ML20027A816
| ML20027A816 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 03/17/1982 |
| From: | Bridenbaugh D AFFILIATION NOT ASSIGNED |
| To: | |
| Shared Package | |
| ML20027A786 | List:
|
| References | |
| FOIA-82-328 NUDOCS 8207140488 | |
| Download: ML20027A816 (13) | |
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a A F F I D A V 1' T
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ng rully sworn and under oath do state:
I res'ide a Illinois.
I was o
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employed as a core driller at the Commonwealth Id'ison IIsalle County' nuclear plar.t construction site from approximately June, 1978 until S
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Fron about June, 1973 until about Februn:y 198O my employar was Corr.ercial Concrete Sawing and Drilling Cor.pany.
My duties were the drilling of holes in concrete.
I drilled holes ranging in diameter frca 1/4" to 3/4" with a small hand drilling nachine.
I also drilled l
larger b:cles, ranging in diar.eter frca 1-1/2 to 8", with a large 1
bcring rachine.
7.ncher bolts fer the small holes were used for
.nanginc ccaduit, cable trays, a.d o'ther electrical ecuipment.
The lar:e holes were used to carry conduit through walls and floors.
l I per'ormed core drilling in all buildines, ct all elevatic'n.s,
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thrcuc.hout.the.olant site, including the reactor buildin.s for Units I and 2.
During raest of the vear 1979,.my par ner and I were 1
nsrigned prir.arily to the two reacter buildi.,gs.
We dril?.ed at a11
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ren:. i th cer_?.e r ci a.1 Concrete E -wing and Dr.11ing, I recei ec ny Wh
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3-1 drilling instructions orally from my foremen, a
Ccomercial Concrete e:.ployee.
My wor.k was observed b
a superintendent ecployed by the general contractor, Fo ey E&cctri-Cal Co.
From the time I began drilling at LaSalle in Jur.e, 1979 until about February, 1980, it was che usual practice, upon contacting metal rei'nfer:ement or rebar during core drilling, to drill through the metal rebar..
I was ir.structed to follow this practice, and to the best of ry knowledge,, it was the general practice arcag the
- her core drillers.
' occasionally we were instructed to stcp and r.elocate the hcles when metal was centacted.
But during most of the time per.cd, we cut through the metal.
Small holes were of the following sizes:
Diameter (inches)
Depth (in cb.es )
1/4 1-1/4 3/8 4
1/2 4-1/2
-,3/ 8 "2
3/4 6
When rebar was contacted in drilling small holes with diameters of 1/2 to 3/4 inches, we changed to the larmer " wet drill.", which was a boring drill with a carbon bit and a water spray.
The we drill was too large for 1/4 cnd 3/3 inch holes.
When rebar was contacted
' in drilling these snellast holes, we relocated the heles.
My pattner and I dri led h.2ndreds of small hcles per week.
'e contacted and I
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s-s through rebEr in less than 1/4 of those holes.
cut During the latter oart of 1979, each s r.all hole cut through rebs wad mark'ei'i i
with r.aint er a felt pen.
I am unable to estimate the (nunber 'cf
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rebar cuts more precisely.
I,arge holes, rangine i: diameter from 1-1/2 inch to 8 inches, were all cut with the large boring machine, The depths of the large holes ecualled the thickr.es of the walls, floors, or ceilings'
' i through which the holes un cut, and ranged frora about 1 foot to 6
eet.
For the largest Mmeters, the bits were carbon with dia-
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e..e.i p s.
During my enSyment in 1978-79, core drillers were
_. _._..c.-ad to cut th-oug: char when it..as contacted during the d
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,e of large holes 37e seldom failed to conta=: rebar with the large boring machin My partner and I followed this' practice; and to the best of my k' ledge the other drillers did also.
Until h,.
d of 1979, I beli that all of the large holes were drilled by CoracrCial concrete >l yees.
When I worked in reactor $uildings 'for Units 1 and 2 during 1979. I drilled large.rs thrcugh the walls between the two re-actor buildings, betwche reactor and the off-gas building,. and between the reactor ane at:cilliary building.
Large holed were
,g 33 a rete of aboufoot per hour through concrece.
When rebar i
C C n ' E C u.g.a and Cut took 1Cnger.
u VaS 0CaniCf.5 50 DO,e5 IC be Cf',l*C We!O AE1C Out Oy 7
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s - mi o TO thG Dest of r.y kncW edce met.il
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detecrcrs were nct used' during the first 7-6 months of r..v en.clov-for locating the holes..' A Qual'ity Assurance inspector in-cent s,cacted my werk beginning several ncnths after I started workin'g.
I ca-. recall two specific incidents concerning the drilling of large holes through metal rebar, on one occasion I drilled a 6" diameter hele through rabar in the reactor building of Unit 1, at an elevat. ion below 710 '.
It was at a place where all the s. teel i
tied together, cad I removed about 25-40 pounds of steel.
It toch me 2 or 3 days to drill this holel tiructed me to kE=p drilling this hole, and added, "If you can't de it, we'll get scr.ecne who can."
t Cn a 3econd occasion I drilled a 7" diar.eter hole in the re-acter building cf Unit 1 at elevation 735.
I hit che 2" rebar, aid as I centinued to drill the rebar was splitting.
I asked m 4
3.d if I could relocate the hole.
- s5id, "No."
That g
v hcie was' drilled to a depth of 6 to 7 feet, where se hit a beam in the f1cer of a room where steam pipes were 1ccated.
This hole was la er. grouted in, because 'it was improperly located.
+le filled out a written report, or drill sheet, en 'each.hcle we drilled. fcr both small and large holes.
The reports showed the location, depth, end diameter of each hble.
Ihey also showed.
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UNITED STATES y
g NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 9.
APR 151981 Docket No. 50-373/374 MEMORANDUM FOR:
Harold R. Denton, Director Office of Nuclear Reactor Regulation FR0ti:
Richard C. DeYoung, Director Office of Inspection and Enforcement
SUBJECT:
RESULTS OF INDEpEllDENT ASSESSMENT OF ALLEGATIONS BY T. FAHNER, ESQ.. REGARDING THE LASALLE STATION OFF-GAS FILTER BUILDING As discussed in our conference call with Messrs. Case, Stello and Keppler on March 29,1982, IE accepted responsibility for certain actions necessary to respond to the March 24, 1982 filing by T. Fahner, Esq., Attorney General, State of Illinois. Specifically, IE was to conduct an independent assessment of actions taken by RIII on the issues related to the off-gas filter building and to independently assess the allegations r. elated to the off-gas filter building structural adequacy.
We have completed both aspects of this assessment, the details of which are contained in two separate reports. Enclosure 1 (Inquiry and Assessment of RIII Actions) will not be finalized until April 16 but its basic conclusions are reflected in this memorandum. You will be provided a copy upon completion. These reports were prepared on the basis of documentation review, interviews, field measurements and field observations.
Our conclusions as a result of this independent effort are as follows:
1.
Region III followed a logical cour se of action in responding to the allegations related to the off-gas filter building. They first determined whether the allegations pertained to safety related structures, systems or components. This was completed,using the reference document of the Safety Analysis F.eport plus the expertise of RIII staff in.BWR plant systems including the off-gas system.
Since the allegations were found to relate to non-safety items, no further action was taken. Our independent assessment is that the allegations did not require additional RIII follow-up. This is due to the fact that the off-gas filter building is a non-safety related building which contains systems and equipment which have no function in preventing or mitigating accidents or accident conditions.
CONTACT:
R..E. Shewmaker, IE 49-29678 JO 1).. p,'
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1 Harold R. Denton ApR 151982 I
Further it was noted during this assessment that when RIII became i
aware of the allegation of the cutting of reinforcing steel in safety-related structures, prompt action was taken by the investiga' tor to alert his supervision which, in turn, alerted the RIII engineering Currently, RIII is working on this matter in conjunction with NRR.
g roup.
2.
The reinforced concrete roof of the off-gas filter building has slight deviations from the thickness of the 12-inch thick slab as specified by the design drawings.
The range of deviations of +1-1/4" and -13/16" will have na significant effect on structural behavior.
3.
The loading imposed by the temporary construction related transformer on the roof of the off-gas filter building did not exceed the design loads and did not cause structural damage.
4 The current condition of the roof of the off-gas filter building, which includes existing cracks, embedments, anchor bolts and nicked reinforcing steel, is acceptable.
There is every reason to expect that the roof system can fully carry the design live load of 100 psf and remain in a serviceable condition while performing its intended function over its service life.
If there are any questions on this assessment and its related details, my
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staff and I will discuss them with you. We are maintaining the various documents which support this assessment and its conclusions.
W Young, /
Richard C.
frector Office of spection and Enforcement
Enclosures:
l 1.
Inquiry and Assessment of RIII Actions 2.
Assessment of Off-Gas Filter Building l
cc w/ enclosures: See page 3 l
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-Harold R. Denton 3-APR '151982 1
cc w/ enclosures:
l J. H. Sniezek, IE E. L. Jordan, IE R. Fortuna, IE D. G.'Eisenhut, HRR R
A. Purple, NRR
^
. Schwencer, HRR l
R. H. Vollmer, NRR J. G. Keppler, RIII R. L. Tedesco, NRR e
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ASSESSMENT OF THE OFF-GAS FILTER BUILDING AT LASALLE NUCLEAR STATION i
by R. E. Shewmaker, P.E.
April 8,,1982
Background:
At noon on March 30, 1982, I was provided with some preliminary information related to statements contained in a petition dated March 24, 1982 from Mr.
. Fahner, Esq. addressing the off-gas filter building roof by E. L. Jordan.
I was alerted that it might be necessary to go to the LaSalle facility to view the structural components in question that same week and provide a written assessment by mid-week the following week.
' At the direction of Mr. E. L. Jordan, mid-morning on March 31, 1982, I was instructed to assess in the field (I) whether the reinforced concrete roof of the off-gas filter building met the design requirements (that is,- does the as-built condition conform to the drawings) and (2) whether the reinforced concrete roof of the off-gas building can meet its service requirements. The need for such an assessment was apparently the subject of a telephone conference call on March 29, 1982 regarding a petition filed by the Attorney General, State of Illinois requesting a Show Cause Proceeding or Other Relief related to this and other issues. This conference call was followed b for assistance from IE by the RIII Regional Administrator (y a written request see Attachment 1).
Initial In-Office Effort (to determine requirements)
During the afternoon of March 30, 1982, I proceeded to review the pertinent portions of the LaSalle Final Safety Analysis Report (FSAR) to determine what the licensee had defined the structural safety requirements to be for the off-gas filter building.
Section 3.2, Classification of Structures, Components and Systems, was examined along with Table 3.2-1 which provides a detailed classification of various plant structures, equipment and components. As noted in the text of Section 3.2, plant structures, systems and components important to safety are designed to withstand the effects of a safe shutdown earthquake and remain functional.
These are known as Category I and include all such items if they are required to ensure:
The integrity of the reactor coolant pressure boundary, a.
b.
The capability to shut dowr. the reactor and maintain it in a safe condition, or k
. ENCLOSURE 2
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The capability to prevent or mitigate the consequences of accidents which could result'in potential offsite exposures in excess of the guideline exposures of 10 CFR 100.
This revealed that all of the equipment housed in the off-gas filter building which is part of the off-gas system is classified as seismic Category II as-well as the building itself. This means that the equipment and structure are
.not required to meet the requirements of 10 CFR 50, Appendix B, Quality Assurance Requirements. The quality group classifications for the various portions of the off-gas system are either C or D as indicated in Table 3.2-1 and defined in Regulatory Guide 1.26 and meet the various quality standards of the pertinent industry codes and standards (see Attachment 2).
The FSAR was then reviewed to determine what documented design requirements had been committed to by Commonwealth Edison for the design and constru'ction of the off-gas filter building. The FSAR in Section 3.8.4 describes the criteria for the seismic Category I structures other than containment which in this case did not include the off-gas filter building. Commonwealth Edison provided for another level of safety-related structures in the design criteria known as "Non-Seismic Category I Safety Related Structures," but it also did not include the off-gas filter building. Therefore,' no defined criteria. existed in the licensee's application which is consistent with the fact that the building was not classified as safety related.
The Safety Evaluation Report (SER) for the LaSalle Nuclear Station was also '
examined to determine if the NRC staff had accepted the classif,ication of structures, components and systems provided by the applicant. Section 3.2.1 indicates that with the exception of the classification of cooling loop of the spent fuel cooling and cleanup system all structures, components were correctly classified by Commonwealth Edison.
In Section 11.2.2 which addresses gaseous, wastes the NRC staff specifically stated that the off-gas system is located in the off-gas filter building which is a nonseismic Category I structure. The NRC staff further stated that the process off-gas system and the structure housing the system were acceptable (see Attachment 3).
I also had a discussion with.an IE BWR systems engineer concerning the proper classification of "an off-gas filter building and the off-gas system." He indicated that the system was not used for accident prevention or mitigation and that the system would therefore most likely be classified as non-safety related.
On March 31, 1982, I discussed the proper classification of the off-gas system l
w3th an IE health physicist who reviewed the typical system's design and function as well as pointing out that many BWR's operate without such 'a system l
though newer plants are installing such treatment systems to conform to the ALARA guidelines of 10 CFR 50 Appendix I.
His conclusion was that, building i
failure should be of no real concern from a radiological viewpoint:
~
Also on March 31, 1982, during a meeting held with Comonwealth Edison at the -
request of the NRC for which a transcript was made, it was reiterated that the 6
, -t NRC, specifically NRR, had considered the off-gas filter building as a "non-safety grade building" which contained no Category I safety related equipment.. In addition, the NRC Region III office also stated that they had treated this as a Category II, or non-safety related building. The licensee stated, however, that they did apply the safety-related quality assurance program to the construction of the off-gas filter building (see Attachment 4).
In conclusion, all information available and assessments ~ made indicated that
.the licensee's classification of the off-gas filter building was in fact correct in that it is not a Category I structure and that the structural requirements governing the design and construction would be -those specified by the owner and his agents and would not necessarily incorporate any NRC requirements.
With such a classification, the off-gas filter building would not be part of those structures that would be inspected by the NRC.
I i
Field Effort (to determine as-built conditions):
On April 2,1982 I visited the LaSalle Nuclear Station facility to obtain information and make first hand observations of the off-gas filter building roof.
Three specific concerns were to be addressed during this field effort:
1.
Facts related the off-gas filter building reinforced concrete slab roof thickness, 2.
Facts related to the external loading of the roof by an electrical transformer, and 3.
Facts related,to the current condition of the reinforced concrete roo.f system such as holes, anchor bolts, embedments and cracking.
Roof Thickness The as-designed roof slab thickness shown on the design drawings is 12" (ref.
S&L Dwg. 5-188, Rev. J). The top of the concrete was established by design at Elevation 726'-6".
Copies of the surveyors field notes for the surveys which the licensee had completed were obtained. These notes reflect three separate surveys with the first being a single point thickness established on March 10, 1982. This established a total thickness of l'-21" which included 1-3/4" to 2" of insulation, asphalt and gravel or a 121" or 121" concrete roof slab. Two separate surveys were completed on March 29, 1982 by a four-ma, survey party using Level No.
2R728.
W. Larson was the survey party chief. The party was made up of personnel of Falsh Construction Company and are classified as Technical Engineers and are unihn personnel. Their survey work was performed at the request of Mr. D.
Shamblin, Construction Engineer, Commonwealth Edison. His directions were to establish the thickness of the roof slab between the integral roof beams.
There are five (5) spans of roof slabs so the team selected three (3) points on approximately each of the spans' centerlines, making'a total of fifteen (15) points on which to establish slab thickness. The approximate locations of the
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points were established from.the column and beam centerlines for reference.
The survey party conducted the field work without direct oversight by Commonwealth Edison representatives or quality assurance' personnel. No independence from the construction company who built the roof existed.
From a previously established benchmark elevation on the off-gas filter building wall a series of vertical elevations were established by a surveyor's level on the outside of the roof and similarly for the ceiling inside the building at the fifteen (15) points.
In the first case the level rod was held at each of the 15 locations on top of the builtup roofing gravel so that the thickness determined at each point reflected the thickness of the concrete roof slab plus the insulation, asphalt and gravel. The concrete roof slab thickness was then determined indirectly by subtracting the approximate design thickness of the insulation, asphalt and gravel which was taken as.2-3/4" to 2".
In the second case a 6" steel spike was driven in the outside of the roof to punctu're through the gravel, asphalt and insulation until the. top of the concrete roof. slab was struck. The top of this steel spike was then the point on which the level rod was held at each of the 15 locations on top of the roof. The thickness of the concrete roof slab was then obtained directly from the difference in elevations for the top and the bottom of the concrete roof slab.
From the first case the thickness at one (1) point was determined to be 11.68"
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assuming a 2" thick built up roof.
In the second case the thickness at three (3) points was found to be 11.16", 11.52" and 11.64."
I consider measurements of 11.88" as within the allowable measuring errors of 10.01'.
Four (4) measurements of 11.88" were obtained.
The reduction of field survey data has been checked to verify the determination of the thickness of the concrete roof slab based on the field data. No discrepancies were noted. No direct measurements of the concrete roof slab thickness were physically possible since no holes are open through the roof thickness. From the design drawings and field observation the six (6) penetrations through the roof are all sealed. They consist of three (3) roof drains, an HVAC vent / intake, I
an electrical conduit and an abandoned 12" diameter sleeve which is sealed i
closed.
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- Discussions with site personnel revealed that during drilling for concrete anchors there was an instance where drilling into the underside of a 24" thick reinforced concrete floor resulted in penetrating the trap of the floor drain
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for the 690'-0" floor level in the off-gas filter building. This allowed the water in the trap to flow down through the hole and the possibility of daylight i
being seen up through the drain via the hole. This occurred in the next floor level below the off-gas filter building's roof.
Based on my review of the facts my assessment is that the reinforced concrete slab portion of the off-gas filter building roof is nominally a 12" thick section with the average thickness, based on fifteen (15) measuremen'ts, being l'-01".
The actual range of values for the fifteen (15) measurement's was +11" and -13/16" indicating the tolerances are somewhat outside the generally accepted values of +i" and -i" as provided in ACI 301, specifications for 9
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provided in the Proposed ACI Standard ACI 117,' Tolerances for Concrete Construction and Materials, August 1980.
These slight deviations of tolerance will ha' ve no significant effect on structural behavior.
External Loading of Electrical Transformer Based on the information available at the site from Mr. D. Shamblin, Construction.
. Engineer, Commonwealth Edison, there was a transformer placed on the roof of the off-gas filter building to provide electrical service for construction sometime probably in the last half of 1976. The concrete in the off-gas filter building was placed in November of 1975. The size of the base of the transformer-
, as 4'-0" by 13'-2" and the assembly had a total weight of 6700 lbs. The w
transformer was removed sometime in 1981.
Or.e end of the transformer was placed on the east wall (known as Ab) which is a 12" reinforced concrete wall with the long axis of the tranformer running in the east-west direction nearly aligned over the centerline of the roof beam just north of Column line 13. Based on my calculations this loading, conservatively assuming no loads are directly transmitted to the wall and that the roof slab and beam system must carry the. load, results in a value' of only about one-third (1/3) of the design live load (100 psf) existing over about 40% of th~e span of the beam.
Therefore, the loading of the construction transformer placed as it was on the reinforced concrete roof of the off-gas building represented less than one-sixth (1/6) of the design live load on the supporting beam.
In addition to this assessment of actual loading vefsus design capability, I l
examined the underside of the reinforced concrete off-gas. filter building roof in this area for indications of distress that could be caused by excessive loading or underdesign as a result of construction deficiencies. No evidence '-
of load distress were found. Minor hairline cracks were visible but in no greater concentration than elsewhere on the structure.
Based cn these facts qy assessment is that the temporary construction l'oading of the electrical transformer was well within the design loads for the reinforced concrete roof of the off-gas filter. building and that no structural distress was c.aused by the loads.
Current Roof Condition In September of 1979, nearly four (4) years after the concrete had been placed for the roof of the off-gas filter building, the Quality Assurance group of Comm,onwealth Edison noted soma surface cracking in the bottom surface of the off gas filter building roof slab. The general area was noted as having a high den (ity, of expansion anchors and some concern was expressed as to whether the cracking was serious and whether it at all related to the anchors. The area in question was examined by the Walsh Construction Company Quality Assurance Supervisor and the General Superintendent as well as the Commonwealth Edison Company Structura.1 Engineer. The decision was made to chip out two of the e
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cracks over some several feet to determine the depth of cracking. After chipping the area was patched when the conclusion was reached that the cracks were surface type cracks and no further action would be required.
During the March 31, 1982 meeting mentioned previously, a representative of Conraonwealth Edison indicated that the cracks did not exceed one-quarter (i)'
inch depth. The cracking was also characterized as shrinkage cracking associated with the slab type construction (see Attachment 4, pp.15 and 16).
On April 2,1982, I examined the underside surfaces of the four (4) main roof beams and the five (5) roof slab areas for cracking, holes embedments, anchor bolts and patches. The area where the largest crack size was found consisted of the slab area adjacent to the two (2) patched areas which were repaired in 1979 as a result of the licensee's evaluation of cracking. From this observation I would conclude that the cracks investigated by Comonwealth Edison in 1979 were in fact the largest ones visible then also.
I observed that at the end of the repaired area there had been no continuation or propagation of cracking since 1979 from the cracked and unrepaired area into the patch.
I did observe a small (probably width of 0.005") crack which crossed the patch (generally at 90') and continued about two feet on one side of the patch and about three feet on the other side of the patch.
I attributed this to minor shrinkage that has occurred since 1979.
Generally one can consider that 80-90% of the shrinkage takes place during the first year after placement and that this crack was a result of the later shrinkage.
The largest cracks I observed were on the order of 0.005" to 0.008" in width based on a wire feeler gauge I utilized.
The maximum depth I was able to insert the probe was about 1/8". The cracks that were. observed appeared to define no particular pattern with respect to embedded anchors, drilled-in anchors or the lines of distress that would develop as a result of excessive load or an understrength condition due to construction errors. Based on my I
I observations I concluded that the roof does not show signs of distress as a result of cracking from any conditions related to external loads, drilling or construction errors. There are cracks, however, and these are to be expected in reinforced concrete construction. The shrinkage effects on the concrete in this particular roof framing system may be somewhat amplified due to differential shrinkage since the slab portions are relatively thin and can lose moisture fairly easy with the resulting shrinkage. The beam portions, on the other hand, are massive (3' x 4' in cross-section) and tend to have fewer losses and changes of moisture.
The embedments which were cast-in-place when the roof system concrete was placed consist of flat steel plates anchored by welded studs in the typical fashion.
I The condition of the concrete adjacent to these embedments showed sorse of the same minor cracks of from 0.005" to 0,008" in width. There f
that would relate to relatively heavily loaded anchors vs. lightly loaded appeared to be no consistency in the location of cracks to define a pattern anchors.
In one instance an anchor judged visually to be relatively heavily loaded had no" crack adjacent to the anchor, whereas an anchor plate with no load (unused) had some adjacent cracking. There was also no evidence to indicate that the unused anchor had ever been loaded.
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7, The anchor bolts which were the drilled-in expansion type were used for attachments where the embedments could not serve as a result of there locat! ion or co'nfiguration.
I observed several locations where drilling had apparently been started and was terminated as a result of apparently contacting reinfoicing steel.
Three specific anchors were examined in detail from the field observations back to the design layout and control of the design (anchors CC-13, CC-93 and CC-CP-7)..
All locations found where drilling was terminated due to contacting rebar were
, apparently patched as indicated by the licensee since no open holes were found.
v Several unused drilled-in anchors were observed and probably were left unused due to relocation of other anchors on a specific anchor plate with multiple i
anchor bolts.
It was stated by Mr. D. Shamblin of Commonwealth Edison that he 1.new of no core holes (cut all the way through) made in the roof slab. All throagh-slab penetrations were cast in place with sleeves or blocked out during concrete l
placement.
I observed no indications of any core holes.
During the drilling operation in the off-gas filter building there were no cuts made through reinforcing steel.. There were only hits or-nicks made on the-reinforcing steel as it was contacted. These hits were ricorded when they occurred and thei were illustrated on S&L Drawing, RHS-188, Rev~. J.
No observations could be made in the field but it is my opinion that these nicks will not have any significant effect on the off-gas filter building roof.
Conclusions As a result of my review of the pertinent documents, discussions with cognizant individuals and my independent field observations and measurements I have concluded the following:
1.
The off-gas filter building is a non-safety related bulding which contains equipment which has no function in,nreventing or mitigating accidents or accident conditions.
The reinforcehconcrete roof of the off-gas filter bulding has slight 2.
deviations of the thickness of the 12 inch thick slab as specified by the
' design drawings. The range of deviations of +11" and. -13/16" will oave no significant effect on structural behavior.
i 3.
The loading imposed by the temporary construction related transformer on
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. the roof of the off-gas filter bulding did not exceed the design loa !s and
- did not cause structural distress.
- 4. - The current condition of the roof of the off-gar filter building which includes existing cracks, embednents, anchor bolts and nicked reEinforcing
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steel is acceptable and.there is every reason to expect that the roof system can fully a serviceable concarry the design live load of 100 psf while remaining in dition in performing its intended function over its service life.
Attachments:
1.
Request for Assistance 2.
Secti,n 3.2 and Table 3.2-1 of the SAR 3.
Extracts from SER (HUREG 0519) 4.
Extgacts from Heeting Transcript (3/31/82) 1
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/ Robert E. Shewmake l, P.E.
IDate Senior Civil-Struc'; ural Engineer Office of Inspection and Enforcement 9
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NtlCI EAR REGULATORY COMMISSION 3 g(l~' :.
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REGION lli
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r, *6 7sa noostvtLT HuAo
%,,.*.v a
ci en > tvw. Ltmors coiar March :10 1982 F
R. C. DeYoung,111 rector. 01hice of Inapaerson and NEnt1RANDilM FDR:
F.nforcement TROM:
. fames G. Keppler. Regional Admf niarratur, Regluu III Stin.1 MIT t 1.A SALLE COUNTY NUCLEAR STATION - TCTITI0t! FROH ILLINDIS ATTORNET GDiERAL As you know, on March 24, 1982. ' the Illinnir. Attorney Cencral petilluned the NRC to suspend licensing proceedings at La Salle. pe.nding investigation et recene alles,stions and to inst 1Lutu = Show Cauwe Hearing with Illinois ss a party to the llearing. The allecations deal with the overall adequacy of safety reisted structures as a result of videspread rebar cutting and npcc.ific nrrur.rnrn1.lefielenelew in the ruuf of the, off-gue building.
A cent ~crence call war. hcH nn Mari h 29 involving Meests. DenLun, Cuze,
,5tello. DeYoung and Keppl. ' rn discuss the handling of thewe luvet;tigallung.
We agreed that. becau'se the petitiva expresa.es concern that the uff-gas buildlug defieleneles had been verbelly com=unicsted earlier to NRC,and that the NRC had concluded an investigation of thcsc 331cgcd deficiencics was not varranted. it vould be prudent to have an independent reviev of thir.
silegation by 1E (since LE was not involved in the, consideration not to investigst e). This review should address both thu'techulcal adequacy of the ntf-gan huilding enne.e.rns as vall as the NRC's hand 11oc of the earlier notitication in this regard. Vith respect to the concerns associated with cutting through rebar this matter will be reviewed by Region III with technical ssr.ir.tance frem N1UL.
1 realite your staff is aliesdy depleted as a result of other investigation owslutance you are C ving us. ditd your willitigniemu to muslut lu this effort i
la Renuinely apprecisted.
1 Asb f( %.
[ James C. Kcppler Regional Administrutur cc: IV. Stello, U1:DRUdR
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- 11. Denton, NRR g.
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fOhllY Y
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t LSCS-FSAR
(
3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS Certain structures, components, and systems of the nuclear plant are considered important to nuclear safety because they perform safety actions required to prevent or mitigate the consequences of abnormal operational transients or accidents.
The purpose of-this section is to. classify structures, components, and systems according to the importance of the safety function they perform.
In addition, design requirements are placed upon such equipment to ensure the proper performance of safety actions when required.
3.2.1 Seismic
- classification Plant structures, systems, and components important to safety' are designed to withstand the effects of a safe shutdown earthquake (SSE) and remain functional, if they are required to ensure:
the integrity of the reactor coolant pressure a.
- boundary, b.
the capability to. shut down the reactor and maintain it in a safe condition, or the capability to prevent or mitigate the c.
consequences of accidents which could result in potential offsite exposures in excess of the guideline exposures of 10 CFR 100.
Plant structures, systems, and components, including their foundations and supports, designed to remain functional in the event of a SSE are designated as Seismic Ca'tegory I, as indicated in Table 3. 2-1.
All Seismic Category I structures, systems, and components are analyzed under the loading conditions of the SSE and operating-basis e.arthquake (OBE).
Since the two earthquakes vary in intensity, the design of Seismic Category I structures, compon en ts, equipment, and systems to resist each earthquake and other loads are based on levels of material stress or load f actors, whichever is applicable, and yield margins of safety appropriate for each earthquake.
The margin of safety provided
[
for such structures, components, equipment, and systems ensures i
that their design functions are not jeopardized.
For further details of seismic design criteria refer to:
a.
mechanical, in Subsection 3.7.3;
.[
b.
electrical, in Section 3.10; c.
structural, in Subsection 3.7.2; and d.
instrumentation and controls, in Section 3.10.
[tl65[//k, 3.2-1
LSCS-FSAR AMENDMENT 54 JANUARY 1981 3.2.2 System Ouality Group classifications each water, steam, or radioactive waste-contain
~
those applicable fluid systems which are relied up n
of on to:
a.
malfunctions originating within the reactor co pressure boundary, b.
permit shutdown of the reactor and maintain it in the safe shutdown condition, and c.
contain radioactive material in large quantity or concentration.
A tabulation of quality' group classification for ea h system
" Quality Group Classification."and component is shown in Table 3.2-1 c
structure, j
diagrams which depict the relative locations of thesFigures 3.2-1 and 3.2 ng, systems and components along with their quality group clas ifi e structures, cation.l s
The implementation of the code requirements' outli 3.2-1, 3.2-2, ned in Tables discussed in Sections 3.7 and 3. 9.3.2-3, and 3.2-4 for fluid system comp A boiling water reacter has a number of structures facility which have no direct safety functioncomponents in th
, systems, and e
connected to, or influenced by, the equipment within thebut which'may be safety-related classifications defined previously nuclear structures, systems, and components are designated as "other "
such The design requirements for equipment intended service of the equipment and expected plan the environmental conditions under which it operates.
n codes and standards.possible, design requirements are based on applica Where When these are not available, ry relies on accepted industry or engineering practibe the designer Structures systems and components whose safety functions require con,formance,to the quality assurance requirement of 10 CFR 5 0, Appendix B,
- heading, are summarized in Table 3.2-1 program is described in Chapter 17.0." Quality Assurance Requirement under the e
e O
3.2-2 e
TABLE 3.2-1 (cont'd) 00AI.ITY (4a)
OUALITY (4b)
SEISMictSI GROUP ASSURANCE ELDCTRICAL (4c) PURCitASE PRINCIPAL COMPONENT (1)
LOCATION (3) CATEGORY CLASS t rICATION REQUIREMENT CLASSIFICATION DATE f 2)
COMMENTS XXII.
Local Panels 1.
Electrical panels with a safety function A.RD I
HA I
1E 4-74*
(15) 2.
Cable, with a safety function A,R8 I
HA I
1E 10-75 3.
Remote shutdown panel A
I HA I
1E 10-74 f"XXIII. Off-cas System (2) 1.
Atmospheric glycol tanks F
II D
II HA 10-71 2.
Heat exchangers r,T II C
II HA 10-74 5
(
3.
Piping and valves (down-n stream of steam jet Y
air ejectors)
T,r,0 II C
II NON 15 9-74 3
6, 4
Piping and valves (up to and including air g
i IE" ejector)
T II D.
II NON IE, 9-74 5.
Valves Ter II C
II NON 1E l
II D
II HA 2-72
{
7 Charcoal vessels F
II C
II NA 10-71 8.
Recombiners T
II C
11 NA 10-71 9.
Filters F
II C
II HA 10-71 10.
Afterfilter F
II C
11 NA 10-71' 11.
Reheater F
II II NON 1E 1-72 XXIV.
Service Water System 1.
Piping RB,0,L,A T 11 D
II NA 9-74 2.
Strainers L
II D
II HA 7-73 3.
Pumps L
II D
II HA 7-73
$g 4.
Pump motors L
II II NON lt 7-73 g%
o 5.
Valves 7,0,L,A RD II D
II HA 6-73 m
6.
Electrical & instru-yg ment' Modules PB,L A II II NON 1E 7.
Cable RD,0,L.A.T II II NON 1E 10-75 0%
me me 7=^ /~ C.*,,, A Sc;o ~a d.w F= (g-
,fkp j I
....is TAntE 3.2-1 (Cont'd)
QUALITY (4a)
QUALITY (4b)
SEISHIC(5) GROUP ASSURANCE ELECTRICAL (4c) PURCHASE PRINCIPAL COMPONENT [1)
LOCATIOH ( 3) CATEGORY CI.ASSIFICATION _ REQUIREMENT CI.ASSIFICATION DATE(2)
COMMENTS X L11. Civil Structures 1.
Reactor building RD I
NA I
HA 2.
Lake screen house L
NA tlA II NA (22,27) l (22) 3.
Radwaste building RW NA NA II HA 4.
Auxillary building A
I HA I
NA i
(22)
(22) 5.
Turbine building T
NA NA II HA (22)
'6.
Off gas f11ter building F II HA II HA (22) 7.
Steam tur.nel A
I HA 1
NA (22) 8.
River screen house G
II HA II NA (22) 9.
Diesel-generator building RB 1
NA I
HA (22)
\\
10.
Aux 111ary Spillway 0
NA NA II NA (22) 11.
Cooling Lake Embankment O I
taA II HA (22) 12.
Subenerged CSCS Pond O
I NA I
NA u
(22) 13.
Biological! Shield PC I
NA I
HA y
e.
a 14.
Primary Containment PC I
HA I
HA (22) y
- a. Vacuum breaker E
piping PC/RB I
a I
HA 9-74 W
- b. Vacuum breaker N
valves PC/R8 I
B I
lE
- c. Maintenance butterfly valves PC/RB 1
B 1
NA
- d. Suppression vent downcomers PC I
NA I
HA X LI II. MSIV Leakage Control Syste' s
1.
Piping, within RCPB
! solation valves R8 I
A I
HA 9-74 2.
Piping, other upstream system lines RB I
8 I
3.
Piping, downstream sys-NA 9-74 st,em from steamline con-nection to first valve RB 1
D+
II NA (7,8) 3 g
4.
P! ping, other downstream p
system lines RD I
e I
NA 9-74 5.
Valves, within RCP8 RD I
A I
1E 12-73
-4 e
6.
Valves, other RB I
B I
IE 12-73 7.
Heater RD I
HA I
1R 5-76 8.
Blowers RB I
B I
18 11-75 9.
Electrical modules with a safety function RB I
HA I
1E 10.
Cables, with a safety 8.w-*lan HR I
MA I
IE 10-75
1 LSCS-FSAR
- AMENDMENT 24 SEPTEMBER 1977 TABLE 3.2-1 (Cont'd)
EQUIPMENT CLASSIFICATION COMMENTS (1)
A module is an ' assembly of interconnected components which constitute an identifiable device or piece of equipment.
For example, electrical modules include sensors (including electromechanical), power supplies, and signal processors; and mechanical modules include filters, strainers, and flow (element) ass,emblies/
orifices.
(2)
Purchase order dates (month / year) are given for equip-ment as a basis for determining certain applicable codes on Tables 3.2-2, 3.2-3, and 3.2-4.
Where two dates are given and indicated with a slash between them (e.g., 9-70/5-71) the first date corresponds to Unit 1 and the second date corresponds'to Unit 2.
Where two dates are given with a comma between (e.g.,
9-70, 5-71),
multiple purchase orders apply.
(3)
PC = within primary containment RB = within reactor building 0
= outdoors onsite L
= lake screen house A
= auxiliary building T
= turbine building RW = radwaste building F
= off-gas filter building
-- = all buildings except 0, L
(4) a.
Quality group classification per Tables 3.2*-2, 3.2-3, and 3.2-4.
Group "E"
components are special engineered components in accordance with l
the codes and standards specified in the notes and comments for this Table.
b.
I
- The equipment meets the quality assurance re-quirements of 10 CFR 50, Appendix B.
II - The equipment is not required to meet the quality assurance requirements of 10 CFR 50, Appendix B.
h c.
lE - Electrical equipment that meets the quality assurance standards of NRC guidelines and IEEE Standard 323-1971.
Non-lE Electrical equipment l
that is not required to meet lE requirements.
NA - not applicable because the equipment.is not electrical.
3.2-17 l
V
~
LSCS-FSAR*
TABLE,3.2-1 (Cont 'd )
(5)
I - The equipment is designed in accordance with the seismic requirements for the SSE.
7 II - The seismic requirements for.the SSE are not applicable to the" equipment.
(6)
The control rod drive insert and withdraw lines from the drive flange up to and including the first valve on the hydraulic control unit are Quality Group B.
(7)
The main steamlines between the outermost containment isolation valve up to the turbine stop valve, the main turbine bypass lines up to the turbine bypass valve, and all branch lines (2-1/2 inch nominal size and larger) connected to these portions of the main steam and turbine bypass lines up to the first valve capable of timely actuation are classified as D+.
These sec-tions of pipes meet all of the pressure integrity re-quirements of code practice for steam power plants plus the following additional requirements:
All longitudinal and circumferential butt weld a.
joints are radiographed (or ultrasonically tested to equivalent standards).
Where size or configuration does not permit effective volumetric examination, magnetic particle or liquid penetrant examination is substituted.
Examination procedures and acceptance stan-dards are at least equivalent to those specified as supplementary types of exami-nation, in ANSI B31.1 Code.
b.
All fillet and socket welds are examined by either magnetic particle or liquid penetrant methods.
All structural attachment welds to pressure-retaining materials are examined by either magnetic particle or liquid penetrant methods.
Examination procedures and accep-tance standards are at least equivalent to those specified as supplementary types of examinations, in ANSI B31.1 Code.
c.
All inspection records are maintained for the life of the plant.
These records include data pertaining to qualification of inspection personnel, examination procedures, and exami-nation results.
O O
3.2-18 e
LdLd-rdan namnununa a,
JULY 1981 TABLE 3.2-1 (Cont'd)
J' (20)
The unprocessed radwaste piping will meet Group D requirements and the following supplementary require-ments:
a.
Piping For sizes over 4 inches nominal, random radio-graphy of 20% of the joints was performed on girth and longitudinal butt-welds.
Sockets and fillet welds in sizes over 4-inch nominal will be given random magnetic particle and liquid penetrant. examination on 20% of the joints.
b.
Pumps and valves Welds in pumps and valves of pipe size over 4-inch was given random magnetic particle or liquid penetrant. examination.
Random examination is defined as examination of the linear dimension of a weld in a pump or valve with piping connecting over 10-inch nominal size or as examination of all of the welding in 20% of the pump and valves with piping connecting 10-inch nominal or less.
(21)
Quality group classification, requirements do n'ot apply to piping and components supplied by the diesel engine manufacturer as an integral part of the diesel,-
generator unit.
In this casel the manufacturer's standards are used with the intent that the piping or component is to function as reliably as possible.
(22)
Civil structures were used in missile analyses as bar-riers.
No individual missile barriers other than civil -
structures were credited.
(23)
Includes Scram Discharge Volume Accumulators.
(24)
Expendables and Consumables are purchased per original specification and stored under controlled conditions.
(25)
Includes raceway installations containing Class lE cables and other raceway installations required to meet Seismic Category I requirements (those whose failure during a seismic event may result in damage to any Class lE or other safety-related system or component.
[
(26)
Subsystems required for post-LOCA monitoring include containment hydrogen monitoring, containment. pressure monitoring, containment temperature monitoring, suppression pool water level monitoring, suppression pool water tempera-ture monitoring, and containment high-range radiation moni-toring.
Subsystems not required for post-LOCA monitoring 1 7.7A c--
ttachment 3
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DESIGN CRITERIA FOR STRUCTURES, SYSTEMS AND COMPONENTS j
3 Conformance with Nuclear Regulatory Commission General Design Criteria 3.1 In Section 3.0 of the Final Safety Analysis Report, the applicant presented an
} j.,
Nj ;
evaluation of the design bases for La Salle against the Commission's General We evaluated the
? p Design Criteria listed in Appendix A of 10 CFR Part 50.
fina.1 design rH the design criteria and conclude, subject to the applicant's 1 f:"
adeption of the additional requirements made by us as discussed in this report,
!g; that La Salle has been designed, can,be constructed an'd can be operated to g{;
ceet the requirements of the General Design Criteria.
}'ji:
3.1.1 Conformance with Industry Codes and Standards h
Our review of structures, systems and components relies extensively on the application of industry codes and standards that have been used as accepted.
]j industry practice. These codes and standards, as cited in this report and the' attached bibliography, have been previously reviewed and found acceptable by g$@3 us; and have been incorporated into our Standard Review ' Plan.
' @l$i 3.2 Classification of Structures, Components, and Systems
'd 3.2.1 Seismic Classification 7
. Criterion 2 of the General Design Criteria requires that nuclear power plant jh structures, systems, and components important to safety be designed to with-ji stand the effects of earthquakes without loss of capability to perform'their Q.
safety function. These plant features are those necessary to assure (1) the y
E integrity. of the reactor coolant pressure boundary, (2) the capability to. shut down'the reactor and maintain it in a safe shutdown condition, or (3) the
?{
capability to prevent or mitigate the consequences of accidents which could-f result in potential offsite exposures compeble to 10 CFR Part 100 guideline 4
1 exposures.
F Structures, systems, and components important to safety that are required to j
i f t
be designed to withstand the effects of a safe shutdown earthquake and remain functional have in general been properly classified as seismic Category I l
items, in conformance with Regulatory Guide 1.29, " Seismic Design Classifica-
.I The applicant's nonseismic i
tion," Revision 2 except for the following system.
l Category I classification of the cooling loop of the spent fuel pool cooling s
l snd cleanup system is not in conformance with Regulato,ry Guide 1.29.
As an.' alternate to a seismic Categury I design cooling loop of the fuel pool cooling and cleanup system, the applicant has provided an analysis in the..
~ Final-Safety Analysis Report that shows the radiological releases, following a
~
postulated failure of this system to function, are a small fraction of the guideline values of 10 CPR Part 100.
A seismic Category I cooling water makeup system to the pool is provided.
For further review of the spen,t fuel,
pool cooling and clear.up system, see Section 9.1.3 of this report.
O!I[3 3-1
-m
~
~
~
s 3
F All other structures / systams, and components that may be required for opera-tion of the facility have been designed to nonseismic Category I requirements, including those portions of seismic Category I systems such as vent lines, fill lines, drain lines, and test lines on the downstream side of isolation valves which 'are not required to perform a safety f. unction.
Structures, systems, and components important to safety that have been designed to with-stand the effects of a safe shutdown earthquake and remain functional are ident'fied in an acceptable manner in Table 3.2-1 of the Final Safety Analysis Report.
The basis for acceptance in our review has been conformance of the.
applicant's designs, design criteria, and design bases for structures, systems,
~
and comopnents important to. afety with the Commission's regulations as set c
forth in Criterion 2 of the General Design Criteria and to Regulatory Guide 1.29, our technical positions, and industry codes and sta'ndards.'
Except for the cooling loop of the' spent fuel pool cooling and cleanup system identified above, we conclude that structures, systems, and components important to safety that are designed to withstand the effects of a' safety shutdown l
earthquake and remain functional have been properly classified as seismic l
Category I items in conformance with the Commission's regulations, the applicable regulatory guides, and industry codes and standards and are accept-
. Design of these items in accordance with seismic Category I requi'rements able.
provides reasonable assurance that in the event of a safe shutdown earthquake, the plant will perform in a manner providing adequate safeguards to the health,
and safety of the public, and is acceptable.,
3.2.2 Systern Ouality Group Classification -
Criterion 1 of the General Design Criteria requires the nuclear power plant systems and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the'importance of, the safety function to be performed.
Fluid system pressure-retaining components important to safety have been designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety function to.
be performed.
The applicant identified those fluid-containing components which are part of the reactor coolant pressure boundary and other fluid systems important to safety where reliance is placed on these systems:
(1) to prevent i
or mitigate the consequences of accidents and malfunctions originating within I
the reactor coolant pressure boundary, (2) to permit shutdown of the reactor l
and' maintain it in a safety shutdown condition, and (3) to contain. radioactive i
These fluid systims have been classified in an acceptable manner in material.
Table 3.2-1 of the Final Safety Analysis Report and on system piping and instrumentation diagrams in' the Final Shfety Analysis Report, based on conform-ance with Regulatory Guide 1.26,' " Quality Group Classification and Standards,"
Revision 3.
~
$ The applicant has applied ' Quality Groups A, B, C, and D in conformance to
- Regulatory Guide 1.26, to the fluid system pressure-retaining components important to safety. Those components that, are classified Quality Group A,' B, C, or D have been constructed to the codes and standards identified in Tables 3.2-2, 3.2-3, and 3.2-4 of the Final Safety Analysis Report.
The basis' for acceptance in our review has been conformance ofIhe applicant's
~
designs, design criteria, and design bases for pressure-retaining compon*ents such as pressure vessels, heat,exchangers', storage tanks, pumps, piping and 3-2 e
e o
- 2
- +. - -
-,.m
,,m
^
- 1
~
L We determined that the liquid radwaste treatment system is capable of reducing the release of radioactive materials in liquid effluents to concentrations below the limits in 10 CFR Part 20, during periods of fission product leakage from the fuel at, design levels.
Based on these findings, we conclude that the design of the liquid radioactive
- waste treatment systems is acceptable.
11.2.2 Gaseous Radioactive Waste Treatment System The gaseous radioactive waste treatment system is designed to process gaseous wastes based on the origin of the wastes in the plant and the expected levels of radioactivity.
~
The gaseous waste treatment system consists of the main condenser offgas treatment system, mechanical vacuum pump 'offgas system, dryw, ell purge system, gland seal condenser offgas system, and building ventilation systems.
The offgas treatment system,' shared by Unit Nos.1 and'2, is designed to collect and delay fission product noble gases removed from the condenser by, the air ejectors.
In the offgas treatment system, the gas from both units flows through a preheater, a recombiner, a condenser /. separator, a 30 minute holdup pipe, a condenser / separator, a reheater, prefilter, eight charcoal beds in series, and an afterfilter.
Except for the holdup pipe and the second reheater, the offgas treatment system consists of'two separate trains.of equipment which* provide 100 percent redundancy in the processing of the gaseous wastes.
The eight charcoal beds are designed to operate at 77 degrees Fahrenheit and 45 degrees Fahrerheit dew point and contain three tons of charcoal each.
We consider the system capacity and design to be, adequate for meeting the demands of the station during normal operation, including anticipatdd operational occurrences. The system design include,s, dual hydrogen analyzers upstream and downstream of the recombiner which will provide automatic dilution or activate an alarm upon exceeding a preset hydrogen concentration and indicate that' switchover to the standby recombiner is required.
In addition to the protective instrumentation, the offgas treatment system is designed to withstand a hydrogen explosion (design pressure, 350 pounds per square inch gauge). We find the design provisions incorporated to reduce the potential of hydrogen explosion and to mitigate the effects to be in accordance with the guidelines of Regulatory Guide 1.143 and are, there. fore, acceptable.
l The seismic and quality group classification of the offgas treatment system is b'ased on criteria which were acceptable during the construction permit licensing stage, i.e, Quality Group C classification, nonseismic design for components.
Although these criteria differ from the current criteria contained in Regulatory,
Guide 1.143, we have determined that the' provisions incorporated in the design of the offgas treatment system are acceptable under the guidelines of Regulatory Guide 1.143.
The process offgas system is located in the offgas filter building whicih is a nonseismic -Category I structure. The charcoal vessels were designed to Quality Group C and meet American Society of Mechanical Engineers, Code Section III, Class 3, 1971. The parameters of the principal components con-sidered in the-process offgas system evaluation are listed in Table 11.4.,We find the process offgas system and the, structure housing the system to be acceptable.
11-16 *.
NUCI.ZAR REGUI.ATORY COM2".ZSSION
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I COMMCINEAI.TH EDISON CONPANY DOCKET NOS.
50-373 and 50-374 LaSalle County Nuclerr C-enerating Station, Unit 1 and Unit 2 CA S:
March 31, 1982 pAczg: 1 - 77' AT:
Bethesda, Maryland t
1 31DERSOX
},
- -(, REPORT 1XG
(...-
400 Virg d 2. Ave., S.W. *dasi-' 79.r.r., D. D. 2 0 0 2 4 d
V Telechc=e : (202) 554-2345 AIS///4-
9 a
i 7
MR. DE1 GEORGE:
What I v'ould like to do is 8 review the allegations presented in the petition as ve 9 understand then, stating the facts and the information to ve have which we think vill resolve the concerns that 11 have been raised i,n your mind.
12 I would like to start with the questions 13 raised reistive to the off-gas building because.ve feel.
T) 14 tha t to be a less conplicated is ue that can be more 15 easily dispositioned..
16 First, there is an allegation that the roof j
17 thickness is eig'ht inches as opposed to the 12 inch t
1s design thickness.
I would like to say at the outset to that although this building is a non-safety related building containing no safety-related equipnent and not l
20
~
21 requiring the implementation of our quality assurance 22 program, we did in fact apply our quality assurance which has h
23 progran to the construction of this building, 24 given us greater confidence in the accuracy of the lI 25 information that we vill be providing to you.
.)
i ALDER $oN REPORTING CoMPAM.INC.
l 15 i
)
s 13 MB. DELOEORGE:
I am ready.
The last 14 allegation sugge'sted that the co'ncrete associated with 15 this slab had been cracked substantially.
Commonwealth 16 Edisen discovered surf ace cracking of the subject slab
~.
17 through its own site quality sssurance department in 18 Sedtember 1979.
As a result of the deficiency 19 identified, an ingdiry was made at that time which 20 included an engineering evaluation and which also 21 included the tracing of the crack depth by chipping at 22 the concre te in th e vicinity of the cracks.
As a result of that review, it was established 23 24 that the crack depth did not exceed one quarter. inch;
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25 that the cracking vas, in fact, surface cracking, a n d'. a s
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l-f ALDERSoN RE?oRTING COMPANY,INC.
i
O 16 1 a result, it was patched.
We have no reason to believe, 2 based on that investigation, that the cracking alleged 3 is the result of drilling of. anchor bolt holes.
It as 4 our opinion, based on that evaluation, that the cracks
{
l 5 observed are normal shrinkage cracks associated with
.i i
6 this type of slab.
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ALDER $oM REPORTING COMPANY. WC.
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l 13 MR. DENTON:
Let me ask the project manager 3's 14 vha t ca tegoriration ve gave that roof.
15 MR. BOURNIA It is a n'on-safety grnde 16 building.
I have the reviewer here.
We did not 17 consider this as a safety grade building.
18 MR. DENTON:
What is under the roof?
19 MR. BOURNIA4 What is this?
20 ME. DESTON:
What is under it?
21 MB. DELOEORCE:
That is described in our 22 report.
The concrete enclosure above-grade as a pa rt of 23 the off-gas roof is a non-safety related structure which 24 houses off-gas building, heating / ventilating / add air 25 conditioning,. air handling units, HVAC, water cooled ALDERSON REPORTING COMPANY.INC.
18 1 condensing units, HVAC exhaust filter units, HYAC 2 control panels and associated motor control centers and 3 switchgear.
4 HR. DENTON:
Does that nean there is no 5 Category 1 safety-related equipment in that building?
6 HR. DEICEORGE:
Yes, sir.
7 Mk. DENTON:
Any qtestions?
'de can cone back 8 to this, but I thought we vould give the conpany a 9 chance.
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ALDERSON REPORTING COMPANY.1NC.
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18 HE. NORELIUS.
We received allegations on this 19 sone nonths ago a.nd evaluated it in-office.
I do not 20 have those with me.
I am not sure that I know they say 21 e xa ctif wha t she said, and I have not read them 22 caref ully.
But we vere aware of the allegation.
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23 evaluated within our office and I think, in recognition 24 of our msnpower c'onsiderations, va chose not to delve 25 deeply in.to this a t the field level because of its ALDERSOgREPoRDNG COMPANY,INC.
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Docket Nos.: 50-373 l'
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'EPaos 711e Tyrone C. Fahner, Esquire Attorney General State of Illinois 160 North La Salle Street Chicago, Illinois 60601
Dear Mr. Fahner:
This letter is to acknowledge receipt of your Request to Institute a Show Cause Proceeding and for MEeTRETTsf-(PetitWn) dated March 24, 1982, filed with the Nuclear Regulatory Commission on behalf of the State of Illinois. The Petition principally seeks institution of a show cause pro-ceeding under 10 C.F.R. 5 2.202, and suspension of further consideration of the operating license application regarding the La Salle County Nuclear Generating Station, Units 1 and 2 of the Commonwealth Edison Company, based on certain allegedly newly discovered safety issues. The issues raised in your Petition are two in number. The first issue relates to the drilling of holes in the concrete walls, floors, and ceilings of certain buildings at the La Salle facility, including in some instances severance of steel reinforcing bar (rebar) with the potential for affecting structural integrity. The second issue relates to the structural adequacy of the off-gas building roof.
Your l
Petition alleges substantial cracking of this roof and the possibility that i
I the roof thickness does not meet design specifications.
Your Petition has been referred to me by the Commission for consideration pursuant to 10 C.F.R. $ 2.206, and appropriate action will be taken on your petition within a reasonable time. The NRC Staff is investigating the allegations contained in your Petition 1/.
I 1/ Your Petition at page 2 states that there is an operating license proceeding presently before the Commission and that no hearing has been requested or noticed in this proceeding. A Notice of Receipt of Application for Facility Operating Licenses; Notice of Availability of Applicant's Environmental Report; and Notice of Consideration of Issuance of Facility Operating Licenses and Notice of Opportunity for Hearing, were published in the Federal Register regarding this f acility on June 3,1977. Consequently, a hearing opportunity regarding the La Salle facility operating licenses was noticed. No hearing was' requested pursuant to the notice and consequently no operating license " proceeding" is before the Commission. I am presently considering Commonwealth Edison's license application, and have under consideration the issuance of a fuel load and low power testing (up to 5% rated power) license in the very near future.
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2-As one step in our investigation, we contacted the Commonwealth Edison Company (CECO) on March 29, 1982, and asked that.they meet with us on March 31, 1982, to present information with regard to the matters you raised. This meeting was held in Bethesda, Maryland on March 31, 1982, and Ms. Judith S. Goodie. Esq.
attended the meeting as a representative of your office. A verbatim transcr'pt was taken of that meeting and has been made available to you.
0._ut egi_g!LilLpificeAdetemining_whethet_the_ applicant _has_adegua_tely R
implemented its rebar. damage idefttification_and.. assessment activity 11acj;pr_ dance _wi.th_ suitable. procedures during the. construction. proc _ess. Also, our Office of Inspection and Enforcement is reviewing Region III's handling of the allegation regarding the off-gas building roof, and is further determining whether the off-gas building roof meets its design requirements as stated by CECO at the March 31 meeting and in the FSAR for the La Salle facility. In conjunction with Region III activities, my office is assessing the technical adeguacy of the applicant's en_gineer_iDg aesign asteisment in the affected areas as e result of these allegations.
I have considered your request at Section III.1 of your Petition for an immediate suspension of consideration of Commonwealth Edison's application for a fuel load and low power testing license for La Salle Unit I until the allegations contained in your Petition are investigated and a decision made regarding instiWtion of the requested show cause proceeding under 10 C.F.R. 52.202.2)
By letter dated February 4,1982, CFC0 has addressed its startup test schedule for La Salle Unit 1.
In this letter, the appifcant estimates approximately 60 to 90 days will elapse from the time of issuance of the low power license l
I, until the time the reactor achieves initial criticality. The applicant's'best current estimate is that this elapsed time will be approximately 61 days.
During this time, there will be preliminary startup activities going on; however, since the reactor will not have been brought critical, there will be essentially no fission products in the core and no significant radiological hazard.
- Also, there will be no significant build-up of residual core activity through zero power physics testing.
Ctring the zero power physics testing, the off-gas building would not receive any radioactive materials and thereby pose no public health and safety hazard should its roof fail. Also, continued structural integrity of concrete which 2/ Your Petition at page 9 also seeks suspension or stay of all " proceedings concerning Edison's applications for operating licenses" for the La Salle facility.
As discussed in footnote 1, supra, there are no such " proceedings" pending.
The only relevant pending matter is trly consideration of Commonwealth Edf. son's license application including its request for a fuel load and low power testing license.
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. may have been affected by the alleged drilling and boring is not essential for even, in the most severe incident which could be postulated, the radioactive releases, would be insubstantial. The low fission product inventory also answers the concern raised in your Petition at Section II.8 that fuel loading of Unit I should be postponed "unti" the Commission fully examines the potential safety hazard presented by the cutting of reinforcing steel as alleged herein" on the grounds that the presence of the fuel in the st'ructure of Unit I will interfere with the investigation of.the allegations and with any corrective measures that might be ordered. The insignificant build-up of residual core activity through zero power physics testing presents no significant impediment to completing the investigative efforts associated with the alleged concerns.
In our efforts to investigate these matters, members of the NRC staff have made visits to the plant and to the Architect / Engineer's offices, Sargent and Lundy; have reviewed documents including structural drawings and several calculations performed by Sargent 1 Lundy and have held discussions with key applicant and Sargent & Lundy personnel (see Enclosures 1 and 2). While the reports on these efforts have not yet been finalized, the allegations, s,e_t_forth in your petition have not been substantiated.
"We expect to complete our ongoing investigation of the allegations raised by.
Wour Petition in the next 30 to 60 days and we will make our findings at that
))ime.
In the interim, because there is no significant threat to the health and safety of the public represented by a core that has not experienced operation' beyond initial criticality and zero power testing, I have determined the immediate suspension of Commonwealth Edison's request for a fuel load and low power operating license to be unwarranted. Any_ ssh ljcense, however. will
_be conditioned to require _ prior NRC staff approvalfor any. power. operation followino initial criticality _and_2ero.. power-physics. testing, Our approval will not be given unless warranted by the results of our current ongoing investigationr.
The NRC Staff will continue to review the remaining matters raised in your Petition, and I will issue a decision with regard to them in the reasonably near future.
I will also consider your Petition in any future licensing actions I take with respect to the La Salle facility.
I enclose for your information a copy of the Notice that is being filed for publication with the Office of the Federal Register.
Sincerely, Harold R. Denton, Director Office of Nuclear Reactor Regulation
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Enclosures:
As stated cc: See next page
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Mr. Louis 0. De1 George Director of Nuclear Licensing Cont.onwealth Edison Company P. O. Box 707 Chicago, Illinois 60690 cc: Philip P. Steptoe, Esquire Suite 4200
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One First National Plaza Chicago, Illinois 60603 Dean Hansell, Esquire Assistant Attorney General 188 West Randolph Stre'et Suite 2315 Chicago, Illinois 60601 Roger Walker, Resident I'nspector LaSalle NPS, U.S.N.R.C.
P. O. Box 224 Marseilles, Illinois 61364 Chairman La Salle County Board of Supervisors La Salle County Courthouse Ottawa, Illinois 61350-Attorney General 500 South'2nd Street Spring. field, Illinois 62701 Department of Public Health ATTN: Chief, Division of Nuclear Safety 535 West Jefferson Springfield, Illinois 62761 The Honorable Tom Corcoran United States House of Representatives Washington, D. C.
20515 Chairman Illinois Commerce Commission Leland Building 527 East Capitol Avenue Springfield, Illinois 62706 e
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ENCLOSURE 1 l
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NUCLEAR REGULATORY COMM10SION wAsmwcrow. o. c. rosas APR 141582 MEMORANDUM FOR: Franz P. Schauer CMef 9
Structural Engine,ering'Dranch Division of Engineering THRU:
Dave Jeng, Section A Leader Structural Engineering Branc}h r
Division of Engineering FROM:
Rxuald E. Lipinski and Sai P. Chan Structural Engineering Branch Division of Engineering
SUBJECT:
TRIP REPORT - VISIT TO LA SALLE PLANT AND MEETING O HOLE-DRILLING AND CUT REBARS IN CONCRETE In accordance with your instruction, we visited t County i
Enginee,rs, Chicago e
Illinois on April 9 1982.
sufficient infomation so as to fairly assess the appli The purpose of this trip is drilling holes through concrete elements in response to the allegation ma by the Attorney General, State of Illinois.
At the plant site, we have idenEified and verified some groups of drilled drawings that the applicant submitted at the Marc at Bethesda.
We have also witnessed measurement and verification of dimensions to support the claim that the thickness of the roof slab of the off gas building is 12 inches.
On April 9 Engineers,1982, a meeting was held at the office of Sargent and Lundy of drilling (S & L), Chicago, Illinois to discuss the program of documentation and coring of holes and cutting rebars.
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The attendance list is
.We reviewed several calculations perfomed by the S & L by which the struc margins of the areas where drilling took place have been assessed.
S & L offices we established the following: basis of the informat On the 1.
The roof of the off gas building is 12 inches thick, and 2.
The controls and engineering evaluation of the effect of drillings were such that there is a reasonable assurance that they will not result in unacceptable degradation of structural elements."
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2-In view of the above we are of the opinion that the allegation filed by the Attorney General of the State of Illinois is not justified.
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S. P. Chan Structural Engineering Branch Division of Engineerir.g
Enclosure:
Attendance List cc:
R. Vo11mer D. Eisenhut J. Knight R. Tedesco A.,Schwencer D. Jeng R. Shewmaker (I & E)
F. Hawkings (Reg. III)
A. Bournia CONTACT:
S. P. Chan, SEB, X29534 l
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i ENCLOSURE ATTENDANCE LIST APRIL 9, 1982 CHICAGO, ILLIN0IS NRC F. Hawkins' Region III R. Lipinski NRR/SEB
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M. J. Morris L. O. Del George C. W. Schroeder S&L T.Linglais S. M. Kazmi V. Reklaitis K. T. Kostal L. P. Dolder I
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APR 141992 MEMORANDUM FOR: Franz F.Schacer,44ef structural Engineering Branch Division of Engineering-
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Dave Jeng, Section A Leader
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Division of Engineering
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Romuald E. Lipinski and Sai P. Chan
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SUBJECT:
TRIP REPORT - VISIl TO LA SALLE PLANT AND MEETING ON HDLE-DRILLING AND. CUT REBARS IN CONCRETE 1
p In accordance with your instruction, we visited the La Salle Plant a La'Salle
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County, Illinois on April 8,1982 and attended the neeting at Sargen and Lundy i
. Engineers,' Chicago, Illinois on April 9,1982. The purpose of this ip is to familiarize ourselves with the general design of the plant and to ather i
sufficient infortnation so as to fairly assess the applicant's practicp of i
drillin holes through concrete elemnts in response to the allegatiot made
.by the ttorney General State of Illinois.
i At the plant site, ye_hne_idhiIified and verified _some_ groups _of dtiged
> holes in the containr ent and auxiliary buildings.as indicated in the set of 4
drawihgs that the applicant submitted at t'te M2rch 31. presentation meeting q
at Bethesda. We have also witnessed measurcs. ant and verification of neveral
. dimensions to support the claim that the thickness of the roof slab of the 6
off gas building is 12 inches.
On April 9 1982, a riaeting was held at the office of Sargent and Luncy Engineers S & L), Chicago, Illinois to discuss the pJtogram of docume tation of drillin attached. gand coring of holes and cutting rebars2 The attendance 1 st is
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. We reviewed several calculations _ performed by_the S & L by which the ructural mergins of the areas where drilling took place have been assessed. O the basis of the information gathered at the site and during the meeting the '
S & L offices we established the following:
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The roof of the off gas building is 12 inches thick, and.
- 2, " The controls and engineering evaluation of the effect of dri111ngs
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were such that there is a reasonable assurance that they wilj not i
result in unacceptable degredation of structural elements.
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i In view of the above we are. of the opinion.that the. allegation. filed:by the Attorney General of the. State of. Illinois i.s.not justified.
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[ Docket Nos. 50-373; 50-374]
COMMOUWEALTH EDISON COMPANY (La Salle County Nuclear Generating Station, Unit I and Unit 2)
Request for Action Under 10 C.F.R. 2.206 Notice is hereby given that. hy its Request to Institute a Show Cause Proceeding and for Other Relief dated March 24,1982 (Petition),
the Attorney General for the State of Illinois requested that certain actions be taken by the Nuclear Regulatory Commission with respect to the la Salle County Units ~1 and 2 of the Commonwealth Edison Campany,.
in light of alleged newly discovered safety issues. Alleged safety issues concern the drilling of holes into structural concrete at the La Salle facility and also concern the structural adequacy of the off-gas building roof for that facility. The relief requested included institution of a show cause proceeding oursuant to 10 C.F.R. $ 2.202 and l
imediate suspension of further consideration of Edison's application for operating licenses. This request is being treated as a petition pursuant to 10 C.F.R. 2.206 of the Conriission's regulations, and accordingly, action will be taken on the Petition within a reasonable time.
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Copies of the Petition are available for inspection in the Comission's pubif e document room at 1717 H Street, N.W., Washington, ll NO g 0-g p
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Plant, Units 1 and 2 located at Illinois Valley Comunity College, Rural Route #1, Oglesby, Illinois '61348.
Dated at Bethesda, Maryland this 17 th day of April, 1982.
FOR THE NUCLEAR REG'ULATORY C0HISSION
- f Harold R. Denton Director Office of Nuclear Reactor Regulation '
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Docket Nos. 50-373 f; Mr. Cordell Reed Vice President, Nuclear Operations Commonwealth Edison Company
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Post Office Box 767 Chicago, Illinois 60690
Dear Mr. Reed:
Subject:
La Salle County Station, Unit 1 - Issuance of Facility Operating License The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed facility Operating License No. NPF-ll to Commonwealth Edison Company for La Salle County Station, Unit 1, located in Brookfield Township, La Salle County, Illinois.
License No. NPF-ll authorizes operation of the La Salle County Station, Unit 1, at five percent power (166 mega-watts thermal).
Authorization to operate beyond five percent is still under consideration by the NRC. The issuance of this license authorizing operation at five percent of full power is without prejudice tg future consideration by the Cgrmiission with respect to operation at po' er levels w
in excess of five percent.
Also enclosed is a copy of a related Federal Register notice which has been forwarded to the Office of the Federal Register for publication.
Two signed copies of Amendment No. 3 to Indemnity Agreement No. B-84 which covers the activities authorized under License No. NPF-il are also enclosed.
Please sign and return one copy to this office.
Supplement No. 3 to the Safety Evaluation Report for La Salle County Station, Units 1 and 2 has been issued.
Two copies are enclosed; 18 additional copies will be forward to you in about a week.
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Division of N.icensing i
Office of Nuclear Reactor Regulation
Enclosures:
1.
Facility Operation License No. NPF-il l
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Federal Register Notice i
3.
Amendment No. 3 to Indemnity Agreement No. B-84 l
4.
Supplement No. 3 to the SER (2)
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a Mr. Louis 0. De1 George Director of. Nuclear Licensing Commonwealth Edison Company P. O. Box 767 Chicago, Illinois 60690 cc:. Philip P. Steptoe, Esquire j
Suite 4200 l
One First National Plaza Chicago, Illinois 60603 Dean Hansell, Esquire Assistant Attorney General 188 West Randolph Street Suite 2315 Chicago, Illinois 60601 MagdrJWahhResident Inspector LaSalle NPS, U.S.N.R.C.
P. O. Box 224 Marseilles, Illinois 61364 D
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I LA SALLE C0dNTY STATION, UNIT 1 i
FACILITY OPERATING LICENSE License No. NPF-11 1.
The Nuclear Regulatory Comission (the Comission or the NRC) having found that:
A.
The application for a license filed by the Commonwealth Edison Company complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's regulations set forth in 10 CFR Chapter I, and all required notifications to other agencies or uodies have been duly made; B.
Construction of the La.sa'le County Station, Unit 1 (the facility),
has been substantially coq,;eted in conformity with Construction Permit No. CPPR-99 and the apo; cation, as amended, the provisions of the Act, and the regulativ of the Commission; C.
The facility wir, s, e. in conformity with the application, as amended, the provisi u of the Act, and the regulations of the Comis sio'n; D.
There is reasonaM e as urance:
(i) that the activities authorized by this operatir, Mce1n an be conducted without endangering the health and safety of y
, and (ii) that such activities will be con-
- h the Comission's regulations set forth in ducted in cortpliarc 4
10 CFR Chapter 1:
E.
The Commonwealth Eoisan Ctapany is technically qualified to engage in the activities autho ino by this operating license in accordance with the Commission's regi.3tiom:,et forth in 10 CFR Chapter I; F.
The Commonwealth Edison Cunpany he satisfied,the applicable provisions i
of 10 CFR Part 140, " Financial F ot-ctnn Requirements and Indemnity Agreements," of the Comission's reg aiutions; i
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G.
The issuance of this license will not be inimical to the common defense and security or to the health and safety of the public;f f
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Af ter weighing the environmental, economic, technical, bnd other i
benefits of the facility against environmental and other costs and I
considering available alternativ'es, the issuance of Facility Operating i
License No. NPF-11, subject to the conditions for protection of the i
environment set forth herein, is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied; and I.
The receipt, possession, and use of source, by-product and special nuclear material as authorized by this license will be in accordance with the Commission's regulations in 10 CFR Parts 30, 40 and 70.
2.
Based on the foregoing findings regarding this facility, Facility Operating License NPF-11 is hereby issued to the Commonwealth Edison Company (the licensee) to read as follows:
A.
This license applies to the La Salle County Station, Unit 1, a boiling water nuclear reactor and associated equipment, owned by the Common-wealth Edison Company. The facility is located in Brookfield Township, La Salle County, Illinois, and is described in the licensee's " Final Safety Analysis Report,"
as supplemented and amended, and in the licensee's Environmental Report, as supplemented and amended.
B.
Subject to the conditions and requirements incorporated herein, the Commission hereby licenses:
(1) Commonwealth Edison Company, pursuant to Section 103 of the Act and 10 CFR Part 50, " Licensing of Production and Utilization Facilities," to possess, use, and operate the facility at the designated location in Brookfield Township, La Salle County, Illinois, in accordance with the procedures and limitations set forth in this license; l
(2) Commonwealth Edison Company, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended;
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(3) Commonwealth Edison Company, pursuant to t'he Act and 10 CFR Parts l
30, 40, r.nd 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron l
r sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and?as fission l
detectors in amounts as required; t
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(4) Commonwealth Edison Company, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components;-
)
and (5) Commonwealth Edison Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations,- and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels not in excess of 5 oercent of full power (166 megawatts themal) in accordance with the conditions specified herein and in Attachment 1 to this license.
The preoperational tests, startup tests and other items identified in Attachment 1 to this license shall be completed as specified. is an integral part of this license.
(2) ' fechnical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environ-mental Protection Plan contained in Appendix B attached hereto are hereby incorporated in this license.
The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3) Conduct of' Work Activities During Fuel Load and Initial Startup The licensee shall review by committee all Unit 1 Preoperational l
f Testing and System Demonstration activities perfomed concurrently with Unit 1 initial fuel-loading or with the Unit 1 Startup Test Program to assure that the activity will not affect the safe i
performance of the Unit 1 fuel loading or the portion of the Unit 1 Startup Program being perfomed,. The review.shall address, as a minimum, system interaction, span of control,, staffing, security and health physics, with respect to perfomance of the activity concurrently with the Unit 1 fuel loading or the portion of the Unit 1 Startup Program being perfomed. The committee for the review shall be composed of a least three members, knowledgeable in the above areas, 'and who meet the qualifi-cations for professional-technical personnel specified by n_
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I section 4.4 of ANSI N18.7-1971. At least one of these three shall be a senior member of the Assistant Superintendent of
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Operation's staf f.
Resolution of Rebar Damage and Adequacy of Off-gas Building Roof f
The licensee shall complete its, assessment of the rebar damaged due to drilling and coring in concrete and.the structural adequacy i
of the off-gas building roof.
T.he results shall. be reported j
to. the NRC staff.for review and approval, prior to power operation i
following initial criticality and zero power physics testing.
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tr.ior.to cr.iticality.. thellicensee.shall submit-for NRC 4 3
approval,,.a.. revised listTof~s'afety-related Tridb'b'irsMo be contained inrTable'.3.7. 9-1xoft the Technical-Specificationsi
'to'TnclirdeTsTch'TriubbeF&o~n71hesWiHchis]6 JdiameTeiF"'6r less.
(% Prior to startup after the first refueling outage, the licensee shall provide, a; necessary, a revision to the Technical Specifications to renove snubbers that are detennined to be unnecessary and replace them with rigid strut and rod assenblies.
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Deferred Preoperational Deficiencies The licensee shall satisfactorily resolve those deficiencies which' we rFafe.f.er.t.e:d ffisiIlth' eye ople;r,a ti o,daHi{t] ng.p rag tam W75&fe'd'uie tha t"dhhlRa s s uFsi?th a t, thE t apaEfli ty tof*av ys tem 7eq'u fwd to be '
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operableibf'Techiittal"SiiscTfitstibn is not' degraded.
i f/t Surveillance of Tendons (Section 3.8.1*, SSER #3)
Prior to full power, the licensee shall supply the predicted lif t-off forces required to complete Tables 4.6.1.5-1 and 4.6.1.5-2 of the Technical Specifications.
% Masonry Wall (Section 3.8.3, SER,SSER #2) l Based on the fsindings of our preliminary review of the licensee's submittals and its commitments related to masonry wall evaluation, the following actions are required by the li.censee:
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(a) The present fixes for modifications implemented shall not preclude the option of implementing additional modifications if directed by our future review of the licensee's design l
criteria.
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- The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.
(
Prior to startup after the first refueling outage, the licensee shall resolve the differences between our interim criteria and the criteria used by the licensee to the satisfaction of the staff and shall implement the required wall fixes or modifica-tions that might result from such a resolution.
% Inservice Testing of Pumps'and Valves (Section 3.9.6, SER) 1 Pursuant to 10 CFR Part 50.55a, the relief that the licensee has requested from the pump and valve testing requirements of 10 CFR Part 50, Section 55.55 (g)(2) and (g)(4)(i) is granted for. that portion of the initial 120-month period during which we complete our review.
(h Dynamic Qualification (3.10, SER, SSER #1, SSER #2)
(k Prior to startup after the first refueling outage, the icensee shall complete any modifications or r'eplacement of equipment as' a ;esult of the fatigue evaluation.
br-6 M w4m;=the, licensee sh alt.d ocume nhthesoscu rren c e mfre ve~ryF s'a fetyneel:i efmv al ve-
.actuationeinto. the suppressionspooly-the=essoodated=eumulative damagedactors calculated-for2typicalwepresentative* equipment and keptmp-to-daterandereport<to NRCvanyanalfunction ofcequipment that occurs +tfue"tokany-safety relief < discharge.
(h Prior to startup after the first refueling outage, the licensee shall replace or modify the NSSS equipment (intennediate range i
monitor, C51-K-601 A/H and two-inch air-operated globe valve, Cll-F011) if the results of the requalification tests indicate either change is required.
(1 ) Environmental Qualifications (Section 3.11', SER, SSER #1, SSER #2) l
}
No later than June 30, 1982, the licensee shall be in compliance with the provisioni, of NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment", for safety-related electrical equipment exposed to a harsh environment.
(y Complete and auditable records must be available and maintained at a central location which describe the environmental qualification methods used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with NUREG-0588.
Such records shall l
be updated and maintained current as equipment is replaced, further tested, or.otherwise further qualified to document i
r I
complete compliance no later than June 30,19E!2.
l (M
The licensee shall complete the corrective actions stipulated in Appendix C to Supplement No. 2 of the Safety Evaluation Report by June 30, 1982.
.,. (12) Seismic and Loss-of-Coolant Accident Loads (Section 4.2.3.4, SER, SSER #1, 55ER #2)
By July 30, 1982, the licensee shall submit to NRC a complete i
description of the analytical methods along with all analytical I
results necessary to show that La Salle fully meets the criteria of Appendix A to the Standard Review Plan, Section 4.2 4
(NUREG-0800) with regard to fuel assembly liftoff.
% Prior to startup after the first refueling outage, the review of the fuel assembly liftoff issue must be satisfactorily resolved to the satisfaction of the staff.
(
Surveillance of Control Blade (Section 4.2.3.14, SER)
IE Bulletin No. 79-26, Revision 1, " Boron loss from BWR Control Blades," describes certain actions to be taken by licenstes to i
detennine boron loss from BWR control blades.
The licenste shall comply with items 1, 2 and 3 of this bulletin and submit a written response on item 3 within 30 days after plant startup following the first refueling outage.
(
) Scram Discharge Volume (Sections 4.6.2, SER and 6.3. 2.3, SSER #2)
Y J )' Prior to startup after the first refueling outage, the licensee shall incorporate the following additional modificaticns into the scram discharge volume system:
'J1 Redundant vent and drain valves, and
}%) Diverse and redundant scram instrumentation for each instrumented volume, including both delta pressure sensors and float sensors.
([ Prior to startup after the first refueling outage, the licensee shall complete system or procedural modifications, if required, as a result of the staff's completion of its review of the licensee's response to NUREG-0803.
(
Low Pressure in Pump Discharge of the Control Rod Drive (Section 4.6.2, SSER #2)
,i Prior to startup after the first refueling outage, the licensee shall install instrumentation for an automatic scram that would shut down the reactor in the event of low control rod drive pump i
discharge pressure to be activated during startup,and refueling modes only.
n
(1 ) Containment Long Term Program Load Specifications (6.2.1.1, SSER #2)
Prior to October 1,1982, the licensee shall submit its confinnatory assessment of the containment design adequacy for pool dynamic loads (chugging, vent lateral and diaphragm reverse pressure)
?
developed in conjuction with the Long Tenn Program and reported in 7
(
) Pressure Interlocks on Valves Interfacing at Low and High Pressure
~
(Section 6.3.4, 55ER #2)
Prior to startup after the first efueling outage, the licensee shall implement isolation protect,on in confonnance to the requirements of Section 6.3 of the Standard Review Plan against overpressurization of the low pressure emergency core cooling systems (RHR/LPCI and LPCS) at the h.igh and low pressure inter-face containing-a check valve and a closed motor-operated valve.
(
) Compliance with Regulatory Guide 1.97 (Sections 7.5.2, SER)
By July 1,1982, the licensee shall provide a plan for implementing modifications necessary to comply with Revision 2 of Regulatory Guide 1.97, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident,"
dated December 1980.
) Additional Instrumentation and Control Concerns (Section 7.7.3.4, SSER #1) j
- The licensee shall resolve the following concerns to the NRC staff's satisfaction prior to startup after the first refueling outage:
( )
whether common electrical power sources or sensor malfunctions j
l may cause multiple control systems failures, and
(
whether high energy line breaks will result in unacceptable consequential control system failures.
) Low and/or Degraded Grid Voltage (Section 8.2.2.2, SER)
The licensee shall install a second level of undervoltage protection prior to startup after the first refueling outage.
5
(
Reliability of Diesel-Generators (Sections 8.3.1.1, SER and 9.6.3.4, SER)
Prior to startup after the first refueling outage; the licensee shall implement tb following design modifications with respect to diesel-generator reliability:
I
.~.
., M A heavy duty turbocharger gear drive assembly be installed on the diesel-generators.
(h A prelube pump, powered from a reliable direct current power supply, be installed in the system to operate in g
parallel with the engine-driven lube oil pump, or an alternative acceptable to the NRC shall be installed to preclude dry-starting of the diesel-engine.
(c Controls and monitoring instrumentation be removed from
~
the engine and engine skid, except instruments qualified for this location. The non-qualified control and monitoring instnaments shall be installed on a free standing floor mounted panel and located on a vibration free floor area.
If the floor is not vibration free, the panel shall be equipped with vibration mounts.
(
) Direct Current Power Systems (Section 8.3.1.2, SER)
Prior to startup after the first refueling outage for the 125 and 250-volt. direct current systems for Divisions 1 and 2 and the 125-volt Division 3 direct current system, the following additional instrumentation shall be provided in the control room:
(1) Battery current (ammeter-charge / discharge), (2) Battery charger output voltage (voltmeter), (3) Battery charger output current (ammeter),
(4) Battery high discharge rate alann, and (5) Battery charger trouble alann. Jmthe2i nte r.imytheti.i cens ee=shal3 nimpi ement-
.apptcyed.dracedu esgo.-monitor battery 3 current;d>atter.yacharger f
e r
r ou tpu tavol t agesa nd zba tte rywh arge r.:,ou tpu tscu rrentra t-a the l ocal -
. panel sne t*l east *once9er.r.ei gh tyh our+shi f t.-
)
Reactor Containment Electrical Penetrations (Section 8.4.1,-
SER)
Prior to startup after the first refueling outage, a redundant fault current device (circuit breakers or fuses) shall be provided on each penetrating circuit that would limit a fault current
(
l surge to be less than the surge for which the penetration is l
qualified except for low energy (milliamps) instrument systems.
(
)
Separation of Class lE and Non-Class IE Cable Trays (Section 8.4.6.1, SER, 55ER #1, SSER #2)
I:
Prior to startup after the first refueling outage, the licensee shall provide adequate separation or barriers between Class lE
~
and adjacent non-Class 1E cable trays.
1 i
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9.-
(15b Fire Protection Program (Section 9.5, SER, SSER #2, SSER #3)
(Q The licensee shall maintain in effect and fully implement all provisions of the approved fire protection plan.
In addition, the licensee shall maintain the fire protection program set forth in Appendix R to 10 CFR Part 50, except for the following deviations:
(f) Hydrostatic hose tests in accordance with NEPA 1962-1979, and
(
)
No automatic fire detection systems in areas 2K/3K and 5B4.
) 4.riorato+4 ni ti alscri ticali.ty,-ahee l i c enseers hal l *i ns tal l.:4.
Ahourar.atedabar.r.ierens al l afour+si destof Aa apa rt-i al 4
+r.otecte d spowe rsca bl e ga n sca n d <as ge n e r al a s pH nk l e~rssy stem,
bothmlocatecti n'theNieselsgenerator. corr.idor.
Prior to startup after the first refueling outage, the licensee
~
shall' provide fire protection systems in fire areas 2C/3C, 4C3 and 6E.
(k) Prior to startup after the first refueling outage, the licensee with respect to fire doors shall implement one of the following:
(1)
Perform an engineering review of the manufacturer's certified doors and door frames by a nationally recognized laboratory to certify that the door and door frames provide the required fire resistance rating, or
.ii) Test a replicate "as installed" door assembly t,y a
(
l nationally recognized laboratory to determine the door rating, or (iii)
Replace manufacturer's labeled doors and door frames with UL rated items.
.?
h) Prior to startup af te the first refueling outage, the licensee shall demonstrate th,e adequacy of its fire protection for record storage.
~
(
) Radiation / Chemistry Technicians on the Backshif t (Section 13.1, SSER#2)
AHMt a die tion /6h emistrymTe c h ni cia ns mon ath e4 a cks h i f.tss h a1L he 1.r.a i n ed =pe r*t h e4: ass al l e.LswTr.a ni ni ng,,Q u a l j.fi ca ti o n d u i d e..All i
6uch=Technici.ans shabalsonhavessatisfactorilyw ompletedrthe s
falJewing= emergency responsentr.aining,.
( i ) =Tasksutoeberperfonned ida ri ngwthefi rs t%0rmi nu t e tT6T35'is eri ou s
. emergency:enAhe.backshi.f t;..
(ii)
Bos t-a c cid e n t, s ampl-i n g -a n d :e n aly si s :fo r:th esf.i rst :th re e ;ho u r,s ofwansemergency; (iii) d nepl a nt-rad.i.a tion rsurvey s-du ri ng,a n :acci dent; (iv)
Use.andMnterpr.etation-of both-por-table.and--f.ixedurea-radiation rnoni. tor,ing-equipment,,4uch as..the.Eberline P,ZNG-3.end-SAM-2; (v) -Inter.pretation.,of,.crJ,tical effluent moni.toringdata forf assisting r
n othe2 Chi efuEngineersduri ng sthetf.i rs tuhou r rof ra n da c' i de nt%in.,
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.,s ta ti o nm en tysnonitorsa nd <sta ndby:ga s ;1trea tmen temoni to r ) ;
w (vi)
Eirst. aid 4and bioassayy. techniques;7and, (vii)
Userof' respiratory equipment during -emergency.situatioqs.
(b) By June 1,1983, the licensee shall have Radiation / Chemistry Technicians onshif t for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day who meet ANSI N18.1-1971 or who are qualified in accordance with a NRC approved alternative program.
(3/4-Industrial Security (Section 13.6, SER, SSERf3)
The li censeechalkmai.ntain,4 nsef.fect,.and aful-lymi!nplement.J11 pro vi s i o ns -of rth e'Commi ssion!s t appro ved phys i c ai necur.i ty r pl a n,
guard training-and-qualification plan, and contingency ~ plan,
.i n c l u di ng =a me n dmen ts ea de.pu rs ua n t-to =the-author.i ty.okl 0 CF R
-50 54(p).* The approved plans which contain safeguard information t
are collectively entitled:
"La Salle County Station Security Plan Units 1 and 2,"
Revision 11, dated December 24,1981; "La Salle County Station Guard Training and Qualification Plan,"
submitted by their letter dated August 16, 1979, as revised in August, 1980; and "La Salle Nuclear Power Station Contingency Plan, dated March,1980, as revised by pages dated June,1980.
f The licensee is exempt from the commitment to fully implement those portions of the Security Plan as described in Items 1 and 2 in the licensee's letter dated April 1,1982, provided that th'e compensatory measures delineated in the above referenced letter are 'in place.
Compensatory measures as described in Item 3 in the April 1,1982 letter are approved with full implementation of the security plan commitments to be accomplished no later than July 1,1982.
[
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11 -
The licensee is exempted from the provisions of 10 CFR 73.55(d)(9),
but shall meet all' other commitments of the physical security plan and the following additional items.
(a)
Change all keys, locks, and combinations and related equipment used to control access to protected areas and vital areas at 3
least every 12 months.
j (b)
Issue keys, locks, combinations, and other access control devices to protected and vital areas only to those individuals who possess access authorization to those areas.
(c)
Change keys, locks, combinations, and related equipment to which an individual had access within 5 days and immediately for card keys after access authorization is withdrawn due to lack of trustworthiness, reliability, or inadequate work perfo rmance.
(
)
Initial Test Program (Section 14, SER)
The licensee shall conduct the post-fuel-loading initial test program (set forth in Section 14 of the licensee's Final Safety Analysis Report, as amended) without making any major modifications of this program unless modifications have been identified and have received prior NRC approval. Major modifications are defined as:
(a)
Elimination of any test identified in Section 14 of the licensee's h Final Safety Analysis Report, as amended as being essential; (b)
Modification of test objectives, methods or acceptance criteria for any test identified in Section 14 of the licens?e's Final l
Safety' Analysis Report, as amended as being essential; (c)
Performance of any test at a power level different from that described in the program; and (d)
Failure to complete any tests included in the described progra,m (planned or scheduled for power levels up to the authorized power level).
(
) Assurance of Proper Design and Construction (Section 17.4, SSER #2)
Prior to exceeding 5 percent of full power, the licensee shall-have. conducted an independent review of the mechanical and structural design of th'e loop C residual heat removal system, excluding all branch piping less than 3 inches, in the func.tioning mode of the low pressure injection system using loads [esulting from
=
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. the actuation of the automatic depressurization system in conjunction with the operating basis earthquake to verify that this system has been designed and constructed in accordance with all pertinent i
NRC requirements.
This verification review shall consider design,'
installation, inspection, testing, and any other aspects necessary to ensure conformance with the design. This review shall be performed independently of the licensee and its contractors who perfonned design and construction activities for the La Salle County Station, and it shall be completed to the' satisfaction of NRC.
(30) NUREG-0737 Conditions (Section 22.2) pedicens.e.e.shalkcompigte.,thedollowjng conditionsJo,_4by satisfactionwofsatheWRCs Mheserconditionsweference_sthe-apprq-pr.i. ate =6 tees mindection 22 27a5MIsAction'P1an=Requirementsw(or,,
Ap pl i ca ntsafo rm0 perati n g TIM c e ns e s ;61 n tth e ;Sa fe ty :Iv a l u atin n > R e po r t anddupplementst172;rsid 3?hUREG-0519a (jd Shif t Technical Advisor (I. A.l.1, SER, SSER #2)
The Shift Technical Advisor (STA) function shall be fulfilled by the Station Control Room Engineer (SCRE) who will be a designated SRO. However, if a SCRE is not available, the licensee shall provide a fully-trained on-shift technical advisor to the shift engineer (shift supervisor).
suer
( f Nuclear Steam Supoly System Vendor Review of Procedures (I.C.7, SER)
Prior to beginning low-power testing, the licensee must assure that the General Electric review of the power-ascension test procedures has been completed and the General Electric recommendations have been incorporated.
% Independent Safety Engineering Group (I.B.1.2, SER)
~
/
The licensee shall have an on-site independent engineering group.
(
Control Room Design Review (I.D.1, SER, SSER #2) i phe licensee shall correct the design deficiencies identifieh in Appendix C of Supplement No. I to the Safety Evaluation
(
t Report, NUREG-0519 on the schedule prescribed. therein.
]
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,. (e)
Training During low-Power Testing (I.G.I, SER, SSER #2)
At least 4 weeks prior to performing the Special Test, Simulated 1.oss of Onsite and Offsite Alternating-Current Power Test, the licensee shall provide a safety analysis for the test and its procedures to NRC for review and approval.
(
Post Accident Sampling (II.B.3, SSER g)
P r.io ndow r4 ticalitypoth e wlic e n se ers h all.d n s t al;lwand de s.t a.h ig h-ra d i a ti o nss am pli n g esy st em>f o r tobta i ni n g vea ctor
. coolant end"toWaTninsn't%tmosphere samplingrunderdegraded w
moregccidentNonditibnsWithout' excessive exposure.
(
Direct Indication of Safety / Relief Valve Position (II.D.3, SER, SSER #2)
Prior to startup after the first refueling outage, the licensee shall replace the safety / relief valve position indicator to a model that meets the IEEE Standards 323-1974 and 344-1975.
Additional Accident-Monitoring Instrumentation (II.F.1, SER, SSER #2), Nobie Gas Effluent Monitor
.Br.ic cdo w:rd ti cali ty r=th e rl ice n se e sh al.1 xi n s tal,ba nd ch a v e
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NRCWfor~ noble <gasteff.luentemonitoring
.proceduresuapproved Vy*the system.mtuplanteeffluent pathwayst
,, Sampling and Analysis of Plant Effluents
- 40 ratoscr:iti c ali tyMh e=1:ic en se e xs h a134 n s t al lmand s h a v e pro c e d u r,e s.A p pr ov e d yby scthe r N RC4o rira di ci o di n e ta n d vp a r ti c u la te samp1.4ng*andeanalysisnystem.at-/plantseffluent pathways.
h)
Instrumentation for Detection of Inadequate Core Coolng (II.F.2, SER SSER #1, SSER #2)
By July 31, 1982, the licensee shall submit a report' addressing the analysis. performed by the BWR Owners Group regarding additional instrumentation relative to inadequate core cooling and that the licensee shall implement the staff's requirements after the completion of the staff's review
~
of. this report.
y e
14 -
Proper Functioning of Heat Removal Systems (II.K.1.22, SER, 55ER #2, and II.K.3.13, SER, SSER #2)
The licensee shall implement the logic to restart automatically the core isolation cooling system prior to startup after -
the first refueling outage.
)
Modify Break Detection Logic to Prevent Spurious Isolation of High Pressure Coolant Injection and Reactor Core Isolation Cooling System (II.K.3.15, SER, SSER #2)
Prior to startup after the first refueling outage, the licensee shall implement a circuit modification to assure that transients monitored by pressure instruments to sense flow in these two systems actually sense continuous high steam fl ow.
Modification of Automatic Depressurization System Logic -
Feasibility for Increased Diversity for Some Event Sequences (II.K.3.18, SER, SER #1, SSER #3)
([)
By October 1,1982, the licensee shall evaluate the alternative design modifications of the BWR Owners Group relative to the logic for the automatic depressurization system, submit such evaluation, and propose modification to NRC for review and approval.
k Prior to startup after the first refueling outage, the licensee shall. implement the approved alternative logic modification of the automatic depressurization system.
% Restart of Core Spray and Low Pressure Core Injection System (II.K.3.21, SER, 55ER #2 )
Prior to startup after the first refueling outage, the licensee shall provide an auto start for the high pressure core spray.
Automatic Switchover of Reactor Core Isolation Cooling l
System Suction--Verify Procedures and Modify Dasign (II.K.3.22, SER) 1 Prior to startup after the first refueling outage, the licensee shall implenent the automatic switchover of the reactor core isolation cooling system suction from the condensate storage l
tank to the suppression pool when the condensate storage tanx level is low.
~
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e 9
, (
Upgrade Emergency Support f acilities (III. A.l.2, SER, SSER #1)
The licensee shall complete its Emergency Response Facilities as follows:
(
Safety Parameter Display System,
October 1, 1982 (M) Emergency Operations Facili ty October 1,1982 (i
)
Technical Support Center October 1,1982 Improving Licensee's Emergency Preparedness - Long Term (III.A.2, SER, SSER #1, SSER #2)
(d Prior to exceeding five percent power, the licensee shall complete a successful emergency exercise with the La Salle facility and La Salle County.
(7)
Prior to exceeding five percent power, a test shall be perfonned to demonstrate an adequate alerting / notification system.
Prior to exceeding five percent power, t'he licensee shall demonstrate that the state of offsite preparedness provides assurance that adequate protective measures can and will be taken in the event of a radiological emergency. The use of 10 CFR 50.54(s)(2) to specify a period within which corrective actions must be taken to assure an adequate state of emergency preparedness will include instances where NRC finds that the lack of progress in completion of the procedures in the Federal Emergency Management Agency's proposed rule set forth in 44 CFR Part 350 is an indication that major substantive problems exist in achieving or maintaining an adequate state of preparedness.
Any corrective period specified will relate to substantive problems identified by the Federal Emergency Management Agency.
$ The licensee shall provide the interim meteorological I
f improvement and shall provide the mechanism for long-tem improvements as follows:
~
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(()
Prior to exceeding five percent power, the licensee l
shall install a process computer with the' capability f
to retrieve meteorological infomation that provides a redundant means for data access'.
(
)
Prior to exceeding five percent power, the licensee
(
shall propose a plan for meeting the meteorological and dose assessment capability guidance of Appendix 2, NUREG-0654, Revision 1 as follows:
i l
installation of hardware and software capability y
described above by July 1,1982; and g
full operational capability described above by January 1, 1983.
(ii Prior to exceeding five percent power, the licensee shall include a description of the dose calculational methodology with a Class A transport and diffusion module, and a description of an acceptable meteorological measurement preventative and corrective maintenance program in the radiological emergency plan.
[ Exemptions from certain requirements of Appendices G, H, J, R and 550.55(a) to 10 CFR Part 50 and 10 CFR Part 73 are described in the Safety Evaluation Report and Supplement No.1, No. 2 and No. 3 to the Safety Evaluation Report.
In addition, an exemption was requested until the completion of the first refueling from the requirements of 10 CFR 570.24 and an exemption from 10 CFR Part 50, Appendix E from performing a full scale exercise within one year before issuance of an operating license, both exemptions are described in Supplemen't No. 2 of the Safety Evaluation Report. These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. Therefore, these exemptions are hereby granted. The facility will operate, to the extent authorized herein, in confonnity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.
[. This license is subject to the following additional condition for the protection of the environment:
Before engaging in additional construction or operational activities which may result in a significant adverse environmental impact that was not evaluated or that is significantly greater than that evaluated in the Final Environmental Statement and its Addendum, the licensee shall provide a i
written notification to the Director of the Office of Nuclear Reactor Regulation and receive written approval from that office before proceeding with such activities.
[. The licensee shall notify the Commission, as soon as possible but not later than one hour, of any accident at this facility which could result in an I
unplanned release of quantities of fission products'in excess of allowable limits for normal operation established by the Commis'sion..
The licensee shall have and maintain financial protection of such type l
and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.
[
l t
. H.
This license is effective as of'the date of issuance and shall expire at midnight on April 17, 2022; provided however that should the Commission, in conjunction with its consideration of the petition for Rulemaking in Docket No. PRM-50-30 determine that the tennination date for operating
- licenses should appropriately run fran the date of the issuance of a licensee's construction permit, the expiration date of this license will be September 10, 2013, effective upon notice to the Licensee of the Commission's action in this regard.
FOR TH NUCLEAR REGULATORY OMMISSION Harold R. Denton, Director Office of Nuclear Reactor Regulation
Attachment:
- 1.
2.
Appendix A - Technical Specifications (NUREG-0861) 3.
Appendix B - Environmental Protection Plan Date of Issuance:
April 17,1982
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l From November 1979 to February 1980 I was out on sick leave.
When I. returned to the site in Februarv 1980.-I worked as: a core, '.
I driller for Foley Electrical Co. until July 1980.
Durino. this. time.
period the precedures for contacting rebar were. chanced. ' h'e were instructeb to relocate small holes when rebar was contacted, and we were only allcwed to cut throuch the 2 char if accroial was civen I
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by an encineer.
Written reports were also made of each hole drilled i
durinc this -time period.
'I stcpped vorkina ht.the llSelle vle.nt as c _g c u. y 3.,:.,
_3 g 2 0, because c:a an 2.n3nry.-
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AFFIDAVII OF DALE G. BRIDE %3.tUGH 2.i
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3 STATE OF CALIFORNIA
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4:. COUNTY OF SANTA CLARA
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6; OAiE G. 3RIDEN3AUGH, being duly scorn, deposes and says 18i f ;; a s, I o.,.c ws :
3 86 1.
I am a Pro fe ssional Nuclear Engineer, technical t
s'Ojjconsultant, and
.1 founder and president of $23 Technical -
n 10 f,. As s o c ia t e s,
t e r' 1 consultants on ene rgy and environ.r.ent,
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- a. _.23 Hanilton Avenue, Suite K, San Jose, n::
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I. nave part.tcapaten as an e.xpert witness in n
13:; li:en:'ng procee dings be fcre th~c U.S. Nuclear Regulatory e.
1.;!! Co.r.is sic?n (NRC) ; have served as a consultant to the NRC; y
15[hiv: :es tiVied a t the request of the Advisory connittee on p'
ig;! Reactor Safeguards ; have a >peared be fore various co,?_e.it tees l1 L.
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g of 'the !,'. S. Congress and testified in varicus state licensing
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33.ianc reculatory croceec..ings.
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I an a graduate engineer thorou;;hly fnciliar with 1
f3 9n : the design, construction, and operation of nucle'ar generating
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_ o... lines anc t e c.nno;o;;y 4..
1 M
9-) in 1953, and havl since been regis te red in the state of Cali-9 %
D.
J fornia as a Prc.fessienal Nuclear E: gineer.
Fur the.r de.t ail s 9 u-
7
~
~
6 1.l of my expdience and o ;alifications a:e. contained in ny resuno 23 At tat.hnent 1.
?.....
. c-
.v 3;,
a.
tae purpose of this' Affidavit is to i den t.if.v 1,v.
.\\
4[, meer'.s regarding the adecuacy and qual'ity c'f construction, -
c:; c: certaan structures which make up an essential portion of G.I Contr.cnweal th ' Edis on Connahy sLaSalle Vucle ar Plant.
i I have s.
7 1 re,viewe d the Affidavit.of which describes' S;'j nuncreus e sses of anchor bolt. hole drilling and conduit -
i; 9: passageway core drilling in' the~ LaSalle Units 1 and 2 reac. tor da 10ll buildings during the period of June, 19 7 8 through.7ul.y, 1989.
3-
))
If, as is reported in avit, such drilling
}2;.wa$ CCnduC!c c so Ina t rein crcing stcel Ir. ccncre te Valls Was I
i t
-33l: danaged and/cr completely seve' red without the: b ene fit of
-4 3: antroariate structural anal.< sis, this would appest to me :to hp 3
. a e a con c i t i on wi t h no t enti al s a r.. ty s ign.r,a.
e cance and cne I.nat to.:
)g,jsheuldbe theroughly investigated at I.25211e prior to plant
~
ti criration.
l i.:
,i igj!
4.
I have no way of knowing whether the reported practice
- i. l.
10]has in fact jeopardized safe tf-related s tructures as I do not I.
0.!: have access to the exact 1ccations of the holes 'ihat w' re drili 9
e tl 21 Sfidavit, however, indicates that such. drilling 22jj r.rac tice s we re " usual" with the.asscciated implicati~on' that the 1
23[ii ?ractice was in connen use 3y a large numeer or :. ectrica7 crew 2.;,i,' working thro ugho u t the plant.
If th= practice was wiresp.;sd 1
\\
9...ar.d used bv all drillers during this tir.e period, it s c e.'.s ne a r g:{ ce rnin Jha t some safe ty-rela ted s tructenes (thc.se a urzi-r e du o
o.
15with systens er components assuring the integrity of the
.,S reactor coolant pressure boundaiv oz.' those necessaYy to
~
2 3;! main:-in the capability to shdt down the reacto r.and main:ain' 4 i, i t in a safe, shutdown condi Ion, or those nee'de t o > c e... a t.
it
~
ay or r_itigate the consequences o f accidents.<hich could result 1.
Gij in potential off-site exposures) would have be sn affected.
' l 7 ' If so, the S'..
associated cc.. age or degradation cf safety nargins of safety e tz.ted structures would appear to haic violated Dj; the quality requirements imposed by the U.S. code of Fedf rat 1
10(, Regulatien, 10 CFR Part 50, Appendi.i A, General Design Criteria n
))ij for Nuclear Power Plants and Appendix 3, Quality Assurance
- i 22, Criteria for Nuclear ?cwcr Plants and Fuel Repregessing F_ ants,
.a 13';_.t is a.f..s o p o s s ib l e,... ta.e practice was widespread to ts.e 14 j extent that it also was useci in the attachment of conyoniats s-.
15.: anc ecu..pment to the primary con tainmen't s trticture, th a t t.ne
.i il 16;!. integrity of tha.t structure could be a ffe cte d.
The I.aSalle g
- m. Nuclear Plant configuration includes a Mark II concrete 33 containment structure designed to contain and mitigate the in ID(: consequences of design-basis ' accidents that could occur durin;;
l
- 20. t'.e perati n of the plant.
The U.'S. NRC reviews the.
'l t
21..i ad:quacy cf this containment to assure its comp.:..ance wit.a 1
22j fe de ral regulations.
Stands.rd Review Plan 3. 5.1, Concrete i
l et mE Ccntainme nt, discusses the points normally cove red by the
'-U*.
- o. "; N?.C in such review.
The ir.p a ct of the drilling oper-ticas e
-l..
9 5 d
- 3 a r i 3 c.i in Affi:'a.-it u ul be rele.act
_,, i p& 3.
a.
' A n1*- m-*U % R ~~ U
~
g P a ge 3. S.1 -14, which covers materials, ' quality control, a 8
2* special. construction techn iencs.o f concre te co.htniniact.t.
a.
alli The concepts expressed in this.Revid Plan. applying to ce 4 ;. containment tvou!.d also apply to the structural in e2rit7 5;l.other concre te safety-related walls and s tructures.
6d 5.
I have been informed that sone of the facts cent e.
7';i in.
.i.
y Affidavit have ocen verbally co.w.unicai s
O ji.t o t.2 e U.S. Nuclear Regulatory Commission as called.for. it 1
0,g10 C.:R Part 21 (U.S. Code of Fede ral Regulations),.but th,
~
~
10!i. inve stiga tion has 3 et been reported.
I have also been in 11:' that the U.S. N.7C has been verbally informed (by an unids a
12 e r.ploye e) t h.=. t the concrete rcof slab :r.nking up the ceili 13:n.t.
n
!.a S a..11e o,t gas builc..ing is below specir:.ej t..nic. cess
- 1 1.il.i contains numerous holes and c_acks.
I have furthe; 'ceen
.i 15j. informed that the.N'RC's response to the report of this a
3g!! condition was that no inves tigation or this condition was i
- 5.,e--,.,
.d wp-t a i
ie 1
n' o " - h- ' v. -ha-
~
t i
a l" o :
-t r
t..
' e o-33.;! gas building roof would result. in a "calacitous" accident
- a..
)gg it coe s cca ta n. e qu:.pment and components handling radicac
.i 90,d g a s e s.
The p rina ry s ign:. 1c ance, how. ve r.,. o f the, tcporrec
,l t..a i s c o nci t i o n oy t r.e
.N.R'.,
97
- allure to investigate is Inc
..i.
99] cuestion that it raises as to the e fficacy of the entire gif quality oversight function conducted by the NRC cn.the 93.:overall construc.icn of the IaSalle Plant.
This question 9 o-p a z k e s m o r e urgent the necessi y to resolve the repcrted 2G:!///// -
~
~
e,.
~
't de ficiencies that may exis t in the re5ctor builcbing (sad 7
2.' cthe r s tructure s)..
3-6.
Prompt action to investisate these c.once rns,is
~
icportant.
It is my unde rs tanding that the LaSalle Unit 1-5',' operating License is about to be issued which would permit r) i-. t..e loadi.ng or fuel into t.ne reactor and init.t a,t c.oer2tions c
it
~
- i.! to begin.
While fuel loading in itself is not likely to chang 3h the loading conditions of the potentially affected ; structures
'lg!jso that a failure would be expected, fuel loading does.
10;i r* p re s e n t a point in time that is. of significance in the
.s
.I 11 a prope r conduct of the investigation that may be ree,uired.
- I 12'i.nen rue,
.osding occurs and low power ope rats.on is po s s :... e,
.o.
e certain areas or the plant, inc, uc
.g pert.ons.e:
t 131 acce ss and reactor building must be con tr.olled
.t.i, prinary cont ainment 1..
-il from co=cartcent tc 10.and/or minimized and the free movement e
l gg,;cc=partment by investigatorf persennel eculd be restricted.
f g,;5ubsequent pcVer operation of the reac:or could mak: physi:11
. as 18:.t ac ce s s to some c.ortions of the racility impossiale or at ic
~
s 1c.,s;, e x t reme ly 1imit e d.
7 The consequences of the degradation cf the' s tructurc 90{
':iaualitv potentially represented by the se.verance o's reinforci 21 i-tisteel in the concrete walls is the pctential failure of the e4
_n, j y
- g. structures anc/or systems to perforn their rafety related l*
- . l..f tr.c tions unde r a ccident or seismic cendir. ions.In cv. coinie
.o. t.'.
2 Inat a tnorcugn :.nve s tigat.on ce nace ey tr.e O
- 1t is essentia, 93 authorities of che alleg.i.tions raised.
This v: <
~
g,apprcpriate s
)
~
.y 2,
a 1
assure that dw.a,qe to the essentisi structures, if it in fact' 1
2jexists, has been properly analyzed by appropr'inte. technical'.
3'!, txperts and repairs er modifications are.c.ade if needed hofore
~
4}these. safety systens are called upon to p: event er mi:igstc.
E:!the consequences of an accident.
.j i
U:
_gt t si S N, Dalb G. Iridenbaugh I
Ab.
?
m 9;l il 10 j'.. March 17, 1982 11....;! Subscribed and sworn to before 2
I2 r.e this /7 day o f ///f77C[ 193 2.
13 !
1; a
14' 84
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15 :'
NOTAR'T PUM I
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10:.l.'%ccr. mis sion expires :
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- c. 2
- 3..:.1 ' 3
- r..t n..e )
s c. :. _. s. o
...e. r.s. _r.r. w.. r..
- . 9 i 6..
- c..o r h r.p_
M~3 Technicci Associate:. S c: J o s a, ' C a li.~ o r n i a.
?residant Cc-f ounde r and partnar of t echnic al cens ulting f irs..
S p e ci Alis t in energy c on s u l t in g to governmen:al and o her groups in:::asted in evalea:Los of nuclear plant sift:y and lic ensin.
C o# sti:a d in this capacity to s: ate agencies in Ccliforn'a, N ew Yo rk, Illi ncis, New Jersey, Pennsylvan'a, Oklah::a and Minnesc a and to : '-
Soracgian Nuclea: Pover Ce:sittee, S w e d is h N u c *.c a: Inspe:ters:e and various c th e r c:ganizacicas and environnental
- cups.
Fe:-
for:ad entansive saf ety analys is for S u cdis h Incr5 y Cc==issien he n < o.,
.n C o,... n.
o :...
s.
. s..
- .u.'>,,..
. e.
c ccc. <....ancs.
ed u..-
n~
. a VASE-1400.
Consultaa to the V.S. NRC - 172 S a f e t y I =p r v e= er. :
Progra=, perfor=ed Cos
- .
- c l y s i s cf Spen: Fuel Dispesai for the N a : u r.1 Resources Da f an s e Council, and con:ribu:ed to the Oept c at cf Incrgy LVA Enfa:7 Improvement ?r.c; a=
for ' an d ' r lab ort-
- cries.
Servec as axpar:
- d. : n e s s in N?.C and s tire u t ility
.cc:=ission hearings.
e C n-vs a
- i. 3...e3
- :.3,.v. s. 2..,
C c.. '.. m-
?.. o, a.__c. 5 ". v ' v c '., ". : ' o Al.c.
Ca'4_'
4'.
i Volu :ee r vo:t en Nuclea.: S a f egu ards I t '.t i s t iv e ca=paigns in Calif o rnia, 0:ege, W a s hin g :ca, A: trona, and Colorado.
.T u = a r c e ;
p res ed t a t ion s en nuclac: power and alternative energy op:i=ns.
civic, govern = cat, and college gr:ups.
!.i s o resource p ercen fo' public sc vice presenta:Lons en r i-d i o and t'el e v.i s io..-
1973 1976 Manager.., a r :.or:ance svalua:. o n-and.spreve=ent, Geners,
-,ec::
- t Cc= :an v - Nuclear Energv Divisien. 5en Jose. Califor.iz, Ma: ged sz.anteen techn':e-a:d sevec clerical perscnnti vi h r e s p on s ib ili:y fer estab*ishncn: ' tad =cntgement of syste=s ::
son-t:: and =4csura 3 o i* in g L c; Reac:c: c-uipmen: cnd syste:
.o ua,
._.. ' a -. *... - =.... -...
ope
.,o
,a, e
r.s.
._.s
.e.
eg s..s
.o p.h-M.;a o d n.#f
- c o r d in a t e d cc:re :1on of caus i.'.:i:1cus in custocar ct.:tr m. a'
.,:'c
- :o in c. r e v e reliah111:v 1:d :e:
cf fpreed f
a w r
1956 - 1953 T i e l d En r i. e e_r, Ceneral Electric Co_::p an r. Iestallation an.1 Service E r.,in e e rin t Department. Chica :o. 1111n s i s.
S u p e rv is e d ins tallation and =aintenance of steas thrbines ef.all sides.
Supervisad crews of from ten to =cre than :n= hun ~ dred men, detending en :he job.
'Jorked p rima r ily vi:h icraa utilities.ht:
had si nificant verk with steel, petroleus and o ther process in du s tries.
Had four years of experion. e a: con s truc tio n, s:artup, t ro ubic-s hoo ting an d re f u elin;;.cf the fir 3. lar.ge-scale con =crcisl nucles: power unit.
1955 - 19,55 Ea t in e e rin g Trainin: Progran, General Electric Oc 5snv, Erie.
P enn sy lvan ia, and Scheneccady, Ne w York.
T ra in in g assig= cents in p lan t facilitics design and is s :e ch.
t u rbin e testing.at two General Electric Tactory locations.
1953 - 1?55 Uni:ed States Ar=y Ordnance S c h o o l,. A b a r_d c e n, M..ryland.
Inst:uctor - Heavy Artillery Rep'eir.
Taught classroom and shop disasse=bly of crtille-ry pieces.
.s.
s..
l
,En e 1: e e rin :; [rainin?
Procras. Ganc ral Elec tric Cc=phnv. Evendelc.
Ohio.
(
T rai ing as sign =e=
vith Aircraft Gas Turbi:e Dep:rt= apt.
E'UCATION & AF FII.l ATIONS :
~
- 3SME - 1953, South Dakota S chool of Mines an'd Technology,,
' Rapid City, Sce:h Dakota, Up p e r *< o f c la s s.
P rof es sional Nu clear Engineer - Caliiornia.
Certificatt No. 0973.
Mesber - A=cricas. Nuclear 3ccia y.
Various Cc=pany Training cou rs es during career in cluding Prof es-sional 3usiness.M an a g e = c: :. Xcp=ar Tregee Dc:ision Making, I!!e::1J Presentatica, and nunercus t n :hnic a l ra = in ats.
~.3-
~73 1975 (can:d) e F.e s p on s ib l e for development of Division Mas:dr P erforhang e
~
Improvecent Plan as well au for nUnercus S t af f spe cial es si:n-
=ents en l o.S g - r a n g e s tu d ie s,.
yas en s p eci?.1 as sign =ent' f o r'ths f...
canagemen: of itvo. dif f erent id'.Voc pioj ec ta - f o r=hd to 'Yaxalve.-
unique technical problens.
1972 1973
~
Manager. Frnduct Service, Gen e ra'l E le ct ric Co=uany Nuclear Enerzy Division. San Jose, Calif ornia.
~
~
Mancgad group.cf twenty-one technical and fcur clerical personnel.
Price responsibility was to direct interface cad liaisen personnel invo.1ved in corrective actions required under contrac e warractics Also in charge of re fue'lin g an d s ervice p l acn in g',
perfor=aned an alys is, an d. s e rvice c om' un ica:1er.. fun c tion s supporting all com-
=
~
placed cc=mercial nucicar p~ ver reactors uupplied by General o
Electric, both domestic And overs ecs (S p ain, G e rmany, I taly, Japsu, Indis, and S witz e rlan d).
Li'E - Ic72 Mencrar, ? roduct_ S ervice._ Cen e ral Else t ric _ Cc=pany - Nuclear Energy Division, San Jcse, California.
Managed sixteen technical and six clericci persennel ith the r e s p on s ib ili:7 for all customer contact, p l ann in g cud execut d ep ar t m en t'.f on of vork required after the customer acceptanca.o!
supplie d plan ts ' and /or e qu ip m en :.,This in clude d quo:x tien, sale and delivery of spara and receval parts.
S ale s volune' of p ar:s in c~rc a s e d f r o=' $ 1,0 0 0,0 0 0 in 1968 :n over $ 3,000,0 00 in 1972.
Lc'S - Lc6S Manecer, Co= plaint and Warran ty S e rv ic e. Ocneral Electric Co::any -
Nuclear Enerry Div i s io n. San Josa, California.
M,anaged g.roup of six persons with the re spon sibili ty f or cu s toccr contsc:s, planning and crecution of verk required after cu s toner accepta=ce of dep ar t=cn t-s upp lied p latis and/or equip = ant--both donestic a:d overseas.
1953 - 1966 Tield En g ine e rin g S up e rvis or, Coneral Electric C o = o any. In s talla tion and Service Eneineeritz Departnent, Los Anteles, Califernia.
Supervised a p p r ox in.'.t e ly e ig h t field represen:ctives vi:h res;cazi-bility for General Ele ctric s:ca a d gas tu rb in e ir.s t all a :ica and
=ain :en s= c e work in Southern California, Arizoni, cnd Scu:her:
N:vada.
During this period was responsible for :h e' in s t alla:iun of eigh: different c'entral station steam turbine g enerator units, plus l
= u c h =.. in ~*. e n a n c e sc:1vity. _ Work in cl'ud e d cys:oner c:::at: ty c; 2 -
b.
t
~
- ';
- : cts & A v a ?_.D.s..:
S igna Tau - ?.on o ra ry En gin c arin g F:s te rnity.
General Managers Avnrd, Cenaral Ele.ctric Conpany.
FIRSONAi DATA:
3 ;:-. N ove=S e: 20, la31, Miller, South Dakota.
Marriad, threa children 5'1",
190 lbs.. he alth - e:ccellen t Honorable d i's ch a r g e fron 'Jnited S ta tes Ar=y Hobbies:
Skiiing, hiking, work with' Cub and Roy Scout Groups.
?GlICATIO:is & T E S TTMONY :
~
1.
O p e r a t i_n g and Maintenance Experience, p re s en ted at Twelfth Annual Seminar for Electric Utility Execu tives, ?ebble 3 each, California, October 1972, published in Gene ral Electric NEDC-10697, Decenber 1972.
2.
Maintenance and In-5crvice Insoection, pre s en ted at IAEA Sy=pos tun on Experience Fron Operating and 7t:eling of ';uclear
?cver P lan t s, 3ridenbaugh, Lloyd & Tu:nc:. Vienna, Austria, O c t o b e r,- 1972.
3.
Opdratinz and Main tenan ce Exn e rienc e, presented at Thirteenth Ann u s 1 Seninar for Electric Utility Ex ecu tives, ?-bble 3aach, California, Novenber, 1973, published in General -lect rie
- 4 E D 0 - 2 0 2 2 2, Jaduary. 1974, F
4 '.
I n o'r o v i n : P lan e Avn11ch tlity, p re s ent ed s t. Th ir t'e en th Annu s1'
- S e = in a r for Electric U t ?.lity Ex ecu tives, P chble 3 e ach, C a'li-fornia, Nove=ber 1973, published in Cen cral Electric NE DO-20222, January, 1974.
.3.
Aeolie.trion of ?lant Outace Exne rience te Inorova ?lan Per-formance, E ridenbrugh and E urdsall, Anerican ?ovar Confarance, Chicago, Illinois, April 14, 1974 6.
Nuclem _ Valve Testine Cuts Cost, T im e_, Ile c ':ric al W o rld,
October, 15, 1974 7.
The Ricks of Nucles: ?over ')cictors-A Review of the NRC
^
Reactor Safety S t u d.: W A E E_- 1 4 0 6, Kendail, Eubberd, Minor &
BrideAbaugh, et il, for the 'Jnien o f C onc e rned S cian tis t s,
August, 1977.
-4
4 s
e
.~
e 3.
Swedi:5 leae:or i a _~ e
$ :. u.'
,,. _ J e abf.i Rink A5:4 s s = en :,
MH3 Tecanical As cocia:2s, January, l 'i,7 3.
( ? 6110 hed by the Svedish C ap a r := an t of Indus:ty as Docunc=t Ds! ! ? 73 : 1) 9.
Testineny of D.G. Bridenhau;h, R.3.
- d u b b a r d,
C.C.
Minor :o.
the Calif.3rnia' Str:c Asse:bly Ce==intee en Resources. L ur.d L* c a, and Energy, March 3, 1,976.
10.
Te s ti=en y of D.G. 3 ridenb augh, 1.3.
!!u b b a r d, and G.C..Mi=or before the Un i t e d S ta tes Con g re s s., Joint Coc=ittee c Aro=ic.
Energy, Fab ruary 15, 19 76, ' a s hin g ton, DC (?ublishad by th e Union of Concer=ed Scientic
, Ca= bridge, Massachuse::n.)
11.
Tr.s t i=o ny by D.C. 3 ridenbaugh 'hef ore the California 2nergy C o c = I s s i o'n, entitled, Initia tion of Catastrcphic Accidents at Diablo Canyon, E c Arin g s oc Z=ersency.?TEdning, Avila
~~
Secch. Califdrnia, Nove=ber 1,
1976.
12.
Testi=ony by D.G.
Eridenbaugh before the U.S. Nuclear Regula-
- ory Cc==is s io=, subject: Diablo Canyon Nucles: ?lant Perfer-
=cade, At==ic S afe ty and Licens ing 3 o ard He a rin g s, Dece=bar, 1975.
13.
Testi=ony by D.G. 3ridenbaugh hefore the Calif o rnia 'In c rgy Cc==1ssion, subject: Incari: Soent Fuel Stors;c C ons id e ra ti an s,;
March 10, 1977.
I4.
Tes tine ny by D.G.
3 ridenbaugh before the Se-York S tata Public Service C o== i s s ion Siting S cird -Kearing s ~ c e n c a rn in g th e James-l por t Nuclear Fouer S ta:ica, subj ect: Effe.ct of Technical and 54fety....D e f i c i en c ie s on Nuclear Plant Cost and ?.e L ia bi l i ty,
i t
l April, 1977.
15. ' Testicocy by D.C.
Bridenbaugh before the 'Califcznia 5ts:e 1
E n e r
.v Cc==1ssion, c ubj ec::
Deco==insionine of Pressuriced o
l Watar Re ac ters, S u= de s er t.N u clear ? l.in : Hearicas, June 9,
?
1977.
1 16.
Testi=ony by D.G. Bridenbaugh b'afore she California State l
E n e r,37 Cc==ission, subj ec t : Econocic Relationshion of De c o==is s ionin g, *S unde s e r t. Nuclear Plant, f o r the Naturci
- Refobrees De f.2 ns e Council, J'uly 15, 1977.
l 17.
Tes sinony by D.G. B ridenb augh b ef o're the Ver=en t State ' card S
of Health, subj ect : Ooera:Lon of Vermont Yankee Nuclear ?lant and Its T=oact on Public Health and S af ety, October 6,'1977.
l
'18 Te s ti=eny b y D. G. Bridanbaudh before the U.S. Mu clear Regula-l
. tory.Consis s ion, A tomic S'af e ty and Lice =s ing 3 card, s ubj e ct :.
Deficic=cies in S af ety Evaluation o f N
=-S ei' =ic 'I n su as, Lack s
of a De f ic i_t iv a Finding of S a f e t y_, Diablo Ceayo: Nuclear Units cc ober 18, 1977, Avila 3each, California.
l l
l -
r' t
e
/
= s.
. e 3,.
3 S. s.. C..
- %... 3 *. 4 J m.
%. a g. *.
- b. a.. f.. -. *. f
- *. -. 4 g.., 4 g=
. s. 4 s.
s.
v.
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.f
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- e... z... p.. s.
..n
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s.
. _...,..u
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ii.
Te s ti. on - by D.C.
3 ridenbaugh b ef o r= :*c ?.cuisia*na.itete.
I.e ;i s l a t u r e Cen=ittet on Natural Rasources, s u'2 f 4.c e : N e c'l 2 a r Pever F lan : Deficiencies I nae:ine :n Etfe:v r. Ecliabilitv, Baten Rou g a,. Lcuis,iana, February 13, 1975.
~
21.
S : e.n : T u e.1 Discosal Costs, report c re p a : e.d by S.G. 3 rid,eu'itup for he :.atural Resources _ De f ens e Ccuncil (NROC),.Wyus-31, Lvi4.
22.
Testi=ony by D.G. 3r'denbaugh, G.C. Minor, and R.3. Euhhard b.efore the A:osic Safety and Licensing B oard, in che =.::ar -
ef the 3 'ack Fox Nuclear P over S tation Con:. ruction ?c =it Hearings, S ep tenb ar 25, 1 9.7 3, Tulsa, Oklahona.
23.
Ta s :i=e ny o f D. G. 3 rid anb aug h and R.3. Hubbard bef a:e the Louisiana Public S c:vice Consis sion, Nudlea: Plant and 'cus:
Genera: ton Costs, Nove=ber 19, 1 7S, Ba:on P.o u g e, icuisicac.
2 '..
Testimony by D.G. 3ridenbaugh befora t h e' C i ty Council s..d Ilectric Utility Cc==ission of Austin, Texas,
_D_e s i ; n. Cen-s t : -i sn, and Oceratine E xpe r iun :: e o f Xuclea r Ganera:in t Facui es, Dece:her 5, 1978,' Austin, Texas, 25.
Te i t hen y by D.C. 3ridenbaugh'for the Cermonw ealth of' has sachus e t: s,. De p a r:=en t o f Public Utilities. :::. c : of Ca r a s c i-ia d S a f e : r Issues. Generic Deficiancies, and Ihree Mile i s l an d-In i t ia t e d Mo d i f ie r. t ion s on ?over Genera-ion. Ces t
- _a : the Preposed Pilgri=-2 Nuclear Plant, Juna 5, 1879.
26.
Incr: in: the Safety o f LL* R P ow e r ? l 2n t s, M33 Technical Aasociates, prepared for U.S.
Dept. of Energy, Sandia
'.'a b e :4 : c r i e s, Sep;csber 28, 1979.
F 27, 3 7 ?. ? i : c. and No::lc C r.'.e k s, EH3 Technical Associatta, fcr
- ha S ve dis h Sue le a r ? cuer In,s p e c:cr a :e (5KI), O'ctebir, 1979.
C 22.
Te s :1:en y of D.G.
3 ridenbaugh and G. C.
Min o r.b cf ore.thu Aconic Safety and Licens ing B o ard, in th m a :'t e r o.f S ac ranen :o Municipal U:ility Dis tric t, 22ncho Seco Nuclear Can e ra ting S ta tica f ollowin g TMI-2 accidan t, subject:
~ ~
Operc:or Trainin: and Hunan Facecrs En 2 inc e rin ', fer the Cali f o rn ia En ergy Commis s ion, February 11, 1930.
29.
Italien Reacco: S a f a. c y Sendr:
Caorso. R i r *<. /. s s e s s= en t, x23
, Tachnical As socia tes, for Friends of the Ear:h,, I cely, March, 1980.
D 30.
Decen:a=ina ton e f I r?o :en-8'3 from ~~nree Mile I slan d ';u cle e r
- Plant, E.
- Kencall, R.
Pollard, &
D.G.
3 rid : b at;h, e:.t l,
The Union of Con.c r=cd S cien tis ts, da li ee :td to -he C itraer ca ] n -;m=M <
?w
s
.. e 31.
'D e.c e n : 2 ina ien of
.rve'ren-?f ron l h.e s e Mile 2212nd Nu-las:
o G.U.
3ridenhaigh, Llan:, ii. Randall, R.
,P o lla r d,
4: 21, Zhe inion of Concernad 3cizati :s. dalivaiad to :ha Ocvarner of ? a= c u y lva t:ia, May 15, 1980.
32..
test 1=ony by D.C.
3:idenbacxh before the rev.Tersey 3 card'uf
?ubli: Utilities, on bahalf cf Nav Jarsey Public AdvocAca's Of f ic e, D iv is i o r, of Ra:e'Couauel, Ansivsis of 1979 Salee-1 R e f u e l_in i Ou.:.ne, A.u g u s t, 1980.
33.
M8 anexa ta Nuclear Plants Gasenus Emissions S t u d '.. ME3 Tacheica:
Annec t a res, for Minnasota ?ollu ton Ce ntro l Ag ency,. S ep te=ber,
1980.
31 Posi:1.ca 5:stacent,
_P r o c o s e d Rule =aking__
of Position cn the S tora 2 e n1 Dis _ e s al_ of Nu clea r Wa s t e, Joint Cross-Statenent of the Nav England Coalition on Nuclear ?ollu:isn and the Natural Resources ~ Defense Council, September, 1980, -
35.
Testineny.by D.C.
3ridenbaugh cud Gregory C.
Minot, before tha New York S ta te ?ublic Service C o=n is s io n, In the Macter of Long Island. Lighting.Conpany Temperary ?. ate Case., prepared for the Shorahan. Opponents Coalition, Septe=ber 22,~ 19'30, Shoreha: Nuclear Plin: Cen s:ru c tio n Ec hedule.
36.
Su2ple= ental Tes timony by D.G.
3ridenbaugh before the New Jersey Board of ?ublic Utilities, en behalf of Neu J ersay P u b l. i Advocate's Office, D iv is io n of Ra:a Counsel, Analysis of 1979 S a l en-1 R e f e e_lin e _ Ou t a u e, Dece=*:er, 1980.
37.
Tes inony by D.G.
Bridenbaugh and Gregory C.
Minor, before the New Jersey 3sard of Public Utilities, en behalf of Scv Jersey Department of tha Public Advocata, Division of late Counsel, Ovstar Creek 1980 Refuelin Outace Inv e s tiz a tiot February, 1981.
33.
Econonic Assessmann:
Ownershin In t er es: in ? alo. 7arda Nuclear 5
~
S ta tio n, M.H 3 TAchnical A s s o c ia t e. s,. f o r T h e. C i ty o f Riversice r September 11, 1981.
39.
Tertinony.of D.'C. 3ridanba'c;h before the Publ'ic Utili:ies Con =ission of Ohio, in the matter of the Regulation of the -
Electric Fu el Compenent Con tained Within ' the Rate Schedules.'
of the Tol' ado Edison Company a=d F.ela t ed Ma :ct s, subj ec::
Da vis-3 es s e Nuclear Peyar Station 1980-S1'Outsee Reviev, October, 1981.
40.
Sup pleu en tal T e s tinony of D.G. ~ 3ridenbaugh before the Public U til'i rie s C o==is s i o n o f O h io,. 15. 't h e' = a t t ar of tha Zagula tisE of tha 21ectric yuel Component Co n ta ined within nc Rate
. i I.
Schedules of the Toledo Idison Co=pany and Rela ted Ma ttees, November, 1981.
Nu c l e_s_: ?over Stztion 1920-S1 Outa2a Rev.
sub.icc:: Davis-3csse d
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.\\ s s u c ia r. e for =ha 3..e( Qs ;ue.gg;
?uwa-Inspec: ors:a (5KI), Ja - :. : 7. ii3 2.
42.-
Tes ti=cny* cf D.c.
3 rid enbaugh and Gre.go ry C.
- ".in o r on behalf of Gov rno: I d u.a n.d.
G.
Brown Jr.,
Sefera tha Atcaic e
Safety and Licenstr.g 3oard, r e n s = 4 f.nz. c on r.o n t io n 10,_
P r e s sur i z er Haaturs, January 11, 1982.'
L3.
'T as:inony of D.C. 3 idar.baugh a r.d. Gr eg o ry C. Mino r on b'5 half of Governor Edmund G.
3:c a,
Jr. bef ora :ha Acca te Stfecj and Lic e nsing Board, regarding C o n t e n t i. e n 12, 51ock ar.d Pilot O c e:a ted Relie f.Valv a s, January 11, 1932.
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- c UNITED STATES NUCLEAR REGULATORY COMt.ilSSION n
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MEMGLEDUM FOR:
R. C. DeYoung, Director, Office of Inspection and Enforcement FROM:
James G. Keppler, Regional Administrator, Region III
SUBJECT:
LA SALLE COUNTY NUCLEAR STATION - PETITION FROM ILLINOIS ATTORNEY GENERAL As you know, on March 24,.1982, the Illinois Attorney General petitioned the h"4C to suspend licensing procecdings at La Salle pending investigation of recent allegations and to institute a Show Cause Hearing with Illinois es a party to the Hearing.
The allegations deal with the overall adequacy of safety related structures as a result of widespread rebar cutting and specific structural deficiencies in the roof of the off-gas building.
A conference call was held on March 29 involving Messrs. Denton, Case, Stello, DeYoung and Keppler to discuss the handling of these' investigations.
We agreed that, because the petition expresses concern that the off-gas building deficiencies had been verbally coramunicated ' earlier to NRC and that the NRC had concluded an investigation of these alleged deficiencies was not warranted, it would be prudent to have an independent review of this allegation by IE (since IE was not involved in the consideration noti to investigate). This review should address both the technical adequacy of the off-gas building concerns as well as the NRC's handling df the earlier notification in this regard. With respect to the concerns associated with cutting through rebar this matter will be reviewed by Region III with technical assistance from NRR.
I realize your staff is already depleteu as a result of other investigation assistance you are giving us, and your willingness to assist in this'ef' fort is genuinely appreciated.
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[ James G. Keppler Regional Administrator cc:
V. Stello, DEDROGR H. R. Denton, NRR h
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GLs._h January 28, 1982.
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t IEMOIMDUM FOR:
Region III Files - IASalle FROll:
Robert F. Warnick, Director, Enforce: ent and Investigation Staff SUI,JECr:
TELF2EONE CALL FRO:-! DOUG LO:; Celi 1E OF caA1!EL 5 TV ALLEGATIO?iS AT LA SALLE Doug Longenie called on January 26, 1982 at about 2 p.n. regarding infor-cation he had been given chile pursuing q g3cgsger.1
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Longenie indicated he vould
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g contact d.bhdon January 27, 1982, c.nd pave the way for us to contact K Q R y to get specific details of the follouing allegations:
- b..hu.l hm,,y@,hg clained that when they were ' putting in conduit, the 1.
grounding wasn't adequate.
They did not do a good job of grinding off the zinc.
[Mbknowswheretuoradiationnonitorsweresupposedto 2.
be installed in the off gas building but vere not actually installed.
3.
All of the reqasred heating pads in the off gas building were not installed.
4.
The ceiling (in the off gas building, I think) was supposed to
-be 12 inches thick.
When they were drilling 8 inch anchor bolts, they penetrated the ceiling and could see sky.
5.
The fire alarn system in reactor building no. 1 does not meet specifications.
A CAR was issued but it has been written off vithout the work being done.
This problem uns also mentioned by a second individual but the second person vill not talk to us about it.
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Dust seals were only pontly installed.
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I told Longenic vc would get in touch withf[I,kT$Nand follow' up on these allegations.
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MEMORANDUM FDR: Region III Files i
FRDM:
James E. Foster, Inve tigator s
SUBJECI:
TELEPEONE CONTACT RE I.A SALLE ALLEGATIONS (Ref. Warnick Heno of 1/28/82)
I contacted on February 4,1982, at approximately 7:30 p.rs.
~
M indicate that Doug _ ngenie the Channel 5 newsman who providedM name to Warnick, had not advised thatEvould be contacted by Region III. Nevertheles was very willing to discussEconcerns with me,,
at length, requested confidentiality..
i I discussed each point enumerated in the Warnick cono, and developed the following information:
1.
Conduit grounding was not properly done f'or nos: LaSalle construction.
During the late phase of construction,Ebad read the grounding speci-fication, questioned a QC inspector regarding specification requirements, pointed out deficiencies, and grounding was properly performed from that point on.
Prior to that time, crews had not cicaned and copper coated conduit threads nor adequately ground off the zine conduit coatings where grounding straps were attached.gestimated that some 80% of the installed conduit was not properly grounded (per specifiestion, developed from a NDIA requirement).
g 2.
In the off-gas building, there is a location where radiation sensors for Unit 1 and Unit 2 are in close proximity.
This was described as being at the 710 foot elevation, East of An wall, between 14 and 13 i
Ifne in the filter building (part of the off-gas building). 'Ihe sensor that is not inntalled is for Unit.2.
feels that the Unit 2 nensor should be installed now, as t ie locat on will be radioactive and the installation difficult af ter Unit 1 is in operation.
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Region III Files-LaSalle,f[8 1 0 b Dontai entity of co dentia ource 3.
stated that approximately one month ago, five heating pads were removed from the re-heat cylinders located in the off-gas building.
'Sese re-heat cylinders are reached via an entrance on the 710' elevation.
and then by going down two elevations.
The heating pads were removed to allow some work being performed by' fitters, andgbelieves that the pads were not replaced when the area was " closed up".
4.
Roles drilled for expansion anchors in the ceiling of the off-gas building (725 foot elevation) penetrated.the concrete and asphalt roof covering.
There va's water accumulation on the roof, and water came in via the anchor bolts. There are cracks in the concrete between holes because the holes were drilled o close to each other.
This was brought to the attention of from Sargent & Lundy, and some patching was performed.
5.
The fire detector modules have been wired without regard to separation criteria. The crews were wiring the detectors from any hanger indis-criminately.
6.
Dust protection was not installed on conduits and conduit boxes as specified for dust protection. Some have seal gaskets but no o-rings.
Eobserved this durine installation of the security system wiring, and the required seals were installed af terEbrought this to the attention of Bill Bags.
.on that provided by In addition t
.information (
ch e n
Doug Longenie),
indicated that was in charge of an area which had noncon ormances written on some o e equipment, but the work was corrected as indicated in the nonconformance report close-out.
not had the nonconformances approved as completed when the regular inspector was absent.
also indicated that some core-drill sheets were found to have information ich had been whited-out.
(This may be related to the TV-5 story "drilli w
j for dollars" which aired on the 10 o' clock news, February 4,1982.
stated that Longenic had advisedEthat the story would air that evening.
l m f ?o;M James E. Foster
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cph dential c h February 26, 1982 XDiORANDU)! FOR: Region III Files - 1.aSa11e
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THED:
Robert F. Warnick, Director, Enforcemnt and Invcotigation Staff FROM:
C. A. Phillip, Investigator SUB.IECT:
TEIIPHONE CONThCT FROM IfRS. JUDITH COODIE, ATIORNEY, II.LINDIS ATTORNEY CDIERAL'S OFFICE On February 22, 1982 I uns inforned that 1!rs. Judit Coodie, an attorney '
in the Illinois Attorney Gener
's office 2-793-2491), had called concerning allegations made by via TV Channel 5 reporter Doug Longenie.
Information regarding t ese 11cgations is contained in tuo ucmoranda to F.egion III Files, one prepared by R. F. Warnick dated January 28, 1982, and the other prepared by J. E. Foster dated February 10, 1982.
Before speaking with 11rs. Goodie I attempted to obtain additional infor-mation reEarding the allegations n'ud any action we had taken or planned to take. The following information was obtained, prinarily through discussions with Roger Walker, regarding the items listed in the above cenoranda.
1.
Conduit grounding is an industrial safety not a nuclear safety concern und therefore need not be pursued by URC.
2.
Since it is not known whether Unit 2 vill be built, the NRC cannot force the licensee to take action to install a Unit 2 sensor.
3.
The absence of the heating pads would become apparent during pre-op testing. This is a radiation safety concern.
In all likelihood the licensee has a means of tracking this matter to assure that the pads are replaced. We could with little effort confirm this.
4.
This natter in not of concern to URC since this structure is not considered safety reinted, i.e. subject to seismic considerations.
5.
No additional inforrmation obtained.
6.
No additional information obtained.
?OT DISCLOc 7j [,
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Region III Files - LaSalle February 26, 1982 During a telephone conversation with Goodie on February 23, 1982, I discussed the above with her and indicated we were evaluating the allegations to decide what action we would take.
She indicated she would call Foster or me on s'
March 1 or 2 to find out what we planned to do.,
Goodie indicated she had talked with andaskedwhetherllllhadmade an allegation during our contacts uit regarding the drilling into rebar whichllllsaid had occurred during the early stages of construction.
I indicated that 'I was unaware of that allegation but that it would be of interest to us.
Tollowing my conversation with Mrs. Goodie I was advised by Walker that the LaSalle Resident Inspector had determined tha,t the heating pad removal was documented on Pre-operational Deficiency No. 402 and they were required to be replaced prior to fuel load.
Regarding Items 5 and 6, Ron Cardner advised that we.do n'ot inspect security equipment wiring.
I believe it is NRC's posltion that security equipment must function as required by the licynses's security plan connitments and if they fail compensatory mea 7ures must be taken until they are repaired and operational.
above d'. appears that there is no need to further On the basis of th pursue the matters has brought to our attention.
This should, be conveyed to at tie same time we should attempt to determine whetherll[]has information regarding the drilling into rebar that warrants ed further action.
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Cerald A. Phillip Investigator cc:
R. Walker J. Creed R. Cardner J. Foster s
DO NOT DISCLOSg Contains identity ol confidential source
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}iEMORANDU11 TOR:
Region III Files T11RU:
Robert F. Warnick, Director, Enforcement and Investigation Staff FROM:
.7ames E. Foster, Investigator
SUBJECT:
ALLEGATION RE REBAR CUTTING AT LA SALLE, DOCKET NO.50-373 (REF. PHILLIP MEMO OF 2/26/82)
On March 6 and 8,1982, I was contacted by Ms. Judith Coodie, of the Illinois Att Cencral'-
ffice.
She indicated that she had been in contact with ardin all ations concerning work at the LaSalle site.
Eb. Goodic indicated had told her that had often cut reinforcement bara (
ar n drillin cores or holes at LaSalle. She also indicated she felt that concerns regarding equipment in the off-gas building should not e 1smissed, as some equipment in the building was intended to reduce or mitigate radioactive releases during an accident.
I advised Ms. Goodic that ad not made _any,a_11cgations to me regar_dinA_ cutt of reb uring _our_ previous conversations, and I wou1(
try to recontatt I also advised that the of f'- gas b'iiilding was a non-
~
safety, non-siesmic structure, and as such should not contain safety-
~
,related equipment (NRC definition of iaTety-re7M).
She in7 teatea she hEd talked to " nuclear experts" who had advised her differently.
I recontacted at approximately 9:18 p.m. on March 8, 1982.
E stated tha, up until a roximately a year and a half ago (September 1980?)
when hit a rebar during core drilling, a special crew was called. This crew operated a special, water cooled, diamond dril1~
rig, which would cut the rebar. This was referred to as a " wet hole" due to the water cooling, and the utility was billed for the extra work entailed
__ (crew, diamond drill use, laborers to clean up the water).
indicated the problem waPgen. rea_clo_t_ bui_1_d_i_ng. and in the worked in the o
of f-gas building, but eric" to LaSalle, as other crews also folinwed the same practice in other areas.gstated that a notice (" work notice") was finally put out by Quality Control stopping the uncontrolled cutting.
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I inquired if any present worker vould be able to provide Region III vith additional information, and indicated that a may still be onsite, and may be able to provfde additional n ornat on an locations of cut rebar.
It appears that the allegation can be checked by a review of billing records for core drilling.
Those with additinnal charges for the diamond drill crew should provide locati ons where reinforcement bars were cut f
by the drilling.
It should also be relativaly easy t'o locate a work notice inforning the creus that uncontrolled rebar cutting was to stop.
' As the LaSalle plant vill be ready for operating license issuance in the l
near future, I reco=cend that this issue receive priority attention.
j James E. Foster Investigator cc:
L. Spessard R. Ualker R. Cardner C. E. Norelius 1
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Technical Trograms
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Robert F. Warnick, Director, Enforc'ement and-Investigation s
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ALU. CATION RE REBAR CUTTING AT LA SALLE-DOCKET No. 50-373;
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50L374 l
,t Tne four attached ecnos } document concerns expressed by an alleger and Ms. Juditi,Goodic of the illinois Attorney General's office regarding core drilling through rebar.
Eecause of the high priority of LaSalle and the unavailability of investi-Sators, this matter is being trar.sferred to your Division as we discussed on March 22, 1982.
g IIS would appreciate receiving a copy of the documentation of your findings and c1csecut.
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l Robert F. Warnick, Director Enforcement and Investigation Staff Attachcent's: As stated 4
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O ASSIS.TANCE REQUEST FORM _
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To:
CharJew E. Norelius (o
(Assigned)
Requested Coepiction Trum:
Robert F. Warnick (Requestor)
Docket No.:50-373 _ Category:
FacMif.y:_ I.aSaue 50--3/4
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Ecquest Descriptjon:j A copy of the documentation of vntii-ri-a4 ppg, :
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NUCLEAR REGULATORY COMMISSION 7
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GLEN ELLYN, ILLINOIS 60W WR 31.1982 D0 h T DIS SE Conta dentity of C "U MEMORANDUM FOR: Region III Files FROM:
James E. Foster, Investigator THROUGH:
R. W. Warnick, Director, Enforcement and Investigation Staff
SUBJECT:
CONTACT WITH JUDITH GOODIE I contacted Ms. Judith Goodie, of the Illinois Attorney General's Office, at approximately 9:10 a.m., on March 26, 1982.
I advised M Goodie that Region III had not been aware of allegations by regarding LaSalle, and inquired why the Illinois Attorney enera s petition did not mention allegations from
- s. Goodie s_ tate _d_that she had " assumed" that Regios III had gotten ame_from NBC (as she had) and had contacted g she, indicated had declined to provide her office with an attidavit for fear o nadie being known, and so was not included in the submitted petition.
I indicated that the ' formation provided by was much more detailed than that provided by and would have assisted Region III in its review.
Ms. Goodie stated that she had not meant to withhold any information, and that "it should have been obvious that we were working on something".
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./ James E. Foster l
Investigator cc:
R. L. Spessard R. Gardner W. Walker C. E. Norelius l
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March 29, 1982 0FF_-CAS BUILClNG R03T PEPORT FURPOSE The purpose of the report is to state inforr.ation regarding the second allegction (Page 6. Request to Institute a Show Cause Proceeding and for other Relief - Tyrone Fchner, Attorney General of the State of Illinois) on the Off-Gas Butiding roof.
The evidence shows that the allegation is false.
BACKGROUND The concrete enclosure chove grade as part of the Off-G3s Building is a non-ssfety related structure which houses Off-Gas Building HVAC Air Handling Units HVAC Mater Cooled Condensing Units, HVAC Exhaust Filter Units, HYAC Control Panels and associated motor control centers and switchgear.
The Unfle specification concrete cccpressive strength is 40]O psi at 90 days.
detailed qu:lity assurance requirerents were not required due to the building being non-safety related, they were applicd as part of the overall Con onwealth Edison / Walsh Construction Company quality effort.
FINDiHG5 The Off-Gas Building enclosure concretc (walls and roof) was poured on Noverber 7,1975. Ualsh Construction Corpany(WCC) Q.C. Forn QCP-9A(Pour Checkout Card) was signed by the appropriate construction and 0.C. personnel and Additionally, countersigned by a Corrnorweealth Edison Conpany ricid Engineer.
L2CC Q.C. Forcs QCP-6A(Reinforcing Steel Placenent Audit) and QCP-93(Concrete Placement Control Audit forra) were utilized and signed by WCC Q.C. personnel.
Concrete testing during the pour by A & H Engineering Cor;raration showed the concrete ras' within specification requirer.cnts for slur.p, air content and The concrete ret cor.pressive strength requirecents, the placing temperature.
loeest cylinder break was 4670 psi at 90 days.
On September 25, 1979, co r onwealth Edison Co pany ?>ality Assurance pointed out sor,e surface cracking in the bottom of the Off-Gss Building roof.
An inspection The area had a high density of concrete expansion anchors.
perforced by HCC Q. A. Supervisor. WCC General Superintendant and CECO Structural Engineer found the cracking to be surface in nature and no further action required.
A tecporary construction power center transforner and switchgear were set The unit weighed approxicately 6700 pounds.
The unit was on the roof in 1976 set. over a concrete beam in the longitudval direction and one end rested on the cast concrete wall.
A check was r.ade to insure the roof would take the unit loading prior to installation.
The unit was re-oved in late 1981 as it was no longer required.
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J On The slab thickness has been checked on twa dif ferent occasio March 10,1982 roofing rsterial.
Rooftng mterial is approxirately 1-3/4' - 2" thick.
Fifteen (15)
Additioral slab thickness checks were cade on' P.3rch 29,1982.
points checked shawed the slab plus roofing materail varied from l' 3/4".
the sich thickness varied from 11-1/4" to l' 1-1/?".
A visual survey of the roof underside was cade by WCC Q. A. and CECO on The area under Tne survey showed no abnormsl concrete cracking.
P.srch 27,1982.the fomer electrical equipment showed rc abnorr.a1 concrete cracking.
StI&ARY_
cifications.
.The Off-G5s Building roof concrete is 12 inches thick per sp3 There is no cbnorcal concrete cracking due to concrete expansion anchces and/
The roof will serve its' intended function.
the forcGr clectriccl equipment.
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