ML20039E754

Forwards Responses to FSAR Questions & Text Changes. Unresolved Safety Issue A-12,fracture Toughness of Steam Generator & Reactor Coolant Pump Supports Info Encl

ML20039E754
Person / Time
Site:	Byron, Braidwood
Issue date:	12/31/1981
From:	Tramm T COMMONWEALTH EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
REF-GTECI-A-12, REF-GTECI-SE, TASK-A-12, TASK-OR NUDOCS 8201110317
Download: ML20039E754 (211)

Text

{{#Wiki_filter:r-4 .s Oi Ons First National Plaza. Chicago, lltinois Commonwealth Edison Address Reply to: Post Office Box 767 Chicago. Illinois 60690 g\\ g,j ' t. v S E.. 5 N % 9q ~ 4 ? ?o' ,f.~ C ^ / m ,\\ ..f j '~. ~ ' 1.. IN [. December 31, 1 8 Q ( \\'OP Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclea r Regulatory Commission Washington, DC 20555

Subject:

Byron Station Units 1 and 2 Braidwood Station Unit s 1 and 2 Advance FSAR In formation NRC Docket Nos. 50-454/455/456,'457

Dear Mr. Denton:

This is to provide advance copies of information which will be included in the Byron /Craiuwooo FSAR in the next amendment. Attachment A to this letter lists the information enclosed. One (1) signed original and fifty-nine (59) copies of this letter are provided. Fi f teen (15) copies o f the enclosures are included for your review and approval. Please address further questions to this o f fice. Very truly yours, 4" // T.R. Tramm Nuclear Licensing Administrator Pressurized Water Reactors At.ta chmen t 3129N ol .s //I bO?No?bb$jf4 PDR

c--- a. Attachment A List of Information Enclosed I. FSAR Question Responses ~ 10.44 40.162 thru 40.183 331.33 130.43 II. FSAR Text-Changes - 1. Text changes in Sections 3.6, 3.7 and 3.9 to document resolution of MEB items listed below: i N1 N12 N2 N16 N3 N18 N5 N21 j N6 N22 i N7 N23 i N9 N24 1 2. Text change in Section 7.3 to resolve ICSB agenda item 97. j 3. T e.. t changes in Section 9.2 (and the response to 10.44) to document the resolution of ASB open item no. 3

4. ' A descripton of the Condensate Cleanup System f or Section l

10.4.6 of the FS4R (Chemical Engineering Branch). 5. Ne

SAR Figures for Section 3.8 corresponding to the respos. e to Question 130.43.

j III. Miscellaneous (Not for FSAR) 1. Unresolved Safety Issue A-12 2. Additional information to close open items for the i Structural Design Audit: 4A 19 23 27 29 30 3129N t-a

C/B-FSAR QUESTION 010.44 "Your response to Q010.9 is not complete. You have indicated that tests of the reactor coolant pumps per-p formed by Westinghouse indicate that the pumps can func- ) tion satisfactorily-for 10 minutes without component a cooling water supply.- Low component cooling water flow alarms and high component cooling water temperature ) alarms from the reactor coolant pump oil coolers are provided in the control room to indicate a loss of com-ponent cooling water supply. Operator action can be p taken within the 10 minutes available to secure the reactor coolant pumps. It is our position that the 4 [ alarm indication of loss of component cooling water p flow to the reactor coolant pumps is safety grade and l meet the requirements for Class la instrumentation. l Verify your response accordingly."

RESPONSE

Redundant safety-related indication of component cooling p L water flow to the reactor coolant purp thermal barrier is provided and alarmed on the main control board. In addition, the followin.J five temperatures are input to the plant computer: b l. RCP motor stator winding temperature, 2. RCP motor upper radial bearing Lemperature, 3. RCP motor upper thrust bearing temperature, 4. RCP motor lower radial bearing temperature, and, 5. RCP motor lower thrust bearing temperature. These temperatures would be alarmed and printed byrthe plant computer should any of them go out of range. These would e provide sufficient indication that the pump was too warm } and the operator would be able to trip the pump. l b f b i l Q10.44-1 l b j

B/B-FSAR j_ QUESTION 040.162 "Your response to 0040.12 on the motor operated valves requiring power lockout is not complete. Subsection 8.1.10 of the FSAR does not describe how the power to the MOVs is disconnected. Provide this information. The acceptable method to ensure that electrically operated l valves will not by themselves, spuriously move for any operating or postulated plant condition is to have the motive power locked out (circuit breaker racked out). However the electrical schematics provided show that j the power to the MOVs is disconnected on receipt of i a safety injection signal. This is not in accordance with BTP EICSB 18 PSB and, therefore, not acceptable. Clarify this discrepancy."

RESPONSE

The design utilizes a power lockout " circuit breaker-starter" in series with the " circuit breaker-starter" for each of the MOV valves. This power lockout starter can be controlled from the main control room. Example: Motor-ooerated valve 1SI8808A 1. The power lockout feature is the starter at MCC 131X2, compartment B2 (which is the upstream starter for MOV , 1SI8800A), see key diagram 6/20E-1-4008E. 2. This starter contact can be opened or closed by the selector switch at.the main control room panel 1PM06J (see schematic drawing 6/20E-1-4030 SI33). This feature meets the branch technical position EICSB 18, paragraph B.3. 3. The starter for MOV 1SI8808A is at MCC 131X2A, compartment A3 which is downstream of the power lockout starter as mentioned in Item 1 (see drawing 6/20E-1-4008E). 4. When the power lockout starter contact opens (as mentioned in Item 24 there is no electrical power supply furnished to valve motor ISI8808A (see drawing 6/20E-1-4008E) or there is no power supply furnished to the control transformer which supplies the power supply to the control circuit. 5. During power lockout, if the safety injection signal (or any signal) is present, it will not change the valve position since there is no power to the motor or control circuit. Example for valve 1SI8802A, see drawing 6/20E-1-4030 Sill. When the safety injection signal is Q40.162-1 l ] t

I ,i - B/B-FSAR present, contact 7-8 of K603 will close but the starter "open" coil (On will not energize since there is no power to the control transformer (480-120v). Also if the 3-phase starter contact (0) gets stuck in j the close position, the valve will not change position ) since there is no electric power to the valve motor. r Because of the design which is explained in Items 1 through 5, there is no need to rack the circuit breaker out or separate the control circuit. However, for each passive valve, the individual breaker (located at its respective MCC compartment) will be manually tripped (Opened) and tagged "Out of Service." Verification ~ will be made that the control switch (located at the main L control board) of each passive valve is in the desired position t and tagged "Out of Service." This additional precaution will assure that the passive valves always remain in the a correct position. Q40.162-2 wm a .)

B/B-FSAR QUESTION 040.1.63 "The response to 0040.8 is incomplete. Describe the L procedures to be followed, when temporary jumpers are installed, that will ensure removal of jumpers after the testing procedure is completed." h

RESPONSE

Temporary jumpers and lifted leads are covered by Byron Operating Procedure; BAP 300-5. Jumper installation and removals are logged on the jumper log; BAP 300-T1. Lifted leads are logged on the lifted lead. log; BAP 300-T2. Jumpera f and lifted leads are documented in the appropriate pre-operational test during the testing program. i L u r ( f \\ V l 6 b Q40.163-1

B/B-FSAR 3 QUESTION 040.164 "You state in Section 8.3.1.1 that automatic transfer of each non-Class lE 6900 V or 4160 V switchgear from the UAT to SAT is provided upon a loss of power from 3 the UAT or vice versa. Provide a detail description of the transfer schemes including test interval." 2

RESPONSE

i 3 The above referenced statement in Subsection 8.3.1.1 was 9 eliminated in Amendment 21 of the FSAR dated July 1979, L to avoid confusion. i Each non-Class lE 6900-V and 4160-V switchgear can be supplied .from either the UAT or the SAT. If a switchgear is supplied from the UAT and the control switch for the SAT feeder breaker is in the trip position and the UAT feeder breaker trips for some reason other than a fault on the bus, then the SAT feeder breaker will automatically close. The converse is true if the switchgear is supplied from the SAT. Schematic diagrams which illustrate the transfer scheme in more detail were provided. The operating time for the breakers is less than 10 cycles. Operating times are as follows: Main contacts open: 6 "a" contacts 2.7 1 0.4 cycles 4.16 kV 2.5 1 0.4 cycles 6.9 kV "b" contacts 3.0 1 0.4 cycles Main contacts close: 5.1 1 0.8 cycles 4.16 kV 6.9 1 0.6 cycles 6.9 kV The second source of offsite power to the ESF buses is proviced by the corresponding SAT of the other unit. The tie breakers are manually operated. 040.164-1 l i i

B/B-FSAR QUESTION 040.165 "Your response to 040.13 is not complete. You do not i describe how you will test the penetration primary and backup protective devices. Also, provide the. test interval." 4

RESPONSE

The last paragraph of our response to Question 040.13 stated the following regarding testing of the penetration primary and backup protective devices. "There are no provisions for periodic testing of pen-etrations or fuses under simulated fault conditions because such testing would be detrimental to the pen-etrations and fuses. The overcurrent relays and circuit breakers that provide the primary and backup protection will be periodically tested under simulated fault conditions to demonstrate that the overall coordination scheme remains uithin the specified limits." The test interval will be at least ohce every 18 months during refueling outages. The following will supplement our response to Question 040.13 which identified each type of typical circuit that penetrates. the reactor containment and described the type of protective devices used to provide primary and backup protection: a4 Circuits Energized from 125 Vdc Distribution Panels The 125 Vdc emergency lighting cabinet is the only such load. The primary protection consists of a 125-Vdc molded case air circuit breaker that feeds the penetration directly. The backup protection consists of another molded case air circuit breaker similar (and connected in series 4 to that provided for primary protection. b4 Circuits (Solenoid Ooerated Valves 4 Energized From 125 Vdc Distribution Panels The primary protection consists of a nonrenewable cartridge-type fuse that feeds the penetration directly. The backup protection consists of another fuse similar (and connected in seriest to that provided for primary protection. c4 The primary and backup protection for the rod control system lLft and gripper coil circuits consists of fuses connected in series. Q40.165-1

B/B-FSAR d) The primary and backup protection for the rod position indication data cabinets consists of two 120 Vac molded case air circuit breakers connected in series. 4 I e O 4 i 4 5 1 e 1 4 Q40.165-2

a B/B-FSAR QUESTION 040.166 "We have not received the qualification test reports for the containment electrical penetrations assemblies requested in Q040.14. Provide this information." }

RESPONSE

The qualification test reports for the.. containment electrical penetrations do not contain the information necessary to i evaluate the penetration failure described in Question 040.14-. We have addressed this failure in our revised response to Question 040.14. l l i 4 040.166-1 w ~,

9 B/B-FSAR QUESTION 040.167 1 "The response to question 040.62 omitted the essential { parts of the question. e; 1. Provide a tabulation of each Class lE system required to bring the plant to a safe cold shutdown. 2. Provide the cable routing locations for the power, control, and instrumentation cables for each system. (, (This is best accomplished with a drawing or a sketch). "

RESPONSE

The complete response to Questions 040.62 and 040.167 is covered by the Applicant's " Fire Protection Safe Shutdown Analysis." 4 m o \\ l Q40.167-1

l B/B-FSAR 1 3 ~ QUESTION 040.168 " Provide,a complete description of the load sequencer that is used to connect'the ESF loads to the emergency buses when power is being supplied _from the diesel generator after a loss of offsite power. Also provide a description of how the load sequencer adds loads to the emergency bus after a lOCA and offsite power is availabin. The description should include details of ho'w loads are i shed and buses are isolated. Discuss the conditions required to connect the diesel. generator to the emergency bus and how the ESF loads are connected to the emergency 4 bus by the load sequencer." 1 i l

RESPONSE

The automatic loading of the emergency bus is accomplished i by the sequencing panels. There is one panel per diesel j generator. The sequencing panels are electrically and phys-ically separated per Electrical Division. i i l The sequence circuit (Ref. Schematic 6/20E-1-4030EF02 typica14 t uses relays and timers to accomplish the function of se-quencing. Upon loss of offsite power, the emergency bus undervoltage relay (via auxiliary relays 4 : a4 trips all loads on the bus except the 4160-480V transformers, b4 trips i the feed breakers to the emergency bus, 'and c4 interrups power to the sequencing control circuit to ensure that the ] circuit resets, if in testing mode. Sequencing will begin j provided: 14 the diesel generator breaker has closed and power is restored to the sequencing circuit, and bl inter-unit (cross-tien feed breaker and SAT breaker are open. S.equencing is accomplished via timers whose contacts energize i relays which in turn interlock the start circuits of the required loads. The time intervals for sequencing loads are as indicated in Table 8.3-1 of the FSAR. i i Because the sequencing circuit will not have power for 10 seconds while the diesel is coming up to speed, there is a 10 second difference between the time settings in the I circuit, and those in Table 8.3-1. j During a safety injection signal with sustained offsite power, the required loads are started immediately. With i sustained offsite power, the timers are bypassed and the L relays are energized directly by the safety injection signal (K6084 via auxiliary relay SARB. The bypass circuit is 4 1 blocked on loss of offsite power by at the Unit 1 SAT breaker (52/a contact 4 and b4 the inter-unit (cross-tie 4 breakers 52/a contacts., eeries with the Unit 2 SAT feed breaker (52/a contacte. Q40.168-1 g

B/B-FSAR The conditions required to connect the diesel generator to the emergency bus are as follows (Ref. Typical S.chematic 6/20E-1-4030DG024: Automatic a4 SAT breaker open and not locked out (486-1422) i b4 Inter-unit feed breaker open (52/b 14246 c4 Tie breaker open (52/b 14214 d4 Diesel generator breaker not locked out (486-14234 e4 Diesel generator reaches rated voltage and speed (DGlBX4. Manual a4 Control switch at Main Control Board closed (CS/C4 b4 Synchronizing switch closed (SS/CN4 c4 Synchro-check relay contact (HACR-14 closed.~ d4 Diesel generator reaches rated speed and voltage (DGlBXI e4 Feed frou SAT breaker not locked out -(486-14224 f4 Diesel generator breaker not locked out (486-14234 i 1 4 4 040.168-2

B/B-FSAR i QUESTION 040.169 l "Concerning the emergency load sequencers which are associated with the offsite and onsite power l sources we require that you either provide a separate sequencer for offiste and onsite power (per Electrical Division 4 or a detailed analysis to demonstrate that there are no credible sneak circuits or common i failtyre modes in the sequencer design that could render both onsite and offsite power sour'ces unavailable. In addition provide information concerning the reliability of your sequencer and reference design detailed drawings."

RESPONSE

1 The Byron /Braidwood design utilizes a load sequencer only when the onsite (diesel-generatore power source is being used. The necessary loads are connected simultaneously (block load 4 when the offsite power source is in use. 4 -t 'I i l 040.169-1

B/B-FL}R j QtTESTION 040.170 "The response to 040.70 attempts to justify sharing of the de systems of Unit 1 and 2, which is not permitted by Regulatory Guide'1.81, with the statement, 'The provision of the interconnection of the power supplies increases the overall reliability of the de system since power could be obtained from the non-redundant de bus of the other unit if necessary.' As noted in the FSAR, administrative controls are the only restrictions provided to prevent the inter-connection of dc bus 111 (112) of Unit 1 to de bus j 211 (212) of Unit 2. "The staff believes that any design having inter-connections between Class lE batteries in a multi-unit plant will affect the reliability and availability of de power to both units. Therefore, the applicant is required to meet the staff position, i.e., 'dc systems in multi-unit nuclear power plants should not be shared' as per P'.G. 1.81, position C.l."

RESPONSE

The Applicant also believes that the provisions of administratively controlled, manually actuated, inter-connections between the non-redundant Class lE-dc buses will affect (i.e., increase) the overall reliability and availability of the de systems for each unit in that it will provide a means for manually providing power to a de bus at a time when it would otherwise have to be out-of-service (e.g., to perform a battery 3 ~ discharge test during a refueling outage, to replace a damaged cell, etc.). We believe that the intent of Regulatory Guide 1.81 (Position C.1) was to disallow " normal" sharing of de systems, not to disallow the temporary connection of one de bus to a source in the other unit during periods of testing and/or main-tenance. That this was the intent is evident from the " Discussion," in Regulatory Guide 1.81, Part B, second paragraph, first sentence, which reads as follows: "S,h.aring of onsite power systems at multi-unit power plant sites generally results in a reduction in the number and capacity of the onsite power sources to levels below those required for the same number of units located at separate sites." Q40.170-1

B/B-FSAR The " interconnection" provided in the Byron /Braidwood design does not result in a reduction in either the number or the capacity of the dc power sources. i.e., the number and capacity of the de power sources for each of the two units are exactly the same as l they would be if the units were located at separate i sites. The Applicant therefore believes (at that the Byron /Braidwood design fully complies with the intent of Position C.1 of Regulatory Guide 1.81, and (b4 that the added i reliability, availability, and flexibility of operation afforded by the interconnection justifies its retention. The Technical Specification will preclude closing the crosstie during any mode of operation except Mode 5 (Cold Shutdownl and Mode 6 (Refueling 4. Closing the crosstie (when not allowed 4 will be administrative 1y l controlled by locking the breaker in the open position. l l t I Q40.170-2 i

7 i B/R-FSAR QUESTION 040.171 "The response to question 040.65 refers to Mr. H. K. Stolt's letter of December 7, 1976 which argues against identification of cables by marking every five feet which was a proposed modification for~IEEE Standard 384-1974. Stolt argues that (14 there is no technical or safety-related basis for requiring the cable to be marked on the fashion described, (26 marking the cable does not justify the costs, and (34 coloring and cable jacket could degrade.the quality of the cable. "Stolt mentions the negative aspects of cable identification without offering any alternative method of cable marking to facilitate visual verification that the cable instal-lation is in conformance with separation criteria. "The FSAR proposes to identify cables with a unique number of color-coded tags at each terminating point inside electrical equipment and in all intervening manholes, cable roons, pull boxes, etc. This system of identification will leave many thousand feet of unmarked black-jacketed safety-related cable installed in cable trays throughout the plants, and it will be extremely difficult to verify that adequate separation is maintained between redundant discussed cables. "The proposed method of cable identification is not an acceptable alternative to the requirements of Regulatory Guide 1.75. The applican,t should review his proposed means of cable identification and submit a method of cable marking that will aid ready visual verification that the cable installation is in conformance with separation criteria."

RESPONSE

As stated in Subsection 8.3.1.3.4 "All power, control, and i instrumentation cables are identified by a unique number of permanent color-coded tags at each terminating point in switchboards, switchgear, motor-control centers, motor conduit boxes, control cabinets, equipment cubicles, etc., and in all intervening manholes, cable rooms, pull boxes, etc. The tags shall be color-coded as in Table 8.3-4, allowing positive identification of safety-related cables." The concern expressed in Question 040.171 is that this system of identification does not facilitate " visual verification that the cable installation is in conformance with the separation criteria." Additional verification can be obtained as described below. Q40.171-1

B/B-FS.AR The cables installed at Byron /Braidwood can be divided into two categories: (a4 those purchased under separate cable specifications (representing approximately 95% of the cablest, and (b4 those supplied by equipment vendors with the equipment (representing approximately 5% of the cables 4. a. The cables purchased by the separate cable specifications are supplied with the following information marked on tne jacket: Manufacturer, number of conductors, conductor size, voltage rating, specification number, purchase order number, cable type code, station reel number and sequential footage. These markings are imprinted on the jacket every 30 inches for power and control cable and every 24 inches for instrumentation cable. After the Electrical Installation Contractor installs a cable, the Contractor records (in addition to other information4 the cable reel number, the beginning footage marker and the end footage marker on the cable pull card. A Cross Reference Index will be available at ^ each site which will list (in sequencel the cable type code, cable reel number, cable footage markings, cable number, and the cable segregation code. This index will allow an inspector to verify the installation (including separationi of any cable by using the following procedure: ~ 14 For the selected cable, note the cable type code, reel number, and sequential footage markings on the jacket. 24 Enter the Cross Reference Index at the cable type code and cable reel number. Locate the corresponding sequential footage, opposite which will be both the cable number and segregation code. 34 Cospsre the segregation code in the index with the segregation code marked on the cable tray in'which the cable was found. 44 If the segregation code of the cable (in the cross-index4 agt.3s with the mark on the tray, the cable separation is correct. If the segregation code of the cable (in the cross-index4 does ~; not agree with the mark on the tray, the cable segregit-ion is not correct. b. The cablps nupplied by equipment vendors are identified as descri5bd in Subsection 8.3.1.3.4 (first paragraph of resp.Unch abovel. Q40.171-2

B/B-FS.AR QUESTION 040.172 "In the response to 0040.93, paragraph 2 states that surveillance testing meets requirements of applicable NRC guides. However, the testing procedures specified for the diesel generator in the Technical Specification subparagraph 4.8.1.1.2.c are not in agreement with the requirements of Regulatory Guide 1.108 positions C.2.2.1, 1 C.2.a.2, C,2.a.3, C.2.a.4, C.2.a.5, and C.2.a.6. Correct the tech specs to conform to the above."

RESPONSE

Surveillance testing will conform to the requirements of Regulatory Guide 1.108 except for the requirements in Regulatory Positions C.2.a.5, C.2.b, C8, C9c, and C9d. 4 Appendix A (Regulatory Guide 1.1084 will be revised to identify the specific exceptions and to provide a detailed description l of Applicant's position and justification for same. The appropriate sections of the Technical Specifications will j be revised accordingly. l 4 4 e l Q40.172-1 i

B/B-FSAR QUESTION 040.173 "The response to 0040.69 is incomplete. Provide the ampere-hour rating of each 125V dc battery and the KVA output reting of the battery charger."

RESPONSE

a. Batteries 1. Each of the (two/ unit 4 125-V Class lE batteries consists of 58 Gould Type NCX-1200 lead-calcium cells. The (nominally 1,200 kH4 discharge characteristics, and other specifications for these cells are shown on the attached Gould Catalog S_h.eets (24. 2. Battery capacity is also addressed in Question 040.182. I b. Chargers Each of the (two/ unit 4 125-V Class lE battery chargers is a Power Conversion Products, Model No. 3S,-130-400. The output rating of each charger is 400 amperes at 130 volts (nominal 4 dc. This_is equivalent to a rating of 52 kW (nominal 4. Y } l Q40.173-1

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i B/B-FSAR QUESTION 040.174 "The FSAR does not provide the time interval for the diesel, generator to attain rated voltage and frequency, and ready to accept load. Provide this information. Also describe the conditions which will initiate automatic starting of the onsite diesel generators."

RESPONSE

The diesel generator is designed to attain rated voltage and frequency and be ready to accept load 10 seconds after the receipt of an autcmatic start signal. Automatic starting is initiated by a safety-injection signal and/or a loss of offsite power signal. The safety-injection signal comes from the reactor protection system logic and the loss of of fsite power signal originates f rom under-voltage relays on the 4160-V ESF bus which the diesel-generator feeds. The conditions which will initiate automatic starting of J the diesel-generator are described in Subsection 8.3.1.1.1, 3 pages 8.3-5 and 8.3-6 of the FSAR. i 5 040.174-1

B/B-FSAR QUESTION 040.175 "Your response to 0040.6 reflects a misunderstanding of the question. The purpose of the 0040.6 was to require that one alarm be provided to explicitly indicate conditicns such that a diesel generator is incapable of responding to an automatic emergency start signal. As outlined, you should separate the alarms so tha' there is a dedicated alarm for the conditions that render a diesel incapable of responding to an automatic start signal. Some of the examples are as follows: a at Diesel generator breaker feeding to the ESF bus is racked out. b4 Diesel generator is started locally for test or maintenance and the operator forgets to return the control to the control room after the activity is completed. 1 cl Failure of DC control power supply to the diesel generator breaker. d4 Failure of the control power supply to the start air train circuits. e4 Af ter testing the diesel generator, the operator forgets to reset the exciter pushbutton. l "You may have more than the conditions as mentioned above which would render a diesel generator incapable of responding to an automatic emergency start signal. "We require that you provide a list of all these conditions and that your design comply with Regulatory -Guide 1.47 ' Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems.'"

RESPONSE

In order to meet the requirements of Regulatory Guide 1.47 and inform the operator at the system. level of conditions that would render a diesel generator unable to respond to an automatic emergency start signal, the following inputs are provdied to the Byron /Braidwood Station's Equipment Status Display (ESD4 systems (the listing is for Unit 1; the Unit 2 listing is identicall: Q4 0.17 5-1

B/B-FS.AR a. Diesel Generator A Output Breaker Not Available. b. Diesel Generator D Output Breaker Not Available, c. Diesel Generator A Engine / Generator DC Control Power Off. d. Diesel Generator B Engine / Generator DC Control Power Off. e. riesel Generator A Control Switch Maintenance Lockout. f. Diesel Generator B Control Spitch Maintenance Lockout. 9 Diesel Generator A Cooling Water Flow Low. h. Diesel Generator B Cooling Water Flow Low. i. Diesel Oil S.torage Tank A Level Low. i. Diesel Oil S,torage Tank B Le/el Low. k. Diesel Oil Storage Tank C Level Low. 1. Diesel Oil S.t.orage Tank D Level Low. m. Diesel Oil Day Tank A Level Low. n. Diesel Oil Day Tank B Level Low. The ESD system level indication for the diesel generators is dedicated to these inputs only. e i 040.175-2 l

B/B-FSAR QUESTION 040.176 "Your response to 0040.5 is not sufficient to allow a fully independent review of this aspect of your design. We require that the adequacy of station electric distribution system voltages should meet the following criteria. A.

Background

Events at the Millstone station have shown that adverse effects on the Class lE loads can be caused by sustained low grid voltage conditions when the Class lE buses are connected to offsite power. These low voltage conditions will not be detected by the loss of voltage relays (loss of of fsite power 4 whose low voltage pickup setting is generally in the range of.7 per unit voltage or less. J The above events also demonstrated that improper voltage protection logic can itself cause adverse effects on the Class lE systems and equipment such as spurious load shedding of Class lE loads from the standby diesel generators and spurious separation of Class lE systems from offsite power due to normal motor starting transients. A more recent event at Arkancas Nuclear One (ANO4 station and the subsequent analysis performed disclosed the possibility of degraded voltage conditions existing on the Class lE buses even with normal grid voltages, due to deficiencies in equipment between the grid and the Class lE buses or by the starting transients experienced during certain accident events not originally considered in the sizing of these circuits. B. Branch Technical Position 1. In addition to the undervoltage scheme provided to detect loss of of fsite power at the Class lE buses, a second level of undervoltage protection with time delay should also be provided to protect 4 the Class lE equipment; this second level of undervoltage protection shall satisfy the following criteria: al The selection of undervoltage and time delay setpoints shall be determined from an analysis of the voltage requirements of the Class lE loads at all onsite system distribution I levels; Q40.176-1

B/B-FS.AR b4 Two separate time delays shall be selected for the second level of undervoltage protection based on the following conditions: 1) The first time delay should be of a duration that establishes the existence of a sustained degraded voltage condition (i.e., something longer than a motor starting transient 4 Following this delay, an alarm in the control room should alert the operator to the degraded condition. The subsequent occurrence of a safety injection actuation signal (SIAS4 should immediately separate the Class lE distribution system from the offsite power system. 2) The second time delay should be of a limited duration such that the permanently connected Class lE loads will not be 4 damaged. Following this delay, if the operator has failed to restore adequate voltages, the Class lE distribution system should be automatically separated i from the offsite power system. Bases and justification must be provided in support of the actual delay chosen. ~ c4 The voltage sensors shall be designed to i satisfy the following applicable requirements derived.from IEEE S,td. 279-1971, " Criteria j for Protection Systems for Nuclear Power Generating Stations": i 1) Class lE equipment shall be utilized and shall be physically located at and electrically connected to the Class lE switchgear. i 26 An independent scheme shall be provided for each division of the Class 1E power l system. 34 The undervoltage protection shall include coincidence logic on a per bus basis 2 to preclude spurious trips of the offsite power source; 44 The voltage sensors shall automatically i initiate the disconnection of offsite power sources whenever the voltage set point and time delay limits (cited in item 1.b.2 above4 have been exceeded; Q40.176-2

B/B-FS.AR 54 Capalility for test and calibration during power operation shall be provided. 64 Annunciation must be provided in the control room for any bypasses incorporated in the design. d4 The Technical Specifications shall include limiting conditions for operations, surveillance requirements, trip setpoints with minimum and maximum limits, and allowable values for the second-level voltage orotection sensors and associated time delay devices. 2. The Class lE bus load shedding scheme should automatically prevent shedding.during sequencing of the emergency loads to the bus. The load shedding feature should, however, be reinstated upon completion of the load sequencing action. The technical specifications must include a test requirement to demonstrate the operability of the automatic bypass and reinstatement features at least once per 18 month's during shutdown. t In the event an adequate basis can be provided for retaining the load shed feature during the ) above transient conditions, the setpoint value in the Technical Specifications for the first level of undervoltage protection (loss of offsite I power 4 must specify a value having maximum and I minimum limits. The basis for the setpoints and limits selected must be documented. 3. The voltage levels at the safety-related buses should be optimized for the maximum and minimum load conditions that are expected throughout the anticipated range of voltage variations of the offsite power sources by appropriate adjustment ~ of the voltage tap settings of the intervening transformers. The tap settings selected should be based on an analysis of the voltage at the a terminals of the Class lE loads. The analyses performed to determine minimum operating voltages should typically consider maximum unit steady state and transient loads for events such as a unit trip, loss of coolant accident, startup or shutdown; with the of fsite power supply (gridl at minimum anticipated voltage and only the offsite source being considered available. Maximum voltages should be analyzed with the offsite power supply (grid 4 at maximum expected voltage concurrent with minimum unit loads (e.g., Q40.176-3 l

t B/B-FSAR cold shutdown, refueling). A separate set of the above analyses should be performed for each available connection to the offsite power supply. 4. The analytical techniques and assumptions used in the voltage analyses cited in item 3 above must be verified by actual measurement. The verification and test should be performed prior 1 to initial full power reactor operatien on all nources of offsite power by: 1 a) loading the station distribution buses, including all Class lE. buses down to the 120/208 y level, to at least 30%; b4 recording the existing grid and Class lE bus voltages and bus loading down to the 120/208 volt level at steady state conditions j and during the starting of both a large Class lE and non-Class lE motor (not concurrently 4; d Note: To minimize the number of instrumented l locations, (recorders 4 during the l motor starting transient tests, the bus voltages and loading need only 4 be recorded on that string of buses which previously showed the lowest analyzed voltages from item 3 above. cl using the analytical techniques and assump-tions of the previous voltage analyses cited i in item 3 above, and the measured existing grid voltage and bus loading conditions recorded during conduct of the test, calculate a new set of voltages for all the Class lE buses down to the 120/208. d4 compare the analytically derived voltage values against the test results. "With good correlation between the analytical results and the test results, the' test verification requirement will be met. That is, the validity of the mathematical model used in performance of the. analyses of item 3 will have been established; therefore, the validity 4 of the results of the analyses is also established. In general the test results should not be more than 3% lower than the analytical results; however, the j dif ference between the two when subtracted f rom the voltage levels determined in the original analyses should never be less than the Class lE equipment rated voltages." l Q40.176-4 I

B/B-FSAR

RESPONSE

There are two redundant and independent 4-kV emergency buses and each has two levels of undervoltage protection: 14 loss of power, and 26 degraded grid-voltage. The relays will be connected to the existing potential transformers on the bus. The first level of undervoltage protection is nrovided by induction disk type undervoltage relays. The second level of undervoltage protection is provided by instantaneous undervoltage relays ylth delayed drop-out. The voltage and time set points will be determined from an analysis of the voltage requirements of the safety-related loads and actual field measurements of bus voltages under various motor-starting conditions. The approximate pick-up voltage for the first level of protection is 70% of rated voltage. The preliminary setting for the second level of undervoltage protection is 92% of rated voltage. There is a 10 second time delay to avoid alarms on transients, and if the degraded voltage is not corrected within 5 minutes, the bus will automatically disconnect from the offsite power source and connect to its onsite diesel generator. During a sustained degraded grid voltage condition, the subsequent occurrence of a SI (accidenti signal will (imme-diately6 trip the offsite power supply to the 4 kV ESF buses. Testing will be conducted during refueling outages so spuriots trips during testing will not affect plant operation. The circuit.will be designed to prevent automatic load shed-ding of the emergency power buses once the onsite sources are supplying power to all sequenced loads on the buses. The load shed interlock feature will use the "b" contact of the respective diesel generator breaker. This interlock will defeat the load shedding feature while the loads are being fed from the onsite power source. The load shed feature will be reinstated when the diesel generator breaker is open and the loads are fed from the offsite source. t l Q40.176-5

B/B-FSAR QUESTION 040.177 "You have not responded to parts (b4 and (c4 of question Q040.85. Provide this information. In part (b4 of 0040.85, you were asked to show the differential relay in the D/G breaker trip circuit and show how this trip is retained on receipt of a safety injection signal. Provide a revised logic diagram 8.3.2."

RESPONSE

As stated in our response to 0040.85, Figure 8.3-2 has been revised to correct the breaker control logic.

However, the differential relay is not shown in the breaker trip circuit since it is not directly in the circuit.

The differ-ential relay actuates the lock-out relay (shown as device 486-14134 which trips the breaker. The logic diagram correctly shows the lockout relay directly tripping the breker while the remaining breaker trips (overcurrent, reverse power, the diesel generator shutdown relay 4 are "ANDED" with a Safety Injection "NOT" signal. In this way, the differential trip is retained during nocmal and accident (safety injection 4 conditions and the remaining trips are bypassed by the safety injection signal. i l l 040.177-1 i

-~ i B/B-FSAR QUESTION 040.178 "Your response to 0040.73 is not adequate. You state i in Appendix A, Page A1.75-1, that,'you comply with Regulatory Guide 1.75. Regulatory Position C.1 of Reg. Guide 1.75 states that breakers that trip on' receipt of a signal other than one derived from the fault current or its effects (e.g., an accident signal 4 are acceptable isolation 3evices.. Your approach of coordinating the circuit breaker (which feeds the Non-Class lE load from Class lE bus 4 with the upstream circuit breaker is not in accordance with the above position and, therefore, not acceptable. Acceptable alternate approaches to the above positions would be two breakers or breaker and a fuse in series. " Describe in detail the isolation devices being used on Byron /Braidwood Stations at all voltage levels (4.16 kV, 480V, 120V ac, 125V dc4."

RESPONSE

i The current issue of Appendix A (Page A1.75-ll does not say that Applicant complies with Regulatory Guide 1.75. We agree that Appendix A (Regulatory Guide 1.75 position description 4 requires clarification. The Applicant's position relative to Regulatory Guide 1.75 (Rev. 2, S.eptember 1978n is described in revised Appendix A. 1 j There are no non-safety-related loads supplied frca the safety-related 4.16-kV buses other than non-safety -related i i buses 143 and 144 which may be manually connected to safety-i related buses (41 and 142 (respectively4 as described in the last paragraph of Subsection 8.3.1.1 (page 8.3-2). Non-safety-related loads supplied from safety-related buses at 480-V, 120 Vac, and 125 Vdc utilize circuit breakers 3 (actuated by fault currenti for isolation, all as described in revised Appendix A. Q40.178-1

J B/B-FSAR QUESTION 040.179 "The response to 0040.49 included single line diagrams for the Byron and Braidwood Stations. Figures 0040.49-3 ^ and 0040.49-4 do not match with the figures 8.2-1, 8.2-2, 8.2-6 and 8.2-7 provided with the FSAR. Correct this discrepancy and provide corrected drawings.

RESPONSE

Figures Q040.49-3 and 0040.49-4 have been corrected and transmitted. 4 4 i 4 'I i I i Q40.179-1 t

e B/B-FSAR. , QUESTION 040.180 "We are still waiting for your response to 040.83. Provide your response."

RESPONSE

Applicant will formally answer Question 040.83. j 4 i i 4 g l t 040.180-1 l l

B/B-FSAR QUESTION 040.181 "IEEE-387-77 Section 5.6.2.2(14 and Regulatory Guide 1.108 position C.l.b.3 recommend that the periodic testing of diesel generator units should not impair the capability of the unit to supply emergency power within the required time. The following discussion provides the necesary guidelines (requirements 4 to meet the objective of the above referred Industry Standard and Regulatory Guide. "The diesel generator unit design should include an emergency override of the test mode to permit response to bona fide emergency signals and return control of the diesel-generator unit to the automatic control system. A design which does not have such a feature would neces-sitate operator action of varying levels of complexity depending on the specific design and plant conditions, in order to enable a diesel generator in the test mode to respond to a bona fide emergency signal. The concern here is the possible consequent disabling of a D/G due to operator's inaction or wrong action thereby reducing the available safety margin with regard to onsite a-c power at a time when it is most needed. "Each diesel generator must be periodically tested at a frequency as specified in R.G. 1.108. This test frequency is normally once per month but could be as high as once every three days. The duration of each test is one hour. During a normal successful test the diesel generator is loaded on its bus with the governor operating in a droop mode, and the load carried by the diesel engine is a function of governor speed and speed droop setting. In order to enable a diesel generator in the test mode to respond to a bona fide emergency signal, the design must incorporate the following features for the stated plant conditions: 1. Accident Conditions During the periodic testing of a diesel generator, if a safety injection signal is generated, the diesel generator breaker should be tripped automatically. This will permit the unit to be cleared from parallel operation with the system and enable the diesel generator to attain the emergency standby mode. In this mode of operation, the diesel governor control should change automatically to the isochronous mode which will maintain the engine running at a synchronous speed corresponding to 60 Hz at the generator terminals. All noncritical protective trips except engine overspeed and generator dif ferential should be bypassed for Q40.181-1 4

B/B-FSAR j i this accident condition. Additionally, the voltage regulator should change to the automatic mode thereby maintaining the generator at a preset constant voltage. With the above actions complete, the diesel generator unit will be ready to accept the required load in the event of a loss of offsite power. 2. Loss of Offsite Power (LOOP 4 Conditions 4 Normally, during periodic testing of diesel generator, the diesel generator is paralleled with the of fsite power system. During such a test, should a LOOP occur, a LOOP signal would probably not be generated because the D/G would attempt to provide power to the bus and to the offsite system through the closed offsite power feedbreaker. In this case, the D/G breaker will trip on overcurrent or underfrequency and in some designs the D/G breaker also locks out for this condition. To assure the continued avail-i ability of the D/G unit it is essential that,the diesel generator breaker should not be locked' out for such overload conditions. At the same time, i the governor is shifted automatically from droop to isochronous mode and the voltage regulator to automatic mode. With the above actions complete, the diesel generator unit will be ready to accept its required load for LOOP conditions. 3.! Accident Conditions / LOOP j For simultaneous accident / LOOP condition or sequential accident and LOOP condition, the requirements stated in items 1 and/or 2 would be adequate to assure the restoration of the diesel generator from the a test n. ode to the emergency mode. Provide a discussion how you meet the above stated guidelines." i

RESPONSE

We agree with the guidelines set forth in this question to meet the requirements of IEEE-387 Section 5. 6. 2. 2 ( 14 i and Regualtory Guide 1.108 Position C l.b.3. The Byron / Braidwood diesel-generator system design includes emergency override of the test mode for both accident conditions (Safety Injec tion 4 and loss of offsite power (LOOP 4 to permit response to bona fide emergency signals and return control of the diesel-generator to the automatic control system. Upon 040.181-2

B/B-FSAR receipt of either a safety injection signal or a loss of offsite power signal, the governor is auto-matically shifted from droop to the isochronous mode and the voltage regulator is shifted to the automatic mode. Therefore,.the diesel-generator breaker controls, will be revised to trip the breaker upon receipt of a safety injection signal concurrent with the diesel generator operating in the test mode.

Also, the Byron /Braidwood diesel-generator breaker control scheme does not lock out the diesel-generator breaker on overcurrent or underfrequency which assures the availability of the diesel generator to accept its required load for LOOP conditions.

1 e e 1 Q40.181-3

B/B-FSAR QUESTION 040.182 "The description of the dc Power System in the FSAR is incomplete and should be expanded. Except for a general statement to the effect that the batteries are sized to carry their connected loads for the time periods shown in Table 8.3-5, there is no information provided about the capacity of the batteries. Another area with insufficient information is the subsection on Non-Safety-Related 125V de loads. Information about how the non-safety loads are separated from the Class lE bus is not included in the FSAR text or drawings. " Expand the section on the ac Power System to provide an adequate description of the battery capacity and how the non-safety loads are connected to and separated from the Class lE de systems."

RESPONSE

a. It is unusual for the NRC to request that specific / detailed electrical equipment rating / capacity data be included in the text of the FSAR (i.e., such detailed system descriptions / rating information is not required by Reg-ulatory Guide 1.70, Standard for SAR4. The following will supplement the detailed description of the de systems contained on pages 8.3-23 through 8.3-26 of the FSAR; however, Applicant does not propose to add same to the FSAR: DC Power Systems The station dc systems supply power-to the plant instru-mentation and control under all modes of plant operation. In addition, upon loss of ac power, the de systems provide power for emergency lighting and certain turbine-generator auxiliary motors. The de systems for each unit consist of one 250-V and one 125-V non-Class lE battery system and two independent and redundant Class lE 125-V battery systems. One (common for two units 4 48-V non-Class lE battery system is provided for the control of the non-Class lE switchgear at the river screen house. The Class LE 125-V battery systems supply power to Class lE loads without interruption during normal operations or DBA conditions. Each Class lE 125-Vdc system consists of one battery, one main distribution bus with molded case Q40.182-1

B/B-FSAR circuit breakers, one static battery charger, and local l distribution panels. Redundancy and independence of components precludes the loss of both systems as a result of a single failure. For Unit 1, Battery 111 supplies ESF Division 11 load requirements; Battery 112 supplies ESF Division 12 load requirements. There are no bus ties, or sharing of power supplies,:between redundant trains. Each Class lE 125-V battery, battery charger, and distri-bution panel associated with one ESF division is located in a seismic Category I room, physically separated from the redundant equipment. Electrical separation is also maintained to ensure that a single failure in one train does not cause failure in the redundant train. There is no sharing between redundant Class lE trains of equip-ment such as batteries, battery chargers, or distribution panels. Each Class lE 125-Vdc system has the capacity to con-tinuously supply all the connected normal running load while maintaining its respective battery in a fully charged condition. Each battery has a nominal rating of 1200 ampere-hours and is capable of carrying the various loads continuously, for the time periods indicated in Table 8.3-5 of the FSAR, in the event of a total loss of onsite and offsite ac power. 4 One 400 ampere capacity static battery charger supplied i by a Class lE MCC, is provided for each Class lE 125-V battery system. Protection is incorporated in'the battery chargers to preclude the ac supply source from becoming a load on the battery as a result of power feedback upon loss of ac input power. Bac'<up protection is incorporated by an overvoltage relay mounted on the charger, which trips the charger supply and annunciates the tripped condition in the control room. ~ Each battery charger is capable of floating the battery on the bus or recharging a completely discharged battery within a 24-hour period while supplying the largest combined demands of the various steady-state loads under all plant operating conditions. All battery areas are ventilated to prevent the accumulation of gases produced during charging operations. Each Class lE 125-V battery area is provided with an independent safety-related ventilation system. A -separate safety-related exhaust fan and duct is provided for each Class lE battery area. Q40.182-2

Leq l B/B-FSAR In addition, the battery cells are provided with explosion-resistant vent caps that prevent the ignition of gases within the cell from an ignition source outside the cell. The Class lE de systems are testable, independent, and conform to the requirements of Regulatory Guide 1.6 and 1.32. These systems meet the requirements of General Design Criteria 5, 17, and 18. b. The capacity of the Class lE batteries and chargers is covered in the response to Question 040.173. c. Non-Safety-Related Loads Connected to Class lE Bus (14 The single line diagram of this distribution system is shown on Drawing 6E-0-4001 (referenced in FSAR, S bsection 8. 3. 2.14 and on the Station One Line Diagrams, 6E-1-4001A and 6E-2-4001A. This connection is described in Subsection 8.3.2.1.2. (24 The isolation of the non-safety-related loads from the safety-related (Class lE4 bus, as the result of a fault on non-safety-related circuits, is performed by circuit breakers (2 in series 4, qualified to 1 Class lE requirements, operated by overcurrent. l 1 Q40.182-3 I

B/B-FSAR QUESTION 040.183 "The FSAR states in the seventh paragraph on page 8.3-6 ~ that ESF buses 141 and 241 cannot be fed from the same SAT except when one of the two SAT's is unavailable and the removable links are manually relocated from the transformer secondary to the bus duct crosstie. "If the above statement is true then there is a discrepancy with the bus ties shown on drawing 0040.49-3.- The bus ties between bus 141 and 241 are made by circuit breakers 1414 and 2414. There is a similar bus tie between buses 142 and 242 by means of circuit breakers 1424 and 2424. "These bus ties are in conflict with statements in the FSAR. Review the FSAR and provide a description of the bus connection that agrees with the drawings."

RESPONSE

The above referenced paragraph of the'FSAR will be revised as follows: The 4160-volt ESF buses 141 (2414 and 142 (2424 will not be fed from'the same SAT (parallel operationi except when one of the unit's SAT's is unavailable, and the removable links are manually relocated from the transformer secondary to the bus duct crosstie. Q40.183-1

D/O-FSAR QUESTION 331.33 " Provide additional information regarding the sensitivity of airborne radioactivity < monitors in accordance with Section 1.3 of RegulatoryrGuide 1.70. Verify that the airborne radioactivity < monitors described in Section 12.3.4 of the FSAR are capable of detecting 10 MPC-hours of particulate and iodine radioactivity in compartments which may be occupied and may< contain airborne radioactivity-(the acceptance criteria in Standard Review Plan Section 1 2. 34. "

RESPONSE

Standard Review Plan Section 12.3 implies an assessment of potential contamination is required in "...any compartment which has a possibility of containing airborne radioactivity i and which may-be occupied by personnel." A system of fixed Continuous Airborne Monitors (CAMsl is provided to monitor for airborne radioactivity <in compartments which may be occupied and may<contain airborne radioactivity. Since there a're too many~ rooms or cubicles to monitor independently, 4 a limited number of CAMS are provided to continuously: monitor the air from selected branch exhaust. ducts of the HVAC system. Thus, the exhaust from a single room may'be diluted by the exhaust from other.cooms before the air gets to the monitoring point. The r e fore, the monitor must be sensitive enough to respond to the diluted activity. The maximum possible dilution factor for any cubicle is: DF = F cubicle (cmf4 F duct (cfmn and using the Detectability Factor for MPC (DMPC given We can write an expression f8)r the in Table Q331.33-1. time, T, it takes to detect the presence of MPC 18Y"18 a in the exhaust ducts: T = 1/(DMPC DM, (hrst (11 a

where, time to detect particulate and iodine MPC*,

T = (hrt detectability factor for MPC (see Table DMPC" = a Q331.33-16 , dilution factor for 10 MPC-HR detectability DF = 1 4 Q331.33-1

B/B-FSAR cubicle = flow in exhaust from cubicle, (cfml F F = fl w in branch duct where monitor is located, < duct (cfm). Table Q331.33-1 shows the sensitivity of the particulate, iodine, and noble gas channels for the isotopes of greatest interest. These sensitivities were compared to maximum perLissible concentrations in air (MPC 4 of the most restric-tive particulate and iodine radionucli8es in the areas and cubicles of lowest ventilation flow rate. The criterion used was that airborne radioactivity from the areas described above and having an activity concentration of MPC, would be detected within 10 hours. Exhaust flow rates $ rom cubicles and in branch ducts were examined to determine dilution factors for this assessment. The exhaust flow rates for the monitored branch ducts and the individual room exhaust flow rates are given on the HVAC drawings in Section 9.4. The location of the radiation monitors are also shown on these-drawings. An investigation using the above data indicates that the system is capable of detecting 10 MPC -brs of airborne a particulate and iodine radioactivity in the rooms, cubicles and areas discussed above which may be occupied and may contain airborne radioactivity. Q331.33-2 1

TABLE Q331.33-1 SENSITIVITY OF CONTINUOUS AIRBORNE MONITORING SYSTEM AVERAGE GROSS SENSITIVITY DETECTABILITY* l ENERGY SENSITIVITY (cpm /hr MPC FACTOR FOR MPC ISOTOPE (MEV) (cpm /uCi) per UCi/cc) (UCi/8c) a I. Particulate Channel (Beta Scintillator)' CO-60 0,.096 4.6h,N10 5 12 -9 2.01 x 10 9 x 10 200 6 1 -9 .SR-90,,N b'.200 x1.20 x 10 5.10 x 10 1 x 10 60 f i -8 ,m.4.7.x 10! # ~1.92q 10f2. 6 x 10 1350 [ C O.085 4 e : 6 CS-137 0.'171 1.15 x 10 4.91x10g2' 1 x 10-8 575 o t W W \\ \\ ,w II. Iodine Channel (NaI ;S.pectrometryfwindowed on I--131 peak) W 5 11 -9 .k '. I--131 . 3 64'(y ) 1.01 x 10 4.29 x 10 9 x 10 850 t III. Moble Gases Channel (Beta Scintillatort 1, -e N t g4 f, KR b'.100 ' 1.84 x 10 ** 1.0 x 10 40 7 -5 ,i,e i 1:. s I 7 5 ~. 0.2.50-s eXE-133% 3.6 x 10 ** 1.0 x 10 80 (a' tivity) concentrations is based on a signal count rate at a i

• The minimum detectable c

( 95% confidence level as given by_the formula in ANSI 13.10-1974 and modified for the is' GA siyst'n as follow's: e .x s (BCKC/20)g++ Sensitivity, (for BCKG < 100' cpm) 2 MDC = (BCKG /2,000)b + Sensitivity, (for 100 cpm < BCKG < 1,x:10 cpm) 2 5 S 2 3 ~ Where BCKG is the t'otal ba'ckground counting rate (cpm).- For the particulates 1905 cpm was used and for iodine and noble gases 100 cpm was used; .i this criterion.will yield an answer that has a 95%' statistical confidence level.

• • cpm per PCi/cc s

i

B/B-FSAR AMENDMENT 36 JANUARY 1982 3.6.2.3.3.4 Pipe Restraints and Locations Pipe restraints and locations are discussed in Subsection 3.6.2.3.1.1 and 5.4.14. 3.6.2.3.3.5 Design Loading Combinations As described in Section 3.9, the forces associated with rupture of reactor piping systems are considered in combination with normal operating loads and earthquake loads for the design of supports and restraints in order to asrure continued integrity of vital components and engineered safety features. The stress limits for reactor coolant piping and supports are discussed in Section 3.9. 3.6.2.4 Guard Pipe Assembly Design Criteria GuIard pipes assemblies were not utilized in the design of either the Byron or Braidwood power stations. 3.6.2.5 Dynamic Analysis Applicable to, Postulated High Energy Pipe Break 3.6.2.5.1 Reactor coolant Icops a. Table 3.6-6 and Figure 3.6-24 identify the design-basis break locations and orientations for the main Reactor Coolant Loop. The primary-plus-secondary stress intensity ranges and the fatigue cumulative usage factors at the design break locations specified in WCAP 8082 (8172) (Reference 1) ~ are given in Table 3.6-7 for a reference fatigue analysis. The reference analysis, has been prepared to be applicable for many plants. It uses seismic umbrella moments which are higher than those used in WCAP 8002 (8172) such that the primary stress is equal to the limits of Equation 9 in NB-3650 (Section III of the ASME Boiler and Pressure Vessel Code) at nany locations in the systam where in WCAP 8082 (8172) one location was at the limit. Therefore, the results of the reference analysis may differ slightly from WCAP 8082 (8172) but the philosophy and conclusions of the WCAP are valid. There are no other locations in the model used in the reference fatigue analysis, consistent with WCAP 8082 (8172), where the stress intensity ranges and/or usage factors exceed the criteria of

2. 4 S, and 0. 2, respectively.

Additionally, the thermal transients used in the ref-erence analysis, although different from those in WCAP-i 8082, are the same as the thermal transients which are specified for Byron /Braidwood in Subsection 3.9.1.1. 3.6-35

~, ' -s a .s B/B-FSAR AMENDMENT 36 's JANUARY 1982 n Actual plant. moments for the Byron /Braidwood Units are also given in Table 3.6-7 at the design basis breekC13 cations so that the reference fatigue ~~ analysis.can be shown to be applicable for this e plant. by showing actual plant moments to be no d ~' greater than those used in the reference analysis, it follows that the stress intensity ranges and usage factors 'for the Byron /Braidwood Units will be less ~ than'those for comparable locations in the reference +e' model. By this means it is shown that there are no locations other than those identified in WCAP 8082 ,-[, (8172) where the stress intensity ranges and/or usage factors for the Byron /Braidwood Units might exceed ths' criteria of 2.4 S and 0.2, respectively.

Thus, m

the applicability of WCAP 8082 (8172) to the Byrfn/BraidwoodUnitshasbeenverified. u. s' .b. Pipe. whip restraints associated with the main Reactor coolant Loop are described in Subsections 3.6.2.3.1.1 and 5.4.14. s o-; c. Jet deflectors associated with the main Reactor ~' Coolant Loop Eure described in Subsection 3.6.2.3.1.2. ~ d. Design loading combinations and applicable criteria ,~ for ASME Class 1 components and supports are provided 's s in subsection 3.6.2.3.3.5. Pipe rupture loads "' include not only the jet thrust. forces acting on the piping but also jet impingement loads on the primary i- -y equipment. and supports. e. The interface between Sargent & Lundy and Westinghouse "~ -concerning the design of the primary equipment sup-ports and the interaction with the primary coolant loop is described in Subsection 3.9.3.4.4.1. s '3.6.2.5.2 Post'ulated Breaks in Piping Other than Peactor 1 Coolant Loop . Tne following material pertains to dynamic analyses completed for piping systems other than the reactor main coolant piping which connects the reactor vessel, the main coolant pumps, and the steam generators. 3.6.2.5.2.1 Implementation of criteria for Defining Pipe Break Locations and configurations The locations and number of design basis breaks, including postulated rupture orientations, for the high energy piping systems are shown in Figures 3.6-25 tc 3.6-44. The above information was derived from the implementation of the criteria delineated in subsection 3.6.2.1. 3.6-36

B/B-FSAR AMENDMENT 36 JANUARY 1982 I Stress levels and usage factors (usage. factors for Class 1 piping only) for the postulated break locations are shown in Tables 3.6-11 and 3.6-12. 3.6.2.5.2.2 Implementation of Criteria Dealing with Special Features Special protective devices in the form of pipe whip restraints and impingement shields are designed in accordance with Subsection 3.6.2.3. Pipe whip restraint locations, configuration, and orientations in relation to break locations for each applicable piping systems are shown in Figures 3.6-25 to 3.6-44. Inservice inspection is discussed in Subsection 3.6.1.2.2. 3.~6.2.5.2.3 Acceptability of Analyses Results The postulation of break and crack loca.tions for high and moderate energy piping systems and the analyses of the resulting jet thrust, impingement and pipe whip effects has conservative 3y identified areas where restraints, impingement shields, or other protective measures are needed and had yielded the conservative. design of the required protective devices. Results of jet thrust and pipe whip dynamic effects are given in Tables 3.6-13 and 3.6-14. 3.6.2.5.2.4 Design Adequacy of Systems, Components, and component Supports For each of the postulated breaks the equipment and systems necessary to nitigate the consequences of the break and to safely shut down the plant (i.e., all essential systems and componentn) have been identified (subsection 3. 6.1). The equipment and systems are protected against the consequences of each of the postulated breaks to ensure that their design-intended functions will not be impaired to unacceptable levels as a result of a pipe rupture or crack. When it became necessary to restrict the motion of a pipe which would result from a postulated break, pipe whip restraints were added to the applicable piping systems, or structural barriers or walls were designed to prevent the whipping of tne pipe. ) i Design adequacy of the pipe whip restraints is demonstrated in Tables 3.6-13 and 3.6-14. Data' in the tables was obtained through use of the criteria delineated in Subsections 3.6. 2.1 through 3.6.2.3 inclusive. The design adequacy of structural barriers, walls, and components is discussed in Section 3.8. 3.6-37

B/B-FSAR 3.7.3.1.1 Seismic Analysis Methods (Westinghouse) l This subsection describes the seismic analysis methods per- ) i formed for safety-related components and systems supplied by Westinghouse. l Those components and systems that must remain functional in the event of the SSE (Seismic Category I) are identified by applying the criteria of Subsection 3.2.1. This equipment is l classified into three types according to its dynamic char-acteristics. The analysis methods used for this equipment l also depended on these classifications. l The first type is flexible equipment. This equipment is characterized by several modes in the frequency range that could produce amplification of the base imput motion. The components which are classified as flexible equipment, i.e., I with more than one mode below 33 Hz, are the steam generators, reactor coolant pumps, pressurizers, control drive mechanisms, reactor internals, and fuel. Dynamic analyses were performed for these components using modal analysis techniques with either the response spectrum method, integration of the j uncoupled medal equations of motion, or by direct integration of the coupled differential equation *of motion. Details of l the methods used for these analyses are described in Subsections ( 5.3.7.3.1.1.1 through 3.7.3.1.1.5. I The second classification is rigid equipment. This equipment has a fundamental natural frequency that is sufficiently high i (greater than 33 Hz) so that base input motions are not ampli-fied. Such equipment is particularly suitable for static analysis as described in Subsection 3.7.3.1.1.6. Finally, the third type of equipment is classified as limited flexible, with only one predominate node in the frequency range subject to possible amplification of the input motion. The fundamental mode of this type of equipment is basically a translational bending mode at a frequency less than 33 Hz.- The second mode is usually a rocking mode with a frequency greater than 33 Hz. Because of the simple response character-istics of the equipment, dynamic analysis techniques that i account for multiple mode effects and closely spaced modes are j not required. Therefore, this equipment was evaluated using j static analysis methods as described.in Subsection 3,7.3.1.1.6. t l 3.7-9

B/B-FSAR

3. 7. 3.1.1.1 Dynamic Analysis - Mathematical Model The first step in any dynamic analysistis to model the structure

] or component, i.e., convert the real structure or component into a system of masses, springs, and dashpots suitable for mathematical analysis. The essence of this step is to select a model so that the displacements cttained will be a good representation of the motion of the structure or component. Stated dif ferently, the true inertia forces should not be altered so as to appreciably affect the internal stresses in the struc.ure or component. Some typical modeling techniques are presented in Reference 3. Equations of Motion Consider the multi-degree-of-freedom system shown in Figure 3.7-54. Making a force balance on each mass point r, the equations of motion can be written in the form: i my + rc u. + k ,u, = 0 (3. 7-3) rr L r1 1 r1 1 where: = the value of the mass or mass moment of rotational mr inertia at mass point r hr = absolute translational or angular acceleration of macs point r c i = damping coefficient - external force or moment r required at mass point r to produce a unit transla-tional or angular velccity at mass point i, maintaining zero translational or angular velocity at all other mass points. Force or moment is positive in the direction of positive transla-tional or angular velccity i 3.7-9a

~. t B M-FSAR 1 -x J n f + (at + At ) Cjx 2 h*n+2 - * +1

• n+1 n

n, l (At + At ) at At 3 y J y y +1 (K) bx+2+X+1+x b (* I ~ F +2- = n n J i i 3 /n n J i and x, and rearranging terms, Equation By factoring x +2 1, l n (3.7-28) is obta 3 follows: C EM3

• C E C 3
• I'# I E 3

+ 5 3 n+2 n+2 fC 7 [ M] - (1/3) [K] x + f-C2 [ M] + C3 E3-( I i n where: C = 2 2 Aty(At = At ) 1 C = 3 j At + At y C5 At ( At + At ) y I C C =C 7 2+ 5 The above set of simultaneous linear equations is solved to i obtain the presant values of nodal displacements (xt} in terms of the previous (r.nown) values of the nodal displacements. Since [M], [C ], and [ K] are included in the equation, they can also be time or displacement dependent. l 3.7.3.1.1.6 Static Analysis - Rigid and Limited Flexible Equipment Rigid equipment and limited flexible equipment as defined in Subsection 3.7.3.1.1 are generally analyzed using the static analysis method. This technique involves the multiplication of the total weight of the equipment or component member by a specified seismic acceleration coefficient. The magnitude of the seismic acceleration coefficient was established on the basis of the excitation level that the component was expected to experience in the plant. For rigid equipment, the seismic acceleration coefficients were compared with the high frequency (greater than 33 Hz) acceleration levels for the applicable response spectra developed fc r the plant to confirm the design analysis. The seismic acceleration coefficients for limited flexible equip-3.7-17

D/B-FSAR ment are compared with the acceleration levels from the applicable response spectra at the calculated fundamental natural frequency of the component. If the design seismic acceleration coefficients for either rigid or limited flexible equipment are exceeded by the actual plant acceleration levels, the design analysis is performed again at the actual level to confirm the equipment adequacy. 3.7.3.1.2 Differential Seismic Movements of Interconnected Supports Systems that are supported at points which undergo certain displacements due to a seismic event are designed to remain capable of performing their seismic category I functions. The displacements, obtained from a time-history analysis of the supporting structure, cause moments and forces to be induced into the piping system. Since the resulting stresses are self-limiting, it is justified to place them in the secondary stress category. Therefore these stresses exhibit properties much like a thermal expansion stress and a static analysis is used to c obtain them. e ) 3.7-17a .w

B/2-FSAR flexible building, the number of cycles. exceeding 90% of the maximum stress was not greater than three cycles. This study was conservative since it was performed with single degree-of-freedom models which tend to produce a more uniform and unattenuated response than a complex interacted syftem. The conclusions indicate that 10 maximum stress cycles for flexible, e quipment (natural frequencies less than 33 Hz) and 5 maximum stresc cycles for rigid equipment (natural frequencies greater than 33 Hz) for each of 20 OBE occurrences should be used for fatigue evaluation of Westinghouse systems and components. 3.7.3.3 Procedure Used for Modeling Procedures used for modeling PAfety-related components and systems within Westinghouse's scope are discussed in Subsection 3.7.3.1.1.1. Rigid valycs (i.e., with natural frequencies greater than 33 Hz) are included in the piping system model as lumped masses on rigid extended structures. If it is shown, by test or analysis, that a valve is not rigid (one or more natural frequencies below 33 Hz), than a multimass, dynamic model of the valve, including the appropriate stiffnesses, is developed for use in the piping system model. The valve model used in the piping analysis is constructed such that its calculated frequencies correspond to those obtained by test. For Class 1 piping systems, the ac tual calculated stiffness of pipe supports are included in tae model of the piping system. In the modeling of Class 2 and 3 piping systems, pipe supports are represented as minimum rigid elements for which the stiffness is predetermined based upon the particular support type. The mathematical model used for the dynamic analyses of the reactor coolant system is shown in Figure 3.9-1. 3.7.3.3.1 Modeling of the Piping System The continuous piping system is modeled as an assemblage of beams. The mass of each beam is lumped at nodes which are connected by weightless elastic members, representing the physical properties of each segment. The pipe lengths between mass points are not greater than the length which would have a natural frequency of 33 Hz when calculated as a simply supported beam. All concentrated weights on the piping system such as main valves, relief valves, pumps, and motors are modeled as lumped masses. The torsional effects of the valve operators and other equipment with offset center of gravity with respect to centerline of the pipe is included in the analytical model. 3.7-19 .m

B/B-FSAR 3.7.3.3.2 Field Location of Supports and Restraints The field location of seismic supports and restraints for Seismic Category I piping and piping systems components is selected to satisfy the following two conditions: a. The location selected must furnish the required response to control strain within allowable limits. b. Adequate building strength for attachment of the components must be available. The final location of seismic supports.ind restraints for Seismic Category I piping, piping system components, and equipment, including the ~ placement of snubbers, is checked against the drawings and instructions issued by the~ engineer. An additional examination of these supports and restraints devices is made to ensure that the location and characteristics of these supports t i l i 3.7-19a

C/B-FSAR rod drive mechanisms (CRDM's) and the fuel assemblies of the nuclear steam supply system, when used in seismic system analysis, are in conformance with the values for welded and/or ~ ~ bolted steel structures (as appropriate). For reactor internals-analysis, Westinghouse uses 2% damping 1for OBE and 4% damping for SSE as given by Regulatory Guide 1.61. Tests on fuel assembly bundles justified conservative component damping values of 7% for CBE and 10% for SSE to be used in the fuel assembly component qualification. Documentation of the fuel assemSly tests is found in Reference 2. The damping values used in component analysis of CRDM's and their seismic supports were developed by testing programs performed by Westinghouse. These tests were performed during the design of the CRDM support; the support was designed so that the damping in Table 3.7-2 could be conservatively used in the seismic analysin. The CRDM support system is designed with plates at the top of the mechanism and gaps between mechanisms. These are encircled by a box section frame which is attached by tie rods to the refueling cavity wall. The test conducted was on a full size CRDM complete with rod position indicator coils, attachment to a simulated vessel head, and variable gap between the top of the pressure housing support plate and a rigid bumppr representing the support. The internal pressure of the CRDM was 2250 psi and the temperature on the outside of the pressure housing was 4000 F. The. program consisted of transient vibration tests in which the' CRDM was deflected as specified initial amount and suddenly released. A logarithmic decrement analysis of the decaying transient provides the ef fective damping of the assembly. The effect on damping of variations in the drive shaft axial position, upper seismic support clearance, and initial deflection amplitude was investigated. The upper support clearance had the largest effect on the CRDM damping with the damping increasing with increasing clearance. With an upper clearance of 0.06 inches, the measured damping was approximately 8%. The clearances in a -typical upper seismic CRDM support is a minimum of 0.10 inch. The increasing damping with increasing clearances trend from the test results indicated that the damping would be greater than 8% for both the OBE and the SSE based on a comparison between typical deflections during these seismic events to the initial deflections of the mechanisms in the test. Component damping values of 5% are, therefore, conservative for both the OBE and the SSE. These damping values are used and applied to CRDM component l analysis by response spectra techniques. i l 3.7.3.15 Analysis Procedure for Damping In instances of the equipment supplied by Westinghouse, either the lowest damping value associated with the elements of the system is used for all modes, or an equivalent modal damping 3.7-29

9 B/B-FSAR transient problem (the dynamic response of the system for the static equilibrium position). The time-history displacement solution of all dynamic degrees of freedom is obtained using subprogram FIXFM and employing 4% critical damping. The loss-of-coolant accident displacements of the reactor vessel are applied in time-history form as input to the dynamic analysis l of the reactor coolant loop. The loss-of-coolant accident analysis of the reactor vessel includes all the forces acting on the vessel including internals reactions, cavity pressure loads, and loop mechanical loads.' The reactor vessel analysis is described in Subsection 3.9.1. 4. 6. The resultant asymmetric external pressure loads on the RCP and steam generator resulting from postulated pipe ruptures and pressure buildup in the loop compartments are applied to the same integrated RCL/ supports system model used to compute loadings on the components, component supports, and RCL piping as previously discussed. The response of the entire system is obtained for the various external pressure loading cases considered. For each pipe break case considered, the equip-ment support loads and piping stresses resulting from the external pressure loading are added to the support loads and piping stresses calculated using the loop LOCA hydraulic forces and RPV motion.. The break locations considered for subcompartment pressuriza-tion are those postulated for the RCL LOCA analysis, as discussed in Section 3.6 and SCAP-5172 (Reference 1 of Section 3.6). The asymmetric subcompartment pressure loads are provided to Westinghouse by Sargent & Lundy. The analysis to determine these loads is discussed in Section 6.2. The time-history displacement response of the loop is used in computing support loads and in performing stress eval.uation of the reactor coolant loop piping. The support loads [F] are computed by multiplying the support stiffness matrix [ K) and the displacement vector [6 ] at the support point. The support loads are used in the evaluation of the supports. The time-history displacements of the FIXFM subprogram are used as input to a WESTDYN2 to determine the internal forces, l deflections, and stresses at each of the piping elements. For this calculation the displacements are treated as imposed deflections on the reactor coolant loop masses. The results of this solution are used in the piping stress evaluation. 3.9-23 T

B/B-FSAR Tran.31ents Operating transients in a nuclear power plant cause thermal and/or pressure fluctuations in the reactor coolant fluid. The thermal transients cause time-varying temperature distributions across the pipe wall. These temperature distributions resulting in pipe wall stresses.may be further subdivided in accordance with the Code into three parts, a uniform, a linear, and nonlinea r portion. The un' form portion results in general expansion _ loads. The linear portion causes a bending moment across the wall and the nonlinear portion causes a skin stress. The transients as defined in Subsection 3.9.1.1 are used to define the fluctuations in plant parameters. A one-dimensional finito difference heat conduction program is used to solve the thermal transient problem. The pipe is represented by at least 50 elements through the thickness of the pipe. The convective heat transfer coefficient employed in this program represents the time-varying heat transfer due to free and forced convection. The outer surface is assumed to be adiabatic, while the inner surf ace boundary experiences the temperature of t. e coolant l fluid. Fluctuations in the temperature of the coolant fluid produce a temperature distribution through the pipe wall thickness which varies with time. An arbitrary temperature distribution across the wall is shown in Figure 3.9-2. O 3.9-23a l i i 7m a

9/B-FSAR 3.9.1.4.4 Primary Component Supports Models and Methods Primary component supports are discussed in Subsection 3.9.3.4. 3.9.1.4.5 Analysis of Prinary Components Equipment which serves as part of the pressure boundary in the reactor coolant loop include the steam generators, the reactor coolant pumps, t he pressurizer, and the reactor vessel. This equipment is Seiamic category I and the pressure boundary meets the requirementu of the ASME Boiler and Pressure Vessel Code, Section III, Scosection NB. This equipment is evaluated for the loading combinations outlined in Table 3.9-2. The equipment is analyzed for (1) the normal loads of deadweight, pressure, and thernal; (2) mechanical transients of OBE, SSE, and pipe ruptures; including the effects of aspnmetric subcompartment pressurization and (3) pressure and temperature transients outlined in Subsection 3.9.1.1 The results of the reactor coolant loop analysis are used to determi,ne the loads acting on the equipment nozzles and the support / component interface locations. These loads are supplied for all loading conditions on an " umbrella" load basis. 'This is, on the basis of previous plant analyses, a set of loads is determined which should be larger than those seen in any single plant analysis. The umbrella loads represent a conservative means of allowing detailed component analysis prior to the completion of the system analysis. Upon completion of the system analysis, conformation is demonstrated between the actual plant loads and the loads used in the analyses of the components. Any deviations where the actual load is larger than the umbrella load will be handled by individualized analysis. Seismic analyses are performed individually for the reactor coolant pump, the pressurizer, and the steam generator. Detailed and complex dynamic models are used for the dynamic analyses. The response spectrum correspor. ding to the building elevation at the highest component / building attachment elevation is used for the component analysis. Seismic analyses for the steam generator and pressurizer are performed using 2% damping for the OBE and 4% dampinq for the SSE. The analysis of the reactor coolant pump for determination of loads on the motor, main flange, and pump internals is performed using the damping for bolted steel structures, that is 4% for the OBE and 7% for the SSE (2% for OBE and 4% for SSE is used in the system analysis). This damping is applicable to the reactor coolant pump, since the main flange, motor stand, and motor are all bolted assemblies (see Section

5. 4).

The reactor pressure vessel is qualified by static stress analysis based on loads that have been derived from dynamic analysis. Reactor coolant pressure boundary components are further qualified to ensure against unstable crack growth under f aulted conditions by performing detailed fracture analyses of the critical areas of this boundary. Actuation of the emergency core 3.9-26 FL

B/B-FSAR total stresses in excess of the yield strength does not affect the conservatism of the results, provided that these thermal stresses are included in the evaluation of the stress intensity factors. Therefore, for faulted condition analyses, LEFM is considered applicable for the evaluation of the vessel inlet nozzle and beltline region. In addition, it has been well established that the crack propagation of existing flaws in a structure subjected to cyclic loading can be defined in. terms of fracture mechanics parameters. Thus, the principles of LEFM are also applicable to fatigue growth of a postulated flaw at the vessel inlet nozzle and beltline region. For the safety injection and charging line nozzles, which are fabricated from 304 stainless steel, LEFM is not applicable because of extreme ductility of the material. For these nozzles, the thermal effects are evaluated using the principles of Miner's hypothesis of linear cumulative damage in conjunction with f atigue data from constant stress or strain fatigue tests. The cumulative usage fatigue defined as the sum of the ratios of the number of cycles of each transient (n) to the allowable number of cycles for the stress range associated with the transient (N) must not exceed 1.0. An example of a f aulted condition evaluation carried out according to the procedure discussed previously is given in Reference 3. This report discusses the evaluation procedure in detail as applied to a severe faulted condition (a postulated loss-of-coolant accident). The pressure boundary portions of class 1 valves in the RCS are designed and analyzed according to the requiremi.its of NB-3500 of ASME III. These valves are identified in Subsection 3.9.3.2. Valves in sample lines connected to the RCS are not considered to be Seismic Category I nor ASFE Class 1. This is because the nozzles where the line connects to the primary system piping are orificed to a 3/8-inch hole. This hole restricts the flow such that lors through a severance of one of these lines can be made up by normal charging flow. 3.9.1.4.6 Dynamic Analysis of Reactor Pressure Vessel for Postulated Loss of Coolant Accident The structural analysis of the reactor vessel and internals considers simultaneous application of the time-history loads resulting from the reactor coolant loop mechanical loads, internal hydraulic pressure transients, and reactor cavity pressurization. The vessel is restrained by four reactor vessel supports under every other reactor vessel nozzle and the reactor coolant loops with the primary supports of the steam generators and the reactor coolant pumps. 3.9-28 W

W/W-7sMm Pipo displacement restroints installed in the primary shield wall limit the break opening area of the vessel nozzle pipe breaks. An upper pump break area is determined from break areas calculated using reactor vessel and pipe relative motions for similar plant analyses. Detailed studies have shown that pipe breaks at the hot or cold leg reactor vessel nozzles, even with a limited break area, would give the highest reactor vessel support loads and the highest vessel displacements, primarily Jue to the influence of reactor cavity pressuriza-tion. By c)nsidering those breaks, the most severe reactor vessel suppert Joads are determined. Load ng Conditions 3.9.1.4.6.1 1 Following a postulaind pipe rupture at the reactor vessel nozzel, the reactor vessel is excited by time-history forces. As pre-viously mentioned, these forces are the combined effect of three phenomena: (1) reactor coolant loop mechanical loads, (2) reactor cavity pressurization forces, and (3) reactor inter-nal hydraulic forces. The reactor coolant loop mechanical forces are derived from the clastic' analysis of the loop piping for the postulated break. The reactions on the nozzles of all the unbroken piping legs are applied to the vessel in the reactor pressure vessel blow-down analysis. Reactor cavity pressurization forces arise for the pipe breaks at the vessel nozzles from the steam and water which is released into the reactor cavity through the annulus around the broken pipe. The reactor cavity is pressurized asymmetrically with higher pressure on the side of the broken pipe resulting in horizontal forces applied to the reactor vessel. Small vertical forces arising from pressure on the bottom of the vessel and the vessel flanges are also applied to the reactor vessel. The cavity pressure analysis is described in Section 6.2. The internals reaction forces develop from as3mmetric pressure distributions inside the reactor vessel. For i vessel inlet nozzle break, the depressurization wave path is through the broken loop inlet nozzle and into the region between the core barrel and reactor vessel. This region is called the downcomer annulus. The initial waves propagate up, down, and around the downcomer annulus and up through the fuel. In the case of a reactor pressure vessel outlet nozzle break the wave pas.es e through the reactor pressure vessel outlet nozzle and directly into the upper internals region, depressurizes the core, and enters the downcomer annulus from the bottom of the vessel. Thus, for an outlet nozzle break, the downcomer annulus is depressurized with much smaller differences in pressure hori-zontally across the core barrel than for the inlet break. For both the inlet and outlet nozzle breaks, the depressurization waves continue their propagation by reflection and translation through the reactor vessel fluid but the initial depressuriza-tion wave has the greatest effect on the loads. 3.9-28a l. - u

B/B-FSAR The reactor internals hydraulic pressure transients were calcu-lated including the assumption that the structural motion is coupled with the pressure transients. This phenomena has been referred to as hydroclastic coupling or fluid-structure inter-action. The hydraulic analysis considers the fluid-structure interaction of the core barrel by accounting for the deflections of constraining boundaries which are represented by masses and springs. The dynamic response of the core barrel in its beam-bending mode responding to blowdown forces compensates for inter-nal pressure variation by increasing the volume of the more highly pressurized regions. The analytical methods used to devel-op the reactor internals hydraulics are described in WCAP-8708. (6) 3.9.1.4.6.2 Reactor Vessel and Internals Modeling The reactor vessel model consists of two nonlinear elastic models connected at a common node. One model represents the dynamic vertical characteristics of the vessel and its internals, and the other model represents the translational and rotational character-istics of the structure. These two models are combined in the DARI-WOSTAS code (Reference 1) to represent motion of the reactor vessel and its internals in the plane of the vessel centerline and the broken pipe centerline. 4 The model for horizontal motion is shown in Figure 3.9-23. Each node has one translational and one rotational degree of freedom 1 in the vertical plane containing the centerline of the nozzle attached to the broken pipe and the centerline of the vessel. A j combination of beam elements and concentrated masses are used to represent the components including the vessel, core barrel, neutron panels, fuel assemblies, and upper support columns. Connections between the various components are either pin-pin rigid links, translational impact springs with damping, or rota-tional springs. The m6 del for vertical motion is shown in Figure 3.9-24. Each mass node has one translational degree of freedom. The structure is repres.ented by concentrated masses, springs, dampers, gaps, and frictional elements. The model includes the core barrel, lower support columns, bottom nozzles, fuel rods, top nozzles, upper support structure, and reactor vessel. The horizontal and vertical models are coupled at the elevation of the primary nozzle centerlines. Node 1 of the horizontal model is coupled with node 2 of the vertical model at the reactor vessel nozzle elevation. This coupled node has external restraints characterized by linear horizontal springs which describe the tangential resistance of the supports and by individual nonlinear vertical stiffness elements which provide downward restraint only. The supports as represented in the horizontal and vertical models are not indicative of the complexity of the support system used in the analysis. The individual supports are located at the actual support pad locations and accurately represent the inde-pendent nonlinear behavior of each support. 3.9-28b V

B/B-FSAR 3.9.1.4.6.3 Analytical Methods The' time-history effects of the cavity pressurization loads, internals loads, and loop mechanical loads are combined and applied simultaneously-to the appropriate nodes of the mathe-matical model of the reactor vessel and internals. The analysis -is performed by numerically integrating the differential equa-tions of motion to obtain the transient response. The output of the analysis includes the displacements of the reactor vessel and-the loads in the reactor vessel supports which are cobmined with other applicable faulted condition loads and subsequently used to calculate the stresses in the supports. Also, the reactor vessel. displacements are applied as a time-history input to the dynamic reactor coolant loop blowdown analysis. The resulting loads and stresses in the piping components and supports include both loop blowdown loads and reactor vessel displacements.

_Thus, the effect of the vessel displacements upon loop response'and the effect of. loop blowdown upon vessel displacements are both evaluated.

In addition, using the results of the RCL analysis, the actual break opening area is verified to be less than the-estimated area used in the analysis and assures that the analysis is conservative. i I 3.9-28c i u

B/B-FSAR AMENDMENT 36 JANUARY 1982 3.9.1.4.7 Stress criteria for class 1 Components l All Class 1 components are designed and analyzed for the design, l normal, upset, and emergency conditions to the rules and l requirements of the ASME Code Section III. The design analysis or test methods and associated stress or load allowable limits that will be used in evaluation of faulted conditions are those that are defined in Appendix F of the ASME Code with supplementary options outlined below. The test load method given in F-1370 (d) is an acceptable method of qualifying components in lieu of satisfying the stress / load limits established for the component analysis. The reactor vessel support pads are qualified using the test option. The reactor pressure vessel support pads are designed to restrain unidirectional horizontal motion in.3dition to l supporting the vessel. The design of the supports allows radial growth of the vessel but restrains the vessel from horizontal displacements since tangential displacement of the vessel is preve,nted at each vessel nozzle. To duplicate the loads that act on the pads during faulted conditions, the tests, which utilized a one-eighth linear scale model, were performed by applying a unidirectional static load to the nozzle pad. The load on the nozzle pad was reacted by a support shoe which was mounted to the test fixture. The above modeling and application of load thus allows the maximum load capacity of the support pads to be accurately established. The test load, LT, was then determined by multi-plying the maximum collapse load by 64 (ratio of prototype area to model area) and including temperature effects in accordance with the rules of the ASME Code, Section III. The loads on the reactor vessel support pads, as ca'1culated in the system analysis for faulted conditions are limited to the value of 0.80 L T. The tests performed and the limits established for the test load method ensure that the experi-l mentally obtained value for LT is accurate and that the support 1 pad design is adequate for its intended function. Loading combinations and allowable stresses for ASME Class 1 components are given in Tables 3.9-2 and 3.9-3, respectively. The methods of load combination for each operating condition are as follows: Design Loads are combined by algebraic sum. i 3.9-29 Y h

B/B-FSAR AMENDMENT 36 JANUARY 1982 Normal, Upset These loads are used in the fatigue evaluation in accordance with the methods prescribed in the ASME code. Loadsets are defined for each transient including the OBE and are combined such that the maximum stress ranges are obtained without regard to the order in which the transients occur. (This is discussed in more detail in Subsection 3.9.1.4.3). Emergency Loads are combined by algebraic sum. Faulted LOCA and SSE loads are combined using the square-root-of-the-sum-of-the-squares (SRSS) method on a load component basis (i.e., the LOCa F is combined with the SSE F by SRSS, the x x LOCA F is combined with the SSE F by SRSS, and likewise for y and Mz). Thesustainebloads, such as weight F,Mx, My, effects, are combined with the SRSS re;ults by algebraic sum. 3.9.1.4.8 Analytical Methods for RCS Class 1 Branch Lines The analytical methods used to obtain the solution consist of the transfer matrix method and stiffness matrix formulation for the static structural analysis, the response spectrum method for seismic dynamic analysis, and static or dynamic structural analysis for the effect of a reactor coolant loop pipe break. The integrated Class 1 piping / supports system model is the basic system model used to compute loadings on components, component and piping supports, and piping. The system models include the stiffness and mass characteristics of the Class 1 piping components, the reactor coolant loop, and the stiffness of supports which affect the system response. The deflection solution of the entire system is obtained for the various load-ing cases from which the internal member forces and piping stresses are calculated. Static The Class 1 piping system models are constructed for the WESTDYN computer program, which numerically describes the physical system. A network model is made up of a number of sections, each having an overall transfer relationship formed from its group of ele-monts. The linear elastic properties of the section are used to define the characteristic stiffness matrix for the section. Using the transfer relationship for a section, the loads re-quired to suppress all deflections at the ends of the section arising from the thermal and boundary forces for the section are obtained. 3.9-30 T

B/B-FSAR AMENDMENT 36 JANUARY 1982 After all the sections have be'en defined in this manner, the overall stiffness matrix and associated load. vector to suppress the deficction of all the network points are, determined. By inverting the stiffness matrix, the flexibility matrix is deter-mined. The flexibility matrix is multiplied by the negative of the load vector to determine th'e network point deflections due to the thermal and boundary force effects. Using tlua general transfer relationship, the deflections and internal forces are then determined at all node points in the system. The support loads are also computed by multiplying'the stiifness matrix by the displacement vector at the support point. Seismic The models used in the static analyses are modified for use in the dynamic analyses by including the mass characteristics of the piping and equipment. The lumping of the distributed mass of the piping systems is accomplished by locating the total mass at points in the system which will appropriately represent the response of the'distri-buted system. Effects of the primary equipment motion, that is, i reactor vessel, steam generator, reactor coolant pump, and pressurizer, on the Class 1 piping system are obtained by model-ing the mass and the stiffness characteristics of the primary j equipment and loop piping in the overall system'model. The supports are represented by stiffness matrices in the system model for the dynamic analysis. Shock suppressors which cesist l rapid motions are also included in the analysis. The solution for the seismic disturbance employs the response spectra method. This method employs the lumped mass technique, linear elastic l properties, and the principle of model superposition. l The total response obtained from the seismic analysis consists l of two parts: the inertia response of the piping system and the repsonse form differential anchor motions. The stresses resulting from the anchor motions are considered to be secon-dary and, therefore, are included in the fatigue evaluation. l Loss of Coolant Accident i The mathematical models used in the seismic analyses of the Class 1 lines are also used for RCL pipe break effect analysis. To obtain the dynamic solution for lines six inches and larger and certain small-bore lines required for ECCS considerations, the time-history deflections from the analysis of the reactor L coolant loop are applied at oranch nozzle connections. For other small bore lines which must maintain structural integrity, the motion of the RCL is applied statically. 3.9-30a

B/B-FSAR AMENDMENT 36 JANUARY 1982 Fatigue A thermal transient heat transfer analysis is performed for each different piping component on all the Class 1-branch lines. The normal, upset, and test.conditio'n transients identified in Subsection 3.9.1.1 are considered in the fatigue evaluation. The-thermal quantities AT, 1 AT and (a T a a, -ab b) are calcu-T lated on a time-history basis,2,using a one-dimensional finite difference heat transfer computer program. Stresses due to these quantities were calculated for each time increment using the methods of NB-3650 of ASME III. For each thermal transient, two loadsets are defined, repre-senting the maximum and minimum stress states for that trans-iont. As a result of the normal modo spectral technique employed in the seismic analysis, the load components cannot be given signed values. Eight load sets are used to represent all possible sign permutations of the seismic moments at each point, thus insuring the most conservative combinations of seismic loads are used in the stress evaluation. The WESTDYN computer program is used to calculate the primary-plus-secondary and peak stress intensity ranges, fatigue reduction factors and cumulative usage factors for all pos-sible loadset combinations. It is conservatively assumed that the transients can occur in any sequence, thus resulting in the most conservative and restrictive combinations of transients. The combination of loadsets yielding the highest alternating stress intensity range is determined and the incremental usage factor calculated. Likewise, the next most severe combination is then determined and the incremental usage factor calculated. This procedure gs repeated until all combinations having allow - able cycles <10 are formed. The total cumulative usage factor at a point is the summation of the incremental usage factors. 3.9.1.4.9 Evaluation of Control Rod Drive Mechanisms and Supports The control rod drive mechanisms (CRDM 's ) and CRDM support structure are evaluated for the loading combinations outlined in Table 3.9-3. A detailed finite element model of the CRDM's and CRDM supports is constructed using the WECAN computer program with beam, pipe, and spring elements. For the LOCA analysis, nonlinear-ities in the structure are represented. The time-history 3.9-30b .A

B/B-FSAR AMENDMENT 36 JANUARY 1982 motion of the reactor vessel head, obtained from the RPV analysis described in 3.9.1.4.6 is input to the dynamic i model. Maximum forces and moments in the CRDM's and support structure are then determined. For the seismic analysis, the structural model is linearized and the floor response spectra corresponding.to the CRDM tie rod elevation is applied to determine the maximum forces and moments in the structure. The bending moments calculated for the CRDM's for the various loading conditions are compared with maximum. allowable moments determined from a detailed finite element stress evaluation of the CRDM's. Adequacy of the CRDM support' structure is verified by comparing the calculated stresses to the criteria given in ASME III, Subsection NF. 1 I t I 4 i Y e 4 3. 9-3 0c W I

B/B-FSAR NSSS (For ASME Code Class 2 and 3 Components) Design pressure, temperature, and other loading conditions that provide the bases for design of fluid systems Code Class 2 and 3 components are presented in the sections which describe the systems. 3.9.3.1.1 Design Loading Combinations Balance of Plant The combination of design loadings is categorized with respect to plant conditions identified as normal, upset, emergency, or faulted as shown in Tables 3.9-5 through 3.9-14 for the major components and piping. NSSS The design loading combinations for ASME Code Class 2 and 3 components and supports are given in Table 3.9-5. The design loading combinations are categorized with respect to normal, upset, emergency, and faulted conditions. Stress limits for each of the loading combinations are presented in Tables 3.9-6, 3.9-7, 3.9-8, and 3.9-9 for tanks, inactive

• pumps, I

active pumps, and valves, respectively. Active ** pumps and valves are discussed in Subsection 3.9.3.2. Design of l primary equipment supports is discussed in Subsection 3.9.3.4. I Inactive components are those whose operability is not relied upon to perform a safety function during the l transients or events considered in the respective operating l condition category. Active components are those whose operability is relied upon to perform a safety function (as well as to accomplish I and maintain a safe reactor shutdown) during and following the transients and events considered in the respective operating condition categories. 1 i 3.9.3.1.2 Design Stress Limits 3.9.3.1.2.1 Stress Level for Class 1 Piping and Components (Balance of Plant) Stress analysis was used to determine structural adequacy of pressure. components under the operating conditions of normal, upset, emergency, or faulted, as applicable. Significant discontinuities were considered such as nozzles, flanges, etc. In addition to the design calculation required by the ASME III code, stress analysis was performed by methods outlined in the code appendicos or by other methods by reference to analogous codes or other published literature. 3.9-45 y

B/B-FSAR l l 3.9.3.1.2.2 Stress Levels for ASME Code Class 2 and 3 Balance of Plant For safety-related ASME Code Class 2 and 3 components and piping, the design stress limits are listed in Tables 3.9-5 through 3.9-14. Inelastic methods as permitted by ASME Section III for Class I components were not used for these components. NSSS The design stress limits established for Class 2 and 3 compo-nents are sufficiently low to assure that violation of the pressdre-retaining boundary will not occur. These limits for each of the loading combinations are component oriented and are presented in Tables 3.9-6 through 3.9-9. The criteria for Class 2 and 3 component supports are as follows: a. Supports for Vessels Procured Af ter July 1, 1974 Class 2 and 3 vessel supports are designed and analyzed to the rules and requirements of ASME III, Subsection UF. For linear supports designed by analysis, the increased design limit for stress identified in NF-3232.l(a) shall be limited to the smaller of 2.0 S or G unless g, otherwise justified by shakedown analysiE. Tba methods for analysis and associated allouable limits that are used in the evaluation of linear supports for faulted conditions are thoGe defined in ASME III Appendix F. Plate and shell supports shall satisfy the following y 5 2.0 S, stress criteria for faulted conditions: o 1 +SME 5 2.4 S. (c and re defined in NP-3221.1 c 0 7 y 2 A III.) ot b. Supports for Vessels Procured Prior to-July 1, 1974 1. Linear a) Normal - The allowable stresses of A.I.S.C.-69, Part 1 are employed for normal condition allowables. b) Upset - Stress limits for upset conditions are 33% higher than those specified for normal conditions. This is consistent with paragraph 1.5.6 of A.I.S.C.-69, Part 1 which permits 3.9-46

B/D-FSAR j one-third increase in allowable stresses i for wind or seismic loads, c) Emergency - Not applicable. d) Faulted - Stress limits for faulted condition are the same as for the upset condition. 2. Plate and Shell a) Normal - Normal condition limits are those specified in ASME Section VIII, Division 1 or A.I.S.C.-69, Part 1. b) Upset - Stress limits for upset conditions are 33% higher than those specified for-normal conditions. This is consistent with paragraph 1.5.6 of A.I.S.C, Part 1 which permits one-third increase in allowable stresses for wind or seismic loads. c) Emergency - Not applicable. d) Faulted - Stress limits for faulted condition are the same as for the upset condition. c. Supports for Pumps The stress limits used for Class 2 and 3 pump supports are identical to those used for the supported component, as indicated in Tables 3.9-7 and 3.9-8. 3.9.3.1.2.3 Field Run Piping (Balance of Plant) No Seismic Category I field run piping system exists. Category II piping, 2-inch nominal pipe size and smaller, and 200* F and colder, are field run. Criteria are provided to the contractor to ensure proper routing and design interface with Seismic Category I systems and equipment, or interfaces are appropriately controlled by guides. ~ 3.9.3.2 Pump and Valve Operability Assurance Balance of Plant Design methods are a combination of analysis, past testing, and operating experience. Active mechanical equipment classified as Seismic Category I has been shown capable of performing its function during the life of the plant under postulated plant conditions. 3.9-46a EI"

B/B-FSAR Equipment with operating condition functional requirements includes " active" (active equipment must perform a mechanical motion during the course of accomplishing a safety function) pumps and valves in fluid systems such as the residual heat removal system, safety fluid injection systems, and the essential service water system. Operability will be ensured by satisfying the requirements of the following programs. Continued operability is ensured by periodic testing. NSSS Mechanical equipment classified as safety-related must be capable of performing its function under postulated plant conditions. Equipment with faulted condition functional requirements includes active pumps and valves in fluid systems such as the residual heat removal system, safety injection system, and the containment. 3.9-46b

B/D-FSAR 1 l If power to the stationary gripper coil is cut of f, the combined weight of the drive rod assembly and the rod cluster control assembly and.the stationary gripper return spring is sufficient to move. latches out of the drive rod assembly groove. The control rod f alls by gravity into the core. The trip occurs as the magnetic field, holding the stationary gripper plunger half against the stationary gripper pole, collapses and the stationary gripper plunger ' half is forced down by the stationary gripper return spring and weight acting upon the latches. After the rod cluster control assembly is released. by the mechanism,_ it f alls f reely until the control rods enter the dashpot section of the thimble tubes in the fuel assembly. 3.9.4.2 Applicable CRDS Design Specifications For those comparable in the control rod drive system comprising portions of the reactor coolant pressure boundary, conformance with General Design Criteria 15, 30, 31, 32 and 10 CFR 50: Section 5~0.55a is discussed in Section 5.2. Conformance with Regulatory Guides pertaining to materials suitability is described in Section 4. 5 and Subsection 5.2. 3. Design Bases Bases for temperature, stress on structural members, and material compatibility are imposed on the design of the reactivity control components. Design Stresses The control rod drive system is designed to withstand stresses originating from varioca operating conditions as summarized in Table 3.9-1. Loading combinations for the Class 1 components of the control rod drive system are given in Table 3.9-2, a. Allowable Stresses For normal operating conditions Section III of the ASME Boiler and Pressure Code is used. All pressure boundary components are analyzed as Class I components under Article NB-3000. b. Dynamic Analysis The cyclic stresses due to dynamic loads and deflections are combined with the stresses imposed by loads from component weights, hydraulic forces and thermal gradients for the determination of the total strescos of the control rod drive system. Control Rod Drive Mechanisms The control rod drive mechanism (CRDM) pressure housings are class 1 components designed to meet the stress requirements for normal operating conditions of Section III of the ASME Boiler and 3.9-68 r L

. g, ;. t p", n . B/B-FSAR i, / In addition, dynamic testing programs have been conducted by Westinghouse andEWestirighouse licensees to demonstrate that control' rod scFam time /is not adversely 'af fec'ted by. postulated sei,smic events. Acceptable scram performance is assured by also including t'he-effe' cts oI the silowable s displacements of the driveIIne comp'onents in the.evaluacion of the test raiults.' .g 4 \\ a,.' M 1 / 4 +# w w. A m [ s m a 4 6 4

• W p

C +

• f

+ 4 D l }. v F T f / .e 4 l l ../ 1 44 f f k [* ? s t - < s 3.9-72a 'l s e = o w* ~

v i - B/B-FSAR Incore Instrumentation Support Structures The incore inctrumentation support structures consist of an upper syst6m to convey and support thermocouples penetrating the vessel through-the head and a lower system to convey and support flux thimbles ~ penetrating the vessel through the bottom (Figure 7.7-9 shows the Basic Flux-Mapping System). - The upper syste.n utiliz es the reactor vessel head penetrations. In'ctrum9ntation port columns are slip-connected to inline columns that are in turn f astened to the upper support plate. These port ~ columnc protrude through the head penetrations. The thbrmocouples are carried through these port columns and the upper support plate at positions above their readout locations. .The theriocouple conduits are supported from the columns of the ,vpper core ' support system. The thermocouple conduits are sealed stainless s'ceel tubes. . - ~ J' -,In addition to the upper incore instrumentation, there are reactor vessel. bottom port columns which carry the retractable, cold worked stainless stcel flux thimbles that are pushed upward r , fihto the reactor core. Conduits extend from the bottom of the c63ctor vessel down through the concrete shield area and up to a Ethimble coal line. The minimum bend radii are about 144 inches and,the trailing ends of the thimbles (at the seal line) are extracted Jrproximately 15 feet during refueling of the reactor '4h order to avoid interference within the core. The thimbles are closed at the leading ends and serve as the pressure barrier '/ '~ .between the reactor pressurized water and the containment atmosphSr e. . ' Mechanical seals between the retractable thimbles and conduits arb ~provided at the seal line. During normal operation, the r'e, tractable thimbles are stationary and move only during refue' ling or for maintenance, at which time a space of approximately 15 feet above the seal line is cleared for the retracti on. operation. The incore instrumentation support structure is designed for ade-quate support of instrumentation during reactor operation and is rugged enough to resist damage or distortion under the conditions imposed by handling during the refueling sequence. These are the only conditions which affect the incore instrumentation support structure. Reactor vessel surveillance specimen capsules are cover ed in Succection 5. 3.1. 6. i. 5.2' Design _ Loading Conditions The design loading conditions that provide the basis for the design of the reactor internals are: l y v- ,,l.'! 3.9-77 t ) 9

B/B-FSAR v a. Normal and Upset -r The normal an'd upset loading conditions that provide ths basis.for the design of the reactor internals are: 1. fuel ~and reactor internals weight. e 2. -fuelbndcorecomponent..springforcesincluding spring preloading forces. 3. differential pressure and coolant flow forces. 4. temperature gradients. 5. vibratory loads including OBE seismic. 6. the normal and upset operational thermal transients listed in Table 5.2-1. i 7. control rod trip (equivalent static load). J . y , :s [ 8. loads due to loop (s) out-of-service. J 9. loss of load pump overspeed. b. Emergency Conditions F The emergency loading conditions that provide the basis for the design of the reactor internals are: 1. small loss of coolant accident. ^ 2. small steam break s 3. ' complete loss of flow Faulted Conditions c. The f aulted loading conditions that provide the basis for the design of the reactor internals are: l 1. the large loss of coolant accident 2. the safe shutdown earthquake. i The main objectives of the design analysis are to satisfy allowable stress limits, to assure an adequate design margin, and to establish deformation limits which are concerned primarily with the f unctioning of the components. The stress limits are established not only to assure that peak stresses will not reach unacceptable values,.but also limit the amplitude of the ~ . oscillatory stress component in consideration of fatigue l 3.9-78 m i ~ j r2..

B/B-FSAR characteristics of the materials.. Both-low and high cycle fatigue stresses are considered when the allowable amplitude of oscillation is established. Dynamic analyses on the reactor. internals are' provided in Subsection 3.9.2. As part of the evaluation of design loading conditions, extensive testing and inspections are performed from the initial selection of raw materials up to and including component installation and plant operation. Among these tests and inspections are those performed during component fabrication, plant construction, startup. and checkout, and during plant operation. 3.9-78a c,-,

B/B-FSAR 3.9.5.3 Design Loading Categories The combination of design loadings fits into either the' normal, upset, emergency or f aulted conditions as defined in the ASME Code, Section III, and as indicated by Figures NG-3221.1, NG-3224.1 and by Appendix P, Rules for Evaluating Paulted Conditions. Loads and deflections imposed on components due to shock and_vi-bration are determined analytically and experimentally in both scaled models and operating reactors. The cyclic stresses due to these dynamic loads and deflections are combined with the streuses imposed by loads from component weights, hydraulic forces and thermal gradients for the determination of the total stresses of the internals. The reactor internals are designed to withstand stresses ori-ginating from various operating conditions as summarized in Table 3.9-1. The scope of the stress analysis problem is very large requiring many different techniques and methods, both static and dynamic. The analysis performed depends on the mode of operation under consideration. Allowable Deflections For normal operating conditions, downward vertical deflection of the lower core support plate is negligible. For the loss-of-coolant accident plus the safe shutdown earth-quake condition, the deflection criteria of critical internal structures are the limiting values given in Table 3.9-14. The corresponding no-loss-of-function limits are included in Table 3.9-4 for comparison purposes with the allowed criter.ia. The criteria for the core drop accident is based upon analyses which have to determine the total downward displacement of the internal structures following a hypothesized core drop resulting f rom loss of the normal core barrel supports. The initial clearance between the secondary core support structures and the reactor vessel lower head in the hot condition is approximately 1/2 inch. An additional displacement of approximately 3/4 inch would occur due to strain of the energy-absorbing devices of the secondary core support; thus the total drop distance is about 1 -1/4 inches, which is insufficient to permit the tips of the rod cluster control assembly to come out of the guide thimble in the f uel assemblies. Specifically, the secondary core support is a device which will never be used, except during a hypothetical accident of the core support (core barrel, barrel flange, etc.). There are four supports in each reactor. This device limits the fall of the core and absorbs much of the energy of the f all which otherwise would he imparted to the vessel. The energy of the fall is calculated assuming a complete and instantaneous f ailure of the 3.9-79

B/B-FSAR must be kept sufficiently small to allow core cooling. The functional limitations for the core structures during the design basis accident are shown in Table 3.9-14. To ensure no column loading of rod cluster control guide tubes, the upper core plate deflection iu limited so as not to exceed the value shown in Table 3.9-4. Details of the dynamic analyses, input forcing functions, and -response loadings are' presented in Section 3.9.2. .The basis for the design stress and deflection criteria is identified below: Allowable Stresses For normal operating conditions, the ' intent of Section III of the ASME Nuclear Power Plant Components Code is used as a basis for evaluating acceptability of calculated stresses. Both static and alternating stress intensities are considered. It should be noted that the allowable stresses in Section III of the ASME Code are based on unirradiated material properties. In view of the fact that irradiation increases the strength of the Type 304 stainless steel,used for the internals, although decreasing its elongation, it is considered that use of the allowable stresses in Section III is appropriate and conservative for irradiated internal structures. The allowable stress limits during the design-basis accident used for the core support structures are based on the intent of the draf t ASME Cade for Core Support Structures, Subsection NG, and the Criteria for Faulted Conditions. The stress criteria for the reactor internals that Westinghouse applied before the existence of Subsection NG of the ASMU code are composed of two parts, and depend upon the nature of the stress state membrane or bending. A direct or membrane stress has a uniform stress distribution over-the cross section. The allowable (maximum) membrane or direct stress is taken to be equal to-the stress corresponding to 20% of the uniform material strain or the yield strength whichever is higher. For unirradiated Type 304 stainless steel at operating temperature, the stress corresponding to 20% of the uniform strain is 39,500 psi. For a bending state of stress, the strain is linearly distri,buted over a cross section. The average strain value is, therefore, one-half of the outer fiber strain where the stress is maximum. Thus, by requiring the average bending stress to satisfy the allowable criteria for the direct state of stress, the average absolute strain may be 20% of the uniform strain. Consequently, the outer fiber strain may be 40% of the uniform strain. The I 3.9-81

B/B-FSAR maximum allowable outer fiber bending stress is then taken to be equal to the stress corresponding to 40% of the uniform strain or the-yield strength, whichever is higher. For unirradiated Type 304 stainless steel at operating temperature, the stress is 50,000 psi. 3.9.6-Inservice Testing of Pumps and valves Inservice testing of pumps and valve's will be done in accordance with a plan approved per 10 CFR 50.55g. 3.9.6.1 Innervice Testing of Pumos All ASME Code Class 1, 2, and 3 pumps requiring inservice testing are list' d in Table 3.9-11. Section III pumps not listed are e those excluded by the scope of IWP-1000. The pump test plan and schedule is included in the technical specifications, Chapter 16.0. 3.9.6.2 Incervice Testing of valves ASME Code Class 1, 2, and 3 valves requiring inservice testing are listed in Table 3.9-12 and identified as to valve category as defined by IWV-2110 of Section XI. Section III valves not listed are those excluded by the scope of IWV-1000. The _ valve test plan 3.9-81a g

B/B-FSAR 3.9.7 References 1. WCAP-8252, " Documentation of Selected Westinghouse Structural Analysis computer Codes," Revision 1, April 19 77. I 2. " Sample Analysis of a Class 1 Nuclear Piping System," prepared by ASME Working Group on Piping, ASME Publication, 1972. 3. W. H. Bamford and C. B. Buc ha let, " Methods for Fracture Mechanics - Analysis of Nuclear Reactor Vessels Under Severe Thermal Transients," WCAP-8510, June 1976. 4. WCAP-8317-A, " Prediction of the Flow-Induced Vibration of Reactor Internals by Scale Model Tests," July, 1975. 5. C. N. Bloyd, W. Ciarametaro, and N. R. Singleton, " Verification of Neutron Pad and 17 x 17 Guide Tube Designs by Properational Tests on the Trojan 1 Power Plant," WCAP-8780, May 1976. 6. K Takeuchi, et al., "Multiflex-A Fortran-IV Computer Program for Analyzing Thermal-Hydraulic-Structure System Dynamics," WCAP-8708, February 1976. l 7. G. J. Bohm and J. P. La Faille, " Reactor Internals Response Under a Blowdown Accident," First Intl. Conf. on Structural Mechanics in Reactor Technology, Berlin, September 20-24, 1971. 8. F. W. Cooper, Jr., "17 x 17 Drive Line Components Tests - I Phase 1B, II, III, D-Loop - Drop and Deflection," WCAP-8446 Proprietary and WCAP-8449, December 19 74. 9. S. Kraus, " Neutron Shielding Pads," UCAP-7870, May 19 72. 10. " Benchmark Problem Solutions Employed for Verification of WECAN Computer Program," WCAP-8929, June 1977.

3. 9-83 A-

TABLE 3.9-3 ALLOWABLE STRESSES FOR ASME SECTION III CLASS 1 COMPONENTS (Note 1) OPERATING CONDITION VESSELS / CLASSIFICATION TANKS PIPING PUMPS VALVES Design NB-3221 NB-3652 NB-3221 NB-3520 (D.es ig n) (Design) (Design) (Design) Normal NB-3222 NB-3653 NB-3222 NB-3525 (Level A) (Level A) (Level A) (Level A) Upset NB-3223 NS-3654 NB-3223 NB-3525 g (Level B) (Level B) (Level B) (Level B) g x Emergency NB-3224 NB-3655 NB-3224 NB-3526 (Level C) (Level C) (Level C) (Level C) Faulted NB-3225 'NB-3656 NB-3225 Note 2 (Level D) (Level D) (Level D) NOTES: 1. Limits identified refer to subsections of the ASME Code, Section III.

B/B-PSAR TABLE 3.9-3_ (Cont ' d) i NOTE 2: CLASS'1 VALVE FAULTED CONDITION CRITERIA ACTIVE INACTIVE-a) Calculate Pm from para. a) Calculate Pm from para. NB3545.1 with Internal NB3545.1 with Internal Pressure Ps = 1.25PS Pressure Ps = 1.50Ps Pm < l. 5Sm Pm 12,4Sm or 0.7Su b) Calculate Sn'from para. b) Calculate Sn-from para. NB3545.2 with NB3545.2-with Cp = 1.5 Cp = 1.5 Ps = 1.25Ps Ps = 1.50Ps Qt2 =0 Ot2 =0 Ped = 1.3X value of Ped Pod = 1.3X value of Ped' from equations of 3545.2(b) (1) from equations.of NB3545.2 (b) -(1) Sn < 3Sm ~ Sn13Sm l 3.9-89 l ~ g __

=- t B/B-FSAR \\ g, i TABLE 3.9 ' i i DESIGN LOADING COMBINATIONS

• FOR ASME CODE CLASS 2 AND 3.

COMPONENTS AND SUPPORTS CONDITION CLASSIFICATION LOADING ~ COMBINATION Design and Normal Design pressure, Design temperature **, l Dead weight, Nozzle loads *** l Upset Upset condition pressure, Upset condition metal temperature **, Deadweight,

OBE, Nozzle' loads ***

Emergency Emergency condition pressure, Emergency condition metal temperature **, Deadweight, Nozzle loads *** Faulted Faulted condition pressure, Paulted condition metal temperature **, Deadweight,

SSE, Nozzle loads ***

• The responses for each loading combination are combined using the absolute sum method.

On-a case-by-case basis, . algebraic summation may be used when signs are known for final design evaluations.

• • Temperature is used to determine allowable stress only.
    • Nozzle loads are those loads associated with the-partic-ular plant operating conditions for the component under consideration.

3.9-91 cc-

B/B-FSAR TABLE 3. 9-6 STRESS CRITERIA POR SAFETY-RELATED ASME CLASS 2 'AND CLASS 3 VESSELS ' CONDITION STRESS LIMITS

• Design and Normal The vessel shall conform to the requirements of ASME Section III, NC-3300 (or ND-3300)

Upset a, $ .1 S 1 (o E "L) + m b1 65 S 1 o Emergency. o < l.5 S m (o, or o ) + g bgl.80 S o Faulted a < 2. 0 S (o, or oL} + m < 2. 4 S

• Stress limits are taken from ASME III, Subsections NC and ND, or, for vessels procured prior to the incorporation of these limits into ASME III, from Code Case 1607.

3.9-92

B/B-FSAR TABLE 3.9-7 STRESS CRITERIA FOR ASME CODE CLASS 2 AND CLASS 3 INACTIVE PUMPS AND PUMP-SUPPORTS CONDITION STRESS LIMITS

• max **

Design and, Normal The pump shall conform to the requirements of ASME Section III,'NC-3400 (or ND-3400) Upset o

1. 1'1 8 1*1 m

(a, or o ) + "b $1.65 S L Emergency o 1 m1 5 S 1.2 (o, or o ) +ob 1 80 S 1 g Faulted om $2.0 S 1.5 (o E "L) + "b 12.4 S m

• Stress limits are taken from ASME III, Subsections NC and ND, or, for pumps procured prior to the incorporation of these limits into ASME III, from Code Case 1636.
  • The maximum pressure shall not exceed the tabulated factors listed under P times the design pressure.

max i 3.9-93 I

r B/B- ?SAR TABLE 3.9 ! DESIGN CRITERIA FOR ACTIVE PUMPS AND-PUMP SUPPORTS l f CONDITION ' DESIGN CRITERIA * ' Design and Normal ASME Section III ( Subsection NC-3400 and ND-3400 I m < l.0 S Upset a 1 b1 5S O,+ Emergency 1 a, $ .2 S b1 65 S 1 o,+o i Faulted cm < l.2 S 1 b1

• 0 0 o, + o
• The stress limits specified for active pumps are more restrictive than the ASME III limits to provide assurance that operability will not be impaired for any operating condition.

l t f I 3.9-94 m.

B/B-PSAR AMENDMENT 20 MAY 1979 TABLE 3.9-9 STRESS CRITERIA FOR SAFETY-RELATED ASME CODE CLASS 2 AND CLASS 3 APTTVR AND TNAPTTVR VALVES CONDITION STRESS LIMITS (NOTES 1-4, 6) Pmax (Note 5) l Design and Normal Valve bodies shall conform to ASME Section III. Upset o,11.1 S 1.1 b 1 65 S 1 (o f I* m L Emergency o 1 m $ .5 S 1.2 (m fG I+U b 1 80 S 1 L Faulted < 2. 0 S 1.5 m (o, or o ) +ob $2 4 S g NOTES: 1. Valve nozzle (piping load) stress analysis is not required when both of the following conditions are satisfied: (1) l the section modulus and area of every plane, normal to the flow, through the region defined as the valve' body crotch are at least 110% of those for the piping connected (or joined) to the valve body inlet and outlet nozzles; and, (2) code allowable stress, S, for valve body material is equal to or greater than the code allowable stress, S, of connected piping material. If the valve body material allowable stress is less than that of the connected piping, the valve section modulus and area as calculated in (1) above shall be multiplied by the ratio of S /S If the d881gn gy P a e unable to comply with this requirement, analysis procedure of NB3545.2 is an acceptable alternate method. 2. Casting quality factor of 1.0 chall be used. 3. These stress limits are applicable to the pressure retaining boundary, and include the effects of loads transmitted by the extended structures, when applicable. 3.9-95 Y

4 B-B/PSAR-TABLE 3.9-9 (Cont'd) 4. Design requirements listed in this table are.not applicable to valve discs, stems,-seat rings, or other parts of valves which are contained within the confines of the body and bonnet, or otherwise not part of the pressure boundary. 5.- The maximum pressure resulting from upset, emergency, or faulted conditions shall not exceed the tabulated factors listed under P times the design pressure or the rated pressure at thE applicable operating condition temperature. If the pressure rating limits are met at the operating conditions, the stress limits in this table are considered to be satisfied. 6. Stress limits are taken from ASME III, Subsections NC and ND, or, for valves procured prior to the incorporation of these limits into ASME III, from Code Case 1635. 4 d e i l 3.9-96 y.

B/B-FSAR to an indicator on the main control board (nominal accuracy will be 1 5%) and on the Remote Shutdown Panel, and to a flow switch which energizes a high flow alarm on the main control board. Transfer switches (REMOTE, LOCAL) are provided on the Remote Shutdown Panel. The auxiliary feedwater flow to the four steam generators is normally controlled from eight mrnual control stations mounted on the main centrol board if the transfer switch is in remote. Each manual control station electrically transmits a flow signal to an electric-to-pneumatic (E/P) converter. The pneumatic output flow signal is transmitted through a permissive three-way solenoid valve (which is deenergized for normal control) to the level control valves. An equipment status display (ESD) " alarm" will be actuated if the manual control station for any of the eight level (flow) control valves is set below 160 gpm while the associated auxiliary feedwater pump drive is not operatiag. Going to local control transfers the pneumatic control at the level control valves from the manual station on the main control board to a local controller mounted on the Remote Shutdown Panel, and energizes a " Valve on Local Control" alarm at the Manual Centrol Board. The pneumatic control of the level control valves from the Remote Shutdown Panel is identical with the control from the main control room. A failure in the control system for the control valve will cause the valve to fail open. The failure analysis is provided in Table 10.4-3. Flow restricting devices are provided upstream of each flow control valve (Figure 10.4-2) in order to limit flow in the unlikely event of a pipe break. f. Diverne Sources of Energy Diverse sources of energy are provided for the auxiliary feedwater pumps, valve oparators, instrumentation, and controls as discussed below. Electrical equipment and cabling for both auxiliary feedwater pumps is physically separated into separate i ESF trains. I 7.3-8 ( T

B/B-FSAR One auxiliary feedwater pump per unit is motor driven. Its source of energy is 4-kV ESF bus 141 ( 2 41) ~. The other auxiliary f eedwater pump for each unit is direct diesel driven. Fuel oil is supplied from its own Category I day tank; all necessary electrical auxiliaries for the diesel-driven auxiliary feedwater pump are' powered from its own battery system. The auxiliary feedwater pump diesel, its auxiliaries and d 4 7.3-8a l 7

B/B-FSAR '9.2.2.2.2.3l component c'ocling surge Tanks There are two surge tanks provided. Each' tank serves one loop of the systems under normal operating conditions. The tanks are connected to' the pump suction lines. 'The tanks purpose is to: (1) accomodate system water expansion and contraction due 'to temperature changes; (2) accomodate inleakage to the system; (3) provide makeup for small system leaks until they can be - isola te d ; and (4) as a' point of chemical addition to the system. Design of the tanks is covered in~ Table 9.2-3. 9.2.2.2.2.4 component ecoling Heat Exchangers l Three heat exchangers serve the system. Each heat' exchanger is sized f or 100% cat.acity of normal single unit heat loads..one heat exchanger serves each loop with the third available as a maintenance spare or for additional heat requirements of a i particular loop. Design of the heat exchangers is covered in Table 9.2-3. 9.2.2.2.2.5 component cooling Pum ps_ There are five pumps serving the system. Under normal conditions, up to two pumps will serve each loop with the fifth j pump availablo as a maintenance spare or for additional load l requirements of a particular loop. Design of the pumps is covered in Table 9. 2-3. 9.2.2.2.2.6 component cooling Instrumentation j The operation of the loop is monitored with the following instrumentation: a temperature detector in the component cooling pump a. suction line; b. temperature detectors in the outlet lines for the component cooling heat exchangers; c. pressure detectors on the lines between the component cooling pumps and the component cooling heat exchangers; d. a temperature and return-flow indicator in the pump suction header from the heat exchangers; e. redundant safety-related flow indicators on the l reactor coolant pump motor and shaft seal cooling water discharge line; f. water-level indicators on the component cooling surge tank; l 9.2-12 '{ l t u

m B/B-FSAR The instrumentation -in the CCWS is provided primarily - for initial system tlow ba. lancing and for monitoring purposes during normal operation. Thus failure of any of this instrumentation has no efrect on. system performance. Exceptions to this are: a. letdown heat exchanger CCWS flow controllers, b. reactor coolant pump thermal barrier outlet tlow controller, and c. component cooling surge tank radiation control valve. The letdown heat exchanger tube side outlet temperature controls a butterfly valve which regulates the CCNS flow to the shell side of this heat exchanger. Should the controller fail in a way to snut off CCWS flow to the circuit, a high temperature alarm will sound in the control room allowing the operator to take corrective action. Redundant safety-related indication of component cooling water flow to the reactor coolant pump thermal barrier is provided and alarmed in the main control board. The reactor coolant pump (RCP) thermal barrier outlet header has a flow controller which causes a motor-operated valve to throttle close in this line in the event of high flow (an indication of a broken RCP thermal barrier). Should the controller not operate properly, an in-creasing level is noted in the CCWS surge tank, resulting in a high level alarm, if not isolated. A second motor-operated valve in series with the flow control valve is available for manual isolation of the line if required. Additionally, two level instruments are provided on each surge tank, both of which will give a high level alarm in the control room. Each component cooling surge tank vent has an air operated valve which will close on 'a high radiation signal f rom the radiation monitors in the discharge headers from the CCWS heat exchangers. This high radiation alarm normally indicates a primary to CCWS leak. Three radiation monitors are provided, any of which will alarm and close the vent valve on both surge tanks. 9.2.2.4.5 Electrical Power Supply The normal power supply to the system is f rom the ESF buses. A full description of the power supply is given in Subsection 8.3.1.1. 9.2.2.5 Tests and Inspections During the life of the Station, the Component Cooling System is in continuous operation and performance tests are not required. Standby pumps are rotated in service on a scheduled basis to obtain even wear. Preoperational tests are perf ormed on the system. The equipment manutacturer's recommendations and station practices are considered in determining required maintenance. 9.2 - 19 m

l' B/B-FSAR 10.4.6 Condensate cleanup System 10.4.6.1 Design Bases The condensate cleanu.p systems at Bpron and Braidwood will be utilized'primarily during plant startup to flush the-condensate, condensate booster, and feedwater systems. -This system will not be operated continuously. The equipment is designed to treat one-third of the condensate system flowrate supplied from the discharge header of the condensate ~ pumps. The treated water returns to the condensate booster pumps suction header. 'The condensate cleanup system is designed to produce an-effluent at the design flowrate within the following limits: 2 i a. Sodium < 1 ppb b. Conductivity < 0.1 pmho/cm c. SO < 1 ppb 4 l d. Iron < 10 ppb ~ All pressurized vessels in the system are designed and con-structed in accordance with the ASME code for Unfired Pressure Vessels of ASME Division 1, Section VIII. No part of the system is safety-related, thus it is designated Safety Category II. 10.4.6.2

System Description

10.4.6.2.1 General Description and System Operation The condensate cleanup system for each station consists of four mixed bed polishers each designed for a capacity of 3750 gpm. Two vessels are normally assigned to each unit, however, the valving arrangement permits operation of the vessels with either unit. Normally the flowrate from each unit is equally divided among two vessels. The external resin regeneration system, common to all four mixed bed polishers, consists of one resin mixing and storage tank, one resin separation and cation regeneration tank, and one anion regeneration tank. Resin is sluiced from a mixed bed polisher to the resin separation and cation regeneration tank. The anion and cation resin are separated and the anion resin in transferred to the anion regeneration tank. The cation resin is regenerated with sulfuric acid, and the anion resin is regenerated with sodium hydroxide. After regeneration is complete, the resins are transferred to the resin mixing and storage tank. 10.4-8 m M

B/B-FSAR l When placed in service, the operation of this system is con-trolled and maintained by a solid state controller. The control system will prevent the initiation of any automatic or sequence of operations that would conflict with any opera-tion already in progress, whether such operation is under automatic or manual control. The operation status of each polisher and each regeneration vessel, including which auto-matic nequence is in progress, is indicated by means of lights on the polisher control panel. Improper operation of the regeneration system and components will cause an' alarm to sound and the system will be shut down. Improper _ regeneration solution strength will sound an alarm and the' system will shut down if the situation is not corrected within five minutes. 10.4.6.2.2 Component Description 10.4.6.2.2.1 Mixed Bed Polisher Each of the four mixed bed polishers are 114 inches in diameter with a 60-inch side seam and are sized for a flograte of 3750 of anion gpm. Each vesseg contains approximately 5 kg/ft resin and Skg/ft of cation resin. The vessels are equipped with viewports on the side shell and an illumination port in the upper head. The mixed bed polishers are designed to Section VIII of the ASME Boiler and Pressure Vessel Code, and are rated at 300 psig. A high pressure resin trap in each polisher effluent line is designed to retain particles larger than 50 mesh. 10.4.6.2.2.2 Resin Separation and Cation Regeneration Tank The resin separation and cation regeneration tank is 84 inches in diameter with a 174-inch side shell and equipped with four viewports in the side shell and an illumination port in the top head. The design pressure of the tank is 100 psig. The resin is backwashed to separate the anion and cation resins. The anion resin is drawn off before the cation resin is regenerated. A 3-foot diameter by 5-foot side shell resin hopper is located above the resin separation and cation storage tank to make up for any lost resin. la.4.6.2.2.3 Anion Regeneration Tank Anion resin is transferred to this tank to be regenerated with caustic. The anion regeneration tank is 78 inches in diameter witha 120-inch side shell. The tank is equipped with two view ports in the side shell and one illumination port in the top head. The design pressure is 100 psig. 10.4-8a 7

F -v B/B-FSAR I 10.4I6.2.2.4 Rosin Mix and Storage Tank The resin mix and storage tank is 96 inches in diameter with a 102-inch side seam and.the design pressure is 100 psig. Three viewports are located in the side shell and one illumina-tion port is located in the top head. The tank is sized to contain a complete change of resin for one mixed bed polisher. The anion and cation resin is sluiced from their respective regeneration tanks to this storage tank. The resins are mixed and stored until being transferred to a mixed bed pelisher. 10.4.6.2.2.5 Regeneration Equipment The acid regeneration skid consists of a 200-gallon acid storage, tank, two metering pumps, and a dilution station. The storage tank is sized for two regenerations. The caustic regeneration skid consists of a 700-gallon caustic tank, two metering pumps, and a dilution station. A hot water tank provides dilution water for regeneration of the anion resin. Both regeneration systems are equipped with the necessary instrumentation and controls to automatically provide regenera-tion chemicals in the required amount, temperature, and concen-tration to the respective regeneration tanks. 10.4.6.2.2.6 Sluice Water Pumps Two 400-gpm, 100-psig pumps are used to supply water from the condensate storage tank for sluicing the resin between the various tanks. The pumps also supply the required dilution water to the acid and caustic regeneration systems. 10.4.6.3 Safety Evaluation The condensate cleanup system is a non-safety-related system and is not requir.ed for safe shutdown of the plant. 10.4.6.4 Testing and Inspection All pressurized tanks are designed in accordance with the ASME Code for Unfired Pressure Vessels of ASME Division 1, Section VIII. All equipment is factory inspected and tested in accordance with the applicable equipment specifications and codes. Preoperational tests will oc performed on this system. The equipment manufacturer's recommendations and station pactices are considered in determining required maintenance. 10.4.7 Condensate and Feedwater System The purpose of the Condensate and Feedwater System is to provide feedwater from the condenser to the steam generators. This subsection discusses the Condensate and Feedwater System from the condenser to the connection with the steam generators. 10.4-8b l 7 e e

B/B-PSAR 10.4.7.1-Design Bases 10.4.7.1.1 Safety Design Bases The only part of the Condensate and Feedwater System classified l as saf ety-related (i.e., required for safe shutdown or in the event of postulated accidents) is the main feedwater piping from l the preheater section of the steam generators to, and including, I the outern. cat containment isolaticn and check valves; the l tempering feedwater lines between the steam generator preheater l bypass connections and the outermost check and isolation valves; the interconnecting piping between the tempering lines and the auxiliary f eedwater system, and the chemical feed piping from the l interface into the tempering piping to, and including, the l shutoff valves. These parts of the system are designated as safety Category I, Quality Group B. 10.4.7.2

System Description

The Condensate and Feedwater System consists of the piping, valves, pumps, heat exchangers, controls, instrumentation, and the associated equipment and subsystems that supply.the steam i l generators with heated feedwater in a closed steam cycle using. regenerative feedwater heating. The system is shown in Figure

10. 4-1, Sheets 1 and 2.

There are four 1/3-capacity centrifugal condensate pumps per unit with motor drives and ccmmon suction and common discharge headers, and four 1/3-capacity condensate booster pumps per unit with common suction and discharge headers. Each condensate and condensate booster pump set is driven by a single motor. Three sets of pumps are normally in operation. The fourth set of pumps will automatically start on low pressure at the feedwater pump suction to assure adequate flow to the feedwater pumps. The Feedwater System is of the closed type, with deaerating i accomplished in the condenser. The condensate pumps take suction-from the condenser hotwell and pump condensate through the air ejector condensers and the gland steam condensers to the suction of the condensate booster pumps. These pump the condensate through six stages of low-pressure feedwater heating to the feedwater pumps. The water discharge from the feedwater pumps l 10.4-8c l ~ []

BYRON /BRAIDWOOD UNITS 1&2 UNRESOLVED GENERIC SAFETY ISSUES A-12 Fracture Toughness of Steam Generator and Rcactor Coolant Pump Supports During the course of the licensing action for North Anna Power Station Unit No. 1 and 2, a number of questions wela raised as to the potential for lamellar tearing and low fracture toughness of the steam generator and reactor coolant pump support materials for those facilities. Two different steel specifications (ASTM A36-70a and ASTM A572-70a) cove ed most of the material used for these supports. Toughness tests, not originally specified and not in the relevant ASTM specifications, were made on those heats for which excess material was available. The toughness of the A36 steel was found to be adequate, but the toughness of the A572 steel was relatively poor at an operating temperature of 80 F. Since similar materials and designs have been used on other nuclear plants, the concerns regarding the supports for the North Anna facilities are applicable to other PWR plants. It was therefore necessary to reassess the fracture toughness of the steam generator and the reactor coolant pump support materials for all operating PWR plants and thosain CP and OL review. Fracture toughness and lamellar testing has been adequately addressed for the Byron /Braidwood NSSS supports. The mater-ials used in the Byron /Braidwood supports were, as a minimum charpy V-notch tested in accordance with NF-2300. In general, items subject to through thickness stresses were ultrasonically tested to preclude lamellar tearing. Further-more, the Byron /Braidwood NSSS supports employ the same materials and more stringent fracture toughness requirements than Zion Units 1 and 2. The staff, has stated in its Safety Evaluation Report " Fracture Toughness of Steam Gener-ator and Reactor Coolant Pump Supports" transmitted January 1981, that ..."the steam generator and reactor coolant pump supports at Zion Units 1 and 2 possess adequate fracture toughness to merit a Group III rating per NUREG-0577. Therefore, we consider the fracture toughness of the steam generator and reactor coolant pump supports acceptable." Thus, Ceco has concluded that Byron /Braidwood Units 1 and 2 can be operated until there is an ultimate resolution of j this generic issue without undue risk to the health and safety of the public. I

B/B-FSAR s QUESTION 130.43 "The descriptive information of Category I structures ~ othet than containment is not in accordance with the provisions of the R. G. 1.70. Provide sufficient infor-mation, illustrated by sketches in the FSAR Section 3.8.4 to enable the' staff to perform a meaningful review. . Referencing the FSAR Section 1.2 which illustrates the ceneral layout of the plant is insufficient in detail for a structural review."

RESPONSE

A plan view of auxiliary-fuel handling building shear walls, elevation views, shear wall-slab diaphram connection (above grade and below grade), and a typical uall corner are'shown in new Figures 3.8-52 through 3.8-58. i Byron river screen house

• foundation plans, floor and roof

,_,.. framing plans, and. elevation views are shown in new Figures 3.8-59 through 3.8-64. A Byron ESW cooling tower foundation plan, an air inlet . plan, a fill support beam plan, a distribution support beam ~ plan, a roof framing plan, and tower sections are shown in new Figures 3.8-65 through 3.8-73. A Braidwood lake screen house foundation plan, floor framing plans, and elevation views are shown in new Figures 3.8-74 through 3.8-78. A plan view and section of a typical Byron deep well enclosure are shown in naw Figure 3.8-79. A safety valve' room plan view and elevation views are shown in new Figures 3.8-80 through 3.8-82. An elevation view and typical dome section of the refueling water storage tank are shown in new Figures 3.8-83 and 3.8-84, respectively. These figures, along with the informati'on in Subsection 3.8.4, provide sufficient detail for a structural review. Q130.43-1

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r Q \\ BYR 1/BRAIDWCOD STRUCTURAL DESIGN AUDIT (SUPPLEMENTAL INFORMATION) OCTOBER, 1981 Action Item 4A: Additional-information in response to Question 130.06 (Seismic Reassessment) 1. Average actual material strengths. S&L to provide methodology of calculating the average actual material strengths. Rebar - The average ac'.ual material strength was calculsted from the yield strength values given in the certified material test reports. The heat number and the yield strength values of all the reinforcing steel was listed and averaged. To assure this average value did not exceed 70% of the ultimate strength, 100 randomly selected heat numbers were listed. The ultimate strength values of these heats were recorded and averaged. The average actual material strength did not exceed 70% of the average ultimate strength. Structural Steel - A random sample of 287 of the approximate total of 700 certified material test reports for A-36 material were listed with the heat number and yield strengths. These yield strength values were averaged. A similar sampling of 107 of the approximate total of 200 reports were listed and averaged for A-588 or A-572 Grade 50 material. The averaged yield strength values were compared to the respective average ultimate strength values of approximately 100 test values. The average yield strength values did not exceed 70% of the average ultimate strength values. Concrete - The average compressive strengths at 91 days for both 3500 psi and 5500 psi concrete has been calculated. This average is computed based on all the in-process tests taken during construction. t i.

I 4 s BIRON/BRAIDWOOD STRUCTURAL DESIGN AUDIT (SUPPLEMENTAL INFORMATION) OCTOBER, 1981 i Action Item 4A: Additional information in response to Question 130.06 (Seismic Reassessment) 2. HVAC Supports S&L to assess the behavior of the HVAC supports which have been determined to exceed the yield strength under an SSE' event but will not collapse. HVAC support structure number 3902 (sketch attached) as a whole will behave elastically and will not collapse under the specified loads for the following reasons: a) The maximum deflection at the free-end of the structure is.353 inch, which gives a reasonably small value of 1/270 for the A /L ratio. b) The structure as a whole will experience stresses below yield, except for the 5-1/2 inch-long leg near the support, where the member will experience a local inelastic joint rotation due to an increase in the i moment of 47% above the elastic limit. This increase results in a joint rotation ductility factor of 1.95, which is considered very low. c) The largest b/t rat.io of the members is 7, compared to the 8.5 permitted by Section 2.7 of the AISC Specification for projecting elements in compression. ~ d) - An increase of about 10% to 12% in the yield strength _of t e mater ai l under dynamic loading has not been h utilized. 1 s I /- i I e pc se 1 of 2 i. ~

g s. o ?'. BYRON /BRAIDWOOD ACTION 4A HVAC SUPPORTS (Cont'd.) 0 i

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BYRON /BRAIDWOOD STRUCTURAL DESIGN AUDIT (SUPPLEMENTAL INFORMATION)~ OCTOBER, 1981 a Action Item 4A: Additional-information in response to Question 130.06 (Seismic Reassessment) 3. - Unique differences for the Marble Hill Model S&L to list the unique structural features on the Marble Hill plant. The unique structural differences are as follows: 1) One 40-foot long wall in the Auxiliary Building at elevation 401'-0" is l'-0" thick at Byron /Braidwood and 2'-0" thick at Marble Hill. 2) Select columns in the Marble Hill Auxiliary Building at elevations 346'-0", 364'-0", and 383'-0" have a concrete strength of 5500 psi rather than 3500 psi. e 1 F' e e e o 9 9 6 4 e E e !^ i l

i BYRON /BRAIDWOOD STRUCTURAL DESIGN AUDIT (SUPPLEMENTAL INFORMATION) OCTOBER, 1981 l Action Item 4A: Additional information in response to Question 130.06 (Seismic Reassessment) 4'.. Selection Method for OBE Reassessment of Structural Steel Beams S&L to provide methodology in selecting structural steel beams for reassessment due to an OBE event. The eleven structural steel beams in the Auxiliary Building and the eight structural steel beams in the Containment Building were selected for the OBE comparison because these beams showed the highest stresses under the SSE reassessment. I W O 4 e e a t l -...m.- _m -_

t P BYRON /BRAIDWOOD. STRUCTURAL DESIGN AUDIT (SUPPLEMENTAL INFORMATION) ~ OCTOBER, 1981 Action' Item 4A: Additional information in response to Question 130.06 (Seismic Reassessment) 5. Figure 130.06-2 S&L to provide the bedrock elevation for the Byron, Braidwood and Marble Hill sites. S&L also to provide the extent of backfill under the Fuel Handling Building at the Braidwood site. Refer to attached Figure 130.06-2 for bedrock elevations. At the Braidwood site, compacted fill with a minimum relative density of 85% was placed on the glacial till which is at elevation 378'-0". Refer to FSAR Figure 2.5-76 Section 2-2. ( e e O k D

,0 S K E Q 4 LRO O 9 LAT R H

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BYRON /BRAIDWOOD STRUCTURAL DESIGN AUDIT (SUPPLEMENTAL INFORMATION) OCTOBER, 1981 Action Item 4A: Additional information in response to Question 130.06 (Seismic Reassessment) 6. Unique areas of the Auxiliary-Fuel Handling Building Complex S&L to explain the unique areas of the Auxiliary-Fuel Handling Building mat design as shown in Figure 130.06-3. The unique areas of the mat are unique only for the reinforcing steel. Concrete strengths and thicknesses are the same. The Marble Hill design assumptions used were more conservative; therefore the reinforcing in these areas was increased. None of these areas using the Byron /Braidwood design show any over- . stress during reassessment. I f e e e 4 5 9

C ~.., BYRON /BRAIDWOOD STRUCTURAL DESIGN AUDIT (SUPPLEMENTAL INFORMATION) . OCTOBER, 1981 Action Item 19: Two-inch thick premolded filler joint between - Containment and Auxiliary Building i S&L to provide physical properties of Ethafoam. A copy of the manufacturer's product literature is being pro-vided to the NRC which gives the physical properties of Etha-foam. Based on these properties, the forces generated on the i structures due to relative seismic displacement were accounted - for in the design of the Containn.ent and Auxiliary Building walls. e i e i G '**%-ca s

Important Advantages Of ETHAFOAM* Brand Polyethylene Foam Resilient-resists multiple impacts o o Lightweight-saves handling and shipping costs-and it's buoyant o Moisture resistant-closed cell structure resists water pickap and moisture transmission o Chemical resistant o Available in fabricated shapes through extensive fabricator network o Stable over wide temperature range o Compressible-select compressive strength o Workable-easy to make prototypes I Reusable o

• Trademark of The Dow Chemical Company s'

ETHAFOAM Brand Polyethylene Foam Because of its unusual This booklet is intended to provide combination of characteristics and basic information on the wide properties, ETHAFOAM brand range of applications and the polyethylene foam is used outstanding properties of this extensively in: unique foam material. o Cushion Packaging ~ ~ " " ' ~ " " ~ o Transportation Applications ~N o Water Sports Equipment o Industrial Flotation o Sporting and Recreation Goods o Construction j in addition, engineers and designers continue to find new applications for ETHAFOAM where f~ f' - __(N ~ ' ' no satisfactory material was I previously available. -~ ' g ETHAFOAM expanded polyethylene foam is available in black and g natural white, and in a selection of C 4 shapes and densities. A tough, 1 flexible closed-cell material, it is: I d o Energy-absorbent o Resilient o Lightweight o Flexible over a wide temperature

• -"cF^

-"E range o Resistant to chemicals a.,,,, :1 o Buoyant s 1 i o Easy to fabricate r s ~%.. - J / j ~ - - -

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Morkats And Applications The past and present applications Government Construction listed below,m which ETHAFOAM his been selected for use, illustrate Applications Seeiani Becker the versatility of ETHAFOAM brand Pressure Relief Joint Filler products, and may suggest Munition Wadding and Plugs Closure Strips potential new uses. Any Filler Pads Curing Blanket cpplications, however, should be Missile Containerization Gym Floor Underlayment tested thoroughly by the user to Instrument Cushioning Underground Cable Wrap determine the suitability of the Equipment Vibration Dampening m'terial for a specific use. Seismic Joint Filler Water Sports Coid Wate, gipe insuiation Cushion Equipment Appliances Life Vests Packaging Ski Beits Corner Pads Kickboards Gaskets Pads and Saddles a Vibration Pads Encapsulation a e Lounges Case Inserts Rings Grommets Overwrap Sheeting Canoe Liners, Seats, Sponsons Tape Bracing and Blocking Pool Cove and Liner Backing Display Cases Boat Flotation Transportation Industrial Flotation Dust and Water Gaskets Booms p Carpet Underlayment Protective Separators Buoys p Aircraf t Seating Child Seat Cushioning Car Roof Underlayment g g Recreation Goods Gym and Floor Exercise Mats tVrestling Mats Wall Pads Padding -Sports Equipmnt Toboggan Pads Ski Lift Seat Pad Golf Bag and Backpack Str 3ps All-Terrain Vehicle Flotatior. Handlebar Padding Athletic Field Pads and Ma kers Archery Targets Sleeping Bag Pads 2

Packaging and reduced waste. Names and addresses of experienced An increasing number of fabricators are available from any manufacturers are using of the Dow sales offices listed on ETHAFOAM as a cushioning the back cover. medium in protective packages. For a complete discussion of the ETHAFOAM brand pMyethylene use of ETHAFOAM expanded form consists of tiny -!osed cells' polyethylene in packaging, and which absorb impact shocks while information on designing packages h::! ding the packaged product in with this material, consult Dow's position. Because it is flexible and technical bulletin Form No.172-221, risilient, the foam also dampens Designing Packages to Survive vibrations during handling and Shipping and Handling with shipping. ETHAFOAM Brand Polyethylene The resistance of ETHAFOAM Foam. brand plastic foam to moisture and water vapor transmission is --- \\~- 1-- I_- rnother important advantage when moisture-sensitive products are to p 4.. be packaged. In addition the foam k j h, 8 is noncorrosive to the packaged 'g .j -[ product. [ w# s The savings gained by using i ..[.

• f--

ETHAFOAM can be significant. The 'j l i,_ material is clean and lightweight and generally may be used / repeatedly with minimal loss of effectiveness. Also,it can be g t ~ lie xn _J fabricated with little tooling cxpense. These characteristics all '" ~ ~ -? I i contribute to lower costs through rrduced transportation charges, a lower expenditures for tooling, l { picking labor and housekeeping, m ,t x. s ~' l ; (Q - Ql s. , d4 ('} / f f. xdmM[ [ \\. pt \\ \\ i J 3

Snortinn And ""*'97~^ b' "9 'v$'* * $ '" P' P*"'^i"'"9 "d "" P * ' ' " r a sports equipment seek to absorb can be dangerous and should be Recreation impacts and reduce the prospect of discouraged. injury. The suitability of any Equ.ipment materiai for use in sucn systems Water S orts And h and the system itself should be ETHAFOAM brand polyethylene form is widely used in recreation evalu ted carefully and thoroughly, A@atic equipment. Among its advantages since no cushioning system can c.ro strength, toughness, light provide absolute protection. In any Accessories w ight, good moistere resistance ti@ iming sti@ m hei# and cushioning properties. an individual may receive impacts ETHAFOAM 220 is about thirty ETHAFOAM has been selected by that exceed the capability of the times as light as water. manufacturers for use in cushioning system or may be Because of this excellent cushioning systems for sports and injured from landing in an out-of-buoyancy, and its resilience and recreation equipment to help control position. toughness, ETHAFOAM brand foam soften impacts or blows. Protective equipment and padding is widely used in equipment for ETHAFOAM has also been selected can help reduce the possibility of water sports and boating. for use in athletic equipment, and injuries, but is imperative in any Planks and rounds of ETHAFOAM to provide cushioning in sports activity to promote safety 220 and sheets of ETHAFOAM 221 snowmobile seat composites. awareness in the participants. and ETHAFOAM 222 expanded Participation in sports without polyethylene have been approved i by the United States .[A v x? ~r V ~' ~ } r r-v; ;-

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Coast Guard Underwriters Transnortation m king ci sure strips used with Lc boratories, Inc. for use in r corrugated metal sheet. It is also personal flotation devices. The automotive industry is another effective as thermal insulation in user of ETHAFOAM brand plastic specialty low-temperature gcmes includes tether ball, ring tiss, and water basketball. f am. Components of cars, trucks, applications such as pipe, tanks ETHAFOAM plastic foam is used in aircraf t, and other vehicles are and heat exchangers. producing floating lounge chairs fabricated from this tough, resilient When applied a' an underlayment matenal. end marine devices, such as floats in various floor systems, and buoys. It also finds use in The flexibility and compressibility ETHAFOAM reduces noise. cinoe parts, such as liners, seats of this foam permit it to conform to in dam construction, its toughness, cnd sponsons. surface variations. water resistance, ability to insulate For swimmers, water skiers and Dust and water gaskets of and light weight have established boaters, water s'ei belts, buoyant ETHAFOAM are econoinical and ETHAFOAM 221 and ETHAFOAM jickets, surfboards, and efficient. Such gaskets are durable, 222 plastic foam sheet as an ideal kickboards are made using even in contact with grease, oil, concrete-curing blanket. ETHAFOAM brand polyethylene and most chemicals. fum. Construction G# Jr Tr - q' ~ r ETHAFOAM brand polyethylene i foam is the logical choice for many construction applications because 9, of its insulating value, flexibility ,e' N e 4 i and compressibility. For certain 87 INT other applications, its resistance to h, water or chemicals or to the i [f I f g\\ adverse effects of temperature [j L f g variation are factors of prime / importance. In building construction, the shape _- ~ and depth of sealants applied to joints can be controlled with resilient, non-staining backer rods of ETHAFOAIC SB. ETHAFOAM is an excellent nonporous, flexible material for / 5

Preportios And Characteristics

• l I

Compression Characteristics Compressive strength FIGURE 1-Typical Compressive Strength Compressive strength is a measure Curves for Expanded Polyethylene Plank of the firmness of the product or og, the amount of compression it go experiences under a given. load. Typical compressive strength values for various ETHAFOAM brand products at different 70 percentages of deflection are shown in Figure 1 and Table 1. 60 3 ( _. su n ,. 7, *y, y v g-x l ETHAFOAM 900 )! .[ l 4 (it j 3:8hw) f;l l 2

.s ~-,: Y f

l - t n - 3 sSR A~iC \\ j g g 40 j j i i 1 j h f L a 3 ) 30 } ETHAFOAM 600 20 ETHAFOAM 403 l 10 ; p~~".,,.._' ETHAFOAM 220 10 20 30 40 50 STRAIN. % 6

• Nc to to conseered specincetion values.

TABLE 1-Typical Properties

• of ETHAFOAM Products p

,.n e x -, m - ~f >~ g. w. y m m n - n.,- r- - - -, m:, l i-v. t a. a.w. ~. # 24.J A,a- .h,; A .J. ;-17 ~ .A',._ A., ,hh. ajA.h.J % a.- nm.j Density (PCF) ASTM D 3575 Test C.. 1.6 2.7 2.2 4 6 9 %'=2.0 %'= 1.0 Cell Size (MM) ASTM D 3576 %'=2.2 %'= 1.2 1.3 1.2 1.1 1.0 Dow Modified %*=2.4 % '= 1.4 Compressive Strength (PSI)..... ASTM D 3575 Test B at5% 23 3.i 6.5 15 44 at 10%. 3 5 8.5 17 48 at 25%... 3 6 8 12 20 54 at 50 %. 15 15 - 21 33 75 Compressive Creep (% Deflection). ASTM D 3575 Test BB Loaded @ specified PSI static load for 1000 hrs. 75*F. <5 @ 1.5 PSI <5 @3 PSI <5 @ 4 PSI <4 @23 PSI 160*F, <5 @.25 PSI <6 @ 1 PSI <5 @ 2 PSI <7 @ 7 PSI Tensile Strength (PSI). ASTM D 3575 Test E 60 100 50 65 80 130 Tensile Elongation (%) ASTM D 3575 Test E 60 80 60 70 70 100 Teer Strength (Ib./In.). ASTM D 3575 Test D. 30 15 25 30 50 Flexural Modulus (PSI) ASTM D-790 300 500 700 1100 Buoyancy (PCF). ASTM D 3575 Test AA 55-60 5!Hi0 5540 5!Hi0 53-58 50-55 Thermal Conductivity (BTU-inJhr. ft* *F) at %*=0.3 75*F mean temperature ASTM D 3575 Test EE %'=0.4 %'=0.3 0.4 0.4 0.4 0.4 Method B % * =0.4 %"=0.4 Thermal Stabildy (% Shrinkage). . ASTM D 3575 Test F Conditioned @ specified ternp.with no foad: 24 hrs -1.0 @ 165'F -10 @ 165'F -1.0 @ 165'F -1.0 @ 165*F -10 @ 165'F 48 hrs -1.5 @ 165*F -2.4 @ 165*F -1.5 @ 165*F -1.0 @ 165'F -1.0 @ 165'F s' 7 w to i. coa m.o p.cie=.e n..w.s

i FIGURE 2-Typical Compressive Creep of FIGURE 4-Typical Compressive Creep of e ETHAFOAM 220 Plastic Foam 4' x 4" x 2* ETHAFOAM 600 Plastic Foam 4" x 4" x 4* Sample

• Sample
• 20 20 4.0 psi (?73*F 1.0 psi @160*F j

/ i 0 g o 6 psi @73'F 9 'i I h10 5 g 0.5 psi @160*F + 1.0 psi @73*F "] $l @73*F + 2.0 psi @100*F f 0.5 psi @73*F E 5 3.0 psi @73*F 2.0 psi 0 i 1.5 psi @73*F @ 73*F 0 200 400 600 800 1000 ,..... _ } _ .~, i _ TIME, HOURS 0 0 200 400 600 800 1000 FIGURE 5-Typical Compressive Creep of ME, HOURS ETHAFOAM 900 Plastic Foam 4* x 4* x 2~ Sample

• 20 FIGURE 3-Typical Compressive Creep of ETHAFOAM 400 4" x 4* x 2" Sample
• 20

/f 15 [ 2.0 psi @160*F 15 / b 10 f g j 20 psi @73*F (2' x 2" x 1' O 4.0 psi @73*F W U 10 0 8 psi @160*F

• ' ~ -

uj 1.5 psi @150*F l* 5 8 'f'- u. J 3.0 psi @73*F I. __ 5'f 1.0 psi @160*F + 2.0 psi @73*F 5 psi @160*F 10 psi @73'F 1.0 psi @73*F 0 0 200 400 600 800 1000 0r TIME. HOURS 0 200 400 600 800 1000 TIME, HOURS 8 m u, v. cmw.a w.e.sme.m

Compressive creep When low. F!GURE 6-Typical Recovery of density materials are loaded ETHAFOAM (At Room Temperature) After continuously over a period of time, 50% Compression for 22 Hours @ Room they tend to " creep"; that is, they Temperature

• lose some of their original 100 thickness. Creep characteristics, which are detc/ mined by the variables of load, time and E 90 temperature reflect the long-term g

load-carrying ability of a material E ,r ETHAFOAM 900 I[ end affec: its cushioning ability.

/

Typical compressive creep s 80 characteristics of ETHAFOAM E brand products are shown in O Figures 2 through 5-h Recovery Characteristics of ETHAFoAM 400 and ETHAFoAM 600 Fall Between Those of Compressive Set Recovery g ETHAFoAM 220 and ETHAFoAM 900 w Compressive set is a measure of the thickness a material fails to O recover after it has undergone E 60 compressive creep and the load has been removed. To express this relationship in another way-so I percent recovery plus percent 0 20 40 60 80 100 compression set equals 100 RECOVERY TIME, HOURS percent. Figure 6 shows the recovery characteristics cf ETHAFOAM 900 and ETHAFOAM 220 after compression to 50 o of [ ~ ' ~ .~ their original thickness for 22 l } hours at room temperature. e e r -;;a -/ a l~ e 3h a ' h, / 1 /V.4 ~r y f.m ,9 ,~ s e, 4 i I-I m% /, j WE .) /r

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Tonsile and Tear Thermal Water Vapor Transmission erg,go,y p,,,,,,,,,,,,o,,ee, Strength Properties an exceiient barrier to the transmission or penetration of ETHAFOAM brand plastic foam is a Thermal Conductivity The amount water vapor. ETHAFOAM products t; ugh material. Typical tensile and of heat transmitted through a foam have perm ratings of less than 0.4 t:Ir properties for ETHAFOAM is determined by many factors. perm-inch, as shown in Table 1. brand products are listed in including the nature of the base When a skin is left on the surface, Tcble 1, page 7. material, the cell size and the the water vapor transmission can degree of closed-cell structure. drop to 0.2 perm-inch. -{~-~ i i~- f 71 y'- Permanence of good thermal g insulating value often depends on Buoyancy The light weight and O' the ability of the foam to resist the excellent water and water vapor f. M7

• j,#

effects of water and moisture resistance of EMM MaM 'y ses'rbIn vapor. Typical conductivity values plastic foam make it an excellent b.: I.i 'tI are listed in Table 1, page 7. material for flotation applications. 0 r k' '. ' 4 See Table 1 and page 4. I,Y %. '. , }I f je' Thermal Stability The thermal J-7 NNE i ' * 'J stability of ETHAFOAM plastic foam 7* M is suitable for most applications. WARNING-ETHAFOAM brand ,1 1<~ 4 polyethylene foam is combustible and ' *; f;. Without load, ETHAFOAM products y I can be ignited by contact with a fra-e qN Q'j, are used successfully,in many Therefore it snould not be execsed to A 'j applications in which they are flames or other ignition sources. The bumng charactens ics f EWCW exposed intermittentiY to vary significantly depending on the temperatures up to 180*F. Typical amount of matenal present and other 'g' ~ N ^~j dimensional stability data are given combustion conditions. Large quantities ] g- - in Table 1. Like other thermoplastic to$9e e n rao$y and hoc :e _.LL-materials, polyethylene foam dense smoke. Under normat comoust40, ~ becomes more flexible at high conditions. carbon rnonox ce is generated. Additional toxic st.bstances temperatures and more rigid at low may be released under less tnan full temperatures. combustion concitions. In f:refigntmg situations. dense smoke should be avoided and resoirators used. Water Resistance Water Absorption Even when it is totally immersed. ETHAFOAM plastic foam absorbs a negligible amount of water. In fact, essentially all the water pickup measured in immersion tests is accounted for by the open cut cells on the surface (see Table 1). / 10

l i Chemical-Solvent Liaht Stabilitu p lyethylene film to improve the O J acceptance of paints and inks. can Res tance exteneed exposure to soniignt be used to prepare ETHAFOAu for is causes degradation of ETHAFOAM. the application of coatings and ETHAFOAM plastic foam is This degradation is first noted in a adhesives. Coatings such as vinyl, resistant to solvents and most yellowing of the surface. After to protect foam against ultraviolet cther chem,ct.s at room lor'ger periods of exposure, some degradation, increase toughness i temperature E. id contains no loss of physical properties will and weather resistance or to create i water-soluble constituents. As a occur. The degree of aging decorative effects can be applied buoyancy material in water, it,s depends on the climate. For to ETHAFOAM. Such coatings are i unaffected by contact with fuel oil example, three months of exposure available from: and other hydrocarbons. However* to the summer sun in Arizona will p; if the foam is immersed for a long produce some degradation of period of time in certain solvents. ETHAFOAM, while exposure to including gasoline, it will swell-sunlight for a whole year in tbsorb solvent, and lose strength. Michigan has very little effect. United Coatings Also, at elevated temperatures the 1130 E. Sprague foam becomes mere susceptible to For applications under direct Spokane, WA 99202 I cttack by certain - vvents. Acids sunlight, where long-term l and elkalis norma.-.. ave no effect performance is required, a Bee Chemical Co. on ETH AFOAM polyethylene foam, protective coating should be Lansing, IL 60438 but strong oxidizing agents may applied over the foam. The surface cause degradation, especially at of polyethylene is not naturally Food and Drug l high temperatures. No universal receptive to coatings, but flame j solvent for ETHAFOAM plastic treatments and other treatments, pggkggjgg foam is known. such as those used on C a Compatibility ,*r h [ ' Some forms of ETHAFOAM 220, [/4 ETHAFOAM 221 ETHAFOAM 222. p[,.,rdk ETHAFOAM 400. ETHAFOAM 600, f jMC fQM',/,q and ETHAFOAM 900 expanded C, / J polyethylene foam comply with the

g d

Food Amendment of the U.S. Food - Q,c9 % .. =.. )# 7 'En j Drug and Cosmetic Act when used J. 4 5 unmodified and according to good Y:;7,~1 %p' /' { b. %- q _u manufacturing practices for food [ ,; f /,f O [pg }1 packaging applications. Contact your neorest Dow sales office for 6 ( /..g i 7 & j,. ( further information. p, / g g i y' b us. - / -v y__- /_ y 11 ~

? Fcbrication + Ease of fabrication is an important advantage of ETHAFOAM brand Bondinn r""h*"a'*'m**'aa oa U ETHAFOAM products is available plastic foam. It can be skived to ETHAFOAM plastic foam can be from it.e Dow Sales Offices listed precise thickness, cut and shaped adhered to itself by use of heat on the back cover of this bulletin. to form custom parts, and joined to alone. The two surfaces are heated For more detailed information on itself or other materials without a simultaneously with hot air or by the use of ETHAFOAM in major investment in equipment. use of a plate heated to about packaging, ask for Form No. 350'F. When the surface of the 172-221, Designing Packages to Author.ized foam begins to soften, the pieces Survive Shipping and Handling are quickiy joined witn moderate with ETHAFOAM Brand Fabricators pressure. Only short cooling time Polyethylene Foam. is necessary. This simple method Generally, the most economical, produces an excellent durable quickest and most reliable way to bond. The surface of the heating secure special shapes or custom plate can be coated with Teflon parts made from ETHAFOAM is to fluorocarbon resin or Dow Corning cmploy the services of an 1890 protective sealer to facilitate authorized fabricator. Fabricators release of the melted foam. are equipped not only with conventional tools, but also with Note When heat is used to cut or special devices, such as machines form ETHAFOAM plastic foam, with blades or bits that operate adequate ventilation must be with a slicing action, splitters with provided to carry smoke or fumes scalloped blades that cut up to 50 away from the breathing zone of workmen. feet per minute without producing dust, electrically heated resistance CAUTION When large quantities of wires, and contoured heated ETHAFOAM plastic foam are stored molds. For names and addresses of or fabricated, small quantities of fabricators please consult the the blowing agent released from nearest Dow sales office. the foam may tend to accelerate corrosion of heaters and boilers. Corrosion can eventually create holes in the combustion chamber, leading to the release of combustion gases and carbon monoxide, which would endanger the health of persons in the area. Heating equipment should be inspected regularly during every heating season to check for pinholes or larger defects in the combustion chamber. 12

TABLE 2 - Adhesives for ETHAFOAM g._.-.- ... ~ m..,,,.,.,, _,.,,. g-- .l g3 , ;., 77 a* +F m _. -:._. __m., u.~...._ ' _... a.w. A,. r _..a.- , ___i i. m. _ _ , _ ~

1. Contact Adhesive ~

Structural Pliogrip AD965; Neohte Goodyear Tire & Rubber Compary Applications all purpose adhesive Akron. Ohio Requinng Polyfcamstix 1579 Adhesive Products Corpora!.cn High Bond Bronx, New York Strength Foam . Bondmaster G-590 PPG To Foam, Wood, ' Bloomfield, N J. Metal etc. Scott.h Grip industnal 3M Company Packaging Adhesive 4693 or 4729 St. Paul, Minnesota Applications. Armstrong $20 Armstrong Cork Company . Lancaster, Pennsylvania i ' 2. Pressure Sensitive Gaskets & ' Dri Tac 1519 Adhesive Products Corporation Sealer Stnps Bronx. New York

3. Double Faced Self.

Gaskets & Fox Film Morgan Adhesives Ccmpany Adhesive Films Sealer Stnps Stow Ohio S-510 Series Fasson Products Painesville. Ohio

4. Release Paper Used with Press.

Riegel Paper Company Sensitive Adhesives N.Y, N.Y.

5. Hot Melt See Contacts Thermogno 212; USM Corporation Easy to apply Thermognp 1398 Cambndge. Massachusetts Short open time Hot Gnp Adhesive Products Corporatan Bronx. New York i

NOTE. Achesive toit coaters avadable from Biack Bros Co. inc.. Menoota inino.s Hot Mell apphCators avasiaDie from USM Corporal @n

2 BYRON /BRAIDWOOD STRUCTURAL DESIGN AUDIT (SUPPLEMENTAL INFORMATION) OCTOBER, 1981 Action Item 23: Masonry Walls The NRC provided required action for the masonry walls at Byron /Braidwood Stations in a letter from B. J. Youngblood to L. O. Del George of Commonwealth Edison Company, dated November 30, 1981. The following required action was re-quested: a) All unbuilt walls will be reinforced concrete or reinforced masonry designed to comply with SEB Masonry Wall Criteria, Revision 1. A list of unbuilt walls as of October 20, 1981 should~be provided. b) Perform a systematic mapping of structural cracks on all safety related walls and discuss safety implications of any existing cracks and possible disposition. c) Confirm in a letter stating full conforman'ce to QA requirements of Appendix B to CFR Part-50. Actual audit by NRC QA Branch may be necessary for key walls. d) Provide calculations for each type of Category I masonry walls for staff's review to verify conformance - with the intent of "SEB Criteria for Safety-Related Masonry Wall Evaluation" July 1981, Revision 1. e) Provide detailed assessment of behavior of walls assuming existence of reasonable crack patterna for the safety related walls for the staff's review and acceptance. The assessment should be realistic and adequate in describing actual wall behavior under SSE shaking. The assessment should utilize any existing test data pertinent. Pending staff's review, need for additional analysis, limited tests, and fixes may arise to demonstrate the adequacy of walls. The CEC OL should not be contingent upon completion of any additional analysis testing or fixes. This will be treated as a license condition and any remedial action (testing and fixes) will have to be completed to the satisfaction of the NRC staff prior to the resumption of power operation after the first refueling. Page 1 of 4

g LYRON/BRAIDWOOD STRUCTURAL DESIGN AUDIT (SUPPLEMENTAL INFORMATION) OCTOBER, 1981

Response

a). lists the masonry walls not built at Byron Station as of October 20, 1981. lists the masonry walls not built at Braidwood Station as of October 20, 1981..All of these walls will be reinforced concrete or reinforced masonry in accordance with "SEB Interim Criteria for Safety-Related Masonry Walls Evaluation", Revision 1. b) At this time there are no masonry walls with structural cracks at either Byron Station or Braidwood Station. If structural cracks appear in safety-related masonry walls in the future, an inspection program will be ini-tiated to survey these walls and map any structural cracks. c) The design' drawings and specifications fcr safety-related masonry walls require all work to be performed and inspected in accordance with an approved quality assurance program that is in conformance with the requirements of-Appendix B to 10 CFR Part.50. Specific QA/QC documentation is available for NRC review at Byron Station and Braidwood Station. d) Calculations for the following three representative walls are attached as Enclosure 3 for the staff's review: 1. Wall 4A-88, Elevation 392'-6", Auxiliary Building 2. Wall 7A-19, Elevation 439'-0", Auxiliary Building 3. Wall 5A-134, Elevation 401'-0", Fuel Handling Building a These walls, and all other safety-related walls for Byron /Braidwood Stations, have been designed in accordance with the requirements of NCMA-1974 and the load and load combinations as described in Table C of the calculations. There are no significant deviations between NCMA-1974 and the "SEB Criteria for Safety Related Masonry Wall Evaluation", Revision 1, July 1981. e) Commonwealth Edison Company has reviewed the data obtained from the Shaking Table Study of Single Story Masonry Houses performed at the Earthquake Engineering Research Center (1) at the University of California, Berkeley, to determine crack patterns under dynamic Page 2 of 4

STRUCTURAL DESIGN AUDIT (SUPPLEMENTAL INFORMATION) -OCTOBER, 1981 loading. The test results for vertically spanning non-load bearing walls subject to out-of-plane loads, as summarized in Ref erence 1, indicates that a horizon-tal crack develops in the top third height for walls without openings. This horizontal crack is attributed to tension i-the wall perpendicular -to the bed joint due to the vertical wall moment. For walls with open-ings, diagonal and/or horizontal cracks were initiated either near the top or bottom corner of the opening. l It should be noted that horizontally spanning walls were not tested at Berkeley. The vertically spanning n >n-load bearing walls subj ect to in-plane loads de-veloped horizontal cracks near the base of the wall and horizontal and/or diagonal cracks at the ends of the door.and window lintels. This cracking was as-sociated with the rigid body rocking of the wall. l All saf ety related concrete masonry. walls at Byron / l Braidwood have been designed to span horizontally. It is the opinion of Commonwealth Edison Company that f horizontal cracks would not develop under dynamic condi-tions due to out-of-plane loading, since the stresses l per pendicular to the bed joint will always be less than one-half those parallel to the bed joint. How-l ever, if a horizontal crack is postulated in the safety j related walls at Byron /Braidwood, the walls will retain their integrity, since the walls are designed to span horizontally. Should a diagonal crack occur locally adjacent to an opening, the joint reinforcement which is.provided in alternate masonry courses would prevent crack propagation, and would assure the structural integrity of the wall for out-of-plane loads. The dynamic test data indicates that horizontal or diagonal cracks may f orm due to in-plane loading on the masonry walls. Diagonal cracks may form locally at the openings. However, the joint reinforcement l provided in alternate masonry courses would prevent crack propagation, and would assure structural integrity of the walls. It is the opinion of Commonwealth Edison Company, theref ore, that only horizontal crack f orma-tion may have an eff ect on structural integrity of masonry walls used at the Byron /B raidwood S tations. However with the presence of a horizontal crack due to in-plane loads, ti.e behavior of the wall remains unchanged from the initial design, and the wall will - retain its structural integrity to carry out-of-plane loads, since the walls are designed to span horizontally. Page 3 of 4 p 9 WA. u: _, W- ~-

~ BYRON /BRAIDWOOD . STRUCTURAL DESIGN AUDIT (SUPPLEMENTAL INFORMATION) OCTOBER, 1981 l LIST OF REFERENCES 1..

Clough, R. W.,

Mayes, R. L, and Gulkan, P.,

" Shaking Table Study of Single-Story Masonry flouses", Volume 3: Summary, Condlusions and Recommendations, Earthquake and Engitiecring Research Center, Report No. UCB/EERC-79/25, September, 1979. 1 6 Page 4 of 4

g .L4d@l%5BsV 7s Byron Station Unbuilt Walls Wall Location Drawing Elevation On Between A-208 346'-0" 25 N.1 & N.8 N N.24 & 24.8 24.8 N& P.9 24.5 P.8 & Q Q.1 24 & 24.5 A-210 346'-0" S.5 19 & 19.6 19 3.5 & Q Q.8 19 & 19.6 A-211 346'-0" S.7 21 & 23 .U.S 21 & 23.4 A-212 343'-0" X.5 21.5 & 23.5 X.6 21.1 & 21.5 A-213 354'-0" 15 U.4 & U.5 U.4 14.9 & 15.3 15.3 U & U.4 U 15.5 & 15.6 U.2 15.5 & 15.6 15.5 0 & U.2 20.3 U & U.2 U 20 & 20.3 U.2 20 & 20.3 21 U.4 & U.S U.4 21.1 & 20.8 20.8 U & U.4 A-220 364'-0" 21.8 N.9 & P.5 N.9 21.8 & 23.5 24 N.5 & N.9 N.5 24 & 24.8 25 N.9 & N.5 N.5 25 & 25.6 25 L& L.2 25.3 L& L.2 L.2 25 & 25.3 A-223 364'-0" 19 0& Q.9 19.6 0.8 & 5.3 19 S.5 & U 19 U&V U.2 21 & 18.6 18.6 S.9 & U.S 18.6 U.9 & V.3 19.1 v.3 & V.5 19.6 v.3 & W A-225. 364'-0" Y 12.5 & 13.8 Y 21.2 & 18.5 1 of 4

Byron Station .,j Unbuilt Walls Wall Location ^ Drawing Elevation On Between - i A-229 383'-0" L.5 10 & 10.6 L.4 10.6 & 10.8 "~' 11.3-L.5 & Q.1 ~ 12.1 N& P.9 13 M.1 & P.6 14 M.'5 & P.5 ~~ P.5 14 & 14.5 i 15 P.2 & P.6 14.8 M&N A-230 - 383'-0" 18.8 L &.M.1 M.1 18.8 & 20 M.1 20 & 21 20.1 L.8 & M L.8 20.1 & 21 i 24 P& P.3 P.3 24 & 24.2 24.2 P.3 & P.5 l-A-232 383'-0" 19.8 V.7 & V.9 ^ U.5-19.2 & 21 19 S.5 & S.9 I S.5 19 & 19.8 19.8 08& S.5 S.2 19 & 19.8 i A-233 391'-0" 12 N&Q 13 M& Q A-234 392'-6" U 15 & 17 1 394'-6" Q 15 & 21 375'-6" Q 15 & 21 A-239 401'-0" N 23 & 24 23.8 N.7 & P.8 24.8 N.7 & P.8 25.2 N.7 & P.8 A-242 401'-0" V.2 15 & 15.5 15.5 U.8 & V.2 V.2 20.5 & 21 20.5 U.8 & V.2 17.4 Q& V.6 18.6 0 & 'I 9 19 0 & V.2 '~ Q 16.5 & 16.9 9 2 of 4 s g 4 .,,, ~

y ye x. ~ Enclos ure-1 / =e ( 7 Byron Station Unbuilt Walls r. ~ Wall Location Drawing Elevation On Between s'

1. t j

A-242 401'-0" 16.5 Q & Q.4 j Q 19.1 & 19.5 19.5 0 & Q.4 i Q.1 15 & 15.5 05 15 & 15.5 15.5 0 1 & Q.5 8 01 20.5 & 21 Q.5 20.5 & 21 20.5 0 1 & Q.5 A-243 401'-0" Q.1 21 & 20.8 05 21 & 20.8 20.8 Q.1 & Q.5 23 01& S S 23 & 24.5 23.4 Q.1 & Q.2 23.6 0.1 & Q.2 0.2 23.4 & 23.6 A-246 409'-6" 17.4 0 & V.9 18.6 0 & V.9 ~ P.5 26.4 & 26.6 26.4 L.5 & P.5 L.5 26.8 & 28.8 a 28.8 L.1 & L.5 i 4 A-249' 414'-0" 22.2 U.'6 & V.9 U.6 22.2 & 23.8 A-251 414'-0" L.5 24.3 & 24.6 5 L.8 24.2 & 24.5 i 24.5 L&N l A-254 426'-0" N-23.1 & 23.9 M 24 & 24.3 L.9 24 & 24.5 24.3 L.9 & M 'A-257 426'-0" U.8 16.2 & 17.1 17.1 U.4 & U.8 19 01& S.2 S.2 19 & 19.9 19.9 Q.1 & S ~ S 19.9 & 20.8 ' ' ^ 16.2 U.4 & U.9 E d 3 of 4 h ,3. a

7 ", 1 Byron Station Unbuilt Walls I Wall Location Drawing Elevation On Between A-267 451'-0" L.1 11 & 11.2 11.1 L& L.1 ,A-268 451'-0" 23 0& P.8 P.9 23 & 23.5 23.5 P.9 & Q 24.8 L& L.1 L.1 24.8 & 25 A-270 451'-0" S 12 & 15 A-272 451'-0" S 21 & 24 A-277 467'-0" S 12 & 15 A-279 467'-0" S 21 & 24 f 5 e 4 of 4 i Braidwood Station Unbuilt Walls Wall Location Drawing Elevation On Between A-206 330'-0" Q.3 13.1 & 14 Q.3 21.9 & 23 A-208 346'-0" 22.2 L & L8 23.2 L & L.4 01 24 & 24.5 A-209 346'-0" U.4 12.5 & 15 S.7 13 & 15 A-211 346'-0" U.5 21 & 23.4 S.7 21 & 23 A-212 343'-0" X.6 14 & 14.7 X.5 12 & 14 X.6 21.2 & 21.5 X.5 21.5 & 23.5 A-219 364'-0" P 10.5 & 10.9 I 10.5 P& P.4 A-220 364'-0" 25 L& L.2 L.2 25 & 25.3 25.3 L& L.2 A-225 364'-0" Y 12.5 & 13.8 Y. 21.2 & 23.5 A-226 355'-4" L.6 19.3 & 20 20 L.7 & L.7.5 L.7.5 20 & 21.9 21.9 L& L.7 A-229 383'-0" P.1 10 & 10.6 L.5 10 & 10.6 L.4 10.6'& 10.8 L.6 10.6 & 10.7 11 N.1 & N.9 11.3 L.5 & Q.1 12 L& L.5 12.1 L.8 & P.9 12.5 M & N.5 13 M.1 & P.9 L.8 13.1 & 14.1 L.8 15.1 & 15.9 15.1 L& L.8 1 of 4 4

I Enclos ure. 2 ^ Braidwood Station Unbuilt Walls Wall Location Drawing Elevation On Between A-229 383'-0" M 13.2 & 14.3 14.3 M&N 14 M.2 & P.5 P.5 14 & 14.5 15 P.2 & P.6 18 L.5 & M A-230 383'-0" 19.9 N.5 & P 21.9 N& P 22.9 N& P 'P.1 23.2 & 23.9 24 N.1 & 0.9 18.8 L & M1 18 L.5 & M L.8 20.1 & 21 M.1 18.8 & 20 M.1 20 & 21 N 22.9 & 23.9 N 24.2 & 24.7 20.1 L.8 & M A-232 383'-0" 16.5 V.7 & V.9 4 19.8 V.7 & V.9 A-233 391'-6" P.1 10.1 & 10.5 L.7 10.1 & 10.5 11.4 L.6 & Q 12 L.0 & Q N.4 11 & 11.4 A-234 375'-6" Q 17 & 19 394'-6" Q 15 & 21 A-237 401'-0" 7.5 N.4 & P.4 P.5 7.5 & 7.7 P 7.5 & 7.7 N.5 7.5 & 7.7 A-239 401'-0" 20 L&M 23.8 N.7 & P.8 24.2 L.7 & M M 24.2 & 24.5 24.9 N.7 & P.8 25.2 N.7 & P.8 2 of 4 e =

1 Enclos uro. 2 Braidwood Station Unbuilt Walls Wall Location Drawing Elevation On Between A-229 383'-0" M 13.2 & 14.3 14.3 M&N 14 M.2 & P.5 P.5 14 & 14.5 15 P.2 & P.6 18 L.5 & M A-230 383'-0" 19.9 N.5 & P 21.9 N& P 22.9 N& P 'P.1 23.2 & 23.9 24 N.1 & 0.9 18.8 L & !!1 18 L.5 & M L.8 20.1 & 21 M.1 18.8 & 20 M.1 20 & 21 N 22.9 & 23.9 N 24.2 & 24.7 20.1 L.8 & M t A-232 333'-0" 16.5 V.7 & V.9 19.8 V.7 & V.9 1 A-233 391'-6" P.1 10.1 & 10.5 L.7 10.1 & 10.5 11.4 L.6 & Q 12 L.8 & Q N.4 11 & 11.4 A-234 375'-6" Q 17 & 19 394'-6" Q 15 & 21 A-237 401'-0" 7.5 N,4 & P.4 P.5 7.5 & 7.7 P 7.5 & 7.7 N.5 7.5 & 7.7 A-239 401'-0" 20 L&M 23.8 N.7 & P.8 24.2 L.7 & M M 24.2 & 24.5 24.9 N.7 & P.8 25.2 N.7 & P.8 2 of 4 l

1 ~ Braidwood Station Unbuilt Walls Wall Location Drawing Elevation On Between A-240 401'-0" 28.7 N.5 & P.5 P.5 28.3 & 28.7 P 28.3 & 28.7 N.5 28.3 & 28.7 A-241 401'-0" V 12.5 & 14.7 A-242 401'-0" Q.1 15 & 15.5 .Q.5 15 & 15.5 15.5 Q.1 & Q.5 17 S&V 17.4 0& V.9 i 18.6 0& V.9 19 S&V Q.1 20.7 & 21 05 20.7 & 21 20.5 01& Q.5 V 21 & 24 A-243 401'-0" V 21.2 & 23.6 A-247 415'-0" 7.1 L& L.5 L.5 7.1 & 7.6 28.9 L & L.5 j A-251 414'-0" M 24.1 & 24.5 L.8 24.3 & 24.7 A-255 426'-0" 28.3 L & L.4 A-256 426'-0" V 12 & 15 14.9 U.2 & V.1 A-258 426'-0" V 21 & 24 21.2 U.2 & V.1 A-264 439"-0" P.6 29.2 & 30 A-208 451'-0" 18.5 P.5 & P.9 A-271 451'-0" S.3 15.2 & 17.8 S.3 18.2 & 19.8 9 3 of 4 h

L4iBht@JeJud 73 Braidwood Station Unbuilt Walls Wall Location Drawing Elevation On Between A-274 451'-0" S 12 & 15 A-275 459'-2" S 17.7 & 18 08 17.7 & 18 S.3 15.1 & 16.7 S.3 16.7 & 17.8 20.5 0& S S.3 18.2 & 19.2 S.3 19.2 & 20.8 15.3 0& S A-276 A59'-0" S 21 & 24 A-277 467'-0" S 12 & 15 A-278 467'-4" S.3 16.7 & 17.8 S.3 15.3 & 16.7 S,3 18.2 & 19.2 S.3 19.2 & 20.2 S.4 20.2 & 20.8 15.5 -Q & S 17.7 S & Q.9 18 S& Q.9 S 17.7 & 18 Q.9 17.7 & 18 S 18 & 21 20.4 0& S A-279 467'-0" S 21 & 24 A-282 477'-0" 17 L&Q 18 L& N.5 'A-283 477'-0" 19 L&Q Q 18 & 19 0 18.& 18.9 A-312 401'-0" 18 Y& Z5 18.5 Y & Y.9 19 Y& Z.5 Z.6 17.1 & 18.5 Z.6 18.6 & 20.1 i 18.6 Z.6 & Z.9 Z.9 17.9 & 18.6 Y.4 20.3 & 20.7 Y.5 20.5 & 21. l Y.4 15 & 15.2 Y.3 15.1 & 15.3 Z.5 18.3 & 18.7 4 of 4 t

1 Design Summary I. Material Properties A. Masonry Units Densi ty: 145 pcf (solid block) 10 5 pcf (hollow block) Compressive strength of unit (f c') : 1800 psi (solid block) 1000 psi (hollow block) Ultimate Compressive strength of masonry (f'm): 1350 psi (solid & hollow) B. Mor tar : Type M Compressive strength of mortar (m ) : 2500 psi g II. Allowable S tresses (per NCMA) A. OBE Load Combination Flexural tension parallel to bed joint (f t g ): 78 psi (solid block) 46 psi (hollow block) Flexurel tension perpendicular to bed joint ~(f tl): 39 psi (solid block) 23 psi (hollow block) Shear (fv): 34 psi (solid & hollow block) B. SSE Load Combinatici. (overstress f actor = 1.67) Flexural tension parallel to bed joint (f t a ): 130 psi (solid block) 77 psi (hollow block) Flexural tension perpendicular to bed joint (f ti) : 65 psi (solid block) 38 psi (hollow block) Shear (fv): 57 psi (solid & hollow block) III. Frequency Calculaticns (Ref. S& L SD&DD Eepor t #25) A. Simply supported Walls 1 T f = 56 EmIm (cps) y(Lw)4 2 if 144 W 1 of 18

ebIsD14EMr7; B. Ccntilivered Wallo x 20 EmI m g =2 v 14 4W,( Lw) 4 cps) IV. Nomenclat ure ~ A = Not unit area of wall in horizontal section (in /f t) 'Em = Modulus of Elasticity = 1000 f m' (psi) fc' = Canpressive strength of individual masonry (psi) f'm = Ultimate compressive strength of masonry (psi) hw = Height of wall (f t) Im = Moment of inertia of masonry (in 4/f t) Lw = Horizontal length of wall (f t) M = Compressive strength of mortar (psi) g S = Section modulus of masonry (in3/f t) tw = Thickness of wall (in) Ww = Weight of masonry wall (lb/sf sq wall) V. Load Combinations Table C Load Load Factors Combi na ti on-D L Ro Po To Eo Ra Ta Pa Ess Notes Normal 1.0 1.0 1.0 1.0 1.0 Normal / Severe Env. 1.0 1.0 1.0 1.0 1.0 1.0 (1) Abnormal. 1.0 1.0 1.0 1.0 1.5 Normal /Ex - treme Env. 1.0 1.0 1.0 1.0 1.0 1.0 (2) Abnormal / Severe Env. 1.0 1.0 1.25 1.0 1.0 1.25 Abnormal /Ex-treme Env. 1.0 1.0 1.0 1.0 1.0 1.0 Notes (1) G ove rni ng OBE Load C omb. (2) Governing SSE Load C omb. l Controlling Load combinations are normal / severe environmental and normal / extreme environmental. No walls are subj ected to LOCA or SRV loads. 2 of 18 .~

VI. Damping Values A. Seismic Damping Values For Walls OBE: 4% SSE: 7% B. Seismic Damping Values for Attachments to Walls OBE: 4% SSd: 7% i VII. Allowable Stresses for Steel Wall Supports (Columns & Tees) OBE: AISC Allowable SSE: 1.6 x ( AISC Allowable) not to exceed 0.95 Fy VIII. ..ttachments to Walls A. Contingency load to account for 2" p' and under control instr umentation piping and 4"# and under electrical condui t: P = 180 lbs per foot of height f or solid block walls P = 135 lbs pe r f c;o's of hei gh t for hollow block walls B. For design purposes, consider the load as follows: P r This configuration results in maximum moment, and o mQy .,Gw maximum shear at the supports. +----. l = C. Use peak seismic acceleration f or attachment loads. ~ ~ t/ss 13 j '\\ l 4- \\ W:_ No llA 68 ' ' ~ flecR EL. 9 ?' 4* d '_r I Daum .%. A.2w, s us 60stornneo d. G I I,, g l s.c' _.we Oc*.m-6'- 2 * --* I NORTil 1 PV.vst PLAd VIC W z pm., ev 3 of 18 _-e

~ Enclosure 3 Well Pnrcmetern l'-ll 5/8" (solid) tw o Lw = 8'-3" hw 6'-2" = Density = 145 pcf f'c 1800 psi = f' m = 1350 psi l 2 A e ( 23 5/8 ") ( 12 ") = 283.5 in /f t (12" ) (23 5/8") 3. I' 4 12 = 13186 in /f t (12")(23 5/8) 2 8 3 6 = 1116.3 in /ft (14 5 pc f ) ( 23 5/8) Ww = 12 in/ f t = 285.5 psf 6 1000 f' m = 1. 3 5 x 10 pgg Em = Considering l'-0" beam strip f rom the 8'-3" simply supported span: f = 56/ Em I 23,744 gw gw4 6 56 (1. 35x 10_) ( 13186) _ 21r (144)(285.5)(8.25)4 = 86.2 cps Period, T = 0. 0116 seconds "gh" values f or wall f requency 86.2 cps > 33 cps, rigid zone. Use response spectra values for E. 401'-0" 4% 1Dmnpi ng (OB E) I 7% Damping (SSE) l Allowable S tress Rigid Pea k Fb Fv Spectra No. 10 7 -OB -EW 0.195 0.53 78 psi 34 psi Spectra No. 107-SS-EW 0.38 1.0 130 psi 57 psi P p k tw Lv Simply Supported Span Li 2 4 I 1 ,= W, P = 18 0 lbs/f t ht. i 11' I1 I I I E W = 285. 5 psf L Lw = 8.2 5 f t i l 1. OB E Mmax = ( 0.195) (1/8) ( 285. 5) ( 8.25) 2+(0. 53) (1/4) (180) ( 8. 25) = 67 0. 4 f t-l bs Vmax = ( 0.195) (1/2) ( 285. 5) ( 8. 25) + ( 0. 53) (180) = 3 25.1 lbs 4 of 18

f b = f = ( 67 7.2 psi < 78 psi (OK)

4) 12)

= { = 3 2 5.1 fy

1. 2 psi < 3 4 psi (OK)

= 3, 2. SSE Mmax = ( 0. 3 8) (1/8) ( 285. 5) ( 8. 25) 2 + (1. 0) (1/4) (180) ( 8. 25) 12 94. 3 f t-lbs = i Vmax = ( 0. 38) (1/2) ( 285. 5) ( 8.25) + (1. 0) (180) = 62 7. 5,1bs 1294

13. 9 Psi <; 130 psi (OK) iS

= ll16 3 EV 2.2 psi < 57 psi (OK) = 283.5 3 CHECK COLUMNS W8X13 Sx = 9.91 in 'PT/" Allowable Bending Stress (F ) b C OBE 24ksi (Continuous Lateral Support) --r SSE 0.95Fy = 34.2 ksi _4 q Wobe = 325.1 + 220. 7 = 545. 8 #/f t W 1 Wss e = 62 7. 5 + 4 24.1 = 1051. 6 p/f t __3 6 _4 1 ( 0.195) ( 1/2) ( 285. 5) ( 4. 5) + 0. 53 (180) 4.5 Ft. Span Vmax (OB E) = = 220. 7 lbs Vmax (SSE) ( 0. 3 8) ( 1/2) ( 285. 5) ( 4. 5) + 1. 0 (180) = = 4 24.1 lbs OBE Mmax = 1/8 ( 545. 8) ( 6.167) 2 = 2594. 7 f t-lbs fb = 2594. = 3.1 ksi < 24 ksi (OK) n 999 x 9.9 SSE Mmax = 1/8 (1051. 6) ( 6.167) 2 = 4 999. 3 f t-lbs i fb = 4999.3 X 12 1000 x 9.91 = 6.1 ksi <s 34.2 ksi (OK) a 5 of 18

L4nTs952UWT3 Ch:ck %T Section' cnd Expansion Anchors .2w 6'. 4 WT9x42. 5 with 3/4" 9 anchors 0 3'-10" O.C 2 A = 12 " x tw = 12 x 0. 52 6 =. 6 in RTEM 1 1 OBE Vmax = 325.1 ft. f h f7 8'-3" Span i 1 SSE Vmax = 6 27. 5 t Check Shear on WT Stem, Use 1 Ft. Strip 1. OB E f v = 3 5. 1bs

51. 5 psi < 14,400 psi = 0.4 FY (0:i)-

2 = 2. SSE f v = 6 7. 5 1 bs. = 9 9. 6 ps i < 23, 20 0 ps i = 1. 6 x 0. 4 x Fy 60 K) 2 Anchor Bol ts 3/4 "7 Anchor Allcuables lbs 1. OB E 325.1 x 6.167 ft = 501.2 lb/ Bolt 3,400lb (OK) lbs 2. SSE 627.5 f x 6.167 ft = 9 67. 4 lb/B ol t <C 4,230lb (OK) Check Shear on Reduced Wall Area Due to WT Connecti on (1 f t. Str ip) 1. OB E 325.1 lb/'Ft x 1 Ft = 2. 3 psi < 34 psi (O K) 1/2(23.625) (12 in) i 2. SSE 627. 5 lb/Ft x 1 Ft = 4. 4 ps i < 58 psi (O K) 1/2(23.625) (12) 6 of 18 i

Floor EL. 439'-0* Drawing No. A-262 n / Wall No. 7 A-19 Orientation E-W 7A 26. wcit tril M M 1 I I I t , _s I i o'. o' 5'. q d n 3'. y a 4'.8K" 24' - o" N W8vi3 v/Br 3 5 2u CE. 5 V Stxoou [- Syrru cave"b NcRTal t L. 'li'l'- o" i l 1 I I T l l 8 l l l I

l 3-i l

i t g s l N g I I a l I i I I g \\ l I t L 4W. o* 1. e WALL PROPERTIES: t = 13 5/8" Horizontal Walls: y 2 L, = 2 4 ' -0 " A = 36.0 in /ft b = 8'-0" I = 929.4 in /ft y Dens i ty = 105 pcf S = 159. 9 in~/f t g f' = 1800 psi W, = 42.6 psf 6 f'

1350 pai E,

1000 f' = 1.35 x 10 psi 7 of 18

L#ic)losure 15 10'-0" Section I Horizontal Simply Supported Span L = Using l'-0" Beam S trip

• dontrols Analysis
• 56

/ EmIm 4 W V 144W L y g ( 3 x1 ) 29*4I 4 = 40. 3 cps, T

0. 025 s ec6 f

22 { 0 i g Values f rom response Spectra @ EL. 439'-0" (Base of Wall) H RIGID (f 33 CPS) PEAK Spectra No: 109-OB-NS 0.25 1.3 4% damping Spectra No: 109-SS-NS 0.55 1.7 7% damping ' Allowable S tress OBE SSE P 46 psi 77 psi b F 34 psi 57 psi y P P lbs For Attachment w a P,= 135ft. of ht. Load to Hollow L M y w 2 I Block Walls y_ v3 I i1 1 i 1 W = 42.6 psf 10'-0" L = y (0. 2 5) ( 42. 6) (10) 2 + (1. 3) ( 135) (10) = 571. 9 f t-lbs. 1. ODE M = max 12 fg= f= ygf9 = 42.9 psi <P = 46 psi (OK) bu ( 0. 2 5) ( 42. 6) ( 10 ) + 1.3 (135) = 2 2 8. 8 lbs. V = max 2 = 6.4 psi < 34 psi (OK) f = y 3 (O. 55) ( 42. 6) (10) 2 + (1. 7) (135) (10) = 8 6 6. '6 f t-l bs 2. SSE M = max '5 6 5 psi <C P

77 psi (OK) fg

159. = bu ( 0. 55) ( 42. 6) (10) + 1. 7 ( 13 5) = 3 4 6. 7 lbs.. V = max 346.7 = 9.6 psi < 57' psi (OK) f = y 8 of 18

Check Masority Steel Columns w/ 10 '-0" 5'-4" Shaar Forces e Controls Analysis, Worst Loading s v lbs W = 22 8. 8 + 203. 9 = 4 32. 7 OBE f' V/ N s -+ W = 3 4 6. 7 + 2 91. 9 = 6 3 8. 6 ggg ]> , --\\/S: 13 '1 F OBE = 24 ksi AISC (C on t. Lateral b Support) F SSE = 0.95 Fy = 34.2 ksi b 5'-4" Span V (0.25) ( 42. 6) ( 5. 33) + 1. 3 (13 5) = 203. 9 lbs OBB = max Ss V (O. 5 5) ( 4 2. 6) ( 5. 3 3) + 1. 7 (13 5) = 2 91. 9 1 bs = max h (432.7) (8) 1. OB E M = 3 4 61. 6 l bs. f t. = fb* = 4. 2 ksi 4. Fb = 24 ksi (OK) x .91 f (638.6) (8) 2 2. SSE M= = 510 8. 8 f t. l bs, f = C.2 ksi < 34.2 ksi (OK) b" 0x 9.91 Check WT4x10 Section i s _4 l'lr" v>tL 3 & C'CFACl*lS t* EL.*l'3'o o 'a. 'C v A4 'g C,,, { ',e / -d t-- . / T [Locs e.L L p' 'L.0 . ' / /, /

• z4 4-(

. }l, ' l 7

d.a 1

C , e-

• FMr. CF v ~{.

~ j y_ j 1 E L W 2 ) Emmoco R l unoro stoo[ }_ ?. - g, m. g dl' check Shear on WT4x10 Web A=L r*

• web = 7.66x12x0.248 = 22.8 is 9 of 18

1 1. OBE k/ft 8, x02 14.4 ksi (OK) 0.08 ksi <. F = f 2 = = y 2 in y 9 2. SSE k/ft 8, x 0 = 0.13 ksi < F = 1. 6 x 14. 4 = 2 3 k s i M f = y 2 8 in y Check Shear Stress on Reduced Wall Area Due to WP Insert 2 = 12. 7 psi < 34 psi (O K) 1. OBE V max /2( 6 n) 2 = 19.3 psi < 57 psi (O K) 2. SSE Vmax " /2 36 n) Check Weld Capacity 4-1 1/2" Long 1/4" Welds Per Plate k 1. OB E P=8' x 0.2288k/1 = 1. 83k < 0. 707x 21x1/4x 2x 4x1. 5 = 4 4. 5 (OK) k/1 k 71.2 (OK) 2. SSE P=8' x 0.3467 = 2.77 < l.6 x 44.5 = Wall No. 5A-134 Floor El. 401'-0" Drawing No. A 312, S-812 Orientation: Nor t h-S out h "tr-- gr. g * >ln}s' 84

1. t W Esi3 wBsl3 p"

7 '. q " , (- N l l s v }f all Properti es 2 t, = 1'-7 5/8" (Solid) A = 19.625x12 235.5 in /f t = L, = 9 ' - 2 " (12)(19.625)2 3 770.3 in /f t g, = 6 h, = 8 ' - 6 " 3 Density = 145 pcf ,(12)(19.625) 4 y = 7558.4 in /f t 12 f'

1800 psi (14 5 pcf) (19. 6 2 5 in ) /12 i n/f t

1350 psi W f = = W 237.1 ps9 m 6 1.35 x 10 psi E, = 1000 f' = l l 10 of 18 - ~ _ _ - _~ ___m

gnR@sseurw a Consider o l'-0" Thick Hori zontal Beam S trip f rom the 7'-4" Simply Supported Span and the 11" Cantilever Span I E = 4 Simply Supported Wall g b ww 4 = 90.7 cps T = 0.011 Seconds ss " k4237 1) 3 3). i f=h I Mhd M 144 w I1* l 58. cantilever

• 37 0gt 4

= 2070. 8 cps, T = 0.00048 secs Both f and f are greater than 33 cps theref ore both are in the 2igid 28SO "gH' values are determined f rom the response 8 spectra of the base of the wall, EL. 401'-0". Peak values are used for attachment loads. 4% Damping (OB E) 7% Damping (SSE) i Allowable Stresses Rigid Peak F gv 1 Spectra No: 106-OB-EW 0.28 1.85 78 psi 34 psi 2 Spectra No: 106-SS-EW 0.24 1.2 130 psi 57 psi Since "gH" values for OBE are larger than gH SSE and OBE has smaller allowable stresses, only OBE need be considered. Simply Supported Span P -- P lbs For Solid k W. P = 180 ~ j { f t. ht. Masonry Walls W = 257.1 psf D* c y br, r$- L,= 7.33 ft. U N ( 7. ( 1. 8 5) M = 0.26 + = 10 56.1 f t-lbs. max ( I I 16.5 psi < 78 psi (OK) b, " 77b f = V = 0. 2 8 23 7._ p + (1.85) (180) = 576. 3 lbs max 2 h h = 2.45 psi < 34 psi (OK) f = l y 1 11 of 18

- Enclosure 3 Cantilevered Span M,,x = ( 0. 2 8 ) (237.1) (1/2) (0.9167) 2 = 27.9 f ts. Ibs. I ~ (237.1) ;(0.9167) '(0.28) = 60.9 lbs. g., n - V = max E' w. m.: fb" 7673 = 0.43 psi < 78 psi (OK) j 3 3 11 60 0.3 psi < 34 psi (OK) f = = 3 5 Check Masonry Steel Columns W8x13 Sx " 9*91 I" 4 ontinuous ., F' - i '-- Fb (OBE) = 24 ksi Lateral Support lbs. W = 576.3 + 60.9 = 637.2 = ft 4 (6 37. 2) (8~. 5) 2 (1/8) 5754.7 ft-lb@ M = = max (12) = 7 ksi < 24 ksi (OK) b, (5754.7 f 12 of 18

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lloris.cntel Floor Rerpence Snectra 2 u.. n i East '.!cnt Co::ponent (0.il.D.) El. l.01'-0" ~ Fuel Handling Bu11 dint; 17 of 18 ~,

.m-m C L Le.,,,,.. <.. - -..... n. ..s i 'Id : ' M. ' . Enclosure 3 a ih '* 9. PROJECT (2 fb(.'(2 l:**I i :'. E 'U).N Gir 8ad ~ DATE ~' ~> ~' 7 d.- 933.c3 gy iIIt< C!rG.:IP.3 CHECKED BY DATEj. -11.2.2 SHEET.:.~. C. !) O (...' '[, a a v. N o. I i l 2 l l 1. I j Daft I?-7-7d.I N 3 l'! I su TsAts l t-I ' l k S> l l I I ~ frepexy. CP5 y 3) 20 10 10 11 10 10 5 -, r : 3, ;,.... ..6 iii. 15.0 in i 4 '- le.o i - 3 i i i i 8 i 4.. to i i i I l I i.L.Ll l 6e2?s D!d PII:C ~ u 4355 D!d~PI::G t' l l' f 4 ?s D!dIPII:G i,S-e n.=...n-..a _ : /t .a m to ,4 p Da..P..,.ti.s n t s s,. 10 s g .iI U 2.0 l iIII k .'f i ~ [ ~~~ ' N.M\\-'T 7 ~ ~3 1 ** ~'i~ ~ t i 1 6 l 6: iTT\\! I. i '.{'. t / -- p Mey, .;u u 3 I-

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a -g p BYRON /BRAIDWOOD STRUCTURAL' DESIGN AUDIT (SUPPLEMENTAL INI'ORMATION) OCTOBER,.1981 i Action Item 27: URC Question No. 130.53-(Spent Fuel Pool Racks) The NRC requested comparison of the method used for seismic analysis of the spent fuel storage racks versus the SRP 3.8.4 method using spectral velocity. Enclosed is a letter from NUS Corporation dated December 18, 1981 providing the information requested. The method used by NUS is more con-servative than the SRP method. 4 g e O e e G W t a

u t A NUS CORPORATION E 0 7 MAW NO 2OO-c 5106/5107-NUS-208 December 18, 1981 s Mr. M. Amin Sargent & Lundy Engineers .,c 55 East Aionroe Street Cnicago, Illinois 60603

Subject:

New and Spent Fuel Storage Racks Byron /Braidwood Units 1 & 2 P.O. Nos. 205996 and 205997 S&L Spec. F/L-2743

References:

1. Appendix D to SRP 3.8.4, " Technical Position on Spent Fuel Pool Racks", Rev. O, July 1981

2. NUS Internal Correspondence EMD-HJE-187

Dear Mr. Amin:

Per your request, the following information is provided for your use in responding to NRC on their question 130.53. This question basically asked if the racks were designed in accordance with appendix D to SRP3.8.4 (Reference 1). Our response was that the only exception of note concerns the basis for calculating the maximum velocity of the fuel assembly for use in the fuel-can impact analysis. Our procedure, and the reasons for not following Reference 1, were given in Reference 2. The maximum velocity of the fuel during the SSE as calculated by NUS is 12in/sec. This velocity is equal to the SRSS of the maximum floor velocity and the maximum velocity of the fuel rack with respect to the floor. The NRC position is that "the maximum velocity of the fuel assembly may be estimated to be the spectral velocity associated with the natural frequency of the submerged fuel assembly". The natural frequency of the submerged fuel assembly is a function of the support, or restraint, afforded the assembly. In our design, as well as the typical design in the industry, the fuel assembly is restrained laterally by frictional forces at the bottom and by the walls of the guide tube, or can, along the upper portion of the assembly. During the postulated seismic event, this upper support changes from no restraint during the time the assembly moves from one side of the can to the other, and the length and location of contact of the support changes after initial contact as the . assembly deforms. Since the assembly support is a function of time, the natural frequency of the assembly is also a function of time. The fundamental frequency of a submerged PWR fuel assembly with simply supported end conditions is on the order of 3 Hz. For the case where the assembly is suppo::ed (pinned) only at the bottom, its first nonzero frequency is approximately 5 Hz. For the'ase of maximum lateral support, i.e., when the upper one-third of the fuel assembly is in contact with the can, the frequency is approximately 9 Hz. The table below gives the spectral velocities corresponding to these frequencies for the appropriate SSE floor response spectrum at 4a6 damping. s

I s

• e 5106/5107-NUS-208 December 18,1981 Page Two Frequency (Hz) Velocity (in/sec) 3 9.6 5

8.1 9 6.8 The velocity, as calculated using the above procedure, varies from about 6 to 10 ~ in/sec., depending upon the conditions assumed. The value of 12 in/sec. used by NUS is therefore greater than that which would be calculated by using the procedure of Reference 1. Very truly yours, fc ~ ... y Howard 3. Eckert, Jr. Manager, Engineering l Mechanics Depa.tment /ac cc: P. D. Arrowsmith B. 3. Reckman

3. L. Renchan S. B. Gerges G. Antonucci, Jr.

DCC File f 4 s NUS CORPORATION

e f- '<- .1 4 1 d .~ s s .i ~- 2' BYRON /BRAIDWOOD ~. STRUCTURAL DESIGN AUDIT (SUPPLEMENTAL IhTOEMATION) OCTOBER, 1981 l } s s 1 Action Item 29:' Buried Piping 2 S&L to pro' vide clarification of the' calculation's showing the 1: j effects of a joint displacement at'the 900 bend _of the Essential Service Water Pipo Concrete Encasement. Calculations should t also be provided for the effects of axial load in.the concrete pipo encasement. s Refer to the attached sheets; calculation number 2.3.3. BY, Revision 1, pages 45 and 51. f ~, I e i. i i l l ene l I f I I I 1 5

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ry 3 g BYRON /BRAIDWOOD STRUCTURAL DESIGN AUDIT (SUPPLEMENTAL INFORMATION) OCTOBER,-.1981 f Action-Item 30: Missile. Protection for Manhole Cover S&L to provide information for the manhole cover to comply with missile protection. ' Ductile iron with 100 ksi tensile strength is being provided for all Category I manhole covers. These manhole covers have been designed in accordance with SRP 3.5'.3 for tornado missilei effects.

• .n 1

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