ML19319D593

Forwards Request for Addl Info Re FSAR Sections.Info Should Be Submitted by 720214 to Maintain Review Schedule

ML19319D593
Person / Time
Site:	Crystal River
Issue date:	01/17/1972
From:	Deyoung R US ATOMIC ENERGY COMMISSION (AEC)
To:	Rodgers J FLORIDA POWER CORP.
References
NUDOCS 8003170763
Download: ML19319D593 (40)

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m - determine whether the ground water environment would allow future usern to ne water f rom beneath the site. 2.3 It is Indicated on page 2-2 that the Gulf of Mexico is used for barge traffic to deilver coal to Units 1 and 2 adjacent to the nuclear site. Units 1 and 2 have been converted to oil with approxinutely 1,000,000 gallons on site storage capacity in four tanks. 2.3.1 Assume an oil barge releases oil to the Gulf of Mexico in the vicinity of the site. Provide a description of the ef fects of the released oil on the operation of the main condenser circulating water system, the engineered safety feature cooling water system, and other systems that might be affected. 2.3.2 Describe the design provisions which prevent or nitigate the safety consequences of on site oil tank ruptures which might be caused by a hurricane. or earthquake. 2.3.3 Assume an oil tanker explodes in the closest normally used channel in the vicinity of the site. Provide a description of the effect of the explosion on the opera-tion of plant systers that might be affected. i 2.3.4 Assume a barge >r ship becomes uncontrolled during a storn j in the vicinity of the turning basin or fuel dock. Provide a description of the effect of the impact of the uncon-trolled barge on the plant intake structure, the emergency cooling water intake structure, and any dikes and intake canals that might be affected. 2.3.5 Assume a barge releases chlorine gas in the vicinity of'the site. Provide a description of the effect of the released gas on the performance of the reactor control room ventila-tion system and other systems that might be affected. In-clude a discussion of the protection that is provided for the reactor operators if chlorine can get into the control room and other regions of the plant that might be affected. Provide a description of the location, quantity, chemical composition, and physical state of the chlorine stored on the site and the effects of potential chlorine gas releases from these sources on the safety of reactor operation. 1 2.4 Provide a description of existing airports and missile bases within 10 miles of the plant and of proposed airports within 5 miles of I l

n \\ . Engineering Research Center; and " Estimation of Hurricane Surge llydrographs," by' G. Marinos and J. W. Woodward, ASCE paper 5945, May, 1968, WW2. This is a more accurate method than was generally available during the /SAR stage. The probable maximum hurricane parameters are from NOAA Report ilUR 7-97. Provide the following information to support the Pt!H surge and associated wave estimates: 2.1.1 Detailed computations of the PMH surge stillwater level using probable maximum parameters and the referenced bathystrophic storm' tida theory (or another method which can be substantiated), a cross section along the assumed fetch, and the hurricane track; 2.1.2 Estimate the critical significant and maximum (one percent) wave heights and resulting runup on safety related facilities in the spectrum of waves which can be associated with the PMll; 2.1.3 Document the ability of safety related structures to with-stand the static and dynamic effects of the PMH and associated wave action without loss-of-function, and the ability of safety related equipment to operate during such an event; 2.1.4 Provide detailed computations of a PMH along a critical traverse which could produce the minimum water level at the-site, and verify that suffic'ent pump suction would exist during such an event; 2.1.5 Provide documentation of the model studies of wave runup such as illustrations showing the runup as a function of stillwater level and wave height; 2.1. 6 Determine whether wave action within the intake and dis-charge canals for the PMH can have an adverse effect on safety related structures and equipment. 2.2 Provide a map which shows the location of the public water supplies tabulated on Figure 2-22. Also show the location of any existing or known potential future private wells within two miles of the site. Tabulate similar data for the private wells. Determine whether it is possible for existing ground water users, under adverse conditions, to pump water from beneath the site. Similarly,

s <m a REOUEST FOR ADDITIONAL INFORMATION FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NO. 50-302 2.0 SITE AND ENVIRONMENT 2.1 We and our consultant, the U. S. Army Coastal Engineering Research Center, do not agree that the estimated hurricane surge discussed in section 2-4.2.1 is the highest that can be generated at the site. Furthermore, we believe that a large radius, relatively fast moving hurricane with probable maximum characteristics could produce a significantly higher surge. For example, ilurricane Camilie (1969) induced a water level elevation of 24.2 f t. (MSL) measured at Pass Christian, Miss issipp i. Although it was not in the class of the probable maximum hurricane (PMil), the resulting surge level is about equivalent to the estimated maximum surge presented in section 2.4.2.1. Our consultant has made some preliminary parametric computa-tions which indicate a surge stillwater level in excess of +35.4 ft. (MLW) based on PMll parameters as follows: PARAMETERS Central Pressure 26.70 Inches Hg Peripheral Pressure ~ 31.25 Inches IIg Radius of Maximum Winds 24 Naut. Miles High Spring Tide Elevation 4.3 ft. (MLW) Initial Rise 0.6 ft. Track of Hurricane N63*E(True) PRELIMINARY SURGE ESTIMATES Forward Speed of Translation / Surge Elev. 4K/29.3 ft. (MLW) Forward Speed of Translation / Surge Elev. ilk /32.1 ft. (MLW) Forwa-i Speed of Translation / Surge Elev. 20K/35.4 ft. (MLW) The surge stillwater estimates were based on bathystrophic store tide theory as discussed in our consultant's Technical Memorandum No. 35, " Storm Surge on the Open Coast: Fundamentals and Simpli-fied Prediction" by U. S. Army Corps of Engineers, Coastal - - + - -

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3 ,s _4_ l the plant. For existing airports, provide a description of the aircraft using then including the weight of the ai rc ra f t, the number of takenf fs and landings, and the flight paths. 2.5 Page 2-2 indicates a railroad track spur comes into the plant site. Assume a railroad accident releases hazardous chemicals in the vicinity of the site. Provide a description of the effects of the released chemicals on the operation of plant systems that might be af fected. Describe any limitations that might limit the nature and likelihood of such assumed releases. 2.6 Appendix 2A, the Wind Dif fusion Program Computer Printout for Crystal River Unit Number 3, tabulates approximately 60% of the data that could be obtained if the meteorological instruments and the NUS wind variance computer had functioned continuously. We expect at 1 cast 90% data recovery. Provide a tabulation of time periods when no data were recorded and state the reasons for not recording these data. Provide a tabulation of the meteorological parameters calculated by the WIND VANE computer code using meteorological data for a 12 month period (the table attached in Appendix 2A uses meteorological data for a 25 month period) and provide a tabulation of the joint frequency distribu-tion of wind speed and direction from the 30-foot level by wind directional variance (sigma theta) classes from the 150-foot le ve l. for data obtained during the 365 consecutive day period that has the best data recovery. If the range of wind direction obtained from strip chart recorders and computer data is used to improve data recovery, compare the results obtained by each method for a time period when both the computer and the strip chart recorders were operating. If a reanalysis of the meteorological . data indicates any significant change in the accident or annual average X/Q values used in those analyses, these revised y/0 values ~ should be supplied. 2.7 The original Crystal River Unit 3 Environmental Report Indicates that a second set of meteorological instrumentation was installed on the 150-foot meteorological tower in July 1970. The FSAR does not include information on this instrumentation. Please include the data from this elevation in your analysis of both routine release and accidental release meteorology used for the Crystal Hiver facility. 2.8 The population statistics (FSAR page'2-1) are based on the 1960 cens us. Provide revised population values based on the 1970 census, and appropriately revise Figure 2-6.

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. 2.9 Clearly delinence, including appropriate indication on a map, the boundaries which you propose for the restricted area for purposes of 10 CFR Part 20.106. Is there public access immediately adjacent to the boundary of the restricted area? If so, describe the n,ature and extent of use of such access. 2.10 So that the staff may complete its review of the foundation CoRditions following the foundation exploration and treatment program described in the PSAR and later amendments, we request that you submit for review: Gilbert Associates (C. A. I.), Inc. SP-5500 Speci fications for Subsur. face Grouting, Crystal River Unit No. 3, Florida Power Corporation, Feb. 28, 1968. C. A.I., SP-5785, Proof Testing of Chemically Consolidated Foundation Elements, Crys tal River Unit No. 3, Florida Power Gorp., Jan. 7, 19 70. The independent engineering evaluation made of the entire chemical grouting program by an "outside consultant to the owner" that was mentioned on page 2-46 of the FSAR. 2.11 On page 2-36 of the FSAR you state: " Comparison of the imposed loading with the conservatively estimated shearing strength of the foundation materials indicated that an adequate factor of safety against a bearing capacity failure would be achieved under the most unfavorable conditions which could be reasonably postulated. This conclusion, however, was predicted on the assumption that all significant volds occurring above elevation +30 would be filled so as to minimize local overstressing and possible future progressive failure." 2.11.1 Provide assurance of the conservatism of the predicated assumption. 2.11.2 State the computed tactors of safety for all adverse conditions analyzed, including seismic loading conditions. 2.12 Page 2 of the report, Foundation Investigation. Florida Power Corporation. Proposed Nuclear Power Plant. Unit No. 3. by Woodward-Clyde Associates states: "The Reactor Building foundation mat is reported to 0mpose an average unit load under operating conditions of about 7. 8 Ks f. The maximum contact pressure of the Reactor Building mat under operating and under 1.5 accident pressure conditions is estimated to be 10.3 and 23.4 Ksf, respectively." i

,. 2.12.1 Demonstrate how the values 10.3 and 23.4 Ksf were obtained. 2.12.2 What Is meant by 1.5 accident pressure conditions? 2.12.3 Show that the ' foundation will support the 10.3 and 23.4 Ksf pressure from the Reactor Building mat. 2.13 Page 5 of the Woodward-Clyde report states: "It is also pertinent to note that no evidence of active subsurface sub-sidence (sink holes) due to solution activity has been reported within the proposed plant site." Is there evidence of active or inactive subsurface subsidence outside of the plant site, that is, within several miles of the site? MS i

- 1.0 RI: ACTOR 3.1 The bases for Technical Specification 15.2.1 indicate that the limiting Fy for DNB is 1.50 with a symmetrical modified cosine axial power shape. 3.1.1 Provide details on configurations analyzed to indicate why symmetrical power distributions with an axial peaking factor in excess of 1.50 cannot occur. Is there an experimental basis? 3.1.2 What is the margin between the worst expected axial peaking condition and the 1.5 limit? 3.1.3 What is the axial peaking factor for the worst condition with the part length rods in the bottom of the core and the maneuvering bank positioned at the top of the core, or what restrictions will prevent such a configuration? 3.1.4 How will the symmetrical axial peak be monitored during operation of the power plant, since the out-of-core instrumentation senses only tilts? 3.1.5 What are the axial peaking factor limits for peaks in the top of the core? 3.1.6 In view of the sensitivity of the core limits to axial locationoftheaxialpeak(thedesignFEisstatedto be 1.70, not 1.50), have you considered more comprehensive definitions of the design and limiting peaking factors? t

s s 5 , 4.0 REACTOR COOLANT SYSTEM 4.1 To evaluate the adequacy of the proposed heatup and cooldown limits for this plant, provide the following information: 4.1.1 For all pressure-retaining ferritic components of the reactor coolant pressure boundary whose lowest pressurization tempera-ture* will be below 250*F, provide the material toughness properties (Charpy V-notch impact test curves and dropweight test NDT temperature, or others) that have been reported or specified for plates, forgings, piping, and weld material. Speci fically,,for each component provide the following data for materials (plates, pipes, forgings, castings, welds) used in the construction of the component, or your estimates based on the available data: (a) The highest of the NDT temperatures obtained from drop wetight tests, (b) The highest of the temperatures corcesponding to the 50 ft-lb value of the C fracture energy, and y (c) The lowest of the upper shelf Cy energy values for the weak direction (WR direction in plates) of the material. 4.1.2 Identify the location and the type of the material (plate, forging, weld, etc. ) in each component fo'r which the data listed above were obtained. Where these fracture toughness parameters occur in more than one plate, forging, or weld, provide the information requested in 4.1.1 (a), (b), and (c) for each of them. 4.1.3 For reactor vessel beltline materials, including welds, j provide the information requested in 4.1.1 and 4.1.2 and in addition specify: i

• Lowest pressurization temperature of a component is the lowest temperature at which the pressure within the component exceeds 25 percent of the system normal. operating pressure, or at which the rate of temperature change in the component material exceeds 50*F/hr., under normal operation, system hydro-static tests, or transient conditions.

-w _9_ (a) The h ighest predicted end-of-li fe trans ition temperature corresponding to the 50 ft-lb value of the Charpy V-notch fracture energy for the weak direction of the material (WR direction in plates), and (b) The minimum upper shelf energy value which will be acceptable for continued reactor operation toward the end-of-service life of the vessel. 4.2 For the predicted NDT temperature shif t o f 2 50* F (7S AR, page 4-25) at least 5 capsules are required by the AEC proposed " Reactor Vessel Material Surveillance Program Requirements " 50.55a, Appendix 11, published in the Federal Register on July 3,1971. Each of these surveillance cap- <ules should include specimens from the base metal, heat-af fected zone and the weld metal, as recommended in the ASTM E-185, Section 3.3. Section 4.3.3 of the FSAR refers to the report BAW-10006 for the description of the surveil-lance program consisting of 6 capsules, only 3 of which contain weld metal specimens. In effect, the proposed surveillance program consists of only 3 capsules containing ) the required number. and type of impact specimens. Provide the proposed capsulp withdrawal schedule, and describe how the fracture toughness of the weld metal will be monitored throughout the life of the reactor vessel. J 4.3 Describe the plans which were followed to avoid partial or local severe sensitization of austenitic stainless steel during heat treatments and welding operations for core structural load bearing members and component parts of the reactor coolant pressure boundary. Describe welding methods, heat input, and the quality controls that were employed in welding austenitic stainless steel components. 4.4 If nitrogen was added to stainless steel types 304 or 316 to enhance its strength (as permitted by ASME Code Case 1423 and USAS Case 71), provide justification that such material will not increase the tendency to sensitization at heat af fected zones during welding. i

ss . 4.5 Since the process of electroslag welding will be used in the f abrication of components within the reactor coolant boundary (i.e., steam generator shells and pressurizer shell), describe the process variables and the quality control procedures applied to achieve the desired material properties in the base metals, heat affected zones, and welds. 4.6 Indicate whether electroslag welding will be used in fabrication of any other components, particularly those made from stainless s tee l. 4.7 Provide the following information regarding the primary coolant pump flywheels: 4.7.1 State the type of material used for the pump flywheel and its minimum specified yield strength. Indicate the nil-ductility transition (NDT) temperature specified for the materials, as obtained from drop-welcht tests (DWT), the minimum acceptable Charpy V-notch (Cy) upper shelf energy level in the weak direction (WR orientatien in plates), and the fracture toughness of the material at the normal operating temperature of the flywheel. 4.7.2 Describe the nondestructive examinations to be performed on the finished flywheel. State if the flywheels will he subjected to a 100 percent volumetric ultrasonic inspection using procedures and acceptance criteria equivalent to those specified for Class A vessels in Section III of the ASME not ler and Pressure Vessel Code. Describe the surf ace inspections to be performed on flywheel bores, keyways, and drilled holes. 4.7.3 State the design stress specified for the flywheel as a percentage of the minimum specified y? 'd strength, for the normal operating speed and the des _gn overspeed condi-tion. 4.7.4 State if the calculated combined primary stresses in the flywheel, at the normal operating speed, will include the stresses due to the interference fit of the wheel on the shaft, and the stresses due to centrifugal forces. 4.7.5 State the highest anticipated overspeed of the flywheel and the basis for this assumption. l 1 l 1

m i . 4.7.6 State the estimated maximum rotational speed that the flywheel attains in the event the reactor coolant piping breaks in either the suction of discharge side of the punp. In addition, describe results of any studies directed towards: (1) determining the maximum speed the pump or motor can reach due to physical limitations (e.g., the speed at which the pump impeller seizes in the wear rings due to growth from centrifugal forces or the speed at which notor parts come loose and grind or bind to prevent further increase in speed); (2) establishing speed and torque for various pipe break sizes; (3) devising means to disengage the motor from the pump in the event of pump overspeed; (4) verifying that pump casing fragments generated at maximum speed do not penetrate the pump casing and that any missiles leaving in the blowdown jet do not penetrate containment; (5) estab-lishing failure speeds for motor parts and whether they will penetrate the motor frame and if so with what energy; and (6) defining a mininum rotor seizure time. 4.7.7 State the rotational speed that will be specified for the preoperational overspeed tests of the flywheel. 4.7.8 Describe the inservice inspection program proposed for the flywheels, state the areas to be inspected, access provi-slons, type and frequency of inspections, and the acceptance criteria. 4.7.9 Provide the name of the manufacturer of the motor and sketches of the arrangement of the flywheel on the motor.

4. 8 Describe the umthods that will be used to determine coolant leakage f rom the reactor coolant sys tem.

Provide sufficient detail to Indicate that a minimum of two systems with diverse modes of opera-tion will he installed in the plant. 4.9 Describe the methods used to provide positive indications in the control room of leakage of coolant from the reactor coolant system to the containment. 4.10 Discuss the adequacy of the leak detection subsystem which depends on reactor coolant activity for detection of changer in leakage during the initial period of plant operation when the coolant activity may be low.

m y . i 4.11 Estimate the anticipated normal total leakage rates and major leakage sources on the basis of operational experience from other plants of similar design. 4.12 Describe the tdequacy of the leakage detection systems to differentiate between identified and unidentified leaks from com-ponents within the primary reactor containment and indicate which of these systems provide a means for locating the general area of a lenk. 4.13 Describe the proposed tests to demonstrate scnsitivities and opera-bility of the leakage detection systems. 4.14 The air sampling system, RM-Al, asgociated with the containment purge valves appears to possess 10 times more continuous operating sensitivity in detecting containment atmosphere radioactivity than air sampling system, RM-A6, proposed to detect primary coolant leakage into the containment atmosphere. Consider adapting System RM-Al to detect primary coolant leakage or adding a continuous reading air particulate monitor to System RM-A6. 4.15 Describe the design and arrangement provisions for access to the reactor coolant pressure boundary ac required by Sections IS-141 and IS-142 of Section XI of the ASME Boiler and Pressure Vessel Code - Inservice Inspection of Nuclear Reactor Coolant System. Indicate the specific provisions made for access to the reactor vessel for examination of the welds and other components. 4.16 Section XI of the ASME Boiler and Pressure Vessel Code recognizes the problems of examining radioactive areas where access by person-nel will be impractical, and provisions are incorporated in the rules for the examination of such areas by remote means. In s'ome cases the equipment to be med to perform such examination is under development. Provide the following information with respect to your inspection program: 4.16.1 Desc' ribe the equipment that will be used, or is under I development for use, in performing the reactor vessel and nozzle inservice inspections. 4.16.2 Describe the system to be used to record and compare the data f rom the baseline inspection with the data that will be obtained from subsequent inservice inspections.

.m 4.16.3 Describe the procedures to be followed to coordinate the development of the remote inservice inspection equipment with the access provisions for inservice inspection afforded by the plant design. 4.17 The list of transients that have been used in the design of com-ponents within the reactor coolant pressure boundary as specified in Table 4-8 of the FSAR appears to be incomplete. 4.17.1 Identify all design transients and their number of cycles, such as' control system or other sys tem malfunction, component mal-functions, transients resulting from any single operator error, inservice hydrostatic tests, etc., which are speci-fled in the ASME Code-required " Design Specifications" for the components of the reactor coolant pressure boundary. 4.17< 2 Categorize all transients or combination of t ransients with res pec t to design Cases I through IV of paragraphs 4.1.2.5.1 and 4.1.2.5.2 (comparabic to the operating condition cate-go ri es identified as " normal," " upset." " emergency," and " faulted" as defined in the Summer 1968 Addenda of the ASME Section III Nuclear Vessel Code). 4.17.3 Describe the program which will be used to record and main-tain an accounting of significant transients occurring during plant operation and to compare the service number of transients accumulated with those specified as the permissible number of transients for which the plant is designed. 4.18 Specify the Code Case Interpretations '(Special Rulings) that have been implemented in conjunction with the component codes delineated in Table 4-2 of the FSAR. ~~ 4.19 Paragraph 1701.5.4 of the ANSI B31.7 Nuclear Power Piping Code requires that piping shall be supported to prevent excessive vibra-tion under startup and initial operating conditions. 4.19.1 Submit a discussion of your vibration operational test pro-gram which will be used to verify that the piping and piping restraints within the reactor coolant pressure boundary have been designed to withstand dynamic effects due to valve closures, pump trips, etc. l

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• 4.19.2 Provide a list of the transient conditions and the asso-ciated actions (pump trips, valve actuations, etc.) that will be used in the vibration operational test program to verify the integrity of the system.

Include those t ran-sients introduced in systems other than tSose within the reactor coolant pressure boundary that will result in significant vibration response of reactor coolant pressure boundary systems and components. 4.20 Specify whether the design criteria which have been used to examine the effects of pipe rupture have considered postulated pipe breaks to occur at any location within the reactor coolant pressure boundary, or at limited areas within the system. Provide confirmation that both longitudinal and circumferential type ruptures were evaluated and describe the basis for your design approach. 4.21 Provide a more detailed description of the measures that have been used to assure that the containment liner and all essential equip-ment within the containment, including components of the primary and secondary coolant systems, engineered safety features, and equipment supports, have been adequately protected against blow-down jet forces, and pipe whip. The description should include: 4.21.1 Pipe rsstraint design requirements to prevent pipe whip impact. 4.21.2 The features provided to shield vitsi equipment from pipe whip. 4.21.3 The measures taken to physically separate piping and other components of redundant engineered safety features. 4.21.4 A description of the analyses performed to determine that the failure of lines, with diameters of 3/4 inch or less, will not cause failure of the containment liner under the most adverse design basis accident conditions. 4.21.5 The analytical methods which were used. 4.22 Provide the stress or deformation limits and the design codes or standards applicable to the principal reactor coolant system com-ponent supports (i.e., supports, restraints, " snubbers," guides, etc., as applied to vessels, piping, pumps, and valves). l-i i

4 .m i 4.23 Reported service experiences of PWR steam generators have demon-strated that flow induced vibration and cavitation effects can i cause tube thinning, and corrosion and erosion mechanisms both l from primary and secondary side may contribute to further structural degradation of the tube integrity during the service life-time. The failure of a group of weakened tubes as a consequence of a ( design ~ basis pipe break in the reactor coolant pressure boundary could impair the capability of emergency core cooling systems to i perform their intended function. In erder to evaluate the ade-l quacy of design bases used to prevent such conditions from devel-oping in the steam generator during service, the following additional information is required: 4.23.1 State the design conditions and transients which were specified in the design of the steam generator tubes, and I the applicable design stress intensity limits associated with Cases I through IV of paragraphs 4.1.2.5.1 and 4.1.2.5.2 (comparable to the " normal," " upset," " emergency," and " faulted" operating condition categories). Justify the basis for this selection. 4.23.2 Specify the margin of tube-wall thinning which could be i tolerated without exceeding the allowable stress limits identified in 4.23.1 above, under the postulated condition of a design basis largest pipe break in the reactor coolant pressure boundary during reactor operation. 4.23.3. Describe the inservice inspection which will be employed to examine the integrity of steam generator tubes as a means to detect tube wall thinning beyond acceptable limits -and 1 whether excess material will intentionally be provided in the tube wall thickness to accommodate the estimated degra-dation of tubes during the service lifetime. 4.24 Submit a copy of the summary technical report on overpressure pro-tection which has been prepared in accordance with the requirements of paragraph N910.2 of the ASME Section III Nuclear Vessel Code. i 4.25 The applicable code summary of Table 4-2 for components of the j reactor coolant pressure boundary only addresses safety and relief valves. Provide the design' criteria and identify the design codes applied to all other valves within the reactor coolant pressure boundary (e.g., ASME Code for Pumps and Valves for Nuclear Power. ANSI B16.5, MSS SP-66, etc.). I 1 1 e i

m . -4.26 Describe the design and installation criteria for the mounting of the pressure-relieving devices (safety valves and relief valves) within the reactor coolant pressure boundary and on the main steam lines outside of containment. In particular, specify the design criteria which have been used to take into account full discharge, loads (l.c., th rus t, hend ing, torsion) imposed on valves and on connected piping in the event the valves discharge concurrently. Indicate the provisions made to accommodate these loads. 4.27 . I'aragraph 4.1.2.5.2 of the FSAR states that design cases III and IV, comparable-to the " emergency" and " faulted" conditions de fined in Section lit, have been applied to reactor coolant pressure boundary (RCPB) components. Identify any other components or systems that are not a part of the RCPB for which design Case IV stress limits will be applied. In the event that Case IV stress limits have been.spplied to any system or component exclusive of the RCPB, provide the bases for such application. 4 4 t

"3.0 CONTAINMENT SYSTEMS AND OTilER SpECIAL STRUCTURES 3.1 Discuss the manner in which the specified general shell analyses and finite element analyses recognize and address t'se cracked state of the concrete, and the influence of the prestressing tendon holes on the surrounding concrete for the containment structure. 5.2 To permit our evaluation of the application of the static shell solut ion uqing Kalnins 's digital computer program to this shell problem, provide the information required for program input by the analyst for a typical loading case. This will require the physical properties of the materials, the geometry in detail, and other information such as the breakdown.of the shell into the various elements and segments, the applied loads, etc. 5.3 For all Class I items which will be located within or adjacent to Class II or III structures, describe in detail what precautions are taken to ensure that the failure of Class II or III structures will not adversely af fect Class I items. 5.4 Describe in detail the procedures used to design the roof stru tural steel for the auxiliary building so that it cannot damage the Class I portions of the building under the loads imposed on Class I st ruc tures. 5.5 liresent additional details regarding the deficient concrete that was noted on page 5-15 of the FSAR. Include the reanen for con-ciuding there was a deficiency, how the dericiency wa 4 disco ve red, the program used for correcting the deficiency, and the verifica-tion procedure used to finally assure that the original requirements had not been compromised. 5.6 Present the data obtained in verifying that the steel liner plate ~ was erected within the tolerance limits defined on page 5-52 of th e FS AR. 5.7 Specify the nondestructive testing procedures employed to detect laminations in load carrying steel plates welded in~co the contain-ment liner which are required to transfer loads normal to their surface. 5.8 For the Cadweld splice test program that was utilized as described 4 on page 5-17, present the following information: N

',s 18 - 5.8.1 Justify the inclusion of 21 specially selected pilot splice test sampics with 30 production splice test samples as a group to determine "hether the sampling frequency can be reduced. 5.8.2 Describe what has been done as a result of the outlined test program and indicate whether sampling frequencies were reduced. 5.8.3 Compare the test program to AEC Safety Guide No. 10, entitled " Mechanical (Cadweld) Splices in Reinforcing Bars of Concrete Containments." 5.8.4 Present the results of the Cadweld testing completed to date. 5.9 Instrumentation such as embedded strain gages which would continue the meridional lines of instrumentation (at 90*, 200*, and 333' 55') from the ring girder to the apex appears not to have been placed in the dome of the structure. Justify this approach and indicate by what means the level of strain in the dome can be monitored during the structural acceptance test. 5.10 Indicate the degree of compliance of the containment leakage testing program with the AEC proposed " Reactor Containment Leakage Testing for Water Cooled Power Reactors," 550.54(o), Appendix J, published in the Federal Register on August 27, 1971. 5.11 Describe the design features of the containment airlocks that will permit testing of the airlocks. at the calculated peak containment pressure. Describe the test method that will be used to verify 'eak tightness of air lock doors, door penetrations, and door ~ gaskets. 5.12 Submit the seismic design spectra specified for the Maximum Hypo-thetical Earthquake. 5.13 Identify the method of seismic analysis (modal analysis response spectra, modal analysis time history, equivalent static load, etc.) or empirical (tests) analysis which has been employed in the design of the listed Class I structures, systems and components. 5.14 Provide a more detailed descrl.it leni of nli u s h...in i h.it weee c...i. l.. y e.1 for seismic analysis including damping valuen,.len. r i pt l anin (nici. hens of the mathematical models employed and the procedure f o r l umpisig masses.

.3 - 19. T 5.15 Submit the basis for the methods aed to determine the possible combined horizontal and vertical amplified response loadings for the seismic design of structures, systems, and components including the following: 5.15.1 The possible combined horizontal and vertical amplified response loading for the seismic design of the building and floors. 5.15.2 The possible combined horizontal and vertical amplified response loading for the seismic design of equipment and components, including the ef fect of the seismic response of the building and floors. 5.15.3 The possible combined horizontal aro vertical amplified response loading for the seismic design of piping and instrumentation, includiny -5t *ffect of the seismic response of the building, s, supports, equipment, component, etc. 5.16 if a fixed base has been assumed in the mathematical models for the dynamic system analyses, confirm the validity of this assump-tion by providing summary analytical results that indicate that the rocking and translational response are insignificant. Include a brief description of the mathematical model and damping values (rocking, vertical, translation, and torsion) that have been used to consider the soil-structure interaction. 5.17 . Describe the method which was employed to consider the torsional modes of vibration in the seismic analysis of the Class 1 building structures. ~' 5.18 Provide the structure material properties and soil structure interaction which were used in seismic design analysis and the bases for selection of these properties. Describe the measures which will be taken to assure that the calculated response of Class I structures will conservatively reflect the expected variations in the periods of vibration of the structures. 5.19 The use of both the modal analysis response spectrum and time history provide a check on the response at selected points in the ~ station structure. List the response obtained from both eethods at selected points in the Class I structure to provide the basis for checking the seismic system analysis. 5.20 Describe the evaluation performed to determine seismic induced effects of Class II or III piping systems on Class I piping.

m 3 5.21 Indicate the manner in which the crane located in containment is held on its rails to preclude dislodgment during seismic excitation. 5.22 With respect to Class I (Seismic Design) piping buried or other-wise located outside of the containment structure, describe the seismic design criteria employed to assure that allowable piping and structural stresses are not exceeded due to differential move-ment at support points, at co'ntainment pene trations, and at ent ry points into other structures. 5.23 Describe the seismic design criteria employed to assure th( 'r-quacy of Class I mechanical components such as pumps, heat exchangers, and elect rical equipment such as cable t rays, bat tery racks, in tru-ment racks and control consoles. Describe the measures taken for scismic restraint to meet these criteria, the analytical or testing methods employed to verify the adequacy of these rest 'aints and the methods utilized to determine the seismic input to these components. 5.24 Describe the criteria employed to determine the field location of seismic supports and restraints for Class I (seismic design) piping, piping sys tem components, and equipment, including place-ment of snubbers and dampers. Describe the procedures followed to assure that the field location and characteristics of these supports and restraining devices are consistent with the assump-t ions made in the dynamic analyses of the sys tem. 5.25 Describe the procedures used to account for the number of earth-quake cycles during one seismic event and specify the number of loading cycles for which Class I systems and components are designed in this event, as determined from the expected duration of the seismic motions or the number of major motion peaks. 5.26 The following information requests are specifically related to gal Report No. 1729 " Dynamic Analysis of Vital Piping Sys tems Subject to Seismic Motions," which is referenced in the application: 5.26.1 The comparison for a single degree-of-freedom model method identified in " Seismic Analysis of Equipment Mounted on a Massive Structure," J. M. Biggs and J. M. Roesset and referenced in CAI Report No. 1729 may not be suf ficiently conservative for multi-mass systems. Provide the basis for and the conservatism in the use of this method, by demonstrating equivalency to a multi-mass time history method. Alternatively, other theoretical methods or experimental analyses and testn may he-nub-mitted in justification of the use of the proponent design method.

o q ,. 5.26.2 Provide the design criteria which apply to the dynamic analysis of valves and other in-line system components which by virtue of their geometry and size introduce significant torsional moments into the piping system. 5.26.3 Provide the design criteria and analyt ical procedures which apply to piping and which take into account the differential movement between floors at different elevations. 5.26.4 Predominant mass and compliance of structures, systems and components will affect the response of the system. Provide the criteria used in formulating the system analysis which takes into account the coupling of these predominant mass compliance ef fects in the system dynamic analysis. 5.26.5 In-phase or out-of-phase displacements at dif ferent ele-vations between equipment, piping and tne building may 4 produce high stresses or deformations in connecting piping. Provide the criteria and analytical procedure used in the system analysis to account for in-phase or out-of-phase displacement of the building, equipment, and piping. 5.26.6 Provide the design criteria used to compute shears, moments, stresses, deflections, and/or accelerations for each mode as well as for the combined total response, including the criteria for combining closely spaced modal frequencies. 5.26.7 The presentations of the methods of dynamic analysis are general and the precise techniques used for the system and piping analysis are not suf ficiently described to permit evaluation of their applicability and design conse rvatism. Provide sufficient detail concerning the methods and procedures used including the mathematical models and analytical results for major piping systems. 5.27 With respect to seismic instrumentation, submit a statement of your intent to implement a program such as described in AEC Safety Guide 12, " Instrumentation for Earthquakes," (April 9, 1971). Provide the basis and justification for elements of the proposed program which differ substantially from Safety Guide 12.

n .) 7.0 INSTRUMENTATION AND-CONTROL SYSTEMS 7.1 Provide in fo rma t ion requested in the Commission's In fo rmat ion Guide 2, " Ins trumentat ion and Elect rical Sys tems" (10/27/71). If the requested Infomation is presently contained in the FSAR or its amendments, the response should identify the speci fic location of the Information and augment same, as requi red, to meet the requi re-ments of the Information Guide. The response to the following items of the Information Guide should: 7.1.1 Item I.C. - Describe the degree of conformance of the protection system to the provisions of IEEE Std 338-1971, "IEEE Trial Use Criteria for the Periodic Testing of Nuclear Power Generating Station Protection Systems." Describe and justify all exceptions to this standard. 7.1.2 Item 3 - Describe the degree of conformance of your seismic testing program to IEEE Std 344-19 71, "IEEE Guide for Seismic Qualification of Class I Electric Equipment for Nuclear Power Generating Station." Descrlhe and justify all exceptions to this standard. The response should include the results of tests of the batteries (cells), in addition to those of auxiliary equipment such as racks, and any necessary extrapolation that accounts for all degradation due to time. 7.1. 3 Item 4 - Describe the degree of conformance to IEEE Std 336-1971, "IEEE Standard Installation, Inspection, and Requirements for Instrumentation and Electric Equip-ment During the Construction of Nuclear Power Generating Stations" during the construction phase of the plant. Describe and justify all exceptions to this standard. 7.1.4 Item 5.a. - Discuss your criteria regarding maximum percentage fill in trays and wireways, and minimum spacing between same. Describe and justify all exceptions to the criteria. 7.1.5 Item 5.c. - Describe the methods used to preserve the l independence of safety related loads served by E.S. Bus 3B. The response should account for all potential conflicts created by usipg E.S. Bus 3B to serve safety and non-safety related loads. 1 ~.

s' ) 7.1.6 Item 11 - Discuss the indications available to the control room operator which allow him to recognize that a protection system or subsystem has been placed on test, bypassed for operation or maintenance purposes, or removed from service for any cause. The response should list all such bypasses and verify that they have been designed to meet the requirements of Paragraph 4.12 of IEEE 279. 7.1.7 Item 12 - Refer to IEEE Std 308-1971 in lieu of IEEE 308 as stated in the Guide. (Note: Where a conflict exists between the "eight hour" provision of Section 5.2.3.4 of IEEE Std 308-1971 and General Design Criterion 17, the applicable provisions of Criterion 17 govern.) Also discuss the use of automatic transfer switches (Figure 8-9) and the ef fects of the stated voltage dips (30 percent) of the diesel-generators on loads that require high starting torque such as motor operated valves. 7.2 Identify and provide justification for any aspects of the design that do not conform to Safety Guide 11, " Instrument Lines Pene-trating Primary Reactor Containment" (3/10/71). 7.3 Describe the methods for periodically testing the reactor protec-tion system's response time for the trip parameters. The response should he parametric in nature to show the change in response time function of the level of the parameter and as a function of the as a rate of change of the parameter. Include a discussion of the re-sponse times in relation to the safety limits and state the worst case margin in terms of time. 7.4 State the criteria and design basis which established the heat ~ tracing requirements, temperature control, monitoring, and power-requirements for the boric acid tanks, borated water storage tanks, and spray additive tanks and related piping of the chemical addi-tion system. Discuss the consequences of a single failure in the heat tracing of each of the above mentioned systems. 7.5 The FSAR states that pump interlocks insure against accidental startup of a cold loop if reactor power is greater than 15 percent. Discuss the consequences of an interlock f ailure and the criteria to which these interlocks are designed. +-

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} . 7.6 Identi fy all trip set points of the reactor protection system and engineered safety feature system instruments which are within 10% of the high or low end of the calibrated range. Provide a worst case error analysis that verifies that each output signal is always conservative when viewed from a safety standpoint. 7.7 The FSAR, in describing the core flooding system, states that the electric motor-operated stop valves between the core flooding tanks and the primary coolant system are open during reactor power opera-tion, and that valve position is indicated in the control room. Discuss the design features of the control circuits of these valves, including the assurance provided by the design that the valves will be open when required. The inclusion of the fol-lowing features would provide an acceptable design: Automatic opening of the valves whenever the primary coolant a. system pressure exceeds a preselected value. h. Automatic removal (override) by a safety injection signal of any hypass feature that may be provided to allow a motor operated valve to be closed, for short periods of time, when the primary system is at pressure. c. Valve position visual indication that is actuated by sensors on the. valves ("open" and " closed"), d. Audible alarms, independent of item c., that are actuated by sensors on the valves when the valves are not in the fully open position. I f the design, or planned modifications, are based on criteria d i f fe ren t from those stipulated above, submit the criteria and necessary documentation that verify that an equivalent degree of protection is afforded. 7.8 The FSAR states that flexibility of controlling reactivity rate is provided by a patch panel that permits the patching of any rod into any group except those of Group 8. Describe the interlocks and/or administrative procedures used to insure that _he resultant changes of rods between various groups are correct, and the con-sequences of a failure in the interlock system.

o ~l ,Q ~ ?, 7.9 Several sections of the FSAR (e.g., 1.4.19, 1.4.38, 7.1.1.1) restrict the scope of General Design Criterion (CDC) 21 and IEEE 279 by using terms such as " single component failures" and " single failures of an active component" rather than the mo re extensive term "s ingle failures." All relevant sections of the FSAR should be revised to conform with CDC 21 and IEEE 279. 7.10 The FSAR (Section 1.6) states that the protection and control systems are being analyzed for possible common failure modes. Identi fy and discuss the results of the analyses. If the analyses have not been completed, state the scheduled completion date and indicate your conmitment to modify the design, if required, as a consequence of the analyses. 7.11 The discussion of the operational sequence of the engineered safeguards system is incomplete and not sufficiently detailed to permit evaluation. Submit a more comprehensive description and logic diagram, with nomenclature that is consistent with the other subordinate drawings, that delineate all the required ' inputs and sequential actions. The discussion should be explicit with regard to the timing network for the various load blocks, including the effect of a sustained low voltage condition on the diesel-generator (i.e., low voltage for a period longer than that of the 15 second timing network). Discuss the consequences of such a condition and the means used to prevent the simultaneous energization of load blocks 2, 3, and 4. 7.12 In your design, either low reactor coolant pressure or high reactor building pressure will initiate safety injection; however, of these two signals, only the low reactor coolant pressure ini-tintes reactor trip. Since your analysis of the ef fectiveness of safety injection takes credit for reactor trip, discuss the bases i for not using high building pressure to trip the reactor, l 7.13 Identify and discuss the selected alarm conditions mentioned in Section 7.1.3.2.5 that are annunciated. 7.14 Describe more completely the rod position-display system. The discussion should as a minimum address: (1) the capability to display simultaneously the position of each rod including any constraints therein, and (2) the display and annunciation of an out-of-symmetry condition.

y . 7.15 The FSAR states that control rod-drive motor rotation in the wrong direction is detected by a motor rotation fault detector. Discuss the capability to detect failure to rotate upon a demand signal, and the provisions which prevent such an occurrence from being erroneously dispinyed and from being cumulative. 7.16 Section 7.4.6 states that several actions occur automatically upon loss of of.fsite power, including the tripping of the main feedwater pumps and the reactor coolant pumps. In light of the above, discuss the consequences of loss of offsite power and how full load rejec-tion is accomplished. 7.17 Discuss more completely the protective system with regard to partial loop operation. As a minimum, the response should address the following: (1) the envelope of the permissible operating limits for the three pump and the two pump (one per loop) case as a func~ tion of flow and power imbalance; (2) the method used to automati-cally reset reactor trip to a power level commensurate with the number of pumps,in operation, actual flow, and power imbalance; (3) the method used to discriminate between only two pumps oper-ating in one loop and one pump in each loop; (4) whether the operating limits (trip) delineated in Table 7-4 should be re-st ricted when operating in a partial loop mode; and (5) the basis for apparently not using breaker auxiliary contacts to provide the logic for reactor trip and operating limits. { 7.18 The schematic diagram for motor operated valves (Figure 7-10) shows that overload protection (0/L) is used. Discuss the bases for the 0/L's and any provisions for bypassing the 0/L's, especially during emergency conditions. Discuss the effect of low bus voltage (e.g., during diesel operation) on motor torque, and the possibility for causing O/L trip prior to valve operation.

.m, %. 8.0 ELECTRICAL SYSTLM 8.1 Describe more fully the auxiliary equipment of the emergency diesel generator system. The description should include the fuel storage and trans fer system; the nource of power for con-trol; the starting system and number of start at tempts provided; method of cooling and warming the engine; and the control and protectica system including relevant schematic diagrams. 8.2 The tabulation of connected loads to the diesels, as shown in Table 8-1 of the FS AR, is not suf ficient to evaluate the adequacy of the design with respect to regulatory position 4 of Safety Guide 9. Data that are required include actual starting KVA loads rather than the nameplate data and the time required for the various loads to reach full speed. Te correlate these data with the capability of the diesel generator, the following information should be provided: 8.2.1 A load profile during a LOCA showing the timing sequence and the time duration of the various loads subsequent to diesel start. The anticipated starting KVA requirements for the various loads should be superimposed on the load profile at the appropriate time, and the competed effect of each load transient (system voltage and speed, and voltage and speed recovery time for each step) should be indicated. 8.2.2 The maximum loads that can be incrementally added to the various block plateaus, without exceeding the recommenda-tions of Safety Guide 9. 8.2.3 The continuous, 2000 h r. and 30 min., rating of diesel engine. The generator's X, X, X2, and SCR and the unit's WR. 8.2.4 d d The type of excitation system provided should be discussed, including the response time of same for voltage regulation during the various step load changes. 8.3 Discuss the design criteria for the diesel generator rooms with respect to ability to preclude missiles, explosions and fires in one diesel generator unit frem affecting the redundant counterpart. 8.4 Identify the sources of control power to the 230 and 500 KV switchyard breakers. Submit an analysis to show that no single failure in these power sources, control circuits, and protective relaying will negate the ability to provide of fsite power to the engineared safety features.

-N 3 . 8.5 Describe the monitoring features provided to continuously ensure that the capability of a battery to supply power is not degraded. Consider the relevance of the monitored parameters to the actual charge stored in tha battery, and discuss the limitations of the system to ensure disclosure of battery degradation, including protection against overcharging. 8.6 Discuss your plans for converting the 500 KV bus from a ring configuration to a breaker-and-a-half system. 8.7 The design of the 6900 volt auxiliary system initiates automatic transfer from the preferred source to an alternate source upon loss of the preferred power source. Discuss the effect on avail-ability of offsite power resulting from switching four 9000 llP pump motors from one source (dead) to another (live). Address the ef fect of the transfer occurring before the motor field has collapsed but at such time that the back EMF does not oppose that of *.he source. The concern here is that such transfer could approach that of paralleling generators that are out of phase.

~ ~ y 9.0 AUXILI ARY AND EMERCENCY SYSTEMS 9.1 Describe the fire protection capability provided for Unit 3, inclu-ding both the alarm and suppression aspects of the protection system. Indicate areas of automatic coverage and sources ef emergency power. Consider a fire in one or more of the on-si.te oil storage tanks. 9.2 Describe the instrument and service air systems. Provide a flow and instrument diagram of the systems. Specify the sources of emergency power for the safety aspects of the systems. 9.3 Describe the floor drainage system, including level instrumentation and alarms, limits of Icak detection, and surveillance requirements for radioactivity. 9.4 in numerous instances in Chapter 9, such as 9.6.1.2, the term " fuel handling building" is used. In the figures of Chapter 1 showing the general arrangement of plant equipment and structures, this " fuel handling building" is not identified. Provide a description of this " fuel handling building" including plot and elevation views. If there is no " fuel handling building" correct the text of the i FSAR accordingly. 9.5 Safety Guide No. 13, published March 10, 1971, addresses the special considerations involved with the fuel storage facility in implement-Ing General Design Criterion 61. For the items listed below, discuss the extent to which your design conforms to the recommendations of the guide and indicate what changes, if any, you intend to make to obtain a greater measure of conformance. 9.5.1 The seismic design of the fuel handling ventilation and makeup coolant systems. 9.5.2 The watertight integrity of tie spent fuel pool as af fected by high -winds and missiles generated by these winds. 9.5.3 Interlocks preventing cranes from passing over stored fuel. 9.5.4 Movement of heavy capacity cranes in the vicinity of the spent fuel pool. 9.5.5 ' Ability of the pool to withstand the impact of-the . heaviest load to be carried over it. l l i

.n. . 9.5.6 The location and number of the fuel handling area monitors. Also, the frequency of testing of this equipment. 9.5.7 A makeup system with appropriate backup to add ' coolant to the spent fuci pool. 9.6 in 9.7.2.1.f you state that the supply and exhaust fans of the auxiliary and fuel handling buildings are inoperative during an emergency. Describe the basis for which the fuel handling venti-lation system is turned off following a fuel handling accident. Specify the procedures that would be utilized to contain the released radioactivity in the fuel handling area and control its exhaust. 9.7 On Page 9-12 you state that if all cooling is lost to the spent fuel pool for 1-1/3 cores of spent fuel, approximately 7 hours a*:e available prior to the pool reaching a temperature of 205'F. Specify the arrangements that would be made in this seven hour period to provide emergency coolant. 9.8 Discuss the design basis of the fuel cask, such as weights and dimensions, to assure adequate lif ting capacity and clearances in handling of the cask. Specify the lift distance between the top of the spent fuel pit and the cask loading area for the spent fuel shipping cask. If this distance is greater than 30 ft., provide an analysis of the radiological dose at the site boundary resulting from dropping the cask this distance. For this analysis use the maximum cask size which you anticipate employing in the future. 9.9 What provisions have been made to protect operating personnel from radiation emitted from the spent fuel pit and the apent fuel pit demineralizers and filters? 9.10 A review of the decay heat removal system description in Section 9.4 of the FSAR indicates that overpressure protection of the 450 psig system from the 2500 psf:t reactor coolant system has been accomplished by valves interlocked to prevent opening with reactor coolant pressure above 450 psig and safety relief valves. ~ We have concluded that your present design and operating procedures do not provide for an acceptably low probability for inadvertent high pressurization of the low design pressure system. In our opinion the two valves should be provided with a highly reliable control system to assure that they will close automatically when-ever the pressure in the primary coolant system exceeds the low pressure system maximum operating pressure. Although the FSAR

m, -) -3l-states the valves will be interlocked to prevent opening at high pressure, there is no statement that the interlocks are designed to safety system standards and on the basis of different principles. Both the valve automatic closure system and the interlock system should conform to 1000-279 requirements. Provide either a description of the modifications you plan to make to the system to reduce the probability of-an inadvertent pressuri-1 zation of the low pressure system, or a comprehensive assessment of the facility design and operating procedures to convincingly demonstrate that the present design and proposed procedures provide an acceptably low probability for the overpressurization of the low pressure system. ').1 1 Discuss the redundancy features of that portion of the ultimate heat sink which connects the water source (Gulf of Mexico) with the Unit 3 Intake structures. Include diagrams and plot views as appropriate. Discuss the capability of these features to withstand the effects of severe natural phenomena such as earth-i quakes and hurricanes, i = 1 1 I l l I

+ rm y . 11.0 1[AD_IOACTIVE WASTE A'ID RADI ATION PROTECTION 11.1 Provide a discussion of how the radwaste system will be designed and operated so as to comply with the amendments of 10 CFR Parts 20 and 50 of the AEC's regulations that require releases of radio-activity to be reduced to their lowest practicable levels. Provide a tabulation of the total anticipated maximum yearly discharge of each isotope (including tritium) expected to be released f rom all sources to the environment for ~both' liquid and gaseous e f fluents and compare them with the limits of 10 CFR Part 20. Provide a separate tabulation for the releases that will result from con-tainment purging. Assume water leakage f rom the reactor coolant ' system and fission product Icakage from the fuel rods consis tent with your proposed technical specifications and with data from operating plants, such as Cinna. Describe the basis for these estimates, including the assumptions made with regard to decon-tamination factors used for holdup, filtration, evaporation, and demineralization. 11.2 Provide any test or operational data from similar radwaste treatment components in operating plants. 11.3 Include a description, using numerical values, to show the anticipated use of the containment vent and purge system to relieve pressure buildup in the containment, and to permit access to the containment by plant personnel for routine inspection and maintenance, 11.4 Provide additional information in the form of process data on process load diagrams for both liquid and gaseous wastes. On the ILquid waste process load diagram, show the microcurie content and flow rate in each flow path. On the gaseous waste process load diagram, show the flow rates, temperatures, pressures, and specific isotope radioactivity for the 90 day minimum holdup time in the decay tanks, at appropriate steps of the process. 11.5 Provide a description of your procedures to determine and record the activity of specific isotopes that will be in the liquid waste discharged to the Gulf of Mexico. Include in the description the frequency of periodic determinations of isotopic composition and changes in condition that would require these determinations, such as changes in the radwaste process or unexpected changes in gross activity measurements. 11.6 Provide the following information regarding tritium:

m .h 11.6.1 A comparison of the anticipated maximum tritium concentra-tion in the discharge canal and in the Gulf of hiexico, with current tritium concentrations in the Gulf of Mexico as determined by your preoperational monitoring program. 11.6.2 Your analysis of the fraction of tritium in the primary coolant that will be released to the circulating water discharge canal. 11.6.3 A discussion of the uncertainties in your estimate of the amount of tritium that will be generated. 11.6.4 A description of the methods to be used to monitor tritium concentrations prior to discharge from the plant.

11. 7 Provide the following information with respect to the instrumenta-tion, sampling techniques, and laboratory analvtical procedures to be used for evaluation of gaseous and liquid ef fluents and other in plant and environmental radiation and radioactivity levels:

11.7.1 A description of the instrument types, sensitivities, ranges, set points, and calibration methods and frequencies, 11.7.2 A discussion to demonstrate the capabilitics of the instru-mentation and procedures to detect, measure, and control effluent releases, by appropriate radionuclide, within design objectives and expected for routine operation and for expected operational occurrences. 11.7.3 An identification of each path by which liquid and gaseous ef fluants can be released f rom the plant, and of the instrumentation to be used for monitoring each path. 11.7.4 An evaluation of the possible annual releases of radioactive materials if ef fluents having radioactivity levels equivalent to the monitor setpoint are continuously released. 11.8 'Page 11-16 briefly describes the reactor building purge exhaust duct monitor RN-A-1. Provide a description of the capability of this instrument to detect specific isotopes, including the capability to detect gaseous iodine 131' in the presence of noble gases. Provide a description of the method of instrument calibration for measuring the quantity of radioactive noble gases,- iodines, and particulate matter. If a filter tape is used in this system, state the filter ef ficiency of the paper for radioactive iodine. l

rm . - 11.9 Page 11-18 describes the monitor in the plant discharge line. Provide an analysis of the maximum concentrations of radioactivity that could exist in the discharge canal due to activity, the steam generator blowdown system based on a radioactivity level equal to the setpoint of the liquid radiation monitor, and in the condenser vacutcn pump exhaust due to the continuous release of steam and gases from the flash tank having a radioac:.ivity level equal to the setpoint of exhaust monitor RM-A-12. Assume for the analysis that the maximum coolant activity and maximum coolant leakage to the steam generator are as specified in the Technical Specifications. ' 11.10 Indicate the radiation source te rm used in the design of the control complex emergency charcoal filter shown in figure 9-12. Provide data on the charcoal filter system as to type and weight of charcoal used, flow ra te, face velocity of the filters, and the expected efficiency for both elemental and organic forms of iodine. 5 i l i f

9 m . 14.0 SAFETY ANAI.YSIS 14.1 It is noted on page 14-21 that the fuel handling accident analysis is based on the failure of 56 out of the 208 fuel rods in a fuel assembly. It is further noted that a water decontamination factor of 1000 is taken for Lodine. Our current staff model assumes a water decontamination factor of 100 for elevental iodine and failure of all rods in a fuel assembly. Your refueling accident dose should be reanalyzed based on these revised assumptions. 14.2 Table 14-52 includes the results of leakage doses resulting from the MilA. Provide the details of your analysis, including the fission product source term contained in the leakage, the parti-tion factor assumed as the liquid flashes to steam (figure 14-61 shows the temperature of the sump water is above 212*F for the first two hours following the LOCA), and the flow path to the point of release to the environment. 14.1 Indicate the calculated thyroid dose to the reactor operators in the control room following the design basis accidents. 14.4 Extend the data of Figures 14-60, -61, -62, -63, -64, and -65 of 6 your FSAR through 10 seconds so as to it.clude the long term containment pressure decay. 14.5 For both the continuous and intermittent purge concepts which you discuss in Appendix 14B of the FSAR we need the following in fo rmation: 14.5.1 Using the assumptions contained in Table 1 of Safety Guide 7, calculate the time at which initiation of venting would be required and the rate at which venting must continue to keep both the hydrogen and oxygen con-centrations below the limits listed in Safety Guide 7. If you assume 1000 lbs of aluminum react to form hydro-gen as in Appendix 14B, provide a detailed list of all-aluminum sources including containment coatings and their weights in the reactor building. From this summary justify your assumption of 1000 lbs of alum-inum reacting to form hydrogen. 1 14.5.2 Using (a) the fission product release fraction assump-tions given in Safety Guide 3, (b) the same fission product removal rates and/or radioactive decay rates used in the evaluation of your design basis LOCA, 1

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• (c) the accident meteorology applicable for your site, and (d) an equivalent containment leak rate composed of the allowed containment leak rate plus the venting rate, calculate the infinite-time incremental thyroid and whole body doses due to venting alone at both the site boundary and low population :.one distances.

14.5.3 Repeat the analysis described in 14.5.2 above omitting the containment venting. W 1 i )

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