ML14192A319
| ML14192A319 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 01/21/1982 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | Jackie Jones Carolina Power & Light Co |
| References | |
| IEB-80-04, IEB-80-4, NUDOCS 8202160196 | |
| Download: ML14192A319 (8) | |
Text
DISTRIBUTION JAN 2 1 1982 Docket NRC PDR L PDR TERA NSIC Docket No. 50-261
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ORB#1 Rdg
~EC DEi senhut 1 198 0OELD Mr. J. A. Jones RS-10 Senior Vice President WRoss Carolina Power and Light Company CParrish 336 Fayetteville Street Gray File Raleigh, North Carolina 27602
Dear Mr. Jones:
By letter dated May 9, 1980 you responded,to the concerns stated in IE Bulletin 80-04 "Analysis of a PWR Main Steam Line Break with Continued Feedwater Addition."
Your responses are being reviewed by the NRC's contractor at Franklin Research Center (FRC).
Enclosed is a request from FRC for additional information.that is needed to permit completion of their review. I request that this information be provided to the NRC staff within 30 days of receipt of this letter.
Sincerely, Originalsignedby:
S. A. Varga Steven A. Varga, Chief Operating Reactors Branch #1 Division of Licensing
Enclosure:
As stated cc w/enclosure:
See next page 82021609 82012i
-PDR ADOCK 05000261 PDR ORB#1:
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OFFICEL............
SURNAME
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82 DATE1.
NRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY USGPO: 1981-335-960
Mr. J. A. Jones Carolina Power and Light Compary cc: G. F. Trowbridge, Esquire Shaw, Pittman, Potts and Trowbridge 1800 M Street, N.W.
Washington, D. C. 20036 Hartsville Memorial Library Home and Fifth Avenues Hartsville, South Carolina 29550 U. S. Nuclear Regulatory Commission Resident Inspector's Office H. B. Robinson Steam Electric Plant Route 5, Box 266-1A Hartsville, South Carolina 29550 Michael C. Farrar, Chairman Atomic Safety and Licensing Appeal Board Panel U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Richard S. Salzman Atomic Safety and Licensing Appeal Board Panel U.'S. Nuclear Regulatory Commission Washington, D. C. 20555 Dr. W. Reed Johnson Atomic Safety and Licensing Appeal Board Panel U. S. Nuclear Regulatory Commission Washington, D. C. 20555
REQUEST FOR ADDITIONAL INFORMATION PWR MAIN STEAM. LINE BREAK WITH CONTINUED FEEDWATER ADDITION CAROLINA POWER AND LIGHT COMPANY H. B. ROBINSON STEAM ELECTRIC PLANT UNIT 2 NRC DOCKET NO. 50-261 NRCTACNO. 46856 1
? rnklin Research Center A Division of The Franklin Institute The Benjamin Franidin Parkway, Phila.. Pa. 19103 (215) 448-1000
BACKGROUND Evaluation of the information contained in the May 9, 1980 letter [1]
from Carolina Power and Light Company (CP&L) to the U.S. Nuclear Regulatory Commission (NRC) relating to IE Bulletin 80-04, "Analysis of a PWR Main Steam Line Break with Continued Feedwater Addition," revealed several items of concern. Additional information relating to these concerns is needed before a final evaluation can be made regarding the potential for exceeding containment design pressure or worsening of -reactor return-to-power response.
The concerns and the additional information needed to resolve the concerns are identified in this Request for Additional Information.
ITEM 1 CONCERN IE Bulletin 80-04 directs the Licensee to review containment pressure response to a main steam line break (MSLB) accident to determine the impact of runout flow from the auxiliary feedwater (AFW) system and other energy sources. CP&L's response concerning the MSLB analysis for the H. B. Robinson Steam Electric Plant Unit 2 indicated that continued feedwater addition was at the design AFW pump flow rates, main feedwater (MFW) flow was assumed to be isolated 10 seconds after the start of the accident, and manual isolation of the AEW system was assumed to occur by operator action 10 minutes after the initiation of the accident.
CP&Ls response is not sufficient to allow FRC to complete the evaluation of the potential for exceeding containment design pressure. The AFW flow assumed for the analysis is the design flow rate at design head; it does not assume a significantly lower head, which would occur during a MSLB. It is not apparent that the analysis considered the effects of a.single active.failure of the MFW system. The analysis also takes credit for operator action to identify the affected steam generator and isolate AFW flow to that generator within 10 minutes of the start of the accident. In the light of studies performed on operator response to stressful situations, this time may be unrealistic..
TJJFranklin Research Center A Divisn at The Franidin Insdtute
REQUEST Please provide the following information concerning your analysis of containment pressure response to a MSLB with continued feedwater addition:
- 1. A determination of runout AFW flow to the affected steam generator.
This should be determined from the manufacturer's pump curves at zero backpressure, unless the system contains reliable anti-runout provisions or an actual backpressure value has been conservatively calculated.
- 2. An evaluation of the potential for a single active failure in the EMW system which could cause the greatest feedwater flow to the affected steam generator during a MSLB accident and a determination of EMW flow rate to the affected generator if a single active-failure were to occur.
- 3. An evaluation of the potential for exceeding containment design pressure using the feedwater runout flow rates identified in Item 1, Requests 1 and 2.
- 4. The time after the start of a MSLB that containment design pressure will be exceeded if no operator action is taken to terminate the accident. If the containment design pressure will be exceeded in less than 30 minutes, provide the magnitude of the peak pressure and the time at which the peak occurs.
- 5. If operator action is required to terminate the accident, provide justification for the time at which credit is taken for operator action. The criteria given in draft ANSI N660, "Time Response Design Criteria for Safety-Related Operator Actions," March 1981, should be addressed and any difference between the time taken for operator action and the ANSI N660 criteria should be justified.
- 6. A verification that the steam line and feedwater isolation systems meet the requirements of IEEE Std 279-1971.
- 7. A verification that the MFW isolation valves are Seismic Category I and safety grade.
- 8. A schedule reflecting the proposed completion of corrective actions and appropriate actions implemented for the interim period until corrective actions are completed if any corrective actions are required based on the above evaluations.
-2 TU Franklin Research Center A Division of The Franklin Inshad
ITEM 2 CONCERN IE Bulletin 80-04 directs the Licensee to review the reactivity increase which results from a MSLB inside or outside containment.
The Licensee stated that the worst case MSLB was assumed to occur at hot, zero power condition, outside containment, with offsite power available.
The most reactive control rod was assumed to be stuck out.
The assumptions did not state whether a single active failure to the safety injection system which could delay the injection of boron to the reactor coolant system (RCS) was considered or if the time in core life was chosen to maximize the negative moderator temperature coefficient.
REQUEST Please provide the following information concerning your analysis of reactivity response which results from a MSLB with continued feedwater addition:
- 1. If the assumptions used in the core reactivity response analysis are bounded by those found in Section 15.1.5 of the Standard Review Plan (SRP) provide:
- a. An evaluation of the most restrictive single active failure in the safety injection system and its effect of delaying injection of boron to the RCS.
- b. An evaluation of the time in core life which produces the most limiting moderator temperature coefficient for the MSLB accident.
Note: A statement that the assumptions of SRP 15.1.5 are not considered part of the licensing basis will not be.considered responsive to this request.
- 2. An analysis of the core reactivity response to a MSLB considering the effects of Item 1, Requests 1 and 2; Item 2, Request 1; and assuming no operator actior6 (.2c-o Jo m'na/es
- 3. If operator action is required to mitigate the core reactivity response, provide:
-3 UDranidin Research Center A Dvsion of The Franhdin InsAne
- a. The time at which peak reactivity and minimum departure from nucleate boiling ratio (DNBR) are obtained assuming no operator action. f'rto o
r
- b.
Justification for the time at which credit is taken for operator action. The criteria given in draft ANSI N660, "Time Response Design Criteria for Safety-Related Operator Actions," March 1981, should be addressed and any differences between the time taken for operator action and the ANSI N660 criteria should be justified.
- 4. If, as a result of the above analysis, it is determined that the reactor return-to-power worsens so that a DNBR less than 1.30 can occur, the Licensee should provide the following:
- a. The number of fuel rods predicted to fail, assuming all rods with a DNBR less than 1.30 fail.
- b. Confirmation that the core will remain in place and intact with no loss of core cooling capability.
- c. Confirmation that calculated radiological consequences are below the 10CFR100 guideline values.
- d. Confirmation that the integrity of the reactor coolant pumps will be maintained so that loss of ac power and containment isolation will not result in pump seal damage.
- e. Confirmation that the auxiliary feedwater system is safety grade and, when required, automatically initiated.
- f. Confirmation that tripping of the reactor coolant pumps will be in accord with the resolution of Task Action Plan item II.-K.3.5.
- g. If confirmations b through f above cannot be given, the Licensee should also provide:
(1) Proposed corrective actions to preclude a DNBR less than 1.30 and a schedule for completion.
(2) Interim actions to be taken until the proposed corrective action is completed.
Note:
Lower values for minimum DNBR may be acceptable if justified for certain fuel designs and DNBR correlations.
-4 TOE tranldin Research Center A DiMsknof The Frakn Insulute
REFERENCE
- 1.
L. W. Eury (CP&L)
Letter to J.
P. O'Reilly (NRC)
May 9, 1980
-5 IuFranklin Research Center A DMsion of The Franidin Instiute