ML20155H753
| ML20155H753 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 11/04/1998 |
| From: | Geoffrey Edwards PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9811100271 | |
| Download: ML20155H753 (5) | |
Text
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station support Department
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10 CFR 50.46 (a)(3)(i) and (ii)
PECO NUCLEAR ecco e,e,ox comgee, 1
A Unit of PECO Energy NavIeI'NsIli 1
t November 4,1998 Docket Nos. 50-277
[
50-278 License Nos. DPR-44 DPR-56 U. S. Nuclear. Regulatory Commission Attn.: Document Control Desk' j
Washington, DC 20555 i
Subject:
Peach Bottom Atomic Power Station, Units 2 and 3 10 CFR 50.46 Reporting Requirements
References:
Energy)) to U. S. Nuclear Regulatory Commission (USNRC),
dated January 30,1997 i
.2) Letter from G. D. Edwards (PECO Energy) to USNRC dated March 20,1998
Dear Sir / Madam:
In accordance with 10 CFR 50.46 (a)(3)(i) and (ii), the following is a revision to the -
. licensing basis Loss-of-Coolant Accident (LOCA) peak clad temperatures (PCTs) for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3. Additionally, PECO
/
Energy is revising its method of reporting changes in the licensing basis PCT.
Rather than identifying changes in licensing basis PCT by a single value based on the most limiting fuel type, changes will be reported by each fuel type. This change I
in reporting will ensure greater accuracy in reporting changes in the peak cladding temperatures. Tables 1 and 2 (attached) provide the revised PCT values and the j
applicable changes for PBAPS, Units 2 and 3, respectively. Based on the a
accumulated changes which result in a temperature difference of greater than 50 Fg l
from the calculated baseline temperature, this report is being submitted within 30 days.-
A change to the PCT was previously reported in a 10 CFR 50.46 report as
. discussed in the Reference 1 letter The Reference 1 letter discussed a 45 F increase in the licensing basis PCT, which was conservatively applied to all the fuel
. types analyzed for PBAPS, Units 2 and 3. This 45 F change represented the composite of three previous changes (identified in the attached General Electric Nuclear Energy (GENE) letters as MFN 090-93, 278-95, and 088-96). Each of 9811100271'9811N "h
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' ADOCK 05000277 PDR
. PDR. h L
Nov:;mb:r 4,1998 Page 2
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these changes does not necessarily apply to all fuel types. Table 1 (PBAPS, Unit 2) and Table 2 (PBAPS, Unit 3) reflect the appropriate application to each fuel type.
Table 1 reflects the impact of the incorporation of the Recirculation Pump Trip (RPT) modification on the baseline PCTs for PBAPS, Unit 2. The installation of the RPT modification results in a 45 F PCT increase for the limiting fuel type (P8X8R).
This change was discussed in the Reference 2 letter. The analysis performed by 1
GENE in 1994 to support this modification is considered to be a new baseline as shown in Table 1. This modification was not installed on PBAPS, Unit 2 until the most recent refueling outage (2R12, October,1998). The recirculation pump trip l
modification is currently scheduled to be installed in PBAPS, Unit 3 during refueling outage 3R12 (October,1999). Therefore, Table 2 does not reflect the effect of RPT on the current Unit 3 baseline PCTs.
l In a letter dated October 5,1998, GENE provided PECO Energy Company a summary of the revisions in the PCTs. As identified in the attachments, these
)
changes and errors have been previously reported to the U. S. Nuclear Regulatory Commission (USNRC) in accordance with 10 CFR 50.46(a)(3)(ii) as identified in the Attached letters. The 50 F change resulting from the input parameter study was not previously identified to PECO Energy Company. However, this change is reflected in this report.
The attached Tables provide, by fuel type, the new baseline (calculated) PCTs, the i
applicable errors, and the resultant estimated licensing basis PCTs. The estimated licensing basis peak clad temperatures for the most limiting fuel types at PBAPS, Units 2 and 3 are 1795 F (P8X8R) and 1755 F (P8X8R), respectively. This represents more than 400 F margin to the 2200 F limit specified in 10 CFR 50.46.
Additionally, the upper bound peak clad temperature remains below the 1600 F limit specified in the USNRC acceptance of the SAFER /GESTR methodology.
If you have any questions, please do not hesitate to contact us.
Ve truly yours, G. n Edwards Director - Licensing Attachments cc:
H. J. Miller, Administrator, Region I, USNRC A. C. McMurtray, USNRC Senior Resident inspector, PBAPS l
l
November 4,1998 Docket Nos. 50-277 50-278 TABLEI BASELINE PCT VALUES AND APPLICABLE CHANGES i
(PBAPS, UNIT 2 - WITH THE RPT MODIFICATION)
P8x8R GE8 GE9 GE 11/13 BASELINE PCT F 1735 1624 1624 1645 CHANGES j
i
- 1. MFN 090-93*
0 0
0 5
(Flow initialization / sign error) i
- 2. MFN 278-95*
10 10 10 10 (Bottom head drain)
- 3. MFN 088-96*
0 0
30 30 (Incorrect number of fuel rods) i
- 4. MFN 090-93
- 50 50 50 50 l
(Input parameter sensitivity) 1 TOTAL 60 60 90 95 i
ESTIMATED LICENSING 1795 1684 1714 1740 j
BASIS PCT l-
- See attachment for the associated GENE letter which reported this change. The GENE letter is identified with a "MFN" designation.
i
November 4,1998 Docket Nos. 50-277 50-278 TABLE 2 BASELINE PCT VALUES AND APPLICABLE CIIANGES (PBAPS, UNIT 3 - NO RPT MODIFICATION)
P8x8R GE8 GE9 GE 11/13 BASELINE PCT *F 1690 1575 1575 1645 CHANGES
- 1. MFN 090-93
- 5 5
5 5
(Flow initialization / sign error)
- 2. MFN 278-95*
10 10 10 10 (Bottom head drain)
- 3. MFN 088-96*
0 0
30 30 (Incorrect number of fuel rods)
- 4. MFN 090-93*
50 50 50 50 (Input parameter sensitivity)
TOTAL 65 65 95 95 ESTIMATED LICENSING 1755 1640 1670 1740 BASIS PCT-
- See attachment for the associated GENE letter which reported this change. The GENE letter is identified with a "MFN" designation.
I l
l
t i
t ATTACHMENT i
GENE LETTERS MFN 090-93 MFN 278-95 MFN 088-96 l
l i
p-1 (D
t GENuclear Energy June 30,1993 MFN #090-93 1
Of5cc ofNuclear Reactor Regulation US Nuclear Regulatory Comnussion Mail Station PI-137 Washia-tan DC 20555 ATTN: De-* Control Desk 1
SUBJECT:
REPOR11NG OF CHANGES AND ERRORS IN ECCS EVALUATION MODELS
REFERENCE:
- 1) Letter. SJ Stark to the Office of Nuclear Reactor Regulation,
" Reporting of Changes and Errors in ECCS Evaluation Models" dated June 26,1992 (MFN # 058-92)
The purpose of this leuer is to report, in accordance with 10CFR50.46 (a) (3) (ii), the impact of changes and errors in the Emergency Core Cooling Systems (ECCS) evaluation methodology used by GE. This report covers the period from the last report (Reference 1) to the present. It is noted that Peak Cladding Temperature (PCT) vanations resulting from plant specific system or fuel changes are not addressed in this letter. These should be treated, as appropriate, on a plant specific basis in accordance with other sections of 10CFR50.
There have been no changes er errors identified for the SAFE /REFLOOD model described in NEDE 20566-P-A " Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K" Two minor coding errors were corrected in the SAFER Code. The SAFER /GESTR methodology is desenbed in NEDE 23785-1-P-A, "The GESTR-LOCA and SAFER Models for the Evaluation of Loss-of-Coolant Accidents", and NEDE 30996-P-A.
" SAFER Model for Evaluauon of Loss-of-Coolant Accidents for Jet Pump and Non Jet Pump Plants"..The first error corrected was improper upper plenum flow initialization.
L This error caused a flow discentmuity at the beginning of the transient. A second error was corrected that impacts the latter part of a small break LOCA. A sign error m the l
pressure drop balance caused the top of the hot channel to remain uncovered es en after the I
upper plenum and b> pass were full. The impact of these errors on predicted pct is 15 F.
r # >ff' y n. n
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I The observation that ECCS evaluauon models can be sensitive to small input parameter changes under some ciremnstances was reported in Reference 1. Based on the SAFER case analyzed at the time, the range ofimpact on the predicted PCT was reported as i
- 50*F. Recent studies have indicated that the impact could be slightly larger than : 50'F for some BWR/4 plants with LPCI injection into the lower plenum using the SAFER model. 'Ibese studies indicated a total variation ofless than 85* F for most cases but with one case showing a range of 102' F (i.e.. greater than 50 F).
The identified sensitivity is related to the explicit numerical treatment in S AFER combined with rapid and simultaneous variations of multiple parameters. Work is underway to limit this sensitivity through better control of time steps in the computation. 'Ihis will provide l
assurance that such sensitivities are well within the previously stated 50* F. Any changes resulting from this activity will be reviewed with the NRC at the appropriate time.
It should be noted that existag PCT prMW are valid (i.e., within the stated uncertamty band) and no change to any plant specific evaluation is required.
By copy of this letter, Licensees utilizing the GE ECCS methodology in their plant licensing are informed of the status of changes in the evaluation methodology. Since no reanalysis or technical speci6 cation modifications are required, this submittal is believed to sansfy 10CFR50.46 (a) (3) (ii) for evaluation model changes without further reporting on the part of the individual utilities.
Ifyou have any questions, please call me or HC Pfefferien at (408) 925-3392.
l Sincerely, G2.c.%&J e 00%
RC Mitchell, Manager Safety & Communications (408) 925-2755 M/C 487 CC: HC Pfefferien
v)
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GENuclear Energy Generaf Beceso comparer P. o. Bos 180. wmeenpeen. NC 20402 D - o!,ei15,1995 RJR-95-118 MFN-278-95 Doomm Control Desk U. S. Nuclear Regulatory Comnussion WasMa-tm DC 20555-0001 Attenoon R. C, Jones, Jr.
Subject:
Reporting of Changes and Errors in ECCS Evaluation Models l
Reference:
- 1. Letter, J. F. Klapproth to the Document Control Desk (R. C. Jones. Jr., Reporting of Changes and Errors in ECCS Etnluanon Models, dated June 24, I995 (MFN-087-95).
- 2. Letter, R. C. Mitchell to the Office of Nuclear Reactor Regulation. Repornng of Changes and Errors in ECCS Enduation Models, dated July I, t994 (MFN No.
088-94).
GE is submitting this letter which revises the Reference I letter. Revisions are marked by change bars in the margm.
He purpose of this letter is to report, in accordance with 10 CFR 50.46 (a) (3) (ii), the impact of changes and errors in the Emergency Core Cooling Systems (ECCS) evaluation methodology used by GE. This repon covers the period from the last report (Reference 2) to the present. It is noted that Peak Cladding Temperature (PCT) variations resulting from plant specific system or fuel changes are noi addressed in this letter. 'Ibese should be treated, as appropriate, on a plant specific basis in accordance with other sections of 10 CFR 50.
There have been no changes or errors identified for the SAFE /REFLOOD model desenbed in NEDE 20566-P-A, AnalyncalModelfor Loss-of-Coolant Analysts in Accordance wath 10 CFR SO
\\
Appendix K.
Here have been no changes or errors identified for the SAFER /GESTR model desenbed in NEDE :3785-l-P-A. The GESTR-LOCA and SAFER Models for Evaluanon of Loss-of-Coolant Accidents, and NEDE 30996-P-A. SAFER Modelfor Evaluanon ofLoss-of-Coolant Accidentsfor Jet P:.mp and Non-Jet Pump Plants.
l gw & >w
i Document Contml Desk U. S. Nuclear Regulatory Comnussion Page 2 a
i In March 1995, a domestic utility requested that GENE review a concern regarding the RPV bottom head drain (BHD) impact on the LOCA analysis. The concern was that because the bcttom j
head drain line is directly enaed o the reactor recirculation loops, that a recirculation line break t
LOCA would also break the BHD, and the vessel would depressurize to the drywell faster than j
assumed in current models. Also, upon such an event occumng, some water required to keep the core covered to the 2/3 core height would exit the core due to either gravity or core pressure via the i
intercc==:ei recuculation and bottom head RWCU suction lines.
l A GENE evaluation concluded that wlule no analysis had been pfu. 4 to precisely evalumee the PCT impact of the recirnilatian line break LOCA including the BHD, it is believed that the imped is j
{
less than 10*F based on engmeenng judgment and mL.pcietion of previous LOCA analyses. Since an event is considered by the NRC to be signi6 cant if the PCT is increased more than 50'F (10CFR50.46 j
(a)(3)(i)), this arannat of increase can be considered insignificant and well within the margms of the safety analysis.
The impact of the BHD exiting flow on mamininmg RPV level inside the shroud is simdarly insipF==
It was determmed that a slightly higher muumum makeup flow will be required, however, the increased malraip is well within the margas of available ECCS systems The minen==
makeup flow corresponds to that necessary to makeup for decay heat and the drain rate from the BHD.
By copy of this letter, Licensees utihnng the GE ECCS methodology in their plant licensing are informed of the status of changes in the evaluation methodology. Since no re-analysis or techecal specification modifications are required, this submittal is believed to satisfy 10 CFR 50.46 (a) (3) (ii) for evaluation model changes without further reporting on the part ofindividual utilities.
If you have any questions, please call me or J. L. Embley at (910) 675-5774.
Sincerely, Original signed by R. J. Reda, 12/15/95 R. J. Reda, Manager Fuels and Facilities Licensing (910) 675-5889, MC J26 cc: W. J. Sependa J. L. Embley
___..m 4
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GE Nuclear Energy a=wa c
y r.o.s,n ea. wona,en.n. w senoa June 28,19%
RJR-96-071 MFN-088-%
Don===t Control Desk US Nuclear Rpey Comnussion Wastungeon, DC 20555-0001 Atteme==
- R. C. Jones, Jr., Chief Remerar Systems Branch Subjea:
Reporting of Changes and Errors in ECCS Evaluation Models Reference.
Latter, J. F. Klapproth to the Documet Control Desk (R. C. Jones, Jr.), ReporWng ofChanges andErrors in ECCS Evaluation Models, dated June 24,1995 (MFN-087-95), and revised by Imer, R. J. Rada to the Document Control Desk (R. C.
Jones, Jr.), Reporting ofChanges and Errors in ECCS Evaluation blodels, February 20,19% (MFN-020-96).
. De purpose of this letter is to report, in accordance with 10 CFR 50.46 (a) (3) (ii), the impact of changes and errors in the methodology used by GE to demonstrate compliance with the Emergency Core Coohng System (ECCS) requiremenu of 10 CFR 50.46. His report covers the penod from the last report (Reference) to the present. It is noted that Peak Cladding Temperature (PCT) vanatmos resulting from plant specific system or fuel changes are not addressed in this letter. These should be treated. as appropriate. on a plant specific basis in accordance with other sections of 10 CFR 50.
Dere have been no changes or errors identified for the SAFE /REFLOOD model described in NEDE 20566-P-A. Analytical Afodelfor Loss-of-Coolant Analysis in Accordance with 10 CFR SO
' Appendix K.
Here have been no changes or errors identified for the SAFER /GESTR model desenbed in NEDE 23785-1-P-A, The GESTR-LOCA and SAFER Afodelsfor Evaluation of Loss-of-Coolant Accidents. and NEDE 30996-P-A, SAFER Afodelfor Evaluanon ofLoss-of-Coolant Accidentsfor Jet Pump andNon-Jet Pump Plants.
Dunng the reporting period an error was discovered in some applications of the GE LOCA evaluation model SAFER /GESTR. It was determmed that in some analyses cases an algorithm used to compute the number of fuel rods in a BWR lattice was incorrectly specified. As a result, S AFER input prepased in accordance with the automation process may have had incorrect data. The only impact was on the SAFER analyses for fuel designs containing large water rods where the input generation was automated. His finding does not impact plant safety.
%'M "3
- q. 9 3 '
_y Dwst Control Desk a
U. S. Nuclear Regulatory Commission Page 2 This incorrect value for the number of active fuel rods resulted from a specification error in an automated SAFER /GESTR basedeck generation procedure. As a result of this specification error, the SAFER /GESTR beh for those fuel types containing large water rods (GE9/10/11/12/13) enatmaad both an incorrect number of fuel rods and inappropriate values for the bundle flow areas and hydraulic danwters. Calculations performed to assess the significance of this error indicate that the unpact on the calculated claddag temperature is less than 30*F.
Until recently, the limiting fuel types had s et been associated with the large water rod designs and the base decks asnerated with the automated procedure were cormet.
The lav== ^ y was ducovered as part of a normal GE quality assurance review of the SAFER /GESTR analysu for a specific plant with a large water rod limiting bundle. Actions have been taken to correct the pmbian and to ensure that the correct vanable is used in all future applications. It should be noted that the PCT unpact was small compared to the avadable margm to specified limits haeated by the SAFER /GESTR results and no impact on taaie=! specification limits was found All utilmes using these evaluation models have been noti 6ed of this error.
If you have any ?_e, please call me or J. L. Embley at (910) 675-5774.
l Sincerely, l
l l
R. J. Reda, Manager Fuels and Facility Licensing i
(910) 675-5608 l
i cc:
W. J. Sependa J. L. Embley l
ll-i
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