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I* l WASHINGTON, D.C. 20E45 ., ACES Committee Members SHB- < - HW -- ~ i U $ i . m acn.v ncrea m i k i Docket' No. 50-220 HE{ K-g g, n :L-- FEB 2 d}vg {- G lj ns _ gr_ AiG -- - _, , SH3 N- U- ~~ d JMH- P - N1P. k- o ,2 $ Niagara Mohawk Power Corporation gg y ,,g, wns_fc_ ,, j ' 300 Erie Boulevard West g7 y^ f4; t Syracuse, New York 13202 Mailed- b2 M ij' - l
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< b Attention: Mr. Minot H. Pratt J j / ,
F . Vice President and Executive Engineer 'Q./g n r Gentlemen: This refers to Amendment No. 2, dated June 1, 1967, to your application Tor a construction permit and operating license for the Nine Mile Poin_t 1( Nuclear Station located in the town of Scriba, New York. l During meetings held on December 19, 1967 and January 22 and 23,1968, we discussed various technical aspects of the facility design with your [l
' representatives. We indicated at these meetings that additional informa-tion would be required to allow us to continue our review. In general, ]
this information is related to design changes made during facility , construction, the engineered safety features , the instrumentation and " l control systems, your safety analysis, and the conduct of station operations. < VK D Examples of design changes that have been made since the construction permit was issued include: the arrangement of isolation valves on the feedvater #/ and emergency condenser lines, the core spray nozzle configuration, the < j initiation signals for the ECCS, the automat:ic isolation capability on the J mechanical vacuum pump and the check-valves and a buffer system included in the liquid poison system. The instrumentation and control system is not described in sufficient detail to allow assessment of the independence of the safety channels. The t capability of instrumentation and equipment located within the drywell to withstand the accident environment is not described. . The consequences of certain types of equipment malfunctions have not been considered in your safety evaluation. For example, neither the consequences ' of an inoperative check valve in a feedwater line following n loss of off-site power nor the effect of failures in the reactor building closed loop cooling system are discussed.
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n-- _- - 1 \ - l Niagara Mohawk Power Corporation FEB 2 01968 The assumptions and equations used to calculate the off-site doses resulting from potential accidents and routtae operation are not included in your safety analysis. The iodine separation factors assumed in your
, analysis are not adequately supported.
The capability to isolate the reactor building normal ventilation system in l the event of a refueling accident to prevent excessive doses is not properly included in the analysis. 1 We vill continue our review of your application with respect to the fore-Boing matters upon receipt of the additional information. A lb t of specific comments in regard to the above. items and others is attached. This list is not complete; however, it illustrates the kind of information needed. We vill be available to discuss and clarify any of the foregoing matters.. Sincerely yours, y Original Signed by Fet:r e P- . h Peter A. Morris, Director L Division of Reactor Licensin6 i
Enclosure:
Request for Additional Information
; cc: Mr. Arvin E. Upton, Esquire LeBoeuf, Lamb & Leiby i '
1821 Jefferson Place, N. W. Washington, D. C. 20006 i i t l ( _m_________m_ . _ _ . _ _
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. ADDITIONAL INFORMATION REQUIRED NINE MILE POINT NUCLEAR STATION 'l
[ l[ > NIAGARA MOHAWK POWER CORPORATION ' Ej m a .k DOCKET NO. 50-220 C E L g$ c. le Ud I. GENERAL a {- ~.E w
- 1. Describe, in detail, the quality control program follwed during the construction of the plant. This report should include an
' identification of the responsible organization (s) with regard to:
establishing performance requirements, selection of appropriate codes, preparation of specifications (design and/or procurement), l-fabrication and installation for all plant systems and structures. For all of the organizations responsible for the quality control of a system or structure, a description of the quality control l r organization ehould be provided.
? 2. It is our understanding that the arrangement and fabrication j'
i methods of the control rod stub tubes used in the Nine Mile Point l T . ( reactor vessel are similar to those used in the Oyster Creek Vessel, i J Describe L:alyses, non-destructive test results and other special s precautions taken to assure that cracking in the control rod stub { tubes vill not occur. This information should te ritmitted as periodic
;. status reports until the evaluation is completed.
l l I 3 Provide the design and fabrication details of the isoletion valves , on the emergency condenser steam supply lines. For other lines which penetrate the containment that do not have automatic isolation valves incide of the containment, prWide the basis and details for the
, design along with an appropriate safety evaluation.
- k. Describe the operating cycle of the remote control valves for l
[l recirculation and minimum f1w for the steam jet air ejectors under loss of off-site power and loss of coolant conditions. E 5 Provide a list of all Class I systems and structures for which a l
} dynamic seismic analysis was conducted. Indicate which of the Class I systems and structures were analyzed for earthquake loads using only the static methods described in the Final Safety Analysis Report.
For each Class I system and structure, discuss the bases used to establish the type of analysis performed. 1 l t ( l
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4 ( , , II. ENGINEERED SAFETY FEATURES 1. Describe the expected performance of the Emergency Core Cooling System (ECCS) with an analysis that gives curves of:
- a. peak clad temperature versus break size, i b.
per cent of fuel rod perforations versus break size, and [ 1
- c. per cent metal water reaction versus break size.
This analysis should consider.the following combinations of
) ECCS operation: - )
- a. auto relief and core spray-b.
one motor driven feedwater pump and core spray. The above analysis should be performed for breaks in both steam and liquid lines.
, 2. If the plant is operating and the electrically driven feedwater ; pumps are not in service, describe the necessary procedures and instrumentation required to put them in service to cope with a ; loss of coolant accident.
( 3 We understend that changes to the core spray nozzle configuration f have been made to improve the distribution of the water over the g core. Provide an . evaluation describing the safety implications [ of the changes. I i 4. Provide an analysis that demonstrates that an adequate net positive i suction head will be avEilable to all pumps connected to the torus j for all postulated operating conditions. State the energy sources j used to calculate the absorption pool temperature, including a curve 1 of integrated decay heat versus time. Describe the allowance made i for deterioration of piping and filters with regard to pressure loss. p 5 Describe the provisions that have been made to cool the pumps located in the lowest elevation of the reactor building. State the required [ ambient conditions for the equipment in the area and an analysis to g demonstrate how this ambient condition will be met. Provide a s f Piping and Instrumentation diagram for the reactor and turbine building i closed loop cooling systems.
- 6. We understand that a high drywell pressure signal can initiate the ECCS. Provide the design details and criteria for this signal.
Describe the instrumentation, controls and interlocks that allow i the' operator to assess the operation of engineered safety feature systems under post-accident conditions. The description should include location, redundancy and testability. (
i { . 8. Provide the basis used to establish the frequency of radiation
+ monitoring of the raw water discharge from the containment spray heat exchanger. ,
i 9 ! 1 Describe the procedure for testing the liquid poison injection
; .. system. Describe the action to be taken if the immersion heater fails. l i In addition, we understand that check valves and a buffer '
i system have heen added. Describe these changes.
., 10.
i The Final Safety />nalysis Report states that a delay of 60 seconds in achieving rated flow in the core spray loop would be acceptable ~ i Provide an analysis to support this statement. - I 11. Certain engineered safety feature pumps are located on the bottom l level of the reactor building. These pumps may be inundated by either en influx of Ground water or failure of a pump suction line. i Describe the measures that have been taken to insure that core
; cooling will not be lost due to flooding. . Discuss provisions for alternate sources Of core cooling.
I* 12. ( Provide design details of the core spray piping and sparger ring f located in the reector vessel. Describe the analysis performed, 4 l including load combination equations as well as the resulting stresses and deflections. The effect of seismic, ncrmal and blowdown loads, { alone and in combination, should be included. 4, l 13 Describe the design provisions to prevent actuation of the auto relier system in the unlikely event that alternating current power is not
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I available. Pre-programmed blowdown of the reactor pressure vessel should be prevented unless alternating current power is available to the emergency bucaes. 1 i 14. What reactor coolant leakage, directly in so the drywell, can be tolerated I' h'
- without exceeding a high drywell pressure of 1, 3 and 5 psig? Assute the plant is operating at rated power conditions. t
/ III. REACTOR C007. ANT SYSTEM i
1. Describe the time-sequence of events and the variations of parameters j pertinent to core cooli,ng which will occur as the various protection , j systems' respond to the loss of off-site power because of: { i I
- a. the system frequency increasing
- b. the system frequency decreasing, and
- c. the loss of all transmission lines.
( Equipment operation should be assumed that results in the minimum L loss of water from the reactor vessel. ..argins available for pre-venting actuation of the auto-relief ar; tem should be identified. ___m__.__._-._._..____- - - - - - - - - - - ~ ~ ~ ~ ~
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l l Consideration of primary system leakage should be included in the 6 analysis as well as acceptable variaticus in initial drywell pressure. I 2. Describe the time-sequence of events and the variations of" parameters pertinent to core cooling which vill eccur if the level control valve ) for an emergency condenser fails in the full open. position for the [ cases described in 1, above. How long is cooling available froc. this system with the level control valve failed open? How is such a failure detected? What instrumentation presents the inventory of water I available for this system to the operator? Describe how the condensate transfer pump is used to provide make up for this system, 3 For each syctem containing reactor coolant describe: a.
- b. the neemal expected leak rate and the " abnormal" lean rate,
' the leak detection system used ,
- c. the sensitivity, the time required to detect the "abno; mal"
. I leak rate, and the amount of reactor coolant lost during l this time because of the abnormal leak,
- d. the operators response to an-abnormal leak rate,
{ ( e. the procedure for isolating and repairing the leak. $ 4. Describe the time-sequence of events and the variations of parameters f - pertinent to core cooling which will occur if during the worst case of 1, above, the feedwater isolation valve fails full open. Describ'e 4 how such a failure could be diagnoced by the operato.r. j 5
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Describe the time-to-automatic-blowdown as a function of reactor coolant leaka6e to tt." dryvell assuming only on-site power is available. 1 1
- 6. Tabulate the quality control tests and inspections performed on
(, reactor coolant system ecxnponents end the installed system. Include ! type ct test, . manufacturing or' construction stage when test was performed, j agent that performed the test, basis of acceptance, .and agent that reviewed or evaluated the test results. 1: 7 List possible sources by which oil could contaminate the primary 3 j coolant assuming single failures; describe how oil in the primary a coolant is detected and the time interval to detect it. For the' - case where the maximum quantity of oil enters the coolant describe l the effects of oil contamination of the primary coolant on reactor l j operations and process such as core reactivity, fuel heat transfer characteristics, instrumentation. l (
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- 8. Describe how the pressure vessel material surveillance program will be used to evaluate the nil ductivity transition temperature of the pressure vessel. -
9 Specifically describe the areas of the Reactor Coolant System which f will be available for periodic inspection, the types of inspection methods to be used, and how the results of'such inspections will { be ana'lyzed and interpreted. t i 10. Provide a description of sequence of operation and curves of reactor pressure, coolant inventory and clad temperatures, versus time
' following the instrument air failure assumed in Appendix E-55-3 20.2 with and without' loss of off-site power. The time period should extend to shutdown and depressurization of the reactor. Indicate how long the emergency condensers will provide cooling and the actions required thereafter.
- 11. What is the hydrogen inventory in the primary coolant during normal
' operation? Spacial distributions should be indicated; i.e., concentration ; in the pressure vessel, hot well, rate of discharge out the stack, etc.
f ( 12. Provide an evaluation of plant transients initiated at or near the rod I
! block limit. Discuss the basis on which the normal operating line (power to flow) is established. '
n l* 13 We understand the core thermaland hydraulic characteristics for the Nine Mile Point are similar to those for the Oyster Creek plant. Therefore, the sensitivity analysis conducted for Oyster Creek plant would appear to
; be applicable to the Nine Mile Point Plant. This point snotld be clarified and the following information should be provided:
L i a. Bases for selecting the confidence levels used in the analysis.
- b. Justify use of "best fit" for the critical heat flux data rather than l
- i. - the " lower limit" curves used in APED 3892.
fp c. For each of the following parameters, provide the probable error and the basis for its determination and how such errors affect the minimum critical heat flux ratio. Inlet enthalpy Local Peaking Factor I Reactor Core Power Channel Flcw Radial. Peaking Factor Core Flow Integrated Peaking Factor Pressure
- d. Provide the peaking factors within a fuel pin used to obtain the local power distribution within a central and peripheral fuel assembly.
( e. Provide the method used in calculating standard deviation in the MCHFR using parameters listed in c, above.
1 + 1 ( . IV. INSTRUMENTATION AND CONTROL l
- 1. Identify which reactor protection and engineered safety feature actuation circuits are identical to those installed at Oyster Creek.
l 2. Evaluate the portions of these systems which are not identical to
; Oyster Creek for compliance with the Proposed IEEE Standards for
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Nuclear Power Plant Protection Systems, f 3. Identify, describe and justify any interconnection between reactor protection or engineered safety feature actuation circuits and control
; circuits.
- 4. Describe each type of channel which provides signals to the reactor protection or engineered safety features actuation circuits. This description should include:
- a. Type of channel (pressure, temperature, condenser vacuum, etc.)
- b. Number of channels monitoring parameter
( c. Sensor readout capability (blind or indicating)
- d. Sensor power supply (self or from external source) i
- e. Functien (reactor protection or engineered safety
; features) t l f. Failure modes and effects
- g. Instrument error (complete channel)
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- h. Locations where two or more sensors are on a j f, common line or penetration l l 1. Routing and protection of wiring from sensor to amplifier units to reactor protection or engineered
(( safety feature panel related to preventing a common mode failure I
, J. Power supply. f I l l 5. Identify components or equipment which must function during and following i the design basis accident (DBA). Specify the design ratings of the components and equipment pertinent to their operation while coping with the DBA.
( Describe the qualification or preoperational testing that has or will be accomplished to determine that equipment will function in the accident j environment.
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_I t' [ , . 7, 6. Describe.the isolation of the annunciators and/or data logging computer - interconnections.with the . protective system.
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l Describe how protective system and engineered' safety feature equipment I will be' physically identified as safety system equipment in'the plant.
;i 8.
Specify the safety function of the radiation monitors described in i, Section VIII of the FSAR. )
-level safety action will occur.For each monitor, specify at what preset I
- ) For each monitor, justify th,. sensi-tivity and preset level chosen. Justify the use of two-of-two or one-of-one. logic. Justify the lack of recording of any radioactive effluents <
9. Identify other instrumentation and control systems that-are identical to those installed at Oyster Creek. For example:
- a. Rod Worth hinimizer
- b. Control Rod Drive
- c. .' Recirculation Flow Control
- d. Pressure and Turbine Control
- e. Reactor Feedwater Control
- j. f. Rod Block v
f 10. i Describe the facilities provided to permit continuous occupancy of the control room following DBA. For example:
. It ! a. Shielding }_ b. Communication systems ; c. Fire prevention precautions j d. Radiation decontamination clothing i e. Lighting } f. Ventilation system l- 11.
Justify the manual system to isolate the control room ventilation system. 12. q Explain how the' control room radiation monitor provides adequate warning to personnel of radioactive contamination which enters the control room via the ventilation system. i
- 13. Describe existing ability to shut a the reactor and maintain it in a safe condition from station (s) exterior to the control room. Identify meters, ability.
valves, switches, etc., presently available which provide this , 14. l- Outline the test program for reactor protection and emergency safety ' feature systems. This outline should identify the qualification, field installation, preoperational start-up, and periodic in-service ! 1 I
( . tests which have been or will be performed. In addition, it should discuss any mandatory codes, " check and balance" procedures or inspections, and record requirements. V. ELECTRICAL SYSTEMS I
- 1. Describe any right-of-way which is common to both the two 345 kv transmission f lines..
{ 2. Perform a failure mode and effects analysis of the AC system. Include the effects of all failure modes of the 115 kv reserve bus and the 4160 volt board #101. l f-
- 3. For both the AC and DC power systems, describe the sequence of operation of the circuit breakers which connect the various power sources to the distribution syste2.
! 4. Describe the interlocks between the diesel-generator breakers and the i breakers to power board #101.
J 5. Describe the sources of DC power to the switchgear for equipment required p5 for the accident conditions. Show that the switchgear which connects ( equipment to one essential bus takes its DC power from one source while It the switchgear which connects equipment to the second essential bus takes its DC power from a separate tource, n 6, { Provide the basis for the protection circui*s which can shutdown the diesel-generators. l' ( 7. Provide a flow diagram for the diesel-generator fuel system. Discuss i Precautions to preserve quality of fuel (e.g., replacement time, additives, ll sampling, etc.). Discuss availability from off-site sources under post-
- l. accident and loss-of-power conditions.
- 8. Describe the location of the 48 volt batteries which provide power for the l nuclear instrumentation.
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- 9. Evaluate the relative advantages and disadvantages of starting engineered
~ [f . safety feature equipment in each loop simultaneously versus the proposed i
procedure to start this equipment sequentially.
- 10. Provide a current table of the magnitude and duty cycle of the diesel i
generator loads. Discuss the basis for both the magnitude and the duty cycle. L
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- 11. Provide a current table of the magnitude and duty cycle of the battery loads following a loss of off-site power. Include two cases; 1.e.,
(a) with only maximum technical specification leakages and (b) with ( DBA conditions. 1
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VI. ACCIDENT ANALYSES
- 1. What is the total frequency associated with the 0 - 10 mph wind roses?
- 2. Provide the equations used to calculate changes in diffusion due to
! wind speed and elevation. Det.cribe the method used to calculate the weighting factor due to this effect.
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l 3. List the sector data used to calculate the average chi (X) and dispersion factors.
$ 4. List the X/Q values used in calculating the doses for accidents and normal operation.
- 5. In order to evaluate how the activity in the Dresden I primary coolant can be correlated to the stack release rates, complete data on flow paths, flow rates, volumes etc. should be provided for the Dresden I ;
plant at the time the measurements were made (see Figure 9.1-1 in Vermont Yankee Amendment 3). Similar data is required for the Nine i Mile Point Plant. This information is required to understand the L claims for the noble gases and iodine transport factors and correlation : l to radioactive gasecus release rates. The effect of stopping primary i [ system cleanup for various times should be presented. j j l j 6. Figures E-31, 32 .2nd 34 show that the drywell and suppression chamber
, pressures are equalized before 65 seconds and that clad rupture will not occur untf3 about 100 seconds, i 4 ; In view of this, your model for fission product transport to the sup- ! pression pool, and the removal of iodine there does not appear to be ! applicable. Provide a radiological analysis of the accident which j more closely follows the behavior of other factors involved. This !
should include data on the fraction of leakage which will occur ! directly from the dryvell to the reactor building, and its dependence l {, on the performance of the containment spray system. l
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[ 7. The total delay in tripping the spent fuel storage pool radiation monitor to closing the damper in the ventilation system appears to be over 60 seconds. It would appear that in the analysis of a refueling accident, a substantial portion of the radioactive iodine mig,ht be
; released without benefit of flitration. Provide an analysis to show l how these effects have be a accommodated in the design. Provide a ? detailed schematic showing duct sizes, length of duct runs, flow rates ; and equipment operating times. l l
- 8. We understand that automatic isolation capability has been provided on the mechanical vacuum pumps. Provide design details, including the
( instrumentation arrangement. Further, we understand that a second radiation monitor has been added to *.he discharge canal and that wind , velocity indication is to be provided in the control room. Describe l these changes. _-___-____-______ __ 0
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9. Describe the interrelationship of exfiltration, building wake and atmospheric dfspersion. 10. khat is the shortest distance from the facility stack to the site boundary (two conflicting numbers are stated.in the application)? h 11. F. Provide the design details for the charcoal filters, including' dimensions, I? filter type and method of cooling. Describe the' assumptions that were 1 used to calculate the filter heating values.
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j Calculate the control room doses assuming TID release assumptions, 90% filter efficiency and no other lodine removal. Include calculations for the thyroid dose due-to intake of.cutside air. Have the primary con-4 tainment leak rates discussed in the FSAR been considered in evaluating control room doses? { VII. CONDUCT OF OPERATIONS
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g 1. 3 Provide a detailed outline of the emergency plan that describes: [ .a. Delegation of levels of authority and assignment
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(' - of responsibility necessary to meet the plan's objectives. 1 a' b. Actica levels used for initiation of emergency
,h proceduree.
't' I c. Instrumentation from which necessary data can be obtained to safely and efficiently implement the plan.
- d. Responsibilities and authorities of each group that would be used in plan implementation,
~ particularly the-capabilities expected by personnel j responding from local, state and federal agencies and the coordination of the efforts of all groups, y
- .g e. Action required by all personnel at the station, including visitors, during plant and site ' evacuations.
- f. Designation of evacuation routes. ,
- a. g. Capability of emergency communication systems.
e -3 L ;; h. Methods used to assure personnel accountability following plant evacuation. (, i.
' Methods used to detect and isolate contaminated areas.
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- j. Provisions made for medical treatment of serious l- injuries and contaminated patients.
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- k. M ilization of reserve personnel, i 1.
; Methods used to insure that _ each individual's responsibility and authorities are defined in a' simple fashion so that y;7 individual actions will not be burdened with complexity of requirements of the entire plan. \:
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'$' Clarify the responsibilities of the audit committee to review and approve
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acceptance tests, procedures;and plant modifications.
.j 3.
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hfine the division of responsibility snd authority between Niagara l, Mohawk operating staif during the preoperational and acceptance periods. Clarify the duties'of the Shift Control Operators and other members of a shift during these periods as well as during power operation. Identify
' ' positions held by' licensed personnel.
[> 4.
' Describe the qualifications required of the on-site engineering staff'
[ as for well replacement as members of these of the review.and audit bodies. Descriho the' criteria individuals.
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Provide a'detciled outline'of the proposed station operating procedures. ( This outline'should include the operating,~ai,ninierrative, and emergency
'[ procedures.
The scope and objective of each plant. operating procedure should.be given. Particular emphasis should be given to those procederes
'to be used in the event of an accident which may lead to an off-site
;; 4 hazard.
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