ML19319C251
| ML19319C251 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Midland |
| Issue date: | 08/01/1969 |
| From: | TOLEDO EDISON CO. |
| To: | |
| References | |
| NUDOCS 8002110715 | |
| Download: ML19319C251 (200) | |
Text
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D-B TABLE OF CONTELTS VOLUME h Section M FORMAL QUESTIONS FROM DIVISION OF Corresponds To REACTOR LICENSING AND APPLICANTS' RESPONSES Question Number INFORMAL QUESTIONS FROM DIVISION OF Corresponds To _ REACTOR LICENSING AND APPLICANTS' RESPONSES Question Number IQ-1 Thru IQ-16 , DISCUSSION OF ITEMS CONTAINED IN ACES-1 Thru ACES-12 ACRS LETTER, DATED AUGUST 20, 1970 0002 ( Amendment No. 11
1 D-B (, , s' 1.0 General 1.1 Table h-10 of the PSAR established that the Code specified for the design and fabrication of the reactor coolant piping and valves is USASI B31.7 including Errata up to June 1968. However, the inter-faces between seismic Class I piping fstems are not identified in terms of which systems are designed to B31.7, Classes I, II, or III as applicable and which systems are designed to USASI B31.1. Provide this additional information in a format similar tc that which was provided as drawing h.1-1 of Amendment 5 to the Midland application, Docket Nos. 50-329 and 50-330.
RESPONSE
Figures 1.1-1 and 1.1-2 identify the ANSI B31.7 piping classification and class interfaces for the primary coolant piping and valves. None of the piping and valves are designed to ANSI B31.1.
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I D-B i 1.2 In section 1.6 of the PSAR the asterisked items of areas of con-cern indicated in Advisory Committee on Reactor Safeguards letters are discussed for pressurized water reactors. Provide an updating of these matters of concern as referenced in ACRS letters on pressurized water reactors through January 1970, and discuss how they vill be considered in the Davis-Besse plant design.
RESPONSE
The following information updates PSAR Section 1.6 including the addition
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of new areas of concern expressed by the ACRS for FWRs since the PSAR was submitted. Item 1 - Fuel Clad Failure During a LOCA and Its Effect on ECCS Performance - The ACRS has requested further evidence, both analytical and experimental, that bulging and/or perforation of the fuel clsdding following a LOCA vill not affect the heat trsnsfer capability of ECCS to the extent that clad melting can occur. New Information: The experimental work described in 1.5.3 is complete and the analysis is in pregress. A topical report (BAW-10009) on this subject vill be submitted to DRL during the second quarter of 1970. Preliminary results indicate that tb ECCS will effectively cool the core, even if substantial fuel rod swelling occurs. Item 2 - Partial Melting of Fuel Assembly During Normal Operation - The ACRS desires further inc omation that the celting and subsequent disintegration of a portion of a rue, rod by inlet coolant orifice blockage or by other means vill not lead to ur.#cceptable conditions in ter=s of fission product release, local high pressure productiori, and possible initiation of failure in adjacent fuel elements. New Information: BAW-100lk, " Analysis of Sustained Departure from Nucleate Boilinc Operation" was submittad to DRL on August 8, 1969. No comments have been received en the report other than questions associated with the Oconee Nuclear Station appli-cation. No additional work, beyond answering the previously mentioned questions, is being done on this subject. Item 3 - Ability of Fuel to Withstand Expected EOL Transients - The ACRS recommends fuel burnup tests at linear heat generation rates higher than those calculated for the worst anticipated transient and to burnups comparable to the maximum expected. New Information: No additional information at this time. 9 , 000s 1
~ l 1.2-1 l
D-B Item h - Primary System Quality Assurance and In-Service Inspection - The ACRS vants every precaution possible taken to assure that a high quality primary system is fabricated and they further desire that critical areas of the primary system be inspectable during service life. New Information: See answer to Question h.6 for additional informa: ion. Item 5 - Effects of Thermal Shock on Pressure Vessel Integrity - The ACRS is concerned that the thermal shock resulting from actuation of the ECCS following a LOCA vill cause a flaw in the vessel to propagate through the vessel thickness. If this happened, the coolent vould be lost and the core vould be uncovered. New Information: BAW-10018, " Analysis of Structural Integrity of the Reactor Vessel Subjected to Thermal Shock", has been reviewed by DRL on another application. No additional information is available at this time. Item 6 - Effects of Blevdown Forces on Core and Internals - The ACRS desires that calculational mode.s be developed and used to analyze in detail the effects on core and internals of the blevdcvn forces experienced by those components during a LOCA. New Infor=ation: BAW-10008, Part 2, " Fuel Assembly Stress and Deflection Analysis Due to Loss-of-Coolant Accident and Seismic Excitation", was issued on October 21, 1969. BAW-10008, Part 1 and Part 2 have been reviewed by DRL on another application. These reports are being revised to include additional information and will be reissued by the end of the second quarter,1970. Item 7 - Separation of Control and Protection System Instrumentation - The Cemmittee believes that control and protection instrumentation should be separated to the fullest extent practicable. Interconnection between control and protection instrumentation should be eliminated or reduced to a mini =um. The applicant should review the design for common failure modes , taking into account the possibility of systematic, non-random, concurrent failures of redundant devices , not considered in the single failure criterion. New Information:
- B&W has completed an evaluation of the control and protection system instru-mentation in the current product line to assess its vulnerability to commen ,
l mode failures not considered in the single failure' criterion. Analytical studies of the following accidents or abnormal transients have been =ade and presented to DRL:
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0009 1.2-2
- D-B
- 1. A rod withdrawal from power accident
- 2. A startup accident
- 3. A loss-of-coolant flow accident (caused by loss of power to 4 pumps) 4 A r.ad ejection accident
- 5. A rod drop accident
- 6. Secondary side load changes (caused by turbine trip, loss of feedwater flow and high feedvater flow)
- 7. A loss-of-coolant accident (caused by a primary coolant system rupture)
In all the analyses but that of the LOCA, the common mode failure was assumed to render the primary reactor trip signal inoperative. In the LOCA analysis ~ the common mode failure was assumed to render the primary safeguards actuation signal inoperative. In addition, B&W has agreed at the request of the DRL staff to undertake an additional study of the LOCA where trip is required to determine the consequences of the failure to trip. The analyses to date indicate that the B&W design will adequately accommodate systematic failures and no design changes are being made at this time. The results of this work are to be reported in topical report BA'110019 which is to be submitted during the second quarter of 1970. Item 8 - Instrumentation for Prompt Detection of Grosu Fuel Failure - The ACRS desires that consideration be given to the development of instrumentation to detect gross fuel element failure. This instrumentation should be capable of rapidly detecting fuel failure in the presence of fission products already in the coolant due to " leakage" through the clad and other nor:sily expected sources and to scram cn the signal. New Information: The B&W scoping study report was transmitted to Toledo Edison Company in August,1969 Item 9 (New) "The Committee believes the applicant should evaluate all problems which may arise from hydrogen generation, including various levels of Zircalcy-water reactions which could occur if the effectiveness of the emergency core cooling system were significantly less than that predicted." < Information:
'See the answer to Question 12.6.9.
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D-B Item 10 (New) "The Committee recommends that the applicant study ,;ossible meanc of in-service monitoring for vibration or for the presence os loose parts in the reacter pressure vessel as well as in other portions of the primary system, and implement such means as are found practical and appro-priate." Information: See the answer to Question 3.2. _ Item 11 (Nev) "The Committee recommends that a study be made of the possible consequences of hypothesized failures of protective systems during anticipated transients, and of steps to be taken if needed." Information: At the request of the DRL staff, B&W has undertaken studies to examine the ~ consequences of certain reactor transients if reactor trip does not occur. Transients examined vere:
- 1. A rod withdrawal accident
- 2. A startup accident
- 3. A loss of one reactor coolant pump
- k. A stuck rod accident
- 5. A rod drop accident
- 6. A turbine trip
- 7. A loss cf feedvater flov
- 8. A loss of off-site power This work is essentially complete and vill be presented to DRL soon.
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0011-A 1.2 h
D-B
,e s 1.3 List those systems which contain an interface between Class I and a lower class or a transfer of responsibility between the nuclear steam system supplier and A&E. Indicate the location of the interfaces and discuss the manner in which they are considered in the design.
RESPONSE
A. Systems which contain an interface between seismic Class I and Class II design are listed belov vith the appropriate figure number which delineates the location of these interfaces: Reactor Coolant System Figure 1.3-1 Emergency Injection Safety Features Figure 1.3-2 Contairanent Spray System . Figure 1.3-3 Makeup and Purification System Figure 1.3-k - Decay Heat Removal System Figure 1.3-5 Component Cooling Water System Figure 1.3-6 . Service Cooling Water System Figure 1.3-T Auxiliary Feedwater System Figure 1.3-8 Sampling System Figure 1.3-9 Steam and Power Conversion System Figure 1.3-10 Where interfaces exist, the boundary is indicated by the letters I & II at the stop or a remotely operated valve which is designed to the higher seismic classification, i.e., Seismic Class I. s B. Listed below are the systems which are the design responsibility of the NSS supplier (B&W). For many of these systems, the major equipment is furnished by B&W, while piping and valves are specified by Bechtel. B&W provides design requirements for equipment which is specified by Bechtel in these systems.
- 1. Reactor Coolant System
- a. B&W is supplying the reactor and pressurizer vessels, heat exchangers, pumps, pressurizer safety and relief valves, pressurizer spray control valve, the main reactor coolant (28" and 36") piping, pressurizer surge and spray piping and all process instrumentation and controls,
- b. Bechtel is specifying the pressurizer quench tank.
- 2. Makeup and Purification System
- a. B&W is supplying all system vessels, heat exchangers, pumps, <
power operated valves and process instrumentation,
- b. Bechtel is specifying all system piping and manual valves.
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D-B
- 3. Decay Heat Removal and Core Flooding Systems
- a. B&W is supplying all system vessels, heat exchangers, pumps, power operated valves and process instrumentation.
- b. Bechtel is specifying all system piping and manual valves,
- h. Chemical Addition System 8 a. B&W is supplying all system vessels (except for the two boric acid addition tanks), pumps, and process instrumentation and controls.
- b. Bechtel is specifying all system piping and manual valves.
5 Fuel Handling System
- a. B&W is supplying the containment and fuel pool fuel handling bridges, the fuel transfer tubes, the fuel transfer mechanism, controls, and various equipment storage racks and handling tools.
- b. Bechtel is specifying other equipment which is used for refueling.
- 6. Containment Spray System Bechtel is specifying the BWST, spray pumps, all system piping and valves, spray no::les and process instrumentation.
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Interfaces in the reactor coolant system are at no::les where Bechtel specified piping connects with B&W supplied vessels and piping and/or supports. Within the remaining systems, interfaces are at the supporta and no::les of pumps, vessels, and heat exchangers. The allevable forces and moments are determined for these equipment interfaces through the coordinated efforts by B&W, the vendors, and Bechtel. The piping is thus designed so that the loads imposed by.it vill not exceed those which have been determined to be the maximum allowable. In addition, total forces on the support (s) are determined, considering seismic loadings, to assure that the support (s) vill accept the loads imposed. I C. There are a few other systems for which Bechtel has design and specification responsibility, but for which some furetions and equipment are defined j and/or furnished by B&W. i
- 1. Main Steam System l
B&W is supplying the main steam safety and turbine bypass valves. l l
- 2. Main Feedvater Jystem
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4 B&W is supplying the main and start-up feedvater control valves, and ' the Integrated Control System (ICS) which provides signals to these ~ ) valves.
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D-B 1.k Provide details of the intake and discharge canal structure and indicate which sections of these canals vill be Class I. Include the design criteria and a discussion of the capability to achieve and maintain safe shutdown of the facility under all conditions in the event access to Lake Erie is not available due to failure of the canals.
RESPONSE
This question is discussed in detail and complete in the PSAR Amendments No. 2 and 3 Section 5 9.8. Also, several sketches and a discussion on the soil stability are given in response to the question 5 10. Fig. 5 10-1, 2, 3, h show the plan and sections of the Intake Canal Structure and the limit of Class I structure. Fig. 5 10-2 also indicates the volume and surface. area of the available water in different water surface elevation in the event of canal failure. I i 1 0034 B , 1.k-1 l
D-B 1.5 With respect to the discussfon of brittle fracture control for ferritic steels in compliance with General Design Criterion 35 published on July 10, 1967, discuss the extent to which your design criteria conform to the follewing statements:
- a. Those pipes with wall thicknesses less than 1/2 inch need not have material property tests (such as Charpy V-notch) if (1) they are austenitic stainless steel, or (2) the ferritic material is normalized (heat treated), or (3) the ferritic material has been fabricated to " fine-grain practice."
- b. Pipes with wall thicknesses greater than 1/2 inch must have nilductility transition temperature 60 F below anticipated temperature when the system has a potential for being loaded to above 20% of the design pressure.
Ferritic material with an NDTT of -20 F or austenitic stainless steel will also fulfill the requirements. ,
RESPONSE
.a. All pipes with a vall thickness less than 1/2 inch are stainless steel. Therefore, the design criteria conforms to 1.5 (a) (1).
- b. The Davis-Besse Technical Specification vill contain a
' requirement that, at no time during heatup and cooldown of the reactor coolant system during the plant life, will any component (including piping) be pressurized above a pressure equal to 20% of the design pressure of the reactor coolant system while the system is below DTT (NDTT + 600F) .
$ 0035 1.5-1
/
D-B ,, 1.6 Previous development efforts (such as those under FVRC auspices) have been devoted to the determination of stress distributions and effective elastic constants of perforated plates under thermal inplane and transverse loadings. Although a major application of this theory is to the analysis of the tube sheets in heat exchangers, these efforts, to date, have included no serious attempt to determine the stress state and effective elastic constants in an actual structure, consisting of tubes svaged and velded into the perforated plate. The ASME Section III stress analysis procedure is based upon the perforated plate method described above. In light of this background, discuss the analytical treatment of the tube-tube sheet complex (as an integral structure) of the steam generators for the Davis-Besse Station.
RESPONSE
The " Analysis of Tubesheet on Elastic Foundation for Once-Through Steam ~ Generator" is proprietary. The analytical treatment of the problem is described in the following paragraphs. The once-through steam generator has been analyzed by considering the secondary shell, tubesheets, tubes and primary heads as a single integral structure. The purpose of the analysis was to determine the structural response of the tube-tube sheet complex by considering the tubesheet resting on an elastic foundation provided by the straight tubes. The results of the analysis were redundant forces and moments which were used to find stresses and displacements in the tubes, tubesheet, secondary shell and primary head for various steady state conditions. The analysis of the tubesheet on an elastic foundation was performed by writing a series of equations of continuity for each juncture of the vessel geometric configuration at or near the tubesheet-head-shell intersection. These equations considered continuity of radial displacement and rotation at the geometric juncture. An additional vertical continuity equation was written to account for the vertical force effect on redundants resulting from interaction between the tubesheet, supported on tubes, and the steam generator shell. Influence coefficients for elements of the geometric con-figuration were found for unit values of shear forces, bending moments and vertical forces. The tubesheet influence coefficients were found by con-sidering the tubesheet as a circular perforated plate resting on an elastic foundation. The deviation was based on work by M. Hetenyi in Chapter 5, Article 30, of " Beams on Elastic Foundation",1st Edition, 5th Printing, University of Michigan Press, Ann Arbor, Michigan. The foundation modulus was derived by associating the load intensity relation P = KY from Article 1, Page 2, of " Beams on Elastic Foundation" to the load intensity produced by ; deflection on a long tube. The classical equation of axial strain was used. J Equations of continuity of displacement were vritten for each juncture of the configuration. 0036 1.6-1
D-B Boundary conditions resulting from primary pressure, seccndary pressure, and thermal expansion effects were used to solve the system of equations for unknown redundants. The system of equations were solved by a digital computer program. The redundants were used to calculate stresses in the conventional manner. Tubesheet displacement was found by applying redundant loads to the elastically supported plate. Final tube displacement was de' ermined by superposing pressure, thermal, and tubesheet motions of the tubes and tubesheet. Tube loads and stresses vere then calculated from the final displacement. The results of the analysis of the tube-tubesheet complex (as an integral structure) compare closely to results obtained from the tubesheet analysis in use by E&W for tubesheets without elastic foundation type support. All comparisons were made for steady-state conditions. Tube-sheet stresses were calculated using ASME Section III procedures. The conventional tubesheet analysis gave conservative results and was used for . vessel design purposes.
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0037 J 1.6 Y k
D-B f l.7 Proposed amendment to 10 CFR Part 50 published November 25, 1969 would require that piping majk fittings within the reactor coolant pressure boundary =ees the requirements for Class I piping of USASI B31 7 dated F . ary 1968 and Errata dated June 1968, in-cluding +he require, nts of Appendix IX - Quality Control and Non-destructive Examination lethods, of the 1968 edition of ASMF Code, Section III as nandatory supplement to B31.7. Compare the quality contro' ;uirements specified in Section h.5 of the PSAR vith the requ. 4eents of Appendix IX. Specify all significant differences between the two sets of requirements for the reactor coolant piping.
RESPONSE
There are no significant differences between quality control requirements in Section h.5 of the Davis-Besse PSAR and Appendix IX of ASME Section III. There are also no significant differences between quality control requirements in Appendix B of ANSI B31.7 and ASME Section III Appendix IX. The Babcock & Wilcox quality control program used in fabricatin6 piping vill meet requirements of both ANSI B31.7 and ASME Sectien III Appendix IX. l.7-1 0038
/
D-B 1.8 Will any Class I components, in whole or in part, be designed and/or fabricated in a foreign country? If so, which components vill be fabricated by whom? ,
RESPONSE
- The following Class I Components vill be designed or fabricated in a foreign country.
Some Remote Operated Valves Velan Montreal, Canada High Pressure Injection Pu=p B&W Ltd. Galt, Ontario Decay Heat Pump B&W Ltd. Galt, Ontario
- There is no other equipment that is known to be or expected to be designed or fabricated in a foreign country; however many Class I components are not purchased as yet.
/
/
j e 0039 M 1.8-1 < /
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19 To which edition of the ASME Code Section III and Addenda vill applicable Class I co=ponents be designed and fabricated?
RESPONSE
Class I co=ponents of the Davis-Besse systems will be designed and fabricated in accordance with the applicable section of the Boiler and Pressure Vessel Code of the ASME Section III, including Su==er 1968 Addenda. As detailed design and fabrication of these components proceed, later addenda and/or interpretations of the ASME Code may be envoked to ensure a quality product considering tt7 safety functions being performed. O l 4 00d0 1
)
N 1 9-1 _ a
D-B 1.10 Appendix 2A of the PSAR discusses the type of activities which are expected to be carried out in the restricted areas adjacent
\-
to the site. Provide a discussion and analysis of the margin of safety to be provided in the facility's structural design to protect vital equipment and components against possible missiles generated from the restricted areas. The analysis should include all areas which require missile protection to assure safe shutdown of the facility is not jeopardized and no uncontrolled release of , radioactivity will result. EESPONSE As stated in Appendix 2A, the activities carried out in the restricted areas adjacent to the site do not significantly affect the safety of the Davis-Besse Station. However, the design of the station, to provide protection against missiles resulting from tornadoes provides more than adequate protection to vital components and systems from missiles generated from the restricted areas in the unlikely event this should occur. , The most penetrating missile resulting from the activities related to the use of the restricted areas, is the 25mm armor piercing projectile which is capable of penetrating 16.9 inches of reinforced concrete. The minimum thickness of any wall or roof structure to provide tornado protection to systems within the auxiliary building is 18 inches while the minimum thick-ness of the shield building dome is 24 inches. l 4 o l l 0041 l i 1.10-1 s. L .
D-B 2.0 Site 2.1 In order that we may determine the appropriate atmospheric diffusion factors for the design basis accident analysis, provide seasonal vindroses for each atmospheric stability condition, which include vind directions and speeds at a 20-foot height and describe the ranges of te=perature differences between the 5-foot and 145-foot elevations which were used to determine the various stability conditions. Describe the location of the meteorological tower and the heights and distances of objects that =ay affect the diffusion conditions.
RESPONSE
An on site meteorological program at the Davis-Besse site was conmitted for in mid-1968 and was installed and operating in September of 1%8. The initial instru=entation arrange =ent included wind speed sensors at the 20-foot and 300-foot level. These locations were chosen because the type of , reactor had not been determined at that the and it was anticipated that the 20-foot elevation data would be appropriate for a PWR installation with the 300-foot elevation data appropriate for a BWR installation. During the preliminary design and PSAR preparation period, the containment concept as presented in the PFAR was decided on. This concept provides for a shield building sur ounding the containment building with the space between being ventilated by fans during IOCA and this air including any leakage from the containment vessel being passed through appropriate filters and dis-charged through the station vent at an elevation 10 feet above the highest point in the shield building or 246 feet above ground elevation. With this containment arrangement it was our opinion that the appropriate atmospheric diffusion factors for the design basis accident analysis should be those supported by the meteorology data recorded at the 100-foot height. On this premise, the 20-foot vind sensors were relocated to the 100-foot elevation on June 11, 1969 Since there was no 100-foot recorded data available for the original sub-mission of the PSAR, estimated 100-foot data was given which was derived from the 300-foot data. Following a meeting with representatives from the Division of Reactor Licensing on November 7, 1969, in which an interest was expressed in 20-foot data, an additional recorder was installed on December 3, 1969 at this elevation. It is still our opinion that the appropriate atmospheric diffusion factors for the design basis accident analysis should be those supported by the meteorology data recorded at the 100-fcot height; however, in response to this request the 20-foot data is attached for the periods: J Fall (October and November) 1968 Spring (March, April, May) 1969 l Winter (December, January, February) 1969-1970 I s 2.1-1
D-B The te=perature criteria for each stability class are indicated in degrees C per 100 meters on each windrose. The equivalent English units for 140 ft. difference on the tower am: Moderately Cable - Iapse rate exceeds 1.2 F/140 ft. Slightly Stable - Lapse rate between -0.4 and 1.2 F/140 ft. Neutral - Lapse rate between -1.2 and -0.4 F/140 ft. Unstable - Lapse rate less than -1.2 F/140 ft. The tower is located approximately 000 feet southwest of the location of the containment building in an open field. The nearest obstructions are a residential building and trees located approx 1=ately 500 feet NNE of the tower. These are shown on Figure 2.1-1, which is an aerial view looking due north. The tower location can also be identified as being in the center of a the roadway circle shown on Figure 2-3 in the PSAR. The 100-foot reconied data is included in Amendment No. 3 to Appendix 23 of the PSAR. 0043
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IDLEDO d4TA FALLIUCiutEd,NOVt46ERI 19f. d LLVtl 2., LAPSE NATE LESS THAN -1.5C PER lac NETEd5 UNSTABLE . .._ . JPEED IN N.P.H. i~ DIRtCTION 1-3 4- F S-12 13-10 19-24 25-31 32-38 39-50 51-75 76-100 TJT Al. . NME 0.J 3.70 2.31 J. 3 1.39 _ c.4 . . s.O 0. C- 0.C C.) 7.61 NE 0.0 0.0 C.O 0.C J.3 0.0 u.0 L.0 0.( t.) G.9 ENE 0.0 4.17 C.93 1. 0 0.0 C.0 3.0 _ 0.C (.( i.i 5.~9 E C.46 1.85 (' . 4 6 1.c ?.D . _ t, . q _ _ . it . Q _.__ 3.0 C . O. a. . ) 2.78. ESE 0.0 0.93 0.C 0.e 0.0 v.0 Q.0 . 0.0. C.C 0.0 0.93 SE O.0 0.46 C . C- J.0 0.0 . peu _ 0.0 _ 0.u C.C L.0 7.46 SSE ,_ 0.0 Og3 (.C s.O .R,Q.__.__.0.C u.0 _.___ C.0... __ Q.D _ u.9 D 91 . n) 5 0.0 0.0 4.63 3.24 J.46 C.0 0.u _ 0.t 0.0 0.6 d.33 SSW 0.0 2.31 3.70 3. U _ _. L.31 . . . __.121? Q,u _ __. 0.C. 0.0 .u.u 13.43 __ Fd
* ' SW . C . 0_ _ 3.24 4.1F 2 31__ _ . J.24._ _ 1,}9 h u_ _ . '. " . _ _ . C.c r.a 14 31 WSW 0 46 , 0.93 1.39 6. 44 2.JI 0,0 .
Q,9 _G.C C . (- 0.0 12 04 , W C.0 0.0 3.24 4.63 ..._ _l.mL ____Q,.44_ _ _ 0,0 -. c . c_ . 0.0 0.0 1..19 WNW. . . _ _ .
- 0. 46 _,, . 0. 0 , _ , 2.31 _ 6 38 0.0 0.0 ..
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,,,0.0 0.46 2.78 4.17 0.0 c.L O.0 u.0 0.0 0.0 T.41 NNW _,
0.3 ,0.46 1.39 0.0 __ 0.0 0.0 _ 0.0 . 3.t (.C (.4 1.85 N 0.0 1.35 1.39 0.0 . _ _ 0.93 .. _C., G J.G q.0 C.C C.s 4 17 VR8L C.0 0.0 C.C 1.C O.0 0.C 1.0 C . ;' C.( , t, . 0 C.; TOTAL 1.39 21 30 2e.70 31.48 12.50 3.24 , 0.0 c.L u.e C.0 PERCENT CALN= 1.39 NUMH;R (J NISSING OJ5tRVATIONS. O I.llAL NUNoER f.F V AL I C DhSE RV Al lut.S = 216 C C C C -
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SPFtD IN M.P.H. DIRECTION 1-3 4-7 ft- 12 11-18 8 i-24 25-31 .32-38 39-50 51-75 T6-103 TOTAL NNE G,26 .l.01 1. f 2 3. 6 0.. c.6L Q.0 0.0 0.6 0.0 C.J 4.99 NE 0.0 0.26 v.17 1.46 d . 0 '4 .C . c 2.9 0.C *( 0.7
.. 1.98 .
ENE C.0 1.29 c.34 0.34 u.O 0,9 9.D . 0.0 0.0 C.0 1.98 E G.77 1.72 1. 2 Q . 1.52 a.6 _O d . 0,0 C.0 0.0 0.0
. .s.22 ESE C.52 0.95 0.17 J . -) 0.Q 0.e u. 0. . C.L . C.0 0.0 1 64 SE 1.c3 1.29 c.09 0.0 0.0 0.0 _j.0 C.c r.0 0. ') 2.41 .
SSE 2.OT 3.10 1,12 0.0_ u s.E .9 0 h C.____ .7, . .- 0.C . 0.R . __.6 11_ _ . 80 s 1.55 5.34 3.27 1.64 0,34 ._ 0,0. _0.0. .. .60 0.0 e.0 12 13._. P e 55W 0.60 2,24 3.27 1,55 t,12.__.__D,11 to . . _ _ _Q.0 .Q.9..__._ Q.0 . 0. 0 .. . . L 12__ .. O SW 1,12 . 2 58 .4 22 2 93.. .1.55_ . ___QaAD .. 9.17 _ 0.c __ C.0 0.0 13.12 WSW 1.38 2.93 3.27 5.59 2.41 0,26 _ 40 0.0 _. 0.0 0.0 15.83 W 0.34 1.98 2.58
. 2.50 .l.03 Q.09_____2 0 J.0 0.0 0.0 0.52.
WNt! . 0. 3.4 0.69 . 1,q3_ _ 2,5Q. ._ _ Dip? . .. . OR . __ _ D D 0. '; ._ 0.0 ') . O 4.56 NW 0.09 0 14 2.24 3 27 n.43 0,0 .0,0 0.0 0.0 0.0 6.37 NNW 0.0 0.26 1.38 0.43 0.86 0.0 0.0 0.0 .00 0.0 2.93 N _ . , 0,0 0.52 .O.86 Q.52 J , 9}. .Q.Q _.._y,0..__. 0.9. C.0 8. 0 2.44 VRBL b.C 0.0 0.0 1. 0 0.0 C.0 . 0.0 ,, 0... G.0 0.0 0.0
*iOTAL 10.07 27.2d 26.94 23.84 9.47 1,29 J.17 0.Q ... b.C U.0 PERCE NT l'.ALN= 0.95 NUMBER OF MIS $1NG Da$EkVATIONS= 2 TOIAL NUM8EP a1F V AL10 Ot 52RVAT IOh5= 1162 N
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006Z : 2.1-21
D-B 2.2 We are reviewing your anal,ysis of the high water protection level for the plant, and thus far we have determined that the following information is needed.
- a. An analysis of the maximum probable water level at the plant site rather than the once-in-100-years water level. The analysis should include static (e.g., seiche) and dynamic
-(vave) water levels.
- b. A hydrological or mathematical model study which determines the greatest increase in the maximum probable water level that could be caused by the convergence of water via the gap between the intake and discharge canal.
- c. An analysis of the greatest probable effect of a Toussaint River flood on the maximum probable water level at the plant site.
- d. An analysis of the maximum probable flood levels of the Toussaint River at the plant site.
RESPONSE
- a. An analysis of the maximum probable static and dynamic water levels at the station's site is given in the PSAR Section 2.4.1.2 " Lake Levels" as revised by Amendment No. 8. l8
- b. Lue to application of Cooling Tower System, the discharge canal is no longer 8 required.
- c. An analysis of the greatest probable effect that a Toussaint River flood could have on the maximum probable flood level is given in the PSAR Section 2.h.2 " CHARACTERISTICS OF STREAMS" as revised by Amendment No. 8.
l8
- d. An analysis o'f the maximum probable flood levels of the Toussaint River is given the PSAR Section 2.h.2 " CHARACTERISTICS OF STREAMS" as revised by Amendment No. 8.
l8 - M 0063 2.2-1
D-B
~
2.3 Describe the systems which will assure that radioactive liquid waste will be diluted in circulating water discharge, or confirm that the radioactive liquid waste _ released into the discharge system will not exceed the limits of 10 CRF 20 for unrestricted l8 areas. Your description should include the following information:
- a. The instruments and equipment that will prevent the dis-charge af radioactive liquid-vaste while circulating water is not being discharged.
- b. The minimum circulating water discharge rate (gpm) that will permit the discharge of radioactive liquid waste.
- c. The maximum discharge rate (gpm) of radioactive liquid waste.
- d. A drawing showing how the radioactive liquid waste will be mixed with circulating water discharge. :
RESPONSE
A station discharge water system utilizing service water and cooling tower blowdown will be used for radioactive liquid waste disposal. 8 Samples are taken and analyzed of all waste before it is released from the station. If these show that the specific activity is sufficiently low, the liquids are passed from the radwaste system to the station discharge system. On the way out, they must pass at least one radiation monitor which 8 controls two (2) downstream isolation valves in series. These valves are set to close if excessive radiation is detected. The minimum station discharge water flow rate will be approximately 20,000 gpm normally provided by 8 the two operating service water pumps. The maximum discharge rate for waste will be determined by eli,her the physical capacity of the system (lh0 gpm) or by the need for dilution to meet 10 CFR 20 limits. Processing through the radwaste system should reduce all isotopes, ex-cept tritium to concentrations far below those allowed for release to unre ' 5 stricted areas. For these, no dilution is needed. In the case of tritium, however, processing has no effect. Therefore, if it builds up to a point where credit for dilution is needed, the maximum discharge rate of the waste will be regulated accordingly. For preliminary calculational purposes, about 80% mixing is assumed in the discharge for water and waste. A design for accomplishing this has not been developed yet, l6 e c 0064 2.3-1 O
D-B ~ 2.h Provide an analysis of the maximum average and instantaneous concentrations of radionuclides in Lake Erie at the discharge canal and at potable water intakes that could result from normal operations or an instantaneous rupture of a tank. The analysis should include the following information and all factors , assumptions , and references used to obtain the information.
- a. A list of the radionuclides and the maximum quantity of each that would be released to Lake Erie in a year of normal operations.
- b. A list of all tanks that would contain radioactive liquids and the maximum volume and concentrations of radionuclides that would flow into the lake in the event of the rupture of any tank.
- c. The maximum average and instantaneous concentrations of radionuclides in Lake Erie at the discharge canal and potable water intakes.
- d. An explanation of how the temperature, volume, velocity, and direction of the released water were considered in determining the lake dilution factors.
RESPONSE
- a. In order to compute the maximum quantities of radionuclides that would be released to Lake Erie in a year of normal operations, it is necessary to know what concentrations of radionuclides are normally discharged into Lake Erie. For nor=al operations , the maximum concentrations of radio-nuclidesinLakeErieatthedischargesystemwillbewellbelow10CFR20l8 limits.
The estimates for normal concentrations, with the exception of tritium, are based on primary reactor coolant which has been processed through the clean radwaste system. The initial fission product activities in this coolant v12L be the maximum values attained in an equilibrium cycle with 1% failed f uel. A ' list of these is available in Section 11 of the PSAR. The corrosi on product concentrations used were derived from operating plant data (see question 11.6). The efficiency of the radwaste system is based on the following assu=p-tions:
- 1. Purification Demineralizers Decontamination factors of 1 for Cs , Kr, Xe, Mo, and Y and 100 for all other isotopes (ionic) have been used. These were supplied by BW and are the same as those used in calculating the list of specific activities for .he primary coolant mentioned above. In addition, a decontamination factor of 1 has been chosen for insoluble corrosion products even though demineralizers make good filters.
. R 0065 2.h-1
D-B
- 2. Radvaste Mixed Bed Demineralizers :
A decontamination factor of 103 is used for both cations and anions. This is probably very conservative, sineg calculations and literature surveys have indicated that values of 10 or 105 could be expected (seequestiog11.6). Also$ experiments have produced values of at least 6 x 10 and 1.4 x 10 for cesium and iodine, respectively. Again, a decontamination factor of 1 is used for insoluble corrosion
-products.
- 3. Evaporators A decontamination factor of 106 for soluble and insoluble nonvolatile isotopic species has been given in commercial literature (i.e. AMF Model 750 Ray-Di-Pak liquid waste concentration system) . Conservatively, no credit is taken for degasification.
- h. Degasifier .
A decontamination factor of 105 for gaseous activity is taken from manufacturer's literature. 5 Filters A decontamination factor of 10 is assumed for insoluble corrosion products and 1 for dissolved rpecies. No credit is taken for concentration reduction through the dilution of
'8 l vaste in'the discharge system.
In passing through the letdown coolers, the primary coolant is assumed to go from an operating temperature of 600 F to 120 F. This results in a 1.457 decrease in liquid volume and a corresponding increase in impurity concentrati ns. The primary coolant is let down through coolers and the purification demineralizers into the radwaste system. It is then passed successively. through a degasifier, a filter, a mixed bed demineralizers, and an evaporator. A polishing demineralizer will follow the evaporator, but for this conservative analysis, no credit is taken for it. Also, no credit is taken for a final filter. The cycle just described has the following total decontamination factors: Kr and Xe . . . . . .. . . . . . . . . . . . . . 6.86 x 100 Cs, Mo, and Y . . . . . . . . . . . . . . . . . 6.86 x 107 Cr, Mn, Co, and Fe (insol, corrosion products). 6.86 x 105
- All others . . . .. . . . . . . . . . . . . . . .6.86 x 109 8 -l The concentrations of radionuclides in the lake at the discharge system
, are found by dividing the above total decontamination factors into the specific activities of the primary coolant. These values are listed by isotope in Table 2.h-1. 006G 2.h-2 g
D-B
- ' It is estimated that a maximum of about 150,000 ft3 of vaste are processed during a complete cycle. Normally, a large portion of this would be reused; but for this analysis, all of it will be assumed to be released from the station. With this assumption and the specific activities of the radionuclidea in the processed waste that have already been calculated the total isotopic releases to the lake are determined. These are also presented in Table 2.h-1.
Tritium is treated separately since, at present, there is no feasible way of removing it from the vaste. The only way of reducing its concentration is by dilution. This is done in the station by mixing the vaste in the 8 discharge system before releasing it to the lake. The mixing ratio will be controlled to ensure that the concentration in Lake Erie at the dis-charge point.will be below the 10 CFR 20 limit of 3 x 10-3 ue/ml. Based [8 on the conservative assumption of 30% leakage through the fuel cladding, about 3000 curies of tritium vill be released per cycle.
- b. The following is a list of all major tanks that may contain radioactive liquids:
- 1. Pressurizer Quench Tank
- 2. Makeup Tank '
3 Boric Acid Addition Tank (2)
- h. Borated Water Storage Tank 5 Primary Water Storage Tank
- 6. Reactor Coolant Drain Tank 7 Clean Waste Receiver Tank (2)
- 8. Concentrate Storage Tank 9 Clean Waste Monitor Tank (2)
- 10. Miscellaneous Waste Drain Tank
- 11. Detergent Waste Drain Tank
- 12. Evaporator Storage Tank 13 Miscellaneous Liquid Waste Monitor Tank All of these, with the exceptions of the borated and primary water storage tanks , are housed in Class I structures which vill contain any liquid released by a rupture. The two exceptions are located in the open and are vulnerable to tornadoes and, in the case of the primary water storage tank, earthquakes. The only radionuclide of any significance in either of these should be the tritium which has baen picked up through contact with the primary system. If a tritium level of 6 pc/ml is maintained in the primary system (@ 600 F) it is estimated that the upper limits in the borated and primary water storage tanks would be about h pc/ml and 9 pc/ml respectively. Assuming the possibility of only a single tank rupturing at one time, the worst accident for this analysis would be that involving the primary water storage tank which has the higher activity level. In <
this case the contents of the tank are assumed to go down yard drains into the discharge system and then to the lake at a rate itsited by the lg size of the drainage pipe (12 ft3/sec.). Taking no credit for dilution by the discharge system, this vould result in a maximus accident tritium 6 g concentration in Lake Erig at the discharge point of 9 u c/ml and a dis- a charge rate of 3.058 x lob uc/sec.
~
0067 2.h-3
D-B
- c. The maximum average-and instantaneous concentrations of radionuclides in f 8l . Lake Erie at the station discharge that could result from normal operations or an instantaneous rupture of a tank are shown in responses (a) and (b)
, above. 8l The nearest public potable water intakes to the station discharge serve Camp Perry, the Erie Industrial Park, and surrounding residences and are located approximately 2.8 miles (h.5 km) away. Table 2.h-2 shows that, at this distance, a continuous point source (unit /sec) dilution factor of 6.3 x 10-9 (unit /cm3 )~can be expected with regards to station effluent. Based on the normal maximum discharge rate of vaste from the station of 140 gpm (2.33 gps); thic represents an effective dilution factor of 5.565 x 10-5 which can be applied directly to discharge concentrations. Using this with any type of normal effluent release that has been postu-lated gives radionuclide c: ncentrations at the potable water intakes 6 that' are negligible. - For the case of the accidental rupture of the primary water storage tank where its contents are assumed released to the lake at a rate of 12 ft 3 /sec, an effective dilution factor of 2.1h x 10-3 must be used. Appling this to the 9ue/ml tritium concentration in the tank gives a level at the potable water intake of 1.93 x 10-2 yc/31
- d. The following derivations and calculations were done by Dr. C. K. Huang of the Scripps Institution of Oceanography.
Most mathematical models describing the distribution of conservative material in a plume emanating from a continuous fixed source in the atmosphere or ocean are based on the assumptions that the turbulent field is homogeneous and stationary. The theoretical steady-plume models are deduced from the super-position of an infinite number of patch distributions in the presence of a mean current. If the flow field has a detectable mean velocity the diffusion in the direction of the current can be ignored. Furthermore, if the material distribution within any individual disk-element in the plume is assumed Gaussian, which is in general approximately the case, then the concentration at any point in a plume is estimated by Gifford's (1959) two-dimensional model. In the lake, the mean concentration at any point downstream from the continuous point source is given by
~
- C (x, y, z) = -+
77(d'd,* y )"* O i. 71
- 14* * (1) where x, y, z are coordinates, x is in the direction of mean current, y is horizontal and perpendicular to the current direction, z is vertical; Q is the steady rate of discharge of conservative material from a point source in units /sec; 8#b are the coordinate variences of the material distribution in em ;2 U is the mean current speed in em/sec. Note that the above diffusion model is anisotropic, s a _ m 2.h-h
D-B s The peak concentration on the surface of the lake is q
-max C
(*) *g24]g 8 (2) In a stationary homo 6eneous turbulent field, after a long period of time the diffusivity is considered to approach asymptotically a constant. Csanady (196h) and Okubo and Farlow (1967) studied the turbulent diffusion in the West Basin of Lake Erie and have shown the effective lateral eddy diffusivity is about 10 cm/see to 6 x 10 cm/see and the vertical eddy diffusivity is about 1 - 10 cm/sec. Knowing the mean velocity of the current and the longitudional distance from the source, the mean coordinate variances can be estimated from
-2 2 Kx 0 =
3 .(3) where K is the diffusivity. During the summer of 1968, we ran patches of Rhodamine B dye near Locust Point in Lake Erie. At the same time the mean currents were measured by surface drogues. The peak concentrations of the dye patch as a function of time (or distance) were recorded from the fluorometer readings. The mean concentration distribution across the patch is approximately Gaussian. As we are more interested in the concentration distribution of the conservation material in the effluent under the worst conditions, that is diffusion under da along-shore slow current, the lowest observed mean current about 10 cm/see along the lake shore is used in this study. The lower limit of coordinate variances for the continuous point source are taken from the variances calculated by equation (3) of the dye patch study with a lower limit value of diffusivity. Equivalently the concentrations predicted by equation (1) using the dye patch variances are the upper limit of the material concentration distributions. Conservatively we are using the following data for the calculation of the point source concentraticn distributions: Q = 1 unit /sec U = 10 cm/sec Ky = 103 em2 /sec Kz = 1 cm2 /see Then from equation (3), the variances are
}
= 2 x 102 X,
= 0.2 X.
The concentrations along the beach (maximum corJ.) and 100 m. away from the beach for each successive 1 Km downstream are listed in Table 2.h-2. 7- m W 006S ~ 2.h-5 }
D-B In treating the large scale diffusion phenomena, such as in this case with a large volume of discharged effluents from the power plant, it is . more realistic to use the two-dimensional volume' source model. In the volume source equation the variances at the origin is an essential parameter in describing the concentration distributions. Since we have no similar survey to estimate the original variances of the volume source effluent, we cannot but use the point source equation which results in higher concentration distributions than the volume source (Foxworthy, et. al. 1966). Note that the point source equation is not valid at the origin. S 0070 2.h-6
f i D-B REFERENCES
- 1. Csanady, G. T., Turbulence and Diffusion in the Great Lakes, Publ.
No. 11, Great Lakes Research Division, University of Michigan, 326 (1964). l
- 2. Foxworthy, J. E., R. B. Tibby, and G. M. Barsom, Dispersion of a i Surface Waste Field in the Sea, Journal of Water Pol'.ution Control Federation, Vol. 38, No. 7, 1170 (1966).
- 3. Gifford, F., Statistical Properties of a Fluctuating Plume Dispersion Model, Adv. Geophys., y, 117 (1959).
- h. Okubo, A. and J. S. Tarlow, Analysis of Some Great Lakes Drogue Studies, Proc. 10th Conf. on Great Lakes Research, 299, (1967).
S / N- 00?1 2.4-7 6 l
D-B 2.4' TABLE 2oh-1 8l RADIONUCLIDE LIQUID CONCENTRATIONS AND RELEASES TO THE LAKE ISOTOPE MAXIMUM CONCENTRATION NORMAL CONCENTRATION NORMAL ANNUAL . g IN PRIMARY LOOP AT STATION DIS- RELEASE FRCM (uc/ml) CHARGE STATION (ue/ml) Ci Insoluble Corrosion Products Cr-51 .1h 2.00 x 10-I 8.51 x 10-
-0 Mn-5h .016 2.30 x 10 9.78 x 10 -5 Co-58 .8h 1.22 x 10- 5 19 x 10-3 Co-60 .0045 6.6 x 10-9 2.81 x 10-5 Fe-59 .016 2.3 x 10-8 9.78 x 10-5 ,
, Fission Products (gaseous)
Kr-85m 17 2.h8 x 10-5* 1.05 x 10-1
-1 Kr-85 11.9 1 73 x 10 h* 7.36 x 10 Kr-87 9h 1.37 x 10-5* 5.83 x 10-2 Kr-88 3.00 h.37 x 10-5* 1.86 x 10'-l Xe-131m 2 70 3 93 x 10 -5* 1.67 x 10-1
-1 8 Xe-133m 3.10 h.52 x 10-5* 1 92 x 10 Xe-133 280. h.08 x 10-3* 17.35 1.53 x 10 -5*
-2 Xe-135m 1.05 6.51 x 10 Xe-135 6.70 9 76 x 10-5* h.15 x lo-8.3 x 10 -6 *
-2 Xe-138 57 3.53 x 10
- These values assume that Kr and Xe remain in solution in water after pro-cessing through the degasifier with a conservative decontamination factor of 6.86 x 104.
Fission Products (solid-ionic) Rb-88 3.00 4.37 x 10-10 1.86 x 10-6 3 Sr-89 .0h5 6.6 x 10-12 2.81 x 10 Sr-90 .0041 6.0 x 10 -13 + 2.55 x 10-9 -. Sr-91 .052 7.6 x 10-12 3.23 x 10-0 Sr-92 .019 2.8 x 10 -12 1,19 x yg-8 0072 6l 2.h-6
D-B O. 2.k Y-90 1.05 1.53 x 10-8 6.51 x 10-5 Y-91 .25 3.6 x 10-9 1 53 x 10 -5 Mo-99 6.00 8.7h x 10-8 3 72 x 10-0 l8 I-131 3.60 5.25 x 10-10 2.23 x 10-6
-6 I-132 5.h 7.87 x 10-10 3.35 x 10 I-133 h.2 6.12 x 10-10 2.60 x 10-6 i I-134 56 8.20 x 10-11 3.h9 x 10-I I-135 2.2 3.21 x 10-11 1.37 x 10-0 .
-0 cs-13h 5 7.29 x 10 3.10 x 10-5.57 x 10 -5 4
cs-136 9 1.31 x 10
-3 cs-137 53 7 72 x lo -I 3.28 x 10
-5 cs-138 .82 1.2 x 10-8 5,1o x 10 Ba-137m h7 6.85 x 10-9 2.91 x 10-5 Ba-139 .091 1.30 x 10-11 5.53 x 10-8
-11 Ba-lho ,
.073 1.10 x 10 h.68 x 10-8 La-140 .024 3 50 x 10-12 1,gg x 10-8 ce-1kh .0031 4.50 x 10-13 1.91 x 10 -9 a
g
, g 0073 I 6 2.h-9 L:
D-B 2.h TABLE 2.h-2 Surface concentration distribution along the beach and 100 meters away from the_ beach in the downstream direction from a unit /sec continuous point Source. Conc. 100 m. away from Distance, X in Km Conc. along beach the beach 1/10 2.5 x 10-7 3.5 x 10-17 2.1 x 10 -0 1 2.5 x 10 2 1.3 x 10-0 3.6 x 10 -9 3 8.4 x 10 -9 3.6 x 10-9 h 6.3 x 10-9 3.h x 10-9 5 5.0 x 10-9 3.1 x 10
-9 6 h.2 x 10 -9 2.8 x 10-9 7 3.6 x 10 -9 2 5 x 10 -9 8 3.1 x 10 -9 2.3 x 10 -9 9 2.8 x 10-9 2.1 x 10
-9 10 2.5 x 10-9 2.0 x 10 -9
. . m )
6l 2.h-10
D-B
-~
3.0 Reactor Design
~
3.1 Section 3.1.2.h.2 of the PSAR, which delineates the stress and strcin limits for fuel assemblies under normal and abnormal operating conditions, does not sufficiently define those limits nor the manner and extent to which the cited limits provide an assured margin of :afety against failure under these loadings. Provide the following additional information.
- a. Confirm that the type of atresses referred to in Paragraph (a) are in the " primary" category as defined in Article h of ASME Code, Section III. Describe the basis for estab-lishing 75% of the r tress rupture life of the material as a numerical limit and indicate whether that limit is constructed upon the average stress or the minimum stress to produce rupture at the end of 100,000 hours.
- b. Clarify whether the type of stresses referred to in :
Paragraph (b) are in the secondary category in the same context as above. Where stresses exceed yield, are they calculated on an equivalent elsstic basis, i.e. , pseudo-elastic basis as in Section III? Identify the source of the fatigue curves used for each material of concern, e.g., Article 4, Section III. Where fatigue data are employed which are not included in any codes or standards, specify whether a basic data or design curve
! is used or a design curve which incorporates design /
correction factors, e.g. , 2 on stress, 20 on cycles and correction for maximum effect of mean stress. The state-ment that strain limits will be set using no more than 90% of the material (s) fatigue life implies that you may use less. Clarify this statement and in addition, outline the exact procedure (s) used in setting the strain limit (s). Specify the number and type of cycles that have been established for design purposes and indicate the margin of safety that exists over the expected number and type of operational cycles to be experienced.
- c. For the combination of stresses in (a) and (b), above, specify the stress limit (s) that apply, e.g., 3 Sm rS.t
- d. What are the stress and/or deformation limits which are specified for the fuel assemblies under emergency and faulted conditions, i.e., normal plus pipe rupture loads, normal plus maximum earthquake loads, and the simultaneous occurence of these loads? .
y
. . 0075 3.1-1
D-B
RESPONSE
Inasmuch as the operating conditions, materials, lifetime, and safety requirements for fuel assemblies are significantly different from pressure l vessels, and in view of the lack of a fuel assembly design code comparable ' to that of Section III of the ASME pressure vessel code, the stress limit criteria as given in Section 3.1.2.4.2 have been established within B&W for use until such time as an industry accepted code for fuel assemblies is available. As indicated in the above questions, there are several useful definitions of stress terms and operating conditions which have been developed in the pressure vessel code since the criteria in B&W's safety reports were written. In response to DRL's request to clarify our criteria, the following information is submitted. A. Stress Limits for Normal and Urset Conditions 9 The design of the fuel assemblies shall be such that stress and/or strain shall not exceed the following limits for normal and upset operating conditions. Tems relating to stress analysis which are defined in Article N hl2 of Section III of the ASME Pressure Vessel Code, including addenda for Su=mer 1968 are used herein, unless other definitions. are given. Ceramic fuel materials are specifically excluded from these limits, except as they influence cladding stress. N
- 1) Primary stresses shall not exceed yield strength of the material, nor )
shall the design life exceed 75%* of the stress rupture life of the ' caterial. The stress rupture life is based on the minimum stress to produce rupture. An e ample of primary stress is the membrane stress in the cladding due t a pressure differential.
- 2) Combinations of primary and secondary stress may exceed the yield strength. Tne limits on these stresses (strains) are as follows:
a) The average tensile strain shall not exceed 1% (e.g., the average diameter shall r.ot exceed 1.01 times the initial diameter). b) The equivalent elastic stress _ce to cyclic loading shall be limited suct that the accumulated fatigue damage fraction shall not exceed 0.9 (90% of the design fatigue life) . The design fatigue life shall include a correction for mean strain equivalent to the maximum accumulated strain calculated. The accumulated fatigue damage fraction shall be determined by su" ing the fraction of the number of applied cycles to the t ilovable number of cycles for each specified strain (or equivalent stress) range. _ Examples of secondary stress include thermal stress, discontinuity stress at end caps, and cladding strain which is induced by and/or is limited by the fuel.
- The value 75% of the rupture life provides a positive margin to failure for the case where stress rupture may occur at stress below yield strength, and ,
ductility is less than 1%. {JQg 3.1-2 ne
~
D-B 3\
- 3) Minimum clad collapse pressure margins vill be required as follows:
a) 10 percent margin over system design pressure, on short-time collapse, at end void, b) End void must not collapse (must be either freestanding or have adequate support) on a long-time basis , c) 10 percent margin over system operating pressure, on short-time collapse, at hot spot average temperatare through the clad vall. d) Clad must be freestanding at design pressure on a short-time basis at the local hot spot average temperature through the clad wall. B. Stress and Deformation Limits for E=ereency and Faulted Conditions The effect of emergency and faulted conditions on fuel assemblies are evaluated for other B&W plants in Topical Report BAW-10008, Part 2. Similar fuel asse=bly analysis vill be done for Davis-Besse which includes the effect of differences in the structure, (e.g., pressure vessel supports and vent valve deletion). The design of the fuel assemblies shall be such that deformations shall not exceed the following limits for emergency and faulted conditions.
- 1) Spacer Grid Limits Deformation of the grids shall be limited such that the periphery is not perLanently deformed more than 0.15 inches. This limit, based on results of crush tests, is selected to assure that control rod insertion into the fuel assembly guide tubes vill not be impeded.
l
- 2) Fuel Rod Deformation Limits The deformation of fuel rods shall be limited such that the emergency coolant has adequate access to cool the fuel rods such that gross i fuel relocation is prevented. !
C. Cladding Fatigue Analysis j Variable stress in the cladding is caused by pressure changes, the?bal ' stress, and strain induced by the fuel. The minimum value from fatigue data for irradiated _or unirradiated Zircaloy (Ref. a) is used in evaluating l the fatigue life of the cladding. The lower bound of the data with a correction for effect of mean strain, is given in Figure 3.1-1. Evaluation W of the endurance limit has shown that small variations in stress due to pressure variations during steady state operation contribute negligible fatigue damagg to the cladding. Since Figure 3.1-1 includes a correction for 1% mean strain, strain due to pressure and fuel swelling may be omitted - Ref. (a) " Fatigue Design Basis for Zircaloy Components", W. J. O'Donnell and B. F. Langer, Nuclear Caience and Engineering 20, 1-12 (196h). k 0077 3mR-3
D-B . from the analysis. Variable stro :s significant to a fatigue study includes effects of gross system pressure changes (e.g., reactor shutdown), thermal stress, and clad cycling imposed by fuel thermal cycling associated with changes in power. In the fatigue analysis it is assumed that the maximum value of this stress is superimposed on the thermal stress for each power cycle. For system depressurization cycles,the hoop stress due to system pressure is also superimposed, since this represents a departure fran the mean stress. A factor of two on stress is included for stress concen-tration effects at surface scratches. The design cycles used in the fatigue analysis are given in Table 3.1-1. This is the upper limit of cycles expected in service. These cycles apply to a ad following reactor, and are therefore conservative for base loaded re ; ors. The fatigue life fraction calculated for fuel cladding is 0.85. The margin of safety for the vorst case rod as cc= pared to minimum fatigue data is therefore 18%. J
. )
3.1 h u_
D-L 4 TABLE ,3.1 DESIGN THERMAL TRANSIENTS FOR 3 YEAR CCRE LIFE Transient Power Design Number Transient Description Variations Cycles 1 Heatup, 70 F to 532 F 0 8-0% 18 Cooldown 5320F to 700F 2 Startup, 0-15% power 0-15% 18 Shutdown,15-0% power 15-0% 3 Plant Loading and 8-100 8% 75 Unloading 15-50-15% 188 50-100-50% 1,125 h 10% Step Increase and f;10% 3,600 Decrease 5 Step Lead Reduction 100-8% 12 6 Turbine Trip to Hot Standby 100-3% 12 7 Reactor Trip 100-3% 23 8 Hydrostatic Tests 0 -- 9 Steady State Fluctuations f; 1% 2.6 x 10' 10 Rapid Depressurization 100-3% 6 11 Change of Flow 100-3% 1 50-3% 1 20-3% 1 12 Complete Loss of 100-3% 6 Feedwater Flow 13 Co=plete Loss of 100-3% 3 Station Power lh Loss of Feedwater to One Steam Generator (a) Pressurized 100-50% 2 (b) Depressurized 100-50% 1 15 Loss of Feedwater + 2% 3 Heater 16 Drop of One Control Rod 100 60% 3 17 Control Rod Withdrawal 0-1125 3 Accident 18 Main Steam Failure 100-3% 1 c)(ry3 3.1-5
106 i i iiiis i i i iiiis i i i iisai i i iiisis 1 105 - a _ d - 3 E
=C m
= 104 m _
, , , , , , , , i i i , , , , , i i i ii.. i , , ,,,i, i 103 102 103 104 105 106
' O Cycles to Failure, N e DESIGN CURVE FOR 1RRA0 LATED AND UNIRRADIATED CD ZlRCALOY 2 AND 4 - ROOM TEMPERATURE 10 6000 F C FIGURE 3.1-1
D-B
' Discuss the status of the development of possible means for in-3.2 service monitoring for vibrations or detection of loose parts in the reactor pressure vessel and other parts of the primary system.
RESPONSE
The feasibility of in-service monitoring for vibration and the detection of loose parts is being explored by B&W vith a vibration consultant. The application of such sensors as accelerometers, strain gages and load cells to monitor vibration of internals, and inertia 11y loaded force pickups to monitor for loose parts has been discussed with the consultant. Additional discussion with the consultant and instrumentation vendors is planned. No formal development program has been decided upon at this time. . l l l 0061 3.2-1 J e
D-B ~s 3.3 Provide a discussion and analysis to show that the failure of the control rod assembly pressure housing vill not cause failure of adjacent control rod assemblies. Indicate the minimum control rod assembly pressure housing rupture size which would result in ejection of the control rod.
RESPONSE
The control rod drive assembly housings are designed to the same design criteria as is the reactor pressure vessel. Accordingly, the roller nut drive housings comply with Sectica III of the ASME Boiler and Pressure Vessel Code under the subsection for Class A vessels. The operating transient cycles, which are considered for the stress analysis of the reactor pressure vessel, are also considered in the housing designs. Quality standards relative to material selection, fabrication, sad inspection are specified to insure safety function of the housings essential to accident prevention. Materials conform to ASTM or ASME, Section II, Material Specifi-cations. All velding shall be performed by personnel qualified under ASNE Code, Section IX, Welding Qualifications. These design and fabrication pro-cedures establish quality assurance of the assemblies to contain the reactor coolant safely at operating temperature and pressure. For vibratory and seismic loadings , the assemblies are restrained with a series of contoured plates that are bolted to the main support structure. These plates are contoured to restrain the upper outside diameters of the roller nut housings. The main support structure is bolted to the reactor closure head. The con-toured plates vill provide lateral support on-j and vertical motion of the housings resulting from thermal expansion will not be restricted. In the highly unlikely event that a pressure barrier cc=ponent of the centrol red drive assembly s'hould fail catastrophically, i .e. , a complete rupture, the following results would ensue:
- a. Control Rod Drive Nozzle For the failure of this component, the assembly would be ejected upward as a missile until it was stopped by the containment vessel missile shield.
This upward motion would have no adverse effect on adjacent assemblies,
- b. Roller Nut Drive Housing, Motor Section, Dissimilar Metal Welds (2)
The failure of this component anywhere above the lower flange would result in a missile-type ejection into the missile shielding of the contain=ent vessel. There would be no adverse effect on adjacent =echanisms. The control rod drive assembly has been analyzed for jet reaction loading due to rupture of the control rod drive housing. The jet force was con-sidered to be a dynamic load calculated by multiplying the initial housing pressure times the ares. The crack area was assumed to be the same as that of the flow area from the reactor vessel head into the centrol rod assembly housing. All stress intensities due to this condition plus normal pressure loads were below allowable at all locations . The maximum deflection under this loading will not allow the ruptured assembly to contact or damage other control rod drive assemblies. The load and deflec-3.3-1 0082
D-B I l l I tion of any assemblies hit by the jet will be no more than that on
+he ruptured assembly,
- c. Roller Nut Drive Housing Welded Closure Cap The failure of this component anywhere above the lover flange would result in a missile-type ejection into the missile shielding of the containment vessel. There would be no adverse effect on adjacent mechanisms.
The rod ejection time was determined assuming a co=plete and instantaneous s,everance of the control rod assembly pressure housing. This assumption leads to the greatest possible pressure differential acting on the control rod and hence the shortest possi'.le ejection time. For an assumption of a rupture si::e less than complete severance, the pressure differential would - be less and the rod ejection time would be longer. A longer rod ejection time would lead to less severe results than those presently calculated for the rod ejection accident. c
. 0083 r
3.3-2 N
D-B 3.4 Provide a discussion and any new analyses which have been completed to justify the increase in the core design power from 2h52 to 2633 megawatts thermal. Include any experi= ental information which supports the power level increase.
RESPONSE
The increase in power capability is a result of changing from the " canned" (E-A) fuel assembly to the "canless" (MK-B) fuel assembly. The hot channel of the MK-A' assembly is the corner cell which has the " canned" vall on two sides. Coolant flow between assemblies is not available for heat transfer because of the physical boundaries separating the assemblies. The E-B assembly is substantially different and superior to the MK-A assembly. The MK-B fuel rod diameter is 0.430 inches (0.h20 inches for MK-A) wh17h results in a larger heat transfer area. The coolant flow between the MK-B assemblies is available for heat transfer since no physical boundary separates the assemblies. The absence of the " canned" vall also results in the hot channel ~ being a unit cell. The increase in power capability is best shown in terms of 'he minimum DNB ratio for the typical types of cells. The folleving table is a chronological presee.ation of this information. FUEL DNB CELL DNBR ASSEMBLY POWER, MWt CORRELATION UNIT CORNER WALL MK-A 11h% x 2h52 W-3 (with 1.h6 1.3h 1.38 unheated vall effect) MK-B 11h% x 2h52 W-3 1.71 1.81 1.78 114% x 2568 W-3 1.5" 1.66 1.65 112% x 2633 W-3 1.50 1.63 1.61 The minimum DNBR (based on W-3 correlation) increased from 1.34 (corner cell) for the MK-A aasembly to 1.71 (unit cell) for the MK-B assembly at 11h% of 2452 MWt. This is a substantial i=provement in power capability. The minimum DNBR for the Davis-Besse core design at 1.12 x 2633 MWt is well above the minimum design limit of 1.3. The attached Figur, 3.h-1 shows the DNBR versus % power for the unit cell. The minimum design 13 *: of 1.3 occurs at 118.7% of the nominal power level. The second poi t, shown on this figure is the minimum DNBR of 1.50 for the design overpower of 112%. 3.h-1 S 00S4
l 2.0 l l l s i l 1 l 1.8 - l 1 l
- 1.6 _
7 m 1.50
=~ l
- l 1.4 -
= 1 1.30 I e _______.;_____.
l l 1.2 - ! l l l l 1 l xl e zi al 1.0 - :l l l l l l I
' II I
0.8 100 110 120 130
~
Power (% 2633 MWt) MINIMUM CORE DNBR VERSUS POWER 00S5 g Figure 3.4-1 l .
D-B 3.5 In reference to the discussions of the reactor internals in Section 3 of the PSAR and the criteria of Appendix 5A, list the ec=penents for which buckling is a possible mode of failure when censidering the case for the ec=bined concurrent design basis earthquake and the postulated less-of-coolant accident. For the lT most critical items, i.e. , these closest to failure, state the analytical method and give the margin between the condition censidered to ecnstitute failure and the as-calculated condition for the combined lor. dings. Relate this margin to the degree of uncertainty involved in the loads used to perform this analysis. The uncertainties of the selection of the basic seismic ground motion need not be discussed in this context.
RESPONSE
Buckling vill be evaluated as a possible mode of failure for the internals shells, the upper control rod guide tubes, and the lover grid support colu=ns. The internals shells would be subjected to external pressure during the postulated less-of-coolant accident. The collapse pressure for each shell vill be calculated based on the theory given in Section IIT of the ASME Code. Additional details on the theory may be found in Precess Eouipment Desien, by Lloyd E. Brownell and Edwin H. Young. The upper guide tubes and lower grid support columns vould be subjected to axial IDCA and earthquake loads. Their critical buckling loads will be determined using standard column equations. The allcwable external pressure or co=pressive load will be limited to 85% of the collapse pressure or critical buckling 1 cad, respectively, i e e W 0086 3.5-1 Amendment No. 7
D-3 3.6 Identify the extent, method and findings of the analyses of thermal stresses which would result in the core barrel and core support structure in the event of loss of coolant and subsequent operation of e=ergency core cooling equipment.
RESPONSE
In the event of a LOCA, the reactor is shut down by void formation, control rod insertion, and boron injection. During the time between blevdcun initiation and emergency coolant injection, the resident coolant and its steam forsation and flow will prevent a large temperature increase of the internals structural members. A comprehensive and extensive analysis is not required when the internals material is considered. The 30h SS material, unlike high carbon ferritic material, is insensitive to thermal cracking. The conventional carbide solutiou anneal for austenitic stainless steels is a full soak heat at 1850-2000 F follcwed by a quick cold water or air quench to prevent carbide precipit ation. This treatment is more severe than the transient expected following a LOCA. Also, austenitic stainless steels do not exhibit a signifi-cant loss of ductility or NDTT shift from radiation exposure; therefore, a brittle fracture type f ailure would not be credible for the magnitude of the anticipated gradient. Support lugs are provided in the pressure vessel to maintain the core in place in the event the core support structure does fail. Such a failure would not prevent the flow of emergency coolant to the core. The structural member of the core barrel assembly is effectively shielded by the ther=al shield such that there vill be negligible direct contact with cold emergency coolant. Thus severe thermal transients in this region is prevented. The region above the thermal shield h'as less than 5 x 1019 fast fluence during its h0-year life. This material therefore retains ample ductility to withstand the application of severe transient thermal stress. Mixing of resident coolant and the cold emergency coolant will result in a limitation of the severity of the thermal transient. The requirement to cool the core following a LOCA results in ample cooling for internals to prevent excessive thermal stress for emergency and faulted conditions. It is concluded that thermal stress in the core support structure during a LOCA transient vill not result in an unsafe condition which has not been considered in the design of the reactor. oos7 3.6-1
D-B
- 3.7 Describe the analytical and/or test procedure that will be used
( \ to ensure the functional integrity of the reactor internals in the event of a loss-of-coolant accident. Indicate the loading functions and mathematical models that will be used in the analysis.
RESPONSE
The ability of the reactor internals to retain their functienal integrity will be established using procedures empicyed previously for reactor internals of similar design. Topical Report BAW-10008, Part 1, entitled " Reactor Internals Stress & Deflection Due to Loss-of-Coolant Accident and Maximum Hypothetical Earthquake", outlines the methods of analysis which will be empicyed. The LOCA and seismic loadings and mathematical mocels used will be those for the Davis-Besse station. Changes in plant arrangement and elimination of internale vent valves vill result in' diffe"ent LOCA loadings than given in the Topical Report. The nozzle-supporte . setor vessel requires a new - dynamic model to determine seismic loadings. These changes may result in minor revisions to the internals design documented in the Topical Report. m W 0088 3 7-1
D-B l 3.8 The PSAR does not present sufficient information concerning the specific nondestructive examinations and inspections to be per-formed for the reactor internals. List all such examinations and inspections and identify the codes or standards which apply in each case. 3 ESPONSE The following list of all reactor internals nondestructive examinations and inspections with applicable coder or standards applies to all structural support material of various forms. In addition, one or more of these examinations is performed on ma*,erials or processes which are used for functions other than structural support (i.e., alignment dowels, etc.) so that virtually 100% of the completed internals materials and parts are included in this listing. Internals rav materials are purchased to ASME Code Section II or ASTM caterial specifications. Certified material test reports are obtained and retained to substantiate the caterial chemical and physical properties. All internals materials are purchased and obtained to a low cobalt limitation. The ASMS Code Section III, as applier.ble for Class "A" vessels, is generally specified as the requirement for reference level nondestructive examination and acceptance. In isolated instances when the ASME Code Section III can-not be applied, the appropriate ASTM Specifications for nondestructive testing are imposed. All velders performing veld operations on internals are qualified and certified in accordance with ASME Code Section IX,1968 Edition through the Su=mer 1968 Addendum. The primary purpose of the follcwing list of nondestructive tests is to locate, define, and determine the size of material defects to allow an evaluation decision for acceptance, rejection, or repair. Repaired defects are similarly inspected as required by applicable codes.
- 1. Ultrasonic Examination
- a. Wrought or forged raw material forms are 100% inspected throughout the entire material volume to ASME Code Section III, Class "A",
1968 Edition through the Su=mer 1968 Addendum,
- b. Personnel conducting these examinations are trained and certified.
- 2. Radiographic Examination (includes X-ray or radioactive sources)
- a. Cast rav =aterial forms are 100% inspected to ASME Code Section III Class "A" or appropriate ASTM requirements .
- b. All circumferential full penetration structural veld joints which support the core are 100% inspected to ASMS Code Section III s Class "A" requirements.
00SS 3.8-1
- c. All radiographs are evaluated by certified personnel who are trained in their interpretation.
- 3. Liquid Penetrant Examination
- a. Cast form raw material surfaces are 100% inspected to ASME Code Section III Class "A" or ASTM requirements.
- b. Full penetration, non-radiographic or partial penetration, structural velds are inspected by examination of root, and cover passes to ASME Code Section III Class "A" requirements,
- c. All circumferential, full penetration, structural veld joints which support the core have cover passes inspected to ASNE Code Section III Class "A" requirements. *
- d. Personnel conducting these examinations are trained and certified. -
- k. Visual (5X Magnification) Examinatien This examination is perfor=ed in accordance with and results accepted on the basis of a B&W Quality Control Specification which complies with NAV-SHIPS 250-1500-1. Each entire veld pass and adjacent base metal are inspected prior to the next pass, through, to, and including the cover passes.
- a. Partial penetration non-radiographically or non-ultrasonically -
feasible structural veld joints are 100% inspected to the above B&W quality control specification,
- b. Partial or full penetration attach =ent veld joints for non-structural materials or parts are 100% inspected to the above B&W quality control specification. .
- c. Partial or full penetration veld joints for attachment of =echanical devices which lock and retain structural fasteners are 100%
inspected to the above B&W quality control specification.
- d. Personnel conducting these examinations are trained and certified.
0030
-)
3.8-2 3E L
D-B f ~'s 3.9 The stress limits for the ioading combinations as given in Table 5A-2 require further clarification and more exact definition. The brief discussion of those limits in section h.l.2.h does not provide this clarification. We have identified several items of concern which are described as follows:
- a. Case II - Design Leads Plus Maximum Hypothetical Earth-quake Loads is defined as faulted condition and reference is made to paragraph N bl2(t)(h) of ASME Code, Section III. We consider this particular loading combination as an emergency rather than a faulted condition. There-fore, the Sm values for this leading case should conform to paragraph N kl7.10 of Section III instead of paragraph N kl7.11.
- b. An additional loading case alcng with corresponding stress limits is required for the occurrence of the loads due to pipe rupture (loss-of-coolant accident) plus normal design loads,
- c. One of the Case III loading combination stress limits utili::es ultimate strength curves published by U.S.
Steel which are adjusted to cdnimum ultimate strength values by using the ratio of ultimate strength given by Table N h21 of Section III at room temperature to the room temperature strength given by U.S. Steel. This ultimate strength ratio, which is calculated at room te=p-erature, cannot be verified as a conservative estimate of the actual ratio (or margin) at the temperature of con-cern unless a comparison to the value of the minimum code ultimate strength of each material at temperature is obtained. Discuss the above items in detail and provide the following specific information:
- d. Clarify whether the stresses to be ce= pared to 2/3 of the ultimate strength of material limit for loadirg Case III are calculated on an elastic basis , i.e. ,
pseudo-elastic stress calculation as in Section III. Pro-vide the elastic stresses corresponding to this limit for each of the materials of concern. Furnish the corresponding strain limits for each material.
- e. It is stated that the Case III design stress limits vill be based en either the 2/3 of ultimate strength criteria or paragraph N h17.11 of Section III. Discuss the basis J for using either set of limits addressing: (1) the reason for the choice of limits, (2) to what components or areas thereof each would apply and (3) the basis of l comparisen between the two limits including the margins I of safety that each provides.
& 0091 3.9-1 J
RESPONSE
- a. The basic criteria for the reactor vessel internals for loading
. Case II, no loss of functional integrity, assures that the reactor may be safely shutdown. This criteria and the conservative acceleration level of the maximum possible earthquake best fits the definition of a " faulted" condition according to the ASME code, and the higher stress levels are justified. From Section III, ASME Code, Paragraph N h12(t):
"(h) Faulted Conditions - Those combinations of conditions
, associated with extremely low probability postulated events whose consequences are such that the integrity and operability of the nuclear energy system may be impaired to the extent where con-
. siderations of public health and safety are involved. Such con-siderations require compliance with safety criteria as may be .
specified by jurisdictional authorities. Among the faulted - conditions may be a specified earthquake for which safe shutdown is required."
- b. An additional loading case for the occurrence of loads due to pipe rupture plus normal design would be defined as a " faulted" con-dition according to Section III of the ASME code. The same stress limits as loading Case III would then apply. But , since maximum stresses for each component due to earthquake and pipe rupture loads are added directly for Case III, the stresses for Case III would always equal or exceed those for the proposed additional loading case. This additional loading case is therefore not required to assure safe shutdown of the plant in the event of such an cccurrence.
- c. The ultimate strength of 304 stainless steel obtained by adjusting the U.S. Steel data and the corresponding 2/3 Su allowable stress for design of the reactor internals are given below.
Tensile tests at temperature have been performed on material samples from actual reactor internals for another plant, showing that the actual ultimate strength exceeds the tabulated value.
- d. Stresses compared to the 2/3 Su allowable ge calculated on an elastic basis.
If a plastic analysis of.a reactor internals component is performed,
- he plastic strains are limited to the strain corresponding to 2/3 Su. For 30h stainless steel, this strain limit is approximately -
equal to 20% of the uniforn strain. 0032 t l 3 9-2 s L. .-
D-B ( The following is a tabulation of the elastic stress limit at operatin5 temperature for materials of concern: Ultimate 2/3 S Material Stress (usi) (psi)u SA-2bO, Type 30h 56,250 37,500 SA-516, Gr 70 70,000 h7,000 SA-105, Gr 2 70,000 h7,000 SA-302, Gr B 80,000 Sh,000 SA-508, Clase 2 80,000 54,000
- e. The 2/3 uS criteria or the stress limits of N h.7.11 (a) 6 and (b) are used.
When the 2/3 uS criteria for stresses calculated en an elastic basis is used, the plastic deflections also will be determined for co=parison with allowable values. The method of analysis used and results, including margin of safety, obtained will be supplied for each component. The =argin of safety provided by each limit depends on the actual geometry of a component, material properties, etc. 0093
. 1 3.9-3 A=endment 6 l
D-B
~'
3.10. It has been indicated in Appendix 5A that " reactor vessel inter-nals must satisfy deformation limits which are more restrictive than the stress limits." The deformation limits, though re-ferenced in other sections of the PSAR, are not presented. Provide:
- a. The deforsation limits for reactor internals.
- b. The deformation limits that correspond to the stress limits in Table SA-2 for essential reactor internals.
- c. The deformation limits at which " loss of function" for essential reactor internals is anticipated to occur.
RESPONSE
The reactc- internals vill not deflect or deform to an extent that would result in a loss of functional integrity, i.e., prevent the incertion of control rods or prevent the flow of coolant to the core. An "allovable deflection", less 3han the
" loss of function" deflection, has been established for each internals component and vill be used as the design limit. The allowable and loss-of-function deflection for each major component are tabulated in Topical Report BAW-10008, Part 1 (Table 2) .
The deflection of each internals component will be calculated for
. the Davis-Besse station and meet the above limits.
e 0094
=
W 3.10-1 L
D-B 4.0 Reactor Coolant System Design h.1 Discuss the means to be used to determine if Class I mechanical ecmponents qualify for service under seismic leading condit_ ons ; e.g. , analysis or shake table tests. Relate the methods used to frequency spectrum and a=plitude calculated to exist at :he equipment mounting. State the basis for assuming that iters such as emergency core cooling pu=ps and drives will start and run, if needed, under seicnic loading. Relate your response to any tests or analyses to be perforned on equipment in the running mode as well as the static mode. Provide a su= nary of all such pieces of equipment, such as tanks and pumps.
RESPONSE
The active components in the reactor coolant system which are classified as Class I mechanical ecmponents consist of the reactor coolant pumps with , motors and the control rod drive mechanisms. These components are modeled in the seismic analysis of the reactor coolant system and the appropriate response spectra imposed at the support points of the major contributing components of the system. The resulting seismic displacements and accelera-tions of mass points will be used to check bearing loadings and shaft deflections. As the pump-motor shaft is designed to have a natural frequency at least 20 percent above the critical speed, the shaft is too stiff to respond to any of the lower seismic frequencies. *he pump and motor bearings are designed to be. capable of meeting the seismic design criteria. The design specification for the control rod drives requires that the drives be capable of withstanding the seismic loadings within the stress limits for Class I equipment. The purchase specifications for the ECCS pumps require that the pu=p be capable of operating under the seismic loads predicted to exist at the building elevation where the pumps are located. The equipment supplier is required to certify that his equipment does adequately meet all requirements of the purchase specification. The center of gravity for this type of equipment is lov and both the pump and the driver are rigidly c >nnected to a structural baseplate which in turn is i bolted to the building. This type of equipment is structurally quite rigid i and vill accommodate much higher "g" 1cadings than required to meet the seismic loads found in a central station power plant. Tests are planned to determine the performance of the equipment throughout , its operating range under steady state conditions unless duplicate equipment has been previously tested. 0095 h.1-1
- . .- ._ . . -. = . . . -
D-B The following-list summarizes the reactor coolant and emergency core cooling systems Class I tanks and pumps: ! Core flooding tanks 1 Decay heat pumps High pressure injection pumps i
- Containment spray pumps Service water pumps Component cooling surge tank m
..a e=
0096 p h.1-2 .
D-B
,_ h.2 Describe in more detail the analysis procedures that vill be
/ used.to determine that the nuclear steam supply system will moet seismic Class I criteria (Section 3.1.2.h, page'3-3 PSAR). Include in this discussion the following:
- a. A detailed description and sketch of the proposed mathematical model(s) of the system, including a discussion of the degrees-of-freedom and methods of lu= ping masses and determining section propertieu.
- b. The mathematical model(s) to be used' for the reactor vessel internals.
- c. A discussion of the analytical procedures to be used, including the methods of computing periods, mode shapes, design accelerations, displacements, shears and moments.
- d. An explanation of which " actual earthquake records" are ,
to be used in the time-history analyses and a comparison of the response spectra from these earthquakes and the spectra postulated for the site (PSAR Figures 111-5 and 111-6, pages 20-47 and 2C-h8) .
- e. An explanation of how it will be determined that the LOCA and maximum earthquake time-histories are conser-vatively applied such that the maximum structural response is obtained. Are the LOCA and earthquake time-histories assumed to start at the same time? If so, would it be possible to obtain greater response if the earthquake were started at scme increment of time, such as 10 seconds, either before or after the start of the LOCA?
- f. A listing of the damping values to be used.
RESPONSE
- a. The analysi3 to be used to determine that the core will meet seismic ,
& Class I criteria (Section 3.1.2.h, page 3-3 PSAR) vill be similar to
- b. that discussed in BAW-10008, Part 2, " Fuel Assembly Stress and Deflection Analysis for Loss-of-Coolant Accident and Seismic Excitation:.
The first segment mcdel vill te similar to that shown in Figure 22 BAW-10008, Part 1 and Figure 5 BAW-10008, Part 2. The major difference vill be that the reactor vessel will be supported at the nozzle rather than at the skirt. The main purpose of this first segnent model vill be to determine the responses of the upper and lower grid plates and the core barrel wall. This vill be the same model described in < Question 4.2.b. The second segment model vill be similar to that described in BAW-10008, Part 2. The model censists of five (5) fuel assemblies, the upper and lower grid plates, and the core barrel vall. An engineering sketch of l# the model is shown in Figure h.2-1. The mass of each individual fuel assembly is concentrated at three (3) discrete points, each mass value p; 0037 4
D-B being one-fourth (1/4) of the total mass of the fuel assembly. One (1) mass point is located at the center of the fuel assembly and the other two (2) are located a distance equal to one-sixth (1/6) of the total . .') length of the fuel assembly from either end. These mass points are connected by massless flexible beam elements, having the section properties of an actual fuel assembly. This model is shown in Figure k.2-2. The excitation of this model and the type of results obtained are discussed in BAW-10008, Part 2. The mathematical model of the primary reactor coolant system shown in Figure 4.2-3 vill consist of an equivalent lumped mass for the steam generator, reactor vessel, two (2) pumps, the 36-inch and 28-inch piping lines, and the secondary shield vall. The secondary shield vall is included because a structural support will exist between the steam generator upper tube sheet and the vall. Sy=setry of the sy: tem vill be evaluated to confirm that the system can be analyzed as a single loop. The lumped mass model which vill have about thirty (30) masses will consist of a three-dimensional assemblage of massless beam elements with either straight or curved centroidal axes. Each joint between elements vill have six (6) degrees of freedom, three (3) translations and three (3) rotations. The equivalent lamped masses of the beam elements will be determined by taking one-half of the mass for each of the two (2) beam elements adjacent to the mass joint. Some mass joints will have concentrated mass, e.g., the pump casing, in addition to the beam masses. Each mass will have three (3) degrees of translational freedom. Masses will be located so that the most significant modes will be defined. F: 2xural and torsional moments 3 of inertia vill be determined for each beau element. Shear area vill '
/
also be determined since the analysis will acu~ tnt for the effects of shear deformation. The seismic loads acting on the reactor .2ternels will be calculated using conservative dynamic analysis methods. Tli stresses due to these loads will be calculated using classical static a:"ess analysis methods. The seismic stresses will then be combined with stresses from other loading sources in the manner described in Table 5A-2, and compared to the allowables. The seismic dynesic analysis will be performed using the response spectra approach. Two seismic re.ponse spectra - horizontal and vert. cal - are specified for the building elevation at the interface between the vessel nozzle supports and the concrete support structure. Thelumped-massdyramicmodelusedforthehohiznn+=1 earthquake analysis will be similar to that shown in Figure 22 of Topical Report BAW-10008, Part 1, but will reflect the changes made for the Davis-Besse station. The major difference is the nozzle-support of the < reactor vessel. The distributed mass of components of the structure will be concentrated at discrete points, and these mass points connected by massless flexible elements. The model vill include representation of the reactor vessel, internals, drives, and core. In formulating the model, enough masses will be used to sufficiently si=ulate the dynamic characteristics of the h.2-2
D-B actual structures. The model vill include one lateral degree-of-freedom per mass point. The shell structures are treated us beam elements, with both flexture and shear flexibility. Actual area, coments of inertia, elastic modulus , etc . , are used. The internals loads due to the vertical seismic motion vill be determined using one-dinensional spring-mass uodels of the plenum assembly and of the core-core support structure, similar to Figure 19 of Topical Report RAW-10008. This simplified approach was adopted because all of the vessel and internals components are stiff (high frequency) for vertical motion.
- c. To determine the seismic response of the reactor coolant system, a flexibility matrix corresponding to the three-dimensional lumped-mass model vill be obtained from a s,tructures program. The flexibility matrix vill be used in a dynamics program which calculates natural ,
frequencies, mode shapes, participation factors, the composite model damping for each mode, and the response of the system in the "X", "Y", and "Z" directions for the prescribed response spectra. The inertial forces determined from the response accelerations vill be calculated for each mode. The significant modes, usually 10 to 20 in number, vill be determined. The reversed effective forces equal to the inertial forces for these modes will be used in the structures program to determine the moments and forces for each mode. The RMS of the moments and forces at each joint in the piping system vill be obtained from the modal values. These values for each of the ten horizontal directions , "X" and "Z" will be combined with those for the vertical "Y" direction. Stresses at each joint vill be computed frem the conbined EMS moment and forces. Once the mod 1 to be used for analysis of the reactor internals is fornulated, he various dynamic parameters natural frequencies, mode shapes , modal participation factors , sodal effective masses , etc. , vill be deten ined. The computer code employs the trial vector iteration proc 3ss to calculate the frequencies and mode shapes. The acceleration, displacement, and effective static forces for each mass and the shear force and bending moment for each flexible element vill be determined for all modes of vibration using the required response spectra. The values for cl1 modes are ecmbined on an RMS basis, and stresses on the internals components calculated using the RMS shears e and moments. The stesses on each internals component for horizontal and vertical seismic loadings are added d.irectly.
- d. PSAR Amendment No, h, page 2.C-h2, Time-history accelerogram describes the actual earthquake recerds to be used in the analysis. The computed spectral motions vill not be significantly smaller than those of the response spectra shown in Figure III-5 for damping ratios of 0, 0.005, 0.01. ^.02, 0.05 and 0.10.
h.2-3 cess
f Dy s D-B ( e. The max'imum possible earthquake and LOCA are not assumed to start at the same time. An investigation vill be performed to determine the -
}
relative timing between the start of the maximum possible earthquake and the start of the LOCA to give the maximum structural response of the core. This is done by varying the start of the LOCA with respect to the start of the maxi.92m possible earthquake and observing the responses of the core,
- f. The damping values used in the analysis are as follows:
Reactor Vessel, Drive - 1% Support Structure Reactor Internals, (core barrel, core support shield, plenum assembly) - 2% Fuel Assembly Damning
- The fuel assembly damping values vill be established from several test programs in which full-sized test specimens are used. Tests vill be performed in air, in still water at temperatures up to 200 F, and in still and flowing water at reactor operating conditions (650 F and 2200 psi).
Both displacement loading (pluck tests) and steady state, sinusoidal, excitation vill be used. This extensive testing vill detert'ne the damping range depending on the environment. Damping is also depeaat.at on the amplitude of vibration. This dependence is due to fuel rod slippage in the spacer grids, and the slippage is the prime source of the damping values. Previous tests have established that damping increases with the coolant flow velocity owing to the effect of coolant flov'on the Spacer grids. Damning Values The test program established that damping values vary with the coolant flow rate. Because of this, three different levels of damping have been established depending on the accident condition being investigated. Maximum Possib'.e Evthquake A value of 121 critical damping is uset ' or th. conc tion. This value is based on t ests described in B&W Topical Repc. t BA' -1008. This is a
? zero-coolant flav level since the earthquake c a m ur with the reactor ashutdown and the criterion is based on minimal plastic deformation.
1 s i 1 0100 ) O h.2 h
D-B C i k.3 Hov vill flow induced vibration loads be considered in the design of the primary system? State the extent, methods and findings of the analyses or tests which will be made. In this statement include responses tc the following specific considerations:
- a. Will both normal and emergency modes of operation be considered?
- b. What design limits, a=plitude and frequency apply to these conditions?
- c. Discuss the tests which are considered necessary for the Davis-Besse plant. In these discussions include comments en the type and extent of instrumentation -
planned.
RESPONSE
Flcw-induced vibratica anabses for normal modes of operation have been made for resetor internals such as: the ther=al shield, fuel assembly, fuel rods, surveillance tube and specimen holder assembly, control rod guide tube assembly and piping for the in-core monitors. The thermal shield analysis for vibration problemr showed that the flow-induced pressure fluctuations acting on the scrface of the shield resulted in modal amplitudes less than 0.002 inch. hrallel and cross-flow responses are determined from the Burgeen correlatien and the expression for the Von Karman vortex shedding frequency, respectively. Parallel flow amplitudes and associated stresses are minimized by design so that the cyclic stresses vill not cause fatigue problems. Components subjected to cross flov are checked for response during design, so that the fundamental frequencies associated with cross flow are above the vortex shedding frequencies. Stret ., and deflection analyses for the reactor insernals and fuel asse=blies g due to a LOCA are described in B&W Topit a'. Report 3AW-10008 Part 1 and Part 2. , No vibration tests for measurement of flow-induced vibrations in the reactor coolant system are planned or considered necessary for Davis-Besse. Reactor internals similar to those to be used in Davis-Besse vill be vibration tasted in an early B&W-designed reactor. These tests will f be completed prior to the completion of Davis-Besse. The design of the l Davis-Besse zeactor vessel and internals does not preclude vibration 6 l
-testing cr the removal of the reactor internals for visual inspection l folleving preoperational testing.
0$.0$. h.3-1
D-B s 6 The vibrati n tests in an early B&W reactor are expected to assure the vibration resistance of the Davis-Besse unit for the following reasons:
- 1. The Davis-Basse internals will use a design configuration comparable g to the unit which is vibration tested and to the units started up previously. There are no design differences which would significantly affect flew-induced vibrations.
- 2. The matetials for Davis-Besse internals are comparable to those in the unit receiving the vibration test.
- 3. The initial flow conditions for Davis-Besse and the unit to be tested are the same.
- h. A high level of quality assurance on Davis-Besse vill assure that the internals are built as designed.
- 5. The above factors should assure that the natural vibration frequencies of the internals components for Davis-Besse are very similar to those for tne unit tested.
c
.0102 4.3-2 -
c4
D-B r~ h.h Provide the following information concerning the design and design criteria for the reactor vessel:
- a. Special requirements, if any, imposed by local or state regulation on the reactor vessel design.
- b. The acceptance standards intended for nondestructive testing procedures for the reactor vessel. Indicate if these meet or exceed the requirements of Section III of the ASME Boiler and Pressure Vessel Code,1968 edition.
- c. Indicate whether transients such as loss of reactor vessel flow (one or two loops) and loss of load vill be considered in the transient stress analysis. If not, provide a justification for your position.
- d. The design conditions for the core flooding water nozzle.
- e. A list of stainless steel component parts in the reactor vessel and the reactor coolant system that will become furnace-sensitized during the fabrication cycle.
- f. Provide a summary discussion and enumeration of results of transient stress analyses, illustrated by sketches showing points of analysis, and a list of associated cumulative fatigue usage factors for the reactor vessel.
I
- g. Discuss item 2 in Table b-9 with respect to design cycles versus actual cycles (the design cycles appear to be larger by a factor of 10 than for previous plants).
- h. Discuss the magnitude of the stress in the reactor vessel membrane induced by ga=ma ray heating.
- i. Will ring forgings be used for reactor vessel shell sections other than the closure flanges?
- j. Provide summary results of Charpy V-notch and Drop l Weight tests for the reactor vessel plates and forgings.
RESPONSE
- a. The State of Ohio imposes no additional requirements.
- b. The nondestructive testing procedures and related acceptance standards for the reactor vessel will meet or exceed the requirements of the ASME Boiler and Pressure Vessel Code, Section III,1968 Edition. -
- c. Loss-of-reactor coolant flow in one and two loops and loss of load are included in the transient stress analysis.
- d. The core flooding nozzles are designed for 2500 psig and 650 F and for
( the following operating conditions. 33 03.oa h.h-1
Nor=al Operation: (Cooldown and decay heat removal) To pass up to 3000 gpm of reactor coolant through each nozzle when reactor coolant system temperatures are equal to or less than 280 F and 300 psig, respectively. The nozzles are designed for h80 occurrences. Emergency Oteratien: (LOCA) To pass up. to 7000 gpm of 90 F veter per nozzle from the core flooding tanks when reactor coolant pressure drops to 600 psig. The nozzles are designed for one occurrence,
- e. No pressure boundaries will contain f Irnace-sensitized stainless steel.
The stainless' steel cladding is the only part of the reactor vessel And
'the associated reactor coolant system that vill become furnace-sensitized '
during fabrication cycle.
- f. The points of stress analysis are shown en Figure h.h-l. A summary of the results of the transient stress analysis and cumulative fatigue usage factors for the Oconee vessel which may be considered as preliminary information for Davis-Besse is listed as follevs:
Maximum Primary Plus . Component Part of Seco2dary Stress Maximum Reactor Vessel Intensity Range Umage Factor -] (1) Control Rod Housing 2h,800 psi 0.0 (2) Closure Flanges (a) Closure 63,600 psi 0.03 (b) Studs (axial load
& bending) 85,600 psi 0.57 (3) Primary Inlet and Outlet Nozzles 2h,000 psi 0.06 (h) Reactor Vessel Shell 32,270 psi 0.0 (5) Core Flooding Nozzle 23,660 psi 0.02 (6) Instrumentation Nozzles 10,150 psi 0.0 m
The maximum primary plus secondary stress intensity range is compared to 3 Sm for the particular material at its operating temperature for all components. The maximum allowable usage factor is 1.0 per ASME, Section III. The values tabulated in this summary are the maximum values obtained in each regicn of analysis. l y 0104 J h.h-2
- g. The lh,h00 design cycles for item 2 in Table h-9 is in error and should be lhh0 cycles ,
- h. The maximum steady-state stress resulting frem ga=ma heating in the reactor vessel shell has been calculated to be 316h psi (tension) . This is a relatively lov value, and no problems are anticipated frem thermal stresses in the reactor vessel vall.
- 1. The Davis-Besse reactor vessel vill be constructed of hollow shell forgings from belev the closure ring forging devn to the bottom head as shewn en Figure h.h-2. Ring forgings and hollev shell forgings are similar except for length. Whereas, the ring forging has a small length (or depth), the shell forging length (or depth) vill be greater and may be in the range of 1/8 to 1/2 of the outside diameter.
J. A listing of the Davis-Besse reactor vessel Charpy V-notch test results is not yet available. As an example of the information to be furnished in the FSAR, we have attached Table h.h-1 from a previous FSAR. This table shows measured values for a si=ilar reactor vessel but with rolled plate core shell courses. 03.05 g h.h-3
Table h.h-1 Typica] Reactor Vessel -- Material Physical Properties Ultimate Yield Impact Strength Strength Elong. in Test Temp. Impact Item Heat No. (103 psi) (103 psf) 2 in. (%) ( F) Values Closure llead Center Disc C 2311-2 90.3 69.5 31.0 +10 44-38 43 Bottom Head A 0973-2 87.2 65.0 24.5 +10 35-30 47 Upper Shell Course C 2197-2 91.5 70.0 25.0 +10 39 45-26 Middle Shell Course C 3265-1 87.0 66.2 28.1 +10 3h 64-27 Middle Shell Course C 3728-1 84.5 63.5 28.1 +10 35-29-53 Bottom Shell Course C 2800-1 85.0 60.5 29.0 +10 36-39-39 Bottom Shell Course C 2800-2 90.5 69.0 25.0 +20 32-33 49 Core Flooding Nozzle 94894 98.0 74.0 21.5 +10 45-53 40 e Core Flooding Nozzle 94894 92.5 71.0 2h.0 +10 37-50 45 p 90.0 67.5 25.0 +10 104-94-142 7 Inlet Nozzle 123S346VA1 123S346VA2 92.7 72.5 26.0 +10 104-121-106 Inlet Nozzle 12hS502VA1 97.2 76.0 25.0 +10 120-106-101 Inlet Nozzle Inlet Nozzle 12hS502VA1 94.0 73.5 23.5 +10 110 85-77 Outlet Nozzle 122S316VA2 90.0 67.0 26.0 +10 131-110-94 Outlet Nozzle 122S316VA1 90.0 68.5 25.0 +10 92 86 82 Upper Shell Flange hP16373P1566 82.5 57.h 29.0 +10 49 41-71 liead Transition Piece 122S3h7VAI 9h.5 74.5 24.0 +10 92-70-70 Closure Head Flange 125S535VAI 102.0 81.0 23.5 +10 59 47-70 Closure Head Ring 99392D-2 96.5 75.5 26.0 +10 73-79-88 ZV-2888 82.0 57.0 30.5 +3h avs 30 avg Upper Nozzle Shell Course Imver Nozzle Shell Course ZV-2861 85.0 63.5 29.0 +26 avg 30 avg Q p b O . O . J
D-B p h.5 Discuss the extent to which electroslag velding vill be used in the fabriction of Class I systems. If electroslag velding is to be used, describe the process, its variables, and the quality centrol procedures to be employed.
RESPONSE
Electroslag velding vill not be used in the fcbrication of the B&W lev-alley steel and/or stainless steel ec=penents as classified by Table N h2h, Sectica III, ASME Code. Also, electroslag velding vill not be used in the fabrication of the primary pu=p casings. Electroslag velding vill be utilized en longitudinal seams of the T-inch shell courses of the steam generator, resulting in four velds per generator. . In addition, there are three longitudinal sea =s in the pressurizer that are electreslag velded. Each veld vill be subjected to radiographic inspection, ultrasonic inspection and the finished surfaces of the veld will be magnafluxed. In addition, each plate is ordered with excess vidth so that test specimens may be re=oved after heat treatment. Physical property test specimens including tensile and impact specimens of the base material heat affected zone and veld metal vill be obtained from this access =aterial in accordance with Section III of the ASME Code. Radicgraphie ultrasonic and magnetic particle inspection vill be per-formed in accordance with Section III of the ASME Code and as required by Code Case 1355 which permits such velds for Class A vessels. Physical tests will be performed per Paragraph N-511 of Section III of the ASFS Code. For example:
- a. All veld metal tensile specimens from each heat of veld wire, batch of flux and for each combination of heat of wire and batch of flux used vill be obtained and tested after heat treatment.
- b. Charpy impact test specimens representing veld =etal and heat affected base material for every heat of wire, batch of flux and combination of heat of wire and batch of flux used vill be tested.
- c. Charpy V-notch impact specimens and tensile specimens vill be tested for 25 per cent of all production velds . Included in this 25 per cent are ,
the tests required by (a) and (b) acove. The justification for =easuring ' physical properties, by Charpy and tensile tests, en a portion of the i production electroslag velds is based en the folleving: J (1) There is no ASME Section III requirement for cenducting physical 6 l property tests en each and every production veld. The cnly time such tests are required is when a new lot of veld filler metal or lot of flux is used and when the veld procedure qualification test is conducted. Y 01.07 h.5-1
D-B (2) The policy of testing 25% of the production velds was established to provide assurance. that there was no reductice in =aterial strength attributed to the electroslag veld. (3) Experience indicates that testing of 25% of the production electro-6 slag v lds is adequate. To date all tensile specimens have met minimum base metal requirements. Also, all Charpy test specimens have shown impact values greater than base netal minimum require-ments. BW has been utilizing the electroslag process for some time and all results to date indicate that this process results in velds that equal or surpass the integrity of submerged are velds. All electroslag velds are made in the vertical position. Two men, one on the inside and one on the outside of the vessel, are used to check the progress of the veld and to ensure that the prescribed welding procedure is being followed. The veld is started in a U-shaped starting fixture about six inches deep attached to the bottom of the joint. The veld stabilizes in this etarting tab which is later cut off and discarded. The veld receives a heat treatment which consists of a water quench frem 16250F and a temper of 11500F followed by an air cool. This post-veld heat treatment refines the grain of the veld and the base material heat affected zone such that it is virtually indistinguishable from the unaffected base material. The microstructure is the same throughout the veld. d olo% 1 1 h.5-2
D-B k.6 Section h of the PSAR describes your plans for inservice in-spection of the reactor coolant system. Compare your proposed program with the ASME Code for Inservf ae Inspection of Nuclear Reactor Coolant System, ASME Section AI. If any of the reactor coolant system pressure boundary area initial baseline examin-ations called for in this code are to be omitted, identify the areas and discuss your reasons for not testing them. Should any
, areas required in the code be precluded from an inservice in-spection due to inaccessibility, discuss the reason for such inaccessibility.
RESPONSE
It is intended that in-service inspection of the reactor coolant system
- will be in accordance with the ASME Code for in-service inspection of nuclear reactor coolant systems. Provisions in the design of the station ;
to provide adequate access and clearances for inspection systems which can reasonably be expected to be developed will be provided. As detail design develops, there could be areas where accessibility could be limited or impossible but it is intended that accessibility will be maintained to the fullest extent practicable. There are no known base line examinations that are to be omitted. , Toledo Edison is, and vill follow programs, dealing with the development of techniques and equipment to perform these inspections and is currently undergoin6 discussions concerning active participation in one such project. 0109 h.6-1
D-B C h.7 Describe your inservice inspection program for the Class I (seismic) mechanical systems outsid; the primary system pressure boundary, including itams to be inspected, inspection schedule, and types of inspection. Some items to be considered are primary system components, support, primary pump flywheels, and Class I (seismic) mechanical components in the engineered safety features.
RESPONSE
The primary vessel supports vill be inspected as required by the ASME Code for Inservice Inspection of Nuclear Reactor Coolant Systems. The primary pump flywheels will be ultrasonically tested either during =anufacturing of the motor or in the, field prior to initial starting. Each primary pump flywheel vill be further inspected once in each ten-year interval by ultrasonics or equivalent method.
- Mechanical systems outside the primary system boundary will be inspected on a routine basis for leaks from pump seals, valve packing, flanged joints and relief valves. The inspections will be visual for signs of distress or potential trouble. In addition, components will be periodically tested to demonstrate equipment readiness and operability. For exa=ple, the emergency injection equipment performance testing is itemized in Table 6-3 of the PSAR.
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D-B k.5 Discuss the extent to which electreslag velding will be used in
'N the fabrication of Class I systems . If electroslag welding is f
to be used, describe the process, its variables, and the quality control procedures to be employed. RESP 0NSE Electroslag velding will not be used in the fabrication of the B&W low-alloy steel and/or stain 1 css steel components as classified by Table N k2h, Section III, ASNE Code. Also, electroslag welding will not be used in the fabrication of the primary pump casings. Electroslag velding will be utilized on longitudinal seams of the 7-inch shell courses of the steam generator, resulting in four velds per generator. In addition, there are three longitudinal seams in the pressurizer that are electroslag velded. Each veld will be subjected to radiographic inspection, ultrasonic inspection and the finished surfaces of the veld will be magnafluxed. In addition, each - plate is ordered with excess width so that test specimens may be removed after heat treatment. Physical property test specimens including tensile and i= pact specimens of the base material heat affected zone and weld metal will be obtained from this access material in accordance with Section III of the ASME Co de . Radiographic ultrasonic and magnetic particle inspection will be per-formed in accordance with Section III of the ASME Code and as required by Code Case 1355 which permits such velds for Class A vessels. Physical tests will be performed per Paragre7h N-511 of Section III of the ASNE Code. For example:
- a. All weld metal tensile specimens from each heat of weld wire, batch of flux and for each combination of heat of wire and batch of flux used will be obtained and tested after heat treatment.
- b. Charpy impact test specimens representing weld metal and heat affected base material for every heat of wire, batch of flux and combination of heat of wire and batch of flux used vill be tested.
- c. Charpy V-notch impact specimens and tensile specimens will be tested for 25 per cent of all production welds. Included in this 25 per cent are the test required by (a) and (b) above.
All electroslag welds ale made in the vertical position. Two men, one on the inside and one on the outside of the vessel, are used to check the progress of the veld and to ensure that the prescribed welding procedure is
-being followed. The veld is started in a U-shaped starting fixture about six inches deep attached to the bottom of the joint. The veld stabilizies in this starting tab which is later cut off and discarded. <
0116 M h.5-1
D-B i The veld receives a heat treatment which consists of a water quench frem 16250F and a temper of 11500F followed by an air cool. This post-veld heat treatment refines the grain of the veld and the base material heat affected zone such that it is virtually indistinguishable from the unaffected base material. The microstructure is the same throughout the veld. 3 C 0117 b ast k.5-2
D-B 4.6 Section h of the PSAR describes your plans for inservice inspection
<3 of the reactor coolant system. Compare your proposed program with the ASME Code for Inservice Inspection of Nuclear Reactor Coolant System, ASME Section XI. If any of the reactor coolant system pressure boundary area initial baseline examinations called for in this code are to be omitted, identify the areas and discuss your reasons for not testing them. Should any areas required in the code be precluded from an inservice inspection due to inaccessibility, discuss the reason for such inaccessibility.
RESPONSE
Provisions for inservice inspection of the Davis-Besse reactor coolant system vill be generally in accord with ASME Section XI, limited only by physical features inherent within the design and utilizing inspection techniques and equipment expected to be available in the near future. As appropriate in-spection systems are developed, it is anticipated provisions for their use will be incorporated into the Davis-Besse design. . The entire subject of inservice inspection is currently under review for the Davis-Besse project and this question vill be more fully answered at a later date. ( 0118 - 4.6-1
D-B 4.7 Describe your inservice inspection program for the Class I 7 (seismic) mechanical systems outside the primary system pressure boundary, including items to be inspected, inspection schedule, and types of inspection. Some items to be considered are primary system components , support, primary pump flywheels , and Class I (seismic)' mechanical components in the engineered safety features.
RESPONSE
See answer to question h.6 for general inservice inspection criteria. l l J c l O.1.1.9 M l i h.7-1
D-B 4.8 Describe the provisions used to prote :t the reactor primary system,
. other vital systems, and structural supports for these systems from missile hazards. Include a discussion of design criteria, missile shields, missile size and masses, missile velocities and the penetration formulas used for design.
BESPONSE The reactor primary system, other vital systems and structural supports for these systems are provided with protection from both internally and extern-ally generated missiles. The shield building provides protection for the containment vessel and systems within the containment vessel from external missiles while the auxiliary building provides protection for the vital systems located withi it from external missiles. The containment vessel, primary system and other vital systems within the containment vessel are protected against internally generated missiles by the primary and secondary shield walls, missile shields, or by physically locating components capable of generating a missile in such a manner that the potential missile cannot cause damage. The general criteria for missiles generated within the containment vessel is as follows:
- 1. No missile will M pemitted to cause a IDCA.
- 2. No missile genersted by a IDCA vill be permitted to damage the reactor vessel or core flooding systems.
3 No missile generated by a IDCA vill be permitted to damage the containment vessel. The primary and secondary shield walls are constructed of reinforced con-crete and the design is such that there vill be no loss of function from any IDCA. Secondary shield walls vill be capped fith heavy grating or similar arrangement to provide for pressure venting ut designed to restrain missiles crqable of damaging the containment vessel. Removable missile shields are located over the top of the reactor vessel areal T to provide protection to the containment vessel from any missiles generated from the reactor vessel head area. Core flooding tanks, valves and lines are located outside of the primary and l secondary shield wall areas to provide isolation frcs any missiles generated I8 from other parts of the primary system. Valves will be located in such a manner that valve stem and stud missiles 9 would be restrained from damaging the primary system, containment vessel or other vital systems.
. All safety features systems are located in the auxiliary building which is a Class I building and which is also designed to provide protection from .; tornado' generated missiles. The enclosure of this building having exposure to tornado generated missiles will be reinforced concrete with a minimum thickness of 18 inches.
4.8-1 03.20
D-B The attached Table 4.8-1 is typical of the potential internally generated G missiles that are being considered in the design of the Davis-Besse Station ! to meet the above criteria. Tornado missile data being used in the design for tornado missile protection is given in paragraph 5 2.2 3 5 on page 5-16 of the PSAR. In addition to these missiles an additional tornado generated missile consisting of a steel pipe 10 feet long, 3 4 inches 0.D. , weighing TO lbs., travelling 100 mph and striking end on is being used. Penetration formulas being used are those found in Report No. SRIA-17 from Stanford University and in NAVDOCICS P-51 (Modified Petry Formula). l f 07,3$ d 6 4.8-2
D-B - Table 4.8-1 MISSILE SOURCES, MASSES, AND ENERGIES The physical layout of the Reactor Coolant System including the various mechanical attachments and accessory equipment within the containment vessel include items which have been defined as potential credible missiles. These potential missiles can be categorized into three classes. I. Missiles generated by the conversion of stored strain energy to kinetic energy, e.g., nuts, bolts, studs, and items not directly under system pressure. II. Piston type missiles, e.g., valve items or other items acted on by a constant force for some distance in a . piston type action. III. Jet propelled missiles, e.g., head covers, valve bonnets and items in direct contact with system pressure. A list of the potential missiles by class giving their weight, impact area, velocity and kinetic energy are given in Table 1 below. The table has been divided into parts, each part covering the potential missiles from a major component of the Reactor Coolant System. The table does not include the velocities or the kinetic energies for the Class III type missiles. The maximum velocity obtained by ~he Class III missiles depends on the dis-tance of travel before impact. Therefore, physical layout of equipment, position of missile shielding, etc. must be considered when calculating maximum velocity and kinetic energy for each missile. 0122 4.8-3
r CREDIBIE MISSILES TABIE 1 A. REACTOR VESSEL & CONTROL ROD DRIVE Missile Impact Maximum Kinetic Missile Weight Are Velocity Energy Class Description (1bs.) (in.g) (ft/sec) (ft-lbs) I 1. Reactor Closure head nut 80.0 38.o 577 0 11,680
- 2. ReactorClosureheadstudv/ nut 660.0 T1.0 97 0 96,hoo 3 1" Valve bennet stud 05 0.6 T3 5 42
- 4. Control Rod (CR) nozzle flange bolt 30 31 97 0 438 II 1. CRD closure cap 8.0 T.O
( B. STEAM GENERATOR Missile Impact Maximum Kinetic Missile Weight Area Velocity Energy Class Description (lbs.) (in.2) (ft/sec) (ft-lbs) I 1. lh" Vent valve bonnet stud 2.0 0.8 T3 5 167
- 2. Feedwater inlet flange bolt 03 0.6 67 5 21 3 16" I.D. manway stud, tube side 8.0 2.1 67 5 566
- 4. 5" Inspection-opening cover stud 15 1.2 73 5 125
. 5 1" Valve bonnet stud 05 0.6 73 5 42 II 1. Vent valve stem & wheel 50 0.45 44.5 154
- 2. Sample line 1" valve stem & wheel 4.0 03 35 8 80 3 Sample line 1" EMO valve stem & wheel 4.0 03 35.8 80 O
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C. PRESSURIZER
-~
MiFsile Impact Maximum' Kinetic Missile Weight Area Velocity Energy Class Description (1bs.) (in.2) (ft/sec) (ft-lbs) I 1. 4" Valve bonnet stud 30 1.8 73 5 250
- 2. 5" Valve bonnet stud 30 2.4 73 5 250 3 16" Manway cover stud 75 31 67 5 530
- 4. Heater bundle stud 25 0 70 73 5 2,100 ,
5 3/4"Valvestemstud 0.8 0.45 73 5 67 II 1. Spray line 4" Electric Motor Operated 9 1.0 135 0 2,560 (EMO) valve sten
- 2. Sample line 3/4" valve. stem 4 03 72 7 330 y 3 Sample line 3/4" EMO valve stem 4 03 72 7 330 D. REACTOR COOLANT DRAIN TANK Missile Impact Maxi 2num Kinetic Missile Weight Area Velocity Energy Class Description (lbs.) (in.2) (ft/sec) (ft-lbs)
I 1. 1 " Drain valve bonnet stud 0.6 0.2 73 5 50
- 2. 4" Valve bonnet stud 2.0 03 73 5 167 II 1. li"EMOdrainvalvestem 50 0.45 11.0 9
- 2. 4" EMO valve stem 90 1.0 21 5 65 E. INSTRUMENTS o
Le Missile Missile Maximum Kinetic Weight Impact Velocity Energy Is) Class Description (1bs.) Area (in.2) (ft/sec) (ft-lbs) M III 1. Resistance Temperature Element 1.0 0.2
- 2. Resistance Temperature Element & Plug 2.0 4.0
F. SYSTEM PIPING Missile Impact Maximum Kinetic Missile Weight Are Velocity Energy Class Description (1bs.) (in.g) (ft/sec) (ft/lbs) Core Flooding Line I 14" Check Valve (C.V.) bonnet stud 2.0 17 73 5 167 I 14" Valve bonnet stud 35 4.0 67 5 248 II 14" C.V. check pivot stud 10.0 1 75 2h9 0 9,650 II 14" Power operated (P.O.) valve stem 98.0 50 143 0 31,100 Low Pressure (L.P.) Injection Line I 12" C.V. bonnet stud 2.0 17 73 5 167 I 12" C.V. check pivot stud 10.0 1 75 249 0 9,650 y Reactor Vessel Outlet Line to L.P. System g , I 10" Valve bonnet stud 25 17 73 5 177 g I Relief valve bonnet stud 05 03 73 5 42 I Relief valve stem assembly 40.0 12 5 35 3 768 II 10" EMO valve stem 50.0 31 130.0 13,200 R.V. Inlet Line from High Pressure (H.P.) System I 4" C.V. bonnet stud 1.0 0.8 '(3 5 83 5 II 4" C.V. check pivot stud 30 0.8 158.0 1,170 Steam Generator Outlet Line to Pump Inlet I 1" Drain valve bonnet stud 0.8 0.6 73 5 67 II 1" Drain valve stem assembly 4.0 03 84.0 438 Chemical Addition System Pressurizer Line I 3/4"Valvebonnetstud 1.0 0.45 T3 5 83 II 3/4"Valvestem 4.0 03 73 0 330 II 3/4"EMOvalvestem 4.0 03 73 0 330 C u v1 4 .
D-B h.9 Reactor vessel Material Surveillance Program You state that you are participating in the B&W integrated surveillance program as presented in the topical report, BAW 10006,
" Reactor Vessel Material Surveillance Program." We understand that extensive revisions are being made to this program. Pro-vide the following in regard to the Davis-Besse vessel:
- a. The total number of capsules and capsule locations related to the core midplane.
- b. The number of capsules of each type to be withdrawn and tested over the life of the vessel.
- c. A listing of the archive material reserved. Indicate if sufficient archive material has been set aside to fabricate enough specimens for a minimum of two capsules.
- d. A listing of the chemical centent of the vessel materials, including the residual element centent in weight percent to the nearest 0.01%.
RESPONSE
Since the submittal of the PSAR; the decision has been made by B&W to construct the Davis-Besse reactor vessel using forgings rather than plate. This alcng with the decision to utilize forgings on some of the other B&W fabricated vessels has necessitated an integrated surveillance program for forgings in addition to that described for plate in the B&W topical Report BAW-10006. B&W Topical Report BAW-10006 is being revised consistent with the B&W infor-mation presented to AEC-DRL cn January 21, 1970, which included separate with-drawal schedules for plate and forging material. The follcwing is in accord with the information presented and being included in BAW-10006, Revision 1:
- a. A total of eight capsules vill be installed in three holder tubes. Six capsules as 3 pairs are positioned near the peak azimuthal neutron flux location and each pair vill straddle the core midplane. The remaining thermal aging pair (2 capsules) are located where the neutron flux level is essentially zero. WOL (1-X) specimens have been replaced with additional Charpy specimens.
- b. Two flux sone capsules or one pair of capsules (1 upper and 1 lover) vill be withdrawn and tested during the vessel 40-year design life.
- c. A separate listing of archive material is not available at this time and vill not be available until after the required quantity of surveillance 1 specimens have been finish machined for the Davis-Besse station. All specimen material has been ordered (or reserved) from the vessel fabrication shop as described in Topicel Report BAW-10006, " Reactor Vessel Material Surveillance Program", in the form of rough machined plates. These plates contain more than sufficient material to produce the quantities of finished
_l__, NY 4.9-1
D-B specimens listed in BAW-10006, to provide for cutting or machining opera-tions, to allow for fabrication scrap, and to allow an ample remainder for archive material. After all finished specimens are leaded into the capsules, the archive material vill be listed and retained in the form of rough machined plates until such time as it may be required to produce additional specimens. Therefore, an exact determination of the archive material available cannot be made until after the test specimens have been machined, but we are confident that sufficient material vill be available
. to fabricate enough specimens for a minimum of five capsules, if ever required, for the Davis-Besse station.
- d. A listing of the chemical analysis of the Davis-Besse reactor vessel material is not yet available and vill be provided in the FSAR. As an example, we have attached Table 4.9-1 showing measured values from the FSAR for a similar vessel. The core shell materials for the Davis- :
Besse vessel have been purchased on the basis that the residual element content will be reported in weight percent to the nearest 0.01 percent. Section h.h.3 of the PSAR has been amended in accordance with this response. a I h.9-2 .- 1
s [' , Table 4.9-1 Typical Reactor Vessel - Chemical Analysis Element Heat Number C Mn P S Si Ni Mo Co V Cr c 2311-2 0.22 1.35 0.009 0.018 0.22 0.61 0.41 0.005 - - A 0973-2 .21 1.34 .011 .016 .18 46 .47 .010 - - c 2197-2 .21 1.28 .008 .010 .17 .50 46 .021 - - c 3765-1 .21 1.42 .015 .015 .23 .50 49 .016 - - c 3728-1 .19 1.26 .010 .016 .23 60 47 .016 - - c 2800-1 .20 1.40 .012 .017 .20 .63 50 .014 - - c 2800-2 .20 1.40 .012 .017 .20 .63 .50 .014 - - 94894 .22 0.62 .006 .009 .23 .87 .60 .016 - 0.33 1233346VA1 .22 .61 .010 010 .20 .69 .56 .01 0.01 .27 123S346VA2 .21 62 .010 008 .20 .69 .57 .01 .01 .28 e, ', 1248502VA1 .22 .010 e
.65 010 .22 .75 .59 .02 .01 .35 m 1243502VA2 .23 .68 .010 .014 .22 .78 60 .02 .01 .31 322S316VA2 .20 .62 .010 .009 .28 .73 .57 .013 .01 .33 122S316VA1 .18 .58 .010 .014 .28 .68 61 .015 .01 .32 kP16373P1566 .20 .72 .010 .012 .28 .74 .55 .'011 .03 .34 1223347VA1 .20 .63 .010 .008 .25 .66 .55 .021 .02 .32 125S535VA1 .21 .63 .010 .011 .23 .72 60 .010 < .02 .39 99392 D-2 .25 .72 .010 .025 .22 .78 .64 .010 -
.38 Zv-2888 .22 .74 .012 .010 .31 .71 .56 .007 .02 .36 f) ZV-2861 0.22 0.64 0.006 0.010 0.29 0.65 0.57 0.01 0.01 0.31 C'
RY GD
m-D-B A. 4.10 With regard to the capability for detection of leaks from the k primary system, provide additional discussion of the following:
- a. The proposed leak detection instrumentation, including a dis-cussion of sensitivity, response time, control roor alarms, diversity and redundancy.
- b. The maximum leak rate from an identified or unidentified source that will be permitted during operation, and the predicted crack size that can be related to the allowed leak rate from an unidentified source. Include a discussion of the bases for the selected leak rate, and a description of the analytical methods used to establish the relationship between the unidentified leak rate and the crack size.
- c. The sensitivity of the leak detection for the primary coolant pressure boundary. ,
- d. The leak detection systems provided for other Class I fluid systems. List those Class I fluid systems for which no special leak detection system is provided.
RESPONSE
- a. Primary system leakage will be detected by one or more of the following methods:
- 1) Pressurizer and makeup tank water levels will be continuously moni-tored. Pressurizer level is controlled to a ecnstant level; coolant leakage will result in a decrease in makeup tank level. The makeup tank capacity is 31 gallons per inch of tank height, and each gradua-tion on the level recorder represents two inches of tank height.
Coolant loss over a time interval can therefore be determined. High and low level alarms are provided from redundant level transmitters on both the pressurizer and makeup tank.
- 2) Radiation monitoring of the containment atmosphere and selected areas within the containment will provide control room recordings and alarms.
- 3) Containment atmosphere relative humidity will be monitored for changes I which can reflect primary system leakage. .
s I
- 4) Containment sump level will be monitored and alarmed on high level.
- 5) Project design has not yet proceeded to the point where details such <
as system sensitivities and response times can be stated. l I
- b. Allowable leak rates which will permit centinued plant operation will be l established during preparation of the Technical Specifications to be sub-mitted with the Final Safety Analysis Report. These Technical Specifica-( tions vill take into consideration leakage detection systems available, their sensitivity, availability, . makeup system capability, and radiological consequences resulting from such leakage.
4.10-1 ' ~ ~ M- ola
t D-B The leakage flow areas will be determined by using the orifice equation with a critical pressure ratio of 0 55 and a discharge coefficient of '3 0.61. ;/
- c. Refer to part "a".
- d. Radiation monitoring and/or sump level monitoring along with applicable flow metering will be used for leakage surveillance of the engineered
, 6 safety feature systems which directly handle the reactor coolant. No
- special leak detection systems are provided for the other class I fluid
! systems. 9 5 I b b 1 0130 di
- . u.10 2 .gr ;
i
D-B / h.11 Several statments in the PSAR indicate or imply that Class I systems will be protected against damage from failure of other ; systems. Specify the design criteria which will be applied to l protect against damage of these Class I systems by pipe whipping. RESPONSE l The basic design criteria for pipe whipping being used provides that:
- 1. The primary coolant system is to be protected from any pipe whip source that could cause a LOCA..
- 2. The containment vessel is to be protected fran any pipe whip source resulting from a LOCA and from lines required to be in service during a LOCA.
- 3. Each redundant engineered safety features system is to be protected .
from damage by pipe whip from its parallel system. lg' To meet this design criteria, the following assumptions are being used:
- 1. A severed pipe, operating at 300 psig or greater, may experience significant translocational motion in any direction, if unrestrained, and cause physical damage to other piping or components within close proximity.
- 2. Lines 3/4" nominal diameter and smaller, will not cause physical damage to large, adjacent components and piping.
To more fully categorize this criteria, table 4.11-1 is included.
.e t .
0133, Amendment No. 9 h.11-1
- 4 D-B Table 4.11-1 !
PIPE MIIP PROTECTION CRITERIA Lines A 3 C D E High Pressure Safety X X X X Injection Lines Low Pressure Safety X X X X Injection Lines Core Flooding Lines X X X X X Makeup Line X .X - Cooling Water Lines t X Containment Air Coolers Reactor Coolant X X X - Letdown Lines Decay Heat Removal Lines X X X X X
~S 9l 6{
M"i" 8U*** ""d X X X - X J) Feedvater Lines Primary Coolant 9:I System Lines X X(1) - X - Containment Spray Lines X X X X ! PROTECTION CATEGORIES 1 7. A Lines that will be restrained from damaging the primary coolant system. B Lines that vill be restrained from da-aging the containment vessel. 1 C Lines that will be protected from da2:: age by ruptured primary system piping. D Lines that vill be protected from da.nage by main steam and feed-water lines. E Lines that .will be protected from da . age by or restrained from damaging their parallel redundant lines, j (1) Protection vill be provided if results of evaluations indicate a need.0132 O Amendment No. 9 4.11-2
D-B t 4.12 Failure of a primary pump flywheel could result in the generation of missiles capable of severely damaging equipment within the containment . Pr6 vide the results of an evaluation assessing the potential consequences from possible missiles generated by failure of a flywheel. Describe the program to be followed to minimize the probability for experiencing a flywheel failure, including the consideration to be given to material selection, design margins, fabrication, failure analyses , acceptance testing, it.2ervice inspection requirements, and other special quality assurance measures. What practical measures can be taken to provide missile protection to vital equipment that could be damaged by missiles generated by failure of the flywheel?
RESPONSE
Reactor Coolant Pump Motors will be made from steel plate which will be in - accordance with ASTM Specification A-533 and has a mini =um tensile strength of 80,000 psi. In addition to chemical and tensile tests, impact tests will be performed to determine that the nil ductility transition temperature (NDT te=p. ) is a minimum of 30 F below the minimum motor operating te=perature. The flywheel material vill be vacuum degassed and vill be subjected to 100% 6 volumetric ultrasonic inspection from the flat surface in accordance with Paragraph N321 of the ASME Boiler and Pressure Vessel Code (Section III). The material vill also be subjected to 100% angle beam ultrasonic inspection using a 3% - 1" calibration notch with any indication greater than that for the calibration notch being cause for rejection. The flywheel plate surfaces vill be subjected to a magnetic particle or liquid penetrant examination in accordance with Paragraph N322 of the ASME Boiler and Pressure Vessel Code (Section III) to a radial distance a minimum of 8 inches beyond the final largest machined bore diameter. Indications of crack or linear defects are cause for rejections, a linear defect being defined as one in which the length is 3/16 inch. Primary stresses in the flywheels vill not exceed 50% of the yield stress at normal operating speed excluding stress concentrations. There vill be no stress concentrations such as stencil marks, center punch marks, or drilled or tapped holes vitnin 8 inches of the edge of the largest flywheel bore. Where keyvays are used, the minimum fillet radius vill be 1/8 inch. The motor vill be designed with the flywheel at the top of the motor, above the upper bearing bracket so that the top surrace and the periphery of the flywheel vill be readily accessible for inspection. Because of the very high quality of the flywheel, failure is considered j incredible, and the analysis of the consequences of failure is not being carried out. 0133 M h.12-1
D-B h.13 Provide a summary description of the reactor vessel stress analysis which includes simple sketches showing the location and geometry of areas of discontinuity or stress concentration. Identify the controlling critical loading conditions.
RESPONSE
A summary of the results and location of points of reactor vessel stress analysis has been provided in response to Question h.k(f) . The location of structural discontinuity or stress concentration points has been pointed out on the sketch attached with the response. The critical loading condition or transient is the heat-up and cool-down transients. Other transients are of either short duration or small temperature change. ( J l L M 0134
)
h.13-1
D-B (Oh k.lk Provide a summag of the maximum intensities and cumulative o damage usage factors calculated for the steam generators and pressurized vessel acecmpanied by sketches illustrating the points of analysis.
RESPONSE
A summary of the maximum primary plus secondary stress intensity ranges and cumulative damage usage factors for the steam generator and pressurizer are presented in Tables h.1h-1 and h.lk-2, respectively. Figures h.1h-1 and 4.lk-2 show the points of analysis for the steam generetor and pressurizer, respectively. i l l 1 l l l l 1 l a 1 i
' 03.35
. . g l
u.lu 1 ! l
TABLE h.14-1 STEAM GEIERATOR STRESS RANGES AND USAGE FACTORS Pri. + Sec. Stress Usage Intensity Range Factor Support Skirt 121.5 Ksi* 0.96* Upper & Lover Tube Sheet 35.0 Ksi 0.13 Primarf Inlet Nozzle 18.0 Ksi 0.01 . Primary Outlet Nozzle 2h.0 Ksi 0.01 Steam Outlet Nozzle 27.0 Ksi 0.0 Auxiliarf Feedvater Nozzle kh.5 Ksi 0.0 OTSG Shell 25.5 Ksi 0.0 Feedvater Nozzle 50.T Ksi 0.56 s)
- These results are fro = an elastic-plastic analysis.
c e k.14-2 t
s TABLE 4.1h-2 PRESSURIZER STRESS RANGES AND USAGE FACTORS Primary Plus Secondary Stress Usage Intensity Range Factor External Supports (In Shell) 22.2 Ksi* Not Checked Shell h3.0 Ksi 0.0 I Surge Nozzle 27.0 Ksi 0.5h Spray Nozzle 63.0 Ksi 0.01 Heater Bundle Closure 31.0 Ksi 0.016 Heater Stud 66.0 Ksi 0.50
- Bijlaard Analysis 0137
([S W h.1h-3.
INLET N0ZZLE
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N g 4 SUPPORT SKIRT g a /, OUTLET N0ZZLE # l I lyf1 LF ANALYSIS
- PRIMARY PLUS SECONDARY _
4 PRIMARY PLUS SECONDARY AND FATIGUE POINTS OF STRESS ANALYSIS FOR STEAM GENERATOR F i gure 4.14 -1 , g 0138 l
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I SURGE N0ZZLE = - TYPE 0_F ANALYSIS 4- PRIMARY PLUS SECONDARY 4 PRIMARY PLUS SECONDARY POINTS OF STRESS ANALYSIS . AND FATIGUE F OR PRESSURIZER ,, g 0139 Figure 4.14 2
D-B p h.15 The PSAR states that the reactor coolant pump casings vill be designed and fabricated to meet the intent of ASME Code, Section III, Class A vessels as applicable. Outline the stress analysis procedures to be used for the pump casing, furnishing references as appropriate, and provide a summary of stress intensities and cumulative damage usage factors obtained. Specify any deviations from Code requirements other than lack of Code stamping.
RESPONSE
The reactor coolant pumps for Davis-Besse vill be designed and fabricated in accordance with the requirements for Class A vessels of the ASME Code, Section III,1968 Edition and Addenda through and including 'Jinter 1966 Addenda with the only exception being the omission of Code Stamping.- The procedure to be used in the stress analysis of the reactor coolant pump casing vill be con-sistent with those procedures indicated in the ASME Code, Section III. The loads imposed on the pump casing are non-axisymmetrical and the normal " methods for calculation of elastic stresses which handle symmetrical loadings on cylindrical vessels (Reference ASME Section III, Par. I-720) an not suitable for use in the stress analysis of the volute type pump casing. In recognition of this problem, a 3-dimensional finite element, matrix displacement computer code has been developed by the pump manufacturer which is capable of performing a stress analysis on the pump volute casing under non-axisymmetrical loading. The code derivations, to a large extent, follov those of J. S. Przemieniecki, in Theory of Matrix Structural Analysis, I i McGraw-Hill, New York, (1968). l l d Oh10 h.15-1
D-B f' h.16 With respect to formulation of the operating pressure-temperature i relationship limits insofar as a brittle fracture mode of failure, discuss the adequacy of using a stress concentration factor of k (page h-19) on assumed flaws in calculating stresses, cularly in the presence of a crack or cracklike defect (parti-either present initially or developing in service).
RESPONSE
The intent of Item (a) of Section h.3.1.1.3 (Page h-19) was to cover velds which cannot be radiographed where a flav was assumed in the analysis. Since all velds in the irradiated region can be completely radiographed and surface inspection is to an indication free criteria, Item (a) is not applicable to the reactor vessel. Accordingly, this item is deleted in - Amendment 3 to the PSAR. I \ g 0141 4.16-1 J
D-B l t i i, ' ,.m-j h.17 State the relief capacity of the spring-loaded safety valves in ' j the main steam lines.
- i
RESPONSE
j the capacity of the spring leaded safety valves in the main steam lines is
- 14.17h,922 lb/hr at 1155 psig. 6 1
1 1 5 i .. 1 4, l 4 I j i i t 1 i l w 0142 h.lT-1
D-B 7 5.0 General Structural Criteria and Design Loads 5.1 With regard to the design basis loads for consideration of containment stability, indicate the extent to which a combination that includes seismic loads but excludes accident pressure leads vill be considered.
RESPONSE
Par ~~rar. 6.1.2.1 of Appendix 5A of the PSAR indicates that the Shield Building vill be designed for normal operating loads in combination with the maximum possible earthquake. The Containment Vessel vill be designed for the maximum probable (smaller) earthquake in combination with each of the following: (1) Pressure test loads, (2) Leak test loads, and (3) Construction loads. In addition, the vessel vilJ le designed for the maximum probable (smaller) earthquake loads or the maximum possible (larger) earthquake loads in combination with operating conditions loads. . In addition the vessel vill be designed for the =aximus probable (smaller) l 6 earthquake loads or the maximum possible (larger) earthquake in combination
, with other loads but excepting the stabalizing effects of internal pressure, f
o 4 e u I Cll/l.1 i a 4 5 1-1 L
D-B 5.2 Specify the design criteria for the flexible closure of the space between a penetration and the shield building. Does this closure meet the single failure criterion? What design basis loads (earthquake, maximum accident pressure, tornado pressure differentials, etc.) is it designed to accomodate?
RESPONSE
The design criteria for the flexible closures of the personnel lock, emergency lock and the equipment hatch will allow for all temperature and pressure transients that could be experienced during the life of the station, the postulated tornado phenomena or the LOCA. The flexible closures will also accommodate the differential movements caused by either earthquakes and/or expansion during normal operating cycles or LOCA. The . closure vill meet the single failure criterion. The closure vill be ' designed for a temperature range of 30 F to 150 F, a pressure differential of + 3 psi batveen the annular space and the penetration, both acting concurrently with the maximum differential movements that could be experienced with either earthquake and/or LOCA. t 4 1 e
~ ~
03A4
.5.2-1
D-B i 5.3 With regard to reinforcement lap splicing, provide the minimum lapping in terms of bar diameter that vill be used, the splice stagger that vill be employed, and data to show that the splicing proposed vill be sufficient to develop reinforcement ultimate tensile strength especially for areas of high seismic tensile forces.
RESPONSE
The following is the Table 5 3.1 used for the minimus lap splicing. TABLE 5.3.1 TYPICAL REINFORCING SPLICES f'c = 4,000 or 5,000 psi ~ fy = 60,000 psi (U.S.D) Bar_ TOP BARS i OTHERS
# L. IN L. DIA. l L. IN. ' L. DIA.
3 17 h5 1T i h5 8 4 22 4h 18 36 ( 5 28 h5 23 37 6 33 kh 27 3o 7 h2 - h8 ' 30 3h l 8 55 55 40 ho 9 70 62 50 kh l 10 88 69 63 ! 50 11 108 77 TT ! 55 e l l 1 l l l
@ OM5 1
5.3-1
D-B l'^ 5.h The design criteria for containment penetrations are not as explicit as is desired. Provide the design criteria:
- a. to be applied for forces on the penetrations resulting from pipe rupture or relative displacement of internal or external systems,
- b. for design for isolation against vibrational loads.
RESPONSE
All penetrations connected to the Containment Vesse shall be designed in accordance with the ASME Boiler and Pressures Vessel Code, Section III for Class B Nuclear Vessels. The larger penetrations such as main steam and feedvater vill be anchored in the Shield Building vall. A flexible bellows type connection vill be . provided at the Containment Vessel shell thus allowing for all differential movements between the two structures. All significant loads including vibrational loads will be isolated from the Containment Vessel by the bellows. Penetration sleeves for smaller pipe vill be anchored in the Containment Vessel shell. These penetrations will be designed to take the maximum resisting moment, the maximum shear, and the maximum thrust which can be transcitted to the shell by plastic yielding of the sleeve. Bending and ( torsion vill be considered to act simultaneously with pipe thrust or shear force in accordance with the interaction formula shown below:
"B(design) MT Design MB (maxi um) M Maximum where MB = Bending moment Mt = Torsional moments l
Pipe thrusts vill be developed for the particular penetration service. : The Containment Vessel penetration vill then be designed in accordance j vith Bulletin No.107 of the Welding Research Council. Vibrational loads vill be treated in accordance with paragraph N-415 of the
, ASME Boiler and Pressure Vessel Code.
l l l l @ l OMG i 5.h-1
D-B
,s
( 5.5 With regard to a pipe break in the annular space between the containment and the shield building, provide:
- a. The maximum local pressure load that can be developed.
- b. The =aximum local stress that will result therefrom.
- c. The margin of safety with regard to steel shell stability from such a local load.
RESPONSE
Main steau and feedvater lines which pass through the annular space vill be provided with guard pipes attached to the lines which will direct the effects of the pipe rupture into the Containment Vessel, thereby protecting the annular space from the effects of large pipe rupture. Smaller pipe ruptures have been investigated and found to produce pressure build-up effects less severe than the criterion used to size the vacuum breakers. The Containment Vessel shell vill be designed for the combined effects o pressure buildups, jet impingement forces and pipe reaction forces acting concurrently caused by a double-ended failure of any single unshielded pipe passing through the annular space. The allowable local stresses in the Containment Vessel vill be within those allowed by the ASME Boiler and Pressure Vessel Code for Class B Nuclear ( Vessels. In the event that any pipe does not meet the above criteria, guard pipes will be provided to direct the effects of the pipe rupture into the Containment Vessel. 1 ( O OM7 5 5-1 w 1
E-B
/ 5.6 Provide a description of the protective coatings and paints to be used within the containment. Include the following items:
- a. Identification of =aterial to be used, location, and function.
- b. Physical and chemical characteristics.
- c. Performance under accident conditions including washdown, radiation, steam, temperature, and jet impingement effects.
RESPONSE
- a. Three coating systems are currently planned for use within the containment. .
The first system will consist of epoxy or modified phenolic coatings, such as A=ercoat No. 66 manufactured by the Amercoat Corporation, or Phenoline No. 305 finish manufactured by the Carbonline Company-or Val-Chem Hi-Build Epoxy manufactured by the Mobil Chemical Company, or approved equal coating. This system will be applied to concrete floors, walls and ceilings. The second system is an inorganic zine primer followed by an organic i topcoat, such as Dimetcote No. k or No. 6 primer followed by A=ercoat No. 66 epoxy topcoat as manufactured by Amercoat Corporation, or Carbo-Zine No. 11 primer followed by Phenoline No. 305 modified phenolic finish, as manufactured by the Carboline Company, or Mobilzinc 7 primer followed by Val-Chem Hi-Build epoxy, as manufactured by Mobil Chemical Company, or approved equal systems. This system will be applied to ferrous metal surfaces, such as structural steel, liner plate, piping, and equipment up to wainscot height above the floor levels in areas subject to hard usage or to contamination. The third system is an inorganic zine primer with an inorganic topcoat, such as Dimetcote No. h or No. 6 primer as manufactured by the Amercoat Corporation, or Carbo-Zine No. 11 primer as manufactured by the Carboline Company, or Mobilzinc 7 primer as manufactured by the Mobil Chemical Company, or approved equal. This system will be applied to the remaining ferrous metal surfaces, such as structural steel, liner plate, piping, and equipment above the wainscot height described in the preceding paragraph in areas not subject to hard usage or to contamination. m The function of the materials is to provide surfaces which will resist exposures due to both normal operating and LOCA conditions. Exposures include ionizing radiation, high temperature, impingement from sprays, and abrasion due to traffic. Q 01AS 5.6-1
D-B s
/
- b. Physical characteristics of the materials are as follows:
- 1. Application characteristics: Pot life, drying and recoating times, all at 70 F, and contents of solids are as follows:
Manufacturer Pot Life, Drying Recoating Content of and Material Hours Time, Time Solids per Hours Hours volume Amercoat 8 05 When color 62.2% Inorganic Primer changes to Dimetcote No. k blue-gray Amercoat 24 0.25 24 43.7% Inorganic Primer Dimetcote No. 6 emercoat 8 2.0 h 33 5% Epoxy Primer No. 66 Amercoat 8 4.0 6 56% Epoxy Topcoat No. 66 Carboline 8 8 to 12 8 to 12 80% Inorganic Primer Carbo-Zine No. 11 Carboline 1.5 18 18 72% Modified Phenolic Phenoline 305 Finish Mobil 8 0.5 72 43% Inorganic Primer Mobilzinc 7 Mobil 8 4 16 53% , Epoxy Val-Chem Hi-Build
- 2. Resistance to Ionizing Radiation: All materials listed above resist
- as systems accumulated gamma radiation of 3 x 109 rads from a Cobalt 60 source.
j dlll> flfg 5.6-2
D-B [
- 3. Mechanical resistance: Mechanical resictance of the materials is as follows:
Manufacturer Sward Abrasion Adhesion Impact and Material Hardness Resistance Resist.- Coefficient to Falling Sand ance Amercoat N.A. N.A. N.A. N.A. Inorganic Primer Dimetcote No. h Amercoat N.A. N.A. N.A. N.A. Inorganic Primer . Dimetcote No. 6 Amercoat N.A. N.A. N.A. N.A. Epoxy Primer No. 66 Amercoat N.A. 23.0 N.A. N.A. Epoxy Topcoat Liters / mil No. 66 Carboline N.A. 6.9 Excellent 130 inch-lbs. Inorganic Primer Liters / mil drops minimum Carbo-Zinc No. 11 Carboline 25 Over 10.0 Excellent 130 inch-lbs. Modified Phenolic Liters / mil drops minimum Phenoline 305 Finish Mobil over 350 Excellent 70 inch-lbs. Inorganic Primer 6H Liters / mil drops minimum Mobilr.inc 7 Mobil h8 to 56 80 Excellent 60 inch-lbs. Epoxy Val-Chem Liters / mil drops minimum Hi-Build
- 4. Fire Resistance: The materials listed are inflammable during application. After curing, they exhibit flame spread ratings as follows: -
Amercoat No. 66: 21.8 Carbolice Company: 39, approximately Mobil Chemical: 40, approximately
- 5. Resistance to high temperature: All material listed above resist without harmful effect, individually and as a system, intermittent temperature of up to 300 F. They are not designed for use at 5.6-3 O.1.50
l f' /\ D-B i kl ~ s
)
continuous temperatures in excess of 200 F. Chemical characteristics of the material are as follo..s: Manufacturer Material Binder Inhibitive Pigment Amercoat Inorganic Primer Inorgarte Zine Dimetcote No. k Silicate (Water-based) Inorganic Primer Inorganic Zine Dimetcote No. 6 Silicate . (Solvent-based) Epoxy Primer Epoxy - - No. 66 Polyamide Epoxy Topcoat Epoxy - ----- No. 66 Polyamide Carboline Inorganic Primer Inorganic Zinc 'h Carbo-Zine No.11 Silicate s/ (Solvent-based) Modified Phenolic Catalyzed ----- Phenoline 305 Modified Finish Phenolic Mobil Inorganic Primer Inorganic Zine Mobil-zinc 7 Silicate (Solvent-based) Epoxy Val-Chem Epoxy - Hi-Build Polyamide
- c. Materials were tested by Franklin Institute Research Laboratories, Philadelphia, Pennsylvania and ORNL under simulated LOCA conditions and performance was rated as follows:
Amercoat system: No visible attack Carboline system: No visible attack. Some loss of gloss. < (
Reference:
FRIL project 181-C-2697-01, November 1969). Amercoat System: "Some evidence of undercutting and some evidence of brittleness".
.J 03.51 5.6-4
D-B Carboline System: Some slignt undercutting around score mark; excellent on unscored surface. Mobil System: Not listed under " Defects Observed" and " Coating Looks Good". (
Reference:
ORNL Report for USASI Subcommittee N 101 5). Materials were also tested by ORNL for gamma radiation resistance and performance was rated as follows: A=ercoat 66 system: Excellent Phenoline 305 system: Fair (
Reference:
ORNL Report No. 3589 dated February 1965). m 0152 ( 5.6-5
D-B l ,e S 5.7 With regard to the shield building and connected penetration rooms, l provide:
- a. The manner in which negative pressure vill be assured in the shield building base slab area,
- b. The type of connection that is being used for the penetration room / shield building connection.
- c. A discussion of specific lines penetrating the steel contain-ment and shield building structures which vill not be enclosed within the containment system complex. Describe the manner in which potential leakage through these lines is assessed and factored into dose estimates for accident conditions.
RESPONSE
- a. Athough a negative pressure cannot be assured in the base slab area, the base 17 slab is a 3'-6" thick concrete slab and also the mean lov vater level which exerts a continuous hydrostatic pressure on the external surface is at eleva- 6 tion 568'-6". Under these conditions we can see no mechanism by which the airborne activity leaks to the enviroment through the base slab area.
- b. The penetration room to shield building connection vill allow for all dif-ferential building motion that can be anticipated during the life of the station. A one-inch water-stopped gap will be provided between buildings where basement or secondary containment boundaries interface with the J
\ shield building. The continuous split-type center bulb water stop will be designed to withstand 3 psi differential pressure and will provide both water-tight and air-tight connections between the structures, thereby main-taining secondary containment.
- c. Following a LOCA, the pipe penetrations which do not open into the contain-ment will not beccme the leakage paths for radioactive liquid or gas out of the containment. These penetrations are for closed systems which are also i Class I seismic inside the containment. This includes the following:
(a) Feedvater and Auxiliary Feedvater (b) Main Steam (c) Main Steam Sample (d) Steam Generator Drain and Sample (e) Service Water to the Containment Air Coolers (f) Component Cooling Water for R. C. Pumps, CRD Coolers, Quench Tank Cooler, Letdown Heat Exchangers, etc. (g) Fuel Transfer Tube (Blind-flanged) The penetrations which open into containment vessel during a LOCA can be classified as follows:
- 1. Penetrations for systems which are closed loops outside the contain-ment vessel. These vill not cause any ground release.
O (a) Demineralized Water 0153 (b) Letdown Line to Purification Demineralizer (c) Reactor Coolant System Make-Up. 57-1
D-B (d) Reactor Coolant System Drain to Reactor Coolant Drain Tank (e) Pressurizer Sample (f) Normal Containment Su=p Drain (g) Quench Tank Vent and Sample (h) Core Flooding Tank Sample and Fill (i) Nitrogen Supply (J) High and Lov Pressure Injection (k) Decay Hea't Cooling (1) Emergency Sump Recirculation (m) Service Air (n) CoMainment Spray (o) R. C. Pump Seal Return (p) Waste Gas Header
- 2. Penetrations for systems which are open outside the conttinment are the only possible in-line leakage paths through which radioactive liquid or gas could leak out of the containment and escape the 8 emergency ventilation system. Only this release can become I
a ground release. This includes the following: Line Size Leakage Rate ec/hr (a) Purge Air Inlet h8" 480 (b) Purge Air Exhaust h8" 480 , (c) Containment Air Samples 1/2" 5 (d) Equipment Hatch 1. , x 10 4 #) All the valves specified in Bechtel specifications will require seat leakage no't to exceed 10 cc/hr per inch of the diameter across the valve seat. This criteria is used in calculating possible in-line leakeage of isolation valves (lioted in 2) above.
~~
3 The in-line leakage paths which could by pass the emergency ven-tilation system and become an unfiltered ground release source, as given in (b) above, constitute an effective leak rate of only 1.05 x 10-6% containment volume per day. The O to 2 hour doses at the site boundary due to these unfiltered inline leakage paths are 0.0h rem thyroid and .053 m rem whole body. These dose basedon0to2hourgroundreleaseX/Qvalueof2.16x10garesec/m3 I The inline leakage rate and site boundary doses are insignificant in comparison with the LOCA leak rate and doses. The additional dose contribution from this unfiltered leakage does not change the doses described in Sec+1on lh of the PSAR. j a A positive hydrostatic seal vill be maintained by the ground water which has a minimum elevation of 568 feet. This hydrostatic seal acting to-gether with the continuous water proof membrane and the considerable thickness of concrete vill prevent airborne or waterborne leakage escap-ing from the Shield Building. T M 0154. 5.7-2
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g t! 1 8 f k})[JL'*L'? L,..~.., DAVIS-BESSE NUCLEAR POWER STATION GENERAL ARRANGEMENT SECTIONS C-C & D-D FIGURE 5.7-8 M 03.70
D-B l
-- 5.8 The liquefaction potential for the site and the margin of safety
(' against its occurrence under the specified seismic ground motion (s) has not been presented. Provide:
- a. Identification of those structures that will be located on foundation soils and, hence, subjected to potential displacement due to soil liquefaction.
- b. An evaluation of the stress or strain intensity for which seismic liquefaction is estimated to occur for the various soil strata at the site. Include the extent to which variability in relevant soil properties, foundation preparation, ground water variations, and structural loads has been censidered.
- c. The analytical procedures used to predict stress intensity under seismic loading, the preliminary results achieved and the margins of safety with respect thereto. -
RESPONSE
- a. The main building structures (containment v< ssel, shield building, auxiliary building and turbine building) are to be located on the sound rock. The presently known structures to be located outside the main station and on soil are shown on PSAR figure numbers 1-6, 1-7, 1-8, and 1-9. These structures include the start-up transformers, auxiliary transformer, main transformer, borated water storage tank, and primary water storage tank. The diesel oil storage tank is located outside the vicinity of the main station buildings adjacent to the railroad track.
The borated water storage tank and the diesel oil storage tank and pipe line are the only structures, presently planned outside the main Class I buildings, that are required for nuclear operaticn and safe shutdown; ilus they are the only structural foundations evaluated for liquefaction and other seismic foundation considerations. If other nuclear critical structures are placed outside as layout and design progress they too will be evaluated for the seismic foundation conditions. I
- b. A preliminary evaluation of the stress induced by the operating basis and design basis earthquakes has been made for the existing soil profile at the site and the proposed backfill areas. In the preliminary analysis we used the yard grade at 58h' and the top of rock at 560'.
An acceleration of 10% g was computed for the top of rock (560') using realistic soil damping parameters and a ground surface (58h') " , acceleration of 15% g. L The induced shear stresses in the soil at the top of rock was determined to be approximately 260 psf using the E-W Helena' October 31, 1935 accelerogram with the acceleration scaled to a peak value of 15% g. The shear stress values will vary essentially linearly from 0 psf at
, the grodnd surface (58h') to.approximately 260. nsf at top of rock (560').
M 0171
, 5.8-1
In yard areas where a liquefaction susceptible granular fill may be used a relative density of about 75% or t density of about 95% or 92% of maximum dry density as defined by tne Standard Proctor test
}'
(AASHO T-99; ASTM D 698) and Modified Proctor test (AASHO T-180; ASTM D 1557) respectively will be adequate at the top of rock where the earthquake induced stress is highest. These soil densities will have a higher margin of safety above the top of rock zone because the earthquake stress deminishes nearer to the surface. ; i In areas wherr the on-site clay soils (glaciolacustrine or till) are used in e_ ner an undisturbed or remolded and compacted state ) the liquefaction phenomenon cannot occur due to the moderate to high cohesive nature of the clay soil. Furthermore, the hypothetical-seismic ground motions will not induce any strength reduction of the site clay soils because the clay soils are insensitive. In areas where structure loads are present near the ground surface or where they are present and imposing a stress equal or higher than a ,, plain column of soil the margin of safety will be equal or greater than fill without the load at the top of rock zone. Under such structural conditions the induced earthquake shear stress will increase as the structural load increases; however, the resistance to liquefaction (increase in effective stress) will increase at a greater rate thus causing an equal or increased margin of safety.
- c. The analytical procedures used for these evaluations and to be used for any continuing studies are those developed by Dr. H. B. Seed and Dr. I. M. Idriss. A Bechtel computer program was developed by Dr. I. M. Idriss using the Seed /Idriss methods of analysis.
l The input data (damping, Poisson's ratio, modulus of elasticity, ! shear modulus and unit weight) for site soils, either remolded or l undisturbed, and proposed granular and clay fills came from the j evaluation of the site soils and rock. l e l g 0172 5.8-2' L
D-B 5.9 With regard.to the extent to which the structures of the facility may be subjected to differential settlement due to seismic ground motion, provide:
- a. The differential displacements predicted as the result of design basis ground motions.
- b. The safety margins that will be incorporated in the design of vital structures of the facility for accommoda-tion of such displacements.
RESPONSE
The main station structures (containment vessel, shield building, auxiliary building and turbine building) will be founded on rock and the outside accessory structures critical to nuclear operation (diesel fuel oil storage and borated water storage tank) will be located on in-situ or , compacted cohesive soils. There vill be no permanent uniform or differential settlement or displacement as a result of seismic ground motion. The structural motion due to seismic ground motion will be an elastic response. The structures may move during an earthquake relative to a fixed datum in space but they will return to their original position immediately following the termination of the earthquake motion. In the analysis of class I structures supported on soil, the mass models vill include springs representing the dynamic properties of the soil. 6, l I l i 03.73 Y ., , 5 9-1 i i t.__. w-
i D-B r 5.10 With regard to site slope stability under seismic ground motion, provide:
- a. The seismic ground motions that are being considered as input into the analyses.
- b. A sketch orasketches showing the extent of embankment and slope areas at the facility and their slope ratios.
- c. The criteria for the design. l
- d. A rietailed description of the design methods used, in-cluding an example calculation illustrating each pro-cedure used.
- e. A tabulation of results achieved for each vital soil -I l
structure to include factors of safety under the design seismic ground motions.
- f. Specific details concerning possible failure modes con-sidered, the extent to which pore pressure effects were considered in the analysis, and indication as to whether vertical earthquake excitation was included concurrently with lateral excitation.
[ RESPONSE Excavation and embankment slopes will be made in the yard area and in the vicinity of the intake canal and structure. The recently submitted g PSAR Amendment No. 1 and 2 discusses the intake canal's excavation and embankment which is to be designed and constructed as a Class I structure; see Section 5 9.8. Approximately 700 feet of the intake canal's excavated portion and dike in front of the intake structure will be ' designed and constructed as a Class I seismic structure. The other areas at this site where slopes will.be present are not Class I because their stability is not related to the nuclear safety of the station. The aforementioned Class I area at the intake structure is to provide water during an emergency shutdown. Within the Class I canal area limits, ' for a canal water level at elevation 560 (I.G.L.D.) approximately 7 7 l8 l million gallons will be available to the service water pump inlet. The 7.7 million gallons of stored water in the intake canal forebay will l8 j provide sufficient cooling surface to continue cooling the station by , evaporation for at least ik days. This will provide sufficient time to re-establish direct water flow communications between the lake and the station intake structure via the intake water system, if it is closed by seis- 8 mic or other eventsi The design incorporates a Class I return line from the service water system to the intake canal Class I area forebay. [ Four sketches are herein included to show the Class I intake a; Fig. 5 10-1, 2, 3 and h. A plan view of the entire intake "ater system and station areas 0 i g 01.74 l 5.10-1
D-B r is shown on PSAR Figure 1-2 and site and station profiles are shown on PSAR Figure 1-3 The general soils, foundation, geology and seismology are presented in Appendix 2C and PSAR Section 2. The seismic ground motion used for the analysis is based on the Helena, Montana 31 October 1935 earthquake scaled to a surface acceleration of 15% gravity for the operational basis earthquake. For the analysis of the intake slopes a static and pseudo-dynamic slope stability analysis similar to those methods described in references 1 through 5 have been used. The static analysis included slip circle and sliding vedge and analyses. The dynamic analyses include the slip circle and sliding vedge methods with the dynamic force applied as an extra static stress in the direction of slope instability. The design method of analysis of possible failures and the examination of existing failures of finite slopes in soils having cohesion, and in which there are no unusually weak strata or pockets, indicate that the failure occurs by circular rotation. The surface of failure is an elongated segment of a sphere. The plan of the failure shows that the surface is wider at the bottom of the slope than at the top, while the cross-section shows that the curvature is greater at both the top and bottom than in the middle of the failure sarface. Three forms of circular-type failure can take place. The most common is the toe failure which passes through the toe of the slope. This is most > likely to occur when the slopes are relatively steep or when the soil below the toe of the slope is strong. The base failure occurs when the slopes are flat or when the soil below the toe is relatively weak. The third form, the face or slope failure occurs only when there is a relatively weak zone in the upper part of the slope or where there is a very strong stratum at the toe of the slope or above that level. The general method of analysis is to approximate the actual failure surface by an are of a circle. Moments are developed about the circle center by the internal forces and external loads including the weight of the mass, the force exerted by water pressure on the face of the slope, water pressure in cracks, and any structures resting on the zone of rotation. The algebraic sum of these moments is the overturning moment tending to cause failure. The resisting moment is provided by the shearing resistance developed along the failure surface plus any resistance offered by piling of structures which extend across the failure surface. If any tension cracking occurs in the soil at the top of the slope the length of the failure arc which develops shear is correspondingly reduced. Water pressure in the soil acts normal to the failure surface. It does not develop any moment but it does influence that part of the shear strength which is caused by internal friction. The safety factor for the slope is the ratio of the resisting to the overturning moment. The most critical circle for failure is unknown unless the analysis is for a slope which actually failed. In determining the stability of a proposed ,,)
- y 0175 5.10-2
D-B
)
slope, such as a cut or an embarkment, it is therefore, necessary to try many different circles. The circle which yields the minimum safety factor is the most critical. Theoretically, it is necessary to try an infinite number of possible failure circles before the most critical one can be located; in practice a limited number of circles, selected on the basis of past experience, are usually sufficient. The basic circular analysis as described may be used for any soil or combination of soils where shearing resistance is independent of the normal pressure on the failure surface. In soils where the shear ta affected by the confining pressure, the effective stress on ead4 segment of the failure are must be determined in order to compute the shear strength. No exact analysis for this is available. Instead, various approximations are employed of which the method of slicca described below is the most widely used and is also the most reliable. The above described . analysis was used for the site slopes using the method of slices as a computational convenience for both the manual and computer calculations. The method of slices was developed to compute the nonnal pressure on the shear surface. The failure zone is divided into vertical slices. These slices need not be of equal width and for convenience in computation the boundaries should coincide with the intersections of the soil strata with the circle and with the slope face. The vertical force acting on each slide is the weight of the soil in the slice plus the weight of the water above the soil surface. If there is any external load, such as a structure, on the slice, its weight is also included. The overturning moment contributed by the slice is the product of the total weight and the moment arm from the circle center. The net or effective downward force acting on the curved bottom boundary of the slice is the total weight minus the upward force due to water pressure. If .he slice is sufficiently narrow, the curved boundary can be approximated by a straight line. The effective vertical force can, therefore, be resolved into a component parallel to the surface and one normal to it. The shear resistance due to internal friction along that section of the failure surface is the product of the effective normal force and the tangent of the angle of internal fricticn. The moments caused by the horizontal water pressure, pressure in cracks, and the shear resistance due to cohesion are computed as for the circular analysis and added algebraically to the overturning and resisting moments. If tension cracking should develop at the top of the slope, the shearing
- resistance along that part of the failure surface is zero. If water can get into the tension cracks an additional overturning moment is developed, which should be added to the overturning moments previously described.
, The method of slices described above is adapted to all combinations of soil strata,-friction, water pressure, and external loads. Numerous variations
.of this procedure are in use.
5 10-3
D-B A
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When the soil is non-homogeneous without a definite pattern for its variations, the failure surface tends to maintain the circular or near-circular cross-section. In such cases the circular method of analysis is employed. If the soil includes zones or planes of pronounced weakness, the failure surface is distorted so that a greater part of it will occur in the weaker zone and the smaller part in the stronger. If the weak zone is a plane, the failure surface vill flatten and follow it. The most common form of non-homogeneity consists of weak strata in the foundation. Ordinarily these are horizontal and not too thick. The failure occurs primarily by translation along the weak plane with a curved zone of shear at each end. In other words, the failure surface is a plane with curved ends or a very flattened circle. Two forms of failure are possible. The first may occur when the plane of weakness is in a soft clay or similar soil where strength is not influenced materially by the ~ normal pressure. The embankment moves horizontally slowly as a unit, pushing up a series of bulges beyond the toe. The top of the embankment drops and the surface of the embankment develops an irregular profile. The second type may occur when the plane of weakness is provided by cohesionless strata or seams whose strength has been reduced by water pressure. The source of the water pressure can be seepage through porous strata or fissures in the foundation. Because the confining pressure on the cohesionless seam is greater beneath the center of the embankment ~. than at the toe, the soil is weakest at the toe and stronger toward the ) center. At failure the outer part of the embankment moves suddenly as a unit leaving the central part unsupported in a steep slope. This drops into the gap so formed leaving a trough or lov area at the point where the embankment had been the highest. Mud waves also develop below the toe of the slope. This type of failure also occurs with an embankment resting on a smooth but pervious rock foundation, but without any mud waving. After the canal is filled, the first form of failure vill be limited to the outside face. The second type of failure could occur on either face; but it is more likely to occur on the outside under conditions of steady seepage and on the inside only when a sudden drawdown occurs. The linear failure is analyzed by considering the forces acting on the wedge-shaped zone. Active earth pressure acts on the high end while , passive earth pressure develops on the low end. Their difference is the force tending to cause movement. If there are seepage pressure within the embankment or the foundation these are to be computed. The earth pressures computed must be the effective pressures. The horizontal water preasures are added to them to obtain the total forces acting. The resistance to motion is provided by the shear or friction developed along the plane of sliding. This is computed from the soil shear strength, using the normal pressure produced by the weight of the embankment above. If water pressure is present on the plane of sliding, it must be subtracted 'T from the total to obtain the effective normal pressure for computing shear _s/ strength. g 0177 5 10-h
D-B As in the case of circular failures, the position of the boundaries of the zone of movement of failure are not known and must be found by trial. Ordinarily the most critical vertical boundaries are through the top of the slope and the toe, while the horizontal beundary is determined by a weak plane. If several planes of equal weakness are present the shallowest is often the most critical. . When the zone of weakness is fairly thick and is a clay where strength is a constant and independent of the confining pressure a somewhat different analysis applies. This is based on the plastic flow or squeezing of a layer of material sandwiched between more rigid horizontal boundaries. This type of failure, a plastic flow, is applicable when the thickness of the soft stratum is less than half of the base width of the embankment. For thicker soft strata the circular analysis must be employed. For the site foundation soils in the Class I intake area the plastic failure mode cannnt occur because the till clay is of low plasticity. The strength dats also shows that a plastic flov vill not occur. The foundation data for in-situ material is shown in Appendix 2C of the PSAR Volume I. A field vane shear investigation was made in November 1969 to determine the cohesion or shear strength of dikes presently existing at the site and at other locations in the Toledo area. The purpose of f this investigation was to obtain information concerning: (1) the in-situ ( shear strength and sensitivity of the glaciolacustrine deposit, and (2) the shear strength and sensitivity of the coil in dikes constructed of the glaciolacustrine deposit. Five sites were investigated: They were: (1) the Davis-Besse Nuclear Power Station site, (2) Carp fish pond No. 1 near Bono, Ohio, (3) Carp fish pond No. 2 near Bono, Ohio, (h) Bayshore Generating Station of the Toledo Edison Company and (5) Fermi Nuclear Power Station Unit No. 2 presently under construction. All the vane shear tests in this investigation were made with Acker vane shear test equipment. The equipment used consisted of: (1) standard A-rod sections, (2) a 2-ft section of 3-in. casing with a casing head and ball bearings, (3) a 2-in.-0.D. Acker vane, and (h) two torque vrenches with capacities of 0 to 200 in-lb and 0 to 600 in-lb. After the hole was augered to the required test depth, the 3-in. casing was placed in the top of the boring. Then, the vane was lowered down into the boring, the ball bearing was slipped over the top of the rod into the casing head, and the vane was pushed approx. 0 5 - 1 ft into the undisturbed soil. The torque wrench was attached, torque was applied by hand, and as the vane started rotating slowly, the maximum torque was recorded. Based - on the geometry of the vane the shear strength in k/ft.E.the A maximum torque vane shear in is test in-lb was converted illustrated in to Fig. 5 10-5 ' After the maximum torque was recorded, the vane v'as left in the ground for approximately 2 minutes , and a second vane shear test was made. The remolded 's. shear strength was obtained from this sec.ond, test. 0178 5.10-5
9 D-B J Vane shear tests were made in the borings at intervals varying from one to three feet. At the depth where there was a vane test made, a soil sample was taken. The samples were visually classified and water contents were determined. FIELD INVESTIGATION AND DATA
- 1. Proposed Davis-Besse Nuclear Power Station A total of 26 vane shear tests were made in eight (8) borings which were augered at the site of the proposed Davis-Besse Nuclear Power Station.
Nine tests were made in three borings which were augered a total of four to six feet in natural deposits, i.e., one to two feet of organic deposit and then terminating in the glaciolacustrine deposit. The remaining 17 tests were made in five borings which were augered a total of six to ten feet into four different existing dikes constructed of variable soil deposits including one boring made into a small dike constructed of the organic deposit. Two borings were augered into dikes constructed of the organic deposit and'containing a considerable amount of silt and fine sand. Two borings were angered into a dike constructed of the glaciolacustrine deposit overlying the in-situ organic and , glaciolacustrine deposits. ; It appears that dikes were constructed by draglining the adjacent in-situ material and dumping it in place. Little or no compaction and grading work was done. Based on available information three dikes were at least several years old and one dike was built in the Fall of 1969
- 2. Carp Fish Pond No. 1 Netr Bono, Ohio A total of four vane shear tests were mau' in one boring which auguered eight feet into a dike at carp pond No. 1. The dike was constructed by deepening an existing carp pond. The glaciolacustrine deposit was excavated in water with a dragline, and the clay was dumped in place onto the dike. No -compaction and grading work was done during or L= mediately after construction. The dike was approximately two months old (November 1969).
3 Caru Fish Pond No. 2 Near Bono, Ohio A total of four vane shear tests were uade in one boring which was augered six feet into a dike at carp pond No. 2. c According to available information, this carp pond was constructed during the first week of October 1969 The dikes of the pond were constructed in the following manner: (1) the top soil was stripped off with a bull-dozer, (2) the pond was excavated in the dry with a dragline, (3) the dikes were formed by dumping the clay in place, and (k) a bulldozer graded the slopes and compacted the different clay layers. The dike slopes varies from 1 5 to 2 hor. to 1 vert. s, )
. M 0179 5.10-6 s
D-B
- k. Bayshore Generating Station A total of nine vane shear tests were made in three borings which were augered 6.5 -T ft. into a dike on the north side of the intake canal at the Bayshore Generating Station.
The dike material is a gray sandy silty clay with some fine gravel and appears to have originated from a till deposit, and consequently all vane shear tests made in these borings do not apply to the subject investigation. Based on information provided by the Toledo Edison Company, the dike material was hydraulically dredged from the bottom of the intake canal and deposited on its north side; this construction was done in the Summer of 1953 5 Fermi Nuclear Pcver Station Unit No. 2 A total of eleven vane shear tests were made in four borings which were augered at the site of the Fermi Nuclear Power Station Unit No. 2 northeast of Monroe, Michigan. Two borings were made in the devatered marsh area. Three tests were made in one boring which was augered a total of four feet througn two feet of organic deposit and terminating in the upper glaciolacustrine deposit. One test was made in one boring which was augered one foot into the glaciolacustrine deposit (the existing top of the ground surface was in an excavated area 1 t five feet below the top of the organic deposit). Three tests were made in one boring which was augered six feet into a dike. The dike was built in the Summer of 1969 with soils . rom the organic and near surface glaciolacustrine deposits in the marsh area. A dragline excavated the organic and glaciolacustrine deposit s and dumped the 1 material in place. There was no compaction or grading work done. Four tests were made in one boring which was augered six feet into a second dike. This dike was built in the summer of 1969 with excavated glaciolacustrine deposit, a brown silty clay with some sand, obtained from a local borrow pit (rock quarry). For the construction of the second dike, road dump tru ks brought the soil from the borrow pit, and a bulldozer pushed the dumped material in place. The organic and near surface glaciolccustrine deposits in the marsh area vere excavated with draglines on mats. Track front-end loaders were used to excavate the remaining glaciolacustrine deposit and lead it on road dump trucks. It appeared that the trafficability of the glaciolacustrine deposit was such that it was impossible to operate road dump trucks directly < cn the surface of the glaciolacustrine deposit. The road dump trucks moved on a 6-8 in.-thich crushed stone roadway placed on the glaciolacustrine l deposit. i
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A ,ummary of all field vane shear test results is precented in the attachment l (. to this report on Table 5~.10-1. K 0180 5.10-7 j
D-B As previously stated, the purposes of this investigation were to obtain information concerning: (1) the in-situ shear strength and sensitivity of the glaciolacustrine deposit, and (2) the shear strength and sensitivity of the soil in dikes constructed of the glaciolacustrine deposit. Based on: (1.) the observed soil conditions and the apparent method of construction for the various dikes, and (2) the visual description:: and water contents of the soil samples taken, the field vane shear tests applicable to the subject investigation and proposed intake dike
- onstruction were determined and analyzed.
The table gives: (1) the minimum, maximum, and average in-situ shear strength of the glaciolacustrine deposit, and (2) the shear strength of the soil in dikes constructed of the glaciolacustrine deposit: The average in-situ shear strength for the upper six feet of the glaciolacustrine deposit varied from 1280 psf to 2580 psf. The average shear strength for the upper eig't feet of the soil in dikes constructed of the glaciolacustrine - deposit varied from 1180 psf to 25h0 psf. Sensitivity is calculated by dividing the initially determined shear strength by the remolded shear strength. The average sensitivity values are given in the table. They varied from 1.10 to 1.2h for the glaciolacustrine deposit, and from 1.16 to 1.kh for the soil in dikes constructed of the glaciolacustrine deposit. Based on the results of the field vane shear tests made in dikes constructed of the glaciolacustrine deposit, it appears that a minimum shear strength s of 500 psf in a dike constructed of the glaciolacustrine deposit is easily obtainable if the excavated soils are placed in layers and compacted at the preconstruction in-situ water content. We expect that the actual average shear strength will exceed 1000 psf. It is planned to make vane shear tests of the Class I dike during construction. Usage of till will mean even higher etrengths. Conclusion The Class I dikes will be constructed of soil excavated in the dry. The dike slopes are designed to be 3 (hor.): 1. Analysis of the stability of the dike slopes have been made under static and dynamic conditions. Even under the worst, hypothetical conditions, that is assuming that a rapid draw down of the. canal water to elevation 561' (I.G.L.D.) simultaneously occurs with the maximum possible (larger) earthquake, the factor of safety of the dike slopes was calculated to be 1 50 for a shear strength of the dike so'.1 equal to about 500 lb/ft.2 Based on an analysis of the site soils, the representative shear strength of the soils in the completed dikes will exceed this value. h , J 0181 5 10-8
D-B
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References:
- 1. " Stability of Earth and Rockfill Dams", U. S. Army Corps or Engineers, Engineering and Design Manual, EM 1110-2-1902, 27 Dec. 1960.
- 2. Casagrande, A., "No es on the Design of Earth Dams", Jcurnal of Boston Society of Civil Entineers, Vol. E , October 1930.
- 3. Newmark, N. M., " Effects of Earthquakes on Dams and Embankments",
Fifth Rankine Lecture, 1965, British Geotechnical Society, Published i in Geotechnical Society, London, England 1965
- h. Sherard, J. L., et. al., " Earth and Earth-Rock Dams", John Wiley & Sons, Inc.,
New York, 1963 .
- 5. Sovers, G. F., " Earth and Rockfill Dam Engineering", Asia Publishing House, New York 1962.
- 6. " Soil Stability Analysis", G. E. Information Systems, Time-Sharing Service, Program I.ibrary Users Guide, January 1968.
- 7. " Slope Stability Analysis", Bechtel Corporation Program CE 533, November 1966.
J
- 8. " SLOPE: Slope Stability Analysis System", ICES Computer Program, McDonnel Automation Co., St. Louis, Missouri, 1968.
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D-B oil Location Boring Shear Strength Shear Strength Sens[tivity h er Remarks e"crip- No. remolded Content (lon su sr av v min max av min max av av 2 2 2 2 k/ft k/ft k/ft k/ft k/ft2 k/ft2 % Davis-Besse(1) VT1 1.80 2.32 2.15 1.56 1.80 1.72 1.24 31 Intake Canal Davis-Besse(1) VT6 1.68 ? 10 2.24 1.54 2.32 1.87 1.19 31 flortheast Marsh Area l8 h4 Davis-Besse(1) VT16 2.06 3.10 2 58 1.54 2.06 1.80 1.2h 20 Station Area o f o p. Fermi (3) VT13 1.16 1.42 1.29 1.02 1.30 1.16 1.12 38 In dewatered marsh area M 6 fj Fermi (3) VT17 1.28 1.28 1.28 1.16 1.16 1.16 1.10 In excavation, 5 ft. below g.s. O g Davis-Besse(1) VTh 1.28 2.06 1.67 1.02 1.28 1.15 1.44 29 Several years old, g draglined o f , Davis-Besse(1) VT5 1.80 1.80 1.80 1.42 1.42 1.42 1.27 31 Several years old, draglined x
- N Carp Pond #1(2) VT8 1.34 2.70 2.36 1.80 2.32 2.03 1.16 24 Approx. 2 mos. old, g draglined w
h Carp Pond #2 ,VT15 0.h6 1.68 1.18 0.46 1.42 1.01 1.16 24 Approx. 3 wks. old, g draglined u f ,'j Fermi (3) VT14 2.20 3.10 2.54 1.80 2.32 2.02 1.25 20 Sev. mos. old, dumped in place o TABLE - 5.10-1 C pe RESULTS OF ANALYSIS OF FIELD VANE SHEAR TESTS M - 5.10-10 _ _ - - - -------- ------ - x
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INTAKE souuoouomuuuuoo< wL STRUCTURE ' TT TTTTTTTTTTTTTTTTTT'"'s O,c
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4 ' N APPROACH CHANNEL 540' f-INTA KE IN V ERT CLASS I AREA - 0+00 7+00 l WATER EL. VOL (GALS.) SUR AREA (SE) J 568.6' 19.530,000 200,000.__, 562.O' 10,200,000 173,500 56 0.O' 7,700,000 . 165,000 ___ Ft<,.Sao-2 (' ' 557.6' 4. 76 0,0 0 t. 155.000 ._ 5 55.U'. I , Sou,q00 144,20Q _ 03.85 554.O' _. 740,000 22,200 O 553.O' 582,000 19,700 54 8.U' 'f34u V 7,42O 5
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D-B l CINTAKE C DIKE , CANAL , C DIKE 19 9 19 9 _ 30'_ _3 0'_ l l FINISHED GRADE l EL. 5 8 3.O7 ! II CAN AL BOT T. ' EARTH FILL EL. 554.O', g E L. 5 50.9 200' _ ST A. 7+ 00 HOR. SCALE l=60' VEft SC ALE l*=3O' s 1 Fta. s.lo 4 o n.. . 1 Y ' l l
D-B 5.11 The description in the.P.SAR' indicates a possibility for relative motion between the various structures of the facility. Provide a listing of adjoining and interconnected structures, components, and systems where differential motion is of possible safety s consequence. Provide a description of the analysis methods through which possible motions at these locations vill be quantitatively evaluated and the design margins that vill be em-ployed with respect to the computed differential motions. With special regard to penetrations bridging between the containment and shield building, what differential motions are co=puted and what differential motions are tolerable without loss of function?
RESPONSE
The following is a list of adjoining and interconnected structures, components, and systems where differential motion is of possible safety consequence:
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Adjoining Structures Containment Vessel and Shield Building (co= mon monolithic foundation) Shield Building and Auxiliary Building (separate foundations) Interconnected Components and Systems Main Steam Lines Main Feedvater Lines Fuel Transfer Tubes Equipment Hatch
., Personnel Lock Emergency Lock Other Class I Piping All major Class I structures are supported on competent rock, thus eliminating differential movement due to settlement.
Differential displacements due to earthquake were analyzed using the response spectrum technique. Points of interest such as the main steam and feedwater lines, fuel transfer tubes, emergency and personnel locks, equipment hatch and the top of the Auxiliary Building were included as mass points in the c mathematical models of the various structures. The displacements at each point were calculated and the maximum differential displacements determined. A one inch expansion joint is provided between the Auxiliary Building valls, floors, and foundation and the Shield Building cylindrical vall and foundation. The expansion joint material has a compressibility of 90% which is sufficient to accommodate the maximum differential displacerents between the two buildings as shown in Figure 5.11-1. g 0188 5.11-1
D-B
'3 The main steam and feedwater lines and the fuel transfer tubes will be anchored in the Shield Building wall and will be provided with bellows expansion joints at the Containment Vessel shell permitting free movement axially and vertically. The bellows will allow sufficient movement due to the combined effects of earthquake and thermal displacements.
All other Class I piping which cross the interface between structures will have sufficient flexibility to allow for differential movements between structures and components. The equipment hatch, personnel lock, and emergency lock will be provided with expansion joints which will permit differential movements between-the Shield Building and the Containment Vessel. Figure 5 11-2 shows the maximum differential displacements between the Shield Building and Containment Vessel. . l l g 03B9 ) 5.11-2
DIFFERENTIAL DISPLACEMENT BETWEEN SHIELD BUILDING & AUXILLARY BUILDING ELEVATION FEET EL660'-O' i O.Ol 69' EL.654'- O / O.015 7' EL.643'-0" O.0134' EL623'-0" - O.0096' EL.603'-0" O.0059' EL.585'-O" O.0030' EL.565'-0" O.0002' I
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E L.54 5'-O" O. O' Maximum Possible (Larger) Earthquake -0.15g 03.90 M k. Figure 5.11-1
DIFFERENTIAL DISPL ACEMENT BETWEEN SHIELD BUILDING & CONTAINMENT VESSEL ELEVATION FEET EL.803tO" O.0582' E L.783'-O" O.0539' EL.7 56'- 6" O.0497'_ EL.74 8t O" O.0452' EL.720'-0" O.0384' EL 692' 0" O.0307'. E L.664'- 0" O,0229' EL.636'-6" O.0155'
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EL.6 09'-0" ! O.0087' E L.5 89'- 6" O.0045' E L,570 '- 9" O.0009' - El 565-O" f 0.O' Maximum: Possible (Larger) Earthquake -0.15g lt 0131 l Y Figure 5.11- 2
D-B 5.12 The design attention that is being given to vital interconnects and exterior piping runs due to static and seismic differential settlement and relative seismic motion is not described. Pro-vide:
- a. The design procedures and/or detailing that are being used to accomplish the design of interconnects.
- b. A typical example of how these design procedures and/or detailing are being accc=plished.
RESPONSE
- a. Piping conforming to ANSI B31 7 Class I will be designed with adequate flexibility to accept the movements,of connected equipment. Flexible connectors such as bellows joints will not be used in Class I piping.
Piping conforming to ANSI B31.7 Class II and Class III and ANSI B31.1.0 " vill also be designed with adequate flexibility, however flexible connectors vill be used where necessary.
- b. The analysis of a piping system to determine the effects of relative terminal movements will be done as part of the thermal expansion analysis. Flexibility calculations vill be made for each case of relative movements and the results vill be evaluated to determine the magnitudes of bending and torsional stresses at critical points.
Forces and moments acting on connected equipment and structure vill be evaluated for acceptability. l 1 l 1 i 03.92 l 5.12-1 3
D-B t 5.13 Class II structures, systems, and equipment are defined on page 5A-3 of the PSAR as those whose failure vould not result in the release of radioactivity and would not prevent safe reactor shutdown. Provide a listing of all Class II systems which may contain radioactive material and discuss how they meet the above criterion.
RESPONSE
The definition of Class II structures, systems, and equipment in the PSAR should have been stated "Those whose failure would not result in the uncontrolled release of radioactivity. . . , " etc., consistent with the definition for Class I. This has been corrected in Amendment No. 3 of the PSAR. One of the Class II systems containing radioactive material is the radvaste system; a discussion of how this system meets the above criterion is , contained in the answer to Question 11.2. Other Class II systems containing radioactive material are: the makeup and purification system (ref. answer to Question 5.1h), and the fuel pool cooling and purification system. All of these systems are contained within Class I structures, and liquid releases due to postulated component or piping failure would be contained within these structures. The makeup tank in the makeup and purification system contains the most significant inventory of gaseous radioactivity in these systems. The maximum inventory is estimated at 25,000 curies, which is a small fraction of the gaseous release analyzed in the answer to Question 11.2. d l g 019a J
- 1 5 13-1 1
D-3 m 5.lb Your definition of Class I structures, systems and components on page SA-2 of the PSAR states that certain portions of these systems may be composed of Class II components. Clarify your intention in this regard. Provide a list of Class II components of Class I systems.
RESPONSE
Systems which contain an interface between Seismic Class I and Class II design are shown on the figures included with the answer to question 1.3.
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5 15 Describe the' containment building vacuum breaker system and
\. provide the following information:
a.- Crit,erion used to establish the system's capacity l7 b.- A failure analysis of the system.
- c. Would the vacuum breaker system be adequate for a p pe break in the shield building annulus or penetration rooms?
RESPONSE
As stated in paragraph 5 2.1.3.h of the PSAR, the containment building vacuum breaker system consists of multiple vacuum relief devices which will assure that the external pressure differential on the containment vessel does not exceed 0.50 psi.
' a. The. system is. sized to provide the required relieving capacity with
-three containment air recirculating units operating under winter.
conditions although any two units will satisfy the requirements of 1 nomal station operation. The containment building net. air volume is approximately 2,800,000 ft 3. In the event of three coolers operating with .40 F_ inlet water temperature, the containment atmosphere temperature would decrisase from an initial temperature of 120 F (14.h psia) to a final temperature of 100 F (13 9 psia) in 315 seconds. To limit the' !
'. containment vessel differential pressure to 0.50 psi the vacuum breaker system capacity must be 1,270,000 ft3/hr. Ten (10) vacuum relief valves: including one (1) spare will be.provided set to open at .
0.25 psi external pressure differential on the containment vessel,
- b. Since each vacuum breaker will be pre.ided with an isolation valve, each vacuum relief device may be tested without impairing'the integrity of the system.
'c. As stated in the answer toLquestion 5 5, the vacuun breaker system will be adequate for a pipe break in the shield building annulus or penetration room.
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D-B
,,' 5.16 Discuss the criteria for design of the pressure vessel cavity shield valls. Include the following infornaticn:
- a. Discuss design provisions that vill be made to reflood the reactor pressure vessel cavity following a loss-of-coolant accident.
- b. Indicate any changes in design which result from the pressura vessel being supported from the inlet and outlet nozzles.
- c. Describe the maximum capability of the cavity to withstand a pressure transient and indicate largest rupture of primary system considered. Include a discussion of the loading conditions considered and the associated stress levels in the steel and concrete.
- d. Discuss the heat removal require =ents for the cavity shield vall.
RESPONSE
- a. Figure 5.16-1 shows the various floor drains at elevation 565' draining into the normal sump. The normal sump is in direct communication with the reactor cavity through an opening at the bottom of the reactor cavity. Following the LOCA, the normal sump will not be in operation, therefore, the water vill flood the reactor cavity through the opening.
There vill also be some communication through the in-core instrumentation tunnel. The drain piping and the opening to the reactor cavity will be sized to give the mini =al time lag. Should these drains be found inadequate, during the detailed design, additional drains or opening vill be provided.
- b. The arrangement of the primary system has always called for the reactor vessel to be supported from the nozzles. The support arrangement and cavity design vill be sinilar to that employed for other nozzle supported installations.
- c. The reactor vessel cavity will be designed for the temperature and pressure transients caused by a LOCA acting concurrently with earthquake and operating loads. A 14.1 square foot break in the primary coolant system vill be considered for design. Load combinations, load l8 facters and yield capacity reduction factors will be in accordance with paragraph 6.1, Class I Structures, Systems and Equipment, Appendix 5A of the-PSAR.
Preliminary analyses indicate a maximum pressure of 235 psig associated a with a lb.1 square foot break. It is anticipated that this pressure 8 vill be reduced through further design efforts. I i y ons 5.16-1
D-B
- d. The reactor cavity shield vall vill be cooled by the circulating air. q The forced air cooling provided by the containment coolers vill maintain J the shield valls at or below 150 F. The shield vall cooling coils vill be provided, if needed, to dissipate heat generated by the neutron and gamma flux. The design has not proceeded far enough to provide beat removal rcquirements at this time.
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+ D-B k' 5.17 Provide a listing of all valves which will be required to operate on a containment isolation signal and include the bases for establishing requirements for valve closure times. Include the steam line isolation valves and feedvater system valves.
RESPONSE
NUMBER OF NORMAL CIS CLOSING SERV!CE DESCRIPTION ISOLATION VALVES POSITION POSITION TIME
- 1. Closed Pressurizer Sample Lines 3 losed 30sec.l8
- 2. (DELETED) l8
- 3. Component Cooling Inlet 1 open Closed 60 sec.
- h. Component Cooling Outlet 1 Open Closed 60 sec. -
5 Containment Air Recire. 3 (1 on each line) Open Open --- Cooling Unit Water Outlet
- 6. Containment Air Recire. 3 (1 on each line) Open Open --- 8 Cooling Unit Inlet 7 Containment Air Samples 8 (2 on each line) Closed Closed 10sec.l8
- 8. Containment Normal Sump 2 Open Closed 30 sec. l8 Drain Line 9 Containment Emergency 2 (1 on each line) Closed Closed
- 60 sec. 8 Sump Recirculation Lines
- 10. Letdown Line to Purifi- 3 Open Closed cation Demineralizer l8
- 11. (DELETED) 8
- 12. (DELETED) 8 13 Steam Generator Secondary 2 (1 on each) Closed Closed 30 sec. g Water Sample Lines
- 14. High Pressure Injection h (1 on each line) Closed Open 10 sec.
Lines y 15 Containment Spray Lines 2 (1 on each line) Closed Open 10 sec.
- 16. Low Pressure Injection - 2 (1 on each line) Closed Open 10 sec.
]
l 17 Decay Heat Removal Pump 1 Closed Closed --- Suction 0 l
- opens during recirculation mode. hh 5 17-1
D-B eI NUMBER OF NORMAL CIS CLOSING SERVICE DESCRIPTION ISOLATION VALVES POSITION POSITION TIME
- 18. Reactor Coolant Sys. 2 Open Closed 60 sec.
8 Dr. Line to Reactor Coolant Dr. Tank 19 Contafnment Purge Inlet 2 C1csed Closed 10 sec.
- 20. Containment Purge 2 Closed Closed 10 sec.
Outlet
- 21. Pressurizer Quench 1 Open Closed 60 sec.
8 Tank Circulating Inlet Line Closed Closed 8 l 22. Service Air Supply Line 1 30 sec. 23 Core Flooding Tank 1 Closed Closed 60 sec. Sample Line 8 24. (DELETED) 25 Pressurizer Quench 1 Closed Closed 60 sec. Tank Nitrogen Supply s) J 8 26. Pressurizer Quench 1 Closed C1csed 60 sec. Tank Sample Line 8 l 27 Auxiliary FW Line 2 (1 on each line) Closed Closed ---- Main FW Line 2 (1 on each line) open Closed 8l28.
=
29 Main Steam Line 2 (1 on each line) open closed 5 sec.
- 30. Containment Equipment 2 open Closed 10 sec.
Vent Header
- 31. Normal Make-up to 1 open Closed 30 sec.
Reactor Coolant Sys-tem 8
- 32. Instrument Air Supply 1 Open Closed 30 sec.
Line 33 Pressurizer Quench 1 Open Closed 30 sec. Tank Circulating Out-let Line 3h. Waste Disposal to Misc. 2 Open Closed 30 sec. Waste Drain Tank 5 17-2 ,
D-B m \ NUMBER OF NCRMAL CIS CLOSIN" SERVICE DESCRIPTION ISOLATION VALVES POSITION POSITICN TIME 35 Reactor Ccolant Pump 1 Open C1csed 30 sec. Seal Wcter Supply O
- 36. Reactor Coolant Pump 5 Open Closed 30 sec.
Seal Water Return 37 Steam Generator 2 (1 on each line) Closed Closed 30 sec. Drain Containment isolation valves are provided in lines penetrating the containment vessel to assure that no uncentrolled release of radioactivity from the
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containment can occur, particularly following a radiation release type accident. Criteria for establishing cicsure times for normally open isolation valves are such that the requirements of containment integrity are met trior to peak containment pressure and temperature for the largest credible pipe rupture. The normally closed valves vill receive a closure signel to close them if they are open, otherwise the signal serves as a "make-sure signal". t c W 0202 5 17-3
D-B O 5.18 Describe the criteria that will be used in the collapse analysis of those Class II structures that enclose Class I equipment / systems or other Class I structures.
RESPONSE
The listed Class I Equipment / Systems or Structures below are protected from the collapse of enclosing Class II Structures in the fc11oving manner:
- a. The Class I service water pumps and piping system in the intake concrete substructure is belov both the top concrete slab and the Class II intak steel superstructure. The top concrete slab will be a minimum of 18 6 inches thick and vill be the protecting membrane against unlikely fail-ure of the Class II structure. I
- b. Class I service whter piping system in the Class II turbine building vill be enclosed in a Class I reinforced concrete pipe tunnel that vi. be completely surrounded by a granular compacted fill with a minimum top cover of four feet. The 8 inch concrete ground floor slab will bear on the compacted fill. The reinforced concrete unnel, k feet of compacted fill cover and the 8 inch concrete ground floor slab will protect the Class I piping against unlikely failure of the Class II turbine building l 7 superstructure.
- c. A small portion of the Class I reinforced concrete auxiliary building vill support the structural steel fr=ning for the heater bay of the Class II turbine building. The loads from this framing vill be incorpo-rated in the dynamic analysis for the Class I structures as Class I 6 applied loads. Multi-level steel floor framing, the elevated and ground floor concrete slabs in the heater bay, and the reinforced concrete aux-iliary building valls and slabs will protect the Class I structure from the unlikely failure of the Class II structure and/or equipment. l7 The minimum degree of separation between foundations of the auxiliary and turbine building are as following:
Slabs and column caps at and above the ground floor 1 inch Column foundations at the top of rock 12 inches
- d. The turbine building heater bay loads that transfer to the auxiliary building structure are considered as class-1 loads in idealizing the = ass model. The mass model vill consist of lumped masses at the point of turbine building support and at the auxiliary building floor supports.
No spring effects vill be considered for the leads from the relatively 7 flexible turbine building structure that transfer to the rigid aux-iliary building concrete superstructure. This portion of the turbine " building relics solely on the auxiliary building for lateral support. No additional steel crossbracing exists in this portion of the turbine building. _- 0203 5.18-1 _
-f D-B
, 5.19 Describe the instrumentation (such as strong-motien accelerographs and relative displace =ent =easuring devices) to be provided to assess potential damage in the event of streng earthquake ground motion.
RESPONSE
One strong motion accelerograph and three peak-recording accelerometers will be installed at the points which vill be indicated later. The exact type of instruments have not been selected. However, it is anticipated that they will be of the type, or similar to, the Teledyne Model RFT-250 Strong Motion Accelerograph and Mcdel FRA-100 Peak-Recording Accelerometers. C 0204 5 19-1 l
D-B m 5.20 Specify the containment design leak rate and conditions that will be met during plant preoperational testing. Various construction tests requiring leak rate determination at various stages should be identified as well as the acceptance criteria established for each test. The leakage specifications for penetrations and valves should be included in your response.
RESPONSE
The Containment design leak-rate is specified in paragraph 5.2.1.2 of che PSAR to be not more than 0 5 percent of the Containment free volume in 2h hours when tested at h0 psig. This is more than adequate to meet the guidelines of 10CFR100. The Shield Building leakage rates are specified in paragraph 5.2.2.2 of the PSAR for all elements of the secondary containment, the equipment doors and personnel doors. Tests which require leak rate determination at various construction stages are (1) Soap Bubble Test 7: (2) Halide Leak Detector Tests, (3) operational Tests of Locks, (h) Equipment Hatch Door Test (seals) and (5) Penetration and Valve Leak Tests. All construction ' tests of the Containment Vessel such as Soap Bubble or Halide Tests of the bottom head shall be conducted to no leak specifications. In the event that a leak is found in the shell, the leak shall be repaired and the test repeated until leaks are eliminated. Locks shall be shop pressure tested at 125% of design pressure and then leak tested at 40 psig pressure. The maximum allowable leak rate shall not exceed 0.10% of the Containment allowable leak rate over a 2k hour l8 period. Test connections shall be provided on each door to allow pressure testing of the seals at h0 psig pressure. The equipment hatch door seals will be leak tested et h0 psig. A pre-operational integrated leak rate test shall be made of the completed Vessel. The initial test shall be conducted at h0 psig following the completion of the soap bubble test. Pressure shall be maintained for whatever ~12e is required to demonstrate that the integrated leakage rate does not exceed 0 5 percent of the total contained weight of air at test pressure and ambient temperature. H l l s 0205 1 1 5.20-1 j
D-B -s 5.21 In sec+, ion 5.2.2 of the PSAR, it is stated that the pressure in the shield building cavity will be vented to the atmosphere in the event the pressure exceeds six inches of water. Provide a discussion and analysis to show the a=ount of leakage that would be vented following a design basis accident with subsequent heating of the shield cavity air volu=e.
RESPONSE
"Following a loss of coolant accident, heat transferred to the air in the annular space could cause a pressure rise. These temperature induced pressure transients will be limited to less than 6 inch H O2by venting the annular space".
This statement in Section 5.2.2, page 5-13 of the PSAR was erroneous and has been corrected in Arendment No. 2. The correct statement reads as follows: -
"The emergency ventilation system shall be designed to Itnit these temperature induced pressure transients to less than 6 inch H 0." l8 2 "The emergency ventilation system shall be designed to limit these l8 It passes through the filter system, as shown in Figure 6-T of PSAR, before being discharged to the vent stack. - 0.aos 5 21-1
D-B
<- m f 6.0 Engineered Safety Features 6.1 Identify the available and required net positive suction head (NPSH) for all emergency core cooling system pu=ps and spray pumps following a loss-of-coolant accident. De scribe the marmer in which the available NPSH vas calculated and provide a breakdevn of the factors and assumptions used to assure minimum NPSH is met even with containment pressure 3 psi belcw atmospheric precsure as stated in section 6 of the PSAR.
RESPONSE
The 3 psi pressure differential stated in section 6.1.2, page 6 h of the PSAR is an erroneous statement which has been corrected in Amendmut No. 3. The steel containment vessel cannot withstand 3 psi negative 'fressure. As discussed in section 5.2.1.3.h of the PSAR, Vacuum Breakers are provided to limit this negative pressure to less than 0 5 psi. Therefore, the NPSH calculations are based on a maximum negative pressure of 0 5 psi. The table below shows the required and available NPSH for the various safety fee w ?s pumps. Required NPSH Calculated Min. Available, Ft. Pumus Ft From BWST From CV Sumn Decay Heat 14.0* 67 17' H.P. Injection ** x 69 Cont. Vessel Spray 4 15.0** 67 15
- See Figure 6-3 PSAR
** Pump not selected at this time X If required in recirculation mode, the hip.I. Pump takes suction from the discharge of the decay heat pump.
The available NPSH,is calculated by the following general expression: NPSH a = Pa + ( E a -E)-Hf-N p Where Pa = Pressure above water surface, ft. (E, - E p) = Elevational difference between water surface and Pump center line, ft. Hg = Friction losses in suction piping, ft. - N = Vapor pressure at the pump inlet, ft. To be conservative in calculating the available NPSHa during the recirculation mode following the LOCA, it is assumed that Pa is equal to Pv. s p 0207 6.1-1
D-B C 6.2
.; With regard to the calculations of the sump coolant temperature following a loss-of-coolant accident, state the time after shut-down when (1) the heat removal capability of one decay heat ex-changer equals the reactor decay heat and (2) the capability of the two heat exchangers equals the decay heat.
RESPONSE
Following are the times when the decay heat removal capability equals the reactor decay heat: (1) 25,000 seconds (2) 4600 seconds 4 2 0208 6.2-1 ,
D-B 6.3 Indicate the maximum primary system leakage rate which could be accommodated by the normal charging system without initiation of the emergency core injection system. Relate this leakage rate to that resulting from the double-ended rupture of any small lines connecting to the pri=ary system.
RESPONSE
The high pressure injection and =akeup and purification systems have been changed and are now two separate systems which use two different sets of pu=ps. The makeup pump has not been selected at this time; therefore, the load capacity curve for the pump is not known, and this question cannot be answered at this time. This question vill be answered at a later date when the required pu=p characteristics are known. ( m A 0209 g 6.3-1
D-B 6.h To permit determination of the adequacy of isolation provided between the high pressure primary system and the 1cv pressure portiens of the emergency core cooling system, list those portiens of the ECCS which are not designed for reacter cperating pressure and temperature and indicate the means by which isolation is provided.
RESPONSE
Those portions of the ECCS which are not designed for the reactor coolant system design conditions of 2500 psig at 650 F are shown on attached Figure 6.4-1. c 0210 6.h-1
(~. A l l l DESIGN CON 0lil0NS j CORE 3050 PSIG @ 300F FLOODING ! 2500 PSIG @ 650F 2500 PSIG @ 300F v g LOW PRESSURE j e INJECTION
+ -
a:x ( = 4
@O HIGH PRESSURE INJECTION RC INI.ET PIPING EMERGENCY CORE COOLING INJECTION INJECTION PlPING DESIGN CONDITIONS Figure 6.4-1 O
0211
D-B (' 6.5 State your criteria regarding the sizing, design, and construction of the borated water storage tank. What concentration of boron vill be maintained in the borated water storage tank and fuel storage pool?
RESPONSE
The borated water storage tank (BWST) is designed and constructed to seismic Class I standard and AWWA D-100 Code. This tank is sized to meet the larger of the following two requirements:
- 1. Volume adequate to fill the refueling canal, incore instrumentation tank and portions of the steam generators and reactor coolant piping up to the refueling canal vater level.
- 2. To provide cufficient volume of borated water (including the volume in -
core flooding tanks) to flood the reactor cavity and the contain=ent floor so that adequate NPSH is available for the decay heat removal pumps through the emergency sump following a LOCA. The boron concentration in the BWST and the spent fuel pool 'ill be maintained at 1800 ppm. ( 0212 6.5-1
D-B l g ( 6.6 Provide the analysis which shows that the h psig containment pressure signal for ECCS initiation provides adequate backup for the 1500 psig primary system pressure signal for ECCS initiation for the spectrum of primary break sizes of 0.05 ft2 to 14.1 ft 2. Initiation by either of these diverse signals, assuming failure of the other signal, should provide the required ECCS capability. If the analysis of the effectiveness of the Davis-Besse safety injection design takes credit for a reactor trip, show how the initiation by only the high containment pressure signals meets this requirement.
RESPONSE
The analysis presented in the PSAR covered a spectrum of rupture sizes from 0.h ft2 to 14.1 ft2 (double-ended rupture of the 36" pipe) . For all of these caser, the actuation of the ECCS came as a result of a RCS pressure signal. Also, it was assumed in all of these cases that heat removal by the steam - generators ceased at the time of the rupture. This causes a delay in the time to reach the RCS pressura signal (1500 psig) which was used to actuate the HPI system. In all cases, the h psig containment vessel pressure signal occurs either before the 1500 psig RCS pressure or within one second cf that signal as can be seen from the table below. Therefore, either of the signals provides the required signal for ECCS actuation. Table 6.6-1 Time to ECCS Actuation vs Ereak Size 2 Time to 1500 psig Time to h psig Break Size, ft RCS Press., See C.V. Prest 'e c 1h.1 <1 <1 8.5 <1 <1 5.0 1.6 1 3.0 h.0 1.8 1.0 2h.0 5.0
.h 52.0 % 13.0 The instrumentation used to provide reactor tly has been tested in an environ-ment that could result should a loss-of-coolant accident occur and has been shown to function as specified. The reliability of the low reactor coolant pressure signal makes a containment vessel pressure signal unnecessary.
a The subject of co= mon mode failures and particularly the reactor trip following the LOCA is continuing to receive a detailed evaluation to determine if diverse reactor protection signals are required. If diversity or some similar form of added protection is found to be necessary for some particular class of abnormal transients as a result of this generic analysis, the need for reactor trip following loss-of-coolant accidents vill be reassessed and resolved pricr to the operating license review. 6.6-1 0213
D-B . r' t, 6.7 Provide the annlysis and discuss the results which support the reduction from 1800 psig to 1500 psig for the primary system pressure ECCS initiation signal. Include a discussien to show the effects of this pressure reduction for primary system rupture sizes frem 0.05 ft2 to 1h.1 ft2
RESPONSE
As stated in the answer to Questien 6.6, the spectrum of breaks analysis was performed frem 0.h ft2 to 1h.1 ft2 If no credit is taken for the containmen' vessel signal for actuation of the high pressure injection system, then a delay of as much as 50 seconds (0.h ft2 break) occurs between the time to 1800 psig and the time to 1500 psig. In this pressure range and using only 1 high pressure injection pu=p; which was used in the analysis, S 30 ft3 of injection water could have been punped in during the time interval of 50 sec. This volume of water is censidered to be within the accuracy of the calculations since this volume is N 1/h% of the RCS volu=e. Furthermore, if credit had been taken for heat removal by the steam generators, the time difference between 1500 and 1800 would have been even less. I d ( g214 s_. 6.7-1
D-B 6.8 Provide the design bases for the shield building emergency ventilation system. Include the following infor:ation:
- a. Expected conditions (such as temperature and humidity) of the air in the annulus and penetration rooms.
- b. Capacity and type of activated charcoal filters.
- c. Consideration to prevent possibility of fire.
- d. Failure analysis of the system.
RESPONSE
- a. Under normal operating conditions, the temperature of the air in the annular space vill be above outside ambient temperature, varying from l 6 about 80 F vith outside temperature 0 F, to about 11h F vith outside temperature 95 F. The heating of the air within the annular space and penetration rooms will reduce the relative humidity below that of the outside air introduced whenever the penetration and annular spaces are purged. Regardless of ambient conditions, the relative humidity within the space vill be less than 70%. In the event of a LOCA, the temperature within the annular space vill rise rapidly, further reduc-ing the relative humidity.
( b. The charcoal filter bank vill be designed on the basis of 333.3 CFM (nominal flow) per element. The filter beds vi l be 2" in depth and have an air flow velocity of approximately h2 't/ min. Impregnated charcoal (activated coconut shell) vill be useu in filters.
- c. Based upon the leak rate of the containment vessel limited to 0.5 percent per day, it is not considered credible that, during any single anticipated run of the emergency system, there could accumulate sufficient radioactive its ignition temperature of 680 F. material to raise the charcoal to
- d. emergency ventilation system is redundant with respect to all components including ductwork up to modulating va]ve as shown l7 in Figure 6-7 of PSAR. Failure of one system to operate as required vill automatically place redundant system in operation. Both systems l 7 shall be seismic Class I, up to the modulating valves.
M ( - 0215 6.8-1
D-B (~ 6.9 The =akeup tank is available to =akeup-injection pumps during emergency conditions. De=enstrate that the cover gas in . 2 makeup tank cannot override the berated water head and feed gas into the =akeup pu=p suction.
RESPONSE
This situation does not exist in the Davis-Besse design because the design has been changed to provide separate =akeup and high pressure injection systems. It is now physically i=possible to feed makeup tank cover gas into the high pressure coolant injection pu=p suction. C E . 0216 1 6.9-1 i i v.
D-B
, 7.0 Instrumentation, Control and Power 7.1 In regard to the protection syetens which actuate reactor trip and engineered safety feature action, the follcwing information is requested:
- a. A listing of those systems designed and built by Babcock l7 and Wilcox that are identical to those of the Three Mile Island Unit #2 (as. documented in the PSAR); discuss any design differences;
- b. Identification of the supplier for those systems that are designed and/or built by suppliers other than Babcock and Wilcox, and
- c. Identification of those features of the design which differ from the criteria of IEEE 279 and the Commission's General Design Criteria. Explain the reasons for any ,
differences.
RESPONSE
I 6
- a. The reactor protection system, consisting of a Nuclear Instrumentation i Reactor Protection System and a Control Red Drive System, is supplied by B&W. The Nuclear Instrumentation Reactor Protection System and the Control Rod Drive System supplied by B&W for Toledo Edison vill be sLnilar to those supplied for Three Mile Island Unit #2. Detail design of these systems may result in small changes, but this is not known at this time,
- b. The safety actuation system is being designed by Bechtel. The suppliers for the various components in this system have not been selected at this time.
- c. None 0u7 7 1-1 L
D-B 7.2 In regard to the Babcock and Wilcox designed control systems, the following information is requested:
- a. Identification of the major plant control systems (e.g.,
primary temperature control, primary water level control, steam generator water level control) which are identical to those in the Three Mile Island Unit #2; and
- b. A listing and a discussion of the design differences in those systems not identical to those used in the Three Mile Island Unit #2. This discussion should include an evaluation of the safety significance of each design change.
RESPONSE
The major plant control systems (e.g. , primar/ temperature centrol, primary
- vater level control, steem generator water level control) are similar to those in Three Mile Island Unit #2. Detail design of these syste=s may result in small ca._;s, but this _s not kncun at this time.
I l 1 1 H 0218 7.2-1
D-B 7.3 What are your seismic design bases for the reactor protection l system (as described in Section 1 of IEEE 279) and the emergency electric power system? Will the systems be designed to be capable of actuating reactor trip or engineered safety feature action during the maximum peak acceleration? If a seismic disturbance occurred after a major accident, would emergency core cooling be interrupted? What tests and analyses vill be performed to assure that the seismic design bases are met? What seismic specifications are included in the instrumentation and control purchase orders?
RESPONSE
The reactor protective system and the emergency electric power system are designed as Class I systems as defined in PSAR Amendment No. 3. The systems are designed to function before, during or after the maximum peak acceleration. A seismic disturbance following a major accident would not interrupt emergency core cooling. Purchase specifications for instrumentation and control components will contain requirements for submission of test data, operating experience or calculations which substantiate that the components will not suffer loss of function under design seismic loadings due to the maximum hypothetical accelerations determined by the seismic analysis of the structure.
~
oais 7 3-1
D-B T.h Describe the quality centrol procedures which apply to the equipment in the reactor protection syste= (as described in Section 1 of IEEE 279) and the emergency electric power system. This description should include, but not necessarily be limited to: (a) quality centrol precedures used during equipment fabricatien, shipment, field storage, field installation, and system component checkout; and (b) records pertaining to (a) above.
RESPONSE
All safety and protection equipment will be subjected to quality assurance lT programs which include the folleving quality control methods and procedures:
- a. Control of the Quality of Purchased Parts - From inspection by the manu-facturer through receipt inspection and testing at the site,
~
- b. Control of Inspection and Test Equipment - Including calibration standards and recalibration frequency.
- c. Control of Packaging and Shipping - Including final inspection releases, inspection of packages, and maintenance of cleanliness.
- c. Control of Changes to Documents Affecting Quality - Including drawings, specifications , procedures, and other related procurement documents,
- e. Disposition of Non-Conforming Items - Including repair, rework, and retest procedures and the steps taken to assure that non-cenfor=ing items are pcsitively identified and not inadvertently used.
- f. Control and Storage of Acceptance Inspection and Test Records - Including their immediate availability for review of Toledo Edison and its agents.
- g. Site Receiving Inspection - Including the procedures for site receiving inspection, identification and storage.
- h. Erection and Installation Inspection - Including a system for verifying the quality of erection and incta11ation work.
- i. Field Operations Audits - Including procedures f:r conducting audits of field operatione to verify their adherence to approved procedures and processes, j l
a 0220 7.h-1
D-B 7.5 Submit your cable installation design criteria for preserving the independence of redundant reactor protection system and i engineered safety feature circuits (instrumentation, control and power) . For the purpose of esble installation, the protection system circuits should be interpreted in their broadest. sense to include sensors, signal cables, control esbles, power cables (both a.c. and d.c. ), and the actuated devices (e.g. , breakers , valves,pu=ps).
- a. Cable separation should be considered in terms of space and/or physical barriers between redundant cables. Please address (1) the separation of power cables from those used for control and instrumentation, (2) the intermixing of control and instrument cables within a tray (or conduit, ladder, etc. ), (3) the intermixing within a tray of cables for different protection channels, and (h) the intermixing of non-vital cables with protection system cables.
- b. What are your criteria with respect to (1) the separation of penetration areas, .?) the grouping of penetrations in each area, and (3) the separation of penetrations which are mutually redundant?
- c. Discuss cable tray loading, insulation, derating, and overload protection for the various categories of cables.
- d. Discuss your criteria with respect to fire stops, pro-tection of cables in hostile environments, temperature monitoring of cables , fire detection, and cable wire-way markings.
- e. Discuss the administrative responsibility for, and control over, all of the foregoing (a-d) during design and installation.
- f. Discuss your design criteria for locating the process instrumentation inside contain=ent to include (1) separation of redundant sensors and sensing lines, (2) protection provided to sensors and sensing lines, and (3) protection provided to cables between sensors and electrical penetrations.
RESPONSE
- s. In general the separation of redundant cables of the reactor protection system and engineered safety features system circuits will be accomplished by spatial separation where it is practical or by physical barriers. J Generally the separation vill not be less than four (h) feet vertical and 18 inches horizontal.
If, in areas of natural convergence of cables, these minimum separation distances cannot be maintained, physical barriers will be provided 0221 7.3-1 -
D-B between redundant cables. These minimums vill not necessarily apply to embedded conduits. However, they will apply at the points where embedded conduits leave the material in which they are embedded. Wherever practical equip =ent associated with redundant channels will be located in separate cabinets. Whenever this is not practical, a minimum of 12, inches of separation vill be provided inside the cabinete or physical barriers vill be installed. In addition, the following specific criteria vill be applied:
- 1. Separate tray and/or conduit systems vill be furnished for all classes of cable; 15 kv, 5 kv, 600 volt power, 600 volt unshielded control and shielded instrument cable. Wherever possible, power cables will be run in the top trays; control and shielded instrumentation cables will be run in lover trays.
- 2. Shielded instrumentation and unshielded control cables will not be '
in the same wirevay. 3 Different parameter signal cables may run in the same wirevay as long as they do not belong to separate redundant channels.
- 4. Vital and non-vital cables may be mixed. However, both the vital and non-vital cables vill be of the sa=e type and have similar overload and short circuit protective features. This will ensure that the non-vital cables will not 'eopardize the integrity of the vital cables.
- b. 1. Four separate rooms, two on the east and two on the vest side of the containment vill be provided for electrical penetrations. The two rooms on each side vill be on different floors. Wherever possible, redundant circuits vill be run through separate penetration rooms. The center to center spacing betweea penetrations at the outside of the containment will be h8 inches horizontally and 60 inches vertically. This arrangement vill give a minimum distance of approximately h0 inches vertically and horizontal 1; between the closest conductors of two adjacent penetrations.
- 2. Separate penetration canisters vill ' provided for power, control
& and int.;rument cables and in all can :s for redundant safety feature
- 3. equipment. In every case, whenever there are two redundant systems in a penetration room, such systems will be located as far apart as possible. No barriers will be provided between penetrations or sections rf penetrations in a penetration room.
- c. All power cables in the sa=e ladder tray, which are rated for continuous ,
service, will have as a minimum a separation of 1/h of the diameter of of larger of the two cables. All electrical conductor insulation vill be flame resisting material in accordance with IPCEA standards. Except for vital circuits within the , containment, all insulation vill be either oil base rubber, ethylene propylene rubber, Kerite type HT or cross-linked polyethylene. Vital
- 0222
~
7.5-2
D-B p- s circuits within the containment that must be operable under accident conditions will be high temperature Kerite, silicone rubber covered with asbestos braid or mica insulated covered with a fibreglass braid. Power and control cables for services required to operate for extended periods after a loss of coolant accident will be suitable for operation under the conditions of high ambient temperature, radiation, pressure and humidity prevailing after an accident. Base ampacity rating of cables will be established as published in IPCEA P-46-k26 and in accordance with the manufacturer's standards. To this basic rating, a grouping derating factor also, in accordance with IPCEA-k6-h26 will be applied. Credit will be taken for load diversity factor where applicable. However, motor feeder power cables will be selected on the basis of 100% load factor and continuously rated for 125% of the motor full load current. - All power and control circuits will have circuit protection provided by breakers except for some control circuits, which will have fuses. In accordance with normal practice on three phase circuits, all three phases will be simultaneously interrupted by the breakers, and on a single phase grounded circuits, only the hot side will be interrupted.
- d. All classes of cab,le except power cable will be run in closed trays
& when not in embedded conduit or exposed rigid metal conduit. Whenever
- f. a tray passes through a floor or fire wall, a fire stop will be provided to prevent the spread of fire through the wall or floor.
Power cables not in embedded conduit will be in rigid metal conduit or be armored cable in ladder type trays. Cables for reactor protection systems and engineered safety feature equipment will be installed in Class I structures. Outside the shield building, cables will be installed in either trays or condaits. Redundant wirevays will be separated or have barriers between them to protect against common failure from a single event. Temperature monitoring of cables will not be provided. All cables (except lighting, receptables, and small power) will be properly tagged at their terminations with an identifying number. Coding to identify redundant channels , (e.,g. , a color band or marking), will be applied at the terminations and at appropriate intervals along the length of safety related raceways,
- e. The cable installation is on the Q-List and as such will be subject to the quality assurance procedures to insure that the design criteria is #
met through all stages of design, procurement and installation. 0223 M T.5-3
~
D-B S 7.6 What are your design criteria for reactor protection system and engineered safety feature related electrical and mechanical equipment located in the containment or elsewhere in the plant which take into account the potential effects of radiation on these components due to either normal operation or accident conditions (superimposed on long-term normal operation)? Describe the analysis and testing performed to verify compliance with these design criteria.
RESPONSE
Prototypes of Reactor Protection System and Engineering Safety Features related transmitters required for use within the containment following a LOCA or steam line break have been tested under conditions simulating an accident radiation environ =ent superimposed on the normal operating radiation , environment. The design criteria established maximum allowable output signal errors of + 12 percent and + 10 percent for the pressure and differential pressure transmitters respectively. , A three (3) phase test was performed to simulate the post-accident reactor contain=ent building environment. Each phase is described belev: Phase I - Irradiation test at the Babcock & Wilcox Nuclear Development Center (NDC) to simulate the environmental dose to the transmitters associated with the 40-year plant deuign lifetime. Phase I consisted of irradiating the transmitters while the units were in a nonoperating mode. The transmitters were placed in a sealed aluminum box with two dosimeters e.ttachment to each transmitter and positioned over two reactor fuel elements in the NDC storage pool. Phase II conservatively simulated tLe reactor containment post-accident pressure and te=perature environment. This phase vill be discussed in more detail in the answer to 7 7 Phase II_I_- Irradiation test at NDC to simulate the maximum expected dese to the transmitters associated with a LOCA. Phase III consisted of irradiating the transmitters while the units were in the operating mode in ; much the same manner as Phase I except that the box was lowered into position beside one fuel element from the reactor. l A constant input of l approximately 2/3 of full range was maintained throughout the test. l The design criteria for all electrical control cable and safety related equipment power cable are that the cable shall not fail when subjected to associated accident conditions after the long-term normal operating conditions. Cables to be used in high radiation fields are of the cross-linked polyethylene or ethelene propylene insulation which has been subjected to the following tests:
- 1. Rated current and voltage concurrent with 130 C temperature and 100%
humidity. ' / 7.6-1
D-B 76
- 2. Irradiated at twice the DBA plus expected 40 year integrated dose concurrent with 50 C temperature.
All other material and equipment associated with safety related systems have been specified for construction and materials to be suitable for the appropriate h0 year integrated dose plus a DBA dose. 6 One feature which should be considered for evaluation of long-term effects is the requirement for periodic tests and calibration of the subject equipment. These tests will allow detection of gradual equipment deterioration and will assure capability of the system to operate as required by the original design basis since the interval between such tests vill be short compared to the time required for significant deterioration. , t e c (32: set; 3 s 7 6-2 i - -_
D-B
- s T.T Identify all equipment and components (e.g. , motors, cables, 1
~f ilters, pump seals) located in the primary containment which are reqr' red to be operable during and subsequent to a loss-of-coolant ur a steam-line break accident. Describe the qualifi-cations tests which have been or vill be performed on each of these items to insure their availability in a combined high temp-erature, pressure and humidity environment.
RESPONSE
Electrical and mechanical equyment within the containment vessel which are required to be operable during and subsequent to a F-3 and/or a steam line break are:
- 1. Reactor coolant system pressure transraitters.
- 2. Pressurizer level transmitters. ,
- 3. Steam generator level transmitters.
- h. Some containment vessel isolation valves and associated drives and remote position indication.
5 Containment air cooling unit fans and cooling coils. 6
- 6. Instrument cables for radiation, pressure, level and valve position instruments.
T. Power cables for the containment air cooling fan motors. Prototypes of the transmitters in items 1-3 have been tested successfully as Phase II of the test. referenced in answer to question T.6. Phase II was an environmental +est at Franklin Institute and consisted of exposing transmitters in the operating mode to the steam environment in a test autoclave for 2h hours. The temperature and pressure test environment is shown in Figure T.T-1. The test maintained an adequate margin above the environmented conditions predicated for any rupture. The units were supplied with a constant input of approximately 2/3 of full range and the resultant output / input ratio was measured for the test duration. The resulting output signal errors were analyzed and found to be acceptable. All essential electrical equipment will be specified for operation in the 6 post accident environment. Specifically the motors of the air recirculation units vill be specified to be operable in the high pressure, temperature and humidity environment equivalent to post accident containment conditions for a period of time greater than that necessary to reduce containment
- pressure and temperature to normal operating conditions. The vendor vill 6 be required to submit data which shows that the equipment meets these l requirements.
I The cables associated with the cooling units and instrument cables will also be specified for operation in post accident environment and prototype test results will be required for verification of this capacity. 6 T.7-1 0226
^
60 300 50 , 250 Reactor Buiiding Temperature 40 , 200 Reactor Suilding Pressure E 30 150 4 . B 3 2 4 h r ---+ 2
,y 20 100
's 10 , , 50 0
0 1 milJ 104 105 100 10 102 Time, sec N RE ACTOR BUILDING ENVIROMENT A: PRESSURE AND TEMPERATURE TEST ENVELOPE Figure 7.7-1
l l D-B ( 7.8 What criteria have you established relative to assuring that loss of the air conditioning and/or ventilation system will not adversely affect operability of safety related control and elec-trical equipment located in the control room and other equipment rocms? Describe the analysis performed to identify the worst case environment (e.g. , temperature, humidity) . What is the limit-ing condition with regard to temperature that would require reactor shutdown, and how was this determined? Describe any testing (factory and/cr onsite) which has been or will be per-formed to confirm satisfactory operability of control and elee-trical equipment under post-accident environmental conditions.
RESPONSE
- a. Redundant refrigerating units are provided for the control room air conditioning system. When cooling is required, one unit will be operating and the other unit is available for manual actuation in the event of failure of the operating unit. .
The ventilation system fans are redundant seismic Class I. Failure is not considered credible. Should a failure of both refrigerating units occur during warm weather and while plant is operating normally, conditions within the control room could be maintained close to outside ambient, since the ventilation system will have 100% outside air capability. In the unlikely event that both refrigerating units fail following an accident such that outside air may not be used for ventilation, temperature within control room will rise very gradually due to the large heat storage capacity of concrete walls, floor, and ceiling. The reactor protective system furnished by B&W is designed for continuous operation in a room environment of 40 F to 110 F and up to 75% relative humidity. Since the reactor protective system will perform its design functions within seconds, and since the control room will not reach 110 F before substantial time has elapsed, it follows that the reactor protection system is not affected by loss of air-conditioning and/or ventilation as postulated. All portions of the safety features systems (not including reactor protective system) that are located in the control room will be designed to operate in ambient conditions of 110 F and 80% relative humidity. The maximum temperature that the control area vill reach is estimated to be about 130 F and this temperature will not be reached for at least h8 hours. Sensible heating of the space will reduce the relative humidity below outside ambient.
- b. The worst case environment is based on the following assumptions:
(1) outside summer design conditions of 95 F DB and 76 F WB, (2) a temperature of 110 F in the surrounding auxiliary building, and (3) an internal heat load consisting of control equipment, emergency lighting, 7.8-1 0228 ! l
D-B r and ventilation fans in recirculation mode.
- c. If, for any reason, the control room temperature should reach 110 F and continue to rise, during normal station operation, administrative procedures vill require a station shutdown.
No special tests, except for those described in answer to questions T.6 and 7 7 have been planned for the equipment and instruments in the control room. However, post-accident environmental conditions vill be specified as being the design conditions in the specifications for the equipment and components connected with the reactor protection and the safety features related systems. 4 0329 s p (~ 7.8-2 l
D-B [ 7.9 Describe how reactor protection system and engineered safety
-. feature equipment will be physically identified as safety equip-ment in the plant.
RESPONSE
The Engineered Safety features equipment and instrumentation vill be identified with unique color coded tags prior to or during installation. Control Room safety features devices will have colored labels, and cabinet doors will be provided with colored labels specifying the cabinet function. O l N ! l l C 0230 9-1 E
D-B ! p' 7.10 On Page 8-8 of the PSAR the electrical and physical independence ( of the diesel-generators is discussed. Provide the design criteria for the diesel-generator cooling water system which vill insure a fully redundant cooling vater system.
RESPONSE
The diesel-generator cooling water system is a closed loop system circulating through the diesel jacket and the diesel lubrication system oil cooler. The closed loop system consists of a shaft driven circulating pump, a surge or expansion tank, piping, a thermostatically controlled by-pass valve and a thermostatically controlled heater, all of which are designed to maintain the engine at proper temperature during operation and standby. The circulating water is passed through the shell side of the diesel l generator cooling water heat exchanger. The cooling medium to the j diesel generator cooling water heat exchanger is furnished by the component cooling water system. Each heat exchanger is served by a separate line from the component cooling water system. 'i 1 0231. l M ! 7.10-1 l
N l T--- PS I l l o i i DIESEL wATen JACKET l WATER PUMP h '__ t
.g A I
m d O AUX WATER Pu bAP . L e....., , y m l SURGE TANK h--,--h IMMERSION HE AT ER FROM DIESEL LU BRIC ATIO N SYST E M FROM FIRE # 2 OlESEL PROTECTION SYSTEM ,
^-
A A Y_ Y. 3 M N > EMERGENCY DIESEL EMERGENCY DIEGEL GENERATOR G ENE.R ATO R COOLI NG WATER X X COO LING W AT E R
> H E AT EXCHANGER HEAT EXCH ANGER y 1 i TO COMPOWENT FROM COMPONENT COOLING SYSTEM COOLI NG SYSTEM Y
O'lESEL GENERATOR 1 C00 LlN G WAT.ER .5Y.STE M g FlGNA,By O 7.10-1
D-B ,c 7.11 Identify the emergency diesel-generator protective interlocks
- i: and discuss the basis for their selection.
RESPONSE
The emergency diesel generators vill be equipped with mechanical and electrical interlocks necessary to assure personnel protection and to prevent or limit' equipment damage during operation or electrical fault and serious overload conditions. i Under emergency conditions, the diesel generators will be tripped ! automatically only under the following conditions: ; Diesel Engine High Temperature , Overcurrent (Fault) l Generator Differential Relay Action I Engine Overspeed ~l Crankcase High Pressure l Low Lube Oil Pressure Trips during emergency conditions are minimum in number but important for equipment protection. Emergency power availability is thereby enhanced by tripping the unit off for these faults which enables prompt repair and return to service. _ _ g 0233 7 11 . -
D-B 7.12 The text states that during normal operation station power leads [r, are supplied from the auxiliary transformer. Upon failure of this power source, startup transformers 01 and 02 provide the required loads in what appears to be a " split bus" concept. Describe the design concepts utilized to protect this " split bus" concept and show that the design is not compromised by any of the following:
- a. Multiple power sources to switchgear buses A and B.
- b. Cross feeds to switchgear buses CD, F3 and E3.
- c. The use of transformer between buses C1 and Dl.
RESPONSE
- a. The main ~ reason for the " split bus" concept was to insure two outside power sources, not only for the emergency power supply but for the -
entire plant auxiliary power system. The control circuits of the feeder breakers vill not allow multiple power sources to be connected to either bus A or B except for a momentary paralleling of sources by manual transfer at plant startup or shutdown.
- b. 1. No single failure can parallel emergency buses C1 and D1 through emergency bus CD. The two (2) breakers connecting CD with C1 (D1) m2st be open before the two breakers connecting CD with D1 (C1) can close.
- 2. Neither bus E3 nor F3 (E2 nor F2 in exhibit 8-1 Amendment No. 3) can be supplied from two buses since the closing of one incoming feeder breaker vill trip the other incoming source breaker.
- c. The common transformer between C1 and D1 buses has been deleted. ro separate transformers have been substituted connected as shown on Figure 8-1 of Amendment 3 i'
" 0234 7 12-1
D-B y- 7.13 Paragraph 8.2.3.3 of PSAR discusses but does not clearly identify
- l. the diesel starting signals. Identify the diesel start signals and clarify whether the circuit is an "and" circuit.
RESPONSE
The diesel start signals are: Manual start (Local or Remote); main generator tripping due to the detection of serious electrical trouble in the main generator, main transformer or unit auxiliary transformer; turbine trip, reactor trip; loss of voltage of the off-site power sources; a safety actuation signal, and supervised routine test signals to test integrity of the system. Any of the above conditions will initiate a start signal to the diesel; however except for manual synchronizing during supervised routine testing, the emergency buses will not be isolated and the diesel generator breakers vill not be closed unless the loss of outside power is confirmed ct the 4 kv - emergency buses. Paragraph 8.2.3.3 of the PSAR has been changed in Amendment No. 3 to clarify this information. ( l l 1 f \ 0235 7.13-1
D-B T.1h Provide the design criteria for the switchyard 125 v de system and f._ show that a single failure vill not negate the ability to supply offsite power.
RESPONSE
The design of the 125V L1 system for the switchyard shall consist of two independent 125V DC sys+. ems. Each of the two systems shall be co= posed of a separate 125V DC Lattery, battery charger, and distribution system. Cal ' 2 separation will be maintained between the two systems. A single failure caused by a malfunction of either of the two 125V DC systems will not effect the performance of the other system. The ability of the switchyard to supply off-site power to the plant will not be effected by the loss of one of the two 125V DC systems. C 0236 7.1h-1
D-B 7' 7.15 Provide a loading table for the essential buses for the safe shutdevn and for the accident conditicus. This table should provide the timing sequence for starting the loads and the re-sulting margin for the worst case load with reference to the diesel continuous rating.
RESPONSE
A loading table showing the timing sequence for starting the diesel generators and the essential loads vill be provided at a later date. This will indicate a diesel generator accelerating time of approxi=ately 10 seconds and a total loading time of approximately h0 seconds, including the diesel generators and all essential loads. To calculate this in detail requires a kncvledge of final values of inertia, motor sizes, impedances, and speed torque curves as well as the impedance, and transient response of the emergency generator voltage and frequency which are not now available. e oaa7 T.15-1
.n.
D-B p 7.16 Perform a system stability study 'md show that neither loss of this unit nor loss of the largest generating unit on the system (either CAPCO or Toledo Edison) will negate the ability to pro-vide offsite power.
RESPONSE
System stability studies are performed on a periodic basis and particularly when major transmission and generation additions or changes are planned. Completed studies have shown that we meet this criterion, criterf.on 39 and the criteria in Sectirn 8.2.1.2 of the PSAR. e 6 e 2 0238 7.16-1
D-B b' 7.17 Paragraph 7.1.1.23 of the PSAR discusses the use of isolation amplifiers. Provide the design criteria for these amplifiers.
RESPONSE
The design criteria for the isolation amplifiers are those stated in para 6raphs 3 and 4 of Section 7.1.1.2.3. of the PSAR. The reactor pro-tection system is designed so that it is incredible to get greater than 6 410 volts on the input side of the isolation amplifiers. I l am M 02as 1 7.17-1
D-B p 7 18 The neutron detectors which are part of the Plant Protection ( System are temperature limited at 175 F. Provide the design criteria for detector cooling and temperature monitoring which allows the detectors to meet the requirements of IEEE 279, Paragraph 3(g).
RESPONSE
Neutron detectors are contained within detector thimbles which in turn are surrounded by the concrete of the primary shield. Cooling for the primary shield is provided to limit the maximm concrete temperature to less than 150 F. Becauce of the margin between maximum concrete temperature of 150 F and the upper limiting temperature of detector operation at 175 F, no special detector cooling is required. Temperature detector elements will be provided to verify that actual concrete temperatures do not exceed design limits. [ 1 l 1 l l 0240 7 18-1
D-B T.19 Paragraph T.6.h of the PSAR describes the com=unication systems (^ as having independent power supplies. Describe these sources of power. What radio communications do you have from the station to your system (offsite) in the event of an emergency?
RESPONSE
Communication system amplifiers will be supplied from the 120 volt AC bus provided for the communication system and computer. This bus is a non-interruptable bus normily fed from the DC batter /' system through an inverter with an alternate feed from the 120 volt AC instrumentation bus. No " broadcast" radio ccmmunications are centemplated; however micro-wave channel t<lephone communications vill be installed for communications with the Toledo Edison system dispatcher. e l l J l 7.19-1
. l
D-B A -T.20 Provide the design criteria for supplying power to the ,k radiation monitoring subsystems, specifically those which monitor releases to the atmosphere and those which provide control functions.
RESPONSE
Where radiation monitoring is provided to monitor and control releases to the atmosphere, a minimum of two redundant sensors are provided, each supplied from a separate vital bus. g 0242 7.20-1 L
D-B g 7 21 Provide the design criterion for sizing the containment air recirculating feal motors. Do the required ratings represent nameplate ratings or nameplate times a service factor?
RESPONSE
The estimated brake horsepower requirement of each fan at full (normal) speed is 152. The estimated brake horsepower at half (emergency) speed is h5 horsepower. A two speed, constant torque, water cooled motor will be used to drive each fan. Motor rating will be 150/75 HP with a 1.15 service factor based on a 40 C ambient. The equivalent service factor under emergency conditions would be 1.15 times 75 divided by 45 or approximately 1 9 The temperature of the air discharged from each motor cooler will be continuously monitored. 4 I 1 1 1 1 k l N 0243 7.21-1
D-B (s - 7.22 Discuss explosion proof requirements for instrumentation (local mounted detectors and transmitters) and electrical equipment (motors, switches and other arcing devices) '.*hich are located in potentially hazardous areas. (A hazardous area is defined as one in which flammable or explosive concentrations or material may exist).
RESPONSE
According to the National Electrical Code, a hazardous location is one having a hazardous atmospheric mixture containing any one of the following:
- a. Flammable vapors from volatile ficmmable liquids
- b. Flammable gases
- c. Combustible dust
- d. Easily ignited fibers or flyings Flammable liquid is defined in NFPA No. 321-1965 as any liquid having a flash point below 140 F (60 C) and having a vapor pressure not exceeding h0 lbs. per square inch absolute at 100 F (37.8 C).
On the basis of the above the only potentially hazardous areas appear to be the battery rooms where explosive hydrogen mixtures could accummulate. f The battery rooms will be provided with individual ventilation systems and explosion proof lighting fixtures. No other instrumentation or electrical equipment will be allowed into these rooms apart from the batteries themselves and the connecting cables. t 0244 i 7.22-1 )
]
D-B T.23 Generally the design criterion followed for instrument air provides (' guidance in the design of a reliable source of clean dry air. Past experience has indicated that the use of instrument air valve operators (piston operators) can result in sticking or galling due due to lack of lubricatien. State your design criterion regarding air supplied to piston operators. _ RESPONSE
- 1. All instrument devices, including cylinder actuators vill be supplied dry, filtered, oil-free air.
- 2. The potentie*. for galling or sticking of cylinder actuators is eliminated by selection of materials proper for service with dry air supplies.
l 1 1 0345 7.23-1 l
D-B 8.0 Conduct of Oceration . 8.1 Discuse your plans regarding the develop =ent of an overall site emergency plan including scope, organization, protective =easures, technical bases , action of offsite agencies , medical support arrangements, training of personnel, recovery following accident, and ot'aer applicable =aterial.
RESPONSE
This subject is discussed in detail in the PSAR At:end=ent No. 3 Section 12.h.1 and Section 12.h.2. O allik . 0246 8.1 -1
D-B 8.2 Describe what provisions will be made to ensure plant security
.m from unauthorized entry both during construction and operation.
Indicate the extent of perimeter fences, lighting, guards, employee screening procedures, visitor control, control of con-tainment access, and other site surveillance methods.
RESPONSE
Figure 8.2-1 indicates the extent of Toledo Edison plans to ensure Davis-Besse station area security during construction and ultimate operation. North, south and west permanent station perimeter security-lighted fence lines vill be installed following initial site clearing, grading and excava-tion. Certain openings across the east side of the station area vill be temporarily fenced during construction of the on-shore waterways and associat-ed structures. Eventually the east side vill be securely enclosed with permanent fencing, completing the station perimeter. Security-lockable controlled access gates are provided only where construction and ulti= ate station operation demands dictate their location. - Toledo Edison vill utilize a security force to implement their written l procedures regarding controlled station access covering all categories of employees, visitors and vehicular traffic, during the construction and operation phases. The guard house overseeing the construction access roadway will be perpetually canned throughout the station construction period. Tha guard 1 house for the pemanent station access road will be established on a timely basis before station startup to cover Toledo Edison operating personnel.and permit phasing out of the construction gate facilities. It is the intent of Toledo Edison to maintain a continuing record of all individuals and vehicles within the station security-fenced area. Individuals passing the controlled access gates will be screened and logged through by means of coded and serialized badges. 'lisitors will be approved for site access only after meeting prescribed qualifications and then agree to escorting while within the controlled area. Vehicular traffic will be continually monitored and logged along with ocergants. Toledo Edison vill limit and approve site access for only certain registered and qualified vehicles. The reactor containment and certain areas of the auxiliary structure vill I be set up under a special access control when construction and pre-operation-al check and test stages demand closer supervisian, especially regarding cleanliness. Following criticality and actual startup, these areas will be monitored by health-physics personnel with the usual controlled access check points for all individuals. In addition, cer'ain openings will be equipped with panic-type hardware, allowing opening from the inside only for emergency exits and sounding control room alarms when so operated. Contain-ment access lock and emergency hatch are equipped with interlock devices notifying control room personnel if these openings have been operated. General site surveillance vill be made possible by utilizing periodic tours l f of inspection by security force personnel following site roads and top of dikes. k 8.2-1 .. i
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DAVIS-BESSE NUCLEAR POWER STATION STATION AREA SECURITY PLAN FIGURE B.2-1 0249
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._. _ - - _ - _ _ _ _ _ _ _ _ J
D-B 8.3 - Describe the. locations of hospitals, schools an1 detention institutions within the low population zone (LPZ) and state the (S number of beds in the hospitals, the number of students in the schools and the number of individuals in the institutions.
RESPONSE
There are no hospitais, schools or detention institutions within the low population zone radius of two miles. The nearest institution of any kind is 3-3/h miles southwest of the station on State Route 19 It is Carro11 Grade School (approx 3mataly 300 pupils and teachers) including grades one throu6h six and serves Benton, Carroll and Salem Townships. k x I 1 1 lC , M l 0250 a.3-1 a
D-B p 90 Auxiliary and Emergency Systems 9.1 On page 9-1 cf the PSAR, you have provided a listing of the fifteen codes and standards used in the design, fabrication, and testing of components, valves, and piping of the auxiliary and emergency systems. Provide a cross-reference identifying the components to which the various codes and standards apply.
RESPONSE
The following is a cross reference giving the components and the applicable design codes and standards. Components are listed by system with reference to the list in PSAR section 9, page 9-1. SERVICE WATER SYSTD4 Pumps a,e,j,p Piping e,g,f Valves a,e,p Motors h, i, r, s COMPONENT COOLING WATER l Pumps a, e, j , p Heat Exchangers a,c,d,e,k Tanks e, o Piping e,g,f Valves a,e,p Motors h, i, r, s SPENT FUEL POOL COOLING SYSTEM Pumps a, e, j , p Heat Exchangers a,b,c,d,e,k Filters a,b,d,e,g Demineralizer Tanks a,b,d,e Piping e , g, f Valves a,e,p Jglgggy Motors h, i, r, s ();)(( 9.,1-1
D-B SAMPLING SYSTEM
,]
Piping e,f,g
)
Heat Exchangers e, k Valves a,e,f.p HEATING, VENTILATION AND AIR CONDITIONING Fans 1,t,q Filters 1,t,q Ductverk 1,t,q Heating and Cooling Coils , 1,t,q Motors h, i, r, a INSTRUMENT AND SERVICE AIR SYSTEM Receivers a, c, d, e Piping e,f,g Vc.1ves a,e,f,p N Motors h , 1, r , s AUXILIARY FEEDWATER SYSTEM Pumps e, j Tanks e, o Piping e, f, g Valves a,e,p Turbines e, i FIRE PROTECTION SYSTEM Pumps e, j , q Piping e, q i Valves e, q Motors h,i,r,s,q Ow*.n 8 Diesel q j 9.1-2 - e
D-B MAKEUP AND PURIFICATION ANT CHEMICAL ADDITION SYSTEMS
- Pumps a, e, j , u
- Heat Exchangers a,b,c,d,e,k
# Filters a,b,d,e
*Demineralizer Tanks a,b,d,e
# Tanks a,b,d,e,m Tanks e, m Piping e, f, g
' Valves e, f, v -
Valves a,e,f,p
- Motors h, i, r, s
- EQUIPMENT .( NSS SUPPLIER DECAY HEAT RD40 VAL SYSTEM
,(
- Pumps a, e, j , u
- Heat Exchangers a,b,c,d,e,k Piping e,f,g
- Valves e , f, y Valves a,e,f,p
- Motors b , i, r, a
- EQUIPMENT BY NSS SUPPLIER m
g 0253
.9 1-3
D-B 9.2 Makeup and Purification System 9 2.1 Discuss the possibility of the leakage of hydrogen from the takeup tank. If local concentrations of hydrogen in excess of two volume percent can occur in the auxiliary building external to the makeup tank, describe the means taken to prevent ignition of the hydrogen.
RESPONSE
The tank and connected piping vill be hydrostatically tested to demonstrate leaktightness during preoperational testing. Thereafter, the only likely source of leakage vill be at the safety valve, which is piped to the radwaste system. e cas4 9.2.1-1
I D-B 9.2.2 Discuss your criteria regarding the prevention of crystallization [.- in those portions of the system containing concentrated boric acid. Discuss the design criteria for the tracing system pro-posed considering (1) the type of tracing proposed, (2) redundancy, (3) the minimum te=perature margin between fluid temperature and the saturation temperature for the concentration involved, (h) means used to assure adequate coverage of critical areas (e.g. , valve bodies and elbevs), and (5) plans for preoperational testing. Discuss the consequences of solidification at any single point in this system, indicating if cold shutdown could be achieved.
RESPONSE
Two separate and redundant systems are provided for boric acid addition. Both the boric acid addition tanks have electric heaters to maintain the temperature 15 F above crystallization temperature and the tank temperature vill be monitored during operations. Also the tanks, the boric acid pumps and the associated suction piping are located in a room which will be heated and maintained at a temperature 15 F above the crystallization temperature. All other pipes and valves carrying concentrated boric acid which are out-side this room vill be heat traced to maintain the 15 F temperature margin. One line is installed as the primary addition route, while the other is only used for standby purposes in the event the primary path is not available. Flush connections are provided that will permit cleansing of the addition lines. Preoperational tests vill be conducted to ensure operability of each heat tracing circuit. No adverse consequences result from a solidification at any single point in the system because sufficient boron can be added from the berated water storage tank via the high pressure injection pumps to ec=pensate for reactor coolant system shrinkage. There is a separate line from the tank to each of two pumps , thereby assuring this path of addition. C 9.2.2-1 N u .
D-B
If the co=ponent cooling water system fails downstream of the 93 reector building isolation valve, the function of the shield coolers, reactor coolant pu=p coolers, and the letdown coolers ate lost. Since under this situation the makeup system could not be used for bleed-and-feed operations owing to the loss of the letdown coolers, discuss the ability to achieve cold shutdown.
Indicate the margin between the volu=e contraction in the primary system during cooldown and the volume of borated water required to ec=pensate for the reactivity gained during cooldown. RESPON'SE The ability to achieve cold shutdown is maintained even with the loss of the co=ponent cooling water system. Boric acid may be added in combination with the required demineralized water so that the total added quantity injected vill produce the required soluble poison concentration level as well as the required makeup for contraction. The total contraction of the reactor coolant system from 582 F to 140 F is approximately 3250 ft3. Assuming the CRA of highest worth stuck out of the core, the required volume of 7 w/o boric acid solution that must be added is approximately 460 ft3 C 4 s-0256 9 3-1
D-B 9.h Service Water System (,_ 9.h.1 Identi.^f the number of auxiliary feedvater pu=ps needed to dissipate decay heat i==ediately after shutdown.
RESPONSE
Either of the two auxiliary feed pu=ps meets the capacity requirements necessary to dissipate decay heat i==ediately after shutdown. This information is given on page 9-30 in the PSAR. ( M ( 0257 l 9.h.1-1 1
D-B
.9. h 2 - Providt- a process diagram of the auxiliary turbine-driven feed-water pump steam supply system.
RESPONSE
The steam supply for the Turbine Drives for the auxiliary feed pumps is taken from the main steam headers as shown on Figure 9.h.2-1. i l 9 i : l 4 ' 0 t 1 4 4 4 5 1 ' .i. [.
?. :
9.h.2-1 . . 04
'e
,_ r - .- . _ , . . , _ . _ ,,.,---,,._.__,--v_- -
=-
M AIN d " TURBINE
= =
A 1 p n b a a 1 O V STEAM GEN ER ATORS
^
TURBINE I l
/ ,
I TO ATMOS. = f55 =>< M IrROM - I AUXI LI ARY y y p g STEAM SOURCE 9my -b > m 9 , -g>s b e 1 m4 m / I k TO ATM O S. " AUX. FEED PUM P , TUR8IN E 2 .
D-B 9.5 List the seismic design classification of the various components of the fire protection system. Indicate to what extent this system can function with any single failure. To facilitate understanding, provide a diagram of the system. Identify those portions of the fire protection systems designed to Class II seismic standards whose failure could damage Class I structures and components. Would failure of a Class II portion of the system prevent fire protection to any Class I structures or components? RESpCNSE All components of the fire protection system are designated as Seismic Class II and will be designed and installed in accordance with the Uniform Building Code. The entire fire protection system will be in accordance with the recommendations of the National Fire Protection Association and Nuclear Energey Property Insurance Association. The motor driven fire pump and the jockey pump are installed in a pump house located adjacent to the fire water storage tank. The fire water storage tank and the pump house are designated as Seismic Class II. The diesel driven fire pump will be installed on the intake structure which is a Class I structure, however, 8 the fire pumps are not tissile protected. A single failure in some portions of the fire protection system would result in loss of fire protection of some Seismic Olass I cc.ponent or an accidental release of water; however, the system will be designed with sectionalizing valves such that such a failure can be isolated and the re-mainder of the system can function. ~
~
02:f3() 3-9.5-1
M M 3 3 3 3 /^ >< >< X >< d % TURBI NE BLOG HOSE RACK LOOP ELE CT R!CA L
+ PE N ETRATION Ro0M h SERVICE BLDG :>4-. CABLE SPREADING AUX. FEEDWATER ST RT-UP BACKUP TRANSFORMER 1URBINE
~
BASEMENT >4-> AUX. BOILER ROOM
% TURBINE OIL >4-* MAIN TRANSFORMER ROOM h AUX. TRANSFORMER
-94+ LUBE OIL A% BUG TIE 4t i STORAGE ROOM
>4-o- BUS TI E
- 2 HYDRO PNEUM ATIC h START-UP TRANSFORMER STORAGE TANK
-X @ >4-*- SE AL OI L ROO M n >4-* DIESEL GEN. ROOM DIESEL COOLING
] WATER BACKUP
>< :>< X ><
h a g g M _.-N_ 1
~
FIRE WTR. STORAGE
>0-V' TANK d c =
FIRE PU Af P ( MOTOR DRIVEN) y n = >o-LA h JOCKEY PUMP h INTAKE STRUCTURE FROM WATER
/7 / /7 / / / /7 777/ 77 // TREATMENT 5 SYSTEM W FIRE PUMP DIESEL DRIVEN g
'_ . DAVIS BESSE NUCLEAR POWER STATION : FIRE PROTECTION SYSTEM FIGURE 9.5-1 02RIl
D-B g 9.6 In Section 9.7.2 the normal liquid and gas sampling capability
< is described. Provide a discussion t-d similar listing of the sampling capability folleving a loss-of-coolant accident; include the capability to monitor hydrogen gas concentrations throughout the facility.
RESPONSE
In addition to the liquid and gaseous process samples listed in PSAR Section 9.7.2 the atmosphere inside the containment building, penetration areas, and reactor coolant and vaste gas processing areas vill be monitored by local diffusion type catalytic combustibles detectors. These detectors vill provide a warning of the accumulation of hydrogen or other potentially cabustible gases or vapors. A sample of the containment vessel atmosphere vill be monitored for both gaseous activity and combustibles content. Following LOCA samples of process stream fluids may be obtained as required by opening sample line isolation valves. Following LOCA the containment - at osphere local corbustible detectors may be inoperable, therefore, a sample of containment gaseous contents may be obtained by opening the double isolation valves in the containment atmosphere sample lines. The gas sample so obtained may be analy::ed in the laboratory. After LOCA, a sample of the containment sump liquid will be obtained by withdrawing a sa=ple frem the downstream of the decay heat cooler.
~
9.6-1 u
D-B 10.0 Steam and Power Conversion Equipment
.A t 10.1 Discuss the consequences of inadvertent opening of the feedwater startup valve during hot zero power cperation. Indicate the maximum reduction in the primary coolant system temperature which would be experienced.
RESPONSE
Inadvertent opening (full) of a feedwater startup valve at hot zero power results in an increase in feedwater flow to the affected steam generator to lh% of full flow (795,000 lb/hr) . This flow rate would cause a reduction in pressure in the steam generator to a value 1e.ss than 900 psia, and the turbine bypass valves would remain closed. Heat transfer would be limited to heating the feedwater to - saturation temperature at the existing pressure. Assuming that saturation temperature does not decrease, and remains at 532 F, the heat transfer from . the reactor coolant system results in a maximum cooling rate of approximately 5 F/ min. Under these conditions the maximum steam generator water level limit end alarm would be reached in about 45 seconds. At this point the ICS would close the startup valve or, the operator would close the startup valve if the valve is in hand control. The maximum reduction in reactor coolant temperature would be 10F assuming one minute for the high level alarm, and one minute for the operator to take action to close the valve. ( m E i 10.1-1 0 243,'1
D-B 10.2 Describe the provisions made for steam generater blevdown and f_; secondary system cleanup, indicating the magnitude of the blow-down flow and the disposal of the secondary coolant discharged.
RESPONSE
Due to the once-through steam generator design incorporated in the Babcock & Wilcox NSS pressurized water reactor system, no steam generator blevdown is required. Full flow condensate polishing demineralizers are provided as described in Section 10.2.h of the PSAR. b e
,4 0264 10.2-1
D-B
"' s 11.0 Radioactive Waste Distosal System 11.1 Provide an estimate of the amounts of radioactivity on an isotopic basis (including tritium) that would be released daily and annually from the proposed Davis-Besse facilities. Consider both liquid and gaseous vastes , assuming that the activity of the primary coolant is that corresponding to operation with an activity level to be stated in the Technical Specifications.
List all assu=ptions made and explain their bases. Indicate bow a load-following mode of operation vill affect releases. Based upon the estimated available annual average dilution factor, provide an estimate of the maximum gaseous release limit for this facility which would be within the limits of 10 CFR 20. Compare this estimated release limit with the estimated gaseous releases frcm this facility.
RESPONSE
DAILY AND ANNUAL RELEASES OF RADI0 ACTIVITY
- a. Liouid Wastes The estimated daily release of radioactivity will be based on the following assumptions:
- 1. The amount of vaste that will be discharged from the station in a day is 5775 ft 3. This represents the daily capacity of the clean radvaste
(. system assuming the use of two 15 gpm evaporators. This is a con-servative assumption because (1) the radvaste system is not scheduled to operate on a regular basis and (2) a large portion of the processed wastes =ay be reused.
- 2. All vastes will initially have the naximum radionuclide concentrations found in the reactor coolant assuming 1% failed fuel.
3 All vastes vill be processed through the radvaste system described in the answer to Question 2.h.
- h. The tritium concentration of the vastes will be 6 ue/ml @600 F or 8.Th2 ue/ml @l20 F. Since the radvaste system removes none of this, the latter figure vill be the level in the discharge. Based on 5775 ft3, this amounts to about lh30 Ci in one day. The conservatism of this amount is indicated by the fact that the reactor only releases about 3000 Ci to.the coolant in an entire year.
The estimated annual release of radioactivity will be based on the following assumptions: < 1 The tritium discharged vill be limited to 3000 Ci since this is the total assumed released to the coolant in a year.
- 2. 150,000 ft3 of vastes vill be released in a year (see question 2.k).
3 All other assumptions are the same as in the daily case. 11.1-1 oms ~
D-B G
/
Tables 11.1-1 and 11.1-2 are, rescectively, lists of the daily and annual isotopic releases from the stat i based on the preceding assumptions.
- b. Gaseous Wastes There are many different ways the gaseous vastes can be disposed of involving various ce=binations of decay and release times. Although few of these would involve continuous release, each generally has a period during which it occurs. For this analysis, the daily releases were calculated for two modes of operation which were based on the folleving assumptions:
Mode 1: 7l 1. Activity equal to the maximum enount external to the fuel elements
~
vill be accumulated in a decay tank. Except for Kr-85, this is approximately equal to the maximum total activity levels found in the primary system assuming 1% failed fuel. The Kr-85 present vill be taken as 8500 Ci which is the total leakage from the fuel in one year. The initial tank activities on an isotopic basis are listed below: Kr-85m 550 7 Ci .- Kr-85 8500.0 Ci Kr-87 30h.5 Ci Kr-88 971.8 Ci Xe-131m 874.6 Ci Xe-133m 100h.2 Ci Xe-133 9070h.h Ci Xe-135m 3h0.1 Ci Xe-135 2170.h Ci Xe-138 18h.5 C1
- 2. The gases are allowed to decay for 3 days.
- 3. The gases are released at a uniform activity rate over 27 days. The total decay and release time was chosen to achieve a 30 day turnover for the decay tanks.
While the total activity release rate vill be kept constant, those of the individual isotopes will vary considerable. That is, the release rates of the shorter lived isotopes would be greatest at the beginning 4 of the release while those of the longer livad nuclides would peak later. Calculating these individual rates is a difficult problem but rough estLnates have been made and are given in Table 11.1-3 as a function of time. These were based on a constant release rate.of 725 Ci per day. O'Z*% Mode 2:
./
- 1. Activity equal to the maximum amount external to the fuel vill be accumulated in a decay tank (see Mode 1).
m
D-B (' s
- 2. The gases are allowed to decay for 25 days.
- 3. The gases kre released at uniform activity rate over 5 days. Again, the total decay and release time was chosen to achieve a 30 day turn-
-rer for the decay tanks.
Table 11.1-h gives the isotopic release rates as a function of time. These were based on a constant total daily release of 2275 Ci. These two methods of disposing of the gases show how, on an isotopic basis, the release rates can vary considerably with operating procedures. The total decay and release time for both were chosen as 30 days because this is considered to be the minimum time that will be available. Actually, during normal operations, up to 60 or more days could be used. . The estimated annual release of radioactiv'ity will be based on the . following assumptions:
- 1. A decay tank will be filled every 30 days during a cycle with activity equal to the maximum amount external to the fuel elements.
This will be a total of about ten (10) tanks.
- 2. The gas in each tank will be allowed to decay for an average of 30 days before it is released.
I i Table 11.1-5 is a list of the estimated annual release of the gaseous fission products to the envircnment. LOAD FOLLOWING MODE OF OPERATION
- a. Liquid Wastes The quantity of liquid wastes that are released from the station should change little i? a load-following mode of operation is used. This occurs because, normally, processed coolant is reused unless it is released' to keep the tritium level down. Since the amount of tritium leaked into the primary system 1s basically a function only of the core power generated, it does not vary significantly with operating mode.
- b. Gaseous Wastes The gross quantities of primary coolant letdown to the radwaste system can increase greatly when load-following is done. Since the coolant is assumed to be degassed at the rate of h0 cc-H2 /11ter, the amount of gaseous vaste created will also increase. For daily 100%-50%-100% power <
transients (the projected design basis), it is estimated that up to 22,50c ft3 of additional gas may be created during a cycle. Although holdup periods for decay may have to be shortened, the radwaste system will be able to handle this vaste. Due to the conservative assumptions made-in its compliation, Table 11.1-5 should still give good estimates [v .s', for annual release of the various radionuclides. J 11.1-3
D-B
/
MAXIMUM ALLOWABLE ANNUAL GASEOUS AC"'IVITY RELEASE The various gaseous isotopes have several different 10 CFR 20 limits for release to unrestricted areas. However, after three (3) days of decay, the only significant nuclides left are Kr-85 Xe-131m, Xe-133m, and Xe-133. All of these have a release limit of 3 x lo-i ue/ml or greater (Xe-131m is 7 l 4 x 10-I) . Therefore, to simplify the problem, all icotopes will be grouped together with the single limit in force at the site boundary. Using thig assum tion and the annual (X/Q) value for the site boundary of 1.69 x 10-0 sec-m , the maximum allowable release for one calendar year is: Maximum Release = 5.61 x 106 Ci The total estimated annual release (27541 Ci) is well below this figure. N
')
)
a 0 u.1-4
D-B
- _ TABLE 11.1-1 Daily Release of Activity (liquid vastes)
~
Isotope Maximum Concentration Daily Release from in Primary Loop Station (ue/ml) (Ci) Tritium H-3 6.0 1.h3 x 10 3 Insoluble Corrosion Products cr-51 0.1h 3.27 x 10-5 . 0.016
-6 Mn-sh 3 76 x 10 co-58 0.84 2.00 x 10-
-6 co-60 0.00h5 1.08 x 10 Fe-59 0.016 3 76 x 10-6
(- Fission Products (gaseous) ' Kr-85m 1.T h.06 x 10 -3
-2 Kr-85 11 9 2.83 x 10
! Kr-87 0 94 2.24 x 10 -3 Kr-88 -3 3.00 7.15 x 10 Xe-131m 2 70 6.h3 x 10-3 Xe-133m 3.10 7.39 x 10-3 Xe-133 280.0 6.67 x 10-1 Xe-135m 1.05 2 50 x 10 -3
-3 Xe-138 0.57 1.36 x 10 ,
d Fission Products (solid-ionic) ' Rb-88 3.00 7 15 x 10-8 l Sr-89 0.0h5 1.08 x 10 -9 m y 0269 l 11.1-5 l l l
D-B O Isotope MAXIMUM CONCENTRATI0L Daily Release from in Primary Loop Station (uc/ml) (C1) Sr-90 0.00h1 9.81 x lo U Sr-91 0.052 1.2h x 10 -9 Sr-92 0.019 h.58 x 10-10 Y-90 1.05 -6 2 50 x 10 Y-91 0.25 5.89 x lo-T. Mo-99 6.00 1.h3 x 10-5 I-131 3.60 8.59 x 10-0 I-132 5.h 1.29 x 10 -7 I-133 h.2 1.00 x 10 -7 I-13h 0 56 1.3h x lo-I-135 2.2 5 25 x 10-9 ' Cs-13h 50 1.19 x 10-5 CS-136 09 2.1h x lo-Cs-137 53.0 1.26 x lo-b Cs-138 0.82 1 96 x 10-6 Ba-137m hT.0 1.12 x lo Ba-139 0.091 2.13 x 10 -9 Ba-lho 0.073 1.80 x 10 -9 La-lho 0.02h 5 72 x 10-10 ce-1kh -11 i 0.0031 7.36 x 10
~ '
0270 11.1-6
D-B ,~ TABLE 11.1-2
~
Isotope Maximum Concentraticn Annual Release from in Primary Loop Station (ue/ml) (Ci) Tritium 3 H-3 6.0 3 00 x 10 l7 Insoluble Corrosion Products cr-51 0.1h 8.51 x 10 h Mn-5h 0.016 9.78 x 10-5 - Co-58 0.8h 5 19 x 10-3 co-60 0.0045 2.81 x 10 -5 Fe-59 0.016 9 76 x 10 -3 Fission Products (gaseous) Kr-85m 1.7 1.05 x 10-1 Kr-85 11 9 7.36 x 10-1 Kr-87 -2 0.9h 5.83 x 10 Kr-88 3.00 1.86 x 10-1 Xe-131m 2.70 1.67 x 10-1 Xe-133m 3.10 1.92 x 10 -1 Xe-133 280.0 17.35 Xe-135m 1.05 6.51 x 10-2 Xe-135 6.70 h.15 x 10 -1 Xe-138 0 57 3.53 x 10-2 Fission Products (solid-ionic) Rb-88 3.00 1.86 x 10-6 Sr-89 0.0h5 2.81 x 10-8 Sr-90 0.0041 2.55 x 10 -9 02.71 11.1-7
D-B
- D Isotope Maximum Concentration Annual Release from in Primary Loop Station (Uc/ml) (C1) __
Sr-91 0.052 3.23 x 10-8 Sr-92 0.019 -8 1.19 x 10 Y-90 1.05 6.51 x 10-5 Y-91 0.25 1 53 x 10-5 Mo-99 6.00 3.72 x 10-4 I-131 3.60 2.23 x 10-6 I-132 5.h -0 3.35 x 10 I-133 4.2 2.60 x 10 ' I-13h 0 56 3.h9 x 10-7 I-135 2.2 1.37 x 10 -6 Cs-13h 5.0 3.10 x 10-h Cs-136 0.9 5.57 x 10 -5 Cs-137 53.0 3.28 x 10-3 Cs-138 0.82 5.10 x 10-5 ! Ba-137m 47 0 2.91 x 10 ' Ba-139 0.091 5 53 x 10-0 Ba-lho 0.073 h.68 x 10-0 La-140 0.024 1.h9 x 10-8 Ce-14h 0.0031 1.91 x 10-9
'J 0272 11.1-8 -
W
D-B TABLE 11.1-3 DAILY RELEASE Day Isotope (Ci) Kr-85 Xe-131m Xe-133m Xe-133 Xe-135 1 87.0 75 h.2 626.2 0.0 2 97.6 8.0 35 616.2 0.0 3 109 2 8.h 2.9 60h.5 0.0 h 122.0 8.9 2.h 591.8 0.0 5 135.8 9.3 19 577 9 0.0 . 6 150.8 9.8 1.6 562.8 0.0 7 167 0 10.2 1.3 Sh6.h 0.0 8 18h.3 10 7 1.1 528.9 0.0 9 202.8 11.1 09 510.3 0.0 i 10 222.3 11 5 07 490 5 0.0 11 2h2.8 11.8 0.6 h69.8 0.0 12 264.2 12.1 0.5 hh8.2 0.0 13 286.3 12.k O.h h25 9 0.0 1h 309.0 12.7 0.3 403.1 0.0 15 332.0 12.8 0.2 379.9 0.0 16 355.3 13.0 0.2 356.5 0.0 17 378.7 13.0 0.1 333.1 0.0 18 401.8 13.1 0.1 310.0 0.0 19 h24.6 13.0 0.1 287 3 0.0 20 hh6.9 13.0 0.1 265.1 0.0 21 h68.h 12.8 0.1 2h3 7 0.0
- 22. h89 2 12.6 0.0 223.2 0.0 g
;~" --
02?3 u.1-9 g
D-B Day Isotore (Ci) Kr-85 Xe-131m Xe-133m Xe-133 Xe-135 23 508.9 12.h 0.0 203.6 0.0 2h 527 7 12.2 0.0 185 1 0.0 25 5ks.h 11 9 0.0 167.8 0.0 26 561.9 11 5 0.0 151.6 0.0 4 27 577 2 11.2 0.0 136.6 , 0.0 b 0274 N 11.1-10
D-B m I TABLE 11.1-h DAILY RELEASE Day Isotope (C1) Kr-85 Xe-131m Xe-133m Xe-133 Xe-135 1 1597 0 39.0 0 638.9 0 2 1655 9 38.1 0 580.9 0 3 1711-3 37.2 0 526.h 0 h 1763.1 36.2 0 h75.6 0 5 1811.4 35.1 0 h28.5 0 02';3
- b. ,
M 11.1-11
D-B TABLE 11.1-5 )~1 ANIMAL RELEASE Isotope Release (Ci) Kr-85m -0 Kr-85 8h54 Kr-87 0 Kr-88 0 Xe-131m 15h6 Xe-133m 1 - Xe-133 17540 Xe-135m o Xe-135 0 Xe-138 0
]
.J 02.76 l
11.1-12 '
n-B 11.2 Provide a list of the seismic Class I and Class II equipment, components and/or structures in the radwaste system. Summarize the design capacities, pressures and temperatures for each of the vaste disposal tanks. Provide an estimate of the =aximum radioactivity levels in each of the tanks in the liquid and gaseous radvaste systems and the minimum available holdup times. Evaluate the consequences of the simultaneous release of all the radioactive liquid contained within the seismic Class II storage tanks. Indicate whether these amounts vill be limited, by design provisions or operational restrictions, to values within 10 CFR 20 limits.
RESPONSE
Radwaste System Seismic Classification and Design Parameter The radwaste system is entirely centained in the auxiliary building which is a Class I structure. The surge and gas decay tanks are designed to Class I standards. All other radwaste system equipment and piping is Class II. Tank Design Parameters: Design Capacity Design Pressure Design (ft3) (psig) Temp. ( F) Reactor Coolant Drain Tank 80 50 200 Clean Waste Receiver Tanks lh,000 15* 200 Concentrates Storage Tank 100 Atmos. 200 Clean Waste Monitor Tank 3,000 15* 150 Spent Resin Storage Tank 300 100 200 Detergent Waste Drain Tank 1,100 Atacs. 200-Misc. Waste Drain Tank 1,700 Atmos. 200 Misc. Liquid Waste Monitor Tank 1,100 15* 150 Evaporator Storage Tank 100 Atmos. 200 Waste Gas Surge Tank 1,000 15 200 Waste Gas Decay Tanks 1,000 150 200
'plus hydrostatic head
(' . l y- 11.2-1 - ' M
"} y
D-B Radvaste Tanks Activity Content
- 1. The following are conservative estimates of each tanks' activity inventory, based on equilibrium reactor coolant activity levels corresponding to one (1) percent failed fuel and no credit for decay (except for N16)"
Additional calculational bases are listed with each tank
- 2. Clean Liauid Wastes
- a. Reactor Coolant Drain Tank (1 - 80 ft3 )
This tank can receive unprocessed reactor coolant with the radioactive isotopes at their equilibrium level. G (gaseous activity content, curies) = 1040 D (dissolved or suspended, curies) = h50 T (tritium, curies) = 20* -
- b. Receiver Tanks (2 - lb,000 ft 3)
Each of these tanks receives primary coolant letdown after it has successively passed through a degasifier, a filter, and a mixed 5 beddemineralgzer. Conservatively, these have a total D.F. of 10 for gases, 10 for cations and anions, and 10 for insoluble corrosion products. G=2 D = 100 T = 3h66*
- c. Monitor Tank (2 - 3,000 ft3 ea,)
Each of these receives vaste from the receiver t p after it has been processed through an evaporator with a D.F. of 10 . G=1 -h D = 3 x 10 T = Th3*
- d. Concentrate Storage Tank (1 - 100 ft3)
The contents of this tank are essentially the same as those in the receiver tanks except that they are assumed to have been concentrated 1500 times and then passed through a mixed bed demineraliser with a D.F. of 103 for ions and 1 for corrosion products. All gases are assumed to have been removed. G=0 D = 600 T = 25* N
- Based on the maintenance of a tritium level of 6 mc/ml @600 F in the primary .y system or 8.742 @l20 F external to it. -
0278
, 11.2-2
D-B n
- 3. Miscellaneous Liquid Wastes Determining, with any degree of accuracy, the activity in the Miscellaneous Radwaste System's tanks is almost impossible since the origin of their contents are many and varied. An upper limit can be found if it is assumed that all miscellaneous vaste is primary coolant with equilibrium radionuclide concentrations. It will also have to be assumed, for the present, that no degasification occurs even though the vastes are exposed to the air and pass through an evaporator. Both of these processes remove gas but it is difficult to assign a numerical efficiency.
- a. Miscellaneous Waste Drain Tank (1 - 1700 ft3 )
This tank is assumed to receive reactor coolant G = 21,900 D= 9,500 T= 421*
- b. Detergent Waste Drain Tank (1 - 1000 ft3)
This tank is assumed to receive reactor coolant G = 14,1h0 D= 6,100 T= 272*
- c. Miscellaneous Waste Monitor Tank (1 - 1100 ft3 )
This tank receives the vaste from the two drain tanks after it has been processed through an evaporator with a D.F. of 10 . 5 G = 14,1h0 D = .06 T= 272*
- d. Evaporator Storage Tank (1 - 100 ft 3)
The total activity in this tank vill probably be limited by restrictions on off-site shipments. Fortgepresent,thislimit will be set at 10,000 Ci of equivalent Co 5 G=0 D = 10,000 T= 25* "
- Based on the maintenance of a tritium level of 6 mc/ml @600 F in the primary system or 8.742 @l20 F external to it.
1 Y 11.2-3
D-B
- k. Gaseous Wastes
' q.)
- a. Waste Gas Surge Tank (1 - 1000 ft3 )
Except for Kr-85, the acximum amount of activity that can be contained in this tank is that which is in the primary coolant at equilibrium conditions. For Kr-85 the activity will be taken as 8500 Ci which is the total amount released from the fuel in a cycle. The tritium content is assumed to be negligible. G = 106,000 D=0 T=0
- b. Gas Decay Tanks (3 - 1000 ft3 ) ,
The maximum activity for any one, or all, of these tanks is the same as in (a). G = 106,000 D=0 T=0 Minimum Hold-Up Times Except in the case of the contents of the gas decay tanks, there is no ~ anticipated need for, and/or no significant advantage to be gained from, taking credit for any decay in the radvaste system. This means that minimum hold-up times will be determined primarily by equipment capacity and/or operational procedures (i.e., pump sizes, need for analysis, need for diluting tritium, etc.). The minimum hold-up time in the gas decay tanks is governed by the ability to release the contents without violating 10 CFR 20 limits at the site boundary. Assuming that no credit is taken for decay and using a 30 day (X/Q) value, a tank containing the maximum amount of gaseous activity (106,000 C1) can, with evenly distributed discharge, be released in less than two weeks. This means a two week turnover is possible. Normally gases vill be held for up to 60 days to allow for decay. This hold-up period may decrease during load following operations due to the larger quantities of vaste created. This loss of decay time vill be partial compensated for by the increased dilution of the fission products.
- Consecuences of Assumed Radvaste Tank Failures All liquids released from the postulsted rupture of all Class II radvaste e tanks vill be contained within the Class I auxiliary building.
Any gaseous activity would be released to the building ventilation system and ultimately to the station stack. The amount of activity released will be less than the full power equilibrium amount in the primary system. The s
& J 11.2-h
D-B J release of this amount via a decay tank rupture has been analyzed in Chapter lh of the PSAR and it was shown that the 2 hour whole body dose resulting at the site boundary was well below the guideline values of' 10 CFR 100. It is not intended to limit the activity content of any of the radvaste
- system tanks to the 10 CFR 20 values.
e a
+ 9 0281 11.2-5
D-B
<-; 11.3 Describe the Water Radiation Monitoring System, including the type of monitors used and their sensitivity.
RESPONSE
The design of the Water Radiation Monitoring System has not been finalized au this time. Radiation monitors will be provided to detect the presence of the activity within the component cooling water system. The prceessed g liquid radwaste effluents will be monitored prior to its discharge from l8 the station. The water radiation monitoring system design has not progressed far enough to provide the type and sensitivity of the monitors at this stage. N c 0282 En 11.3-1
D-B 11.h Identify the systems and tanks that vill contain radionuclides (* ' . and vill not be designed to prevent the release of radionuclides in the event of the design basis tornado and flood. List the maximum quantity of radionuclides that would be contained in each system and tank. Provide analyses that show the maxiinum whole body and critical organ doses that could result from the release of radionuclides to unrestricted t reas as a result of the tornado or flood. Identify all ft : tors and assumptions used in the analyses.
RESPONSE
All tanks and equipment comprising the radvaste system are located inside the auxiliary building and therefore, are protected from the design basis tornado and flood. The borated and primary water storage tanks are located in the open and : are vulnerable to tornadoes and, in the case of the latter, earthquakes. The only radionuclide of any significance in either of these should be the tritium which has been picked up through contact with the primary system and which must be taken into the body to produce a significant dose. As shown in answer to the Question 2.4, the worst accident involving one of these tanks vould result in a tritium concentration at the nearest potable water intake of 1.93 x 10-2 ue/ml. To determine what dose might result 6 from this, the following conservative acsumptions were made:
- 1. For 2 days a " standard man's" entire water intake (2500 ml/ day) contained 1.93 x 10-2 uc/ml of tritium.
- 2. After 2 days no more tritiated water was consumed and the existing body burden decreased with an effective half-life of 12 days.
Under the above assumptions a whole body dose of less than .02 rem would result from the accident. 1 4: g 0283 11.h-1
t-D-B 11.5 Provide the folleving information regarding the radioactive gas
\ , vaste system.
- a. The maximum volume of primary coolant that vill be bled and/or degassified each year for nor=al and refueling operations.
- b. A list of the noble gas radionuclides and the maximum quantity of each that vill be released from the primary acolant each year.
- c. A list of the fractions and quantity of each noble gas radionuclide that vill remain after 3, 30, and 60 days holdup in the vaste tanks.
RESPONSE
- a. The maximum amount of primary coolant letdown to the radvaste system per refueling cycle for degasifi ation and other processing is estimated to be about 108,000 ft for normal operation. An additional one (1) system volume (11,hh0 ft3 g 580 F) is degassed in place prior to refueling,
- b. Estimates of the total annual release of the noble gas radio nuclides, assuming 1% failed fuel and no credit for decay, are:
Isotope Annual Release (Curies) Kr-83m 2.8 x 10 5 Kr-85m T.9 x 105 Kr-85 8.5 x 103 Kr-87 1 5 x 10 6 6 Kr-88 2.2 x 10 Xe-131m 1.6 x 10 N 5 Xe-133m 1.1 x 1.0 Xe-133 h.3 x 10 6 5 Xe-135 'T.9 x 10 c
- c. The following is a list of fractions and quantities of each noble gas that vill remain after 3, 30 and 60 days holdup in a decay tank.
Except for Kr-85, the activities of the isotopes initially present vill be equal to the maximum levels attained in the primary system during
~
0284 11 5-1
D-B. an equilibrium cycle (1% failed fuel). This approximately corresponds to the amount of each that can be found external to the fuel at any one time. The activity of Kr-85 vill be taken as 8500 Ci which is the total released to the primary coolant in a year. e C e N. A 0265 . 1, 11.5-2 3 x .
T D-B 3 DAY HOLDUP 30 DAY HOLDUP 60 DAY HOLDUP
. ISOTOPE ACTIVITY INITIALLY FRACTION OF ACTIVITY FRACTION OF ACTIVITY ORACTION OF ACTIVITY PRESENT IN DECAY ORIGINAL IN TANK ORIGINAL IN TANK )RIGINAL IN TANK TANK AMOUNT AMOUNT WOUNT LEFT (Ci) LEFT (Ci) E.FT (C1)
(Ci) Kr-85m 550 7 1.06 x 10-5 .0058 1.86 x 10 -50 10 -100 Kr-85 8500.0 99946 8495.16 99462 8454.3 98928 8408.9
-11 6.18 Kr-87 304.5 2.03 x 10 x lo-II lo- 10 -334 Kr-88 971.8 1.49 x 10-8 1,1,5 x 10-5 1g -79 10 -157 Xe-131m 871.6 4 .8409 735.5 .17677 151.6 4
.031249 27 3 Xe-133m 1004.2 .4049 406.6 1.18 x 10 -4 .1185 1.40 x 10-8 1.41 x 10-5 Xe-133 90701.14 4 .67396 61, 131.1 .01934 175!. 2 3 74 x 10-N 33.9 Xe-135m 340.1 2.4 x 10-84 1g-836 10-1672 Xe-135 2170.4 4.11 x 10-3 4 9 57 2.75 x 10-2t 5.97 x 10 -2.- 10-48
- Xe-138 184.5 2.93 x lo-Il 10-765 10 -1530
'l d
O ' N 03 C) 11 5-3 A
D-B 11.6 Provide the following infomation regarding the radioactive liquid waste system.
- a. A list of the maximum primary coolant concentr.ations of radionuclides that are activated erosion or corrosion products and a statement of all factors, assu=ptions, and references used to determine the maximum concentrations.
- b. The minimum efficiencies of the primary demineralizer for the removal of the various activation products and the minimum efficiencies of the spent fuel pool, miscellaneous waste, and polishing demineralizers (Figures 9-7, 11-1, and 11-2) for the removal of the various fission and activation products,
- c. The types of resins to be used in the demineralizers and factors, assumptions, and references used to determine the minimum efficiencies. .
- d. The smallest sieve or mesh designation of the filters (Figures 11-1 and 11-2) that will remove solid matter from the liquid radioactive vaste prior to discharge into the lake,
- e. An analysis of the possible accumulation in lake bottom sediments of the radionuclides discharged from the liquid waste system.
(
RESPONSE
- a. The erosion and corrosion product or crud activity concentration in the primary coolant is based on measured levels from Connecticut Yankee operating data since there is no acceptable means of calculating these concentrations in the coolant. Connecticut Yankee operating data was selected because inconel steam generator tubes are present. Canecticut Yankee found circulating erosion and corrosion product levels to be approximately 0.10 milligrams per liter during operation and with bursts of 1.0 milligram per liter during cooldown. Based on the Connecticut Yankee data, the erosion and corrosion product level is not expected to be greater than 10.0 milligrams per liter in the Davis-Besse Station. The activity concentrations in the coolant based on this data is listed as follows:
Co-58 8.h x 10-1 ue/ml
-3 Co-60 ~h.5 x 10 uc/ml _
Cr-51 1.h x 10-1 uc/ml
-2 Mn-5h 1.6 x 10 ue/ml
-2 1.6 x 10
( Fe-59 ue/ml 11.6-1
D-B
.G
- b. of all mixed-bed demineralizers The decontamination has been factor conservatively set(D.F.)3 at 10 for both cations and anions. This applies to the concentrate, primary, and spent fuel pool demineralizers.
The type of demineralizers, if any, that will be used to polish distillate from the evaporators, has not been determined yet. Therefore, no credit , is taken for them in deriving the theoretical efficiency of the system. Also, no credit is presently assumed for the removal of insoluble corrosion products by the resins, even though they actually make very good filters.
- c. The mixed-bed resins used vill be a combination of any of a number of commercially available nuclear grade cation and anion. re. sins (i.e. , Rohm and Haas IRN-78 and IRN-77, Duolite C-20ER and A-30R, etc.). Literature (1,2,3) on the use of such resins, as well as theoretical studies based on the work of Thomas, Klinkenberg, and Fletcher (4,5,6), have indicated that a D.F. greater than the 103 being used is achievable. '
- d. The minimum sieve or mesh designation of the filters will be determined by the particulate size distribution found in the waste produced by the operating stations. The filter. housings vill be designed to accept cartridges down to a very low rating, probably 1-3 microns. Ior analytical purposes, the filters are conservatively assumed to be only 10% effective in removing all insoluble corrosion products.
- e. The liquid vaste discharged from the station vill contain tritium as well as minute quantities of various soluble and insoluble radioactive '
species. Of these, only the latter have significant potential for 7l accumulating in the lake bottom sediments. They will consist mainly of corrosion products and perhaps, some resin fines from the polishing demineralizers. Before reaching the lake, the corrosica products will have been processed through one or more filters and/or demineralizers and an evaporator. Any particles that get through will probably be
-small enough to remain in suspension for some time. The resin fines, which contain little activity, have to pass through a final filter before release. This tends to remove any particulate matter large enough to settle quickly in the lake. Therefore, due to the small quantity and size of the insoluble radioactive species discharged, there seems to be no mechanism by which they can accumulate to any extent in the lake sediments.
a
,d was 11.6-2
D-B
References:
- 1. Dickinson, B. N. and Higgins, I. R., " Selective Ion-Exchange Sorption of Cesium-137 and Iodine-131 from Borated Solution", Nuclear Science and Engineering, 27, 131-139 (1967).
- 2. Rodger, W. A. , Lavroski, L. , and Burris, L. , Nuclear Engineering Handbook, Chapter 11, pp. 96,' Etherington, H., Editor, McGraw-Hill Book Company, 1958.
3 Rodger, W. A. and Fineman, P., Reactor Handbook, Vol. II, Fuel Reurocessing, pp. 500, Stoller, S. M. and Richards, R. B., Editors, Interscience Publishers, Inc. 1961.
- h. Thomas, H. C., J. Am. Chem. Soc., 66, 1665-1666 (19hh).
5 Klinkenberg, A., Ind. Eng. Chem., h6, 2285 (195h).
- 6. Fletcher, W. D. , " Ion Exchange in Boric Acid Solutions with Radioactive Decay", WCAP-3716, Nov. 1962.
I u "OZSS 11.6-3
i D-B C 4 12.0 Safety Ana3yses 12.1 Reactivity Transients 12.1.1 We note that your design does not incorporate a pressurizer level trip. Provide the bases for your conclusion that this trip is not required. State the pressurizer level assumed in your analyses of the startup accident and the accident resulting from rod withdrawal at rated power. State the = ass of steam dis-charged from the secondary system during these transients. In view of the fact that no pressurizer level trip is provided, give assurance that the integrity of the primary system would not be jeopardized if these transients were to occur with the pressurizer full. Indicate the discharge capability of the safety valves for releasing liquid water.
RESPONSE
Calculations have been performed for a range of rod withd{aval rates extending from very slow rates up to greater than 4 x 10- ak/k/s, a value more than four times as great as the maximum rod withdrawal rate for one rod group. These calculations show that the pressurizer will not fill due to insurge from the primary system because the high-pressure g trip will terminate the transient well before all the steam in the pres-surizer has been expelled. The normal level of the pressurizer is 18.33 feet of water at normal operating conditions. This value is used in the accident analysis. The secondary system is assumed to remove no extra heat from the primary system during a startup accident or a rod withdrawal from power. Thus, no secondary relief occurs. This very conservative assu=ption assures that the calculated insurge to the pressurizer is the maximum possible reactor coolant system volume expansion for the transient investigated. In the unlikely event that either a startup accident or a rod withdrawal from power were to occur with the pressurizer water at an abnormally high level, the high-pressure trip would occur much faster than normal. Only
. a very small amount of energy would be required to raise the system pres-sure to the trip point under these circumstances.
The pressure relief valves, should they be required to do so, are capable of relieving approximately 1,080,000 lb/h of water at 2500 psia. _ A pressurizer level trip is not required because the reactor is safe without such a trip. It is not justified to assu=e that eith0r a rod withdrawal accident from power or a startup accident should be ecmpounded by a full pressurizer. To provide the basis for this belief, it is necessary to [ discuss the two accidents separately because of the different circu= stances related to each accident. - i 0230 [ 12.1-1 >
DB O
~ The pressurizer contains a total volume of 1500 cu ft, its inside diameter )
is 7 ft and the distance between its upper and lower level sensing nozzles is approximately 35 ft. The pressurizer normally will contain 800 cu ft of water and 700 cu ft of steam. The volume in the upper portion of the pz essurizer (even when the level is at the high level alarm point) is 400 cu ft. If the pressurizer were full to the point where the upper level indicating nozzle vere covered, there would still be approximately 150 cu ft of steam in the pressurizer. During normal operation, the pressurizer level is automatically maintained at approximately 18 ft. It is not considered credible that the pressurizer could become full during power operation and then be followed by a rod withdrawal accident. Tbia is based upon. the fact that several malfunctions , and operator errors must occur to achieve a full pressurizer. If the normal level control system fails such that full makeup is provided through the "2" flow control valve, the flow rate would be limited to - approximately 136 gpm by the pu=p characteristics. With only one makeup pump running against a head of 2130 psia, the total available flow is 136 gpm. Normally, there would be so=e letdown so that the net makeup would be considerably less than 136 gpm. If two makeup pumps are operating, the capacity could be conservatively estimated at no more than 272 gpm. Abnormal makeup would produce a high pressurizer level alarm from any one of three independent sensors when the water inventory reached approxi- > mately 2240 gallons above normal. After approximately 2990 gallons total had been pumped from the letdcvn storage tank, a 1cv level alarm would be actuated in that tank. If it is assumed that the pressurizer level controller fails, with the flow rates already postulated, the times required to reach points with the pressurizer initially at normal level and the letdown tank at high operating level are tabulated below: No. of Time to Alarm (Min) Pu=ps Flev Failure Mode 3i Hi-Press Lo-Letdown Time to Fill Operating Gpm Water Source Level Tank Level Pressurizer (Min) 1 136 Single - 16.5 22 38 Letdown Tank 2 272 Double - 8.2 11 19 Letdown Tank
- In conclusion, compounding the assumption of a rod withdrawal accident from power with a full pressurizer is not justified on the basis of the single failure criterion.
3
)
j 12.1-2
D-B r8 i, In connection with the discussion of the startup accident and why a high pressurizer level trip is not required to protect against that accident, it should be pointed out that there are actually very few situations that involve filling the pressurizer completely. One of the few situations which does involve a solid system is a plant hydro test. Good operating practice vill call for a known pressurizer level to be established prior to performing any rod withdrawal operations . For example, when the reactor is being prepared for startup, the written pro-cedures vill call for a minimum pressurizer level rather than a normal or above normal level. The procedures will be written this way to accommodate the coolant sy' stem expansion and to minimize the heatup time for the pressurizer. To conduct any rod motion operations without a known level in the pres-surizer would violate written procedures. If the system is to be hydrotested, the pressurizer is first filled solid by venting at low pressure. Then the reactor coolant system is isolated and pressure in the system is increased. During the low-pressure portion of the hydrotest, a low-pressure trip input to the reactor protection system exists cnd, during the hi6h -pressure portion of the test, a high-
, pressure trip input to the reactor protection system exists. Thus, \ control rods cannot be withdrawn.
In view of the above, it is not considered credible to compound a startup accident with a full pressurizer. 1 l l 1 1 M 1 1 I V Y 09 a7, i ns.s: i 12.1-3
D-B l 12.1.2 Discuss the consequences of the startup accident occurring with
~-
minimum core flow and indicate how this operating condition will be treated in the technical specifications.
RESPONSE
Technical Specifications vill require that during reactor startup and power operation, a minimum of two reactor coolant pumps be operating before more than a single control rod can be removed from the core. There is no single control rod of sufficient worth that will render the reactor critical in a startup condition. Hence, if a startup accident occurs, a minimum of two reactor coolant pumps will be in operation. If one or more reactor coolant pumps are operating the flux-flow monitor and the flux-pump monitor vill act to terminate this accident well before system damage can occur. . ( c C . h~0293 12.1.2-1
D-B A 12.1.3 State the basis for your conclusion that a cold-water accident is not credible. Discuss the amount of backflov which may occur through inactive pu=ps when operating with ene or more primary coolant pu=ps idle. Indicate the volume and minimum temperature of the water in the idle portion of the loop.
RESPONSE
There are no check valves or other isolation valves in the primary coolant system and as a consequence, significant temperature differences cannot exist in the cold leg piping. The following statements su=marize system flow conditions for less than four pumps operating (a. minimum of two pu=ps will be operating whenever the reactor is made critical or in service): Three Pumps Operating
- a. Flow in the normally operating loop 68.8 x 10 lb/hr.
Steam generator flow in the one pump loop - 28.0 x 10 6 lb/hr. b.
- c. Flow through the operating pu.g in the one pump loop kl.7 x 10 6 lb/hr.
- d. Backflow through the down pu=p - 13.7 x 106 lb/hr.
One Pumo Each Loop 6 g a. Flow through each operating pu=p h2.5 x 10 lb/hr.
- b. Flov through each steam generator - 31.0 x 10 6 lb /hr.
- c. Backflow through each down pump - 11.5 x 10 6 lb/hr.
Two Pumus in One Loop
- a. Flow in operating loop - 71.6 x 106 lb/hr.
- b. Backflow through down loop - 11.5 x 106 lb/hr.
The volume of each 28 in. pipe branch including the coolant volume in the pump is 258.3 ft3 The temperature of any idle loop or branch vill be the core inlet temperature which depends on the number of steam generators in operation and the power level. 0294 12.1.3-1
D-B C 12.1.h \nalyze the consequences of scramming or inadvertently moving one or all of the xenon control rod assemblies.
RESPONSE
The xenon control rod assemblies cannot be tripped. The control rod drive mechanisms - for the xenon control rods are modified so that the roller nut assembly will not disengage from the lead screw on a loss of power to the stator. This is accomplished by the addition of two friction buttons to the segment ams. These buttons not only restrict the radial movement of the segment arms but also develop sufficient friction restraint to prevent any rotation of the segment arms relative to the lead screw such that the xenon control rod assemblies will not drift downward following a loss of power to the stator. Inadvertent movement of the xenon control rods provides a reactivity change. However, the fastest rate of resulting reactivity change can be handled by the automatic control system which would remove or insert control rods to : maintain the power level and reactor coolant system temperature. If the reactor is not in the automatic mode, the positive reactivity insertion due to inadvertent movement of the xenon control rods would be less than for a single control rod group withdrawal presented for the startup accident (PSAR Section 1h.1.2.2) and the rod withdrawal accident (PSAR Section Ih.1.2.3). Inadvertent movement of the xenon control rods might produce abnormal axial flux peaking. This will be analyzed for the final design of the Davis-Besse . station in the event that further automatic action is desirable. l 02.95 12.1.h-1 v -
D-B
- c. 12.2.1 Calculate the two-hour whole body and thyroid doses at the site boundary and the course-of-the-accident doses at the outer boundary of the low population zone from the rod ejection accident analyzed which results in failure of 4.1% of the fuel rods.
Consider (1) ECCS operation, (2) depressurization within thermal stress cooldown limits, (3) leakage to the reactor building from the pressure housing failure having the minimum possible flow area, (h) primary-to-secondary leakage occurring at the anticipated technical specification limit prior to the accident and continuing in accordance with orifice flow assumptions there-after until the decay heat removal system can be placed in oper-ation and (5) less of offsite power, thus making the condenser unavailable for decay heat removal. Terminate the analysis when the steam generators are no longer required for heat removal. To aid our understanding of your analysis, supply the following: a4 Mass of fluid discharged to the reactor building at two
, hours and at termination of steam generator boiloff. .
- b. Mass of fluid transferred to the secondary system for ths time period in 1 above.
- c. Justification for any partition factors employed.
- d. A plot of primary system pressure versus time.
RESPONSE
The loss of integrity in the reactor coolant system which results in the postulated rod ejection is assumed to result in a very small leak which releases a negligible amount of coolant to the containment vessel. The leak rate is too small to produce any significant pressure rise in the containment vessel, so radioactive release from this source can be neglected. It is, therefore, conservatively assumed that the depressurization of the reactor coolant system is not enhanced by this leak, so that a maximum of the radioactivity in the reactor coolant is leaked to the secondary system because of the assumed defective steam generator. Figure 12.2.1-1 depicts the depressurization of both the reactor coolant system and the secondary system for several hours after the rod ejection. The cooldown rate of the reactor coolant system is about half of that permitted by the thermal stress limits. Assuming an initial reference leak rate of 1 gpm from the reactor coolant system to the secondary, 600 lb. leak after 2 hours and 1600 lb. have leaked at the termination of the steam generator boiloff. The reactor coolant which leaks to the steam generator is assumed to contain fission products from:
- a. long term operation with 1% defective fuel. Concentrations of fission products in the reactor coolant is as shown in Table 11-3 of the PSAR.
v
- b. gap activity in all fuel rods which experience a DNB (4.1% of fuel rods).
12.2.1 u._ J
D D-B The amount of radioactivity released to the steam generator is: - Curies of Radioactivity Released Isotope In 2 Hrs. At Termination of LeaksuLr Kr-83m 0.38 .89 Kr-85m 2.57 6.02 Kr-85 21.5 50.4 Kr-87 1.39 3.26 Kr-88 4.59 10.8 Xe-131m 4.05 9.h8 Xe-133m h.91 11 5 Xe-133 h26.4 998.4 Xe-135m 1.41 3.31 Xe-135 3 55 8.30 , I-131 57.8 135.h I-132 9.46 22.2 I-133 13.0 30 5 I-13h 0.88 2.05 I-135 4.66 10.9 Based on this data the does are calculated to be as follows: ' 0-2 Hrs. Termination of Leakage s Thyroid 6.825 Rem 2.33 Rem Whole Body 0.009 Rem 0.00h6 Rem No partition factor has been used in the calculation of these doses. s
.**D 12.2.1-2 0 97
J 2400 2000 1600 E 5 i
$ 1200 ,
[ Reactor Coolant 800 Y l
\ Steam
\
c n x N x N 1 e - % s 4 0 0 1 2 3 4 5 6 7 8 l DEPRESSURIZATION FOLLOWING A R00 Time, Hours EJECTION ACCIDENT
- FIGURE 12.2.1-1 i i
D-B
~
- 12. 3 Equirment Malfunction 12.3.1 On page lh-16 of the PSAR, it is implied that the auxiliary feedvater pump can be started within 83 minutes folleving a loss of offsite pcuer. Assuming relief through the safety valves, what is the calculated maximum primary system pressure during the 83-minute period in which reactor coolant boiloff is required to dissipate decay heat?
RESPONSE
The auxiliary feedvater pu=p can be started within much less time than 83 minutes folleving loss of offsite power. However, if the pump was not started, the maximum primary system pressure during the 83 minutes following the loss of offsite power is the pressure setpoint (2500 psig) of the pressurizer relief valves. Even though the pressurizer fills with water, the pressurizer relief valve capacity for either steam or water is greater than the maximum surge flow (reactor coolant system expansion) ~ reaulting from decay heat addition.
]
O mss 12.3.1-1 l _l
D-B (' 12.3.2 Describe the techniques used to analyze steam generator blewdown following a steam line failure. Indicate the pressure drop and heat transfer assumptions employed. RESPONSI: An analog hybrid computer program has been developed to study the transient characteristics of the reactor coolant system and the steam generator during secondary load variations. The model includes a detailed analog description of the secondary side of the steam generator, including a simulation of most of the secondary system valves with appropriate flow rates and opening and closing times: (1) =ain feedvater and startup valves with associated stop valves, (2) emergency.feedvater valves, (3) steam bypass and relief valves, and (h) turbine stop valves. The model also includes energy balances for the principal steam generator components, the entire reactor coolant system (core, loops and steam generator) and the pressurizer (with both mass and energy transfer). The reactor kinetics, trip logic and action, and a fuel - pin simulation with Doppler and moderator temperature feedback are also features of the model. Further information.concerning this model can be found in Proprietary Topical Report BAW-10002, "Once-Through Steam Generator Research and Development Report". The heat transfer coefficients have been calculated to produce a high rate of cooldown of the reactor coolant system. The film coefficient on the reactor coolant side and in the boiling region on the secondary side are kept constant during the transient. The heat transfer caefficient in the superheat region has no effect on the accident results. The pressure drop is appropriately considered in the continuity of = ass, energy and momentum equations e= ployed. M 0300 12.3.2-1 i
D-B (-' 12.3.3 In the event of a steam line failure, what is the amount of water which must be boiled off in the steam generator not affected by the failure in order to depressurice the primary system? Assuming that the steam generator with the failed steam line was experiencing primary-to-secondary system leakage at the anticipated technical specification limit, indicate the volume of primary system water which would be released to the atmosphere during this accident. State the basis for your assumption that unit cooldown and leakage termination can be accomplished in three hours.
RESPONSE
In the unlikely event of a steam line failure, 7 76 105 lb of water must be boiled off in the unaffected steam generator dur g the 3-hour cooldown of the reactor coolant system. If the rupture occurs in the containment vessel, there vill be no reactor coolant release to the atmosphere. If the rupture occurs outside the isolation valves, the following sequence of events occurs: The reactor trips in about 6 seconds on low pressure or high neutron power. The operator then closes the steam isolation valves on the affected steam generator. The isolation valve closes in 5 seconds. Assuming 10 minutes for operator action, the atmospheric release is terminated in 10 5 minutes. Technical specifications will define the limits of secondary coolant activity under which reactor operation may continue regardless of the quantity or rate of leakage from the primary to the secondary system. Assuming an arbitrary rate of 1 gpm leakage into the affected steam generator, less than 17 gallons of primary coolant will be released to the atmosphere during this accident. The cooldown time is based on two reactor coolant pumps operating and a 25 percent capacity turbine 8 bypass. The decay heat removal system is initiated at 300 F after the reactor ecolant system is cooled down at 100 F/h. c 0301
<]llll 12.3.3-1
D-B 12.3.4 State the maximum stresses experienced by (1) the steam generator tube sheets in the event of a steam line rupture, and (2) the steam generator tubes and tube sheets in the event of a loss-of-coolant accident.
RESPONSE
The naxiv..m pressure differential from the reactor coolant side to the secondar/ side in the steam generator following any steam line break will be less than 2500 psi. The maximum primary membrane plus primary bending stress for this pressure differential is 39,700 psi. This is much less than the allevable stress of 55,900 psi. The added stress due to any temperature difference in the tube sheet is negligible. A loss-of-coolant accident could cause a maximum pressure differential from the secondary side of 1050 psi. In this condition, there is no rupture of the reactor coolant to secondary boundary (tubes and tube sheets) . The maximum primary membrane plus primary bending stress in the tube sheet is 15,900 psi, which is well below the allevable limit of h0,000 psi at 650 F. '( 0302 1 . 12.3.h-1 L --
D-B (~ 12.3 5 Your analysis of a steam generator tube rupture assumes that the affected steam generator vill be isolated as soon as possible. Discuss the means available to detect which steam generator experienced the failure and estimate the time required to detect and isolate it.
RESPONSE
A double-ended rupture of one steam generator tube with unrestricted discharge from each end will result in a leak rate greater than the normal makeup to the reactor coolant system. The reactor vill trip on low pressure in about eight minutes. The operator can recognize the problem from the low reactor coolant system pressure, the low pressurizer level and the earlier increase in radioactivity in the steam line from the affected steam generator. Approximately 15 minutes is then required to cool dovn the reactor coolant system to the temperature corresponding to the saturation pressure at which the atmospheric dump valve is set. If it is assumed that "' the operator takes corrective action after the reactor trips, the total time required to isolate the secondary side of the affected steam generator is about 23 ninutes. I l l l 1 l I l l l l l m aoao l l I 12.3.5-1 1
D-B 12.3.6 Assuming that the unaffected steam gens.ator is experiencing I primary-to-secondary system leakage at the anticipated technical specification limit prior to the tube rupture, state the volume of primary system water transferred to the unaffected steam generator during the cooldown of the primary system.
RESPONSE
Technical specifications vill define the linits of secondary coolant activity under which reactor operation may continue regardless of the quantity or rate of leakage from the primary to the secondary system. Assuming an arbitrary rate of one gpm leakage into the unaffected steam generator, the volume of reactor coolant system vater transferred to the unaffected steam generator during cooldown to 280 F is 80 gallons. Only 18 gallons are transferred during the time required to cool the reactor coolant system down below the secondary system atmospheric dump valve set point and isolate the affected steam generator. (See Question 12.3.5 above. ) l l l M 0304 12.3.6-1
D-B
,_s 12 3 7 To illustrate the safety margin which exists due to the inherent design of the facility, identify the largest missile originating from the turbine which would not penetrate the containment, control room, or spent fuel pool. To indicate the sensitivity of the analysis to your assumptions regarding energy absorption by the turbine casing, present this information assuming casing energy absorptions of 0, 25, 50 and 100 percent of your best estimate of absorption.
RESPONSE
Experience and tests have shown that turbine missiles could be generated only from the rotating parts of the low pressure turbine. The h3" wheel fragment (see Fig. 1 & 2) is the largest missile that vill not penetrate the contain-ment, control room or spent fuel pool. LOW-PRESSURE TURBINE h3" WHEEL The wheel capable of producing the largest missile is the last stage wheel of the low pressure turbine. See Fig. 1 & 2. Using the analysis techniques described in G. E.'s Report TR6TSL211, "An Analysis of Turbine Missiles Resulting from Last Stage Wheel Failure", it has been concluded that a 120 wheel fragment is the largest missile. The physical properties of the 120* wheel fragment are s - ,avized below TABLE 1 120 Wheel Fragment Missile Properties Wheel fragment weight 826h lbs. Fragment angle 120 Minimum Proj. area 5 17 ft 2 Marimum Proj. area 11 7 ft2 Failure speed 3190 rpm Initial translational velocity 676 FPS Energy translational 58.7 x 10 ft. lbs. Energy rotational 34.1 x 10 ft. lbs. Estimated velocity after leaving casing h09 FPS Estimated energy after leaving casing 21 5 x 10 6 ft. lbs. \ M05 l 12.3 7-1
D-B f ANALYSIS D L The analysis of the above mentioned turbine missile penetrations was carried ~' out using the following modified Petry formulas: D=KA p V' (Equi. tion 1) l' V' = log 10"1* E"" 215000 , D' = D 1 + e- " - (Equation 3) c' = T_ (Equation k) D MI SSILE ENERGY LOFdES Along the projeed.ile path, there is a certain amount of energy loss due to air drag which is proportional to the square of the velocity. Fp = Cp pAV 2 (Equation 5) 2 Upward Flight: dy = - gdt - C pAV2 y dt (Equation 6) Downward Flight: 2 dy = gdt - C pAVP-p dt (Equation 7) 2 The Turbine Manufacturer has advised that 6h". of the total energy of the turbine missile will be absorbed by the casing. We have adopted this as our best estimate of absorption and this has been used in our analysis. f a 3 12.3 7-2 0306
D-B TABLE 2 (
SUMMARY
OF PENETRATION CALCULATIONS
- 7. or best Estimate or Energy Absorption Exposed t by cu4. e
~
Surface Thickness 0 25' 50 100 of . . . . . , Building V D V D g D V D Containment Dcne 2h" 581 27 2" 543 24.9" 499 17" 375 7.h Control Room 18" Roof 670 32 5" 621 29.75" 560 25.9" 400 10.9'
& 12" Ceiling 30" Spent Fuel Pool Roof 18" 670 32 5" 621 29.75" 560 25 9" h00 LO.9" CONCLUSION:
It is concluded that the largest turbine missile vill not penetrate the shield building, control building or spent fuel pool area. See k- Table 2. The penetration values have been calculated without considering the energy lost in penetrating the turbine roof. The probability of formation of a turbine missile is remote. However if it should occur, the vertical trajectory must be between 89* and 90 to land on the shield building roof. In order to make a direct hit on the shield building vall, the trajectory ralst be between 16 and 50 from the horizontal. The designs will be such that a missile vill not cause a LOCA or prevent shutdown of the reactor. The spent 02e1 pool roof vill have hb" of Q-Deck en bottom of the 18" of l8 reinforced concrete slab. Beams 10 ft.c/c will be used to support the . Q-Deck. The total thickness of roof including Q-Deck vill be 22 ". This 6 ) vill restrain any spalled concrete from the under side from falling into the pool. W 1 1 J n307
- 6.. l l
l 12.3 7-3
D-B-NOTATIONS:
.3 D = Depth of penetration in a slab of infinite thickness in FEET. ')
D' = Penetration in a slab of a finite thickness T. T = Thickness of slab in feet. A p = by its crossectional area (PSF). Sectional pressure obtained by dividing the weight o V = Velocity of missile (FPS). K = Penetration material co-efficients experimentally detemined. V' = Velocity co-efficient factor. a' = Ratio of slab thickness to penetration thickness. , FD = Drag force in lbs. Cp = Drag co-efficient dimensionless used as 1.0. p = Air density lb-sec 2/ft . v = Air density, 1b/ft3 used as 0.074 lb/ft3, A = Projected area,
'N W = Weight of missile in lbs.
M = Missile mass, lb-sec 2/ft. g = Acceleration of gravity, ft/sec. REFERENCFS:
- 1. General Electric Report, TR6TSL211, "An Analysis of Turbine Missiles Resulting from Last-Stage Wheel Failure".
- 2. Design of Protective Structures by Archam Amirikian "NovDocks P-51".
3 6N 12.3 7-4
- oaos
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i FIGURE 12.3.7-3 l 1 l
D-B I 12.h Fuel Handling 12.h.1 Discuss the provisions that vill be made to prevent the dropping of the spent fuel element cask into the spent fuel storage pool. If the spent fuel element cask or other heavy objects must be moved over the spent fuel storage pool, analyze the consequences of dropping a cask or other object into the pool. Consider the possibility of (1) fuel clad damage, (2) loss of pool water and ability to continue cooling the spent fuel, (3) damage to other equipment by flooding if the integrity of the pool liner is lost and (h) damage to the gas decay tanks located beneath the pool if the cask can penetrate the base of the pool.
RESPONSE
As shown in Figure 1-7 of PSAR Amendment No. 2, the spent fuel cask pit is . ecmpletely separated from the spent fuel pool by a three (3) feet thick concrete vall. The bottom of the cask pit is solid concrete down to the foundation. The cask handling crane hook travel vill be limited to the centerline of the cask pit with the help of mechanical stops and/or electrical interlocks. The cask pit is approximately 14' square, therefore, should the cask topple over by hitting the vest vall of the cask pit the center of gravity of the cask will be vell within the cask pit area thus eliminating any possibility of its falling into the spent fuel pool. casa 12.h.1-1
D-B l 12. 4.2 Describe the manner in which containment integrity is maintained during refueling operations. Describe the mode of operation of the containment building ventilation system. If the containment building is not isolated during refueling operations, provide the time required to detect and isolate any potential radioactivity releases during refueling. Indicate the potential fraction of the total fission products reaching the containment building atmosphere that could escape prior to isolation.
RESPONSE
During refueling, the fuel transfer tube between the containment vessel and the auxiliary building vill be open and under water. The containment vessel personnel lock will generally be open and the equipment hatch may also be' open. The containment purge air system, as well as containment air coolers will be in operation. The high radiation monitor of the containment vessel and the station vent stack which controls the isolation ! valves for the containment purge air system vill bc in service. Based on the analysis of the fuel handling accident in the auxiliary building as described in Section 1h.2.2.1 of the PSAR, it is not anticipated that the containment integrity need be maintained during refueling. However, following such an accident the containment purge air system can be isolated within 10 seconds. Assuming that all the released radioactivity mixes with the containment atmosphere instantaneously, it is estinated that about 0.3% of the activity . released would have escaped to the environment prior to isolation of the purge valves. There should be no significant release of radioactivity through the open equipment hatch and the personnel lock. e a ..... W 0314 12.h.2-1
D-B f 12.h.3 Analyze the refueling accidents in the containment and auxiliary buildings assuming damage of an entire fuel assembly, a release to the refueling water of 10% of the iodines and 20% of the noble gases from the hottest fuel assembly, a refueling water retention factor of 10 for iodines, Pasquill Type F meteorological con-ditions, an atmospheric diffusion factor determined by F. A. Gifford's Equation (Nuclear Safety Vol. 2, No. 4, June 1961) and a building shape factor of c = 0.5 Using the results of the analysis, discuss the adequacy of the facility to limit the accident doses to within *he guidelines of 10 CFR 100. PESPONSE We do not agree that the assumptions set forth in this question should be the basis for analyzing this accident, however the following analysis is based on the above conditions. Dilution factors for the accident situation were computed from X = Q _,_ 8 (v oo y + CA) U Where C = 0.5, A = 2600 2m , U - 2m see-1 for two (2) hour release. Sigma values are for Pasquill "F" category. Distance 2-Hour h 7 0 73 km 2.31 x 10 1.0 1.79 2 9.0h x 10-5 h 3.85 6 2.23 8 1.56 13 8.58 x 10-6 16 6.62 2h h.17 , 32 3.00 ho 2.32 In calculating the. dose at the site boundary the fuel assembly is assumed to have been in operation at 2h.6 MW for 930 days. A decay period of 72 hours has been . allowed before the rupture of all 208 pins occurred. All the activity in the fuel assembly is released to the pool. No credit for any ( decay beyond 72 hours has been taken. Atmospheric dilution is calculate ' sing 2 hour disperion factor. The total integrated dose to the whole body at the exclusion distance is 2.06 Rem, and the thyroid dose at the same distance is 6h9 Rem. 12.h.3-1 g 03J5
D-B
/ 12.5 Emergency Ventilation System
( 12.5.1 In order that we may determine the adequacy of the auxiliary and shield building ventilation systems which may be necessary to reduce the potential doses from several classes of accidents, provide (1) the radiation source terns and heat loads used for the design of air cleaning equipment in these systems, (2) a discussion of the ability of the filter and charcoal units to withstand desorption and/or ignition with loss of air flow, with these heat loadings, (3) the anticipated accident atmospheric
- enditions to which these systems vould be exposed, (h) a de-
> cription of the design of the equipment to withstand these conditions, and (5) the method of testing the equipment to demonstrate that it vill function as designed under accident conditions.
RESPONSE
- 1. There are two ventilation systems provided for the shield building.
These are the purge system which can be used during normal operation and the emergency system which starts automatically following a containment isolation signal. The purge system is described in Section 5.h of the PSAR. The emergency system is described in Sections 5 5 and 6.3 of the PSAR and in answer to question 6.8. [ With respect to the auxiliary building ventilation system, two accidents resulting in activity release are considered. A fuel handling accident is discussed in Section 14.2.2.1 of the PSAR; rupture of a gaseous radvaste decay tank is analyzed in section lk.2.2.5 In both cases, the corresponding heat loads are negligible.
- 2. On the basis of the analysis of accidents to which the auxiliary ventilation system might be exposed, together with the fact that fans in system vill be redundant, the possibility of ignition of roughing or HEPA filters is not considered credible. The ability of the shield building emergency system to withstand anticipated heat loadings is discussed in part c of question 6.8.
3 The post-accident atmosphere within the auxiliary building vill not differ significantly from the previously existing ambient conditions of temperature and relative humidity. The post-accident atmosphere in the shield building is discussed in part a of question 6.8.
- h. All fens and filters are located in the auxiliary building. Therefore, the maximum amount of the radioactive iodine that the charcoal filters receive is 0 5%/ day of the amount presert in the containment atmosphere. ~
The heat generated by the deposited radi> active iodine vill not raise the temperature above the desorntion and/or ignition limits. The equipment is designed to operate in maximum ambient canditions of 110 F and 90% humidity. s. a 12 5.1-1 i
D-B
/3 3 5 The auxiliary building, purge and emergency ventilation systems are accessible for testing and maintenance during normal station operation.
)
The charcoal filters will be tested and qualified with Freon 12 or 112 under the test procedures described in USAEC Reports DP 870, DP 910, DP 950, DF 1053, and DP 1082. The high efficiency filters vill be tested for penetration with the efficiency-penetration test (DOP) using homogeneous particles of dioctyl phthalate. t 0 J i
+ ,
12 5 1-2 t
D-B f~~ 12.6 Containment Pressure Response 12.6.1 Heat transfer to the heat sinks from the reactor building atmo-sphere is determined by a " Modified Tagami" heat transfer coefficient in the COPNITA program. Provide a discussion of the methods used to correlate this coefficient with the Tagami, Uchida, and Kolflat data. Present a plot of the value of the condensing heat transfer coefficient as a function of time.
RESPONSE
Please refer to the Bechtel proprietory information submitted in the Midland PSAR Amendment No. 5, Response to Question 13 7 3.1. e s-e e I q 0318
. W 12.6.1-1 1
D-B (~ 12.6.2 Discuss the manner in which the variation in turbulence in the ( various compartments of the containment building is considered in applying the " Modified Tagami" coefficient to the entire containment building.
RESPONSE
Please refer to the Bechtel proprietory information submitted in the Midland PSAR Amendment No. 5, Response to Question 13.7.3.2. S 0319 12.6.2-1 t
D-B r~'- 12.6.3 Indicate how painted surfaces within the reactor building are I w considered in the determination of heat transfer to painted heat sinks. Discuss the effect en the condensing heat transfer co-efficient and state the thickness and thermal conductivity of the paint layer assumed. List the areas of both painted metal and painted concrete surfaces within the reactor building.
RESPONSE
Please refer to the Bechtel proprietory information submitted in the Midland PSAR Amendment No. 5, Response to Question 13 7.3.3. Areas of painted containment vessel interior surfaces for the Davis-Besse station are as follows: Metal Concrete Painted 191,000 ft2 93,000 ft2 Unpainted 0 0 l 0320 12.6.3-1
D-B 12.6.h To provide a better understanding of the sequence of events in (;N your analysis of the design basis accident; provide a chronology indicating the time after the pipe rupture occurs that the following events occur: (1) core flooding tanks start injection, (2) reactor building pressure reaches peak pressure, (3) primary system blowdown is completed, (4) core flooding tanks empty, (5) emergency injection starts, (6) containment spray starts, (7) air cooling units start, (8) recirculation mode starts, and (9) containment returns to atmospheric pressure, assuming (a) two air cooling units and one spray pump operate for containment heat removal, and (b) three air cooling units and two spray pumps operate.
RESPONSE
- a. 3.0 SQ FT BREAK (3000 gpm injection, 2 coolers, and 1 spray)
Event Time After Rupture (sec) *
- 1. CFT Start Injection 22
- 2. Peak Pressure in Containment Vessel 30 3 Primary System BlowCown Complete 35
- h. CFT Empty 47.5 l
5 Emergency Injection Starts HPI 25 LPI 47 5
- 6. Containment Spray Starts 35 T. Air Csolers Start 30
, 8. Recirculation Mode Starts h928
- 9.
- Containment Vessel Returns to 1.0 psig Pressure k.0 x 100
- b. 3.0 SQ FT BREAK (7000 gpm injection, 3 coolers, and 2 sprays)
Event Time After Rupture (sec)
~
- 1. CFT Start Injection 22
- 2. Peak Pressure in Containment Vessel 30
. 3 Primary System Blowdown Complete 35
- h. .CD Empty 0:121 hT.5 a.2.6.h-1
D-B 5 Emergency Injection Starts h,
/
HPI 25 LPI 47,5
- 6. Containment Sprays Start 35 7 Air Coolers Start 30
- 8. Recirculation Mode Starts 2225
- 9.
- Containment Vessel Returns to 1.0 psig Pressure ~ 1.34 x 10h "The containment vessel pressure analysis computer program could not take
~
the pressure down to zero. D 1 0322 s 12.6.4-2
D-B 12.6.5 Provide a table indicating energy distribution prior to and (1 following the loss-of-coolant accident, the ar ount of energy \ generated and absorbed from the time of pipe rupture to the time of peak pressure, and the distribution of energy at the time of peak pressure. This listing should include at least the following for the break sizes indicated in the PSAR: (1) primary coolant internal energy, (2) core flood tank internal energy, (3) energy initially stored in the core, (k) energy stored in core internals, (5) energy stored in reactor vessel metal, (6) energy generated during shutdown and decay heat, (7) energy transferred to the steam generators, (8) energy stored in piping, valves, and pumps, (9) energy in steam generator metal, (10) secondary coolant internal energy, (11) energy content of water vapor in the reactor building, (12) energy content of air in.the reactor building (13) energy content of water in the reactor building, (14) energy content of steel structures, (15) energy content of concrete internal structures and the reactor building valls and dome, (16) energy removed by the air handling units, and (17) energy removed by - the containment spray.
RESPONSE
Energy distribution for 3.0 ft 2 break, assuming 3000 gpm injection, 1 spray, and cooler. Energy Energy Generated Distribution from at Time of Energy Time 0 to Peak Pressure, Distribution Time at Peak (30 Sec), Btu Prior to Pressure, LOCA, Btu Btu 0
- 1) Reactor Coolant 306.15 x 10 0 37 7 x 10 Internal Energy
- 2) CFT Coolant Internal Energy 9.1 x 106 0 6.5 x 10 6 6
- 3)
- Energy Stored in Core 29 1 x 10 0 8.3 x 10 6 6 h)
- Energy Stored in RV 10 5 x 10 0 7.6 x 10 Internals
- 5) " Energy Stored in RV 30.h x 10 6
0 30.2 x 10 6 Metal 0
- 6) Energy Generated During 0 9.6 x 10 0 Shutdown + D.H.
- 7) Energy Transferred to 0 0 0
-' Steam Generator
- 8)
- Energy Stored in Piping 22.3 x 10 6
0 21.6 x 10 6
& Pumps y oaea 12.6.5-1
D-B
- 9)
- Energy Stored in Steam 31.3 x 10 0 31.3 x 10 6 Generator Metal 6 6 -
- 10) Secondary Coolant Internal 72 x 10 0 72 x 10 Energy (Both steam gen.)
6
- 11) Energy Content of Water 10.14 x 10 0 251 7 x 10 Vapor in Containment Vessel 6
- 12) ** Energy Content of Air in 0 0 h.1786 x 10 Containment Vessel
- 13) Energy Content of Water in 0 0 hh.75 x 10' Containmer.t Vessel
- 14) ** Energy Content of Steel 0 0 h.172 x 10 6 and Steel Sheathed Structures -
6
- 15) ** Energy Content of Unsheathed 0 0 9 583 x 10 ,
Internal Concrete Structures and Containment Vessel Walls and Dcme
- 16) Energy Removed by Air Handling 0 0 0 Units
- 17) Energy Removed by Containment 0 0 0 Spray
- Stored Energy Above 280 F Datum Level
** Initial Energy Content (@ 120 F) Set at 0 0324
- s I
1;. 6.5-2
D-B l 12.6.6 Provide plots of the temperature of the steam-air mixture within
- s. the reactor building vs. time and of the reactor building sump water temperature vs. time assuming:
(1) reactor building heat removal assuming (a) two air j cooling units and one spray pump operate and (b) three ; air 4.coling units and two spray pumps operate. (2) emergency core cooling assuming (a) one independent pumped injection train operates and (b) the two independent pumped injection trains operate.
RESPONSE
Figures 12.6.6-1 to 12.6.6-6 give the containment vessel steam-air mixture and sump temperatures as a function of time folleving 14.1, 8.5, 5.0, 3.0, . and 1.0 ft2 hot leg ruptures. Plots are given for all break sizes assuming
- two air cooling units, one spray pump, and one independent pumped injection train in operation. In addition, the 3.0 ft2 design case is given with three cooling units, two spray pumps, and two independent pumped injection trains operating.
( 4 N 3 i l ! i-- 0325 l 12.6.6-1
D-B ( 12.6.9 Provide your analysis of the potential hydrogen evolution to containment during the period following the post-loss-of-coolant accident. Clearly state your assumptions. Describe your plans regarding use of combustible gas control =easures, including consideration of measures which do not entail the intentional release of fission products from the containment to the atmosphere.
RESPONSE
There are four (h) sources of hydrogen inside containment vessel during accident conditions. These are as follows: I. Hydrogen released by radiolytic decomposition of water. II. Chemical hydrogen released by Zirconium - water reaction. III. Evolution of dissolved hydrogen contained in the reactor coolant. IV. Hydrogen released by zine-boric acid reaction. A discussion of each of these sources and the magnitude of the hydrogen release is as follows: I. Radiolytic Hydrogen The calculations for the radiolytic hydrogen generation following the LOCA vere based on the following assumptions:
- a. The core has operated at 2772 MWt for an average irradiation period of 620 days.
- b. Hydrogen generation constant ("G" value) Core - 0.h3 molecules H2 /100 eV absorbed energy.
- c. Contribution due to the noble gases in the containment is not significant by comparison to the other sources, therefore it is neglected.
Calculations show that the hydrogen generation rates vary from 0.5 cfm at 500 hours after the accident to 0.2 cfm at 2000 hours. Figure 12.6.9-1 gives a curve of total radiolytic hydrogen generation versus time after LOCA. II. Zr-Water Hydrogen The extent of zirconium-water reaction is contro1~ed by the emergency core cooling system performance. The Davis-Besse ECCS system is designed to limit the hot spot fuel temperature to 2300 F or less during a LOCA, thereby limiting the zirconium-water reaction to less than 0.1 percent. However, for conservatism one (1) percent zr-water reaction has been assumed.
\_ ,
A one (1) percent reaction produces 3,320 cubic feet of hydrogen. 12.6.9-1 0326 L
D-B
,O g
)
III. Dissolved Hydrogen The reactor coolant system is operated normally containing 15 h0 cc of dissolved hydrogen per liter of coolant. This hydrogen is partially released as the coolant is depressurized and cooled during the accident. The total amount of dissolved hydrogen contained in the coolant is only h72 scf, therefore, this is considered an insignificant source. IV. Zine-Boric Acid Hydrogen Tests at the Franklin Institute performed under LOCA conditions to study the hydrogen evolution due to chemical reaction between inorganic-zine paints and H 30 3 -N3 S223 0 -Na0H solution have shown insignificant amounts of hydrogen generated. Since there are no . chemical spray additives in the Davis-Besse containment spray system, the hydrogen generation vill be even smaller. 8l A1 can be seen, the total hydrogen generated over a period of ten weeks is approximately 56,000 cubic feet or about 2% of the containment free volume. It is recognized that a potential hydrogen problem may exist following a LC CA. Since there is no one tested and acceptable solution available, at
^
this stage, we are not in any position to specify a definite design of the hydrogen control measures for the Davis-Besse Station. We are studying the following three methods as possible solutions.
- 1. A review of the containment air recirculation system to provide best circulation in order to avoid any possibilities of creation of hydrogen pockets.
- 2. Purging of containment atmosphere.
- 3. Catalytic recombiners.
A definite design recommendation vill be made at a later stage. I i gy 12.6.9-2
D-B
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