ML20040E954
| ML20040E954 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 01/20/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Kay J YANKEE ATOMIC ELECTRIC CO. |
| References | |
| IEB-80-04, IEB-80-4, LSO5-82-01-052, LSO5-82-1-52, NUDOCS 8202080128 | |
| Download: ML20040E954 (8) | |
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O DISTRIBUTION Docket NRC PDR January 20, 1982 Local PDR ORB Reading NSIC Docket No. 50-29 i
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Mr. James A. Kay 2
Senior Engineer-Licensing S-SEPB s' p Yankee Atomic Electric Company g
1671 Worcester Road ies Framingham, Massachusetts 01701
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Dear Mr. % :
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SUBJECT:
PWR MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDITION -
i IC BULLETIN 80-04 We are continuing our review of your submittals regarding the above sub, ject, and we find that additional information is necessary to complete our safety evaluation.
Please provide the requested infomation within 45 days of receipt of this letter.
The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not required under P.L.96-511.
Sincerely, Original signed by l
l Dennis M. Crutchfield, Chief l
Operating Reactors Branch #5 Division of Licensing l
cc:
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. January 20, 1982 Mr. James A. Kay CC Mr. James E. Tribble, President Yankee Atomic Electric Company 25 Research Drive Westborough, Massachusetts 01581
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Greenfield Community College 1 College Drive Greenfield, Massachusetts 01301 Chairman Board of Selectmen Town of Rowe Rowe, Massachusetts 01367 o
Energy Facilities Siting Council 14th Floor One Ashburton Place Boston, Massachusetts 02108 U. S. Environmental Protection Agency Region I Office ATTN: Regional Radiation Representative JFK, Federal Building Boston, Massachusetts 02203 Resident Inspector Yankee Rowe Nuclear Power Station c/o'U.S. NRC Post Office Box 28 Monroe Bridge, Massachusetts 01350 Ronald C. Haynes, Regional Administrator Nuclear Regulatory Commission, Region I Office of Inspection and Enforcement 631 Park Avenue King of Prussia, Pennsylvania 19406 9
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REQUEST FOR ADDITIONAL INFORMATION PWR MAIN STEAM L!NE BREAK WITH
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CONTINUED FEEDWATER ADDITION YAL'KEE ATOMIC ELECTRIC COMPANY YANKEE R0WE ATOMIC POWER PLANT NFC DOCKET NO.50-029 i NRC TAC NO. 46868
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Franklin Research Center A Division of The Franklin Institute The Benjarrun FrankLn Park.ay, Ptula Pa. 19103 (215)448 1000
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BACKGROUND Evaluation of the infornation contained in the May 8,1980 letter [1]
f rom the Yankee Atomic Electric Company (YAEC) to the U.S." Nuclear Regulatory Commission (NRC) relating to IE Bulletin 80-04, " Analysis of a PW'R Main Steam Lit.e 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 e
CONCERN IE Bulletin 80-04 directs.the Licensee to review containment pressure response to a main steam line break (MSta) accident to determine the impact of runout flow from the auxiliary feedwater (AFW) system and other energy sources.
YAEC's response _ concerning the MSLB analysis for the Yankee Rowe Atomic Power Plant stated that in the event of a MSLB the condensate and main fee'dwater (MFW) pumps would automatically trip at power levels greater than 1,5 MWe.
The Licensee also. stated that even if feedwater flow is not terminated within 10 seconds, the containment pressure response analysis would remain valid because
" operating experience and OP-3000 actions (emergency-shutdown-from-power procedures)" would keep the steam release from exceeding the predicted value.
YAEC's response is not sufficient to allow FRC to complete the evalua-tion of the potential for exceeding containment design pressure.
It is not apparent that the analysis considered the effects of a single act.ive failure l
in the MFW or condensate system which would allow runout feedwater flow to the af fected steam generator for a period in excess of the predicted value.
A b) Franklin Research Center A h of The Frank 6n insotwee
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4 Isolation of feedwater to the affected steam generator is assumed to occur within 1 minute of the accident.
In light of studies performed on operator response to stressful situations, this time may be unrealistic.
. REQUEST Please provide the following~information concerning your analysis of containment pressure response to a MSLB with continued feedwater addition:
1.
An evalu'ation of the potential for a single active failure in the MFW.
or condensate systems which could cause the greatest feedwater flow to the af fected steam generator during a MSLB accident, and a determination, of the feedwater flow rate to the affected generator if -
a single active failure were to occur.
2.
An evaluation of the potential for exceeding containment design pressure,' using the feedwater runout flow rate identified in Request 1 above.
e 3.
The time af ter the start of a MSta 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, 4.
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.
5.
Verification that t.1e steam line and feedwater isolation systems meet the requirements of IEEE Std 279-1971.
6.
A schedule for proposed corrective actions, if any corrective actions are required due to the above evaluations, and for appropriate actions to be taken during-the interim period until corrective actions are completed.
ITEH 2 CONCERN i
IE Bulletin 80-04 directs the Licensee to review the reactivity increase which results from a MSLB inside or outside containment.
4.
b000 Franklin Research Center A Dres on of The fraima insaw
t In regard to the core reactivity response to a MSLB at Yankee Rowe, the Licensee stated:
"A detailed reactivity calculation for the " current cycle, Cycle XIV, under the most conservative reactivity conditions was performed. This analysis essentially shows that for both the full
- power and zero power cases, the reactivity level in the reactor remains subcritical for RCS cooldown to 70*F.
This calculation included modetator temperature defect, fuel temperature Doppler defect, the most reactive control rod stuck out, boron insertion with safety injection, and conservative uncertainties on all the physics data.
"In addition to this calculation, it was determined that even without boron, the reactor would remain subcritical down to 70*F when the uncertainties w'ere not included. With reactivity uncertsinties, boron was.
required.
However, the amount of boron required to maintain a s'uberitical condition was quite small.
In fact, the time required to inject enough boron with the safety injection system to maintain r;uberiticality is substantially shorter than the time required to cool the reactor coolant system to 70*F, thereby precluding the return-to-power."
e The Licensee's statement in regard to the time required to inject enough boron to maintain suberiticality could be invalid since the reactor may return to power significantly before its temperature reaches 70*F.
In addition, the assumptions do not include the effect of a single active failure which would
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delay the injection of boron into the reactor coolant system.
REQUEST Please provide the following information concerning your analysis of the reactivity response which results from a MSLB with continued feedwater addition:
A determinat, ion ok the most restrictive single active failure of the 1.
safety injection system which causes the losgest delay in the delivery of high concentration boric acid solution to the reactor coolant system.'
2.
A quantitative analysis of the core reactivity response to a MSLB which. incorporates the results of Item 1, Request 1 and Item 2, Request 1.
The analysis should initially be performed. assuming no operator action for 20 minutes.
3.
If operator action is required to mitigate the core reactivity
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response, provide:
nklin Research Center A Dresson of The Franen m
t a.
The time at which peak reactivity and minimum departure from nucleate boiling ratio (DNBR) are obtained assuming no operator action.
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, ass'uming' all rods with
- a DNBR less than 1. 30 f ail.
b.
Confirmation that the core will remain in place and intact' with no loss of core cooling capability.
Confirmation that calculated radiological consequences 'are 'below c.
the 10CFR100 guidelines values.
d.
Confirmation that the integrity of the reactor coolant pumps will be maintained so that loss of ac power and containment isolation
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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 ths resolution of Task Action -Plan item II.K.3.5.
g.
If confirmations b through f above cannct be given, the Licepsee 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.
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nklin Research Center A Dunnon of The Freen hannae
REFERENCES 1.
D. E. Moody (YAEC)
Letter to B.
H. Grier (NRC)
May 8, 1980
. 00h Franklin Research Center A Come.on et The Fem lascue
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