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.j-fl U. S. ATOMIC ENERGY COMMISSION J
REGION I DIVISION OF COMPLIANCE Report of Inspection
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CO Report No. 219/68-10 N{
Licensee:
JERSEY CENTRAL POWER & LIGHT COMPANY h
(OYSTER CREEK -1) 3 Construction Permit No. CPPR-15
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Category B Dates of Inspection:
November 21 and 22, 1968 December 2, 3 and 10 - 12, 1968 Dates of Previous Inspection:
October 15 - 18, 1968 Inspected by: 37 de-lt C bol R. T. Car'Ison, Senior Reactor Inspector Date Reviewed by : hIM
/!/4 /67 Jl/ G. Kepp1'er, Senior Reactor Inspection Date Specialist Proprietary Information:
None
SUMMARY
Most outstanding previously identified plant quality issues have been adequately resolved.
Present indications are that those items remaining will be resolved in a timely manner.
A review of the scope and results of GE's recent audit in the electrical separations area, and the results of CO's_ followup i
sampling audit, showed that GE has been adequately responsive to CO's concerns in this area.
Correction of identified deficiencies l'
has been initiated and is scheduled to be completed prior to fuel loading.
As a result of the above experience, GE is reportedly initiating a separations criteria review at all other domestic turnkey projects.
Additional assurances of plant quality by GE were evidenced-in CO's review and sampling inspection of the results of GE's construction records inventory and review program, and the punch list system in effect at this facility.
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GE has reported that the results of their investigation into cable tray loading indicate no heating problems due to cable tray over-f loading at Oyster Creek 1.
The detailed results of this investiga-p tion, unavailable at the time of these visits, together with GE 's g
plans for in-plant verification, will be reviewed by CO at an early
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date.
j Sample inspections of conduit loading showed no deviations from applicable standards.
Examination of the control and,instrumenta-l-
tion features of the air ejector off-gas system revealed no items of nonconformance with the provisions of the Application.
A detailed followup review of vendor fabrication and testing records for the double isolation valves in the isolation condenser system indicated that the special quality control provisions outlined in the Application had been complied with.
Questionable indications on the radiograph of one related weld have reportedly been identified as a film artifact as the result of further radiography.
Cracks were detected on an installed core spray system valve during a cursory examination by a CO metallurgist.
GE's followup investiga-tion of the condition has resulted in rejection of the valve.
The cracks have been attributed to faulty final heat treatment by the vendor.
Lack of specificity in purchase specifications as to the timing of valve casting radiography and actual performance of the radiography prior to final heat treatment appear to account for failure of the vendor to detect the condition.
GE has determined that all other valves supplied this project by the subject vendor, Anchor Equipment'Co., are similarly open to question.
GE has under-taken an investigation of certain of these valves.
Indications of p
varying significance have been detected on four additional valves.
The adequacy of GE's investigative program and the broad applicability aspects of the problem are the subjects of current Regulatory concern.
Meetings with JC and GE are planne6 by DRL in this regard.
i Some question exists regarding the adequacy of the ventilation in the battery room to cope with hydrogen generation under loss of air flow conditions.
GE has undertaken a re-evaluation of the need for the containment penetration restraints.
Initial fuel loading is now scheduled to begin on March 4, 1969.
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Scope of visits i
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A meeting was held with management and technical representatives'
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of Jersey Central Power and Light Company (JC) and General Electric.
Company (GE) at the site of the Oyster Creek Unit 1 facility on
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j November 21, 1968.- A list of attendees is given below.
The El meeting represented a continuance of the construction phase inspec-l-
tion program.
Subjects discussed included the following:
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4 resolution of outstanding plant quality issues; GE's punch-list
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system;_the scope and results of.GE's most recent audit of plant.
. i electrical systems; the status and results of GE's investigation j
of cable tray loading; and the resolution of outstanding-issues relating to the preoperational testing program.
I A follow-up inspection of several of the areas discussed above
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, was performed by the writer on November 22, 1968.
Messrs. D. F.
Sullivan and O. D. Parr, DRL, were also present at the site at this time.
The site was visited _ by the writer on December. 2 and 3,1968, for the purpose of follow-up inspection in certain mechanical areas.
Technical assistance was provided during this visit by j
Mr. W. J. Collins, Metallurgist, CO:HQ.
7 The site was also visited on December 10 - 12, 1968, by Mr. R.
T. Dodds, Reactor Inspector, CO:V, and the writer.
Mr. F. U.
t Bower, Reactor Inspector (Construction), CO:II, was also present i
during the latter two days.
Mr. Dodds was present for orientation preparatory to his inspection involvement to commence with initial fuel loading.
Mr. Bower was present in conjunction with the writer for the purpose of reviewing further the results of. GE's most recent audit of plant electrical systems, and to perform a sampling inspection in this area as a measure of the effectiveness.
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of the GE audit.
The systems inspected by CO were the core spray system and the instrumentation and controls associated With the air ejector off-gas system.
The results of GE's investigation.in the area of cable tray loading, and conduit loading, were also reviewed at this time.
Portions of the report relating to the December 2 - 3 and 10 - 12 inspections were contributed to by Messrs. Collins and i-I Bower, respectively.
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j The principal persons contacted during these visits were as 5-follows:
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Attendees at November 21, 1968 Meeting AEC-CO t
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r J. P. O'Reilly, Chief, Reactor Inspection & Enforcement Br.
l J. G.'Keppler, Senior Reactor Inspection Specialist I
R. T. Carlson, Senior Reactor Inspector AEC-DRL D. F. Sullivan, Instrumentation and Power Technology'Br.
j O. D. Parr,-Instrumentation and Power Technology Br.
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J D. Ross, Technical Supervisor Trainee 9.E, J. Barnard, Manager, Licensing Activities - Domestic Turnkey Projects (San Jose)
D. K. Willett, Project Manager,(San Jose)
L. M. Loeb, Manager, Materials and Quality Services -
t Domestic. Turnkey Projects (San Jose)
M. R. Lane, Systems Electrical Engineer (San Jose) 1 N.'M. Strand, Construction Site Manager i
K. W. Hess, Site Operations Manager l
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Other Visits
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T. J. McCluskey, Plant Superintendent 4,j '
D. E. Hetrick, Operations Supervisor y
D. Ross, Technical Super, visor Trainee E. J. Riggle, Instrument Foreman 9.E.
L. M. Loeb, Manager, Materials.and Quality Services -
Domestic Turnkey Projects (San Jose)
A. G. Mellor, Electrical Engineer, Balance-of-Plant Engineering (San Jose)
W. Scott, Project Engineer (San Jose)
N. M. Strand, Construction Site Manager K. W. Hess, Site Operations Manaijer R. Green, Instrumentation Construction Engineer W. Fletcher, Electrical Construction Engineer Burns & Roe d
G. A. Lari, Project Engineer (Oradell) t II. Results of Visits A.
Plant Quality - Outstanding Items The status of resolution of all outstanding plant quality deficiencies as of the last visit covered by this report is summarized below.
For purposes of comparison, the pertinent references to previous inspection reports are given in parenthesis.
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Recirculation ~ System Vibration Testino (219/68-8. Paragraph II.A.l.)
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[The inspector determined'that.GE has issued'a memorandum:
to the responsible field engineersfoutlining the need-i i
to establish'zero loading conditions'at the system 1 5"
sway.. braces during the preoperationalj (ambient) tests..
This matter is considered resolved.
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Pipe Hangers : (219/68-8, Paragraph II. A.2. )
JC has still'to generate the maintenance procedureffor.
- a. recirculation system pump motor replacement.:.There-fore, verification of the incorporation 1 1n the: procedure of precautionsfrelat'ing to the blocking of the; system hangers remains outstanding.
It is not required.that.
this action be completed prior to issuance of~the
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provisional' operating license; however; the inspector will confirm.that it has been completed within a-
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reasonable time thereafter.
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3.
Main Steam System Piping Material Certification-
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(219/68-8, Paragraph II.A.3.)
l Reference to ASTM A106, which applies to tlus 6 inch piping in question,- shows that the. supplemental ' tests'~
Si through S6 apply only to 8 inch piping and above.-
i The question regarding the applicability of the' subject.
tests to the 6~1nch piping in. question is considered to be resolved.
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With respect to the previously noted failure of -the Burns and' Roe vendor-inspector to witness the performance s
of the~S2 through S6 tests for the.8 inch and above piping for this system (CO Report No. 219/68-5, Para-j graph' II.B.l.) i the inspector observed formal disposition of this matter in the form of~a waiver of.this require-ment by Mr. Huggins, Principal Project Engineer, dated October 29, 1968.
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(219/68-8, Paragraph II.A.4.)-
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The GE engineering disposition of.this question, dated-F october 29, 1968, and signed by Mr. Huggins,.was 4,. 39 reviewed by, the inspector and noted to contain the 4? I following statement:
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. The vendor certification for the subject bellows
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(main steam and feedwater penetrations) indicate
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that they are certified for ASME Section III Class C.
Normally Class C would indicate no requirement for impact testing. which is in conflict with the letter 4
of requirements for containment vessel penetrations.
L However, the body of these bellows are fabricated-from austenitic stainless steel which is exempt
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from impact: testing.under Class B requirements, t
Furthermore, we have obtained approval from APED Engineering for use of 1967' Winter Addenda to l
Section III-B, ' paragraphs N-1210 and N-1211, which exempt materials of 5/8" or less thickness from impact testing.
Since the ' safe-ends' for these bellows are of 1/2", it is concluded that the bellows meet the intent of the penetration impact testing criteria."
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Based on the above, the original question raised by CO regarding the certification of the bellows expansion joints is considered to be resolved.
i Further-discussion on. th'e subject of containment
'j penetrations revealed that as a result of these original questions by CO plus other questions that i
developed during GE's own records. inventory and review-Dj (discussed further in~ Paragraph II.B.l. of this report),
GE conducted an extensive review of penetrations in t
general.
As a result, GE has satisfied themselves~that all the penetrations at this facility meet the i
L applicable code requirements with the exception of a question relating to the stress relieving of.'some carbon steel welds on the penetrations for the isolation condenser system.
Mr. Loeb stated-that the subject a
welds will be stress relieved if found necessary.
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5.
Protection of Critical Electrical Equipment 7
(219/68-8, Paracraph II.A.6. )
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G Discussions ou this subject during the November.21' meeting indicated to CO that with the possible exception gg of the solenoid valves for the main steam isolation
, ', 'C valves, GE had properly addressed themselves to this area of concern and were satisfied that all vital
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j' electrical equipment was adequately protected against1 l
the accident environment.
With respect to tlue solenoid valves for the' main steam isolation ~ valves, GE initially maintained that no-I special protection was required in that the isolation l
valves.close prior to the accident and that there-was-l no plausible event that would cause them to open.
l Mr. Loeb stated that on this basis, testing ~ of 'the subject solenoid valves;against the accident environment' j
was considered not to be required.
However,.during the ACRS meeting on December 5, 1968*, in response to a question by the committee, GE representatives present J
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stated that t2un-subject valves had been tested against
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the accident environment'and_found to be adequate.
d Staten. cats to this effect are 'also contained in j
Amendment No. 48, Paragraph.4.0.
During subsequent H
l-discussions Mr.'Loeb stated.that his rational against testing was still valid, and that the test was performed in order to put to rest concerns in this area by the ACRS and Regulatory.
'i Subsequent to the visits' covered by-this report, Mr. Loeb provided the inspector with a copy of a. report on the results of the above tests.
The test environ-ment was steam at 3000 F and 62 psig.
The results j
show that the solenoid valve under test survived this
~3 environment for.the required maximum 10 minutes, but that it failed after about 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> of testing.' The i
significance of this latter point, which is not
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reflected in Amendment 48, is currently under review
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by CO and will be' reported on in a future report as is appropriate.
Otherwise, the questions regarding j
the protection of critical electrical equipment in the k
dry well are considered to be resolved.
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UI 6.
Steam Line Cold Spring (219/68-8, Paragr,aph II.A.7.)
t GE provided documentation for review by CO that indi-cated that this matter had received adequate investiga-tion by GE and that no problem existed.
This item is considered to be resolved.
7.
Questionable Welds in Main Steam Lines
_(219/68-8, Paragraph II. A.8. )
Documentation made available by GE and reviewed by the inspector showed that the two main steam line penetra-tion welds in question by GE, guard pipe welds, were eventually declared acceptable by GE Engineering on the basis that they were basically structural in nature.
On the basis of this observation and earlier statements by GE regarding other main steam line penetration welds questioned by CO, to tne effect that GE's original review of radiographs showed the questioned welds to be acceptable, this matter is considered to be resolved.
The previously planned visit to the vendor shop by CO is therefore no longer necessary and is cancelled.
f The welding, inspection and stress relief heat treatment records of weld repairs on the main steam piping and reinforcement welds on the main steam flow restrictors j
were reviewed in depth by Mr. Collins to determine that I
GE's special instructions and applicable code require-ments were met.
The records reviewed consisted of the welding data, weld procedure and personnel qualifica-tions, liquid penetrant tests and radiography, and the heat treatment temperature charts.
No deficiencies were identified which indicated a deviation from requirements or that an unfavorable condition exists in the welds.
For reference purposes, these welds are identified on isometric drawings as follows:
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Weld repair, main steam piping joint No. S-ll24.
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Reinforcement welds, weld Nos.. N-1351 and S-1351 L
(north and south piping system arrangements.)
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8.
Auto-Depressurization System - Adequacy of Exhaust sp Piping (219/68-8, Paragraph _II.A.9.)
T The subject piping consists of 8", 12" and 14" sizes.
A review of site records and discussions with Mr. Loeb showed that the 8" and 12"' pipe met the design requirements, whereas the 14" pipe was good for 580 psig vs a design requirement of 600 psig.
GE considers the latter pipe to be adequate for this information.
The inspector concurs.
This matter is considered to be resolved.
9.
Feedwater System Isolation Valves (219/68-8, Paragraph II.A.10.)
d Documentation made available by GE for review by CO indicated that the subject valves had been investigated and determined to be acceptable for this application.
This matter is considered to be resolved.
10.
Containment Spray System - Heat Exchanger Improperly Labeled (219/68-6, Paragraph II. B. l. )
l The correct nameplate was determined to have been installed.
Back-up documentation from the vendor supporting this action was noted to be present in the i
site files.
This matter is closed.
1 11.
Isolation Condenser System - Shells of Condensers Not l
Code Stamped (219/68-6, Paragraph II.B.3.)
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contrary to previous indications, the subject vessels were determined to be code stamped - on the head flanges.
It was also noted that the appropriate ASME forms U-l have since been added to the site record files.
This issue is considered to be resolved.
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12.
Reactor Protection System - Failure to Meet Separations f
Criteria (219/68-7, Paragraph II.D. & 219/68-8, Para-K graph III.B. )
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The inspector visually confirmed that the physical re-routing of the reactor pressure impulse lines to "I
conform with the description in the Application had been implemented.
The inspector also visually t
confirmed that the electrical conduits from the sensors for the drywell pressure system had been rerouted to conform with the minimum 25' separation specified in the field drawings.
These items are considered to be resolved.
With respect to the routing and location of the drywell pressure impulse lines and sensors, Jc has since amended the Application
- to reflect the actual installation, and DRL has evaluated the design and found it to be adequate.**
This matter is considered resolved.
13.
Isolation condenser Tank Supports (219/68-8, Paragraph IV:E.1.a.)
Visual inspection of the subject supports by the inspector during the November 22 visit revealed the presence of a "lubra plat,e" under each support.
Also, close examina-tion of the detailed drawing of the supports, made available earlier by GE, confirmed that the installa-tion was in accordance with design requirement.
This matter is considered to be resolved.
14.
Isolation condenser System suspension (219/68-8, Paragraph IV.E.1.b. )
The deficiencies previously noted here were determined to be still outstanding as of the November 22 visit.
However, the inspector confirmed that these items had been picked up on the GE punch-list system, discussed
- Amendment No. 45, Page V-6-1.
- DRL Supplemental Report to Report No. 5 to ACRS, Docket No. 50-219, dated 11/29/68.
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Consequently, j
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- 15. : Isolation Condenser System - Gouces Across Weld At~
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Flued Head (219/68-8, Paragraph IV.E.1.c.)
AS The inspector determined that-this' item.was still-out-l j.
standing as' of the November' 22 visit; however, it was j
observed to have been added to GE's punch list.
On this basis,-this item is considered to be resolved.
t 16.
Isolation Condenser System - Questionable Adequacy of Weld at Valve V-13-35 (219/68-8, Paragraph IV.E.1.c.)
Mr. Loeb informed the inspector that the subject weld was examined by GE and considered to be adequate. 1This j
item is considered resolved.
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Isolation Condenser System - Double Inlet Isolation Valves (219/68-8, Paragraph IV.E.1.c.)
l A follow-up review of the fabrication and testing records
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relating to'the. subject valves.was~ conducted by.
i Mr. Collins during the December 2 - 3 visit.
The i
results are summarized in the following paragraphs.
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The isolatidn valves were supplied by_.the Anchor l
Equipment Company. (AEco), to meet the ASA standard i
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rating for conditions'of service up to.1500 psi'at 6000 F.
Amendment No. 32 of Ene FDSAR and the Burne and Roe specifications S-2299-61; contain the. specifica-tions and quality control provisions for the' valves.
A brief summary of these follows:
a.
Materials are to be cast 304 or 316 stainless j
steel conforming to ASTML A-351, Grade CF8 or CF8M specifications.
b.
Pressure containing castings shall be liquid penetrant examined in accordance with paragraph N-627,Section III of ASME B&PV Code.
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Complete radiography of all valve bodies and all A
1 weldst including welds' joining. valves 'in series
.I and welds joining valves and valve extension
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pieces (flued fittings) in accordance with para-graph N323.1 (castings) and paragraph N624 (welds) 4 of Section III of ASNE B&PV Code.
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Hydrostatic testsoof completed valve body extensions
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and final welded assembly of valves at a test pressure of 2175 psig.
The pressure containing components of the valves consisting of body, bonnet, discs, retaining rings and pressure seal gaskets were manufactured by Roemer
- Electric Steel Foundry for AECo.
Roemer casting inspection reports showed that cast stainless conforming to ASTM A-351, Grade CF8M (Ty 316 s.s.) was used for r-these components.
The inspector reviewed the chemical.
analysis and mechanical properties test data and determined that the components were compliant with the specifications.
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The valve components cast by Roemer were radiographed by Conam Inspection, Inc., under contract to X-Ray Engineering Corporation.
Conam's radiography examina-tion reports provided information on:the techniques i
employed and specified that valve components were 100%
radiographically examined to acceptance standards of ASTM E-71 for Class 2 castings, as specified in para-graph N-627 of Section III of ASME B&PV Code.
All I-defects beyond acceptance standards were identified (location and type) and acceptability of repairs of i
these defects, as determined.by re-radiography, was provided.
The inspector also determined that Conam, X-Ray Engineering and GE representatives reviewed the j
radiograph file on separate occasions and have independently concluded that the valve components are acceptable for the intended service.
The radiographs were not available at the facility for review by the inspector.
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kc Roemer. Foundry inspection' reports indicate;all-valve l
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. components.were'100$ liquid penetrant examined;in' iP accordance with ASTM standards.E165-60.
According to L
these reports, no defect-indications requiring repair:
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were observed by. this. inspection. : It should be noted
^j i that ASTM E165-60 is substantially equivalent to para-graph N-627 of-Section_III of ASME B&PV Code.
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AEco test reports specified that individual valves and-
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.1 valves ~ welded in series were hydrostatically tested, y
both in: assembled and disassembled condition,:at 2175
'l psig for 10 minutes.
Concurrently, valve. seat-leak rate' l
tests were performed hydrostatically at 735 psig. for - 3 ;
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' minutes. ~No visible sign of leakage resulted1from l
these tests, according to the reports.. The ABCo further.
2 alluded to the. structural quality:andtintegrity of the valves through the satisfactory performance of:a 1440:
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~ V psig " shock test" - (Hi Impact) ' on ' a valve of identical.
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design at the Hughes Aircraft Company.
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the shock test and-post-shock inspections is available~
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Following the individual' valve' tests, two valves were -
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welded into each in-line assembly by the Advance Heli-.
Welders & Manufacturing Company,. Berkeley, California..
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A report of welding data,. including procedure. used an'd:
.I welder qu'alifications, provided by Advance Heli-Welders was notedto be in accordance with Section. IX of the ASME B&PV Code as,specifiedfin the B&R a
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specifications.
The report also specified that root'
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i and final passes of weldswere. liquid penetrant'. tested 1
and that no repairs resulted from this inspection.-
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After completion of welding, AECo performed a liquid penetrant examination of critical areas on the valve assemblies.
A review of inspection reports indicated that no repairs were necessary as a result of welding.
The areas examined were:
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l Seat springs By-Pass piping welds Main In-Line valve welds According to certified inspection reports,.X-Ray Engineering Corporation radiographically examined the shop welds which join valves in series; and valves to flued fittings, to requirements of paragraph N-624, 1
Section III of the ASME B&PV Code.
A review of the I
radiographs revealed the radiographic quality was excellent.
In some instances,1-lT penetrameter l
sensitivity was obtained.
While closely examining weld soundness qualities exhibited by.the radiographs, a film condition was observed which was characterirstic of a 2-1/2" long linear discontinuity.
This film was noted I
to be the 0-1 location on the shop weld joining Valve No.
V-14-32 to Valve No. V-14-33 as shown on the isometric drawing of the system.
Discussions with GE and Branch Radiographic Laboratory representatives failed to produce convincing arguments as to whether the film condition was 9
attributable to a weld defect or a film artifact result-ing from the radiography process.
Further, no evidence was presented that would indicate the radiographs of this weld were examined by GE during their overall review of facility records.
In contrast to this fact, i
radiography reports of other valve weld joints indi-cated GE acceptance, as evidenced by signatures of GE representatives involved.
The soundness quality of these welds, as exhibited by the radiographs was determined to be within ASME B&PV Code acceptance standards.
The deficiency in the radiography was referred to the GE representative, Mr. Loeb, for resolution.
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Subsequent to this inspection,.Mr. E. Franks'of GE.
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telephoned Mr. Collins in regard to the action taken
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by GE in' resolving ~the.above deficiency.
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information was-obtained:
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GE concurred with the inspector's' radiographic -
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Using two. porosity indications in the original b
radiograph as locators, the area was' re-radiographed -
1
.with Ir-192 and double film techniques by Branch Radiographic Laboratory. -Radiographic quality obtained was equal to the origital radiograph '(1-lT penetrameter sens.).
i
- i I
c.
The.new radiographs did not exhibit the linear dis-i
- continuity.
Hence, GE concludes.that the original radiograph contains a film artifact.
Records will-q j.
be changed'to denote this condition.
L 1
d.. GE believes the film artifact resulted from film-
{
screen pressure marks.
Further, these pressure marks were caused by the ' radiographer's fingernail
'j
~
l during a placement:of the film. cassette in close j
contact with the weld, i
e.
The new radiographs have been retained by GE and are available for compliance review, if-considered I
necessary.
I Based on the information presented above, it is concluded that the two sets of double inlet isolation-valves in the isolation condenser. system'were fabricated and tested in accordance. with commitments t.
in Amendment 32 and the applicable Burns and' Roe z
specifications.
Examination by CO of the radiographs of the; valve castings and the new' radiographs on the one questioned weld is planned as verification of the
- i 1
information presented above; however, its completion j
prior to the issuance of the provisional operating j
license is not considered to be necessary.
I i
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18.
Electrical separation -' Core Spray,' Auto-Depressurization j
- {-
and 125 V DC Control-Systems (219/68-8,- Paragraph IV.E.
b, & c.)
i l
.j i.
An examination of the results of ' the ' November '11
. 22,
[Q' 1968, electrical audit by GE, discussed ~further in
, i ;
paragraph II.B.2. of this report, revealed that the
.j'~.
deficiencies identified in' the subject systems by Co J.
during the October 15 - 18, 1968,< inspection were l
included on the list of items to be corrected by GE.
Satisfactory completion of this work will be verified on a' sampling. basis by the inspector during subsequent visits.' otherwise, these items are considered to be.
resolved.
19.
Testino of Sensor Cable Conductcr Insulation (219/68-8, Paragraph IV.E.2.d. )
Checks made by the inspector during the December.2 3
1 visit confirmed that this work was in the process of satisfactory completion.
This item is considered resolved.
t 2
4 20.
Separation of 125 Volt DC Batteries (219/68-8, Paragraph IV.E.2.e.)
As noted in the referenced report, the separation
.[
between the two battery systems meets the' definition of a two-hour fire door but does not meet the. fire wall 2
classification 'as described in the Application.* 'At the time, GE proposed.to solve the discrepancy-administratively.
Subsequently, however, they took I
the position that the installation was in accordance with the Application.
i This matter has been discussed with DRL and the ACRS.
Both groups have been made aware of GE's position.
i
- Amendment No. 20, Page II-4-1, Paragraph II.4.
Also, Uniform Building Code, Volume I, 1967 Edition, Section 42 - 43, Fire Resistance Standards for Fire Protection.
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Based on the comments received during these discussions, it.is concluded that although the Linstallation:does 'not l
- [
meet the words in the Application, it is considered to d
be: acceptable.
1 1
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s u 21.
Battery Room ventilation (219/68-8, Paragraph IV.E. 2.e. ).
m,7I M.
i The ' adequacy of the ventilation.in' the subject room.
j l'
was first questioned by Co during the inspection on-l October 15 - 18, 1968, and has been followed up on l
during subsequent visits.
The principal concern is the i
ability of the system to cope with hydrogen generation -
j from the station batteries, which are lead-acid.-type.
1 Visual examination of the subject equipment and related --
field drawings, a review of.the results of related-l Burns and Roe and GE studies, and discussions with j
t c6gnizant GE and Burns and Roe. representatives has j
provided the following pertinent information:
l 3
I a.
The two battery systems share a common-room-i separated only partially by the two-hour fire:doo V l
b.
A single ventilation system, with 'one supply and.
one exhaust fan, serves both the' battery room and
[
the M G. room (located.directly below.the battery i
room).
l c.
The design air flow through the subject rooms 'is l
~
1 l-i 84,670-cfm, 4470 cfm to the battery room and j
[
80,200 cfm to the M-G room.
1 i
-.j d.
The ventilation system is set to provide 500 F air
- l1 (outside temperature permitting) to both rooms.--
U l
The system is designed to allow 100% recirculation of the air when the temperature falls below the 500 F set point.
e.
Failure of the. fans is indicated. in the control room with a pilot light controlled by auxiliary o
[
contacts on the blower motor controllers.- There are no provisions to detect loss of flow per se (Ex: a failure of the damper system while the F
fan motor ' is still running. )
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The;results'of a Burns.and Roe study of the l
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adequacy of the ventilation system
- shows that V HI when the batteries-are being charged,.a' maximum T
of 176 cfm air' flow-is required to maintain.the' hydrogen concentration?at or below 2% based on-i Y
107 0 m.,Nl -
battery temperature and "end of; life" battery F
condition.
Also,ithat for 1/2 life' condition and
. xY T.
770 F-battery temperature, the ventilating air.
I requirements would be 16.4 cfm.- The documents j,
indicate that assistance-in these computations was
~
obtained from the battery supplier,LGould-National.
The same study states' that the worst condition i
would exist when no heat load is being. generated 1
in the M-G room (stated as being "very conservative"),
the battery changes are in operation, and it is 0
cold outside (assumed'+ 10 F design).
On this j
I basis, the study shows that.1675 cfm of outside q
air would be required to absorb the heat generated.
~j in the battery room"(assumed design heat load in room - 21.2 kW) and still-maintain a 500 F dis-
~
charge temperature.
The study concludes that t p amount, 1675 cfm, "is considerably more than the.
amount required to maintain. less than 2% hydrogen -
concentration."
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A separate but related study by GE**Tshows~that 22 cfm of outside air would be required to maintain a 4% hydrogen concentration.
This is based on 0.877 cfm hydrogen evolving from the batteries charging I
0 l
at equalizing voltage at -77 F and "end of life" q
condition, which'information was documented as "I
also having been obtained from the battery supplier.
This study goes on to state that the actual hydrogen rate would be less than the 0.877 cfm a
a
- Transmittal memo from Burns and Roe to GE, dated 7/10/68, with attached calculation sheet dated 7/1/68.
- GE Engineering Calculation Sheet by J. C. Longwell, dated 8/28/68.
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since the. battery temperature is estimated to' i
l be 52
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The' study' _
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sh ws thatJtheL1675 cfm;outside air' requirement di t
computed by L Burns : and Roe is 76.11 times that -
j n
. required to maintain 4% hydrogen.
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A third study, also"by.GE. shws that assuming the' j
Tj9 0.877 cfm hydrogen. generation rate,- 7-1/4 hours j
1 would be required to reach a.4% hydro' gen ;
j concentration -(assuming complete distribution and i
thorough mixing) follwing total. loss of flw.;
j i
- 10. Discussions with Messrs. -Loeb and-Mellar during -
the December 10 --12 visit
- regarding the adequacy-of the ventilation system, including specifically l
operation at 100% recirculation flow with a condi-tion of maximum hydrogen generation,. elicited a' positive response that'in their opinion there was-no hazard.
The same-situation was postulated'to-Mr. Lari, Burns,and Roe Project: Engineer, in afphone' communication held subsequent to the" December 10 -
12 visit.
Hisl response, following; consultation.
j with others in his office,. was that-system leakage -
{
during 100% recirculation,' conservatively estimated p
to be 2% of total flow (~:1700 cfm), was more than
-l enough to maintain the hydrogen. concentration well' 1
=j within safe limits even-during periods of maximum hydrogen generation.-
1 Based on the above'information, it' appears that the' l
ventilation _ system serving the battery room is
]
[
adequate to cope with the hydrogen -generated during l
[
normal operation, including when operating in the 100%_
recirculation flow mode.
It is not' apparent at this l
time, however, that the system will adequately cope with the hydrogen in a situation where maximum hydrogen generation -is combined with a loss of
- Paragraph II.D.4. of this report.
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- ventilating air flow. (Ex: Loss of all' off-site power j
y accompanied' by a loss of coolant.: accident.*)' This, j
~
matter is currently being pursued and will be reported h
ll on in the next report.
j 4
m sa i
22.
Reactor Pressure Vessel Internals
~M4, (219/68-1,- Paragraph IV.C. and 219/68-2, Paragraph'II.C2I I
,I A visual examination of the--shroud head and steam' 2
4 separator assembly and. discussion with cognizant GE
{
site' representatives,by Mr. Collins, revealed that the l
necessary steps were taken to rework and: restore:all j
structurally deficient' welds on this equipment tot the.
design requirements.
New welds', made necessary~by-
]
cutting of standpipes for access to. rework' areas,.
1 exhibited _ good welding practices.
It was also-noted;
]
that every effort was made-to remove-objectionable 1
(
surface defects (weld. spatter, arc strikes, : undercutting; i
etc.) on the structure and to properly contour restored.
]
weld reinforcements.
According to the GE site' construc-i tion supervisor, the cutting, welding 'and inspection -
I tasks involved in the rework program was under direct' GE supervision and was performed by the. original fabricators, the P. F. Avery. Corporation of Bellerica,-
d Massachusetts. 'The inspection records' associated with-J f
the rework' program had been' shipped to theLGE office (San Jose) ~ and were not reviewed by the inspector.
Modifications to the steam separations. (pre-dryers) comparable to'the' changes being made at.Nine Mile Point **
i-were being implemented at this facility.
These modifications were discussed with DRL by JC-GE at a meeting held on October 31, 1968, attended by the writer.
. 2
.i
- Amendment 32, Page 3-1, indicates that in this situation the battery changers are manually started but the ventilation fans are not.
- CO Report No. 220/68-8, Paragraph II.A.l.
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j Through a sampling visual ~ examination of equipment,*
Eg a review of site records and discussions with cognizant
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GE. construction engineers,-the. writer determined that GE had adequately inspected the remainder of'the' y
reactor pressure. vessel. internals supplied by'P. F..
,, C -
Avery and Knapp Mills, and that' corrective action had
].
been or was being implemented.as appropriate for the J
j deficiencies identified including.those identified'in' the referenced inspection' reports.
Based on the above observations, the subjectcof deficiencies in the reactor pressure vessel internalsJ is considered to be resolved.
B.
Plant Quality - Supplemental Efforts by GE-
- 1. - Construction Records Inventory and Review **
4
- The results of GE's efforts-in this area, including
'the status of resolution of deficiencies identified,.
were reviewed.during.the November-21 and 22 visits..
a m **
I some of the items identified were confirmed as being comparable in significance to those identified by Co~
in their own. independent' audits.
An examination by j
the ~ inspector,on.a sampling basis, of the status of resolution of some.of these~1tems_showed that they-either had been resolved through proper engineering dispositions or, where 'still outstanding,: were appropriately covered by the GE punch-list; system (discussed further in paragraph II.C.'of this report.)
As a~ result of this review, CO concludedJthat the records inventory and review program represented.a.
reasonably meaningful effort on the'part of GE'towards?
gaining additional assurance regarding the'overall'
~-
quality of the plant.
i
- Conducted over a period of sevetal:pievious! inspection visits. -
- Previously discussed in CO Report No. 219/68-8,- Paragraph II.B.
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2.
Electrical Audit - Cable Separation p
As a; result of the October 15 18,_1968,. inspection
- p~ -(-.
.during which a number of significant, deficiencies i
wererfound in the installed electrical system, CO-informed JC-GE that additional assurances, of plant-i
.y%
quality, wich emphasis in the electrical area,. musir R
be provided before CO could arrive at a favorable position regarding c6mpletion of the'fac111ty.
~
subsequently undertook another audit in.the electrical area.
This was done from November 11
.22, 1968..The scope and results of that audit were reviewed during the November 21~and December 10 - 12 visits.
The high '
lights are summarized below:
a.
A task force was established for this purpose.
The task force was comprised of 9dognizanteperags,.
4 3 on-site and 6 off-site.
- b.. A total of 80 mandays were devoted to the project. -
~'
c.
The systems audited include the following:
1 (1)
Core' Spray 1
(2)
Containment. Spray (3)
Containment Isolation
-l l
(4)
i (5)
Auto-Depressurization f
'}
(6)
Reactor Protection i
d.
Simplified operational block diagrams were prepared from the various' installation drawings.
Only~those components and cables important to the successful' operation of the system were included.
Components,
.)
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- CO Report No. 219/68-8.
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key panels and the ~ interconnecting cables were identified...The method of cable installation, r
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l 1.e., conduit or tray, was also ' indicated.
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e.
A; copy.of the separations criteria utilized is j
j given in Attachment A.
Subsequent.to informal.
.j lid review and acceptance by DRL,.these same criteria
) N]
- were used by CO as a basis' for the December 10 -
.)-
12 inspection of the containment spray system, i
discussed in Section II.D. of'this' report.
f.
The ' completed block diagrams were then taken to the field and the cables on the diagrams physically
.]
located.
Those cables not meeting the separations j
criteria were identified and listed on specially 1
prepared sheets showing their present routing and 1
the required rework..Those cables not yet installed j
were also identified and the proposed routing.
specified to meet the separations criteria.
t I
g.
A summary report of the results of the task force's' efforts, dated December 2,1968, was reviewed by.
the writer and noted to contain the following H
j tabular summary:
1 Deficiencies Old Items New Items Total Cables to be rerouted 25 37 62 Cables not yet l'
installed (to be checked) 58 1.j' The old items given in the table were those y
!f identified prior to this most recent audit.
h.
The following statement was given in the task force report.
"The task force concludes that with the correction of the items identified, the OC-1 electrical system will meet the separations.
criteria dated 11/27/68 in every respect."-
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The task force recommended that similar auditing-L programs be implemented, expeditiously, at all i
L other domestic turnkey projects.- Also, that the i
audits include not only cabling but-all non--
i s
j electrical items of: safeguards systems as well..
i
' {p&f:j%
upper GE management,' including Mr. Dickeman, i
j.
The subject report was noted to be directed to-Manager, Domestic Turnkey Projects.
,i i
k.
Engineering orders have been issued to the' field' to implement the necessary modifications.
Site j
j management is responsible for this implementations.
2 however, per Mr. Loeb, GE - San Jose will followup to confirm satisfactory completion of the work..
1 A subsequent comparison of the task force report with the engineering orders by the inspect' ors during the-December 10 - 12 visit, indicated that more cables were identified as being discrepant than had been listed for-installation or relocation.
Subsequent discussion with GE indicated an apparent problem in semantics.
The. inspect' ors concluded from the discussion l
that -the persons involved knew what' they were doing and
{
that the end results would be satisfactory.
Based on the observations noted above, the inspectors concluded that this latest effort by GE in the electrical separations area was adequately responsive to the concerns' of CO in this area.
This-subject'is also discussed in Paragraph II.D.4. of this report.
Cable Trav Loading
- J According to Messrs. Loeb and Mellon at the time of a
the December 10 - 12 visit, the report resulting from GE's investigation into this subject area was in the s process of being prepared and would not be available j
to the field at that time.
However, both claimed
- CO Report No. 219/68-8, Paragraph IV.E.2.g.
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O, knowledge of.the contents of the report, in general, and stated that it. reported that their investigation
[
had revealed there were no heating problems-due>to
.-s cable -tray. overloading at ' Oyster Creek.. Upon questioning ~ Mr. Mellor revealed that-early analytical d
' evaluations of cable heating;due to cable tray loading' had indicated that they would-have some problems if IPCEA - (Insulated Power Cable Engineers Association)-
(( C, approach
- to analysis were used.
To achieve the L
j results obtained, they modified the IPCEA standards by i
changing the. circular displacement of. the cables in the tray with the highest current carrying cable in. the l.
center, to a rectangular configuration which they felt more nearly met the condition in the plant.
This rectangular confinguration produced more optimistic results than the earlier analysis.
In addition to the analytical approach to the cable tray loading problem, they had conducted experimental tests which seemingly P
i substantiated their-analytical approach to the problem.
This information is consistent with. that presented by '
GE-at the December 5,~1968, meeting of the ACRS, attended by;the writer.
Mr. Loeb-indicated that the subject report would be t
available for Co' review ~at the site by early January,-
[
1969.
The inspector reiterated Co's desire to review.
i the report as well as GE's plans, if any, regarding l
l in-plant verification.
This matter. remains outstanding
-[
pending satisfactory results from this followup action.
The subject'of cable tray loading is also discussed in paragraph II.D.4. of this report.
C.
GE Punch List System i
Because of CO's previously stated need for additional 1
assurances of plant quality,** attention was focusedLon l
1
- Used at San Onofre.
3
- CO Report No. 219/68-8, Paragraph IV.E.
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l the punch list system as one means of providing.such.
1
- assurance.- As previously arranged, GE - Messrs. Loeb w4 and Strand - presented an outline of'the system in effect l
~~
at Oyster Creek -1 during the November -21 meeting. ; The
. consensus-of the CO representatives present at the. meeting was that the system as described, when combined with ' the.
j
%. ? M;,,
preoperational test program, appeared adequate to provide 4
a means of. detecting significant deficiencies in plant-
~
If installation (as designed), and to ~ maintain follwup-l I
control of identified deficiencies.
+
.i As a means of evaluating the effectiveness of the punch list system, the writer, during the December 2_ - 3 visit,'
1i a-conducted a walk-through of the system.
The status,
'j including degree of followup control being exercised, of.
)
eight previously identified deficiencies, was examined as j
l part of the walk-through.
Six of the eight items were
(
- a.
determined to be either completed or adequately identified g'
on the punch list system.
The remaining two items, which j
p were still outstanding (and which were of lesser significance f
in the judgement of the inspector), although known to some T"
GE field engineers present, were not clearly under
}
controlled followup.
Action was initiated by GE at that
)
time to affect a greater measure of control on these latter I
I items.
Based on the above. observations, the punch list system-at y
this facility, although not as tightly controlled as might be desired, is considered to be reasonably adequate.
D.
CO Followup Audit - Electrical Areas This section of the report summarizes certain results of the December 10 - 12 inspection visit, specifically the observations made by Mr. Bower regarding the containment spray system (electrical separation)', the air ejector off-gas system (instrumentation and controls), and conduit loading (compliance with applicable standards).
As previously discussed in paragraph II.B.2.e. of this report, the separations criteria given in Attachment A H
were used as a basis for inspection of the containment a--
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spray system.
Items discussed and~ comments made at the l
^
1,
' management' interview held at the conclusion of this visit Tl 6
are also presented-here._
l
.2 l
i,
'l..
containment spray System i
r
_[2-A GE ' field ' engineer, Mr. - Fletcher, - and Mr. Mellor, l
assisted the inspector during the inspection of the i
4 containment spray system installation.
Mr. Fletcher-I 1
is the field engineer with the primary responsibility
~!'
for the electrical parts of'this system.
To get a' working understanding of the system, the
)
elementary control diagram and the piping and instrumentation drawings-were reviewed with Mr. Fletcher i I
and Mr. Mellor.
It was determined that the electric'al design was predicated on the fact that -the containment
{
-spray system was an extension of the containment vessel.
i since the valves provided are not__ isolation valves, they j
are operated in.the open position _at all times'except iwhen. closed for testing. purposes. 'The-test circuitry-
+
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is arranged in such a. manner that any signal initiating
~
operation of the containment spray' system will cause 1
j these valves to be' driven to the open position over- -
)
l riding the test circuitry.
System operation is-initiated when a reactor vessel low water, signal j
i-and containment vessel highgressure signal is experienced.. The containment vessel _ pressure switch is
~
det to. close upon two psig increase above operating i
pressure.
It'was during the review of the system _ opera -
]
tion elementary drawingr that it was observed that system-i operation was dependent upon the main. test valves being opened.
This is accomplished by a-limit switch on the valve motor operator which closes when the valve is
.4 driven to within 2% of completely.open.
Without this j-valve in the open position, the pumps will' fail to start.
In GE's review and preparation of block
'i diagrams, they had failed to note this particular..
function, consequently, the cables associated with this contact were not included in their review.:. Detailed i
physical inspection by the inspector revealed, however, 1
that these cables follow divergent routes from the valve operator to logic panels BA and 8B'in the
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1 460 volt distribution room and do meet the' separations.
f
'4 criteria.
It was pointed out to the GE. representatives l
'that these, cables must be includedLin their reworking'
-plans for this' system..The representatives; signified..
g 1
their understanding that a_new dimension had been:
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added to their problem and they would take the' Qlp !=
necessary action to include it when the're-routed tho' j
p-. q cables.
{
Although corrective work had' not commenced on those" i
c1
?
cables-identified.as requiring relocation and those cables identified as not having yet been installed, it.was determined from the action proposed and from-the hardware that had been installed, that the' separations-criteria from the containment spray system had or would:
be met.
d 2.
Air Ejector Off-Gas System Although not'one of.the engineered safeguard systems, the air ejector off-gas system performs an'important y..
function related to the public safety of the plant.
j The control features of this system, along with the.
physical installation,.were examined to determine if
,j.
the provisions'of the FDSAR had been' met.
3 l
The elementary control ~ diagrams were reviewed with the assistance of Mr. Green and Mr. Loeb.' - This review'of the latest documents available revealed that additional k.
controls had been installed since the issuance of'the original FDSAR flow diagram.. This added control initiates isolation of the off-gas system when high-radiation levels are detected in the main steam lines.*
i b
The original controls of this system, which. included radiation monitors and controllers.which scan the gas.
i in the off-gas system, remain.as before.
Either one-of these controls will close the main isolation valve j
at the effluent stack.
Subsequent to closing this
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valve, pressure will rise in the hold-up piping upon j
- Discussed in Amendment No. 32, Page 16-1.
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continued operation of the ejectors.
When either a y
high pressure or high temperature is sensed in this
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hold-up system by instruments located therein, the p
isolation valve behind the ejectors will close, thus g
effectively isolating the entire system.
Examination of piping and instrumentation drawings at M
the site, as well as the physical installation, revealed I
additional piping and automatic controls beyond what was
]
shown on the P&ID.
These controls and piping, in general, proved to be supplemental to those shown and were provided as additional details to make the text and installatien agree.
The complete operation of the safety features of this system were not entirely apparent from those drawings available and reviewed at the site.
Discussions with the responsible GE engineers revealed the operations as described.
Questions posed about returning this system to service after clearing an anomaly indicated l
that the system resets automatically upon sensing the reduction in radiation to safe levels.
This resetting operation automatically opened the valves permitting effluents to discharge through the stack.
Physical inspection of the installed system revealed it to be approximately 98% complete.
Hardware and wiring is available to complete the job at any time chosen by the contractor.
Since it is not an j
essential system prior to actual power operation, they.
have not chosen to complete it at this time because of the press of other work.
However, discussions with
}
responsible GE engineers indicated that it is currently scheduled to be completed prior to initial loading.
3.
Conduit Loadinq To determine conduit loading throughout the plant, sample inspections were made at several different locations.
The standard checked against was the
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t conduit loading chart: contained in the National Electrical Code.
The sampling reviewLdid not reveal-
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-any conduits' loaded beyond the established limits in I
the aforementioned code.
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4 Management Interview
- 'nai An interview was held with JC-GE management representa-2 tives at the conclusion of the visit on December 12, 1968.
The pertinent issues discussed and comments made are summarized below.
Those Present -
Name Organization D. Ross JC E. J. Riggle JC t
L. M. Loeb GE A. G. Mellor GE R. T. Carlson AEC F. U.
Bower AEC t.
R. T. Dodds AEC Mr. Carlson conducted the'. int 6rview.
Mr. Bower commented as requested and as appropriate during the meeting.
Mr. Dodds was an interested observer during j
the meeting but did not contribute since his inspection
{
function was not related to those subjects discussed j
during this meeting.
Upon opening the subject' of the electrical audit just completed by GE (discussed in paragraph II.B.2. of this report), Mr. Carlson commented that the work performed by GE was apparently quite extensive and seemingly accomplished in a professional manner.
The fact that Co's review of the containment spray system had i.
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revealed an essential control that they had failed.to
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recognize, was considered as a "minus" quant'ity that i
jt appeared not to be related to the review results of
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the other systems in any known manner.
It was pointed out that the inspection had shown that the cables associated with these controls did meet the separations
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criteria, however, their presence introduced a new dimension in the routing of other essenti,a1 cables of I
l this system which would require new field instructions.
Mr. Loeb and Mr. Mellor agreed and stated that the new conditions would be considered in their corrective action.
They also agreed to assure themselves that the aforementioned oversight in their audit of the containment spray system indeed did not have applicability to the other systems audited.
Mr. Carlson commented on the apparent discrepancy between the GE field audit report of the cables as
~
installed in the plant and the field instructions i
issued to correct those condiudons found discrepant.
A comparison of these two documents indicated that more cables were identified as discrepant than had been listed for installation or relocation.
The discussion results indicated a certain difference in language and some misunderstanding as just what was meant between j
the two documents.
Since there was seemingly a problem
}
of semantics, CO concluded that the people involved in j
the solution to the problem knew what they were 'doing i
and the end results would be satisfactory.
In any event the end results could be easily checked and i
CO would satisfy themselves in this regard at some future date.
I i
i 1
Mr. Carlson concluded that, based on the examination of the scope and results of GE's audit and the results of CO's followup audit, GE had been adequately responsive to CO's concerns in the area of electrical separations.
In response to a question by Mr. Carlson, Mr. Loeb stated that the GE San Jose office was initiating a separations criteria review of all domestic turnkey 9
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projects that they'had.'in work.
It was the. opinion 1
of those.present that such:a review at'an earlier.date.
than that performed at-Oyster Creek should increase, the possibility of reducing problems'at the other:
,a I plants.
Mr. Loeb was requested to report on an' investigation-E
"[i he had agreed to make to determine'if the containment spray piping was an extension of the-containment vessel.
His report stated that the. containment spray piping was indeed an extension of the containment vessel,
- However, it had not been, constructed to Class I requirements because it was designed as a 150 psi system.
Mr. Loeb spoke of an agreement with DRL to provide Class.I piping for this system to the first valve on the~ suction side only.
From that point on, the installation.had.been made as a standard industrial system with visual inspection only of the welding.
Mr. Carlson stated that the' significance of this.latter fact would require further evaluation by CO and that.JC - GE would be informed of the results thereof.
Mr. Carlson expressed his regrets that the. report-on cable tray loading, had. not been.available~ to CO at -
the time of the inspection.
Mr. Loeb and'Mr. Mellor both professed to know the contents'of.the report and:
said that in general it reported: that. there would 'be.
no problems of overheating of cables at any place within.
the plant.
They further stated that this report was~in t
the process of preparation and would be made available to CO at the earliest-possible date.
Mr. Carlson i
reiterated the desire of Co to review the subject report' l
_l as well as GE's plans regarding.in-plant _ verification 1
of-the results of the investigative--atudy.
q 1
It was reported to those present that the review of
~
conduit loading throughout.the plant, made during the inspection, had revealed no' conditions of conduit over-loading using the criteria for conduit loading as set
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forth in the National Electrical Code, p
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A. review of.the battery room ventilation' system.
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drawingsxhad revealed that.the system was,. designed to-1 4
close -the exhaust damper andi open. a: bypass danper when:
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the room'temperatureifellLbelow'50 F,1thu'sLproviding.-
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'100% recirculation of air. 1The. inspector's' question l
lM regarding safety conditions under'this mode of' operation q
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coupled with a condition of maximum hydrogen generation
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elicited a positive responsa from Mr.' Loeb-and Mr. Mellor i Tj:
that;in their opinion there.was,no hazard..
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',f E.
Cracks in. Emergency Core Spray System' Valve i
n During a cursory ex'mination of the emergency core spray;
~l a
system at the time of the December 2 - 3 visit, Mr. Collins:
visually observed several linear indications on one'of the.
1 solation valves in - the south side piping arrangement of.
s the system. -These indications were about-1/2" to 2" in u
i length and were located on the valve body where it was machined for seating the yoke clamp.
The Burns and Roe purchase specifications indicate the valve was procured from Anchor Equ'ipment Company.
Markings on the valve identify it as Valve No. V-20-41,: Grade CF8 cast. stainless-4 steel (Ty,304), Heat No. 4-12, Radiography No. 7588. -
The above findings were brought to the' attention of Messrs..
McClusky and Loeb.
Mr. Loeb. expressed concern over the-inspector's: findings and stated that an immediate investiga-
. tion would be initiated.
The scope and results of that-F
.i investigation, as of December 13, 1968, are documented in Inquiry Memorandum 219/68-B (copy attached as-Attachment -B).
The highlights of subsequent developments, obtained.in
'l telephone' communications with Messrs. Loeb and Strand up' to January 9, 1969, are summarized in the'following para-graphs:
1.
-The task force examination of installed valves has.
'j.
been completed.
A total of 21 valves, including..the original defective valve, were examined - three. (core spray) using dye penetrant'and the remainder visually.
All were from the same' supplier,' Anchor Equipment--
Company.
The following is a tabulation of valves l
checked, by system-l
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Isolation Conde'nser 4
I' CRD Hydraulic Return 1
Total - 21
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2.
Questionable indications were detected'on the bodies-of four valves, all by' visual means followed with confirmation by dye penetrant examination.. The rest of'the valves were said to be " clear". 'The indications were removed on two of. the four by light grinding.
1 Both valves are in,the cleanup demineralizer system.
l The. third valve, still being investigated, is a 10".
return valve (No. V-14-34) in the isolation condenser i;
system.
_l 3.
The fourth valven No. V-17-54, is a 14" isolation l
i valve in the reactor shutdown cooling system located i
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inside the drywell.
Here, an ~ 3/4" linear indication-l was detected in the valve body, approximately 3" from.
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the downstream weld, and was said to' be at an area of' dimensional transition (latter point also-true of original defective valve).- According to Mr. Loeb, the cause of this defect was'apparently the same as'
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that experienced in the original-instance.. The i
indication was removed by grinding to a depth of 3/8".
1 The wall thickness was reported to be ~ 1-3/8" at thir,
'j point.
Current plans are'.to weld repair and rehydro' when this work can be fitted into the schedule' with-1 minimum time loss.
In any case, it will be completed j
prior to loading.
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4.
According to Mr. Loeb, the only stainless steel valves
.j inside containment that were not examined were the L
four air operated check valves in the core spray system.
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1 GE's: rational for not checking these was that.they.
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final' heat treatment.
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- 'i 5.- Otiher stainless steel' valves.from suppliersLotherithan j
d those named are in use at;thisffacility.
InLresponse j
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tota specific question, Mr. Loeb' stated'that none of-
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these' were examined nor were they ~ researched regarding (f
the timing of. vendor-radiography.
GE's basis was that q
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none of-these was;in the. original category of! critical-'
valves ~to be checked of safety-significance and.non.
isolatable.
j 6.
In answer to another question, Mr.. Loeb stated with :the
' exception of the original case, none of the 21. valves examined were re-radiographed.
His rational was;that because of the nature of the cause of the condition, the defects being' observed would evidence themselves j
-first on the surface and therefore could be detected.
visually.whereas they might not be detected with.
radiography.
7.
The original defective valve, No. V-20-41 (8") will-be given a " post-mortum", including metallographic examinations, following replacement.
i The significance of the-cbove described observations was' discussed by CO in a meeting with representatives of DRL 3
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and DRS on January 8, 1968.
As a result, the Regulatory plans to hold meetings with JC and 'GE, tentativ$ly. scheduled d
for January 17, 1968,_ to discuss both the specific'situa-tion at this facility as well as its possible applicability;
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to other facilities.
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Containment Penetration Restraints
'I During the December inspections covered by this. report,
.l the inspector' learned that GE has underway a re-evaluation:
of the need for the subject restraints.- Discussions with 1
Messrs. Loeb and Scott provided the~following pertinent j
information:
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Originally, GE arbitrarily decided to install the restraints both inside and outside of the penetrations without first ev luating the construction-type consequences.
This approach has been reconsidered i
after seeing the size hardware henerally huge) that
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has resulted.
2.
Per Mr. Scott, the re-evaluation is being made with r
respect to the determination.of loads and the evaluation of stresses, but not with respect to the intent of meeting the conditions discussed in the applicable criteria.*
3.
Paragraph I.B. of the referenced criteria. states, in part, that for the pipe rupture design condition,.
"---the resultant stresses in the drywell penetration shall not exceed 90% of the material yield stress of 33,000 psi."
Per.Mr. Loeb, the study may result in a higher percentage figure in'the foregoing quote.
Mr.
Loeb is aware of the possible resultant need to revise the Application.
i 4.
Installation of the restraints has been stopped pending completion of the study.
5.
The final approach decided on for this facility will be applied at all other facilities..
l GE's re-evaluation of the subject restraints is expected to be completed by early January, 1969.
The inspector will' j
review the results when they become available.
G.
Preoperational Testing - Outstanding Items The status of the items considered to be outstanding as of the last review of this subject ** were discussed during the November 21 visit and are summarized below:
- Amendment No. 11, Page III-15-1.
- CO Report No. 219/68-8, Paragraph II.C.
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,i The subject; procedure has been found acceptable by h
DRL ' and CO without the inclusion of the. dryer in the--
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This matter is considered resolved.
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- I l procedure, reviewed by the inspector, incorporates-i 4
provisions for testing at operating. conditions in l
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addition to the ambient tests as announced earlier l
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by GE.*
2.
Auto-Depressurization System According to Mr..Hess, methods.of testing.this system are still being evaluated by GE.
This matter remains
.l 1
j outstanding.
I The actual preoperational tes' ting of the plant equipment' and systems is underway.
Co followup in this area is currently being provided by Mr. F. J. Nolan, CO:HQ.
The
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results of his observations will be documented in future j
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inspection reports.
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Exit Interviews-1
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The exit interview held with pertinent JC and GE j
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representatives at the conclusion of the December 10.-
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t 12 visit, is discussed; in. Paragraph II.D.4.
In the'other j
j instances, exit interviewsLwere not required due to the nature of the visit or because the. persons involved were i
1 in the company.of the inspectors for all or most'of the j
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visit.
In each case, the pertinent items. discussed'andi
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the significant comments made by those interviewed are-contained in'the body of the report.
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- CO Report No. 219/68-8, Paragraph II.C.3.
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- ENTRAL PCMER & LIGHT IEPORT NO. 219/68-10)
ATTACHMENT A T
I ig for the ECCS is in accordance common process tap' may run in 0YSTER CREEK PROJECT an one variable for a given ECCS wireway tray.
Strol and power wiring to all devices fl0NS PRACTICES FOR SAFEGUARDS 1g a loss of coolant accident are ns r by redundant functions by SYSTElS three feet between trays (except in limited cases in the 480V 1r by a steel barrier equivalent fall and separated sufficiently that likely to disable both systems.
d satisfied by having the cables Ens Requirements and Practices stems in rigid conduit.
ical and Instrumenta' tion Equipment that all' cables operate below rating.
,,' of Safeguards Systems 1 ave not been required to have tparation but must demonstrate iting.
(Any temperature above the of cable insulation, during normal test expected ambient temperatures.)
By APED Engineering
'l room panels are closed to provide nes and equipment separation is to Revised Nov 26,1968 2.3 - D. above.
d control power for services requirec 2nditions is ' attained as follows:
run in independent steel enclo-ts to redundant equipment.
leer cables are run in rigid steel utpment are run in separate trays.
t olt emergency AC switchgear Atomic Power Equipment Dept.
ys.
General Electric Co.
San Jose, Calif.
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The redundant emergency Diesel generators,- their controls, and starting batteries are physically independent of each -
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other and wiring for them is run in separate. cable trays.
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Separation is-in accordance with 2.4 - C. above.
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2.5 Primary Containment Isolation System j
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- A.
The~ primary containment isolation system has features of both s the -
Sensors are de-energized to operate and employ a one-out-of-two twice logic and are in some _ cases shared by the RPS. Wiring for these: sensors -
l is separated *in the same way as that for the RPS (see above). Valves are redun-dant and thus operate on a one-out-of-two logic (like the ECCS).. Cabling fors redundant M.0. valves is physically separated outside of the main controlLroom and cable spreading room and the separation-is' at least three feet in open. trays.:
or at least one of a pair is in rigid conduit.
- i B.
For the solenoid pilots on solenoid' operated main-steam isolation 1
I
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valves, each valve has its wiring enclosed _in conduit except through the drywell penetration.
Cabli.ng to other redundant " fail closed" solenoid piloted a
air operated valves need not be separated.because shorts or grounds in wiring will cause closure of such valves.
)
)
C.
Valve position lights and signals to the event recorder are not-required to be separated.
-l r
D.
No separatiors of cabling is required within the individual control room pancis nor in the_ ebnduits interconnecting the Protection Panels and with 4
other control room panels but inboard and outboard valves control: circuits ~ within__
this zone are in separate jacketed cables and the circuits are designed such that shorts to ground will cause. isolation.
The main steam line excess flow trips are allowed to operate from e single pair of flow nozzle' taps on each of the two main steam lines because these " steam leak" signals are backup to the: steam tunnel-1 temperature switches, which are more sensitive. The excess flow switches are 4.-
mounted in. individual weatherproof cases on a singlelocal rack and are wired out!
through independent conduits.
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- 3.
- Evolution of Separation Requirements
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4 The Oyster Creek plant electrical and control systems wiring (including i
protective equipment wiring) was original;ly designed in accordance with good,
- and commonly accepted, power plant practices.
The entire complex of protective
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' equip-ent was considered in the light of' relative reliability of its constituent parts and was designed with greatest concern,for failure given to those devices or parts which, by their nature, could be expected to have a higher degree of.
m vulne rabili ty.
Thus, the allowance for single active _ component failure, separa-tion of sensory equipment around the Reactor Vessel. and provisions for periodic.
i l
testing has produced a coordinated reactor protection system that is compatible with the service demanded of it.
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It was recognized that the possibility of a damaging fire, although ex-i tremely unlikely in view of past history, could not be completely ruled out.
However, there is no correlation between fire in the plant outside of the drywell and loss of primary coolant system integrity.
(i.e. No fire outside the drywell could cause a loss of coolant accident and conversely no loss of coolant l accident could cause a fire in cable ways outside the drywell.1 The p fire is considered negligible.
l The protection system (scram system) was designed to give no failure or a safe failure for all credible events (but not necessarily for all conceivable events) and carried the single failure criteria to the point which included complete functional failure of a single electronic drawer.
J Great effort has been expended on this plant since.special emphasis has been placed on physical separation by the licensing reviews. The changes have been towards providing such incremental improvement in real safety in the pro-tection system and in the engineered safeguards systems' as could be realized by physical separation attainable, within the limits of practicability, on a i
retrofit basis.
The follo" ting is a list of some of the major changes that have been I
l made in the electrical equipment and installation in a sincere effort to respond' to and to alleviate the concerns of the licensing authorities:
l I
1.
Divided controls for ECCS systems to provide independent control power from the two sections of the station's batteries.
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2.
Placed fire resistant barrier between the two 4160V emergency switchgear sections, i
3.
Added a second emergency diesel generator system.
4.
Changed the core spray system to a split bus arrangement.
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Panels that could not be rebuilt. to meet the new channel separation l
criteria are:
1.
Neutron. Monitoring Panel l
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Core and Containment Cooling Manual Control Panel I
3.
The Protection Relay Panels l
4.
The' Isolation Valve Panel 5.
The Process Radiation Monitoring Panel
'i Building areas where arrangement could not be altered to the preferred '
criteria are:
1.
Dattery Room Arrangement 2.
Motor Control Center Arrangement l
3.
Main Control Room i
l
-i 4.2 Basis for limitations on separation:
I i'
l.
The development of a nuclear power plant design requires that certain major arrangement details and equipment p
j design concepts be est-blished three to four years before completion of the plant.
Examples of this are major electrical equipment arrangement, control _ room panel arrangement and design specifications.
Departure-from
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original concepts 'at any time after plant construction has started is difficult, if not practically impossible, to implement.
1 1
Any such concept change should be the result of a compelling technical need rather than a desire' to make incremental improvements.
j i
2.
Original plant design was based on the premise that fires in 4
~h' cable raceways concurrent with need for emergency equipment was t
not credible and need not be considered.
3.
Fires in cable raceways used for engineered ' safeguards have a j
very remote probability of occurrence for the following reasons:
f a.
Cables are sized with sufficient copper that heat losses per foot will not raise the insulation temp-1 crature above allowable design limits.
Design limit on the power cables is voltage drop instead of ampacity, j
considering lengths of the cables involved.
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liigh temperature insulation (Vulkene) is being used..
Vulkene has a normal working temperature of 90 C and -
a short time (days) rating of 125 C.
Therefore, insulation breakdown from gradual deterioration due
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a to operating close to design temperature limits is L
extremely unlikely, especially on the engineered safe-9.:
guards cables which do not see continuous duty.
4 c.
Cable trays loading.is such that. adequate heat dissipation is achieved and, in addition, many of the cables in the raceways that carry the valve motor and control cables are either nonna11y de-energized or carrying-very small currents.
(Evenstalled-rotor-currentforthevalve.
motors cannot heat the cables ' dangerously.)
l d.
Most of the valves in question are nonna11y open and/or do -
not have to move when core cooling is called for..Thus, t
a short of several selected ) airs of control wires accom-plished without concurrent s1orting of power wires would be required to disable both core cooling systems.
j e.
Zones of common exposure are limited and in locations where there is little, if any, probability of a fire from any non-electrical source.
i f.
There is no apparent mechanism whereby fire in the cableway could precipitate a loss of coolant accident.
f g.
There is no apparent mechanism whereby a loss of coolant accident could precipitate a fire that could destroy any of the common cabling outside the drywell.
y h.
Good wiring and cabling practices have been followed in the installation of the cables.
They are securely fastened l
to trays and have ade4uate support. Trays are securely anchored to the build,ing structure.
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4.
Compatibility of Electrical Equipment:
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The Oyster Creek power plant has been designed in accordance with conventional wiring practices and special attention has been given. to a
protection against ~ component failure by providing redundant active t
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components.
It is not believed possible that a fire of damaging-proportions could go undetected for very long and plant shutdown.is i
j always'possible regardless of the extent of any single postulated fire.
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Special Cases I
The following is a listing of items which after re-examination as a-r result of reviews of the Protection and Engineered Safeguards' Systems are considered to be acceptable exceptions to the general rules.
Technical. rationale l
for the decision to allow these cases to stand is given for each item.-
i 5.1 The position switches on the two main steam isolation valves inside i
the dryriell (NS03A & B) have their wiring sharing the same penetration. This could lead to the supposition that both steam lines could be shut off without i
3 directly causing a scram. This would ' result from the selective shorting of two pairs of wires without grounding, j
Resolution:
Although this particular scram function could be jeopardized, the condition postulated can be readily tested for, and reactor high pressure scram and/or neutron flux would provide backup protection for the reactor.
In addition, a scram would be assured in the event of an automatic isolation valve closure by virtue of the position switches on the valves NSO4A & B outside the drywell.
Based on the above rationale, no redesign action is considered necessary.
5.2 Certain redundant pairs of solenoid piloted air-diap'iram operated isolation valves have a comon teminal box and a common cable containing circuits to both valves.
Resolution:
Such valves are de-energized to close so damage to cables (shorts or grounds) will cause closure of the valves.
Shorting of two circuits in cable or between terminals within the box could require that both closure initiating circuits open on comand in order-to close the valves, but this is inherent.
in the control circuit because the coil circuits for two relays controlling the two valves are actuated by contacts on the same four logic relays connected in i
one-of-two twice logic.
The logic circuitry can.5e tested for shorts between cir-cuits. Therefore, no single failure is considered capable of holding botii valves j
open.
i Based on the above rationale, no redesign action is considered necessary.
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5.3-The emergency diesel generator manual start circuits that come_into the main control panels 8F/9F & 9 Fx. are in separate-trays but these trays are running one above the other for some distance in the cable room.,
Pesolution:
The lower tray is to be covered by, a solid steel cover of 14 gauge to reduce probability of fire in one tray from propagating to the other.
-5.4 Relays NK3 in the emergency switchgear which provides reset of the con-tainment spray automatic control circuits after power loss have contacts common to both A & B Control Systems.
Resolution:
The automatic reset function is only effective during return of system from test mode to normal and is, therefore, not essential to system automatic operation.
Shorting of relay NK3 contacts is tolerable.
Shorting of line side of contacts to ground would give a ground on the D-C bus or buses but would not knock out the control circuits. There is no mechanism for shorting out the containment spray control circuits so the operation of the containment spray system cannot be negated.
No redesign action is considered necessary.
5.5 The APRM channel bypass selector switch cables between Panel 4F and 3R and 4F and SR present a potential for multiple channel bypass if shorted.
Resol ution:
Such multiple bypassing is obvious to the operator by means of indicating lights on each channel bypassed.
The zone encompassed is essentially all within the control room where panel internal wiring separation has not been a design basis.
No gross separation could be accomplished without panel redesign and overall reliability would not be significantly improved by so doing.
Based on the above rationale, no redesign action is considered necessary.
5.6 The sensing lines for the drywell high pressure switches RE04, IP15, RV 46 are in close proximity for a short distance where they come through a common opening in the floor chroute to their separate taps and separate instru-ment racks.
Resolution:
Examination of the area in question shows the instrument lines to be well protected and no probable mechanism for damage during either nonnal or accident conditions.
Thus, a change is not considered to increase overall plant safety and, therefore, no redesign action is considered necessary.
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CO REPORT NO.
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DEC 161909 i
q J. P. O'Reilly, Chief, Reactor Inspection and Enforcement 1
Dranch, Division of Compliance, Hsadquarters INQUIRY MEMORANDUM
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.ILR3";7 CENTRAL PCMER & LIGHT. COMPANY, OYSTER CREEK 1, 219/60-B i
- $5 APPARENT FAULTY VENDOR STRESS RELIEF OF EMERGENCY CORE SPRAY
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SYSTEM VALVE RESULTING IN VALVE REJECTION - INADEQUACY IN VALVE PORCHASE. SPECIFICATIONS RELATING TO NONDESTRUCTIVE TESTING REQUIREMENT RESULTING IN FAILURE TO DETECT CONDITION AT VENDOR' i
PLANT l
As you are aware, several linear indications were visually _
l actccted on the body of an emergency core spray system isola-tion valve' (No. V-20-41) by J. Collins, CO HQ, during an inspection at the subject facility on December 2 and 3, 1968.*
It should be noted that this observation was incidental to the primary purpose of the visit.
The indications observed were-
~ 1/2 - 2" in length and were in the vicinity of the machined surfaces-for the yoke clamp.
During a subsequenP visit to the site on December 10 - 12,-'1968,
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the writer determined that'GE's follow-up investigation of this-condition as of that time included dye penetrant and radio graphic examination of the subject valve,-and visual examination of two other comparablo valves in the same system.
According to informed GE personnel'at the site, the dye penetrant check confirmed the linear indications noted earlier by Mr. Collins, and at least one of these indications was also detected radio-graphically.
No deficiencies had been noted on the visually enoched valves.
The following additional information was provided to CO in a l
phone call from Mr. L. Loeb, Manager, Materials and Quality. _
Sorvices, GE Domestic Turnkey Projects, to the Region I office; on December 13, 1968:
1
- Renults of inspection to be discussed in Inspection Report No.
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1 The defective valve had bee l
valve company in accordance,n fabricated by the Anchor j
with purchase specifications
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by Burns & Roe, the archithct-engineer for this project.
The valve casting was mad for Anchor by Roma' Company.
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Doth firma are located in California.
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~E i The purchase specifications included.a requirement for radiographic examination of the valve casting; however, 4' [;!
j they did not clearly state whether the_ radiography should j
be performed before or after the final stress relief opera-tion ~
The formor was actually the case in this instance.
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A re-examination"of the radiographic records for the subject valve by GE failed to reveal any of the indications discussed above.
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GE has concluded that the defects probably relate to the
.. final stress relief operation.
The subject valve will be replaced before initial foal loading.
4.
The situation regarding the performance of radiographic examination of valve castings before completion of final I
stress relief applies to all the valves supplied by Anchor for this project.
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An inspection team comprised of representatives of GE, Burns & Roe and Anchor is being formed to look at these valves.
This will include witnessing dye penetrant examinations of 4 to 6 valves.
Based on an evaluation of thcoo results, and assuming no further indications of defects, the team will visually examine all of the Anchor valves located inside the dry well, and up to and including the second isolation valve - about 20 valves total.
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GE acknowledges that this investigation may lead to dye penet{ ant checking of additional valves.
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GE is checking the specifications for all valves on the other Domestichturnkey Projects in this regard.
Thus far it has been determined that those currently made for Monticello were radiographed following final stress relief.
The valves for the Millstone Point 1 and Consnonwealth Edison Company projects, reported to have been supplied -
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radiographed following final stress relief.
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Region I will continue to follow closely the situation at the Oyster Creek facility.
You will be kept informed of any 3,I significant developments in this regard.
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.4 jD R. T. Carlson, i
Senior Reactor Inspector C
li i-cc E. G. Case, DRS R. S. Doyd, DRL (2) 4 S. Levino, DRL D. J. Skovholt, DRL (2) l 1
L.
Kornblith, Jr., CO REG files CO II I k CO III i
CO: IV CO:V i
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R. S. Boyd, Asrictant Director for Reactor Project
'>i Division of Renctor Licensing (2) 2 JVHSEY CENTRAL POWER AND LIGHT COMPANY (OYSTER CREEK)
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DOCKET NO. 50-P19 P
The enclosed summary report, which compiles the results of our in-npcetion effort related to the Oyster Creek pressure vessel repair
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program, is forwarded for information. The significant results were discussed prompt.ly with concerned regulatory personnel as informa-
- u tion vns developed and was sumnarized in periodic status reports.
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[5 As a result of this inspection effort, Compliance concluded that the detected defects were repaired in accordance with the an{ ended appli-5."
cation and thnt the repairs met applicable code requirements.
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_k R. H. Engelken, Assistant Director
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for Inspection and Enforcement P
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Division of Compliance
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Enclosure:
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Co Rnt No. p19/68-9 dtd 1/6/69 it f
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- 11. 11. Marn, DR P. A. Morrin, DEL E. G. Caro, DRS l
O. Levine, DRL (6) hz..)
D. J. Gkovholt, DRL (3)
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L. Kornblith, Jr., CO fi R. W. Kirknan, CO:I
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J. C. Davis, CO:II l l B. H. Grier, CO:III Q"
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D. I. Walker, CO:IV N
R. W. Smith, CO:V O,9 J
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