ML20127N510
| ML20127N510 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 02/18/1977 |
| From: | Mayer L NORTHERN STATES POWER CO. |
| To: | Stello V Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9212010292 | |
| Download: ML20127N510 (3) | |
Text
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r NSED pejulatory Docket File NORTNERN STATES POWER COMPANY M f N N E A PObt e. MIN N E G OTA 99408 February 18, 1977 Tele 9
Mr Victor Stello, Director cOPled l
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Division of Operating Reactors
- 18/77
< ',9 U S Nuclear Regulatory Cocunission Washington, DC 20555 Ty s.
Dear Mr Stello-b
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,]y-MONTICELLO NUCLEAR GENERATING P1 ANT g
Docket No. 263 License No. DPR-22
- E"IS NRC Inquiry Regarding ECCS Analysis Changes This letter is written in response to a verbal request from the Monticello NRC Project Manager, Mr Dick Snaider, on February 14, 1977 concerning ECCS analysis changes. We have been in contact with General Electric, the vendor who performed the Monticello ECCS Analysis and have received the following information.
The aspects of the ECCS model identified on the attached pages have been under review and discussion between General Electric and your staff. TVo groups of changes are identified. Of the first group entitled "ECCS Input Changes and l
Detrimental Model Changes", items a, b, e-1, e-ii, f and g are applicable to the Monticello ECCS analysis filed July 9, 1975. The remaining items are not applicable. The cumulative effect of the applicable changes is that the MAPLHGR q
limits in the Monticello ECCS analysis and in the current Technical Specifications are conservative by an amount of 27..
Tbc attachment continues to identify
" Beneficial Model Improvements" for which an additional margin of conservatism can be expected. Since the corrait MAPLHGR limits were shown to be conservative in light of the above changes, there was no estimate made of the additional conservative effects of beneficial model improvements to the Monticello ECCS analysis.
After the model changes have been incorporated into an ECCS model which is formally approved in accordance with 10CFR50.46, we will submi* an analysis using the revised model. In the interim, operation is supported by the July 9,1975 esaluation using the currently approved ECCS model and the reevaluation reported above which shows compliance of the existing Technical Specification operating limits with the criteria of 10CFR50.46.
Yours very truly,
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~ y'g L 0 Mayer, PE
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Manager of Nuclear Support Services LOM/MIV/ak 1babb O
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Attn: J W Feman h'K 9212010292 770218 PDR ADOCK 05000263 P
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DESCRIPTION OF ECCS REEVALUATION hs:Qsd witi cafd-Y* N The attached table contains the results of all emergency core' coolfng'"""
system (ECCS) input and model changes that have been identified and indicates those which :are applicable to the plant.
These changes have been divided into two groups, ar.d the cumulative effects on maximum planar linear heat generation cate (MAPUICR) of each group conservatively e s t ima t ed.
The total ef fect on MAPLitCR has then been determined. The groupings and a description of the individual changes are given below.
The ceferences identified following each change description provide further amplificatior. of the change.
1.
ECCS Input Changgs and Detrimental Model Changes a.
New suction Bicak Area in the SAFE calculation - This item refers to a new recirculation suction line break area if it has changed from previous analyses.
(Reference 1) b.
Vaporization Calculation - The calculation of the steam gen-eration in the E EFLOOD code has been made consistent with the approved ECCS evaluation model.
In previous anclyses, the vapor generation was underestimated which provided a reduced effect on counter current flow limiting (CCFL). A lower value of CCFL gives shorter reflood time and lower values of peak clad temperature (PCT).
(Reference 1)
Elhminate Structural Absorption Double Credit - In the original c.
calculation doubic credit was taken for the effects of structural absorption in the decay heat calculation. (Reference 1) d.
Credit for Suction Line Friction - The approved ECCS evaluation model allows for reduction of blowdown due to piping friction.
In the previous analyses on plants incorporating the low pres-sure coolant injecticn system (LPCI) modification, no credit was taken for friction in the recirculation system suction line for the discharge break calculations. (Reference 1) e.
Others.
1.
Reactor Internals Thermal Characteristics - In the REFLOOD code, the reactor internals are modeled as heat sources which ircrease stnam generation.
Since the magnitude of these sources has been revised, there will be an effect on CCFL and refload time.
(Reference 1)
- 11. Bypass Area Adjus tment - The bypass area provides a path i
for core spray flow around as opposed to through the fuel assembly.
For a larger bypass area, there is a reduced CCFL eff 3ct on srray water entering the bypass region. More precise bypass area calculations have been completed and ur ed for REFLOOD code inputs. (Refer' ace 1) i i
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iii. Discharge Valve Closure Assumption - For LPCI modi-fied plants, the effective pipe break area is dependent on whether or not the discharge valve is assumed to close.
It is conservative to assume no valve closure since this maximizes the break area for a discharge break. In previous analyses, it was assumed that the valve closed, f.
Pressure Rule - In the SAFE code, there is a non-conservative spike in pressure when the reflooding flow recovers the bottom of the active fuel.
In previous analyses, the calculated pres-sure before the spike was assumed to remain constant for duration of the event.
As a result of discussion with the NRC, it was agreed to use the lower of the SAFE calculated value or conbtant pressure calculation.
(Reference 1) g.
Increased CCFL Differential Pressure - Some experimental evidence exists that the differential pressure in a fuct assembly during periods of CCFL may be higher than previously assumed. Ihis could cause a delay in reflood time.
(Reference 5) 2.
Bene ficial Model Improvements a.
REFLOOD 04 - In the approved ECCS evaluation model (REFLOOD 03) for certain conditions, the steam split between the jet pumps ar.d the fuel was incorrectly calculated.
REFLOOD 04 revised this calculation.
(Reference 3) b.
Partial Drill - For plants with plugged bypass flow holes and some but not all fuel assembly lower tie plates drilled, no credit has been given for the reflood flow through these holes.
The partial drill change to the ECCS model conservatively accounts for this change.
(Reference 3)
CHASTE 05 - In the approved ECCS evaluation model (CRASTE 04), there c.
is a very conservative treatment of radiation and conduction heat transfer.
In CHASTE 05, the heat transfer effects are treated more consistent with the actual phenomena and experbnental data.
(Reference 4)
REFERENCES 1.
" Meeting Summary January 12, 1977 GE Meeting Concerning Effects on Operat--
ing BWR's of Input Errors to ' Appendix K' LOCA Analys!.s".
2.
Letter, R E Engel to Victor Stello Jr, " Supplemental Information to Eliminate Significant In-Core Vibrations (NEDE-21156-1)". September 16, 1976.
3.
Letter, A J Levine to D F Ross, "BWR ECCS LOCA Evaluation Model", Oct 13, 1976.
4.
Letter, A J Levine to D F Ross, " General Electric (GE) Loss of Coolant Accident (LOCA) Analysis Model Revisions-Core Heatup Code CHASTE 05", January 27, 1977.
5.
Letter, G G Sherwood to Victor Stello, Jr, " Additional Changes to GE Emergency Core Cooling System Analyses". February 14, 1977.
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