NUREG/CR-3935, Forwards Util Comments on Facility Pressurized Thermal Shock Study.Requests That Document Be Placed in PDR W/Existing File on Unresolved Safety Issue A-49.NUREG/CR-3935 Encl
| ML14176A216 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 03/29/1985 |
| From: | Shotkin L NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Bailey I NRC OFFICE OF ADMINISTRATION (ADM) |
| Shared Package | |
| ML14176A217 | List: |
| References | |
| REF-GTECI-A-49, REF-GTECI-RV, RTR-NUREG-CR-3935, TASK-A-49, TASK-OR NUDOCS 8504100499 | |
| Download: ML14176A216 (22) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 MAR 2 9 1985 MEMORANDUM FOR:
Inez K. Bailey, Chief Record Service Branch Division of Technical Information and Document Control FROM:
Louis M. Shotkin, Chief Reactor Systems Research Branch Division of Accident Evaluation
SUBJECT:
UTILITY COMMENTS ON THE H. B. ROBINSON PTS STUDY FOR SUBMISSION TO THE PUBLIC DOCUMENT ROOM) ;
/4 Enclosed is a copy of Carolina Power and Light's review of the H. B. Robinson pressurized thermal shock study. We request that these documents be submitted to the Public Document Room to be incorporated with the existing file on Unresolved Safety Issue A-49 Pressurized Thermal Shock.
If you have any questions, please contact Jose' N. Reyes on 427-4422.
Louis M. Shotkin, Chief Reactor Systems Research Branch Division of Accident Evaluation
Enclosure:
Ltr fm Phillips to Shelby dtd 10/12/84 8504100499 850329 PDR ADOCK 05000261 P
Carolina Power & Light Company October 12, 1984 A2-TS-115 CNS-84-265 Mr. D. L. Selby Oak Ridge National Laboratories Post Office Box X Building 6025 Oak Ridge, Tennessee 37830
Dear Mr. Selby:
The purpose of this letter is to document our review of the August, 1984
- report, "Thermal-Hydraulic Analyses of Overcooling Sequences for the H. B. Robinson Unit 2 Pressurized Thermal Shock Study,"
NUREG/CR-3935, EGG-2335, Draft Report.
The report appears very satisfactory.
We do however, have the following comments.
The basis for various analysis assumptions critical to the T&H analysis should be addressed and documented.
Some of these are the reason for the selection 2.5" Hot Leg Break, and why 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> shutdown was assumed for hot standby.
These were specified by ORNL, are reasonable, but need to be documented in this report or elsewhere in the final report.
Similarly, event definitions and how they were arrived at should be documented somewhere in the final report.
Attached also are comments from our consultant, Dr. D. M. Speyer.
Concerning the remainder of the project, our understanding is that CP&L will continue to comment on the draft chapters as they are transmitted.
After review of the individual chapters when the draft report is complete.
The analysis details and results will be presented separately to the NRC's Research Review Group and CP&L.
After the presentation CP&L will provide comments on the final report.
Please call me if you wish to discuss these comments.
Sincerely, Jerry H. Phillips JHP/laD1 Attachment cc:
Mr. J. G. Hammond Dr. J. D. E. Jeffries Mr. C. E. Johnson (NRC)
Mr. D. S. Lucas Mr. D. M. Ogden (INEL)
Mr. R. E. Oliver 0
F e
Box 1551
- Raieigh. N C 27602
DMiS-84-212 page 1 of j COMMENTS ON
SUMMARY
SECTION pv. lot Paragraph. Clarify that assessments with regard to severity in the present report are somewhat qualitative -
e.g.,
add sentence as follows's control actions, and uncertainties. As part of this report Jimited aqsequment have been made wiT reRud to trmanglent qqverity and thege-are (1) qualitative and 2) based solely on thermal-hydraulic presaure and downcomer tmiperature. Computer sim ulations D.v.
2nd Paragraph. Clarify that the severity of sequences refers to thermal hydraulic severity as apposed, for example, probability or consequences.- e.g.,
indicate that the most severe thermal-hydraulic sequences for PTS
- In this and subsequent write up changes to existing draft report are underlined.
DYS-84-212 page 2 of 10 COMMENTS ON 1. INTRODUCTION SECTION p.1 2nd Paragraph, The present study deals with a specific plant, accounts for minimal operator corrective actions and utilizes tran sients that were based on a more extensive basis than may be in ferred from "postulated."
The purpose of the thermal-hydraulic analyses presented in this report is tocquantify the behavior (downcomer pressure, temperature and to a lesser extent heat transfer coefficient versus time) for a specific plant during various kinds of postulated severe overcooling transients, with multiple failures of equipment and with only minimal operator corrective actiong. The transients were specified by Oak Ridge National Labor atory (ORNL) based on Probablistic and other related studies carried out as part of the program.- OR will calculate both the reactor vessel...
p.2, 2nd Paragraph.
Current status as of the time period the analysis was completed is probablyfbre appropriate - e.g., as
- followst, The reactor is of Westinghouse three-loop design and is currently undergoing steam generator replacement.
Prior to the SG replacement outage the unit was operated at reduced power, due in part ft steam generator plug ging.
Other anticipated (during the SG replacement outage) changes were also incorporated into the anilysis in order to more realistically model the plant.
DMS-84-212 page 3 of 10 COMMENTS ON 2. SEQUENCE DEFINITIONS SECTION p.3, 1st Paragraph.
It would be beneficial to clarify "highest probability of vessel failure" and to identify that Tables 13 and 14 provide clarification/description of the entries-found earlier in Tables 1 thru 12 - e.g.,
Oak Ridge...
those sequences expected to have the potential to rapidly cool the RCS, coincident or followed by moderate to high RCS pressure. and of event frequency greater than 10-7 per reactor year.
The sequences...
Groups 1 through 12., respectively (Tables 13 and 14 provide additional clarification/
description for the entries in Tables 1 through 12).
p.8 (Table 5).
Heading MSIV CLOSE not required and not used on other small SLB Tables. If left in, change to be cohsistent with Table 13 (or add entry to Table 13).
1.9 (Table 5).
Under AFW SG ISO the entry "occurs on demand" is included. Either change to occurs as required or clarify with ad ditional description in Table 13.
p.11 (Table 7).
Same as Table 5, p.8 above.
p.12 (Table 8).
Same as Table 5, p.9 above.
p.14 (Table 9),
Potential conf usion on MSIV CLOSE as per text.
the signal (steam flow and low T-AVG) is not generated for 5 SDVs failing to close - but per Table 14" probably not demanded unless at least 4.SDVs fail to close."
p.26 (Table 13)1 For MSIV closure (MSIV RPCL and RM MSIV CL) the phr'ase "not required for small break" is confusing - e.g., change tol heading not used for small break as MSIV closure gignal not generated.
Also the NA is used (e.g., Table 9) for cases where signal is not generated.
Add to description(s) - e.g., as follows:
Closure not required if check valve operates properly OR sinal not generated due to size of break,
DMS-84-212 page 4 of 10 AFW ACT Description states minimum AFW was used. This is not correct -
e.g., change as followas AFW aliways actuates on demand and delivers flow based on two motor driven aux. feedwater pumps.
and turbine driven AFW where 2 of 3 low SG level indications were reached, SDV CLOSE -
add basis -
e.g.,
were not considered based on probablistic considerations.
p.27 Table 13).
THR AFW and THR HPI CH Descriptions appears superfluouspotentially confusing. and incorrect -
i.e.,
the 10-7 is event frequency (not failure probability) and could imply a similar criterion was not used for other sequences.
D.28 (Table 13).,
same as p.26 (Table 13)
SDV CLOSE, above.
p.29 (Table 14).
FW REG VLY and SDV CLOSE Description(s) -
same as p.26 (Table 13) SDV CLOSE, above.
MSIV CLOSE Description -
see p.14 (Table 9),
above.
AFW ACT Description -
see p.26 (Table 13), above.
p.30 (Table 1W).
AFW Heading -
Tables used heading AFW AUTO.
THR AFW -
see p.27 (Table 13), above.
P*31 (Table 14).
see p.27. (Table 13)., above.
DMS-84-212 page 5 of 10 COMMENTS ON 3. METHODS SECTION p.35 (Table 16),
Note C is correct but potentially confusing e.g., change as follows:
C - Since only on feedwater isolation valve fails open, the failure of one or two feedwater regulating valves are equivalent.
Note E says sequence 9-24 (and 9-191 have "NA" for SG ISO (i.e.,
This is not the case -
see Table 9, page 14. Correct Table 9 or Note E.
p.38, 3rd Paragraph. Previous INEL analyses investigated con densation effects - that were partially resolved in MOD2. This should be discussed to provide basis for comments on page 134.
o.40, 3rd Pargraph. Identify which ANS standard (1973 or 1979) and value of To used.
p858, let Parazraph.
Provide additional basis for the transit time t80 sec.) used.
p 58, 2nd Paragraph.
In view of the simple model it appears ad ditional discussion on the break fluid conditions (how obtained) may be needed.
,62 (Figures 6 and 7).
Taken out of the report the figure would mply a simplified and detailed model yield identical results for times less than 1390 sec. Include in legend, Simplified Model (t, 1390s)
P.6 3 (Figures 8 and 9).
As for p.6 2 (Figures 6 and 7) above except t >-75 S.
D.65 (Figures 10 and 11). As for p.62 (Figures 6 and 7), above, except tW200 s.
p.6 6 (Figures 12 and 13),
As for p.62 (Figures 6 and 7), above, except t 390S.
p.68 (Figures 14 and 15).and p.69 (Figure 16).
As for p.62 (Figures 6 and 7), above, except t,400 s.
p.69 (Figure 17) and p.71 (Figures 18 and 19),
As for p.62 Figures 6 and 7). above, except t7200s.
DMS-84-212 page 6 of 10 p.70, 2nd Paragrph.
It may be desirable to identify the reason the start of the simplified calculation and detailed calculation were not over"lays as regards RCS/Downcomer temperature, since the former case is started using a state point from the latter.
DMS-84-212 page 7 of 10 COMENTS ON 4.
GROUP A RESULTS
... SECTION p.74, 1st Paragraph, In view of the limited number of points used to generate the figures in Appendix A (as opposed to the results shown in Section 3.) and the straight line segment appearance of the figurges some explanation is appropriate - e.g.,
add footnote as followst To facilate referencing of data, plotted results
- showing pressure and temperature...
- The figures in Ampendix A were computer generated using a limited number of state points, thus the discontinuties in slope are not necessarily indicative of specific events in the scenarios.
p.75, last paragraph Some additional discussion of the hand calculations may be Ln order - e.g., add following paragraph at end of page 75.
It should be noted hand calculationq were carried out in a simple conservative fashion. For examle in the above case the heatup was based on specific heat constant. More importantly the repregaurization was based on the time to reach the PORY aetpoint and did not include the more detailed behavior with time (i.e.. initially a slower Dressure increase until the level increased significantly).
p.81. Last Paragraph. it is not clear by inspection that sequence 9-28 is more severe than 9 e.g., change as followes Sequence& 9-28 and 9-32, involving failure to throttle AFW to the USGs and failure to throttle charging, were the most severe sequences of the subgroup, p.87, Last Paragraph, As above, p.81 Last Paragraph e.g., as followsa Sequences 9-28 (failure to throttle AFW and charging) and 9-32(9-28 plue AFW overfeed) were the most severe sequences 412 K (282 0F),
16.35 MPa (2371 psia).
An important...
DMS-84-212 page 8 of 10 D.92 1st Paragraph.
Suggest restating that no operator action to close MSIYs was credited. e.g., change as follows.
Since the MSIVs are not demanded, and no credit was taken for operator action to close MSIVs, AFW is not isolated...
P.99, Last Paragraph. Suggest adding information on operator in it ated MSIV closure - e.g.,
change as followss the maximum subsequent pressure was 16.35 MPa (2371 psia).
In these and other steam line breaks operator action to close MSIs would gignificantly reduce the severity. As an example sequence 9-22 would have experienced a minimum temperature about 1300F highwr if MSIV closure by 10 minutes was assumed.
D.101 and 103 (Section 6.2) and p.102 (Table 27).
Previously 0RNL indicated they would review cases 7-9, 10 and 11 and may discard these - for the same reason(s) sequence 3-3 was discarded.
We recommend these be discarded on probablistic grounds -
as was done for sequence 3-3.
12.106. Last Paragraph.
It is not clear that PTS severity would be increased significantly vs. T severity and potential PTS be severity. Suggest changing to T/
severity, as followss For these sequences, thermal-hydraulic severity would be...
D.107, 1st Paragraph.
As above, p.10 6, Last Paragraph. Suggest change as aboves the thermal-hydraulic severity of these sequences would be increased significantly.
P.117 (Table 31).
Typo on detailed model calculation used.
Should be 1-1 (not 11-1).
p.122. Section 10.1.
It would appear appropriate to review/discuss the assumptions on MSIV closure for sequences 1-8 and 2-8. Per Table 1 and 2 it would be inferred that MSIV closure occurs which apparantly is not the case. In addition it may be that the coin cident LOCA provides sufficient cooling to satisfy both high steam flow and low TAG, particularly for the M BLOCA. This should not be significant for the study since the probability of a LOCA with out 5 stuck SDVs is of considerably higher frequency and is addressed by the LOCA results.
DMS-84-212 page 9' of 10 p,122, Last Paragraph.
Sequence 3-2 should be 3-3.
.29. 1st Paragraph.
The use of the phase "and may not be real" is redundant. It is covered by the uncertainty statement and is also consistent with the phenominologic arguments. Suggest delet ing the phrase.
p.132 (Table 35).
Sequence 12-1 and 12-2 identify sequence 2-1 as the detailed model calculation used. This is presumably correct but could be interpreted as a typo. (i.e., the entry as 12-1 would also be correct as the sequences are identical to 600 seconds).
Suggest clarify - such as a sentence earlier in the text.
P.134. 2nd Paragraph.
The complete condensation in SBIOCAs, as not reasonable, would benefit from some discussion on the differences in the code MODs and the investigation carried out by INEL relative to same.
See p. 38, 3rd paragraph, above.
2.17 Tale 36).
Sequence 19-5 minimum downcomer temperature given as 4060F -
should be about 370oF (see Figure A-350).
DMS-84-212 page _U of _U COMMENTS ON 14. HEAT TRANSFER COEF.
SECTION p.140,14 1 (Table 37).
It appears sequences 1-5 through 1-12 and 2-5 through 2-8 are better represented by scenario no. 4.
Comparing sequence 9-15 (equivalent to scenario no.4) to sequences 1-7 and 2-7, the temperature response is similar and (as discussed in the section 10.2) is not strongly influenced by the concurrent LOCA (excepting pressure).
The principal effect of interest is flow rates(s) and the steam line break at power is more representative.
COMMENTS ON 15. UNCERTAINTIES SECTION P.145 (Table 38),
Large Steam Line Break uncertainty at 200OF appears excessive -
as the calculation should be reasonably accurate at the lower temperatures where essentially.only a simple energy balance is involved, and + 250F would be more reasonable.
COMMENTS ON APPENDIX A SECTION Figures A-192.
A-326 (sequences 9-13 and 9-80 Figures are not labeled and scaling changed see Figure A-296 (Seg.
9-65).
Figures A-357, A-358 (seauence 11-4).
Figures are missing from report.
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P.O. BOX 1625, IDAHO FALLS, IDAHO 83415 August 13, 1984 Mr. F. L. Sims, Director Reactor Research and Technology Division Idaho Operations Office -
DOE Idaho Falls, ID 83401 TRANSMITTAL OF PRESSURIZED THERMAL SHOCK STUDY REPORT (A6047) - TRC-75-84 Ref.:
Thermal-Hydraulic Analyses of Overcooling Sequences for the H. B.
Robinson Unit 2 Pressurized Thermal Shock Study (Draft NUREG).
Dear Mr. Sims:
- The referenced report (see attachment) documents RELAP5 thermal-hydraulic analyses performed in support of the. U.S. Nuclear Regulatory Commission's (NRC) investigation into unresolved safety issue A-49, pressurized thermal shock (PTS) for the H. B. Robinson Unit 2 Pressurized Water Reactor. The report summarizes the analyses. of 183 PTS scenarios which were developed at Oak Ridge National Laboratory (ORNL), the integrator of the PTS study for the NRC.
The report also includes descriptions of the RELAP5 thermal hydraulic and control systems models which were developed specifically for these analyses and a discussion of a unique method developed to generate reactor vessel downcomer pressure and temperature responses.
The results shown in the referenced report represent part of the information required by ORNL for the assessment of the PTS issue. These results are not to be used directly as an indication of PTS severity for the scenarios in vestigated. Following additional analyses of multi-dimensional and fracture mechanics effects, ORNL will integrate all results and publish a report estimating the likelihood of reactor vessel failure and identifying important event sequences, operator and control actions, and uncertainties.
F. L. Sims August 13, 1984 TRC-75-84 Page 2 The referenced report is transmitted as part of the assistance to the NRC as provided in the PWR application task of FIN No. A6047 and satisfies our committment to the Nuclear Regulatory Commission to provide a draft report documenting the results of our extrapolations for the H. B. Robinson plant (NTPD Milestone 3-06).
Recipients in the distribution-list are requested to submit written comments on the Draft.NUREG by October 1, 1984. Comments may be submitted to:
Don Fletcher EG&G Idaho, Inc. (TSB)
P. 0. Box 1625 Idaho Falls, ID 83415 (208) 526-9859 FTS:
583-9859 Very truly yours, T. R. Charlton, Manager Reactor Simulation and Analysis Branch CDF:sb
Attachment:
As Stated cc:
C. Johnson, NRC J. Koenig, LANL R. Oliver, CP&L J. Phillips, CP&L (3)
J. Reyes, NRC (3)
P. Saha, BNL D. Selby, ORNL (3)
D. Speyer, CP&L Consultant T. Theophanous, Purdue E. Throm, NRC R. L. Turner, Westinghouse (2)
J. 0. Zane, EG&G Idaho (w/o Attach.)