05000346/LER-1995-001, :on 950126,determined That LOCA Analysis Results for 2 Ft Core Elevation Potentially non-conservative Due to Modeling Errors.Vendor Evaluated Current Fuel Cycle to Determine If Operating Limits Valid
| ML20080M835 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 02/24/1995 |
| From: | Peterson N, Jeffery Wood TOLEDO EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| AB-95-0006, AB-95-6, LER-95-001, LER-95-1, NP-33-95-001, NP-33-95-1, NUDOCS 9503060055 | |
| Download: ML20080M835 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition |
| 3461995001R00 - NRC Website | |
text
_ -_
,,;} * ^
TOLEDO EC180N A Centerior Energy Company E
O 1'
AB-95-0006 g
NP-33-95-001 1
Docket No. 50-346 License No. NPF-3
)
. February 24, 1995 United States Nuclear Regulatory Commission Document Control Desk
}.
Vashington, D. C.
20555 Gentlemen:
LER 95-001 Davis-Besse Nuclear Power Station, Unit No. 1 Date of Occurrence - January 26,1995 Enclosed please find Licensee Event Report 95-001, which is being submitted to provide 30 days written notification of the subject oc~ -~ce.
This LER is-being submitted in accordance with 10 CFR 50.73(a)(2)(uno) and 10 CFR l
50.46(a)(3)(li).
Very truly yours,
]"
J n K. Vood Plant Manager Davis-Besse Nuclear Power Station JKV/ eld Enclosure cc Mr. John B. Hartin Regional Administrator USNRC Region III Hr.-Stan Stasek DB-1 NRC Sr. Resident Inspector i
Utility Radiological Safety Board 9503060055 950224 PDR ADOCK 05000346 S
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMd NU. 3150-0104 AC3 EXPIRES 5/31/95 EST! MATED BURDEN PER RESPONSE TO COMPLY WITH THIS l"l$"^ "REGARC,'STRoEN E'ddio""*i NF6"*C LICENSEE EVENT REPORT (LER)
ANO RECOROS MANAGEMENT BRANCH (MN98 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20655@01. AND TO THE PAPERWORK REDUCTION PRWECT 13150 0104), OFFICE OF (See reverse for required number of digits / characters for each block)
MANAGEMENT ANO BUOGET, WASHtNGTON, DC 20603.
FAC8LITY NAME (1)
DOCILET NUMBER (2)
PAag (3)
Davis-Besse Unit Number 1 05000-346 1 OF 06 TITLE (4)
Potentially Non-Conservative LOCA Analysis Due to Modeling Errors EVENT DATE (5)
LER NUMBER (6 REPORT NUMBER (7)
OTHER FACILITIES INVOLVED (8)
MONTH DAV YEAR YEAR MONTH DAY YEAR g
FAClurY NAME DOCKET NUMBER 01 26 95 95
~
001
~
00 02 24 95 05000 EIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 0 (Check one or more)
OPERATING MODE (9)
[
20 402(b) 20.405(c) 50 73(a)(2)(iv) 73 71(b)
POWER 20 405(a)(1)(i) 50.36(c)(1) 50 73(a)(2)(v) 73.71(c)
_ LEVEL (10) 100 20 405(aH1Hn) 50.36(c)(2) 50 73(a)(2)(vu) y OTHER 20.405(a)(1)(m) 50.73(a)(2)(i) 50,73(a)(2)(viii)(A) 88 20 405(a)(1)(iv)
X 50.73(a)(2)(ii) 50.73(a)(2)(vin)(B)
F,,,,3eeA, 20.405(a)(1)(v) 50.73(a)(2)(m) 50.73(a)(2)(x) 50.46 LICENSEE CONTACT FOR THIS LER (12)
NAM 6 TEMPHONE NUMBER Moe Area Code)
Norman K. Peterson. Sr. Engineer - Licensing (419) 321-8450 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIDED IN THIS REPORT (13)
CAUSE
SYSTEM COMPONENT MANUFACTURER
CAUSE
SYSTEM COMPONENT MANUFACTURER RO SUPPLEMENTAL REPORT EXPECTED (14)
EXPECTED MONTH DAY YEAR YES SUBMISSION No pf yes, co<were EMPECTED SUOM!sSION DATE)
X DATE (15)
ABSTRACT (Limit to 1400 spaces, i e., approximately 15 single-spaced typewritten hnes) (16)
The Loss of Coolant Accident (LOCA) analysis for Davis-Besse is performed by B&V Nuclear Technologies (BVNT).
This analysis determines the allovable linear heat rate limits as presented in the Core Operating Limits Report (COLR), which is referenced by the Technical Specifications (TS). The COLR provides acceptance criteria for some TS surveillance requirements.
BVNT has determined that the LOCA analysis results for the 2-ft core elevation are potentially non-conservative.
BVNT initiated a Preliminary Safety Concern (PSC 5-94) on November 28, 1994 and nctified the Nuclear Regulatory Commission (NRC) of this concern by letter dated January 27, 1995. The NRC was notified by Toledo Edison via Emergency Notification System (ENS) at 1522 hours0.0176 days <br />0.423 hours <br />0.00252 weeks <br />5.79121e-4 months <br /> on January 26, 1995 under 10CFR 50.72 (b)(1)(ii)(B) because the non-conservatism could result in the predicted peak clad temperature exceeding 10CFR50.46 limits at the 2-ft core elevation.
Existing plant operating restrictions are sufficient to preclude limits from being exceeded.
This event is being reported under 10 CFR 50.73(a)(2)(ii)(B) as a condition outside the plant's design basis. This report also fulfills the reporting requirements of 10 CFR 50.46(a)(3)(ii).
NAC FORM 366 [$ 921
NRC FOM 384A U.S. NUCLEAR REGULATORY CC^^^ TE9N APPROVED BY OM8 NO. 3150 0104 EXPIRES 5/31/96 ESTIMATED OURDEN PER RESPONSE TO CoWPLY wrTH TMS LICENSEE EVENT REPORT (LER)
Elf.""* "J $'R"J!$0 Jio L IOr%
E TEXT CONTINUATION -
MC " 'cMg"'MEgo so.o roy 7c"M@"
w g g atoucron,p omes or 4
eaceurv name m oocarv muussa m osR avseen m paos m SEQutiNTiA.
fEVISCN.
y NUMSER NuuSER 05000 -346 02 OF 06' Davis-Besse Unit Number 1 95 001 00' snus cu me nom a==== a. --
z uo, an,ec ram, sesy on Description of Occurrence:
Recently, the Loss of Coolant Accident (LOCA) analysis applicable to Davis-Besse, performed by B&V Nuclear Technologies (BVNT), has concluded that the analytical results for the 2-ft core elevation are potentially j
non-conservative. The non-conservatism is the result of the discovery of a more adverse combination of Core Flood Tank (CFT) level and pressure than was previously believed to be bounding.
Specifically, operation with CFT at.the maximum' permissible level and the minimum permissible pressure results in the l
most rapid CFT pressure decrease and lowest rate of core flood injection.
It was determined that only the 2-ft core elevation is sensitive to this reduction in flow rate.
j.
Additionally, during the investigation of this condition, a computer code interface concern between the CRAFT 2 and THETAl-B computer codes used in the LOCA analysis was discovered.
Coarse noding in the-core model and internal CRAFT 2 treatment of refilling nodes combined with a low data sampling rate produced a non-conservative enthalpy input into the THETAl-B code. This caused significant changes in the analytical results to occur only at the 2-ft core elevation. These computer codes and methodology are part of the Nuclear Regulatory Commission (NRC) approved 10 CFR 50, Appendix K LOCA Evaluation Model.
BVNT initiated an internal Preliminary Safety Cencern (PSC 5-94) on 2
November 28, 1994. The preliminary evaluation in response to PSC 5-94 concluded that a reduction in the 2-ft allovable linear heat rate of 1.3 KV/ft is sufficient to maintain Peak Cladding Temperature (PCT) within the established limits for all B&W plants. The B&V Fuel Company has reviewed the Davis-Besse core operating limits-for the present cycle and has concluded that power tilt, power imbalt.nce, and control rod insertion limits, as provided in the Technical Specifications (TS), are not adversely impacted, hovever, the Core Operating Limits Report (COLR)~allovable linear heat rates (LHRs) are adversely affected and vill be revised.
Apparent Cause of Occurrence:
The apparent cause for this condition was that the assumption made in the original analysis was not conservative at the 2-ft elevation.
During original licensing of the LOCA evaluation model, nominal CFT initial conditions (level and pressure) were used.
Sensitivity studies done by BVNT at the core midplane (6-ft. elevation) indicated that initial conditions 1
Nac romu ssea ng
.. _.. _ _ _ _ _, _ _, _ _.,. ~. _ _
_m
_m.
NRC FORM 364A U.S. NUCLEAR REGULATORY CC^^^^:5?!Ni APPROVED BY OMB NO. 3150 0104 N
EXPIRES 5/31/96 1
ESTIMATED OVROEN PER RESPONSE TO COMY WffM TM4 LICENSEE EVENT REPORT (LER)
C'#"E$E'5".u.EliM?oLL"O TEXT CONTINUATION
@"co^"Tg','M*,f,"7c"M";*T M M o*,'%",":'"
"a;': *
- A 1
raciufv name m ooenET aumeen m un numeen m n
Paea m
. stove.ta, aEveoN g y,,,
NyWSER NUMOER g 05000 -346 g 03 oF 06 Davis-Besse Unit Number i 95 001-00- l mTp-
.cm:
we me9 n -
Apparent Cause of Occurrence (Continued):
using minimum CFT level and pressure vere the most conservative for evaluating PCT.
In an attempt to bound all permissible operating conditions, these CFT parameters were adopted for use in subsequent analyses.
- However, it was recently discovered that the most limiting initial conditions for the 2-ft elevation PCT analysis are the maximum CFT liquid inventory and minimum pressure.
The apparent cause for the second issue concerning computer code behavior was less than ideal familiarity with the internal operation of the CRAFT 2 code used in the currently approved Emergency Core Cooling System (ECCS) evaluation model. Earlier extensive reviews by all involved parties failed to detect the noted deficiencies. A contributing factor is that the existing evaluation model is cumbersome to use, placing practical limits on extent'and frequency of use. Thus, experience and expertise in its application is limited.
Analysis of Occurrence:
When the CFT input parameter changes were ana'lyzed with the current evaluation model and existing allowable LHRs, the calculated PCT change at the 2-ft elevation was found to be greater than 50' F and exceeded 2200*F.
However, if the LOCA LHR limits at the 2-ft elevation are reduced by-1.3 i
KW/ft, all PCTs vill meet the acceptance criteria contained in 10 CFR 50.46.
3 i
In November 1994, a sensitivity study performed by BWNT for another utility using the RELAPS/ MOD 2-based evaluation model for the Mark-Bil fuel design i
revealed that the maximum CFT liquid inventory would produce the highest PCT at the 2-ft elevation.
It was found that the combination of minimum gas volume and minimum pressure would result in the lovest CFT flow during the j
adiabatic heatup period. This lov flow decreased the liquid remaining in the reactor vessel lower plenum and extends the time period to refill the lover plenum. The net result was an increase in the adiabatic heatup period by approximately 20 percent. The PCT was found to be higher than would be produced with the minimum CFT initial liquid volume.
On this basis, Preliminary Safety Concern (PSC) 5-94 was initiated.
This PSC 2
related to the anticipated PCT variation associated with the input of the minimum versus the maximum CFT liquid inventory in the approved Large Break LOCA ECCS analyses.
i t
.U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 315o-o104 de2>
EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS LICENSEE EVENT REPORT (LER) 5 Z I?"RE S $'NG ?u"A"' ES$JioLl "ITU N
TEXT CONTINUATION 7,$u"'AEc"OMNENTE!'cEi"ooi No's o
THE PAPE5fWOAK REDUCTION PROJECT (3 50-0104), OFFCE OF MANAGEMENT AND BUDGET, WASHINGTON, De 20501 FACIUTV MAME (i)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE13)
SEQUENTA.
REV@ON YEAR 05000 -346 op Davis-Besse Unit Number 1 95 001 00-Tw tu m. se.ce ne ree-we.aomana e oo Nnc Fan,, my o r, Analysis of Occur: ence (Continued):
The adverse effects from the CFT parameter changes were compounded by an observed variation in the CRAFT 2 core path inlet enthalpy supplied by the CRAFT 2 code to THETAl-B for the hot fuel pin thermal analysis. The enthalpy was supplied to THETAl-B on a coarse data frequency (one point every 0.5 seconds).
However, the CRAFT 2 output was subject to high frequency-oscillations that corresponded to the changes in instantaneous flow direction. The calculated PCT was found to be sensitive to the enthalpy sampling because a large enthalpy difference existed between the two nodes surrounding the 2-ft elevation.
In addition to the oscillations, it was observed that the enthalpy could also be skewed in the non-conservative direction by a nonhomogeneous treatment of the core nodes following total dryout with subsequent return to two-phase conditions. The Large Break LOCA blovdown model is constrained to homogeneous flows calculated by homogenous node conditions.
So long as the nodes remain continuously in two-phase Large Break LOCA conditions the homogeneous conditions are correctly calculated.
After dryout and return to two-phase conditions, however, the homogeneous condition is not met.
Under these conditions, CRAFT 2 allows the node to separate the steam and liquid phases.
Because of the flow path connections, this configuration can artificially cause the inlet flow path enthalpy to be lover than the model homogeneous enthalpy.
Since this enthalpy is transferred to THETAl-B, it can cause nonconservative conditions to be used for the hot fuel pin analysis.
While this mechanism could potentially affect results at all elevations, significant affects are limited to the 2-ft elevation.
Corrective Actions
After the initial 2-ft LHR limit reduction estimates were available, BVNT evaluated the current fuel cycle to determine if the operating limits remained valid. Technical Specification operating limits are not affected and remain valid. Therefore, this condition has minimal safety significance.
Administrative reductions of 1.5 kV/ft have been temporarily applied to the affected LHR limits as provided in the COLR. These COLR limits are applicable only under certain core flux imbalance or tilt conditions (e.g.,
dropped rod).
This assures that PCTs would remain acceptable even if a LOCA should occur during unusual operations outside the TS limits.
Based on the most recent preliminary results, required reductions in allowable linear heat rates for Davis-Besse are considerably less than the worst case generic B&W plant predicted required reduction (1.3 kV/ft).
The 1.5 kV/ft administrative reduction in LHR limits applied at Davis-Besse is conservative.
NRcFORu mAis,
~
.eU.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 Dem EXPlRES S/31/95 ErnMATED BURDEN PER RESPONSE TO COMPLY Wf7H THl3 LICENSEE EVENT REPORT (LER)
C Z E "REl %'S ER E E'E J *o O.F % Ir%
s TEU CONTINUATION M Q8"^*'8'"'[w'! @ g ">c' Z 8, M c u 3
THE PAPERWORK REDUCTION PROJECT (31500104), OFFCE OF MANAGEMENT AND BUDGET, WASHINGTON. OC 20503 FACILffY MAME (t)
DOCKET NUMBER (2)
LER NUMSER (4)
PAGE (3) 6EQUENT A REYSION g
- 8'"
05000 -346 03 OF 06 Davis-Besse Unit Number 1 95 001 00' TxxT ou om sou, maa me emana m>
or w<uam, asy vn Corrective Actions (Continued):
The adjustment of the traditional CFT liquid inventory input assumption and non-conservative enthalpy specified to the hot fuel pin thermal analysis vill produce changes in the Large Break LOCA peak cladding temperature in excess of 50' F.
Reductions in the LOCA LHR limits are required to continue to meet the 10 CFR 50.46 acceptance criteria. The most conservative CFT liquid inventory input will be used in analyses that conservatively adjust the enthalpy transferred from the CRAFT 2 results to THETAl-B for the hot fuel pin analysis. The work performed to date supports the validity of a simple i
change in the enthalpy data transfer to the THETAl-B analysis, without input model changes or use of a new CRAFT 2 code version. The current analyses vill continue to calculate the CRAFT 2 blowdown transient without any changes. The 1
inlet enthalpy for the CRAFT 2 analysis will be adjusted before the input is
)
supplied to THETAl-B. The inlet enthalpy will be conservatively set to envelope on the upstream homogeneous nodal enthalpy based on the filtered flow direction. THETAl-B will be run with this conservative enthalpy to determine a PCT. The new 2-ft LOCA LHR limits vill be used in maneuvering analyses to determine if any changes are needed to the core operating limits.
l The required analyses are in progress. The results of the LOCA analyses, including changes the COLR, are scheduled to be completed by August of this year.
A report vill be sent to the NRC by BVNT within 30 days of the completion of the aforementioned reanalyses. Toledo Edison vill also provide the COLR revisions to the NRC in accordance with T.S. 6.9.1.7 within 30 days of the completion of the BVNT analyses.
The existing CRAFT 2/THETAl-B based evaluation model is in the process of being replaced by a RELAP5 based evaluation model. A topical report (BAV-10192) describing the new evaluation model was submitted for NRC review in February, 1994. The new model does not suffer from the same technical shortcomings as the CRAFT 2 based model and is more economical to use.
It utilizes modern thermal-hydraulic techniques and requires fever manual interfaces.
The core noding in the new model is much finer, leading to more reliable results.
NRCroRM m A mea
.=
NRC FORM 386A U.S. NUCLEAR REGULATORY COMMISSION APPROVED SY OME NO. 3180 0104
@st EXPlRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS LICENSEE EVENT REPORT (LER) lI O 2 ko*uo"','a'iURENlfSE d k O !="! E TEXT CONTINUATION
^"
"'C " * "^*^ ". ' 'ASHINGTON, DC
- "C""""'""""^"
REGULMORY COMMISSON W 306664301, AND TO THE PAPERWORK PEDUCTION PRCk!ECT (319001041, OFFICE OF MANAGEMENT AND SUDGET, WASHINGTON. DC 20003 PACI W NA84E0)
DOCaLET NUMSER (3)
LER NUteggR (6)
PAGE Q) i SEQUENTA.
MFviscN gg NUMBER NUMBER 05000 -346 Davis-Besse Unit Number 1 06 OF 06 95 001 00' sex, tu aa..P a. a o.ea.a a. :
e.P., a NRC ram, m o n t
Failure Data In the past five years, there has been one Licensee Event Report (LER) submitted describing an error in the Davis-Besse LOCA analysis.
LER 91-006 describes an event where the analysis of boron concentration in the core after a LOCA in the Reactor Coolant System cold leg was non-conservative, f
i NP-33-95-001 PCA0R: 94-1327