ML20197D420
| ML20197D420 | |
| Person / Time | |
|---|---|
| Issue date: | 10/30/1985 |
| From: | Voglewede J NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | Vance J VANCE & ASSOCIATES |
| References | |
| REF-WM-53 NUDOCS 8605140265 | |
| Download: ML20197D420 (3) | |
Text
-
V.i Rxa4 File WM Pr$ct N C&t NL PDR V N 3 0 1985 LPDR -. _
WMEG/JCV/ JEAN VANCE DWjMg ;
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_______i
(&tum ta W1 GSS3)
Mr. Jean N. Vance Vance & Associates P.O. Box 997
{
Ruidoso, New Mexico 88345
Dear Mr. Vance:
It was a pleasure meeting you curing your 30 September 1985 visit to the NRC in Washington, D.C.
The description of your waste stream analysis program IS0 SCALE, c
was both interesting and informative.
L To summarize my conclusions and those of other members of the NRC staff, your computer program appears -to have' the greatest potential benefit in (1) providing a better understandfr.g of the waste stream process at a given plant, and (2) possibly providing an integrated inventory of waste streams (i.e., an accounting or curie comparison of the radionuclides generated and the radionuclides deposited within various waste media). However, we do not believe that such a system is an effective substitute for sampling individual waste streams. We are also concerned that a great deal of expertise provided by your company is not actually emoedded in the program software. The value of your system appears to lie largely with the fact that you, rather than a utility, are interpreting the results. Therefore, there is a potential for blindly using scaling factors and other calculated results of the system in the absence of your direct support.
During your visit we discussed a nutoer of documents describing the release of radioruclides from the fuel. The conclusions of the American Nuclear Society Standards Committee Working Group ANS-5.4 are provided in the ANSI /ANS-5.4 Standard (Ref. 1). A copy of this document was provided to you during your visit.
A ' supporting document, NUREG/CR-2507 (Ref. 2), is enclosed with thi's letter.
The acti.vity of the ANS 5.4 Working Group, which addressed the release of radio-nuclides from the fuel to the fuel-cladding gap, ceased with the publication of the ANSI /ANS-S.4 Standard. A similar working group, ANS_5.3, is pursuing a-related issue (Fission Product Release to the Coolant of Light Water Reactors from Failed / Defective Fuel) and may be of more interest to you. Dr. Stan Turner, o
(
the chairman of both the ANS-5.3 and ANS-5.4 Working Groups, has informed me that ANS-5.3 is awaiting the results of an EPRI-sponsored contract with Battelle Pacific Northwest Laboratories (RP-2229-1, Collection and Fonnatting of Data on Reactor Coolant Activity and Fuel Rod Failures) before proceeding further with their work. Dr. Steve Gehl, the EPRI Program Manager, has informed me that the data collection activities of this contract have been completed and that a report has been submitted to him for review.
e605140265 851030 PDR WASTE JJM-53 PDR
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______:______ g..._:___ ___ ____:____________:__________..:____________:___________
NAME :JCVoglewede :GWRolef
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OCT 3 0 Igg 5 WMEG/JCV/ JEAN VANCE
_2-You may contact either of these individuals at the addresses given below:
Dr. Stan Turner Black & Veatch/ Southern Science Office P.O. Box 10 Dunedin, Florida 33528 (813)733-3138 Dr. Steve M. Gehl Electric Power Research Institute 3412 Hillview Avenue P.O. Box 10412 Palo Alto, California 94303 (415)855-2770 A number of documents that describe the relationship between fuel failures and reactor coolant activity levels were also mentioned during your visit. A review of our files indicates that the majority of these documents are proprietary.
However, I have included some non-proprietary information on this subject that may be of interest. The enclosures from Duke Power Company (Ref. 3) and Virginia Electric and Power (Ref. 4) describe the assumed relationship between fuel failures and reactor coolant activity under accident conditions. The calculation of releases under these conditions (Ref. 5) is commonly based on the use of escape rate coefficients. These calculations are routinely ignored in license applications in favor of Regulatory Guide assumptions (Refs. 6-9),
but there has been renewed interest in this area because of the TMI-2 post-accident monitoring requirements. As an example of the work reported under normal operating conditions, I have included a letter from Portland General Electric (Ref.10).
Should you require additional information concerning these comments, please contactmeat(301)427-4275.
Sincerely, ORigDEL SimR@ BT John C. Voglewede Engineering Branch Division of Waste Management OFC :WMEG
- WMEG
- WMEG
'N#4E :JCVoglewede :GWRoles
- TCJohnson DATE :10/ /85
- 10/ /85
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REFERENCES 1
"American National Standdrd Method for Calculating the Fractional Release of tolatile Fission Products fror Oxide Fuel," American Nuclear Society Stendard ANSI /ANS-5.4-1982.
- 2. American Nuclear Society Working Group 5.4 (Compiled by Southern Science Applications, Inc.), Background and Derivation of ANS-5.4 Standerd Fission Froduct Release Model, U.S. Nuclear Regulatory Comrission Report NUREG/CP-2507, January 198?
- 3. William G. Parker, Jr. (Duke Power Company) letter to Harold R. Denton
'NFCI dated May 4, 1961.
4 E.R. Sylvia (Virginie Electric and Power Company) letter to Harold R.
Denton (NRC) on "Supplen entary Informiation for Proposed Technical Specifications Change" dated March 26, 1981.
c Core Performance Bre' ch, "The Pnle of Fission Cas Release in Reactor 4-ticensing." U.S. Nucleer Regulatory Commissien Report NUREG-75/077, November 1977.
- 6. " Assumptions Lsed for Evaluating the Potential Padiclogical Consequences c' a Loss of Coolant Accident for Boiling Water Reactors," U.S. Atomic Energy Commission Regulatory Guide 1.3, Revision 2, June 1974 7
' Assumptions Used for Evaluating the Potential Radiological Consequences ci a Loss of Coolant Accident for Pressurized Water Reactors," U.S. Atomic Energy Comnission Regulatory Guide 1.4, Revision 2, June 1974
- 8. "f ssumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boilirg and Pressurized Water Reactors," U.S. Atomic Energy Commission Safety Guide 25, March 23, 1972.
- 9. " Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors," U.S. Atomic Energy Commission Regulatory Guide 1.77, May 1974
- 10. C.P. Yundt (Portland General Electric Company) letter (CPY-796-80) to R.H. Engelken (NRC) dated August 18, 1980.
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DUKE POWER COMPANY Powra Best.oswo 422 Socin Cnuncu Srazzr. CnAntoriz. N. C. asa4a wi t um u o
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May 4, 1981
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"C' Mr. Harold R. Denton, Director 1*' I /4 g
Office of Nuclear Reactor Regulation U
U.S. Nuclear Regulatory Co= mission p
j, Washington, D. C.
20555 ga% y O
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N Attention:
Ms. E. Adensam, Chief
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Licensing Branch No. 4 4
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, ;Il Re: McGuire Nuclear Station Docket Nos. 50-369, 50-370
Dear Mr. Denton:
Attached are ten copies of McGuire procedure, AP/0/A/5500, " Estimate of Failed Fuel Based on I-131 Concentration". This is one of the implementing pro-cedures for the McGuire E=ergency Plan and as such should be included with the other i=ple=enting procedures previously submitted on February 13, 1981.
By copy of this letter, three copies of this implementing procedure are being provided to NRC, Region II.
Very truly yours, 1.
/
Win ism O. Parker,.Jr.
GAC:p.r Attactment cc:
M. J. Graham (w/o attach.)
Mr. J. P. O'Reilly, Director (w/3 cys.)
Resident Inspector U.S. Nuclear Regulatory Cocaission McGuire Nuclear Station Region II 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 300/
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~ * (1)"D ' NO. s'AP /D /N/ 5 PROCEDURE PREPARATION Changa(s).
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PROQURE TITLE:
Estimate of Failed Fuel 3ased on I-131 Concentration PFJJARED 3Y:
Ps Od.$. Utf e,, b DATE:
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I DATE:
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Cross-Discipid"ry Revd ew By:
,s.
V TDIPORARY IJPROVAL (I? NECESSARY)
EY:
(SRO)
DATE:
3Y:
DATE:
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A2?EOTD 3Y:
DATE:
/
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.T S CE* ~.A.YEOUS :
i Reviewed / Approved By:
DATE:
3 i'i/F1 s
Reviewed / Approved 3y:
DATE.
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NUCLEAR SAFETY EVALUATION CHECK LIST (1) STATION:
Mc Go.,
1 -x = r@C.'..W * * "uFmW UNIT:
F. i..-
OTHER:
(2) CHICK LIST APPLICA3LE TO:
A P / of Al gSbe.l 31 (3) SAFETY EVALUATION - PART A The item to which this evaluation is applicable represents:
Yes Ne K A change to the station or procedures as described in the T:
i or a test or experiment not described in the FSAR7 l
If the answer to the above is "Yes", attach a detailed description of the ice being evaluated and an identification of the affected section(s) of the FSAR.
(4) SAFETY EVALUATION - PART B l
Yes No X Will this iten require a change to the station Technical Specifications?
If the answer to the above is "Yes," identify the specification (s) affected and/or attach the applicable pages(s) with the change (s) indicated.
l (5) SA::. i EVALUATION - PART C As a result of the item to which this evaluation is applicable:
\\
Yes No X Will the probability of an accident previously evaluated in the FSAR be increased?
Yes No X Will the consequences of an accident previously evaluated in the FSAR be increased?
l Yes No X May the possibility of an accident which is different than any already evaluated in the FSAR be created?
X Will the probability of a malfunction of ecuipnent Yes No important to safety previously evaluated in the ISAR be increased?
X Will the consequences of a malfunction of equipment Yes No i=portant to safety previously evaluated in the FSAR be increased?
)C May the possibility of malfunction of equipment Yes No important to safety different than any already evaluated in the FSAR be created?
Yes No Y Will the margin of safety as defined in the bases to any Technical Specification be reduced?
If the answer to any of the preceding is "Yes", an unreviewed safety question is involved.
Justify the conclusion that an unreviewed safety question.is or is not involved.
Attach additional pages as necessary.
(6) PREPARD BY: Midd Sbdah 1 DATE:
4l*/8/
I (7) REVIEWED BY:
/
3 M DATE:
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/
U (8) Page 1 of l
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DUKE POL *ER COMPA:~i McGUlRE NUCLEAR STA!!ON ESTIMATE OT TAII.ID FUEL BASID ON ;-131 CONCENTRATION 1.0 Sv=otoms 1.1 13MT-48 reactor coolant radiation =onitor has alarmed.
1.2 1DE-18 reactor coolant filter lA radia:1on monitor has alarmed.
1.3 1DG-19 reactor coolant filter 13 radiation monitor has alarmed.
- 1. t.
Any plant condition in which the operator vould suspect failed fuel or vant an esti= ate of the amount of failed fuel.
2.0 I: mediate Action 2.1 Automatic None 2.2 Manual 2.2.1 Obra1= a che=istry sa=ple f :he res:::: ceclan: in crder to deter =ine the I-131 concen: ration of the coolant.
2.2.2 Once the I-131 concen:ra:ica is known for :he reactor coolant de:er '"a which of the following four cases bes: describes
- he presen: fuel conditions.
NOTE:
A.
- he su=bers ch:si=ed by using this procedure are at best, esti= aces only.
B.
All for=ulas ouo:ed are based upon equilibrium full s
power core iodine.
If fuel damage is suspected to have occurred during ti=es of reduced power or near the :i=e of significant power change, the core iodine investory
=us: be compensa:ed accordingly by using E: closure i. 2.
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nis is :he correc:ics fae: r Y.
C.
All values given are norsali:ed to vole =es of coolant a: normal reactor cec' ant system pressure and re=pera-ture.
To correct fer other NC syste= :emperatures or reduced NC sample te=peratures, use Enclosure 4.1.
This is -Jie corree:ica fac:or I.
D.
he decay of T 131 :: I-131 has been negle::ed as e
insignificant in this analysis.
E.
Isdins epiking may cec:: after a shutdown or significant power change. Data from other Wes:inghouse plants has shown that :he iodine spiking process has been observed to occur during i
a period of 1 to 3 days af:er the change or shut-I down. However, the spike seems to peak during. the period from 4 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> af:er the. change.
I-131 concentrations can increase by a factor of 2 to 25 above the equilibri=n levels during :hese times, although an increase over a fac:or of 10 is unusual and would only be seen a: a shutdown.
Increasas l
]
by a factor of 2 to 3 are typical for a significant power decrease (i.e.,100 to 50: power).
Do not misinterpret this te=perary change for fuel failure if there is no other evidence of fuel da= age. Other l
evidence of fuel damage can be constituted by any l
indication of inadequate core cooling, loose par:s indication, high incere thermoccuple indicatien, etc.
F.
If esti=ates for fuel failure are needed for fuel I
conditions other than those covered by the four cases described below, or if more accurate fuel failure data is needed, see See: ion 2.2.7 of :his i
procedure.
G.
The following four cases cever a very broad rznge of core conditions.
Choose :he one that bes: sui: :he existing conditions.
H.
Che=istry samples should be :aken as soon as da= age is suspected.
2.2.3 Case I - Nor=al Operation 2.2.3.1 The condi: ions which per:ain to Case ! - ::ormal operation are as follows:
2.2.3.1.1 Nor=al rese:or operation at any power or shundevn with no unusual condi: ions prior to shu:down. Adequate core cooling has been =aintained.
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2.2.3.2 If th2 abova boet dscerib 3 :ha c=ra c:nditiens, use the following for=ulas to calcula:e the range of failed fuel values.
Evaluate correction factors I and Y by using Enclosures 4.1 and 4.2.
2.2.3.2.1 (Measured I-131 concentration uCI/ml
-3
+ 3.5 x 10 u CI/ml)
I. Y l
= Number of failed pins (Max. expected and i
best esti= ate) 2.2.3.2.2 (Measured I-131 concen::ation uC /ml
-3
+ 4. 9 x 10 u CI/=1)
I Y
~
= Nc=ber of failed pins (Min. expec:ed) 2.2.3.2.3 (Measured !-131 concen::a: ion u CI/ml
+ 1.8 aCI/=1)
IY
= Percent failed fuel (Max. expe::ed and best esti= ate) 2.2.3.2.4 (Measured I-131 concen: ration uCI/c1
+ 2.5 u CI/ml)
IY
= Percen: failed fuel (Min. expec:ed)
NOTI:
Typical values for I-131 concen::ation in uCI/ml for
-2 a nor= ally operating plan: are berveen 1.0 x 10 and' 4.0 x 10 uCI/ml. These values are based on the reae:or coolant I-131 activi:1es experienced by the 2 ion and Troj.
Plants.
2.2.4 Case II - Macroscopic Clad Damage 2.2.4.1 The conditions which per:ain :o Case II -
FMeroscopic clad damage are as follows:
2.2.4.1.1 Nor=al reac:or operation at any ;cuer, or shutdown where se=e =echanical clad failure (i.e., a loose par: =eni:or indica: ion) or a flow induced failure is suspected.
The core has adequate cooling and no significan: fuel overta=pera:ure is observed.
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2.2.4.2 If tha abova b2st d2cerib:s ths cora condi:icus, una tha follcwing for=ules to eniculate th2 rangs of failed fuel values.
Evalua:e corree:1on factors I and T by using Enclosure 4.1 and 4.2.
2.2.4.2.1 (Measured I-131 concentration uCI/ml
-2
+ 5.5 x 10 uCI/=1)
I T
= Number of failed pins (Max. expected) 2.2.4.2.2 (Measured I-131 concentration u CI/ml
-2
+ 16.5 x 10 uCI/=1)
I T
= Number of failed pt s (3es: es:i=a:e) 2.2.4.2.3 (Measured I-131 conce ::a: ion uCI/=1
+ 27.4 x 10' uCI/=1)
= Number of failed pins (Min. expected) 2.2.4.2.4 (Measured I-131 concen::ation CI/=1
+ 27.9 uCI/ml)
I T
= Percent failed fuel (Max. expected) 2.2.4.2.5 (Measured I-131 concentration u CI/=1
+ 83.7 u CI/ml)
= Percent failed fuel (3est esti= ate) 2.2.4.2.6 (Measured.I-131 concentration aCI/mi
+ 139.5 uCI/ml)
I T
= Percen: failed fuel (Min. expected) 2.2.5 Case III - Severe Fuel overta=perature 2.2.5.1 The c=ndi:1ons which per:ain to Case III -
Severe Fuel Overta=perature are as follows:
2.2.5.1.1 TMI type acciden where :here has been 1
an abnormal shutdown and it is suspec:ed tha: :he fuel has been at least partially uncovered for a peri:d :f ::=e grea:e: tha:
a few minutes. Voiding 1: :he core is detec:ed by high incere thermocouple readings and loss of =argin to sa:uration.
Fuel clad oxidation is detected by excess hydrogen in the contai==ent or in the reactor coolant sa=ple; however, no fuel mel:ing is suspected.
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2.2.5.2 If th2 abava bsst doceribas th2 caro conditicus, uso th2 following formulas to calculate the range of failed fuel values.
Evaluate correction factors I and Y by using Enclosures 4.1 and 4.2.
2.2.5.2.1 (Measured I-131 concentration u CI/ml
+ 2.4 u CI/ml)
IY
= Nu=ber of failed pins (Max expected) 1 2.2.5.2.2 (Measured I-131 concentration u CI/ml
+ 2.9 u CI/ml)
IY I
= Number of failed pins (3es estimate) 2.2.5.2.3 (Measured I-131 concentration u CI/ml
+ 3.2 u CI/ml)
I.T
= Nu=ber of failed pins (Min. expected) 2.2.5.2.4 (Measured I-131 concentration a CI/ml
+ 1255 u CI/ml)
IY
= Percen: failed fuel (Max. expected) 2.2.5.2.5 (Measured I-131 concen::a: ion u CI/ml i
+ 1535 :CI/ml)
IY
= Percent failed fuel (Bes: esti= ate) 2.2.5.2.6 (Measured I-131 concentration u CI/ml
+ 1675 uCI/ml)
IY
= Percent failed fuel (Min. expected) 2.2.6 Case IV - Fuel Melting 2.2.6.1 The condizions which per:ain to Case IV - Fuel Melting, are as follows:
2.2.6.1.1 Severe accident where there has been an abnor=al shutdown and the core is uncovered for a long period of time.
Incore ther-moccuple :empera:ure readings are above 2300*? for a 1cus period of tine.
Fuel mel:ing is suspected (i.e., fuel :empera-ture exceeds 5000 F) and is verified by the inabili:v to operate the incere instru-mentation system properly.
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2.2.6.2' If':h2 ebova b22: daccribes tha caro condi:1' ens, uco tha fo11cwing formulas to esiculcro tha fciled fuel values.
Evaluate orrec:fon fae: ors I and Y by using Enclosures 4.1 and 4.2.
2.2.6.2.1 (Measured I-131 concentration u CI/ml
+ 5.5 uCI/ml)
IY
= Number of failed pins (Best es:1 mate) 2.2.6.2.2 (Measured I-131 concentration u CI/ml
+ 2790 uCI/ml) *I T
= Percent of failed fuel (3es: estimate) 2.2.7 If fuel condi:1ons other than those described above exist, or if a more detailed failed fuel es:imation is desired for either emergency or normal operation, con:act the appropriate Westinghouse people below in the Order listed un:11 contact i
i is nada.
2.2.7.1 Energency Plant Condi:icas - Energency Response Team Westinghouse, Pittsburgh, Pennsylvania 2.2.7.1.1 Dire :or:
Eank Ruppel 412/256-3611 Wo:
i 412/366-6781 Ho:
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2.2.7.1.2 Deputy Direc:er: Ron Lehr 412/256-5401 Woz I
2.2.7.1.3 Technical Suppor: Manager:
Tom Anderson 412/373-5766 Work; 412/327-8289 Ecme 2.2.7.1.4 Materials Design:
Wally Chubb 412/373-4364 Wo 2.2.7.2 Normal Plant Conditions a
2.2.7.2.1 Southern Regi:nal Manager - Steve longdon -
404/885-5900, Work 2.2.7.2.2 Westinghouse - Duka Represen:a:ive - Mike Miller - 412/273-5160, Work 2.2.7.2.3 Materials Design - Wally Chubb - 412/373-4364, Work 1
3.0 Subsecuent Actions 3.1 Follow up as necessary with Westinghouse - Fi::sburgh, Pennsylvania depending on the plant situation.
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4.1 Density Correction Factor, I, for NC Temperature Changes 4.2 Iodine 131 Inventory Correction Factor, Y, for reduced power operation or for times of power change 4.3 Examples a
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5 AP/0/A/5500/33 Inclosure 4.1 Density Correction Factor, I, for NC Temperature Changes Find the appropriate NC System temperature at the time of accident.
Find the approxi= ate temperature at which the NC samples are taken. The intersection of both numbers is the density correction factor, I.
NOTI:
Normal NC System sample temperature is approximately 90 F.
Use this temperature if no other infor:ation is available.
NCS Sample Te=perature ?
80 90 l
100 g
100
.996 998 1
150
.983 985
.987 E
200
.966 968
.970 I
E 250
.945 947
.949 a
92 300
.921 923
.924 s
- 3 $
350
.894 895
.897
!b 400
.862 864
.865 4
e o e
$. E 450
.827 82S
.830 1.
200
.787 755
.790 g:
.5 ]
550
.739 74 0
.74 1 5*
560
.728 729
.731 j*
570
.717 718
.719 E
580
.706 708
.708 c*
590
.693 694
.695 600
.680 681
.683 l
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AP/0/A/5500/33.2 Iodine 131 Inventory corree:1on, Y, for Reduced Power Opera:1on or for Times of Power Change Si:uation 1 To conect for core Iodine Inventory if fuel damage is suspected to have occun ed during times of any power level except 0 where the power level has not changed greater :han 110% within the last 22 days, use the following ecuation.
100 i
T=
l Full Power at time of failure where Y is the correction factor to be used in Section 2.0.
7 ple: The plan: has been at 35% full power for the las: 30 days when fuel damage is suspected. Therefore:
100 Y=
= 2.86 35 Situa:ic= 2 To correct for core iodine if fuel da= age is suspected to have.cccun ed at ti=es c:her :han fi: Situatica 1 above, use :he following equa:1on.
4 100 cid power level in (e I ) + new power level in : (1-e I)
E where:
Y = correction fac:c to be used in Section 2.0
- ld pcwer level in *; = the full power before the power change new power level in : = :he : full power after :he power change at which
- ime :he fuel failure has occuned A
1 1 = is the decay constan for 1 1
131 which equals.0864 day 4
= is :he median time to nake a power change plus the time af:er :he power change until damage is suspected to have occurred, in days.
Exa=ple:
If it took 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to =ake a power change and da= age vas suspec:ed
, 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> af:er the power change.
9
=g-10=11 hours i
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i AP/0/A/5500/33 Inclosure 4.3 Examples Problem 1 a.
Power level has been decreased from 85: to 50%.
b.
This power change took four hours and occurred between 1200 and 1600.
T at 50: is 570 s.
AVG c.
A: 1800 a loose par: moni:or alarm goes off indica:ing a loose objec:
in.he core.
The reactor is no: tripped.
i d.
A Chemistry team is i=medir.:aly dispatched to take a sample NC System as failed fuel is suspec:ed.
a.
Che=is:ry sample indica:es I-131 concentration is 10.0 aCI/ml.
Par: 1.
Deter =ine the bes: esti= ate of the nu=ber of failed pins.
Par: 2.
Deter =ine the bes: es:i= ate of percen: failed fuel.
1 Selution n is is Cas'e II, Step 2.2.4 Use ecus: ion 2.2.4.2.2 for Par: 1 Use equazion 2.2.4.2.5 for Par: 2
[ Measured I-131 concen::a:1en ac/=1\\
.ar:
. I Y = S.c=ber of failed pins
(
16.5 x 10 u CI/=1 l
Deter =ine I:.1 T
is 570 ? at 50%.
AVG Assume NCS Sa=ple Te=perature is 90 ?
Therefore, I =.718 De:er=ine Y:.0 1
1 4
A
.0864 day -
=
7 1
(4j + (2) = 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
- =
2 Remember, t is the median time to =ake a power change plus the difference herwee:
- he :i=e when :he damage is suspected and the ti=e the new power level is reached.
dav Conver: : to days
- = 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> x
.16 davs
=
hrs.
1 9_-..____7-,_
.~,
,-,..,m y
~
Fog 2 2 of 8 i
AP/0/A/5500/33 l.3 Examples 100
-l)(.167 day)) + 50 [i,l-ef(.0864 day-1)(.167 day)];
C 85 - (.0864 day l
e 100 Y=
= 1*183 85(.9857) + 50 (.0143) 10 u CI/ml Par: 1.
(.718) (1.183) = 51.5 =52 failed pins Answer
-2 16.5 x 10 u CI/ml i
I Measured I-131 Conces::a:ic= uCI/=1
- 1..,
I Y=
failed fuel 83.7 u CI/ml i
10 u CI/b=1 (.718) 1 (1.183)
= 0.1% failed fuel 53.7 u C A swer i
9 s
Pago 3 of 8 AP/0/A/5500/33 Inclosure 4.3 Examples Problem 2 The reactor has just tripped instantly from 100% power due to a a.
malfunctioning i=s:rument. There were no unusual conditions prior to the trip.
o b.
TAL, is now 557 ? at 0 power.
c.
The operator, while having no reason to suspec: failed fuel, is curious about the a=ount of failed fuel present now following the trip.
d.
A Che=istry team is sent :o take an NC sample 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af:er the trip.
e.
he Chemistry sa=ple gives an I-131 concentration of 2.0 x 10 uCI/ml.
~
(A typical value for e normally operating plant.
See Note under Case I, Step 2.2.3.2.)
1
)
f.
Che=istry personnel also indicate that NC sa=ple temperature is 100 F.
Par: 1.
Dete:- *"e the -=vd j
-- expected nu=ber of failed fuel pins, Par: 22.
Deter d"e the maximum expected percent failed fuel in the core.
Selu: ion y,
This is Case I, Step 2.2.3 Use equazion 2.2.3.2.1 for Par: 1
)
Use equation 2.2.3.2.3 for Par: 2 Measured I-131 cencentration uCI/nl
,ar: 1.
I Y = Nus.eer at. failed pins
~3 3.5 x 10 uCI/ml Deter =ine I:.1 NC Temperature is 557 7 at0 NC sample temperature is 100 F i
Therefore, I =.732 l
1 pgm-
..-.e
,y--s_,
.--epwgn 9
?cga 4 of 8 AP/0/A/5500/33 Enclosura 4.3 Examples Determine Y:.2 i
i i
i dav situation 2:
e = 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> x
=.5 days 4
24 hrs.
i l
100 Y=
.00 -l (.0864) (.5) + 0
- [(.0864) (.5)l-l 1-e e
i Y = 1.044 N0YE:
If t = 0 or a sample was taken i=medir.tely, Y = 1.0.
~2 2.0 x 10 uCI/mi (.732)(1.0M)
Part 1.
= 4.4 3
3.5 x 10 u CI/=1 or =4 to 5 failed pins Answer Part 2.
Measured I-131 Concentration u CI/ml 4
- I
- Y =. failed :.uel 1.8 uCI/mi
~.0 x 10~2 u CI/ml
'3
(.732)(1.044) =.0085
- failed fuel 1.8 u CI/=1 "he above nu=bers are indicative of nor. a1 operation.
.usver NCYE:
I-131 spiking =ay be a problem here.
See 5:ep 2.2.2, Note E.
i i
l l
Pass 5 cf 8 AP/0/A/5500/33 Innlosure 4.3 Examples Problem 3 a.
Power level has been between 50% and 65% for the las: 30 days and is presently at 60% a: 1800.
b.
Y is =575 Y at 6'0 power.
yg Itisdesi$adtoseeifanysignificantfailedfuelexistsinthe c.
- core even though no abnormal occurrences have taken place.
d.
At 2200 the same day, a che=is:ry sa=ple is taken of the NC system.
The chemistry sa=ple indicates I-131 concentration is 3.9 x 10
- uCI/ml.
e.
Par: 1.
Determine the bes: estimate of the nu=ber of failed pins.
Par: 2.
Determine the best estimate of the : failed fuel.
Solution 1
i This is case I, Step 2.2.3 Use equation 2.2.3.2.1 for Par 1 Use equation 2.2.3.2.3 for Part 2 Measured I-131 Ccncentration uCI/ml
,1..- 1.
,3 X
Y = nu=,cer c:.. iled pins ta 3.5 x 10 uCI/ml j
Dete=ine X:.1 T
is 575 F at 60* power gyg s
1-Assure NCS sample temp. of 90 F o
Therefore I =.713 Deter =ine Y:.2 Situation 1 Y = 00 1
1.67
=
60 i
1
- - - + -
---y-
-,_,.-,.,m,_
~ - -.
,,,_,m,-.-.-,..-,,7,,
-,-.-,,,.y-
- --9 g
p-
,,.%-y-v e
==
rgg.w--w--
y_c w
Pago 6 of 8 i
AP/0/A/5500/33.3 Examples
-2 3.9 x 10 u CI/ml Par: 1.
(.713) (1.67) = 13.27
_3 3.5 x 10 u CI/ml
=14 failed pins Answer 1
Measured I-131 Concentration uCI/=1
,a.,
I Y = ~ failed fuel 1.8 uCI/ml 3.9 x 10 u CI/ml
(.713)(1.67) =.026 : failed fuel Answer 1.8 u CI/ml The above numbers are acceptable for a normally operating plan:.
i F
e i
8
Pago 7 of 8 AP/0/A/5500/33 j.3 Examples i
l problem 4 The unit has been at 97% power for a month when a depressuri:ation a.
of the NC systes occurs.
b.
The reactor trips.
~
, Heavy vibration is observed in the NC pumps.
c.
d.
Thermocouple temperatures over 1000 F are indicated in the core.
1DfF 48 and 1DfF 13 have gone off.
e.
f.
Safety Injection was delayed and it is suspected the core as uncovered between 30 and 60 minutes before sufficient reactor vessel water level was regained.
g.
The incore instrumentation system is still operable.
h.
The NC sample indicates an I-131 concentration of 3800 u CI/nl.
1.
A Chemistry sample is taken imnediately (within the hour) after the trip.
Part 1.
Determine the m=rimum expected number of failed pins.
Part 2.
Deter =ine the mmv4 expected of failed fuel.
na Solution This is case III, Step 2.2.5 s
Use ecuation 2.2.3.2.1 for Part 1 Use equation 2.2.5.2.4 for Par: 2 Deter =ine I,:.1 NC Temp. ! yg at 0% power is 557*F Assume sample ce=perature of 90 F Therefore, X =.730 3eter=ine T:
Y = 100- = 1.03 9:e y
p w
p---
--s7
i 1
Paga 8 of 8 AP/0/A/5500/33 Enclosura 4.3 Examples 3800 uCI/ml Part 1.
(.730)(1.03) = 1190.5 2.4 uCI/ml l
1191 number failed pins, max. expected Ansu l
=
l i
3800 uCI/ml.
p8-2-
(.730)(1.03) = 2.28: failed fuel, =ax. expected Ans e i
l 7
e e.e
=m rn, y~-
,w,-
p-g
-+w,,
g--,,
r
.--,--,-,--,-w,--w--.wg
m._
o
,L{
~
VIRGINIA EI.ECTRIC AND POWER COMPM nacmwown vsmarwzA sones March 26,1981 Mr. Harold R. Denton, Director Serial No. 195 Office of Nuclear Reactor Regulation N0/RMI:sav U. S. Nuclear Regulatory Commission Docket Nos. 50-280 Washington, D.C.
20555 50-281 50-338 50-339 License Nos. DPR-32 DPR-37 NPF-4 NPF-7
Dear Mr. Denton:
SUPPLEMENTARY INFORMATION FOR SURRY POWER STATION UNITS NO. 1 AND 2 AND NORTH ANNA POWER STATION UNITS NO. 1 AND 2 PROPOSED TECHNICAL SPECIFICATIONS CHANGE In our letters dated May 15,1980 (Serial No. 400) and March 6,1981 (Serial
)
No.109), we proposed changes to the Technical Specifications for Surry Units No. I and No. 2 and North Anna Units No. I and No. 2 to permit an increase in
{
the enrichment limit for new fuel.
In response to a verbal request from a member of your staff, Vepco agreed to provide answers to several questions on the effects of extended fuel burnups on accident source terms and iodine spiking.
These answers are provided in.
If you require any additional information, please contact this office.
Very truly yours,
$C B. R." Sylvia Manager - Nuclear Operations and Maintenance Attachment cc:
Mr. James P. O'Reilly
/n N/
Office of Inspection and Enforcemen 4
o Region II
\\
j Mr. Robert A. Clark, Chief Operat'ing Reactors Branch No. 3 g4 fy""%
If fh I IOSI > d' Division of Licensing Y "henecos, )A Mr. Steven A. Varga, Chief y *\\h[
Operating Reactors Branch No. 1 Division of Licensing
(
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=
PAGE 1
Suestion 1:
What is the effect of burnups above 33,000 GWD/MTU on accident source terms?
Response
Table I shows core activities for Morth Anna 17x17 fuel at 33 and 38 GWD/MTU.
These activities were determined from ORIGIN calculations using the ENDT B IV Library.
The fuel was assumed to be enriched to 4.1% and the plant was assumed to be operated at 2900 MWt for the calculation.
Table II lists the gap fractions calculated with the ANS 5/4 model at burnups of 20 and 38 GWD/MTU.
The Table II calculations are consistent with core activities shown in Table I and the parameters assumed in the calculation of these fractions are also shown in Table II.
In Table I
it can be seen that the burnup increase causes either a
decrease in activity or an insignificant increase I
for most isotopes.
In the case of the noble gas isotopes, either the small inventory present or the small contribution to the total dose result in a
minor impact on accident consequences.
- However, more limiting core activities were assumed in the North Anna TSAR for the noble gases, as shown in Table I.
The fission product core iodine inventories given in Table I are higher than those in the North Anna TSAR due tot a) the use of a 4.1% enrichment b) the use of the ENDT B/IV fission product data library.
The values in the North Anna TSAR were taken from the
- m. m PAGE 2
TID-14844 document.
- However, the overall radiological consequences of postulated accidents should not change to any significant degree if another model update is also used.
The North Anna TSAR used iodine dose conversion factors (DCTs) from TID-14844 uhile current practice is to use those values in the NRC Regulatory Guide 1.109.
The decrease in DCT values for iodines will offset the minor increase in the iodine inventories and thus leave the TSAR results virtually unchanged.
Table II shows a large drop in the gap fractions with burnup. ~
The parameters which drive diffusion of fission gases from the fuel to the gap are fuel temperature and burnup.
Tual temperature reductions with increasing burnup more than i
offset the effect of increasing burnup on diffusion rates.
Tables I and II show that accident source terms (core and gap) either decrease or have an insignificant increase with higher burnups.
In either case, the radiological consequences of the TSAR accidents remain unchanged.
While the specific calculations were performed for a Morth Anna 17x17 core, the minor changes with burnup are representative of both plants and therefore
- Suri, specific calculations are not needed.
Specifically, Surry fuel would exhibit the same phenomena exhibited by the Morth Anna extended burnup calculations.
i.e.
minor increases in the iodines th,at can be offset by louer DCT's and louer gap fractions due to reduced fuel I
1
4'h l
,e
?
/
b PAGE 3
temperatures.
se k
4 f
4
- .=
e o
re"
i o
PAGE 4
Table I Noble Gas and Iodine Core Inventories at Shutdown (Values are in Curies)
Basis Assumed in 33000 MWD /MTU 38000 MWD /MTU North Anna Radionuclide FSAR Noble Gases Kr-45 5.33 x 105 6.02 x los 8.14 x 105 Kr-85m 2.42 x 107 2.33 x 107 3.22 x 107 Kr-87 4.54 x 107 4.34 x 107 6.19 x 107 Kr-88 6.44 x 107 6.17 x 107 8.82 x 107 Xe-133 1.57 x los 1.58 x 108 1.66 x 10' Xe-133m 2.35 x 107 2.35 x 107 4.22 x 10' Xe-135 3.85 x 107 3.74 x 107 4.54 x 107 i
Xe-135m 3.11 x 107 3.17 x 10 7 4
4.45 x 107 Iodines I-131 7.74 x 107 7.84 x 107 7.17 x 107 I-132 1.138 x 10e j,143 x jos 1,og x joe I-133 1.654 x 10' 1.652'x los 1.61 x los I-135
~
1.538 x los 1.534 x 10' 1.46 x los n
y
,e,..,-.-
,7-1 PAGE 5
r s
i g..
TAULE II 1
s
- \\
r Fuel Rod. Gap Fractions For Lead Burnup Fuel Assenbly Discharged at Refueling s
^, Muclide AMS 5/4' AMS 5/4 Model Model 20,000 MWD /MTU 38.000 MWD /MTU I-131 0.086 0.0012 I.133 0.029 0.00038 I-135 0.016 0.00022 1
Xe-133 0.027
0.00036 Xe-135 0.0073 0.000096' Xe-138 0.0013 0.000017 Kr-85m 0.0051 0.000066 Kr-85 0.12 0.080 Kr-87 0.0027 0.000036 Kr-88 0.0041 0.000653 parameters used to calculate fuel rod gap, fractions'of lead assembly Core Average Pouer 5.44 kd/ft Radial Peaking Factor 20,000 MWD /MTU 1.25 38,000 MWD /MTU 1.07 i
i I
(
5
,---r-
+-
ww--
w wt'-
i p
o PAGE 6
TABLE II (CONT)
Fuel Teaperature 1 20,000 MWD /MTU, 1422'K 38,000 MWD /MTU 977'K
~
Gap Traction Model*
AMS 5/4 Taken as fuel centerline temperature at lead rod power.
2 AMS 5/4 model used except the entire fuel assembly was modelled to be at one uniform temperature.
The temperature chosen to conservatively apply this model was the fuel centerline temperature of the lead rod in the assenbly with the burnup noted.
Y O
i
PAGE 7
Suestion 2:
What are the bases for any decontamination factors used in consequence evaluations?
Response
The decontamination factors (DFs) used in the North Anna and Surry TSAR analyses are discussed in Sections 15.4.5.2.3 and 14.4.1.2 of their respective FSAR's.
The correlation used to calculate the North Anna DF was developed from small and large scale tests.
The DF is an exponential function of'the the bubble rise time (or pool depth) and the effective bubble diameter (calculated from the total volume of gas released from the fuel assembly gaps).
With consideration given to the total volume of gas released from a fuel assembly, i.e.,
6.9 SCF for the North Anna 15x15 array (the volume of release would be much smaller for a 17x17 array due to its lower fuel temperatures),
the pool decontamination factor was indicated in the FSAR to be a minimum of 760 for the 26 foot pool.
- However, for conservatism a
DF of 100 was used in the evaluation of the fuel handling accident.
An even more conservative value of 10 was used in the Surry FSAR analysis.
As previously discussed, the gap fractions shown in Table II are much decreased due to the effect of decreasing temperature with burnup.
Consequently, previous margins in the DF and the reductions in the gap fractions offset any uncertainty associated with higher back pressures, bubble size, etc., at higher burnups, and the fuel handling accident analyses presented in the FSARs remain conservative.
i i
.. _ ~
~ ~. - - -
1 -
PAGE 3
Suestion' 3:
What are 1
the radiological consequences of accidents?
Response
- Since the changes in the source terms and gap fractions discussed in the reponses to questions 1
and 2
are insignificant, the radiological consequences of the licensing analyses remain bounding.
e l
.we-
-we, ya
,,,w
=,
+
,e PAGE 9
Question 4 How does the iodine spiking behavior for fuel burnups greater than 33,000 MWD /NTU compare to that for present models?
s
Response
The prevailing theory on the thermal and hydraulic mechanisms producing the iodine spiking phenomena is the result of independent investigations in both the United States 'and abroad.
The theory describes the probable source of the spike inventory as cesium iodide salts which are deposited on the inner surfaces of the fuel rod cladding and to a lesser degree on the outer surface of the fuel pellets.
A fuel rod with a cladding defect will admit reactor coolant liquid to contact the inner surfaces of the fuel rod only when the local power is belou approximately 2 ku/ft.
When reactor coolant enters the fuel rod, it will dissolve the cesium iodide salts deposited there.
The' dissolved cesium and iodine are then free to be transported to the reactor coolant system where the iodine is seen as an iodins spike.
A similar hydraulic mechanism occurs during reactor coolant depressurization wherein the. trapped gases within the fuel rod above and/or below the defect location will be at a higher pressure than the coolant as the reactor coolant system is depressurized.
This creates a driving head to expel the iodine ladden water from the fuel rod thus producing another iodine spike.
These spikes have become known as power spikes and pressure spikes.
1 o
a
~
PAGE 10 Since the source of iodine has been shown to be the fuel rod cladding gap activity, any spike activity resulting from fuel rods with high burnup(i.e.
greater than 33,000 MWD /MTU) would be less than that from defected rods at lower burnup due to the decreasing gap inventory as shown in the answer to question 1.
Additionally, the spiking data upon which the Present quantatative NRC model has been developed does not include local burnup of the defected fuel rods.
It is.not possible to ascertain the burnup or temperature history of the defected fuel rods producing the spikes on which data is available.
Therefore, it is believed that it is inappropriate to attempt to model the iodine spiking Phenomena for variations in burnup.
The only effect of higher burnups on iodine spiking would result from the louer linear power of high burnup fuel rods.
If a
defected fuel rod were at a high burnup level, the linear power of the fuel rod would be closer to 2 ku/ft compared to lower burnup fuel rods.
Thus a smaller decrease in reactor power would be necessary in order to produce an iodine spike of the power spike variety.
This implies tuo observations 1.
UPon reactor shutaoun, the spike contribution from high burnup fuel rods would be produced earlier in the shutdown transient than that contribution from defected rods at lower burnups t
9
,z-:.
o i
J o
PAGE 11 2.
Iodine spikes would be produced with smaller power variations (e.g. smaller load follow suings) if defected rods were at high burnup as compared to defected rods at louer burnups.
(
- However, since the use of the ' iodine spiking phenomena in saf e t'y Analysis Roports and Safety Evaluation Reports is in the area of radiological consequences of accidents, this effect would not affect the results of the analyses.
In accident scenarios, a reactor trip occurs thereby resulting in all fuel rods dropping below 2 ku/ft at roughly the same time.
In this case. high burnup rods would not produce an early spike component.
In conclusion, it is believed that the MRC spiking model remains conservative in dealing with iodine spiking when considering fuel rod defects in high burnup fuel rods.
G I
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Portland General Electric Company Trojan Nuclear Plant P.O. Box 439 August 18, 1980 CPY-796-80 Rainier. Oregon 97048 Mr. R. H. Engelken, Director Nuclear Regulatory com:nission, Region V 1990 North California Blvd.
Walnut Creek, California 94596
Dear Sir:
On May 8,1980, in accordance with the Trojan Plant Operating License, Appendix A, US NRC Technical Specifications, License Event Report No.
80-06 was sent to you regarding abnormal degradation of two nuclear fuel rods. This transmittal is a follow-up report regarding the danaged fuel, its cause, and corrective action.
Sincerely, C. P. Yundt General Manager L
CPY/GGB:bb Attachnents c: LER Distribution
'V 1 e
~
f%
e l
~ ~ ~ ~ ~ ~
- ;,, g,
F
- ~ ~
Backgrcund - Chemictry i
Through Trojans first cycle of operation, radiochemistry data maintained a consistently low reactor coolant system gross activity (.45 uc/ml) and Iodine-131 conceatration (0.01 uc/ml) attesting to very stable fuel clad intergrity/ defect conditions. It was felt that tramp uranium constituted a small portion (approximately 10%) of the activity and that some small fuel defect percentage existed, presumably from manufacturing defects in the fuel. During late March, 1979, the coolant activity changed in a manner suggestive that open, easy-access type fuel clad failures had occurred (I-131/I-133 ratio changed from approximately 0.8 to approximately 0.35s gross activity increased, Iodine-131 increased to 0.02 uc/ml and noble gas increased to 1-2 uc/ml)..
To attempt to quantify the fuel defect percentage, the concentrations of various isotopes in the coolant was compared with the predicted concentrations with 1% failed fuel found in FSAR Table 11.1-2, using the 1% fuel defect as defined as 2.5 uc/ml I-131 and 274 uc/ml I -133 and the activities of I-131 and I-133 in the equation:
1.12*I-131 - 0.079*I-133
% Failed fuel = "
2.35
- Power /100 1
Thus the Iodine-131 and Xenon-133 concentrations yield fuel defects of 0.00824 and 0.00034% respectively and the Iodine ratio showed 0.006% defect.
Before March of 1979, the Iodine-131 indicated less than 0.003% fr.el defect.
It is estimated that a defect in one roo can raise the reactor coolant Iodine-131 activity by as much as a 0.08 uc/ml. After March of 1979, there was at least one more rod failure.
Since April 1979, the activity levels and ratios remained essentially -
constant, indicating no additional fuel exposures had occurred.
In December 1979, during steam generator tube plugging, crud samples were obtained from the steam generators that indicated the presence of fission product isotopes not expected to be found in the coolant.unless fuel was in direct contact with the coolant. Indications of plutonium were found in the coolant..
To attempt to determine which fuel region might have sustained the March 1979 fuel rupture, the ratio of Cesium-137 to Cesium-134 during transient conditions of a reactor trip or power reduction was examined. To relate the defect to fuel region and burnup it must be considered that Cesium-134 builds up slowly as a trans-mutation product of Zenon-133. Cesium-134 is not a fisson product like Cesium-137. Therefore, Cesium-134 builds up more slowly and is related to neutron flux and fuel burnout. Generally, the. ratio of Cesium-137 to -134 i
is very high at beginning of life and very low at and of life. If a fuel I
defect comes from fuel that has only undergone a single cycle, then the Cesium-137 to -134 ratio will normally be greater than or equal to 2.0.
If the fuel has seen additional burnout, this ratio will decrease. After two completed fuel cycles, the ratio will be approximately 0.9 to 1.0 and after three complete fuel cycles the ratio will be 0.6.
Since these values reflect end of fuel cycles only, any value in between may be interpreted based on the neutron l
flux the fuel rods have actually seen. However, multiple defects from j
l 1
f j
... _ _....s "different fuel loads, will cause the ratios to change proportionately to the Only the transient magnitude of the defects in the respective fuel loads. data is used data as a result of the respective production mechanism.
Cesium-138 is'a good indicator of fuel defects during power operation.
Cesium-138 is usually five Cesium-138 is a daughter products of Zenon-138.
times the concentration of Iodine-133 and about 10 times less than the Because of Cesium-138's short half-life Xenon-133 in the reactor coolant.
A new defect of 32.2 minutes, it would indicate a fuel defect only at power.
would be indicated by a rise over approximately a 30 day period of Cesium-138.
New defects have a step-wise increase during approximately a 30 day period.
Tission gases are completely They then show a very stable radio-chemistry.
lost out of the fuel ro.Is no matter what the size of the defect, however, salts such as Cesium, Iodine, and Rubidium are more dependent on the size of Some very small defects the defect since these are washed out of the fuel.
show a high percentage of fission gas and almost no iodine.
O I
I i
i t
t 4
-I
- .~
.w
Backarrund - Chemintry Through Trojans first cycle of operation, radiochemistry data maintained a consistently low reactor coolant system gross activity (.45 uc/ml) and Iodine-131 concentration.(0.01 uc/ml) attesting to very stable fuel clad intergrity/ defect conditions. It was felt that tramp uranium constituted a small portion (approximately 10%) of the activity and that some small fuel defect percentage exis'ted, presumably from manufacturing defects in the fuel. During late March,1979, the coolant activiry changed in a manner suggestive that open, easy-access type fuel clad failures had occurred (I-131/I-133 ratio changed from approximately 0.8 to approximately I
0.35: gross activity increased, Iodine-131 increased to 0.02 uc/mi and noble l
I gas increased to 1-2 uc/ml)..
To attempt to quantify the fuel defect percentage, the concentrations of
' various isotopes in the coolant was compared with the predicted concentrations with 1% failed fuel found in FSAR Table 11.1-2,.using the 1% fuel defect as defined as 2.5 ue/ml I-131 and 274 uc/ml X*-133 and the activities of I-131 and i
1-133 in the equation:
1.12*I-131 - 0.079*I-133
% Failed fuel =
2.35
- Power /100 Thus the Iodine-131 and Xenon-133 concentrations yield fuel defects of 0.00824 and 0.00034% respectively and the. Iodine. ratio showed 0.006% defect.
Before March of 1979, the Iodine-131 indicated less than 0.003% fuel defect.
It is estimated that a defect in one rod can raise the reactor coolant Iodine-131 activity by as much as a 0.08 uc/ml. After March of 1979, there w'as at
~
least one more rod failure.'
Since April 1979, the activity levels and ratios remained essentially constant, indicating no additional fuel exposures had' occurred.
In December 1979, during steam generator tube plugging, crud samples were obtained from the steam generators that indicated the presence of fission product isotopes not expected to be found in the coolant unless fuel was in direct contact with the coolant. Indications of plutonium were found in the coolant.
To attempt to determine which fuel region might have sustained the March 1979 fuel rupture, the ratio of Cesium-137 to Cesium-134 during transient conditions of a reactor trip or power reduction was examined. To relate'the defect to fuel region and burnup it must be considered that Cesium-134 builds up slowly as a trans-mutation product of Xenon-133. Cesium-134 is not a fisson product like Cesium-137. Therefore, Cesium-134 builds up more slowly and is related to neutron flux and fuel burnout. Generally, the ratio of Cesium-137 to -134 is very high at beginning of life and very low at end of life.
If a fuel defect comes from fuel that has only undergone a single cycle, then the Cesium-137 to -134 ratio will normally be greater than or equal to 2.0.. If the fuel has seen additional burnout, this ratio will decrease. After two completed fuel cycles, the ratio will.be approximately 0.9 to 1.0 and after three i
completa-fuel cycles the ratio will be 0.6.
Since these values reflect end of I
fuel cycles only, any value in between may be interpreted based on the neutron flux the fuel rods have actually seen. However, multiple defects from
1 Radi'slegicel cud' Srfety Conneurnees of Opnrnting With Damated Funi ~
(Based on Westinghouse Report P'R-80-537)
O s
The concern here is mov ng po entially damaged assemblies inte the inner i
t regions of the core for usa in subsequent cycles.
Since the fuel rods would no longer be,in the vicinity of high momentum flux fuel damage associatad with baffle jetting, it is anticipated that no gross I
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would occur and no release of fuel pellets. The fuel then would contribute to the overall coolant. activity. Because of the relatively few rumbers of rods, the activity increase would not be significant, especially when the activity release mechanisuf is diffusion controlled.,,
For a fuel assembly that was damaged by baffle jetting and then inadvertently returned to the core interior, there'would be no adverse effect on potential.
i design base accidents.
The one or more cycle burnup of.the assembly would assure that the LOCA peak, clad te perature would not occur in the damaged asse bly. Therefore,.there would be no impact on LOCA peak clad tec:perature.
The operations Staff at Trojan may be alerted to the existence of reds that have been recently caused to defect or are currently being actively fretted by baffle jetting by changes in the radioisotopic content of the ccolant as described below.
When a new defect of any kind first appears in a reactor core, there will be a step increase in the steady-state activities of many fissica product
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isotopes such as Iodine-131 and Iodine-133 CesiuS138, Krypton-87 and Xenon-133. Steady-state requires essentially constant pcwer operation for three half-lives of each of these isotopest the 1,ongest half-life is that of eight-day Iodine-131. These isotopes show changes when there are changes in power levels, as well as when there are changes.in defect level. Changes in defcct level must be evaluated under conditions that represent equivalent pcwcri e
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s There is no way of distinguishing rods that are damaged from an ordinary cladding defect frem one that has just breache'd its cladding from ba,ffle jetting. This apparently was the case for the defect formed in Trojan Cycle 1 (Assembly C-18). However, when the fretting proceeds to the point where fuel is being ground away and introduced into the coolant, there will These be step increases in certain isotopes normally found only in the fuel.
isotopes tend to associate with crud and crud deposits; but of ten they can also be detected in the coclant.
4 The most positive indicator of the presence of active fretting of fuel is This the presene.e of significant amounts of Neptunium-239 in the ecolant.
isotope is an activation product of Uranium-238. The presence of Neptunium-239 in the ' coolant is sufficient evidence that Uranium-238 and other fuel
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Neptunium-239 has a half-life cf cor::penents are present.in the ecolant.
2.4 days and produces ga:m:a rays wi,th energies of 277.6, 228.2, 334.3 key and abundances of about 22,12, and 13 percent, respectively.
l Heptun'ium-239 is difficult to detect in the presence of short-lived activity;
'so that sa:ples may have to be allowed 'to decay for a day or two before they are exa=ined for !!eptunium-239.
When !!eptunium-239 is present, large amounts of P.uttenium-103 and -106, Cerium-141 and -14a, Lathanum-140 and Zirconium-fliobium-95 are also present.
These radioisotopes are often easier to detect than rieptunium-239; but their presence should not be interpreted to positively indicate the presence of Uranium in the coolant. One of the best of these preliminary indicators is Cerium-144 which produces a.gama ray energy of 133.5 key with an abun-dance of 11%.
Defects formed by fretting of rods by baffle leakage, when they are in posi-tions next to the baffle, will have additional characteristics which distin-guish their location. Their burnup will usually be ccmparatively low; so
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that the ratio of Cesium-137 to Cesium-134 released by such a defect will The relative b'e high, corresponding to an activity ratio of two cr more.
power.of these rods will usually be ccmparatively low; so that activity spikes associated with reacter power changes can be 2::pected to appear at i
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Fretting defects tend t3 be sufficirntly large and pcwer icvals cv;r,50".
open as to be plainly visible, and exposed fuel is easily washed by the As a consequence, the activity ratio of Iodine -131 to Iodine-133 coolant.
approach 0.1 when these defects are present; and activity spikes, be tends to produced by release'of stored activity when power is decreased, te None of these indications is unique to fuel fretting small, less than 10X.
from baffle leakage; but r.ost of them were present in the case of Troj Cycle 2.
The pre:cnce of the fuel-type fission products noted above is the c:ost The other indicaticas pelling evidence for active fretting of the fuel.
Both of these provide evidence of where-that fretting is taking place.
indicators are available to the operations staff at Trojan to signal sub-sequent failures..
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