ML20127N008
| ML20127N008 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 08/15/1966 |
| From: | Case E US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Pack D COMMERCE, DEPT. OF |
| References | |
| NUDOCS 9211300536 | |
| Download: ML20127N008 (1) | |
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Mr. Seas 1d M. Paek Office ef Metoerological Researeh U
W. S. Westbor Buress U. 8. Depertaent ef Osmasroe
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Washington, D. C.
20 40 Dear Mr. packt e
Enclosed herewith are Volumes I and 11 of the Facility Descripties and Safety Analysis Report filed by Northern States Power Ceapany with its applicaties for authorisaties to seastrust and operate ai
- bettias water amolear resster on its site loested on the west beak of the MiseLasippi River, apptesimately three miles metthuest ef Montleelle in Wright County, Minnesota.
We unuld appreciate receiving a report from the Offies of Mateerelegical Research en the meteorelegical apposts of the proposed reacter leasties which may have a bearing spen sur ' safety review.
ca the basis of our tentative _ schedule for this - project, it would be desirable to have your report available for ou; use by movember 1, 1966.
Sincerely yours, i
i Edsea-G. Case Assistaat Director-Diviston of teacter Licensing Enciesurest l
As stated above i.
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Docket No. 50-263 C. Esmeeroom i
E. S. Onse L. Eornblith (2)
N. D. Mason K. Eniel K. Woodard Northern States Power Company 411. Uicollet Avenue Minneapolis, Minnesota 55kO1 p
Attention:
Mr. D. F. McElroy Vice President, Engineering Gentlemen:
On October 11-12, 1966, and again on December 2 and 12,1966, representatives of Northern States Power Company met with the staff of the Division of Reactor Licensing to discuss your application for a Construction Permit and Facility License for a nuclear power plant at a site near Monticello, Minnesota.
During the earlier meeting, it became evident that changes to the engineered safeguards, as described in Facility fescription and Safety Analysis Report, u re being developed.
To assure an accurate understanding of the proposed engineered safeguards, as well as related systems and enslyses, please pro-vide ansvers to the questions listed in the attachment.
The staff vill be available to discuss and amplify the meaning of any of these questions, should this be necessary.
A supplement to the original application on the subject of field erection of the reactor pressure vessel was recently submitted and is currently being reviewed. Specific staff questions on the supplement are not included in this letter, however we request that as a further supplement to informa-tion on field erection, you provide a comprehensive history and evaluation of operating experience (at least U.S. experience) with field-erected pressure vessels. Please emphasize high-temperature high-pressure vessels.
Sincerely yours, Origir.21 Siped by Nur A. vo ris I
Peter A. Morris, Director Division of Reactor Licensing i
Enclosure:
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" ns PAMorris List of questions 12/2,/@j 12/g/66 12/'2). /66
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. REQUEST NO.1 FOR INFORMATION ON
- NORTHERN STATES POWER COMPANY MONTICELLO UNIT NO.1 n
1.0 Accident Analysis Sections 5 1 3 1, 5.1 3 2.13 3.4.2. and 13 3.4.4 of the Preliminary safety -
and Analysis Report should be supplemented.
The following is required:
1.1 Provide e technical description of the analytical blowdown model with an -
essessment of the adequecy of the model.
Provide en evaluation of the blowdown forces on the reactor vessel internal 1.2 structures including thermal transients which would occur during a primary coolant blowdown or post MCA condition.
Discuss the potential-forces and structural capability for the various internals considered.
13 Provide a quantitative evaluation of the blowdown forces generated in the upper reactor vessel region following maximum steam line rupture, i. e., 1x1 separators, baffles, driers, etc.
1.4 Provide an integrated presentation of the system behavior following the MCA (double-ended recirculation line breek) with particular attention-paid to the following espects, including appropriate assumptions, justifications,-
identifications of operating modes and the particularly sensitive error bands on the calculated values.
Reactor Internal Pressure Differences Reactor Coolant Water Levels Containment Thermodynamic Response Core Fuel Thermal Characteristics (including percent fission gas-released from fuel as a function of fuel temperature and time).
1.5 Discuss the accident recovery conditions from the MCA (no core cooling case)
E for the situation in which the containment spray coolers vere inoperative, particular. attention being paid to the pressure and temperature character-istics of the containment.
i 1.6 Provide an evaluation of the behavior of the control-rods and the. fuel
' bundles for a spectrum of depressurization events.
Attention should be paid to the ability of the' control rods to be inserted and other possible j
reactivity effects.
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-2 1.7 Provide e discussion of the design mergins (with respect to deleyed cooling effects) involved with the provided engineered safeguards of core flooding-and core sprey with respect to the MCA, 1.8 Provide the following information with respect to the liquid boron injection I
system:
Minimum boron concentration in the liquid poison and in the total primary system volume es well es a reactivity belence for the spectrum of reactor conditions under which liquid poison should insure reactor shutdown.
Include coolant temperature, coolant void pressure, reactivity and nuclear power as a function of time.
1.9 Provide en snelysis of the loss-of-coolant accident where the core remains covered with veter and radiolytic decomposition of the veter occurs.
Include en snelysis of the potential hydro 6en and oxygen buildup and a discuscion of any hazard potential -that might result.
This enelysis should. include justification of the g-values used.
1.10 Discuas the damage level and design margins of the primary system end internals with respect to reactivity transients as influenced by the release of molten fuel into the coolant.
Discuss the inportance of void collapse reactivity effects due to pressure pulse and cepebility of nuclear overpower circuits to respond to rapid power bursts and cause the control rods to screm.
What is the effect of fuel burnup-on these snelyses?
2.0 Engineered Safeguards 2.1 Discuss the differences in concept and degree of redundancy, if any, be-tween the safeguards for this plant and those proposed for the Quad-Cities reactor.
2.2 Provide the design basis for each of the safeguards systems including flow rates, system descriptions, heet transfer cepecities, initiating signals end instrumentation lo6ie.
2.3 Provide the design basis for the standby diesel-generator including e-system description, eccident power loads, initiating si6nels, instrumentation logic, reliability and eccessibility during accidents.
Show typical pump motor starting leeds vs. time. Justification for not providing redundant on-site power cepebility (e.g., e second diesel-generator) should be included in this discussion.
2 '. L Provide a description of instrumentation used to monitor rest accident' recovery to assure control of release of radioactivity tu the environs.
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' Identify the range of primary system breake-for which each of the engineered' 2.5 safeguards systems will provide adequate. core cooling.
I 2.6 Discuss the capability of the automatic-depressurization system and the conditions which require its use.
2.7 Provide the design basis for end the 7eliebility requirements for the mein steem isoletion valves.
Idscuss prototype tests to demonstrate closure cepobility during eccident conditions and leak tightness during post MCA -
recovery.
2.8 clarify the following:
2.8.1 On page II-6 h a statement is mede about using the spectre of Figure II 6-5 and demping values from Table II-6-3 Later in the same peregraph of Section II 6-31 the statement is made, "if computerized methods of dynemic enelysis are used, the mathematical model-may be subjected to en excursion through the Taft earthquake of July 21,1952 North 69 West' component with en
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applied factor of 0.33."
The statement then goes on to indicate that the structure should be examined under values of twice those given in Fig. II 6-5 as well, or a dynamic excursion with an applied factor of 0.66.
Clarify the meaning of this factor in terms of its use in the procedure, and whether the maximum earth-quake corresponds to valaes twice those indicated in the spectre of Fig. II-6-5 2.8.2 On page II-6-5 it is noted that for Type 2 structures and equipment a minimum seismic horizontal coefficient of 0.10 with e one-third ellevable increase in basic stress vill be used in the design.
State the reason for selecting this value end its consistency as compared to the procedures adopted for the Type I structures.
Provide the basis of the response acceleration spectrum of Fig. II 6-5 In order that we may enelyze more readily the short period range of this spectrum, please provide this -portion of the spectrum on en How -
expanded scale or provide e-logarithmic pl ot of the spectrum.
is the uncertainty considered in calculatiot of period using e response spectrum showing a large change in ecceleration response for e small change in period?
2.8.3 Ibmping values are listed in Table:II-6-3 on page II-6-5.
The demping level for reinforced concrete structures is listed en 5%eritical. We believe that the demping value is a function of the stress level permitted either under design conditions or for -
cefe shutdown. What is the basis for the 5% figure?
What provisions are teken to insure stability or cranes during en earthquaket r
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2.10 Provide the design basis, reference design, structural integrity analysis and proposed surveillance program for the facility atack.
2.11 Some clarification of the treatment of Class I equipment contained within Class II structures is requested.
Are there.any such items required for safe shutdovn?
If so, how is the snelysis handled for these items and how ic their response effected by the response of the Class II structures to I
which they may be etteched?
2.12 Provide a description of the inspection procedures to be followed during the construction of the containment end other critical structures, and identify the organizationel responsibility for quality control and' inspection.
2.13 Provide e description of the design considerations which reflect the require-4 ment that the containment well accommodate the stresses end deformations (v
which might be imposed by earthquake loeds or pipe breaks without impeiring conteinment integrity.
2.1h Describe the design criteria for the biological shield around the drywell.
2.15 Iescribe the design basis for protection of the containment and engineered cafeguer s against internal end external missiles.
a 2.16 Provide the design basis and functional description for the reactor building to torus vacuum breakers.
i 2.17 Provide e discussion comparing the containment design bases for the Monticello facility to those used for Dresden 2 and 3 and Quad-Cities 1 and 2 facilitiec.
Include such considerations as pressures, free volumes, energy inputs (containment capability) from blowdown and metal-water reactions, and typical dryvell end torus unll thicknesses.
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2.18 Provide the' criterie for reactor vessel and coolent system inspection and descrite how thece criteria vill be satisfied.
In particular, identify those areas of the vsssel which can be inspected and outline tentatively e program to be followed over vessel lifetime.
Discuss the requirements for hydrostatic tests above design or operating pressure and any plans to revelidate system integrity over vessel lifetime.
Is there any reason why such tests could not te performed?
30 site Analysis 31 Provide a statement end evaluation, based on the 1980 population projections,_
of the low population distence for this site.
3.2 Provide e description and evaluation of any anticipated meteorological program to be initiated at the Monticello site.
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33 Provide en analysis shoving the minimum dilution to he expected between the condenser discharge outfall and the intakes for the nearest public drinking veter supplies for both an secidentel slug release and continuous release -
of radioactive e'ffluent associated with (1) normal, and (2) lov flow in the river.
State the maximum amount of liquid radiotetive veste which vill be stored in the various on-site containers in relation to the cepecity of these same containers end the maximum rete et which these amounts could be released.
3.4 Provide date on storage cepecity and en estimate of the length of time withdrevel of drinking veter can be suspended for the municipal veter supplies down river from Monticello. Provide date on the veter storage cepecity and the rett of connumption of population centers within 50 miles downctreem of the site.
35 rescribe the cooling tower complex end its operation.
Also discuss any nuclear safety considerations associated with the operation of the cooling towers.
3.0 Provide the justification for the estimated liquid radioactivity discharge rates of 1 mc/ day normally and 250 mc/ day with fuel leeks. Please discuss liquid effluent discherge in relation to 10 CFTt 20 limits.
37 1:escribe the control room ventilation system and criteria for design.
3.8 Provide the Justification for the exfiltration rates in high vinds described in Section 5.2.2.2.
Iescribe procedures and test frequency to periodically demonctrete leek tightness of the reactor building.
39 Iescrite the design bases for the plant to withstand the vind and pressure effects of tornadoes.
3 10 Provide design criterie and a detailed description of the reactor building standby gas treatment system.
h.O Power Plant k.1 Discuss the increase in the boron concentration (5000 ppm vs. 3800 ppm for Quad-Cities) in the temporary control curtains.
What experience is related to this change?
Discuss the structural integrity of the curtains as a function of the proposed irradiation period.
h.2 rescribe the work which assurec that cavitation in the jet pumps vill cause no edverse vib 9tional effects during normel or abnormal operation such es locs of feet se cer, t
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, k.3 Lescribe the procedures used to determine the individual engineered hot channel factors. What is the accuracy or error bend? How vere these values used in the core thermal performance calculations?
k.h Section II.2.2 of the Facility Description and Safety Analysis Report states that the main condenser vill accommodate a 15% load rejection.
Section 11.2 3 3
i recognizes that the bypass valve vill pees up to 15% of the throttle steam dilectly to the mein condenser, and that the combined capacity.of the bypasc velves (three 5% capacity valves) and relief velves is sufficient to keep the reactor safety valves from opening in the event of a sudden loss of full load on the-turbine generator.
The increase in neutron flux, however, eccording to Section 1131 causes a reactor scram. With reference to the foregoing what is the maximum step change in load that can be tolerated without serem, and what effect does the 15% bypass cepebility heve?
lescribe reactor power level, pressure, recirculation flow, bypass steam flow, es a function of time efter step power reductions.
L.5 Identify all components (e.g., power sources, transmission lines, protection devices, protective relaying and communication systems) which are pertinent to maintaining euxiliary site power following en abrupt shutdown of the Monticello plant from 100% power.
Describe the expected behavior of the NSP power network and relate this behavior to the reliebility of site power.
List the minimum out of service equipment, failures, and mal-operations which vill result in the loss of normal auxiliary site power following this transient.
L.6 With regard to the core thermal analysis and considering the effect of flow distributions produced by orificing, what vill te tLe maximum exit quality in the hottest channel? Identify the MCHFB of 1.5 with regard to quality, flov lete, and location in the core.
'wnet confidence is there that this ratio vill not be lessened by exiel power dis tributions other then the.
reference dJ stribution cor sidered in your analysis? If the MCHFB of 1.5 is reached, what vill be the increase in power necessary to reach the critical heat flux?
k.T Provide en enelysis of the buildup of tritium in the primary coolant over the life of the plent.
Consider such sources of tritium as diffusion of fission product tritium throuEh the cledding, activation of additives or impurities in the primary coolant, if any, neutron reactions with boron and photonuclear reactions.
Evaluate the hazard from tritium inventory in the primar: coolent in terms of a steem line rupture.
What meanc of tritium monitoring will be'provided to ensure that excessive concentrations are l
not reached in the reactor coolant or redweste system?
h.8 State the excess reactivity for the hot critical end hot full power (free and equilibrium xenon) condition.
Show net reactivity, xenon reactivity coolant temperature and power level as a function of time if, as e result of equipment failures, only half of the withdrawn control rods scram.
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-7 4.9 What is the maximum allovable reactor vessel cooldown and heatup rates for a normal reactor shutdown and startup, respectively?
k.10 What instrumentation senses a steam line break and causes the steam line isolation valves to close?
k.11 It is stated that the plant may be controlled either by recirculation flow control or control rod movement.
Please describe the relationships and interdependency of these controls.
4.12 Discuss possible courses of action in the event of individual LPRM (in-core neutron sensors) failures to assure that effective core protection is mainteined, k.13 How is accidental control rod movement out of the approved operating pattern detected automatically and what corrective action occurs?. Fbr example, if a control rod drifts or unintentionally moves when other selected rods are actuated while in automatic mode of operation and is undetected by the operator, what vould prevent automatic featurec from compensating the local disturbance and thus propagating or extending an unevaluated opereting condition?
4.1h rescribe the reactor sefety considerations associated with fire in the control room, k.15 State the criteria for detection of primary system leakage within 'the dry-well and the bases for corrective action, rescribe the methods and instrumenta-tion that vill be used, k.16 Provide. a comparative analysis of the natural _ circulation capability _of the Monticello reactor and the Oyster Creek (or Nine Mile Point) reactor to support the following statement:
"The higher natural circulation potential of the jet pumps system also tends to improve the 'hea t-flux-coolant flov' relationship which results from a complete loss of power to the recircule-tion pumps" (pp. IV-2-3 FrfcSAR).
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