ML19323J225
| ML19323J225 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 06/18/1980 |
| From: | Ferguson J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Harold Denton, Youngblood B Office of Nuclear Reactor Regulation |
| References | |
| 537, NUDOCS 8006190463 | |
| Download: ML19323J225 (60) | |
Text
s VIRGINIA EI.ECTHIC AND Powan COMPANY Itxennon1>,v2BOINIA 20261 June 18, 1980 Mr. Harold R. Denton, Director Serial No. 537 Office of Nuclear Reactor Regulation LQA/ESG:rab Attn:
Mr. B. J. Youngblood, Chief Licensing Branch No. 1 Docket No. 50-339 Division of Licensing License No. NPF-7 U. S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Denton:
SPECIAL (LOW POWER) TEST PROGRAM SAFETY ANALYSIS AND TECHNICAL SPECIFICATION EXEMPTION REQUESTS NORTH ANNA UNIT No. 2 Enclosed (Enclosure 1) is the safety analysis and discussion of required exemptions to Technical Specifications for the special low power test program to be conducted on North Anna 'Jnit 2.
The procedures for conducting these tests were submitted with our letter dated June 13, 1980.
Pursuant to 10 CFR 50.90, we request an amendment to license NPF-7 which would permit the performance of the special test program by temporarily exempting certain Technical Specifications. Table 3-1 of lists the_ specifications which will require exemptions and the tests for which they cannot be met.
Section 3 explains the reasons these specifications must be exempted, and provide the basis for continued opera-tion during the tests.
The enclosed safety analysis and requests for Technical Specification exemptions have been reviewed and approved by the NSSS vendor, Westinghouse Electric Corporation. In addition, they have been approved by the Station Nuclear Safety and Operating Committee and the System Nuclear Safety and Operating Committee.
In accordance with Technical Specification Table 3.7-4, also cnclosed (Enclosure 2) are revisions to the listing of safety-related hydraulic snubbers. These additional snubbers have been added as a result of recent reanalyses, and bring the affected systems into compliance with FSAR commitments. The addition of these snubbers to Table 3.7-4 has also been approved by the Station and System Nuclear Safety and Operating Committees.
Since the proposed test program and safety analysis are based on the program reviewed by the Staff for the TVA Sequoyah plant, the Staff should be able to conclude that the enclosed Technical Specification requests do not involve a significant hazards consideration. Accordingly, per 10 CFR 170.22, _ this request has been determined to be in Class III, and our check in the amount. of $4,000 is enclosed in payment of the amendment fee.
80061gJOM63 k
Vsmusa Etnereic ano Powsm Conraxy to Mr. Harold R. Denton 2
As has been discussed with the Staff, we will be ready to begin the test program in the very near future.- We stand ready to answer any questions or provide any further information you may require in order to promptly approve the program.
Very truly yours, g._~
J. H. Ferguson Executive Vice President-Power
Enclosures:
1.
Special Test Evaluation 2.
Revisions to Table 3.7-4 3.
Check for $4,000
COMMONWEALTH OF VIRGINIA )
)
- s. s.
CITY OF RICHMOND
)
Before me, a Notary Public, in and for the City and Common-wealth aforesaid, today personally appeared J. H. Ferguson, who being duly sworn, made oath and said (1) that he is Executive Vice President-Power of the Virginia Elcetric and Power Company, (2) that he is duly authorized to execute and file the foregoing Amendment in behalf of that Company, and (3) that the statements in the Amendment are true to the best of his knowledge and belief.
Given under my hand and notarial seal this I 8* day of 3ane
, neo.
My Comission expires Itintictd a 9O,6B\\
O W /w / v?9'/L'/
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Notary Public~
(SEAL)
- b R
o ENCLOSURE 1 NORTH ANNA UNIT 2 SPECIAL LOW-POWER TESTS SAFETY EVALUATION JUNE 1980 5366A
1.0 INTRODUCTION
AND
SUMMARY
In an attempt to move the licensing process for North Anna, VEPCO proposed to the NRC the same special tests to be performed at reactor power levels at or below 5% of Rated Thermal Power as TVA proposed for Sequoyah. These tests would demonstrate the plant's capability in several simulated degraded modes of operation and would provide opportunities for operator training.
The basic mode of operation to be demonstrated is natural circulation with various portions of the plant equipment not operating, e.g., pressurizer heaters, loss of offsite power (simulated), loss of onsite AC power (simulated), and RCPs for plant cooldown.
Westinghouse has reviewed the proposed tests and has with the exception of TVA proposed tests 8 and 9 (startup from stagnant conditions and boron mixing and cooldown) has determined that with close operator surveillance of parameters and suitable operator action points in the event of significant deviation from test conditions, the tests as outlined in the VEPCO Special Test procedures are acceptable and can be performed with minimal risk.
It is recognized that in order to perform these tests some automatic safety functions, reactor trips and safety injection, will be defeated. Westinghouse has determined a set of operator action points which should replace these automatic actuations.
It is also recognized that several technical specification requirements will not be met while either preparing for or performing these tests.
Again Westinghouse has determined that the low power levels and operator action will suffice during these time periods.
Westinghouse has reviewed the effect of the proposed test conditions on the incidents and faults which were discussed in the Accident Analysis section of the North Anna Final Safety Analysis Report.
In most cases, ths FSAR discussion was found to bound the consequences of such events occurring under testing conditions.
Consequences of an ejected RCCA have not been analyzed because of the low probabilities.
For some incidents, because of the far-of f-normal conditions, the analysis methods available have not shown that, with reliance on automatic pro-tection system action alone, the FSAR analyses are bounding.
In those cases reliance is placed on expeditious operator action.
The operator action points as defined will provide protection for such events.
1.1
%3AAA
e 2.0 UESCRIPTION OF TESTS 2.1 NATURAL CIRCULATION TEST (TEST 1)
Objective - To demonstrate the capability to remove decay heat by natu-ral circulation.
Method - The reactor is at approximately 3% power and all Reactor Cool-anc Pumps (RCP's) are o'perating.
All RCP's are tripped simultaneously with the establishment of natural circulation indicated by the core exit thermocouples and the wide range RTD's.
2.2 NATURAL CIRCULATION WITH SIMULATED LOSS OF OFFSITE AC POWER (TEST 2)
Objective - To demonstrate that following a loss of offsite AC power, natural circulation can be established and maintained while being powered from the emergency diesel generators.
Method - The reactor is at approximately 3% power and all RCP's are operating.
All RCP's are tripped and a station blackout is simulated.
AC power is returned by the diesel generators and natural circulation is verified.
2.3 NATURAL CIRCULATION WITH LOSS OF PRESStlR1ZER HEATERS (TEST 3) i Objective - To demonstrate the ability to maintain natural circulation and saturation margin with the loss of pressurizer heaters.
Method - Establish natural circulation as in Test 1 and turn off the pressurizer heaters at the main control board. Monitor the system pres-sures to determine; the saturation margin, the depressurization rate, and the effects of charging / letdown flow and steam generator pressure on j
the saturation margin.
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2.4 EFFECT OF STEAM GENERATOR SECONDARY SIDE ISOLATON ON NATURAL CIRCULATION ("TST 4)
Objective - To determine the effects of steam generator secondary side isolation on natural circulation.
Method - Establish natural circulation aditions as in Test 1 but at 1%
Isolate the feedwater and steam itne for one steam generator and power.
establish equilibrium.
Return the steam generator to service.
2.5 NATURAL CIRCULATION AT REDUCED PRESSURE (TEST 5)
Objective - To demonstrate the ability to maintain natural circulation at reduced pressure and saturation margin.
The accuracy of the satura-tion meter will also be verified.
Method - The test method is the same as for Test 3, with the exception that the pressure decrease can be accelerated with the use of auxiliary pressurizer sprays.
The saturation margin will be decreased to approxi-mately 20 F.
2.6. COOLDOWN CAPABILITY OF THE CHARGING AND LETDOWN SYSTEM (TEST 6)
Objective - To determine the capability of the charging and letdown system to cooldown the RCS with the steam generators isolated and one RCP operating.
Method - With the reactor shutdown, trip two of the RCP's and isolate all three of the steam generators. Vary the charging and letdown flows and monitor the primary system temperatures to determine the heat removal capability.
-~
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1.7 SIMULATED LOSS OF ALL ONSITE AND OFFSITE AC POWER (TEST 7)
Objective - To demonstrate that following a loss of all onsite and offsite AC power the decay heat can be removed by natural circulation using the auxiliary feedwater system in the manual mode.
Method - The reactor is at approximately 1% power and all RCP's are running. All RCP's are tripped and a total station blackout is simu-lated.
Instrument and lighting power is provided by the backup batteries since the diesels are shutdown.
2.8 ESTABLISHMENT OF NATURAL CIRCULATION FROM STAGNANT CONDITIONS Westinghouse does not believe that it is advisable to pc?*orm this test as noted in a letter from T. M. Anderson, Westinghouse, to H. Denton, NRC, N'-TMA-2242, April 29, 1980.
2.9 FORCED CIRCULATION C00LDOWN 4
This test is. performed as preparation for the Boron Mixing and Cooldown Test.
Since Westinghouse does not believe it is advisable to perform the Boron Mixing Test as defined using core heat, it is not necessary to perform the Forced Circulation Cooldown Test.
2.10 BORON NIXING AND COOLDOWN Westinghouse does not believe that it is advisable to perform this test utilizing core heat as noted in NS-TMA-2242, T. M. Anderson, Westinghouse, to H. Denton, NRC, 5366A e,
3.0 IMPACT ON PLANT TECHNICAL SPECIFICATIONS In the evaltatise of the proposed tests Westinghouse has determined that approximately cl..rteen technical specifications will be violated, and thus require exceptions, during the performance of the tests.
Table 3-1 lists the technical specifications that will require exceptions and the tests for which they will not be met.
The following notes the reasons these specifications must be excepted and the basis for centinued opera-tion during the tests.
3.1 IMPACT
SUMMARY
3.1.1 T.S. 2.1.1 REACTOR CORE SAFETY LIMITS The core limits restrict RCS T,yg as a function of power, RCS pressure (pressurizer pressure) and loops operable.
These limits provide protec-tion by insuring that the plant is not operated at higher temperatures or lower pressures than those previously analyzed.
The core limits in
.the North Anna tech specs are for three loop operation.
Obviously when in natural circulation with no RCP's running these limits would not be met.
However, it should be noted that the tests will be performed with limits on core exit temperature (< 610 F), T
(< 580 F) and avg Loop AT (< 65 F) such that no boiling will be experienced in the core and the limits of specification 2.1.1 for temperature will be met.
The limits will not be met simply because less than three RCP's would be running.
3.1.2 T.S. 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip System provides protection from various transients and j
faulted conditions by tripping the plant when various process parameters exceed their analyzed values. When in natural circulation two trip
-functions will be rendered inoperable, Overtemperature AT and Over-power AT.
There is a tanparature input to these functions which originates from the RTD bypass loops. Due to the low flow conditions, 5% or less, the temperature indications from these loops will be highly 3.1 15366A-
suspect. To prevent the inadvertent tripping of the plant when in the natural circulation mode these functions will be bypassed.
Their pro-tection functions will be performed by the operator verifying that Pres-surizer Pressure and Level, Steam Generator Level, and subcooling margin (T,,g) are above the operator action points for Reactor Trip and Safety Injection.
Steam Generator Level-Low-Low is the third trip function that can be affected. When at low power levels it is not uncommon for this function to be difficult to maintain above the trip setpoint.
This function assures that there is some volume of water in the steam generators above the tops of the U-tubes to maintain a secondary side heat sink.
The amount of water is based on the decay heat present in the core and to prevent dryout of the steam generators. With the pl.nt limited to 5%
RTP or less and being at BOL on Cycle 1 there will be little or no decay heat present.
The heat source will be the core operating at the limited power level. Tripping the reactor on any of the different operable trip functions or the operator action points will assure that this require-ment will be met.
Thus, Westinghouse finds that it is acceptable to lower the trip setpoint from 18% span to 5% span for all of the special tests.
3.1.3 T.S. 3.1.1.4 MODERATOR TEMPERATURE COEFFICIENT The Moderator Temperature Coefficient is limited to O pcm/ F or more negative. When performing tests with the plant critical below 540 F this coefficient may be slightly positive. However, it is expected that the Isothermal Temperature Coefficient will remain negative or approxi-mately zero.
The tests will be performed such that this is the case and thus minimizing any impact from rapid heatups or cooldowns.
In addi-tion, the effect of a small positive Moderator Temperature Coefficient has been considered in the accident analyses performed for the test conditions.
3.2 5366A
3.1.4 T.S. 3.1.1.5 MINIMUM TEMPERATURE FOR CRITICALITY 0
The Minimum Temperature for criticality is limited to 541 F by spec.
3.1.1.5 and 531 F by spec. 3.10.3.
To perform test 4 it is expected 0
that the RCS average temperature will drop below 531 F.
Wes tin ghouse as low cs 500 F is has determined that operation with T,yg acceptable assuming that:
1.
Control Bank D is inserted to na deeper than 100 steps withdrawn, and 2.
Power Range Neutron Flux Low Set. point and Intermediate Range Neutron Flux reactor trip setpoints are reduced from 25% RTP to 7% RTP.
This will considerably reduce the consequences of possible transients by
- 1) reducing individual control rod worths (Bank D) on unplanned with-drawal, 2) reducing bank worth (Bank D) on unplanned withdrawal, 3) maximizing reactivity insertion capability consistent with operational requirements, 4) limiting maximum power to a very low value on an unplanned power excursion, and 5) allowing the use of the "at power" reactor trips as back up trips rather than as primary trips.
3.1.5 T.S. 3.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION The reactor trips noted in Section 3.1.2 will not meet the operability requirements of spec. 3.3.1.
Specification 3.3.1 can be excepted for the reasons noted in Section 3.1.2 of this evaluation.
3.1.6 T.S. 3.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION To prevent inadvertent Safety Injection and to allow performanes of the special tests, all automatic Safecy Injection functions will be blocked.
Indication of partial trips and manual initiation will be operable, however, the automatic functions will be inoperable by forcing the logic to see that the ceactor trip breakers are open. Westinghouse believes that this mode of operation is acceptable for the short period of time these tests will be carried out based on the following:
3.3 5366A
l.
'Close observationoof the partial trip indication by the opsrator, 2.
- Rigid adherence to the operator action points as defined by Westinghouse, see Section 3.2.
3.
Little or no decay heat it present in the system, thus Safety Injec-tion serves primarily as a pressurization function (shutdown margin capability is considerably more than 1.77? ^K/r for control rods at or above the insertion limits).
Blocking these functions will allow the performance of these tests at low power, pressure, or temperature and close operator surveillance will assure initiation of Safety Injection, if required, within a short time period.
3.1.7 T.S. 3.4.4 PRESSURIZER The Pressurizer provides the means of maintaining pressure control for the plant. Normally this is accomplished through the use of pressurizer heaters and spray.
In several tests the pressurizer heaters will be either turned off or rendered inoperable by loss of power.
This mode of operation is acceptable. in that pressure control will be maintained through the use of pressurizer level and charging / letdown flow.
3.1.8 T.S. 3.7.1.2 AUXILIARY FEEDWATER SYSTEM i
)
The auxiliary f eedwater system will be rendered partially inoperable for The two tests simulate some form of loss of AC power, i.e.,
f two tests.
motor driven auxiliary feedwater pumps inoperable. Westinghouse has determined that this is acceptable for these two tests because of the little or no decay heat present allowing sufficient time (# 30 min-
)
utes) for operating personnel to rack in the pump power supplies and regain steam generator level.
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f 3.1.9 T.S. 3.8.1.1, 3.8.2.1, 3.8.2.3 POWER SOURCES These_ specifications are outside Westinghouse control, however it is acceptable to alter power source availability as long as manual Safety Injection is operable and saf ety related equipment will function when required.
3.1.10 T.S. 3.10.3 SPECIAL TEST EXCEPTIONS - PHYSICS TESTS This specification allows the minimum temperature for criticality to be as low as 5310F.
Since it is expected that RCS T will be taken 3yg as-low as 5150F this specification will be excepted.
See Section 3.1.4 for basis of acceptability.
3.1.11 TECHNICAL SPECIFICATIONS NOT EXCEPTED While not applicable at power levels below 5% RTP the following tech-nical specification limits can be expected to be exceeded:
1.
3.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (g) q At low temperatures and flows F (Z) can be expected to be above q
normal for 5% RTP with RCPs running.
However at such a low power level no significant deviations in burnup or Xe peaks are expected.
2.
3.2.3 RCS FLOWRATE AND R - (F6H)
At low temperatures and flow For t..in be expected to be higher than if pumps are running. However, no significant consequences for full power operation are expected.
3.
3.2.4 QUADRANT POWER TILT RATIO With no, one, or two pumps running and critical, core power distri-butions resulting in quadrant power tilt may form.
At low power levels and for short periods of times these tilts will not signifi-cantly influence core burn-up.
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6 O
4.
3.2.5 DNB PARAMETERS In the performance of several tests the plant will be depressurized
~
below 2230 psia.
At low operating power levels this depressuriza-tion is not significant as long as subcooling margin is maintained.
3.1.12 SPECIAL TEST EXCEPTIONS 1.
Special Test Exception Specification 3.10.3 allows limited excep-tions for the following:
3.1.1.3 Moderator Temperature Coefficient 3.1.1.4 Minimum Temperature for Criticality 3.1.3.1 Movable Control Assemblies 3.1.3.5 Shutdown Rod Insertion Limits 3.1.3.6 Control Rod Insertion Limits 2.
Special Test Exception Specification 3.10.4 allows limited exception for 3.4.1.1 Reactor Coolant Loops - Normal Operation.
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3.2 OPERATIONAL SAFETY CRITERIA
~
During the performance of these tests the operator must meet the fol-lowing set of criteria for operation:
1.
For All Tests a) Primary System Sub-cooling (T Margin)
> 20 F sat b)' Steam Generator Water Level
> 33% Narrow Range Span c) Pressurizer Water Level (1) With RCPs running
> 22% Span (2)' Natural Circulation
> Value when RPCs tripped d) Loop AT
< 65 F e) T
< 580 F av8 f) Core Exit Temperature (highest)
< 610 F g) Power Range Neutron Flux Low Setpoint and Intermediate Range Neutron Flux Reactor Trip Setpoints
< 7% RTP h) Control Bank D 100 steps withdrawn or higher
- 2. Reactor Trip and Test Termination must occur if any of the following con-dicions are met:
a) Primary System Sub-cooling (T M*#8 ")
< 15 F I
sat b) SL:eam Generator Water Level
< 5% Narrow Range Span or Equivalent Wide Range Level c) NIS Pouer Range, 2 channels
< 10% RTP d) Pressurizer Water Level
< 17% Span or an unexplained decrease of more than 5% not concurrent with a T change e) Any Loop AT
> 65 F f) T,yg
> 580 F g) Core Exit Temperature (highest)
> 610 F i) Uncontrolled rod motion 3.7 5366A.
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- 3. Sefcty Injcetien must ba manus 11y initiccad if any of the following con-dicions are met:
a) Primary System Sub-cooling (T Margin)
< 100F sat b) Steam Generator Water Level
< 0% Narrow Range Span or Equivalent Wide Range Level c) Containment Pressure
> 17 psia d) Pressurizer Water Level
< 10% Span or an unexplained decrease of more than 10% not concurrent with a T,y change.
e) Pressurizer Pressure Decreases by 200 psi or more in an unplanned or unexplained manner.
Safety Injection must not be terminated until the Weatinghouse criteria as defined in E01:E-2, Loss of Secondary Coolant are met.
s These operating and function initiating conditions are selected to assure that the base conditions for safe operation are met, i.e.,
1.
Sufficient margin to saturation temperature at system pressure to assure adequate core cooling (highest temperature fuel not boiling),
2.
sufficient steam generator level to assure an adequate secondary side heat sink, 3.
sufficient level in the pressurizer to assure coverage of the heaters to maintain pressure control 4.
sufficient control rod worth to ensure adequate shutdown margin and minimize impact of uncontrolled bank withdrawal, and 5.
limit maximum'possible power level in the eveat of an uncontrolled power increase.
1 1
3.8 5366A
TABLE 3-1 TECHNICAL SPECIFICATION IMPACT Test Technical Specification 1
2 3
4 5
6 7
2.1.1 Core Safety Limits X
X X
X X
X X
2.2.1
- Various Reactor Trips Overtemperature AT X
X X
X X
X Overpower AT X
X X
X X
X Steam Generator Level X
X X
X X
X X
3.1.1.4 Moderator Temperature Coef-X ficient 3.1.1.5 Minimum Temperature for X
Criticality 3.1.1 Various Reactor Trips Overtemperature AT X
X X
X X
X Overpower AT X
X X
X X
X Steam Generator Level X
X X
X X
X X
3.3.2 Safety Injection - All X
X X
X X
X X
automatic functions 3.4.4 Pressurizer X
X X
3.7.1.2 Auxiliary Feedwater X
X X
3.8.1.1 AC Power Sources X
X
.3.8.2.1 AC Onsite Power Distribu-X X
tion System 3.8.2.3 DC Distribution System X
X 3.10.3 Special Test Exceptions -
X Physics Tests l
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4.0 SAFETY EVALUATION In this section the safety effects of those special test conditions which are outside the bounds of conditions assumed in the FSAR are evaluated. The interaction of these conditions with the transient analyses in the FSAR are discussed.
4.1 EVALUATION OF TRANSIENTS The effect of the unusual operating conditions on the transients ana-lyzed in the F3AR are evaluated.
4.1.1 CONDITION II - FAULTS OF MODERATE FREQUENCY 4.1.1.1 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from a Suberitical Condition Restriction of control rod operation to manual control, and constant operator monitoring of rod position, nuclear power and temperatures greatly reduces the likelihood of an uncontrolled RCCA withdrawal.
Operation without reactor coolant pumps, and in some cases with a posi-tive moderator temperature reactivity coefficient, tend to make the consequences of RCCA withdrawal worse compared to the operating condi-tions assumed in the FSAR. For these reasons the operating procedures require that following any reactor trip at least one reactor coolant pump will be restarted and the reactor boron concentration will be such that it will not go critical with less than 100 steps withdrawal on D Bank. An analysis of this event is presented in Section 4.2.1.
t 4.1.1.2 Uncontrolled Rod Control Cluster Assembly Bank Withdrawal at Power
'Ihe same considerations discussed in Paragraph 4.1.1.1 apply here.
In addition, the low operating power and the Power Range Neutron Flux Low and Intermediate Range Neutron Flux trip setpoints act to mitigate this 4
m 4.1 5366A
.,y
incident, while lack of che Overtemperature AT trip removes some of the protection provided in the FSAR case.
An analysis is discussed in Paragraph 4.2.2.
4.1.1.3 Rod Control Cluster Assembiv Misalignment The FSAR discussion concerning static RCCA misalignment applies to the test conditions.
The consequences of a dropped RCCA would be a decrease in power.
Thus no increase in probability or severity of this incident is introduced by the test conditions.
4.1.1.4 Uncontrolled Boron Dilution The consequences of, and operator action time requirements for, an uncontrolled boron dilution under the test conditions are bounded by those discussed in the FSAR.
The fact that the control rods will never be inserted to the insertion limits, as well as the Power Range Neutron Flux Low Setpoint and the constant operator monitoring of reactor power, temperature and charging system operation, provides added protection.
4.1.1.5 Partial Loss of Forced Reactor Coolant Flow Because of the low power limits the consequences of loss of reactor coolant pump power are trivial; indeed they are bounded by normal opera-ting conditions for these tests.
4.1.1.6 Startup of an Inactive Reactor Coolant Loop When at least one reactor coolant pump is operating, the power limit for these tests results in such small temperature dif ferences in the reactor coolant system that startup of another loop cannot introduce a signifi-cant reactivity disturbance.
In natural circulation operation, inadver-tent startup of a pump would reduce the core water temperature and thus provide a change in reactivity and power.
Because of the small mod-erator reactivity coefficient at beginning of life the power increase in the worst condition would be small and gradual and the flow-to power 4.2 5366A e
ratio in the core would be inerensing. Tha Powsr Rings Nsutron Flux Low Setpoint reactor trip prevides an upper bound on power.
Because of the increase in flow-to power ratio and because of the low setpoint on the reactor trip, DNB is precluded in this transient.
4.1.1.7 Loss of Ex:ernal Load and/or Turbint' Trip Because of the low power level, the disturbance caused by any loss of load is small.
The FSAR case is. bounding.
4.1.1.8 Loss of Normal Feedwater Because of the low power level, the consequences of a loss of feedwater are bounded by the FSAR case.
In the case of loss of all feedwater sources, if the reactor is not shutdown manually, it would be tripped on Low-Low Steam Generator Water Level.
Ample time is available to re-institute auxiliary feedwater sources.
4.1.1. 9 Loss of Offsite Power to the Station's Auxiliaries (Station Blackout)
Because of the low power level, the consequences of a loss of off-site power are bounded by the FSAR case.
4.1.1.10 Excessive Heat Removal Due to Feedwater System Malfunctions The main feedwater control valves will not be used while the reactor is at power or near criticality on these tests. Thus, the potential water flow is restricted to the main feedwater bypass valve flow or auxiliary
+
feedwater flow, about 15% of normal flow.
The transient is further mitigated by the low operating power level, small moderator temperature reactivity coefficient, the low setpoints on the Intermediate and Power Range Neutron Flux Low setpoint trips, and close operator surveillance of feed flow, RCS temperatures, RCS pressure, and nuclear power.
The case of excess heat removal due to feedwater system malfunctions with very low reactor coolant flow 1s among the cooldown transients discussed in more detail in Sec, tion 4.2.3.
4.3 5366A
4.1.1.11 Excessiva Load Increnso Incidsnt The turbine will not be in use during the performance of these tests, and load control will be limited to operation of a single steam dump or steam relief valve.
The small moderator temperature reactivity coef-ficient also reduces the consequences of this transient. Close operator surveillance of steam pressure, cold leg temperature, pressurizer pres-sure, and reactor power, with specific initiaties criteria for manual reactor trip, protect against an unwanted reactor power increase.
In addition, the low setpoints for Power-Range and Intermediate-Range Neu-tron Flux reactor trips limit any power transient.
In addition, modifi-cation of the High Steamline Flow setpoint allows a reactor trip on Law Steam Pressure only. Analyses are discusssed in Section 4.2.3.
4.1.1.12 Accidental Depressurization of the Reactor Coolant System Close operator surveillance of pressurizer pressure and of hot leg sub-cooling, with specific initiation points for mar.ual reactor trip, pro-vides protection against DNB in the event of an accidental depressuriza-tion of the RC3.
In addition, automatic reactor trip caused by the Low Pressurizer Pressure Safety Injection signal would occur when core out-0 let subcooling reached approximately 8 F as an automatic backup for manual trip. During test 3 and 5, when this trip is bypassed to allow deliberate operation at low pressure, the pressurizer PORV block valves will be closed to remove the major c. edible source of rapid inadvertent depressurization.
(The Lov Pressure trip is automatically reinstated when pressure goes above 1990 psi and the PORV block valves will be reopened at that time.)
4.1.1.13 Accidental Depressurization of the Main Steam System The FSAR analysis for accidental steam system depressurization indicates that if the transient starts at hot shutdown conditions with the worst RCCA stuck out of the core, the negative reactivity introduced by Safer, Injection prevents the core from going critical. Because of the small moderator temperature reactivity coefficient which will exist during the 4.4 5366A
test period, the reactor would remain suberitical even if it were cooled to room temperature without Safety Injection. Thus the SAR analysis is bounding.
4.1.1.14 Spurious Operation of the Safety Injection System at Power In order to reduce the possibility of unnecessary therms 1 fatigue cycling of the reactor coolant system components, the actuation of high head charging in the safety injection mode, and of the safety injection pumps, by any source except manual action will be disabled.
Thus, the likely sources of spurious Safety Injection, i.e., spurious or most
" spike" pressure or pressure-difference signals from the primary or secondary systems, have been eliminated.
4.1.2 CONDITION III - INFREQUENT FAULTS 4.1.2.1 Loss of Reactor Coolant from Small Ruptured Pipes or from Cracks in Large Pipes Which Actuates Emergency Core Cooling A review of the plant loss of coolant accident behavior during the low power testing sequence indicates that without automatic Safety Injection there is sufficient cooling water readily available to prevent the fuel rod cladding from overheating on a short term basis.
The system inven-tory and normal charging flow provide the short term cooling for the l
small break transient. A sample calculation for a 2 inch break in a four loop plant shows that the core remains covered for at least 6000 seconds. While that analysis is not directly applicable to North Anna, it does provide an order of magnitude estimate of the time available for operator action. The time to uncover the core in the three loop care should be in excess of one hour using the four loop calculation as a guideline. This is sufficient time for the operator to manually initi-ate SI and align' the system for long term cooling.
It must be noted that the magnitude of the resulting clad heatup tran-sient during a LOCA event from these conditions is significantly reduced from-the FSAR basis scenario by the low decay heat and core stored energy resulting from the low power level and short operating history.
4.5 5366A
.l
4.1.2.2 Minor Secondary System Pipe Breaks The consequences of minor secondary system pipe breaks are within the bounds discussed in Paragraph 4.2.3.
4.1.2.3 Single Rod Cluster Control Assembly Withdrawal at Power The FSAR analysis shows that assuming limiting parameters for normal operation a maximum of 5 percent of the fuel rods could experience a DNBR of less than 1.3 following a single RCCA withdrawal. As the FSAR points out, no single electrical or mechanical failure in the control system could cause such an event.
The probability of such an event happening during the test period is further reduced by the short dura-tion of this period, by the restriction to manual control, and by the close operator surveillance of reactor power, rod operation, and hot leg temperature.
4.1.2.4 Other Infrequent Faults The consequences of an inadvertent loading of a fuel assembly into an improper position, complete loss of forced reactor coolant flow, and waste gas decay tank rupture, as described in the FSAR, have been reviewed and found to bound the consequences of such events occurring during test operation.
4.1.3 CONDITION IV - LIMITING FAULTS 4.1.3.1 Major Reactor Coolant Pipe Ruptures (Loss of Coolant Accident)
A review of the plant loss of coolant accident behavior during the low power testing sequence indicates that without automatic safety injection there is sufficient cooling water readily available to prevent the fuel rod cladding from over heating on a short term basis. During the large break event the system inventory and cold leg accumulators will have removed enough energy to have filled the reactor vessel to the bottom of the nozzles. Following the system depressurization there is enough 4.6 5366A e
e*
e e
water in the reactor vessel below the nozzles to keep the core covered for over one hour using conservative assumptions.
This is sufficient time for the operator to manually initiate SI and align the system for long teen cooling.
At no time during this transient will the core be uncovered.
It must be noced that the magnitude of the resulting clad heacup tran-sient during a LOCA event from these conditions is significantly reduced from the FSAR basis scenario by the low decay heat and core stored energy resulting from the low power level and short operating history.
4.1.3.2 Major Secondary System Pipe Rupture The small moderator temperature reactivity coefficient, close operator surveillance of pressurizer pressure, cold leg temperature, and reactor pe-eer, with specific initiation criteria for reactor trip; low trip secpoints on the Intermediate-Range and Power-Range Neutron Flux trips; Low Flow Mismatch setpoint for Reactor Trip and MSIV closure on High Steamline Flow in coincidence with Low Steam Pressure; and Low Pressurizer Pressure trip (S.I. initiation) assure a Reactor Trip without excessive reactor power following a cooldown transient caused by the secondary system. Following reactor trip, assuming the worst RCCA stuck out of the core, the reactor would remain suberitical even if it were cooled to room temperature. Transient analyses for a steam pipe rupture are provided in Section 4.2.3.
The consequences of a main feedline rupture are bounded in the cooldown direction by the steam pipe rupture discussion. Because of the low operating power, the heatup aspects of a feedline rupture are bounded by the FSAR discussion.
I 4.1.3.3 Steam Generator Tube Rupture 4
The steam generator tube rupture event may be categorized by two dis-tinct phases. The initial phase of the event is analogous to a small LOCA event. Prior to operator-controlled system depressurization, the steam generator tube rupture is a special class of small break LOCA
- l 4.7 5366A
o e
transients, and the operator actions required to deal with this situa-tion during this phase are identical to those required for mitigation of a small LOCA. Hence, evaluation of the steam generator tube rupture during this phase is wholly covered by the safety evaluation of the small LOCA.
After the appropriate operator actions have taken place to deal with the initial LOCA phase of the event, the remainder of the steam generator
. tube rupture accident mitigation would consist of those operator actions required to isolate the faulted steam generator, cooldown the RCS, and depressurize the RCS to equilibrate primary RCS pressure with the faulted steam generator secondary pressure. These actions require util-ization of the following systems:
1.
Auxiliary feedwater control to the faulted steam generator.
2.
Steam line isolation of the faulted steam genarttor.
3.
Steam relief capability of at least one non-faulted steam generator.
4.
RCS depressurization capability, f
Evaluation of the VEPCO special test procedures has verified that all of the above systems are immediately available for operator control from the control room. Therefore, it is concluded that the ability to miti-gate the steam generator tube rupture event is not compromised by the modifications required for operation at 5% power during the proposed tests, and that the analyses performed for the SAR regarding this event remain bounding.
4.1.3.4 -Single Reactor Coolant Pump Locked Rotor Because of the low power level, the locking of a single reactor coolant pump rotor is inconsequential.
4.8 j
5366A
F
~
N 4.1.3.5 Fuel Handling Accidents The FSAR analysis of fuel handling accidents is bounding.
4.1.3.6 Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection)
The control rod bank insertion will be so limited (i.e., only Bank D inserted, with at least 100 steps withdrawn) that the worth of an ejected rod will be substantially less than the delayed neutron frac-tion. Thus, the power rise following a control rod ejection would be relatively gradual and terminated by the Power Range and Intermediate Range Neutron Flux reactor trips. While the core power transient and power distribution following an RCCA ejection at this time would be less severe than those shown in the FSAR, the result of combining these ame-liorating effects with the effect of the natural circulation flow rate on clad-to-water heat transfer and RCS pressure have not been analyzed.
The extremely low probability of an RCCA ejection during tits brief period in the test sequence does not warrant such an analysis.
4.2 ANALYSIS OF TRANSIENTS 4.2.1 ANALYSIS OF RCCA BANK WITHDRAWAL FROM SUBCRIT.ICAL CONDITION An analysis was performed to bound the test transients. The methods and assumptions used in the FSAR, Section 15.2.1 were used with the foi-lowing exceptions:
1.
Reactor coolant flow was 0.1% of nominal.
2.
Control rod incremental worth and total worth were upper bound values for the D bank initially 100 steps withdrawn.
3.
Moderator temperature reactivity coefficient was an upper bound (positive) for any-core average temperature at or above 500oF.
4.9 5366A
4.
Th3 lovar b:und for tocc1 d31cysd nsucron frcction for th2 baginning of life for Cycle 1 was used.
5.
Reactor trip was initiated at 10% of full power.
6.
DNB was assumed to occur spontaneously at the hot spot, at the beginning of the transient.
The resulting nuclear power peaked at 68% of full power, as is shown in Figure 4.2.1.
The peak clad temperature reached was under 1200 F, as is shown in Figure 4.2.2.
No clad failure is expected as a result of this transient.
4.2.2 ANALYSIS OF RCCA BANK WITHDRAWAL AT POWER Analyses of RCCA bank withdrawal transients were performed for natural circulation conditions. The transients were assumed to start from steady-state operating conditions at either 1% or 5% of full power, and with either all steamline isolation valves open or two of those valves closed. A rmage of reactivity insertion rates up to the maximum for two banks moving was assumed for cases with all steamlines open, and up to the maximum for one bank moving for the cases with two steamlines iso-lated. Both maximum and minimum bounds on reactivity feedback coef-ficients for beginning of life, Cycle 1, were investigated.
In all cases, reactor trip was initiated at 10% nuclear power.
Reactor conditions at the rime of maximum core heat flux are shown in Figures 4.2.3 and 4.2.4 as functions of the reactivity insertion rate f or three three-loop active cases.
For high reactivity insertion rates, the minimum reactivity coefficient cases give the greatest heat flux after the trip setpoint is reached, and have the lowest coolant flow rate at the time of peak heat flux.
For these cases even the slowest insertion rates studied did not result in any increase in core inlet temperature at the time of peak heat flux. For maximum feedback cases, however, the transients for very low insertion rates go on for to long 4
o 4.10 5366A
th0t th3 cero inlot temp:rature finally increcsss bafore trip, i.e.,
after approximately one and one-half c!.nutes of continuous withdrawal.
Thus, the cases shown bound the worst cases.
4.2.3 ANALYSIS OF C00LDOWN TRANSIENTS Cooldown transients include feedwater system malfunctions, excessive steam load increase, accidental depressurization of the main steam sys-tem, and minor and major secondary system pipe ruptures. Attention has been focused on the possibility and magnitude of core power transients resulting from such cooldowns before reactor trip would occur.
(Follow-ing reactor trip, no cooldown event would return the reactor to a critical condition.)
During natural circulation operation, approximately one to two minutes would elapse following a secondary side event before cold water from the steam generator reached the core; thus, considering the close and con-stant surveillance during these tests, time would be available for the operator to respond to such an event.
Analyses were aise performed to determine the ex'ent of protection provided by automatic protection t
systems undh-trip conditions.
4.2.3.1 Load Increases A load increase or a small pipe break, equivalent to the opening of a single power-operated steam pressure relief valve, a dump valve, or a safety valve, would cause an increase of less than four percent in reac-tor power, with a corresponding increase in core flow with natural cir-culation, assuming the bounding negative moderator temperature coeffi-cient for the beginning of life, Cycle 1.
Thus no automatic protection is required, and ample time is available to the operator to trip the reactor, isolate feedwater to the faulted steam generator, and isolate the break to the extent possible. Calculated results for the sudden opening of a single steam valve, assuming the most negative BOL Cycle one moderator reactivity coefficient are shown in Figures 4.2.5 and 4.11 5366A
4.2.6 (5% initial power, steam dump valve), and in Figures 4.2.7 and 4.2.8 (1% initial power, one steam generator isolated, pressure relief valve.)
4.2.3.2 High Flux Protection L
Reactor trip on high nuclear flux provides backup protection for larger pipe breaks or load increases. Analyses were performed to determine the worst core conditions that could prevail at the time of high-flux trip, independent of the cause. The following assumptions were used:
1.
Upper-bound negative moderator isothermal temperature coefficient, vs. core average temprature, for beginning of life, Cycle 1.
2.
Lower-bound fuel temperature - power reactivity coefficient.
3.
Initial operation with core inlet temperature 555oF.
4.
Initial powers of 0% and 5% of full power were analyzed.
5.
Hot leg coolant at incipient boiling at the time of reactor trip.
This results in some boiling in the reactor. The negative reac-tivity introduced by core boiling would effectively limit power; this negative reactivity was conservatively neglected.
6.
Uniform core inlet temperature and flow.
7.
Reactor trip equivalent to 10% of full power at the initial inlet temperature.
The power as measured by the NIS is assumed to be diminished from the true power by 1% for each loF decrease in reactor inlet temperature, resulting in a true power of greater than 10% at the time of trip.
8.
Core flow rate as a function of core power was assumed equal to the i
predicted flow under steady-state operating conditions.
4.12 5366A.
O
Analyses of core conditions based on these assumptions indicate that the DNB criterion of the FSAR is met.
4.2.3.3 Secondary Pressure Trip Protection Large steamlina ruptures which affect all loops uniformly will actuate reactor trip end steamline isolation on Low Steamline Pressure signals in any two lines, because the required coincident High Steamline Flow signal is set to zero flow. Low Pressurizer Pressure and Power Range Neutron Flux low setpoint trips serve as further backups.
An example is
-the double ended rupture of a main steamline outside of the steamline check valves and isolation valves, with all isolation valves initially open. Figures 4.2.9 and 4.2.10 show the rasponse to such an event, with an initial power of 5% and natural circulation. The Low Steamline Pres-j sure trip occurs almost immediately.
In the example shown, the main steamline isolation valve on loop one was assumed to f ail to close. No power excursion resulted, and the reactor remained suberitical after the trip.
For large steamline ruptures inside one of the steamline check valves, pressure: difference between steam generators would normally actuate reactor trip and main steamline isolation. For test 4, the desired pressure difference is greater than the setpoint range. This system j
will be disabled for all tests.
Power Range Neutron Flux low setpoint thus serves as backup to manual protection. An example case is shown in Figures 4.2.11 and 4.2.12.
For this case operation of 1% power was a
assumed with steam generator B isolated and with natural circulation. A double-ended rupture of the main steamline upstream of the steamline venturi was assumed. Reactor trip and steamline isolation would be initiated by steam generator pressure difference in approximately fif-
~
teen seconds. For this calculation, however, the transient was allowed to continue without trip until the Power Range Neutron F}ux low setpoint trip was reached at 104 seconds. The reactor remai ae t suberitical after the trip.
4.13 5366A
4.3 ADDITIONAL' CONSIDERATIONS In the great majority cf ctses it was concluded, either by reanalysis or by comparison with previously analyzed FSAR conditions, that fuel clad integrity would be maintained without need for operator mitigating action. For the LOCA or steambreak events, it was concluded that the operator would have more than ample time (> 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) to respond by manual action, e.g., manually initiate safety injection, to preclude fuel damage.
Finally, in certain other cases, primarily associated with certain inadvertent RCCA withdrawal events, the postulated accident conditions were neither amenable to direct analysis nor credit for operator inter-vention.
In particular, the postulated accident conditions were outside the bounds of accepted analysis techniques so that fuel damage was not precluded either by analysis or identified operator action. For these cases, the basis for acceptability was primarily associated with the low probability of an inadvertent rod withdrawal event during the limited duration of the special tests.
This section provides an additional assessment relative to the potential for and consequences of fuel failure for these "unanalyzed" accident conditions associated with certain roc withdrawal events.
This assessment is partially based upon an attempt to bound certain ef fects which may exist for conditions removed from the range of direct model applicability. Additional information (attached) is provided for four areas:
1.
Thermal margin associated with normal test conditions.
2.
The potential for DNB during accident conditions.
3.
The clad temperature response assuming that DNB occurs.
4.
Radiological consequences associated with presumed gross fuel failure.
4.14 5366A
Tha c:nclusiona of-this cescssE:nt era se follows:
1.
DNB is not expected for the limiting thermal condition associated with any RCCA withdrawal event.
2.
Even assuming DNB, there should be adequate heat transfer to
. prevent clad overheating.
3.
Fuel clad failure is not expected.
4.
Even assuming 100% clad failure and other extreme conservatisms, the resulting offsite dose would be small.
4.3.1 DESIGN CONSIDERATIONS Margin to hot channel boiling has been incorporated with all normal test conditions by establishing a lower bound requirement on the degree of reactor coolant subcooling. This test requirement assures that postu-laced accidents are initiated from a condition of excess thermal margin.
4.3.2 DNB CONSIDERATIONS For certain cooldown transients, the conclusion that DNB is precluded was drawn based on use of the W-3 critical heat flux correlation.
Although the analyses for the cooldown events discussed in section 4.2.3.2 result in mass velocity below the range of direct applicability of_the correlation, the reactor heat flux was so low relative to the predicted critical heat flux that even a factor of 2 would not result in serious concern for DNB for this event.
For the non-cooldown transients the limiting conditions, with respect to DNB, are farther away from the W-3 range of applicability because the coolant temperature is higher and the power-to-flow ratio is larger.
Comparison ~of the W-3 DNB cbrrelation to low flow DNB test data and correlations (references 1 and 2) indicate that it will conservatively 4.15 5366A
'l
-l'j prddict critical heat flux at low pressure (# 1000 psi) conditions with low coolant flow. Pool boiling' critical heat flux values (reference 3) at these pressures are higher than those predicted by the low flow correlations. Further review of the data in reference 1 indi-
~
cates that the critical heat flux at higher-pressure is significantly icaer-than the above data at 1000 psi. The minimum critical heat flux of the data set is.16 x 106 BTU /hr-f t2 for a data point at 2200 psia at a mass -.ocity.of.2 x 106 lbm/hr-ft2, Since the exit quality for this data point was 64%, it is unlikely that the reactor would be able to maintain a heat flux of that level due to
'the nuclear feedback from voiding. The power distribution would tend to peak towards the bottom thus further reducing the local quality at the peak flux locations.
Also the pool boiling correlations in reference 3 show some decrease in critical heat flux above 1000 psia to the maximum pressure of applica-bility of 2000 psia. However extrapolation of the correlations to a value of zero critical heat flux at the critical pressure (3206.2 psia) would not result in lower critical heat flaxes than shorn in the data set from reference 1.
Since the core average heat flux at 10% of 4
nominal power (highest expected power for heatup events) is only on the order of.02 x 106 BTU /hr-ft2 a large peaking factor would be required to put the reactor heat flux as high as the critical heat flux.
For the transients considered the only ones that lead to significant off normal the peaking factors are the rod motio'n transients.
The rod with-drawal from suberitical is a power burst concern. As such, it is expected that even if DNB occurred, the rod surface would rewet.
For the rod bank withdrawal, the combination of maximum power and peaking factor would result in a peak power lower than the data referenced above. Given the lack of data, it-is difficult to completely preclude DNB, although a prudent judgement indicates that it is indeed remote.
4.16
-5366A'
4.3.3 CLAD TEMPERATURE CONSTDERATIONS Should DNB occur, the peak clad temperature reached would depend pri-marily on the local _ nuclear transient following DNB and on the behavior of the post-DNB heat transf er coe fficient.
For a rapid power transient, as is illustrated by the SER analysis for RCCA bank withdrawal from a'subcritical condition, the fuel temperature reactivity feedback and reactor trip on a nuclear flux signal would shut down the reactor bef ore sufficient energy could be generated to cause a damaging rise in clad temperature.
In that case, the maximum clad temperature calculated was under 1200 F even assuming an extremely low 2
heat transfer coefficient (# 2 BTU /hr-ft OF).
A possibly more limiting condition for RCCA withdrawal would be the case in which a power increase causes DNB but would either not result in reactor trip on high nuclear flux or the trip is delayed.
In the former case, a steady state condition with hot spot DNB could be postulated.
In this state the clad temperature could be calculated given only the total core power, local heat flux channel factor, heat transfer coefficient and saturation temperature.
The core power is postulated to be essentially at the power which would cause a reactor trip on high Power Range Neutron Flux low setpoint. The trip setpoint is at 7% for these tests. To allow for calorimetric errors and normal system errors trip is assumed to occur at 10% of rated thermal power (RTP), unless a large decrease in downcomer coolant temperature occurs during the test.
In tests 3 and 5, depressurization to less than approximately 1450 psia could require temperature reduc-tion, as is indicated in Figure 4.3.1; however, such low pressures are not expected.
Figure 4.3.2 shows the relationship of peak clad temperature, local heat transf er coefficient, and the product of heat flux hot channel factor (F ) times core power (fraction of RTP). For the event of an q
uncontrolled RCCA bank or single RCCA the upper bound of this heat flux 4.17 5366A.
i i
=
.ve e
c product is approkimately 0.34.
Using this value, the heat transfer coefficient required to keep the peak clad temperature below 1800oF, the threshold of significant heat flux increases due to zirconium-water reaction, can be found from Figure 4.3.2.
Various film boiling heat transfer correlations have been received to evaluate the heat transfer coefficient for post-DNB conditions.
Although no correlations were found which cover the complete range of conditions being tested, some data exist which can be extrapolated to obtain representative heat transfer coefficients. The Westinghouse UHI film boiling correlation (reference 4), was developed at low flow con-ditions similar to those postulated for incidents occurring during the VEPCO tests.
This correlation was extrapolated to the higher pressure conditions of the tests to obtain representative film boiling coeffi-cients.
This resulted in a heat transfer coefficient in excess of (100 BTU /hr-ft2-oF)a,c at 2200 psia and 5% flow with quality between 10-50%. Other film boiling heat transfer correlations, devel-oped at higher pressures, were also examined.
These correlations were extrapolated down to the lower flow conditions of the VEPCO tests as another approach to obtain representative film boiling coefficients.
.Using both the Mattson et al (reference 5) and the Tong (refarence 6) film boiling correlations resulted in post-DNB heat transfer coeffi-cients in excess of 150 BTU /hr-ft2-oF at the conditions given above.
These results indicate that a clad temperature excursion resulting in fuel damage is not likely to occur even if DNB is assumed.
4.3.4 DOSE ANALYSIS CONSIDERATIONS The dose analyses were performed for a hypothetical accident senario using conservative assumptions so as to determine an extreme upper bound
.on postulated accident consequences. The analysis assumed a reactor accident involving no pipe-break with a coincident loss of condensar vacuum. This accident scenario is representative of the Condition II type events analyzed in the FSAR. The bounding were assumptions made in the analysis which include:
4.18 5366A'
139 Mut (5% power) 1.0 dose-equivalent I-131 RCS activity (tech spec limit)
'500_gpd steam generator leak in each SG (tech spec limit) 100% clad damage and gap activity release 10% iodine / noble gas in gap space 100 DF in steam generators 500 iodine spike factor over steady state 509,000 lb. atmospheric steam dump over 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.7 x 10-3 sec/m3 x/Q percentile value The results of the analysis show that the two hour site boundary doses would be 5 rem thyroid, 0.9 rem total body and 0.4 rem to the skin.
The analysis of the accidents has incorporated some very conservative assumptions which goes beyond the normal degree of conservatism used in FSAR analyses. The most prominent of these assumptions and a brief description of the extreme conservatism includes:
- 1) Equilibrium radionuclide inventories established at 5% power. For iodines, this requires # 1 month of steady state operation at 5%
uninterrupted.
- 2) Fuel clad gap inventories at 10% of core inventory, this is a time dependent, temperature dependent phenomona. At 5% power, very little diffusion to gap space is expected for the short test period.
3) 100% fuel rod clad damage.
- 4) Primary to secondary leakage to tech spec values.
Since North Anna is a new plant, no primary to secondary leakage is expected.
If leakage were present, it would most likely slowly increase in steps up to tech spec levels.
= 5) Percentile meteorology, there is 95% probability of better diffusion l
characteristics and thus lower offsite doses.
l t
4.19 5366A-8 9
.9 66
For these reasons, in the unlikely event of a potential accident during the tests, the resulting dose is small, even assuming 100% clad damage and other extrees conservatisms.
4.3.5 OTHER CONCERNS The LOCA analyses presented indicate that there is over 6,000 seconds for the operator to take action.
This is more than sufficient time for the operator to take corrective action.
Some transients wete not analyzed or discussed in this supplement due to the combination of the low probability of the transient occurring and the very short time period of the special tests. This is true for the rod ejection accident. The combination of the low probability of occurring and the bounding dose evaluation for a condition II transient given here in-dicate that these events do not need to be analyzed.
Similar dose calculations have been done for the steamline break accidents which j
results in somewhat higher doses than the condition II analysis.
These dose results indicate that the fact that the NIS channels are not completely qualified does not alter the conclusion that the results are bounded.
1 l
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4.20 5366A
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REFERENCES 1.
J. S. Gellerstedt, R. A. Lee, W. J. Oberjohn, R. H. Wilson, L. J.
Stanek, " Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water," Symposium on Two-Phase Flow in Rod Bundles, Code H27, ASME Winter Annual Meeting, November,1969.
2.
Hao,15.
R.,
Zielke, L.
A.,
Parker, M.
B., " Low Flow Critical Heat Flux," ANS 22, 1975.
3.
Lahey, R.
T., Moody, F.
J., "The Thermal-Hydraulics of Boiling Water Nuclear Reactor," American Nuclear Society, 1977.
4.
WCAP-8582-P, Vol. II, " Blowdown Experiments With Upper Head Injection in G2 17x17 Rod Array," McIntyre, B.
A., August, 1976.
(Westing' house Proprietary) 5.
Mattson, R.
J., Condie, K.
G., Bengston, S. J. and Obenchain, C.
F.,
" Regression Analysis of Post-CHF Flow Boiling Data," paper B3.8, "Vol. 4, Proc. of 5th Int. Heat Transfer Conference, Tokyo, September (1974).
6.
Tong, L.
S., " Heat Transfer in Water-Cooled Nuclear Reactors," Nuc.
Engng, and Design 6, 301 (1967).
4.21 5366A 9
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1 10 100 REACTIVITY INSERTION RATE (PCM/SEC) i Figure 4.2.4 Uncontrolled Rod Bank Withdrawal at Power.
Time of Reactor Trip vs Reactivity Insertion f
Rate 4.25 t
~
.10000 5 s
.08000 --
d g-O
.06000 --
< z Y
O
.04000 --
w a
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me
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,3
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u
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50.0 m
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< R Cold Leg v e w
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=
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,z w
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.05000 --
v
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w r
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m TIME (SEC)
FIGURE 4.2.5 TRANSIENTS IN THE REACTOR CORE AND COnLANT LOOPS FOLLOWING THE OPENING OF A STEAM DUMP VALVE FROM 5% POWER, ALL LOOPS ACTIVE 4.26
g 2400 0 S
w M
2200.0 --
q
=
2000.0 - r
~
-m 2e o
1800.0 --
w fr i
b i
2 o>
1500 0
=
5 " 1000 00 --
u w m
= 3 500.00 g-o 0.0 E
1200 0 l-1000 00 --
All toops
_2 800 00 a
waa 600 cc --
w Y
a.
X 5
400 00 --
- e 200 00 --
0.0 a
e e
e a
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o.
e o
e o
o e
o o
o o
o o
o o
o o
ou m
a e
w e~
m o
TIME (SEC)
FIGURE 4.2.6 TRANSIENTS IN THE PRESSURIZER AND STEAM GENERATOR FOLLOWING THE OPENING OF A STEAM DUMP VALVE FROM 5% POWER, ALL LOOPS ACTIVE 4.27
.10000 c
55
.08000 -
d=
p@
.C5000 - -
g
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a w
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- 0. 0 D
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~
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u 0.0 W
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l l
l Core Avg Loof 3 o
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Loop 2 w w m
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l
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w o
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o o
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u m
8 m
o cu m
TIME (SEO
~
FIGURE 4.2.7 TRANSIENTS IN THE REACTOR CORE AND COOLANT LOOPS FOLLOWING THE OPENING OF A SINGLE STEAM RELIEF VALVE AT 1%
POWER,One STEAM GENERATOR ISOLATED 4.28
W 2400.0 2200.0 --
I
~:-
=
.y "E
2000.0 --
=
3mm 1800.0 --
w CC e
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w1:s gg r
1500.0 --
=
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o
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- 0. 0 l
l l
x L
1200.0 l
Loop 3 1000.00 -
Loop 2 800.00 --
2:
w=
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Loop 1 m
wa*
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e--
m 200.00 -
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o o
o ci o
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o o
o o
o o
o o
o o
o o
o fu M
e e
up N
CD Ti.p**
e f..* 9 FIGURE 4.2.8 TRANSIENTS IN THE PRESSURIZER AND STEAM GENERATOR FOLLOWING THE OPENING OF A SINGLE STEAM RELIEF VALVE AT l
1% POWER, One STEAM GENERATORS ISOLATED 4.29
i 1
.10000 O
55
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06000 <-
EE
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w a s
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m a
TIME (SEC)
FIGURE 4.2.9 TRANSIENTS IN THE REACTOR CORE AND COOLANT LOOPS FOLLOUING A DOUBLE ENDED RUPTURE OF A MAIN STEAM-LINE DOWNSTREAM 0F THE STEAMLINE ISOLATION AND
, HECK VALVES AT 5% POWER, ALL LOOPS ACTIVE C
4. 30.,
y 2400.0
- )
M M
M 2200.0 a.
. -(
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2000.0
~ ~
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1800.0 -
w
=.
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1500.0 --
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0.0 o
o o
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o o
a o
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o 5
d d
o n.
m n
w o
TIME (SEC)
FIGURE 4.2.10 TRANSIENTS IN THE PRESSURIZER AND STEAM GENERN
(
FOLLOWING A DOUBLE ENDED RUPTURE OF A MAIN IAhLINE DOWNSTREAM OF THE STEAMLINE ISOLATION A1T. DECK VALVES AT 57. POWER, ALL LOOPS ACTIVE 4.31
10000 z.
E s
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d2
- o
.06000 -
<m w
=m
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.02000 --
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C.02000 --
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7.
e 7
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600.00 Loop 3 awm w
e C 500.00 --
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w O<
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av 200.00 l
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f
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2 I
10000 -- Loop 1 Loop 2
\\
u
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a Loop 3
$ ~
0.0 g
g g
g g
8 o
o o
6 5
o e
o 5
5 5
o 2
C 2
o
~
o
~
TIME (SEC)
FIGURE 4.2.11 TRANSIENTS IN THE REACTOR CORE AND COOLANT
~1 LOOPS FOLL0%'ING A DOUBLE NDED RUPTURE OF A MAIN STEAMLINE UPSTREAM JF THE STEAMLINE VENTURI AT 1% P0h'ER, LOOP ISOLATED 4.32
y 2400 0 g
^
0 2200.0
=. _
a g 2000 0 m~
e E3 1800.0 3
E 1600.0 Y
- 3 g
1500 0
=-
OC
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$ 0 1000 00 -
,u w ao
~ =
500 00 --
0 0.0 a
G.
1200 0 Loop 3 1000 00 E
p 800 00 Loop 2 w
E3 600 00 -
e E
a.
1 5
400 00 --
toop 1 G
200 00 -
0.0 o
o e
o e
o e
o o
o o
o o
o o
o o
o o
a o
e o
e o
d C
S C
C O
TIME (SEC)
FIGURE 4.2.12 TRANSIENTS IN THE PRESSURIZER AND STEAM GENERATOR FOLLOWING A DOUBLE ENDED RUPTURE OF A MAIN STEAMLINE UPSTREAM OF THE STEAMLINE VENTURI AT 17. POWER, One LOOP ISOLATED 4.33 98 e
i e
i i
.i 2500 2400 -
Limited by 1600 psi P
-P
=
Primary Secondary 2300 -
- 11. 2, 7 r
g___
._ (1 & 3%)
2200-L' p
/
2100
/
,/
J 2000-3 j
Test No.
4 i
wer Level N)
RCS 1900
\\
Pressure.
.{
Psig 1800 J (3 + 1 1/2%)
~
j 3, 5 l
/
a 1700 1600.
1 1500.
)
/
l 1400.
/
/'4 Limited by Hot Leg Subcooling = 20*F at 1.5%
1300
\\
/
i 1200 460 480 500 520 540 560 580 Active Cold Leg Temperature. *F FIGURE 4.3.1 NATURAL CIRCULATION TEST CONDITIONS i
120 '
100 Peak Clad Temperature = 1800*F 80
_Ileat Transfer Coefficient Pressure a 2600 psia Btu /llr F Ft.2 60
.w
-1600 psia m
40 20 0-0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 F x Power Fraction q
FIGURE 4.3.2 IIEAT TRANSFER COEFFICIENT VS. IIEAT FLUX FOR CLAD TEMPERATURE OF 1800 F-I
e ENCLOSURE 2 ADDITIONS TO TECHNICAL SPECIFICATION TABLE 3.7-4 e
a
4
- E5-TABLE 3.7-4 (Continued)
.x
];
SAFETY RELATED HYDRAULIC SNUBBERS
- 5 h
HIGH RADIATION 35 SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR DURING SHUTDOWN ESPECIALLY DIFFICULT NO.
ON, LOCATION (ELEVATION-AREA)# INACCESSIBLE ZONE **
TO REMOVE h)
( A or-I)
(Yes or No)
(Yes or No) 518-2 RC-260-B I
'Yes Yes 523-1 RC-242-B I
Yes
-Yes 444-3 RC-256-C I
Yes Yes 445-3 RC-256-A I
Yes Yes u,
)2 446-3 RC-256-8 I
Yes Yes
[
447-1 RC-256-A I
Yes Yes 447-2 RC-256-A I
Yes Yes 448-1 RC-256-B I
Yes
-Yes 448-2 RC-256-8 I
Yes Yes 449-1 RC-256-C I
Yes Yes 449-2 RC-256-C I
Yes Yes 453-29 RC-236-A I
Yes Yes 453-30A RC-238-A I
Yes No 453-308 RC-238-A I
Yes No 453-31 RC-243-A I
Yes No 455-1 RC-243-A I
Yes No 458-1 RC-242-C I
Yes No 458-2 RC-242-C I
Yes No 459-2 RC-243-C I
Yes No 4
h.
E
~ 25 TABLE 3.7-4 (Continued)
.x
' ]E SAFETY RELATED HYDRAULIC SNUBBERS
- E e
c.
HIGH RADIATION 25 -
SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR DURING SHUTDOWN ESPECIALLY' DIFFICULT
~ -d NO.
ON, LOCATION (ELEVATION-AREA)# INACCESSIBLE ZONE **
TO REMOVE
^*
(A or I)
(Yes or No)
(Yes or No) 466-1 CH-224-RCB I
No Yes' 482-2 CH-226-RCB I
No Yes 483-2 CH-226-RCB I
No Yes 633-1A CH-236-RCB I
No Yes u,
'32 633-1B CH-236-RCB I
No Yes
[(
633-4 Cli-236-PCB I
No Yes 801-25 CH-237-RL3 I
No Yes 801-26 CH-237-RCB I
No Yes 492-42A CH-255-C I
Yes
.Yes 492-428 CH-255-C I
Yes Yes 304 QS-255-RCA I
No Yes 301 QS-255-RCA I
No Yes 706 QS-267-QS A
No No 800 QS-274-SG A
No No l
303A RS-260-RCA I
No Yes 303 RS-260-RCA I
No Yes 300 RS-333-RCA I
No Yes
_ TABLE 3.7-4_(Continurd)
SAFETY RELATED HYDRAULIC SNUBBERS SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT 35 NO.
ON, LOCATION (ELEVATION-AREA)#
INACCESSIBLE ZONE **
TO REMOVE
[,
_(A or I)
(Yes or No)
(Yes or No)
. E..
8 800 SI-258-SG A
No No
-. g 801 SI-258-SG A
No Yes 802A SI-258-SG A
No No 803 SI-258-SG A
No No 8 04 SI-258-SG A
No No 806 SI-259-SG A
No Yes 807A SI-259-SG A
No Yes w3 8078 SI-259-SG A
No Yes y
h 808 SI-253-SG A
No Yes 809A SI-258-SG A
No No 809B SI-258-SG A
No E
No 810A SI-263-SG A
No Yes 8108 SI-263-SG.
A No Yes 6
., _