ML20027D539
| ML20027D539 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 10/13/1982 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20027D538 | List: |
| References | |
| IEB-80-04, IEB-80-4, NUDOCS 8211050155 | |
| Download: ML20027D539 (7) | |
Text
--
2-_.--
SAFETY EVALUATION REPORT BY THE OFFICF 0F NUCLEAR REACTOR REGULATION CONCERNING, MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER OPERATION ARKANSAS NUCLEAR ONE, UNIT NO. 1 DOCKET NO. 50-313
1.0 INTRODUCTION
In the summer of 1979, a pressurized water reactor (PWR ) Licensee submitted a report to the NRC that identified a deficiency in its original analysis of containment pressurization resulting from a postulated main steam Line break (MSLB).
A reanalysis of the co'ntainment pressure response following a MSLB was pe rf ormed, and it was determined that, if the auxiliary feedwater (AFW) system continued to suppt,' f eedwater at runout conditions to the steam generator that had experienced the steam Line break, the containment design pressure would be exceeded in approximately 10 minutes.
In j
other words, the Long-term blowdown of the water supplied by the AFW system had not been considered in the earlier analysis. -
On Octobe r 1,1979, t he f o regoi ng inf ormation was provided to all holders of operating Li c*e ris e s a nd construction pe rmits in IE
^
Information Notice,79-24 [23.
Another Licensee performed an oc cide nt analysis r'eview pursuant to the inf ormation 'f urnished in the above cited not; ice and discovered that, with of f site elect rical power availabLe, the condensate pumps would f eed the af fected steam I
generator at an exc'essive rate.
This excessive feed had not been considered in the analysis of the postulated MSLB a c ci de nt.
"' ~
8211050155 821013 PDR ADOCK 05000313-L 0
~
A third Licensee informed the NRC of an error in the MSLB analysis for their plant.
For a zero or low power condition at the end of core Life, the Licensee identified an incorrect postu$ation that the startup feedwater control valves would remain positioned "as is" during the transient.
In reality, the startup feedwater control valves wilL ramp to 80% full open due to an override signal resulting froa the Low steam generator pressure reactor trip signal.
Reanalysis of the events showed that the rate of feedwat,er addition to the affected steam gene-rator associated with the opening of the startup valve would cause a rapid reactor cooldown and reruttant resctor return-to power response, a condition which is beyond the plant's design basis.
Following the identification of these deficiencies in the original MSLB a c ci de nt analysis, the NRC issued IE ButLetin 80-04 on Feb rua ry 8,1980 This butLetin required alL Licensees of PWRs and near-t ern PWR ope rating License applicants to do the following:
1.
Review the containment pressure response analysis to determine if the potential for containment ove rp r es sur e in 'the event of a MSLB inside containment included the impact of runout flow from the auxiliary feedwater system and the impact of other energy sources such as continuation r
of feedwater or condensate flow.
In your, review, consider the ability to detect and isolate the damaged steam generator f rom these sources and the ability of the pumps to remain operable after extended operation at runout flow.
..-__u.-
-.... ~.
2.
Review your analysis of the reactivity increase which results from a MSLB inside or outside containment.
This review should consider the reactor cooldown rate and the t-potential for the reactor to return to power with the most reactive control rod in the fully withdrawn position.
If your previous analysis did not consider atL potential water sources (such as those Listed in 1 above) and if the reactivity increase is greater than previous analysis indicated, the report of this review should include:
a.
The boundary conditions for the analysis, e.g'.,
the end of Life shutdown margin, the moderator temperature coefficient, power Levet and the net effect of the associated steam generator water inventory on the reactor system cooling, etc; b.
The most restrictive single active failure in the safety injection system and the effect of that failure on delaying the delivery of high concentration boric acid solution to the reactor coolant system; c.
The effect of extended water supply to the affected steam generator on the core criticality and return to power; and d.
The hot chadndl factors corresponding to the most reactive rod in the fully withdrawn positions at the end of Life, and the Minimum Departure from Nucleate Boiling Ratio (MDNDR) values for the analyzed transient..
A
3.
If the potential for containment overpressurization exists or the reactor return-to power response worsens, provide a proposed corrective action and a schedule for completion of the corrective action.
If the unit is operating, provide P
a description of any interim action that wilL be taken until the proposed corrective. action is completed."
Following the Licensee's initial response to IE ButLetin 80-04, a
request for additional information was developed to obtain aLL the information necessary to evaluate the Licensee's analysis.
The results of our evaluation for Arkansas Nuclear One Nuclear Plant, Unit 1 (Arkansas I, 1) are provided below.
2.0 Evaluation Our consultant, the Franklin Research Center (FRC), has reviewed the submittals made by the licensee in response to IE ButLetin 80-04, and prepared the attached Technical Evaluation Report.
We have reviewed this evaluation and concur in its bases and findings.
3.0 conclusion Based on our review of the enclosed Technical Evaluation Report, the fotLowing conclusions are made regarding the postulated MSLB with continued f eedwater addition for Arkansas I,1:
1.
There is no potential for containment overpressurization f
resulting from a MSLS with continued feedwater addition because the main feedwater system is isolated..
O y
2.
The EFW pumps wiLL remain operable when subject to effects of runout flow and therefore can be expected to carry out their intended function during the MSLB event.
3.
A,LL potential water sources were i dentitied and, although a reactor return-to power is predicted, there is no violation of the specified acceptable fuel design Limits.
Therefore, the FSAR MSLB reactivity increase analysis remains. valid.
4.
No further action regarding IE ButLetin 80-04 is required.
O e
I
......w-w--s
- ~ '
_..__ _....~._.~.
4.0 References 1.
" Analysis of a PWR Main Steam Line Break with Continu~ed Feedwater Addition," NRC Office of Inspection and Enforcement, February 8, 1980, IE ButLetin 80-04 2.
E "Overpressurization of the Containment of a PWR Plant After a Main Steam Line Break,"
NRC Office of Inspection and i
Enf o rcement, O ctobe r 1,,1979, IE Inf o rmation Notice 79-24 3.
D.
C. T rimble (APL)
Letter to K. V.
Seyfrit (NRC, Region IV )
Subj ect :
IE ButLetin 80-04 May 27, 1980 4.
D.
C. T rimble (APL)
,(NR C, Region IV )
Letter to K. V.
Seyfrit
Subject:
IE ButLetin 80-04, Emergency Feedwater Pump Analysis July 9, 1980 5
J.
P. Marshall (APL)
Letter to J.
F. Stoltz (NRC, ORB No. 4)
Subj ect:
Additional Inf ormation Rega rding IE ButLetin 80-04 J uly 30,1982 6.
Arkansas Nuclear One Unit 1 Final Safety Analysis Report, through Amendment 49 Arkansas Power & Light Company, Septembe r 1975 l
7.
Technical Evaluation Report
(
"PWR Main Steam Line Break with Continued Feedwater Addition - Review of Acceptance Criteria" t
Franklin Research Center, November 17, 1981 T E R-C 5 506-119 8.
" Criteria for Protection' Systems for Nuclear Power Generating Stations" Institute of Electrical and Electronic Engineers, New York, NY, 1971 IEEE Std 279-1971 9.
Standard Review. Plan, Section 4.2
" Fuel System Design NRC, July 1981 NUREG-0800 10.
Standard Review Plan, Section 15.1.5
" Steam System Piping Failures Inside and Outside of Containment (PWR)"
NRC, July 1981 NUREG-0800.
e---
-m w-
--,,,-,-e--
w
v 11.
" Criteria for Accident Monitoring Functions in Light-Water-Cooled Reacto rs" American Nuclear Society, Hinsdale, IL, December 1980 A N S / A N S I-4. 5 -19 80 12
" Ins trumentation f o r Ligh t-Water-Cooled Nuc le ar Powe r Pla nts to Assess Plant and Environs Conditions During and Following P
an Ac cident,"
Revision 2, NRC, December 1980, Regulatory Guide 1.97 13
" Single Failure Criteria for PWR Fluid Systems,"
American Nuclear Society, Hinsdale, IL, June 1976, ANS-51.7/N658-1976 14 "Guality Group Classifications and St anda rds for Water,
Steam, and Radioactive-Waste-Containing Components of Nuclear Power Plants" Revision 3, NRC, Feb rua ry 1976, Regula to ry Guide 1.26 15.
" Interim Staff Position on Environmental Qualification of Safety Related Electrical Equipment," Revision 1, NRC J u ly 1981, NUR EG-0588 16.
D.
C. Trimble (APL)
Letter to R.
W.
Reid (NR R )
Subject:
Submittal of EFW Upgrade Proposal Design Information October 15, 1980 07.
D.
C. Trimble (APL)
Letter to T. M. Novak (NRR ORB)
Subject:
Final Submittal of EFW Upgrade Design Inf o rma tion D e c emb e r 1, 19 81 l
l O
9
?
.,. ~.