ML20031B357
| ML20031B357 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 09/24/1981 |
| From: | Little L, Remick F, Wolf J Atomic Safety and Licensing Board Panel |
| To: | |
| References | |
| ISSUANCES-OLA, NUDOCS 8110010271 | |
| Download: ML20031B357 (95) | |
Text
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DocmtD usunc SEP2 91981!
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G1fice et t?;e Secretary [7 McUng & Senice UNITED STATES OF AMER [CA
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4 fiUCLEAR REGULATORY C0tG1ISSION p
ATOMIC SAFETY AND LICENSING BOARD 3ERVED SEP 231981 Before Administrative Judges:
John F. Wolf, Chairman Dr. Linda W. Little Dr. Forrest J. Remick
)
Docket Nos. 50-237 Ot
- In the Matter of
)
50-249 OLA
)
4 COMM0NWEALTli EDISON COMPANY
)
(Spent Fuel Pool Modifi-
)
cation)
(Dresden Station, Units 2 and 3)
)
)
September 24, 1981 PARTIAL INITIAL DECISION MODIFYING OPERATING LICENSE TO PERMIT IflSTALLATION OF FIVE HIGil-DENSITY SPENT FUEL STORAGE RACKS AT DRESDEN UNIT 3 Appearances
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Davf d Stahl, Esq. Phil'a Steptoe, F3q., and /
v Robert Fitzgibbons, Esq, for the -
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Cormonwealth Edison Company, c.
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Applicant.
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Richard J. Goddard, Esq., and Charles k Y ti, Mr g)4, A. Barth, Esq.,.for the Nuclear 4
/
Regulatory Conraission Staff.
Susan N. Sekuler, Esq., and Mary Jo Murray, Esq., for the In'oervenor, the % i.e of Illinois.
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8110010271 810924 '.
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TABLE OF CONTENTS Page I.
PRELIMIt1ARY STATEMENT 1
II. FItIDINGS OF FACT 6
A.
Board Questions 6
B.
Criticality Analysis 16 C.
Quality Assurance 21 D.
Transportation Damage 36 E.
Corrosion 41 F.
Radioactive Waste Manitoring and 47 Health and Safety G.
Accident Analysis 62 11.
Fuel Channel Deformation 68 I.
Environmental Impact Appraisal 81 and Safety Evaluation III. CONCLUSIONS OF LAW 84 IV. ORDER 85 Appendix A.
Exhibits 90 d
9 0
..m
. I. Preliminary Statement n n
.c.u.~..
1.
This partial initial decision is issued in response to Commonwealth Edison Company's (" Applicant's") motion for a partial initial decision modifying the operating license to permit at Dresden Unit 3 installation of five high-density spent fuel storage racks and the withdrawal of thirteen of the present spent fuel racks. The Applicant has pointed out that time is of the essence regarding the Board's response to the motion since Dresden Station must begin to shift fuel to prepare for the required forthcoming refueling outage no later than September 21, 1981 unless applicants five rack proposal isapprovedbythattime.M
-1/ There appears to be some flexibility in date for starting to shift fuel to prepare for the forthcoming refueling outage.
Certain NRC comnitments arising in the wake of the Three Mile Island Unit 2 accident must be satisfied by January 1,1982 or the Dresden Units 2 and 3 will be shut down. Applicant is shutting down on that date to meet that requirement.
(Tr. 1030-31.)
During the shutdown a number of corrective modifications will be made to the feedwater system.
( Tr. 1027-1028. )
To accomplish the modifications during the projected refueling outage period, the Applicant states that it must be able to accommodate the fuel core discharge by January 1.
(Tr. 1029.)
To neet that deadline Applicant contended at the September 11, 1981 hearing that rack removal and installation processes must begin by September 21,1981 (Tr.1044) and added that if the January 1,1982, outage date is not met the resulting additional shutdown beyond the scheduled outage would cost Applicant one h# to one million dollars per day (Tr.1045).
The Applicant also contends that if the 5 rack proposal is not adopted, the only available alternative woulo !;e to transfer fuel from the Unit 3 pool to Unit 2 pool.
This procedure would result in an occupational exposure of about 19 manerem.
It would also entail risks involved in moving heavy loads on the refueling floor.
(Testimony of Robert F. Janecek (Janecek, Five Rack Testimony) following Tr.10?l, at pp.10-11; Tr.
1025-27.)
2.
At the time the motion, dated August 13, 1981, was filed seeking a partial initial decision and approval of the 5 rack project, the Board had before it Comonwealth's application for amendments of the operating license for Dresden Station Units 2 and 3 relating to the modification of the spent fuel pools.
The Applicant sought to install naw storage racks whereby the storage capacity of the spent fuel pools would be increased from 1400 fuel assemblies for Dresden Unit 2 pool and 1420 fuel assemblies for Dresden Unit 3 pool to 3537 fuel assemblies for each pool.
3.
Notice of the proposed amendments was published in the Federal Register on August 11, 1978 (43 Fed Reg 35763).
In a notice of hearing dated March 29, 1979, the Board granted the State of Illinois' ("Intervenor") petition to intervene.
4.
Evidentiary hearings were held in Morris, Illinois from '
Hovember 19, 1980 through November 21, 1980 and in Chicago, Illinois on April 20, 1981.
However, issuance of a decision is being withheld pending receipt of answers to a Board No'tification, dated May 20, 1981 raising questions regarding the effect, if any, of a seismic occurrence on the Dresden 2 and 3 spent fuel pools.2] The NRC Staff (" Staff") subsequently requested the Board not to issue a final 2/ The full history of this proceeding will be set forth in the
~
Board's final Decision which will not issue until resolution of the seismic issue.
initial decision pending Staff's review of this issue.3/
Answers to the seismic question are expected during the course of the next two months.
S.
The Board deems it possible that the seismic answers may have to be reflected in some of the Findings of Fact.
It is for this reason that the Partial Initial Decision sought by the Apolicant in a motion, dated July 24, 1981, was not granted.
6.
The seismic questions raised by the installation of the five storage racks have been answered by Applicant.
The Staff supports approval of installation of 6 racks. However, the Staff has asked that the P1ard withhold, pending further analysis, a decision on the requested license amendments, i.e.,
installation in each pool of 33 high density storage racks to increase the storage capacity of each pool to 3537 fuel assemblics. Questions ' are yet to be an' wered s
regarding the effect of the installation of 33 storage racks on the stability of the pool structures during a possible seismic occurence.
7.
The NRC Staff response to Applicant's Motion for a Partial Initial Decision, dated August 13, 1981 stated:
"On July 24, 1981, Licensee moved the Atomic Safety and Licensing Board for a partial initial decision in the captioned proceeding. As stated therein, the NRC Staff does not object to the issuance of a partial initial decision on all issues not affected by Board Notification (BN-81-10),
dated May 20, 1981.
Licensee subsequently, on August 10, 1931, requested that the NRC Staff analyze proposals to install five of the proposed racks (Enclosure 1). The Staff has completed its analysis of this proposal (Enclosure 2) and
-3/ Letter dated June 29, 1981, to Board Members from Gus C. Lainas, Assistant Director for Safety Assessments, Division of Licensing, USNRC.
has concluded that the installation of five racks poses no safety issue.
Therefore, the Staff would not oppose an Order of the Licensing Board authorizing the installation of five racks, upon issuance of the requested partial initial decision, deferring its order on the installation of the remaining racks."
8.
NRC Staff's technical experts testified in favor of the "fivestoragerackproject."O The Intervenor, the State of Illinois, illed a response opposing Applicant's Motion.
The reasons stated for its opposition were not persuasive.
9.
At the close of the hearing on September 11, 1981 on Applicant's Ibtion for a Partial Initial Decision Approving Installation of Five Racks, the Board orally granted the Motico in 1
part and denied it in part.
It suranarized orally the relevant Findings of Fact and announced that it would subsequently issue a written order.
1.
A representative of the NRC Staff stated on the record that it took issue with the Board's decisicn that a partial initial decision need not be made.
He cited 10 CFR 2.760(r) and stated that 10 CFR 2.730 does not affect 2.760(c). Another staff representative questioned the Board's procedural decision.EI 11.
After the hearing on Friday September 11, 1981, it was clear from the record that the "5 rack project" should be approved.
4/ Tr. pp. 1127 et seq.
1/ Tr. 1183-84; Tr. 1188-90.
t However, since the _ Board's oral order was questioned by Staff Attorneys, it was determined that the Board's order would be issued in the form of a partial initial decision.
This conclusion was reached in order to avoid a waste of time over a procedural matter.
Such wasted time would increase the outage time at greet expense to the Applicant and tiie consumers.
Further, other methods of accommodating the required fuel discharge, i.e., alternatives to the action sought, would increase the occupational exposure and wculd not avoid the additional outage time and expense.
12.
On bbnday September 14, 1981, in a telephone conference.the Chairman of the Board advised counsel to the parties in this matter of the form its order would take.
3
. II.
Findings of Fact
.-_....m..
Fol!owing the first two hearing sessions, on the application to 13.
modify the spent fuel pools at Dresden Station, proposed findings of fact were submitted by Applicant, Staff and Intervenor.
In many instances, Applicant's proposed findings were adopted by the NRC Staff and Intervenor.
The Board has evaluated all proposed findings by the parties.
It has relied on those findings in part in preparing this decision.
A.
Board Qeestions 14.
Board Questions 1 and 2 were the subject of extensive testimony during the proceeding.
Board Question No. I asks:
A.
What is the current status of the spent fuel unfilled storage capa::ty at Dresden Station Units 2 and 37 B.
When will full core discharge no longer be possible?
C.
When will normal refue?ieg discharga no longer te possible?
D.
What alternatives, if any, exist to shutting down the Unit (s', when the spent fuel pool (s) is ' ~e) fi' led to capacity?
E.
Which, if any, of these alternatives would require subsc. tent license amendments?
Board Question 2 states:
Based on a review and analysis of the various generic unresolved safety U. sues under continuing study, what relevance is there, if any, to the proposed spent fuel pool modification? Further, what is the potential health -nd safety implication of any relevant issues remaining unresolved?
.?-
.,,, 15.
Board Questions ' and 2 are not addressed in this pr.tial idtial decision.
Board Questions 4 and 10 deal with those aspects of Board Questions 1 and 2 which are applicable to the installation of five racks.
Consequently, the full discussion of Board Questions 1 and 2 is reserved for the Board's final decision on the installation of 33 racks.
16.
At the hearing session of September 11, 1981, the Applicant and Staf f presented testimony on eight additional Board questions (Goard Questions 3-10) relating to the request by Applicant to permit installation of five high-density racks prior to the final resolution of the seismic
'issueidentifiedbytheStaff(i.e.,theissuewhichledtheStaffto request the Board to cafer issuing a decision on installation nf 33 racks).
Intervenor State of Illinois particip&ted in cross-examinatfon.
The Board did not direct the parties to present proposed findings on this narr,w issue nor have the parties chosen to file any.6f
- 17. Board Question 3 asked:
What is the history and current status of the seismic issue tthich led to Board Notification BN-81-10, dated May 20, 19817 l '8.
Robert F. Janecek testified on behalf of the Applicant. Kenneth S. Herring t,tified for the NRC -taff (Staff).7/
- 19. The seismic issue was identified during Staff's perforrance of the Systematic Evaluation Program (SEP), Topic IV-1 assessment of the Dresden 2 6/ 10 CFR 2.754(a).
-7/ Testimony of Robert F. Janecek (Janecek, Five Rack Testi,nony) pp.1-4 following Tr. 1021.
Testimony of Kenneth S. Herring (Herring) on Board Question 3, following Tr.1134.
l t
i L
. and.3. spent fuel pools 'SFPs) with high density racks.
Unlike. the racks
. _.. >,. m currently in use, the proposed rccks are free-standing, rather than attached to the SFP structure.
In the course of the SEP assessment it was discovereo that Applicant had not evaluated certain motions of the racks, i.e., rocking and tipping, during postulated seismic events. Consequently, Applicant was directed to provide information to demonstrate compliance of ths proposed Ci' modification to show "that any sliding and typing motien will be contained within suitable due to the cleart.nces is incorporated."garances, and that any impact.
geometric constraints such as ti ' mal c 20.
Staff witness testified that the major problem delaying resolution of this issue is the lack of adequate demonstration, by detailed nonlinear dynamic analysis, that the SFP floor can withstand the impact loads thich would be imparted by the installation of all 33 proposed racks should they rock (tip) during seismic events.AI 21.
Staff notified Applicant in April,1981, of the neej for further analysis of this issue, as well as of an error in the Licensing Report (Applicant Ex. 2) on minimum spacings between racks and between racks -and pool walls.
The Staff issued Board Notification BR-81-10 on May 20, 1981.5I Since April,1981, Staff and Applicant has ioined in extensive analysis of the seismic issue.EI All of th; Staff concerns 8/ Herring at p. 2.
9/
Id.
_10/ Janecek Five Rack Testimony at p. 2.
11/
Id. at pp. 3-4.
1
_g_
_ _.,m.Ile.ta11.ed in BN 10 have been resolved with one exception, the. potential stresses on the pool structures during an earthquake should all racks tilt up and fall back simultaneously.
Staff arid Applicant havc. disagreed as to the likelihood, of such an event and as to the appropriate analytical methods which should be used to calculate uplift rid impact energy of the tiltedracks.EI Because of the complex calculations required, coupled with time constraints in regard to preparation for pool reracking, the Applicant requested the Staff to inform it as to what assumptions would provide sufficient conservatism for impact analysis of five. racks.
A'pplicant then performed this analysis,EI upon which Staff approvi.d the limited five-rack proposal. EI
- 22. O.: second phase,
.ich will require further lengthy and complex calculations, will be to demonstrate the adequacy of the ipstallatioir of all the hirn density racks in the SFP's.EI On receipt and review of this analysis, the Staff will issue a supplement to the SER conta aing its evaluation of the 33-cack full re. aciing proposal.EI H/
Id.
-13/ Seismic Anaiysis for Installation of Five and Ten High Density Fr?
Racks, by letter cf T. J. Rausch, dated August 10, 1981.
-14/ NRC Staff Response to Applicant's Motion for a Partial Initital Decision dated August 13, 1981; Janecek F_ive Rack Testimony at
- p. 4.
H/ Herring at~p. 3.
16/
Tr. 1152.
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" 2 3.
Board Question 4 asks:
Is the current seismic question related to Soard Questier 2 on Unresolved Safety Issues, or is it a separate issue related to the Staff's Systematic Evaluativ Program (SEP)'?
24.
Witnesses for both Staff and Applicant testified that the currunt question is not related to any Unresolved Safety Issue but rather b oerceived inadequacies in Applicant's Li.e:~. sing Report identified as a result of tN Staff's SEP.
The question must be resolved to easure compliance with the criteria against which the proposed 33-rack installation was initially reviewed.EI 25.
30ard Question No. 5 asks:
What specifically is being proposed for Board considention?
For example:
A.
What size racks are being proposed for installation.
(i.e., 9 x 10 or 9 x 13 arrays)?
B.
Are there any special risks associated with fuel or rac) (old or new) movement which are different than previously testified to?
C.
What will be the disposition of the removed racks?
l 26.
The Applicant proposes to install five new 9 x 11 array storage racks in the northernmo:.
part ;f Unit 3 pool.
To make space for these racks thirteen existing racks will be removed from the north end of the pool.
27.
The installation of five new racks and the removal of thirteen existing racks will result in a net increase in storage capacity of 23F spaces.
Because the Unit 3 ' pool is presently 104 spaces short of full core i
E/
Janecek Five-Rack Testimony at p. 5; Tr.1135.
j discharge capability (FCDC) "r proposed installation of five racks will reestablish FCDC far the January 1982 shutdown.EI l
28.
The only special risk associated with the t'.o-step installation (five racks folicwed by twenty-eight versus thirty-three at one time) is an additional exposure of less than one-half man-rem associated with approximately 650 additional fuel moves caquired. EI 29.
The racks removed from tha pool will be decontaminated by rinsing or hydre.ining, then stored temporarily in the Dresden Unit 1 fuel' butidicg.
Some of these racks may be shipped to the Quad Cities Nuclear Stat.or for further use.
The remaining racks which have been removed may be either shipped whole or shredded to reduce their voluTie and then shipped to a aurial site. Additionally, a method of chemically cleaning the, acks to remove surface e.cntamination is bein~g ir,vestigated.
If successful, the decontaminated racks may then be sold as scrap.20/
30.
Board Question 6 states:
. Compare the relative accuracy of Method I and Method II for calculation of responses of the s.oent fuel pool and rack structures to seismic stresses.
- 31. Method I and IIEI are approximate methods for determining impact energy which use an energy balance approach e a substitute for H/
Janecek Five-Rack Testimony at p. 6; Tr.1135.
E!
Janecek Five-Rack Testimony at p. 7; Tr.1135-36.
20/
Janecek Five-Rack Testimony at p. 7; Tr.1023-25.
_2_1/ Methods. I and II are methods for calculating the kinetic energy.0,f..,_._
an uplifted storage rack upon its impact with the pool floor.
Applicant's Ex. 4; Staff Ex. 2.
,.__..,m.non-ljnear time-history analyses. The response of the spent fuel pool racks and structures to a seismic event is a complex phenomenon.
It is extremely difficult to discuss the accuracy of either Method 1 or Method II for the calculation of e:smic resosises.
32.
In Method I, the angula: velocity is computed by equating the restoring moment of the tilted rack to the product of its moment of inertia and the angular acceleration. The degree of conservati n in the application of M%d I is difficult to estimate.
33.
Method II assures that all of the potential energy. of the uptilted rack is transformed into the kinetic energy of vertical impact.
This method is conservative for assessing the effects of a half cycle of impact.
arealistic motion is assumed in the application of Method I.
34.
Given the conservatism thought to be inherent in Method II, the margins'in the rack and pool structures calculated by the Applicant through applying Method II for the five rack case, and considaration of the lesser seiWc hazard specified for Dresden 2 by the "P site specific spectra, there is reasonable assurance thac the integrity of the rack and pool structure will be maintained, giv,en the occurrence of an SSE, without the necessity fcr comparing the relative accuracy of the two methods utilized or for further c;uantitation of the uncertainties of the methods.E 35.
Board Question 7 asks:
What evidence, if any, already received into the record in the context l
of replacing all racks should be revised, struck or otherwise modified j
in the satext of the proposed installation of five racks?
-22/ Testimony of Dr. Quazi A. Hossain following Tr.1071; Response to Board Question 6 by Kenneth S. Herring following Tr.1138.
l 3.6..
Little evidence, already received into the record in th,e.ontext
,,_.,u,_,..
of replacing all racks in botn pools, will need to be revised in the context of the proposed installation of five racks / Revisions include those relating to the date when Full Core. Discharge Capability will be lost, the number of racks being installed and thus additional fuel storage capacity, additional fuel movements, and additional occupational exposure.EI The necessary revisions are not significant and have been incorporated in this Partial Initital Decision, where applicable.
~
37.
Board Question 8 asks:
Will Applicant's previous commitments and/or proposed license conditions, made in the context of replacing all racks, be applicable in the context of the installation of five racks? For example:
(a)
Quality receipt inspection procedures (b)
Handling of loads over stored fuel
.(c)
Corrosion surveillance progran (d)
In-situ neutron attenuation tests (e)
Removal of rack if more than one Boral plate is found missing (f), Plug gauge testing and lead-in clip grinding 38.
All commitments ano/or proposed license conditions, made in the context of replacing all racks in both pools, will be applicable in the context of the installation of five racks in Dresden Unit 3 spent fuel pool.El
_23/ Janecek Five-Rack Testimony at pp. 7-8; Tr.1139.
M/
Janecek Five-Rack Testimony at p. 9; Tr.1139-41.
39.
In addition to the six coamitments or conditions listed,-the condition that fuel stored in the spent fuel pool shall have a U-235 loading less than or equal to 14.8 grams per axial centimeter is also applicable.
This Lindition and relt..ej technical specification was presented in the ccntaxt of replacing all racks', but is also applicable to the installation of ffve racka.S5I 40.
Board Questica 9 asks:
Is there any possibility that the resolution of the currently unresolved seismic issue could result in need for removal of the five high density racks and reinstallatico of the current
- racks, i.e., in reversal of the decision being sought?
- 41. Witnesses for both Stair and Applicant testified that they saw no possibi'ity c.iat the old racks would ever be installed in place of the five new racks since it has been fully deconstrated that the pool floor can-withstand the impact of five racks.
In the unlikely event that they had to be removed, they could be.2p/
42.
Board Question 10 asks:
What alternativec to the proposed installation of five racks are available to the Applicant to achieve the proposed full core discharge capabilty (FCDC)?
43.
Appliant's witness presehted testimony, with which Staff concurred, that the only available alternative is to transfer fuel from the Unit 3 pool to the Unit 2 pool using GE IF-300 Casks.
This operation would ressic in an occupational exposure of about 19 man-rem (compared to less 4
than half a man-rem for installation of five racks) and involve movement of i
25/ Staff Ex.1, Safety Evaluation Report at p. 3; -Tr.1186-88.
2p/ Janecek Five-Rack Testimony at pp. 9-10; Tr.1141.
...,..,.J1eavy. loads on the.? fueling floor.
Further, the fuel would eventually have to be transferred back to the Unit 3 pool if new racks cre eventually approv._ 'or the Unit 2 pool.2_7/
- 44. Applicant has considered transferring fuel from the Unit 3 pool to the GE-Morris facility, but it does not have a contractual agreement with GE-Morris for either permanent or temporary storage. Furthermore, there would be no advantage to this transfer since it is unlikely that GE-Merris would permit permanent ttorage.28/
45.
Intervenor raised the possibility of putting on additional shifts in order to transfer fuel between pools more rapidly, so that inst'allation of the five racks could be delayed beyond September,1981.
46.
The Staff could see no reason to get involved with this question, considering schedule arrangements made by Applicant in carrying out its licensed activities as not a mutter of concern as long as these activities are performed safely and in accordance with its license and applicable NRC reguictions.EI
- 47. The Board finds that the installation of five high density racks is the best available alternative for achieving full core discharge capacity for Unit 3.
27f Janecek Five-Rack Testimony at pp. 9-11; Tr.1027; Tr.1142-43.
28/ Tr. 1025-26; Tr. 1142.
g/ Tr. 1163-64.
J.
Criticality Analysis
_..m.,.
- 48. A criticality analysis was performed on the proposed storage racks.
It is contained in the Applicant's Licensing Report.E! Since criticality is related to the geometry of the fuel assemblies in the racks, such analysis is applicable to the high density racks, whether five or 33 are at issue.
49.
The proposed storage racks consist of a vertical array of rectangular stainless steel tubes welded together at the corners to-form a checkerboard pattern.
Spent fuel is stored within the tubes and in the space.e between tubes (inter-tube) formed by the checkerboard pattern.
There are four neutron-absorbing Boral plates within each stainless steel tube, one on each side.EI 50.
For full reracking two sizes of racks have been designed to provide'the additional storage.
The size applicable to the instant "5-rack" modification will store 99 fuel assemblies in a 9 x 11 array.
The other wil4 store 117 in a 9 x 13 array. Full rerac;.ir g, if permitted, will use 18 racks with a 9 x 11 array and 15 ncks with a 9 x 13 array making a total of 3537 storage, paces for each generating unit.
Each rack has a checkerboard pattern with an absorber tube at each corner.
The tubes are made by Brooks and Perkins Company. 2 The checkerboard pattern is u.11ike that
-30/ The Licensing Report was prepared by Applicant's architect engineer, Nuc' ar Servi e Corporation (NSC).
The document introduced as Applicant's Ex.
I was revision 4 of the Licensing Report.
Subsequently, revision 5 of the Licensing Report was submitted to show the location of the vents in the storage tubes.
Tr. 460-91, 495-97, 499-500, 584-97.
g/ Applicant's Ex.1, pp. 3-1, 3-30; Tr. 467.
_3_2/ Applicant 's Ex. 1, pp. 3-6; Tr. 478.
2
~
-,-~-., -of the Applicant's Zion absorber racks which have a neutron abscrbing tube for each storage position The minimum boron-10 content of the Boral plates 2
in bot'h the Zion and the Dreeden racks is 0.02 gm/cm.
However, BWR fuel assemblies such as those used at Dresden are much smaller, contain fewer fuel rods and have lower enrichment than the PWR assemblies used at Zion.
Dresden fuel assemblies will be limited by Technical Specifications
. to less than half the U-235 content authorized for Zion fuel.E
~
51.
The effective multiplication factor, K,g, is a measure of g
how close an array of 1 assemblies is to being a self-sustaining auclear chain reaction.
When K is equal to 1.00, the reaction is self--
eff sustaining (that is, " critical").
The proposed racks are designed to keep in the spent fuel pool below 0.95 in accordance with the NRC v
e f f Standard Review Plan and ANSI Stanoard N18.2.
The limit of 0.95 is an becomes equ'ai to or greater'than 1.00 in important criterion.
If K eff the spent fuel pool a criticality accident with serious consequences could 33/ See ' Commonwealth EdOcn Company (Zion Station, Units 1 and 2), LBP-80-7, 11 NRC 245, 269, 77 D 79-80, 295 (1980); Affirmed ALAB-616,'12 NRC 419 (October 2, 1980); NRC Staff Ex. 1, Safety Evaluation at
- p. 3; Tr.
467-68.
r-
9 result. Maintaining K below 1.0 is important in i spent fuel pool eff because there are no control rods in the pool to stop the chTin reaction.E!
- 52. The Applicant's criticality calculations conservatively assume that the fuel is clean (i.e., no fission products), there has been no fuel burn-up and there is no burnable poison gadolinia remaining in the fuel.El 53.
Subcritical mcitiplication experiments could be conducted ~during the full reracking of fuel in the 3resden 2 and 3 spent fuel pools.
However, there k no indication in the record that this would be done, or that it should be required.EI At any rate, it is not applicable to the "5 rack" modification.
- 54. Applicant has made a conmitment' to conduct an in-pool neutron-attenuation test of a sufficient nuai,er of storage locations to ensure to a 95% con."idence level that no'more than 1 Boral plate out of 32 is missing.
If one plate is found missing, the tube location (but not the 34/ App'licant's Ex.1, pp. 3-9; Tr. 473-77; 586-80.
In boiling water reactors such as Dresden 2 and 3 the snent pool water is unborated.
While it might be possible to pump bor :'1 into the pool from a remote location, there are no existing procedures to do so.
Tr. 587, 593.
M/Tr.469,47,,585-86,591-93.
36/ Tr. 591-93.
The Board takes notice of the fact that reracking of the spent fuel would provide an excellent opportuni'a to conduct a subcritical multiplication experiment, from whicn Keff for fuel in the storage racks, as built, could be readily determined. The results could assist in determin".g the magnitude of conservatism in calculated versus the actual valve for Keff.
This should be of interest to the Applicant.
Such data also might assist the Staff in its evaluation of calculations for future fuel storage proceedings.
.,m_, adjacent inter-tube storage location in the checkerboard pattern) will be blocked to prevent insertion of a fuel assembly.
Further, every tube in the ?ool then would be subjected to the neutron attenuation test.3_7 /
w uld be more
- 55. Applicard s Licensing Report shows that Keff than 0.95 if more than one c1t of thirty-two Boral plates is missing.
Therefore, the Board was co1cerned that, if more than one Boral plate is missing out of thirty-two, blocking the. associated storage tube m:, it not be a sufficient measure to maintain K less than 0.95.
The opinion of eff Applicant's witnesses was that the decrease in reactivity wou M be much more than the increase in reactivity due to the missing plate.
However, no specific analysis of this Ituation had been conducted.
Therefore, Applicart modified its comitment to 'the Board to provide that, if more than one missing Boral plate is detected, Applicant will remove the racks containing such additional missing plate or plates from the pool.
Such racks will not be replaced in the pool until a specific criticality analysis will n't exceed 0.95 has been o
of the situation showing that Keff submitted to and approved by.the f1RC.
There. fore, there will be no more than 37/ TiRC Staff Ex.1, Safety Evaluation at p. 3; Pickens at p.16.
To achieve a 95% confidence level, 63 tubes would have to be checked.
However, because Applicant's contract with the testing contractor requires that a minimum of 300 tubes must be checked per visit to the station by the contractor, a higher confidence level will be achieved.
Tr. 227-29, 483-84, 495-97.
In subsequent testimony in the hearing on the installation of five racks, it was indicated that the Applicant is charged for a minimum of. 300 tubes for each visit and more than 63 tubes may be checked during each visit.
Janecek Five Rack Testimony at p. 9.
,one missine alate allowed in each pool, and that missing plate will have the associated tube blocked. EI
- 56. Applicant's witness testified that a c-iticality event in the spent fuel pool could only occur through poor quality manufacture, design and testing.
He did not believe that such an accident is credible because of the design of the proposed racks.E 57.
The Applicant informed the Board in January 1981. that it nad purchased,o 8 x 8 fuel from r,xon Nuclear Corporation, This fuel is proposed'for use in future reloads at Dresden Units 2 and 3.
At that time of 0.95 will not the Applicant submitted ca affidavit showing that Keff be exceeded if the Exxon Nuclear Fuel is stored in the proposed racks.
Applicant's witness testified with respect to this affidavit at the evidentiary hearings held on April 20, 1981.
The Exxon Nuclear fuel has not yet been appi aved by the Staff for use at the Dresden Station.
Such approval will require further licensing action by the NRC, including a criticality review by the Staff addressing the storage of Exxon Nuclear fuel in the Dresden storage racks.EI The potential use of Exxon fuel does no' affect the instant partial initial decision which addresses 5-racks t
which will receive fuel presently in the core.
t 38/ Applicant's Ex.1, pp. I 17, 3-27, 3-28; Tr. 229, 484-87, 595-96.
39/ Tr. 587-88.
40/ Affidavit of Kin W. Wong dated 21 January 1981 (subsequently bo" d in the transcript following Tr.1013); Tr. 827-29.
4 m
-w e
,p
,,n, r
, 58.
The Board finds that the pending application to use Exxon Nuclear fuel at the Dresden Station does not present an impediment to the issuance of th'e proposed 'icense amendment in this proceeding for the installation of five racks for storage of existing General Electric spent fuel assemblies at Dresden Station.
- 59. The Board finds that, with'the quality assurance program for the
~
manufacture of the racks and the described comitment to neutron attenuation testing of the racks there is reason &ble assurance that a criticality event will not occur in either Dresden spent fuel storage pool.
60.
Further, the Board finds that the criticality analysis performed will not exceed by Applicant provides reasonable assurance that Keff 0.95.
C' Quality Assurance 61.
Contention 2 states:
ihe Application does not show that the quality control and quality asst rance programs of /.pplicant and its contractors are adequate to assure that tube and rack construction and the boron-10 loading of the Boral in the tubes will meet specifi-cations.
62.
The proposed storage racks are to be fabricated in.two stages.
First, the stainless steel tubes containing Boral are manufactured by Brooks and Perkins Company, Livonia, Michigan.
Second, the tubes ~are shipped to Leckenby Corporation, Seattle, Washington, where they are welded together in a checkerboard pattern onto base plates to form the storage racks. Finally, the completed storage racks are shipped to the Drc M en Nuclear Station for installation.
Applicant, NSC, Brooks & Perkins, and Leckenby Corporation
. 4 each.have quality' assurance programs to ensure that the proposed storage racks when completed and installed meet safety-related design requir.ements.
- 63. Applicant's quality assurance program meets the requirement of 10 CFR Part 50, Appendix B, " Quality Assurance Criteria for Nuclear Plants;"
Section III of the ASME Boiler and Pressure Cod.t; ANSI Standard N45.2 -
" Quality Assurance Program Requirements for Nuclear P0wer Plants;" and applicableNRCRegulatoryGuides.SI The NRC's Office of Inspection and Enforcement has found that Applicant's qulity assurance program has been satisfactorily implemented at Dresden. Additionallyg the quality assurance programs of Applicant's contractors and subcontractors meet the applicable portions of 10 CFR Part 50, Appendix B, as required by Applicant's comitment in the Comonwealth Edison Company Quality Assurance Topical Report.S In connection with this' project, Appli -
cant's (v iity Assurance Department and the quality control group-in Applicant's Station Nuclear Engineering Division reviewed the NSC, Brooks d
& Perkins, and Leckenby qual'.ty assurance manuals and found them to be acceptable.SI 64.
In its Proposed Findings 64-71, Intervenor criticized Staff and Applicant witnesses' testimony on the quality assurance program, in that the 41/ Testimony of Walter J. Shewski (Shewski) at p. 2, following Tr. 239; Supplemental testimony of William L. Belke (Belke) at p. 2,
~
following Tr. 422; Tr. 429-30.
g/ Belke at pp. 2-3; Tr. 424.
fl/ Shewski at p. 3, Tr. 240-41.
.., _ _ g itn, esses proffered were not primary reviewers of all quajity assurance documents but are the supervisors of the primary reviewers. The Board does not agree with Intervenor that the testimony was flawed as g consequence.
- 65. The quality assurance programs of Applicant, NSC, Brooks &
Perkins, and Leckenby are designed to assure that materials and processes utilized in fabrication of the racks will meet safety-related design requirements and to assure the quality and correctness of the manufacturing process.N
'and stainless steel materials to be used in the neutron absorbing tubes are certified by the suppliers of these materials as meeting applicable American Society for Testing and Materials (" ASTM") standards as required by the
^
procurement specifications.
The certification documentr, which are' traceable to specific lot numbers of the supplied materials, are supplied to Brooks & Perkins.
Brooks & Perkins quality assurance personnel review the certification documents to ensure that the materials conform with the procurement specifications.
Additionally,' Brooks & Perkir; audits the supniier of the boron carbide to ensure its certifications are acceptable.S 67.
The Brooks & Perkins certification review and verification are documented in a " Nuclear Material Review Report,' prepared by Brooks &
Perkins Quality Assurance personnel.
This document is forwarded to NSC, 44/ Shewski at p. 5.
45/ Shewski at p. 6.
- 24 =
.__.. _,.,,Wic,h is required to review it and ascertain thg acceptability of the certification documents and Brooks & Perkins review thereof. Only when such a finding is made are materials released by NSC to Brooks & Perkins for fabricationintotubes.$
>B.
Quali',y assurance in the manufacturing process is achieved by inspection and :ampling at several points during fabrication of the tubes and racks.
The tub..
are double-walled structures into which Boral plates are inserted. To ensure that the neutron absorbing quality of the boron carbide in the Boral plates has not been altered in the manufacturing process, all plates are inspected for proper thickness at six Iccations and a sample is taken from each end of the plates.
Of those samples, 10% are analyzed, chemically or by neutron attenuation measurement, for boron content.$
- 69. The storage tubes are fabricated at Brooks & Perkins by folding and welding stainless steel plates to f'ra the inner and outer walls.
Brooks and Perkins quality control personnel visually inspect the inner and outer full-length seam welds of each tube; in addition, dye penetrant tests are performed on 10% of the outer tube seam welds. After insertion of the Boral plates,'the completed tute assembly is subjected to inside and outside i
visual and dimension tolerance check.
Further,'10% of the completed tubes 4y Shewski at p. 6.
4// Shewski at p. 7.
l l
.ar.e given a full-length check with a simulated fuel element to verify straightness and proper clearances and to ensure no binding occurs.$
70.
Brooks & Perkins utilizes a new computerized system to chec'K chemical analysis, materials, fabrication, and personal inspection activities to verify acceptability of Boral sheets and other materials and to verify identification of each tube.
Only if a specific tube meets all these quality-related requirements is it approved by computer print-out.
The computer checkout has in at least one instance erred in failing to reject tubes for inadequate Boral content.
However, further quality assurance measures are taken, in that following its own approval, Brooks &
Perkins is required to forward data, inspection, and weld reports to NSC for review and acceptance.
Only upon determination by NSC that Brooks &
Perkins' quality requirements have been met and that design and fabrication' requi'rements have been met, are tubes released for shisuent to Leckenby forrackfabrication.N 71.
Leckenby conducts its own quality assurance inspection and review during fabrication of the tubes into storage racks.
NSC reviews the data sheets and weld reports documenting these activities and releases the completed racks for, shipment to the Dresden Station on determination through its inspections ano documentation review that the racks' design, fabrication, and qua.lity requirements are acceptable.
g Shewski at p. 7.
49/ Shewski at pp. 7-8; Tr.193-94.
.,,...w
- - +
i
- 4
- _ _ m,_ m 72.
Finally,'on _ receipt of the storage racks at the Dresden Station and their immersion in the pools,-a neutroa attenuation test will be conducted on a sampiing of storage tubes to confirm presence of the Boral plates.ES/ Applicant's commitment in the event missing-plates are i
detected is described supra 1 54 & 55.
i 73 To comply with 10 CFR 'Part 50, Appendix B, Criterion XVIII, Applicant, NSC, Brooks &.'erkins, and Leckenby periodically conduct self-audits, and in addition Applicant has hired NSC to perform audits, surveillances, and' inspections of Brcoks & Perkins and Leck.enby during rack fabrication.
Applicant conducts audits and surveillance of NSC, Brooks &
-Per< ins, and Leckenby.
The NR0 has the authority.to audit Applicant, or NSC, Brooks & Perkins, or Leckenby.<E'! Board Exhibits 1-3 consist of Applicant's purchase orders to NSC, Brooks & Perkins, and,Leckenby, which i
set fortn responsibilities of.each entity.
- 74. Applicant audits NSC's quality assurance activities to ensure that the NSC audits are conducted in accordance with NSC QA programs.
f Applicant's audit includes review of NSC audit reports.
Deficiencies found l
50/ Shewski at pp. 8-10.
51/ Shewski at p. 3; Tr. 244-47, 254-57, 331-32; Board Exs. 1-3.
9 i
f f.
....... _ -. _,,,,,, - _.. - -. -., -... - - ~ - - _ _ _ _... _,,, _...,.,,.... _. _,, _.......... _ _.,....,, -.,,....,
..,__.~, by.NSC which cannot be resolved with Brooks & Perkins or Leckenhy are reported to Applicant for corrective action.j2/
75.
Several quality assurance terms were defined as used in this proceeding:
" audit" - a function.done in accordance with specific formally approved checklist questions; " surveillance" - review on a continuing basis of activities without a formal checklist; and " inspections" - those specific detailed inspections rcquired under contract during and after fabrication to establish that items are acceptably built.
Checklists may be standerdized, specifically designed for the activity to be. audited, o,r a combination thereof.
Types of deficiencies disclosed in audits are classified in audit reports as follows:
" finding" - violation of a rule, such as a commitment or one of the 10 CFR Part 50 Appendix B Criteria; " observation" - a variance considered by the auditor as less severe'then.a finding, generally an item which is almost, but not completely, implemented; and " comment" - a nonenforceable suggestion.
76.
When one of Applicant's audits discloses either a finding or an observation, the audited firm must indicate corrective action and commit to a date for its implementation., Shortly af ter that date Applicant conducts a follow-up audit to co.nfirm satisfactory completion of the corrective action.
Applicant's quality assurance manager maintains a list of open findings and observations, updates it monthly, and reports it to Applicant's upper 52/ Shewski at pp. 3-4; Tr. 244-54.
. management. Applicant keeps on file audit checklists, responses from audited firms, and associated closeout reports.E 17.
If a deficiency.is detected during a surveillance, a deficiency letter is written in two to three days, and an audit of that item is required within two weeks. E 78.
Generally' audits, surveillances and inspections are not conducted on an unannounced or surprise basis.
The system of audits, surveillances and inspections is designed to discourage circumvention of requirements andprgcedures.E
- 79. Several technical problems which arose during fabrication of the Dresden racks were addressed.
Slight bending of tube arrays due to shrinkage of full-length welds on cooling was identified by Leckenby and satisfactorily resolved by Applicant, NSC, and Leckenby by " flipping" the racks after completing welding of each row of tubes, thus allowing the bow l
of each subsequent row to cot'nteract bow from the previous row.
Calculations by 11SC show that with this welding technique bowing is within allowable tolerances.
As part of final rack inspection a mandrel test is conducted to confirm adequacy of individual storage location dimensions.N 53/ Tr. 310-11, 315-26, 311-13.
54/ Tr. 314.
j 55/ Tr. 326-28. 331.
(
56/ Pickens at pp. 14-15; ir. 200-01, 224-26, 702-05.
[
_ 80.
Accumulation of pitch spacing tolerances from cooling and shrinking of corner to corner welds produced an unanticipated systematic effect on the first rack fabricated, i.e., the rack pulled together into a more dense configuration than specified, causing overall dimensions of the rack to be slightly too small (25 mils shorter in one lateral dimension than specified).
NSC's criticality analysis, assuming that the specified pitch spacing of 6.3" + 0.060", nonaccumulative, was decreased to 6.24" (6.3" -
0.060") center-to-center spacing, demonstrated that the rack would still less than 0.95) (Applicant's satisfy criticality requirements (Keff Exhibit 1).
Leckenby initiated tooling changes to i m ove control of pitchspacingonsubsequentracks.E 81.
Problems have arisen at Brooks & Perkins due to loss of proper identification of Boral stock during rolling and stamping, occasioned by deformation of identification marking during fabrication. Also, documentation establishing acceptability of neutron attenuation properties for some plates has been lost after their insertion into the tube.
In the first case, reidentification was made by neutron attenuation testing at the University of Michigan's research reactor.
In the second case, physical limitations prevented tests of completed tubes at the University so Brooks &
Perkins contracted with National Nuclear Corporation to perform a neutron attenuation test.
Brooks & Perkins has a new computerized system to reduce identification problems. NSC is required to inspect documentation to ensure each tube has adequate Boral content prior to shipment from Brooks & Perkins 57/ Pickens at pp.15-16; Tr. 202-03, 226-27.
,to,_Lec kenby.
In this capacity NSC rejected two tubes determined to have inadequate Boral content.- Applicant's witness believes that no tubes having inadequate Boral have been shipped to or accepted by Leckenby.El 82.
In addition to the fabrication problen described above, therc
~
have been some deficiencies in implementation of quality assurance programs for the storage racks.
Design of the proposed racks began in August or September 1977 and fabrication at Leckenby began April 10,1980. The racks were initially designated "non-safety related," hence, the original purchase order did not specify that the fabrication of the tubes was " safety-related".
In late 1977 Applicant upgradeJ the project to safety-related and verbally so notified NSC at that time. Applicant failed, however, to amend the NSC purchase order until October 1980.
Reverification of the design was completed by NSC in September 1980.
Applicant testified that despite the documentation error, the quality of tube and rack fabrication was not compromised, since as early as October 1977 all work at NSC was done as safety-related in accordance with their quality assurance progam.
Brooks &
Perkins and Leckenby's contracts were identified from the beginning as safety-related 'and all work was done in accordance with their own quality assurance programs.EI
--~58/ Pickens at pp.13-14; Tr.192-94, 219-23; Intervenor's Ex. 2; see V 84, infra.
59/ Tr. 185-90, 234, 318; Pickens at pp.11-12; Intervenor Ex. 4.
._ _.83.
Several documents relating to quality assurance and obtained from Applicant during discovery were introduced into evidence by Intervenor.
84.
Intervenor's Exhibit 2 (see paragraph 81, supra) is a " Trip Report" dated September 2, 1980,. indicating rejection by NSC of two tubes at Brooks & Perkins because of inadequate Boral content.
85.
Intervenor's Exhibit 3 is a report dated December 26, 1979, of results of an NSC audit of Brooks & Perkins on December 12-13, 1979. The report noted one finding, relating to violation of 10 CFR 50, Appendix B, Critcron II in tiiat. the QA procedure for compliance with 1,0 CFR 21 was not
' established, and five observations.
Intervenor's Exhibit 38, consisting of the closecut documents from this audit, indicates corrections as necessary of noted deficiencies. This exhibit, supplied by Applicant on December 1, 1980, to complete the evidentiary record pursuant to stipulation, includes:
Brooks >& Perkins' response to the NSC audit, dated January 31, 1980; NRC's evaluation of this response, dated March 13, 1980, and the cover letter accompanying NSC's transmission of these documents to Applicant, dated
(
April 28,1980.N 86.
Intervenor's Exhibit 4 is a report dated September 18,1980, of an i
audit by Applicant o,f Brockc & Perkins on September 11-12, 1980.
This audit i
resulted in four observations of deficiencies in specification and documentation of dutics of some QA personnel; review of purchased items and l
6_0/ Intervenor's Exhibit 3; Tr.195-200; Tr. 511-12.
I t
..,. ~
___.,_nlatef als for conformity to purchase order requirements; trainir,ig of some i
personnel performing activities affecting quality; and timeliness of corrective actions after notification of deficienc.ies.
Identified deficiencieshavebeenclosedoutsatisfactorily.E 87.
Intervenor's Exhibit 5 is a report dated September 1980, of an audit by Applicant of NSC on September 17-18, 1980, together with the related close-out documents.
This audit resulted in eight findings, four observations, and one coment.
Findings noted deficiencies in documenting indoctrination and technical training of project personnel;, establishing position descriptions; traiaing audit team leaders; conducting surveillances of project activities conducting internal audits during 1979; and certification of the QA Services Manager in.accordance with ANSI N45.2.6.
~
Observations noted among other things, that NSC had not performed a QA audit of Leckenby during 1980.
88.
In reference to the fii. ding that NSC did not conduct an internal audit in 1979 for the Dresden and Zion storage r'ack projects and to the observation that NSC had not audited Leckenby du ing 1980, testimony indicated that NSC completed an internal audit in September 1980, and in response to Applicant's finding developed an internal audit pim for the future.
Applicant further indicated that fabrication of Dresden racks at i
Leckenby began April 10, 1980, and that NSC audited Leckenby in October 1980, within six months of initiation. Applicant audited Leckenby on-March l
-61/ Intervenor's Ex. 4; Tr. 214, 280-86, 323-24; note:
followup documents for this audit are not part of the record in this proceeding.
,,, _, _ 1 1 1980, just prior to start of fabricat';n, with a followup audit in May 1980 and another audit in September 1980.
During this general period, April '- October 1980, NSC.did three surveillances and four insnections of
.Leckenby. While Applicant believes that N"",'s failure to conduct an audit during this period did not compromise quality of the racks, Applicant has taken steps to ensure timely audits by NSC in the future.SI 89.
Intervenor's Exhibit 6 is an audit report und associated close-:.ut i
documents reflecting an audit by Applicant of Leckenby on September 24-25, 1980.
The audit resulted in two findings, dealing with changes i the organization chart and traceability of purchased weld filler material, and four observations in regard to approval of suppliers, logging of QA-related documents, verification and documentation that qualified personnel and approved procedures were utilized in performing special processes, and documentation of welders' training.
Close-out documents indicate time.1y correctionofthesedeficiencies.SI 90.
Intervenor's Exhibit 7 concerns another audit by Apol:eant of Leck e by, dated March 1?, 1980, and the associated close-opt documents.
This aud'it resulted in one finding, dcMing with failure to transmit quartcrif reports on QA to the vice-president of Leckenby, and two g/ Tr. 261-70, 278-79, 317-23.
g/ Intervenor's Ex. 6; Tr. 286-89.
observ,s 7ns concerning lack of a training program for QA persor,inel and missing resumes of certain arsonnel. E 91.
Intervenor's Ex.10 is an NRC audit report of Lecke"# y conducted Marc..17- ??, b80.
The audit dealt with three of the eighteen criteria of
Twenty significant deviations of com itments were disclosed, leading the auditors to conclude that Leckenby had not implemented an effective QA program consistent with requirements of 10 CFR 50, / iendix B, and ANSI N-45.2 as contractually imposed by Leckenby customers.
Although this NRC audit disclosed many more deficiencies i
than did Applicar,t's audit of the previous week, Applicant's audit was limited to its own project; none of the deficiencies identified by the NRC audit related specifitally to the Diesder project.E R
Because of the NRC audit, Leckenby had Olympic Engineering.
i Corporation conduct r internal audit and recommend improvems.ts in Leckenby's QA program.
By reviewing Leckenby's file Applicant's auditors learned of the deficiencies identified by Olympic Engineering and assuredthattheywerecorrected.E i
93.
Brooks & Perkir.s conducted an internal audit on June 11, 1980. As is customary, Applicant's auditors reviewed this internal audit ~to assure 64/ Inte enor's Ex. 7, Tr. 289-90.
--65/ Intervenor's Ex.10; Tr. 332-37; Board Ex. 3 (Leckenby ;itract);
T:. 715-16.
-~66 'Tr. 291-98. The recomendations were contained in Intervenor's Exhibit 8, marked for identification but not admitted into evidence due to its hearsay nature.
35 -
i ti f t l
~
._m,,.sa.. s ac ory c ose-out of all deficiencies.
The findings identi.ffed in this audit were corrected and on inspection the quality of the end product was determined to be acepDble.6,7/
04.
The Board has reviewed the quality assurance documents introduced 1
by Intervenor.
The deficiendes reflected in those documents have been closedout.SI The documents reflect the system of inspections, surveillances, and audits which' assur as the quali;y of the completed racks.
95.
Intervenor (Proposed Finding 96) urged the Board to find inadequate the quality assurance and quality control procedures deiscribed by Applicant and Staff, based on the concern identified in 164 (supra) and on previous history of the Dresden quality assurance program.
The Board finds that the quality centrol and quality assurance programs of Applicant and its contractors cre adequate to assure that tube f.nd rack construction 'nd the boron-10 loading of the Boral in the tubc3 will meet specifications. The Bwrd also, however, takes note of Intervenor's concern.
We stress to the Applicant that' implicit in our finding that the current programs are adequate is our expectation that the written procedures will be implemented i
67/ Tr. 580-83.
The Brooks & Perkins internal audit was marked for identification (proprietary matter) as Intarvenor's Ex. 9, but was not introduced into evidence; Tr. 302..
68/ Tr. 323.
.m
-m m---
m.
m
.m
.. _ m.,f.uny and on a timely basis.
We request of the Staff that the Applicant be firmly held to strict adherence to the programs it has described.
D.
Transportation Damage 96.
Contention No. 3 states:
The Application does not demonstrate that rack and tube packaging, transportation, and receipt inspections are adequate to prevent and detect transportation damage.
97.
The packaging and transportation procedures initially used by Brooks & Perkins involved packing tubes o cardboard boxes (.4 tubes to ihe box) which were banded together onto wooden sk'ds (6 boxes per skid). These package units were' transported by closed trailers with two p Kkage units placed along eacn wall of the trailer with bracing between adjacent wood skids to prevent sliding.
This procedure proved unsatisfactory and resulted in a transportation incident in early August 1979.
Consequently, Brooks &
Perkins' transportation procedures were moditied in September 1979, to in rove the load configuration and to p rvide additional banding.
Transpor-v tation proceoures were again improved in January 1980, to provide additional load stability and' weather protection.
With this modified procedure, as of
~
mid-September 1980, a total of 1579 Dresden tubes had been transported in 9 l
f shipments with no known transportation damage.SI 1
98.
At Leckenby each shipment is inspected on receipt by Leckenby QC personnel to detect any transportation damage.
In connection with the
_69/ Pickens at pp.17-10; Shewski #. pp.10-11; Tr. 397-98.
l L
,__tra,nsportation incident cited supra, Leckenby and Appl'. cant inspected the boxes, removed those suspected of damage, identified the tubes therein and shipped the boxes back to Brooks & Perkins for reinspection.
Of the 80 tubes returned, only three required minor rework and these subsequently f
passed inspections.
The tubes which were not shipped back to Brooks &
Perkins wer? subjected to and passed the standard receiving inspection at
. Leckenby.EI 99.
For shipment of the fuel racks from Lecker? j to Dresden, special packaging, loading, tie-down, and bracing m'ethods are used, to prevent
' transportation damage and provide weather protection.
Shipments are made using dedicated tractor-trailer units with drivers who are instructed in the relevant precautions and shipment requirements.EI 1
100.
On their arrival at Dresden and prior to unloading, the racks are subjected to preliminary visual ~.:.spection by storeroom personnel to detect I
any transportation damage.
Storeroom personnel document their findings in a Receiving Inspection Notice which is forwarded to the Station QC n partment.72_/
e 10'l.
Applicant intends that QC personnel will in turn perform a Quality Receipt' Inspection which will include visual inspection of acccasible welds by a certified Level II instructor.
Documents accompanying 70/ Shewski at p.11; Pickens at pp.18-19.
A/ Shewski at p.12.
7_?/ Testimony of Ron Ragan (Ragan) at p. 2 following Tr. 412.
2
_.~
38 -
the racks will be reviewed to ensure conformance of the racks to all applicable specifications ann standards and to verify completion of all eq'i red weld examinations and chemical and physical tests.23/
f 102.
The written procedures for the Quality Receipt Inspection were being formulated at the time of the first hearing.
On September 11, 1981, they v.ere still being reviewed prior to final approval. They must mee; the requirementsofANSIStandardN45.2.2S/
103.
On completion of the receipt inspection the Receiving Inspection Notice and Quality Receipt Inspection will be sent for rev,iew and approval-to Applicant's Quality Assurance Department, which wh:Ir physically located at Dresden Station is independent of station management.
The racks cannot be released from storage for installation without the approval of Quality Assurance.25/
104.
Prior to a rack's installation-in a spent fuel storage pool, each storage location in the rack will be subjected to a drag test. 'A dummy fuel assembly having dimensions identical to those in use will be inserted and withdrawn from each storage location.
If the drag exceeds 50 pounds, indicating a physical defect.in the. contours of a tube, that storage locationwillbepluggedbywekdingstrapsacrossitstop. While preventing 73/ Ragan at pp. 2-3; Shewski at p. 9.
74/ Tr. 397; 405-08; 414-15; 1046-48; 1112-14.
75/ Ragan at p. 3.
J i
-e n
39 -
fuel insertion this plug will still allow circulatica of cooling water in that location. EI
'105.
At the time of the November 1980 hearing, four racks had arrived at Dresden.
Due to unavailability of written procedures for the Quality Receipt Inspection, quality re.eipt inspections did not take place.
Therefore, the racks were segregated and stored on 'he site by storeroom personnel in accordance with instructions of the Station "'iclear Engineering Department and pursuant to the written temporary hold area procedure applied to all safety-related equipment. Storeroom' personnel are responsible for periodic verification of the condition of the racks and their storage location.
The racks are "on-hold" by Quality Assurance and cannot be moved or used until t'ter a quality receipt inspection has been done and QA has release ' them.
No additional racks will arrive at Dresden Station prior to completion of the written q'Jality receipt inspection procedure.b 106.
Due to inadequate storage capacity at *;he Station the remaini.ig racks may not be shipped directly from Leckenby to Dresden.
Applicant is currently considering storage of the racks in a warehouse near the Station or near'Leckenby. k'hile there are no written storage procedures for interim off-site storage, such procedures are unnecessary since the racks would be subject to quality eceipt inspection at Dresden.7_8/
107.
Applicant has adequately documented those packaging, transportation, storage and receipt procedures which have been implemented 76/ Ragan at p. 3; Tr. 705.
6 f_7/ Ragan at pp. 5-6; Tr. 391-92, 394, 405, 408-09, 414-15, 578-80.
7 78/ Tr. 393-96.
e
t_o, prevent and detect transportation damage, and has described, corrective measures taken to improve these procedures. Applicant has stated that t')e final quality receipt inspection at Dresden Station will ensure that amaged racks not in conformance with specifications will not be installed in the spent fuel pools at the Station.
As set forth above, Applicant states that (a)
Quality receipt inspection procedures will be formulated and written (para. 101, supra);
(b)
No' additional racks will arrive at Dresden Station prior to completion of (a) above (para. 105, supra);
(c)
No racks, either those currently onsite or those awaiting shipment, will be released for installation until comp 13-tion of the quality receipt inspection and release by Quality Assurance (para 105, supra).
108.
Therefore, the Board finds tha't Applicant has demonstraced the adequacy of those procedures now in effect to prevent and detect transportation damage, and the Board further finds that added assurance of safe operation of the spent fuel pools is provided by proposed procedures to ensure that damaged racks not in conformance with specifications will not be installed in the spent fuel pools at Dresden Station.
m
.a.
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Co r ro s i o n
,109.
Contentica 7 asserts that:
The Application doet act adequately assess the possibility of general corrosion and galvan.ic corrosion in the racks, in that:
A.
The life expectancy of the Boral tubes is unsubstantiated.
B.
Swelling of the Boral in the tubes and its effect on removal of fuel assemblies have not been analyzed.
C.
The corrosion surveillance program will not assure detection of corrosion in the racks because the samples to be inspected will not be representative of the actual tubes in the racks, because the-sample environment will not represent pool conditions in and near the racks, and because the program does not require a dummy fuel test snortly before placement of fuel in each tube.
D.
There is no plan for steps to be taken should corrosion be discovered in the racks.
110.
Contention 8 states:
The Applicant should develop criteria for the racks defining when their use to store fuel would be proscribed.
These criteria should be the acceptable amount of' corrosion, limits on dimensional changes and strength tolerance.
111. As described earlier tne proposed storage racks consist of a vertical array of rectangular stainless steel tubes welded together at the corners'to form a checkerboard pattern.
Within each stainless steel tube are four neutron-absorbing Boral plates, one On each side.
On each side of each tube near the top, 's a 1/4-inch vent hole v.hich penetrates the inside stainless steel sheath.
In addition, the stainless steal inner and outer sheaths of each tube are not welded together at the botiam corners.
Therefore, water may enter the tube at the bottom and through the holes at the top and come into contact with the Boral.
Borai is a product manufactured by Brooks & Perkins, Inc.
It consists of boron carbide
. ""~"-(BjC) particles embedded in a matrix of commercially pure (1100) aluminum formed into a plate and clad with 1100 aluminum on both sides.2EI 112.
The life expectancy of the Boral plates and stainless steel tubes encapsulating the plates is in excess of forty years.
This assumes that the present.high water purity in the Dresden spent fuel pools is maintained.ES/
113.
Contamination of the pool water might occur through spillage or immersion a.f something containing chloride into the pt-f, waiter. A smali quantity of chloride ion would not have a discernable effect on the t
stainless steel.
However, as little as 1 or 2 ppm in the water actually in contact with the Boral might lead to the occasional formation of pits. Such pitting would not lead to the loss of a significant quantity of boron from the Boral.
The corrosion product, aluminum oxide, would hold the boron carbide in place.
The effect of other contaminants such as hydroxide ion or sulphate would be to cause the thickness of the aluminum oxide film on the surface of the Boral to increase.SA/
/9/ Applicant's Ex. 2; Licensing Report: Dresden Nuclear Power Plant Units ?
~~-
and 3 Spent Fuel Rack Modif: cation (Revision 5) at pp.
3-6; Testimony of J. E. Draley (Draley) at p. 2 following Tr. 341; Tr. 358, 466-67.
80/ Draley at pp. 3-5; Supplemental Testimony of John R. Weeks (Weeks) on Contention 7 and 8 at pp.1-2, following Tr. 434; Tr. 345-348, 372, 375-377, 440, 81/ Weeks, Attachment to Supplemental Testimony, BNL-NUREG-25582, " Corrosion
~~~
Consideration in the Use of BORAL in Spent Fuel Storage Pool Racks," at pp. 2-3, 4, 6, following Tr. 434; Tr. 347, 352-53, 372.
t
~
_ 43 -
~ ~ ~ ~ ~ 114.
The quality of the water in the pools is maintained t'hrough the operation of the Spent Fuel Pool Clean Up System filters and demineralizers which remove such contaminants.
The quality of the water is checked regularly. Thdre have not been extended periods of loss of water qual ity.8_2,/
115.
Three mechanisms could lead to swelling of the Boral within the tubes.
First, if the quality of the Boral is poor so that there is porosity, there could be a path for permeation of the core material by
[
water.
It would then be possible for this water to react with the alu'minum to 1,aduce hydrogen gas in quantities sufficient to expand the Boral in the form of an internal blister.
However, this swelling should be self-limiting, because expansion of the blister should doform the plate to allow release of the hydrogen pressure. Moreover, such swelling 'would be local in
~
nature, related to some unexpected defect in the Boral.
Because of good j
experience with commercial grade Boral, no swelling of this type is expected in the Dresden pools.El 116 A second mechanism would involve local corrosion, or pitting, i
induced by galvanic interaction between the aluminum cladding of the Boral I
and the stainless steel tubes. Because the solid corrosion product has a greater volume than the metal, local-swelling could result.
The extent of.
J t
I 82/ Testimony of Don Adam (Adam) at pp.2-7 following Tr. 550; Applicant's -
r
(
Ex.1, Table 3.7-1:
Draley at p. 9; Tr. 376-80, 706-8.
l 8_3/ Weeks ate p. 2; Draley at pp. '5-8, Attachment 5 ' llowing Tr. 341; Tr. 354-55.
i
_m,
,....m.,
- 44
""~'" 7aTvanic corrosion is limited by the poor conductivity of the water, tha poor electrical contact between the Boral and the stainless steel, and by the protective oxide films forming on both metals.
The degree of perfection of the oxidc determines the rate of corrosion.
If galvanic corrosion is not so limited, ' he maximum swelling of the Boral sheet would be 0.180 inch.
This was calculated by converting the entire thickness of the Boral plate to the aluminum corrosion product.
This amount of swelling should not interfere with the normal fuel a sembly within the proposed Dresden racks.
Such swelling would be local in nature.
The only mechai: ism'which would lead to such swelling iould be some unexpected defect in the Boral. b 117. A third mechanism for swelling of the Boral would be the
~
accuT.ulation of gas trapped between the Boral and the stainle:S steel. 'The gas would be a mixture of the air originally in the stainless steel tubes and hydrogen produced by the initial corrosion of aluminum when exposed to water.
This mechanism is believed to explain the swelling of some tubes in the spent fuel storage racks at the Monticello Plant in 1978.
It should not
-84/ Draley at pp. 6-8; Tr. 339-40, 350-58. However, Dr. Draley noted that such swelling, when combined with the phenomenon of fuel channel bowing, could lead to a possible impediment to insertion or withdrawal of a fuel assembly, depending on the location of the swelling.
The Board at Applicant's request agreed to continue the evidentiary hearing to allow further analysis of the possible clearance problem caused by fuel channel bowing. - Tr. 380-84.
,y g
y,,
. """'BTcifr~at Dresden, if the vent holes in the tubes provided to allo' such w
gas to e. scape remain unplugged.EI 118. The corrosion surveillance pr: gram proposed by Applicant includes the installation' of eighteen small test samples and two full-length vented tubes in each pool.
The samples will be representative of the materials in the tubes of the racks.
The samples will be inspected periodically evec forty years.
The sample environment will be that of the spent fuel conditions in or near the racks. The number of samples and planned schedule for examinatio.1 of the samples are adequate.5/
119. Damaging cocrosion processes that might be anticipated should be slow and gradual, developing over a number of years.
There shou ' be adegaate time after the corrosion process is discovered to make plans for repairing the corrosion damage or replacing tne. Om roded mat'eriai <'thout significant risk to the fuel being storec' or to the environment.
This assumes that no corrosive contaminants are put into the pool water in substantial quantity.
However, because of the routine analysis of pool water and the efficiency of the spent fuel pool clean-up system, such contaminants should not remain undetected in the pool water for long periods of time.EI
-85/ Draley at p. 8; Weeks at p. 2, and attached report BNL-NUREG-25582,
" Corrosion Consideration in the Use of Boral in Spent Fuel Storage Pool Racks"; Tr. 358-359, 372.
-86/ Draley at pp. 8-10 and Attachment 6, " Neutron Absorber Sampling Plan-In Pool"; Testimony of James D. Gilcrest (Gilcrest) at pp.1-5 follo' wing Tr. 447; Weeks at pp. 3-4; Tr. 342, 362-67, 370-71, 436-39, 459-62.
87,/ Weeks at p. 4; Dr9ey at p.10; Adam at p. 6; Tr. 367-68, 379-90, 439-42, 707-08.
4 46 -
-?20, Contention 8 claims that criteria should Le jeveloped for the
=-s racks which define then their use to store fuel would be proscribed.
Such criteria would need to be developed if the surveillance program at Dresden, in combination with surveillance programs or experience at othe. reactors, should show significant deterioration of such racks However, considering that" deterioration is i;ot anticipated, that a turvaillance program will be established, and that modes of deterioration are not expected to be rapid, 1
such criteria can ce formulated if a specific problem develops.
It is rot neccssary to define in advance the maximum possible damage that the racks could withstand from a range of hypothetical corrosion or other prob-lems.88I 121.
The Board finds that the life expectancy of the Boral plates and stainless steel tubes encapsulating the plates should be in excess of forty
' Farther, the Board finds that swelling of the Boral in the tubes has
/ ears.
been analyzed.
It is not anticipated to occur.
Thus, it should not affect-the renoval of fuel assemblies from the stor.aSe tubes.
122.
The Board finds that the proposed corrosion surveillance samples will be representati/.* of the m'aterials in the actual tubes in the proposed storage racks.
The sample ;nvironment will represent pool conditions in or near the racks.
The Board finds that the proposed surveillance program should be adequate to ensure detection of corresion in the storage racks.
---88/ Weeks at pp. 4-5; Draley at, p.10; Gilcrest at pp.1-5; Tr. 367-68, 440-41, 461-62.
. -- 123.
The Board finds that the corrosion processes that might be
" ~ '* "
anticipated ar
.xpected to be slow and gradual, developing over a nu..oer of yeaa.
Therefore, there should be adequate time af ter any corrosion process
~
is discovered to make plans for repairing the corrosion damage or replacing the corroded material without significant risk to the fuel being stored or to the environment.
124.
The Board finds that it ia not necessary to develop criteria proscribing the use of the racks in advance as a result of a range of hypothetical corrosion problems. As the modes of deterioration are not expected to be rapid, such criteria can be developed if and when a specific problem develops.
F.
Radioactive Waste Treatment, Radiation Monitoring, and Health and Safety Of Workers at Dresden Station.
125. Contention 1 reads:
The Application gives no os3urance that the radioactive waste treatment system for the speat fuel pools is adequate for the proposed increase in spent fuel storage capacity.
126.
Contention 4 reads:
Applicant has not provided adequate monitoring equipment in the spent fuel pool water to detect abnormal releases of radioactive materials from the increased numbers nt i
spent fuel bundles. Absence of such monitoring and elarms
.I could result in undue exposure to workers in excess of ALARA, specifically:
A.
There 's no description of m'onitoring devices, and therefo.e, no assurance exists that workers in each pool area will have adequate warning of possible hazardous conditions.
B.
The Applicant should demonstrate that the radiation monitoring equipment has adequate range and sensitivity to
48 -
indicate accurately the rates and magnitudes of radiation
- - - ~ ~
releases that could occur in the reracked pools.
127. Contention 5 reads:
There is no assurance that the health and safety of workers in the spent fuel pool areas will be adequately protected during rack removal and installation, in thac:
A.
The Application does not supply adequate information to assess the occupationa': radiation dosage to workers involved in removing and installing racks and rearrang-ing spent fuel in the pools, and to other workers who may be in the pool areas.
B.
There is no consideration of the occupational radiation hazards from accidents that may occur a; a' result of rack removal and installation, e.g., flooding of the pool area and water spraying on workers.
128.
.ne two systems which treat radioactive waste from the spent fuel pools are the fuel poo,1 cooling and cleanup system end the plant radioactive waste disposal system.
Dresden Units 2 and 3 each have an independent fuel pool cooling and cleanup system consisting of a closed loop in which water is pumped from the pool through a heat exchanger, then through a filter and demineralizer, then back to the pool.
The fuel pool cooling and cleanup systems are " full flow," i.e., all of the water that passes through the pump and heat exchanger also goes through the filter and d>mineralizer.
Each system is designed to filter an amount of water. equal to the spent fuel pooi volume every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The fuel pool cooling and cleanup systems can v
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+-be-connected so that Unit 2's filter and dec.ineralizer can be used to treat Unit 3 fuel pool water, and vice-versa.5/
129.
The pre-coat type filter, coated with clay-like filter aid, removes particulates from the pool' water.
Filter condition is assessed by monitoring and recording differential pressure across the filter; at 30 psid an alarm sounds to indicate.need for backwashinj and recoating the filter.
'C;'erational experience at Dresden Stat.on does not indicate that filter cnange frequency has increased with increased spent fuel storage.EI 130.
Soluble contaminants except tritium are removed in the fuel-pool demineralizer, whic'h has an efficiency of about 99.9% or 99.93% of ter one pass.
The resin in the demineralizer requires replacement after depletion of its ion exchange capacity. OP 'ating experience at Dresden Station indicates that required resin replacement frequency has remained at.once or twice,r:t year despite increasing numbers of spent fuel assemblies stored in the pool.N 131.
If storage of additional amount.s.of spent fuel in the pools caused increased radioactivity in the pool water, more frequent changes of filter aid material and demineilizer resins might be necessary. Hoadver, only slight increases -in overall quantity of radionuclides in the water are expected.
Introduction of radioactivity into the pool water is a function 89/ Supplement 4! testimony of Valentine Malafeew (Malafeew) at p.1, following Tr. 521; Tr. 530;. Tr. 537-38.
_90/ Adam at pp. 2-4 and Attachment 1.
-~~91/.'. dam at pp. 5-7; Tr. 707-12; NRC Staff Ex.1, Environmental Impact Appraisal, pp. 4-6.
i
. ---~of-leakage from stored spent fuei and of mixing of poolwater with' re'ctor coolant system water during refueling.
Since the proposed full reracking modification will not affect frequency or methcd of' refueling it will not increase the amount of impurities introduced into the pool from the reactor coolant.
Experience indicates that there is little radionuclide leakage from stored spent fuel after several months of cooling.
Consequently, the amount of radionuclides in the pool water due to leakage from the stored fuel is expecteci to increase less than linearly with the number of spent fuel assemblies stored in the pools and to be relatively mino'r.
For these reasms, the existing spent fuel pool cleanup systems should be adequate tne proposed full reracking modification,N and certainly for the for limited modification permitting installation of 5 racks.
132.
The proposed increase in spent fuel storage capacity is not expected to increase frequency of filter and resin replacements and thus 6he proposed full reracking modification is not expected to result in any significtnt increase in solid radwaste generation.
Nevertheless, as a
,:onservative estimate the NRC Staff assumed that two additional resin beds would have to be changed out each year for each unit.
This conservative assumption would result in an increase of about 720 cu. ft./yr. in solid l
l rajwaste shipp;d from Dresden, or en increase of less than 0.8% in the total amount of wastes shipped from Dresden Units 2 and 3.93/
92/ Malafeew at pp.1-3; Adam at p. 7; NRC Staff Ex.1, pp. a-b; Tr. 528, 538-41.
93/ Malafeew at p. 3; Adam at pp. 7-8; Tr. 562-67.
... l
~ ~ ~ ~~ ~ 133.
Contention 4 challenges the adequacy of radiation monitoring devices in and around the Dresden spent fuel pools. Ten separate area radiation monitors are locaed throughout the refueling floor which houses the spent fuel [ mis. The detect 6rs for these monitors ar9 of the Geiger-Mueller type, and their location, range, current trip setting, L..d alarm and meter readout locations were described in Applicant's testimony.
9 ase monitors would quickly warn workers of any increase in direct radiation levels, and they wnuld aise respond to increases in gaseous radioactive contaminants released from the pool water.
In ~ addition, a portable Continuous Air Monitor (" CAM") located on the refueling floor contains.a scintillation type detector with a local meter, recorder and alarm.
If airborne particulate or gaseous activity increase to a preset level, the local alarm will sound.
In addition, the CAM is used to collect particulate and iodine samples wtkh are removed to a counting room for daily analysis.
During refueling outages, additional CAM's may be placed on the refueli,*.g floor as recommended by the Dresden Station Rad-Chem Departmer t.
Finally, the reactor building ventilation monitoring system utilizes, in addition to four of the area monitors de, ribed above, four more monitors in' the reactor buildig ventilation ductworks.
Abnormal releases to the environment are prevented by switching the ventilation exhaust to the standby gas treatment system on alarms by either the area or ventilation monitors.
The vent duct monitors have a range of 0.01-100 mrem /hr, an alarm point of id. mrem /hr., and they alarm in the main control room. The vent duct monitors therefore also serve to protect workers by
. 4
~~'~'"n'otifying control room oersonnel of increased radiation in the fuel pool area.b 134 Th:.re are no radiation monitors which ccitinuously and directly measure radioa'ctivity concentrations in the p'ol water.
None are needed because the existing system of area radiation monitors and CAM's, as well as the portable monitoring instruments and personal monitocing devices described bclow, are adequate to detect radiation in the area around the pool and thereby to protect workers, who work around the pool, not in the 4
pool.
If a diver is needed, a continuous radiation monitor ~ will be lowered into the pool with him.EEI 135.
Contention 5 states that there is no assurance that the health and safety of the workers will be adequately protected during rack removal and installation.
In addition to the system of area monitbrs, ventilation monitors, and CAM's described ' bove, workers at Dresden Station are' a
protected from unsafe radiation exposure by numerous other measures including personnel monitoring, which involves the use of film badges, pocket dpsimeters, timekeeping in high radiation or airborne areas, and periodic whole-body counting and isotopic analysis to check for ingestion of 94/ Adam at pp.10-12.
---95/ Supplemental testimony of Seymour Block (Block), Tr. 600-02, 631-34; 639-40, following Tr. 638; S47.-49.
In Block's opinion, a water monitor might give a more prompt resptnse to any increase in radioactivity in spent fuel pool water than an area monitor or air monitor would, but this did not change his conclusion that there is no need for such a water monitor.
Tr. 649.
. "~~'7ad fo' is otopes.
Dose-rates and contamination levels in all work ' areas are routingly measured, and access to high radiation areas and airborne areas is controlled. Station procedures for control of occupational exposure conform to all federal standards.96/
136. The proposed full reracking will be accomplished through a step-wise procedure in whicn all the fuel stored in each pool will first be moved to the south end of the pool. The old rach at the north e.ed will be removed, the vacated pool floor will be vacuumed, the new racks will be installed, the neutron attenuation tests will be conducted f.o verify the presence of Boral, and the fuel will be placed in the new racks at the north end of the pool.
The process of moving fuel, removing old racks, vacuuming, installing and testing new racks will proceed north ~to south until all but I
six of the new racks are installed. These s% racks will be stored indoors at the Station to leave room for the control blade storage until additional fuel storaga space is needed. The racks will not be carried over stored spent f6el at any time, and this prohibition will be incorporated in the Technical Specifications accompanying the proposed license amend-ments.N The proposed modification limited to ' installation of five racks in the Unit 3 SFP will be accomplished similarly by removing 13 old racks at the northernmost end of the Unit 3 SFP and replacing them with i
5 new racks (il 26, supra).
2 137. The Applicant originally estimated the occupational exposure
,96/ Ragan, at p. 7.
E/ Ragan at p. 9; NRC Staff Ex.1, Safety Evaluation at pp. 5,10.
~
54 -
associated with the entire rack replacement operation to range from 18 to 47 man-rem.El which represents a small fraction of the total annual man-rem burden from occupational exposure it Dresden Station.
Installation of five racks followed by twenty-eight will result in an additional 1/2 man-rem.
Subsequent operation of the spent fuel pools with increased quantities of stored spent fuel assemblies will cause only a negligible (less than one per cent) increase in annual occupational doses. Although the Applicant does not have a formal written'ALARA program governing Station operations, the proposed spent fuel pool modification and su'sequen,t b
i operation of the pools will be performed in a manner that will maintain exposures as low as reasonably achievable (ALARA).EI 138.
For full reracking the two alternative methods of disposing of the old racks discussed in the NRC Staff's June 6,1980 Saf'ety Evaluation were cutting the old racks into small sections to significantly reduce the j-volume to be shipped to the burial site, or crating the racks whole to reduce the man-rem exposure.
The matter was left open in the safety l
evaluation, allowing the Applicant to make the choice between these alternatives based on actual measurements of dose rates when the racks are
-98/ NRC staff Ex.1, Sataty Evaluation at p.10, Environmental Impact-Appraisal at p.
7; Ragan at p. 8 and Attachment 2.
This estimate ~was based on boxing the existing racks.
--99/ HRC Staff Ex. 1, Safety Evaluation at pp. 10-11, Environmental Impact Appraisal at pp. 4-8; Block at pp. 3,5, Tr. 639, 644, 650-53, 656-57.
=. -.,,
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= m
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. removed from the pools.100/ Prior to the hearing Applicant pulled one of the old racks out of one of the Dresden pools and measured the dose rate.
Based on this, Applicant changed its estimate of the dose rate required to crate the racks whole for disposal from a maximum of 0.6 man-rem to a total of 5.67 man-rem.
It estimated the occupational exposure associated with shredding the racks and barreling the shreddings for disposal to be 14.7 man-rem, or an increase of about 9 man-rem over disposing of the racks whole.
Crating the old racks and burying them whole would cost about
$300,000.
Shredding the old racks and disposing of them in barrels would cost about $135,000; including the cost of the shredding machine.
Therefore, Applicant estimates a savings of $165,000 associated with the shredding alternative.
Moreover, shredding the old racks will reduce t' e
~ h
~
volume of waste thereby conserving low level waste Mrial site space.
Shredding the racks will also reduce from about thirty-five to seven the number of shipments of radioactive waste required, thereby decreasing the chance of transportation accidents.101/
139. The Appeal Board has observed, "The ALARA Standard contained in Part 20 is more easily stated th'an applied." Northern States Power Company (Prairie Island Nuclear Generating Plant, Units 1 and 2), Vermont Yankee Nuclear Power Corporation (Vermont Yankee Nuclear Station), ALAB-455, 7 NRC 100/ NRC Staff Ex.1, Safety Evaluation, at p.11.
101/ Tr. 551-55.
. 41,57(1978).
This case presents two technically feasible methods of disposing of the old racks.102/ Shredding, and barreling the racks for disposal instead of burying them whole results in 9 additional man-rem, but it also involves an ec6nomic savings of $165,000 and socioeconomic benefits associated with conserving burial ground space and minimizing shipments of radioacti n wastes on the public highways.
140. Altnougu 1.he proposed method of disposal results in slightly higher doses than the disposal of the racks intact, the Board is satisfied
'that the proposed method is nevertheless acceptable under the ALARA criterion embodied in 10 CFR Part 20, 141.
Contention 5 raises the possibility that occupational radiation hazards could arise during the rack replacement operations due to flooding of the pool area or spraying of water on workers. The Dresden pools have high water level alarms.
While it is possible to overflow a pool by adding water at a sufficiently high rate, this does not result in flooding on the refueling floor. Excess water beyond the capability of the skimmer surge tanks would flow into air intake vents located abat 3 inches above the high water level and cause low level contumination of the floors below.
An event of this type occurred in the Unit 3 fuel pcol on October 25, 1979 when an 102/ By affidavit of Robert F. Janecek, dated July 9, 1981, this Board was informed by Applicant of another potential alternative of disposition of the current Dresden Units 2 and 3 spent fuel storage racks, namely, relocation to Applicant's Quad Cities fluclear Station for installation in that Station's SFP's.
Additionally, a method of chemically cleaning the racks to remove surface contamination is being investigated.
If successful, the decontaminated racks may be sold as scrap.
Tr. 1023-25.
! equipment attendant tronee inadvertently opened a valve supplying condensate water to the pool. do apparent damage was caused and appropriate corrective actions were taken to preclude repetition. The incident was minor in nature and could not have resulted in serious conse,quences.
It is not possible to mistakenly open a wrong valve and drain a fuel pool.103/
)
142.
It is unlikely that any water could be sprayed on workers during the proposed reracking.
The racks v.ill be carried over the pool with the I
main overhead crane. system, previously reviewed and approved by the Staff as single failure proof.104/
It is possible that water could be sprayed on workers during hydro-lazing of the old racks after their removal from the pool.
During the reracking workers will be wearing protective clothing designed for the task. The dose rate if a worker's face were to be sprayed with Dresden pool water is estimated to be on the order of 10 tc 20 millirem per hour.
If this were to occur, a decontamination procedure would be executed. The efore, there would be no integrated dose of any consequence.105/
lt3. TheInstituteofNuct$rPowerOperations(INP0)recentlyissued 4
103/ Ragan, at p.11 and Attachment 4, at pp. 5-6.
104/ Supplemental testimony of Millard L. Wohl (Wohl) at p. 2, following l
Tr.
674; Pickens at o. 25.
105/ Tr. 649-50; Ragan, at pp.11-12. The dose estimate provided by Mr.
Block did not consider inhalation or ingestion of the pool water by the worker.
Tr. 654.
Report No. EA 80-01 dated September 12, 1980 entitled, " Evaluation of Dresden Nuclear Power Station. 106/ The INP0 audit team determined that, within the scope of their evaluation, the plant is being safely operated by an experienced, capable and dedicated staff.
- However, they noted opportunities for improvement in a number of areas, incluJing the Station's ALARA program.
In response to this recommendation, in May 1980, Applicant hired Scientific Applications, Inc. to develop a formal ALARA program for its nuclear stations. The project is divided into four sections:
evaluation of existing activities, recommendation of one or more ALARA organizations, implemention of the program, and testing and training of personnel. Applicant produced as witness the head of the Health Physics Program at the Dresden Station to responc to questions about this project.107/
As of November 21, 1980 the evaluation had been completed and recommendations for an ALARA organization made to Applicant's l
corporate office.
The ALARA organization would be a functional group fnr implementing the ALARA program at the Station. The ALARA program was due to i
f be implemented at the Dresden Station by December 31, 1980. Subsequent to l
l 106/ Intervenor's Ex. 12.
107/ George Arthur Myrick (Myrick); Tr. 609-19.
1 l
l
t '
that tire, as part of the final test phase, a formal training program in regard 'to that ALARA program will be developed and implemented.108/
3 144.
The INP0 audit team also recommended improvements in' training. in the maintenance, radiochemistry and technical. staff departments at the j
Dresden Station.
In response, Applicant has establ'ished responsibilities in
.each dep' rtment and initiated other actions to meet training needs.
In the a
j long term a reorganized Production Training Center scheduled for operation in late 1982 will have responsibility for the review and development of al1 training programs.
Standardized training programs are being developed
[.
over a two year period commencing January 1981.109/
145.
There presently is a health physics train.ag program, and workers receive training in accordance with 'O CFR Part 19 before they are allowed i
to go into radioactive materials areas. Applicant indicates that there will be detailed training of all workers involved prior to the fuel rack l
replacement operations, however, the procedures for this training had not.
yet been written at the time of the hearing.110/
14.
A recent appraisal by USNRC Office of Inspection and Enforcement,.
Region III, indicated that numerous weaknesses exist in the Dresden Station health physics program, including insufficient management support for l
l 108/ Intervenor's Ex. 12 at'pp. 2,'35-36; Tr. 615-16.
109/ Intervenor's Ex. 12 at pp. 14-15; Tr. 617-18.
~
j 110/ Tr. 616-20.
I
.- ~,...
m
.s
_ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ - _.-~.w.._..
professional health physicists, radiation chemistry technician training, access control, contamination control, abnormal condition surveillance, monitor operability surveillance and emergency response.
The NRC Staff's assessment indicated that the identified weaknesses required correction to enable Applicant to perform well in normal and abnnrmal conditions.
However, the present Dresden Station health physics program was considered adequate for continued operation while schieving acceptable. corrective action.
Staff's witness on Contention 5 had no knowledge of the I&E Health Physics Aopraisal.
Applicant has responded to each of the weaknesses in the NRC Health Physics Appraisal, including corrective steps taken or to be taken and schedules for com,letion.IIII 147.
The Board has carefully reviewed Intervenor's Exhibits 12,13A, and 138. While these documents show there are deficiencies in Dresden Station operations, particularly in the health physics area, they also show that the Applicant is tat.ing action to correct the deficiencies so identified, and that the NRC Office of Inspection and Enforcement is monitoring Applicant's performance in this regard.
It appears that thi new formal ALARA program is being implemented on schedule and will be in place by the time the racks are installed.
Similarly, while overall training 111/ Intervenor's Exhibit 13A (cove' letter); Intervenor's Ex. 13B; Tr.
644.
646.
.._,..~.~.s..-~
activities at the Station need improvement, the evidenus shows that workers involved in the rack replacement will receive adequate train-ing.112/
148. The Board finds that the Dresden spent fuel pool clean-up system and the radioactive waste Osposal system are adequate to support the '
proposed increase in spent fuel storage capacity with 5-racks or with the full 33-rack installation.
149. The Board finds that there is sufficient monitoring equipment with adequate range and sensitivity in the vicinity of the Drerden spent fuel pools to detect abnormal releases of radioactivity-from the pools as modified, to provide adequate warning to workers of hazardous conditions, ar.1 prevent undue cxposure to workers in excess of ALARA.
150.
The Board finds that there is renonable assurance that the health and safety of the workers will be protected and that occupational exasures will be as low as reasonably achievable during the proposed rack replacement arid subsequent operation of Dresden Station with increased quantities of tored spent fuel.
The Board also finds that adequate consideration has been given occupational radiation hazards due to accidents occuring during rack removal and installation, such as flooding of the.
pool area or water spraying on workers.113/
151.
The Board shares the concerns of Intervenor in regald to the
. health and safety of workers at Dresden Station and stresses that implicit 112/ Intervenor's Ex.12, at pp. 35-36; Tr. 612-20, 67.9-30.
113/ See also 1 141 and 142, supra.
h I
- 6?. -
l l
p
.-,..__,.w...in.1bese findings is~ the exiectation of the Board that the'ALARA progran f
will be rigorously implemented.
G.
Accident Analysis-l l
D 152.
Intervenor's Contention 6 asserts:
l The Application inadequately addresses the increased consequences of accidents considered in the FSAR, SER and FES associated.with l
the operating license review of Dresden Units 2 and 3 due to the j
increased number of spent fuel assemblies and additional amount of de ^r.tive fuel to be stored in the spent fcel pool as a result of l
tr difications.
l i
1 l
153.
The Final Safety Analysis Report (FSAR) was prepared in November l
l f
1967 and submitted as part of Applicant's operating license application.
Two Safety Evaluation heports (SER) for Dresden Units 2 and 3 cnd a single 5
Final Environmental Statement (FES) for both units were issued by the Atomic l
l Energy Comc.ission in.1969,1970 and 1973 to document the aperating license stage safety and environmental review.114/.
154.
Four design basis accidents were considered in the FSAR and SERs:
l control rod drop, main steam line break outsi V the drywell, loss of reactor coolant accident, and refueling accident.
Only the last accident is I
relevant to th ', proceeding.
155. A fuel. handling accident postulated to occur in the spent fuel pool would have consequences similar to those of the refueling accident.
{
The refueling accident considered in the FSAR and SER assumed the drop of a i
i 7 x 7 fuel assembly onto the reactor core from the maximum height allowed by l
l the refueling equipment (less than thirty feet) twenty-four hours af ter 114/ Pickens at pp. 26-27.
l
[
i i
T.;;x 2.f
. 63 -
m.......
., ~.- reactor shutdown.
dsing a kinetic energy analysis, it was concluded in t, hee"'
p> v FSAR that ninety-two fuel rods could be perforated.
However, even assuming 445 fuel rod failures, the radiological effects of the accident were calculated to be a small fractiSn o~fj0'YNPaY100 limits.
The Staff, utilizing more conservative assumptions, also concluded that the doses would' remain well below 10 CFR Part 100 guidelines.115/
156.
The FES prepared,.at the1'perating license stage covered a number of fuel handling accidents involvfag 7 x 7 fr ' assemblies in the reactor core and in the pools.
These used assumptions defined in former 10 CFR,Part 50,' Appendix 0.
The' accidents considered were:
fuel bundle drop (in core) heavy object drop onto fuel in core fuel assembly drop in fuel rack and heavy object drop onto fuel rack.
In each case the consequences were calculated to be a small ' fraction of regulatory limits.116/
158.'
The proposed spent fuel poc1 modifications will not change the manner orsfrequency of refueling.
Therefore, the probability of accidents involving dropping fuel or heavy objects on,to the reactor core, as discussed in the FSAR, SER and FES, is not increased. However, rack replacement 3
involves about 3700 additional fuel movements in the two storage pools 'ct contemplated when Drescn Units 2 and 3 were licensed.
This is an increase of about 5-6% over the 66,000 fuel moves anticipated over the lifetime of 115/ Pickens at p. 27, Attachments 4, 5 and 6; Supplemental Testimony of Millard L. Wohl (Wohl) on Contention 6 at pp.1-2.
116/ Pickens at pp. 20-27, Attachment 1 at p. 7-4, and Attachment 3.
10 CFR Part 50, '.ppendix D was revoked effective July 18, 1974, and replaced by 10 CFR Part 51.
l m
.,-.--- both. units. The number of additional fuel movements is increased by about 650 for Applicant's five rack proposal which represents a further increase of 1% of the total number of' fuel moves anticipated over the lifetime of Dresden Station.
A corresponding slight increase in the probability of a fuel assembly drop in one of ti.c? pools will result. However, the fuel to be moved during the proposed rack replacement wil' have been stored for a period longer than the decay period assumed in the FES for the fuel assembly drop in fuel rack accident.
Not granting the proposed amendments might also i
r;;11t in an increased number of fu'el moves ove-the lifetime of the Station. Applicant then would probably.have to shift fuel among the Dresden pools to prolong Station operation pending availability of an away from reactor (AFR) facility.117/
158.
The proposed rack replacement wil1 not alter the fuel authorize'd to be used in the Dresden reactors.
Therefore, it will not affect the consequences of the fuel drop accidents considered in the FSAR, SERs, or FES.
However, in 1974 the HRC a'uthorized the Applicant to use 8 x 8 fuel also at Dresden Units 2 and 3.
Tnere are +' a same number of fuel assemblies in the reactor core when 8 x 8 fuel assemblies are used as when 7 x 7 feel assemblies are used.
The average 8 x 8 fJel assembly operates at the same power level as a 7 x 7 fuel assembly resulting in the same average activity per fuel assembly.
Therefore, the average activity per fuel rod in an 9 x 8 l
c 117/ Pickens at pp. 21-23; Wohl at sp. 2-3; Janecek Five Rack Testimony at
- p. 8.
1 -..-assembly is less than in a 7 x 7 fuel assembly.
The refueling accident described in the FSAR and SER was reanalyzed in 1974 for 8 x 8 fuel.
It was found that the consequences of 8 x 8 fuel are less than for 7 x 7 fuel. The accidents described in the FES have not been specifically reanalyzed for 8 x 8 fuel.
However, because the activit'; per rod in 8 x 8 fuel assemblics is less, the consequences of these accidents should also be less.118/
159.
The.yplicant also considered the structural effects of a dropped fuel assembly hitting one of the new storage racks. This analysis showed that the struck storage rack would withstand the damage and stT maintain K
below0.95.119_/
eff 160.
The main overhead crane system will be used to move the fuel racks during the replacement operation.
This crane was approved for up to.
100 ton loads by the NRC Staff in 1976.
The crane meets the intent of.
NUREG-0554, entitled " Single Failure Proof Cranes for Nuclear Power Plants."
During the proposed rack replacement administrative controls and technical specifications will be implemented to prevent, the racks from being carr.ed ove:r stored fuel.120/
161.
I' a storage rack is dropped in one of the spent fuel icols during the proposed replacement operation, the pool liner might be torn.
9 118/ Pickens at pp. 23-28; Wohl at p. 3; Tr. 689-700.
119 Pickens at p. 22 and Attachment 3; Applicant's Exhibit 1, Section 3.4.3.5; Tr. 453-54.
120/ Pickens at pp. 25-26; Wohl at p. 2; NRC Staff Ex.1, Safety Evaluation at p. 10, Environmental Impact Appraisal at pp. 8-9; Adam ; Tr. 665-66, 674-77.
However, the concrete and steel structure of the pool should nct suffer significant damage. W9ter le.uing through the liner would be collected by drainage troughs leading to the reactor building floor drain system. There are four outlets from each fuel pool, enh of which is valved closed.
Therefore, no pool water should escape to the environment.121/
162.
There asa three different methods by wh'ch make-up water can be put into the storage pools, other than by using hoses.
A manual valve at the pool can be used.
This valve is normally used to make up for evapor-oration losses. A second method is provided by a six inch line from the concensate storage system. This line joins the spent fuel pool cooling and cleanup system at the pumps downstream of the heat exchangers. Tne spent fuel pool pum?s and heat exchangers are located about two floors from the stwage pool floor.
They would be accessible in the event of high airborne activity or high contaminatica levele ca the storage pool floor, but not in the event of extremely high radiation levels on every floor of the reactor building.
Further, a manual valve in the radwaste facility can be used to tie the p!rnt condensate water system into the fuel pools directly. L,e radwaste f acility is on the far side of the turbine building from the reactor Wilding.
Thus, the manual valve should be accessible ader any accident circumstance. There are large supplies of make-up water at the Dresden O ation.122/
_1_21/ Pickens at pp. 23-26; Tr. 658-60.
2 122/ Tr. 588-95.
-._.,_._.. J.6 3.
The spen, fuel storage pools are Class 1 seismic structures.
They are designed to withstand the Operating Basis Earthquake (0BE) and Pfe Shutdown Earthquake (SSE) defined for Dresden Units 2 and 3.
Each pool was analyzed individually. The structural analysis did not consider any other event occurring at the same time as the seismic event.
The new storage racks are designed to withstand these seismic loadings. The existing structure of the spent fuel pools is adequate to withstand the additional loads due ta five storage racks. Therefore, the consequences of the occurrence of the Dresden OBE and SSE earthquakes would not be increased by.
the' proposed installation of five rac! 1.
Resolution of whether the existing structures are adequate to withstand the additional loads of 33 racks during the SSE must await further analysis. Similarly, the reactor building is designed to withstand thc impact of tornado-driven missiles.
Because the installation of new storage racks will not require structural modification to the reactor building, the consequences of the design-Lcsis tornado will not be increased.123/
164.
The Board finds that the consequences of the accidents considered 1
in the FSAR, SER and FES associated with the operating license review of Dresden Units 2 and 3 will not be increased as a result of issuance of the proposed license amendment permitting installation of five high density storage racks in the Unit 3 SFP.
123/ Pickens d. pp. 28-29; Applicant's Ex.1, Sections 3.4 and 3.5 NRC Staff Ex. 1, STfety Evaluation Analysis at pp. 6-9; Affidavit of Robert F.
Janecek correcting inconsistencies in Applicant's former testimony, of May 6,1981; Tr. 661-62.
Staff Ex. 2 at p. 4.
. f H.
Fuel Channel Dt.ormations 165. A fuel assembly for a mailing water reactor has two major components, a fuel bundle and a fuel channel.
The fuel bundles presently used at Dresden Units 2 ar.d 5 have 64 rcds in an 8 x 8 array.
The rod; are held in position by an upper tie plate, a lower tie plate, and seven grid 124/
spacers.
166. A typical fuel channel is a scaare with an inside diameter of 5.278 inches, a wall thickness of 0.080 inches and_a length of about 13 1/2 feet. The' fuel channel is placed over the bu' dle of fuel rods.
It n
completely surrounds the array of fuel. rods on the four latera!_ sides. The channel is attached to the fuel bundle at one corner of the upper tie plate assembly by a channel fastener bolt.
The dry weight of a fuel assembly N approximately 680 lbs., including ti:e channel which weighs approximately tl i
lbs.125/
167. When a fuel assembly is irradiated in the reactor, normal hydraulic pressure gradients and neutron flux gradients cause the dimensions fuel of the channel to changa from the original dimensions. The potential safety concern raised by Applicant in November 1980 was that these channel 124/ There are also fuel bundles. stored in the Dresden pools with 49 fuel rods in a 7 x 7 array.
Testimony of Carl.R. Mefford, (Mefford)
Related to Cr.pability of Fuel Assemblier. to Accommodate Loads Applied During Insertion and Removal from Spent-Fuel Storage Pot
, at p.
2, following Tr. 1013.
125/ Testimony on Dimensional Changes of BWR Fuel Channels as a Result of Irradiation and o hon-GE Fuel Bundles and Channels by Dennis 0'Boyle (0'Boyle) at pp. 2-4 and figures 1-4, following Tr. 1013; Tr. 738-41.
.m
. m.
.m
._,,,.._=
..de. formations might be large enough to affect tif clearances between some fuel channels and the calls of storage locations ir, the proposed racks.126/
168. An analysis was made of. the worst-case combinations of the largest measured channel deformations and the minimum size storage locations in the proposed racks.
P is included all allowable manufacturing tolerances for the proposed racks compounded in the most adverse way.
The analysis showed that there would be a potential for a maximum interference of about 1/4 inch. 27f 169.
If this interference occurs, it would' cause the channel to rub against the wall of the storage tube during insertion, storage, and removal of the fuel assembly. Ap.'sicant addressed the loads which this rubbing could impose on the channe. There would be no damage'to the fuel l
cha1nel, the fuel rots or the proposed racks.
This is true even if the maximum 1/4 inch 'nterference occurs.I 170.
The drag loads for the maximum interference case are not sufficient to cause a fuel assembly to become stuck in the proposed racks.
126/ 0'Boyle at pp. 1-2.
127/ Testimony of James D. Gilcrest Related to Fuel Channel Bowing dated January 16, 1981 (Cilcrest supplemental testimony) at p. 4, following Tr.
1013.
128/ Gilcrest supplemental testimony at p. 7-8; ha.f tord at p. ?-4.
y-
-e e
2 3
-f a
w
_., _. Joth the Applicant and the Staff testified that it would not be a safety problem, even if the maximum or worst case interference occurs.129/
171.
The three modes of reactor-induced fuel channel deformation are twist, side-wall bulging, and longitudinal bowing.
Measurements and analysis indicate that the amount of channel twist is small and does not significantly affect the clearance between the fuel channel and the fuel storage rack.130/
172. Bulging of the side-wall of a channel occurs as a result of a pressure differential in the coolant across the channel wall.
This can produce a slight outward displacement of the fo': sides of the channel.
The outward displacement of the walls (bulging) is usually less than 0.060
. inches. This is small compared to the overall cross-sectional dimension of the channel.
Therefore, the-channel remains essentially square.
It is possible to get bulges greater than 0.060 inches in Dresden-2 channels.
Side-wall bulge is largest about 5 to 6 feet from the bottom of the channel and the magnitude of the bulge decreases toward.both the top and the bottom of the fuel channel.131/
~
129/ Sur06emental Testimony of Ronald M. Ragan (Ragan supplemental testimony) at p. 3-4 following Tr.1013; Supplemental Testimony of Horace K. Shaw on Fuel Channel Bowing ',Shaw) at p. 4, following Tr.
1013; Tr. 956-59, 989.
130/ 0'Gcyle at p. 5; Tr. 757-62, 812-13, 819 20.
131/ 0'Boyle at p. 5; Tr. 747-49, 781.
_.__m_._.
, 1
-__ m 173. Fuel channel bowing results from fast neutron flux gradients-across the walls of a channel when the fuel assembly is placed in certain core locations.
This can cause a displacement of the mid-elevation of the channel with. respect to the upper and lower ends of the channel.
The largest channel bowing ;.:nerally occurs when channels reside for. several cycles of reactor operation in locations near the pe ', Sery of the core.
where neutron flux gradients are hi hest.1-
^
9 174.
Between July and November of 1980 Applicant measured the dimensions of 875 irradiated channels at its Quad Cities Nuclear Power j
' Plant The Quad Cities reactors are BWR/3 reactors like Dresden' Units-2 and.
?-
The fuel bundles and fuel ~ channels used at Quad Cities and Dresden are i
similar.
The Quad Cities measurements were made to detarmine whether the
~
in-core life of fuel channels could be extended without 1eading to channel l
deformatiois so large as to cause interference between channels and reactor control blades.
175. As these measurements at Quad Citie, were made, it was recognized.
that the combination of fuel channel bow and bulge for some channels was greater than the minimum clearance dimensions for storage.locationi allowed l~
by'the engineering drawings for the-proposed Dresden storage racks.
176. After disclosing the potential problem to the Board in December-1980, the Applicant analyzu a total of 1736 channel sides measured h Quad
(
Cities in order to evaluate tha fit of the: irradiated channels in the proposed Dresden storage racks. Appr 3ximately 86% of the channel sides / had
!~
a total deformation (bow plus bulge) o' less than 0.150 inches over the i
132/ 0'Boyle at pp. 5-6; Shaw at p. 2; Tr 7:^.
i I
..,-.-m.,
.. -.,_...,.-.. - -,, _. _, -. _,., - -. _,, - -., -. _. - - - -, _, -. ~, -
,... 162.2 inch length of the channel. A bow plus bulge deformation of less than 0.%0 inches was found in 94.5% of the channel sides measures. Less. than 1%
(15) of the surfaces measured had a total bow plus bulge deformation of greater than 0.300 inches. Only two channel surfaces had a total deformation exceeding 0.350 inches.
The maximum bow plus bulge was 0.420 inches. This was found in only one channel side.
The next largest bow plus oulge was 0.390 inches.
This was measured again on only one channel side.133/
177.
The minimum clearance between a straight, unirradiated fuel channel and the w&1l of any storage position in the proposed Dresden spent fuel storage racks is 0.346 inches total.
The clearance is 0.173 inches on each side of a stored channel, assuming the channel is centered in the storage location.
This clearance exists in the inter-tube storage posit' ions.
The corresponding clearance inside a storage tube location is
.ches total or 0.248 inches on each side).134/
greater (0.496 178.
The " worst case" situation would occur if a Dresden fuel channel with bow plus bulge equal to the maxiTum value measured at Quad Cities (0."20 inches) was placed in a, storage location in the Dresden racks with 133/ 0'Boyle at p. 6-9; Snaw at p. 2-3; The charnel with bow plus bulge of 0.420 inch had gone through five reactor cycles, four of them on the periphery of the core. Tr. 747, 752, 774-78, 809-12.
134/ Gilcrest supplemental testimony at pp. 3-4; Tr. 739-40.
The storage racks are a checkerboard pattern of stainless steel tubes containing Boral.
Fuel assemblies can be stored inside the tubes ana in the inter-tube locations. Applicant's Ex. 2.
-y.
,e
.~.--. -.....the minimum allowable clearance of 0.173 inches.
'ne resulting interfere, c e would be approximately 0.25 inches.135/
179.
Applicant initialll described another potential ir.urference, unrelated to. fuel channel deformation, which might exist at the top of ~ the inter-tube storage positions between 1.N channel spacer button at the top of thu fuel channel and the lead-in clips which create the minimum dimension of each storage location. Applicant has committed itself to checking each storage location in each rack with a plug gauge prior to installation in the pools to ensure that the dimension between the lead-in clips is no less than the maximum dimension of the channel at the spacer button, 5.76L inches.
N necessary, Applicant will grind down the lead-in clips to achieve this dimension, thereby eliminating any interference at the top of the storage location.136/
180.
The load required to remove a fuel assembly from the proposed racks would be composed of the drag due to such interference and the dead weight nf the fuel.
The drag force to overcome the worst case interference of 0.25 inches was calculated to be 310 pounds.
This assumed a conservative coefficient of friction at 0.5.
Therefore, the maximum load which would be necessary to remove a fuel as:embly from one of the proposed Dresden storage racks would be 990 pounds. T5is would include the weight of the fuel 135/ Gilcrest suplemental testimony at p. 4.
136/ Gilcrest supplemental testimony at p. 3, and figures 1 and 2; Tr. 736, 796-97, 888-89, 920, 923, 944-47.
I 4
L,
assembly (680 pounds ignoring a buoyancy force d 80 pounds) plue. the drag
^ - - -
due to channel bow interference (310 pounds).137/
181.
The maximum li't that the Dresden fuel grapples can exert in the fuel assembly is limited to 1100 pounds by an electrical intericd.1_38/
182. / 'ifting load of 990 pour.ds would not damage the fuel assembly or the proposed storage racks. The only components of the fuel assembly which would undergo significant loading changes w..,'d be the upper tie plate lifting bail and the channel corner gussat.
The lifting force exerted by the crane grapple is transmitted to the fuel assembly by the lifting bail.
The design load of the upper tie plate lifting bail is 2040 pounds.
The actual load at which the lifting bail would fail is much greater. Of. the 990 pounds applied to L.ie bail by the grapple, only 374 pounds (the d-ag force of 310 pou.1ds and the channel weight of 64 pounds) would be A
transmitted to the fuel cheanel through the channel corner gusset.
General Electric performed a test showirg the d,eformation of the channel corner gusset, as essentially elastic up to 3240 pounds and did not faii up to 4080 pounds.139/
137/ Gilcrew, supplementui testimony at p. 5-7; Shaw at p. 3; Ragan supplemental testimony at 1; Tr. 546-50, 957-58; 979.
138/ Ragan, supplemental testimony at p. 2; Tr. 887-88, 907-09.
~
139/ Mefford at pp. 3-4; Gilcrest supaiemental test.nony at p. 7; Tr. 866-67, 949.
J
._ }83.
The drag force of 310 pounds would also be transmitted to the dffect storage rack. This would not tip the 18,000 pound rack or exceed allowable stresses as defined in the U.S. NRC Standard Review Plan 3.8.4.140/
184.
Insertion of a fuel assembly under the worst case 1/4 inch interference would be resisted by a drag force of 310 pounds.
The fuel assembly weight exceeds this drag force. Therefore. 'ne fuel assembly would insert fully into the iack by its own weight. The only fuel assem51y component which would be loaded during insertion is the thannel fastener g
bolt.
The load on this bolt would be 246 pounds (the drag force rf 310 n,nds minus the channel weight of 64 pounds). The 246 pounds plus the d
tensile load of 1280 pounds, produced when the bolt is tightened, is less than the certified breaking load of the channel f astener bolt (3150 -
pounds).141/
A number of conservative assumptions were made in analyzing the 185.
channel deforiaation issue including the following:
a maximum channei bow plus bulge of 0.420 inches was assumed (a) ' to occur, even though
- new fuel channels with improved heat treatment and 140/ Gilcrest supplemental testimony at p. 8; Shaw at p. 3; Tr. 933-34, o62.
141/ During normal insertion of fuel assemblies in storage racks at Cresden, an additional 500 pound weight of telescoping cans, which provide rigidity to the fuel gapple hebt, rests momentarily on.the This would q
upper tie plate af the fuel assembly being inserted.
further assure full insertion of the fuel assembly.
Gilcrest supplemental testimony at p. 8; Mefford at p. 4; Tr. 867, 878-79, 905-09.
j
76'-
f abrication processes are being used at Dresden.142/
- fuel cha.'.nel measurements are underway at Dresden which will be used to prevent reuse of channels having bow plus bulge greater than 0.125 inches.143/
- in-core locations for fuel channels will be selected such that bowing is not compounded by multi-cycle irradiation on the core periphery.144/
(b) c11 manufacturing. tclar ences for the proposed racks were assumed to combine in the most adverse way to result in the minimum allowable storage dimensions.145/
(c) the channel with the maximum bow plus bulge measured at Quad Cities was assumed to be placed in the minimum sized storage location.146/
(d) the fuel assemblies were assumed to be centered in the storage location, although the top of the fuel assemblies is free to tip away from the direction of bow reducing the possible interference.147/
(e) a coefficient of 'rittion.of 0.5 was assumed although a more realistic value for Zircaloy against steel may be more nearly 0,15.148/
(f) a fuel assembly dry weight of 680 pounds was assuated rather than the submersed weight of approximately 500 pounds.149/
142/ 0'Boyle at p.10; Tr. 781-82, iftcamera 790, 792.
143/ 0'Boyle at pp.10-11; -Tr. 798, 987, 992.
144/ 0'Boyle at pp. 10-11; Tr. 790,198.
145/ Gilcrest supplemental testimony at p. 4; Mefford at p.1.
146/ Gilcrest supplemental testimony at pp. 4, 8.
\\
147/ Gilcrest rupplemertal testimony at pp. 2,4.
148/ Gilcrest supplemental testimony at p. 5; Tr. 957-60, 978-79.
149/ Gilcrest supplemental testimony at p. 8; Ragan supplemental testimony at p.1; Tr. 864, 958.
s
-n e
y y, t y
w v 186.
The ana' lysis did not include the effects from the formation of blisters in the stcrage tubes due to hydrogen gas bubbles because it is cor.sidered unlikely to occur. The effect'of creases formed in the Boral during fabric.ition were not included in the analysis. Such creases occur near the end of the charnels, which is not the area of bowing. The creases will not create interference Cth the storage tube lead-in clips.150/
187. The analysis of fuel assembly design and channel dimensional measuremeite, applies to fuel bundles and channels supplied,by the General Electric Company. Exxon Fuel and associated fuel channels from Carpenter Technology Corporation ("CarTech") purchased in 1970 which may be used at Oresden in the future, need not be considered in this partial _ initial decision dealing wita racks fcr the current fuel.151/ -
188.
Should it be necessary to place Exxon-type fuel assemblies in these racks in the future, the materials used in the upper tie plates of the General Electric anc Enon fuel bundles are nearly identical and the design; are similar. The materials and dimensions of the General Electric and CarTech channels are nearly identical. The loads that the new fuel 1
150/ Tr. 930-32.
151/ 0'Boyle at p.12; Tr. 781; in camera Tr. 790, 792.
n
I.
components can withstand are not significantly different than similar fuel components su, pl'ied by General Electric Compny.152/
189.
The effect of manufacturing tolerances in the CarTech and Gene,al F.iectric fuel channelsbas addressed. This included a brief in camera
~
session on certain proprietary information.
Any increase in potential t
i'nterference between the. fuel channels and the storage racks, and therefore any increase in the loads imposed on the fuel assemblies and racks, due to these tolerances will be negligible.153/
19 0.
Applicant's witness said that he expects galvanic corrosion between the Boral and the stainless steel walls of the storage tubes to be quite limited.
The limitation would be due to the low conductivity of the pool water and by the naturally occurring oxide films on the Boral and on the stainless stcel.
If galvanic corrosion were not so limited so that the entire thickness of the Boral was converted to corrosion products, the tube wall could swell by a maximum of 0.180 inches.
This swelling'would affect the stnrage locations within the proposed storage tubes, rather than the inter-tube locations.
The ' inner-stainless steel walls of the storage tubes are thinner than the cuter walls. Thus the inner walls would tend to bulge more readily from the corrosion product.
Such swelling would be localized.
This swelling, in combination with channel bow and bulge, might present a possible impediment to insertion or withdrawal of a fuel assembly.
Although 152/ 0'Boyle at pp.12-13;A camera Tr. 784-87; Tr. 950-54.
153/ Gilcrest, supplemental testimony at p. 9; Tr. 735-36; in camera Tr. 784-92; Tr. 800-01; 803-05; 807-09; 929-30, 950-54, 957.
I A
a the witnass considered this highly improbable he recommended periodic mandrel testing of unfilled storage tubes in the proposed racks.154/
191.
Subsequently the Applicant informed the Board that it had determined not to accept the recomendation for mandrel testing. A maximum ir.terference of 0.352 inches would occur if 0.180 inches of localized swelling 'in a storage tube wall occurred opposite a channel having a maximum bow plus bulge of 0.420 inches. To remove a fuel assembly with this-much interference would require a force of 436 pounds.
A maximum fuel assembly handling bail load of 1116 pounds would be encountered.
Thi.s is well within the capability of the affected fuel components.155/
192.
A properly designed mandrel test would help determine if a particular storage location could accomodate a bowed fuel assembly.
Such tests would probably not cost very much.
However, to con @!;t mandrel testing would require three men verking above tha pools for at least 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> per year.
Exposure rates of about 3 to 5 millirem per hour.*ald be encountered. Although the resulting occupational exposures would be small, they vould not be as low as reasonably achievable (ALARA).
The Staff said exposure of workers during testing is not justified.156/
193.
Channelled fuel assemblies are'not expected to become stuck in the proposed storage racks., However, such an event was revi2wed.
A stuck fuel assembly would not be a scfety problem unless efforts to free the 154/ Draley, testimony at pp. 7-8, Tr. 353-57, 372-73, 377-78.
155/ Ragan, sue?lemental testimony at op. 4-5, Tr. 954-57.
156/ Tr. 893, 898, 901-03, 911-13, 990-92.
assembly led to perforation of the fuel rods and a release of radioactivitys To ensure that excessive loads are not imposed on a stuck asy mbly, Applicant's Dresden Fuel Handling Procedures are being revised.
The revised proceduras will provide that, if the 1100 pound fuel grapple interlock trips as a fuel handler is attempting io lift an assembly out of the racks, he will call for the assistance of the licensed fuel handling foreman.
The foreman would notify station management and obtain any technical support needed. The grapple interlock can be raised to 1500 pounds by a station electrician if edditicnal lifting force is required by special circumstances approved by station management. There is no way to force a partially inserted fuel assembly down into the storage location.
Once the affected fuel assembly is removed, it could be inserted in another, larger storage location in the proposed racks, or the fuel assemoly could be dechannelled and the fuel bundle stored separately from its channel in the racks.
A-channel which would not fit in any storage location could be stored separately fre:.. the pool beside the racks, with no effect on criticality.157/
l 194. The Board finds that the possibility of reactor-induced fuel channel deformations does not pose significant health and safe'ty problems for the proposed spent-fuel modifications for Dresden Units 2 and 3.
f 195.
The Board finds that even the worst credible interferences that might occur between the fuel channel with the largest reactor-induced l
l 157/ Ragan. supplemental testimony at pp.1-2; Tr. 842-43, 868, 886, 890, 907-09, 989.
l l
deformations (bulge, bow and twist) and the minimum sized storage location in the proposed Dresden fuel storage racks will not lead to damage of the fuel assembly or the pron 6 sed racks.
196.
The testing of storage locations with a full scale mandrel would be a feasible and conservative test.
However, the Board finds that such test 4.g is not required to be reasonably assured that fuel channel
' deformations will not pose a health and safety problem with the proposed fuel storage racks.
I.
Environmental Impact Appraisal And Safety Evaluation 197. The Staff's Safety Evaluation Report (SER), Environmental Impact Appraisal (EIA) and affadavit of Walter Brooks amending the SER were 158/
received into evidence as Staff Exhibit 1.
198.
Intervenor in its proposed findings of fact (PF 55-61) submitted that the SER and EIA should be given no weight because in Intervenor's view the Staff did not submit for cross-examination witnesses knowledgeable, able, or qualified to testify in regard to these documents.
- Further, Intervenor faulted the Staff for not modifying the SER after June 6, 1980 though the Staff subsequently received information regarding changes in rack design, including a supplement to the design report on October 29, 1980 and information on fuel channel assembly bowing on or about November 7, 1980.159/ In Intervenor's view, the Staff'should have presented for 158/ Tr. 117-18.
159/ Tr. 151-57.
cross-examination the Staff personnel who actually authored each sect'.i of the SER and EIA (Intervenor's PF 57-59).160/
199.
The EI A and SER were received in evidence after testimony by Paul O'Connor, the Staff project manager, who testified that in his role he knew the Staff members who prepared the documents, and had interacted with the reviewers of the documents. Upon cross-examination he named the reviewers who prepared the various sections and stated that in his capacity as project manager and sponsor of these documents he adopted the summary and conclusions set forth in the SER as his own.161/ Absent' requests from parties or the Board, the usual Staff practice is to have the project manager present the SER and EIA rather than have all the project personnel available for cross-examination.
200. Mr. O'Connor was available for cross-examination on relevant portions of the document, as were the witnerres presented by the Staff on each admitted contention and board questions. No requests were made during tha evidentiary hearing for additional witnesses who had been identified as l
Staff reviewers by Mr. O'Connor.
Acc~rdingly, the objections which o
Intervenor poses in regard to the sponsorship of the EIA and SER, after the c'osing of this portion of the evidentiarj record,- are lacking in merit.
201.
In regard to the current status of the SER the Board finds that the SER should be updated to reflect the information received between Ju,ne 6, 1980 and tha close of the evidentiary record, i.e., after receipt l
160/ Tr. 129-135.
l 161/ Tr. 132.
l l
. of the analysi3 of the currently unresolved seismic issue (e.g., impact of the new racks on the spent fuel pool and walls during seismic events).162/
202.
Based on the record befors it, the Board finds that issuance of the license amendment requested in this proceeding, installation of five high density r:cks in the Dresden Unit 3 spent fuel pool, is not a major Co' mission action significantly affecting the quality of the human m
environment and therefore it does not require the preparation of an environmental impact statement (EIS). The Board finds that tFe SER is adequate to support the installation of the five racks, but that it should be supplemented to reflect information received subsequent to its date of preparation, June 6, 5 380, as indicated in the immediately preceding paragraph.
162/ Tr. 1152.
III.
CONCLUSIONS 0? LAW 203.
The Board has reviewed the es!Jence submitted by the parties in regard to Applicant's motion for approval of the 5 rack project, and in response to the Board's questions 3 through 10.
The Board has also considered the proposed findings of fact and conclusions of law submitted by the parties on contested issues at the close of the hearings held on Commonwealth's application to modify the Dresden spent fuel. pools 2 and 3.
Consideration has been given the record which was made at the September 11, 1981 hearing on the motion to approve the 5 rack project. The Board makes
_the follcwing conclusions of law:
1.
The issuance of the license amendment requested in this proceeding, installation of five high density racks in the Dresden Unit 3 spent fuel pool, is not a major Commission action significantly affecting the quality of the human envircnment and therefore it does not require the preparativ.. of an environmental impact statement under the National Envircnmantal Policy Act of 1969, 42 U.S.C. Section 4321, et seq., and Part 51 'of the Commission's regulations,10 CFR Part 51.
2.
The Licensing Board in this case is not requireo to consider the five factors set forth in the Commission's " Notice of Intent to Prepare Generic Environmental Impact Statement on liandling and Storage of Spent Light Water Power Reactor Fuel," 40 Fed.' Reg.
42801 (September 16,1975).
See " Notice of Finality of Co.mmission Action with Regard to Final Generic Environmental Impact Statement on
. _ _ Handling and Storage of Spent Light Water Power Reactor Fuel (NUREG 0575)," 46 Fed. Reg. 14506 (February 27,1981).
3.
There has been no showing by Intervenor through filing a timely contention meeting the requirements or the Commission's Rules of Practice or otherwise, that there is a reasonable nexus between
" systems interaction", and the subject matter of this proceeding.
4.
There is reasonable assurance that the activities authorized by the requested operating license amendment can be conducted without endangering the health and safety of the public provided that the conditions set forth in the Order, below, a're inem porated into the license, and provided that the commitments set forth below are followed.
5.
The activities authorized by the requested operating license amendment will be subject to compliance with the Commission's regulations.
6.
The issuance of the requested operating license amendment will not be inimicable to the common defense and security or to the health and safety of the public provided there is compliance with the conditions and commitments set forth in the order belr.
IV. ORDER In accordance with the Atomic Energy Act, as amendcd and the regulations of the Nuclear Regulatory Commission, and based on the findings and conclusions set forth herein it is 8s
- =.
~
ORDERED that the Director of Nuclear Reactor Regulation make appropriate findings in accordance witF the Commission's regulations and issue the appropriate license amendment authorizing the requested replacement of 13 spent fuel storage racks by 5 high density storage racks at Dresden Station Unit 3 spent fuel pool.
The aforementioned license amendment shall contain the following conditions:
1.
Fuel stored in the spent fuel pool shall have a U-235 loading less than or equal to 14.8 grams per axial centimeter. 63/
2.
No loads heavier than the weight of a single spent fuel assembly shall be carried over fuel stored in the spent fuel pool.164/
In. deciding to grant the aforementioned license amendment, the Board has relied upon the following commitments by the Applicant:
1.
A corrosion surveillance prngram for the racks to insure that any loss of neutron absorber material and/or swelling of the storage l
tubes it detected.165/
2.
In situ neutron attenuation tests to verify that tubes and racks contain a sufficient number of Boral plates such that l
163/ NRC Staff Exhibit 1, Safety Evaluation at p. 3.
l 164/ NRC Staff Exhibit 1, Safety Evaluation at p. 10.
l 165/ Draley, prepared testimony attach. cat 6, following Tr. 341;. Weeks, supplemental testimony at p. 3, following Tr. 434.
I i
i 1
1
87 -
K-effective will not be greater than 0.95 when the spent fuel is in place.lj6,/
3.
If one Boral ' plate is detected missing, the associated tube will be blocked to prohibit ~ insertion of a fuel assembly.
If mere than one missing Po al plate is detected per pool, Applicant will remove the storage rack or racks containing any additional missing Boral plates from the pool.
Such storage racks will not be replaced in the pool until a specific criticality analysis covering the proposed corrective action has been submitted to and approved by the NRC.167/
~
4.
Before any storage rack is placed in the Dresden pools, Applicant will check each storage location with a plug gauge to confirm that the minimum dimension between th'e lead in clips at the top of each storage location is at least 5.758 inches.
If necessary, Applicant will grind down the storage clips to ensure this dimension is achieved.168/
The Board finds inat these commitments by the Applicant add to the assurance of safe operation of the spent fuel pool, and therefore they contribute to the Board's conclusion that the application to modify the Dresden Unit 3 spent fuel pool should be granted.
Accordingly, the Board hereby orders the Applicant to keep these commitmc.?ts until it is released 166/ Tr. 595-596.
167/ Tr. 595-596.
168/ Gilcrest, Tr. 920.
-n..
+
. from them by the NRC, and further, Applicant is ordered to include these commitments in the Dresden FSA? when it is updated.169/ Failure to implement to.se commitments is subject to any appropriate sanctions found in the Comission's regulations.
It is further ORDERED in accoreance with 10 CFR 2.760, 2.762, 2.764, 2.785 and 2.786, that this partial initial decision shall be effective immediately and shall constitute the final action of the Commission forty-five (45) days after the issuance thereof, subject to any review pursuant to the above-cited Rules of Practice.
Within ten (10) days after service of this partial initial decision any party may take an appeal to the Commission by.
the filing of exceptions to this decision or designated pa; thereof.
A brief in support d che exceptions shall be filed within thirty (30) days thereafter [ forty (.40) days in the case cf the Staff]. Within thirty (30) days' of the filing and service of 169/ See 10 CFR 50.71(e)0, as amended, effective July 22, 1980, 45 Fed.
Reg. 30614 (May 9,1980).
l L
39 -
the brief [ forty (40) days in the case of the Staff] any party may file a brief in support of, or in opposition to, the exceptions.
THE ATOMIC SAFETY AND LICENSING BOARD 4
/k 4
Linda W. Little ADMINISTRA11VE JUDGE or J. Remick ISTRAllVE JUDGE
~
L>-&-
JohnF. Wolf, Chairman ['
ADMINISTRATIVE JUDGE Dated at Bethesda, Maryland this 24th d>iy of September, 1981.
4 4
r
,.e
APPENDIX A LISTS OF EXHIBIT",
EXHIBIT ADMITTED IN EVIDENCE A.
Applicant's Exhibit Number:
1.
Licensing Report Dresden Nuclear Report Dresden Nuclear Power Plant Units 2 and 3 Spent Fuel Rack Modification (Rev.4).
Tr. 451 2.
Licensing Report Dresden Nuclear Power Plant Units 2 and 3 Spent Fuel Rack Modification (Rev. 5).
Tr. 965 3., Five page letter dated June 12, 1381 signed by Mr. Janacek and addressed to the Administrative Judges.
Tr. 1092 4.
Letter dated August 10, 1981 to i
Mr. Dennis Crutchfield of NRC frpm Mr. i. J. Rausch of Commonwealth Edison.
Tr. 1093 5.
Seven page response by Applicant to NRC Staff questions 6 and 7.-
Tr. 1125 B.
Staff's Exhibit Number:
l.
Safety Evaluation Report and Environmental Impact Appraisal Relating to the Modification of the Spent Fuel Storage Pool Provisional License No. DPR-19 and Facility Operating License No. DPR-25.
Tr. 118 2.
Affidavit of Kenneth S. Herring evaluating 5 rack project.
Tr. J
_ 91 EXHIBIT ADNITTED IN EVIDENCE C.
Iatervenor's Exhibit Number 1.
Memorandum from Henry E. Bliss to D. J. Scott and W. L. Stiede regarding Clearances on Dresden's i
High Density Spent Fuel Storage Racks, dated October 31, 1980.
Not IJaitted 3
2.
NSC " Trip Report," dated 4
September 2,'1980.
Tr. 511 3.
NSC Report and associated close-out documents of NSC Audit of Brooks & Perkins, dated September 26, 1979.
Tr. 512 4.
Commonwealth Edison Company Audit Report of Commonwealth Edison Company's audit of Brooks & Perkins dated September 13, 1980.
JTr. 285 5.
Commonwealth Edison Company Audit Report of Commonwealth Edison Company's audit of NSC, dated September 25, 1980.
Tr. 268 i
6.
Commonwealth Edison Company Audit Report and associated close-out
' documents of Commonwealth Edison i
Company's audit of Leckenby Company,
-dated September 29, 1980.
Tr. 287 i
/.
Commonvealth Edison Company Audit Report and associated close-out documents of Commonwealth Edison Company's audit of Leckenby Company, dated March 13, 1980.
Tr 290 8.
Internal Audit Summary Report from T. L. Sumter to P. D. Moore, dated June.19, 1979.
Not Admitted m
l " VIBIT ADMITTED IN CIDENCE 9.
Internal Audit conducted by Brooks & periins, Inc.,
dated Junc 11, 1980.
t'or,<!mitted
- 10.. Nuclear Aegulatory Comission Audit Report of Leckenby Company, dated April 14, 1980.
Yv. 334 11.
NSC "Tr.p Report," dated
'May 5, 1980.
Tr. 511 12.
" Evaluation of Dresden Nuclear Power Station," dated September 12, 1980.
Tr. 607 13.
Nuclear Regulatory.Comission's Health Physics, Appraisal, dated Scotember 12, 1980 and Comonwealth Edison Company's Response, dated October 6, 1980.
Tr. 627 5 C 14.
One page sketch en+.itled
" Deformation of Edison's BWR '3' 80 Mil Channels," undated.
Not Mmitted 15.
General Electric Company Specification 22A5866, Revision-0 " Fuel Storage Requirements," dated November 3, 1978.
Not Admitted 16.
Comonwealth Edison Company Handwritten Notes, undated.
Not Admitted 17.
Comonwealth Edtstn Company Handwritten Notc, dated November 14, 1980.
Tr. 803 18.
General Electric Company Document entitled "Recomendations for Mitigation of the Effects of Fuel Channel Bowing," dated Decembei 1979.
Tr. 862 1.
4 93 -
i i
EXHIBIT ADM*ifDINEVIDENCE 19.
NSC I4emort;., tum to Q. Hossain from J. Gilcrest entitled "Dresdea Fuel Racks (Com-0219) - Fuel Channel Bowing," dated December 9, 1980.
Not Mmitted D.
Board Exhibit Number:
1.
NSC Purchase Order.
Tr. 713 2.
Brooks & Perkins, Inc.
Purchase Order.
Tr. 713 a
3.
Leckenby Co. Purchase Order.
Tr. 713 -
t 9
4 9
4
.