ML20005A979
| ML20005A979 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 07/01/1981 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| TASK-15-13, TASK-RR NUDOCS 8107060160 | |
| Download: ML20005A979 (12) | |
Text
,
e Consumers Power Company General Offices: 212 West Mkhleen Avenue, Jackson, Michigan 49201 * (517) 78g4650 July 1, 1981 Director, Nuclear Reactor Regulation Att Mr. Dennis M Crutchfield, Chief Operating Reactors Branch no 5 US Huclear Regulatory Commission Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT - SEP TOPIC XV-13, SPECTRUM OF R0D DROP ACCIDENTS (BWR) - SYSTEMS AND RADIOLOGICAL PORTIONS The proprietary copies of this document are not included in this mailing.
They are being mailed separately and will follow shortly.
D37 s
f./t 8107060160 81070 '
DR ADOCK 05000 p
-ATTACHMENT 1 a
Toph T7-13: Spectrum of Rod Drop Accidents (BWR), Systems Portion Evaluation:
The. reactor core response in a rod drop aci dent va,s recently reevaluated by Exxon Nuclear Company using N3C-approved arslysis mathodology. The.results of this analysis are contained in ENC Report XN-NF-78-51, " Exxon Nuclear Con-trol Rod Drop (pcident Analysis for Big Rock Poin," January 1979 The XTRAN computer code vas used in this' analysis. Thr yr.'AN computer code has been approved by the Staff for application to pressur_:ed water reactors. Although the application of this methodology to BRP as described in XN-NF-78-51 has not been formally approved by the Staff, the methods employed (ie, XTRAN) are equally applicable to BWRs and FWRs.
A' comparison of methods used in this analysis with Staff criteria identified in SRP 15.4.9 follows:
Acceptance Criteria 1.
Reactivity excursions 'should not result in radially averaged fuel rod en-thalpy greater than 280 cal /gm at any axial location in any fuel rod.
For Cycles 15, 16 and 17, the maxi 39,insequencedroppedrodworthcalcu-a three-dimensional nodal boiling lated using the GROK computer code water reactor core simulator with full thermal and hydraulic feedback, was T mk.
The nodal peaking factor associated with the maximun insequence dro; ped rod was 2.23 The maximum infinite lattice local peaking factor was 1.21.
The minimum end-of-life doppler coefficient for type G-2 fuel was -1.6x10-5 Ak/k/0C.
Using figures 1.1 and 1.2 of XN-NF-78-51, the peak deposited enthalpy for this case is calculated to be approximately 138 cal /gm, and total radially averaged entaalpy (including initial enthalpy of 18 cal /gs), would be 156 cal /gn. This is much less than the 280 cal /gm allowed per the SRP and in fact is less than the fuel failure threshold of 170 cal /gm. Thus, no fuel rods would be expected to experience clad damage in such an event.
2.
The maximum reactor pressure during any portion of the assu=ed transient should be less then the value that will cause stresses to exceed the
" Service Limit C" as defined in the ASME Code.
As peak fuel enthalpies for insequence rod drops are predicted to be much less than the incipient fuel melting threshold of -280 cal /gn, prompt fuel rupture and dispersal of molten UO int the coolant will not occur. Hence 2
the pressure surge associated with the insequence rod drop will be easily limited by the large thermal capacity of the fuel and primary coolant and the primary safety relief valves which are sized to pass more than 200% of rated steam flow.
5 2
Review Procedures 1.a.
The reviewer verifies that the applicant has considered a spectrum of initial conditions for this evels that covers the range of time-in-cycle and initial power levels.
Initial full power conditions should include the uncertainties in the calorimetric measurement of power.
Only hot standby initial conditions were considered in XN-NF-78-51.
Generic analyses presented in Reference 3 have shown that rod drops from significant power levels are much less severe than drops from hot zero power for a number of reasons; including much more rapid void and doppler feedback and a more rapid scram. Time-in-cycle differences are accounted for by parametric analysis of the more important reactor kinetics para-meters including dropped rod worth, doppler coefficient, power peaking factor, and delsyed neutron fraction.
1.b.
The reviewer verifies that the maximum expected individual rod worths are used. In developing rod worth criteria, the nominal control rod withdrawal pattern must be considered, as well as those abnormal patterns that are not precluded by an instrumentation system acceptable to ICSB.
Maximun insequence dropped rod worths are considered in the design of each reload core to assure that the 280 cal /gm limit on radially averaged fuel rod enthalpy is not exceeded. BRP has no rod worth minimizer (RWM) or rod sequence control system (RSCS). The Staff concluded in Reference h that the probability of an out-of-sequence rod drop accident causing fuel enthalpies greater than 280 cal /gm is about 2x10-8 per year which is acceptably low. Thus installation of a RWM or RSCS is not considered necessary.
l.c.
The reviewer determines that an acceptable and conservative function is used to describe the control rod worth as a function of control rod posi-tion and that the control rod position as a function of time is suitably conservative.
Dropped rod worth and scram bank worth as a function of position are ex-plicitly calculated in XTRAN which is a two-dimensional (r-: cylindrical geometry) computer program. The code calculates the rapidly changing flux distribution as a control rod travels out of the core and the scram bank subsequently enters the core. The dropped rod is conservatively as-suced to be a free-falling object accelerated by gravity, since BRP has no control rod velocity limiters.
1.d.
The reviewer determines that conservative reactivity coefficients, notably the Doppler coefficient, are used and that they are compatible with those described in SRP Section 4.3 The results of parametric analyses of the importanc reactivity coefficients are provided in ZH-NF-78-51.
The calculation of maximun deposited enthalpy presented above assumed the minimum end-of-cycle doppler coefficient for any ENC fuel type in the core.
. = -
~
3 1.e.
The reviewer assures that the scram action is conservatively represented in the use of the integral scram worth curve (SRP Section h.3) and in the use of the scram delay time.
Refer to item 1.c.
A total scram bank worth of 8.9" Ak/k was assumed in
'the analysis. The actual scram bank vorth is much higher than this value.
Scram delay and insertion rates used in the analysis are as specified in the Plant Technical Specifications, l.f.
The. reviewer checks the analytical methods or assures that they have been reviewed and approved previously. The reviewer may also perform an in-dependent audit calculation using methods acceptable to the Staff. The I
applicant's methods should account conservatively for all major reactivity feedback mechanisms.
The analytical methods used in XN-NF-78-51 have been reviewed and approved by the Staff to the extent desc ribed above.
. 2.
The reviewer inspects the results of the calculation of maximum reactor pressure to determine compliance with the second criterion listed in sub-section.II (the reviewer may do an audit calculation when appropriate).
S'ee discussion of acceptance criteria.
3
' The number of fuel rods experiencing clad failure is determined (for use in evaluating the radiological consequences) by the following procedures:
a.
The reviewer determines the transient critical power ratio (CPR) has been computed by an acceptable technique (either previously reviewed or reviewed de novo during this review).
b.
The reviewer determines that the number of rods with enthalpy exceeding 170 cal /gm has been computed by.an acceptable method.
c.
The reviewer determines that the amount of fuel exceeding melting con-ditions has been computed by an acceptable method.
See discussion of acceptance criteria.
I Conclusion C
It is concluded that the analyses and results for the spectrum of rod drop accidents are acceptable and are in compliance with current NRC criteria.
i 1
I i. -,
Re ferences :
(1) XN-CC-32 P-A, "XTRAN-PWR: A Computer Code for the Calculation of Rapid Transients in Pressurized Water Reactors with Moderator and Fuel Temperature Feedback", October 7,1975 (2) Letter from D A Bixel (CPCo) to D L Ziemann (NRR) dated October 30, 1978.
-(3) C J Paone, et. al., " Rod Drop Accident Analysis for Large Boiling Water Reactors", NED0-10527, March, 1972.
(h) Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No.15 to Facility License No. DPR-6, Consumers Power Company Big Rock Point Plant, Docket No. 50-155, October 17, 1977 (5) XN-NF-78-51, " Exxon Nuclear Control Rod Drop Accident Analysis for Big Rock Point", January 1979 i
i f
L
ATTACHMENT 2 Topic XV-13: Spectrum of Rod Drop Accidents (BWR), Radiological Portion Evaluation Based on the eva'uation of an in-sequence -control rod drop accident (RDA) presented in Attachment 1, no fuel failure due to a RDA is expected.
However, fuel failure may occur in the unlikely event of a rod drop ecmbined with a single active failure (i.e. an out-of-sequence rod drop).
It has been conservatively estimated that no c: ore than (h6h) fuel rods
-(all of the rods in the four bundles surrounding the dropped control rod) vill experience cladding failure in such an event.
If there is fuel failure, the gap activity contained in the failed rods is released to the reactor coolant. If there is coincident loss of off-site power (per SRP 15.h.9), then the MSIV will close automatically. This vill isolate the containment and no release to the atmosphere vill result.
If the MSIV does not close automatically and if the reactor operators do not manually close the MSIV or the off-gas valves, then some activity may be released to the atmosphere.
(Note that the operators are required to terminate off-gas if the.h7/:; Ci per see limit set by the Technical Specifications is exceeded. Sote also that no atmopheric dumping vill be required to cool the plant down after shut-down.)
The projected doses for the unlikely event of fuel failure followed by a release to the atmosphere are 0.23 rem whole body and 1.h5 rem thyroid at the site boundary, assuming F-stability at 1 m/s meteorology. These doses are well belov (less than 1%) of the 10CFR100 exposure guidelines.
Calculations Assumptions stated in SRP 15.h.9 vere followed in calculating the projected doses. All of the noble gas activity in the failed fuel rods and 10% of the iodines are assumed to reach the turbine and the condenser. All noble gases and 10% of the iodines reaching the condenser are available for release to the atmosphere. Release is assumed to take place over a short time period ( =2 hrs.) and radiological decay is neglected in order to obtain the most conservative estimates.
NED0-2hT82 was used to obtain the total core inventory data listed in Column 2 of Table I.
Gap activities listed in Column 3 are calculated using the estimated percentages of core inventory given in the Big Rock Point Probabilistic Risk Assessment (FRA Table V.3-5, page V-62) previously submitted to the NRC by our letter dated March 31, 1981. Gap activities for the various isotopes are listed below:
Gap Activity as Isotope 5 of Total Core Inventory Xe, Kr 3%
Halogens 1.7 Organic Iodine 7%
All Iodine 1.9%
- Cs, Rb 5%
Ba 0.00015
- The organic iodines are h5 of the total iodine content (Reg. Guide 1.3).
Thus 1.7% of 96% I2 plus 7". of k% Organic I gives 1.9% of total iodines.
2
- Colu=n k of T*ble I lists the activities released to the primary coolant as a result of fuel clad failures on h6h rods (55 of total). 100% of the noble gases and 10% of halogens reach the turbine and condenser. Particu-lates ~are not considered per SRP 15.k.9 Table II lists the total activity available for release to the atmosphere by isotope (100% of noble gases and 10% of iodines reachin6 the turbine / condenser).
The projected doses were calculated as follows:
Phole Body Dose Ascu=ing a semi-infinite radioactive cloud, the ga=ma dose is:
D =.25 E X (Reg. Guide 1.3)
This can be written as:
Y"*
(
) (f)
(f Q )
q i
or
=.25 ([E Q
) 1 D
1g y
Q Where D = gamma dose in rods Y
Ei = average gn==a energy per disintegration for isotope 1
-Qi = Curies of isotope i released to the atmosphere 3
1 = atmospheric diffusion factor (2.5E-5 sec/m for F-stability Q
at im/see vind at site boundary (825 m) from Reg. Guide 1.3) r The su=mation i EQ is calculated from the energies i and amounts of g4 listed in Table II and is found to equal 3.73E+h MeV-Ci release Qi dis This gives:
=.25(fEQ)1 D
Q 7.=.25 ( 3.73x10 ). ( 2.5x10-3 )
h D
D =.23 rads y
Thyroid Dose The thyroid dose is calculated using the equation:
thyroid
( [ Q B} (3) g
3 Where x = at=ospherie diffusion factor (2.5E-5 see )
Q as before m#
Qg = curies of iodine
%pe i released to the atmosphere R = dose conversior_ factors from TID - 1h8hh listed in Table II g
B = breathing rate (3.hTE-k m /see from Reg. Guide 1 3)
I The su=mation i Qg is. calculated from the amounts of release Qi and dose conversion factors listed in Table II and is found to equal 1.67E+8 rem.
This gives:
thyroid "
(f Q R() (B) i
)
D thyroid 3
see m
Dthyroid "
Conclusion The bounding case analysis of a postulated rod drop accident (one in which no credit is taken for reactor trip, containment isolation, off gas isola-tion time and flow rates, etc.) shows that the projected doses are vell
'within the 10CFR 100 exposure guidelines.
i While the projected doses vould be substantially reduced if a more realistic analysis is performed, it is concluded that the projected doses for the bounding case are sufficiently lov so as not to warrant a more detailed analysis at this time.
4 I
It is therefore concluded that the results of this very conservative analysis of rod drop accidents for Big Rock Point are acceptable and are in compliance with current NRC criteria.
l t
TABLE I Big Rock Point Core Inventory (t=0), expected gap activities, and curies released to reactor coolant as a result of a 55 fuel rod failure.
5% of Gap Activity released to Isotope Inventory (Ci)
Gap Activity (Ci)
Reactor Coolant (C1)
.Xe-131m h.23 + h 1.27 + 3 -
6.35 + 1 Xe-133 1 36 + 7 h.08 + 5 2.04 + 4 Xe-133m 4.68 + 5 1.h0 + 4 7 00 + 2 Xe-135 2 35 + 6 7 0h + 4
-3%
3 52 + 3 Xe-135m 4.07 + 6 1.22 + 5 6.10 + 3 Xe-137 1.13 + T 3 39 + 3 1 70 + h Xe-138 1.06 + 7 3.19 + 5 -
1.60 + h Kr-85 7 21 + 4 '
2.16 + 3 -
1.08 + 2 Kr-85m 1.56 + 6 4.68 + h 2 34 + 3
-3',,
Kr-87 2 71 + 6 8.1h + h h.07 + 3 Kr-88 h.0h + 6
- 1. 21 + 5 6.06 + 3 I-131 7 05 + 6 1 35 + 5 -
6.77 + 3 I-132 9 98 + 6 1.91 + 5 9.56 + 3 I-133-1.15 + 7 2.22 + 5 1.9%
1.11 + h I-134-1.h9 + 7 2.T8 + 5 1.h3 + h I-135 1.18 + 7 2.26 + 5 --
1.13 + k 3r-84 1.13 + 6 1.02 + h -1 75 9.61 + 2 i
Eb-88 4.10 + 6 2.05 + 5 y 1.03 + h Cs-13k h.68 + 5 2.34 + h 1.17 + 3 t
Cs-136 2.15 + 5 1.08 + 4 5.38 + 2
-5 ',
2.14 v 3 Cs-137 8.57 + 5 4.28 + h j ~
Cs-138 1.24 + 7 6.20 + 5 3.10 + h Cs-139 1.19 + 7 5.97 + 5 --
2.99 + 4 3a-lho 1.13 + 7 1.13 + 1 -0.0001%
5.67 -1 h
._..,..__...._..,,,,.,.-.,_..,..___,._,_....-,._,.m,-._.
I TABLE II Activities available for release to atmosphere, average gamma energies for each isotope, and dose conversion factors.
Average Gamma Energy Activity Released Thyroid Dose Conversion Isotope Per Disintegration (MeV) to Atmosphere (C1)
Factors (rem /Ci)
Xe-131m 3.27 6.35 + 1.
Xe-133 3 02 - 2 2.04 + h Xe-133s 2.h0 - 2 7.00 + 2 Xe-135 2.h5 - 1 3.52 + 3 Xe-135=
k.28 - 1 6.10.+ 3 Xe-137 1.37 - 1 1 70 + 4 Xe-138 9.87 - 1 1.60 + 4 Kr-85 2.21 - 3 1.08 + 2 Kr-85m 1.57 - 1 2.34 + 3 l
Kr-87 7.16 - 1 h.07 + 3 Kr-83 1.81 + 0 6.06 + 3 I 'i -
3.75 - 1 6.77 a-1 1.48 + 6 3
I-32 2.22 + 0 9.56 + 1 5.35 + 4 I 33 5 77 - 1 1.11 + 2 4.00 + 5
.-134 2.h6 + 0 1.43 + 2 2.50 + 4 I-135 1.42 + 0 1.13 + 2 1.24 + 5 Br-84 1.63 + 0 9.61 + 0
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l 3
REFERENCES 1.
Regulatory Guide 1 3 2.
SRP 15.h.9 3
HEDO - 24782 4.
TID - 148hh
j.
t i
4 ATTACHMENT 3 EXXON NUCLEAR CORP. REPORT XN-NF-78-51 EXXOT NUCLEAR CONTROL ROD DROP ACCIDENT ANALYSIS FOR
}
BIG ROCK POINT
]
4 i
e l
l
-Note: This report contains information proprietary to Exxon i
Nuclear Corp. and is therefore exempt from public j
disclosure. This report is not included in general distribution of this letter, but has been transmitted i
separately.
..h-