ML19269D212
| ML19269D212 | |
| Person / Time | |
|---|---|
| Issue date: | 02/15/1979 |
| From: | Mattson R Office of Nuclear Reactor Regulation |
| To: | Anderson T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| References | |
| FOIA-80-587, REF-GTECI-A-09, REF-GTECI-SY, RTR-NUREG-0460, RTR-NUREG-460, TASK-A-09, TASK-A-9, TASK-OR NUDOCS 7903070473 | |
| Download: ML19269D212 (61) | |
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[8 UNITED STATES NUCLEAR REGULATORY COMMisslON w
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,,j FEB 15 B79 Mr. Thomas M. Anderson, Minager Nuc~ ear Sa fety Department Westinghouse Electric Corporation P. O. Box 355 Pittsburgh, Pennsylvania 15230
Dear Mr. Anderson:
In Volume 3 of NUREG-0460, the Nuclear Regulatory Commission's (NRC's) sta ff report on Anticipa ted Transients Without Scram (ATWS), it ta s recommended that prior to the ;ommission's consideration of a proposed ATWS regulation, certain generic safety analyses should be verfomed.
These analyses are to confim that the proposed modificatione for various classes of Light Water Reacto'.- (LWR) designs accomplish t9 degree of ATWS prevention and mitigation described by the staff u Volume 3 of NUREG-0460. The Regulatory Requirements Review Committee ias concurred with the generic analysis approach and the Director of the Office of Nuclear Reactor Regulation has authorized the staff to proceed. If the generic analysis approach is successful, the rule to be proposed for Commission action will not treat ATWS as a design basis accident and will not require a new safety analysis of ATUS on each licensing case. There might be specific exceptions in the future where an analysis for a particular design would be desirable or necessary because the present generic analyses do not envelop that specific design or some future, unanticipated mode of nomal operation.
Generic questions and guidelines are provided in Enclosure 1 for two kinds of plant modifications recommended in Volume 3 of NUREG-0460.
These are the Alternative 3 modifications for plants receiving a Con-struction Pemit prior to January 1,1978, and the Alternative 4 mod-ifications for plants receiving a Construction Pemit after January 1, 1978. The plants listed in Enclosure 2 which began operation prior to Dresden 2 will be treated according to Alternative 2 of Volume 3 and will be examined on a case-by-case basis after the ATWS rule is prmiulgated in its final, effective fom.
We require that by April 15, 1979, the four LWR vendors provide respon-ses to the questions in Enclosure 1 applicable to their designs.
Responses to some of the questions can be delayed until June 1,1979.
Thes, a.e noted by an asterisk or footnote in the enclosure.
7 9 0 3 0 7 0 4 T3
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Mr. Thomas M. Anderson FEB 15 S79 For this generic analysis approach to be successful, it is imperative tha t: a) the responses be complete; b) the responses cover all LWR designs for each vendor, except the plants in Enclosure 2; c) consider-ation be given in the selection of analysis parameters to envelope the naninal conditions for these designs and their anticipated modes of operation as specified in Enclosure 1 so as to minimize the need for ATWS reaqalysis in the future; and d) applicants and licensees provide the necessary support to the four LWR vendors to complete these generic analyses in the required time frame.
The time available to canplete the generic analyses is short. Therefore, it is important that the questions be fully understood and that the answers be as complete as possible so that ocr review does not bog down with an iteration of questions and answers. To this erd we have scheduled a meeting in Bethesda, thryland, Roan P-118, for all day March 1,1979, to explain and discuss the questions with representatives from the four LWR vendors.
It may be necessary to further subdivide the question list at that time to assure timely submission of the generic analyses necessary for the staff to complete its drafting of the proposed ATUS rule in May. The meeting will be open to interested members of the public. Representatives of interested and potentially affected utilities are also invited to attend by copy of this letter.
Si ncerely,
~~
q....
Roger J. Mattson, Di rector Division of Systems Safety Office of Nuclear Reactor Regulation
Enclosures:
1.
Generic Questions 2.
List of Plant for Alternative 2
s__
TABLE OF CONTENTS Page I
Calculational Models........................................
1 II ATWS Events to be Analyzed..................................
6 III Plant Conditions and Assumptions for Evaluation of ATWS Events............................
10 IV Operative Equipment and Systems.............................
11 V
Sensitivity Studies.........................................
15 VI Classes of Plants...........................................
17 VII Analyses Considerations.....................................
19 VIII Acceptance Limits and Additional Gu.idance on Requirements...
22 IX Information Requirements in Electrical Areas.........'.......
a6 X
Diversity Considerations....................................
57
\\
I.
Calculational Models A.
Analysis Computer Codes 1.
Babcock & Wilcox The computer codes used by B&W to perform the ATWS analyses are described and reviewed in Volume 2 of NUREG-0460.
B&W must provide the following additional informatior.:
a.
Sensitivity study using RADAR code to assess the impact of differences between single node and five or more node models.
The ATWS analysis can be. conducted using a one node model if it yields more severe ra ults than the multinode model.
b.*
Describe and justify how CADDS code will be verified using plant experimental data from start-up tests.
c.
Justification for using B&W low flow high quality heat transfer correlation instead of Mirapolsky correlation.
2.
Westinghouse The computer codes used by Westinghouse to perform the ATWS analyses are described and reviewed in Volume 2 of NUREG-0460.
Westinghouse must provide the following information:
a.
Describe and justify the use of LOFTRAN-TRANFLO in determining steam generator heat transfer degradation.
b.*
Describe and justify how LOFTRAN code will be verified using plant experimental data from start-up tests.
- 3.
Combustion Engineering (CE)
The computer codes used by CE to perform the ATW5 analyses are described and reviewed in Volume 2 of NUREG-0460.
CE must provide the following additional information:
a.
Describe and justify the DN8R calculations using the CE-1 correlation.
b.*
Describe and justify how CESEC-ATWS code will be verified using plant experimental data from start-up tests.
~
4.
General Electric The primary code used by GE in ATWS analyses is entitled "REDY.'
This and other computer codes used by GE are described and reviewed in Volume 2 of NUREG-0460.
The staff evaluation (discussed in Vol. 2, NUREG-0460) concludeo that because of discrepancies between test data and REDY results for pressurization events, analyses performed with the REDY code for these events were inappropriate.
In early 1978 General Electric submitted for staff review a new computer code, ODYN, which includes simulation of pressure pulses in the steam line, thereby permitting a more accurate modeling of pressurization events.
The staff expects to issue its safety evaluation report on ODYN as it is applied in transient analyses before March 1, 1979.
GE is required to (a) modify the ODYN Code as necessary for use in ATWS analyses, (b) conduct overpressurization ATWS analyses using ODYN code, and (c) conduct other ATWS analyses using REDY or ODYN.
General Electric evaluates hydrodynamic stability and reactor core stability through the use of the STABLE and FABLE codes.
These codes are also discussed in Volume 2 of NUREG-0460.
Because af lack of
\\ data and insuffichnt andlyses, the staff expressed its concern in Volume 2 of NUREG-0460 regarding the stability of the reactor core in response to an event wherein the recirculation pumps are tripped.
In particular the staff is concerned with the reactivity feedback of the entire reactor core which could drive the reactor into power oscilla-tions. These concerns and the information required from GE to resolve these issues are summarized below; detailed analyses submittal can be deferred until June 1, 1979, only if the consequences of potential instability can be thoroughly described.
a.
In terms of the attached figure, it is not clear that all opera-tional states of a plant, both steady state and transient, are bounded by the curves presented. GE should provide and define all ATWS events, ooth steady state and transient, which possess a decay ratio that exceeds the bounds established by the 105%
rod line, and the natural circulation line to the intersection with the 105% rod line. Of particular interest to the staff is the stability assessment of
'h transient response of the reactor core to a recirculation put..p trip initiated at design power.
b.
Since ATWS analyses are conducted with best estimate design parameters, the assessment of core stability should also be conducted in the same manner. Since most stability analyses are based on the license value of parame'.ers, additional data is required to quantify the stability margin for best estimate parameters. GE should provide a power-flow stability curve for each class of plants based on:
(i) License value of design parameters.
(ii) Best estimate value of design parameter and list the param-eters used in each analysis.
c.
Based on best estimates of design paramete s, GE should define the decay ratio for all plants and define the class of plants
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IB.
Additional Guidance on Analysis Methods for Calculating Safety Valve Relief Rates in PWRs The following method shall be used in the calculation of discharge rates from the pressurizer safety and relief valves.
a.
For saturated or slightly subcooled water at the inlet, the theoretical mass velocity at the throat under choked flow conditions, G, should g
be obtained from the nomogeneous equilibrium model (reversible, adiabatic expansion from inlet to throat with equal phase velocities).
The maximum mass velocity for these inlet conditions corresponds to a finite equilibrium flow quality and a Mach number of unity at the throat.
For more highly subcooled inlet conditions the theore*.ical maximum mass velocity is taken as that corresponding to zero equili-brium flow quality.
b.
The estimated relieving :.pacity at full lift condition should be calculated (followine, the ASME Section VIII type of approach) from:
W=0.9GAK gtD In this equation W is the mass flow rate, 0.9 is the ASME conservatism factor, A is the flow area at the throat, K is the nozzle coefficient t
D obtained with the steam under the Section III certification tests.
Discharge coefficient (K ) is calculated in ;ccordance with Section D
VIII of the ASME Code.
It is assumed that full lift with water at the inlet is obtained at 110 percent of the set pressure.
\\- II. ATWS Events to be Analyzed Anticipated operational occurrences are those conditions of operation which are expected to occur one or more times during the life of the nuclear power unit.
Provide analyses of ATWS events initiated by the following anticipated operational occurrences assuming that reactor scram does not occur.
1.
PressurizedWaterReactorsW a.
Rod Withdrawal. This event concerns maximum ~ orth control rod withdrawal from zero and from full power resulting from a reacter control system malfunction.
b.
Baron Dilution. This event concerns a boron dilution incident resulting from a malfunction in the enemical and volume control system, causing dilution of the maximum rate.
c.
Loss of Primary Flow.
The loss of one or more coolant pumps caused by a failure in a reactor coolant pump, or by a fault in the power supply to the pumps.
d.
Inactive Primary Loop Startup.
This event concerns an inadvertent startup resulting from an inadvertent operation of an inactive
- pump, e.
Loss of Electrical Load.
These transients include generator trip, turbine trip and loss of condenser vacuum.
Following a M
The most severe transients (Primary Pressure, Fuel and Containment considerations) should be analyzed and submitted by April 15, 1979.
The remaining ATVS events must be analyzed and submitted by June 1, 1979.
turbine trip, the bypass valves open to pass steam to the condenser. However, for a loss of condenser vacuum event the condenser will not be available to receive the steam.
Therefore, the loss of condenser vacuum ATWS event results in higher primary temperature and pressure and has been chosen for loss of load ATVS event analysit.
In addition, discuss plant behavior if the bypass valves are open end the condenser is available.
f.
Loss of Normal Jeedwater.
Loss of normal feedwater could result from a malfunction in the feedwater condensate system or its control system, from trip of condensate pumps, trip of both feed pumps, or closure of feedwater control valves.
The loss of offsite power could also result in the loss of feedwater; however, loss of normal electrical power is analyzed separately (Transient g).
The analysis should consider the most limiting loss of feedwater event.
g.
Loss of Normal Electrical Power. This event covers the simul-taneous loss of power from the unit generator and from the offsite grid, leaving the reactor with the onsite emergency diesel generator sets functioning as the only source of electric power.
h.
Load Increase.
This event concerns excess load increases resulting from excessive loading by the operator, failure of the steam bypass system or an equipment malfunction in the turbine speed control. The largest step load increase in the steam demand due to a hypothetical opening of steam dump and bypass valves is also to be considered.
Analyze the most limiting load increase event and describe why the chosen event is the most limiting one.
. f.
Primary System Depressurization.
This event concerns depressuri-zation of the reactor coolant system resulting from accidental opening of a pressurizer relief valve, a pressurizer safety valve or a failure of an instrument, drain or sampling line.
The largest possible blowdown would occur from an accidental opening of a pressurizer safety valve, because of its higher vent area. Therefore, an accidental opening of a pressurizer safety valve has been chosen for ATWS analysis.
J.
Excessive Cooldown. Heat removal rate in excess of the heat generation rate in the core causes a decrease in moderator temperature which increases core reactivity and can lead to a power level increase.
Excessive cooldown could result from a decrease in feedwater temperature, increase in feedwater flow or inadvertent opening of a steam generator safety or relief valve. Analyze the most limiting excessive cooldown event and describe why the chosen event is the most limiting one.
2.
Boiling Water ReactorsS a.
Primary Coolant Flow Decrease.
These transients include failure of recirculation flow controller causing decrease in core flow and could also result from faults in the power supply. Analyze the most severe primary coolant flow decrease event and describe why the chosen event is the most limiting one.
b.
Reactor Water Temperature Decrease.
These transients include malfunction of the feedwater control in a direction to increase feedwater flow, loss of a feedwater heater, shutdown cooling malfunction and inadvertent activation of auxiliary cold water system, or inadvertent startup of an idle isolated recirculation
$/
See Footnote 1 under Section II.1, Page 6.
9_
loop. Analyze the most severe reactor water temperature decrease event and describe why the chosen event is most limiting.
c.
Reactor Coolant Flow Increase.
These transients inclur e a malfunction of the recirculation flow controller in a manner to cause increasing primary coolant flow, and the startup of a recirculation pump.
d.
Reactor Water Inventory Decrease.
These transients include loss of feedwater, pressure regulator failure, and opening of condenser bypass valves.
e.
Primary Pressure Increase. These transients include loss of load (generator trip, turbine trip and loss of condenser vacuum),
closure of main steam line isolation valves and a malfunction of the pressure regulator.
f.
Rod Withdrawal. This event concerns maximum worth rod withdrawal from zero and from full power resulting from a reactor control system malfunction.
g.
Loss of Normal Electrical Power.
This event covers the simul-taneous loss of power from the unit generator and from the offsite grid, leaving the reactor with the onsite emergency diesel generator sets functioning as the only source of a-c power.
h.
Stuck open safety /.31ief valve.
i.
60 F step loss in feedwater heating. Justify why a higher loss (e.g., 100 F) should not be assumed in the analysis.
j.
Turbine trip without bypass.
III. Plant Conditions and Assumptions for Evaluation of ATWS Events A.
Conditions and Assumptions The evaluation report shall include a section justifying the values selected for plant conditions and other assumptions.
As a minimum, the following liat of parameters shall be addressed. Other parameters not listed below should also be addressed if they have a significant effect upon the evaluation.
If, however, certain parameters have negligible effect on a particular transient, you may so state with a brief justification.
In addition, some operational characteristics (e.g.,
stretch power) of plants or other possible changes in design at some future date could change these ATWS analysis results.
All such characteristics should be identified for each transient analyzed.
It is our intent to condition licenses to require notification of NRC whenever these characteristics change so as to fall outside the range of applicability of the current analyses so that the ATWS conclusions in this generic approach can be reconsidered, if necessary.
8.
Justification of Conditions and Assumptions The value used for each conditiond! and assumption shall be selected by one of the following methods:
1.
Selection of a conservative value as specified in the design basis Final Safety Analysis Report Analysis, or the Technical Specfication limit.
2.
Selection of the design operational value including allowance for control band, but excluding any allowance for measurement uncertainty, S
The selection of the moderator temperature coefficient value is an exception, see Sections VII and VIII.
. for variables controlled either by automatic control systems or manually under administrative contral.
3.
Selection of either the measured or design value excluding any allowance for design margin or measured uncertainty.
In addition bases should be provided for the poison mixing and effectiveness assumptions made in the analyses.
Describe any data from experiments or tests performed or yet to be performed to support the analysis assumptions.
Also, the peak pressure calculations in PWRs should not assume any leakage from prfmary to secondary systems.
The reason for selecting each value of the conditions and assumptions listed in Table 1 shall be described.
IV.
Operative Equipment and Svstems For each ATWS event, identify the systems relied upon to bring the plant to cold shutdown condition and maintain the plant in that condition.
For each system provide the following:
1.
System description 2.
Design parameters and bases 3.
Description of how the systems conform with the guidance given at page 18, fection 2.3, and Appendix C of NUREG-0460, Vol. 3 4.
Info mation described in Section IX af this letter.
. Table 1 Initial and Boundary Conditions for Parameters to be Characterized and Justified in ATWS Analysis (1) Reactivity coefficients.
(2) Core power distribution.
(3) Core power.
(4) Core coolant flow.
(5) Core inlet temperature.
(6) Control rod insertion.
(7) Soluble boron concentration in the core.
(8) Reactor decay heat function.
(9) Pressurizer water level (10) Pressurizer pressure.
(11) Pressurizer safety and relief valve flow rate (both steam and wate.').
(12) Steam generator temperature.
(13) Steam generator pressure.
(14) Steam generator steam flow rate.
(15) Steam generator safety and relief valve flow rate.
(16) Steam generator heat transfer coefficient.
(17) Steam generator secondary side water inventory.
(18) Feedwater temperature.
(19) Equipment performance.
(20) Turbine bypass condition.
(21) Containment ambient coriditions (pressure, temperature, including suppression pool temperature and level, etc.).
(22) Fuel element gap size.
(23) Auxiliary feedwater:
number of pumps available, number assumed in analysis.
. (24) Auxiliary feedwater flow for each pump as a function of system pressure.
(25) Auxiliary feedwater temperature.
(26) Vessel water level.
(27) All water sources inventory.
(28) HPSI and any other high pressure makeup and poison system flow rate as a function of system pressure.
(29) HPCI(s) and any other high pressure makeup and poison system flow rate as a function of system pressure.
(30) Boron concentration of HPSI and/or other borated solution.
(31) Sodium pentaborate concentration of SLCS.
(32) Reactor coolant mass.
(33) Reactor coolant system volume.
(34) Core average temperature.
(35) Pressurizer water volume.
(36) Pressurizer total volume.
(37) Number of relief valves on pressurizer.
(38) Number of safety valves on pressurizer.
(39) Relief valve setpoint on pressurizer.
(40) Safety valve setpoint on pressurizer.
(41) Feedwater flow rate.
(42) Number of S/V per steam generator or steam line.
(43) Number of steam bypass valves.
(44) Setpoint of steam bypass valves.
(45) Capacity of steam bypass valves.
(46) Number of atmospheric dump valves.
(47) Capacity of atmospheric dump valves.
(48) Steam flow rate including bypass flow.
(49) Valve closure times.
(50) Pump trip and start times.
(51) Poison reactivity worth.
. (52) Poi;on-water mixing efficiency.
(53) Signals and setpoints for all automatically actuated systems.
(54) Containment pressure, temperature and colume.
(55) Steam generator tube leakage.
(56) Service water temperature and flow.
(57) Component cooling water temperature and flow.
(58) RHR heat exchanger performance.
(59) Fouling factors in heat exchangers.
(60) Power availability for pressurizer relief valves during loss of offsite power ATWS event.
(61) Core average void fraction.
(62) Heat transfer surface area.
. *If the reliability criterion of Appendix C paragraph H in Volume 3 of
'. ' EG-0460 is used for any function (including offsite power), justify the Juancsry of the data base, including failure rates and downtime due to test and/or repair, and the statistical methodology.
Show how common mode failures are included or justify not including the n.
V.
Sensitivity Studies Realistic calculations are acceptable for these ATWS analyses as described above. However, some parameters in the analysis will vary from plant to plant within a class of plants. Therefore, sensitivity studies must be performed to cover the range of system characteristics or other variations expected for each class of plants analyzed. Otherwise, the analysis for each class of plants must be shown to bound the ATWS consequences for expected variations in these parameters across the class of plants and for all permissible normal states of operations of those plants.
A.
Sensitivity to Single Parameters Variations As a minimum, provide sensitivity studies for five most sensitive parameters selected from the following list.
Provide a qualitative discussion of the effects of the remaining parameters.
PWRS:
Core Power Pressurizer Level Steam Generator Inventory Main Feedwater Enthalpy Reactivity Feedback Parameters Boron Content d#
Sensitivity studies to five most sensitive parameters should be provided by April 15, 1979.
The remaining sensitivity studies must be provided by June 1, 1979.
. Core Exposure (including Power Shape)
RCS Volume Pressurizer Relief Capacity Auxiliary Feedwater Characteristics Rate Initiation Time Fuel Element Gap Size Pressurizer Spray Flow Turbine Trip Initiation Time Average Coolant Temperature Delays in Operator Action BWRs:
Core Power Pump Trip Time Steam Relief Capacity Poison System Characteristics Mixing and Effectiveness Rate Initiation Time Plant Design Parameters Reactivity Feedback Parameters Service Water Temperature Valve Closure Times Ratio of Vessel Volume to Suppression Pool Volume Heat Exchanger (RHR) capability Delays in Operator Action Initial Pool Temperature Initial Pool Water Mass Initial Void Content of Core Core Exposure including Rod Position, Power Shape
. 8.
Sensitivity tv *'ultiple Parameters r
On t'e basis of the single parameter sensitivity considerations in V.A, above, U termine the five most sensitive parameters.
In order to charac-terize the overall sensitivity of the consequence analysis for a class of reactors, provide the following analyses:
1.
Five most sensitive parameters simultaneously assumed to be equal to their reference values plus 10%, where plus 10% means that the refer-ence values are all increased (in more severe direction) simultaneously.
2.
Same as 1, but use minus 10%.
VI. Classes of Plants A.
Analyses of the ATWS events listed in Section III are to be provided for classes of plants. The classes of plants should be derived on the basis of similarity in design. Some examples of considerations which could be made in determining such classification of plants are provided in the attached table 2.
In the analyses for PWRs, include a list of plants for which Category 1 steam generator leakage assumption of Appendix V of NUREG-0460, Volume 2, apply.
If it is expected that steam generator replacements will result in recategorization of these plants, they should be identified with both pre-and post-replacement gr ap'1gs.
B.
Include a list of plants for which each analysis is applicable.
Describe the plant mortifications used in the analysis to conform with Alternative 3 or 4 of NUREG-0460 Volume 3 as appropriate in light of the January 1,1978 cutoff date.
. TABLE 2 Some Considerations in Plant Classifications Plant Analysis Plant Parameter Design Assumption Power Level Design Type (e.g., BWR/4, BWR/5)
Number of Fuel Assemblies Containment Design (e.g., 1. Condenser, Mark I)
Emergency Core Cooling Systems (e.g., HPSI, HPCI)
Long-Term Core Cooling Systems (e.g., RHR)
Number of Loops Steam Generator Design (e.g., Series 44, 51)
Pcwer-Operated Relief Valves Pressurizer Safety Valves Auxilf ary Feedwater System (e.g., Auto / Manual, Capacity, etc.)
Bypass Flow Capacity Upper Head Injection BWR Core Size Integrated Control System Delivery Capability of Poison System (BWR) for Alternatives 3 and 4 Core rotection Calculators
. C.
If the assumed ATWS mitigating system on one or more plants is significantly different than that assumed in the generic analyses, the difference must be described, and its effects on the analysis of consequences must be described.
D.
If the sensitivity studies of Section V do not cover some specific design in a class of plants, then analyses for that specific plant must be provided.
VII. Analyses Considerations A.
Events for Analyses 1.
For Plants Modified Per Alternative #3 Analyze, as a minimum, ATWS events described in Section II using conditions and assumptions as described in Section III with the exception of the initial value of the moderator temperature coefficient (MTC) in the PWRs. The equipment assumed to operate must satisfy constraints of Section IV.
The initial MTC absolute value must be less than that experienced by the reactor 95 percent of the time.
The MTC value calculation shall be as described as specified in Section VIII.
2.
For Plants Modified Per Alternative #4 a.
Same as A.l. above except use a 99 percentile value of the MTC for PWRs.
In addition, the equipment assumed to operate must account for a single failure in any function that does not satisfy the numerical reliability objective or is designed to IEEE-279 criteria (which includes a single failure criterion).
Based on our evaluations to date, the effects on consequences of the following failures must be separately analyzed.
. 1.
Failure of one power-operated relief valve to open for PWRs, and ASME Code requirements on relief valves failing to open at the relief setpoint for BWRs.
11.
One half the capacity of the auxiliary feedwater system for PWRs.
iii.
High pressure coolant injection system or high pressure spray system for BWRs.
iv.
Failure of safety / relief valves (one for PWRs, two for BWRs) to reclose.
v.
Other safety or reliability-based failures.
v-1. Failure of single train in a safety system (e.g., high pressure safety injection in PWRs; one train of RHR or one train of poison injection in BWRs).
v-2. Assume a disabling failure of other systems which do not satisfy the reliability criterion of Appendix C to Volume 3 of NUREG-0460.
B.
Results Documentation For each event the analyses should be carried through until the plant is in a stable condition (e.g., Hot Shutdown).
For subsequent plant behavior simplified analyses should be performed and described to demonstrate the capability to bring the plant to a safe cold shutdown condition and maintain in the cold shutdown condition without dependence on control rod insertion.
As a minimum, the following information must be provided:
. 1.
Describe the sequence of events including definition of the times when any of the mitigating systems are either automatically or manually initiated and the transient behavior undergoes a marked variation (e.g., pressurizer solid, steam generator heat transfer degradation).
2.
Describe operator actions. As a minimum, credit for operator action cannot be assumed within the first 10 minutes of the event and any subsequent operator action must be justified (e.g., credit can be taken after 10 minutes for simple operator actions such as pushing HPSI button if adequate information is disolayed for such an action).
3.
As a minimum, figures must be provided for the following parameters as a function of time up to achieving a hot shutdown condition.
Core Power Core Average Heat Flux Pressurizer (PWR) or Vessel (BWR) Pressure Reactor Coolant Temperature (FWR)
Steam Generator Pressure (BWR)
Feedwater Flow Core Inlet Flow Steam Flow Surge Line Flow (PWR)
Relief and Safety Valves Flow and Enthalpy Vessel Level (BWR), Pressurizer Level (PWR)
Steam Generator Inventory (PWR)
Steam Generator Safety Valve Flow (PWR)
Steam Bypass Flow Steam Generator and/or Primary System Makeup and Poison Injection Flow Reactivity Components (e.g., Void, Moderator, Doppler, and Boron)
9
. Steam Generator Heat Transfer Area (PWR)
Steam Generator Tube Leakage Containment Conditions (Pressure and Temperature)
Suppression Pool Temperature (BWR) 4.
Provide simplified calculations showing how the system capacities, flow rates, and initiation times assumed in the calculations would bring the plant to and keep it in a cold shetdown condition.
5.
Dascribe how the systems satisfy the criteria described on p.18 of Volume 3, Appendix C of Volume 3, and discussed in Sections VIII and IX of this transmittal.
6.
The analyses of recovery from ATWS should also consider (a) the effect of temperature, pressure, and humidity upon equipment and instrumentation necessary for controlling the course of ATWS mitigation, and (b) the potential, if any, for poison dilution from ECCS or cleanup systems.
VIII. Acceptance Limits and Additional Guidance on Requirements A.
Acceptance Limits 1.
Radiological Consequences The calculated radiological doses from postulated ATWS events shall be within the guidelines values set forth in 10 CFR Part 100.
The doses shall be calculated (Section B.3 below gives guidance) in accordance with an acceptable dose calculation model (described in a subsequent section) and shall consider, among other things, the leakage from steam generator tubes and the damage to fuel rod cladding.
. 2.
Primary System Inteority See Section VIII.B.I 'clow.
u 3.
Fuel Integrity Damage to the reactor fuel rods as a consequence of an ATWS event shall not significantly distort the core, impede core cooling, or prevent safe shutdown. The number of rods which would be expected to have ruptured cladding shall be determined for the purpose of evaluating radioactive releases.
4.
Containment Integrity The calculated containment pressure, temperature, and other variables shall not exceed the design values of the containment structure, components and contained equipment, and the systems or components necessary for safe shutdown.
For boiling-water reactor pressure-suppression contain.T.ents, the region of relief or safety valve discharge line flow rates and suppression pool water temperatures where steam queni;.;ag instability could result in destructive vibrations shall be avoided.
5.
Long-Term Shutdown and Cooling Capability The plant shall be shown to be capable of returning to a safe cold shutdown condition subsequent to experiencing an ATWS event, i.e.,
it must be shown that the reactor can be brought to a subtritical state without dependence on control rod insertion and can be cooled down and maintained in a cold shutdown condition indefinitely.
, 8.
Additional Guidance and Requirements 1.
Engineering Considerations
- 3.
- Assurance of Structural Integrity for Plants Conforming to Alternate 4 Provide detailed descr'ption of elastic analyses performed and i
corresponding results that verify that all ASME Cl.1 components comply with Service Level C stress criteria of ASME Code Section III as specified in the Summer 1978 Addenda of the Code.
The integrity of steam generator tubes may be evaluated based on a conservative assessment of tests and the likely condition of the tubes over their design life.
Information to be provided shall include a discussion of all loads that could affect each component if exposed to the ATWS event resulting in highest calculated system pressure i.e., for BWR's this would specifically include SRV induced initial impact and subsequent containment vibratory loads.
Analyses shall be performed for each component design used in this class of plants.
If components designed by more than one manufacturer are used for the same function in plants within this class, e.g., reactor coolant pumps or reactor isolation valves from different manufacturers, analyses shall be performed for each design used.
b.
Assurance of Structural Integrity for Plants Conformino to Alternate #3 For components for which ASME Service Level C criteria will not be exceeded, provide the information required for components under a. above. We expect that most components will be in this class. However, some may not and therefore we have
e developed the following information requirements for components for which ASME Service Level C criteria are exceeded.
The information is needed to allay three gereral types af concerns for components exposed to stresses in excess of the Code Level C Service Limit.
First, structural integrity needs to be demon-strated by tests or analyses.
Second, since stresses above the Level C lin.it can potentially result in considerable inelastic behavior, assurance shall be provided that any resulting deforma-tions will not lead to rupture or in ar.y way inhibit safe plant shutdown.
In general, inelastic analysis is the preferred method for evaluating component response at these stress levels.
Elastic analysis may be acceptable if adequately justified.
Finally, there is a concern with the behavior of bolted and flanged connections. Such areas are normally assumed to be tightly sealed and their structural adequacy analyzed using elastic analysis techniques.
However, many such joints are only torqued for normal operating pressures and not those associated with pressures that would be associated with ATWS.
Based upon the staff review of ATWS analyses submitted to date and through participation in related ASME Code Committee activities, it appears to the staff that components exposed to stress levels higher than Level C may be susceptible to gross leakage at bolted connections and also the basic structural integrity of the connection is open to question because the initial assump-tions of connection rigidity cay no longer be valid. Additional elastic or inelastic analyses shall be performed for components in plants under Alternative #3 that exceed Service Level C to verify connection structural integrity and assesss the potential for leakage. Justification shall be prov'ied that any resulting leakage will have no adverse effect on safe plant snutdown.
. The following specific concerns shall be addressed as applicable for components exposed to stresses 52s in excess of the Code Level C Service Limit. Analysis shall be performed for each component design used.
If components designed by more than one manufacturer are exposed to stress levels higher than the Level C Service Limit and are used for the same function in plants within this class, e.g. reactor coolant pumps from different manufacturers, analyses shall be performed for each design used.
(1) Reactor Ccolant Loop & Shutdown System Branch Line Piping Assurance shall be provided that the effects of the ATWS pressure in combination with other applicable mechanical loads (e.g. operating basis earthquake, dead weight of the piping, thermal expansion load etc.) will not result in deformation of the piping to such an extent that fluid flow would be impaired.
(2) Pressurizers (PWR's)
Any areas of permanent deformation shall be evaluated using appropriate analytical methods and shown not to prevent
-S/*
The staff also intends to require a re-evaluation of current tube plugging limit under ATWS conditions using R.G. 1.121 criteria. Note that as stated in Appendix V of NUREG-0460 tubes must be plugged if tube wall stresses resulting from ATWS induced loading would exceed those permitted by Level D Service Limits of ASME Section III.
-6/*
In addition, the staff intends to require that its licensees commit to perform 100% tube inspection following an ATWS event that produced loads (pressure) that e.xceed those used to establish the plugging limit, including a submittal of ISI results for NRC review and approval before returning to power.
, safe plant shutdown.
In general, inelastic analysis should be performed in order to accurately assess the effect of permanent deformations on component pressure retention capability.
Specific areas of concern include bolted manway covers and pressurizer heater to head welds.
(3) Steam Generators (Other than Tubes)
Effects of all deformations both elastic and inelastic shall be evaluated and shown not to prevent safe plant shutdown. Specific areas of concern include tube to tubesheet welds and bolted manway covers.
In general, as noted above, inelastic analysis should be used to accurately assess the significance of permanent deformations.
(4) Reactor Coolant Pumps Large permanent deformations are of concern.
Inelastic analyses should be used to accurately assess the conse-quences of these deformations.
Effect of deformations on bolted closure and pump impeller are of particular interest.
(5) Primary Loop Isolation Valves (PWR's) and Recirculation Looo Valves (BWR'sl Same as Reactor Coolant Pumps.
(6) Reactor Vessel Pennanent deformations should be accurately assessed using inelastic techniques.
Specific areas of concern to be addressed include:
28 -
(a) Bolted Closure Head to Vessel Flange Connection Sufficient analysis shall be performed to substantiate that structural integrity will be maintained. Assurance shall be provided that primary coolant leakage from the closure joint will not result in inability to safely shut the plant down.
Effect of coolant loss on fuel clad temperature and vessel cavity integrity shall specifically be addressed.
(b) Control Rod Drive and Instrumentation Tube Vessel Penetrations Sufficient analysis shall be performed to demonstrate that these penetration housings will not be ejected.
Analyses shall specifically take into account the additional bending loads imposed on these housings as a result of potential vessel head deformations resulting from ATWS imposed pressure loads.
(c) A Linear Elastic Fracture Mechanics Evaluation to Demonstrate Vessel Integrity The analysis shall include the following elements:
1.
Evaluation of radiation damage including considera-tion of relation between ID fluence and operating time, attenuation of fluence through vessel wall, limiting copper content and unirradiated upper shelf toughness, drop in upper shelf toughness as determined by R.G. 1.99, and method of conversion from C energy to KI V"I"**
y c
\\
' 11.
Calculation of allowable pressure at EOL according to App. G of Section III of the ASME Code, using a factor of safety of 1 instead of 2 on pressure component of KI for 1.5 inch and 1/4 t crack lengths.
iii. Calculate fluence and approximate date when the applied KI corresponding to the maximum credible ATWS pressure equals KI
- 1. and 2. as in b.
c above.
(7) Safety and Relief Valves See Item d on Page 32.
The information supplied for each component that exceeds the Level C Service Limit shall include the following:
(1) Elastic Analysis (Finite Element (F.E.) or Analytical) a.
Sketch of actual component in sufficient detail so that correlation with mathematical models can be determined.
b.
Sketch of mathematical model c.
Description of mathematical model i.
Type of element for F.E. analysis ii.
Description of code and benchmark tests iii. Boundary Conditions
4 d.
Material properties 1.
ASME Code Allowable values 11.
irradiation effects cot,idered as appropriate e.
Evaluation of Results i.
Comparison with Level C and D ASME ccie stress limits ii. Determination of operability based on strains and
.1splacements (See Item II)
(2) Inelastic Analysis a.
Same as la thru le above b.
Material properties i.
Stress-strain curve to ultimate ii. Ductility including biaxial and triaxial states of stress as appropriate.
iii. Discussion of non-linear modelling techniques c.
Assurance of Operability for Shutdown System Isolation and other Active Valves for Plants Conforming to Alternate #3 or #4.
(1) Conventional Check Valves (other than Globe Check Valves)
Demonstrate by analysis (elastic analysis should be sufficient for most designs) that the disc remains intact, the seat does
. not deform inelastically or shear, and the shaft has no permaner,; deformation.
Documentation to be supplied to NRC should consist of a sketch of the valve, a description of the analysis, and the results of analyses. Due to simplicity of analysis, extrapolation of results to larger or smaller nominal valve sizes is not acceptable.
A separate analysis for each size and design is required.
(2) Cate Valves Demonstrate by elastic finite element analysis that there is no permanent deformation in either the disc or seat.
For a valve design where results of elastic analysis clearly show that no permanent deformation results from exposure to the ATWS environment, results of such analyses can be extrapolated to valves of nominal sizes other than that for which detailed analysis has been performed provided that the following criteria are met:
a.
Valve qualified by extt-apolation shall be of the same generic design type and of the same manufacturer as the valve analyzed.
b.
Nominal size of valves qualified by extrapolation shall be no smaller than 50% af the nominal diameter size of the analyzed valve and no larger than 200% of the nominal size of the analyzed valve.
If there is permanent deformation of the disc, demonstrate that it will not prevent operation by either tests or inelastic finite element analysis. The analysis or tests shall be performed for each size and design used.
If permanent deformation of the valve seat results from exposure
. to the ATVS environment, valve operability shall be demon-strated by test.
Tests shall be performed for each size and design used. Documentation to be supplied to NRC should consist of a skstch of valve, a description of analyses or tests performed, and a summary of the results.
(3) Globe, Globe Check, and Globe Stop Check Valves Based on information on file with the NRC, it is the staff's understanding the few, if any, valves of this type are being used in nuclear plants for shutdown system isolation applications.
Since these valve configurations are extremely difficult to model analytically, for those few cases where such valves are used the following requirements apply.
Demonstrate by test that operability is retained.
The following general testing sequence shall be used:
Close valve with maximum motive power (i.e., maximum possible operator output), 2.
Apply maximum applicable ATWS pres-sure at appropriate temperature, 3.
After removal of test pressure, open valves using minimum motive power (i.e.,
power output of operator under minimum permissable operator energy input.) Documentation to be supplied to NRC should consist of sketch of valve, a description of tests per-formed, and a summary of the results.
d.
Miscellaneous Structural Considerations Acplicable for Both Alternate #3 and #4 Plants (1) Safety and Relief Valves Assurance of valve operability while system pressure is above the valve sct pressure and ability to reclose after system
. pressure returns to below valve set pressure shall be provided.
For PWR's specifically, assurance shall be provided that safety valve " chatter" will not decrease valve discharge rate below the rate used in ATWS analyses.
(2) Safety and Relief Valve Discharge Piping Assurance shall be provided that such piping retains suffi-cient functional integrity so as not to impair valve operabi-lity. Description of analyses performed together with results shall be provided to NRC.
(3) Pressurizer Quench Tank (PWR's Only)
Assurance shall be provided that quench tank response to ATWS discharge is such that safety and relief valve operabi-lity are not impaired. Assurance shall also be provided that equipment essential to safe shutdown is protected from any missiles that may be generated such as by bursting of the rupture disc. Description of analyses performed together with results shall be provided to NRC.
2.
Fuel Damage Considerations The results of our earlier technical review (presented in Appendices XIV, XV, and XVII of NUREG-0460, Vol. 2) of PWR ATWS fuels analyses, coupled with further development of understanding of fuel behavior under transient conditions, indicates that the following procedures and considerations should be accounted for in the calculation of PWR ATWS fuel failure.
, Babcock and Wilcox As indicated in Appendix XIV of NUREG-0460, Vol. 2, B&W submitted anal,yses of the expected fuel duty for the loss of offsite power (LOOP) and two pump coastdown ATWS events (the worst-case events in terms of the departure from nucleate boiling ratio (DNBR)), and showed that using a conservative methodology and code called RADAR, DNBR values Fl.3 were obtained. Using less conservative methodologies and codes, DNBP.s J1.3 were obtained.
B&W must describe and justify the applicability of the code utilized to estimate the number of rods, if any, in DNB condition.
If the calculated DNBR value falls below the 95:95 DNBR limit (for any class of plant and fuel design),
the number of failed fuel rods should be taken to be the number which actually experience a departure from nucleate boiling condition as determined by summation of the probability of DNB on individual rods based on accepted statistical correlation.
Combustion Engineering Our review of ATWS fuel duty analyses supplied by C-E indicated, that whereas an early analysis of the partial loss of flow event (.ba
" worst-case" in terms of DNBR) using the STRIKIN-II code produced a minimum DNBR of 0.97, later calculations using a code called TORC-CE-l produced a ONBR of 1.28, which is above the C-E 95:95 DNBR value.
C-E should reaffirm that the minimum DNBR does not fall below the 95:95 limit, using codes approved by NRC for C-E ATWS analyses, for all classes of plants and for all of its fuel designs.
If the DNBR does fall below the acceptance criterion for any ATWS event and with any given combination of plant and fuel designs, C-E should provide the number of rods predicted to be in DNB.
If the calculated DNBR value falls below the 95:95 DNBR limit (for any class of plant and fuel design), the number of failed fuel rods should be taken to be the number which actually experience a departure from nucleate boiling
. condition as determined by summation of the probability of DNB on individual rods based on accepted statistical correlation.
Westinghouse Our understanding of W ATWS analyses is that, although DNBR values below the 95:95 limit were initially calculated to occur for the loss of normal onsite and offsite AC power, a DNBR value below the 95:95 limit is currently calculated not to occur. W should confirm that, using NRC-approved codes, the minimum DNBR never drops below the 95:95 value (1.3 for the W-3 DNB correlation) for any combination of plant and fuel-designs.
If the calculated DNBR value falls below the 95:95 DNBR limit (for any class of plant and fuel design), the number of failed fuel rods should be taken to be the number which actually experience a departure from nucleate boiling condition as determined by summation of the probability of DNB on individual rods based on accepted statistical correlation.
PCI Failure Assumotion (for all PWR vendors)
At the time we completed our earlier review and input to volumes 1 and 2 of NUREG-0460 (early 1978), it was our understanding that none of the PWR vendors were predicting violation of the 95:95 DNBR limits (although some clarification is needed regarding the codes used in tne calculations, the applicability of the calculation to specific fuel and plant designs, etc.). We thus currently do not anticipate the prediction of PWR fuel rod failures due to overheating effects that are cased on violation of DNBR criteria. As shown in Appendices XIV, XV, and XVII of Vol. 2, however, we believe that for power-increasir) events, where the rapid increase in temperature of the fuel pellets may create a thermal expansion that might effect an interaction with the cladding, pellet / cladding interaction (PCI) failures may occur. Therefore we require each PWR vendor to provide
an assessment of the likelihood and consequences of PCI failures during an ATWS event.
The general types of power-increasing events to be considered include control rod withdrawal, boron dilution, inactive primary loop startup, load increase, and excessive cooldown.
We are attempting to develop a PCI model for safety analyses related to licensing as part of a generic study of PCI. Analytical and experimental work is in progress and planned.
It should be understood, therefore, that the PCI fuel failure values estimated by the vendors in this early verification phase of the ATWS licensing effort are interim values.
If the results from our analytical and experimental programs indicate that the PCI failures are more severe than those used in the early evaluation, the NRC staff will require a re-evaluation of ATWS to assess the impact of increased likelihood and/or consequences from PCI on our earlier conclusions.
Clad Collapse (For All PWR Vendors)
Perform analyses to determine whether cladding creep collapse will occur taking into account the fuel and primary system pressures calculated under alternatives 3 and 4.
Estimate the number of rods with collapsed cladding for all classes of plants and fuel designs.
General Electric The results of our earlier evaluation of BWR ATWS fuel duty predictions (presented in NUREG-0460, Vol. 2, Appendix XVI) indicated that the main steam line isolation valve (MSIV) closure ATWS with recirculation pump trip, automatic baron injection, and no additional system failures was the worst-case BWR ATWS from a fuel duty standpoint because it resulted in the lowest minimum critical power ratios (MCPRs) and the greatest number of rods in boiling transition. Because MSIV closure involves a reactivity-initiated power increase, we believe that fuel
. rod failures may also result from pellet / claddlng interaction.
PCI failures are considered to be more likely to occur during power-increasing than reduction-in-flow events because during the former the fuel pellets heat up and expand more rapidly than the cladding, whereas in the latter type of event the opposite thermal expansion effect occurs. On the other hand, we do not believe that the rods in boiling transition will necessarily fail due to overheating effects.
Indeed, there is considerable experimental evidence that rods can operate in non-nucleate boiling for fairly lengthy periods of time without failure.
Therefore we will use the number of rods in boiling transition as an estimate of the total number of fuel rod failures.
In our judgment, the number of rods in boiling transition will encompass the number that would actually fail as a result of both MCPR and PCI combined (because not all of the rods in boiling transition are sure to fail).
The application of a boiling transition criterion for both overheating and PCI-induced failures is made necessary because (a) we do not have an applicable licensing acceptance criterion for PCI and (b) we do not belie"e it is possible to develop an acceptable criterion in the time period consistent with ATWS "early verification." It should be noted, however, that we are attempting to develop a PCI model for safety analyses related to licensing as part of a generic study of PCI.
If the results from our generic PCI activity indicate that the PCI failures have not been adequately addressed in the "early verification" approach, the NRC staff will require re-evaluation of the ATWS conclusions.
Based on GE-supplied analyses, it is our understanding that for BWR/5s and BWR/6s,10.5 and 17%, respectively, of the fuel rods would be in boiling transition, and thus assumed to fail, during an MSIV closure ATWS. We have asked GE to provide the number of rods in boiling transition for the BWR/4 case, but have not yet received this
information. Moreover, t'he MSIV closure ATWS BWR/5 and BWR/6 analyses were presented for 8x8 two-water-rod assemblies only.
Predictions for other BWR fuel assembly designs (both in current operation and planned) have not been submitted. General Electric should, therefore, submit a tabulation of the number of rods predicted to be in boiling transition for an MSIV closure ATWS (unless another ATWS event is more severe) for all classes of plants using acceptable models and the " worst-case" (bounding) fuel design for each type of plant.
3.
Radiological Considerations Radiological dose analyses shall be presented for plants within the scope of Alternatives 3 or 4 of Volume 3 of NUREG-0460.
These scoping calculations shall be presented to show conformance with the requirements of 10 CFR Part 100. These source terms, the proposed scenario, the activity release, and the dose model shall be justified.
The following conditions shall be assumed in the calculation of radiological consequences:
a.
Steam generator tube leakage as specified in Appendix IV, Volume 2, NUREG-0460 (See Section VI.A).
b.
Fuel rods experiencing a cladding rupture shall be assumed failed. Of those rods which are presumed to fail, an instantaneous release to the primary coolant corresponding to 2%
of the rod inventory of xenon and krypton and 2% of the rod inventory of lodine should be assumed.
The inventory should be based on nominal full power, end-of-life core conditions.
9
\\,
c.'
The coolant activity existing at the time of the ATWS should be taken as 50% of the maximum steady state technical specification value.
d.
An iodine release rate increase (iodine spike) of a factor of 500 above the pre-ATWS release rate from the fuel shall be a:samed for any analyses for which zero fuel failures are assumed.
e.
An iodine decontamination factor of 1.0 should be assumed for any RCS leakage to containment, unless it can be demonstrated that flange or seal leakage (to the quantity equivalent to a stuck open relief valve) would be discharged into water where a decontamination factor of 10 should be used.
For relief valve discharges into BWR suppression pools (which satisfy the requirements described herein) an iodine decontamination factor of 10 should be assumed.
f.
Aftercontainmentisolationoccurs,1 the contribution to the radiological dose from containment leakage should be addressed utilizing a leak rate based on containment pressure conditions.S#
1#
It must be assumed that airborne radioactivity from relief valve discharge and pipe leakage (flanges, etc.) is being discharged to the environment before containment isolation does occur.
Credit for mixing in the containment before discharge must be justified for each class of plants.
The fraction of activity in coolant discharged or leaked to the containment air that becomes airborne should be assumed t,ual to the flash fraction of the coolant.
In determining isolation time, the appropriate delay times of instruments, closure times of valves, and transit time of air passage should be considered.
An acceptable method of accounting for the relationship between leakage and containment pressure is as specified in 10 CFR Part 50, Appendix J. III.4A.
. g.
Alternative #3.
If the ATWS event can degrade or jeopardize the performance of seals in ECCS equipment, failure of a seal in ECCS equipment outside containment during post-ATWS hot shutdown shall be assessed, and the radiological consequences of that failure added to the contributions from other releases to the environment.
Alternative #4. The failure of a seal in the ECCS equipment outside containment shall be assumed as a single failure in the long term following the accident. The dose from the resulting airborne release shall be added to the other releases in determining the Low Population Zone dose.
h.
Representative meteorological dispersion conditions should be used for Alternative #4 when consideration of a single failure is included, and for Alternative #3. The meteorological models of Regulatory Guide 1.111 (for intermittent releases) are acceptable for this purpose.
i.
Alternative #4. ATWS evaluations shall include a five percentile meteorological dispersion factor as representative of a single failure.
j.
Alternative #4 ATWS evaluation shall include the dose contribution resulting from a pre-existing iodine spike in the primary coolant equivalent to the maximum transient coolant Technical Specification limit for full power operation as represenative of a single failure.
k.
When considering an additional single failure (Alternative #4) each aspect of the analysis which determines a contribution to the dose (e.g., ECCS seal leakage, containment leakage) should include consideration of the worst single failure (an analysis without the failure may also be performed).
The total dose may be determined by combining the components of the dose to identify
. 4 the combination resulting in the maximum dose where only one component uf that dose includes a single failure.E 1.
Any additional leakages that may occur as a result of high primary pressures as discussed in Section VIII of this transmittal should be considered in the dose calculation.
m.
Radioactivity releases through the BWR steam system must be evaluated unless the releases can be shown to be through the offgas system and the integrity of the steam system can be demonstrated using the guidance in Sections IV and VIII.
Under Alternative 4, single failures should be assumed.
This should include a single failure in the steam bypass system and the condenser can be overpressurized or steam piping integrity lost.
4.
Reactivity Feedback Considerations ATWS calculations may use expected, nominal values of initial conditions (as described in Section III) and transient parameters appropriate as representations of full power (100%) operation initial conditions at a time in (the expected 40 year) plant life when parameters are such that the combined effect of (nominal) parameters are expected to provide the most limiting characteristics for the event analyzed. An exception (in addition to the use of 100% licensed power) will be the moderator temperature coefficient (MTC) of PWRs.
For PWRs falling under Alternative 3 (of NUREG-0460, Vol. 3) calcula-tions should assume a (initial condition, see NUREG-0460, Vol. 2, E#
Items 9 and 10 should be included in the determination of the worst single failure.
\\
- Appendix VIII, page 5) MTC value which is less negative than that experienced during 95% of the time the reactor is at relevant power levels.
For PWRs under Alternative 4 a 99% MTC value should be used.
The determination of appropriate MTC values should consider expected, nominal values as a function of core design and operating states and the effect of expected operations (as discussed in Vol. 2 Appendix VIII, pages 6-8).
A discussion and validation of the MTC chosen to meet these criteria must be presented.
The MTC used in the analysis must represent the MTC which will not be exceeded at either the 95% or 99% probability level over the plant life. As discussed in Appendix VIII the anticipated MTC calculation-measurement error (see NUREG-0460, Vol. 3, Appendix B, Question 3, pages 2 and 3) and modifications for operations do not present a serious problem for MTC use and specification (aithough the latter should be accounted for).
However, core design variations that have occurred in the past and future and potential variations can present a problem. As discussed in Appendix VIII of Volume 2 of NUREG-0460, past variations have indicated a potential in operating reactors for
-5 possibly +3x10 / F greater MTC than generic ATWS analysis values.
This magnitude might yield significant pressure increases in the analyses.
Present trends and future design optimizations for extended cycles and fuel cycles without reprocessing indicate that more positive MTC (e.g., through higher enrichments, larger water to fuel ratios) may be necessary in the future. This trend might be offset by the use of burnable poisons (possibly including the development of more
" efficient" poison design such as gadolinia in fuel).
Thus, it is expected that the potential for, and impact of such changes will be discussed in these generic ATWS analyses to provide a basis for understanding the importance to ATWS considerations of changes in this area in the future.
Furthermore, a commitment on present and future core design and verification must be stated to assure an appropriate MTC is used for present analyses, and that future changes
. in design will be reanalyzed for their ATWS implications for NRC review and approval.
5.
Containment Considerations PWR There is no open issue related to containinent pressure and temperature response following ATWS. However, we require documentation of the analyses relied on by the vendors in references (1) and (2) for Babcock & Wilcox plants, references (3) and (4) for Combustion Engineering plants, and references (5), (6), and (7) for Westinghouse plants.
BWR For all BW2 plants, including BWR 4, 5, and 6, the suppression pool temperature limit remains an open item for ATWS.
However, during the course of our evaluation of this issue, which is a subtask of TAP A-39, the staff has concluded the following:
Based on our evaluation of the test data, we believe tne quencher a.
device results in superior performance over the Rams Head in the steam quenching mode, and thus consideration :hould be given to the use of quencher devices to minimize the concern with steam quenching instability.
b.
The applicants / licensees are required to meet the following suppression pool temperature specification.
It should be noted, however, that this specification was established on the basis of test data that are currently available to the staff. Additional data will be required to justify pool temperatures exceeding the limits, specified below:
44 i.
Pool Temperature Limit The suppression pool local temperature shall not exceed 200*F for all plant transients involving SRV operations.
GE is required to provide justification for establishing the difference between local and bulk pool temperatures for diffcrent types of quencher devices. The definition of local and bulk pool temperature is provided in the following section. This limit was established as a result of our evaluation of the data base provided in the General Electric Topical Report NEDO-ll314-08(8)
The tests were performed on quencher devices with specific hole patterns, and the hole patterns have been identified as the key parameter for steam quenching performance.
Therefore, the pool temperature limit as established is applicable only to the quencher device with the prescribed hole patterns. Any deviation from the specifications in NED0-ll315-08 should be justified.
ii.
Local and Bulk Temoerature Local temperature is defined as the water temperature in the vicinity of the quencher device.
For practical purposes, measurament by temperatt:e sensors, which are located on the containment wall in the sector containing the discharge device and at the same elevation of the discharge device, can be taken to be the local temperature.
Bulk temperature, on the other hand, is a calculated temperature which assumes the pool to be a uniform heat sink. Bulk temperature is calculated on the basis of energy and mass released from the primary system through the safety / relief valves following the plant transients.
, REFERENCES 1.
BAW 10099, " Babcock & Wilcox Anticipated Transients Without Scram Analysis,"
December 1974, 2.
Letter, James F. Malley to Victor Stello, Jr., dated January 2,1975.
3.
CENPD-158, " Analysis of Anticipated Transients Without Scram in Combustion Engineering NSSS's," September 1974.
4.
CENPD-158, Revision 1, May 1976.
5.
WCAP-8330, " Westinghouse Anticipated Transients Without Trip Analyses,"
August 1974.
6.
Letter, C. Eiche1dinger to D. B. Vassallo, dated May 1, 1975.
7.
Response to NRC request for additional ATWS analysis.
8.
NE00-11314-08, "Information Report Mark I Containment Dynamic Loading Condition," July 1975.
, IX.
Information Requirements in Electrical Areas For the " Alternate Plant Modification" requirements of NUREG-0460, Volume 3 (Section 2), the " supplementary electrical equipment to increase t'.e reliability of the scram systems" (for B&W the Back-up scram system; for C-E the supplementary protection system; and for GE the ATWS rod injection system and changes to scram discharge volume) and the "ATWS-mitigating system actuation circuitry" (for B&W, C-E, and W the relief valve, turbine trip, and auxiliary feedwater actuation circuitry, and any other ATWS mitigating system actuation circuitry; and for GE the recirculation pump trip, the Standby Liquid Control System actuation circuitry, and any other ATWS-mitigating system actuation circuitry) are defined to include the equipnent and circuitry from the process sensors through and including, the actuation devices (as defined in IEEE-279).
It does not include the actuated equipment (as defined in IEEE-379-1977; copy attached).
For the sensing portion of the prevention and/or mitigation system, the existing sensors can be relied on if they are shown to have diversity to protect against a common mode failure disabling the current scram system and backup systems.
A.
The following additional information is required for Babcock & Wilcox nuclear steam supply system designs.
1.
For Plants Meeting Alternative 3:
a.*
Provide the preliminary design information for the backup scram system (BUSS), including the system design description, design criteria and bases, functional logic diagrams, schematic wiring diagrams, electrical power distribution diagrams, and physical arrangement drawings.
b.*
Provide a detailed discussion, including diagrams as appropriate, which addresses how each of the requirements (sections) of IEEE-279-1971 are satisfied by the backup scram system.
IEEE Standard Application of the Single Failure Criterion to Nuclear Power Generating Station Class 1E Systerns
- 1. Scope required for the safety system to accomplish its protective functions.
This document interprets the smgle failure criterion, discusses failures, and presents an Class 1E. The safety classification of the elec-acceptable method of single failure analysis for tric equipment and systems that are essential Class IE systems.
to emergency reactor shutdown. containment solation. reactor core cooling. and contain-ment and reactor heat removal. or are other-
- 2. Purpose wise essential in preventing sienificant release of radioactive material to the environment.
The purpose of this document is to interpret channel An arrangement of components and the single failure criterion and to provide modules as required to generate a single protec-guidance in its application. It is intended that tive action signal when required by a generating invokmg system standards which utilize the station condition. A channel loses its identity single failure criterion will result in application where smgle protective action signals are com-of this standard:however, it is not the function bined.
of this standard to identify those standards in ortier to determine where the single failure common mode failure. Multiple failures attribu-criterion is to be applied, or to force compli-table to a common cause.
ance of the single failure critenon on a system.
design basis events. Postulated abnormal events It is the specific function of this document to used in the design to establish the acceptable interpret how the single failure criterion is to performance requirements of the structures, be applied to Class IE systems.
systems, and components.
detectable failures. Failures that wul be identi-
- 3. Definitions fled through periodic testing or will be revealed by alarm or anomalous indication. Component actuated equipment The assembly of prime failures which are detected at the channel or movers and driven equipment used to accomp-system level are detectaMe Mmes.
lish a protective action.
failure. The termination of the ability of an NOTE: The foHowing are eumple of pnme movers:
E turbines, motors. and solenoids. The following are periodic test. Test performed at scheduled in-les of dnven equipment: control rods, pumps, terva's to detect failures and verify operability, actuation device (actuator). A component or protection system. The electrical and mechan-assembly of components that directly controls ical devices (from measured process variables the motive power (electricity, compressed air, to protective action system input terminals) hydraulic fluid, etc) for actuated equipment.
involved in generating those signals associated NOTE: The fouowing are examples of actuation de-vices: circuit breakers, relays. and pilot valves.
clude those that initiate reactor trip, engineered safety features (for example, containment iso-auviary supporting features. Systems or com-lation, core spray, safety injection, pressure ponents which provide services (such as cool-reduction, and air cleaning) and auxiliary sup-ing, lubrication, and energy supply) which are porting features..
.. c.
- Provide additional information, as necessary, to demonstrate how the diversity and independence requirements of Page 18 of Section 2.3 and Appendix C of NUREG-0460 Vol. 3 are satisffed by the backup scram system.
d."
Indicate on schematics and layout drawings where the new equipment interfaces with the existing equipment and where the commonality of the scram system starts.
- e.
- Provide any proposed technical specification changes which correspond to the proposed plant modifications.
2.
For Plants Meeting Alternatives 3 and 4:
Provide a list of all systems, subsystems, and/or components a.
which are required for ATWS mitigation.
- b.
- Provide the preliminary design information for the ATWS-mitigating systems actuation circuitry, including the system design descrip-tion, design criteria and bases, functional logic diagrams, schematic wiring diagrams, electrical power distribution diagrams, and physical arrangement drawings.
- c.
- Provide a detailed discussion, including diagrams as appropriate, which addresses how each of the requirements (sections) of IEEE-279-1971 or Appendix C (of NUREG-0460, Vol. 3) are satisfied by the ATWS mitigating system actuation circuitry.
- d. " Provide additional information as necessary to demonstrate how the diversity and independence requirements of Page 18 of Section 2.3 and Appendix C of NUREG-0460 Vol. 3 are satisfied by the ATWS mitigating system actuation circuitry.
- e.
- Indicate on schematics and layout drawings where the new AMSAC equipment interfaces with the existing equipment.
- f.
- Provide any proposed technical specification changes which correspond to the proposed plant modifications.
B.
The following information is needed for Combustion Engineering nuclear steam supply system designs.
1.
For Plants Meeting Alternative 3:
a*.
Provide the preliminary design information for the supplementary protection system (SPS), including system design description, design criteria and bases, functional logic diagrams, schematic wiring diagrams, electrical power distribution diagrams, and physical arrangement drawings.
b.*
Provide a detailed discussion, including diagrams as appropriate, which addresses how each of the requirements (sections) of IEEE-279-1971 are satisfied by the supplementary protection system.
- c.
- Provide additional information, as necessary, to demonstrate how the diversity and independence requirements of Page 18 of Section 2.3 and Appendix C of NUREG-0460 Vol. 3 are satisfied by the supplementary protection system.
- d.
- Indicate in schematics and layout drawings where the new equipment interfaces with the existing equipment, and where tr.' commonality of the scram system starts.
- e.
- Provide any proposed technical specification changes which correspond to the proposed plant modifications.
. 2.
For Plants Meeting Alte.* natives 3 and 4:
a.
Provide a list of all systems, subsystems and or components which are required for ATWS mitigation.
b.*
Provide the preliminary design information for the ATVS mitigating systems actuation circuitry, including the system design descrip-tion, design criteria and bases, functional logic diagrams, schematic wiring diagrams, electrical power distribution diagrams, and physical arrangement drawings.
c.*
Provide a detailed discussion, including diagrams as appropriate, which addresses how each of the requirements (sections) of IEEE-279-1971 or Appendix C (of NUREG-0460, Vol. 3) are satisfied by the ATWS mitigating system actuation circuitry.
d.*
Provide additional information as necessary to demonstrate how the diversity and independence requirements of Page 18 of Section 2.3 and Appendix C of NUREG-0460 Vol. 3 are satisfied by the ATWS mitigating system actuation circuitry.
e.*
Indicate on schematic and layout drawings where the new AMSAC equipment interfaces with the existing equipment.
- f.
- Provide any proposed technical specification changes which correspond to the proposed plant modifications.
C.
The following information is needed for Westinghouse nuclear steam supply system designs.
1.
For Alternatives 3 and 4:
a.
Provide a list of all systems, subsystems and or components which are required for ATWS mitigation.
. b.*
Provide the preliminary design information for the ATWS mitigating systems actuation circuitry, including system design description, design criteria and bases, functional logic diagrams, schematic wiring dia, grams, electrical power distribution diagrams, and physical arrangement drawings.
c.*
Provide a detailed discussion, including diagracs as appropriate, which addresses how each of the requirements (sections) of IEEE-279-1971 or Appendix C (of NUREG-0460 Vol. 3) are satisfied by the AMSAC.
d.*
Provide additional information, as necessary, to demonstrate how the diversity and independence requirements of Page 18 of Section 2.3 and Appendix C of NUREG-0460 Vol. 3 are satisfied by the AMSAC.
e."
Indicate in schematic and layout drawings where the new AMSAC equipment interfaces with the existing equipment.
- f.
- Provide any proposed technical specification changes which correspond to the proposed plant modifications.
D.
The following information is needed for General Electric nuclear steam supply system designs.
1.
For Alternatives 3 for the ATWS prevention Dortion of the design:
- a.
- Provide the preliminary design information for the ATWS rod injection (ARI) system and scram discharge volume modifications including the system design description, design criteria and bases, functional logic diagrams, schematic wiris, diagrams, electrical power distribution diagrams, and physical arrangement drawings.
\\ b.*
Provide a detailed discussion, including diagrams as appropriate, which addresses how each of the requirements (sections) of IEEE-279-1971 are satisfied by the ATWS rod injection (ARI) system and the scram discharge volume modification, c.*
Provide additional information, as necessary, to demonstrate how the diversity and independence requirements of Page 18 of Section 2.3 and Appendix C of NUREG-0460 Vol. 3 are satisfied by the ATWS rod injection system and the scram discharge volume modification.
- d.
- Indicate in schematics and layout drawings where the new equipment interfaces with the existing equipment and where the commonality of the scram system starts.
e.*
Provide any proposed technical specification changes which correspond to the proposed plant modifications.
2.
For Alternative 3, for the control modifications:
a.*
Provide preliminary design information for the propose, changes in the logic to reduce vessel isolation events and penJt feedwater runback.
Include the description, design criteria and bases and functional logic diagrams and schematic wiring diagrams showing the changes.
b.*
Explain how each of the changes accomplishes its intended objective i.e. reduce isolation events and/or permit feedwater runback.
- c.
- Indicate in schematic arj layout drawings where the proposed logic changes interface with the existing design.
. 3.
For Alternative 3, for the ATWS mitigating syster. actuation circuitry:
Provide a list of all systems, subsystems and/or components a.
which are required for ATWS mitigation.
b.*
Provide the preliminary design information for the ATWS mitigating systems actuation circuitry, including system design description, design criteria and bases, functional logic diagrams, schematic wiring diagrams, electrical power distribution diagrams, and physical arrangement drawings.
- c.
- Provide a detailed discussion, including diagrams as appropriate, which compares the recirculation pump trip (RPT) design to either the 1.
Monticello design as described in NED0-25016, September 1976 2.
Modified Hatch design or 3.
Zimmer design as described in the FFAR.
For any deviations (there may be other acceptable designs) from the above stated comparison designs, provide a discussion which addresses the Appendix C requirements in Volume 3 of NUREG-0460.
- d.
- Provide a detailed discussion, including diagrams as appropriate, which addresses the requirements of NUREG-0460, Appendix C, subparagraphs A through H for the actuation circuitry of the standby liquid control system (SLCS).
Include the discussion which demonstrates that the reliability of this actuation circuitry is equivalent to the remair. der portion of the SLCS.
. e.
- Provide additional information, as necessary, to demonstrate how the special diversity and independence requirements of Page 18 of Section 2.3 and Appendix C of NUREG-0460 Vol. 3 are satisfied by the RPT automatic actuation circuitry and the SLCS actuation circuitry.
1.
If a timed inhibit feature is being considered to prevent unnecessary SLCS actuation, describe alternative inhibit designs, i.e., manual and automatic, for this feature.
For each alternative discuss the benefits and the limitations and justify the selected alternative.
If reliance is placed on indications such as flux level, the discussion should include consideration of the potential for a common mode failure disabling the scram function and inhibiting the SLCS actuation.
- f.
- Indicate in schematic and layout drawings where the new actuation circuitry and its components interface with the existing equipment.
g.*
Provide any proposed technical specification changes which correspond to the proposed plant modifications.
4.
For Alternative 4 Plants:
Provide a list of all systems, subsystems, and/or components a.
which are required for ATWS mitigation.
b.*
Provide the preliminary design information for the ATWS mitigating systems actuation circuitry including the system design description, design criteria and bases, functional logic diagrams, schematic wiring diagrams, electrical power distribution diagrams, and physical arrangement drawings.
.~ c.
- Provide a detailed discussion which addresses the requirements, including diagrams as actuation circuitry for the RPT and hiof Appendix C for th boron system, and any other syst gh capacity automatic listed in item d.l.
ems, subsystems, or components
- d.
- Provide additional information the special diversity and inde, as necessary, to demonstrat by the RPT automatic actuatio pendence requirements ar boron system automatic actuatin circuitry ard the high cap e
on circuitry.
1.
Same as e.1 above as applicabl system.
e to the high capacity boron
- e.
- Indicate where the new actuatio existing equipment.
n circuitry interfaces with the
- f.
- Provide any proposed technical correspond to the proposed plant m dispecification changes w o
fications.
~
. Enclosure 2 List of Plants for Alternative 2 Shippingport Dresden 1 Yankee Rowe Indian Point 1 Humboldt Bay Big Rock Point Connecticut Yankee San Onofre 1 Lacrosse Nine Mile Point Oyster Creek s
. X.
Diversity Considerations The increased assurance of scram reliability attributed to some of the proposed ATWS prevention and mitigation features depends to a large extent upon a careful consideration of common mode failuies in the reactor shutdown system design. Diversity in equipment is one generally accepted method of increasing assurance of high reliability of shutdown and mitigating system (s).
Therefore, include a discussion to demonstrate how the proposed diversity provisions for the mitigating systems protect against the following types of common mode failures which could disable the shutdown system.
1.
External Environment This class of events includes such things as high temperature, moisture, vibration, wear, dirt, and various more severe environmental events such as storms, fires, floods, earthquakes and accident conditions that might act in more or less the same way upon similar components throughout the protection system.
2.
Design or Manufacturing Deficiencies These include common mode failures that might be due to dependence upon a common element unrecognized in the design, to a design error, or to improper manufacture of all components of a similar type.
3.
Ooerating and Maintenance Errors These errors might include incorrect calibration of all of the compo-nents of a given type throughcut the system, inadequate testing, mistakes made in maintenance work that might apply to a series of similar components, incorrect or outdated operating or mair.tenance instructions, or operator errors.
0
. 4.
Functional Deficiencies Such deficiencies might result from an unrecognized deficiency in sensing instrumentation such as not providing needed sensitivity, unanticipated changes in plant operating conditions that leave the protection system inadequate for its purpose, or misunderstanding of the behavior of process variables in the design of the protection system.