ML19326D468
| ML19326D468 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 05/05/1975 |
| From: | Howell S CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML19326D465 | List: |
| References | |
| NUDOCS 8006110497 | |
| Download: ML19326D468 (23) | |
Text
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s CONSUMERS POWER COMPANY APPLICATION FOR REACTOR CONSTRUCTION PEPHIT AND OPERATING LICENSE Docket No. 50-329 Docket No. 50-330 Amendment No. 29 Enclosed herewith, amending and supplementing the above-entitled application, are revised and additional pages for incorporation in the Preliminary Safety Analysis Report submitted with the original applica-tion on January 13, 1969, as amended.
This material consists of the following:
1.
Revised Pages h-iv, h-5 and h-13, and new Page h-36a to incorporate Table h-lla which lists the ASFE Code Cases for which Mid-land requests NRC approval pursuant to Title 10 Code of Federal Regu-lations Part 50.55a (a)(2)(i).
2.
Revised Page 5-17 vhich amends the allowable reactor building membrane compression stress.
3 Revised Page 5-k6 to permit the use of welded splices, only if required, for main reinforcing bars in the reactor building exterior structures.
h.
Revised PaBes 5-51a and 5-56b to clarify the inspection requirements for the liner plate leak chase channel system.
5 Revised Page 5-63 to expand the allovable structural materials used in the auxiliary building.
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Revised Page 50-1 and new Pages 5C ha through SC hd which clarify the procedures used at the Midland Plant for mechanical splicing of reinforcing bar.
7 Revised Page 5E-2 which amends the installation jacking requirements for the reactor building tendons.
8.
Revised Page T-8 which amends the operating procedure for the reactor building isolation valves.
9 Revised Page T-41 to delineate the radiation sensors and monitors that are designed to Seismic Category I requirements.
10.
Revised Page 9-32 to delete the requirement for emergency power backup for the instrument and service air compressors.
11.
New Page 2.2.8-1 responding to Item 2.2.8 of the enclosure to Mr. A. Schwencer's letter to Mr. S. H. Howell dated November 12, 1974.
This response concerns the makeup water supply system.
12.
Divider tabs entitled: " Table of Contents"; " Appendix SI";
"AEC Questions"; " Encl. B - Identification of Problem Areas"; " Encl. A -
Additional Information Required"; " Hydrology Questions"; " Additional Question Responses"; and " Questions Received After C.P. Issued".
t These new and revised pages bear the notation, " Amendment No. 29, 4/T5," and are marked in the margin to indicate where changes have been made.
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We request that Staff review of this amendment be completed by August 1, 1975.
CONSUMERS POWER COMPANY Dated May 5, 19T5 ny dCLi M
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Li Stephen' H. Howell, Vice-President Sworn and subscribed to before me on this 5th day of May, 1975 i f. n.o b/be Y Notary Puslic, Jackson County, Michigan My commission expires May 18, 1976 l
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t INSTRUCTIONS POR INSERTING AMENDMENT NO. 29, 4/75 n
During insertion of new or revised pages, a dash (-) in the remove or insert column of the directions means no action is required.
REMOVE INSERT Volume 1 Amendment No. 2R cover letter at the front of the volume List of Effective Pages, List of Effective Pages, 4/75, 8/13/14 (Located in front TiZ r=gec) of Table of Contents, page 1)
Tatrentitica "laole of Cvutears" Page 4-iv -
Jage kiv
-Pa~gM Page-+-5--
Page -443-Paga 6 P41p 4-30.
Volume II
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s Page-6--17 Page 5-17 ^
Page 5 Page-5-4F Pag &5 91=
Eage-5-51a-Page 5-56B Page-5-56b-Pag &5-53 Page-5"6T Page 30-1 Page-Sel--
PageJ C ';, 4b, 4c, --4 d' Page-SrE=2 Page-5 E Tak titled " Appendix-M2' Page P' Page 7-F Pa;; 7-51 Page-7 41 Paga 4-17 Page 9-32 Volume III Tab entitled "AFC Ouestions" ne the front of the vo'lume
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Tab. entitled " Encl. B - Identifi-c'Ition-of-Probica Aicos" between
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page 13.7.3.10-1 and page 1-2 Tab entitled " Encl. A - Additional
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INSTRUCTIONS FOR INSERTING AMENDMENT No. 29 CONTINUED:
2 REMOVE INSERT n
Tab entitl_ed." Hydrology-Ques-tiotis" between page 12.0-2 and Page I--l__
Tab entitled "Additisnal ques-tiott-Resp ~6n~ses" between Page II-2 and page 1.00-1 Tab entitled " Questions Received Af ter C.P. Issued" follo. wing-page 12.00-1
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Page 2.2.8-1 following the tab entitled "Quesr4==--R er C.P. Issued" O
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LIST OF TABLES (At Rear of Section)
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Table No.
Title Page
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h-1
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Tabulation of Reactor Coolant System Pressure Settings 4-29 h-2 Reactor Coolant Quality k-30 Re\\
h-3 ac$or Vessel Design Data h-31 Press \\
h-h ur r Design Data h-31 h-5 Steam Generator Design Data k-32 h-6 Steam Gener Feedvater Quality h-33
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h-T Reactor Coolanthmp Design Data h-3h
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h-8 Reactor Coolant Pip' ira Design Data h-35 h-9 Transient Cycles 4-35 h-10 Design Transient Cycles h-36
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h-ll Reactor Coolant System Code and Classifications h-36 k-12 Materials of Construction h-37 h-13 References for Figure 4-k - Increase in Transition Temperature Due to Irradiation Ef cts for A302B Steel k-39
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'.that control pressure and temperature during heatup and cooldown.(3) This pro-I cedure vill insure that the stress levels do not exceed those specified in a through c above.
h.l.4.h Quality Manufacture The quality control program for the reactor coolant system is as outlined in h.5 This program vill be organized, implemented and monitored as described in l Appendix 13,andAmendmentsNo.6 nd 8.
4.1 5 CODES AND CLASSIFICATIONS All pressure-containing components of the reactor coolant system are designed, l fabricated, inspected, and tested to applicable codes as listed in Table h-ll.
4.2 SYSTEM DESCRIPTION AND OPERATION h.2.1 GENERAL DESCRIPTION The reactor coolant system censists of the reactor vessel, two vertical once-through steam generators, four shaft-sealed coolant circulating pumps, an elec-trically heated pressurizer, and interconnecting piping. The system is arranged as two heat transport loops, each with two circulating puups and one steam gen-erator. Reactor system design data are listed in Tables 4-3, 4-4, 4-5, 4-7 and h-8, and a system schematic diagram is shown in Figure 4-1.
Elevation and plan views of the arrangement of the major components are shown in Figures k-2 and 4-3 h.2.2 MAJOR COMPONENTS h.2.2.1 Reactor Vessel The reactor vessel consists of a cylindrical shell, a cylindrical support skirt, a spherically dished bottom head, and a ring flange to which a removable reactor closure head is bolted.
The reactor closure head is a spherically dished head welded to a ring flange.
The vessel has six major nozzles for reactor coolant flow, 69 control rod drive nozzles mounted on the reactor closure head, and two core flooding nozzles -
i all located above the core. The reactor vessel vill be vented through the con-trol rod drives. The vessel closure seal is fomed by two concentric 0-rings with provisions between them for leakage collection. The reactor vessel, j
nozzle design, and seals incorporcte the extensive design and fabrication ex-4 perience accumulated by B&W.
Forty-six in-core instrumentation nozzles are located on the lower head.
The reactor closure head and the reactor vessel flange are joined by sixty 6-1/2 in. diameter studs. Two metallic 0-rings seal the reactor vessel when the reac-tor closure head is bolted in place.
Leakoff and test taps are provided in the annulus between the two 0-rings to dispose of leakage and to hydrotest the ves-sel' elosure seal after refueling.
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The vessel is insulated with metallic reflective-type insulation.
Insulation panels are provided for the reactor closure head.
h-5 Amendment No. 8 2/9/70
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-4.2.7 LEAK DETECTION To minimize leakage from the reactor coolant system all ecmponents are inter-connected by an all-welded piping system.
Some of the components have access openings of a flanged-gasketed design. The largest of these is the reactor closure head, which has a double metal 0-ring seal with provisions for dis-posing of leakage between the 0-rings.
With regard to the reactor vessel, the probability of a leak occurring is con-sidered to be remote on the basis of reactor vessel design, fabrication, test, inspection, and operation at temperatures above the material NDTT as described in 4.3.1.
Reactor closure head leakage will be zero from the annulus between the metallic 0-ring seals during vessel steady-state and virtually all transient operating conditions. Only in the event of a rapid transient operation, such as an emergency cooldown, would there be some leakage past the innermost 0-ring seal. A stress analysis on a similar vessel design indicates this leak rate would be approximately 10 cc/ min through the seal monitoring taps to a drain, and no leakage vill occur past the outer 0-ring seal.
The exact nature of this transient condition and the resulting small leak rate will be determined by a detailed stress analysis.
In the unlikely event that significant leakage should occur from the system into the reactor building during reactor operation, the leakage will be de-tected by one or more of the following methods:
a.
Instrumentation in the control room will indicate the addition rate
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of makeup water required to maintain normal water level in the pres-surizer and in the makeup tank. Deviation from normal makeup and letdown to the reactor coolant system will provide an indication of the magnitude of the leak.
b.
Control room instrumentation will indicate an increase in reactor building activity, c.
Radwaste instrumentation will indicate the existence of a large amount of water flow from the reactor building sump.
4.3 SYSTEM DESIGN EVALUATION 4.3.1 SAFETY FACTORS The reactor coolant system is designed, fabricated, and erected in accordance with proven and recognized design codes and quality standards applicable for the specific component function or classification. These components are de-signed for a pressure of 2,500 psig at a nominal temperature of 650 F.
The
. corresponding nominal operating pressure of 2,185 psig allows an adequate mar-gin for normal load changes and operating transients. The reactor system com-l ponents are designed to meet the codes listed in Table 4-11.
Aside from the safety factors introduced by code requirements and quality con-trol programs, as described in the following paragraphs, the reactor coolant system functional safety factors are discussed in Sections 3 and 14.
i 4-13 Amendment No. 8 2/9/70
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f membrane compression stress is limited to 0 30 fc1' whereas in combination with flexural compression the ~ vim m allovable stress is limited to 0.60 fei' per ACI 318-63 b.
For local stress concentrations with nonlinear stress distribu-tion as predicted by the finite element analysis, 0 75 fei', is permitted when local reinforcing is included to distribute and control these localized strains.
These high local stresses are present in every structure but they are seldom identified because of simplifications made in design analysis.
These high stresses are allowed because they occur in a very small percentage of the cross section, are confined by material at lower stress and would have to be considerably greater than the values allowed before significant local plastic yielding would result. Bonded reinfore-ing is added to distribute and control these local strains.
Membrane tension and flexural tension are permitted provided they do not jeopardize At aintegrity of liner plate. Membrane tension is limited to 1.0 y f '. When membrane tension exists, the tensile e
stress in the liner plate is limited to 0 5 f When there is y
flexural tension,.but no membrane tensicn, the'section is designed in accordance with Section 26-5 (a) of ACI 318-63 Shear criteria are in accordance with ACI 318-63, Chapter 26, as modified by the equations shown in 5 1.1.4.6, using a load factor of 1 5 for shear loads.
5 1.1.4.4 Icads Under Sustained Prestress The conditions for design and the allowable stresses for this case are the same as above except that the allovable tensile stress in bonded reinforcing is limited to 0 5 f. ACI 318-63 limits the concrete compression to 0.45 f '
forsustainedprestbessload. Values of 0 3 f ' and 0.60 f ' are used as e
describedabove,whichbrackettheACIallovab$evalue. However, with these same limits for concrete stress at transfer of prestress, the stresses under sustained load are reduced due to creep.
5 1.1.4.5 At Design Ioads The reactor building is designed for the following loading cases on a
" working stress" basis:
a.
D+F+L+T o
b.
D+F+L+P+T A
Where:
D = Dead Load L = Appropriate Live Load
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F :: Appropriate Prestressing Icad 5-17
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the field performs similar user tests on one sample from each 25 50' tons of material. High strength bars are clearly identified
-prior to shipment to prevent any possibility of mix-up with lower y,,
-strength reinforcing bars.
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Mechanical Splices The CADWELD inspection program is detailed in Appendix 5-C.
Ordinary welded splices are not used for_ main bars in the reactor Lbuilding' exterior structure.
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Fabrication-i
_ Visual inspection of fabricated reinforcement is performed to ascertain dimensional conformance with specifications and drawings.
d.
Placement q
Visual' inspection of*in-place reinforcement is performed by the placing inspector to assure dimensional and location conformance
. ith drawings and specifications.
w 5.1.3.3.4 Prestress System a.
Wires Sampling and testing of the tendon material used in construction conform to ASTM Standard A-421 or ASTM A-416.
The following procedure is used:
1.
Buttonhead rupture test from each reel of wire is made.
2.
Each size -of wire from each mill heat and all strands from each manufactured reel that is shipped to the site shall be
-assigned an individual lot number and tagged in such a manner that.each such lot lcan be accurately identified. All unidentified prestressing steel or anchorage assenblies
. received at the jobsite are subject to rejection.
3.
Random sanples as specified in the ASTM Standards stated above are taken from each-lot of prestressing steel used in the work.
With each sample of prestressing steel wire or strand that is' tested, there'is submitted'a certificate stating the manu-facturer's minimum guaranteed ' ultimate tensile strength of j
the sample tested.. Stress-strain curves are plotted and the j
yield and tensile strength verified.
The anchorages develop j
the minimum guaranteed ultLmate strength of the tendon and the i
minimum elongation of the tendon material as required by the applicable ASTM specification.
Field inspection insures that there are no visible machanical i
or metallurgical notches or pits in the tendon material.
5-46 Amendment N. 25 o
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The cxtent of wald cxamination in this lattsr c cs in tha came cs for the magnetic particle method. LPE techniques are specified in section IX-3600 and acceptance standards are in accordance with
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section CC-5534 of the proposed Section III-Division 2.
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Leak Chase Channel Inspection After the liner plate seam welds have been successfully examined through 1) visual inspection, 2) soap bubble vacuum box tests, and
- 3) either random radiography or alternately MPE or LPE examination, the seam welds are covered with a leak chase test channel. These channels allow a final means to assure leak tightness of the seam welds.
23 After attachment of the channel to the liner plate and prior to placement of concrete adjacent to the liner weld, the leak chase channels are pressurized to 80 psi, and a soap bubble solution coated over the attachment welds. Any bubble appearing within 20 seconds is cause for repair of that portion of the weld. The 80 psi pressure is monitored by valving off of the air supply for at least 15 minutes; any pressure decay noted during this period is further cause for rejection.
Any leakage detected shall be repaired and the channels shall be retested.
5.1.4.1.2 Preoperational Integrated Leak Test The design leak rate is not more than 0.1 percent by volume of the contained
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quality control during erection, this is a reasonable requirement.
14l atmosphere in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 67 psig.
It has been demonstrated that, with good The basis of the leak rate test is the Absolute Method as specified in Bechtel 26 Topical Report BN-TOP-1, Revision 1, November 1, 1972, " Testing Criteria for Integrated Leakage Rate Testing of Primary Containment Structures for Nuclear Power Plants." During the performance of this test, the liner plate leak 23 chase channels are open to the reactor building atmosphere in order to ensure that the test pressure is applied directly to the liner seam welds.
The integrated leak test is conducted in accordance with the requirements of 10 CFR 50, Appendix J, " Reactor Containment Leakage Testing for Water Cooled Power Reactors," after the inspection and testing of welded joints, penetrations, and mechanical closures; completion of repair measures for minimizing leakages; 26 and completion of any required containment structure pressure tests for strength.
Because the reactor building is a thick walled concrete structure, short-term temperature or meteorological variations,should not have any appreciable effect on the reactor building ambient temperature and pressure. The duration of the integrated leak rate test is determined by the test duration criteria established in Section 2.0 of BN-TOP-1.
N Amendment No. 26 5-51a 4/74
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' 5 2.2 DESIGN CRITERIA AND GENERAL DESCRIPTION The building is designed in accordance with " Building Code Requirements for Reinforced Concrete - ACI 318-63" for concrete, and " Specification for the
- Design, Fhbrication and Erection of Structural Steel for Buildings," 1963 edition, for structural steel unless otherwise noted.
Live loads used in the design conform to requirements of the " Uniform Building Code," 1967 edition. Wind and earthquake loadings,' load factors and load combinations as specified in Appendix 5-A are used in the design of this building.
The major structural materials used in the design of this building are as follows:
2d Concrete fc = 4000 & 5000 psi at 90 dnys Reinforcing bars ASTM A-15 Structural steel ASTM A-36 High strength bolts ASTM A-325 Stainless steel pool liners ASTM A-167 Type 304L The building is founded on a concrete mat foundation. The reinforced con-crete floor slabs are supported on steel beams and columns.
Exterior valls and the spent fuel pool are reinforced concrete while most interior valls
,f are concrete block. Structural members are designed for dead, live and vind or dead and tornado load combinations, wherever applicable. Some walls in 2
the fuel pool area and in the area adjoining the reactor buildings are de-signed to act as deep beams.
In the design, seismic, vind and other appropri-ate lateral loads have been assumed to be resirted and carried down to the foundations by diaphragm action of the slabs at.d the shear vall action of the walls.
The new und spent fuel pool walls are inherently resistant to tornado and the micciles generated from it.
The spent fuel pool has been designed to withstand temperature stresses caused by the failure of the pool water cool-ing equipment.
High-temperature service piping embedded in the fuel pool walls has been thermally insulated to avoid damage to structural concrete of the fuel pool walls. The enclosure over the fuel storage facilities is not designed to res!st tornado.
A couerete enclosure, with five floor levels, which houses the control room, the ventilation equipment room, the cable spreading room, the access control room, and the switchgear and battery rooms, is designed to withstand the tornado loading specified in Appendix 5-A, and hydrostatic pressure due to the maximum probable flood.
The area containing the engineered safeguards equipment is partitioned into separate rooms to provide protection in the event of flooding due to a pipe rupture. The partition walls are designed to withstand hydrostatic loading over their full height.
s 5-63 Amendment No. 27 8/74
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APPENDIX SC g
MECHANICAL CPLICING OF REINFORCING BAR USING THE CADWELD PROCESS S_ OPE C
This procedure specification is to be used for mechanical splicing of deformed concrete reinforcing bar for tencile loading. The mini =um tensile strength of the splices shall equal or exceed 125 percent of the mini =um yield strength for each grade of reinforcing steel as specified in the appropriate ASTM standard.
PROCESS.
Splicing of concrete reinforcing bar under this procedure specification shall be. done by the CADWELD Process (ERICO Prcducts, Inc) using "T" series and "B" series full tension splice materials only.
"C" series and "C-16" series splice materials shall not be used. Any "special" splice kits used should also meet the minimum tensile strength requirements as stated in " Scope" above unless specifically permitted otherwise.
Co*nection of reinforcing bars to reactor building embedded plates shall be made with "B" series splices except that the sleeves will be supplied by the liner plate fabricator, as shown on attached Fig SC-1.
MATERIALS REBAR MATERIAL This procedure specification shall be used only for splicing the concrete reinforcing bars shown in Table 5C-1.
SPLICE MATERIAL The "T" series splice material shall be as shown in Table 5C-2.
SPLICING GENERAL All splices shall be made in strict accordance with the manufceturer's instruc-tions as presented in ERICO Products Bulletin RB10M-169, CADWELD Rebar Splicing.
A manufacturer's representative, experienced in CADWELD splicing of reinforcing bar, shall be present at jobsite at the outset of the work to demonstrate the equipment and techniques used for making quality splices. He shall also be present.for at least the first fifty (50) production splices to observe and verify that the equipment is being used correctly and that quality splices are being obtained.
O SC-1 Amendment No. 5 11/3/69
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The only significantly new value introduced is %. 0 95 for prestressed
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tendons in' direct tension. Ahigher%valuethanforconventional'reinfore-ing has been allowed because (1) during installation the tendons are each jacked to about 94 percent of their yield strength, so in effect, each -
tendon has been proof tested, and (2) the method of manufacturing prestres-sing steel (cold drawing and stress relieving) insures a higher quality product than conventional reinforcing steel.
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(~'m Air-operated valves are automatically positioned to their engineered safeguards position upon loss of control air. Valves used with active redundant valves are equipped with a single electrical actuator for control by a single engineered safeguards channel as shown in Figure T-2C.
Solenoid valves used with redundant passive valves are provided with two electrical actuating signals, each con-trolled by a single engineered safeguards channel operating in an OR configura-tion.
Engineered safeguards action is initiated when power is applied to the electrical actuator.
7 1.2.2.2 Reactor Building Cooling, Isolation and Borated Water Storage Tank Low Level Figure T-2D shows the initiating sensors and logic for the reactor building cooling and isolation system. The control system is designed to initiate auto-matically the necessary equipment upon the appropriate engineered safeguards signal. To assure reliability, the control system is designed on a two-channel concept with redundancy and physical separation, each channel initiating reactor building isolation and the operation of separate and redundant equi gent trains.
The system is also designed to prevent reopening of the reactor building isola-tion valves unless the reactor building pressure and/or radiation is below a preset level.
Each critical variable has four sensors utilizing a 2-out-of-4 logic to provide reliable operation with a minimum of nuisance tripping. The fcur sensors are physically isolated and operation of any 2-out-of-4 vill initiate the appro-priate safeguards action.
This action is provided by combining the four sensors into a relay matrix which provides a dual channel initiation signal.
Coincident 2-out-of-4 high recctor building pressure signals vill:
(1) close all reactor building isolation valves not required for emergency safeguards service; (2) start reactor building spray pumps; (3) open reactor building spray valves; (4) initiate operation of reactor building emergency cooling and recirculation units.
In like manner, coincident 2-out-of-4 reactor building high radiation signals close all reactor building penetrations open to the reactor building atmos-phere (Type II).
At least three out of the four radiation sensors must sense nor=al radiation level and three out of the four pressure sensors must sense nor=al pressure before the operator can reset the pressure isolation circuits and the radia-tion isolation circuits. This will automatically open the reactor building isolation valves.
Coincident 2-out-of-4 low level sensors in the borated water storage tank will automatically initiate the necessary valve operations to permit shift to the recirculation mode of operation for the low-pressure injection and reactor building spray _ pumps.
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I 7 5 2.2 Waterborne Radiation Monitoring System The waterborne radiation monitoring system monitors the possible sources of radioactive liquids released to the environs as well as the point of release to the environment itself. The gross activity in the dilution water, service vater and radvaste liquids discharge line is monitored.
The point-of-release monitor on the dilution water alams when the momentary level is slightly above the maximum permissible limit, but the monitor is supplemented by a sampling system so that laboratory analysis of the effluent demonstrates that the actual integrated release was considerably below the caximum level.
The service water conitor alarms when the gross activity level in the service water would result in a fraction of the maximum release to the environment.
The radvaste liquids monitor ala ms and terminates the release when the gross activity level in the radvaste effluent line would result in a fraction of the taximum permissible release to the environment.
In addition, continuous conitors are provided for process steam and component cooling water.
Supplementing the continuous monitoring, samples are taken from the component cooling water, service water, radvaste, condensate and primary coolant syste=s for laboratory verification that the gross activity levels are within permis-rk sible limits.
7523 Area Radiation Monitoring " Qtem The multichannel area radiation monitoring system monitors the radiation in-tensity of areas in the plant where it is possible for operating personnel to be subjected to gamma radiation. The selection and number of points are coor-dinated with the plant access control so that operating personnel are not able to enter an unmonitored area in which they could be exposed to a dose in ex-cess of the limits of 10 CFR 20.
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All channels of the radiation monitoring and protection system consist of re-motely mounted detectors connected to panel mounted readout, control and power supply instrumentation in the main control room. Except for the detectors, which contain Geiger-Mueller and photomultiplier tubes, the instrumentation is completely solid state. The operational reliability of each channel can be verified by control room actuation of radioactive check sources at each detec-tor.
Loss of sample flow, loss of signal or loss of power supply causes an alarm annunciation. All detector signals are recorded on a potentiometric strip chart recorder.
T.5 3 SYSTEM EVA WATION T.5 3 1 Reliability Each component of each system is designed to meet its =aximum environmental i
_ conditions of pressure, temperature, relative humidity and seismic loading.
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(2) Makeup Tank (3) Quench Tank (h) _ Radwaste System (Local Grab Samples)
-(5) Main Steam i
Samples are collected in containers designed for full operating temperatures and pressure and at flow velocities which insure transport of suspended parti-cles where appropriate.
Sample lines are purged to insure that representative samples are obtained.
Gaseous leakage is collected by placing the sampling station under a hood pro-vided with an off-gas vent to the radwaste area ventilation system. Liquid leakage from the valves in the hood and sampling effluents are drained to the vaste disposal system.
9 11 INSTRUMENT AND SERVICE AIR SYSTH4 9 11.1 DESIGN EASES The Instrument and Service Air System is designed to provide a reliable con-tinueas supply of dry, oil-free compressed air for pneumatic instrument opera-i k-'
tion and for control of pneumatic valves. The system also supplies air to service outlets throughout the station for operation of pneu=atic tools or other requirements. The compressors supply air at a pressure of 100 psig with pressure reduced as necessary for the various service requirements.
9
11.2 DESCRIPTION
AND OPERATION Figure 9-12 is a schematic diagram illustrating the features of the system.
During normal operation, two of the three full capacity, nonlubricated air compressors operate continuously to supply station instrument and service air requirements. The remaining air compressor is placed in automatic standby from the control room and will start upon decrease of supply air header pressure.
Operation of the standby air compressor is annunciated in the control room.
To insure instrument air supply, the service air header is automatically valved off when the compressed air system pressure drops to a preset value.
Protection against loss of instrument air is provided by redundancy in active components comprising the instrument air system.
In addition, in the event of a loss of all instrument air supply, all pneumatically operated valves are arranged to assume their respective safe positions.
To maintain acceptable purity and low dew point, a dual tower dessicant. type air dryer and two full capacity filters are provided from which the instrument air supply is split into headers; branch lines are taken off to supply all areas of the station.
The power source for the compressor motors is the normal a-c distribution system.
Automatic backup from the emergency generators is provided.
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FILE:
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/ FROM: Consumers Power Company DATE OF DOC DATE REC'D LTR TWX RPT OTHER
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TO:
ORIG CC OTHER SENT AEC PDR XX 7"X Mr ciambusso 3 signed SENT LOCAL PDR CLASS UNCLASS PROPINFO INPUT NO CYS REC'D D
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DESCRIPTION:
ENCLOSURES:
Ler notarized 5-5-75...trans the following.:
Amdt #29 to the License Application:
Consisto ing of revision to the PSAR to reflect mis-cellaneous design changes....(70 cys encl rec ' d)
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