ML18219B003
| ML18219B003 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 06/06/1978 |
| From: | Baer R Office of Nuclear Reactor Regulation |
| To: | Tillinghast J Indiana & Michigan Electric Co, Indiana Michigan Power Co, (Formerly Indiana & Michigan Power Co) |
| References | |
| Download: ML18219B003 (21) | |
Text
Docket
<. 50-316 p
Indiana and Michigan Electric Company Indiana and Michigan Power Company ATTN:
Mr. John Tillinghast Vice President P. 0.
Box 18 Bowling Green Station New York, New York 10004 Distribution l2t%
NRC PDR JUN 6 1978 Local PDR LWR ¹2 File EGCase RSBoyd RCDeYoung DBVassallo BScott OELD IE (3)
BJones BScharf ACRS (16)
CNi les Gentlemen:
JRBuchanan, NSIC TBAbernathy, TIC
SUBJECT:
ORDER FOR MODIFICATION OF LICENSE
{Donald C.
Cook Nuclear Plant Unit 2)
MMMlynczak KKniel TNovak RJNattson ZRosztoczy DRoss ASchwencer EReeves DNeighbors Enclosed is a signed original Order for Modification of License, dated June 6, 1978, issued by the Commission for the Donald C. Cook I'nuclear Plant Unit 2.
This Order amends Facility Op6rating License No.
DPR-74 by modifying the Technical Specification limit for the total nuclear peaking factor
{FA) to 2.11.
This Order also requires"submittal of a corrected ECCS an8lysis as soon as practicable.
A copy of the Order is being filed with the Office of the Federal Register for publication.
Sincer ely, Original signed by Robert Boer Robert L. Baer, Chief Light Hater Reactors Branch No. 2 Division of Project Management
Enclosure:
Order for Modification of License t
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& Michigan Electric Company Indiana, & Michigan Power Company June 6, 1978 cc:
Mr. R.
W. Jurgensen Chief Nuclear Engineer American Electric Power Service Corporation
. 2 Broadway New York, New York 10004 Gerald Charnoff, Esquire Shaw,.Pittman, Potts
& Trowbridge l800 M Street, N. W.
Washington, D. C.
20006 Mr. David Dinsmore Comey Executive Director Citizens for a Better Environment 59 East Van Buren Street Chicago, Illinois 60605
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of
)
)
INDIANA 5 MICHIGAN ELECTRIC COMPANY
)
INDIANA 8( MICHIGAN POWER COMPANY
)
)
(Donald C.
Cook Nuclear Plant, Unit No. 2)
Docket No. 50-316 ORDER FOR MODIFICATION OF LICENSE I.
The Indiana 8 Michigan Electric Company and Indiana 8 Michigan Power Company (the licensees),
are the holders of Facility Operating License No.
DPR-74 which authorizes the operation of the nuclear power reactor known as Donald C.
Cook Nuclear Plant, Unit No.
2 (the facility) at steady state reactor power levels not in excess of 3391 megawatts thermals (rated power).
The facility, usin'g a Westinghouse Electric Corporation designed pressurized water reactor (PWR), is located at the licensees'ite in Berrien County, Michigan.
In accordance with the requirements of the Commission's ECCS Acceptance Criteria 10 CFR 550.46, the licensees submitted on April 1, 1977, an ECCS evaluation for Unit No.
2 for proposed operation using 17 X 17 fuel manufactured by the Westinghouse Electric. Corporation.
This evaluation included limits on the peaking factor.
The ECCS performance evaluation submitted by the licensees was based upon an ECCS evaluation developed
by the. Westinghouse Electric Corporation (Westinghouse),
the designer of the Nuclear Steam Supply System for this facility.
The Westinghouse ECCS Evaluation Model had been previously found to conform to the requirements of the Commission's ECCS Acceptance Criteria, 10 CFR Part 550.46 and Appendix K.
The evaluation indicated that with the peaking factor limited as set forth in the evaluation, and with other limits set forth in the faci lity's Technical Specifications, the ECCS cooling performance for the facility would conform with the criteria contained in 10 CFR 550.46(b) which govern calculated peak clad temperature, maximum cladding oxidation, maximum hydrogen generation, eoolable geometry and long-term cooling.
On March 23, 1978 Westinghouse informed the Nuclear Regulatory Commission (NRC) that an error had been discovered in the fuel rod heat balance equation involving the incorrect use of only half of'the volumetric heat generation'ue to metal-water reaction in calculating the cladding temperature.
Thus, the LOCA analyses previously submitted to the Commission by licensees of plants with Westinghouse reactors were in error.
The staff promptly determined that no immediate action was required to assure safe operation of these plants.
3 The error identified would result in an increase in ca culated peak clad temperature, which, for some plants, could result in calculated tempera-tures in excess of 2200'F unless the allowable peaking factor was reduced somewhat.
Westinghouse identified a number of other areas in the approved model which Westinghouse indicated contained sufficient conservatism to offset the calculated increase in peak clad temperature resulting from the correction of the error noted above.
Four of these areas were generic, applicable to all plants, and a,number of others were plant specific.
As outlined in the attached Safety Evaluation Report (SER), the staff deter-mined that some of these modifications would be appropriate to offset to some extent the penalty resulting from correction of the error.
The attached SER sets forth"the value for each modification applicable to each facility.
Revised computer calculations correcting the error, noted above, and incorporating the modifications described in the SER have not been run for each plant.
However, the various parametric studies.that have been made for various aspects of the approved Westinghouse model over the course of time provide a reasonable basis for concluding that when final revised
/
calculations for the facility are submitted using the revised and corrected model, they will demonstrate that with the peaking factors set forth in the SER operation will conform to the criteria of 10 CFR
~ 50.46(b).
Such revised calculations fully conforming to 10 CFR 50.46 are to be provided for the facility as soon as practicable.
4 As discussed in this Order and in the SER, operation of the Donald C.
Cook Nuclear Plant Unit No.
2 at the peaking factor limit specified in this Order, will assure that the ECCS will conform to the performance require-ments of 10 CFR 550.46(b).
Accordingly, these limits provide reasonable assurance that the public health and safety will not be endangered.
Upon notification by the NRC staff, the licensees committed to provide a
reevaluation of ECCS performance as promptly as practicable to limit operation to achieve a peaking factor not exceeding the values specified herein.
The commitments were confirmed by the licensees'etter of April 6, 1978.
The staff believes that the licensees'ction, under the circumstances, is appropriate and that this action should be confirmed by NRC Order.
II:I:
Copies of the Safety Evaluation Report and the following documents are available for inspection at the Commission's Public Document Room at 1717 H Street, Washington, D.
C. 20555, and are being placed in the Commission's local public document room at the Maude Preston, Palenske Memorial Library, 500 Market Street, St. Joseph, Michigan 48975.
(1)
L'etter from Westinghouse to NRC'dated April 7,'978.
(2)
Letter from Indiana. 5 Michigan Power Company, to Mr. Edson G. Case, Office of Nuclear Reactor Regulation, dated April 6, 1978.
(3)
Letter from Indiana 8 Michigan Power Company, to Mr, Edson G. Case, Office of Nuclear Reactor Regulation, dated April 18, 1978.
Accordingly, pursuant to the Atomic Energy Act of 1954, as amended, and the Commission's Rules and Regulations in 10 CFR Parts 2 and 50, IT IS ORDERED THAT Facility Operating License No.
DPR-74 is hereby amended by adding the following new provisions:
(1)
As soon as practicable, the licehsee,shall submit a reevaluation'f ECCS cooling performance calculated in accordance with the Westing-house Evaluation Model,- approved by the NRC staff and corrected for the errors described herein.
(2)
Until further authorization by the Commission, the Technical Specification limit for total nuclear peaking factor (F~) for the D, C. Cook Nuclear Plant Unit No.
2 shall be limited to 2.11.
FOR THE NUCLEAR REGULATORY COMMISSION Dated at Bethesda, Maryland this 6th day of June, 1978 Roger S.
Boyd, Director Division of Prospect Management Office of Nuclear Reactor Regulation DFPICE~
SURNAME~
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t UNITEO STATES NUCLEAA REGULATORY COiMM)7blOM WASHINGTON, D. C. 20555, SAFETY EVALUATIOit BY TllE OFFICE OF tlUCLEAR REACTOR REGULATIOt<
SUPPORTI!IC ORDER FOR tiODIFICATIOlt OF LICEt/SE RELATEO TO ERROR Ilk 't'ESTIt!CHOUSE ECCS EVALUATIOt! tlODFL Intvoducti on llestinghouse was infor)))ed on March 21, 1978 by one of their licensees that an error had been discovered in tt)eir ECCS Evaluation tlodel.
This el ) or was co!'I!I)on to both the, bl o)idown and heatup codes.
'llestinghouse deter)i1ined by analyses that the fuel rod heat balance equation in the LOCTA IV u'ATAtl VI codes was in error and that the LOCA analyses previously sut)mitted by their customers were incorrect and predicted peak clad temperatures (PCT's) )Il)ich were too low.
Ilestinghouse determined that only hal f of the volum tric heat-generation due to metal-water reaction was used in calculating the cladding temperatures.
Th)IS an I)n. e.iewnd safety:!unstinn existed since preli!...inary I stimates indicated that so5)e plants would ))ot meet the 2200'F li)",itof 10 CFR 50.46 at the calculated
).Iaxir)um overall peaking factor limit. Westing-
'house notified their cI)stoners and tlRC on triarch 23, 19/0 while the utilities notified HRC through the regional Offices of Inspection and Enforce))) nt.
Promptly upon notification by 'testi>>qhouse, the l!RC staff assessed the immediate safety significance of this information.
!Ae noted certain points that indicat..d no i.)I~edate action was required to'ssuve safe operation of tl)e plants.
First, nost plant". op rat at a peaking factor significantly below the max)!)u::" peaking factor used for safety calculatiOns.
By making safety comp<<tations at factors higher than actual operating levels, the facility has a v)ide range of flexibility, wl ttlout th need for hour to hou) vecoi')putations of core status.
The difference between the actual peal;ing factors nd the EIaxi))ur) calculated peaking factovs, for."ost plants, would offset the penalty resulti))g froh the correction of the evror.
Second,'or
','.)ost reactors there are
V n
a number of very plant-specific parameters which bear upon aspects of the ECCS performance calculations.
Utilities do not generally take credit for these plant-specific parameters preferring to provide a
simpler computation iihich conservatively disregards these individually small credits.
Third, the error in tlie Westinghouse computations relates to the zirconium-water reaction heat source.
This is an aspect of Appendix K, which is generallv recognized to be very conservative.
, Hew experimental data indicate that the methods required by Appendix K appreciably over estimate the heat source. Thus,'hile the error in fact entails a deviation from a specific requirement of Appendix K, it does not entail a matter of immediate safety significance.
Westinghouse continued to evaluate the impact of the error on previous plant specific LOCA analyses arid performed'coping calculationS, sensitivity studies and some plant-specific reanalyses.
In addi tion,
. Westinghouse investigated several modifications to the previously approved methods which if approved by the HRC staff would offset some of the immediate impact of the error on Technical Specifications limits and on the plants operating flexibility.
On t1arch 29,
- 1970, Westinghouse and several of their customiers met with members of the HRC staff. in Bethesda.
Westinghouse d scribed i" detail the origin of the error, expiained iiow ii affected the LOCA analyze~,
and liow the error had been corrected and characteri.zed its affect on current plant specific analyses.
In order to avoid reduction in the overall peaking factor (Fq), Westinghouse presented a des'cription of three proposed ECCS-LOCA evaluation model modifications which would contribute a compensating reduction of PCT.
They were characterized as follows:
1.
Revised FLCCiiT 15 x 15 Heat Transfer Correlation This new reflood heat transfer correlation which had been recently developed and sub",.iitted by Westinghouse in Reference
( 1) was prop'osed as a replacement for the currently approved FLECHT correlation.
To determine the benefit, the propos d correlation was incorporated into the LOCTA IV heatup code and >>as,fou!id to result in improved heat transfer during the reflood portion of the LOCA.
2.
Revised Zircalo Emissivitv Based on recent.EPRI data (Reference 2), Westinghouse proposed to modify the presently approved equation for Zircalov cladding emissivity to a constant value of 0.9.
The higher emissivity (previously below 0.8) provides increased radiative heat transfer from the hot fuel pin during the steam cooling period of reflood.
3.
Post-CHF Heat Transfer Westinghouse proposed to replace their present post-CHF transition boiling heat transfer correlation with the Dougall-Rohsenow film boiling correlation (Reference 3)~which they stated was included in Appendix K to 10 CFR Part 50 as an acceptable post-CHF correlation.
These three model modifications were classified as generic, applicable to all plant analyses.
Subsequently, as discussed below, these changes were rejected by the HRC staff as providing generic benefi t.
- However, a portion of the credit proposed by Westinghouse
>~as approved by the NRC staff for certain specific plants, which had provided specific calculations with the new 15 x 15 correlation.
Durinq the period Harch 29 to April 18,
- 1978, Westinahouse provided us with acldi!.ional sensi tivity ana',yses
=.pd >la!>t specific analysis in which they evaluated the effects of some changes to plant-specific inputs in the LOCA analyses.
These were as follows:
1.
Assumed Plant Power Level A reduction of the plant power level assumed in the SATAH VI blowdown analyses from 102% of the Engineered Safeguards
'Design Power (ESDR) level to 1025 of rated power was proposed.
Previously, analyses had been performed at approximately 4.5% over the rated power.
This change was worth aproximately 0.01 in Fq,.and is refered to as LFESDR in Table l.
2.
COCO Code Input A modification to the COCO code input (Reference
- 3) to more
.'realistically model the painted containment walls was proposed.
S'ince the paint on containment walls provides additional resistance to heat lo.s into the walls, the COCO code calculates an increase in containment back pressure, which results in a
3.
benefit to the calculated peal: cladding temperature of 0 to 40'F,
.during the reflooding transient.
The magnitude of the benefit is dependent on the type of plant and the heat transfer properties of the paint, and results in up to 0.03 benefit in Fq, and is referred to as
~FCP in Table l.
Initial Fuel Pellet Ten erature A modification of the initial fuel pellet temperature from the design basis to the actual as-built pellet temperatures was proposed.
In the present LOCA calculations, Westinghouse has assumed margins in the intial pellet temperature.
The margin available is plant-specific and ranges from 28'F to 59'F.
Use of the actual pellet temperature rather than the assumed value results in a reduction in pellet temperature (stored energy) at the end of blowdown, as calculated by the SATAH code, of approx-imately 1/3 of the initial pellet temperature margin.
'westing-house has provided sensitivity analyses which indicate that a
37'F reduction in fuel pellet temperature at end of blowdown is worth approximately O.l in Fq.
This is referred to as aFPT in Table l.
Accumulator Vater ~colum". rons."'oration Westinghouse has evaluated the effect on ECCS performance of reducing the accumulator water volume, and has determined that for those plants for which the downcomer is refilled before th.
accumulators are emptied, there is a benefi t in PCT.
The sensitivity studies have indicated that this benefi t in Fq is plant-specific.
l'his is referred to as ~ FACy in Table l.
5.
Steam Generator 1'ube Pluaqinq Consideration In previous analyses, Westinghouse has assumed values of steam generator tube plugging which were. greater than the actual plant-specific degree of plugging.
Sensitivity analyses submitted in Reference 4 were used to evaluate the benefit, available by realistically representing the plant-specific data.
For the plants affected, the benefit in PCT ranged from 7 to 66'F which was conservatively worth from 0.007 to 0.66 in Fq.
This-is referred to as a FSG in Table l.
Discussion and Evaluation The information provided by Westinghouse was separated into two categories; the generic evaluati'on model modifications and the plant-specific sensitivity studies and reanalyses.
The HRC staff reviewed the peaking factor limits proposed by Westinghouse to verify their conservatism.
The metal-water reaction heat generation error in. the Westinghouse ECCS evaluation model was evaluated by us to determine an appropriate interim penalty.
Westinghouse provided two preliminary separate effects calcula-tions which indicated that a maximum penalty of from 0.14 to 0.17 was appropriate to compensate for the model error.
The staff conservatively rounded this penalty up to 0.20.(Reference 5)
Westinghouse also proposed several compensating generic changes in their evaluation model to offset any necessary reductions in peaking factor due to the error.
These changes were assessed by us as follows:(Reference 5) 1.
Ho credit would be given at this time for the changes in the post-CHF heat transfer correlation and new Zircaloy emissivity
- data.
2.
Partial credit (70%) would be given at this time for the use of the new 15 x 15 FLECHT correlation only for plants which had provided a specific calculation demonstrating that such credit was appr opr iate.
Based on this review we developed recommended interim peaking factor limits for all the operating plants and decided that any other plant-specific interim factors (benefi ts) not related to the generic review should be considered separately.
In addition, the staff reviewed plant specific reanalyses for OC Cook Unit Hos.
1 and 2, Zion Unit Hos.
1 and 2
and Turkey Point Unit Ho.
3 which had corrected the error in metal-water reaction.
In these analyses the Dougall-Rohsenow and Zircaloy emissivity credits. were not consi dered, while the new 15 x 15 FLECHT correlation was included.
He concluded that these reanalyses could serve as a basis for conservatively determining interim peaking factor limits for thos plants.
For most of the operating plants our generic. review resulted in a lower allowable peaking factor 'than Westinghouse had proposed; However, in one case, Westinghouse had proposed ri>ore limiting peaking factors in order to prevent clad temperatures at the rupture node from exceeding 2200'F.
We concluded that it would be properly conservative to use the minimum of these values.
Based or, plant-specific sensitivity studies, performed by Westinghouse, the. licensees have submitted requests for interim plant-specific benefits.
We reviewed these sensitivity studies and recommended that appro-priate credits be accepted.
The results of these analyses are shown in Table 1.
We informed each licensee by telephone on April 3, 1978, that they should admir}istratively reduce the plant's peaking factor limit from the limit contained in the Technical Specifications to the ~interim peaking factor limit contained in the right hand column of Table l.
In those cases where the limit in Table 1 is 2.32, this represents no change from the Technical Specifications limit.
The peaking factor limit.of 2.32 is.
generally suppor+ed and approved for Westinghouse reactors employing constant axial offset control operating procedures (Peference 6).
For the reactors having an interim peaking factor limit of 2.31, we requested no further justification of the limit.
This is because the generic analysis supporting the limit of 2.32 approaches the limit only at beginning of the first cycle.
Since the affected reactors have operated past this point, it is clear t'hat the maximum attainable peaking factor will be less than 2.32.
While this margin has not been quantified, we are convinced it is substantially greater than the 0.01 for which we are reouiring no additional justification from the plants with an interim limit of 2.31.
'or the reactors with an interim limit less than 2.31 we requested that
+he licensee furnish administratively imposed procedures to replace Technical Specifications either:
1.
To provide a plant specific constant axial offset control analysis of 18 cases of.load following which would ensure that the interim limit would not be exceeded in normal operation of the power plant, or, at its option, if such analysis were unobtainable, inappropriate or insufficient, 2.,
To institute procedures for axial power distribution monitoring of the interim limit using a system designed for thi s purpose.'f such systems do -not exist manual procedures could be used as indicated in our Standard Technical S'pecifications 3/4'2.6 and ancillary Specifications.
i
We requested the licensees to confirm by letter that they have adopted the above interim LOCA analyses, interim peaking factor limits and administrative procedures by April 10, 1978, if their reactors were operating, and by April 17, 1978, if the reactors were not operating.
Conclusion We conclude that when final revised calculations f'or the facility are submitted using the revised and corrected model, they will demonstrate that with the peaking factors set forth herein, operation will conform to the criteria of 10 CFR 50.46(b).
Such revised calculations fully, conforming to 10 CFR 50.46 are to be provided for the facility as soon as possible.
As discussed
- herein, the peaking factor limits specified in the particular Orders or Exemptions issued for the affected facilities, with operating surveillance requirements, as applicable, specified in Orders or Exemptions for particular plants, will assure that the ECCS will conform to the perfor-mance requirements of 10 CFR 50.46(b). Accordingly, limits on calculated peak clad temperature, maximum cladding oxidation, maximum hydrogen genera-tion, eoolable geometry and long term cooling provide reasonable assurance that the public health and safety will not be endangered.
Date:
June 6, 1978
References 1.
R.
S. Dougall, tl. tl.'Rohsenow, "Film Boiling on the Inside of
. Vertical Tubes with Upward Flow of the Fluid at Low gualities",
't1IT Report 9079-26, September 1963.
2.
EPRI Report !!P-525, "High Temperature Properties of Zircaloy-Oxygen Alloy", tldrch 1977.
- 3.
WCAP-9220, "Llestinghouse ECCS Evaluation tlodel, February 1978 Version", February 1978.
I 4.
1(CAP-8986 - "Perturbation Technique For Calculating ECCS Cooling Performance",
February 1977.
5.
tlemorandum; Posztoczy to Eisenhut and Ross, "tietal-plater Reaction, Heat Generation Error in 'llestinghouse ECCS Evaluation Hodel Computer Program,'pril 7, 1978.
6.
T. t!orita, et al.,
"Power Distribution Control and Load Following Procedures,"
)!CAP-8385 (Proprietarv) and 1!CAP-8403 (Won-Proprietary),
September 1974.
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TABLE 1
Fq Analysis
~FFLECH OI:0 FPCT FSE Fq,MIN <<ESOR aFCP aFPT AFSG aFACV Fq LIMIT Pt.
Beach 1
Pt.
Beach 2
'inna Kewaunee Prairie Island 1/2 3 Loon fiorth Anna Beaver Valley.
Farley Surry 1
Surry 2
Turkey Point 3
Turkey Point 4
~4Loo Indian Point 2 Indian Point 3 Trojan Sa1em 1
Zion 1/2 Cook 1
Cook 2 20?5 202S 1972 2172 2187 2181 2041 1991 2177 2I77 2019'195 2086 2125 1975 2135 Iu9*'161'190'.32 2.32 2.32 2.25 2.32 2.32 2.32 2.32 1.85 1.85 1 ~ 90 2.05 2.32 2.32 2.32 2.32:
2.07 1.90 2.10
.16
.16
.26
.03
.01
.02
.15
.24
.02
.02
.14
.00
.11
.07
.26
.06
.03
.01
~ 2
~ 2
~ 2
- 2
~ 2
~ 2
~ 2
~ 2
~ 2
~ 2 0
~ 2
~ 2
~ 2
~ 2
~ 2.
0 0
0
~\\
.05
- 05
.06
.06
-.03
,05
.06
-.03'.03 0
2.28 2.28 2.32 2.13 2.18 2.14 2.27 2.32 1.73 1.73 2.01 1.90 2.23 2.25 2.32 2.18 2.04 1.90 2.11 2'2 2.32 2.32 2.2S 2,26
- 2. 32 2.32 2.32 1.84 1.84 2.05 1.91 2.23 2.19 2.32 2.32 1.98 2,28 2.28 2.32 2,13 2.18 2.14 2.27 2.32 1.73 1.73 2.01 1.90 2.23 2.19 2.32 2.18 2.04 1.90 2.11
.01
.01
~)
.01 F01
.01
.01
.01
.01
.01
.02
,02
.005
.03
.03
.036
.025
.025
.03
.037
.024
>029
,066
,053
.023
.023
.020
.01 s03 2,32 2I32 2132 2.16 2,24(+)
2.14 2.31 2'2 1.81 1.81 2.03 1.91 2.24 2.23 2.32 2.21 2.04(+).
1.90
- 2. 11 FT
- Credit in Fq for PCT margin to 2200oF limit.
Fz.02
- Metal Mater Reaction penalty on Fq.
FFLECHT Credi t in FII for improvements to 15xl 5 FLECHT Correlation.
FPCT
- Staff estimated Fq based on 2200 F PCT limit:
'S~
- Mestin9house proposed Fq based on stored energy sensitivity studies.
- Denotes reanalysis at Fq old value error corrected.
- Benotes reanalyses at F~ old value, error corrected, accumulator Vol. Change of 100 ft, accumulator pressure of 650
(+) These limits are applicable assuming licensee modif es accumulator conditions as appropriate.
If not. Prairie Island 1/2 Fq 2.21, Zion 1/2 Fq=1.9 psia
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