ML20058H321
| ML20058H321 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 07/30/1982 |
| From: | Longenecker J ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT |
| To: | Check P Office of Nuclear Reactor Regulation |
| References | |
| MQ:S:82:074, MQ:S:82:74, NUDOCS 8208030540 | |
| Download: ML20058H321 (28) | |
Text
_
Department of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:82:074 JUL 3 0 D32 Mr. Paul S. Check, Director CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Comission Washington, D.C.
20555 l
Dear Mr. Check:
l RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION l
Reference:
Letters, P. S. Check to J. R. Longenecker "CRBRP Request for Additional Information," dated April 30, May 14, and June 9,1982, This letter formally responds to your request for additional infonnation contained in the reference letters.
Enclosed are responses to Questions CS 760.23, 37, 123, 168, 169, 171, 172, and 174 which will also be incorporated into the PSAR Amendment 70; scheduled for submittal later in August.
Sincerely, Oz n us
'W J n R. Longengker Acting Director, Office of the t
Clinch River Breeder Reactor Plant Project Office of Nuclear Energy Enclosures cc: Service List Standard Distribution Licensing Distribution pool 8208030540 820730 PDR ADOCK 05000 t
__________o
NN Ouestion CS760.23 Section 15.2.2.3.1 implies that flux tilts can exist during normal operation (e.g., "the other rods wil I movo in to compensate for the excursion and no transient results.") Aro flux tilts possible?
If so, what offect do these flux tilts have on predicted clad temperatures? On plant protectivo system instrumentation?
Rosconse:
Compensated flux tilts cro not considered to be a credible occurrence due to a rod misalignment rod block system which precludes such an occurrence.
Analysis indicted that the probability of a flux tilt resulting from the withdrawal of a singlegontrol rod with the other rods moving in to compensate is on the order of 10-per year. This probability is so low that this type of flux tilt is considered to be beyond the design basis.
Flux tilts resulting f rom the withdrawal of a singlo control rod without movement of the other rods to compensato has boon analyzed as having a probabit Ity of occurrence of less than 5 x 10-7 per year.
This event is classified as an Extracely Unlikely event and will be analyzed in Chapter 15 of the PSAR on this basis.
The rod misalignment rod block system uses rod position Information provided by both the absoluto and relative rod position Indicating systems in two rodundant logic trains.
If either of the two logic trains detects a rod bank misal ignment condition, rod block signals are sent to each of the individual rod control controllors which wilI prohibit any further rod motion except scran.
An alarm signal is also provided to the operator upon the occurrence of a rod block.
QCS760.23-1 Amend. 70 Aug. 1982 sieusane
wwg.rv wec=wnsu-g.r,macuan-7.7 INSTRUMENTATION AND CONTROL SYSTEMS NOT REOUIRED FOR SAFETY The Plant Control System and other auxiliary Instrumentation and control systems are not safety-related sinco these systems are not required to terminate or mitigate the consequences of postulated events to prevent unacceptable consequences nor will f ailure of these systems af fect the operation or ubility of the Plant Protection System to terminate the results of the PPS design basis events. The Plant Control System. which provides overall control and coordination of the Reactor. Heat Transport. Nuclear Island Auxillary and Balance of Plant Systems for all normal plant operating modes is described in Section 7.7 as are other auxiliary instramentation and control systems.
The Instrumentation and Control Systems not required for safety include:
o Supervisory Control System o
Reactor Control System Primary & Secondary CRDM (Control Rod Drive Mechanism) Controller o
Systems o
Rod Position Indication System o
Rod Misalignment Rod Block System j
e Sodium Flow Control System o
Steam Generator Feedwater Flow Contro: System o
Feedwater Pump Speed Control System o
Recirculation Flow Control System Sodium Dump Tank Pressure Control System o
o Fuel Handling and Storage Control System Nuclear Island Auxiliary Instrumentation end Control Systems o
o BOP instrumentation and Control Systems 7.7.1 Plant Control System Descrlotion The Plant Control System Integrates the manual and automatic controls provided to maintain the plant at the desired power, temperature, pressuro, and flow conditions for startup, load changing, rated power, standby, and shutdown conditions. The Plant Control System equipment includes: the control hardware for the Supervisory, Reactor, Sodium Flow, and Feedwater Control Loops; the Control Rod Drive Mechanism electrical and electronic processors; the Rod Position Indication and Display; and the Inter-system Interlocks and permissives.
Automatic control of the power, sodium and steam temperatures, steam pressure, sodium flows, steam and water flows is provided for, load changing and power operation above 40% thermal power.
Below 40% thermal power the plant is controlled either manually or by flux control by the operator.
7.7-1 Amend. 70 Aug. 1982 m
in cny automatic control mode, or in Group Manual mode, the mechanism controllers are operated in sequence one step at a time to keep the rod bank in required alignment. The sequence rate and direction are determined re:,pectively by analog and digital signals f rom the reactor control system.
If the selector sequence rate is higher than a predetermined trip point, an overspeed detector will alarm and place the controllers in HOLD.
A functional block diagram of the control is shown in Figure 7.7-5.
Ifold Bus A Hold Bus Power Supply and transf er select circuitry are provided to allow any controller to be replaced without a plant shutdown.
In the event of a controller failure, the mechanism controller in question can be switched out and transferred to a Hold Bus. Power to the Hold Bus Power Supply is provided downstream from the scram breakers.
This ensures that if a scram is initiated, a rod on the Hold Bus will also scram.
7.7.1.3.2 Primary Rod Position Indication System l
Two Independent Rod Position Indicating Systems are provided for each primary l
control rod: An Absolute Position Indication System (ARPI) and a Relative Position Indication System (RRPI). These systems assure that the plant operators can continuously determine the position of the control rods.
The ARPI provides a direct measurement of rod position at any time and, unlike the RRPI, does not require re-zeroing after a scram or temporary loss of power. The system is solid state, utilizing ultrasonics and magnetics to provide a D.C. output indicative of rod positlon.
The sensor for this system consists of a tube extending down f rom the top of the motor tube and into the inside diameter of the PCRDM lead screw. A nickel-cadmium wire is stretched axially through the tube.
As the lead screw translates, the flux from a torroidal magnet mounted on top of the lead screw intersects the wire at a point indicative of the rod position.
Electrical pulses sent down the wire generate magnetic fields which, when they intersect thu flux of the lead screw magnet, causes a torsional strain creating a sonic pulse which travels f rom the point of flux Intersection upward. The sonic pulse is dviected at the top of the wire, and the time of propogation is measured electronically. This propagation time is converted to a D.C. signal which is analagous to rod position.
This signal is read out on the main control panel by rod top and rod bottom indicator lights and a vertical bar graph Indicator.
It is also used to operate the rod out of alignment alarm, the rod misalignment rod block system and rod control interIocks.
7.7-6 Amend. 70 Aug. 1982 L
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(Paga 9 In-M [@,@O CM The Relativo Rod Position Indication System provides a digital rod position Indication on a CRT at the Main Control Board.
Two pairs of magnetic coil pick-ups are mounted within each stator Jacket above the stator and on oppo-sito sides.
A 6 polo magnetic section is attached to the mechanism rotor and rotates in the plano of the pick-up coils.
Voltage pulses caused by the move-mont of the poles in the proximity of the pick-up coils are sent to a digital to analog converter.
The D/A converter produces an analog signal which is a measure of rod position.
This analog signal is sent to the PDH&DS and the rod misalignment rod block system.
The resolution of this signal is 0.1 inch.
Unlike the Absoluto Position Indication System, this system must be reset after each scram and in the event of u power failure reset after power is restored.
The pulsos ero also counted by an odometer type roadout in the rod control equipment room.
7.7.1.3.3 Rod Misalignment Rod Block Svstem The rod misalignment rod block system ensures that a row 7 control rod cannot be withdrawn more than a set distance above the average position of the six row 7 control rods when the plant is operating. As shown in Figure 7.7-6, rod position signals f rom the Relative Rod Position Indication (RRPI) and Absolute Rod Position Indication (ARPI) systems aro used by two redundant trains of rod blocking logic.
Each logic train outputs a rod block signal when the position of one of the six row 7 control rods is more than a set distance above the averago position of all the six row 7 rods comprising the operating bank.
A rod block signal from olther of the two redundant logic trains results in all controllers for the six rods of the operating bank switching to the HOLD mode.
Signals are also provided to the unit load controller of the supervisory control system to ensure that a plant loading or unloading is stopped upon the occurrenco of a rod block.
This prevents a reactor trip due to power / flow mismatches which may occur if sodium flow is allowed to change wl1hout a corresponding change in reactor power.
In addition to the redundant logic trains, the rod block system includes:
- 1) Circuitry necessary to convert the pulses of the RRPI signal conditioners to an analog signal.
- 2) Deviation alarms which continually compare the RRPI signal and ARPl signal from each rod and from the rod position overage circuit and provide a position f ault alarm to the Plant Annunicator System when the two signals dif f er by a set amount.
- 3) A Low Power Bypass in each logic train which may be manually Instated at low power to disable the rod block system.
This bypass is provided to allow for control rod movement which is necessary to perform low power physics and startup testing.
This bypass is automatically removed during the ascent to power.
- 4) A momentary manual overrido f oeture to allow the removal of the rod block so that the operating bank may be realigned if a misalignment occurs.
When the manual overrido feature is engaged, the operator may manually insert control rods to real!gn the operating bank.
Withdrawal of control rods while the manual override featuro is engaged is automatically prohibited.
7.7-7 Amend. 70 Aug. 1982
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- 5) Testing and bypass features to allow for the testing and maintenance of the RRPI, ARPI or one train of the rod block system during plant operation.
- 6) System alarm outputs which provide signals to the Plant Annunciator System when either train is bypassed or upon the occurrence of a rod block.
7.7.1.4 Sodlum Flow Control System The Sodium Flow Control System consists of six controllers used to drive the three primary and three intermediate sodium pumps.
Each controller consists of a cascade system with an inner loop using speed as the feedback signal and an outer loop based on a flow feedback signal. The flow control range is 30, to 100% of rated flow.
The flow setpoints are generated either manually or by the Supervisory Control.
Figure 7.7-7 is a block diagram of the flow / speed control loop which is typical of the six controllers in the system.
The Speed Control System is an inner loop and used pump speed, which is sensed via a pump shaft mounted tachometer, as the feedback variable.
The Speed Control System is limited Internally by the torque limit circuit which sets both the accelerating and decelerating torque of the variable speed pump drive.
The demand to the Speed Controller is set by the FLOW / SPEED Mode Select Switch.
in the Speed Mode, pump speed is set by a manually adjusted potentiometer; in the Flow Mode, pump speed is set by the Flow Controller.
The Flow Controller uses the filtered, median select signal of three available redundant flow meter buffered PPS outputs as the feedback signal.
This signal, along with the flow demand, is used to generato the error signal which is compensated through the Control Compensation Network and then limited by the High Speed Limit Circuit prior to being used as the speed demand signal.
The demand to the Flow Controller is set by the MAN / AUTO Select Switch.
In the automatic mode, the demand comes from the supervisory control, while in the Manual Mode, the demand comes from a manually adjusted potentiometer on the control panel.
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I 7.7-7 a Amend. 70 Aug. 1982
7.7.2.2 Reactor Control Svstem The Reactor Control System shown in Figure 7.7-3 contains an Outer Core Exit Temperature Controller with an Inner loop based on flux feedback. Failure in this system could result in erroneous movement of the control rods.
This could result from failure in the sensors or feedback signal conditioning 1
(i.e., flux or tanperature), f ailures in the controller electronics, or a failure in the CRDM controller.
The Reactor Control System has redundant sensors and average / reject or median select circuitry to prevent single sensor l
f ailures from initiating reactivity transients. Even though it is highly improbable, simultaneous multiple f ailures in PPS compensated ionization chamber instrumentation could cause the loss of flux instrumentation channels in both the Plant Protection and Control Systems, the consequences of this,
potential failure will be mitigated by diverse fission chamber instrumentation in the secondary reactor shutdown system.
Rod withdrawal block circuitry and a rod misalignment rod block system are provided which are independent of the normal control to prevent control electronic f ailures from causing reactor trip.
Withdrawal blocks are initiated for both high power and power-to-flow ratio.
These withdrawal blocks operate directly on the Control Rod Drive Mechanism Controllers to stop outward rod motion of all primary rods.
Withdrawal and insertion blocks are initiated by the rod misalignment rod block system to prevent severe misalignment of the control rod bank. The Control Rod Drive Mechanism Controller and Rod Sequencer include overspeed detector and block circuitry to provide assured limitation of rod withdrawal speed even if reactor controf f alIures and f aliures of the rod block or overspeed circuitry are postulated. The PPS acceptably terminates the results as shown in Chapter 15.
It is also considered that a f ailure of the Reactor Control System could result in improper banking of the control rods which is not severe enough to require action by the rod misalignment rod block system.
Under these condi-l tions the reactor operator wou!d have to readjust the out of bank rods manually. To aid the operation, the main control board is equipped with rod i
position indications for each rod and also an alarm if the rods deviate from the proper banking requirements.
7.7.2.3 Sodium Flow Control System l
A block diagram of the Sodium Flow Control System is given in Figure 7.7-7 which is typical of the six HTS flow loops.
The controller contains an outer j
flow loop with an inner loop based on pump speed.
A failure of any of the six flow controllers would result in improper pump speed and, consequently.
undesired sodium flow. Power to flow or primary to intermediate flow mismatch would occur resulting in a plant trip.
Even though it is highly improbable, multiple f ailures in PPS flow instrumentation could cause the loss of flow l
Instrumentation channels such that the secondary RRS fails to trip; the consequences of this potential failure to initiate control system action which requires Protection System action wilI be mitigated by the primary reactor shutdown system.
Pump speed instrumentation; this is Independent of the flow instrumentation and is, theref ore, not af fected by these f ailures.
i 7.7-18 Amend. 70 Aug. 1982 rgwsw
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Figure 7.7-6 General Block Diagram for the Rod Misalignment Rod Block System 1
l 7.7-25 Amend 70 l
l Aug 1982
's Question CS760.37_
Given that the project's " leak before break" philosophy has not been accepted by the NRC, we fool that the analysis of large pipe breaks should be included in the undercooling events.
While the likelihood and size of pipe breaks may be resolved at some future date, it is important to understand the reactors inherent ability to survive even hypothetical cccidents.
Please include results for a spectrum of pipe break aims and locations including a double-ended of f set shear of the hot leg at the pump outlet.
Ecsponse Information on this topic has been provided in prior submittals.
~
Detailed information on the structure integrity of the piping is given in WARD-D-0185 (Ref. 2 of PSAR Section 1.6).
This integrity study concludes that i
large pipe breaks need not be considered for the heat transport system piping in the Reactor Containment Building.
The study further showed that the maximum credible fatigue crack (for the entire CRBRP duty cycle) would be 4 Inches long.
Supporting information on in-service inspection is given in PSAR Appendix G and relevant data on leak detection is given in PSAR Section 7.5.
The leak from a 4 inch crack is the piping Design Basis Leak (DBL) for assessing for functional performance of the HTS and auxiliary systems.
The systems ef fects of leak rates up to 4 times the DBL rate from the PHTS piping are discussed in PSAR Section 15.3.3.4.
In addition, Section 15.3.3.4 Indi-cafes that a leak rate 4 orders of magnitude greater than the DBL must be hypothesized before sodium temperature saturation would be approached in the core (and thus, core coolability begins to be a concern).
Although the evaluations discussed above do not include specific analyses of a double-ended of f set shear of the hot leg at the pump outlet, they do provide an understanding of the reactor's inherent ability to survive even hypotheti-cai events.
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QCS760.37-1
Paga 1 (82-0563) (8,22] #134
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1 Ouestion CS760.123 What considerations have been given to thermal stresses under Natural Circulation conditions for the areas listed below:
(a) outlet nozzles (b) UlS (c) effect on component life i
(d) number of Natural Circulation transients assumed in component life calculation.
Response
(a) The outlet nozzle liners were designed to Code Case 1592 and the Natural Circulation transient has been evaluated. The outlet nozzles in the reactor vessel shell have been evaluated using the low temperature code which excludes the emergency events f rom the f atigue evaluation.
t However, for the Natural Circulation event the temperatures are above the low temperature code limits.
These Natural Circulation (emergency) transients will be evaluated using Code Case 1592 procedures which require the inclusion of emergency events in the creep-f atigue damage evaluation. Due to the short period of time and the small number of events the creep-fatigue damage is expected to be very small.
(b) Thermal stresses due to Natural Circulation in the UlS are considered as part of the 577 duty cycles, it is considered as an emergency event.
(c) The ef fect of Natural Circulation events on life for the outlet nozzles In the reactor vessel shell is expected to be very small. (See item (a) above).
For Uls components this Natural Circulation event is umbrellaed under more critical events. The specific damage summation for this event alone has not been separately assessed.
However, it is estimated to be less than 2% of the total danage.
(d) Refer to PSAR Appendix B General Plant Transient Data for the number of occurrences of events, designated as Emergency Events, that the UlS and outlet nozzles must accommodate.
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pego 4 W82-0586 [8,22] 141 Ouestion CS750.168 What aro the sito boundary No aerosol concentrations for the sodium spill events listed in Tablo 15.6-17 Resoonse:
The only event listed in Tablo 15.6-1 for which the site boundary concentra-tions of sodium reaction product aerosols rould be significant is the Intermediate Heat Transport System (IHTS) piping leak (PSAR Section 15.6.1.5).
The chemical form of concern f or environmontal protection is sodium hydroxide.
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The concentration of sodium hydroxido at the sito boundary can be calculated' as follows:
C=A.Xg.E
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d Whoro 0 = sodium oxido aerosol release from the SGB : 440 lb in 5 min. (1.47 lb/sec)
X/Q = atmospheric dispersal (50%) : 1.01 x 10-3 sec/m3 Fd = depiction f actor for f al lout and conversion to carbonate : 1/100 C = concentration of sodium hydroxide at the site boundary for the IHTS pipe leak = 6.7 mg/W The sito boundary concentration of sodium hydroxido resulting from the other events identiflod in Tablo 15.6-1 have been reviewed and would be orders of magnitudo below 1 hat for the IHTS piping leak.
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l QCS760.168-1 Amend. 70 Aug. 1982
Peg) 2 (82-0563) [8,22] #134 l
Ouestion CS760.169 s
Section 9.13.2.1 states that the catch pan system for air-filled cells is on Engineered Saf ety Feature. However, no inservice inspection requirements are listed olther here or in Section 3.8-C.
Please provide your plans in this area.
Resoonsq The Project is presently developing lospection/survelllance requirements for the Engineered Safoty Feature Catch Pan System. This information wilI be available for Staf f review by January 1983.
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QCS760.169-1 Amend. 70 Aug. 1982
Pagn 3 (C2-0563) [8,22] #134 Ouestion CS760.171 Section 3A.8.4 describos the development testing programs to support the coil linor design.
Please doscribe the dif ferences and similarities in the cell liners used in the sodiun spill design qualification test versus the linor design proposed f or CRBR. Address such items as plato material and thickness, wol d type, vent space and vent path size, linor design and support.
Resoonse The CRBRP coil liner tested in the Sodium Spij l Qualification Test was repro-sentative of the current CRBRP cell liner design.
A listing of the similari-ties and dif ferences between the test article lir.or design and the current proposed CR3RP design is provided below:
Similarities 1.
Liner Plate Material (SA516, Grado 55) 2.
Liner Plate Thickness (3/8")
3.
Liner Air Gap Size (1/4")
4.
Insulating Concreto Panel Naterials (Perlite-Sand-Coment) 5.
Insulating Concreto Thickness (= 4")
6.
Stud Anchor Material and Spacing (A108, Grade 1020); (15" o/c) 7.
Bottom Liner (Floor) Anchorage Material (SA30-Rolled Sections) l 8.
Coll Linor Woldment Configuration and Materials (FCAN - Prime Cell Liner Weldment) l (SMAW - Overhead Repair Weldment) l 9.
Structural Concrote Materials 10.
Coll Liner Configuration and Layout 11.
DBA Sodium Spill Peak Temperaturo (11000F)*
12.
Inerted Cell Atacsphere (N )
2 Differences 1.
Liner Floor Insulation (Mg0 Gravel)
The current CRBRP cell liner floor design utilizes precast insulating concrete panels consisting of the same material utilized in the coli liner wal l and ceil ing panels.
- Note: The test exceeded the DBA temperature by including long term exposure at 16000F (approximate TMBDB tempnrature).
QCS760.171-1 Amend. 70 Aug. 19 L2
FWgo 4 (UZ-0563) L8,22J #134 2.
Linor Panel Stud Anchorage Length (18 inchos)
The coil liners exposed to DBA sodium spill conditions utilizo linor stud anchors having a total longth of 12 inches. Main:
Cell liners exposed to temperatures >10000F, i.e., TNBDB, util ize stud anchors having a total longth of 24 inchos.
3.
Air Gap Formation Method (Ethaf oam)
The test article utilized a thermal plastic material, Ethafoam, to f orm the required 1/4 inch air gap.
The material was designed to shrink and/or vaporize under elevated temperatures and provide the 1/4 inch air gap space.
During the conduct of the test this material condonsed and plugged the opening.
The current CRBRP design elimi-natos the potential for plugging by providing a pre-formed air gap as described in Question CS760.166.
4.
Corner Anchorage The corner anchorago design utilized in the test article was based on the preliminary design.
The CR3RP corner anchorogo has been refined and increased in load capacity based on the final detailed design.
Non-prototypic anchorages were incorporated to anchor the non-con-tinuous (non-prototypic) free edges of the test article. These anchorages were specific to the limited test articlo size and are not utilized in the design.
5.
Stud Anchor Wolds As noted in Ref erence QCS760.171-1 (PSAR Reference 5 in Chaptor 3A) the test perf ormer installed the stud anchors using SMAW fillet wolds 013/16") contrary to the specified automatic full penetration weld process (ASME B&PV Code Section li t, Division 2, Subarticle CC-4540). Those weldments were Incapable of developing the stud I
capacity and resulted in several stud f ailures early in the testing.
The CRBRP cell liner studs will be Installed in accordance with the ASME B&PV Code Section ilI, Division 2, Subarticle CC-4540.
Reference QCS760.171-1 Simmons, L. D., Wierman, R. W.
HEDL-TME-79-35, "Large Scal e Liner Sodium Spil l Tost, LT-1", dated December 1980.
QCS760.171-2 Amend. 70 Aug. 1982
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,Qyentica CS760.172 Please provide simpliflad sketches of all sodium systems (exclusive of the primary and intermediate HTS) showing the sodium volumes in each major portion of the system and the location (cell number) of each system or portion of the system.
E_esponse:
'The simplif ied sketches and additional Information requested are provided in Figures CS760.172-1, 2, and 3.
Both sodium and NaK piping and components are included Inn these figures.
Sodium inventories in the components, and cell locations of these components are Indicated on these figures. All sodium inventories are referenced to a 400 F coolant temperature, except the primary scdium overf low vessel inventory which is also given for the primary system operating temperature of 850 F.
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page 6 W82-0586 [8,22] 141 Ouestion CS760.174 Available test data from the SRI international Scale Model Tests seem to prove the capability of the CRBR vessel as a whole to contain a 661 MJ CDA. However, we do not think the tests performed necessarily prove that the vessel head is capable of taking the simulated loads. From the hydrostatic test of the head, SM-1, it appears that the head f ailure mode would be disengagement of the intermediate Rotating Plug (IRD) from the Large Rotating Plug (LRP) due to bending deformation of the LRP. Because of the way the under-head shielding plates were attached to the head in the scale models, the bending deformation of the LRP was over constrained. In fact, the bending stiffness may have been at least an order of magnitude too high. Therefore, we believ.e that the tests performed to date cannot be directly used to confirm the head-capability or to benchmark analytical methods. The heads from the SM-4 and Shb5 test specimens experienced little, if any, plastic deformation during the previous test series. Therefore, they could be reused to perform additional hydrostatic tests, similar to test SM-1. The SM-4 head could be used as presently designed with the shielding plates in place. The SM-5 head and shielding plates could be modified to make it as prototypic as possible. Both these heads could then be hydrostatically tested giving a comparison between stif f nesses for the model head (SM-1), the model head with shielding plates attached as tested in SM-4 and ShE5, and the prototypic model head with shielding plates attached. These tests could, conceivably, provide the data needed to successfully use the results of the overall SRI international test program in confirming the vessel head design. However, at least one additional transient test may be required to prove the head capability for a 661 MJ CDA. The SM-6 specimen could be modified to mske the head prototypic and then be used for this purpose. This test would be useful in meeting the second overall objective of benchmarking analytical models before they are used to predict response to higher energy CDAs. The staf f and its consultants have performed an in-depth review of the simplified analytical model presented in Appendix H of SRI International's Technical Report No. 4. To derive the simplified model, an equivalent uniform plate was substituted f or a simplified, axisymmetric geometry of the three-plug vessel head. Dynamic response is calculated assuming the plate responds in the first vibrational mode. Therefore, the model applies only to response in the linear range. it predicts that the head would experience a peak displacement of 0.102 in. during a 661 MJ CDA. No test data exists with which we can directly benchmark this model. Analytically it can be modified by eliminating the ef fects of the shiciding plates and then comparing its static response with the SM-1 test. This exercise shows that the model underpredicts peak displacement by f actor of 2.4. Comparison of the volume-pressure relationship for the SM-1 test with a modification of the vol ume-pressure relationship (modif ied to el iminate shiel ding plate stif f ness) QCS760.174-1 Amend. 70 ( August 1982
page 7 W62-0586 [8,22] 141 e used in developing this model provides additional evidence that the model is 2.2-3.2 times too stiff, if its predicted displacement is directly adjusted by a f actor of 2.5, we predict a maximum displacement of 0.25 in., which occurs at e static pressure of 1100 psi for SM-1. This is just under the failure pressure of 1160 psi. This assumes linear response. If the response goes into the nonlinear range, which it will for these high displacements, the response will be even larger. One should be cautious not to draw too many conclusions f rom such a simpl if ied model. However, the points made indicate the reasons for the dif ficulty in accepting either the experimental on analytical evidence now in hand as proof that the vessel head can withstand a 661 MJ CDA (112 MJ slug impact). Also, a suitable test program may provide enough inf ormation to meet NRC's objectives of estimating the ultimate capability of the head.
Response
The applicant has initiated the additional testing. (Tests SM-7 ad SM-8) QCS760.174-2 Amend. 70 August 1982
N .l PSAR RESPONSE APPROVAL FORM Date: 0-/(s-h Response to NRC Question (s) h[a O. 2 3 Reference Letter Number The above NRC Question Response (s) have been reviewed and dispositioned as follows: /] Approved W1.%ot Changes: (list response number) Approved Subject to Incorporation of Changes on Attached Copy: (list response number) ,/~T Disapproved with Comments: (list response number) ,t ~ /7' Approved Subject to Completion of Action Items (list response number) t Comments: Approval: l J l)/8 Z- \\ V date Concurrence: 4*&// ?!/M/hI2-7/11!JZ // 'VCr41 7//6[G h. E 'date LRM d' ate P0' Concurring date Divisions Distribution: P0 Public Safety Division (Action) LRM (Information) AE P0 Lead Division m-- -m.w1. w
se PSAR RESPONSE APPROVAL FORM Date; 3 2 / Response to NRC Question (s) O.37 Reference Lctter Number The above NRC Question Response (s) have been reviewed and dispositioned as follows: 6 Approved Without Changes: (list response number) Approved Subject to Incorporation of Changes on Attached Copy: (list response number) . G Disapproved with Comments: (list response number) u 6 Approved Subject to Completion of Action Items (list response number) Coments: i 1 Approval: lv 7/$2 b date Concurrence: 2[ll/[72- . / LA 6hc/e> bt&7HMar/s2 }). AAAm dele E dhth LRM date P0/C5ncurring dat6 Divisions '7/9/W t ' ss l Distribution: P0'Public Safety Division (Action) LRM (Information) i AE P0 Lead Division 4
g[C m. _.. _ - ~ ~ ~ ~ ' ~ ~ ~ ~ ' ~ ~ ~ ~ ~~ ~ ~ ~ ~ ' ' .s ./ I PSAR RESPONSE APPROVAL FORM CL. (,,fe?5,fP2_. m, Date: M h 11 Response to NRC Question (s) k O. k Y [i+l Reference Letter Number n I The above NRC Question Response (s) have been reviewed and dispositioned as t) follows: 1 i i ) /7 Approved Without Changes: (list response number) Approved Subject to Incorporation of Chan es on Attached Copy: (list response number , /7 Disapproved with Comments: (list response number) i -/ T' Approved Subject to Completion of Action Items (list response number) Comments: Approval: 1 $ tf (s ?_. \\ date' Concurrence: ~ YkA h h f / $ 2 h l h k Y N W B Z-lb, k (k 7ffvfft. 7 AE ~ 'da te LRM date P0 Concurring ~ da t'e J Divisions Distribution: P0 Public Safety Division (Action) LRM (Information) AE P0 lead Division _ _ _ _ _ _ _ _ _ - - - - - - - - - - - - ~ - - - - - - - - - ~'
r PSAR RESPONSE APPROVAL FORM e Date: __ '/ d / - E Response to NRC Question (s) O./ d Reference Letter Number The above NRC Question Response (s) have been reviewed and dispositioned as follows: Approved Without Changes: /\\ (list response number) L7 Approved Subject to Incorporation of Changes on Attached Copy: (list response number) . LT Disapproved with Comments: , (list response number) i ~~ -/ T Approved subject to Completion of Action Items (list response number) Comments: Approval: d Y2ffL l ~- l" date Concurrence: f k m / L 7/2// e 7/22/92.h.YYnEM Y2= 5't AE 'date LRM date P0 Concurring date Divisions Distribution: P0'Public Safety Division (Action) LRM(Information) l l AE P0 lead Division
/ PSARRESPONSEAPPRdVALFORM Date: 7 Mb Response to NRC Question (s) O. /@ Reference Letter Number _ The above NRC Question Response (s) have been reviewed and dispositior)ed as 4 follows: ~/ Approved Without' Changes: ~ (list response number) ~ /_T Approved Subject to Incorporation of Changes on Attached Copy: (list response number) . /7 Disapproved with Comments: ~ (list response number) ~ -/7" Approved Subject to Completion of Action Items (list response number) i Commen ts: Approval: // d 7 /6/ E date Concurrence: I Y!e hE ' ~ Wf/6 L . YYbhoY WAlS2 {AY~(k h<fm 99 date LRM date P0 Concbrring date Divisions l-Distribution: P0 Public Safety Division (Action) LRM (Information) AE P0 lead Division ,,,.r'- --.w,yr. -w y ,m +, 7,----g
=- ' ~ / PSAR RESPONSE APPROVAL FORM Date: 7di L Response to NRC Question (s) _ aO,/7/ Reference Letter Number The above NRC Question Response (s) have been reviewed and dispositioned as folicws: /7 Approved Without Changes: (list response nurr.ber) Approved Subject to Incorporation of Changes on Attached Copy: /4 (list response number) , / 7 Disapproved with Co:rEents: ~ (list response number) t -/7' Approved Subject to Completion of Action Items (list response number) Comments: Approval: (: r E /(0 hh date Concurrence: k N & h k J A.7 he SN c7f 7lN /8 2 _/ S ybt ' d' ate LRM date PO Con ~ curring 'da'te i Divisions Distribution: P0 Public Safety Division (Action) LRM (Information) AE P0 Lead Divisicn w r-
T PSAR RESPONSE APPROVAL FORM 7 /A /. d Date: af Response to NRC Question (s) bO. /72 Reference Letter Number The above flRC Question Response (s) have been reviewed and dispositioned as follows: / Approved Without Changes: (list response number) /_7 Approved Subject to Incorporation of Changes on Attached Cop [: (list response number) . /_7 Disapproved with Comments: (list response number) L7 Approved Subject to Completion of Action Items (list response number) Comments: Approval: & Wz1,/91. V date Concurrence:
- f. ks&sA~- YA//f2-l a Y 1hLh2 th. bc Yk1%
h AE / date LRM date P0 Concurring date Divisions Distribution: P0 ~Public Safety Division (Action) LRM (Information) AE P0 lead Division
t I PSAR RESPONSE APPROVAL FORM Date: 7 / /f,t. / Response to NRC Question (s) ((o[). /'-7h Reference Letter Humber The above NRC Question Response (s) have been reviewed and dispositioned as - follows: -/7 Approved Without Changes: list response number) / Approved Subject to Incorporation of Changes on Attached Copy: (itst response number) , /7 Disapproved with Comments: (list response number) b ~ /7 Approved Subject to Completion of Action Items (list response number) Comments: Approval: ~ ff7,lTL V date Concurrence: ..l. b aAL 2 /k[2-TfU k b dL {A'E ' / /date LRM date P0 Concurring date Divisions' Distribution: P0 Public Safety Division (Action) LRM (Information) ~ AE P0 lead Division j -}}