ML20132C153
| ML20132C153 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 05/15/1984 |
| From: | Houston R Office of Nuclear Reactor Regulation |
| To: | Novak T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20132C158 | List: |
| References | |
| FOIA-84-926, RTR-NUREG-0737, RTR-NUREG-737, TASK-2.E.1.1, TASK-2.E.1.2, TASK-2.K.2.13, TASK-2.K.2.17, TASK-2.K.2.19, TASK-2.K.3.02, TASK-2.K.3.03, TASK-2.K.3.05, TASK-TM NUDOCS 8405290144 | |
| Download: ML20132C153 (21) | |
Text
DISTRIBUTION
- a ' Docket File RSB R/F RSB P/F:
BV-2 ELantz R/F AD/RS Rdg.
BSheron LMarsb MAY 15 1984 Elantz EMORANDUM FDR:
Thomas M. Novak, Assistant Director for Licensing Division of Licensing FRON:
R. Weyne Houston, Assistant Director for Reactor Safety Division of Systems Integration
SUBJECT:
BEAVER VALLEY UNIT 2 DRAFT SER Plant Name; Beaver Valley Unit 2 Docket No.:
50-412 Licensing Stage:
OL Responsible Branch:
Licensing Branch f3 Project Manager:
M. Licitra' Review Branch:
Reactor Systems Branch Review Status:
Awaiting Information Attached is the Reactor Systens Branch's ir.put to Chapter 15 of the draft SER for BV-2. This includes draft evaluations of the THI related issues for which RSB has lead responsibility.
RSB input to SALP is also attached.
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R. Wayne Houston, Assistant Director for Reactor Safety Division of Systems Integration cc: w/o enclosure RSB Section Leaders M. Licitra M. Ley C0!,' TACT:
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'r 15 ACCIDENT ANALYSES The accident analyses for Beaver Valley Unit 2 have been reviewed in accordance with Section 15 of the SRP (NUREG-0800).
Conformance with the acceptance criteria, except as noted for each of the sections, formed the basis for concluding that the design of the facility for each of the areas reviewed is acceptable.
In accordance with SRP 15.1.1, Paragraph I, the applicant evaluated the ability of Beaver Valley Unit 2 to withstand anticipated operational occurrences and a broad spectrum of postulated accidents without undue hazard to the health and safety of the public.
The results of these analyses are used to show conform-ance with GDC 10, 15, 27, and 31.
For each event analyzed, the worst operating conditions and the most limiting single failure were assumed, and credit was taken for minimum engineered safe-guards response.
In questions 440.73 and 440.74 the staff has asked the applicant to:
1.
Supply listings of the single failures which wo.re assumed for each event in the Chapter 15 analyses.
2.
Supply the limiting single failure that results in the peak pressure or limiting performance for each event.
3.
Show the effect of a loss of offsite power on all anticipated operational occurrences and postulated accidents.
When this information is received it will be incorporated into the evaluations of the individual events.
Parameters specific to individual events were conservatively selected.
Two types of events were analyzed 05/15/E4 i
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t (1) those incidents that might be expected to occur during the lifetime of the reactor (2) those incidents not expected to occur that have the potential to result in significant radioactive material release (accidents)
The nuclear feedback coefficients were conservatively chosea to produce the most adverse core response.
The reactivity insertion curve, used to represent the control rod insertion, accounts for a stuck rod; it is in accordance with GDC 26.
For transients and accidents, the applicant used a method that conservatively bounds the consequences of the event by accounting for fabrication and operating uncertainties directly in the calculations.
DNBRs were calculated using the
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W-3 correlation with a modified spacer factor R, with a minimum DNBR of 1.3 used as the threshold for fuel failure.
The applicant accounts for variations in initial conditions by making the following assumptions as appropriate for the event being considered:
3-Loop Doeration 2-Loop Operatinn Core Power (MWt) 2652 + 2%
1724 + 2%
Average Reactor Vessel Temperature (*F) 576.2 4%
566.0 4%
Pressure (psi)
(at pressurizer) 2250 30 2250 30 The staff concludes the assumptions for initial conditions are acceptable because they are conservatively applied to produce the most adverse effects.
These assumed values will form the basis for the technical specification limits.
For transients and accidents used to verify the ESF design, the applicant used the safeguards power design value of 2780 MWt.
The applicant has also analyzed several events expected to occur one or more times in the life of the plant.
A number of transients can be expected to occur with moderate frequency as a result of equipment malfunctions or operator errors in the course of refueling and power operation during the plant lifetime.
05/15/84 2
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l Specific events were reviewed to ensure conformance with the acceptance crite-ria provided in the SRP.
The acceptance criteria for transients of moderate frequency in the SRP include the following:
(1)
Pressure in the reactor coolant and main steam systems should be main-tained below 110% of the design values (Section III of the. American Socie-
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ty of Mechanical Engineers Boiler and Pressure Vessel Code).
(2)
Fuel clad integrity shall be maintained by ensuring that the minimum DNBR will remain above the 95/95 DNBR limit for PWRs.
(The 95/95 criterion discussed'in Section 4.4 of this SER provides a 95% probability, at a 95%
confidence level, that no fuel rod in the core experiences a DNB.)
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(3) An incident of moderate frequency should not generate a more serious plant condition without other faults occurring independently.
(4)
For transients of moderate frequency in combination with a single failure, no loss of function of any fission product barrier, other than fuel ele-ment cladding, shall occur.
Core geometry is maintained in such a way that there is no loss of core cooling capability and control rod inser-tability is maintained.
Cenform.nce with the SRP acceptance criteria for anticipated operational occur-rences constitutes compliance with GDC 10, 15, and 26 of Appendix A to 10 CFR 50.
See Section 6.8 of this SER for a discussion of auxiliary feedwater system con-
'formance to TMI Action Plan Item II.E.1.1 and Sections 6.8 and 7.3.1.7 for a discussion of compliance with TMI Action Plan Item II.E.1.2.
The transients analyzed are protected by the following reactor trips:
(1) power range high neutron flux (2) high pressure 05/15/54 3
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's (3) low pressure (4) overpower AT (5) overtemperature AT (6) low coolant flow (7) pump undervoltage/underfrequency (8) low steam generator water level (9) high steam generator water level Time delays to trip, calculated for each trip signal, are included in the ana-lyses.
See Section 4.6 of this SER for a discussion of the staff review of reactivity control system functional design.
All of the events that are expected to occur with moderate frequency can be grouped according to the following plant process disturbances:
changes in heat removal by the secondary system, changes in reactor coolant flow rate, changes it, reactivity and power distribution, and changes in reactor coolant inventory.
Design-basis accidents have been evaluated separately and are discussed at the end of this section of the SER.
15.1 Increase in Heat Removal by the Secondary System The applicant's analysis of events that produced increased heat removal by the secondary system is addressed in the following paragraphs.
15.1.1 Decrease in Feedwater Temperature The consequences of a decrease in feedwater temperature transient are bounded by those in Sections 15.1.2 and 15.1.4.
The peak pressure is less than that in Section 15.1.2.
The minimum DNBR is greater than that in Section 1.5.1.4.
15.1.2 Increase in Feedwater Flow Increases in feedwater flow decrease the temperature of the reactor coolant water.
Due to the negative moderator temperature coefficient this will insert 05/15/84 4
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o positive reactivity and increase core power.
In Section 15.1.2.1 of the FSAR the applicant states that for these events the high neutron flux trip, overtemperature AT trip, and overpower AT trip prevent any power increase which could lead to a DNBR less than the limit value of 1.30.
However, the only analytical results presented for these events are those where a steam generator hi-hi level trip closes all feedwater control and isolation valves, ' rips the main feedwater pumps, trips the turbine, and initiates a re-t actor trip.
The applicant states that continuous addition of feedwater is pre-vented by the steam generator hi-hi level trip.
This analysis shows that the maximum reactivity insertion rate due to an in-crease in feedwater flow occurs at no-load conditions and is less than the max-imum value calculated for an inadvertent control rod withdrawal, which is
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evaluated in Section 15.4 of this SER." However, this analysis also shows that an increased feedwater flow event can cause a peak RCS pressure of 2270 psia.
This is below the design pressure of 2485 psig, but it is the highest RCS pres-sure the applicant calculated for any of this group of events.
15.1.3 Increase in Steam Flow The consequences of an increase in steam flow transient are bounded by those in Sections 15.1.2 and 15.1.4.
The peak pressure is less than that in Section 15.1.2.
The minimum DNBR is greater than that in Section 15.1.4.
15.1.4 Inadvertent Opening of a Steam Generator Relief Valve or Safety Valve The transient that is most limiting of this group of transients with respect to fuel performance is the inadvertent opening of the steam generator relief or safety valve.
The suddenly increased steam demand causes a reactor power in-crease which results in a reactor trip due to high neutron flux, overtemperature, or overpower signals.
The continued steam flow through the open valve will cause additional cooldown which will, because of the negative moderator temper-ature coefficient, result in positive reactivity.
The safety injection system (SIS) will inject highly concentrated boric acid from the boron injection tank 05/15/S4 5
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7 into the primary coolant system on either two out of tfiree pressurizer low pres-sure signals, or two out of thre,e low steamline pressure signals in any one loop.
This ensures the reactor will be shut down during any subsequent cool-down.
The normal steam generator feedwater would be isolated automatically upon SIS initiation, and then the plant would be gradually cooled down with only safety grade equipment.
DNB does not occur during this transient.
The applicant has provided results of its study for a transient of this group in combination with its limiting single failure.
No credible single failure has been identified that could result in a more limiting peak reactor coolant system pressure or DNBR than that from the events themselves.
The applicant's analyses show that for transient events leading to an increase in heat removal by the secondary system (with or without single failure), the minimum DNBR is 1.3.
Thus no fuel failure is predicted to occur, core geometry and control rod insertability are maintained with no loss of core cooling capa-bility, and the maximum reactor coolant system pressure remains below 110% of design pressure.
The staff finds the results of these analyses in conformance with the acceptance criteria of SRP 15.1.1 through 15.1.4, and, therefore, acceptable.
15.1.5 Steamline Rupture Accident The applicant has submitted analyses of postulated steamline breaks that show no fuel failures attributed to the accident.
These results are similar to those obtained for previously reviewed Westinghouse three-loop plants.
A postulated double-ended rupture at hot standby power with no decay heat was analyzed as the worst case.
Since the steam generators have integral flow restrictors with a 1.4 ft8 throat area, any rupture with a break area greater than'1.4 ft*, regardless of location, will have the same effect on the system as a 1.4 ft8 break; so this was assumed in the analysis. The doubled ended rupture would cause the reactor to increase in power due to the decrease in reactor coolant temperature.
The reactor would be tripped by either reactor i
overpower AT or by the actuation of the SIS.
The SIS wi'll be actuated by any 05/15/84 6
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of the following: two out of three low pressurizer pressure signals; two out of three HI-1 containment pressure signals; or two out of three low steamline pres-sure signals in any one loop.
The transient is terminated using only safety-l grade equipment.
The injection of highly borated water ensures the reactor is returned to and then maintained in a shutdown condition.
The staff concludes that the consequences of postulated steamline breaks meet the relevant criteria in GDC 27,18, 31, and 35 regarding contr61 rod inser-tability and core coolability and TMI Action Plan Items. This conclusion is based upon the following:
(1) The applicant has met the criteria of GDC 27 and 28 by demonstrating that fuel damage, if any, is such that control rod insertability will be main-tained, and there will be no loss of core cooling capability.
The minimum DNBR experienced by any fuel rod'was 21.30, resulting in none of the fuel elements being predicted to experience cladding perforation.
(2) The applicant has met the criteria of GDC 31 with respect to demonstrating the integrity of the primary system boundary to withstand the postulated accident.
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(3) The applicant has met the criteria of GDC 35 with respect to demonstrating the adequacy of the emergency cooling systems to provide abundant core cooling and reactivity control (via boron injection).
(4) A mathematical model, which accounts for incomplete coolant mixing in the reactor vessel, has been reviewed and found acceptable by the staff. This model was used to analyze the effects of steamline breaks inside and out-side of containment, during various modes of operation, with and without offsite power.
(5) The parameters used as input to this model were reviewed and found to be suitably conservative.
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15.2 Decrease in Heat Removal by the Secondary System The applicant's analyses of events that result in a decrease in heat removal by i
the secondary system are presented below.
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15.2.1 Steam Pressure Regulator Malfunction or Failure that Results in Decreasing Steam Flow I
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In Section 15.2.1 of the FSAR the applicant states that any steam flow decrease caused by a malfunction or failure of any steam pressure regulator is conserva-tively bounded by the turbine trip.eyent.and analyzed in Section 15.2.3.
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15.2.2 Loss of External Load
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InSection15.2.2oftheFSARtheappfic'ani.statesthattheresultsofthetur-bine trip event analysis.are more severe than those expected for the loss of external load. The reason given is that a turbine trip actuates the turbine stop valve whereas a loss of external. load _ actuates only the turbine control valves.
Since the stop valve can more suddenly cut off the steam flow to the~
turbine this is a more severe " decreased heat removal" transient.
15.2.3 Turbine Trip Assuming offsite power is availab1'e to run the reactor coolant pumps, the ap-plicant analyzed the turbine trip event for a complete loss of steam load from full power without a direct reactor tr.ip and with only the pressurizer and steam generator safety valves assumed for. pressure relief.
These assumptions
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result in the highest peak RCS pressure _ for.any " decreased heat removal" event.
The calculated peak value is 2560 psia, which is well below the ASME limit of
'10% of the design pressure.
For these assumptions the minimum DNBR is 1.75, which is well above the minimum limiting value of 1.30.
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The applicant's analyses show that if instead of relying on just the safety valves, the pressurizer spray and p0RV's are used to limit the_ pressure during this turbine trip event, the minimum DNBR can go down to 1.60.
If a stuck open 05/15/84 8
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PORV were to be assumed as the single failure during this course of action, it I
appears that the DNBR could go 1ower.
The applicant has not discussed the pos-sibility of a stuck open PORV or atmospheric steam dump valve being the worst single failure during this course of action.
The consequences of a turbine trip without offsite power available are dis-cussed in Section 15.2.6.
t 15.2.4 Inadvertent C1'osure of Main Steam Isolation Valves Consequences are the same as those discussed in Sections 15.2.3 and 15.2.6.
15.2.5 Loss of Condenser Vacuum Consequences are the same as those discussed in Sections 15.2.3 and 15.2.6.
15.Z.6 Loss of Nonemergency AC Power to the Plant Auxiliaries j
A loss of nonemergency ac power event is more limiting than the turbine-trip-initiated decrease in secondary heat removal without loss of ac power because the reactor coolant pumps are lost and the subsequent flow coastdown further reduces the amount of heat the primary coolant can remove from the core.
In this transient, the loss of offsite power is closely followed by a turbine trip and reactor trip.
The reactor trip is assumed to come from low-low steam gen-erator level which is the second safety grade trip. The emergency feedwater system is automatically started and one electric-motor-driven pump is assumed to be feeding all three steam generators.
The applicant's LOFTRAN analysis shows that the natural circulation flow avail-able adequately transfers the decay heat from the core.to steam generators, which are being fed with emergency feedwater flow.
The steam which is generat-t ed is assumed to be relieved through the steam generator safety valves.
The primary system relief valves are assumed not to function.
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The emergency feedwater comes from the primary plant demineralizer water stor-age tank (pPDWST) which, the applicant states in FSAR Section 10.4.9.1, contains sufficient water to reduce the hot leg temperatures to 350'F for this transient.
At 350*F the RHRS can be started to take away the decay heat.
The DNBR remains above 1.30 throughout this transient, and the peak RCS pres-sure remains below 110% of the design pressure.
15.2.7 Loss of Normal Feedwater Flow The consequences of this anticipated operatforial occurrence are more severe if a concurrent loss of offsite power is assumed.
However, if a loss of offsite power is assumed the consequences will be the same as the loss of nonemergency ac power event discussed in Section 15.2.6.
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15.2.8 Feedwater System Pipe Breaks The applicant has provided a feedwater line break analysis for Beaver Valley Unit 2 using assumptions that minimize secondary system heat removal capabili-ty, maximize heat addition to the primary system coolant, and maximize the cal-culated primary system pressure. A double ended. rupture of the largest feedwater line was assumed, as well as failure of the turbine-driven auxiliary feedwater pump to start and supply emergency feedwater to the ' steam generator.
The applicant used the NRC approved LOFTRAN code to do this analysis. The analysis assumed that with a single fcilure of the auxiliary feedwater system, i
emergency feedwater flow is supplied to two intact steam generators by only one electric motor-driven auxiliary feedpump. This' is sufficient feedwater flow to adequately remove the residual heat after reactor shutdown.
The use of only safety grade equipment will mitigate this accident.
No fuel damage was calcu-lated to occur, and the peak calculated pressurizer pressure was approximately 2500 psia.
As required for all other events a list of the single failures that were considered and the most limiting single failure must be provided.
05/15/84 10 BEAVER VALLEY SER INPUT SEC 15
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p 15.3 Decreases in Reactor Coolant Flow Rate 15.3.1/15.3.2 Loss of Forced Reactor Ccolant Flow, Including Trip of Pump and Flow Controller Malfunctions The applicant has analyzed the total loss of forced reactor coolant flow event that bounds partial loss of forced reactor coolant flow.
This event was re-viewed with the procedures and acceptance criteria set forth in-SRp 15.3.1 -15.3.2.
The loss of offsite power and resulting loss of all forced coolant flow through the reactor core causes an increase in the average coolant temperature and a decrease in the margin to DNB.
The reactor is tripped from an undervoltage trip monitoring the reactor coolant pump (RCP) power supply, and a minimum DNBR of 1.47 is reached 3.2 seconds into the transient.
The maximum calculated RCS pressure is 2310 psia during the transient.
15.3.3/15.3.4 Reactor Coolant Pump Rotor Seizure and Shaft Break Accident The applicant has analyzed the reactor coolant pump (RCP) rotor seizure and shaft break events with the LOFTRAN and FACTRAN computer codes.
Since the ini-tial rate of reduction of coolant' flow is greater after an RCP rotor seizure, this is the limiting event.
For the analyses the applicant assumed that the fuel cooling goes into the nucleate boiling regime (i.e., DNB) immediately at the beginning of the transient.
The maximum RCS pressure will occur in the event of an RCP rotor seizure while only two of the three loops are operating.
This maximum pressure is calculated to be 2647 psia with only the opening of the pressurizer and steam generator safety valves.
The applicant states that i
2647 psia is below the faulted condition stress limit of the RCS, In response to a question on a loss of offsite power (LOOP) during these events, the applicant states that a LOOP will have only a negligible effect on the cri-tical parameters of RCS pressure and clad temperature and that it would have no effect whatsoever on the conclusions. The staff finds that a quantitative analy-sis of the worst case, which would have only two loops in operation, with a 05/15/E4 11 BEAVER VALLEY SER INPUT SEC 15
concurrent loss of offsite power is needed for the evaluation of this issue.
The staff's evaluation and finding on fuel damage and consequent control rod insertability and core cooling considerations during this event are included in SER Section 4.2.
The L0r7RAN computer code has been approved by the NRC.
The remaining staff findings are (1) The parameters used as input to the mathematical model are suitably conservative.,
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(2) The use of " Service Limit C" of the ASME Code is acceptable for conforming to GDC 31 and demonstrating the integrity of the RCS during this accident; the maximum pressure is below this limit.
15.4 Chanoes in Reactivity and Power' Distribution 15.4.4/15.4.5 Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature In FSAR Section 15.4.4, the applicant provides the results of an analysis for startup of an inactive reactor coolant pump event.
This event was reviewed with the procedures and acceptarce criteria set forth in SRP 15.4.4.
During the first part of the transient, the increase in core flow with cold water results in an increase in nuclear power and a decrease in core average temperature.
Reactivity addition for the inactive loop startup event is the result of the decrease in core inlet water temperature. This transient was evaluated by the applicant using a mathematical model that has been reviewed and foun'd acceptable to the staff. The maximum calculated RCS pressure is 2310 psia and the minimum DNBR is above 1.3 throughout the transient.
15.4.6 Inadvertent Boron Dilution Various chemical and volume control system (CVCS) malfunctions which could lead to an unplanned boron dilution incident have been review'd.
The malfunctions e
05/15/84 12 BEAVER VALLEY SER INPUT SEC 15
- y that allow the operator the shortest time for corrective action have been ana-lyzed starting from plant conditions of startup, power' operation (automatic and manual), hot standby, and cold shutdown.
The applicant used acceptably conser-vative assumptions in these analyses.
The results show that the operator has at least 15 minutes between the time when an alarm announces an unplanned mod-erator dilution and the time of loss of shutdown margin, i.e., criticality.
The maximum reactivity insertion rate by boron dilution was found to be 1.5X10 5 Ak/k (1.5 pcm) per second.
In the event the operator does not stop the dilu-tion, the DNBR will still remain above 1.49, and the RCS and main steam pres-sures will remain below 110% of design.
In response to 'a question on protection from inadvertent boron dilution during refueling, the applicant stated that during refueling the RCS is isolated from the potential source of unborated water. This isolation is accompiished by having the operators place danger tags on the primary grade water header isola-tion valves, or by locking these valves closed whenever the RCS water is below the normal level.
The operator performing these tasks is required to sign off on each step of a procedural checklist. This long term use of administrative controls to prevent an inadvertent boron dilution during refueling has not been accepted by the staff on other plants, and will be evaluated.
The staff is not at this point, convinced that a design basis event can be eliminated from de-tailed evaluation based on administrative means alone.
We will report the res-olution of this issue in a subsequent safety evaluation.
With the exception of the refueling mode the staff concludes that the analysis for the decrease in reactor coolant boron concentration event is acceptable and conforms to General Design Criterion 10, 15, and 26.
This conclusion is based on the following:
1.
The applicant has met the criteria of GDC 10 with respect to demonstrating that the specified acceptable fuel design limits are not exceeded for this event.
This criterion has been met since the results of the analysis showed that'the thermal margin limits are satisfied as indicated by SER Section 4.4.
05/15/84 13 BEAVER VALLEY SER INPUT SEC 15 4
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The applicant has met the criteria of GDC 15 with respect to demonstrating 2.
that the reactor coolant pr. essure boundary limits have not been exceeded for this event.
This criterion has been met since the analysis showed l
i that the maximum pressure in the reactor coolant and main steam systems did not exceed 110% of the design pressure.
3.
The applicant has met the criteria of GDC 26'with respect to demonstrating that the control rod system has the capability of overcoming the effects of boron dilution events during reactor operation. The applicant has dem-onstrated conformance with these criteria by showing that under the postu-lated accident conditions, and with appropriate margins for stuck rods,
'he ~specified acceptable fuel design limits are not exceeded.
t 15.5 Increases in Reactor Coolant System Inventory 15.5.1 Inadve~rtent Ope ~ratioW of 'the Emergency Core Cooling System During I
Power Operation ECCS operation could be initiated by a spurious signal or an operator error.
Two cases were examined, one in which reactor trip occurs simultaneously as a result-of the safety injection signal, and the other in which the reactor trips later in the transient because of low reactor coolant system (RCS) pressure.
The reactor pressure decreases during the initial phase of the transient and then increases to a peak pres ~sure of 2350 psia at 200 seconds into the tran-
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sient. The DNBR never 4rops below its initial value for either case.
All of these transients are ' terminated by use of only safety grade systems.
If the
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operator fails to turn off the HHSI/ charging pumps the safety valves will open.
Continued operation of these pumps would overfill the Pressure Relief Tank.
However, as stated in Table 6.3-1 of the FSAR the cutoff head of the HHSI/
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charging pumps is 6000 ft (2600 psig); so they cannot create 110% of the reac-tor vessel design pressure (2733 psig) and thus cannot fail the vessel.
15.5.2 CVCS Malfunction That Increases Reactor Coolant Inventory Evaluation of consequences is included in Section 15.4.6.
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15.6 Decrease in Reactor Coolant Inventory r
15.6.1 Inadvertent Opening of a Pressurizer Safety or Relief Valve In FSAR Section 15.6.1, the applicant provides the results of an analysis for inadvertent opening of a pressurizer safety valve.
During this event, nuclear power is maintained at the initial value until reactor trip occurs on low pres-surizer pressure.
The DNBR decreases initially, but increases. rapidly follow-ing the trip.
The minimum DNBR of 1.50 occurred at 31 seconds into the transient. The RCS pressure decreases throughout the transient.
15.6.3 Steam Generator Tube Rupture In response to the staff's concern that 30 minutes is not sufficient to diag-nose and isolate a stream generator tube rupture, the applicant has provided additional data regarding the systems response and radiological consequences aftgr a steam generator tube rupture accident.
This information, however, did not support the isolation time of the affected steam generator at 30 minutes.
Upon receipt of additional information, the staff will complete the review of the consequences of this accident and provide our evaluation.
15.6.5 LOCAs In FSAR Section 15.6.5, the applicant has analyzed the double-ended cold leg guillotine (DECLG) as the most limiting large-break LOCA. The analysis was done for three different flow coefficients.
The results of these show that the
~ DECLG with a Moody break discharge coefficient of 0.4 is the worst case. In this analysis, the peak clad temperature reached is 2179'F.
For the small-break LOCA the applicant-has determined that a cold leg rupture of less than 10-in. diameter is the most limiting.
The analysis was performed for 3-in.,
4-in, and 6-in.-diameter breaks.
The results show that the 3-in.-diameter break is the worst case, and it results in a peak clad temperature of 1985'F.
Both of these accidents are terminated by SIS and ECCS operations.
Only safety grade equipment is used to mitigate the accident.
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l The applicant has performed analyses of the performance of the ECCS in accor-dance with the Commission's regulations (10 CFR 50.46 and Appendix K to 10 CFR 50).
The analyses considered a spectrum of postulated break sizes and locations.
As s
shown in NUREG-0390, these analyses were performed with an evaluation model that had been previously reviewed and approved by the staff.
The results show that the ECCS satisfy the following criteria:
(1) The calculated maximum fuel rod cladding temperature does not exceed 2200*F.
(2) The calculated maximum local oxidation of the cladding does not exceed 17%
of the total cladding thickness before oxidation.
(3) The calculated total amount of hy'drogen generated from the chemical reac-tion of the cladding with water or steam does not exceed 1% of th.e hypo-thetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surround-ing the plenum volume, were to react.
(4)
Calculated changes in core geometry are suen that the core remains amena-ble to cooling.
(5)
After any calculated successful initial operation of the ECCS, the calcu-lated core temperature is maintained at an acceptably low value and decay heat is removed for the extended period of time required by the long-lived radioactivity.
The staff concludes that the calculated performance of the ECCS following pos-tulated LOCA accidents conforn. to the Commission's regulations and to applica-ble regulatory guides and staff technical positions, and the ECCS performance i
is considered acceptable for the postulated accidents.
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15.9 TMI Action Plan Recuirements 15.9.1/15.9.2 II.K.1.5 Review ESF Valve Positions, Controls, and Related Test and Maintenance Procedures To Assure Proper.ESF Functioning II.K.I.10 Review and Modify Procedures for Removing ESF From Service To Assure Opera'bility Status Is Known The applicant states that the intent of these two items will be met when the Operating and Maintenance Procedures are written. They are scheduled to be completed in June, 1985. The acceptability of the measures taken to satisfy these items will be evaluated when these procedures are submitted.
15.9.3 II.K.2.13 Thermal Mechanical Report:
Effect of High-Pressure Injection on Vessel Integrity for Small-Break LOCA With No Auxiliary Feedwater Staff review of this item will be covered in NRC unresolved safety issue A-49,
" Pressurized Thermal Shock."
15.9.4 II.K.2.17 Potential for Voiding in the Reactor Coolant System During Transients Westinghouse has performed a study that addresses the potential for void forma-
' tion in Westinghouse-designed NSSS during natural circulation cooldown/
depressurization transients.
This study has been submitted to the NRC by the
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Westinghouse Dwners Group. As stated in R. Wayne Houston's December 6, 1983 memorandum to Gus C. Lainas entitled, "Multiplant Action Item F-33, Voiding in the Reactor Coolant System During Anticipated Transients," the results of this study have been accepted.
05/15/64 17 BEAVER VALLEY SER INPUT SEC 15
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t 15.9.5 II.K.2.19 Sequential Auxiliary Flow Analysis Sequential auxiliary feedwater flow criteria are only of concern to once-through steam gene.rator designs..Since Westinghouse has inverted U-tube steam genera-tor designs, the analysis requested by Item II.K.2.19 is not needed for Beaver Valley Unit 2.
15.9.6 II.K.3.2 Report on Overall Safety Effect of Power-Operated Relief Valve Isolation System As a response to Item II.K.3.2, the applicant referenced a generic Westinghouse Owners Group submittal.
Should staff generic review of this material conclude otherwise, NRC will request further consideration of modification of Beaver Valley Unit 2.
15.9.7 II.K.3.3 Reporting SV and PORV Challenges and Failures The applicant states in FSAR Table 1.10-1 that it will be responsible for en-suring that any failure of PORVs or safety valves to close will be reported promptly to the NRC and that all challenges to PORVs and safety valves will be documented in the annual report. The staff concludes that the Beaver Valley Unit 2 procedures meet the criteria of this item and are acceptable.
15.9.8 II.K.3.5 Automatic Trip of RCPs During LOCA In response to this criterion, the applicant stated that Westinghouse performed an analysis of delayed RCP trip during LOCA.
This analysis is documented and is the basis for the Westinghouse position on RCP trip (i.e., automatic RCP trip is not necessary because sufficient time is available for manual tripping of the RCPs).
Westinghouse has submitted a generic report which is under review.
The appli -
cant should state whether or not it intends to endorse this report and comply with the criteria proposed in it assuming the NRC finds it acceptable.
05/15/84 18 BEAVER VALLEY SER INPUT SEC 15
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discussed in FSAR Section 15.6.5.
However, this does not constitute a review.
that shows Beaver Valley Unit 2 is in full compliance w'ith 10 CFR 53.46. After the staff's review of this evaluation model is completed a specific submittal en this issue will be required.
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e BEAVER VALLEY SER INPUT SEC 15 05/15/84 20
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SALP FOR BEAVER VALLEY UNIT 2 SER e
t Evaluation Criteria
- Category 1.
Management involvement in assuring 2
quality 2.
Approach to resolution of technical 2
3.
Responsiveness to NRC initiatives 2
4.
Staffing (including management)
N/A 5.
Reporting and analysis of reportable N/A events 6.
Training ef fectiveness and qualification N/A
- Ref NRC Appendix 0516 - Systematic Assessment of Licensee Performance.
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ENCLOSURE 1 JljN 18 1984 MENORANDtM FOR: Thomas M. Novak Assistant Director for Licensing Division of Licensing FRON:
R. Wayne Houston, Assistant Director for Reactor Safety Divi. ion of Systems Integration l
SUBJECT:
BASIS FOR BEAYER VALLEY 2 FEEDWATER ISOLATI0H ON HIGH STEAM GENERATOR LEVEL DESIGN REQllIREMENTS Plant Name:
Beaver Valley 2 Docket No.:
50-412
/
Licensing Status:
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Responsible Branch:. LBf3 l
Pmject Manager:
L. Lazo Review Branch:
ICSB Review Status:
Incomplete In Section 7.3.3.12 of the Beaver Valley 2 draft SER, ICSB expressed a concern that the design of feedwater isolation on a high steam generator level did not meet the requirements of paragraph 4.7 of IEEE-STD-279.
The applicant's response to that concern, dated March 28, 1984, stated that IEEE-STD-279 is not applicable to the issue.
/:
In response, ICSB stated, in our Licensing Position il for Beaver Yalley 2 dated April 30, 1984, that either the design of the feedwater isolation on a high steam generator level be modified to meet the re-quirements of IEEE-STD-279 or an analysis be provided to show that the consequences of feedwater addition not being terminated by the high steam generator level signal are not safety significant.
In a May 30, 1984, response, the applicant claims that ICSB's position is a new requirement and should be processed in accordance with NRR pmcedures for plant specific backfitting.
ICSB has reviewed the applicant's claim and believes that the char-acterization of this issue as a backfit is inappropriate. As stated in ICSB Licensing Position #1, the applicant, in Chapter 15 of the Beaver Valley 2 FSAR, takes credit for feedwater isolation on a high steam generator level signal and identifies this isolation action as an engineered safety feature actuation function.
Consistent with this, the applicant, in Section 7.3 of the FSAR, has identified IEEE-STD-279 as an acceptance criterion for the feedwater isolation func-tion and has further claimed to meet those requirements (specifically including paragraph 4.7 of IEEE-STD-279).
CONTACT:
F. Burrows, ICSB
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