ML20052C974
| ML20052C974 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 03/30/1982 |
| From: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Deyoung R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| Shared Package | |
| ML20052C945 | List: |
| References | |
| NUDOCS 8205060155 | |
| Download: ML20052C974 (18) | |
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, t. j UNITED STAT ES
[,q NLICI EAR REGULATORY ':OMMISSION I p REGION lil
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- {$74 f?
os es > tvu. LLtNors sour March
- 0, 1982
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1 NMMORANDilM FOR:
R. C. DcYoung, Director, OLIice of Inapperion and Enforcement TROM:
James G. Keppler Regional Adminiarratur, Region III Stim.l FUT 1.A SALLE COUNTY NUCLEAR STATION - TCTTTION FROH ILLINOIS ATTORNEY GENERAL As you know, on h rch 24, 1982,'the Illinnfr. Attorney Cenerni petitioned the NRC to suspend licensing proceedings at La Salle. pending investInation et recent slic&stions and to institute a Show Cauwe Hearing with illinois I
ss a party to the Hearing. The allegations deal with the overall adequacy of safety telated structures as a result of widespread rebar cutting and npor.ific nrrunrurn1 deficiencies in the roof of the off-gav building.
i A cont'crence calI var, hcid nn March 29 involving Meesrv. DenLun. Cave,
,Stelio. DeYeung and Keppler rn discust, the handling of thewe Investigations.
L'e agreed that. becau'se the petitiun expresses cuncura that thw off-gas building defielenelva had been verbslly cet=unicated 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 th1A
/ allegation by IE (since 18 was not involved in the consideration not to i
investigste). This review should address both thu' technical.adct{uacy of the t
nrf-gan hutiding conce.rna as vall as the NRC'r. handling of the earlier notitication in this regard. With respect to the concerns associsted with cutting through rebar this cutter will be reviewed by Region III with i
techntcal ssr.ir.tance from NRR.
i rea11tc your staff is aliendy depleted as a resuit of other investigation
)
j a usist.46ce you are giving us, altd your willitignesu to assiuL in this effort a
is genuinely appreciated.
--& s b
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[ James C. Kcppler Regional AdminiwLrutur cc:
V. Stallo, DEDRUGR
- 11. H. Denton, NRR i
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P205060155 820417 PDR ADOCK 05000373 l
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LSCS-FSAR 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS j
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 s t ructures, systems, and components, including their foundations and supports, designed to remain functional in the of a SSE are designated as Seismic Ca'tegory I, as indicated event 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 earthquake (OBE).
Since the two earthquakes vary in intensi t y, 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 factors, whichever is appli' cable, and yield margins of safety appropriate for each earthquake.
The margin of safety provided for such structures, components, equipment, and systems ensures that their design functions are not jeopardized.
For further details of seismic design criteria refer to:
mechanic'el, in Subsection 3.7. 3; a.
b.
electrical, in Section 3.10; c.
structural, in Subsection 3.7.2; and I
d.
instrumentation and controls, in Section 3.10.
l 3.2-1 1
b LSCS-FSAR AMENDMENT 54 JANUARY 1981 3.2.2 System Ouality Group classifications System quality group classifications have been determi those applicable fluid systems which are relied upon ned for a.
malfunctions originating within the reactor coola pressure boundary, 3
b.
safe shutdown condition, andpermit shutdown of the reacto t
c.
contain radioactive material in large quantity or concentration, A tabulation of quality' group classification for each i
system
" Quality Group Classification."and component is shown in Table 3.2-1 l
(
structure, diagrams which depict the relative locations of theseFigures 3.2-1 and 3.2-r systems and components along with their quality group classifi 1
structures, cation.l The implementation of the code requirements outlined i 3.2-1, 3.2-2, 3.2-3, and 3.2-4 n Tables discussed in Sections 3.7 and 3. 9.for fluid system components is A boiling water reactor has a number of structures components in the power conversica or other portions of the systems, and f acility which have no direct safety function, but which'may be connected to, or influenced by, the equipment within the nuclear safety-related classifications defined previously.
structures, systems, and components are designated as "other "
Such The design requirements for equipment intended service of the equipment and expected pla i
e environmental conditions under which it operates.
n possible, design requirements are based on applicable industry Where codes and standards.
When these are not available, the designer relies on accepted industry or engineering practi'ce.
structures, systems and components whose safety functions require conformance,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 Requirements.
under the 3.2-2
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._m TABLE 3.2-1 (cont'd)
OUAI.ITY (4a)
QUALITY (4b)
SEISMICIS) CROUP ASSURANCE ELECTRICAL (4c) PURCHASE PRINCIPAL COMPONENT (1)
LOCATION ( 3) CATECORY CI.ASSIFICATION REQUIREMENT CLASSIFICATION DATE(2)
COMMENTS l
XXII.
Local Panels 1.
Electrical panels with a safety function A,RB I
HA I
IE 4-74~
(15) 2.
Cable, with a safety function A,R6 I
NA I
1E 10-75 3.
Remote shutdown panel A
I NA I
1E 10-74 fXXIII. Off-cas System (2) 1.
Atmospheric glycol tanks F
II D
II HA 10-71 2.
Heat exchangers F,T II C
II HA 10-74 C
3.
Piping and valves (down-n stream of steam jet Y
air ejectors)
T, F,0 II C
II NON 1E 9-74
]
w
'u 4
Piping and valves (up to and including air g
a 5
ejector)
T II D.
II NON 1E 9-74 5.
Valves T,F II C
II NON 1E l
6.
II D
II HA 2-12 7.
Charcoal vessels F
II C
II NA 10-71 8.
Recombiners T
II C
II NA 10-71 9.
Filters F
II C
II NA 10-71 10.
Afterfl1ter F
II C
II HA 10-71 11.
Reheater F
II II NON 1E 1-72 XXIV.
Service water System 1.
Piping RB,0,L A.T II D
II NA 9-74 2.
Strainers L
II D
II NA 7 13 y{
t 3.
Pumps L
II D
II HA 7-73 3g 4
Pump motors L
II II NON 1E 7-73 go 5.
Valves T,0,L,A,RB II D
II HA 6-73 6.
Electrical 6 instru-cr z sent Modules RB,L,A II II NON 1E 7
Cable RD,0,L,A,T II II NON 1E 10-75 G"
U F
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TABLE 3.2-1 (Cont'd)
QUALITY (4a)
OUALITY(4b)
SEISHIC(5) CROUP ASSURANCE ELECTRICAL (4c) PURCilASE PRINCIPAL COMPONENTQ)
LOCATION ( 3) CATEGORY Cf.ASSIFICATION REQUIREMENT CLASSIFICATION DATE(2)
COMMENTS XLII. Civil Structures 1.
Reactor building RB I
HA I
NA (22) 2.
Lake screen house L
NA t3A II HA (22,27) l 3.
Radwaste building RW NA NA II HA (22) 4.
Auxiliary building A
I HA I
HA
)
(22) 5.
Turbine building T
tlA NA II HA (22)
'6.
Off gas filter building F II t1A II NA (22)
'i F
7.
Steam tunnel A
I NA I
NA (22) 8.
River screen house O
II NA II HA (22) 9.
Diesel generator building RD 1
NA I
HA (22) 10.
Auxiliary Spillway o
NA NA II liA (22) 11.
Cooling Lake Embankment 0 I
t1A II HA (22) 12.
Submerged CSCS Pond O
I NA I
13A (22)
(Ultimate Heat Sink) g o
13.
Diological Shield PC I
f1A I
NA (22) n g
a 14.
Primary Containment PC I
f3A I
HA la 5
- a. Vacuum breaker 4
piping PC/RD I
D I
IIA 9-74 N
- b. Vacuum breaker valves PC/RD I
D I
lE
- c. Maintenance butterfly valves PC/RD I
B I
HA
- d. Suppression vent downcomers PC I
13A I
11A XLIII. MSIV Leakage Control Syste'm 1.
Piping, within RCPD l
isolation valves RD I
A I
NA 9-74 2.
Piping, other upstream e
system lines RD I
D I
NA 9-74 3.
Plping, downstream sys-g tem from steamline con-p nection to first valve RB I
D+
II HA (7,8) g 4.
Piping, other downstream e
system lines RD I
D I
HA 9-74 un 5.
Valves, within RCPD RD I
A I
IE 12-73
'd 6.
Valves, other RD I
D I
IE 12-73 7.
Ileater RD I
NA I
lE 5-76 8.
Blowers RD I
B I
IE 11-75 9.
Electrical modules with a safety function RD I
FIA I
lE 10.
Cables, with a safety
- e-*lan nn I
NA I
lE 10-75
LSCS-FSAR '
AMENDMENT 24 oEPTEMBER 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 i
sensors (including electromechanical), power supplies, and signal prccessors; and mechanical modules include filters, strainers, and flow (element) assemblies /
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 fiLst date corresponds to Unit 1 and the second date corresponds'to Unit 2.
Where two dates are giv'en 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 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.
i II - The equipment is not required to meet the quality assurance requirements of 10 CFR 50, l
Appendix B.
I 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 requirementn.
NA - not applicable because the equipment is not electrical.
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3.2-17 J
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LSCS-FSAR*
TABLE 3.2-1 (Cont ' d )
(5)
I - The equipment is designed in accordance with the seismic requirements for the SSE.
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 me: hods.
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 equivalant to l
those specified as supplementary types of examinations, in ANSI B31.1 Code.
I l
All inspection records are maintained for the c.
life of the plant.
These records include data pertaining to qualification of inspection personnel, examination procedures, and exami-nation results.
3.2-10 m
Lat.b-z utut swuwem ao JULY 1981 TABLE 3.2-1 (Cont ' d)
The unprocessed radwaste piping will meet Group D T
(20) 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 Sockets and girth and longitudinal butt-welds.
fillet weldr in sizes over 4-inch nominal will be given random magnetic particle and liquid penetrant examination on 20% of the joints.
I 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 not apply to piping and components supplied by the diesel engine manufacturer as an integral part of the diesel-generator unit.
In this case, 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.
Expendables and Consumables are purchased per original (24) specification and stored under controlled conditions.
Includes raceway installations containing Class lE cables (25) and other raceway installations required to meet Selsmic Category I requirements (those whose f ailure during a l
seismic event may result in damage to any Class lE or other safety-related system or component.
I I
(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-moni-ture monitoring, and containment high-range radiation toring.
Subsystems not required for post-LOCA monitoring
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U il iki OESIGN CRITERIA FOR STRUCTURES, SYSTEMS AND COMPONENTS
- h
' s 3
Conformance with Nuclear Regulatory Commission General Desian Criteria
- 3. I _
Q In Section 3.0 of the Final Safety Analysis Report, the applicant presented an f,
i.!j i evaluation of the design bases for La Salle against the Commission's General We evaluatea the Design Criteria listed in Appendix A of 10 CFR Part 50.
final design and the design criteria and conclude, subject to the applicant's k'
adeption of the additional requirements made by us as discussed in this report, j }p};'
that La Salle has been designed, can.be constructed an'd can be operated to
{Q meet the requirements of the General Design Criteria.
'qt y:g.u 3.1.1 Conformance with Industry Codes and Standards 4
our review of stri.ctures, systems and components relies extensively on the application of industry codes and standards that have been used as accepted.
yd industry practice. These codes and standards, as cited in this report and the, p%
30 attached bibliography, have been previously reviewed and found acceptable by us; and have been incorporated into our Standard Review' Plan.
,hre 3.2 Classification of Structures, Components. and Systems hp 3.2.1 Seismic Classification 7
6 Criterion 2 of the General Design Criteria requires that nuclear power plant structures, systens, and components important to safety be designed to with-g stand the effects of earthquakes without loss of capability to perform their p$
These plant features are those necessary to assure (1) the safety function.
K integrity of the reactor coolant pressure boundary, (2) the capability to. shut capability to prevent or mitigate the consequences of accidents which could T[
down the reactor and maintain it in a safe shutdown condition, or (3) the result in potential offsite exposures comparable to 10 CFR Part 100 guideline 4
j j
exposures.
p Structures, systems, and components important to safety that are required to j
v be designed to withstand the effects of a safe shutdown earthquake and remain functional have in general been properly classified as seismic Category I j
j",
items, in conformance with Regulatory Guide 1.29, " Seismic Desiign Classifica-The applicant's nonseismic r
tion," Revision 2 except for the following system.
Category I classification of the cooling loop of the spent fuel po61 cooling and cleanup system is not in conformance with Regulato,ry Guide 1.29.
As an alternate to a seismic Category 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 spent fuel Pool cooling and cleanup system, see Section 9.1.3 of this report.
f h
l 4
3-1
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3
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i All other structures, systems, 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, i
fill lines, drain lines, and test lines on the downstream side of isolation i
i I
valves which 'are not required to perform a safety function.
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 be. sis 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 safety with the Commission's regulations as set forth in Criterion 2 of the General Design Criteria and to Regulatory Guide 1.29, our technical positions, and industry codes and standards.
~
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 earthquake and remain functional have been properly classified as seismic 4
Category I items in conformance with the Commission's regulations, the 4
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 System Ouality Grouc 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 l
the safety function to be performed.
Fluid system pressure-retaining components 4
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 l
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 or mitigate the consequences of accidents and malfunctions originating within 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 material. These fluid systsms have been classified in an acceptable manner in 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 Classificati,on and Standards,"
4 1
Revision 3.
The applicant has applied Quality Groups A, B, C, and 0 in conformance to Regulatory Guide 1.26, to the fluid system pressure-retaining components important to safety. Those components that are classified Quality Groun A,' 8, C, or 0 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 of'the applicant's designs, design criteria, and design bases for pressure-retaining compon*ents such as pressure vessels, heat,exchangers~, storage tanks, pumps, piping and 1
3-2 E
~a
1 1
s 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, drywell purge system, gland seal condenser offgas system, and building ventilation systems.
The offgas treatment system, shared by Unit Nos. I 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 Fahrenheit 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 anticipatsd operational occurrences.
The system design includes. 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 N 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, therefore, acceptable.
I The seismic and quality group classification of the offgas treatment system is based 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 "3
which 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
NU"M R.IGT ATORY CC.50'_ISSICN
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I COMMON'E.J_LTH EDISON COMPANY DOCKET NOS.
50-373 and 50-374 LaSalle County Nuclear Generating Station, Unit 1 and Unit 2 CAC: March 31, 1982 p ga,.,g : 1 _ 77'
- AT:
Bethesda, Maryland
_Ul)OLSO.Y REPORTLTG
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40 0 Viry d a Ave., S.W.
Was! '
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20024 d
U Telechc=e : (202) 554-2345
S 9
i 7
ER. DELGEORGE:
What I v'ould like to do is 8 review the allegations presented in the petition as we 9 understand them, stating the facts and the information to ve have which we think vill resolve the concerns that i
11 have been raised in your mind.
12 I would like to start with the questions 13 raised relative to the off-gas building because ve feel
,)
14 that to be a less-conplicated issue that can be more 15 easily dispositioned.
16 First, there is an allegation that the roof 17 thickness is eig'ht inches as opposed to the 12 inch is design thickness.
I would like to say at the outset 19 tha t although this building is a non-saf ety related.
20 building containing no safety-related equipment and not 21 requiring the implementation of our quality assura'nce 22 program, we did in fact apply our quality assurance l
which has l
23 progran to the construction of this building, l
24 given us greater ronfidence in the accuracy of the 25 information that we vill be providing to you.
we.
l ALDERSoN REPCRTING COMPANY.INC.
s 15 13 M2. DELGEORGE.
I am ready.
The last 14 allegation suqqe'sted that the co'ncrete associated with 15 this slab had been cracked substantially.
Commonwealth 16 EdisCn discovered surf ace cracking of the subject slab 17 through its own site quality sssurance depart =ent in 18 Sep tember 19 79.
As a result of the deficiency 19 identified, an inquiry was made at that time which t
20 included an engineering evaluation and which also 21 included the tracing of the crack depth by chipping at 22 the concrete in the vicinity of the cracks.
23 As a result of that review, it was established 24 that the crack depth did not exceed one quarter inch; l
25 that. the cracking was, in fact, surface cracking, a n d'. a s
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ALDEASON REPCRTING COMPANY,INC,
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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 is l
1 4 our opinion, based on that evaluation, that the cracks l
i.
5 observed are normal shrinkage cracks associated with 6 this type of slab.
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ALOEASON REPORTING CCMPANY. INC.
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9 13 MR. DENTON:
Let se ask th e project =anager
?,
14 vha t categoriration ve gave that roof.
s 15 MR. BOCHNIA:
It is a n'on-safety grade 16 building.
I have the reviewer here.
We did not 17 consider this as a saf ety grade building.
18 MR. DENTON:
What is unter the roof?
19 MR. BOURNIA:
What is this?
20 MR. DE5 TON:
What is under it?
r 21 MR. DEL"EORGE:
That is described in our 22 report.
The concrete enclosure above grade as a part of 23 the off gas roof is a non-safety related structure which i
I 24 houses off-gas building, heating / ventilating /and air 25 conditioning,. air handling units, F.V AC, vater cooled I
i l
RCER$CN REPoRMNG CCMPMY. INC.
18 e
1 condensing units, HVAC exhaust filter units, HVAC 2 control panels and associated motor control centers and 3 switchgear.
4 MR. DENTON:
Does that mean there is no 5 Category 1 safety-related equipment in that building ?
6 MR. DELGEORGE:
Yes, sir.
7 MR. DENTON:
Any qtestions?
'4e. can come back 8 to this, but I thought we would give the company a 9 chance.
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We received allegations on this 19 som e months ago and evaluated it in-office.
I do not 20 have those with me.
I am not sure that I know they say 21 e xa ctly wha t she said, and I have not read them 22 caref ully.
But we vere avare of the allegation.
It was 23 evaluated within our office and I think, in recognition 24 of our manpower considerations, we chose not to delve I
i 25 deeply into this at the field level because of its i
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