ML20064L005
| ML20064L005 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Brunswick |
| Issue date: | 11/27/1981 |
| From: | Kreger W Office of Nuclear Reactor Regulation |
| To: | Novak T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20064E577 | List: |
| References | |
| FOIA-82-389 NUDOCS 8112300287 | |
| Download: ML20064L005 (13) | |
Text
DISTRIBUTION:
NOV 2 71981 Docket File 50-325 Docket File 50-324 ETSB Reading File 44TSB Docket Files WEKreger Docket 11os. 50-325/324 CSPaul 11EMORNIDU'1 FOR: Thomas !!. !!ovak, Assistant Director for Operating Reactors, DL FRO:1:
William E. Kreger, Assistant Director for Radiation Protection, DSI SUDJECT:
BRUl!SUICK, IllllT !!05.1 AllD 2, SPEllT FilEL POOL EXPAllSION (TAC #43797)
In accordance with TAC 643797, the Ef fluent Treatment Systens Branch (ETSB) has completed the review and evaluation of the April 16, 1980 letter from the licensee, Carolina Power and Light Conpany (CPAL) which included a document entitled " Brunswick Stean Electric Plant, Unit Kos. I and 2, Spent Fuel Storage Expansion Report" and which provided infomation on the pro-posed expansion of the storage capacity of the spent fuel pool (SFP) for an,ending technical specification 5.6 of DPL-71 and DPL-62. The review was perforced by J. S. Boegli, ETSB (Ext. 27634).
The present licenses for Drunswick, Unit Nos.1 and 2, pemit a spent fuel storage capacity of 1386 BWR fuel assenblies in each SFP.
In addition, the technical specifications for each license pennit space to store PUR fuel assenblies from CP&L's H. B. Robinson Plant. There are presently 304 PWR fuel assemblies at the Brunswick Plant and no additional space is expected.
This nodification proposes to increase the licensed storage capacity of the SFP to 1803 BWR and 160 PUR fuel assemblies at Unit 1, and 1839 BUR ar.d 144 PWR fuel assemblies at Unit 2. is suitable for inclusion in the Safety Evaluation. Enclosure 2 is suitable for inclusion in the Environmental Impact Appraisal.
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. arc:;.:r William E. Kreger, Assistant Director for Radiation Protection Division of Systems Integration
Enclosures:
1.
Safety Evaluation Input
/
2.
Envir'onmental Inpact_l Appraisal Inpu( /
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3.
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g Thomas M. Novak liDV 2 71981 cc:
R. Mattson R. Capra W. Gammill F. Congel J. Van Vliet T. Ippolito R. Bangart C. Willis J. Boegli
SAFETY EVALUATION INPUT FR0t4 THE EFFLUENT TREA1 MENT SYSTEMS BRANCH IN THE MATTER OF THE BRUNSWICK, UNIT NOS.1 AND 2 SPENT FUEL POOL EXPANSION APPLICATION 3.5.1.3 Radioactive Waste Treatment The plant contains waste treatment systems designed to collect and process the gaseous, liquid, and solid wastes that might contain radioactive material. The waste treatment systems were evaluated in the Safety Evaluation, dated November 1973.
There will be no change in the waste treatment system or in the conc,lusions given in Sections 9.0 and 11.0 of the evaluation of these systems because of the proposed modification. Our evaluation of the SFP cleanup system, in light of the proposed modification, has concluded that any resultant additional burden on the system is minimal and therefore the existing SFP cleanup system is adequate for the proposed modification and will keep the concentrations of radio-activity in the pool water within acceptably low levels.
3.5.2 Conclusions Our evaluation of the radiological considerations supports the con-clusion that the proposed modification to the spent fuel pool at Brunswick, Unit Nos. I and 2, is acceptable because:
(1)
The conclusions of the evaluation of the waste treatment systems, as found in the Brunswick, Unit Nos. I and 2, Safety Evaluation Report (November 1973), are unchanged by the modification of the SFP.
(2) The existing SFP cleanup system is adequate for the proposed modification.
L
ENVIROMiENTAL IMPACT APPRAISAL INPUT FROM THE EFFLUENT TREATMENT SYSTEMS BRANCH IN THE MATTER OF THE BRUNSWICK, UNIT NOS.1 AND 2 SPENT FUEL POOL EXPANSION APPLICATION 1.3 Padioactive Wastes The plant contains waste treatment systems designed to collect and process the gaseous, liquid and solid waste that might contain radio-active material.
The waste treatment systems are evaluated in the Final Enviionmental Statement (FES) dated January 1974.
There will be no change in the waste treatment systems described in Section III.D.2 of the FES because of the proposed modification.
1.4 Spent Fuel Pool Cleanup System The SFP cleanup system is part of the pool cooling system.
It consists of a demineralizer with inlet and outlet filters, and the required piping, valves, and instrumentation. There is also a separate skimmer d
system to remove surface dust and debris from the SFP.
i This cleanup system is similar to such systems at other nuclear plants which main-l tain concentrations of radioactivity in the pool water at acceptably f
low levels.
We expect only a small increase in radioactivity released to the pool water as a result of the p'roposed modification, as discussed in Section 2.2.1, and we therefore conclude the spent fuel pool cleanup system is adequate for the proposed modification and will keep the con-centrations of radioactivity in the pool water to acceptably low levels.
I 2.2.2 Radioactive Material Released to the Atmosphere With respect to releases of gaseous materials to the abnosphere, the onlj radioactive gas of significance which could be attributable to storing additional fuel assemblics for a longer period of time would be the noble gas radionuclide Krypton-85 (Kr-85).
As discussed pre-viously, experience has demonstrated that af ter spent fuel has decayed 4 to 6 months, there is no longer a significant release of fission products', including Kr-85, from stored fuel containing cladding defects. One hundred forty (140) fuel assemblies are expected to be stored following each March refueling at Unit 1 and each November re-fueling at Unit 2.
Since space must be reserved to accanmodate a complete reactor core unloading operation (nominally 560 fuel assem-blies), and module spaces are reserved for PWR fuel assemblies, the useful pool capacity is 1243 fuel assemblies at Unit I and 1279 fuel assemblies at Unit 2.
At an input of 140 fuel assemblies per year, the storage capacity is approximately 9 years at each unit.
For the simplest cas'e, we assumed that all of the Kr-85 that is going to leak from defected fuel is going to do so in the 12 month interval between refuelings.
In other words, all of the Kr-85 available for release is assumed to come out of the fuel before the next batch of fuel enters the pool. Our calculations show that the expected release 74 s',1 of Kr-85 from a 140 fuel assembly refueling is approximately JHf Ci each.12 months.
As far as potential dose to offsite populations is
. 2.2.2 concerned, this is actually the worst case, since each refueling would generate a new batch of Kr-85 to be released.
As more and more fuel is added to the pool, one might think that this would increase the releases, but according to the terms of our model, this is not the case since all of the Kr-85 available for release has aircady lef t the de-fected fuel previously stored in the pool before the next batch enters, with the result that the annual releases are not cumulative but remain approximately the same.
In other words, the enlarged capacity of the pool has no effect on the total amount of Kr-85 released to the atmos-phere each year. Thus, we conclude that the proposed modifications will not have any significant impact on exposures offsite.
Assuming that the spent fuel will be stored onsite for several years, Iodine-131 releases from spent fuel assemblies to the SEP water will not be significantly increased because of the expansion of the fuel storage capacity since the Iodine-131 inventory in the fuel will decay to negligible levels between refuelings for each unit.
Storing additional spent fuel assemblies is not expected to increase j
the bulk water temperature during normal refuelings above 1S* 150 F used in the design analysis. Therefore, it is not expected n3s there will be any significant change in the annual release of tritium l
or iodine as a result of the proposed modifications from that pre-l viously evaluated in the FES.
Most airborne releases of tritium i
l
.- 2.2.2 and iodine result from evaporation of reactor coolant, which contains tritium and iodine in higher concentrations than the spent fuel pool.
Therefore, even if there were a higher evaporation rate from the spent fuel pool, the increase in tritium and iodine released from the plant as a result of the increased stored spent fuel would be small compared to the amount normally released from the plant and that which was previously evaluated in the FES.
If it is desired to reduce levels of radiofodine, the air can be diverted to charcoal filters for the re-moval of radiciodine before release to the environment.
In addition, the station radiological effluent Yechnical Specifications which are not being changed by this action, limit the total releases of gaseous activi ty.
2.2.3 Solid Radioactive Wastes The concentration of radionuclides in the pool water is controlled by the filters and the demineralizer and by decay of short-lived isotopes.
The activity is highest during refueling operations when reactor coolant water is introduced into the pool, and decreases as the pool water is processed through the filters and demineralizer. The increase of radioactivity, if any, due to the proposed modification, should be minor because of the capability of the cleanup system to continuously remove radioactivity in the SFP water to acceptable levels.
,_ 2.2.3 The licensee does not expect any significant increase in the amount of solid waste generated from the spent fuel pool cicanup systems due to the proposed modification. While we agree with the licensee's conclusion, as a conservative estimate we have assumed that the amount of solid radwaste may be increased by an additional two resin beds a year due to the increased operation of the spent fuel pool cleanup system. The annual average volume, per unit, of solid wastes shipped from the Brunswick Plant during 1978 through 1980 was 15,000 cubic feet.
If the storage of additional spent fuel does increase the amount of solid waste from the SFP cleanup systems by about 160 cubic feet per unit per year, the increase in total waste volume shipped would be approximately 1% and would not have any significant additional environmental impact.
The present spent fuel racks to be removed from the SFP because of the proposed modification are contaminated and will be disposed of as low level solid waste.
We have estimated that approximately 7000 cubic feet of solid radwaste will be removed from the plant because of the proposed modification. Averaged over the lifetime of the plant this would increase the total waste volume shipped from the facility by less than 3%. This will not have any significant additional environmental impact.
o - 2.2.4 Radioactive Material Relesed to Receiving Waters There should not be a significant increase in the liquid release of radionuclides from the plant as a result of the proposed modification.
Since the SFP cooling and cicanup system operates as a closed system, only water originating from cleanup of SFP floors and resin sluice w'ater need be considered as potential sources of radioactivity.
It is expected that neither the quantity nor activity of the floor cleanup water will change as a result of this modification. The SFP demineralizer resin removes soluble radioactive material from the SFP water. These resins are periodically sluiced with water to the spent resin storage tank. The amount of radioactivity on the SFP domineralizer resin may increase slightly due to the additional spent fuel in the pool, but the soluble radioactive material should be retained on the resins.
If any radioactive material is transferred from the spent resin to the sluice water, it will be removed by the liquid radwaste system for processing.
After processing in the liquid radwaste system, the amount of radioactivity released to the environment as a result of the proposed modification would be negligible.
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SFP Modification S
Estimate Release Rate of Kr-85 Data BrunswieK, Unit-I n>cKet too. 50-326 ORK-78 MSG t1tAH Core =-
560 fuel assemblies,
Single Refueling =
/VO core assemblies / 9 ear Cladding =
Z.dc.afg-Y Burnup =
3C,000 HlOd/HT in Core =
/Co NT Weight of UO2
-8 Escape rate Coeff. of Kr-85 = 6.5 x 10 sec Fission Yield of Kr 0.0034 13 26 - S 2.
3.9
' years Present Capacity =
.g Future Capacity =
- 0 f' F.9
~ years 5
9 Failed fuel Fraction (tiUREG-0017) =
.0 012.
Half-life (Kr-85) = 10.7 years Eff. Full Power Days = (
ll6F days)
(
3, 2.
years)
(.80 availability
)
Amt Kr-85 in fuel 5.
Production
'A
- A decay leakage l
atoms /f f/MWsec MWD /Mt 16 0.0034 x 3.12 x 10 x 3f,000 Production =
fl6F days t
di 3.l E X io atoms /Mtsec
=
-8 (A
= 2.05 x 1@/sec, Aleak = 6.5 x 10 /sec) decay Amt Kr-85 in fuel 5 4.74 X/0*
atoms /Mt l
l 26Yf Curies /Mt l
This model assumes that all Kr-85 in the failed fuel assemblies will be released before the spent fuel is removed from the pool. The Kr-85 c.
s release rate is assumed constant with time. The additional capacity allows spent fuel to remain in the pool up to F.9 years.
fleglecting decay and assuming all spent fuel has failed, Kr-85 release rate is IIII 2
=
Mt g, q Kr-85(h)=
29'l Ci/yr/Mt The failed fuel fraction is 0.0 01 %
cladding.
The weight of a single fuel assembly in U0g is 0,/76 Mt. The number of fuel assemblies stored each year is 14o The 'ddit.icnel capacity of the pool because:of the expansion is 12.43-assemblies. The e+ftt%nal Kr-85 release rate with the pool fulTand no decay is:
R(Ci/yr)(Kr-85) =
297 Ci/yr/Mt x 0.178 Mt/assemb x l 2A3 assemb x.coli R=
'18,9 Ci/yr This release rate is conservative because:
1.
radioactive decay was neglected; 2.
release rate of Kr-85 from failed fuel should be exponentially decreasing - the release rate is dependent on the amount of Kr-85 within the fuel; 3.
release rate of Kr-85 should decrease as the spent fuel cools; and 4.
this release rate assumes the pool is always full.
Since 140 assemblies will be added each refuelling and assuming one refuelling each year, the increment in the Kr-85 release rate each year, until the pool is full, is:
78.9 Ci/yr x N 0/l?.43 79 Ci/yr/ refuelling
=
If the decay of Kr-85 is accounted for at the end of each year af ter refuelling, the release rate when the pool is full is:
In2
~
~ "}lE7YF b 2.7 Ci/yr g,g
=
n=1 per refuelling
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s SFP flodification Estimate Release Rate of Kr-85 Data Brunswick, Unit 2 DoeXc-t tJo. 50 - 32.4 DPR-4Z Core = _
[60 Single Refueling =_
fuel assemblies,
/@
_ core assemblies Cladding =
25c.afn - Q V
Burnup = _
3Ggog
/1p)d/g7 Height of U0 in Core = _
/0o MT 2
Escape rate Coeff. of Kr-85 = 6.5 x 10-8 see fission Yield of Kr-85 :P 0.0034 Present Capacity =_ i3%.(4o~ *
- fsf, Ii j --
years future Capacity = - /g31-540 9.)
~
/#
Failed Fuel fraction (tJUREG-0017) =
years Half-life (Kr-85) = 10.7 years
.0012.
Eff. Full Power Days = (__
ll68
(
3
~ days (3CQ0,g) = //6SA years
(.80 ava ability
)
f4 h.8)
/Eit Kr-85 in fue'l 1-Production 3.2 p Adecay
- A leakage atoms /f f/t4Wsec T4WD/ lit Production = 0.0034 x 3.12 x 10 x
/j6g days 3.18 X/O
=
atoms /titsec (A
9 4.7V(IO XM[3%)
decay = 2.05 x 10 /sec, A
-8 leak = 6.5 x 10 /sec)
~
5_
//,Jg Wo**
6.02 (lo")
Ant Kr-85 in fuel a toms / lit M46
_ Curies /t4t released before the spent fuel is removed fro uel assemblics will be e pool.
The Kr-85
' ~.
s release rate is assumed constant with time. The additional capacity allows spent fuel to remain in the pool up to _9./ years.
tieglecting decay and assuming all spent fuel has failed, Kr-85 release rate is y,7 g5 Ci/p,
2W Mt 9,l Kr-85 ( E)=
2.9 )
Ci/yr/Mt The failed fuel fraction is _
d,00/2.
cladding.
The weight of a single fuel assembly in UO2 is o./7; Mt.
The number of fuel assemblies stored each year is
- 4o The additier.al capacity of the pool because:of the expansion is 1276 assemblies.
The additiomil Kr-85 release rate with the pool full and no decay is:
R(Ci/yr)(V,r-85) =
24)
Ci/yr/Mt x 4.178 Mt/assemb x 1174 assemb x.0012.
R=
'77.6 Ci/yr This release rate is conservative because:
1.
radioactive decay was neglected; l
2.
release rate of Kr-85 from failed fuel should be exponentially decreasing - the release rate is dependent on the amount of Kr-85 within the fuel; 3.
release rate of Kr-85 should decrease as the spent fuel cools; and 4.
this release rate assumes the pool is always full.
Since_
140 assemblies will be added each refuelling and assuming one refuelling each year, the increment in the Kr-85 release rate each year, until the pool is full, is:
W.fi Ci/yr x l@/IT74 F. 7 Ci/yr/ refuelling
=
If the decay of Kr-85 is accounted for at the end of each year'after refuelling, the release rate when the pool is full is:
In2 5'-
Gl.3 Ci/yr g,"l
=
n=1 per refuelling
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PRELIMINARY NOTIFICATION OF EVENT OR UNUSUAL OCCURRENCE PN I 2 l This preliminary notification constitutes EARLY notice of ev...Liin 3
entjElf POSSIBIE safety or public intere'st signifirance.
The information is ~as iiif4failyW cation or evaluation > and is basically all that is known by IEhe MWtE60t verifi--
f on trifs date.
- diiin if rolina Epwer.tnd 1.ight Company.... Licensee _Emergencg Clas.k. "sMicat FACILITb __Ca H. B. Robinson
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X Notifica ngof Un, usual Event
- --Do:ket ho,- 50-241 5
Hutsville, South. Carolina Site Area Si gency ' -
General Mrdency p
Not ApplHnMe,, ' [ ~,
.se
SUBJECT:
UNUSUAL EVENT AT H. B. ROBIN 50ft UNIT 2
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'An unusual event was declared on November 30 at H. 4. Ra'binson) l*[Nn~M 1:44 p.in. (EsT), with the plant in hot shutdown for reactor cooTritt system integrity tests prior t6 startup, a leaking valve gasket in the reactor cMa t pump. (RCP) seat in.iection line, incated in the charging purnp room, reso. ted.in W spilli 1
gallons of prhaary coola,nt water tu the auxiliary, build,i_ng,f1pogfi,j,,,,,,,.ng.of.t.500 j
- n. sit it Precautionary!eyacuation of the auxillary building was condecteQ environmental ! release, or significant personnel contar.in,ation ossgr$.o e
e-d.
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Tennination of itCP seal injection flow required the shutddwn o'tiir7 o eagtoicoolant pumps. This z'erminated pressurizer spray resulting in 'a 'rbac c
pressure incrdase.
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Blocit volves for the unit's two power operated relief valves (POW'iJ 'weFe 'dfened to allow PORY acduation for RCS pressure relief.
s However, leakige% ghlth-10RV seats The. block valves failed to Miy@_c19Jie.Jahun caused a decrdase in RCS pressure.
' actuated frun ithe control room, resulting in automatic safety inJQfton.d.nd.startup of
- * * * ] " ".' "..]m: f emergendy1!!csels dud t610w RCS pressure."
- ~~
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- ws....
The NRC reside,nt inspector was present in the control iobm,"and,'tWe gliiii I.I ' Incident Response Center was raanned.
Additional regional support, enroute to thef si te for more detailed reviews.
_inclu gdi supervisor, are l_ _,$..
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The licensee tas repaired the seal injection line gaskat.
Regf 5W21r is reviewing safety concerris associated with the event.
CP&1. has agree'd,~o
{ionIl~hasconfirmed the agreement ;Iin writing, not to restart until safety question @s mtv bee'n' resolved.
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. : ar.
. The licensee i'ssued a news release.
The NRC does not pl'an to fsfue
'neM"r%, Tease. -
m The State of 5:)uth Carolina. has been informed.
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l y ;4.
Region 11 '(Atlpnta) rece'ived notificatt'on of this occurren,ct! bggl,tyhone Ir.:..
. :. -._:.... c: n,-
i
~t e
om the
, resident inspector at 2:45 p.tn. on November 30, 1981.
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This infornation is current as of 2 p.m. on December 1.1181. g.4. 3..y
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Contact:
C. ' A. Julian, RII 242-5538; C. W. Burger, RII 242-5532s.= i
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UNITED STATES D)['^n
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NUCLEAR REGULATORY COMMISSION 3
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December 1,1981, MEMORANDUM FOR:
H. Denton E. Case.
'D. Eisenhut R. Purple.
S. Hanauer R Vollmer R. Mattson J. Kramer DL Assistant Directors J. Sniezek
,Cr Michelsoni x
Di Crut~chfield THRU:
Steven A. Varga, Chief, ORB #1, DL FROM:
William J. Ross, Project Manager, ORBil, DL
SUBJECT:
DAILY HIGHLIGHT H. B. Robinson' Steam Electric Plant Unit 'No. 2 On Monday, November 30, 1981, a water leak occurred in the charging pump area as the result of -50 gpm leak (body to bonnet) in a valve in a line to the reactor coolant pump seals. _Approximately 1000 to 1500 gallons of wat' r spilled
~
e into the charging pump area and overflowed to the floor of the auxiliary building. A local emergency was declamd in these areas and partial evacuation of personnel initiated. The leak was isolated and the event teminated in four hours. Recovery from low (10%) pressurizer water level was aggravated by malfunctioning pressurizer block and relief valves. Safety injection was initiated but restoration of pressurizer level was achieved mainly through.
use of the charging pumps.. During the transient three additional events occurred: one diesel did not function properly upon safety injection initiation; the bellows in the relief valve of the let-downline ruptured; and a telephoned threat of a bomb explosion at 5:00 p.m. was received. The plant had been in hot shutdown mode since No:vember 6,1981, and the licensee plans to continue preparation for startup as soon as failed components are repaired.
IE plans to issue PN4 for the transient and bomb threat.
f
. WI William J. Ross, Project Manager Operating Reactors Branch #1 Division of Licensing l
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