ML19317G711

Forwards Draft Request for Addl Info Re CP Application. Review of Seismic & Geologic Aspects of Site & Containment Design to Continue

ML19317G711
Person / Time
Site:	Rancho Seco
Issue date:	03/15/1968
From:	Morris P US ATOMIC ENERGY COMMISSION (AEC)
To:	Davis E SACRAMENTO MUNICIPAL UTILITY DISTRICT
References
NUDOCS 8003270687
Download: ML19317G711 (14)

Text

{{#Wiki_filter:___ _ .s ./ i( / SUPPLE'stTA!. 5UPPLEBENTAL Docket No. 50-312 FILF. U FILE CO?Y _ w? e t / Sacramento Municipal Utilities District , i " Post Office Box 15830 N Sacramento, California 95813 I, li y f Attention: Mr. E. K. Davis _.f' M' General Counsel s, \\ Centlemen: ,m This letter refers to your application for a construction permit and operating license for the Rancho Seco Generating Station (Unit No. 1) to be located k, in Sacramento County, California. Representatives from Sacramento Municipal Utilities District, Bechtel Corporation, and Babcock and Wilcox Company met with the Regulatory Staff on February 5 and 6,1968, and on February 29 and March 1, 1968, for detailed discussions of your application. As indicated in these meetings, the information submitted in your application, including Amendment No. 1, does not provi,de adequate information for us to continue \\ 3 our review in certain areas. Additional information which we believe e,! D necessary to continue our review is described in the' enclosure. t 7 s Wefare continuing our review M e se'ismic anA geologic aspects of $e 4 site, and the proposed containment (design./ Any Additions 1' questions which J ~ we may have in these areas will be transmitted to you in following letters. / ~ x We urge.that you provide full and complete answers to the enclosed request fo(additional information in orde'r to minimize interruptions in the processing o Fjour application. ~ o J Sincerely yours, N N N N Peter A. Morris, Director Division of Reactor Licensing

Enclosure:

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i REQUEST FOR ADDITIONAL INFORMATION Sacramento Municipal Utilities District (Docket No. 50-312) March 15, 1968 s,. '7 - :, i i 's _ ( 1. GENERAL N.'t 1.1 Update the discussion of your proposed design with respec't to its conformance to the Commission's General Design Criteria. Include in this discussion the impact of the several design changes made in your facility. 1.2 7\\ Defi,neych of your research and development programs with a proposed / if J" schedult for obtaining the desired information. Include, as wt j appropriate, when the design of the associated feature must be > -(h i frozen in order to meet the schedule for construction of the Rancho q q s Seco Plant. j 1.3 If not specifically incleded in 1.2, describe your program for, ,y vibration testing of the core barrel clieck valves. /A -{ 7 m [ 1.4 If not specifically included in 1.2, discuss the programs currenti ~N N

Q in progress that will assure fuel element capability for 55,000

%(I -o% ll s MWD /MTU burn-up at the design power densities.

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'T IJ Submit the staffing and training plans for SMUD's Nuclear Project 3 Engineering Staff. T f g-y %'% Discuss the principal design decisions yet to be made, that require i 1.6 nuclear and steam plant knowledge and which affectypower plant safety. ]j g '; N5 Indicate the approximate dates by which these decisions must be made and to what extent reliance will be placed epon contractors for 4 making decisions. Indicate how the training plans for SMUD personnel are orientated toward these requirements. 3 p questions raised during the review of a similar plant (Metropolitan _ (!:\\ 1.7 Your Amendment No. 1 provided the SMUD response to applicable MdW JM Edison).1 Please update your response to these questions by considering-N ,Y-Qdali g typpckli'tE Nison]4menchme h tiiLEupZtodoQoc lud i_ngj. ,M, ~, r % n sene doe y M P 6{.e :m g par gaa,,e u g qi 2. SITE AND ENVIRONMENT ~ f 2.1 ~ Mnot Uef ne-that methads.usht@bhhe-ihloahtivi yo cQ sectind rya-. nt'isyska. {~An'anilysis should be '~ t I tt me u r ~ ~ ~ * ~ * ~ l['_^***-,

/ F presented which relates primary coolant activity, assumed leakage rate froniJrimarytosecondary/removalandcleanupmechaaismsforthe secondary coolant, and the derived activity contained in the secondary system. 2.2 The PSAR description of the steam generator tube accident includes an assumption that the iodine water to air partition factor is 10,000. Show how this factor was derived, and indicate how concentration, temperature, pressure, and air to water volume ratio which exist throughout the course of the accident may effect this partition factor. l l 2.3 The PSAR calculations of off-site doses due to release of noble gases include an assumption that the average effectiye energy per disinte-gration of noble gases is 0.4 MEV. The origin or justification of that assumption should b,e provided. 2.4 Submit a listing of the radioactive isotopes and maximum activities i of each which may be present in the liquid waste holdup tanks at any one time, and include an analysis demonstrating that failures in the liquid waste system would not cause excessive release of radioactive liquids to the environs. i 2.5 Specify the distance to the low population zone as it is defined in 1 10 CFR Part 100, Section 100.3 (b). f 2.6 Based on data presented in the PSAR, it appears that the Rancho Seco site is subject to a high frequency of inversion conditions with low transport winds. Data presented show a computed frequency of about 257. extremely stable conditions with an average wind speed of 0.9 meters per second; it would appear appropriately conservative to use this condition for calculating the 2 hour off-site doses. Please provide the environmental consequences of hypothetical accidents using this basis.E Lpt Jha. 4;uadfu t A _ in* b Discuss the vite'r flow [patternKand their associated consequences on 2.7 plant operations following a failure of the on-site water storage facilities. 2.8 Provide a p of earthquake epicenters within a radius of 200 miles showing all earthquakes of intensity V or greater at the epicenter. 3. REAC'lVR 3.1 Discuss your plans for providing a negative moderator coefficient of reactivity throughout core-life in the event detailed studies show g this to be a d ign requiremant. 3.2 Describe your hlopment. of the " power doppler coefficient" given g_;. in Table 3.2-3 of the PSAR and discuss thei(regugney;rtaP'engg of this g coef ficient with fespectM"=p'cC-kauelys@ & daS CJlj-,~lh sh, .NLW h 5% s

r > 3.3 Submit the latest available results of those analyses on xenon oscillations described on pages 3.2-21 and 3.2-23 of the PSAR and specify the dates when the remaining analyses will be completed. 3.4 Discuss the detection system for xenon oscillations and indicate the expected minimum sensitivity of this system during power operation. Describe the 2-D analysis method for xenon'ghddE/dd-AC,s+** (gut MM N l esc:144a %. W 3.5 i h, 3.6 Assuming that control rods are used to stabilize xenon oscillations, give the maximum values anticipated for the transient and steady-state errors in local power' density at the hot spots. 3.7 Indicate the margin of xenon stability by givi5g the_ power level at 6 which xenon oscillations are predicted to occur p h l

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0 1 various times A core life. ' M.y \\ 3.8 Discuss the fuel auinagement plans and techniques that will limit maximum fuel burn-up to 55,000 MWD /MTli and describe the associated uncertainties. 3.9 Discuss your calculational model and indicate the error band on the fast neutron flux (E ) 1.0 Mev) at the gressure vessel inner (n/cm2.,,e), surface which was calculated to be 3.4 x 10 l Include in the discussion: l a) How azimuthal variations are treated in the analysis and relate I these to the azimuthal placement of the surveillance specimens. l b) The uncertainties associated with the attenuation factor of 6.0 x 10"/ 3.4 x 10'Uor 1760 and relate their potential consequences to higher values of NDTT for the pressure vessel wall. c) The maximum fast neutron exposure (see pg. 4.1-8) is indicated to be 3.0 x 10" (n/cm ) or, at 607 load factor,1.9 x 10'0 2 2 (n/en -sec). Explain the relationship between this design limit and the data given in Table 3.3-7 of the PSAR with respect to the factor of 2 conservatism indicated on page 3.2-14 (Q{,%2Y Cm ~ 3.10 Discuss the probability for a single fuel pin to unbrgo DNB during \\ the first three years of,nommob.ici3% power operation { (Alternatively, specify the number of fuel pins that have greater than 507. f (N-probability for undergoing DNB during three years of 6 power operation. Include in your discussion: L %Wd h M.Mn> a) The potential consequences of a single fuel pin undergoing DNB during full power operation. b) The time behavior of events that occur in those fuel pellets located in 'the vicinity of the DNB surfaces. L

l c) Definition of the word " jeopardy" as used in the PSAR to describe the conclusions of your statistical analyses. 4. REACTOR COOLANT SYSTEM AND OTHER CLASS I SYSTEMS 4.1 Thermal Shock With regard to thermal shock on reactor compo$ents, induced by operation of the emergency core cooling system (ECCS), provide details of an l analysis which indicates that the reactor vessel and reactor internals can withstand the rapid temperature change at the end of their design life. The analysis should' include both the ductile yielding and the i brittle fracture modes of failure. 4.1.1 The brittle fracture analysis for the vessel should assume an initial crack si*ze just below the critical crack size corresponding to the stresses present during normal operation and transients. Since the initial crack is most likely to exist in a weld or a heat affected zone, the analysis should consider two cases: a circumferential crack, and a crack parallel to the axis of the reactor vessel. The details of the analysis should be provided including specific information on: (a) The critical stress intensity factor (KIC) assumed, and the basis for its selection, (b) The assumed time-integrated neutron flux (nyt) at the reactor vessel inner diameter, (c) The value of residual stresses assumed in the base metal and the weld areas, (d) The initial crack geometry and size assumed in the analysis, (e) Equations used to correlate crack size with the calculated stress intensity factor (K ). y 4.1.2 The details cf the ductile yiesding rode of analysis for the vessel should include the following information: (a) The geometry of the plate and the cooling method assumed in the analysis, (b) The heat transfer coefficient used, its experimental basis, and the degree of conservatism involved, (c) The initial temperature of the vessel as a function of time delay in injecting the cold water, 4 (d) The effect of axial temperature gradient in the vessel, during filling with cold water, on the r;tal stress intensity and the distortion of the vessel, i (e) The temperature profiles and the calculated thermal stress profiles through the thickness of the plate for several times during the cold water injection transient,

1 i l r (f) The magnitude of the axial dead load stresses in the vessel, (g) The magnitude of the stresses in the vessel shell due to potential simultaneous seismic loading, (h) The value of the yield stress used as the failure criterion in the ductile yielding analysis. 4.1.3 Based on the analyses for the vessel provide: 1 (a) An estimate of the mar.imum acceptable initial temperature of the vessel that could be tolerated without failure of j the vessel, (b) An estimate of the maximum neutron flux exposure (nyt) of the vessel that could be tolerated without vessel

failure, (c) An estimate of the maximum allowable pressure stress, when combined with other stresses present in the vessel, which could be tolerated without failure.

) 4.1.4 Evaluate the capability of the piping, safety injection nozzles, and vessel nozzles to withstand the transient. 4.1.5 Evaluate the effec:s of this transient on the core barrel and other internals with regard to assuring that distortion would not restrict the flow path of the emergency core coolant. / 4.2 Seismic Design 4.2.1 For all Class I systems and components provide the design basis load combinations and the proposed stress and deformation limits for each combination. 4.2.2 Supply criteria or specific information on the interaction forces, deformation and stresses connected with the relative motions between the reacter vessel, steam generators or other large components. Indicate how these relative motions will be controlled by snubbers or other means, and what reaction forces (and corresponding stresses) will'be transmitted to the pipes. 4.2.3 Identify specific reactor internals which must maintain their functional performance capabilities to assure safe shutdown of the reactor. Provide calculated (or estimated) maximum limits of deformation or stress, at which inability to function occurs, for each component identified. Also, supply the calculated (or estimated) maximum design limit value, and the expected deformation or stress. In all cases identify the applicable loading combination and state the proposed margin of safety. 4.2.4 For reactor internals provide information that will permit evaluation of the effect of irradiation on the material y operties and on the proposed deformation limits, i n -'^Z_,...__

2 4.3 Discuss the full power radiation environment with respect to corresponding damage thresholds for the control rod actuators and the primary loop pumps and pump motors. Consider the N-16 activity, the fission product activity in coolant, and the radiation streaming contributions.

4.4 Provide a tabulation of all the nuclear pressure vessels in the Class I (seismic design) systems in the facility. The tabulation should inelude a notation of whether the vessel design is complete, the stage of fabrication of the vessel, and the extent to which each of the vessels will comply with each of the 34 supplementary criteria in " Tentative Regulatory Supplementary Criteria for ASME Code-Constructed Nuclear Pressure Vessels", issued by AEC Press Release No. IN-817, dated August 25, 1967.

For each vessel, provide a discussion that represents the reason why total compliance is not. feasible for each criterion not met in its entirety. 4.5 Submit Certified Code Design Specifications for component parts of the Class I systems as required by the ASME Code Section III, paragraph N-141 (passed 6-23-67). 1,. ? 7 6. ENGINEERED Mii m v _ / I1'vi m y1duZ w hp2 ri7V e 6.1 pgTose design revisionsg described at our March 1,1968 meeting.Q7 t G-h m-< L - Provi e} d e 'p st-a cident radiation.,ept in the containment. Compare M Nea)nticipatedgammaexposureswithdamagethresholdsforth 6.2 %3Ateggarps -systW H g~ ulhe test programs that will assure adequate performance of the W-a: e 6.3 t 4 engineered,sa @ %_.s in the post-accident environment. QGf p 6.4 Provide an evaluation of the ultimate iodine removal capability for the proposed spray systems that car. be rigorously supported by presently available experimental evidence. Include a discussion of spray effectiveness in removing aerosols. 6.5 Provide an analysis of the physical aspects of the proposed spray systems, including the fraction of the entire containment volume directly covered by the sprays, the convection circulation into the spray pattern, the range of drop sizes, and the relative temperatures of spray and containment air with their effect on iodine removal rate and efficiency. 6.6 Discuss the extent to which reversible, competitive, and slow chemical reactions have been considered in the evaluation of the effectiveness of the spray systems. Consider the contribution of liquid film mass transfer resistance in the calculation of the overall mass transfer coefficient. w

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1. 6.7 Provide an analysis of the composition and pH of the emergency core cooling solution as a function of time following the design basis loss-of-coolant accident. Consider spray system additives, soluble neutron poisons, fission and corrosion products, elements leached from concrete, etc.

6.8 Provide a discussion of the extent to which exposure to the solution discussed in item 6.7 above will be factored into the procedure for selection of materials for the engineered safety features for the fa'cility. Discuss the systems that will be affected and the nature of the considerations that will be taken into account. 6.9 Discuss the time, temperature, and radiation dependent stability of the spray solution under both storage and p~ost-accident recir-culating conditions and indicate the possibility of forming solid decomposition products o'r precipitates which could potentially interfere with system performance. 6.10 Discuss both the time-dependent radiolytic and chemical hydrogen formation under post-accident conditions for the solution given in item 6.7 above. IncludeanestimateoftotalTand@ activity in both the core and in the liquid, and of the total expected irradiation dose characteristics. Indicate the extent of hydrogen formation by chemical reaction (corrosioa) with exposed reactor materials. 7. INSTRUMENTATION AND CONTROL 7.1 Discuss and evaluate the differences between the SMUD Station, Babcock & Wilcox designed protection systems which initiate reactor trip and engineered safety feature action and those to be incor-porated in the Three Mile Island Station (Docket No. 50-289). The discussion should include preliminary design of the complete circuit from sensors to actuation logic. p4 '.Y ' 7.2 With respect to the reactor protection and engineerea safety feature g actuation circuits to be designed by othgr than Babcock and Wilcox, identify the design features which ---f2*'ejhMMh the proposed IEEE staadard for Nuclear Power Plant Protection Systems. Justification for all (6pfli6ft'should be provided. ,s . t //"'O . (L 4U11 E Cc% fhtt;' f2 r. Describe and evaluate the.cr rion for the physical identification 7.3 of the reactor protection and engine}ered safety feature equipment including panels, components, and cables. 7.4 Describe and evaluate the changes which will be made in the design of the instrumentation and control systems as a result of the ACRS recommendations contained in the Three Mile Island letter. Include in the discussion:

_ _ _. i _ _ _. k r l (a) Diversity of engineered safety feature actuation signals and (b) Separation of control and protection systems. 7.5 Identify the instrumentation and electrical equipment which must function in an accident environment. Discuss and evaluate the i qualification testing which is necessary to insure that this equip-ment will function in the accident environment. Your intentions with respect to obtaining the required data should be discussed. 7.6 With respect to the reactor protection and engineered safety feature signals which feed annunciators and/or a data logging computer, describe and evaluate the design criterion to be used to assure circuit isolation. 7.7 Identify and discuss the differences between the SMUD Station, i I Babcock and Wilcox designed control systems and those to be incor-porated in the Three Mile Island Station (Docket No. 50-289). This discussion should include an evaluation of the safety signifi-cance of each system. l 7.8 Identify, discuss, and evaluate the differences between the SMUD Station in-core instrumentation and that to be incorporated in the Three Mile Island Station (Docket No. 50-289). 7.9 Describe the control room ventilation system and evaluate the need for placing the system automatically in a recirculation mode utilizing an airborne radiation detector which monitors the intake duct. 8. ELECTRICAL SYSTEMS 8.1 In the evaluation of the ability to supply power to engineered safety features from offsite sources, consider the effect of the sudden tripping of the unit. In addition to the effect on system stability, consider coincident failures in the generating station switchyard to assure that none will cause the loss of all offsite power to the station. Consideration should be given to but not be limited to the following: faults, circuit breaker failures, control circuit failures, and battery failures. 8.2 Evaluate the ability of the offsite power to meet General Design Criterion 39 with the proposed single startup transformer. f"2' /GEd 'equence -@_ P41esel ' Q WH& 8.3 Describe and evaluate the automatic i i p /p%&.w l 8.4 Provide an evaluation of loads (HP) required to be powered in the interest of safety and the relationship of the maximum emergency load that may be placed on each diesel generator to the rating (KW) of the generator.

Ju 5, 8.5 Describe and evaluate the provisions to prevent two diesel generators from being connected together and from being connected to another source of power that is out of phase. 9. AUXILIARY AND EMERGENCY SYSTEMS 9.1 Submit the design revisions for the cooling water systems that were described at our meeting on March 1, 1968. 9.2 Discuss the maximum extent (frequency and duration) to which reservoir make-up water will be used in the event of canal water supply system outage. 9.3 Discuss the plant's capability for detecting fuel failure. This discussion should include the detection time as a function of fuel failure severity. 9.4 Submit a brief statement of your provisions in the emergency cooling water supply to cope with the lowest anticipated ambient tempera-tures (ev19 F). 9.5 Discuss the provisions for draining the spent fuel pool. 9.6 Discuss the potential for inadvertant draining of the spent fuel f pool. 9.7 Discuss the potential for draining the water in the fuel transfer canal and tube and specify the required fission product decay period after which the fuel elements do not require water cooling. J ,1 12. CONDUCT OF OPERATIONS N Discussthej$jtt[N$% iN' 5 A t? wo15cT,ngunprOt6ur between SMUD, Bechtel, B&W, WEC, and [ 4 12.1 others. BThis discussion shiould include a list of the subsystems and support functions provided by the principal patties. h& ' < U:W_f 1/m:'lw.ft $ s EdJ M.3 A1 ! '^, !b 50I,.EddQ,.WE0' E' W 12.2 Provide organization" char % @ the construction phase and the e - i operations phase, %bhh 12.3 Expand MQtied organizational charth to show lines of responsi, bility for quality control efforts during the construction phase. 12.4 Submit, organizational chart for the Bechtel Corporation indicating . (_ responsibility channels for Quality Assurance and Quality Control oC @ effortsE Delineate,home office as well as site groups. / Q, \\ l MMMS

4g 12.5 Submit,

organizational chart for the Babcock and Wilcox Company

~ indicating responsibility channels for Quality Assurance and Quality Control efforts. 9

12.6 Submit the staffing and training plans discussed at DRL on February 5, 1968 for the Rancho Seco No. 1 operating personnel.

13. INITIAL TESTS AND OPERATIONS 13.1 Discuss the extent to which test results will be documented. 13.2 Discuss your plans for measuring and/or verifying the threshold conditions for xenon oscillations. Include in your discussion the extent to which data from earlier plants will be used. 13.3 Provide a detailed outline of the test program for each engineered safety system. The outline should provide a set of test objectives for each system, a brief description of the proposed test, and a brief discussion on how achievement of design objectives can be assured. 13.4 Provide the following information in outline form regarding emergency planning for the SMUD facility: (a) Plan objective, (b) Scope, (c) Delineation of responsibility and authority for plan implementa-

tion,

/ (d) Notification liaison to be established with federal, state and local authorities and emergency assistance personnel that they provide. A (e) g3calj provisions made with local hospital and physicians for p treatment of injured persons, including contaminated persons. .7 > (f) Instrumentation to be installed Meh3fatiftswith readouts in ) the control room to be used for assessment of the extent of a radioactive release, both on site and offsite. (g) Proposed training of onsite staff and means to be used to evaluate the plan's effectiveness on a periodic basis. 14. SAFETY ANALYSIS 14.1 Describe the analytical model used to study the reactor s stem response to a 1007. loss of demand load and to dEhrdtgo 29 4 W c]- 0-. C. >ct.+ u, Provide the foll;owing results of your nnalysis of the load loss 14.2 transient: i a) Rise in average moderator temperature, b) Minimum DNB ratio during the transient, c) Rise in reactor loop pressures, d) Extent of turbine over-speed, e) The reactor thermal power transient, and f) Fuel and clad temperatures. m ~

. 14.3 Describe the natural circulation characteristics of the primary loop system. Will operation of primary loop relief valves, due to its dead-band characteristics, affect this flow? 14.4 In Figures 14.2.1 through 14.2.11. of,the PSAR, the reactor kinetic p' I corresponding values for 8 eff ? g [, parameters are given for C(d e Q' , and g. What were the q ,!h; ' J t h*p% i' / 14.5 "'d ef - LL.s :c.+g.g. ; -das:i!E, htj.;,c4'd,.A.p y ca.64/ DiscusT the p;; ::;h used in calculating the, effective delayed, , neutron. frac tion t JJ / 3 ! a s J w a y g1_ ) 1; 14.6 Discuss'the accuracy of the energy yield predictions 'for the rod ejection accident. Your discussion should include the anticipated i$f power profile transients. How is spatial dependence treated for h eff k 1*? t a lu d 22 a.m.cedG,}4 14.7 g, $ % m t {.a t. As f ul jk%w ch. sj% IU ~y ry,c.fO Q 14.8 For theassuuFt.ry ejection accidentg, discuss the predicted pressure 7 pulse in the reactor vessel and the associated uncertainties. N,g

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7 14.9 Discuss the potential for reactivity]and.. associated consequences 's when a repaired pump is returned to service.

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/ ~ ' 15. TECHNICAL SPECIFICATIONS ? \\ c em/ M., / - hMA. W h-s \\ / t 15.1 Identify those items that will ev tud11y be classifie[ as technical specificatiens that now affect plan design.. Examples include the' generators; and in-core flux monitors P h (- M ] emergency minimum conditions of operation on: M -- -- J j i ~

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RESPONSE TO ACRS LETTER ON METROPOLITAN EDISON'S PLANT i y Sever O or Nse / sir,ua,ph of,thesejiteyts appear 21sewhere indhis enc'losuretiops referen b ,a Sg 'Q c ', /. / yJ w c. '9 [# a l Disc ^uss (ctudti,y6ur efforts t@eibp "4fversity ofyrinciple" in the , j \\ -h[\\ * - M re,,v*assehe)ACRS recommendaHon fn_-EKM on 6 Lehe\\ECCS' w to ~ LemdNc/ d '/lvJA,y }sj o e " L,t/ ,4 I.; { g * - 3 - y 2 _ ry W garding potential failure to de-energize the scram i j.] gl bus. g 4 Qy% (16Q~~,Dfbcu#yakr eT.focesT to separatClontroT~from-pr'ot' action ~~+nahume'nts'. v) d'5 d p' pchW 7ko2 Mll #4d A ) 16.g.utscus our _ eioop apability for prompt detection of Ndjy */, D / * '"p* * / '*t 7 " *i. N F ' 4 A h lY 16.3 Discussryour program e6to assure that fuel failures will not \\ I significantly inhibi the ECCS from preventing clad melting. l j l t a-,

. (fkYi4L G.ur a LO.~] i 4u,w i.' zL CL" 16[Discussyour for use fpargJength ods to con % xenon ^ ~' c f. g W b,e /S, Q

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16. { Discuss the effect o blowdown forces on reactor internals by identifying appropriate load combinations and deformation limits, g

r \\B-k oAUN 16 4 Discussy9our program to experimentally study vibrations in check ( valves. A 6 A 3 t t .,. -}}