ML19352A152
| ML19352A152 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 08/31/1980 |
| From: | Soong P NUS CORP. |
| To: | |
| Shared Package | |
| ML19352A149 | List: |
| References | |
| NUS-1761-ADD-1, NUDOCS 8103110540 | |
| Download: ML19352A152 (25) | |
Text
{{#Wiki_filter:- _ .. ATTACHMENT 4 9 ~ NUS-1761 Addendum 1 CRITICALITY ANALYSIS OF THE SPENT FUEL STORAGE RACKS FOR THE VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA UNIT 1 AND 2 , By P. Y. Soong August 1980
?
. Approved by: MV 7 i
.I W. V. Macnabb i Manager, Nuclear-Economics and {
Fuel Services Department i NUS CORPORATION 4 Research Place Rockville,-Maryland 20850 I 8iog13g yQ
4 l l TABLE OF CONTENTS Section and Title Page No. 1 1.0 PURPOSE AND SCOPE OF ANALYSIS 1 2.0 GENERAL METHOD OF ANALYSIS 2
-3.0 INPUT INFORMATION AND SOURCES 3 4.0 MAJOR ASSUMPTIONS 4 :
I 3 5.0 REFERENCE COMPUTER CODES 5 6.0 DETAILED CALCULATIONS- 6 6.1 Reference Design Criticality Analysis 6 6.2 Sensitivity Studies 6 6.3 Tolerances 10
- 6.4 Calculational Uncertainties- 11 Accident Analysis 11
~
6.5
- 7. 0'-
SUMMARY
OF RESULTS 12- /
^# , , ,-% v w -- v- -
CRITICALITY ANALYSIS OF THE SPENT FUEL STORAGE RACKS FOR THE VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA UNIT 1 AND 2
- 1. 0 PURPOSE AND SCOPE OF ANALYSIS The results of criticality analysis for the NUS-designed North Anna spent fuel storage racks for storing 3.5 w/o enriched fuel were reported in NUS-1761 as revised March 1, 1977. This Addendum is prepared to present additional work (NUS Analysis File G-RA-18, June 1980) performed recently in determination of the highest enrichment fuel that may be stored in the same spent fuel storage racks without violating any safety li aits. The highest enrichment as determined in this analysis is 4.1 w/o and the corresponding maximum keff
- of the rack system is 0.9496 at 95% confidence level. A summary of criticality results for the 4.1 w/o enriched fuel
. is given in Section 7.0.
It was originally contemplated to run a few KENO-IV/NITAWL (AMPX) cases with 4.0 and 4.2 w/o fuels and then combi.e these results with the ak, biases due to mechanical toler-antes and calculational uncertainties for the 3.5 w/o fuels to obtain the maximum enrichment. However, the later develop- . ments show that the final k eff f the rack system is very
- close to the design limit of.0.95 and the new ok, biases (for 4.1 w/o enriched fuel) are not necessarily the same as the old values (for 3.5 w/o enriched fuel). In consideration of these developments, it was decided to conduce a full scope of criticality analysis for storing 4.1 s/o enriched fuel in'the rack system. The current results are, therefore, selfconsistent: The work file (NUS Analysis File G-RA-18, June-1980,'Section 6.0) is self-contained.
1 l
2.0 GENEPAL METHOD OF ANALYSIS The method of analysis is similar to the enes used for the several other NUS spent fuel storage rack designs, six cf which have been licensed. The method involves (i) neutzen transport calculations using Monte Carlo code KENO-IV, c: css section processing code AMPX and 123-group GAM-THEF2.OS cross : I
- section library, and (ii) diffusion theory calculations :
using diffusion code PDC-7 and NUMICE-2. The reference rack design is analyzed with the transport method, while small reactivity changes due to variation in material ec= position, 4 physical /.1=ensions and pool water temperature are deter-mined with the diffusion method. The 'inal k e_. ,, of the rack
- system is normalized to the transport result.
- Y 6
3 f' b i , 3 2
. 3.0 INPUT INFORMATION AND SOURCES e Fuel Specifications- Reference was made to Stone &
Webster specification for Spent Fuel Storage Racks, Addendum No. 1, October 4, 1976 for fuel specifications. Both Westinghouse 17 x 17 and 15 x 15 and B&W 17 x 17 and 15 x 15 fuel assemblies were examined for determining the most reactive
.. . type ef fuel.
e Rack Drawings- The following NUS drawings provide
~
rack spacing and storage cell dimensions:
; 5094M2000, Revision 2, Sheet 1 of 1 5094M2001, Revision 4, Sheet 1 of 1 5094M2002, Revision 3, Sneet 1 of 1
- s
. .s s
wh 1-
'.-'T
' .i . r..
- h. . ' . .
- O i .
lA E v3-4 ,. 2 R
# A 3-s,.
v y y + v.N- -
/ ,,y r- ,,----r ,. i e-- 4
. 4.0 MAJOR ASSUMPTIONS (1) All fuels are assumed to be fresh and to have a uniform fuel enrichment of 4.1 w/o.
, (2) Nominal water temperature is 68 F, however the
; final result is adjusted to the most reactive tempcrature condition.
2 (3) No credit is taken for the soluble boron in the spent fuel storage pool except under accident
- conditions.
1 (4)' No credit is taken for the burnable poison and 1 control rods (if any) present in fuel assemblies. 1
.j_ (5) Fuel storage racks are infinite in size in three dimensions (6) No credit is taken for neutron abscrption by structural materials other than stainless steel ]
8 cans. I 1 Assumption 5. removes the difference in physics interpretation J of k,of a storage cell and keff f the rack system. Trese
] two terms.are interchangeably used in this report.
I i: I.7l ! J' 7 L
- l
'}. -J i -l
- 1 c1 l 3 i
;k t
q-4
i 5.0 REFERENCE COMPUTER CODES i WAPD-TM-678 PDQ-7 Reference Manual by W. R. Cadwell, Bettis Atomic Power Laboratory, January 1967. e NUS-894 (Rev. 2) NUMICE A Spectrum Dependent Non-Cp tial Cell Depletion Code by Y. S. Kim, NJS Corporation, March 1976. NUMICE is NUS' version of LEOPARD. ORNL-4938 KENO-IV, An Improved Monte Carlo Criticality Program by L. M. Petrie and N. F. Cross, ORNL, November 1975. ORNL-TM-3706 AMPX - A Modular Code System for Generating Coupled Multigroup Neutron Gamma Libraries from ENDF/B by N. .". Greene, et al., ORNL, March 1976. l-I. . 5
6.0 DETAILED CALCULATIONS 6.1 Peference,gysian Criti'cality Analysis Figure 6-1 shows the mechanical design of a stainless steel can. The.can has an inside dimension of 9.00 ! .06 inches with can wall thickness of 0.125 t .010 inch. Figure 6-2 shows a unit storage cell consisting of a stainless steel can loaded with a Westinghouse 17 x 17 fuel assembly and
~
spaced at 14 x 14 inch center-to-center distance. With reflection boundary imposed en all sides, the geometry shown in Figure 6-2 represents an infinite array of the identical
- storage cells of infinite length. The cell k, as determined by KENO-IV with 123-group cross sections processed by NITAWL (a subroutine in AMPX code) is 0.9158 ! .3047, while the PDQ-7 k, is 0.9073.
~ The same storage cell loaded with other types of fuel was 1 also analyzed. Both Westinghouse 15 x 15 and B&W 17 x 17 fuels were found to have slightly higher k ,by 0.0038 aks.
3 l The final rack reactivity is adjusted to reflect the difference between Mestinghouse 17 x 17 and the most reactive type of
] fuel.
a 6.2 Sensitivity Studies J Sensitivity studies were performed to show change of storage i'
!' cell k= with respect to small variations of such conditions b
as fuel enrichment, pool water temperature, center-to-canter spacing and stainless steel composition. ! e Fuel Enrichment - According to DOE enrichment g service specifications, a variation of.! .050 w/o is allowable for uranium enriched to 2.00 to 5.00 w/o. m An increase of 0.050 w/o in enrichment corresponds to a a-Ak,of +0.0019 which is derived from the following
? PDO-7 calculations:
J f
'.q-1
'l 6
Source: NUS drawing 5094M2001 Revision 4, Sheet 1 of 1 INSIDE SQUARE AT ALL GROSS SEC TlONS. 1-1 1 1 010 e 9.00 06 >'< .I2 5 STK 3 j 2 d g
/
l ' I ' 1' SGUARE " : 8.750 MIN. FREE PATH - j
~ l ; i .
l 4, , l I, j; m [ .. _ 7 i si s is r iis /r /s 7 1 74
.125 R PERMISSIBLE .., ON FP':E PATH a
I ? i
- j. -
FIGURE 6-1 >7 f: -j STAINLESS STEEL CAN a
- 7-
[,
=
" 9" SQ.
- 8.432" SQ.
~
'f J0 o I
W ATE R l . r,- r r , , r i , r , r r , , r, , I- 4 -b 4 + -l - E -t
< - I- 4 +- d I
} , [--[- + -I- h M- h --l l- L - * - f 1- 4. M - l--l 9 - -l- L ? - l- + -l 1 -i--!- -j l- -f - }- + -1 1
~? i--} 4 9 !- f- -1 4 4 1- 4 0 -i- I--i1 i
' _ l- i- -h 1 - y - t- -l - b -! ~1 - F- + 1- - F 4 -F- i
- i- + -! O l- 4 04 l Oi 4 -- + O 4 -l 9HJ t- 4 -l- -i - 1 -i + - h -i - F J-Fd g
] o I -. I- -l- -i- + 4 - F 1- F + + -l- l- 1-y e g
, q i FO-i 104 + l- +- + -i - 1 04-l-4, A h i !I . 5 t- 4 }- -l- 4 i- 1- 4 + -I -4 -4 I I- L + 4 4 - t- + + 4-- t- 4 -i h 4--t O-i - l- -i 9F4-104-4-10-i- -+ -4I S -i--j l I- t- + h 4- 1 -t 4- 1- 1 -j 0 l- +- -1 4 -i 4 i 4- J St - L -I --i . 0 i-4- l- 1 -9 4 -l- t-Q? -t 04 -i - I -+ --J -4 j ' l- + l- 1 L- 4. --I -4 -- 4- y J--l--{
' L L _j _ L J _ J. J-. - .1 -. I L L .L 1 l J - J l
/8" SS CAN
, WATER
' J=0
- 2 14" SPACING
- L FIGURE 6-2
'* UNIT STORAGE CELL 1;ITl!1:ESTINGHOUSE 17 x 17 FUEL ASSEMBLi .j.
ll;J 8
.m Il
w/o PDO-7 Unit Storace cell U-235 k= Ak= 4.0 0.9034 4.1 0.9073 0.0 4.15 (interpolated) +0.0019 4.2 0.9110 +0.0037
; o Pool Water Temperature - The pool water operating temperature is approximately 140 F. The unit
, storage cell behaves in such a manner that its reactivity follows a positive temperature coef ficient until steam is formed in water. As shown below, I}
the maximum ak,due to the pool water temperature
~] - variation is limited to +0.0037.
J
' Void PDO-7 Unit Storace Cell
, Temp. Inside Outside
~ '3 F can can k= aka 4 39.2 0% 0% 0.9050 68 0 0. 0.9073 0.0 140 'O O 0.9103
.j- 212 0 0 0.9110 +0.0037 10 .10 0.8924
"- 20 10 0.8593
.s 20 20 0.8795 30 10 0.8244
"- 30 20 0.8470 m 30 30- 0.8758 e Center-to-Center Spacing The.storace rack design-shows a center-to-center spacing of
'14 .~ 0 .06 inches. By varying the spacing from
~
fl1g. .the nominal l4.0 to 13.941 inches', the unit storage
'~'
cell shows a reactivity rise of 0.0016 ok . Il. :
. b. ':
e 9 - s
PDO-7 Unit Storage Cell Spacing, inch k= ak= 13.88 x 13.88 0.9105 13.94 x 13.94 0.9089 +0.0016 14.0 x 14.0 0.9073 0.0 14.06 x 14.06 0.9058 t e Stainless Steel Composition - A set of stainless steel composition with low nickel and chromium content was established in the original report
' NUS-1761. By comparing the stainless steel having q low content of nickel and chromium with the regular i steinless steel in the following PDQ-7 study, a
,i ak , of +0.0035 is chosen to account for the
, j ' reactivity change-due to the possible variation of stainless steel composition. PDO-7 Unit, Storage Cell Stainless steel k= ak= l Regular 0.9073 0.0
, Change case 0.9108 +0.0035 6.3 Tolerances l-" 'In addition to the variations described in Sectic>n 6.2, the worst tolerances considered in the current analysis include
.the fo13owing:
l , aka e e Variation of the inside dimension of-can from 9.0 to 9.06 inches 0.0005 e Variation of can wall thickness from 0.125 to 0.115' inch 0.0026 e Non-straightness of storage cans 0.0016 e
- e. . Eccentric fuel. loading i.e., a. fuel 1 assembly positioned off-center of a
., stainless steel can.. 0.0064 l
74
In each of these cases, a PDQ-7 unit storage cell with one perturbation at one time was run and compared with the reference cell. Ak,value is obtained from these two PDO-7 Cases. 6.4 Calculational Uncertainties The KENO k,for the reference unit storage cell is 0.9158 which bears a standard deviation of : .0047 after 29,100 At 95% confidence level, neutron histories were traced. 2c = 0.0094. A Ak,of 0.0094 is assigned to cover statis-tical variations. In addition, a Ak,of 0.0086 is assigned as the KENO bench-mark bias factor to cover the uncertainty related to the i
, method as well as the accuracy of the neutron cross sections.
l The'4k,value in this case was derived from a series of ORNL calculations using the same KENO code and the same cross 3 NUS' separate c&lculation of several 3 section library. the bias factor different critical asse'mblies indicates that 1
] established herein above is conservative.
6.5 Accident Analysis 1
.a Several cases of fuel handling accident were analyzed in j
' October, 1977 with 3.5 w/o enriched fuel and 1950 ppm a,
(minimum) of soluble boron in the pool water and reported in Analysis File G-RA-05, Addendum 1. It was shown that the
] 1950 ppm soluble boron is worth more than 20% ak,which is r -
far more than enough to' compensate the reactivity rise in any potential fuel handling accident. Despite the change t,f fuel ' enrichment from 3.5 to 4.1 w/c, in
.the current analysis, the large margin of safety contributed byL the soluble boron is expected. to remain availab'.a. None
' '}^. of the hypothesized fuel handling accidents will, therefore, pose a threat to_ criticality safety of the storage rack system as. long 'aus the minimum amount of soluble boron is The accident analysis was, present in the_ pool water.
.therefore, not repeated.
~
11 e _ q. V
7.0
SUMMARY
OF RESULTS The NUS-designed spent fuel storage racks for North Anna Unit 1 and 2 will have no difficulty in accepting Westinghouse and/or B&W fuel assemblies with cnrichment e.s high as up to
] 4.1 w/o. The rack system will remain suberitical and will
" not exceed the design k limit of 0.95.
eff Results of the criticality analysis for 4.1 w/o enriched fuel are summarized in Table 1. Because of Assumption i in Section 4.0, the k m of the unit storage cell which is 0.9159
}
(Section 6.1) may also be accepted as e.. k ,, of the rack < " system. After being added with bias factors due to the
- =w orst tolerances and calculational uncertainties, the
. maximum k ef, f the rack system becomes 0.9496 at 95%
J confidence level. It is to be noted that this keff value represents, in fact,
"' the reactivity level of an infinite array of storage cells of infinite. length. The actual storage rack is, however, a
]e finite system. Although the final kg ,f value is close to 0.95, it'has built-in margin of safety.
...J
~
]
W e -
"L) 12
~
TABLE 1
SUMMARY
OF CRITICALITY ANALYSIS RESULTS Reference k eff Pa c e
- _
Reference 'init storage cell .9158 34 W 17x17 fuel, 4.1 w/o enrich.aent,
. 680F,14 inch spacing Fuel other than W 17x17 fuel .0038 40
. Pool' water temperature variation .0037 54
. Worst Case tolerances
... Enrichment variation .0019
.. Spacing variation .0016
.a Variation of inside dimension of can .0005 Variation of can wall thickness .0026 Non-straightness of can .0016
+
Variation of stainless steel composition .0035
~
Eccentric fuel loading .0064 a rms .0083 .0083 86
~
Calculational uncertainties l KENO benchmark .0086 2 Statistics .0094
]s- Sub-total .0180 .0180 37 4 Maximum rack k,ff at 95% .9496 J Confidence level 1 .
Refer to the page number.in NUS Analysis File G-RA-18.
.4 13_.
a , U _
ATTACHMENT 5 PAGE 1 1.0 Introduction and Conclusions currently fuel batches at the North Anna power Station are designed such that the batch average discharge burnups range from 28,000 to 33,000 megawatt days per metric ton (MWD /MTU) of uranium. In order to improve uranium utili'ation, reduce nuclear fuel cycle costs and reduce the number of discharge fuel assemblies, batch average discharge burnup extensions to approximately 45,000 MWD /MTU are being proposed. Vepco is participating in a program sponsored by the Department of Energy to demonstrate extended burnup technology. The Department of Energy (Reference 1) has determined that this improved fuel utili=ation program will have no significant impact on the environment. From a general safety viewpoint, Reference 1 states that, "no changes to existing facilities or to any aspects of fuel design or fuel use will be required. Radionuclide release from extended burnups of the current design fuel will be within normal facility design considerations and no l change in the safety and accident considerations of light water reactorr are expected." The purpose of this safety evaluation is to document that the general safety assessment provided in Reference 1 is also applicable l specifically to the North Anna power Station. l The potential safety impacts of higher burnup fuel include the two basic
- areas of fuel performance and safety analysis. Sections 2.0 and 3.0 document the evaluations of the potential fuel performance and safety analysis impacts, respectively.
From these evaluations, the following conclusions can be draun: u-
W PAGE 2
- 1. Westinghouse fuel performance at the North Anna power Station will retain its current high level of reliability at increased fuel burnup levels.
- 2. The safety analyses for the North Anna power Station are not sj nificantly impacted at increased fuel burnup levels.
< 3. Westinghouse reload safety evaluation methodology (Reference 2) assures an effective cycle by cycle check to insure continued plant safety at increased fuel burnup levels. Should any fuel performance changes be identified, they will be appropriately incorporated in the safety analysis as required. l l 4 9 L: t Ll.
PAGE 3 l 2.0 Funi performance i High burnup fuel will not have a significant impact on either fuel design or operation. Westinghouse fuel is designed to meet the fuel rod i design bases / criteria listed below:
- 1. Fuel Centerline Temperature
- 2. Fuel Rod Internal Pressure
- 3. Cladding Stress
- 4. Cladding Strain
- 5. Cladding Fatigue
- 6. Cladding Collapse i
Further description of these design bases / criteria is given in Reference
- 3. At the' present time, . Westinghouse fuel designed for the North Anna Power Station must meet the indicated design criteria. The higher burnup fuel designed for the North Anna Power Station will also be required to meet the design criteria or the designs will be precluded from further consideration. Furthermore, the high burnup fuel performance will be assessed against the current -fuel related Technical Specifications.
Specifically, high burnup fuel will be required to meet the current peaking factor and reactor coolant activity limits or the impact on tha
- ,. current. safety analyses'will be assessed as discussed in Section/3.0.
PAGE 4 3.0 Safety Analysis As indicated in Reference 1, no fundamental change in the safety and accident considerations of the LWR are anticipated as a result of greater discharge exposure fuel. Higher burnup fuel could potentially impact those safety analpses for which radiation releases were originally postulated. As indicated in the North Anna power Station Final Safety Analysis Report (Reference 4), these accidents include Steam Generator Tube Rupture, Steam pipe Rupture, Fuel Handling Accident, Volume Control Tank Rupture, Waste Gas Decay Tank Rupture and the Loss of Coolant Accident. An evaluation of the impact of higher burnup fuel on each of the accidents is discussed in Sections 3.1 through 3.5. Section 3.6 discusses the Westinghouse methodology which verifies the applicable licensing analyses on a cycle by cycle basis. 3.1 Steam Generator Tube Rupture It is assumed that the accident takes place at power and while the reactor coolant is contaminated with fission products corresponding to continuous operation with one percent of the fuel rods defective. The accident leads to contamination of the secondary systems due to leakage i of radioactive coolant from the reactor coolant system and in the event of a coincident loss of offsite power, there will be a discharge of activity. to the atmosphere through the steam generator safety and/or power operated relief valve. Since the release will occur because of leakage of the primary coolant, maintaining the same technical I
PAGE 5 specification limits on coolant activity would insure that the quantity of radionuclides available for release from the coolant would be unaltered. In addition, the offsite radiological consequences of this accident are based on the airborne releases of volatile fission products (noble gases and radioiodine). All of the important radioactive iodine and noble gas nuclides are of short half-life compared to the fuel cycle time, with the exception of Kr-85. For these nuclides, equilibrium inventories in the fuel are attained relatively quickly. Thus, the quantity of these volatile fission products released in the event of a fuel failure is independent of burnup. In the case of Kr-85, both the small inventory present an.d the small contribution of Kr-85 to the total dose result in insignificant changes in the radiological consequences of this accident. 3.2 Steam pipe Rupture ! The limiting steam pipe rupture accident for the North Anna power n i l Station will not result in departure from nucleate boiling. l l Consequently, no radioactivity is released to the environment because of l a steam line break unless there is or has been primary to secondary l ( system- leakage in a steam generator. Current Technical Specifications l limiting primary to secondary leakage and reactor coolant system
. activity will not change as a result of higher burnup fuel. These limits are -well below values assumed in the licensing analysis. In addition, the offsite radiological consequence results will not be significantly impacted as discussed in Section 3.1.
( ~ Y
PAGE 6 3.3 ruel Handling Accident For the purposes of evaluating the radiological consequences of this accident, all rods in the highest rated discharged fuel assembly are assumed to rupture with a sudden release of the noble gas and halogen inventories. Although larger quantities of long lived fission products will be present in the fuel,5 review of the source terms applicable to this accident indicates that the major contributors to the dose result from thort-lived isotopes of such radioactive gases and volatiles as xenon, krypton, and iodine.' Since these short-lived gases and volatiles reach equilibrium concentrations after about one year of irradiution, there is no significant increase in site boundary dose due to the release of radioactive gases and volatiles during a fuel handling accident from high burnup fuel as compared to standard burnup fuel. 3.4 Radioactive Gas Release I The concentration of r adi oac tive waste gases in the primary and auxiliary systems is a function.of the rate of fission gas release to the coolant from defective fuel and the rate of removal via the auxiliary systems. The components which retain significant concentrations of radioac 'cive gases are the volume control tank and the waste gas decay tanks. The radioactive gas release analysis considers the rupture of the volume control tank and a waste gas decay tank with the instantaneous release of the radioactive gas inventories of each to the environment. Since the Technical SFecifications limit on reactor l-
. _ ~ _ _ _
pAGE 7 coolant activity is not changing, the volatile fission products available for the airborne releases are not significantly impacted as discussed in Section 3.1. 3.5 LOCA i The loss-of-coolant accident (LOCA) would not become more adverse if the use of higher burnup fuel is implemented. In the loss-of-coolant accident, peak linear heat rates and the design of the emergency core coolant system must be such as to limit clad temperatures and oxidation to values prescribed by regulations (10 CFR 50, Appendix N). The primary fuel characteristics which can influence clad temperatures and resulting oxidation are the amount of stored energy and the decay heat source immediately after shutdown. The quantity of stored energy present in the l fuel is dependent upon the peak linear heat rate during operation, but is independent of burnup. Since the fuel is designed and the plant is j operated to limit the peak linear heat rate to values which are established to be acceptable for LOCA, stored energy will not increase j as a result of th.e proposed changes. The decay heat source term immediately after shutdown is due almost exclusively to the decay of short-lived fission products which reach a l saturated concentration during the first year of irradiation. Thus, the proposed changes, and particularly the increase in discharge exposure, will not significantly alter the decay heat generation during the loss-of-coolant event. Furthermore, the Nuclear Regulatory Commission i R requires-the use of the1AMS decay heat standard (see 10 CFR 50, Appendix d L
PAGE 8 K) based upon iniinite operation plus 20% uncertainty, which is thus conservative and already encompassen extended burnup.7 Since neither the stored energy nor decay heat generation rate increases, the proposed changes are not expected to result in an increase in the severity (i.e., in peak clad temperatures or oxidation) of the loss-of-coolant accident or to increase the chances of meltdown, and hence improvements or changes in emergency core cooling systems are not required. The proposed changes (in particular the increase in discharge exposure) also are not expectea to significantly alter site boundary doses as a result of the loss-of-coolant accident. Site boundary doses are due almost exclusively to the short-lived fission products which reach saturnted concentration during the first year of irradiation,' so that the inventory (curies) of isotopes which are significant in the calculation of site boundary dose during the loss-of-coolant accident is virtually independent of burnup.5 Although increased fuel failures during the LOCA event are not anticipated, the Nuclear Regulatory commission requires that site boundary doses be evaluated under the conservative assumption of 100% fuel failure.e Thus, any tendency for increased fuel failures during LOCA of high burnup fuel will not alter l the. perceived consequences of the loss-of-coolant accident.
- [ ~
PAGE 9 3.6 Reload Safety Evaluation Methodology The Westinghouse reload safety evaluation methodology will continue to serve as the mechanism employed to confirm the validity of the existing safety analyses for each fuel cycle. The methodology is used to identify technical specification changes and potential unreviewed safety questions. This methodology will readily identify any potential impact higher burnup fuels may have on plant safety and will verify the existing safety analyses. Should the reload safety evaluation process identify a change in the fuel performance (e.g., an increase in the number of rods predicted to have a minimum DNBR less than or equal te design limits), these changes will be reviewed by the appropriate radiation specialists to determine the impact of these changes on the radiological consequences of the particular event in question. Such changes will be documented in accordance with the requirements of 10 CFR 50.59. I f
o . I PAGE 10 t 4.0 References
- 1. DOE /EA-0118, " Environmental Assessment, DOE Program"to Improve
. Uranium Utilization in Light Water Reactors," August 1980'.
- 2. WCAP-9273, " Westinghouse Reload Safety Evaluation Methodology,"
March 1978.
- 3. Docket No. STN 50-572, "RESAR-414, The Reference Safety Analysis Report Documenting the Westinghouse 3820 MWth Nuclear Steam Supply System," October 8,1976.
. 4. Final Safety Analysis Report - North Anna Power Station Units 1 and 2, Virginia Electric and Power Company, December 1969.
- 5. "NASAP Preliminary Safety and Environmental Information Document l
i
-(Volume I, Pressurized Water Reactors) - Responses to NRC Comments",
l _ August,'1979.
- 6. " Reactor Safety Study", WASH-1400 (NUREG-75/014), October 1975.
l l l l 7. Title 10, Part 50, Code of Federal Regulations, Appendix K. l l- 8. Regulatory Guide 1.4, " Assumptions Used For Evaluating The Potential l l Radiological Consequences Of A Loss Of Coolant Accident For Pressurized Water Reactors",' Revision 2, June 1974. i f _- _ _ - - _ - . -}}