05000029/LER-1985-001, :on 850513,current LOCA Analysis Determined Not in Compliance W/Requirements of Section I.A of App K to 10CFR50 Re Axial Power Distribution Assumptions.Restrictive Control Rod Limit Implemented
| ML20138M746 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 06/14/1985 |
| From: | St Laurent N, Williams F YANKEE ATOMIC ELECTRIC CO. |
| To: | NRC OFFICE OF ADMINISTRATION (ADM) |
| References | |
| LER-85-001, LER-85-1, NUDOCS 8512230139 | |
| Download: ML20138M746 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2) |
| 0291985001R00 - NRC Website | |
text
-
U S. NUCLEud LEGULt. TORY COMMISSION penC Form 300 AP9 MOVED oms NO 3150 4 104 LICENSEE EVENT REPORT (LER)
DOCKET NUMSER (20 F8GE81 PACILITV NAME 116
_1 lOFl nl e Yankee Atomic Electric Company 015to[0l0lqigl9 D termination of Inappropriate LOCA Methodology Assumption /Previously Approved EVENT DATE (St LER NyWeEntel REPORT DATE (7)
OTHER F ACiLiTIES INVOLVED ISI WONTH DAY YEAR TEAM n' b
- [
MONTM DAY YEAR F AC4LITV N AMe 5 DOCnET NUUSERi$l IE 0l5l0l0l0l [ [
0l5 1 l3 8
5 8l5 0l 0l 1 0! l 0l6 3 al s oi 510 t 0 1 0 i I t Tnis mEPOnT is susMiTTEo PunsuaNT To THE nEauintutNT: op 10 Can 1. rCa.c eae e,. re er rae a u
, iiil 1
20 403thi 20 406tel 60.73iaH2H vl 73.7 t tbl MODE (St n
20.406teH1Het 60 341silti 50.73'el(2ilst 73.714el (101 I l 20 409teH1Hul 50.341eH2) 60.73teH2Hval OTHE aec na Aa e
20 40steH1H.M 80.73(eH2HJ 60.73(eH2HvenH Al J6dAJ 20 4051sHtHal X
60 73(eH2Het 60 731sH2Hve=Hel 20 4051sitillel 60 73iell2 Heal 60 73teH2Hal LICENSEE CONTACT FOR THis LER 6121 TELEPMONE huusER g1ME ARE A CQQE F. N. Williams, Manager, Reactor Engineering Department 4 i 13 4, 2, 4; ; Si 2; 4 :.
COMPLETE ONE LINE Pom E ACM COMPONENT P AitumE DESCniSED IN TMit REPORT tt3r Tyhc.
on,Tas A
R{0{,TA8LE 0
"^ ^ C.
CAV5E Sy sttu COMPON E NT
CAusE
SYSTEM COMPONENT g
q n0g I I I I I I I
I I I I I f i
1 l 1 l l 1 I
I I f I I I MONTM OAV YEam SUPPLEMENTAL REPORT E XPECTED ties YES tir p.e gemeuere LMPECTiO SUSwS$ ION CATEI NO l
l l
ABSTR ACT Its+t to f e00 spaces e e, masrees#refe/FMff#9 888d e 808CS IWD###8Ff94 888888 Il0I On May 13,1985, at 1132 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.30726e-4 months <br />, while operating at 100 percent power, it was learned that the Loss of Coolant Accident (LOCA) analysis may not be in compliance with the requirements of Section I.A of Appendix X to 10 CFR 50 pertaining to axial power distribution assumptions.
A notification was made pursuant to 10 CFR 50.72(b)(1)(fi)(A) at 1154 hours0.0134 days <br />0.321 hours <br />0.00191 weeks <br />4.39097e-4 months <br />.
To assure that the plant operation was within analyzed conditions, a more restrictive control rod limit was immediately implemented.
The condition was discovered during a review of the LOCA analysis as the result of a notificatien from the NRC of deficiencies in the Exxon LOCA analysis, this report is submitted due to the determination that the approved LOCA evaluation model [see NRC letter to YAEC, dated December 4,1975, Amendment No. 21 to Facility Operating License No. OPR-3] is no longer valid because the assumptions regarding core axial power distribution are inappropriate.
This LOCA analysis is based on a chopped-cosine axial power distribution when in fact a more limiting top-skewed axial profile could exist during periods of plant operation.
As a result, the plant may have operated with an axial power profile which would result in more severe calculated consequences than identified in the LOCA analysis.
A more restrictive control rod limit was imposed during the remainder of Core XVII and evaluation of future cores' operation continued.
LER 77-30 previously reported an unrelated LOCA analysis error.
8512230139 e50614 JEs e PDR ADOCK 05000029 S
PDR I
o Ir N, e,,.
w-U S. NUCLE 13 5 EIULATO~tY COMMIS1 SON NRC form 364A LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPnovfo ove No siso-oves E xPiRES. 8/31/95 F ACILITY NAME Ist DOCKET NUMBER (2)
LER NUMBER 163 PAGE 13) sa g g*,'
- ;<vg Yankee Atomic Electric Company vE.=
o l5 l0 l0 lo ]0 l2 l9 8l5 0 l0 l1 0 l1 0l2 0F 0 l3 itKT iM more space a orquent use eMnonet MC Fonn J66A'sll17)
On May 13, 1985, at 1132 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.30726e-4 months <br />, while operating at 100 percent power, it was learned that the current Loss of Coolant Accident (LOCA) analysis may not be in compliance with the requirements of Section I.A of Appendix K to 10 CFR 50 pertaining to axial power distribution assumptions.
The plant was notified verbally by Corporate Headquarters and a subsequent Manager of Operations directive transmitted to the plant.
The NRC was notified, via the Emergency Notification System, pursuant to 10 CFR 50.72(b)(1)(ii)(A) at 1154 hours0.0134 days <br />0.321 hours <br />0.00191 weeks <br />4.39097e-4 months <br />.
The Resident Inspector was also notified at that time.
Immediate corrective action to assure that plant operation is bounded by the existing LOCA analysis has been effected by implementing a more restrictive control rod limit.
Control rod Group C withdrawal was limited to not more than 84 inches, and subsequently, limited to 83 inches through End of Full Power Life (E0FPL).
This l limit on Group C operation serves to keep power skewed away from the top of the core.
Because the core power distribution varies with core life, as well as control rod position, the need for further correction, as a result of increasing core exposure, was investigated.
The results of this investigation subsequently permitted rod withdrawal of control group C, at or below 80% of rated full power, at a rate not to exceed 2 inches per day until fully withdrawn.
In March,1985, the NRC staff became aware of errors in the Exxon PWR LOCA analysis methods.
Because Yankee uses Exxon fuel, the NRC contacted Yankee Nuclear Services Division (YNSD) on March 22, 1985, and requested that YNSD evaluate concerns which the NRC had regarding Exxon LOCA analysis deficiencies.
Of the deficiencies noted, one was found to be applicable.
This concern dealt with the acceptability of the axial power distribution study which had been submitted in compliance with Appendix K in 1975.
The NRC staff concluded in their safety evaluation of December 4,1975 that use of the chopped-cosine power distribution was an acceptable basis for determining LOCA limits, however, this conclusion has now been questioned.
Discussions with NRC staff in NRR have been ongoing since the onset of this issue.
As a result of these discussions, a subsequent evaluation has been done using the NRC's current interpretation of an acceptable evaluation model.
The results showed that in the unlikely event that the plant was able to achieve top-skewed power distributions with a peak kw/ft equal to the technical specification limit, concurrent with a LOCA, the calculated peak clad temperature would exceed a value of 2200 F.
It was also concluded that bottom-skewed power distributions were covered by the current technical specification limits.
Therefore, instructions have been given to direct the plant to operate with control rods inserted sufficiently to assure mid or bottom-peaked power distributions in order to assure continued compliance with 10 CFR 50.46.
This report is submitted pursuant to 10 CFR 50.73(a)(2)(ii)(A) due to a determination that previously approved assumptions are inappropriate in current LOCA licensing methodology.
The use of chopped-cosine axial power distributions in LOCA analyses, which was approved by NRC for Yankee in December 1975 (Amendnent No. 21), may no longer be valid.
Placing the above restrictions on control red group C will provide reasonable assurance of compliance to the performance requirements of 10 CFR 50.46 in the event of a LOCA.
ec.o.m
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION AmazovEo oMe s.o mo-om E xPIRES. $131/85 F ACluiY NAME ty DOCKET NUM8E R (2)
LER NUM8f R (6)
PAGE(3)
VEAR S&QWENTI AL
. At v is c %
Yankee Atomic Electric Company 01510l0l0l0l2l9 8l5 0 l0 l1 0 ll 0l3 oF 0 l3 f tX1 IJ9 mere space e requeed, one edekoonel NRC Fonn J66KalUll It shoald be noted that there is a large degree of conservatism inherent in Appendix K LOCA licensing methods.
Plant operation for the remainder of Core XVII was conducted in a manner which assured that the power distribution was bounded by the existing LOCA analysis.
Further analyses have been performed showing that top skewed axial power distribution models are more representative of E0C operating corditions than the chopped cosine axial power distribution models.
Core XVIII LOCA analysis recognizes this and uses the " worst case" burnup(-dependent, xenon-induced power distributions in the burnup sensitivity analysis.
See YAEC Letter to USNRC dated August 30, 1985, " Core XVIII Refueling - Cycle-Dependent Parameters" Attachment A). The resulting peak LHGR curve provides a more limiting peak LHGR versus GWD/MTU than the curves generated for previous cores which were derived from chopped cosine axial power distributions only.
It is anticipated that this modification to the LOCA analysis model will be used in all future LOCA analyses.
In addition to the inclusion of top skewed power distributions for E0C calculations in the Core XVIII LOCA analysis, the stean-ECC water interaction model has been modified to replace the 0.8 psid AP penalty with a 0.15 psid AP penalty during pumped ECCS flow after the accumulator has been emptied.
The new value was determined by experiments conducted by EPRI.
It more accurately represent system conditions.
(For additional details, see YAEC Letter to USNRC dated August 16,1985, "LOCA Injection AP" and YAEC Letter to USNRC dated September 16, 1985, "LOCA Injection AP Penalty".)
A summary of an August 8,1985 meeting to discuss proposed revisions to the ECCS codes and code assumptions is also documented in USNRC letter to YAEC dated October 22, 1983.
The staff Safety Evaluation approving these revisions is provided in USNRC letters to YAEC dated November 27, 1985, titled " Core XVIII Reload Safety Evaluation," and
" Inspection AP Penalty for ECCS Evaluation Model."
2e.o M 2.ee
e YANKEE AT030 ELECTRIC COMPANY r i pnan. <<rs> <2<-s2er z
Star Route, Rowe, Massachusetts 01367 1 '. L i
7N K ljg, #
1/
December 16, 1985 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Subject: Licensee Event Report 50-29/85-01, Revision 1
References:
(a) License No. DPR-3 (Docket 50-29)
(b) YAEC Letter to USNRC dated June 14, 1985.
Licensee Event Report 50-29/85-01, " Determination of Inappropriate LOCA Methodology Assumption /Previously Approved" (c) USNRC Letter to YAEC dated May 22, 1985, " Confirmation of ECCS Codes" (d} YAEC Letter to USNRC dated August 16,
- 1985, "LOCA Injection AP Penalty" (e) YAEC Letter to USNRC dated August 30, 1985, " Core XVIII Refueling - Cycle-Dependent Parameters" (f) YAEC Letter to USNRC dated September 16, 1985, "LOCA Injection AP Penalty" (g) USNRC Letter to YAEC dated October 22, 1985, " Meeting Summary - Meeting with YAEC on ECCS Code Revision" (h) USNRC Letter to YAEC dated November 27, 1985, " Technical Specifications for Refueling Related Parameters"
Dear Sir:
Inaccordancewith10CFR50.73(a)(2)(fi),theattachedLicenseeEventReport, Revision 1, is hereby submitted.
This update contains information submitted in References (d), (e) and (f) and supersedes the LER reported in reference (b). Changes are indicated by lines in the right hand margin.
Very truly yours, ht,tk lll &
Normand N. St. Laurent Plant Superintendent DAR/nm Enclosure cc:
- 3] NSARC Chairman (YAEC)
[1]
Institute of Nuclear Power Operations (INP0) 10 Yr