ML20005B467
| ML20005B467 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 07/01/1981 |
| From: | Vincent R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19277A022 | List: |
| References | |
| TASK-15-13, TASK-RR NUDOCS 8107080264 | |
| Download: ML20005B467 (14) | |
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0 sm General offices: 212 West Michigan Avenue. Jackson, Michigan 49201 *(517) 788 0550 July 1, 1981 Director, Nuclear Reactor Regulation Att Mr Dennis M Crutchfield, Chief Operating Reactors Branch No 5 US Nuclear 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 Enclosed is the Consumers Power Company evaluation of SEP Topic XV-13 for the Big Rock Point Plant. Attachment 1 provides the systems portion of the evaluation, and Attachment 2 provides the radiological portion.
As Attachment 3, Exxon Nuclear Corp. report XN-NF-78-51, " Exxon Nuclear Control Rod Drep Accident Analysis for Big Rock Point", is enclosed. This report contains information which is prop-letary to Exxon Nuclear Corp.
and is exempt from disclosure under Section 2.790 (a)(h) of the NRC Rules of Practice, Part 2, Title 10, Code of Federal Regulations.
It is there-fore requested that Attachment 3 be withheld from public disclosure.
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G L E
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Robert A Vincent Staff Licensing Engineer LPD2 /
NP NRC Resident Inspector - Big Rock Point s u s. s u P s s s h~3,n
[ MPH E"&EY CONTROLLED COPY 7h /
f(i PDR ADDCKI n107080264 810701 05000155i
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AFFIDAVIT STATE OF Washington
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COUNTY OF Benton I, James N. Morgan, being duly sworn, hereby say and depose:
1.
I am Manager, Licensing and Safety Engineering, for Exxon Nuclear Company, Inc., (" ENC") and as such I am authorized to execute this Affidavit.
2.
I am familiar with ENC's detailed document control system and policies which govern the protection and control of information.
3.
I am familiar with the report entitled XN-NF-78-51, " Exxon Nuclear Control Rod Drop Accident Analysis for Big Rock Point," referred to as " Document".
Information contained in this Document has been classified by ENC as proprietary in accordance with the control system and policies established by ENC for the control and protection of information.
4.
The Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by ENC and not made available to the public. Based on my experience, I am aware that other companies regard information of the '::. d contained in the Document as being proprietary and confidential.
5.
The Doc ent has been made available to the United States
.l
~ Nuclear Regulatory Commission in confidence, with the request that the information contained in the Document not be disclosed or divulged.
' 6.
The Document contains information which is vital to a competitive advantage of ENC and would be helpful to competitors of ENC when competing with ENC.
7.
The information contained in the Document is considered to be proprietary by ENC because it reveals certain distinguishing aspects of control rod drop accident analysis methods which secure competitive economic advantage to ENC for fuel design optimization and improved marketability, and includes information utilized by ENC in its business which affords ENC an opportunity to obtain a competitive advantage over its competitors who do not or may not know or use the information contained in the Document.
8.
The disclosure of the proprietary information contained in the Document to a competitor would permit the competitor to reduce its expenditure of money and manpower and to improve its competitive position by giving it extremely valuable insights into control rod drop accident analysis methods, and would result in substantial harm to the competitive position of ENC.
9.
The Document contains proprietary information which is held in confidence by ENC and is not available in public sources.
10.
In accordance with ENC's policits governing the protection and control of information, proprietary information contained in the Document has been made available, on a limited basis, tu others outside ENC only as required and under suitable agreement providing for non-disclosure and limited use of the information.
i 11.
ENC policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis. Checks are made routinely to assure the policy procedures are being met.
t
~,.,, - - - -... - -
12.
This Document provides information which reveals control rod drop accident analysis methods developed by ENC over the past several years.
ENC has invested hundreds of thousands of dollars and several man-years of effort in developing the control rod drop accident analytical methods revealed in the Document. Assuming a competitor had available the same background data and incentives as ENC, the competitor might, at a minimum, develop the information for the same expenditure of manpower and money as ENC.
13.
Based on my experience in the industry, I do not believe that the background data and incentives of ENC's competitors are suffi-
-ciently similar to the corresponding background data and incentives of ENC to reasonably expect such competitors would be in a position to duplicate ENC's proprietary information contained in the-Document.
THAT the statements made hereinabove are, to the best of nty knowledge, information, and belief, truthful and complete.
FURTHER AFFIANT SAYETH NOT.
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SWORN TO AND SUBSCRIBED before me this / g day of Q ldet,19ff..
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OT RY PUBLIf/
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's ATTACHMENT 1 e
Topic XV-13: Spectrum of Rod Drop Accidents (BWR), Systems Portion Evaluation:
The reactor core response in a rod drop accident was recently reevaluated by Exxcn Nuclear Company using NRC-approved analysis methodology. The results of this analysis are contained in ENC Report XN-NF-78-51, " Exxon Nuclear Con-trol Rod Drop pycident Analysis for Big Rock Point," January 1979 The XTRAN computer code was used in this analysis. The XTRAN computer code has been approved by the Staff for application to pressurized 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 PWRs. A comparison of methods used in this analysis with Staff criteria identified in SRP 15.4.9 follows:
l 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.
ForCycles15,16and17,themaxigyinsequencedroppedrodworthcalcu-lated using the GROK computer code
, a three-dimensional nodal boiling water reactor core simulator with full thermal and hydraulic feedback, was T mk.
The nodal peaking factor associated with the maximum insequence dropped rod was 2.23.
The maximum infinite lattice local peaking factor was 1.21.
was-1.6x10gAk/k/0C.e minimum end-of-life doppler coefficient for type G-2 fuel Using figures 1.1 and 1.2 of XN-Ni-78-51, the peak deposited enthalpy for this case is calculated to be approximately 138 cal /gm, and total radially averaged enthalpy (including initial enthalpy of 18 cal /gs), would be 156 cal /gm. 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 vould be expected to experience clad damage in such an event.
2.
The maximum recctor pressure during any portion of the assumed transient should be less than 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 /gm, prompt fuel rupture and dispersal of molten UO into 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 2005 of rated steam flow.
O l
2 Review Procedures 1.a.
The reviewer verifies that tne applicest has considered a spectrum of initial conditions for this event that 7 overs 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 conditiona 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 ccounted for by parametric analysis of the more important reactor kinetics para-meters including dropped rod worth, doppler coefficient, power peaking factor, and delayed 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.
Maximum 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 enthaJpy 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 lov.
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 posicion as a function of time is suitably conservative.
Dropped rod worth and scram bank vorth as a function of position are ex-plicitly calculated in XTRAN which is a two-dimensional (r-z 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-sumed 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 important reactivity coefficients are provided in XN-NF-78-51.
The calculation of maximum deposited enthalpy presented above assened tae minimum end-of-cycle doppler coefficient for any ENC fuel type in the core.
'o a
3 1.e.
The reviever assures that the scram action is conservatively represented in t}< use of the integral scram vorth curve (SRP Section h.3) and in the use of the scram delay time.
Refer to item 1.c.
A total scram bank vorth 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.
1.f.. The reviewer checks the analytical methods c; assures that they have been reviewed and approved previously. The revievar may also perform an in-dependent audit calculation using methods acceptable to the Staff. The 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 described 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).
See 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 reviewe:' determines the transient critical power ratio (CPR) has been compuced by an acceptable technique (either previously reviewed or. reviewed de novo during this review).
b.
The reviewer determines that the n d er of rods with enthalpy exceeding 170 cal /gm has been computed by ar ccptable method.
c.
The reviewer detemines that the _ sunt ! fuel exceeding melting con-ditions has been computed by an accey, method.
See discussion of acceptance criteria.
Conclusion 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.
M
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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", NEDO-10527, March, 1972.
(h) ' : Safety. Evaluation by the Office of Nuclear Reactor Regulation Supportin6 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 197:).
3
~.
i.
ATTACHMENT 2 Topic XV-13: Spec' rum of Rod Drop Accidents (BWR), Radiological Portion Evaluation Based en the evaluation of an in-sequence control rod drop accident (RDA) presented in Attachment 1, no fuel failure due to a RDA is expected.
However, fuel fa' lure may occur in the unlikely event of a rod drop combined with a single active failure (i.e. an out-of-sequence rod drop).
It has been ecnservatively estimated that no more 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 da 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 teminate off-gas if the.h7/:: Ci per see limit set by the Technical Specifications is exceeded. " Note 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 deses are well belev (less than 1%) of the 10CFR100 exposure cuidelines.
Calculations Aseumptions stated in SRP 15.h.9 vere followed in calculating the projected doses. All of the noble gas activity in the failed fuel rods snd 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 conservativa estimates.
NED0-2h782 was used to obtain the total core inventory data listed in Colu=n 2 of Table I.
Gap activities listed in Colu=n 3 are calculated using the estimated percentages of core inventory given in the Big Pock 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
% of Total Core Inventorv Xe, Kr 3%
Halogens 1.7%
Organic Iodine T%
All Iodine 1.9%
- Cs, Rb 5%
Ba 0.0001%
- The organic iodines are 4% of the tctal iodine content (Reg. Guide 1.3).
Thus 17% of 96% I2 plus T% of h% organic I gives 1.9% of total iodines.
2 Colu=n 4 of Table I lists the activities released to the primary coolant as a result of fuel clad failures on h6h rods (5% of total). 100% of the noble gases and 10% of halogens reach the turbine and condenser. Particu-lates are not considered per SRP 15.h.9 Table II lists the total activity available for release to the atmcsphere by isotope (100% of noble gases and 10% of iodines reaching the turbine / condenser).
The pro,jected doses were calculated as follows:
Whole Body Dose Assuming a semi-infinite radioactive cloud, the ge=ma dose is:
D =.25 E X (Reg. Guide 1. 3)
This can be written as:
<q Y
(
) ()
{f Q )
q i
or
=.25'(fEQ
) 1 D
gg y
9 Where D = gamma dose in rods Y
Ei = average gamma energy per disintegration for isotope i
'Q1 = 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)
I The summation i EQ is calculated from the energies i and amounts of gg release Qi listed in Tabic 1." and is found to equal 3.73E+h MeV-Ci dis This gives:
=,25([EQ)x D
Q k
D =.25 ( 3.73x10 ) (2.5x10-5) y D =.23 rads 7
Thyroid Dose The thyroid dose is calculated using the equation:
thyroid "
( [ Q Ry (B) g
\\
3
= atmospheric diffusion factor (2.5E-5 see )
Where v Q
as before 7
Qi = curies of iodine isotope i released to the atmosphere j
R = dose conversion factors from TID - 148kk listed in Table II 3
B = breathine rate (3.hTE h m /see from Reg. Guide 1.3)
I QR is calculated from the amounts of release Qi and dose The su==ation i i
conversion factors listed in Table II and is found to equal 1.6TE+8 rem.
This gives:
thyroid (f.QR) (B) ii 3)
D
= (2.5x10~5sec) (1.6T7.10 rem) ( 3.hTx10
~
m hymid 3
see 3
D thyroid Conclusion The Lounding case analysis of a postulated rod drop accident (one in which no credit is taken for reactor trip, centainment isolation, off gas isola-tion time end flev rates, etc.) shovs that the projected doses are well within the 10CFR 100 exposure guidelines.
While the projected doses vould be substantially reduced if a more realistic analysis is perforr.ed, it is concluded that the projected doses for the bounding case are sufficiently low so as not to warrant a more detailed analysis at-this time.
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.
.Li-
k TABLE I Big Rock Point Core Inventory (t=0), expected gap activities, and curies released *,o reactor coolant as a result of a 5% fuel rod failure.
5% of Gap Activity released to Isotope Inventory (Ci)
Gap Activity (CJ)
Reactor Coolant (Ci)
Xe-131m 4.23 + 4 1.27 + 3 -
6.35 + 1 Xe-133 1 36 + 7 4.08 + 5 2.04 + 4 Xe-133m 4.68 + 5 1.40 + 4 T.00 + 2 Xe-135 2.35'+ 6 T.04 + 4
--.3%
3 52 + 3
- Xe-135m 4.07 + 6 1.22 + 5 6.10 + 3 Xe-137 1.13 + 7 3 39 + 5 1.70 + 4 Xe-138 1.06 + T 3.19 + 5 -
1.60 + 4 Kr-85' T.21 + 4 2.16 + 3 1.08 ^ 2 l
2 34 + 3 Kr-85m 1.56 + 6 4.
+4
--3",
Kr-87' 2 71 + 6 8.14 + 4 4.07 + 3 Kr-88 4.04 + 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 + 4 I-134.
1.49 + 7 2 78 + 5 1.43 + 4 I-135 1.18 +'T 2.26 + 5 -.
1.13 + 4 3r-84 1.13 + 6 192 + 4 -17%
9.61 + 2 Rb-88 h.10 + 6 2.05 + 5 -
- 1. 0 3 + 4 Cs-134 4.68 + 5 2.34 + 4 1.17 + 3 Cs-136 2.15 + 5 1.08 + 4 5 38 + 2
--5",
2.14 + 3 Cs-137-8.57 + 5 4.28 + 4 Os-138.
1.24 + 7 6.20 + 5 3.10 + 4 Cs-139 1.19 + 7 5.97 + 5 2.99 + 4 Ba-140 1.13 + 7 1.13 + 1 -0.0001%
5.67 -1
s -
TABLE II
' Activities available for release to atmosphere, average ga mra energies for each isotope, and dose conversion netors.
Average Gamma Energy Activity Released Thyroid Dose Conversion Isotope' Per Disintegration (MeV) to Atmosphere (Ci)
Factors (rem /Ci)
Xe-131m 3 27 - 3 6.35 + 1 Xe-133 3 02 - 2 2.04 + h Xe-133m-2.40 - 2 7 00 + 2 Xe-135 2.45 - 1 3.52 + 3 Xe-135m-4.28 - 1 6.10 + 3 Xe-137 1 37 - 1 1 70 + 4 Xe-138 9.87 - 1 1.60 + h Kr-85 2.21 - 3 1.08'+ 2 Kr-85m 1.57 - 1 2.34 + 3 Kr-87
.16 - 1 4.07 + 3 Kr-88
- 1. 81. + 0 6.06 + 3 I-131 3 75 - 1 6.77 + 1 1.h8 + 6 I-132 2.22 + 0 9.56 + 1 5.35 + 4 I-133 5 77 - 1 1.11 + 2 4.00 + 5 I-134 2.h6 + 0 1.43 + 2 2 50 + 4 I-135 1.42 + 0 1.13 + 2 1.24 d-5 Br-84
- 1. 63 + 0 '
9.61 + 0 a
4
--m-m m.-
l v
REFEREIICES 1.
SRP 15.4.9 3
ITEDO - 2h782 h.
TID - 14844
.