ML19319D987
| ML19319D987 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 03/18/1968 |
| From: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Davis E SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| References | |
| NUDOCS 8003270698 | |
| Download: ML19319D987 (12) | |
Text
Lu SUPPLEMENTAL 1:g_E COP'L.
TilIS DOCUMENT CONTAINS POOR QUALITY PAGES Docket No. 50-312 Sacramento Municipal Utilities Listrict Post Office Box 15830 Sacramento, California 95813 Attention:
Mr. E. K. Davis General Counsel Gentlemen:
i This letter refers to your application for a const ruction permit and operating l
license for the Rancho Seco Generating Station (Unit No. 1) to be located 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 meet our requirements for the contents of applications as specified in 10 CFR Part 50 and elsewhere.
The purpose of this letter is to document the areas of deficiency and to inform you we will take further action on your application until the additional information not is provided.
The information we need is discussed in some detail in the attached " Request
/
for Information."
We urge that you provide full and complete answers to the
/
enclosed request lor.. additional information in order to minimize interruptions in the processing of your application.
Sincerely yours, Peter A. Morris, Director i
Division of Reactor Licensing
Enclosure:
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REQUEST F0ft INFORMATION Sacramento Municipal Utilities District (Docket No. 50-312)
March 19, 1968 1.
CENERAL 1.1 Update the discussion of your proposed design with respect to its conformance to the Commission's Proposed General Design Criteria.
Include in this discussion the impact of the several design changes made in your facility.
1.2 Describe each of your research and development programs with a proposed schedule for obtaining the desired information.
Include, as appropriate, when the design of the associated feature must be frozen in order to meet the schedule for construction of the Rancho Seco Plant.
1.3 If not specifically included in 1.2, describe your program, including schedule and acceptability criteria, for vibration testing of the core barrel check valves.
1.4 If not specifically included in 1.2, discuss the programs currently in progress that will assure fuel element capability for 55,000 MWD /MTU burn-up at the design power densities.
1.5 Submit the staffing and training plans for SMUD's Nuclear Project Engineering Staff.
1.6 Discuss the principal design decisions yet to be made that require nuclear and steam plant knowledge and which affect nuclear power plant safety.
Indicate the approximate dates by which these decisions must be made and to what extent reliance will be placed upon contractors for making decisions.
Indicate how the training plans for SMUD personnel are orientated toward these requirements.
1.7 Your Amendment No. 1 provided the SMUD response to applicable questions raised during the review of a similar plant (Metropolitan Edison).
This response used information that was available through November, 1967.
Please update your response to these questions by considering applicable information that became available in January, 1968.
2.
SITE AND ENVIRONMENT 2.1 An analysis should be presented which relates primary coolant activity, assumed leakage rate from the primary to secondary system, removal and cleanup mechanisms for the secondary coolant, and the derived activity i
contained in the secondary system.
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, 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.
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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 ef fect this partition factor.
2.3 The PSAR calculations of of f-site doses due to release of noble gases include an assumption that the average ef fective energy per disintegration of noble gases is 0.4FEV.
The origin or justification of that assumption should be provided.
2.4 Submit a listing of the radioactive isotopes and maximum activities 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.
2.5 Specify the distance to the low population zone as it is defined in 10 CFR Part 100, Section 100.3 (b).
2.6 Dased on data presented in the PSAR, it appears that the P.ancho Seco site is subject to a high ftequency of inversion conditions with low transport winds.
Data presented show a computed frequency of about 25' 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 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> off-site doses.
Picase provide the environmental consequences of hypothetical accidents using this basis.
2.7 Discuss the water flow patterns in the vicinity of the plant and their associated consequences on plant operations following a failure of the on-site water storage facilities.
2.8 Provide a map of earthquake epicenters within a radius of 200 miles showing all earthquakes of intensity V or greater at the epicenter.
3.
REACTOR 3.1 Discuss your plans for providing a negative moderator coef ficient of reactivity throughout core-life in the event detailed studies show this to be a design requirement.
3.2 Describe your derivation of the " power doppler coefficient" given in Table 3.2-3 of the PSAR and compare the time constant of this coefficient with that of the system in the analytical model.
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.
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. 3.4 Discuss the detection system for xenon oscillations and indicate the expected minimum sensitivity of this system during power operation.
3.5 Describe the 2-dimensional analysis method for evaluation of xenon insta-bilities.
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 stabilit y by giving the pc*.er level at which xenon oscillations are predicted to occur at various times during core life.
3.8 Discuss the fuel management plans and techniques that will limit maximum fuel burn-up to 55,000 K.a/MTU and describe the associated uncertainties.
3.9 Discuss your calculational nodel cnd indi,cate the error hand on the fast neutron flux (f 1.0 Mcv) at the pressure vessel inner surface which was calculated 'to be 3.4 x 10IO (n/cm2 - sec).
Include in the discussion:
a) llow azimuthat variations are treated in the analysis and relate these to the azimuthal placement of the surveillance specimens.
b) The uncertainties associated with the attenuation f actor of 6.0 x 1013/3.4 x 1010 or 1760 and relate their potential censequences 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 19(n/cm ) or, at 807. load factor, 1.9 x 1010 (n/cm2 - sec).
2 3.0 x 10 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 of the PSAR.
3.10 Discuss the probability for a single fuel pin to undergo DNB euring the first three years of power operation at rated conditions.
(Alternatively, specify the number of fuel pins that have greater than 50~' probability for undergoing DNS during three years of power operatica at rated conditions).
Include in your discussion:
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.
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. c) Definition of the word " jeopardy" as used in the PSAR to describe the conclusions of your statistical analyses.
4.
REACTOR C001. ANT SYSTEM AND OTHER Cf. ASS I SYSTEMS 4.1 Thereal Shock With regard to thermal shock on reactor components, induced by operation of the emergency core cooling system (ECCS), provide details of an analysi* % hic h indicates that the reactor vessel and reaetor internals can withstand the rapid temperature change at the end of their design life. The analysis should include both the ductile yielding and the brittle fracture nedes of failure.
4.1.1 The brittle fracture analysis for the vessel should assume an initial crack. size just below the critical crack size corresponding to the stresses present during normal operatien and transients.
Since the initial crack is mast likely to exist in a.cld or a heat affected zone. the analysis sho,uld consider two cases: a circumferent ial 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 (nvt) at the reactor vessel inner dianeter, (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 of the ductile yielding mode 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, (d) The ef fect of axial temperature gradient in the vessel, during filling with cold water, on the total stress intensity and the distortion of the vessel, (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,
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. (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:
(a) An estimate of the maximum acceptable initial temperature of the vessel that could be tolerated without failure of 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 effects of this transient on the core barrel and other internals with regard to assuring that distort ion would not restrict the flow path of the emergency core coolant.
4.2 Seismic Desien 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 reactor vessel, steam generators or other large components.
Indicate how these relative mntions will be cuntrolled 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 inforumtion that will permit evaluation of the ef fect of irradiation on the material properties and on the proposed deformation limits.
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imary loop pumps damage thresholds for the control rod actuators and t e prConsider i
4.3 in coolant, and the radiation streaming contributions.
and pump motors.
in the Class I Provide a tabulation of all the nuclear pressure vesselsThe taSulation should in the stage of fabri-(seismic design) systems in the facility.
4.4 lt a notation of whether the vessel design is comp e e,to which each of the vessels will in "Tentat ive Regulatory cation of the vessel, and the extent comply with each of the 34 supplementary criteria Frcssure Vessels",
1 Supplementary Criteria for ASME Code-Constructed Nuc car 23, 1907.
issued by AEC Press Reicase No. IN-817, dated August he (cason why total For each vessel, provide a discussion that represents t in its entirety.
leasibic for each criterion nut met compliance is not for component parts of the II. paragraph N-141 Submit Certified Code Design SpecificationsClass I systems a 4.5 (passed 6-23-67).
ENGINEEPFD SAFETY FEATURES _
March 1, 6.
Update your PSAR with those design revisions described at our 6.1 1968 nmeting.
levels in the radiation dose Provide the anticipated post-accidentCompare the anticipated gamma expo 6.2 containment.
holds for the engineered safety features.
rformance of the Describe the test programs that will assure adequate pe i nment.
engineered safety features in the post-accident env ro 6.3 bility for the Provide an evaluation of the ultimate iodine temoval capa ted by presently proposed spray systems that can be rigorously supporInclude a discussion of s 6.4 available experimental evidence.
in removing aerosols.
d spray systems, Provide an analysis of the physical aspects of the proposedirectly covered l
including the fraction of the entire containment vo ume ttern, the 6.5 by the sprays, the convection circulation into the spray paf spray and co range of drop sizes, and the relative temperatures o fficiency.
air with their ef fect on iodine removal rate and e d slow chemical Discuss the extent to which reversible, competitive, anf the effectiveness of reactions have been considered in the evaluation oConsider the cont f
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coef ficient.
resistance in the calculation of the overall mass trans er the spray systems.
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. 6.7 Provide an analysis of the composition and pH of the emergency core cooling solution as a function of tine following the design basis loss-of-coolant accident.
Consider cpray system additives, soluble neutron poisons, fission and corrosion products. elements leached fram cen: rete, 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 facility.
Discuss the systems that will be affected and the nature of the considerations that will be taken into account.
6.9 Discuss the tine, temperature, and radiation dependent stability of the spray solution under both storage and post-accident recirculating conditions end indicate the possibility of forming solid decomposition products or 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.
Include on estimate of total / and 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 (corrosion) with exposed reactor materials.
7.
INSTRUMENTATION AND CDNTROI.
7.1 Discuss and evaluate the dif ferences between the SMUD Station. Dabcock &
Wilcox designed protection systems which initiate reactor trip and engineered safety feature action and those to be incorporated in the Three Mile Island Station (Cocket No. 50-289).
The discussion should include preliminary design of the complete circuit from sensors to actuation logic.
7.2 With respect to the reactor protection and engineered safety feature actuation circuits to be designed by other than Eabcock and Wilcox, identify the design features which differ from the proposed IEEE standard for Nuclear Power Plant Protection Systems. Justification for all dif ferences should be provi'ded.
- 7. 3 Describe and evaluate the criterion to be used in providing for the physical identification of the reactor protection and engineered 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 syattems as a result of the ACRS recommendatiods contained in the three Mile Island letter.
Include in the discussion:
(a) Disersity of engineered safety feature actuation signals and (b) ' Separation of control end protection systems.
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. 7.5 Identify the instrumentation and electrical equipment which must function in an accident environme nt.
Discuss and evaluate the qualification testing which is necessary to insure that this quipment will function in the accident environment. Your intentions with respect to obtaining the required data should be discucsed.
7.6 With respect to the reactor protection and engineered safety feature si nals 6
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 nad d.-euss the dif ferences between the SMU:. ' tat ion. Babeoek and Wilcox designed entrol system.4 and those to be incorporated in the Three Mile Island Etation (Locket No. 50-289). This discussion should include an evaluation of the safety significance of each system.
7.8 Identify, discuss, and evaluate the differences betseen the SMUD Station in-core ins t rumentat ion and that to be incorporated in the Three Mile Island Station (Docket No. 50-289).
- 7. 9 bescribe the control room ventilation system and evaluate the need for placing the system automatiently in a recircul..t ion mode util i zing an airborne radiatien detector which c:oni tors the in;ake duct.
8.
E!ECTRICAL SYSTEMS 8.1 In the evaluation of the ability to supply power to engineered safety features from of fsite sources, consider the ef fect of the sudden trippii.g of the unit.
In addition to the effect oc system stability, consider coincident failures in the generating station switchyard to assure that none will cause the loss of all of fsite 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.
8.3 Describe and evaluate the automatic loading sequence for the emergency diesel generators.
8.4 Provide an evaluation of loads (HP) required to be powered in the interest of safety and the relationship of the maximum cmergency load that may be placed on each diesel generator to the rating (KW) of the generator.
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.
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9.
AUXII,IARY AND EMERCENCY SYSTEM 4 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 fuct failure.
This discussion should includ; thn 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 temperatures ( ~ 190 F).
9.5 Discuss the provisions for draining the spent fuel pool.
9.6 Discuss the rutential for inadvertant draining of the spent fuel 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 af ter which the fuct e tements do not require water cooling.
12.
CONDUCT OF OPERATIONS 12.1 Discuss further the relationships between SMUD, Ecchtel, BSL', WEC, and others.
This discussion should include a list of the subsystems and support functions provided by the principal parties.
12.2 Provide organization charts that show the contributions by SMUD Bechtel, B&W, WEC, and others during the construction phase and the operations phase.
12.3 Expand organizational charts in the PSAR to show lines of responsibility for quality control efforts during the construction phase.
Submit an ors aizational chdrt for the Bechtel Corporation indicating 12.4 a
responsibility channels for Quality Assurance and Quality Control efforts for this project. Delineate home office as well as site groups.
12.5 Submit an organizational chart for the Babcock and Wilcox Con any indicating c
responsibility channels for Quality Assurance and Quality Co: Frol efforts.
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.
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. 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 pregram for each engineered safety system. The outline should provide a set of teat 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 fetiusing information in outline form regarding emergency planning fer the 5ML*D facility:
(a) Plan objective, (b) Scope, (c) Delineation of responsibility and authority for plan implementation, (d) Notification liaison to be established with federal, state and local authorities and emergency assistance personnel that they provide.
(e) Provisions made with local hospital and physicians for treatment of injured persons, including cont amin.ited pe r::ons.
(f) Instrumentation to be in.=talled with readouts in the control room to be used for.o er w ont of the extent of a r.ui s oa c t i v e release, both on site and offsite.
(g) Proposed training of onsite staff and means to be used to evaluate the plan's ef fectiveness on a periodic basis.
14.
SAFETY ANALYSIS 14.1 Describe the analytical model used to study the reacter rystem response to a 100% loss of demand load and to total loss of a.c. power.
14.2 Provide the following results of your analysis of the load loss transient:
a) Rise in average moderator temperature, b) Minimum DN3 ratio during the transient, c) Rise in raactor loop pressures, d) Extent of turbine over-hpeed, e) The reactor thermal power transient, and f) Fuel and clad temperatures.
14.3 Describe the natural circulation characteristics of the primary loop system.
Will operation of primary loop relief valves, due to its dead-band charac-teristics, affect this flow?
14.4 In Figures 14.2.1 through 14.2.11 of the PSAR, the reactor kinetic para-Y meters are given for ;(j,c4,, j, and T'. What were the corresponding values for 6 eff ?
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. 14.5 Discuss the technique used in calculating the effective delayed neutron fraction and include a summary of your calculations for the end-of-life value.
14.6 Discuss the accuracy of the energy yield predictions for the rod ejection accident.
Your discussion should include the anticipated power profile transients.
14.7 llow is spatial dependence treated for ' ef f, k t*? Evaluate the uncertainty in peak power densities a ssociated with this approach.
14.8 For the rod ejection accident (Section 16.2.2.2 of the P5AR), discuss the predicted pressure pulse in the reactor vessel and the associated uncertaintics.
14.9 Discuss the potential for reactivity insertion and the associated conse-quences when a repaired pump is returned to service.
15.
TEC11NTCAL SPECTFICATIONS 15.1 Identify those items that will eventually be clase.lfied as technical specifications that now offect plant design.
F xampl es include the minimum conditions of operation on: engineered safety (catures ;
emergency generators; and in-core flux monitors.
16.
RESPONSE TO ACRS LETTER ON METROPOLITAN EDISON'S PLANT 16.1 Describe and evaluate the design changes that will be made in the reactor scram system as a result of the ACRS recommendation contained in the Three Mile Island letter regarding potential failure to de-energize the scram bus.
16.2 Discuss and evaluate your design changes that will provide the capability for prompt detection of gross failure of a fuel element.
16.3 Discuss and evaluate your program of analysis and design directed to assure that fuel failures will not significantly inhibit the CCCS from preventing clad melting.
16.4 Discuss your analysis and design efforts for uf 6? part length rods to control potential axial and diametral xenon 9tilit;.tichs.
Inclur'e in this discussion a description of your latest d.h i n concept and
- s estimated performance characteristics.
16.5 Discuss the effect of blowdown forces on reactor '-'.ernala by identifying appropriate load combinations and deformation lisats.
16.6 Discuss and evaluate your program to experimentally study vibrations in the check valves.
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