ML20127J163
| ML20127J163 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 04/10/1970 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20127J157 | List: |
| References | |
| NUDOCS 9211190217 | |
| Download: ML20127J163 (17) | |
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,i UNITED STATES OF A! ERICA ATOMIC ENERGY. COMMISSION In the Matter of Docket No. 50-263 NORTHERN STATES POWER COMPANY Monticello Nuclear Generating Plant, Unit 1 APPENDIX A DESCRIPTION AND EVALUATION OF PLANT FEATURES WHICH MAY NOT BE COMPLETE DURING INITIAL FUEL LOADING AND LOW POWER STARTUP TESTING l
l April 10, 1970 1
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9211190217 700412 DR ADOCK 05000263 PDR
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j The operating 11censo application and the ensuing six-teen amendments have been reviewed in detail by the DRL Staff of the Atomic Energy Commission.
The Advisory Committee on Reactor Safeguards (ACRS) also reviewed the application, as
' amended, and reported its findings to the Atomic Energy Com-4 mission by letter dated January lo, 1970.
Both the DRL Staff 4
and the ACRS concluded that the proposed reactor can be operated l
at the Monticello site at power levels up to 1670 MWt without undue risk to the health and safety of the public.
These eval-uations were based on a review of the complete plant as de-scribed in the FSAR.
This statement describes those items or portion of the plant which may not be complete when the p3 ant is ready for fuel loading.
It also presents an evaluation, taking into account the expected construction completion status at the time of initial fuel loading, of the safety of' conducting the initial fuel loading and that portion of the startup program which is to be performed at power levels not exceeding 5 MWt and without the reactor vessel head in place.
Such activities are referenced herein as " Initial Fuel Loading and Low Power Startup Testing" and would be conducted subject to the tech-nical specifications contained in Appendix A to the Regulatory Staff's proposed provisional operating license with the excep-tion of Technical Specification 3.7.C.
Those items or portion of the plant which may not be complete when the plant is ready for. fuel loadihg are:
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, 1.
Permanent Security Fence 1
l 2.
Off-site Power Sources 3.
Class II Piping Restraints 4
Installation of Instrumentation which provides testability of Icak detection system in the auto b1cddown valve bel-S lows and modification of the low pres-sure core cooling pumps discharge pres-sure interlock 5
Inerting System 6.
Liquid Radwaste System
-l 1.
Permanent Security Fence Access to the immediate plant area is controlled by a security fence.
The permanent security fence may not be com-plete at the time of fuel loading for the following reasons:
l The final grading around the plant will not be complete; elec-trical yard work needed to operate automatic gates and tele-phone communications associated with the permanent fence should not be done until the grading is complete to preclude possible I
damage to the electrical cable; and the ground should be thawed
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to preclude damage to fresh concrete around the fence posts.
The present construction fence is not of the same quality as the permanent security fence, i.e.,
the construction fence is a lighter gage and not as high.
However, it consists of t
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3-four-foot-high woven wire topped with three strandu of barbed wire to give a six-foot-high fence.
This fence is lighted and patrolled by guards.
The construction fence is outside of the area the permanent fence will enclose.
This makes it possible to build the permanent fence without violating the construction fence., The construction fence incorporates two structures through which all people must pass on their way into and out of the plant construction site.
All construction craft people and visitors are accounted for when they pass through either of these structures.
Unauthorized people will not be permitted to gain entrance to the site.
The construction fence is an i
adequate substitute for the permanent security fence during initial fuel loading and low power startup testing.
2.
Off-Site Power Sources The FSAR deccribes five transmission lines that will cupply power to the plant substation.
Three of these five trancmiscion lines will be energized to. serve the substation before commencement of fuel loading.
Following are the dates on which we anticipate the completion of the balance of the transmission lines which will serve as outlet facilities for the Monticello Plant.
1.
Monticello-Lake Pulaski - Scheduled f
completion date May 15, 1970 115 KV i
Line 2.
Monticello-Coon Creek - Scheduled
. completion date July 1, 1970 345 KV Line
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Any of the three transmission lines which will be ener6ized meet the requirement's of the Monticello Technical Specifications Section 3.9 These lines provide adequate redundancy for initial fuel loading and low power startup 1
tectJnd-3 Class II Pfping Restraints 4
Prior to intital fuel loading al.1 emergency core cooling systems (high pressure coolant-injection system, residual heat removal system and core spray system) and t'he standby liquid control system-piping and other Class I designated piping shall be complete.
R'estraints on the Class II piping systems may not be completed prior to fuel loading.
Tne Class II piping systems will be isolated from the primary system until the restraints are installed and adjusted.
The isolated Class II piping systems are not uced in the course of initial fuel loading and low power startup testing,
i 4.
Installation of Instrumentation which provides testability of leak detection system in the auto blowdown valve bellows and todification sr the low pressure cors cooling pumps i
i discharge pressure interlock a
, Both of these design changes affect the auto prescure relief system which provides a protective funttion only when the i
primary system is pressurized, i.e., af ter the reactor i
vessel head is put in place.
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Jnretbr_pyinem Inertinn to only required after completion of the entire startup test program and demonstration of plant electrical out-l put an cpecified in Section 3 7. A.S.a of the Technical Specifica-l lo concluded that the inerting system need not be f
tione.
It operable during intital fuel loading and low power startup testing.
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6.
Secondary Containment During initial fuel loading and low power startup test-ing, no radioactive gas will be released through the air-ejector and gland seal off-gas cystema since the main condenser will not The only potential courceu for Generation of be in operation.
radioactive caces are from minuto quantities of " tramp" uranium of the fuel curfaces and possible leakage of minute quantities of fission products produced in the reactor fuel rods during While ficsion products may be theoretically released this pahse.
from " tramp" uranium and fuel rod leakage, experience during fuel loading of other power reactors indicates-that such releases are below the threshold of detect ion relative to natural background, and are insignificatn.
Although the secondary containment and standby gas treat -
ment system will be essentially complete, the ' secondary contain-l Trior to initial fuel ment Icak rate tout may not be complete
/
loading.
While the standby gas treatment system would mi,tigate the consequcnces of_ an accidental radioactivity release, for purpocos of this analysis and until completion of_ the cecondary containment leak rate test, the extremely conservative ascrmption in made that neither the cecondary containment nor the staciby
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The worut pousible incident that coubt occur during inttini fuel loading and low power utartup tenting in the cont. col rod drop ncoident discusaed bcjow.
The offcci, of 01.her pontulated accidents cuch no the refueling acci-dent or loss of coolant accidents are insignificant because the fuel will not have been substantially irradiated.
Control Rod Drop Accident Analysis During initial fuel loading and low power startup test-ing, no substantially irradiated fuel will be present.
There-fore, there will be no significant fission product inventory in the ' fuel rodo prior to the transient which would be caused by a postulated control rod drop accident.
Only those fission producto resulting from the postulated transient-in those fuel rodo which are accumed to be damaged could possibly be released to the environc.
For initial fuel loading and low power startup testing the reactor vencel head, primary containment. head and shielding j
blocks would not be installed.
The secondary containment is assumed for this analysis to be ineffective thereby permitting releanc of the nanumed fission products directly to the' environs I
at the elevation of the refueling floor (appr ximately 30 meters l
above grade).
Meteorological conditions.are conservatively as-cumed to be those which would'cause downwash of the released l
flosion producta At the time of the accident.,
During cold critical tests it is not planned to move any control rod which -is worth more than 1.$ k, lione ver, for i
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7-the purpose of this analynia the accumption is made that a 1
l 2 5% k control rod falls from the core at a velocity of 5 I
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ft/sec, which is greater than 10 possible because of the control l
rod velocity limiters.
The postulated nuclear excursion would 1
J result int (a) a total energy release of 2500 MW-sec; (b) 200 fuel rodo having /uel enthalpien hi6h enough to postulate clad-1 ding perforation, e.g., greater than 170 cal /gm; and -(c) fuel melting in act of the 200 roda (those with enthalpies greater than 220 cal /cm having combined energy release of 83 MW-sec and encompaccing 50 lbo of the UO )'
2 The conservative assumption is made that _100% _of the nobic gauco and $070 of the halogens contained in that volume of fuc1 above 220 ca3/gm and which are produced as a conce-I quence of the. rod drop are released -to the reactor vessel.
In addition, it is also assumed that 2To of the noble gases and I
1% of the halogena in the rods which are perforated are also i
l relenned to the venetor vencel.
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A.Ithough the 2500-MW-cec energy release would cause an
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overall average coo. tant tcmperature rise of less-than-4 F, it l
1s contcevatively accumed that the localized-turbulence cauced by the trancient would permit the transfer of all fission pro-ducts (released-from-the fuel) to the reactor water, vessel, and cavity where the halogens experience a combined plateout-l washout-fallout factor of 5_ prior to release to__the ventilation i
air at the refueling floor level.
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6-l Meteorologieni conditionc are acuumed to be wjnd velve.1ty a
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of 1 meter /sec since this is the most conaervative casumption.
liigher wind speeds would cause greater building exfiltration
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rateu; however, the higher wind speedt also recult in better j
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f atmospheric dilution effects, The radiological effects of the above analyoic results i
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-5 in a cloud gamma done of h.1X10 rem and a thyroid inhalation i
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done of 1.6X10 rem at the nearest site boundary.
i Calculations were also perfortred to investigate the
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i radiological consequencec of the release of 100% of the noble i
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gaces and 50% of the halogens from those 200 rods which are accumed to experience cladding perforation and fuel melting.
Daced upon these concervative release assumptions, as well as the previouc assumptions regarding halogen plateout, downwash, l
inctantaneous fission product release to the environment, 1 m/sec l
l winde, and adverec meteorological conditions, the maximum off-
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--5 cite exposures are 6.7X10 rem cloud gamma and 2.4X10 rem d
j thyroid doce.
I As shown, even uith the most conservative set of assump-l tienc, the off-site expoeures - are still many ordere of magnitude l
below the guidelines. set forth in 10CFR100 and would not require Therefore, the any special measures to protect the public.
secondary containment and standby gas treatment system is not
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required during initial fuel loading and low power startup i-r 1
_tcuting.
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7 Liquid Radwnr.te System t
Pending completion of the Atomic Safety and Licencing Board's review of the intervenors' contentions, the applicant has committed, in Paragraph 10 of the motion to which this Appendix A 10 attached, that during initial fuel loading and low power startup testing the liquid waste generated will not be relenced to the env.trons.
The liquid radwaste system will be operable; howevec, it is possible'that the pre-operational tests of this system wii?
not be complete prior to initial fuel loading and low power startup testing.
All draina will be initially flushed using a fire hose or other suitable supply of water and the tanks cicaned.
During low power startup testing with the reactor vessel head removed, reactor power is not expected to exceed 0.1 MWt.
Essentially zero fission products and induced radioactivity will be produced.
Provisions will be made to store the liquid waste Generated during the low power startup testing in the condensate storage tanks and in other plant storage facilities, (e.g. torus, spent fuel pool, waate uurge tank), which will have been pre-viously leak tested.
As stated above, initial fuel loading and low power s
e startup testing will be performed without any discharge of liquid raduaste to the environs.
This will require certain
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temporary changec-to the piping systems and procedural controls.
There are no changes to the system that would prevent the per-formance of the preoperational test except for the transfer of w;
water from the turbine building and.the discharge of water to
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the canal.
These temporary chan',es will be returned to normal and the preoperational test completed prior to operation with l
the reactor vessel head in place.
t The piping changes are intended to isolate the turbine 1
~ building and hot machine shop from all systems having radio-active waste and to provide a means for reclaining water from the reactor primary syetem.
Water that cannot be efficiently l
reclaimed will be stored in available tangs or it may be mixed f
with an abuorbent material and removed as solid waste.
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The procedural changes will assure that increased storage and processing capacity would be available.
In order to accomp-lish this, certain restrictions will be placed on the operation of the various plant systems, i
1 a.
TEl1PORARY PIPIlla CH ANGES 1.
The turbine building sump systems will be temporarily l
routed to the temporary retention pond and a blank l
i flange or other positive isolation provided to isolate l
the turbine building and the hot machine shop from the systems which could have radioactive waste.
2.
Two temporary-filter units with replaceable cartridge
filters shall be installed one each in parallel to the l
existing waste collector and floor drain filters. - -The
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flow from both of these filters shall be directed.to j
j the radwaste domineralizer.
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A temporary line will be run from the chemical waste l
pump discharge to the floor drain sample tank pump i
j recirculation line for holdup of water which cannot I
be eff1ciently reclaimed.
b.
E MPORARY PROCEDURAL CHANGES l
1.
All radiochemistry shall be performed in the hot lab.
l All liquido not used by the analyses will be returned I
to the reactor building or radwaste building.
Liquids 4
used in the analyses shall be routed to the chemical j
waste tank.
Only those end product samples that require
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counting shall be removed to the counting room.
j 2.
One condensate storage tank shall be used to maintain i
normal level and makeup and the other tank kept empty I-for use ao a holding volume.
Both tanks will be isolated i
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from all non-radioactive systems except the makeup system.
l 3.
The torus shall be available for use as a surge volume.
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I Any excess water in the fuel pool can be drained to the t
torus if the radwaste filter demineralizers are unavail-able or if the storage tanks are filled, i
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4.
There will~ae no laundering of contaminated clothing on o
1 site.
5.
The radwaste demineralizer shall be used to process water j
to one.of the two waste sample tanks.
When the tank-is full it.will be sampled and then discharged to a condensate storage tank.
Water that does not meet condensato storage l
tank quality may be reprocessed back through the ayetem.
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Pocitive meana (two isolation valven in scrica with a j
drain between them) will be provided for prevention i
of any water being sent to the discharge canal.
l 7.
The operation of the condensate domineralizero will not I
be permitted.
The system will be isolated from the i
4 condensato service nystem.
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8.
Steam ] inca and feed lines will be isolated and checked l
for leakage periodically to prevent radioactivity in the i
primary nyLtem from entering tne turbine building.
3 t
CONTHOL _OF C011STRUCTIOH PERSONilEL DURING IMITIAL F_UEL
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LOADIHO AND LOW POWER STARTUP TESTIllo -
Prior to fuel loading, atringent controls will be en-
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forced on construction peruonnel for their work activities i
in tho reactor building, the raduaste buildings, the cable aprending room and the diocel engine room.
Normal construc-tion activitica can be continued in the turbino building and i
outsido arcan vince locked doore or barricades will provent unauthorized entrance to either the reactor building or the radMatte building.
4 1.
Acceso to the Reactor Building i
j Double interlocked doors with weather strip type sea 3a
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I are provided for accean to the reactor building for both per-
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connel and equipment.
All doors except airlock #1214 are f
normally locked..Outside.entrancea to the raduaate building
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are locked to prevent entrance to'the reactor building.
Ex-i cept for the normal controlled e'ntrance to the reactor _ building N'
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through the acceau control area, any use of entranceo into the j
t reactor building muut be approved by the GE and NSP chift super-1 1
l visor and be in accordance with established radiation protection i
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policies.
The shift supervisor will control the keyo to the l
locked entrances; however, if evacuation conditions occur in 1
the reactor building, personnel can 1 cave through the double J
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interlocked doort.
Potential high radiation areas within the 4
reactor building will have ceparate locked entrances ao no q
c direct access to these arcac will be possible without permicsion l
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i of the GE and 110P chif t cupervisoro.
t Normal acceau to the reactor building will be through 1
l-airlock #124 (Doora 62 and 63) located at elevation 935'0" 4
of the reactor building.
This airlock.is located at the top I
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of the stairway leading from the acceas control area of the l
odininistration building to the reactor building.
Whenever 1
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a pornon enters the reactor building, he must sign the access 1
l control log provided by the radiation protection group.
This log 10 availabic in the clean area of the acceca control area, i
l An information board is also provided in the clean area of the i
l acccus control area to keep personnel advised of the defined i
potential radiation areas in the reactor building and the All l
l.
protective clothing requirements to work in thene areno.
work in potential radiation areas in the reactor buiN im,-rwnt l
c be in accordance with approved radiation work permits.
All construction. personnel will_ be under the guidance of a Bechtel supervisor or engineer and, if necessary,-qualified l
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a.
Jk+fuelinre Floor.
l Since the refueling floor will be a restricted aron l
after the start of fuel inspection activities, restric-tions for access to the refueling floor will be maintained.
Construction personnel will be required to be accompanied to that location by OE or NSP personnel whenever they have any work to be performed in that area, All personnel will be controlled as required by t
NSP's radiation work permit.
Only necessary personnel will be allowed on the refueling floor during criticality testa.
b.
Drywell Accens to the drywell will be allowed by authorized work permits signed by the GE and NSP supervisor and an NSP radiation protection man.
Normal access will be through the airlock.
Construction personnel must be accompanied by a GE, NSP, or Bechtel supervisor or engineer.
The drywell shall be cleared of all personnel when-ever control rods are-being witndrawn for criticality or to change povier level.
Personnel may be permitted inside the drywell whenever the' reactor is critical but at a constant power level.
Personnel must notify the shift supervisor.on duty during entrance to-and departure from the drywell.
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- 1 As on the refueling floor, all personnel will be 1
I controlled no required by NSP's radiation work permit.
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2.
Control Voom f
The NSP and GE personnel will restrict the number of j
personnel within the control room to those required for the fuel loading of the reactor and the startup test program.
Any construction activities will be authorized by the NSP t
and GE shif,'t supervisor only after a cicar mutual under-ctanding of the work to be performed has been obtained.
Work i
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can only begin after issuance of an approved work permit.
3.
Cable spreading Room Prior to fuel loading, the cable spreading room doors will be kept locked and access will be gained only by a work 4
permit signed by the OE and NSP supervisor.
Any construction personnel must be accompanied by a responsible Bechtel, GE,
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or NSP supervisor or engineer.
I 11.
Diesel Encine Room This area will be locked at the start of the fuel loading i
final readiness checks.
Construction personnel entrance will
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be authorized by a permit sjsned by*the OE and NSP shift super-visors.
S.
Radwaete Building A two door airlock arrangement is provided between the radwaste building and the reactor building.
Although keys may i
be provided to lock one or all of these doors, normally.these l
doors will not be key controlled because of the routine use 1
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of thiu pnunoce; however, the door interlock arrangement must t
1' be maintained at all times.
Outside entrances to the radwaste building are locked to prevent entrance to the reactor building through the radwnste building.
Defined potential high radiation areas within the radwaste building will have separate locked i
j entrances so no direct access to these areas will be possible even though the main airlock entrance is not key controlled, The requirements for access to the radwaste area are similar to those for the drywell and refueling area, 4
CONCLUSION It is concluded'that the initial fuel loading and low
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4 i
power startup testing program can be safely conducted in the absence of completion of the plant as described herein and taking into account the access control procedures described 9
herein.
This Appendix A has been reviewed with the Monticello 4
Safety Audit Committee at its regu3ar meeting on April 1, 1970.
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