ML20234E337
| ML20234E337 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 09/17/1987 |
| From: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| 87-377, NUDOCS 8709220351 | |
| Download: ML20234E337 (20) | |
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g VIRGINIA ELucTarc AND Powna COMPANY RIcnwoxu, VIRGINIA 20061 September'17, 1987 W.L.srm w ar.
Vaca ymmmingny.
. NocLeam OranAtsown 7
United States Nuclear Regulatory Commission Serial No.87-377 Attention:. Document Control Desk E&C/NAS/cdk Washington,'DC 20555 Docket Nos.
50-338 i
50-339 License Nos.
NPF-4 NPF-7 Gentiemen:
VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS NO. 1 AND 2 ACCIDENT REANALYSIS INFORMATION
. Virginia Electric and Power Company is supplying, for your-information, updated ~ analysis.results for two of the design basis postulated events
. discussed in Chapter 15 ' of the North Anna Updated Final Safety Analysis Report.
In projecting the characteristics of future reload cores for North Anna, we have noted a potential for certain core kinntics parameters to fall outside the range of values previously analyzed.
In anticipation of these trends, -Virginia Power has initiated a program to reanalyze some of these accidents off the critical path for reload safety evaluation.
Sections 1~and 2 of the Attachment provide the results of reanalyses of two addit'ional. accidents, the Rod Withdrawal from a Subcritical Condition.and the Control Rod Ejection events, respectively. The results show that all of
'the appropriate' acceptance criteria continue to be met for both events. No Technical Specification changes are required in conjunction with these
. reanalyses.
This analysis summary is being submitted for your information.
These analyses will be reflected in the next scheduled update of the North Anna UFSAR.
The information presented in the Attachment has been reviewed by the Station Nuclear Safety and Operating Committee and the Safety Evaluation and Control Staff.
It has been determined that neither the results.of the analyses nor the design and operation of future reload cores which fall within the range of characteristics. bounded by these analyses involve an unreviewed safety question as defined in 10 CFR 50.59.
Very truly yours, e
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j Attachment 1.
Accident Analysis and Evaluation - Uncontrolled Control Rod 4
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Withdrawal from a Suberitical Condition 2.
Accident Analysis and Evaluation - Rupture of a Control Rod Drive Mechanism Housing cc: U.S. Nuclear Regulatory Commission
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J 101 Marietta Street, N.W.
Suite 2900
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Atlanta, GA 30323 Mr. J. L. Caldwell NRC Senior Resident Inspector North Anna Power Station 1
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- 1. ACCIDENT ANALYSIS AND EVALUATION - UNCONTROLLED CONTROL ROD j
WITHDRAWAL FROM A SUBCRITICAL CONDITION i
Identification of Causes and' Accident Description j
l A rod cluster control assembly accident is defined as an uncontrolled i
addition of reactivity to the core resulting from withdrawal of the rod cluster control assemblics, thereby producing a po'wcr excursion.
Poten-tial causes for the event include malfunctions of the reactor control and control rod drive systems and operator error.
l l
This event could occur with the reactor subcritical, as in one of the I
shutdown modes, at hot zero power or at power. A brief discussion of the i
protection afforded by the reactor protection system in each of these i
modes follows.
A. Protection at Cold Shutdown - When the reactor is at cold shutdown (mode 5) the unit is at least 1% delta k/k subcritical. In addition, the shutdown margin requirement of 1.77% delta k/k must be satisfied for this j
mode.
Thus if the reactor is just at 1% subcritical, at least another 0.77% delta k/k must be associated with withdrawn rods.
During cold shutdown the reactor may or may not be at full flow conditions. At least one residual heat removal or one reactor coolant pump must be in opera-tion.
Protection against uncontrolled rod withdrawal in this mode is provided by the high source range count rate trip. A minimum' of two source range channels must be operable whenever the reactor trip breakers are closed and the rod drive system is capable of withdrawal.
Trip occurs on one of two channels exceeding a preselected setpoint.
The action of the source range trip prevents any significant power generation in the core in the event of an uncontrolled rod withdrawal event.
B. Protection at Hot Standby and Hot Shutdown (Modes 3 and 4) - As in mode 5, the source range trips are required to be operabic whenever the trip breakers are closed and the rods are capabic of being withdrawn. The re-quirements for shutdown margin and subcriticality remain the same as in mode 5.
Temperatures may be higher (up to the HZP temperature of 547 F in mode 3), but the RCS flow requirements are increased in mode 3 (above 350 F): at least one reactor coolant pump must be in operation.
As in mode 5, operation of the source range trip serves to prevent any signif-icant power generation in the event of an uncontrolled control rod with-drawal.
Further discussion of reactor protection against uncontrolled reactivity addition at shutdown conditions is provided in Reference 1.1.
C. Protection during Startup and Power Operation (Modes 1 and 2)- In these modes all three reactor coolant pumps must be in operation in accordance with Technical Specifications, thereby providing full core flow. The P-6 permissive setpoint (1 out of 2 intermediate range current exceeding a preset value) allows the source range trips to be blocked in order for power escaletion to continue.
Protection against overpower for modes 1 and 2 is provided by:
2
+
- 1) Interm'ediate range high neutron flux reactor trip. This trip is actuated by 1 of.2 Intermediate Range (IR) channels. exceeding a current. corresponding to 25% of rated thermal power.
No
-credit is,taken for the IR trip in'the accident analysis. The
/IR trip can be blocked above the P-10 permissive setpoint (2 of 4 power range channels exceeding 10% of rated thermal power).
- 2) Power range high.noutron flux reactor trip.
Reactor trip occurs when two out of the four power range channels exceed a preset power. level.
There are two trip setpoints associated with the-power range channels.
Below 10% of rated thermal power (the P-10 interlock setpoint) the power range trip setpoint is 25% of rated thermal power (nominal; the Technical Specification allowable value is 30%).
Above 10% power this setpoint can be blocked.
The high flux setpoint then increases to 109% power (nominal; the allowable is 110%)..The response of the core to rod' withdrawal events occuring at above 10% power, where the power range low setpoint is blocked, is discussed in UFSAR Section 15.2.2, Uncontrolled Rod Cluster Control Assembly Bank. Withdrawal at Power.
- 3) Intermediate and power range rod stops.
The rod stops serve to discontinue rod withdrawal on either 1 of 2 high IR flux signals or on 1 of 4 high power range flux signals, thereby eliminating the need to actuate the corresponding high flux level trips.
The reactor is normally brought to power from a subcritical condition'by means of rod. cluster control assembly withdrawal.
Boration or dilution may be performed to ensure criticality at a desired control rod position.
During the shutdown modes, all sources of primary grade water are locked, sealed or otherwise secured in the closed position except during planned boron dilution or makeup activities, thereby precluding an uncontrolled
.1dd{ tion of reactivity to the core.
During the approach to criticality (all RCS loops in operation), the maximum rate of reactivity resulting from boron dilution is less than assumed in this analysis (see also UFSAR Section 15.2.4, Uncontrolled Boron Dilution).
Analysis of Effects and Consequences Method of Analysis This transient is analyzed by the RETRAN digital computer code. Virginia Power's RETRAN models and methods are discussed in Reference 1.2.
To give conservative results for a startup accident, tho' following assumptions are made concerning the reactor initial conditions:
- 1. Since the magnitude of the power peak reached during the initial part of the transient for any given rate of reactivity insertion is strongly dependent on the Doppler coefficient, conservative values (low absolute values) as a function of temperature are used.
3 l-
1
]
1
- 2. A conservative value for the moderator coefficient of
+6.0E-5 delta k/k 'F is used in the analysis to yield the maximum peak heat flux.
- 3. The reactor is assumed to be at hot zero power. The initial effective multiplication factor is assumed to be 1.0 since this results in maximum neutron flux peaking.
- 4. Reactor trip is assumed to be initiated by power range high
't neutron flux (low setting). A 10% increase is assumed for the flux trip setpoint, raising it from the nominal value of 25% to 35% of rated thermal power.
- 5. The maximum positive reactivity insertion rate assumed (1.0E-3 delta k/k-second) is greater than that for the simultaneous withdrawal of the combination of the two control banks having the greatest combined worth at maximum speed (45 in./ min).
- 6. The initial power level was assumed to be below the power level expected for any shutdown condition.
The combination of highest reactivity insertion rate and lowest initial power produces the highest peak heat flux.
- 7. Since the magnitude of the peak heat flux increases with increasing effective delayed neutron fraction (beta), a conser-vatively high BOL value is assumed.
l
- 8. The reactor coolant flow rate is assumed to correspond to all three pumps in service.
As discussed previously, during shutdown mode operation when less than full core flow may be available, protection against any significant core power generation resulting from uncontrolled rod withdrawal is provid-ed by the source range high neutron flux trips.
Results Figures 1-1 through 1-3 show the transient behavior for the indicated i
reactivity insertion rate with the accident terminated by reactor trip at 35% of nominal power.
Figure 1-1 shows the neutron flux transient. The neutron flux overshoots the full-power nominal value, but this occurs only for a very short time, llence, the energy release and the fuel temperature increase are relatively small.
The core heat flux response, of interest for DNB considerations, is shown in Figure 1-2.
The beneficial effect of the inherent thermal lag in the fuel is evidenced by a peak heat flux 1ers than the full power nominal value. There is a large margin to DNB during the transient, since the rod surface heat flux remains below the full power design value, and there is a high degree of subcooling at all times in the core.
Figure 1-3 shows the response of the average fuel, cladding and coolant temper-atures. The average fuel temperature increases to a value lower than the nominal full power value.
4 l
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Conclusions l-In the event of a rod cluster control assembly withdrawal accident from l
'the subcritical condition, the. core and tho' reactor coolant system are l
lnot. adversely affected, since'the combination of thermal power and the coolant temperature results in a DNBR value well above the' limit value.
.This' conclusion regarding the DNBR-is based on the fact that the analysis
.shows that' nominal full power values for the coolant temperature and heat flux are not exceeded.
References 1.1 Letter from W. L. Stewart (Virginia Electric and Power) to 11. R.
1 Denton (NRC), Serial No. 85-772A, " Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Response to Request for Additional Information on Core Uprate," February 6, 1986.
1.2 VEP-FRD-41A, " Reactor System Transient Analysis Using the RETRAN Computer Code," Virginia-Electric and Power Company, May 1985.
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- 2. ACCIDENTIANALYSIS AND EVALUATION - RUPTURE OF A CONTROL R0D DRIVE MECHANISN11003ING (ROD CLUSTER COVrROL ASSEMBLY EJECTION) i4? notification of Causes and Accident Description This accident is defined as the mechanical failure of a control rod mechanism pressure housing, resulting in the ejection of a rod cluster control assembly and drive shaft.
The consequen'ce of this mechanical failure is a rapid reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel rod damage.
')
7 The reactor protection in the event of a rod-ejection accident has been described in Reference 2.1 The protection for this' accident is provided by high-neutron-flux trip (high and low setting) and high rate of neutron flux increase trip.
These protection functions are described in detail in Section 7.2 of the UFSAR. Section 15.4.6.1.1 of the UFSAR also provides a detailed discussion of the mechanical and nuclear design features which limit both the probability,sud potential consequences of this event.
Due to the extremely low probability of a RCCA-ejection accidgat, limited feel damage is considered an acceptable consequence.
Comprehensive studica of the threshold of fuel failure and of the threshold of significant conversion of the fuel thermal energy to me-chanical energy have been carried out as part of the SPERT project by the Idaho Nuclear Corporation.
These test results are discussed in Section
- 15. 4. 6.1. 2 of the UFSAR.
To reflect these experimental results, con-servative criteria are applied to ensure that there is little or no pos-sibility of fuel dispersal in the coolant, stoss lattice distortion, or severe shock waves. These criteria are s
s 1.
Average fuel pellet enthalpy at the hot spot below 225 cal /gm for unirradiated fuel and 200 cal /gm for irradiated fuel.
2.
Peak clad. temperature at the hot spot below the temperature at which clad embrittlement may be expected (2700 'F).
3.
Peak reactor coolant pressure less than that which would cause stresses to exceed the faulted condition stress limits.
4.
Fuel melting limited to less than 10% of the fuel volume at the hot spot even if the average fuel pellet enthalpy is below the limits of criterion 1 above.
l
~
Analysis of Effects and Consequences Method of Analysis The analysis of the RCCA-ejection accident is performed in two stages:
.first, an average core nuclear power transient calculation, and then a hot-spot heat transfer calculation.
The point kinetics model of the RETRAN computer code (References 2.2 and 2.3) is used to perform the average core transient analysis.
This code 9
I includes the simulation of prompt and delayed neutrons (using the six-group model), the thermal kinetics of the fuel and moderator and the balance of the NSS primary and secondary coolant system. Thermal feedback effects are modeled via temperature dependent reactivity coefficients with a detailed multiregion, transient fuel-clad-coolant heat transfer model. Reactivity insertion from the ejection of the control rod and the 7
subsequent reactor trip are accounted for.
The average core energy addition, calculated as described above, is j.
' ~ '
multiplied by the appropriate hot-channel factors, and the hot-spot analysis is performed using a detailed fuel and clad transient heat transfer model of the RETRAN code termed the Hot Spot Model. This model calculates the transient temperature distribution in a cross section of a metal-clad UO2 fuel rod, and the heat flux at the surface of the rod, using as input the nuclear power versus time and the local coolant con-ditions. The zirconium-water reaction is explicitly represented, and all material properties are represented as functions of temperature.
A parabolic radial power generation is used within the fuel rod.
A detailed discussion of the method of analysis can be found in Reference 2.2.
System Overpressure Analysis.
Because safety limits for the fuel damage specified earlier are not ex-coeded, there is little likelihood of fuel dispersal into the coolant.
The pressure surge may therefore be calculated on the basis of conven-tional heat transfer from the fuel and prompt heat generation in the coolant. Details of the analysis methodology are given in Section 15.4.6 of the UFSAR.
Calculation of Basic Parameters Input parameters for the analysis are conservatively selected on the basis of values calculated for this type of core.
Table 2-1 presents the pa-rameters used in this analysis.
Reference 2.4 provides a description of the methodology used to calculate the delayed neutron fraction, the ejected rod worth and pre-and post-ejection local peaking factors (Fq).
These values are recalculated for every reload core to confirm that the Table 2-1 values remain bounding.
The feedback reactivity weighting factor is applied to the Doppler feed-back calculeted in the point kinetics model to account for the effects of high local peaking factors in the vicinity of the ejected rod.
A de-J tailed description and qualification of the feedback weighting factors used is provided in Reference 2.2.
The Doppler reactivity defect is determined as a function of fuel tem-perature using a two-dimensional steady-state computer code with a Doppler weighting factor of 1.0.
The resulting curve is conservative compared to design predictions for this plant. The weighting factor will l
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increase under accident conditions as discussed above.
)
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To allow for future fuel cycles, conservative estimates of Beff of 0.52%
at beginning of cycle and 0.43% at end of cycle were used in the analysis.
The trip reactivity insertion is assumed to be 4% from hot full power and 1.77% from hot zero power, including the effect of one stuck rod, (i.e'.,
the ejected rod). The shutdown reactivity is simulated by a conservative curve of trip reactivity insertion versus time after trip. The start of the rod motion occurs 0.5 sec after the high-neutron-flux point is reached. This delay is assumed to consist of 0.2 sec for the instrument
~
channel to produce a signal, 0.15 sec for the trip breaker to open, and 0.15 sec for the coil to release the rods.
The analyse's presented are applicabic for a rod insertion time of 2.2 sec from coil release to gn-trance of the rod at the dash pot, although measurements indicate that this value should be closer to 1.8 sec. The choice of such a conservative insertion rate means that there is over 1 sec af ter the trip point is reached before significant shutdown reactivity is inserted into the core.
This is a particularly significant conservatism for hot full power acci-dents.
Appropriate margins are added to the results to allow for calculational uncertainties, including an allowance for nuclear power peaking due to fuel densification.
The value of parameters used in the analysis, as well as the results of the analysis, are presented in Tabic 2-1 and discussed below.
i Beginning of Cycle, Full Power.
Control bank D was assumed to be inserted to its insertion limit.
The l
worst ejected-rod worth and hot-channel factor were 0.20% delta k and 6.50, respectively.
The peak hot-spot clad temperature was 2493 'F. The peak hot-spot fuel center temperature exceeded the beginning-of-life melt temperature of 4900
'F.
Ilowever, melting was restricted to less than 10%
of the pellet.
Beginning of Cycle, Zero Power.
For this condition, control bank D was assumed to be fully inserted and control bank C was at its insertion limit.
The worst ejected rod was located in control bank D and had a worth of 0.878 delta k and a hot-channel factor of 15.40.
The peak hot-spot clad temperature reached 2489
'F.
End of Cycle, Full Power.
Control bank D was assumed to be inserted to its insertion limit.
Con-servative values of ejected-rod worth and hot-channel factor, 0.21% delta k and 6.20 respectively, were used.
This resulted in a peak clad tem-perature of 2391 "F.
The peak hot-spc t fuel temperature exceeded the assumed end-of-life melt temperature of 4800
'F.
Ilowever, molting was restricted to less than 10%
of the pellet. The variation of melt temperature with burnup is discussed in Section 4.4.1 of the UFSAR.
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End of Cycle', Zero Power.
The ejected-rod; worth and hot-channel factor for this case were obtained assuming control' bank D to be fully inserted and control bank C at its insertion limit. Conservative values used in the analysis for this con-dition were 0.99% delta k and 19.2, respectively. The peak clad and fuel center temperatures reached 2593 'F and 4206 'F.
A summary of the cases presented above is given in Tabic 2.1.
The nuclear power and hot-spot fuel and clad temperature transients for the worst cases (beginning-of-life full power and end-of-life zero-power) are pre-sented in Figures 2-1 through 2-4.
Fission Product Release.
Radiological concerns for the rod ejection event have been addressed ge-nerically by Westinghouse in analyses which conservatively bound North Anna.
These analyses are diMussed in Reference 2.5.
The evaluation presented there shows that moc';ing the Virginia Power imposed limits of allowabic fuel melt and clad embrittlement temperature for the hot channel assures that the radiological limits for the event as specified in Regu-latory Guide 1,77 wiki be met.
Pressure Surgo.
As for the radiological concerns, the RCS overpressure analysis for the
~'
rod ejection event has been performed generically.
This analysis is briefly described and appropriate references given in the UFSAR and in Reference 2.5.
The evaluatior, of potential structural effects of a rod ejection event is provided in Section 15.4.6.2.3.7 of the UFSAR. That evaluation shows that there are no unaccounted-for reactivity insertions resulting from fuel motion or distortion under post-rod ejection conditions. The eval-untion presented there remains valid for the set of analyses presented
]
herein.
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Conclusions Even on a pessimistic basis, the analyses indicate that the described fuel and clad limits are not exceeded. It is concluded that there is no danger of sudden fuel dispersal into the coolant.
Since the peak pressure does not exceed that which would cause stresses to exceed the faulted condition i
stress limits, it is concluded that there is no danger of further conse-
)
quential damage to the primary loop.
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'bD References
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2.1 WCAP-7306,3 " Reactor Protec' tion System Diversity in Westinghouse 7
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Pressurized Water Reactors," Westinghouse Electric Corporation, April 11969.
'2.2 VEP-NFE-2-A,L"Vepco Evaluation of the Control Rod Ejection Transient," Virginia' Electric and Power Company," December 1984.
'2.3 VEP-FRD-41A,'" Reactor System Transient Analysis Using the RETRAN Computer-Code," Virginia Electric and Power Company, May 1985.
2'4 VEP-FRD-42, Revision 1-A, " Reload Nuclear Design Methodology,"
.j Virginia Electric and Power Company, November 1986.
2.5-WCAP-7588, Revision 1-A, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," Westinghouse Electric Corporation, January 1975.
13
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Table 2.1 PARAMETERS USED IN Ti!E ANALYSIS OF THE R0D CLUSTER CONTROL ASSEMBLY EJECTION ACCIDENT Time in Life Beginning Beginning End End Power level 102%
0%
102%
0%
Ejected-rod worth, % delta k 0.20 0.878 0.21 0.99 Delayed neutron fraction, %
0.52 0.52 0.43 0.43 Feedback reactivity weighting 1.58 3.13 1.53 3.78 Trip reactivity, % delta k 4.0 1.77 4.0 1.77 Fq before rod ejection 2.51 2.51 i
Fq after rod ejection 6.50 15.4 6.20 19.2 Number of operational pumps 3
2 3
2 Maximum fuel pellet average 4011 3456 3875 3577 temperature, 'F Maximum fuel center 4903 4078 4803 4206 temperature, 'F Maximum clad temperature, 'F 2493 2489 2391 2593 Maximum fuel stored energy, 185 147 175 154 cal /gm l
14 I
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FIGURE 2-1 1.6
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-3.
10 CFR 50.59 SAFETY EVALUATION The results presented in Sections 1 and 2 demonstrate that design and use of future reload cores which have key safety parameters (e.g., maximum reactivity insertion rate due to RCCA bank withdrawal from subcritical, maximum ejected rod worth and post-ejection local power peaking factor) which are bounded by the assumptions of these analyses will not create an unreviewed safety question as defined in 10 CFR'50.59.
Specifically, l
use of such cores will not:
j
- 1. Increase the probability of occurrence or the consequences of 1
any previously evaluated accident. The core characteristics of I
interest only impact the response of the core once the accident is in progress, so the probability of event initiation is not impacted.
The analysis results show that all of the design and acceptance criteria for these events continue to be met wit.h margin. Therefore the consequences of the events remain un -
changed from those reported in the UFSAR.
- 2. create the possibility of any new or different accident type not previously evaluated. No hardware change or relaxation in in any limiting condition for operation is reficcted in ther.c analyses.
Changes in core physics characteristics of the ',ype being considered here will not introduce any new or uniqu 3 accident precursors.
- 3. reduce any margin of safety as defined in the bases of the the technical specifications.
As discussed above, the analyces demonstrate that the appropriate design and acceptance criteris for these events will continue to be met.
The acceptance criteria applied to these events have been previously estab-lished in the UFSAR and in Reference 2.2.
Therefore no safety j
margin will be reduced.
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19
_ _ _ _ _ _ _ - _