ML19319D716
| ML19319D716 | |
| Person / Time | |
|---|---|
| Site: | Crystal River, 05000303  |
| Issue date: | 08/10/1967 |
| From: | FLORIDA POWER CORP. |
| To: | |
| References | |
| NUDOCS 8003240722 | |
| Download: ML19319D716 (200) | |
Text
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RESPONSE TO LETTER s
P EER A. MORRIS, DIRECTOR, DIVISION OF REACTOR LICENSING, FLORIDA POWER CORPORATION TO MR. J. T. RODGERS, DATED JA'IUARY 19, 1968 "Specifically, the site description does not contain the results of the consolidation grouting program, essential to the evaluation of foundation vorthiness and determination of structural adequacy to seismic disturoances.
The analysis of site response to the maximum probable hurricane is incomplete in the areas of wave run-up (model description), sea-water drawdown (mini-mum vater level), inlet structure details, and service water pump locations."
The results of the consolidation grouting program, essential to the evaluation of foundation worthiness and determination of structural adequacy to seismic disturbancea have been com-piled and are in a report which will become part of the PSAR Appendix.
The analysis of site response to the hurricane effects, wave run-up, etc. was included in the " Plant Protection Against p
Hurricane Wave Action" report which is part of Appendix 20, V
" Flood Studies and Hurricane Effects" as revised with Amend-ment No. 1 on January 15, 1968.
"The reactor description is deficient in the areas of in-core detectors, core-barrel check valves, the primary pump anti-reverse rotation device, and core design."
Additional information in the area of in-core detectors is provided in Supplement No.1, Questions 1.1, 5.2, and 5.11.
The Preliminary Safety Analysis Report has been up-dated to provide additional information tu the core barrel check valves.
Refer to Section 3.1.2.h, Page 3-3; Section 3.2.h, Pages 3-68, 3-72, 3-72a, 3-72b ; Section 3.3.h, Pages 3-103, 3-103a, 3-103b; Figure 3-61-a; Supplement No.1, Question 2.1; Additional information is provided on the preliminary pump anti-rotation device in Supplement No.1 Questions 2.2 and 8.h.
/mU 0323 '
The Preliminary Safety Analysis Report has been up-dated to provide additional information on the core design. Eefer to Section 3.2, Pages 3-9 through 3-14, 3-31; Figures 3-1, h
3-2, 3-5, 3-6, 3-39, 3 ho, 3-57, 3-58; Section 7.22, Pages 7-18, 7-19; and Section 1h.1.2, Page lk-3 through 1h-7, lk-10, lb-23,1k-25,14 41.
"Your design does not meet, in some areas, the recertly published AEC Supplemental Criteria for ASME-III Vessels. We need additional in-fomation on the significance of the criteria with respect to your design and construction. Justification for lack of full compliance should be presented where applicable."
The reactor coolant system components are decigned and classified in accordance with appropriate existing codes as listed in Section h.1 and Table h-9 of the PSAR. The auxiliary system components are designed to applicable existing codes and standards and classified as shown in Section 9 of the PSAR.
The AEC's Tentativt: Regulatory Supplementary Criteria for ASME Code-Constructed Nuclear Pressure Vessels published August 23, 1967, have been reviewed. These proposed criteria have also been reviewed by an ad hoc group of industry repressentatives of the Atomic Industrial Forum. We concur in the results of the evaluation by the AIF ad hoe group as presented in a letter of December 15, 1967, from Mr. Harvey Brush, Chairman, AIF Reactor Safety Steering Committee, to Mr. Harold L. Price, Director of Regulation, USAEC.
h The AIF letter recognizes that certain of the supple-mentary pressure vessel criteria are truly licensing type criteria and are potentially useful to both the AEC Regulatory Staff and the Huclear Industry.
On the other hand, the AIF points out a concern over the fact that some of the proposed criteria are addressed to code type requirements and some to engineering procurement and design specification type requirements. We concur with the AIF that reco=nendation for modification of code requirements should be referred to the copropriate code writing bodies; in this case, the ASME Code Committee for Boiler and Pressure Vessel Code Section III, " Rules for Construction of Nuclear Vessels."
We understand that a comprehensive review of these criteria is underway and vill be submitted to the AEC in the near future by the ASME. We believe this is the most appropriate manner of providing constructive O
comments on code related criteria, since this is an industry wide consideration. We also believe that the engineering
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procurement and design specification type requirements have
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no place in the proposed criteria.
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The results of our systematic review of the proposed supple-mentary criteria are tabulated in Table 1, " Evaluation of AEC's Tentative Regulatory Supplementary Criteria for ASME Code-Constructed Nuclear Pressure Vessels." This Table lists each of the proposed criteria in accordance with the assigned paragraph numbers. We have designated whether or not our design complies completely with the proposed criteria as we interpret them.
In some cases, the lack of compliance with the criteria involves only minor deviations from the stated guidelines.
i We believe further clarifying interpretations will be re-quired before a more complete assessment can be made regard-ing those items in the column designated "Ho" in Table 1.
l TABLE I EVALUATION OF AEC'S TENTATIVE REGULATORY SUPPLEMENT.;IY CRITERIA FOR ASME CODE-CONSTRUCTED UUCLEAR PRESSURE VESSELS O
Criteria Paragraph Incorporated in FPC Design Yes No.
1.10 Classification of Huclear Vessels Note (1) 1.11 Conditions for Design X
1.12 Certification of Stress Reporta X
1.13 Conditions with Unspecified Design Rules X
1.1h Vessel Owners Responsibility for Inspection X
1.15 Manufacturer's Responsibility for Quality Assurance X
1.16 Vessel Fabrication Report X
0325 J +
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Criteria Paracraph Incoreorated in FPC Desien Yes No.
1.17 Boundary Between Vessel and Piping X
1.20 Vessel Material Property Improvemeat X
1.21 Material Test Coupons X
1.22 Non-destructive Examination of Peactor Vessel Plates X
1.23 Non-destructive Examination and Re-pairs of Material X
1.2h Examination of Reactor Vessel Bolts X
1.25 Ductile Brittle Transition Properties X
1.26 Exclusion of Repairs in Bolting Material X 1.30 Fracture Mechanics Analysis X
1.31 Design for Cyclic Loading X
1.32 Bolting Design Requirements X
1.33 Earthquake Loading X
1.3h Design Conditions - Combinations of Loaddng X
1.35 Computer Programs X
1.36 Environmental Effects X
1.37 Design for Inspectability X
1 38 Attachments to Reactor Vessels X
1.39 Beactor Vessel Core Support X
1.h0 Chemical Analysis of Weld Wire X
1.kl Cutting Plates and Other Products X
1.42 Welding Qualification Procedure Requirements X
t' 0s26
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Criteria Paragraph Incorporated in FPC Design Yes ilo,
1.43 Precautions for Welding X
1.hk Welding Requirements X
1.50 Final Inspection and Examination X
1.51 Non-destructive Examination and Responsibilities X
1.60 Hydrostatic Testing Requirements X
1.70 Approval of New Material for ASME Code-Constructed Nuclear Vessel X
Note (1) Full compliance except for classification of letdown coolers "The containment design describes grouted tendons; however, in a recent meeting
(
you have discussed a different design.
In your clarification of this point, b
as well as for other applicable matters, you may, if you desire, reference other applications."
The Preliminary Safety Analysis Report has been updated to include the required information on the unbonded tendon system design. Refer to Section 5 and Appendix 5B.
"We understand that your present emergency core cooling system design departs substantially frcm that now described in your PSAR. Appropriate updating is required."
The emergency core cooling system design has been updated, and the revised description is included in Amendment No. 2 to the Preliminary Safety Analysis Report.
0327 g i
"The instrumentation and control system is different from previous cases, since it does not provide any direct method for measuring primary flow, either absolutely or relative to a nominal value. Justification of this h
lessening in plant protection, relative to the Dockets 50-269/270/287/289 (Oconee and Three Mile Island Unit), vill reouire submittal of detailed design information. Based on the present information, we cannot conclude that there is adequate safety instrumentation for protection against certain loss-of-flow accidents. Specific questions are provided in the attachment to give guidance as to the type of information needed."
The primary system flow monitoring provides operationt_1 characteristics different from the above referenced plants in that our system does not prevent a reactor trip on the loss of one pump if the reactor power level is greater than 75 percent of rated power. This represents an acceptable reduction in ability to keep the plant on the line, but we do not believe our design represents a lessening in plant protection relative to Dockets 50-269/270/287/289.
Ad-ditional detailed information is provided in Supplement No.
1, Questions 5.1, 5.4, and 8.h.
"This description of the electrical system (Chapter 8) is not adequate because the description of the engineered safety features load distribu-tion is lacking, the description of the off-site power connection to the emergency busses is inadequate, and justification of automatic diesel cross-connection was not provided."
The Electrical System (Section 8) of this Preliminary Safety Analysis Report has been updated to include the additional information required.
' Only one emergency feedvater pump is provided. This is not considered acceptable, because the single-failure criterion is not met."
The Preliminary Safety Analysis Report has been updated to include one motor-driven emergency feedvater pump and one steam-driven emergency feedvater pump.
(Refer to Section 10).
" Insufficient details have been included in your application in regard to justifying the turbine stop valves as steam line isolation valves."
The construction of the seating arrangements for the four main steam stop valves used on the turbine-generator unit is identical to that used in the reheat stop valve of fossil type i-G units.
There are approximately 200 T-G units of this type in service since 1950. On each T-G unit there is a minimum of two stop valves thus there are at least h00 valves in actual use.
O
The pertinent facts about these valves are:
ew 1.
The primary purpose of these valves is to protect the T-G unit against excessive over-speed. To insure each valve vill function properly under emergency conditions, it is tested every two weeks. During this test the high pressure fluid is dumped from the hydraulic valve servo-motor actuator and the valve closes due to spring force.
2.
The valves must be tight and must remain tight for years.
This is necessary to ensure that the steam flow is stopped thus permitting the unit to coast down to stand-still. These valves have been used to perform a hydrostatic test on the reheat sections of boilers. Such tests have demonstrated that the valves are tight during initial installation and that they remain tight over the years - this capability has been demonstrated by hydrostatic testing after boiler repair.
3.
The valves are subjected continuously to 1000/1050 F normal operating conditions. With regular testing and routine maintenance, valve stem freedom is assured.
h.
The valve seats and discs are " blued" at the factory to obtain 100 percent contact.
5.
Hydrostatic testing has been performed in the factory on a representative number of valves and those observed have O
exhibited zero leakage.
6.
The valves are of the unbalanced type thus the greater the pressure differential against the valves the greater the force seating the valve.
7.
The valve closes in approximately 0.15 seconds.
8.
The valve opens by hydraulic force and closes by spring force.
9.
These valves are of the highest quality and have demonstrated excellent performance over a period of fifteen years, i
l The main stop valves for the FPC turbine are: of the unbalanced i
design, opened by hydraulic force, closed by spring action, and have a closing time of approximately 0.15 seconds. Furthermore valve seats and discs will be " blued" in the factory to assure 100 percent contact.
Although a steam line stop valve is very reliable, the results
' of a possible failure to close following a steam line rupture have been analyzed.
/~3 LP 0329..
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The simultaneous blevdet-a double ended rurture ::
generator or jeor rd :e system.
The steam renerr stresses are well bele respcnse to each stec 6.3 h of ouestion 6.
50-270.
(Oconee)
The reactor coolant v;_.
510 F.
Assuming a stuck in the core at the time c:
xenon - 1.h percent Zih/i the tripped rod worth ava; and temperature reductier.
minimum moderator coeffic end of core life and conr:
the reactor vill not co er greater than 390 F.
If ti down just prior to the a rod worth for power and >
Zik/k. "nder these cond; about kh0 F vill be remu:
Following the transicr-operator to maintain the decay heat level rr to the steam generat:-
feedvater pu=p unt ;
Removal System plc-In summary, the -
vill result in tn if all the stear coolant systen be cooled suff1.
'Ve note that no radiat.
vaste discharge lines.
inadvertent release of rc Radiation interlocr..
vaste discharge line:
'The safety analysis sectic:
analysis, fuel handling ine a spectrum-of-break analyses.
utable to obsolete descrirt;-
sutply systems.
9
Additional information on the loss of flow accident is presented in Supplement No. 1, Question 8.h.
O-Additional analysis of a fuel handling incident is pre-sented in Supplement No. 1, Question 8.10.
Additional analysis of potential steam generator failures is presented in Supplement No. 1, Questions 8.6 and 8.7.
The Preliminary Safety Analysis Report has been updated to provide additional information on the spectrum-of-break analyses. Refer to Section 14.2.2.3, Pages lk-kla, through lk-h2; Figures lb-kha. Ih-khb, 14-khe, lk-khd, lk-khe, lk-hkf: and Supplement No. 1, Questions 8.5, 8.6, 8.7, and 8.9.
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Dockets 50-302 and -303 Supplement No. 1 February 7, 1968 O
Q QUESTION Describe the extent and manner in which the in-core detector 1.1 assembly vill be used as a tool in determining maximum-to-average ratios during plant operation. Provide the basis for continued plant operation in the event of in-core in-strument malfunctions.
ANSWER The in-core instrumentation system is not connected to the reactor protection system or the reactor control.
The system is provided primarily to collect data for effective administration of the fuel management program and secondarily to provide the operator with confirming information regarding power distribution in the reactor core.
During normal operation the in-core instrumentation is not needed to provide the operator with any information on which he must take corrective control action, because core power distribution should follow previously calculated values.
The incore instrumentation, however, is expected to alert the reactor cperator whenever xenon oscillations exist, because it is only during xenon oscillation that undesirable maximum-to-average power conditions can occur.
Xenon oscillations, however, can only occur at higher power levels under predictable circumstances, which lend themselves to analytical
(_j determinacion. The required analyses vill be performed during the design of the reactor and a xenon oscillation threshold power versus core life curve vill be developed. However, there is some power level belov which significant xenon oscillations can never occur.
In the event that the plant computer which provides the incore r?ad-out is not operating, the alternate system of in-core readout of el independently monitors enough of the system to observe xenon osu__.a-tions will be placed in service.
In the event neither readout system is in operation, the tech specs vill require the operator to reduce power to a level at which xenon oscillations can never occur.
~0332 oV 1.1-1
Docksts 50-302 and -303 Supplement No. 1 February 7, 1968 QUESTION Ictentify those items that vill eventually be classified as 1.2 technical specifications that now affect plant design.
ANSWER The plant design is influenced by technical design criteria, specifically as indicated below.
1.
Power Level.
The unit vill be capable of an ultimate capacity of 2560 5
MWt, (including 16 MWt from the reactor coolant pumps) corresponding to an 885 MWe gross electrical. capability.
The maximum power level for which authorization is applied is conservatively set at 2h52 MWt, corresponding to a gross electrical output of 855 MWe.
2.
Primary System Activity.
Unit design criteria have been established assuming continuous unit operation with 1 percent failed fuel.
3.
Leak Rate.
The unit design vill be implemented utilizing a Reactor Building l
1eak rate of 0.25 percent per day leak at design pressure and temperature for the reactor containment building.
4.
Integrated Neutron Flux.
The reactor vessel is the only reactor coolant system com-ponent exposed to a significant level of neutron irradiation.
The potential radiation at the end of reactor service life from fast neutrons (E 1.0 Mev) has been computed to be a 1
maximum of 3.0 x 10 9 n/cm2 over a 40 year life with a unit l
capacity factor of 80 parcent.
5 Nil-Ductility Temperatura.
F6r the neutron exposure expressed in No. k above, the predicted Nil-Ductility Transition Temperature (NDTT) shift opposite the core is not greater than 250 F, based on an initial NITIT of 10 F and an end of service life maximum NDTT of 260 F.
6.
Minimum Operating Conditions.
The unit vill be designed with no minimum operating condition and will be specifically designed to operate satisfactorily carrying only its own auxiliary load. In the operating range of 0-15 percent full load power, the nuclear reactor system vill be controlled in the remote-manual mode of control. In the 15-100 percent full load power, normal operation vill be by the automatic control mode.
0 133 1.2-1 (Revised h-8-68)
7.
Equipment Redandancy.
The unit design criteria has been established to assure unit operating reliability and safe shutdown capability under all conditions utilizing redundan. aquipment as necessary to meet this criteria.
8.
Core Flooding Tax.
The unit design incorporates two Emergency Core Floeding Tanks.
Lack of availability of either tank vill require immediate shut-down of the reactor system, and therefore the unit.
9 Meteorology.
The unit vill be designed to operate continuously and to be safely shutdown unrestricted by meteorological conditions.
Continuous monitoring of meteorological conditions and use of area radiation monitoring devices (with alarms) vill confirm acceptable site release rates are being met for any discharge of liquid and/or gaseous vastes.
10.
The unit design incorporates two quick-starting diesel-generator units presently estimated at 2850 KW rating each, each connected to a safeguard bus. Any one of the two emergency generators and its associated safeguards bus provides adequate capability for safe shutdown.
O 11.
Off Site and On Site Auxiliary Power.
The plant design provides as the normal source auxiliary power 5
for the generator to feed' its ovn' auxiliary load. Upon a trip separating the substation from the transmission system, the turbine generator is designed te stay in operation upon a load dump from full load to auxiliary load. Normal engineered safeguards power shall be supplied from the 230 KV substation with automatic transfer to the engineered safeguards diesel-generators upon loss of 230 KV power. Transfer to a unit aux-iliary transformer source can be accomplished manually. There are four 230 EV transmission lines, two to Curlev and two to Central Florida to provide back-up off site power. Additional on site power is provided by either or both of units 1 and 2 at this site.
12.
Personnel.
The Production Department of Florida Power Corporation is responsible for staffing the plant to assure safe, reliable and efficient. operation. Centralized control over,the plant operation rests with the Production Superintendent whose offices are in the corporate headquarters in St. Petersburg, Florida.
The Plant Superintendent is responsible to the Production Superintendent for the operation of the plant.
1.2-2 (Revised h-8-68)
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The plant will be manned with an estimated staff of 59 full-v time employees functioning in four main groups: Supervision, Operation, Maintenance and Technical Support. Each position of this staff vill be filled with men who have had extensive operating and Tsintenance experience in fossil fueled plants and who will be given additional specific nuclear education and training.
A detailed description of personnel requirements, nuclear training and licensing, and administrative control of plant operations and personnel is outlined in Section 12 of the PSAR.
13.
Core Cooling Systems.
The units are designed incorporating a reactor coolant system as the primary and normal source of core cooling. This system is designed to contain and circulate reactor coolant at pressures and flows necessary to transfer the heat generated in the reactor core to the secondary fluid in the steam generators.
In addition to serving as a heat transrsrc medium, the coolant also serves as a neutron moderate: and reflector, and as a solvent for the soluble poison utilized in chemical shim reactivity control.
The reactor coolant system is designed to maintain its integrity under all operating conditions. This function serves the safe-guard objective of containment of fission products that escape
)
the primary barrier, the core cladding.
s In addition to the reactor coolant system, there are three emergency core cooling systems which are activiated in sequence on failure of the reactor coolant system. These are:
A.
High pressure injection, B.
Core flooding tanks and C.
Low pressure injection.
- 14. Fuel Burnup'.
The reactor has been designed for an average burnup of 28,200 MWD /MTU and for a maximum burnup of 55,000 MWD /MTr,. Design thermal output is 19.9 KW/FT at maximum overpovel (llh percent of rated pover).
15 Environmental Consideration.
The plant site of h,738 acres is owned by FPC and is character-ized by:
A.
A hh00 foot minimum exclusion radius; B.
isolation from nearby population centers, the closest population center of 25,000 or more is Gainesville, Florida; C.
sound rock foun-dation for etructures; D.
adequate supply of cooling water, Gulf of Mexico; E.
favorable hydrology, the plant site is seaward of any potable water -upply; F.
location is in a rel-atively aseismic zone, stru,; ural design is based on a horizontal acceleration of 0.05 gravity, conservatively; and desirable ti.
meterology, diffusion of vaste gases in the atmosphere is good.
V d)33>b 1,2-3
Docksts 50-302 and -303 Supplement No. 1 February 7,1968
()
QUESTION Describe how your design complies with General Design Criterion 1.3 No. 11.
ANSWER The design of the control rocm and all equipment and furnishings used in the centrol room vill utilize non-combustible and fire resistant saterial where ever practical.
There are four entry points into the control room, insuring separation of entry locations. Control room shielding and ventilation are designed to maintain acceptable radiation levels during the design basis accident. There vill be fire and smoke detecting instrumenta-tion, and portable fire extinguishing equipment, and breathing apparatus located in the control room for use in extinguishing any fire. The portable oxygen breathing apparatus vill be self-contained similar to Mine Safety Appliance Company Chemox apparatus, which uses canisters each good for approximately one hour. The face mask covers the face and eyes and has built-in alarms to alert the verf-r af the depletion of his oxygen supply so that he may exchange canisters. Fire extinguishing equipment and breathing apparatus will also be located immediately outside of the control room.
While control room ata.19r. ment is considered to be incredible in light of the design criteria for the control room to prevent fire, smoke, or radiation hazards, the follow ing describes the operation that would take place in the event the control room were abandoned and denied access:
()
Actions required br the operator before he abandons the control room:
1.
Actuates the reactor trip button which causes all control rods to insert, and trips the turbine generators.
2.
Assuming that tripping the nuclear units results in loss of off-site power, then the following actions will occur or must be per-formed:
A.
The emergency turbine driven feedvater pump will automatically start.
B.
The main turbine driven feedwater pump turbines will trip.
(Inter-locked with condensate and booster pumps.)
C.
The reactor coolant pumps vill trip.
(They vill automatically trip on loss of off-site power.)
D.
The make-up pump and high pressure injection pumps vill trip.
(They will automatically trip on loss of off-site power.)
E.
The operator will close the letdown valve.
F.
The operator vill close the seal return valve.
G.
The operator vill close the reactor coolant pump' seal injection (Q_/
vater valve.
1.3-1 0336
l In order to bring the reactor to a safe shutdown condition from outside the control room, the following ecuitment vill be provided:
1.
A control station remote to the control roo= will provide the following information:
A.
Steam generator level.
B.
Feedvater pressure.
C.
Steam pressure.
D.
Reactor pressurizer level.
E.
Reactor outlet temperature.
F.
Primary system pressure.
G.
Plant communciation system (phones).
2.
Local controls external to the control room vill be provided to permit the operator to perform the following functions:
A.
Operation of the emergency feed pumps with controls available at the pump.
B.
Operation of the feedvater start-up valves by use of handjacks h
at the valve.
(The start-up valves vill automatically control the feedvater to each steam generator to maintain the required level for shutdown operation, and the handjack would not be required except in the event of malfunction'of the automatic feedvater control system.)
C.
Operation of the makeup pumps to maintain pressuriter level.
(These pumps must be lined up with the borated water storage tank for this operation.)
D.
Operation of make-up valves by use of handjacks at the valve.
3.
Ability to remove excess steam from the steam generators.
(This is automatic by action of the main steam safety valve and/or power operated relief valves.)
- h.. Emergency power vill be supplied to essential equipment from the emergency diesel generatore, which vill start automatically.
It is anticipated that the operations in the control room can be executed individually within one or two minutes, or could con-
- -- -.c'eivably be operated from a single emergency switch.
1.3-2
p/s The operation of supplying emergency feedvater to the steam generators 8.n the event of loss of off-site power can be manually x_
controlled external to the control room if required and the reactor can be brought to and maintained ir a safe shutdown condition.
It is anticipated that off-site power vould be restored and that the conditions which required abandoning the control room vill be remedied within a short period of time. The control room complex vill be reoccupied within a short period of time and control of necessary unit functions vill be restored to this center.
The reactor vill initially be brought to and maintained in a hot standby condition until a decision is necessarily made as to whether the unit can be restored to service, maintained in a hot standby condition (with off-site power restored), or must be brought down to a cold shutdown condition.
In the event the unit must be brought to cold shutdown without operator access to the control room, the required instrumentation and control devices will be provided external to the control room to permit safe cold shutdown operation.
i 0338 l
AL}
1.3-3 (Revised 3-1-68)
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Dockets 50-302 and - 303 Supplement No. 1 February 7,1968
(]/ -
QUESTION - Update the description cf your research and development status.
N-1.4 ANSWER The Babcock & Wilecx Ccmpany's FWR research and development program and status have been described in detail and updated in answer to Questions 4.h, 5.13,16.h, and 17.h of the Metropolitan Edison PSAR, Docket 50-289, supplements 1 and 3.
In addition a study of plant failure mechanism associated with a LOCA is presently underway. This study has included identification of the potential failure mechanisms, a search cf the literature to obtain applicable data, evaluation and applicatien of existing data, and scoping tests tc obtain data cn potential failure mechanisms.
The initial results of this study include the identification of the failure mechanisms, a discussion of the infcrmation available in the literature ccncerning these mechanisms, and the effects of these mechanisms on the reactor system design.
Clad Failure Mechanisms The objective of this study is to insure that there are no potential failure mechanisms that might interfere with the ability cf the emer-gency core cooling systems to terminate the core te=perature trans-ient and remove decay heat in the event of a less-of -coolant accident.
Figure 1.4-1 shows the potential failure mechanisms that must be con-sidered in evaluating the loss-of-coolant accident. These include
{^/]
. clad melting, zircenium-water reaction, eutectic formation between the zircaloy clad and the stainless steel spacer grids, the possibility x-of clad embrittlement as a result of the quenching during core flooding, and clad perforation or deformation acecmpanying its failure. Figure 1.h-1 also shows the normal operating temperatere, the peak calculated temperature for the less-of-coolant accident, and the limit that has been established for the peak tesperature at this time.
Clad Melting and Metal-Water Reaction The potential interference with core ecoling as the result cf clad melting is expected to be the ultimate failure mechanism that estab-lishes the limit en maximum allevable temperatures following a loss-of-coolant accident. At this time, hcwever, there are little data available en the characteristics and effects of extensive clad melt-ing. As a result, the present limit has been established well below this temperc.ure limitation.
Considerable experimental and analytical evidence is available in the literature concerning the oxidation reacticn between zirconium and steam or water. (1,2,3,h,5 ) The 6a a available are sufficient to pre-i dict the amount of additional heat generatien that would cecur in the core as a result of a metal-water reaction (olleving a loss-of-coolant accident. However, additional verk will be required to determine l
A l
\\~ I OB9 1.L-1 (Revised 2-7-68)
the ability of the ECCS to remove the excess heat from large amounts of metal-vater reaction. Accordingly, the present preliminary limit on metal-water reaction in the core is 1%.
Analysis has indicated that only 0.15 per cent of the cladding vill have reacted with a maxi-h mum core temperature of 2,300 F.
Above this temperature the amount of reaction begins to increase more rapidly. As a result, a limit of 2,300 F cn the peak allowable clad temperature in the core 'ollowing the loss-of-coolant accident has been established in the design.
Eutectic Formation Between Dissimilar Core Materials The hot spot clad temperature vill be in excess of 1,700 F for a pe-riod of approximately 30 seconds for the maximum hypothetical loss-of-coolant accident. A possible mechanism of clad failure during this period.is the 'ormation of a lov melting eutectic between the zirconium-bared 0;c * * +erial and the iron and nickel-based spacer grid material.
c onium-iron eutectic forms at temperatures above 1,710 F, v_
itconium-nickel eutectic forms at tempera-tures above 1,760 tsible formation of this eutectic during a loss-of-coolant accau.
has been investigated to insure that it vill not lead to a condition that cou11 prevent emergency core cool-ing.
Diffusion bounding studies of the joining of Zircaloy to both stain-less steels and to inconel ere reported in the literature (6,7,8,9) and demonstrate that the formation of a eutectic between these solid materials is a diffusion-controlled mechanism and is thus time-de-pendent as well as temperature dependent. For example, iron-zirco-nium couples exhibited only slight melting after one hour at 1,800 F but exhibited good bonding after one hour at 1,882 F.
These slov ll diffusion rates were attributed to an oxidation layer on the iron.
In other tests at temperatures above 1,850 F, bonding between zirco-nium and iron required anywhere from several minutes to more than
'one hour. Application of this work to a reactor core vould indicate that the eutectic reaction may be expected between the cladding and the ferrules, but would be a slow reaction and would be limited to the small portion of the spacer grid material in intimate contact with the Zircaloy cladding.
Once the temperature of the materials in contact is reduced below the eutectic temperature, the reaction is expected to terminate immediately.
Some specimens were tested to confirm the conclusion that this method of failure is no problem in the ccre following a loss-of-coolant ac-cident. Six sections of Zircaloy clad material vere heated to tem-peratures between 2,100 F and 2,300 F vith sections of grids attached to simulate a portion of the fuel assenbly. The samples were heated above the eutectic temperatures for approximately 15 seconds. This test time conforms with the 15 seconds or less that the spacer grid material vill exceed 1,700 F at the het spot during a loss-of-coolant accident.
In some of th'e test' specimens, 'eutectic reaction occurred ~at the ~~ on-O c
tact between the spacer grid tips and the clad. This resulted in a 12 cal reaction which terminated as the two materials melted at the k}-}b'b 1.L-2 (Revised 2-7-65) p/$b
x point of contact. Although the di=ple in the ferrule reacted, none
(s)
- of the ferrule proper reacted. Both the clad material and the grid material maintained their structural integrity. The eutectic reactiin progressed to a much lesser degree for specimens containing no internal pressure than for those specimens that were internally pressurized.
Figure 1.h-2 shows the clad and ferrules'for a specimen that was heated to approximately 2,300 F.
These tests demonstrate that the degree of eutectic formaticn will not interfere with the flow of emergency coolant to the core following the accidec.t.
Brittle Failure of Clad as a Result of Quenching In the course of a loss-of-coolant accident, cladding is heated to temperatures on the order of 2,000 K for a short period of time fol-loved by flooding with cooled vater whi,ch effectively quenches the clad surface. The clad must be capable of withstanding this quench-ing effect without failures that could lead to interference with emergency core cooling. The clad must also maintain its structural integrity during and after the quenching condition.
At various stages during the manufacture of Zircaloy, heating to tem-peratures on the order of 2,000 F followed by water quenching is a normal procedure. Subsequent to this water quench, the material is subjected to additional working. This demonstrates that heating and quenching at temperatures of approximately 2,000 F is not detrimental to the mechanical strength of the material.
(T-The peak temperature limit of 2,300 F in +,ne loss-of-coolant accident V
is slightly above the temperatures norma *.ly used in the manufacture of Zircaloy.
Consequently, B&W has conducted tests on four samples heated to 2,100 F and three sa=ples heat 3d to 2,300 F.
After soak-ing times between 1 and 5 minutes the sp ecimens were quenched in room-temperature water. Visual examinati:n of the specimens indicat-ed a heavy oxide coating. A shock test of the specimens indicated that they were not brittle. A strength test of the end of each speci-men indicated that the strength of the specimens vaa not reduced, but
.that a reduction in ductility had occurred.
The effects of irradiation on fuel cladding, including slightly re-duced ductility, are expected to be annealed out during the thermal transient. Results of work at B&W(10) indicate that the ductility of specimens Arradiated at 775 F vill increase slightly. Thus, quench-ing of irradiated clad specimens would be expected to show the same results as the unirradiated quench specimens. These quenching tests demonstrate that ' heating and quenching during the loss-of-coolant ac-cident will not lead to a structural failure of the Zircaloy cladding that could interfere with emergency core cooling.
Clad Perforation and Deformation As the temperature of the Zircaloy cladding exceeds 1,200 F, its (34f
. strength begins to decrease rapidly. Since fuel rods that have op-erated in the reactor for some time vill contain fission gases that develop pressure inside the rod, this loss of strength may lead to
()
perforation or deformation of the cladding. If the clad perforetes, 1.4-3
the contained fission gases vill be released. If it deforms, the deformed clad material could rotentially interfere with the flow of emergency core coolant.
In evaluating the radiological consequences of the loss-of-coolant accident 100 per cent of the fission gases conte d in the cladding is assumed to be released. In addition, a hype.hetical accident has been analyzed in which 50 p;; cent of the iodine and 100 per cent of the noble gases are released from the core. Besultant doses are with-in the guidelines of 10 CFR 100. Thus, perforation vill not lead to accident consequences greater than those presented in the PSAR.
The Oyster Creek PSAR has reported in Amendment 10 Docket No. 50-219 the results of tests with prepressurized, 50-inch-long Zircaloy-2 clad fuel rods containing UO2 pellets. The pins were heated using induction heating to simulate the loss-of-coolant accident thermal transient. The report states, "for initial internal gas pressures ranging from 130 psig to 700 psig, clad failures occurred between 1,3h8 and 1,7h2 F.
Both pre-oxidized and un-oxidized rods were studied. " The deformation of the unoxidized rods consisted of local-ized bulges about 1 inch long at one or more axial locations in the region of maximum temperature. Maximum elongations between 21 and 58 per cent were observed.
The preoxidized samples exhibited a much more brittle behavior than those that were not oxidized.
Typically, the mode of failure was a longitudinal crack with little elongation rather than a burst. These tests indicated that the perforation is random in that it will occur at a particular, even a very slight, weak point along the fuel rod length. The distribution of these weak lh points is expected to be random.
B&W has also undertaken a program to evaluate the perforation and de-
. formation of fuel rods during the temperature transient following the loss-of-coolant accident. Preliminary tests have been run on nine samples of Zircaloy-b cladding filled with ceramic pellets. The spec-imena were pressurized at 600 F and then subjected to thermal trans-ients simulating those associated with a loss-of-coolant accident.
Initial internal pressures ranged from 100 to 3,300 psi.
Figures 1.h-3 and 1.h-4 show typical specimens exhibiting the characteristic ductile failure of the cladding.
In the second figure the restrain-ing effect of the thermocouple that was attached to the specimen is evident.
The results of these preliminary tests are generally com-patible with the work that has already been reported in the litera-ture.
Since the rods were not preexidized they failed with the typi-cal local bulge expected of a ductile failure. Elongations varying from 6 to 39 per cent were experienced.
The results of these preliminary tests support the analytical conclu-sien that the less-of-coolant accident could lead to some clad defor-mation ahich reduces the local cross-sectional flow area in scme chan-nels. Although the axial locations of the local failures are.
expected to conform generally to the hottest sections of the fuel rods, the exact distribution of the perforatiens within these hot zones is ex-pected to be random.
9 03,42
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Since the fuel rod perforations vill have the characteristics discussed
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above, the overall geometry of the fuel assemblies, which contain 208 s r-fuel rods, approximately 12 feet long, will be essentially unchanged after a loss-of-coolant accident. The perforations vill not interfere with the ability of the emergency core coolant to limit the temperature following the accident, since core cooling is achieved by flooding the core. Even in the event that the flow area of isolated channels were to be significantly reduced, cross-flov from other channels would permit coolant to cool the fuel rods above...a affected positions.
Additional Testing Planned Additional experiments are pltaned to gain a clearer understanding of the effects of temperature excursions on Zircaloy-clad fuel ele-ments. Current plans include performance of a three-phase program.
The first two phases are experimental and cor.sist of performing single rod excursions to better establish temperature-pressure relationships at the time of clad rupture. The single-rod tests of the first phase vill also investigate the extent of deformation to be expected under the varying conditions associated with simulated in-reactor tempera-ture excursicns.
These vill include the effects of hydrogen concen-tration and oxide films.
The second phase of the program vill con-sist principally of multirod tests to explore the effect of the re-s training action of spacer grids and adjacent fuel rods and to deter-mine the randomization of the localized deformation in an assembly of fuel rods. The third phase of the program vill consist of applying r~(]/
the data obtained from the two experimental phases to the analysis of the effects in a loss-of-coolant accident.
s 0343 (7.
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l.h-5
REFERENCES g
1.
W. A. Bostrum, The High 2emperature Oxidation of Zr-2 in H 0, Westinghouse -
2 AFD, WAPD-10L, March 19, 195L.
2.
A. W. Lemmon, Studies Relating to the Reaction Between Zirconium and Water a+ High Temperatures, Eatte le Memorial Institute, EMI-115h, January 1957 3.
L. Baker and L. C. Just, Studies of Metal-Water Reactions at Eigh Tempera-tures, III.
Experimental and Theoretical Studies of the Zirconium-W er Reactions, Argonne National Laboratory, ANL-65L8,1962.
h.
J. F. White, "Physico-Chemical Studies of Clad UO2 in Potential Meltdown Environment," GE NFTO, Sixth Annual Report - High Temperature Materials Program, Part A, March 31, 1967.
5 B. Lustman, Zirconium-Water Reaction Data and Application to PWR Loss-of-Coclant Accident, WAPD-SC-5h3, May 1957.
6.
Gerken, J. M., Diffusion Bonding of Zircaloy to Stainless Steel Tubing, General Electric Company, Knolls Atomic Power Laboratory, FJJL-M-JMG-10, Schenectady, N.Y., May 28, 1957 7.
Feduska, W., "An Evaluation of the Diffusion-Bonding Characteristics of Zircaloy-2," Welding Journal, July 1959.
8.
Lehner, W. and H. Schwartzboart, " Diffusion Bonding of Zircaloy Plate-Type Fuel Elements," Welding Journal, February 1961.
9 Eckel, Jchn F., " Diffusion Across Dissimilar Metal Joints," Welding Journal April 196h.
10.
E. N. Harbinson and C. J. Earock, Mechanical Properties of Zircaloy-h After Irradiation at 130, 650, and 775 F, TP-299, ASTM Paper No. h6.
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Dock ts 50-302 and -303 Supplement No. 1 February 7, 1968
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QUESTICN Describe the manner in which Unit 4 vill be constructed as relcted 1.5 to operation of Unit 3; consider possible blasting, Unit h Reactor Building pressure-test, utilization of heavy c,uipment in and around the existing auxiliary building, control room, and other shared areas.
ANSWER The constructiwn management and precautionary procedures required to construct additional generating units adjacent to operating units at an existing plant site are well kncvn to FPC. The Mechan-ical Engineering Department personnel have many years of experience in this type of activity in fossil fuel plant design and con-struction.
The sub-surface testing and stabilization programs developed for the foundation design of Units 3 and h have included consideration for construction of Unit h with Unit 3 in operation. We do not anticipate any requirement for blasting or other construction activity that could have any detremental effect on the operation of Unit 3.
Should such activity be required, procedures will be carefally evaluated and the _vork progress continuously monitored for magnitude of any effects. Proper controls over the work and the surveillance program should preclude any detrimental effects occurring.
O V
0349 0
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1.5-1 l
QUESTION Provide additional justification for the assumption that the 1.6 1-to 5-mile zone will not become much more populated during the life of the plant. Will the 5-mzle radius be considered as the low-population zone?
ANSWER Reference is made to the updated Section 2.2.4 of the Preliminary Safety Analysis Report as amended on 1-15-68, which projects no significant change in the population distribution within five miles. The low population distance based on 10 CFR 100 assumptions is shown to be one mile in Figure 1k-60 of the PSAR, and based on actual population density is k' -iles.
For 3
the purposes cf this application, in the intent o..aded safety, the low population zone will be arbitrarily extended to a five mile radius.
O 0350 r
1.6-1 (Revised 3-1-68)
Docksta 50-302 and -303 i
Supplement No. 1 February 7, 1968 O
QUESTION We require assurance that the DNBR will not be less than 1.3 (W-3 2.1 correlation) at design overpower with consideration given to unde-tected loss of one or more core barrel check valves.
Include the instrumentation response available to the operator that will indi-cate such valve defects.
ANSWER The DNB ratio in the hot channel at the maximum overpower with a vent valve disc off vill be high enough to insure that there is a 99 per cent confidence that at least 9h.5 per cent of the population of all such channels are in no jeopardy of experiencing a DNB. This degree of protection is consistent with Paragraphs 3.1.2.3 and 3.2.3.1.1 of the PSAR.
It will be demonstrated in the final design that the DNB ratio in the hot channel with the flow resulting from the loss of one vent valve disc will not be -less than 1.3 using the W-3 correlation.
A preliminary sensitivity analysis using postulated worst-case param-eters has been made for the reduced flow. The results of this analy-sis are described in the answer to Question 5.1.5 in Supplement No. 1 of the Metropolitan Edison PSAR, where the DNB ratios for full and re-duced flow for various reactor powers are as follows:
Per Cent Rated Power DNER, Full Flow DNBR, Reduced Flow sg 100 1.76 1.68 l
107.5 1.53 1.4h
()
112 1.h0 1.30 llh 1.34 1.2h The minimum DNB ratio of 1.24 resulting from the analysis at 114 per cent power for the postulated worst case is large enough to insure a DNB ratio of not less than 1.30 for final design conditions. The postulated vorst case, used for sensitivity analysis, is not the de-sign condition, but a case with heat transfer and mechanical condi-tions much more severe than expected in the final design. This is demonstrated comparing the nominal and postulated worst case as shown in the answer to Question h.3 of Supplement No.1 of the Metropolitan Edison PSAR, where the W-3 DNB ratios at rated flow conditions and llh per cent power are as follows:
Cell Type Nomina) DNB Postulated Worst Case DNB Corner 1.85 (1.71) 1.3h (1.24)
Wall 1.89 1.38 Unit 1.89 1.46 The minimum DN3 ratios occurring in the corner cell for the two con-ditions at reduced flow due to loss of a vent :ralve disc are shown in parentheses above. The final design DNB will be within the limits u
0 5' 2.1-1
of the 1.71 to 1.2h shown. It is expected that a value greater than 1.30 vill result from final evaluation of the combination of the fol-loving significant factors:
1.
A mixing coefficient of 0.03 to 0.07 at design conditions as against the 0.01 used in the preliminary analysis.
2.
Statistical determination of mechanical tolerances in lieu of minimum conceivable dimensions.
3 A more accurate determination of the hot channel local peaking factor of 1.C95 shown in Figure 3-khr of the PSAR considering:
(a) the statistically determined water gap, and (b) the excess metal in the solid can section surrounding the corner pin. The final value is expected to be about 1.06.
h.
Application of final vessel and core flow distribution test re-sults instead of the hot-to-average fuel assembly flow ratio of 85 per cent assumed for the worst postulated case.
5 The statistical comparison of the multiple rod fuel assembly heat transfer test data with the single-channel data that currently form the basis for the W-3 correlation.
A consideration of the final thermal-hydraulic design compared to the
/
preliminary postulated worst case and the mechanical integrity of the vent valve indicates that it is very unlikely that the core vill be 3' h subject to an unsatisfactory heat transfer condition. The loss of a check valve is very unlikely but even if it does occur the core is not in jeopardy of fuel damage. The loss of two or more check valves between inspections is considered incredible but the thermal perfor-mance of the core has been' analyzed for this postulated condition.
The preliminary sensitivity analysis using postulated worst case parameters and reduced flow has been extended to include the case of two open check valves. Under these conditions core flow is reduced 10 per cent and loop flow is increased 2.2 per cent. The minimum DNB ratio for the postulated worst case would be 1.3 using the W-3 correlation at a power level of 106 per cent. Thus, the minimum DNS ratio specified in the question would be met at full power oper-ation even with two check valves cpen.
Under the conditions of two check valves opened, the changes in re-actor coolant conditions could be detected by the operator upon pe-riodic inspection of reactor flow and temperature differential.
The changed reactor coolant system te=peratures and flow would provide g l} })
an indication available to the cperator inasmuch as the 2.2 percent flow change would cause a corresponding change in reactor temperature differential..In the event that periodic inspection indicates a potential failure, reactor power would be appropriately reduced to assure an adequate ENER, and an investigation vill be instituted.
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February 7, 1968 QUESTION Submit additional design data on the primary pump anti-reverse rota-2.2 tion device. We understand that there has been essentially no ex-perience with the proposed device at high speeds such as vill be encountered. Also, (t appears that performance testing cannot be accomplished during operation. Finally we understand that it is your position that safety was not a consideration in the decision to pro-vide the device. Provide as confirmation your present position on the above statement.
ANSWER The Formsprag clutch has been selected as the anti-reverse rotation device for the General Electric motors on the reactor coolant pumps.
Upon reversal o' the torque on the shaft the sprags lock when the shaft attempts to rotate backward, thus preventing reverse rotation.
The device is designed to withstand a torque in excess of any torque anticipated on the motor shaft. Qualification testing of the clutch vill include operation at pump rotational speeds. Performance test-ing can be easily accomplished by turning one reactor coolant pump motor off and observing its coast down to zero speed as indicated by the zero speed detector. When the anti-reverse device is holding there vill be no rotation indicated. In the event of a failure of the anti-reverse device to hold, the shaft will rotate backward and the zero speed indicator vill show that the shaft is turning. The g3 anti-reverse device was provided to contribute to long motor life by g,j reducing rotor heating during ste ting and to provide the ability to allow two or three consecutive motor starts. Prevention of frequent reverse rotation in the pump seals was a secondary consideration.
Safety was not a consideration in the decision to provide an anti-reverse holding device. This aspect is discussed in mere detail in the answer to Question 8.h.
0353 v
2.2-1
.o
Docksts 50-302 and -303 Supplement No. 1 February 7, 1968 QUESTION State the criteria for dividing the auxiliary building design 3.1 into both Class I and Class II zones.
Indicate whether the components required for safe shutdown can withstand loss of the Class II components (collapse during the maximum probable earthquake).
ANSWER
~
seismic design of the complete structure of the auxiliary building is based upon the criteria for Class I structures.
This includes the steel superstructure, which has previousl:r been classified as Class II.
Refer to revised Sections 1.1 and 1.2 of Appendix 5A, Structural Design Bases."
0354
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Dock;ts 50-302 and -303 Supplement No. 1 February 7, 1968 Ob QUESTION Your design provides capability for Reactor Building purge during 3.2 operation. What buildint pressures would result if the purge valves were open during a LOCA? Is there a break size small enough to prevent (with purge valves open) pressure buildup to h psig (isolation pressure), but large enough to incur clad failures?
Justify not using ECCS actuation as a building isolation signal.
ANSWER In the unlikely event of simultaneous Retetor Building purging and LOCA, the purge valves vill close automatically on a 4 psig signal.
If an activity release occurs prior to a Reactor Building pressure of h psig, then high radiation within the Reactor Building vill automatically initiate closing of the purge valves.
(see Question 5.8).
The case of a 36-inch-ID, double-ended rupture LOCA occurring during Reactor Building purging has previously been considered in Section 14.2.2.3.6 of the PSAR. The e-"dronmental consequences of this accident are the release of 3 equivalm curies of I-131 and a total integrated thyroid dose of 0.h8 rem at th? exclusion distance.
If this dose is added to the LOCA 2-hour thyroid dose, then the total 2-hour dose at the exclusion distance is 1.93 rem, which is well within 10 CFR 100 guidelines. Loss of Coolant accidents resulting from the rupture of pipes smaller than 36-inch-ID vill result in lower environmental doses because the pressure rises within the Reactor Building vill be slower, resulting in less flow out the open purge valve.
nd For the case of small leaks which do not result in a Reactor Building pressure of h psig or even a reactor coolant pressure of 1800 psig; the high radiation signal in the Reactor Building purge duct vill assure closing of the r"rge valves and consequently lov dose rates at the site boundary.
1 l
ou 0355 3.2-1
Dockets 50-302 and -303 (q
Supplement No. 1
)
February 7, 1968 QUESTION Provide the following information on the core flooding tanks:
h.1 (a) Method of adding water.
(b) Immunity of the pair of tanks to a single failure of the N2 pressurization system.
(c) Use-rate of.N during normal operation.
2 (d) Estimated sampling frequency for boron concentration.
(e) Projected frequency of a full-scale discharge test of a CF tank into the primary system.
(f) Leak characteristics of relief valves.
ANSWER (a) Water is added to each of the two core flooding tanks indepen-dently through a separate reactor building penetration. Normal fill and makeup water vill be added from the makeup and purifi-cation system. Borated water may also be added from the chemi-cal addition and sampling system.
i V
(b) The two tanks are considered to be totally immune to a single failure which could in any way affect safety. This includes a single failure of N2 pressurization system. As mentioned above, all makeup, including the addition of N, is made to each tank independently. Thetanksarelocatedobtsidethesecondary shield and at considerable distance from each other as indicated by the plan drawings in Section 1 of the PSAR. Each of the makeup lines entering the tank contains a check valve. With this arrangement, even a failure of the common N supply system 2
outside the reactor building would not permit depressurization of either or both of the core flooding tanks.
(c) It is expected that the use-rate of N during normal operation vill be approximately zero.
(See park f below. )
(d) The tentative sampling frequency for boron concentration will be on a monthly basis.
In all probability during the early stages of plant operation, this test vill be conducted more frequently until it is confirmed that this frequency is adequate.
(e) -A discharge test of a core flooding tank into the primary system
. vill be performed during the initial test program. It is ex-j pected that lesser-tests to determine that the core flooding check valves are working properly will be performed preceding each refueling period.
O 0356
~
h.1-1 (2-7-68) l
(f)
The industry standard for relief valves serving a function such as those installed on the core flooding tanks allows a leakage rate of 0 75 S.C.F.D. We expect valve leakage to be significantly gW less than this.
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0357 O
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,.1-2
Dockets 50-302 and -303 Supplement No. 1 February 7, 1968 QUESTION Indicate whether routine testing of the reactor building spray system 4.2 vill include opening of the sodium thiosulfate tank cutlet valves.
A.tSWER Routine testing of the reactor building spray system will include op-ening of the sodium thiosulfate tank outlet valves.
O 0353 l'
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h.2-1
Dockets 50-302 and -303 f) _
M Supplement No. 1 r
February 7, 1968 QUESTION The PSAR states that primary motor status monitors are designed to 5.1 serve as flow monitors, and that no direct flow measuring devices are provided. In this light, provide:
5 1.1 Design details of a' pump monitor, including its mode of opera-tion, range of alarm sensitivity to abnormally high and low currents, independence from other monitors, and response to loss of one of the three phases of motor voltage, change in pump power for range of temperatures, and response to trip of another pump.
5.1.2. A. summary of PWR experience or previous designs wherein this design has been used.
i 5 1.3 Proposed method of determining at the plant startup that de-sign flow rates have been achieved.
5.1.4 Procedures for verifying during the lifetime of the plant that flow rate is not degraded below design values.
ANSWER.
5 1.1 The electrical power to each reactor coolant pump is moni-tored by four independent power-measuring devices. Each pump
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and potential transformers. The output of each pump power monitor is an analog signal directly proportional to the 4
pump electrical power. The signal is transmitted to the nu-clear cabinets where each protective channel receives one independent signal from each pump. Each signal is monitored for both abnormal high and low power. The detection of an out-of-limits pump power signal initiates an immediate step change in the overflux trip point to a lower value.
All elements of one pump power monitor and its protective system logic are completely independent, both physically and electrically, from every other pump power monitor.
The overall accuracy of the pump power monitoring system is I
within 15 per cent of true pump power.
The loss of one phase of electrical power to a pump will not impair the power-measuring capability of the pump power moni-
~
tor and will trip the monitor logic. The same fault will cause the pump power electrical breaker to open, cutting off all power to the pump.
For comparison, consider the power input to each pump, with four pumps operating and the reactor coolant temperature above 500 F, as being 100 per cent of the normal hot load, oO 0359 5 1-1 y
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The cold (70 F) input power is 139 per cent of the normal hot load, the locked rotor starting power is 560 per cent of the normal hot load, and the unload " broken shaft" power is 7 per cent of the normal hot load. With only three pumps oper-ating, the two pumps in one loop drav 101 per cent of the normal hot load, and the single pump in the other loop draws 105 per cent.
The pump power monitoring system is designed to detect sig-nificant changes in pump power. The logie vill consider a pump's power outside of its normal operating limits then the measured power exceeds 155 per cent of the normal hot value (approximately 12 per cent above the normal cold power value) or falls to 50 per cent of the normal hot value.
5.1.2 The Babcock & Wilcox Company has provided pump monitoring systems on two presently operating reactor plants for protection against loss of flow accidents. The present system is a logical ex-tension of the same basic principal of flow change detection by monitoring pumps incorporating present-day technology and criteria for redundancy and separation of protection channels.
5.1 3 The reactor coolant pump manufacturer vill provide head capacity curves for nis pumps based on full-scale pump tests in a test loop. These curves will be used to verify design flow in the reactor coolant system during plant startup. A second verifi-cation technique vill be to divide the steam side heat balance by the reactor coolant system temperature rise.
1.h During the plant startup, the reactor and steam generator pressure drops will be correlated against the design flow rate.
Any degrading of the flow rate below the design value vill show up as changes in the pressure drop measurements. Any changes in presssure drop, whether increasing or decreasing, vill be cause for investigation.
An evaluation of the significance of changing pressure drop measurements can be made by calculating the reactor coolant flow from the steam side heat balance divided by the reactor ecolant system temperature rise.
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5.1-2
Dockets 50-302 and -303 r
Supplement No. 1
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February 7, 1968 QUESTION Provide an analysis on the likelibood and consequences of a failure 5.2 of one in-core instrument tube at the pressure vessel.
ANSWER In order.to fail an 'in-core instrument tube inside the reactor vessel, three barriers must be penetrated: the guide tube, the instrumentatien bundle, and finally the calibration tube. There is no conceivable ac-tion that could take place that would cause penetraticn of these bar-riers. If it should, the maximum leak area vould be that correspond-ing to the calibration tube which has a 0.1h8" ID.
As shown in Figure 7-13 of the PSAR, the guide tube is shielded all
'the way from the reactor vessel to the flange so that again there is very little likelihood that a tube could be damaged external to the reactor vessel. However, should a failure occur outside of the re-actor vessel the maximum leak size that could be obtained is that resulting from a complete shear of the guide tube which has a 0.622" inside diameter.
As explained in the answer to Question 8 5, the normal makeup syst im can be used to maintain system volume and pres-sure for leak sizes up to 0.60" in diameter. With one ruptured guide tube the system pressure would decrease to 2100 psia. At this pres-sure the normal makeup system would keep up with the leak rate.
Therefore, system volume could be maintained and a normal shutdown cooldown could be achieved.
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5.2-1
=
Dockets 50-302 and -303 Supplement No. 1 February 7,1968
, ~mv) i QUESTION Justify the use of a single bus to energize all control rod clutches.
5.3 Show how this satisfies the IEEE Prop. sed Standard.
ANSWER The output circuitry of the reactor protection system associated with the control rod scram bus vill be revised in accordance with the fol-loving criteria.
A.
Safety Criteria 1.
Plant Safety Criteria shall have priority over all other de-sign criteria.
2.
Two diverse means of tripping shall be provided.
3.
No single failure shall disable trip action to more than one rod.
h.
No single failure following undetectable failures shall dis-able trip to more than one rod.
5 At least one means of tripping shall be testable while the plant is operating.
6.
All means of tripping shall be testable while the plant is
-w g down.
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B.
Plant Reliability Criteria 1.
Insofar as practical, no single failure vill cause the plant to trip.
2.
Insofar as practical, equipment shall be accessible to quali-fled technicians for maintenance purposes without entering the reactor building.
3.
Maintenance aids shall be provided to monitor incipient clutch winding failures which could lead to plant shutdown.
C.
Circuit Protection 1.
Circuit protection shall be provided to limit the consequen-ces of. electrical faults.
Description of Trip Circuit A preliminary design, which is subject to revision during the detailed design, is described below.
Fig tre 3-68 shows the revised circuit arrangement.
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Two power sources, station battery 1 and statirn battery 2 provide redundant power sources to the 69 clutches. F ath sources feed the lh clutch bus through two series-connected circuit breakers. Each series-connected breaker is fed oy the reactor protection system 3-and the combination of reactor protection system logic and the series-circuit breaker arrangement provides a two-out-of-four trip circuit as shown in Figure 7-2b.
Reverse power diodes allow the two power sources to be connected to a commcn clutch bus without power exchange between the two sources.
'the clutch bus feeds four branch busses which are protected against electrical faults by overload elements working in conjunction with one set of the main circuit breakers. By splitting the clutch bus into four branches the level of electrical protection permits the use of reasonably sized conductors of about 12 or lh gauge. A cur-rent meter is provided in each branch feeder as a maintenance aid to monitor possible clutch coil open circuits or shorted turns.
The primary means of tripping is through the circuit breakers feeding the clutch bus.
This trip circuit is fully testable while the plant is operating. Through test push buttons (not shovn) operating in series with the undervoltage coil of each circuit breaker, the break-ers can be tripped, one at a time and monitored on the power supply monitor connected between the circuit breakers and the reverse power diodes.
A secondary means of tripping is provided through relays K1 through K69, connected in series with each clutch power line. A trip relay bus is dual fed from the same two battery sources through the trip circuit breakers, reverse power diodes, and power supply voltage monitors. The trip relay bus feeds 69 relay coils, one for each mechanism. Each of the clutch coils is fed through double pole, double break contacts on the relay.
The secondary trip circuit is fully testable when the plant is down.
The circuit breakers feeding the relay bus are fully testable while the plant is operating as previously described for the clutch bus.
To reduce the probability of an inadvertent rod drop, the clutch and relays are provided with dual coils.
t All equipment except the clutches and parts of the clutch power cables are accessible for inspection by qualified technicians during plant operation. The inaccessible parts, clutches and cables, are monitored through ammeters.
Careful attention will be given to selection of parts, fabrication, packaging, housing and testing to maintain highly reliable safety grade equipment.
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f-Dockets 50-302 and -301 (3)
Supplement No. 1 February 7, 1968 QUESTION The PSAR does not adequately describe the preliminary design of the 5.h power-to-flow instrumentation. Please provide a more detailed de-scription and justify the combining of protecticn and control func-tions.
ANSWER The power-to-flow instrumentation consists of high-low pu=p power limit detection, a monitor logic which counts the number of pumps op-erating within limits, and a flux / pump bistable whose trip point is controlled by the monitor logic.
When all four reactor coolant pumps are operating within the limits of 155% to 50% of their normal hot power, the output voltage of the monitor logic is at its maximum value holding the setpoint of the flux / pump bistable to some value above the fixed 107 5% full power trip point of the overflux bistable.
If the power input to one pump goes outside the limits, the output voltage of the monitor logic steps down to a lower value thus forcing the flux / pump bistable trip point to 75% full power.
If one pump in each reactor coolant loop goes out-side of limits, the trip point is stepped to 50% full power.
The monitor logic not only counts the number of normal pumps but also i
determines the number in each loop.
Should the power input to two i, ')
pumps in one loop go out of limits, the monitor logic initiates a
'~'
There are four identical systems as described above, each independent of all others and each associated with one protective channel. The output of. the monitor logic in one of the channels provides pump in-formation to the integrated control system.
Four relays within the logic correspond to the four monitored pumps. The contacts from these relays supply pump information to the integrated control system.
The combining of protective and control systems functions is justified since (1) only one of four redundant channels is involved, (2) the sys-tems are electrically independent being separated by the insulation resistance between the relay coil and its contacts.
0002 I
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L'ockets 50-302 and -303 g) g Sapplement No. 1 February 7, 1968 QUESTION Provide a description of the preliminary design of the operational 55 test system and procedures for the Protection and Engineered Safety Feature channels.
ANSWER The operational on-line test schemes for the reactor protective sys-tem and the engineered safeguards protective system are identical in concept. Every trip function which originates from an analog signal, flux, pressure, temperature, etc., is tested by substituting an analog test signal for the variable. The test signal is manually injected into the instrument channel at the input. of the first active channel element in the protective system's cabinets. For the nuclear instru-ments, the test signal is injected in the neutron detector input while test signals for pressure and temperature are injected at the linea.
input of their associated bistables. Test signals for the pump coni tor are injected at the input from the pu=p power monitors.
Testing an analog channel consists of varying the test signs 1 over the entire dynamic range of the channel and observing that te chea-nel responds properly and that the associated bistable not only trips but trips at the correct set point. The on-line test is designed to detect set point and zero drift or a change in the channels response.
Calibration and corrective adjustments may be made during the on-line n()
test.
Since the on-line test actually results in the instrument channel trip-ping its associated protective channel and since it is prohibited to place more than one channel en test at a time, the coincident logie networka where the protective channels lose their identity are tested by parts. Each logic element is tripped, observed and reset. The test scheme depends upon the coincidence action of the logic to pre-vent a complete protective system trip during testing.
The test scheme includes an independent interlock within each protec-tive channel. The purpose of this interlock is to trip the protective channel before any associated instrument can be tested; thus, any at-tempt to test elements of two protective channels at the same time vill result in a protective system trip.
Part of the test scheme includes providing readouts of the redundant analog inputs to the protective channels. These readouts will permit the operator to compare the performance of like instruments and ob-serve their response to changes in the plant.
Each engineered safeguards valve, pump, fan, etc. vill be on-line tested by manually activating the associated engineered safeguards control line and observing that the individual equipment responds.
Each item vill be activated only long enough to verify that it per-forms its safeguards function when commanded to do so and vill be im-
/O:
00 0 >}iately restored to its normal state.
mec 5 5-1 t
i Dockets 50-302 and -303 Supplement No. 1 4
February 7, 1968 I
i QUESTION Clarify the use of the same temperature instrumentation for both pro-i 5.6 tection and control.
l ANSWER There is separate, independent temperature instrumentation for pro-tection and control. Figure 7-11 has been changed to clarify this point.
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rebruary 7,1968 QUESTION List all-protection system channels which provide bypasses and shov 57
. that the design complies with IEEE Proposed Standard. List all pro-tection system channels which contain variable trip settings and show that the design complies with the IEEE Proposed Standard.
_ ANSWER The low reactor coolant' pressure trip functions vill be by-passed for reactor startup and cooldown. This includes the low pressure reactor coolant trip of 2,050 psig in the reactor protective system, the 1,800 psig high pressure injection system trip and the 200 psig low pressure injection system trip.
Each of the pressure trips bypass functions are identical in their implementation and differ only at operating point.
In the low pressure state at startup, the operator manualy initiates each of the low pressure trip bypasses separately. Actuation of one of three momentary switches bypasses the low pressure reactor trip until the reactor pressure reaches 2,100 psig at which time the by-pass is automatically removed. Likewise, the low pressure injection-system trip becomes active automatically at 600 psig and the high pressure injection system at 1,900 psig.
f-s During cooldown and depressurization, the low pressure bypasses may
( )
be manually initiated only within the deadband of the bypass equip-ment. For the low pressure retctor trip the operator can initiate the bypass only between 2,100 prig and the 2,050 psig trip point:
for the high pressure injection system, initiation can occur only between 1,900 psig and 1,800 psig, and for tae low pressure injection system between 400 psig and 200 psig.
Once a low pressure trip occurs, the bypass cannot be activated with-out first manually resetting the low pressure trip bistable.
Each low pressure trip bistable has an independent bypass circuit,
associated with it, therefore all elements of the bypass functiou are a part of the protective system and designed to meet the IEIE Standard for Nuclear Power Plant Protective Systems.
" Variable" trip bistables are used for the flux / pump bistables as described in'the answer to question 5.h.
The-protective system bistables may be considered voltage comparators in which the measured variable signal voltage is compared against a trip point voltage originating within the bistable. When the mea-sured variable signal equals or exceeds the trip point voltage, the 0005 bieteb1e trips. seceuee of the comperator ection, the bisteb1e trip point. can be controlled from an external source such as the output of the pump monitor logic. Loss of the trip point control voltage
.results in a trip.
O L) 5.7-1
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The variable trip bistables meet the requirements of the IEEE Stan-dard by being part of the redundant and independent pu=p monitor logic and meeting the single failure criteria.
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Docksts 50-302 and -303 Supplement No. 1 February 7, 1968
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QUESTION Provide further details on your radiation monitoring system. For 5.8 guidance as to appropriate details refer, for example, to the Metropolitan Edison PSAR (Docket 50-289), Supplement 1, Question 10.1 through 10.8.
ANSWER The Radiation Monitoring System vill consist of the Area Gamma Monitoring, Atmospheric Monitoring, and the Liquid Monitoring systems as shown on Figures 5.8-1 thru 5.8-6.
All monitoring signals vill be indicated and recorded in the control room.
The Radiation Monitoring System vill receive power from the battery backed inverter red vital instrument busses, and each channel vill have a loss of power and channel failure alarm in addition to a high radiation alarm.
Interlock functions are as indicated on Figure 5.8-5 The Radiation Monitoring System detector locations and sensiti-vities vill be as follows:
A.
Area samma Monitoring PSAR Detector Range Fig. Ref.
Location
.1 to 100 mrem /hr 1-6 Control Room 1-3 Radiochemical Laboratory Auxiliary Building (O
_)
.1 to 10,000 mrem /hr 1-3 Sample Roan (1) 1-3 Entrance Corridor (1) 5 1-3 Gas Decay Tank Area (1) 1-h Make-Up Tank Area (1) 1-2 ' l-h R.C. Holdup Tank Area (El. 95 & 119) (2) 1-4 Near Personnel Access Hatch (1) 1-3 M.U. Pump Arer. (1) 14 Deborating Demineralizer Area (1) 1-3 Spent Resin Storage Tank 0007 Area (1) 1-5 & 1-6 Decon. Pit & Fuel storage Pool Area (2)
Reactor Building
.1 to 10,000 mrem /hr 1-4 Near Personnel Access Hatch (1) 1 ren/hr to 1,000,000 rem /hr 1-7 Reactor Bldg. Dome (1)
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5.8-1 (Revised 4-8-68)
B.
Atmospheric Monitoring The recorders vill be located in the control room; i e detectors g
vill be located as follows:
W PSAR Type of Type of Location Fig. Ref.
Measurement Monitor Reactor Building Purge Duct (1) 5-6 P,G,I Fixed l5 Auxiliary and Fuel Handling Exhaust Duct 9-12 P,G,I Fixed Control Room Ventilation Duct 9-12 P,G,I Fixed Fuel Handling Building Ventilation Duct 9-12 P,G,I Fixed Auxiliary Building Ventila-tion Duct 9-12 P,G,I Fixed Reactor Building Air Sample Line (1) 1-4 Fixed l5 Site Monitors (3) 5.8-1 P,G,I Fixed Sample Room 1-3 P,G,I, Fixed Waste Gas Decay Tank Discharge 11-1 & 9-12 G
Fixed Condenser Vacuum Pump Exhaust (1) 9-12 G
Fixed l5 Radiochemical Laboratory 1-3 P,G,I
_ Movable Spent Fuel Area 1-6 P,G,I Movable C.
Liquid Monitoring Primary Coolant 5
Letdown (1) 9-2
^
Intermediate Cooling Water (1) 9-5 Nuclear Serv 4ae Closed Cooling Water (1) 9-h Spent Fuel Cooling Water (1) 9-8 Plant Liquid Effluent Line (1) 9-4 Liquid Waste D scharge (Prior to Dilution) (1) 11-1 Regarding how the design basis accident releases and other acci-dental releases relate to the radiation monitoring system design, the following analysis has been made:
The Reactor Building dome monitor is intended to indicate the radiation level inside the building following the design basis accident and has a range, of 1 to.106 rem /hr. The total curies j
released into the Reactor Building following the design basis accident is estimated to be 17 x 100 curies corresponding to 5.8-2 (Revised h-8-68)
/
the release of the total gap activity. It is estimated that this activity would produce a radiation level on the order of 2h0,000 rem /hr, which can easily be read by the monitor.
The concentration of I-131 at the hh00 ft exclusion distance is estimated to be about 4 x 10-I ue/cc during the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period following the design basis accident. The site monitors, with a sensitivity of 10-11 ue/cc for iodine, could record and alarm this activity level. The site gas monitors with a sensitivity of 2 x 10-6 uc/cc could rgeord and alarm an esti-mated Xe-133 concentration of 2 x 10-0 ue/cc. For particulates, assuming all Cs-137 released during the accident wi11 be available for leakage from the Reactor Building, the concen-tration of ce-137 at the exclusion distance is about 8 x 10-7 ue/cc. With a sensitivity of 10-11 ue/cc the particulate monitor could record and alarm this activity level.
The control room and auxiliary building gamma monitors are also anticipated 4 remain on-scale following the design basis accident. For the control room, the shielding provided by the Reactor Building and control room walls and roof and the design of the control room ventilating system will prevent exposures in the control room exceeding 3 rem over 90 days following an MP.A.
The radiation level in the control room over the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following an MHA is estimated to be less than O'
100 mrem /hr. The consequences of the design basis accident are less severe than the MHA hence the control room monitor with a range ur. to 100 mrem /hr should provide level indication throughout the accident. The auxiliary building monitors, with a range up to 10 rem /hr, are also estimated to provide level indication in the auxiliary building following the accident.
In addition to the design basis accident, the effects of other accidents, discussed in Section 1h of the PSAR, upon the ra-diation monitoring system have been investigated. These include steam generator tube leakage, steam line failure, loss of electric pove', rod ejection accident, and a fuel nandling accident.
A 1 gpm steam generator tube leak releases 290 ue/cc of Xe-133 into the secondary system. With a stema flow of 107 lb/hr, this gas could be diluted by as large a factor as 1.7 x 10-6, This dilution factor would produce a Xe-133 concentration of 5 x 10-4 ue/cc available for release from the condenser. The condenser exhaust monitor, with a sensitivity of 10-6 uc/ce, will easily record and alarm this activity. For the steam line failure, with steam generator tube leakage,1h.1.2.9.2 of h
the PSAR indicates a thyroid dose of 0.53 rem at the site bound corresponding to an I-131 concentration of about 8 x 10- uc/ce. The site iodine monitor, with a sensitivity of 10-11 uc/ce, could record and alarm this activity level.
Loss O
5.8-3
of electric power, discussed in 1h.1.2.8.1, analyses steau relief to the environment concurrent with steam generator leakage and gives a concentration of 0.03 MPC for iod4 *:: at the sita boundary.
This is estimated to correspond to a Nncentration of 3 x 10-12 ue/cc of iodine at the site boundar:, during the 2 minute period of steam relief, which is well witnin 10 CFR 20 limits.
Section 1h.2.2.3 of the PSAR discusses the rod ejection with loss of coolant accident. 177,000 curies of I-131 are released to the Reactor Building producing a radiation level of 6500 rem /hr. The additional amount of Xe-133 released is estimated to be 5.3 x 106 curies which will produce about 13,000 rem /hr.
Hence the overall radiation level inside the Reactor Building vill be on the order of 20,000 rem /hr which will be recorded and alarmed by the Reactor Building dome monitor. The dose to the thyroid at the site boundary is 2.68 rem over 30 days correspon-ding to an I-131 concentration of 2 x 10-9 uc/cc which could be recorded and alarmed by the site iodine monitors.
The fuel handling accident is analyzed in Section 1h.2.2.1.2 of the PSAR. The fuel handling ventilation system monitor and the exhaust duct monitor vill normally provide automatic isolation of the fuel handling building for the sudden release of 28.h curies of iodine and 2.79 x 104 curies of noble gages. However, in the event that all this activity is released to the environ-ment the resulting concentrations at the site bounda are estimated to be 1 x 10-7 ue/cc for iodine and 1 x 10 ue/cc h
for tre noble gases. The site monitors could record and alarm these concentrations.
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AREA GAMMA R ADIOCHEMICAL LABOR ATORY X
AREA GAMMA AUX BLDG. ENTR ANCE CORRIDOR (1)
X AREA GAMMA AUX. BLDG. 5 AMPLE ROGM X
ARE A GAMMA AUX. BLDG. R.C. W ASTE EVAP. & GA5 DECAY TANK AREA (1)
X ARE A GAMMA AUX. BLDG. MAKE-UP TANK ARE A (1)
X AREA G AMMA AUX. BLDG. DECONT AMIN ATION PIT X
AREA GAMMA AUX. BLDG. R.C. HOLDUP T ANK ARE A (2)
X X
ARE A GAMMA AUX. BLDG. NE AR PERSONNEL ACCESS HATCH (1)
X ARE A GAMMA AUX. BLDG. MAKE-UP PUMP AREA (1)
X ARE A GAMMA AUX. BLDG. DEBOR ATING DEMINER ALIZER AREA X
ARE A GAMMA AUX. BLDG. 5 PENT RESIN STOR AGE TANK AREA X
ARE A GAMMA AUX. BLDG. THEL STOR AGE POOL ARE A (1)
X AREA GAMMA RE ACTOR BLDG. NE AR PERSONNEL ACCESS HATCH (1)
X AREA GAMMA RE ACTOR BLDG, DOME (1)
X ATMO5PHERIC MONITOR REACTOR BLDG PURGE DUCT (1)
X ATMOSPHERIC MONITOR AUX. AND FUEL HANDLING BLDG.
X ATMO5PHERIC MONITOR CONTROL ROOM VENTILATION DUCT X
ATMOSPHERIC MONITOR FUEL HANDLING BLDG. VENTILATION DUCT X
ATMOSPHERIC MONITOR AUX. BLDG VENTILATION DUCT X
ATMO5PHERIC MONITOR RE ACTOR BLDG. AIR SAMPLE LINE (1)
X ATMOSPHERIC MONITOR SITE MONITOR X
X X
ATMOSPHERIC MONITOR SAMPLE ROOM X
ATMO5PHERIC MONITOR WASTE GA5 DECAY TANK X
ATMOSPHERIC MONITOR CONDENSER VACUUM PdMP (1)
X ATMOSPHERIC MONITOR R ADIOCHEMIC AL L ABOR ATORY X
ATMO5 PHERIC MONITOR SPENT FUEL AREA X
LIQUID MONITOR PRIMARY COOLANT LETDOWN (1)
X LIQUID MONITOR INTERMEDIATE COOLING WAT?R (1)
X LIQUID MONITOR NUCLE AR SERVICE CLOSED COOLING WATER (1)
X LIQUID MONITOR SPENT FUEL COOLING WATER X
LlQUID MONITOR PLANT LIQUlG dFFLUENT LINE (1)
X LIQUID MONITOR PLANT DISCHARGE X
RADI ATION MONITORING SYSTEM h
POWER SUPPLY SOURCES FIGURE 5.8 6 AMEP'. 5 (4 8-68) g
Dockets 50-302 and -303 Supplement No. 1
(,.
February 7, 1968 QUESTION Indicate the means of assuring that those instrumentation and control 5.9 items that must sur.1ve part or all of the LOCA environment have prior qualification performance tests.
ANSWER Protective system equipment specifications require type testing in ac-cordance with the previously referred to IEEE standard and assurance that the quality of the delivered product meets the standards and per-formance requirements specified.
The equipment manufacturer is required to provide qualification test data to verify the performance requirements of the equipment. Adher-ence to the equipment specifications and qualification test data is assured through monitoring and inspection of the manufacturer's work.
Instrumentation and control items that must survive part or all of the LOCA environment are subject to these qualification test verification procedures and requirements.
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Dockets 50-302 and -303
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Supplement No. 1 February 7, 1968 QUESTION Provide your position on diversification of sensing devices for ac-5 10 tuation of the ECCS.
ANSWER Dive: sification of sensing devices vill be achieved in the ECCS by the parallel actuation of both the high pressure and ' av pressure in-jection systems from low reactor coolant pressure and high reactor building pressure.
In logic terms the high pressure injection system vill be actuated from either a low reactor coolant pressure of 1,800 psig or high reactor building pressure of 10 psig, and the low pressure injection system vill be. actuated from either a low reactor coolant pressure of 200 psig or a high reactor building pressure of 10 psig.
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1 0018 4
5.10-1
.m Dockets 50-302 and -303 m
Supplement No. 1 February 7, 1968 i
QUESTION Will the part-length out-of-core ion chambers provide a signal useful 5.11 in detection of axial xenon oscillations? If so, provide details such as sensitivity and projected utilization.
ANSWER The out-of-core power range ion chanbers consist of three h-ft sec-tions whose outputs may be read independently but are normally summed.
The readings of the individual sections might provide information rel-ative to a power tilt of some unknown mahnituda; however, no require-ment for such employment is contemplated in view of the capability of the incore instrumentation to detect power tilts as described in the answer to question 1.1.
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5.11-1
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Docksts 50-302 and -303 Supplement No. 1 February 7, 1968 QUESTION Provide your plans, pre-and post. operational, for survey of marine 6.1 ecology. More details are needed on environmental monitoring programs.
ANSWER A pre-operational environmental radioactivity monitoring program vill be conducted in order to determine the magnitude of the radioactivity in the environme't. surrounding the nuclear reactor site and to study fluctuations in the radioactivity levels prior
.J the operation of the nuclear generating plant. The information obtained will serve as a guide and baseline in evaluating any changes in environmental radioactivity levels that may possibly be attributed to the Crystal River Nuclear Units operation.
Sampling stations and sites vill be selected on the basis of popu-lation density, meteorological, hydrological, and topographical conditions.
A comprehensive sampling program vill be initiated at least two years prior to Unit 3 startup and preferably 1 year prior to Unit 2 startup and include such items as: type of samples, number, frequency of collection and method (s) of analysis. The collected samples will consist of Gulf and well vater soil, air particulate, animal thyroids, fish, shellfish and bottom sediments. Sample radioactivity analyses, based on the type of sample and information desired, vill include one or more of the following: Gross Alpha, O
Gross Beta, Gross Beta-Gamma, Potassium 40, Iodine-131, Strontium-90 and others as appropriate.
The State of Florida has an existing program for monitoring radio-activity levels in air, water, and food crops throughout the State.
Preliminary discussions have been held with appropriate Florida State Agencies on the scope of the program. Continuing discussions will be held with State agencies in formulating an acceptable pre-operational environmental radioactivity monitoring program. The resulting program vill be reviewed periodically to assure maximum effectiveness.
The post-operational environmental program vill be similar to the pre-operational program with the sampling and analyses schedule related to the level of activity found in the environmental sample.
Results will be evaluated to insure the effectiveness of plant radiation control and compliance with all regulatory agencies. This program vill also be periodically reviewed and revised as required.
Reconcentration of specific radionuclides by various tropic levels of marine ecology has been considered in planning the environmental monitoring program by including Gulf water, fish, shellfish as available and bottom sediments.
Interpretations of surveillance results will be based on selected controlling radionuclides such as Cobalt-60, Manganese-54, Cesium-137 and Iodine-131. Factors con-sidered in the selection of these radionuclides are:
(1) the esti-mated composition and relative concentrations of the various radio-()
{q]{ } nuclides in the station's liquid effluent, (2) the reconcentration factors for these radionuclides in the species of fish and shellfish 6.1-1
caught by sport or coc=ercial fishers *n in this region of the Gulf of Mexico, and (3) the appropriate 10 CFR 20 concentration limits. Fish and shellfish samples will be analyzed for gross radioactivity and the controlling radionuclides. Recognition of the reconcentration factor involved in conjuncticn with 10 CFR 20 liquid effluent requirements vill assure that discharges are within acceptable limits for environmental radiation exposure.
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Docksts 50-302 and -303 Supplem nt No. 1 February 7, 1968 f
QUESTION Analyze the likelihood and consequences of a failure of the relief 6.2 valve on one of the vaste gas storage tanks.
ANSWER A standard inch safety valve would be shoptested to a leak tightness of about 0.3 scf per day. The environmental consequences due to this leak rate are far below the limits of 10 CFR 20.
For example, continuous leakage at this rate would produce a concentration at the site boundary of about.0001 MPC using the long-term X/Q 3
value of 1 x 10-6 sec/m,
A leak rate 100 times the above mentioned test specification vould thus produce about.01 MPC at the site boundary. However, manufacturers believe that such a high leak rate is most unlikely thus it is concluded that 10 CFR 20 limits would not be approached due to leaking safety valves.
A complete rupture of the tank is assumed to release all the ectivity as a puff from the plant vent. Atmospheric dilution is calculated using the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 7/Q value of 3 x 10-4 see/m3 The integrated whole body dose at the site boundary is 2.2 rem and the thyroid dose is 2 rem both of which are well within the limits of 10 CFR 100.
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Docksts 50-302 and -303 Supplement No. 1 February 7, 1968 i
QUESTION Sumarize emergency procedures planned for the first hour after 7.1 a major loss-of-coolant accident (LOCA).
ANSWER A criterion for required operator action after a major loss-of-coolant accident has been developed and is presented to indicate the extent of operator action during the initial post-accident period.
" Automatic devices are to be relied upon to accomplish all actions required immediately following an accident to pro-vide protection of the reactor core and reactor building.
Operator action is considered quite acceptable 15-20 min-utes following an accident."
In fulfilling the above criterion, activiation of all engineered safeguards systems for both core and building protection and all isolation valves is accomplished automatically. Switching to the recirculation mode is accomplished manually from the :ontrol room; and, depending on the number of pumps in operation following an accident, recirculation is not necessary until at least 20 minutes after the accident has occurred.
In view of the above, aside from switching to the recirculation mode of post-accident cooling, the operator's primary function is to determine that all automatic actions have occurred properly and take follow-up action in the event that it should be required.
o 0023 o
7.1-1
Dockets 50-302 and -303 Supplement No. 1 February 7, 1968 l'
QUESTION Provide a summary of emergency plans, to include:
7.2 (a) shift responsibilities (b) alarm systems (c) communication systems (d) environmental monitoring equipnent (portable)
(e) notification of and liaison with authorities (f) medical facilities (g) critical actions to be performed prior to evacuation (h) initial assessment of damage plans (i) evacuation plans.
ANSWER (a) Shift responsibilities: Section 12.2.h of the Preliminary Safety Analysis Report has been updated to assign to the shi ft superviser on duty the authority and responsibility to initiate any and all emergency plans, which, in his judgment, are necessary.
(b) Alarm systems: Reference is made to Section 11 of the PSAR, and to the written ansvers to Question 5.8 relating to the radiation monitoring system. On the basis of radiation moni-toring equipment permanently installed, and the alarms associated with this radiation monitoring system, the shift supervisor will determine whether or not a radiation emergency exists, and if required, sound the radiation emergency alarm. This alarm vill have a visual and audible indication in the control room.
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The audible signal vill sound throughout the station and the surrounding site area and vill be unique and distinguishable as a radiation emergency alarm.
(c) Communication systems: Section 12.2.h of the PSAR has been updated to include a description of the communication networks available to Crystal River Plant. These include the Florida Power Corporation microwave network, the Florida Telephone network and a proposed industry vide communication emergency network (EPICEP).
In the event the proposed EPICEP network is r.ot installed at the time Unit 3 goes into operation, Florida Power Corporation vill install a base radio station at Crystal River Plant with emergency power supply, and with a capability of communicating to stations in Ocala, Fort White and Crystal River, and radio equipped vehicles within range.
(d) Environmental monitoring equipment (portable): Section 11 of the PSAR and the answer to Question 5.8 outline in detail the radiation monitoring system, including area gamma monitoring, atmospheric monitoring, and liquid monitoring systems as shown in Figures 5.8-1 thru 5.8-6.
Section 12.2.h of the PSAR has been updated to make reference to the radio chemistry laboratory which is being equipped with all necessary instrumentation and equipment needed to handle a radiation emergency. In addition,
[} g j emergency monitoring kits will be placed throughout the station UC.
gg at predetermined locations. These kits vill be inspected and
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used during periodic drills. Each individual assigned to 7.2-1
(d) Crystal River Plant will have a written copy of his specific g
duties during the radiation emergency plan. Training programs will be established and executed to insure that all selected personnel assigned to emergency monitoring squads have a working knowledge of health physics procedures and the use of radiation instruments.
(e) Notification of and liaison with authorities: Section 12.2.h and Appendix 12-A describe the plan for notifying the Florida Highway Patrol in the event of a radiation emergency which presents a hazard to the surrounding population. General Order No. 36 assigns to the Florida Hi.;hvay Patrol the authority and responsibility for notifying all necessary local and State authorities including the radiological emergency team members.
(f) Medical facilities: The plant will be equipped with medical facilities sufficient to administer first aid to accident victi2ns, and to monitor accident victims for radiation exposure, and vill have facilities for decontamination.
The emergency plan vill have detailed procedures on the handling of accident victims. The three hospitals closest to the Crystal River Plant site are:
1.
Citrus Memorial Hospital at Inverness, Fla.
2.
Monroe Memorial Hospital at Ocala, Fla.
3.
Shands Teaching Hospital & Clinic at Gainsville, Fla.
(g) Critical actions to be performed prior to evacuation: The answer to Question 7.1 summarizes the activities planned for the first hour after a major loss of coolant accident. Personnel assigned to the emergency monitoring squads vill have working knowledge of health physics procedures and the use of radiation instruments, and each individual assigned to these squads vill have specific duties to be performed during the radiation emer-gency plan.
(h) Initial Assessment of damage plans: Instrumentation contained within the control room particularly the radiation monitoring instrumentation, vill inform the orerating staff of radiation levels throughcut the plant. The shift supervisor with the assistance of health nhysics technicians and the emergency squads vill assess the damage and radiation levels throughout the plant.
Dams.ge assessment procedures vill be established and executed as part of the radiation emergency plan.
(1) Evacuation plans: The energency clan vill recuire that designated key perconnel and shift workers required for the emergency opera-tion be retained at the site. All evacuees, both operation personnel and visitors, vill be surveyed by an emergency radiation monitor prior to leaving the plant boundery. All personnel O
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7.2-2
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remaining in the plant to perform emergency operating pro-cedures will be surveyed for possible contamination. Evacu-ation procedures for persons beyond the site boundary will be the responsibility of the Florida Highway Patrol under the guidance of the radiological emergency teams. This procedure and the responsibility assignments is described in Appendix 12A.
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- 0026 7.2-3 i
Dockets 50-302 and -303 Supplement No. 1 February 7, 1968 QUEST 1C Provide an estimate of accumulated radiation to the operating 7.3 staff during and after a major LOCA. Include radiation while in t.** control room, ingress and egress, and possible missions t'
~. turbine building, auxiliary building, and borated water storage tank.
ANSWER An estimate of the accumulated radiation to the operating staff at severd locations in the power plant is given in Table 7.3-1.
The location points A to G are shown in Figure 7.3-1.
This ele-vation was chosen because it shows approximately all locations referred to in the question - that is, borated water storage tank, auxiliary building, turbine building, and the control building.
For conservativeness, several assumptions were made in the calcu-letions. These are:
In the event of LOCA the sources of radiation are the gap a.
activity and the coolant activity h.
The activity is uniformly distributed over the volume of the containment building c.
The iodine is removed after two hours of the LOCA.
As is evident from the results in Table 7.3-1, the tolerance of
's 3 rem is not exceeded after 90 days even if one works full shifts in a location at the vall of the building. Similarly, accumula-ted radiation is well below tolerance in the control and turbine buildings, and also, access to the auxiliary building and at the borated storage tank is possible for timed intervals without exceeding tole.ance levels.
Although the ce7 trol center (room) is at a higher elevation than the dose points shown in Figure 7.3-1, the radiation inside the center will be the same as at point D, and the radiation due to ingress or egress from the center vill be the same as at point G.
Based on the answer to Question 8.1 an additional 50 mrem could be received in the control center due to the noble gases. Also, 4
at dose locations A (directly above dose location E), E and F an additional dose rate of about 1 mrem /hr could be expected due to the radiation levels in the auxiliary building. The total accu-mulated dose at all points would still be well below tolerance.
1 0027 C,)
7.3-1
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B C
D E
F G
1.
At the instant of LOCA (dose rate - mrem /hr) 10 8.5
- 8. 5 ~10-5 6.8 6
7.5 2.
Two hours after LOCA (dose rate - mrem /hr) 9.6 8.2 8.2 ~10-5 6.5 5.7 6.2 3.
Two hours total dose -
(mrem) 20 17 17 ~ 10-5 13.6 12 13.1 h.
One week after LOCA (dose rate - mrem /hr) 516 k.8 h.8 ~10-5 3.8 3.h h.0 5.
One week total dose (8 hrs / day shift) 312 265 265
~10-3 212 187 20h 6.
Ninety days after LOCA (dose rate - mrem /hr) 3.1 2.6 2.6
~10-5 2.1 1.9 2.2 7.
Ninety day total dose (k shift / day basis -
mrem) 2740 2325 2325
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Docksts 50-302 and -303 Supplement No. 1 February 7, 1968 QUESTION Provide an outline of the pre-operational testing of the engineered i
7.h safety features that vill ensure that design criteria have been met or exceeded.
ANSWER Engineered Safety Features of the plant vill be tested as part of the overall precritical test program. The following describes the Engineered Safeguards portions of planned tests.
1.
Hydrostatic or Pneumatic Tests All fluid systems including safeguards portions of systems vill be pressure tested in conformance with code requirements.
Included in the Hydro Tests are leak tests of building isolation valves for the systems served.
Where multiple isolation valves exist each vill be leak tested.
2.
Instrumentation Pre-Operational Calibration Tests Instruments, controllers, components (pover operated valves) vill be tested in response to normal and safeguards signals.
Tests vill include logic, interlock, automatic functions, and over-ride of normal functions by safeguards actuation. Instru-ment strings will be calibrated for correct response from sen-sor to control or readout.
3.
Electrical Tests Tests will be performed on motors for safeguards pumps, valve controllers, valve actuators to verify correct direction of ro-tatt.on, starting characteristics and running characteristics.
Correct phasing from all power sources will be verified, k.
Functional Tests Functional tests vill be performed to verify that systems or equipment are capable of performing the function for which de-signed.
Safeguards functions such as starting sequence, and valve se-quence are included in system functional tests.
Functional Tests will include flow through installed system test lines.
o 0030 7.h-1
5 Queration Test g
This test verifies operation of systems under design or simu-lated design conditions including safeguards operation.
It is not applicable where design conditions constitute destructive testing such as flooding of the Reactor Building.
6.
Integrated Safeguards Test This test demonstrates operational sequence of active components in emergency core cooling, Beactor Building cooling, and Reactor Building isolation syste=s in response to actual or simulated signals.
Systems with installed test lines vill be operated with flow through test lines.
Isolation system vill operate.
High pressure injection vill take place into the RC System for sufficient time to demonstrate injection of water.
Emergency power sources and sequence of starting vill be demon-strated as part of the Integrated Safeguards Test.
7.
Emergency Power and Diesel Loading Test Capability of engineered safeguards busses, power sources, including diesels to start and support safeguards loads, vill be verified prior to fuel loading.
g Detail system test requirements vill be prepared for all safe-guards systems when the systems detail design are finalized and the specific system components are identified.
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7.h-2
Docksts 50-302 and -303 Supplement, No.1 February 7, 1968 QUESTION State the means by which safety-oriented design or construction 7.5 changes vill be implemented for the time period after construction permit but before operation. Outline the internal review process and the decisional line of authority.
ANSWER The organization of the PSAR follows as closely as possible the AEC's Guide announced in the Feddral Register on August 16, 1966.
As the plant design progresses from concept to final detailed design, the plant description and analyses vill be subject to internal review and if approved by the FPC Nuclear Project Manager changes in design and construction vill be incorporated. Any such approved changes to the plant criteria or reference design vill be handled, as appropriate, by up dating or supplementing Safety Analysis Report pages to the AEC with the intended results, during the period questioned, being a Final Safety Analysis Report acceptable to the AEC for the operating license procedure.
GAI is responsible by contract to FPC for the overall plant design ena;ineering for this project and for the original reference plant design incorporating the B&W NSSS presented to the AEC up to the construction permit issuance. All suggested changes relating to safety, regardless of origin, vill continue to be subjected to review by the FPC nuclear project management staff and technical consulting group.
C8 Recommendations furnished by technical consultants shown as a part of the GAI project function are subject to review by the FPC nuclear management staff, as well as its technical consultants, prior to incorporation into the plant design.
As shown on the organizational chart, FPC's Nuclear Project Manager has been designated by the Company's Board of Directors as having complete authority in the management of the Crystal River Plant design and construction, subject by direct line of authority to the Company President.
The FPC Nuclear Project Manager and Chief Mechanical Engineer responsibilities are both held by a single person who is final authority for both design and construction decisions.
The internal review process and decisional line of authority proceeds from the bottom to the top of the Nuclear Project Organization Chart shown graphically on Figure 1-12.
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Docksts 50-302 and -303 Supplement No. 1 February 7, 1968
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8.0 SAFETY ANALYSIS QUESTION Provide an analysis of the whole body dose rate from noble 8.1 gases in the control room following the design bases accident.
ANSWER The control room whole body dose from noble gases during the 30 day period following the design bases accident is estimated to
.05 Rem. This calculation assumes the following:
1.
Release of Noble Gases to the Reactor Building is as per TlD-lh8hh (See PSAR, Page 14-54).
2.
Reactor Building leak rate is 0.25 percent for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 0.125 percent per day thereafter (PSAR, Page 14.52).
3.
Atmospheric dilution takes place according to the site dispersion factors given in Table 2-3 of the PSAR. Val assum d for%/Q are 1 x 10-3 sec/m3 (0-2h hrs), 2 x lo ges sec/m (24-72 hrs) and 6.9 x 10-5 sec/m3 (3-30 days).
h.
The control building ventilation system vill automatically switch to full recirculation following the accident by closure of isolation dampers.
In-leakage (and out-leakage) of air past standard isolation dampers is 1 to 5 percent of the recirculation flow of about h0,000 cfm. For the purposes of this calculation, a leakage rate of 5 percent or 2000 cfm is assumed. Total air volume being recircu-lated is 243,000 cubic feet.
5.
Occupancy of the control room is based on four shifts so that any individual vill spend approximately 7-l/2 days in the control room during the 30 day period of the accident.
The noble gas concentration in the control room as a function of time has been calculated under the above assumptions and an integrated whole body dose of.2 Rem was obtained by converting to dose rate and integrating over 30 days. The whole body dose to an individual is one quarter of this or.05 Rem.
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Dockets 50-302 and -303
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Supplement No. 1 L
February 7, 1968 QUESTIOU 8.2.
Submit a thermal performance of a control rod assembly (CRA) follow-ing a LOCA.
Include energy deposition rates, heat transfer modes, melting point of control alloy, and cladding performance following fusion of control alloy.
ANSWER The thermal performance.of a CRA following a double-ended 36" ID pipe rupture has been analyzed. A very conservative approach was taken in this analysis. In spite of the fact that the etre flow is as high or higher than normal core flow during the first 2 seconds following the rupture, normal steady state cooling of a control pin is assumed. In 2 seconds, the core power is essentially down to de-cay heat levels. Therefore, the following assumptions were made:
1.
The average core power after 2 seconds is 8% of full power and and remains at this level.
2.
All decay heat is absorbed in the core and 50% of the decay heat is in the form of gamma rays available for absorption in the con-trol rod. By ratioing the control pin density to the average core density, an average energy deposition rate in the control assem-blies of 7.87 vatts/cc was obtained.
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3.
The maximum activation product energy in the control pin itself was estimated to be 2.77 watts /cc.
4.
The highest energy deposition rate was assumed to be the average times the ratio of peak-to-average power or 33.52 watts /ce.
5 An adiabatic heatup of the control pin with a heat rate of 33.52 vatts/cc was assumed until the water level reached the point in the core at which the highest peak-to-average power occurs.
6.
The temperature of the control alloy is s650 F at the time the pin is assumed to be insulated (2 seconds).
Using the above assumptions, the average temperature of the Ag-In-Cd goes up to 1007 F at 17 seconds at which time the water level in the core reaches the elevation of this assumed hot spot on the control 4
pin. The temperature of the pin would then rapidly decrease.
Since the melting point of the alloy is %1472 F, a margin of h65 F exists between the conservatively calculated maximum temperature and the melting point of the alloy. The lowest temperature for an eutec-tic formation is that for Zr-Fe which occurs at 1710 F.
Therefore, the integrity of the control rod assemblies is maintained during and following a loss-of-coolant accident.
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8.2-1
Dockets 50-302 and -303 Supplement No. 1 i-Febrevy 7,1968 QUESTION Provide additional details on the restrictive device on the crane 8.3 that limits the height of fuel elements during a fuel transfer.
ANSWER All spent fuel transfer operations are made using the fuel transfer bridges. The handling mechanisms mounted on these bridges employ a telescoping mast which is raised and lowered by a deck vinch. The upper limit of lift will normally be governed by a limit switch.
Should this switch fail, the mast vill travel only a short distance before hitting a positive mechanical sto.r..
This vill cause an over-load condition which will stop the hoist motor. Should the overload cutoff fail, the hoist motor vill stall.
It is not possible to raise the fuel to an unsafe position with the equipment supplied. Note that building cranes are not used for irradiated fuel transfer either in the reactor building or in the spent fuel ctorage area.
8.3-1
Dockets 50-302 and -303 Supplement No. 1 O
February 7,1968 QUESTION Provide a flov loss analysis based on this sequence:
8.h 1.
The reactor is operating at Th% power; four primary pumps are aperating.
2.
A pump with a failed anti-reverse rotation device trips.
3.
The other pump on the same steam generator sends flow back through the failed pump.
4.
Primary flow draps, perhaps below 50% (provide complete details of your analysis).
5 Reactor power stays at Th% (as no trip will detect this incident).
Provide the minimum DNBR vs Time, number of rods, if any, going through DNB, actions by the ICS, and comparison of this incident to the PSAR expressed thermal design criteria for flow loss. Indicate the relative merits of the reverse rotation device. What means are available to indicate to the operator that this incident has occurred?
What would terminate this incident? Is reactor protection sensitive to the rotational inertia of the pump-motor combination? Consider other flow - loss combinations also as related to the thermal protec-tion criteria.
ANSWER a.
Section h.3 7. page 4-25, of the Crystal River PSAR lists the reactor flow rate following loss of one reactor coolant pump.
Section 7.1.2.h lists reactor trip levels for the pump combina-tions.
The reactor flov listed in Section 4.34 of the PSAR is the mini-mum va3ue resulting from runaway reverse rotation of one reactor coolant pump without power. The thermal condition reflects a DNBR equal to or greater than that for loss of all coolant flow.
Figure 8.h-1 shows the flow response in the reactor coolant sys-tem following loss of electrical power to one reactor coolant pump. Table 8.h-1 below lists equilibrium flow conditions for locked rotor (zero rpm) and for a free rotor at equhibrium re-verse rotation (-780 rpm) for one idle pump.
o, 0036 a
8.4-1
(' -
(1) The reactor is tripped if the power was above 75% prior to
\\
loss of one pump.
(2) The unit comes to steady state conditions if the power was below 75% prior to loss of one pump.
f.
Reactor protection is related to inertia of the fluid and the pump-motor combination when loss of all coolant pump power is considered. The rotational inertia of the pump-motor combina-tion was selected to provide protection in excess of that re-quired by the thermal protection criteria.
(e.g., the thermal protection criteria calls for maintaining a DNBR of 138 for loss of all coolant pumps starting at 100% power. Figure 14-20 Crystal River PSAR shows that initial power may be as high as 112% and maintain the DNER at 1.38.)
With respect to assumed loss of flow conditions involving less than all reactor coolant pumps, the protective system senses the electrical conditions of the pump motors and steps the reactor trip level to the steady state power level consistent with the thermal protection criteria for that pump combination. If the power is greater than this value, reactor trip occurs. This mode of protection is therefore not sensitive to rotational in-ertia of the pump-motor combination.
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.120
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REACTOR FL0s
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=
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8 8
12 I4 18 18 20 22 24 28 28 30 32
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- 1 1
.. 0039 REACTOR COOLANT SYSTEM O
rt0i DiSTRi0uTi0s OuRina LOSS OF ONE PUMP 1
FIGURE 8.4-1
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Dockets 50-302 and -303 Supplement No. 1
(')
February 7, 1968 W
QUESTION ' Indicate the size of break in the primary coolant system such that 8.5
' the nomal makeup system could maintain volume and a normal shutdown cooldown could be achieved.
ANSWER The nomal makeup rate is approximately h0 gpm. Assuming that the system pressure remains at its normal value of 2,200 psia, the sys-tem volume can be maintained with a leak area which is equivalent to'a hole 0.25 inches in diameter.
However, by valving off flow to the pump seals, system volume can be maintained with a leak area corresponding to a hole 0.60 inch in diameter.
O t
'. - O 00'40 a.s- -
+
Docksts 50-302 end -303 Supplement No. 1 February 7, 1968
/N QUESTION Provide an analysis of chronic iodine release through the combin-d 8.6 ation of 1 percent failed fuel, leaky steam generator tubes, and leaky safety valves. Include justification for assumed leak rates.
Indicate your applicable concentration limits at the site boundary.
How will the release be monitored so as to stay within applicable limits.
ANSWER Figure 8.6-1 shows the estimated gaseous concentrations at the site boundary resulting from various modes of operation. The release of iodine via the steam safety valve leak does not appear to be a major contributor to the site boundary concentrations.
The principal effect is due to release of noble gases through the condenser vacuum pump exhaust. For example, a 10,000 lb/hr safety valve leak would only increase the site boundary concentration from 0.8 MPC (due to condenser off-gas alone) to 1.0 MPC (due to condenser off-gas plus iodine and cesium released with the steam leak). These calculations are based on the following assumptions:
1.
One percent failed fuel as the design basis in the reactor coolant system.
2.
One steam generator leaking up to 10 gpm.
3.
Steam safety valve leakage up to 25,000 lb/hr at the leaking steam generator. A steam safety valve leak of 25,000 lb/hr is recommended by the manufacturer as an upper limit on valve leakage, which could result from a badly scored seat, 6
h.
Steam flow from one steam generator is 5 x 10 lb/hr.
5.
Atmospheric dispersion is calculated using the long-term T/Q value of 1 x 10-6 sec/m3, 6.
Site boundary concentrations are expressed as the fraction of 10 CFR 20 limits (Appendix B Table II).
As far as plant monitoring is concerned, it is estimated that the condenser off-gas monitor, with a sensitivity of about 2 x 10-6 pc/cc for Kr-85, will provide an alarm far in advance of the de-velopment of excessive fuel failures and/or steam generator tube failures. For example, operation with 1 percent failed fuel and 1 gpm steam generator leak is estimated to produce a Kr-85 concen-tration of about 0.07 pc/cc in the condenser off-gas, which will be well in excess of the alarm point. It is therefore concluded that ample time vill exist to institute corrective measures.
-U 0041 8.6-1
ANNUAL CONTRIBUTION FROM OTHER SOURCES OF GAS: 0.025 MPC i.1 10 GPM STEAM GEN. LEAK 10 CFR 20 1.0
=
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Dockets 50-302 and -303 Supplement No. 1
~'
b' 'i Febru.ry 7, 1968 QUESTION Provide an analysis for the accident involving the double-ended rup-8.7 ture of one steam generator tube. To be consistent with other loss-of-coolant accidents, e.ssume coincident loss of off-site power (and thus loss of circulating water to the condenser).
ANSWER Failure of a steam generator tube is not expected to occur as a re-sult of a loss of off-site power because of the safety margins used in the design of the steem generators. However, an analysis has been made of the environmental effects of a stean generator tube failure coincident with loss of off-site power and circulating water to the condenser.
In this postulated series of accidents, it is necessary to cool the reactor by venting steam to the atmosphere. If the reactor has been operating with 1 per cent failed fuel, then fission products will be released to the atmosphere until the steam generator containing the ruptured tube can be isolated. For this analysis, it is conserva-tively assumed that leakage continues until the reactor coolant tem-perature falls to 200 F, a time period of about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Below 200 F, release of gaseous fission products from the reactor coolant is governed by the gas / liquid partition coefficient and is essentially terminated. Above 200 F all of the gaseous fission products in the leaked coolant are assumed to be released to the atmosphere. At O
normal reactor coelant system temperature and pressure, a double-O ended tube failure produces an initial leak rate of h35 gpm. The leak rate is conservatively assumed to remain constant for the 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, but in fact, it must decrease because of a decreasing pres-sure differential.
Fission product dispersion in the atmosphere is assumed to occur as by the 2-hour dispersion model. A breathing rate of 3.47 predigteg/seeisassumed.
x 10- m The total doses at the hh00 ft exclusion distance are lh8 rem thyroid and 0.78 rem whole body, both of which are well below the 10 CFR 100 guideline values. Therefore, it is concluded that this accident does not produce-an undue hazard to the general public.
0043 k
g 8.7-1
,_ i Dockets 50-302 and -303 i\\~#
Supplement No. 1 February 7, 1968 QUESTION During the course of a major LOCA the sump water may be hotter than 8.8 the building atmosphere.
A.
Provide justification for your position that the iodine vill be picked up in the circulation spray under this condition.
B.
Also indicate your plans for spray water cooling, should your research indicate the need.
ANSWER A.
The temperature of the spray water may exceed that of the reactor building atmosphere after 1,500 seconds following an MHA. This condition occurs when three cooling fans and two sprays are as-su=ed to be operating. When this condition exists, the hot spray tends to evaporate. The maximum evaporation effects occur at about 7,300 seconds when a dynamic equilibrium is reached in the reactor building. At this time the maximum thermal gradient ex-ists between the spray drop and the ambient atmosphere. To evaluate the effect of evaporation on the iodine removal effici-ency, the amount of evaporation from the spray drop was calculated.
The amount of spray evaporation is dependent upon (1) the differ-
,~_ )
ence in temperatures between the spray and the atmosphere, (2) the enthalpy difference between the spray drop and the ambient
(\\"'
atmosphere and (3) the initial size of the drop.
Evaporation and condensation are calcalated on the same basis, which is based on accepted methods. (3,5)
Evaporation and con-densation are reversed directions of the same mass transfer phe-nomenon, i.e., the mass transfer for both evaporation and con-densation for the same T is thus eaulvalent in size, but in opposite directions.
Table 1 compares data for both condensation and evaporation cases.
Diameter Mass Ta - Ts,
- Change, Change, Case Condition F
1 Maximum Evaporation
-h5
-1.6 h6 2
Evaporation @ 2,200 seconds
-24
-0.8
-2.h 3
ORNL Hot Spray Test (Run 28) 18
<0.5
=2 h
Maximum Condensation on the Spray Drop 190
=6
=20 g3 5
ORNL Cold Spray Test (Run 32) 190-200 5-6 18-20 LJ 0044 8.8-1
~
!As indicated in cases 3 ana 5 (Table 1), ORNL conducted spray
! ests (1,2) into a steam atmosphere (at 266 F) with both hot and t
cold spray solutions. The respective spray temperatures were 2h8 F and ambient. The measured iodine removal rates were very high (As = 55 and 59 hr-1 for cases 3 and 5, respectively) and they agreed within the range of experimental varia*, ion. Moreover, both tests were in good agreement with the calculated rates which as-suced no condensation or evaporation effecta.
The mass transfer in case 1 which represents the greatest poten-tial effect due to evaporation is approximately 1/5 of the mass transfer in case h which represents the greatest potential effect from condensate film formation on the drop. The mass transfer in case L is almost equal to the mass transfer in case 5.
As noted, no significant change in iodine removal was observed due to con-densation. Figure A graphically illustrates the relative im-portance of the condensation effect on the rate of diffusion of the iodine. By comparison of these results, it is concluded that neither condensation on drop nor evaporation of drop will produce a significant change in the iodine removal efficiency of the so-lution.
In addition, the heat transfer and the evaporation from the 1000 micron drops vill reach equilibrium in about the first 10 ft of fall, and thus the remaining 90% of the fall height is available for iodine absorption without evaporation effects.
In addition, the radiciodine vill have been effectively removed by the time the spray temperature exceeds the reactor building temperature. The maximum rate of removal is assumed since full operatignofthespraysisassumedinthiscase ind the A is s
>25 hr. This is adequate to reduce the radioi; dine by a fac-ter of >103 within 1,000 seconds of the release. Recirculation of spray solution through the core occurs before the evaporation case arises. This will scavenge the iodine as it is released to the coolant and thereby prevent delayed releases.
The iodine is retained by the spray solution. Tests conducted in Gernany (3) measured the tendency of 131 I to escape from various solutions under several modes of evaporation. This included evaporation to dryness. No iodine loss was detected from solu-tions containing sodium thiosulfate in concentrations as low as 10-5 M.
Preliminary tests conducted by B&W at the Nuclear Development Center indicated that iodine is retained by very dilute solutions of thiosulfate which were depleted with iodine addition, then subjected to irradiation.
ANSWER 9
Although cooling of the spray water is not anticipated, should the research inaicate the need for cooling, it will be provided.
In the event that adequate data is not available from other pro-3 grams with regard to the effectiveness of iodine removal by hot spray water in a cool atmosphere, sufficient tests will be per-8.8-2 (Hevised 3-1-68) 004D
formed by BLW to verify that the necessary information is made' 3
available.
-.teferences:
(1) Parsley, L. F. and'Franzreb, J. K. " Spray Program in the Nuclear 3
Safety Pilot Plant," ORNL-TM-2057, November 27, 1967 i
'(2). Parsley, L. F. and Franzreb, J. K. " Spray Program in the Nuclear Safety Pilot Plant," ORNL-TM-2095, January 1968.
(3) Moldenhaver F. and Uhlig, B.. " Activity Losses During Handling of Aqueous 131 I Solutions" Bulletin of the Federal Center for Radiation Protection, Berlin-Friedlichshagen (translated from 1 -
German for B&W Co.).
(h) Bird, Lightfoot, and Stewart Transport Phenomena, John Wiley and Sons, Inc., New York (1960).
(5)
Ranz,-W. E. and Marshall, W. R. " Evaporation From Drops," Part
^ I - Chem. Engr. Prog., bl, No. 3, pp 1hl-146,. March 1952; Part II - Chem. Ener. Prea., bl,No. 4, pp 173-180, April 1952.
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Docksto 50-302 cnd -303 Supplem:nt No. 1 February 7, 1968 QUESTION Expand your spectrum-of-breaks analysis, for response of ECCS to 6.9 a less-of-coolant accidents, for break sizes less than 0.h ft2 ANSWER The analysis presented in the PSAR indicates that larger leak sizes impose the most stringent requirements on the ECCS. The analysis also shows that the core hot spot is neven uncovered for leak sizes less than 1 ft2 For all leak sizes examined, down to 0.h ft2, the system pressure decreased below 600 psig so that the core flood tanks start discharging borated water into the reactor vessel.
For smaller leak sizes, system depressurization is slower which in turn results in a longer period of cooling by the residual reactor system coolant plus the high pressure injection coolant prior to the tLae that the core ficod tanks start discharging into the reactor vessel.
For rupture sizes down ?o approximately h inches in diameter, calcu-lations show that the system pressure vill decrease belov 600 psig so that the core flood tanks vill begin disenarging into the reactor vessel and the core is never uncovered.
Calculations were made for a rupture size of 3.25 inches (0.0575 ft )
2 in diameter, which indicate that the pressure could hold up above 600 psig for a short period of time until the core heat decayed to a lower level. However, the analysis shows that the high pressure injection system vill keep the core covered even if the pressure remains above 600 psia. Table 1 gives a sunmary of events for this qj break.
Preliminary calculations indicate that core cooling can be adequately maintained without the action of the core flood tanks for leaks sizes of 6 inches (0.186 ft2) and smaller.
However, no credit is taken for core protection by the action of the high pressure injection system acting alone for break sizes above 3.25 inches (0.0575 ft2) in diameter.
Table 1 Loss-of-Coolant Accident 3-1/4" ID Rupture (One H.P. Pump)
- Time, Pressure, see psia Level h5 1,325 Top of Loop (Pressurizer Empty) 660 1,100 Ihaetor Vessel Nozzles 1,260 600 0.3 ft Abt 7e Top of Core (Minimum Level)
U 0048 8.9-1 i;
4
Dockets 50-302 and -303 Supplement No. 1 February 7, 1968 Ab QUESTION Provide an anclysis of the dose consequences at the exclusion boundary 8.10 for the refueling accident. Use as input parameters:
56 fuel rods damaged (from PSAR), 20 percent of noble gases released,10 percent of fuel rod iodine inventory released, 90 percent iodine retention in the pool water, and 90 percent iodine retention on the charcoal adscybers in the building exhaust.
ANSWER The environmental consequences of the fuel handling accident postu-lated above have been calculated. At the exclusion distance the total integrated thyroid dose is 39 rem and the whole body dose is 2.7 rem.
l3 A review of the literature indicates that the 90 percent iodine re-tention is not supported by experimental data (Ref.1 through 7).
Typical results are those of Stinchcombe and Goldsmith (1) who dem-onstrated that at iodine concentrations similar to those which would occur in the refueling accident, the water-to-air partition coef-ficient is 105 to 106 instead of 70 as predicted by Henry's law coefficients. Based on the referenced data, it has been concluded that a water-to-air partition coefficient for iodine of 103 to 10h may be expected. However, to conservatively estimate the environ-mental effects of the accident a partition coefficient of 100 has been assumed for the analysis that follows.
SLnilarly, it is believed that the fissicn products released to the
(/'}
vater from the damaged fuel rods will be the activity within the w
fuel rod gap which has been released by the fuel. This activity has been calculated using experimentally determined escape rate coefficients.
This calculation procedure has previously been described in detail in the Metropolitan Edison PSAR (8).
In this calculation, no credit has been taken for the reduced fuel rod te=perature at the time of the refueling accident, thus overestimating the amount of fission products released.
Using the above procedures, the environmental consequences of the refueling accident have been calculated. At the hh00 ft exclusion distance, the total integrated thyroid dose is 0.k2 rem and the whole body dose is 0.42 rem, both well within the guidelines of 10 CFR 100.
References:
(1) Stinchcombe, R. A. and Goldsmith, P. G. " Removal of Iodine from the Atmosphere by Condensing Steam," Journal of Nuclear Energy, Parts A/B, Vol 20, pp 261-75 (1966).
(2) Watson, 8017~et al. "Iedine Containment by Dousing in NPD-11," AEC-1130 (19 (3) Klepper, D. H. and Bell, C. G. "Under Water Caisson Containment of Large Reactors," ORNL h073, June 1967 OV 0049 8.10-1 (Revised 3-1-68)
(4) Nippon, G. Journal of the Atomic Energy Society of Japan, Vol h
7,1965, pp 563-69, HP-TR-1537.
(5) Barthoux, A.,
et al.
" Diffusion of Active Iodine Through Water with the Iodine Being Liberated in CO 2 Bubbles at High Tempera-ture," AEC-TR-61hh, June 1962.
(6) Styrickovich, M. A., et al. " Transfer of Iodine from Aqueous So-lutions to Saturated Vapor," Soviet Journal of Atomic Ene.,o,
r Vol 17, June-Dec. 1965.
(7) Diffey, H.
R., et al. " Iodine Clean-up in a Steam Suppression System," AERE-R-4882, May 1965 (8) Metropolitan Edison Co., Three Mile Island Nuclear Station Unit 1, Preliminary Safety Analysis Report, Supplement 2, November 6, 1967, Docket 50-289.
O 0u50 9
8.10-2
Dockets 50-302 and -303 (q
Supplement No. 1
/
February 7, 1968 QUESTION Provide a thermal shock analysis for the response of the pressure 8.11 vessel to action of the ECCS. In particular, provide for the frac-ture mechanics approach.
8.11.1 What is the critical stress intensity factor being used in your analysis?
8.11.2 What are the initial crack size and the geometry of the crack assumed in your analysis?
8.11.3 What is the equation used to correlate the crack size with the stress intensity factor?
ANSWER The thermal shock analysis for the reactor vessel due to rapid tem-perature changes at the end of its life was presented in the answers to the Metropolitan Edison Company's questions submitted in Supple-ment No. 2 of Docket 50-289 The specific information vill be found in the answer to Question 11.1 of the reference above. The fracture mechanics analysis as presented on page 11.1-2 of Supplement No. 2 is amended by a technical approach as described in the answers to the questions above.
(~]
8.11.1 The critical stress intensity factor being used in our anal-N/
ysis for evaluation of the thermal quench problem is 30 Ksi in.h.
8.11.2 The acceptance standard for nondestructive examination of detrimentally orientated indications in plate is a notch with a 60-degree bevel 1 inch long by 3 per cent of the plate thickness. This means that no larger indication could exist below the surface of the vessel vall when shipped. No lin-ear surface indications vill be present on the vessel vall.
The complete vessel vall is magnetic-particle-inspected be-fore shipment, and the acceptance criterion for this in-spection is that no linear indication is acceptable. Indi-cations found during inspection are ground out, and the plate is veld repaired if necessary. Therefore, the surface of the vessel vall is shipped in an indication free state, i.e., no crack vill exist on the surface of the vessel vall.
8.11 3 The method and equations to be used in evaluating the crack size with the stress intensity factor are as postulated by Dr. P. C. Paris. This method is based on a plain strain two-dimensional approach utilizing superposition of pure tension and vedge force with correction factors for the plastic and surface effects.
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Docksts 50-302 and -303 Supplement No. 1 February 7,1968
(]
9.0 SEISMIC DESIGN ys QUESTION The foundations for the proposed units receive considerable atten-9.1 tion in the PSAR. It is noted on page 2-1 that the principal structures vill be founded on limerock. On page 2G-8 it is stated that the Reactor Building mats or spread foundations will rest on lean concrete or grouted structural fill replacing incompetent rock or soil above the base co=petent rock. In view of the nature of the cap rock and overlying sediments, as well as the discussion of bedrock solutior. studies, a detailed description of the locations and types of foundstions to be employed for the containment vessels and auxiliary buildings is required.
The additional information supplied concerning the foundations should include a discussion of the steps taken to preclude differential settlement and tilt under both static and dynamic loading conditions.
ANSWER The foundations for all major structures will consist of either cats or spread footings. The foundations for both the Reactor and Auxiliary Buildings are structural mats with the folleving approxi-mate dimensions and elevations :
Reactor Building Auxiliary Building 1.
Thickness 12 5' 5'
T 2.
Plan dimension 1h7' dia.
220' x 185'
~'%
3.
Underside elevation -
80 ' -6" 88'-0" (V
Prior to placing these foundations, all surface fills, natural 3
sediments and any underlying highly-decomposed lime rock through-out the plant area vill be excavated in the dry. In localized deep lying areas where complete devatering may not be practical, subaqueous excavation may be required. Where the excavation can be kept dry, mass concrete backfill vill be placed to the under-side of the foundations. For the deep lying flooded areas, higher strength cass concrete vill be tretied into the area and up to the underside of the foundation, or alternatively, Brooksville limestone may be placed subaqueously and subsequently consolidation grouted. Sandfill in lieu of limestone vill be placed where the deep components such as tendon galleries and heat su=pa are located. Access fill vill then se placed on top so that the consolidation grouting can be performed, after which the fill vill be stripped. The Reactor Building =at will be constructed on the exposed mass concrete backfill. Placement and conpaction procedures will be controlled.
h 0052 1
v 9.1-1 (Revised 7-15-69)
.=
DockI;ts 50-302 and -303 Supplement No. 1 February 7, 1968 Any settlement derived from compression of the foundation materials g
is expected to be primarily elastic, occurring almost instantaneously with load application. Further tests are being conducted to confirm the conservatism of the foundation modulus used in the preliminary analysis of the Containment Vessel.
When final properties of in-situ rock and structural fill are known, an analysis will be made of the effects of the structure-soil interaction using the methods suggested in " Building-Foundation Interaction Effects" by Richard A. Parmelee, Journal of the Engineering Mechanics Division, Proceedings of the American Society of Civil Engineers, April 1967 Based upon assumed soil properties, a preliminary analysis indicates that the change in natural fre-quency of the Reactor Building vill be small.
O 0053 l
9.1-la (Revised 7-15-69)
1 Dockets 50-302 cnd -303 Supplement No. 1 February 7, 1968 QUESTION With reference to the table of damping values on page 5A-3, confirm 9.2 that these values will be used in the analysis for both the design and maximum earthquake conditions.
ANSWER The damping values which will be used for the maximum ead hquake will be the same as those used for the design earthquake and are as stated in revised Section 5 of Appendix 5A, " Structural Design Bases."
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0054 9.2-1 i
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Dockets 50-302 and -303 Supplement No. 1 February 7, 1968 i O QUESTION From a description of the proposed design, it appears that the contain-93 ment structures will undergo rocking on their foundations. What value of damping is to be employed fc,r the rocking of the structure on the j
foundation for both the design and maximum earthquakes?
I ANSWER The rocking effect analysis vill be based on " Building Foundation Interaction Eff.ects," Parmelee, Journal of the Engineering Mechanics Division, ASCE, No. D42. A damping value of 5 percent of critical will be used for both design and maximum earthquake.
i O
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9.3-1
Dockets 50-302 and -303 Supplement No. 1 February 7, 1968 QUESTION The method of analysis to be employed in the seismic design is 9.h described briefly in Appendix SC. A more detailed description of the method of analysis is required, and should include a description of the manner in which the containment structure is modeled for the analysis.
ANSWER The dynamic solution for the containment vessel including the manner for developing the analytical model is described in revised Section 2.2 of Appendix 5C, " Dynamic Golution."
AV 0056 i
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9.h-1 t --
Dockets 50-302 and -303 Supplement No. 1 February 7, 1968 OV QUESTION On page SA-3 it is stated that the vertical a d horizontal components 9.5 (of seismic motion) are assumed to occur simultaneously and their effects added algebraically. It is recommended that the effects associated with the horizontal and vertical earthquake excitation be added directly cnd linearly as appropriate for the item under consideration, and moreover added directly to the applicable dead-load, liveload and operating loads. Clarification of the manner in which the seismic loading effects vill be combined with other loadings is required. The load factors for the various inputs should be defined for all components.
ANSWER Further explanation of the combination of vertical and horizontal seismic components is given in the revised Section 6 of Appendix SA,
" Structural Design Bases."
l The combination of seismic load factors with other load factors is stated in revised Section 1 of Appendix 5B, " Design Program For Reactor Building."
n v
0057 g
9.5-1 l
i
Dockets 50-302 and -303
-Supplement No. 1 February 7, 1968 O
QUESTION On page 5-16 of the PSAR there is an indication that the tendons 9.6 vill be grouted. No other discussion of this point was found in the PSAR. Further information on the details of grouting and the long term surveillance program for the prestressed tendons is re-quired.
ANSNER The tendons will not be grouted, but will be protected as described in Section 5.1.2.8.
The long term surveillance program for the unbonded tendons is discussed in Section 5.6.2.2.
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Dockets 50-302 and -303 Supplement No. 1 February 7, 1968 QUESTION The liner is noted in the PSAR to consist of 3/8-inch steel plate 9.7 in the cylinder and dome and 1/h-inch thickness in the base. Ad-ditional information concerning the fastening of the liner is required. Additional information must be provided in the manner in which the liner is to be attached to the shell, the stresses under which buckling may occur, and the design provisions that are made to ensure that the buckling can occur without distress or difficulties that will endanger the function of the lines.
ANSWER Additional information relating to structural analysis of the liner is provided in the revised Section 8 of Appendix 5B, " Design Program For Reactor Buildings." Figures showing liner anchor details were included as part of the revised Appendix.
O 1
(
0059 r
9.7-1
Dockets 50-302 and -303 Supplement No. 1 February 7, 1968 O
QUESTION The provisions for carrying shear in the concrete containment vessel 9.8 are discussed on pages SC-h and 5.
Further discussion is required of the manner in which the code provisions in Chapters 17 and 26 of ACI 318-63 v111 be applied to the containment structure, in view of the fact that the containment structure is not an element of the type for which the code was originally written.
ANSWER A more complete discussion of the treatment of shear in the contain-ment vessel is provided in the revised Section 2 of Appendix 5B,
" Design Program For Reactor Buildings."
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L) 9.8-1
- ~.... _.,..
Dockste 50-302 and -303 Supplement No. 1 February 7, 1968 QUESTION The design of the penetrations receives limited description in the 9.9 PSAR. Provide additional information on the design technique to be employed with discussion as to how secondary effects arising from thermal loadings, secondary bending, etc., will be handled in the design.
ANSWER Additional information relative to penetrations design has been provided in the revised Section 7 of Appendix 5B. " Design Program For Reactor Buildings" and in Section 5.1.2.6.1.
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1 0
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O 9.9-1
Docksts 50-302 and -303 Supplement No. 1 February 7, 1968 QUESTION Cranes in the structure are noted to be Class I components. Additional 9 10 information is required concerning the design provisions that will ensure that the cranes vill not topple during an earthquake or other-vise cause darsge which could endanger the safety of the plant.
ANSWER The heactor Building crane information is provided in the revised Section 2 of Appendix 5A, Structural Design Bases."
The handling bridges (referred to above as cranes) are no longer considered Class 1 equipment (failure cannot endanger safety of plant or public). They are, however, provided with antiderailing devices which are more than cdequate to prevent bridges from being disengaged from track under earthquake conditions. All structural components far exceed earthquake loading requirements due to the extreme rigidity required during operation.
(The conventional factor of safety of 5 used for structural design of hoisting equipment has been exceeded to provide rigidity needed for indexing.) These bridges will be " parked" during plant operation away from the reactor and may be secured in place by auxiliary attachments if required.
O 0062 O
9.10-1
.-. ~
Dockets 50-302 and -303 h] -
February 73 1968 Supplement No. 1 QUESTION With regard to the piping, reactor internals, reactor vessel, and ves-9.11-sel supports, little information is noted in the PSAR cencerning the design of these-items for seismic and other loadings.
For each of these items provide detailed information concerning (a) the loadings that are applicable to the design, and the manner in which the loads are to be combined; (b) the method of analysis to be followed; (c) stress and/or deformation limits for normal operating conditions as well as those conditions involving earthquake loadings; (d) a discus-sion of the basis for installation of snubbers and/or dampers and their locations; (e) the location of critical valves and their design to preclude difficulties under seismic and other loadings.
ANSWER a.
TYPES OF LOADS 1.
Load Combinations Reactor Vessel and Internals - The reactor and internals are subjected to two fundamentally different types of loading due to LOCA. The first is horizontal shaking due to the side thrust resulting from a rupture that occurs in a radial pipe leg. The majority of the thrust is applied directly to the vessel and is transmitted to the internals through the upper p
flange of the core support shield. The reactor responds as
-()
a complex mass vibrating on a spring composed mainly of the vessel support skirt. This mode is illustrated in Figure 9.11-1.
Since the load duration is much larger than the natural period of the reactor, the response is typical of a structure subjected to a suddenly applied load.
Initially, the reactor deflects twice as far as it would under an equal but slowly applied load, and then vibrates between this double deflection and its initial position, at its natural frequency.
(This' description is somewhat over-simplified, since the actual motion vill be modified by reduction of thrust load with time, and damping.) Inertia loads result through out the reactor. Also, since all components do not respond in phase, some impacting may occur between adjacent parts.
The second type of LOCA loading is that resulting from tran-sient pressure differentials which occur across various com-ponents within the reactor. The Tressure differentials are cyclic, generally having the appearance of a damped sine wave n plotted against time. When the period of the pressure-time history is less than or of the same order as the natural period _of the pertinent structure, dynamic effects predomi-nate in the response. On the other hand, when the period of the pressure-stime history is larger than that of the perti-nent structure, the response is predomintnaly static. The pressure-time histories of' interest within the reactor vessel have finite rise times and, therefore, the internals are not L) 9.11-1 i
0061
_m
subjected to the 2:1 suddenly applied load factor discussed above in the section on response of the reactor vessel to side thrust.
Similar loads result from an earthquake. Horizontal ground motion produces horizontal shaking of the reactor, differing only in magnitude and in that the excitation is applied at the junction of the vessel support skirt and the foundation.
The vertical ground motion produces vertical inertial loading within the reactor which has an effect similar to vertical pressure differentials.
Pressurizer - Qualitatively, the pressurizer loads are simi-lar to those on the reactor vessel. The LOCA loads are much smaller because of the smaller pipe size involved.
Steam Generator - Although the loads are similar, the re-sponse of the steam generator to a LOCA is greatly limited (reduced) by a lateral support at the top.
2.
Maanitudes of Loads Reactor Vessel - Preliminary thrust load (vs time) on the re-actor vessel due to a nearby hot leg rupture is shown in Fig-ure 9.11-2.
An earthquake of design intensity will impose a 0.09 g hori-zontal and 0.0336 vertical load on the support skirt. The resulting forces and moment are 190 kips horizontal, 70 kips vertical, and h,200 ft-kips.
Depending on the location of the rupture, one of two possible accident condition loads could occur simultaneously with an earthquake:
1.
A horizontal force of 2,2h0 kips, a moment of 62,720 ft-kips and a twisting moment of h6,h80 ft-kips.
2.
A vertical force of 2,2h0 kips and a moment of h6,h80 ft-kips.
Internals - Final pressure differential t_me histories within Ohh the reactor vessel will not be available until mid-1968.
Final response anc inertial loads will'not be available until later. However, preliminary response and inertial loads due to LOCA and earthquake are expected to be available before mid-1968 and will be sucnitted as available.
Pressurizer - An earthquake of design intensity will impose a 0.05 g horizontal and 0.033 g vertical load on the support lugs. The resulting forces and moment are 20 kips horizontal, O
9 11-2 v
'(v3 15 kips vertical, and 135 ft-kips. One of two possible acci-dent condition loads could occur simultaneously with the earthquake:
1.
A horizontal force of 125 kips, a moment of 2,281 ft-kips and 1 tvisting moment of 1,038 ft-kips associated with a rupture in the horizontal leg of the surge line.
2.
A vertical force of 125 kips and a moment of 1,038 ft-kips associated with a rupture in the vertical leg of the surge line.
Steam Generator - An earthquake of design intensity will im-pose a 0.16 g horizontal and 0.033 g vertical load on the support skirt. The resulting forces are 325 kips horizontal and 70 kips vertical, and for a cold plant a moment of 12,200 ft-kips at the base.
For an accident at hot plano condition, one of three possible conditions of load on the support skirt may occur with earth-quake forces only, no earthquake moment:
(~)
1.
A horizontal force of 1,355 kips.
%./
2.
A twisting moment of 22,h00 ft-kips.
3.
A vertical force of 2,2h0 kips.
An accident condition can cause a horizontal load of 2,540 kips at the lateral support at the top of the steam gener-ator.
Conbination of loads are discussed further in the answer to Sections b & c of this question.
b.
METHODS OF ANALYSIS TO BE EMPLOYED FOR REACTOR INTERNALS AND CORE All-reactor internals and core components (including control rods )
vill be analyzed separately for stresses and deflections result-ing from accidents and earthquakes.
Static or dynamic analyses will be used as appropriate. In gen-eral, dynamic analysis will be used for the subcooled portion of the LOCA and earthquakes, and static analysis will be used for i
the relatively steady state portion of the LOCA.
Dynamic analysis vill include the response of the entire system (as applicable in each case) to the various excitations.
For LOCA, the excitation vill be applied at the appropriate nozzle l
or internals component.
Where appropriate, the response of the
(~N reactor vessel on its support skirt will be used as input to the
(-)
i 4
9.11-3 i
- ~ - -
w
- ~ - - -
internals. The response of the internals will then be used as input to the core. Seismic excitation will be handled in a simi-g lar manner, except that the ground motion will be input at the junction of the support skirt and the vessel foundation.
Lumped parameter simulation will be used generally.
For LOCA, predicted pressure-time histories will be used as input.
For earthquake, actual earthquake records, normalized to the appro-priate ground motion, will be used as input.
Output will be in the form of internal's motions (displacements, velocities and accelerations ), motions of individual fuel assemblies, i= pact loads between adjacent fuel assemblies and impact loads between peripheral fuel assemblies and the core shroud.
Seismic analysis will also be performed using the response spec-tra approach.
The relative timing of the various aspects of a given LOCA will be considered only as indicated by the various local time histo-ries associated with that particular accident, although sensitivity to the time duration of the pulses and other calculated input will be investigated.
Where simultaneous occurrence of LOCA and the PJE is considered, it is intended that both excitations will be input to the system simult aneously.
Relative starting times will be changed until maximum structural motions, indicating maximus stresses, are ob-tained.
Alternately, the maximum stress from one component of the combination will be added to the square root of the sum of the squares of the other components.
C.
STRESS LIMITS AND IN'I'ERNALS DEFOPRATION LIMITS Stress Limits - The loading combinations and the corresponding design stress criteria for the internals, vessel, supports, and piping are given in Table 1.
Each of the four cases of loadin6 combinations vill be discussed separately, and justi-fication for its stress limits will be 6 yen, 1
Case I - Design Loads Plus Design Earthauake Loads - For this, the reactor must be capable of continued operation; therefore, all components includine piping are desi ned to 6
Section III of the ASME Code for Reactor Vessels.(1) This code's applicability and conservatism for these require-ments are well known an need no elaboration.
\\);2 J Case II - Design Loads PDu Two Times Design Earthan=M; Loads and Case III - Design Loads Plun Pipe Rupture Loads - In estab-lishing stress levels for these two cases, a "no-loss-of-function" criterion applies, and higher stress values than 9.11-h
Table 1 Stress Limits for Internals
^
-(
)
Vessel Supports & Piping Case Loading Combination Stress Limits I
Design loads + design earthquake P 1 1.0 S loads P
+P 1 1*5 b (5
b m
II Design loads + two times design P 1 1.2 SM earthquake loads 1 1.2 (1.5 S,)
f5 P
+P III Design loads + pipe rupture loads P 1 1.2 S 5
1 1.2 (1.5 S,)
P
+P b IV Design loads + two timec design P 1 2/3 S earthquake loads + pipe rupture 1 cads P
+P 12/3S b
u P = Primary general membrane stress intensity.
P = Prica:'/ bending stress intensity.
b
/7 Q
S = Allowable membrane stress intensity.
S = Ultimate stress for unirradiated caterial at operating temperature.
P = Primary local membrane stress.ntensity.
{5 y
(1) All symbols have the same definition or connotatien as those in ASME B&FV Code Section III, Nuclear Vessels.
(2) All components will be designed to insure against structural instabilities regardless of stress levels.
(3) The limits given in Table 1 are for primary stresses, general membrane, 5
local membrane, and bending. Based on stress-strain curves available in the literature, the corresponding strains will be about 20 percent of the uniform strain or less. This applies to all pertinent materials in the unradiated conditions.
O 0067 9 11-5 (Revised h-0-66)
n in Case I can be allowed. The multiplying factor of (1.2) has bec' selected in order to increase the code-based stress limits and still insure that for the primary structural ma-g terials; i.e., 304 SST, 316 SST, SA30 3, SA212B, and SA106C, l5 an acceptable Margin of Safety will always exist. To illus-trate this point, two of the primary materials, 30h SST and SA302B, have been selected, and the Margin of Safety has been calculated for each. These two materials are fairly repre-sentative of the others, since one is a stainless steel and the other a carbon steel.
The Margin of Safety (MS) between the design stress limit (S )d and the ultimate stress (S ) is defined as Su-3 MS =
-x 100%.
3d For 30h stainless steel at 600 F
= 2.75 S (2)
Su y
= minimum yield strength.(1}
where Sy For the stated design limits O
Sd = 1.2(1 5 x S,)
where S,= 0.9 S.
y 3
Therefore, J
rs Q\\
Sd = 1. 2 x 15x 09S = 1.62 S y
y and MS = 2.75 - 1.62 100% = 70%.
x 1.62 For SA302B at 600 F
= 1.43 S (2)
Su y
S,= 1/3 Su " 1/3I1* k3) S = 0.h8 S.
y y
Therefore O
9.11-6 (Revised h-8-68)
n
'Sd = 1.2(1 5 S ) = (1.2 x 1. 5 x 0.48) S = 0.86 S m
y y
.(n)s and
.MS = 1.L3 - 0.86 x 100% = 67%
0.86 It is shown below that the 67% margin calculated above vill apply for all materials whose yield strength equals or ex-ceeds 50". of the ultimate strength. Since the yield strength of most " code" carbon steels exceeds 50% of the ultimate, the margin calculated for SA302 grade B may indeed be considered typical.
For Sy _ 50% S ui S"
Sm=3.-
S
= 1.2(1 5 S )
d m
max.
S" - Sd 3Sm - 1.8 S MS =
=
m 100%
Sd j 1.8 S*
['s MS = 1*2 = 67%.
1.8 Where Sy < 50% S, the value of S vill equal 2/3 S, and the u
m y
margin of safety will exceed 67%, as indicated by the calcula-tion for 304 SST above, where Sy = 36% S, and the margin of safety is 70%.
u Margins of safety are required to cover uncertainties in load, structural performance, and material properties.
In view of the detailed and extensive engineering practices and analyses used on these components, and the use of minimum values of ma-terial strength properties, per ASME Section III, a margin of safety of 50% provides adequate conservatism.
(This margin
.till be used for Case IV, below, covering simultaneous occur-rance of accident and earthquake.) When considering Cases II and III, however, since only one of these two severe accident conditions is applied, a slightly higher margin on maximum stress (P,+ P ) is achievable without penalty. From prelimi-b nary work already completed, it is apparent that a much greater margin on membrane stress is possible without penalty, since these stresses are not high. Taking advantage of these condi-tions permits representation of the Case II and Case III lim-its in a familiar form, simply 120% of the code stress V
9.11-T 0069
~
allowables. This latter form has been adopted for present purposes. Future work is expected to justify much higher stresses.
It should also be mentioned that in applying this criterion, elastic equations are used for calculating all stresses. In the case of plastic bending, the maximum normal stress cal-culated by the elastic equation Sb = MxC I
where Sb = maximum normal stress M = applied moment C = distance from neutral axis of S b
I = moment of inertia of cross section about neutral axis will always yield a normal bending stress greater or equal to the true maximum plastic stress. Therefore, ACTUAL MARGINS OF SAFETY will always be greater than or equal to the CALCU-LATED Margins of Safety. Where bending stresses are signifi-cant, the conservatism of elastic formulas is considerable.
(See Appendix)
Case IV - Design Loads Flus Two Times Design Earthauake Loads Plus Pipe Rupture Loads - As in Cases II and III, the "no-loss-of-function" criterion applies.
Also, in the discussion of Caces II and III, it was stated that a margin of safety of 50% was adequate. To insure that for this case the margin of safety will always be greater than or equal to 50%, the design stress level has been se-lected so that Sd = 2/3 Su with a resulting margin of safety:
MS = 3/3 - 2/3 (s ) x 100% = 50%.
u 2/3 Figure 9.11-3, the design limit curves for a hollow circular cylinder of 30h stainless steel at 600 F, is offered as an ex-ample of the type of curves to be used in the limit design for the four types of load combinations.
Curves of this type for the various structural materials with different parameters 9.11-6
g_s offer graphical representation of the stress limits for each
.i 1
case,'and afford the designer a relatively simple ~ visual
)#
check of stress limits. These limits are expressed in the form of " boxes," and a safe design is one where the combina-tion of calculated stresses for a given case falls inside its
-corresponding " box."
This figure is based on S, = 0.2 S.
y Boxes may be obtained for other values of S by moving the m
left side of the box an incremental distance equal to the in-cremental change in S.
3 The Appendix also discusses in detail the effect of using elastic equations in calculating stresses and a method for accounting for the difference between these elastically cal-culated stresses and the true stresses in the case of plastic bending.
In Figure 9 11-3, the so-called " pseudo-elastic" effect of bending is indicated by the dashed-lined curve above the stress limit " box" of Case IV.
However, pending further investigation into this area, B&W vill.not take ad-vantage of this increase in stress levels, but vill restrict its present design to fall within the box itself even-though elastic equations are used. This thus introduces a further degree of conservatism to the present design criteria.
In Cases II, III, and IV, secondary stresses are neglected, since they are self-limiting. Design stress limits in most cases are in the plastic region, and local yielding has oc-curred. Thus the conditicas that caused the stresses can be
,_( )
assumed to have been satisfied.
9,.
n 0071 o
9.11-9
m w-Deformation Linits of Reactor Interna 2s - Two primary safety con-h siderations govern the deformation limits of the internals.
De-fonnation shall not prevent insertion of control rods, nor shall deformation prevent the flow of coolant to the core. The specific deformation limits given in Table 2 represent the limiting deflec-tion of each component listed. Other considerations were not included, since the values given represent the deflection at which a safety limit is first reached. The "no-locs-of-function" defor-mations could cause a safety problem. The " allowable" deformations are those that are used as design limits.
Table 2 Modes of Deformation of Reactor Internals No Loss of Component Safety Implication Allovable Function Core Support Mode 1--Outward deflection of Shield the shell vill reduce the effec-tiveness of the internals vent valves, resulting in increased probability of uncovering the core during blevdown.
(a) Uniform radial expansion 1/h" 3/8" of the shell at the vent valve.
g (b) Outward local radial dis-1/2" 1"
placement of two valves (ellip-tical deformation of the shell).
Mode 2--Invard deflection of 1"
1-15/16" the shell at the nutvard noz-zles to prevent contact with guide assemblies.
b({l b Mode 3--Deformation limit of t,he upper flange to insure that the core support assembly does not drop.
Uniform decrease in diameter.
1" 1-7/8" Mode b--Axial elongation of See Note 1 the core support shield shall be limited to insure engage-ment of the fuel assemblies in the grid plates.
Core Barrel Mode 1--Decrease in diameter 3/L" 15/16" (local or average) to prevent distortion of fuel assembly spacer grids.
lh 9.11-10
Table 2 (Cont'd)
No Loss of Component Safety Implication' Allowable Function Mode 2--Axial elongation of the See Note 1 core barrel shall be limited to insure engagement of the fuel l
assenblies in the grid plates.
Upper Plenum Mode 1--Limit radial expansion 1 5" 3"
Assembly to maintain clearance between the shell and the internals vent valve wedge ring to insure valve operation.
Mode 2--Limit radial compression 2.h" h.8" to maintain clearance between the.shell and upper guide tube structure.
Mode 3--Bending of the cover See Note 1 as a plate shall be limited to insure engagement of the upper grid plate and the fuel assem-blies.
("T Control Rod Mode
-Limit axial deflection See Note 1
(_/
Guide Assembly to insure engagement of the fuel assemblies in the upper grid plate.
Mode 2--Bending as a beam--
(Later) contact between the guide tube and the control rod resulting from deflection of the guide tube must be limited so that the resultant frictional drag on the control rod will be small enough to permit control rod insertion.
Mode 3--Cross-sectional distor-
'0.01h" 0.029" tion of individual tubes shall be limited to maintain clear-
-ance between the guide tubes and the control pins.
Fuel Assembly Mode 1--The cross-sectional 0.013" 0.027" Guide Tube distortion of guide tubes shall
~
be limited to maintain clear-ance between the tube and the 00 8 control pin.
r0
- (.,/
9.11-11
Table 2 (Cont'd)
O No Loss of Component Safety Implication Allevable Function Lover Grid Mode 1--Downward deflection as See Hote 1 Plat.2 Assembly a plate shall be limited to in-sure engagement of the fuel as-semblies in the grid plates.
Thermal Shield No safety implication Flow Distributor No safety implication NOTE 1 The combined axial displacement of the lover grid plate, the core barrel, the core support shield, the upper plenum assembly and the control rod guide assembly shall be limited to prevent disengagement of the fuel assem-blies,from the. grid plates.
d.
SNUBBERS AND DAMPERS The present dynamic analysis of the piping system indicates that shock repressers, if needed, will only be required at the loca-tion of the primary pumps and that no dampers er enubbers will be required.
e.
CRITICAL VALVES The only valves in the reactor coolant system are the reactor internals vent valves. These valves are discussed in the sec-tions above.
0074 0
9.11-12
REFERENCES 1.
ASME Boiler and Pressure Vessel Code Section III, Nuclear Vessels, The American Gociety of Mechanical Engineers,1965 2.
R. Wiesemann, R. Tome, and'R. Salvatori, Ultimate Strength Criteria to Ensure No Loss of Function of Piping and Vessels Under Earthquake Leading, d
Westinghouse Electric. Corporation, WCAP-5890 Revision 1,1967.
3.
W.' Stokey, D. Peterson, and R. Wunder, " Limit Loads for Tubes Under Internal
-Pressure, Bending Moment, Axial Force and Torsion," " Nuclear Eneineering j~
and Design," Vol. k, North Holland Publishing Company, Amsterdam,1966.
I P. Hodge, Jr., Plactic Analysis of Structures, McGraw-Hill Book Company, 4
Inc., 1959, p. 201.
a 4
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s 1
4 1
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+
0075 h,,g i
9.11-13
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w
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--_4m44wa_.w.-
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_.u aa ma m.->._-se ya.
muw.m
.eao s_s ma aa m e m.m na-m-w..e.em.p.4 w-a.a.m_w-
.-.-a.-----_-emA-A--..
--a_.-w-ei I
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APPENDIX g
Method of Determining Allowable Stresses
'r l
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J!'
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+
f t
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007G
! e f
9 ll-lk k
i.
e, y
..,. : l-g-tH
- -=-*A%
M'E
- "9"
Consider a thin-valled hollow circular cylinder subjected to 2 combination of internal pressure, bending moment, torque, and axial force as indicated in g
Figure 4(a). For an element in the tensile region, the triaxial stress state, as shown in Figure h(b), would yield the following principal ctresses:
2 S
+S (S
-S1 c =
+
[
y 2
2 S
g S
+S IS
-S
\\
t m,
t m
2 o =
3 (2) 2 2
2 S
g
= -P/2 3
(3)
It should be mentioned that Figure h(b)(gges not represent a statically admis-sible stress state as defined by Hodge, l since the radial stress has been assumed to be equal to -P/2 throughout the thickness of the cylinder vall, whereas it actually varies from a value of -P at the inner vall to zero at the outer vall. However, because this radial stress is smaller than the others, causing little appreciable effect on failure, this assumption is considered reasonable. With the exception of the radial stress, all other stresses sat-isfy the conditions of a statically admissible stress siate.
In employing the maximum shear theory, use is made of stress intensities. Fcr stresses acting at a point, the stress intensity is defined as the absolute value of the diffe e between the algebraic largest principal stress and the algebraic srallest Or when 2 * '3 (S -Sf 1
t n
2 SI =
c
-c
=2
,g (g) y 2
(
)
when 2
3 S
+S (S -ST SI =
c
+
+S
+ P/2 (5) y 3
2 2
(
)
In Case IV, the maximum allowable value of stress intensity as specified in Table 1 is (SI) max = 2/3 S (6) u O
goll 9 11-15 L
. = _ -.
"U*#*
S = ultimate stress of the material In addition, in the tensile region, the combination of stresses in the B&W re-actor vill yield values of principal stresses so that c2 > 03; therefore, equa-tion (5) is applicable, and combining it with equation (6) yields S
+S (S -SI SI
=
+
+ Sg + P/2 = 2/3 S.
(7) 2 2
y
(
)
Assuming constant values for S and SS and solving equation (7) for the allow-m able tensile stress (S )
we obtain t allow (S )
t,11c, =,(2/3 S,) - P/2,-S
,(2/3S - P/2) - S,,,
(8) u and when S r 0 g
(S )
= 2/3 S - P/2.
(9) g u
Or, if all the etresses are expressed in terms of their ratio to the yield stress of the material (S ), then equations (8) and (9) yield the following 7
normalized equations:
2 g
g -
(S I
(
S T
S
\\[Y A l 2/3 [
2S ]I -
2/3 [
2S (10)
N Y
Y Y
7 7
Y_
g11oy t
and when Sg=0 f[S T
= 2/3 [
2S Ill)
S P
k 7l I
7 allow For an element in the compressive region, as illustrated in Figure h(c), a similar expression for (S /S ) allow can be derived; however, in this region e y components in the present B&W reactors have stress combinations which cause 2<
3' 0
0078 9.11,
i
.m.
J
Therefore, for 2
3 O
2 (g f '
(S S
(
s s
3 f-12/3 p
-L1 i
(12) 7 a1?.ov Y
N Y
\\ Y]
I and when S
=0 g
(S S
S
=[-2/3[.
(13) i 7 allow 7
7 By using equations (6), (10) and (12) or equations (6), (11) and (13) one could construct the outer " box" for Case IV as indicated by the sample design limit curves in Figure 3.
In this case, equation (6) divided by S would y
yield the top horizontal line, or the maximum permissible stress intensity, and depending on the value of shear stress (S ), equations (10) or (11) would 3
yield the right-hand boundary or the maximum allowable tensile stress, and equations (12) or (13) would yield the left-hand boundary or maximum allow-able compressive stress.
For design purposes, the " box" described earlier would be all that would be required if the desi stresses at a point,gner in his calculations always solved for the true even if the structure possessed stresses greater than the yield stress, i.e., if he performed a plastic analysis every time a stress exceeded the yield stress. However, in most cases the designer utilizes the various elsstic equations at his isposal in performing his analysis rather than resorting to the more cumbersome and detailed plastic analysis methods.
The development below will demonstrate that, where stresses due to the normal force are not great, much higher stress intensities than those permitted by the design " box" could be permitted (using elastic equations) without true stress intensities exceeding the " box" salues. As mentioned previously, this is because, under the conditions described, elastic equations give stress in-tensities that are higher than the actual values.
Of primary concern are the tensile or compressive stresses resulting from a normal force ecmbined with a bending moment, for it is here that e.astic equa-tions can yield stress values that differ greatly from the true stresses.
The first step is to derive expressions for the Iiornal Force (II) and the Bend-ing Moment (M) for an arbitrary plastic stress distribution as indicated in Figure 5(b).
It should be noted that the stress distribution indicates a ma-terial that possesses an amount of stain-hardening rather than one that is Or, as in the B&W design procedure, if elastic equations are used.
/
9.11-17
(3 rigid-plastic._ Also the neutral axis of the crcss section may not coincide
%)
with the centroidal axis of the cross section. This shift in the neutral axis is defined by an angle 0 as shown in Figure 5(a).
Referring to' Figures 5(a) and 5(b) the expressions for (N) and (M) may be writ-ten as follows:
i
[A 847)dAI7)
Ilk)
N=
M=[A S(y)
- y
- dA(y)
(15) where S(y) = stress distribution as a function of y dA(y) = differential area of the cross section y = distance from centroidal axis A = Total area of cross section Converting to polar coordinates and treating the arbitrary stress distribution of Figure 5(b) as-the sum of stress _ distributions of Figures 5(c) and 5(d) greatly simplifies the performing of the indicated integrations; however, once this has been done, and N and M of each case have been determined, combining of the two quantities of each case yields N
= -4S rt 0 - Amr sin 0 (16)
M
- kS r t cos 0 + mI (17) d y
where S = yield stress of the material A = area of the cross section of the cylinder = 2nrt I = moment of inertia about centroidal axis of the cross section of the cylinder = wr3t m = ratio of stress to distance from neutral axis in the esse of Figure _5(d);
i.e.,
the slope of the linear stress distribution in Figure 5(d).
Bearing in mind that N and M are the normal force and moment based on true stresses, the next step is to determine the corre'sponding " pseudo-elastic" stresses which are. derived by applying the following elastic formulas:
S *
(10)
N pV 0080
. 9.u-18 4
==t-
,p-f===
' F =P T" " Mt
-f e 7
---% -+
re-**
t-FTv-'***f
e M
=j (19)
S g
S = n rmal stress due to N N
S = " pseudo-elastic" bending stress of elements in the outer fibers due to M N = normal force calculated by equation (16)
M = bending moment calculated by equation (17) c=r Substituting the values of N and M calculated by equations (16) and (17), re-spectively, into equations (18) and (19), respectively, simplifying, and divid-ing each equation by S yields 0+fsin0 (20)
=-
y y
S
[=
ces0+f.
(21) y y
O Therefore, the total axial stress (S ) would be S
S S
f=f+f.
(22) y y
y Therefore, for an element in the tensile region; i.e., S > 0, the stress in-tensity calculated from elastic equations would be
~
o o
(S S
~
(S S
S f= f-h =h1h+f
+
h1f-f h
+ gg (23)
+
y y
y (y
y (y
y y
y and for an element in the compressive region, i.e., Sg _ 0, 2
%=
[o -[o
/S S ' -
[S
/-f
=2 (24)
+
y y
y y
y,
y 9.11-19
s.
To apply equations (20) through (2h) in establishing stress limits for elasti-cally calculated stresses, reference is made to Figure 6.
Represented are various stress distributions which have been limited by either (S )all as cal-t culated in equation (8) or by (S )all as calculated in equation (12). Consid-c ering each case separately and solving equations (20), (21), (22), and either (23) or (2h) depending on the resulting value of S, a value of SI/Sy can be A
obtained. Plotting these stress intensities versus the stress due to the cor-responding normal forces (Sn/S ), one obtains the dashed-lined curves indicated y
in Figure 3 These curves are the limiting values of the stress intensity when stress values have been calculated by elastic formulas. From these curves, since the dotted line curve calculated using elastic equations always lies above the design " box," one can see that the limits for elastically calculated stresses can always be greater than or equal to the limits imposed on the true or real stresses.
O 0082 9.11-20
O 6 STATIC DEFLECTION DuE TO SLOWLY APPLIED LOAD 4
g
/-2S STATIC j
i I
VIBRATORY I
MOTION
[
s
/
\\
/
\\
TaRuST
/
1 I
I O
l I
/
/
I I
I I
I I
I I
j I
I kf
/
/
a i
4 4
f f
] !
YHH/un '////un) 0083 i
O BASIC REACTOR RESPONSE MODE FIGURE 9.11-1 L
O reno 2coe isoc t
E 200 soo O
doo o
,,,,,,,,,,,,i,,,,,,,,,,, i
,,,,,,,,,,,,i col of 1
1 to TIME SECoND$
i l
l l
l t
Ot THRUST TIME CURVE FOR CIRCUMFERENTIAL W
OR LONGITUDINAL BREAK OF A 36" 1.D.
REACTOR COOLANT PIPE.
h FIGURE 9.11-2
O NAT*L.304 STAINLESS STEEL CROSS SECil0N HOLLOW C I R C'J L A R N00? STRESS.O 20 Sy
$s 0 03y
- 3. 0 TEMP.600 F CASE IV 2.5 i
CASES II & Ill j
~,
s
\\#'
CASE l E
/
j e
/
q 1.5 E
s
/
y
/
/
/
~
s s
l l
C 1.0
/
/
/
O' s
s
/
/
/
/
J
/
s s
/
0.5 s
s
/
s s
y
/l f
/
/
f f
t f
f f
2.0 1.5
-1.0 0.5 0.0 0.5 1.0
- 1. 5 2.0 STRESS FROM AIIAL FORCE /YlELD STRESS (INCLUDING PRES $URE END FORCE) 0085 O
DESIGN LIMIT CURVES FIGURE 9.ll-3
O S.
- S.
St 3,
2 2
f p+p/p m
V (a) 1---
N v
I P. Internal Pressure 7.P /. hoop membrane stress 2
3 rt S. Tensile axial stress
/
t
/
S. Compressive axial stress c
o 5,. Shear stress e,
S..E. radial atress S
r m
S Se
$e LOADING COMBINATIONS AND STRESS DISTRIBUTIONS hl FIGURE 9.II-4
i l
O i
TEllSILE II0M S -5, h
8t 3,
t 1
l l
.s-
_e W
E b
ff fa\\
e'--
s 4
i'
\\
t'
=
1 g ~
~__/
S
- 5, S.8 c
c 7 COMPRESSIVE REGIN (a)
(b)
(c)
(d)
' ROT E: STRAINS A:E ASSUMED TG VARY LINEARLY O
FROM TN! SEUTRAL Al'i 4
4 4
4 0087 d
.fD TYPICAL PLASTIC STRESS DISTRIBUTION FIGURE 9.ll-5
.... ~. - _,.... _ _ _ _.. -
_m__
O
[IIALLOW.
II'I ALL0s. I*I ) allow.
I*3 ) ALL0s.,
i ALL0s.
II'!!LL0s.
I c
c
_[
_/
( 5 I ALL0h.
I8) ALLOW.
t 1
-3,
-5,
-5
-3, 3'
3, E
I IIcIALLOW.
II)ALL0s.
I8 I4LL0s.
IIIALL0w.
f c
c c
I*3 I ALL0s. -5, (S) allow.
c c
~
-5 5,
h IIIALLOW II ) ALL0s c
c 0
0086 ASSUMED STRESS DISTRIBUTIONS AT THE DESIGN LibilT h
FIGURE 9.11-6
Dockets 50-302 and -303 Supplement No. 1 February 7, 1968 (3
\\~sJ QUESTION In view of the possib3e concern about flood conditions that might 9.12 exist at the plant site, information is required concerning the location of any critical pumps and motors that might be necessary for safe shutdown of the plant and the relationship to the possi-bility of flooding that might render them inoperable. Also of concern is the seismic design adequacy of the interconnecting line from the service water intake.
ANSWER Flood and missile proter
'n are provided on plant safeguard systems; in addition sup... ting structures and piping systems related to safeguard systems will be seismic analyzed as noted below. The Nuclear Services Cooling System protection features are specifically reviewed under Question 9.14 Dynamic Piping Analysis The dynamic analysis of critical piping system (i.e. Class I systems) vill be a modal analysis based upon either a distributed or lumped mass solution depending upon the complexity of the system. The two approaches are performed as follovs:
a.
Distributed Mass Analysis The system is represented by a number of b!raight uniform beams with a distributed mass and stiffness. First, the O'
transfer matrix for each of the straight beams is determined and the rotation transfer matrix for each joint calculated.
Next, the equation of motion is written in matrix form.
Previously determined traasfer matrices are u:-d.
Consider-ing the appropriate boundary conditions, the characiwristic determinant is generated. When the natural frequencies are known, the corresponding mode shapes are determined.
Then, by using the response spectra for a single-degree of freedom system, the maximum displacements are obtained as the root-mean-square sum of the modal maxima. Finally, after the maximum displacements are known, forces and moments are calculated at the structural joints.
b.
Lunped-Mass Analysis The system is represcated by a series of concentrated masses.
2 First, the space coordinates are established for the system and coordinates of mass points are detere.ned.
Using a static analysis, flexibility matrices corresponding to these mass points are computed. Next, the equations of motion are written in matrix form. Force influence coefficients method is used. Natural frequencies and mode shapes are btained assuming harmonic motion of the system. Finally, using the same technique as for the distributed mass analysis, maximum internal forces and moments are calculated at the structural joints.
a 0089 9.12-1
m-In addition to the earthquake response for the pipe system, the models described above vill be used to determine forces and moments with resulting stresses for any transient or permanent displacements which vill be induced at the support points.
The foregoing description applies to the dynamic analysis of all piping systems supported nt local points (i.e. not including h
buried pining). This includes local supports which can be defined as affordinc either structural fixity or as a string. Buried piping vill be analyzed by one of the following methods:
a.
Lumped-Mass Analysis The system is again represented by a series of concen-trated masses. The analysis is performed in a manner similar to that described here before for a lumped mass system insofar as attachments to the structures are concerned. The supports for the line within the soil are identified by sprints which has a constant which is a function of the soil properties. All other aspects of the analysis are the same. The analytical model described in this manner can again be used to determine forces and moments with resulting stresses for transient or permanent displacements which vill be induced at the suppc-t points. This provides a basis for considering differ 1. t motion between the soil and structure.
b.
The more common analytical technique vill be on the basis of describing the model to permit analysis as a semi-infinite beam on an elastic foundation. The transient and, if applicable, permanent displacements of the structure can be determined. The displacements of the structure are determined independent of the piping system. The transient displacements of the buried pipe some distance from the connection to the structure are based upon direct application of the response spectra. With a known maximum displacement and soil properties the seismic pipe stresses can then be determined as described before by the analogy of a semi-infinite beam on an elastic foundation.
Both techniques for analyzing seis at stresse in buried pipes are valid only if the soil surro'.adin
.he pi e experiences t
uniform motion due to the eart'q'.ake el itat.4 an.
This requires the use of a homogeneous backfill mate
- 31 '. the complete length of the pipe. This vill be prov. tet.
0090 9.12-2 (Revised 3-lh-68)
Docksts 50-302 and ~303 Supplement No. 1 February 7, 1968 T
QUESTION A review of the control instrumentation section reveals no mertion 9.13 of the operation of critical controls under seismic loadings.
Provide infor=ation concerning the design provisions that are taken to ensure that the critical controls can operate to ensure safo shutdown under seismic loading.
If a battery system is required for emergency shutdown, describe the design of the battery support system, and the provision incorporated to ensure that no damage will occur during an earth-quake.
AdSWER Controls by their nature are most sensitive to high frequency vibrations and high shock loads. Since seismic disturbances are of lov frequency and create lov shock loads they will not affect the critical controls. The components in the reactor protection system and safeguards actuation system vill suffer no loss of function at accelerations of 0.lg. horizontal and 0.067g in vertical condition.
One of the two installed station batteries is required for shut-down upon loss of all a-c power. The batteries vill be mounted in steel racks and bolted to the floor. The individual cells vill be of a tough, plastic construction and will be strapped in place. The rack cell and internal elements will withetand 0.lg horizontal and 0.067g vertical without loss of function or perma-nent deformation.
O b
009l a
9.13-1
r Dockets 50-302 end -303 Supplemennt No. 1 February 7, 1968 O
QUESTION The cooling water intake structure and associated piping are 9.lk critical to the safe shutdown of the plant. Additional information is required as to the design of these items, particularly under dynamic loading conditions. and for conditions associated with high water and hurricanes.
ANSWER The cooling water intake stru:ture refers to the raw sea water circulating water system for the main surface condensers. This system is in no way associated with the Enrineered Safeauards and is not required for the safe shutdown of the plant.
The only raw sea water cooling required for Engineered Safeguards is the Nuclear Services Cooling system. This is classed as a Vital system and its intake structure is located in part of the Auxiliary Building. The entire Auxiliary Building vill be designed as a Class I structure; moreover the associated piping vill be designed as a Vital Cooling system.
The intake structure is part of the Auxiliary Building and the two sides not protected by the Auxiliary Building and Reactor Contain-ment Structure are surrounded by fill up to the roof slab. The associated piping is underground and therefore, the intake structure and piping are protected from missile penetration.
The fill surrounding the plant is high enough to prevent flooding
()
from rising water due to a Maximum Probable Hurricane. Also the Nuclear Services Cooling pumps and the Decay Heat Service Cooler Sea Water pumps are sealed at their bases and the internal clear-ances are such that it is not likely that rising water vill-cause flooding.
r
(,/
/ t 9.14-1 v-y y
d Dockets 50-302 and -303 Supplement No. 1 February 7, 1968 QUESTION With regard to the prestressed containment building and other 9.15 critical ecmponents, describe in detail the long term surveillance program that is planned to ensure the continuing adequacy of the facility.
ANSWER Postoperational testing and surveillance are described in Section 5.6.2 of the PSAR.
4 6
a 4
4 i O 4
1 0
0093 i
1
~
9.15-1
.,__m
-