs. Catton and Tien, in response to Mr. Ward, agreed that full pressure is not really needed. RES noted that funding constraints limited consideration to a scale no larger than 1/10th, and that less than full pressure design makes analysis of transient behavior very difficult. An RES-sponsored Experts Meeting will be held to obtain final agree-ment on the best approach to persue, t
On moving existing test facilities to INEL (Figure 8), RES recom-mends against it.
- 4. The B&W Operational Methods Development study was detailed by D.
Solberg. This is a new effort being proposed by RES. Key objec-tives of this effort include: (1) to collect the available methods and data needed to perform an integrated Risk Analysis, human reliability analysis and thermal hydraulic analysis of a B&W plant and demonstrate the ability to integrate these methods and provide expertise needed in their application to plant assessment; and (2)
T/H Phenomena Meeting Minutes May 21, 1986 to assure that the modeling of one B&W plant (0conee) is of the highest reasonable quality and to provide a baseline risk value for Oconee.
It is planned to complete development of this effort in early 1987 so this capability is available for NRR's mandated reevaluation of B&W plants. A schematic of the methods development concept and schedule are given in Figure 9 and 10.
Mr. Reed decried the attempt to always " upgrade" plant operators and suggested NRC focus more on design flaws and vulneribilities that make plant operation very challenging. Mr. Ward observed that there are some utilities that can successfully run B&W plants and i that other utilities evidence room for improvement. Mr. Jones (NRR) said this Program will not address what fixes should be made to B&W plants. He said it will be an aid in evaluating the impact of any proposed fixes for B&W plants. Mr. Jones speculated that some combination of hardware and procedural fixes will be needed to address NRR's concerns with B&W plant transients.
In response to Mr. Ward, RES said that this development program will be " tested" on Oconee. Oconee was chosen because of the data and methods (PRA) available. Oconee will provide the baseline risk value. Figures 11-12 detail the milestones and products of the Program.
Mr. Ward expressed concern that the focus of the program is on Oconee and that Oconee has not evidenced any major problems. This may skew the results of the effort. The focus should be on a less optimal plant (ex: Davis Besse, Rancho Seco). RES expects that the Program can be modified with input representative of other plants to model their respective problems. In response to Mr.
Reed, L. Marsh (NRR) said this effort is independent of the USI 1
o
T/H Phenomena Meeting Minutes May 21, 1986 A-45 resolution effort. Mr. Reed suggested RES coordinate with NRR on the study of the B&W plant selected for the A-45 analysis (ANO-1).
R. Jones (NRR) commented on this Program vis-a-vis the NRR mandated B&W reevaluation effort. He said the two efforts are independent and NRR had not submitted a request for the RES Program. NRR is in-effect taking a wait-and-see attitude towards the RES effort.
NRR understands that for this Program, RES is linking T/H with operator evaluation and factoring these into a PRA to assess the integrated plant risk. In response to Mr. Ward, NRR said they do not know if they will be able to use the RES Product. It may be possible to use it if RES keeps to their schedule. Since the B&W Owners have the lead on the NRR effort, it may be that the Owners-proposed fixes will resolve NRR's concerns. Dr. Catton expressed concern that NRR does not have detailed knowledge of the RES Program. T. King (NRR) noted NRR has not approved funding for this Program because of differences between NRR and RES vis-a-vis the MIST follow-on effort. The problem is one of incompatible schedules, i.e., NRR and RES have different deadlines to finish their respective programs. Dr. Catton said this situation raises )
concerns with the performance of NRC Management.
- 5. R. Lee overviewed the proposed MIST follow-on Program. Currently, the B&W proposal is to run a follow-on (Phase III) test series for
$4M over 18 months without a power upgrade. Figure 13 shows the proposed tests.
At the last PMG Meeting (5/12-13/86), Toledo Edison (TE) proposed a series of tests concurrent with Phase III. Figure 14 (arrows) details the TE tests. Figure 15 shows the feed and bleed tests proposed in Phase III.
T/H Phenomena Meeting Minutes May 21, 1986 The MIST-related OTSG separate effects tests proposed at INEL were noted. Two facilities are under consideration: (1) low pressure visual air-water loop; and (2) a high pressure steam-water / freon loop. Item 2 is contingent on obtaining additional Industry funding: Item I requires additional funding from Congress. The Consultants noted that study of the physics of the OTGS AFW behav-ior is what is really needed.
N. Lauben, NRR, provided comments on the MIST follow-on proposal.
NRR noted that: (1) NRR endorses the new RES proposal, and (2) however, they believe that alternatives dependent on funding should be clearly spelled out, and the effect on code uncertainty snould be quantified. Further discussion clarified the point that NRR endorses the entire Program (with Industry funding) with the proviso that if funding restrictions result in a smaller Program -
RES should evaluate the impact of a reduced effort on code uncertainties.
- 6. The status of the International Code Assessment and Applications Program (ICAAP) was reviewed by D. Bessette. ICAAP will attempt to l obtain a quantitati,ve measure of code accuracy. Dr. Schrock asked how noding will be handled. RES indicated that each code is !
accompanied by a set of noding guidelines. Dr. Catton suggested that the noding be frozen to assure good comparisons. F. Odar said 1 RES plans to check the code results vis-a-vis different noding schemes. Dr. Tien said some type of independent noding guidelines are needed. The codes are still quite user-dependent.
The ICAAP status is shown on Figure 16. To date, France is the only major nuclear-user Country not yet in the Program. The methodology to be used for quantifying code uncertainty is under !
development. It should be finalized in September 1986.
9 O T/H Phenomena Meeting Minutes May 21, 1986 The key parameters to be used for quantifying code accuracy were noted (Figure 17). The Consultants suggested that RPS inventory distribution and quantity should be included in this list. RES indicated that this comment is valid and they will give it study.
- 7. C. Troutman provided the status of the Plant Analyzer / Data Bank program. The scope and status fcr FY 87 and 88 are given on Figures 18-19. Both efforts are scheduled to end in FY 1988. In response to Mr. Ward, RES said that possible development of a NPDB for BWRs must await completion of the TRAC-B code.
- 8. F. Odar reviewed the plans for future code development and assess-i ment. Key points noted included:
1
- TRAC PF1-/ MOD-1 was frozen in FY 84 and is under ICAAP assess-ment. PF-1/M00-2 will be released in FY 1987. M00-2 includes changes in: entrainment/deentrainment modeling, multisource connection capability, steam / water separator model, upper tie plate CCFL model, and core void distribution model. A M00-3 version, primarily designed to model B&W plant behavior, is under development and should be released in FY 90 (Figure 20).
RELAP-5/ MOD-2 (Cycle 36) was frozen in January 1985 for l on-going ICAAP assessment. A M00-3 version will be developed l based on errors and user convenience features identified under ICAAP. Release of MOD-3 is expected in FY 87.
TRAC BD-1/M00-1 was frozen in December 1984 and is also under
> ICAAP assessment. The BF-1 version is currently under devel-opmental assessment. It should be released for use on June l 15, 1986. A BF-1/ MOD-1 version is planned to be issued in FY !
- 88. l l
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1 l
T/H Phenomena Meeting Minutes May 21, 1986
- 9. The results of the SASA Program applicable to the issue of feed and bleed (F&B) capability was discussed by B. Agrawal (RES). This work was conducted under USI A-45 support activities at LANL. Key conclusions of the study were:
The availability of high-pressure SI delivery capacity greatly enhances the effectiveness of the feed-and-bleed operation.
Plants with only low-pressure or intermediate-pressure SI systems must initiate feed and bleed no later than the loss of secondary heat sink. Plants with high-pressure SI systems can successfully use feed and bleed no later than the loss of secondary heat sink, and can successfully use feed and bleed until the time of primary system saturation.
PORV capacity becomes important duri g the transition from reactor trip to either residual-heat-removal or low-pressure-injection entry conditions.
Simple inspection is a useful technique for extending the limited set of detailed plant-specific calculations to a broader set of plants. There is less confidence in the accuracy of conclusions reached by simple inspection than those based on detailed plant-specific calculations.
Mr. Ward asked if NRR can use TRAC-PF1 to evaluate a bleed and feed system such as is contemplated for Davis Besse (DB). N. Lauben gave a personal opinion that the codes are adequate for such an evaluation.
Messrs Ebersole and Reed raised concerns with the details of the ,
analyses needed to support the DB blowdown system. In response to l 1
Mr. Ward, Dr. Zuber said he does not have a great deal of confi- )
dence in the predictive capability of TRAC for the above analyses. I
T/H Phenomena Meeting Minutes May 21, 1906
- 10. G. Rhee (RES) overviewed the status of the 20/3D Program. Key items noted included:
CCTF tests at JAERI were completed in March 1985. SCTF tests with core-III are on-going and will be complete in August 1987. UPTF shakedown tests are underway. Testing will begin in October of this year and end in FY 88.
The TRAC analyses status and model improvements associated with 2D/30 are given on Figure 21.
1 Some investigation of B&W plant design features will be conducted in 20/30. This includes experiment / analysis of 3
reactor vessel vent valve behavior, and downcomer injection.
A test based on best-estimate LOCA code parameters run in CCTF resulted in hydraulic shocks and facility shaking. Analysis indicates such phenomena are not expected in a full size plant due to runback of water from the steam generators to the i vessel.
I
- 11. Y. Chen discussed the status of the ROSA-IV Program. Key results are:
i Eleven tests have been run in LSTF (Large Scale Test Facility). A counterpart test to the Semiscale SB liquid holdup test failed due to a vessel leakage problem. Na tural circulation tests showed: (1) in the single-phase NC, flow between S.G. tubes is non-uniform and a symmetrical flow occurs in both loops; (2) in two-phase NC, the natural circulation flow increases as the vessel mass inventory decreases below 90%, and the maximum loop flow occurs at i 75-80% mass inventory; and, (3) in the reflux condensation
e o T/H Phenomena Meeting Minutes May 21, 1986 mode, the circulation flow stops and reflux cooling begins when the vessel mass inventory decreases to 55-65%.
The LSTF test matrix for Japanese FY 1986 (4/86 - 3/87) is given in Figures 22-23. The last three tests listed are rated j a low priority by NRC, and may not be run given the ambitious schedule.
Figure 24 gives the status of the TRAC analysis effort sup-porting the LSTF tests.
The ROSA-IV Program experiments will be complete in 1989 with the associated analyses complete by 1990.
- 12. The details of the separate effects research program were presented by J. Hopenfeld and N. Zuber. Mr. Hopenfeld discussed the programs associated with steam generator tube rupture (iodine dose calcu-lations, loss of feedwater, steam line break, and steam ex-plosions).
Results of code calculations applied to result of MB-2 tests show potentially high iodine doses for B&W and 3800 MW(t) CE plants (Figure 25). In response to Subcommittee questions, the large differences in doses between the above plant types and B&W and W plant types apparently results from the differences in stean.ing rates. Dr. Catton and Mr. Ward questioned the validity of these results.
The results of steam explosion studies based on industrial experi -_____ _
ence (usually paper industry steam explosions) were noted. Analy-sis of the results show explosion efficiencies are quite low (0.01
- 0.9%). Scaling to full size should reduce the efficiency even further. Mr. Schrock questioned the methodology used by RES to
T/H Phenomena Meeting Minutes May 21, 1986 calculate these efficiencies, noting this methodology is different from that used by most other researchers in this field.
Prior to his departure, Mr. Ward asked for a written report from the Consultants before June 1, 1986 in order to support the ACRS MIST follow-on topic.
N. Zuber reviewed other separate effects programs. These programs include a study of boron injection behavior during a BWR ATWS (Figure 26 details the related tasks); a study of water hammer phenomena based on recent WH events at San Onofre, Indian Point and Millstone Unit 1. (This Program is under development - Figure 27-29 provide details.)
RES is exploring establishment of resource centers at Universities to perform basic T/H analyses and experiments for NRC.
A Program is also under study to study check valve dynamics. The goal is to develop a check valve operability map (Figure 30). M r .'
Reed questioned the necessity for such a program. Rather, he is concerned with the, design of the valves. Mr. Ebersole said one must study the upstream dynamics a valve is subject to given a transient or accident.
Dr. Shotkin said the RES-proposed program on water hammer is in response to ACRS comments on the topic. He solicited Committee comments on the Program. Mr. Reed said the focus of the Program must be on minimizing major safety related water hammer events.
Mr. Ebersole said he is concerned with reliance _on_ administrative _._____ _ _
controls to prevent water hammer.
- 13. J. Reyes (RES) discussed the status of PTS research. Figures 31-32 overview the status. Three thermal hydraulic areas to be studied
e o T/H Phenomena Meeting Minutes May 21, 1986 further include: (1) a review of the TRAC code's ability to predict the onset of loop stagnation (LANL/SAI/UCSB) (FY 86-87);
(2) thermal fluid mixing due to side entry HPI at low Froude number (FR 1) (UCSB) (FY 86); and, (3) effect of noncondensible gases on thermal fluid mixing (FY 87-88).
- 14. T. King (NRR) discussed NRRs user needs for research. He noted that NRR has not yet agreed with RES on the provisions of the B&W Methods Development Program, but they are working with RfS on the Program details.
NRRs four priority areas include (in descending order):
Experimental data to validate B&W codes (TRAC-P/RELAP-5) for B&W plants.
NPA/NPDB analytical capability and input data.
Separate effects data as necessary for code validation.
Although NRR has no identified testing needs at this time, beyond that necessary to validate B&W codes, it is their opinion that it would be prudent to have available a facility or facilities for obtaining additional experimental data.
- 15. The meeting was adjourned at 5:50 p.m.
r NOTE: Additional meeting details can be obtained from a transcript of this meeting available in the NRC Public Document Room, 1717 H Street, NW, Washington, DC, or can be purchased from ACE-Federal Reporters, 444 North Capital Street, Washington, DC 20001, (202) 347-3700.
9 P
1987 1988 1989 1990 fMC 111: 1 INCIDENT RESPONSE (NPA/DB) SCO 9m l 500 500 fMC 115: CHANGES TO LICENSE CONDITIONS 2 900 (2D/3D, ROSA-IV, ECCS RULE REVISION) 5300 5700 2600 1000 1200 1300 1300 fMC 121: LICENSE CURPENT DESIGNS (CODE IMINTENANCE) fMC 122: PREPARE FOR FUTURE LICENSING 700 800 850 800 (CODE ASSESSKNT 8 IMPROVER NT) fMC 131: EVALUATE SAFETY SYSTEMS UNDER ACCIDENT CONDITIONS (CEC, TIC, MIST, SEPARATE EFFECTS, ICAP) 8800 7450 3 17500 16700 16400 16100 22700 20100 TOTAL PROPOSED CONGRESSIONAL ADDITION MAC 131: B8W TESTING (OTSG, INEL SITE llPGPADE, MIST) 72m 7400 0 0 236CO 23500 22700 20100 TOTAL POSSIBLE INDUSTRY SUPPORT 1000 2500 3200 3400 IMC 131: FOLLOW-0N MIST 8500 300 0 HIGH-PRESSURE OTSG 9000 NOTES: 1. COMPLETE PWR NPDB CAPABILITY
- 2. COMPLETE 2D/3D PROGRAM
- 3. CEC: 0TSG TESTING 8 ANALYSIS: B8W 0.1 INTEGRAL. CONSTRUCTION:
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RSPB FUNDING CROSS-CllT BY REACTOR TYPE 1988 1989 1990 1987 11.8 14.7 13.4 13.0 BaW 1.7 1.9 2.0 2.1 GE 8.9 9.8 6.0 4.6 W/CE 23.5 22.7 20.1 23.6 TOTAL 16.4 16.1 22.7 20.1 BRG 7.4 0 0 7.2 CONGRESSIONAL ADDITION
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9 Work Structure Related to the Technical Integration Center Act it s nd NRC i ~~~~~~~~~
i i Other r---~~~~~ Coordination Across Contact Technical Disciplines l
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Priority Problem A{o Vendor { plicability Program Studies Support Designs &
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OPERATIONAL ggg Wh SAFETY STUDIES (B&W) p . .-t CE tz':'- ri Appendix K Rev'Islon pr GE -Expwinests _
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'1 CODE APPLICABILITY STUDIES OBJECTIVE: TO PROVIDE FOR EACH VENDOR TYPE OF REACTOR DESIGN A CONCISE REPORT OF THE TO WHICH THE APPLICABILITY OF EA01 MAJOR T/H CODE HAS BEEN DEMDNSTRATED S01EDULE: B8W REACTOR FY 1986-1987
}!/CE REACTORS FY 1988 GE REACTORS FY 1989 THEREAFTER EA01 STUDY UPDATED REGULARLY TE0lNICAL BASIS: CODE'S CONSTITlfflVE MDDELS EVALUATED FOR:
PHENCEN0 LOGICAL REPRESENTATION SCALING LIMITATIONS OF DATA BASE GE0TTRIC SIMILITUDE OF DATA BASE RANGE OF FLUID CONDITIONS IN DATA BASE CODE COMPARED WITH APPLICABLE SEPARATE EFFECTS AND INTEGRAL EXPERIENTAL DA d
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EXKRIKNTAL PROGRAMS BASIS ,
FIST SHlffDOWN IN 1984, SLOWLY DEGRADING, NEEDS CORE SENISCALE SHLlTDOWN IN 1986, IN A COLD, PRESERVED STATUS
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- MIST TESTING IN 1986, PROPOSED FOLL0ki-ON TESTING IN FY 1987-1988 p) GREATEST REMAINING NEED FOR DATA IS RELATED TO WIDE RANGE OF TRANSIENT RE AND PARTICULAPLY AUXILIARY FEEDWATER EFFECTS ON RESPONSES
- TO ENHANCE UNDERSTANDING OF SYSTEM RESPONSES AND TO IMPPDVE CODE ASSESSTENT, FUTURE INTEGRAL I! FACILITIES SHOULD EMPHASIZE MULTI-DI K NSIONAL RESPONSES WITHIN THE SYSTEM i
I ADDITIONS TO NRC'S T/H BUDGET OF $7 MILLION IN EACH OF FY 1987 AND 1988 ARE PEQUIRED TO PROVIDE e
A MINIMALLY ACCEPTABLE EXPERIMENTAL PROGRAM INDUSTRY SUPPORT WILL BE SOUGHT FOR EXPERIMENTAL PPDGRAM ENHANCEMENT FOR B8W REACTOR : TYP
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NRC EXPERIMENTAL PROGRN4 1986 1987 1988 1989 1990 isEW GE0tGRY MIST A MIST FOLLOW-ON n OTSG VISUAL SEPAPATE EFFECTS d 0TSG STEAM / WATER FACILITY g FULL POWER MIST OTSG TEST a INEL 3 1/10 LINEAR DESIGN -ji CONSTRUCTION A WESTINGHOUSE GE0 ETRY SEMISCALE 4 1/10 LINEAR DESIGN
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8 TECflNICAL INTEGRATION CENTER 6/86
-- - _ - _ _ _ = -- - - . . _ _ . . _ _ - - _ . - - _
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'I MDVING EXISTING FACILITIES T0 INEL
.i ADVANTAGES: ALL FACILITY TYPES COULD BE UNDER TEST OR MAINTAINED IN A lEST-READY STATUS 4
COULD SUPPLEMFET DATA FROM NEW FACILITY CONCEPTS DISADVANTAGES: SIGNIFICANT INITIAL AND ANNUAL COSTS WITH NO DEFINED PRODilCT NEEDED WITH THE PROPOSED MIST FOLLOW-ON PROGPAM, MOST OF USEFUL DATA FROM TliESE ,
l FACILITIES HAS ALREADY BEEN OBTAINED ,
COSTS: MOVE AND INSTALL FIST AT INEL 51300 -
(W/NEW CORE) i MOVE AND INSTALL MIST AT INEL $3,500K j
- RECCMENDATION: KEEP FACILITIES INTACT AT CURRENT SITES AND WITH MINIMAL DEGRADATION UNTIL ,
REPLACED BY THE 0.1 LINEAR FACILITY i (
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Operational Safety Integration Concept Phenomenological iluman Performance P rf man e Risk Assessment Methods Behavior Methods Methods Methods i 8 Integration j 8' 9f' 4 go* e Operalional Safety
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dchedule for Integrated Methodology Assessment of B&W Plant Safety 1986 Month 19s7 MAR e APR I MAY l JUNE I JULY l AUG 1 SEPT i OCT l NOV i DEC'I JAN FEB MAR 1 APR l MAY l JtJE l bltiri Intayotion f, O@ N W M late 7otM A p,,,, gog,,,ng
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Y 0 ICS Pleture fa WA l l R .. imio so Oconee Assessment Method Application Contraf FLCA l l ,
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PROGRAM MILESTONES AND PRODUCTS LIST OF I
MILESTONE NUMBER - PRODUCT DESCRIPTION 1 A SET OF . INTEGRATED METHODS CAPABLE OF ASSESSING PLANT OPERATIONAL SAFETY l 2 A DRAFT NUREG REPORT DESCRIBING THE 1
INTEGRATED METHODS AND THE ASSESSMENT OF OCONEE-3 RELAPS-MOD 2 DECK FOR THE 3 A COMPLETED DAVIS-BESSEE PLANT A COMPLETED RELAP-MOD 2 DECK FOR THE 4
OCONEE-1 PLANT (OCONEE-1 AND OCONEE-3 ARE E.SSENTIALLY IDENTICAL PLANTS) 5 AN INFORMAL REPORT DESCRIBING COMPARISONS B&W, OF THERMAL HYDRAULIC RESPONSE OF '
WESTINGHOUSE, AND COMBUSTION ENG'INEERING PLANTS '
6 AN INFORMAL REPORT DESCRIBING THE CODES APPLICABILITY OF THERMAL HYDRAULIC TO B&W PLANTS RESULTS 7 AN INFORMAL REPORT DESCRIBING THE lOF'THE OCONEE CONTROL SYSTEM FMEA
LIST OF PROGRAM MILESTONES AND PRODUCTS (continued)
MILESTONE NUMBER PRODUCT DESCRIPTION g AN INFORMAL REPORT DESCRIBING IDENTIFICATION OF SIGNICIANT OPERATIONAL SAFETY ISSUES IDENTIFIED BY A FUNCTIONAL APPROACH 9 AN INFORMAL REPORT DESCRIBING THE.
'. RESULTS OF THE OCONEE TASK ANALYSIS AND HRA 10 AN
~ INFORMAL REPORT DOCUMENTING T.H E RESULTS OF THE OCONEE PLANT ASSESSMENT -
jf. A PRA CATALOG FOR B&W PLANTS
%g '
s j * *
. = . . . .... =- .
e MIST FOLLOW-ON WITHOUT POWER UPGRADE (4/11/86)
TYPE OF TEST # OF TEST SG1 SG HEAT TRANSFER VARIATIONS WITH 2 CONDITIONS IMPOSED BY A FEED LINE BREAK SG2 ABNORMAL SG CONDITIONS DUE TO A LOSS OF 2
FEEDWATER OFILL OVERFILL AFFECTED SG 2 ECCS1 HEAD AND HLUB SPRAY 3 ECCS2 USE OF LOCA VALVES 2 ECCS3 USE OF PRIMARY BLOWDOWN SYSTEM 2 F&B4 F&B WITH INCREASED VENTING CAPACITY 1 F&B5 OPEN HOT LEG HIGH POINT VENTS WITH PORV 1 SGTR6 INTERMEDIATE SGTR SIZE 1 NCCOLL1 NATURAL CIRCULATION WITH SOLID PRESSURIZER 1 SBLCCA1 SBLOCA WITHOUT HPI 3 SBLOCA3 STATION BLACKOUT 1 SBLOCA7 LOWER RV-DC BREAK 1 MAP 2 EXTENDED MAPPING CONDITION 1 1
I l
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- - - ~
i MIST PMG MEETING 12
- HELD ON MAY 12-13, 1986 AT EPRI 0FFICE IN WASHINGTON, D. C.
- CHARACTERIZATION TESTS WILL BE DONE BY END OF MAY
- TOLEDO EDIS0N TESTS
- TOLED0 EDISON REQUESTS TO PERFORM 4 TESTS IN MIST, CONCURRENT WITH PHASE III y - CONTINUOUS VENT LINE TESTS (3 TESTS)
(1) CHARACTERIZATION (11) NATURAL CIRCULATION C00LDOWN (iii) SBLOCA WITH NCG OBJECTIVE IS TO PROVIDE EXPERIMENTAL DATA TO SUPPORT INSTALLATION OF A CONTINUOUS VENT LINE y - HPI-PORV TEST (1 TEST)
OBJECTIVE IS TO PROVIDE EXPERIMENTAL DATA FOR COMPARISON WITH AN EXISTING CODE PREDICTION OF A PLANT TRANSIENT
l .
FEED AND BLEED TESTS (PHASE III)
VARIABLE 1 2 3 ECCS ECCS RCP CAPACITY DELAY j SETTING 1 Full PORV Lift Off 2 EM 20 mins On
- a. 330100 1 1
- b. 330201 2
- c. 330302 2
- d. 360499 2
- e. Toledo Edison HPI-PORV FEED AND BLEED TEST (OTIS)
TEST 220899: HPI-PORV C00LDOWN b*
+
w-- ,- , ---%-- - - - - - -- ._ - , -
y w ,, _-39
a ICAP STATUS o EVISED GUIDELIES AND PROCEDURES ISSUED IN APRIL,1986 TO PROGRAM PARTICIPNITS 4 o PROGRAM MANAGEENT GROUP ET INITIALLY IN OCTOBER,1985 AND WILL KET AT LEAST ANNUALLY o SRCIALIST EETINGS AND TEGINICAL GROUP IETINGS WILL BE IELD AT LEAST RI-ANNUALY, ONCE IN LNITED STATES AND ONCE IN EUROPE o FIRST SECIALIST KETING OF CODE USERS TO E ELD 9-11 JtNE,1986 AT KRAFTWERK UNION, ERLANGEN, F. R. GERMANY o SECIALIST EETING TO BE FOLLOWED BY TEONICAL GROUP EETING ON 11-12 JtfE,1986 0 13 COLNTRIES ATTENDING: ELGUIM, FINLAND, GERMANY, ITALY, JAPAN, KOREA, NETERLANDS, SPAIN, SLEDEN, SWITZERLAND, TAIWAN, UNITED KINGDOM, UNITED STATES o FRANCE TIE ONLY MAJOR COUNTRY STILL NOT IN ICAP. AGREEENT MAY E NEARING FRi r
. i l
TABLE 5 KEY PARAMETERS FOR QUANTIFYING THE ACCURACY OF THE CODE
~
Applicable to .
Key Parameter ,
PWRs BWRs Primary system pressure , X X Primary system break and/or valve flow X X ECC injection flow rate (each operative system) X X Vessel or core collapsed liquid level X X Core cladding temperature at i, i, 3_ and peak power region X X Core inlet mass flow rate X X Nuclear core power (reactivity transients only) X X Pressurizer liquid level X Secondary side pressure X Secondary side break and/or valve flow X Steam generator primary and secondary liquid levels X Hot leg mass flow rate X Cold leg mass flow rate X Steam line and recirculation loop flow rates X Sincie values Blowdown peak cladding temperature X X Reflood peak cladding temperature X X Minimum core liquid level X X Maximum or minimum primary system pressure X X Peak core pcwer (nuclear simulation only) X X Maximum or minimum secondary side pressure X Times of core uncovery X X Times of CHF at the cladding hot spot X X Times of peak cladding temperature (blowdown and reflood, or other) X X Time to final core quench X X Time to rewet and/or quench at the hot spot X X Times of ECC initiation (each operative system) X X Time of MSIV signal generation X X Time of loop seal clearing (i.e. time of first significant vapor flow through seal) X X Time when primary pressure is less than steam generator pressure X Time to pressurizer empty or full X Times of pressure maxima and/or minima (primary and secondary) X X t
25
[/8* l
NPA/NPDB FY 1987 NPDB NPA TASKS
- COMPLETE RiASE I NPDB DESCRIBED UPDATE USER'S MANUAL UNDER FY 86 PROGRAM DEVELOP 1 MDPE INTERACTIVE /PEDUCED DECK AfD MASK
. DEVELOP 2 COARSELY f0DED FAST-RUNNING NPA DECKS FROM EXISTING HIGHLY f0DALIZED DECKS CONSULTATION TO USERS. MAlffTAIN NPA SOFTWARE AT INEL AND LANL
~
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~
NPA/NPDB P
FY 1988 NPDB NPA l
j TASKS l
- COMPLETE PHASE II NPDB EfalANCE EASE OF CREATIf0 f%SKS AfD MASK DRIVERS
- EXPAND CAPABILITIES OF NPDB l DEVELOP a DEMONSTRATE THREE PLANT DECKS AfD TO E8W AND CE TYWS PLANTS MASKS / DRIVERS TO OPERATE IffTERACTIVELY
- EXPAND CAPABILITIES TO MAINTAIN AND UPDATE CALCULATION INDEX DATA BASE WESTItGIOUSE TWO AND THREE
- LOOP PLANTS ADD FIVE CALCULATI0f6 TO NPA REPLAY LIBRARY
- EXPAND CAPABILITIES FOR COMPLETE PROVIDE NPA EXEClfTIVE SOFTWARE MAINTENANCE ON MDDELLING 0F CONTROL SYSTEMS AND THE IEL AfD LAft COMPLITING SYSTEMS SEC0f0ARY SIDE PROVIDE ON-CALL USER ASSISTANCE PROVIDE DOClPENTATION FOR PHASE
- II SOFTWARE UPDATE AND REISSLE USER'S MANUAL DEVELOP FIVE C0ARSELY f0DALIZED DECKS FPfM FY 88 IS ENDP0lf(f 0F DEVELOPENT OF NPDB EXISTING HIGLY NODALIZED DECKS FOR PWRS AfD It(IEGRATION WITH PWR/NPA'S FY 88 IS ENDPOINT'0F PWR SOFTWARE DEVELOPE NT m
A i .
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- a TRAC-PF1/?0D3 TO INCLLDE MODEL Iff9(A001TS NOT CURRENTLY FLAPNED FOR FY 1987 RELEASE ,
- CllRIBIT MAJOR DEFICIENCIES IDENTIFIED THROUGli NRR, ICAP, 2D/Sd, ROSA IV, Af0 MIST
- TO INCLt0E MODELS FOR Iff90VED PREDICTION OF BgW SPECIFIC REN0KNA SUOf AS
- OTSG IFAT TRANSFER l
- Al!XILIARY FEEDWATER BEllAVIOR
/
[ '
- 2. U. S. EFFORT (00flTIMED)
B. TRAC Af!ALYSES AND PDfEL IFPPIADEfffS FISCAL YEAR: 1986 1987 1988 1989 I. TRAC ANALYSES (fD OF Af!ALYSES) TOTAL CCTF 2 0 0 0 2 SCTF 7 3 1 0 11 UPTF 2 8 8 2 N) na 1 ? 2 0 5 TOTAL 12 13 11 2 38 II. DATA ANALYSES Aff) RILS CCTF 3
IPI -
A SCTF-I a CCTF-II A SCTF-II a SCTF-III 3 UPTF A III. TRAC MOOFL IFPROVDENTS TIE ft. ATE CCFL. STEN MIATER SEPARATOR, filLTI-S0llRCE CONNECTION CAPABILITY A CORE VolD DISTRIBUTION A v
D .
(
TABLE I. LSTF TEST MATRIX IN FY 1986 IDENTIFICATION OBJECTIVE DESCRIPTION DATE SB-CL-08 BREAK AREA EFE CT 5% COLD LEG SIDE-BREAK WIllDJT HPI 6/ /86 SB-IL-01 BREAK LOCATION EF E CT 5% 10T LEG SIDE-BREAK 7/ /86 TR-LF-01 OKRATIONAL TRANSIENT C0ffLETE LOSS OF FEEDWATER RECOVERY WITil PORV LATO1ED OPEN ANDliPI 8/ /86 TR-LF-02 OERATIONAL TRANSIENT STATION BLACKDLIT 9/ /86 SB-SG-01 BRFliK LOCATION EFFECT FIVE SG/U-TUBES RUPTURE 9/ /86 SP-PV-01 BREAK LOCATION EFFECT PV LOWER PLENLN BREAK 10/ /86 05/19/86 7 STATUS OF ROSA IV PROGRAM
TAPLE I. LSTF TEST MATRIX IN FY 198G (CON'T)
OBJECTIVE DESCRIPTION DATE IDEffilFICATION SB-CL-09 BREAK AREA EFITCT 10% COLD LEG SIDE-BREAK W/0llPI 11/ /86 SB-CL-10 BREAK AREA EFFECT 2.5 COLD LEG SID-BREAK i W/0llPI 11/ /36 I SB-CL-11 BREAK AREA EFRCT 0.1 COLD LEG SIDE-BREAK W/0llPI 12/ /86 SB-SG-02 BREAK LOCATION EFFECT ONE SG/U-TIFE RUPTURE 1/ /87 SB-SG-03 BREAK LOCATION EFFECT FIVE SG/U-TUBE RUPTURE AFTER MAIN STEAM LINE BREAK 2/ /87 SB-IL-02 BREAK LOCATION & AREA EFRCT 10% IDT LEG SID-BPEAK 3/ /87 SB-IL-CB ' BREAK LOCATION 8 AREA EFFECT 2.5% 10T LEG SIDE-BREAK 3/ /87 SB-llL-Oli BREAK LOCATION 8 APEA EFFECT 0.1% ICT LEG SIDE-BREAK 11 / /87 k 05/19/86 g STATUS OF ROSA IV PROGRAM
/ ,
i STATllS OF TRAC ANALYSIS AT LANL AND IfEL N
GEffEIRY INRIT FOR TIE LSTF TRAC DECK COPPLETE INDEPENDENT REVIEW OF IffUT COMPLETE
~
V0 lites OF l%JOR C0f00NENTS OECIID AGAINST DATA STEADY STATE RUN TO 1000s LOOP FLN IS CORECTLY PREDICTED If0lCATING TilAT TIE SYSTEM LOSSES AE CORRECT 10T LEG TEPPERATUE IS HIGER TilAN KASURED (1 K PROBABLY DUE TO IEAT LOSSES W11101 AE NOT ACCOUNTED FOR IN TIE TRAC ff) DEL)
TEED FURTIER DATA TO BEGIN FOST-TEST CALCULATION
- 1. LEAVAGE FLN DATA FROM SKCIAL TEST I 2. , EATER ROD IEATUP TO VERIFY PR0KRTIES (MG0 CONDlETIVITY)
- 3. ECC INJECTION FLOWS Y
14 N 05/19/86 'l STATUS OF ROSA IV PROGRAM
CITADEL RESULTS TABLE 4. REACTOR DESIGNS AND STEAM GENERATOR TUBE ICDINE* ,
a -
RUPTURE CASES CONSIDERED IN THIS STUDY R '
U
- 1. Westinghouse 4-loop pressurized water reactor (PWR) a) Steam generator tube rupture with loss of offsite power 0.3 b) Steam generator tube rupture with a stuck open safety valve for 7.0 accident duration c) Steam generator tube rupture with only a loss of condenser 0.2
- 2. Combustion Engineering 2-loop PWR (2700 Class) a) Steam generator tube rupture with loss of offsite power 0.08 b) Steam generator tube rupture with only a loss of condenser 0.05 3.
Babcock & Wilcox 2-loop PWR' ' /b '/ t a) Steam generator tube rupture with loss of offsite power 38 b) Steam generator tube rupture with a stuck open safety valve for 47 accident duration c) Steam generator tube rupture with a PORV stuck open for 20 minutes 27 after the PORV opens initially
- 8. Combustion Engineering 2-loop PWR (3800 Class) a) Steam generator tube rupture with loss of offsite power 8.0 b) Steam genera *nr tube rupture with a stuck open atmospheric dump 27.0 valge (ADV) for 20 minutes then isolated c) Steam generator tube rupture with a stuck open ADV for the accident 32.0 duration DF BR CURIE DOSE - [ 1.49 x 10 6
) ( 2 x(X/Q)'
10' ) ( 3.47 x 10~4 ) 1 - O.I man-rem
- BASED ON CONCENTRATION OF i MICROCURIE /GR. IN PRIMARY WATER.
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, TASKS:
1.
PEW 0fM CALOLATION(S) TO OBTAIN VELOCITIES ENC 0lWTERFT) IN A BOILING WATER EACTOR j (BWP) LOER PLEMM DlRING ANTICIPATED TRANSIENTS WITil0llT SCRAM (AINS) 2.
IITILIZE TIE SCALING ET110DS LEARED FORM TIE PTS TERMAL MIXING S110Y TO DEFIE EXPERIENT(S) 141101 WILL YlELD DATA TilAT IS APPLICABLE TO FILL SCALE EACTOR
- 3. PASED ON TASK 2, PEWOFM TIE ECESSARY MJDIFICATIfW TO TlE EXISTING 1/2 SCALE l FACILITY IF IT IS FEASIBLE. MEIFICATIONS SHALL BE E0 TESTED IN WRITING AND MET BE APPROVED BY TIE NRC PROJECT OFFIER PRIOR TO COPfENCEENT OF WORK.
l f.t EWOPM EESSARY EXPERIENTS APO COLLECTION OF DATA 1
l 5.
ANALYZE DATA AM) CONSTf4LT B0fDi TRANSf0RT MEEL FOR ADAPTATION IN TIE TRAC-BFI CODE l
c l
6.
l INTERACT WITil IDNO NATIONAL ENGIEERING LABORATORY STAFF ON TRAC-F1 DEVELOFENT l SD TASK 5 C0lLD BE SUCttsSFtLLY ACC0PFLISED i
(
4
_ . _ _ _ . _ _ . . . . . _-. ~_ _ _ _ . - . . _ _ _ -. _ . . . _ . _ _ _ _ . _ - . _ - _ , _ _ _ _ . _ . _ . _ . . _ . . _ . _ _ _ - _ . ~ ._. _
- N .
INDIAN FOINT SGMI
- 1. PROESSES Ato OY0ITIONS LEADit0 TO Mi ilAVE MEN DETERMINED OECK LIST AVAILAP1E TO DETEINitE A PRIORI MIEN M1 CAN OCOJR EXPERIKHTAL DATA Ato TIEORY ARE AVAILABLE TO DETERiltE MiAT GIANGES IN SUBC00 LING LIQUID FLO3 SYSTEM GE& ETRY APE TEEDED TO ELIMINATE PROBLEN e
I l
n -- _ __ .-_
f?]iL.L S T o h { EVENT .
llil PROCESS DIFFERENT FROM SCHI NO C0lNIER CURRENT FLfW INTERFAE NOT ECESSARILY HORIZONTAL BUBBLE ALRFADY TRAPPED IN PIPIT!G SYSTEM CONDENSATION OCCURS ONLY PECAUSE OF REPRESSURIZATION DIE TO PlW START-tP PROCESS VARIABLES
- DRIVING PRESSUE BUBBLE W1.lPE NON-CONDENSIf1E GAS CONCENTRATION SYSTEM E0KTRY LIQUID TEITERATUE Y >
PLATTED PROGRAM OIUECTIVES !#/4AU##8 COLLECT PLANT DATA CONSTRUCT SCALED FACILITY OBTAIN TEST DATA DEVELOP STABILITY MAP IN TEINS OF PROCESS VARIABLES USE P00EL TO DEVFLOP STRATEGIES DEVELOP OECK LIST TO ELIMINATE POSSIBILITY OF Mi TO OCClR k.
b -__ _ _ _ _ _ _ _
N '
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OIARACTERISTICS:
BENIGN START ,.
SEQUENE OF EVENTS AND CONSEQUENES NOT F0ESEEN MLTIPLE FAILURE r APPRDA01 .
ASSURE GECK VALVE RELIABILITY DETEft1INE CONDITIONS /ACTInflS MlI01 MAY LEAD TO MI FIND A WAY TO ELIMINATE TIESE CONDITIONS Af0/0R FOR AN ORRATOP TO INTERVENE PLANT SKCIFIC/0KRATOR TRAINING
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II. STATUS OF PTS T ERMAL l'YDRAltlC RESEAROI 1
f
- FORTY-FIE OVERC00 LING TPANSIENTS HAVE EEN CALCILATED FOR OCOEE (B a W),
CALVERT CLIFFS (E) API) 11. B. ROBINSON W) (GING ELAPS AND TRAC.
AFF90XIMATELY 300 ADDITIfrAL OVERC00 LING TRANSIFNTS IIAVE EEN ESTIMATFD ISING TIE EXISTlHG TRAC AMI ELAPS CALOLATIONS AND EITER A SIPPLIFIED ELAP CODE OR A MASS APO EERGY BALANCE.
I .
ALL EPORTS ON TlDMAL HYDPAltIC APALYSES HAVE BEEN PIB.ISED.
FINAL 4LE PlH.ISID ON JtLY 23,1985.
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- DRAFT EGLATORY Gile ISSlO FOR COPMN ON JAftJARY 17, 1986.
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- III. STATilS OF TIEmAL FilllD MIXIfE RFSEARDI RBIX CODE WAS lEED WITil TRAC OR RELAP T0. PREDICT iENCOMER FLUID Cff0ITI0fG DIRIfE LOOP staff!ATICfl SITUATI0f6 FOR OCOTE;. CALVEitT CLIFFS AfD 11. B. r0BIfEON EACTORS.
S, TIFPM,'i FLtilD MIXIf6 EXPERIFUITS IIAVE PID' KRFDPKD TO ASSESS TIE PEMIX CODE.
. EffTEST PRE 0!CT10f6 ilAVE IIAD EXCELLEllT ACMEWilT WITil DATA.
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CREAPE 1/2 SCALE 011R1 PESS!!RE)
FUPRE 1/2 SCALE (TRAfSFAEffT) l\0 7/5 SCALE 011TILOOP)
IDR FlLL llEIGIT IPTF FILL SCALE UCSB WILL PffPARE AN ASSESSPUIT EPORT IN JtLY 19Pf3 l gD n
_