IR 05000302/1976022
| ML19308D301 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 12/17/1976 |
| From: | Jape F, Robert Lewis, Rogers R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19308D290 | List: |
| References | |
| 50-302-76-22, NUDOCS 8002270746 | |
| Download: ML19308D301 (24) | |
Text
- _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ - - _ _ _ _ _ _ _ _ _
.
. I.
_.
__ _
.
>R 88 00g4 UNITED STATES
\\
,y jo, NUCt. EAR REGULATORY COMMISSION..
,
,
,
,
)'
3-o REGION li
.$
230 PEACHTREE STREET, N.W. SUITE 818
ATLANTA, GEORGIA 30303 o%, "' W4l
~
+....
IE Inspection Report No. 50-302/76-22 Licensee:
Florida Power Corporation 3201 34th Street, South
,
P. O. Box 14042
~
St. Petersburg, Florida 33733 Facility Name: Crystal River 3 Docket No.:
50-302 Licence No.:
CPPR-51 Category:
B1 Location:
Crystal River, Florida Type of License:
B&W, PWR, 2452, Mwt Type of Inspection: Routine, Announced (November 3-6, 1976)
Routine, Unannounced (November 10-12 and 15-19, 1976)
Dates of Inspection: November 3-6, 10-12, and 15-19, 1976 Dates of Previous Inspection:
October 31, November 2, 1976 Principal Inspector:
F. Jape, Reactor Inspector Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch j
(November 15-19, 1976)
i
!
Accompanying Inspectors:
R. F. Rogers, Reactor Inspector
'
Reactor Projects Section No.2
,
Reactor Operations and Nuclear Support Branch i
(November 3-6, 1976)
'
-
,
,
S. D. Ebneter, Reactor Inspector
.
!
Engineering Support Section No. 2 l
Reactor Construction and Engineering Support Branch (November 10-12, 1976)
=
G. L. Troup, Radiation Specialist Radiation Support Section Fuel Facility and Materials Safety Branch (November 9-10, 1976)
Other Accompanying Personnel:
None v
l
.
!
8002270 2 %
_
l
,
.-
-
-
- - -
_- -
.
~
_
_
~
~1
-
.. -..
_.
. - ~. -. -.
.
..
)
'
.
IE Rpt. No. 50-302/76-22-2-
.
/
IM Principal Inspector:
W F. J(pe, Reactor Inspect'or Date Reactor Projects Section No. 2 Reactor Operations and Nuclear Sdpport Branch
.b. /p M
/2//7/7b
!
Reviewed by:
Date R. C. Lewis,' Chief Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Br'anch j
.
.
{
.
.
i
.
l
,
/
-
.
.
'
.
'
.
.. _. _.. -..
.
.,.
.. - - -
_ - -
.. - _ _,,
.
-...
.
_ _ _ _
_
_,
_, _ _ _ _ _ - _
_
-
-
. _ _ _..
.
.
.
IE Rpt. No. 50-302/76-22-3-
,
SUMMARY OF FINDINGS I.
Enforcement Matters
~
Infraction Failure to Evaluate Examination Data Contrary to 10 CFR 50.55a and Article 1S-300 of the ASME Code,Section XI, Winter 1972, the licensee failed to evaluate an indication in the baseline data that exceed acceptable standards of ASME Section III when performing the preservice inspection.
'
(Details III, paragraph 3)
II.
Licensee Action on Previously Identified Enforcement Matters All previously identified enforcement matters have been closed.
III. Nov Unresolved Items None IV.
Status of Previously Identified Unresolved Items 75-8/2 Safeguards Systems Pump Runout Testing of flow control method to control pump runout has been completed. Item is closed.
(Details IV, paragraph 8)
75-19/4 Testing of Radioactive Waste Sample Lines A draf t procedure for the testing of the lines has been prepared but testing.has not been accomplished. This item remains open.
(Details II, paragraph 2)
75-19/6 Evaluation of Sampling Media Collection Efficiencies The licenseehas initiated modification of 'certain air samplers to improve the collection efficiency for
'
,
radioiodines.
The overall evaluation of the collection efficiencies is incomplete. Tuis item remains open.
(Details -II, paragraph 3)
.
.
$
-
r-.
-
,
- -
_. - _ - _ _ _,,,.
,
_,,__
_
_
,
_,
__
_ _ _. _ _ _--.
_
_
_
-___
.
-
-
..
.
.
.
.
.
.
.
IE Rpt. No. 50-302/76-22 4-
-
.
76-16/1 Annunciator Alarm Procedure (AP-102)
Annunciator alarm procedure has been completely rewrit-ten.
This new procedure was found to be satisfactory.
Item is closed.
(Details IV, paragraph 4)
76-20/1 Conduct of Preoperational_ Radiochemistry Testing The licensee has completed the review of the test procedure and has corrected the deficiencies identified by the inspector.
Iten is closed.
(Details II, paragraph 4)
V.
Unusual Occurrences
.
None
!
VI.
Other Significant Findings a.
Followup on IE Circular 76-05 Licensee's response, dated November 5,1976, reported that ITT Grinnell snubbers are not used at Crystal River 3.
No further effort is planned for this circular.
B.
Containment Dome Delamination Repairs Repairs have been completed and the reactor building structural integrity and integrated leak rate tests have been successfully performed.
.
VII. Management Interview
The inspection was conducted in' four parts and a management inter-view was held at the conclusion of each part.
The first was on November 6, the second on November 10, the third on November 12, and the fourth on November 19, 1976. Each management interview was attended by J. Alberdi or members of his staff.
- The infraction concerning failure to evaluate an indication in the baseline data was discussed. Other inspection findings were also discussed.
s a
.
e e
-- -
y w.---e--
.aawe--T-
-e---
- ---
.mir
---c-e- - -
--u-r
-- - - - --- -,
w e e si, e--
e v%
y ts
- + * -
--
e
-
.
-
.
.
IE Rpt. No. 50-302/76-22 I-1
_
/' IY/7d8 DETAILS I Prepared by:
.
R. F. Rogers, skactor Inspector
'Date Reactor Projects Section No. 2
,
Reactor Operations and Nucle.ar Support Branch Dates of Inspection: November 3-6, 1976 Reviewed by: [.C e,,-b
/2 J/76 R. C. Ledis, Chief Date Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch 1.
Personnel Contacted G. P. Beatty, Jr. - Nuclear Plant Superintendent J. C. Hobbs, Jr. - Manager Generation Testing R. S. Dorrie - Quality Engineer D. B. Black - FPC Test Supervisor R. E. Shirk - GAI Test Engineer T. C. Lutz - GAI Test Engineer D. J. Mitchell - GAI Instrumentation Engineer R. F. Ely, Jr. - GAI Data Analyst
.
2.
Containment Integrated Leak Rate Test - TP 71-150-02 The inspector witnessed portions of the containment integrated l'eak rate test (CILRT) over a four day period. The initial pressurization, the stabalization period at peak pressure, the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> drop test, and the controlled leakage test were observed. Test. procedure results and test personnel actions were evaluated against the requirements of 10 CFR 50 Appendix J, " Primary Reactor Containment Leakage Testing for Water-Cooled' Power Reactors," ANSI N45.4 -
1972Property "ANSI code" (as page type) with input value "ANSI N45.4 -</br></br>1972" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., " Leakage-Rate Testing of Containment Structures for Nuclear Reactors," and FSAR Article 5.6.3, " Initial Integrated Leak Rate Tests." The test was performed at a containment pressure of approximately 49.7 psig. The leakage rate was calculated by three different methods and the results compared favorably.
The leakage
,
rate limit was.187 wt %/ day (.75 La). The actual results and method used follow:
l
/
-
.
.
e a
w
.
-
.
4
.
.
.
.
IE Rpt. No. 50-302/76-22 I-2
-
24 Hour Method Leakage Rate Total Mass
.057 wt%/ day.
Total Time
.063 wt%/ day Point to Point
.068 wt%/ day The controlled leakage test demonstrated measurement agreement within the 25% acceptance criteria. The inspector verified select-ed system valve lineups and actual test results through independent inspection and calculation. hta collection techniques and the performance of the test personnel was satisfactory. Acceptance
~
criteria for the test were satisfied.
i i
.
.
?
!
.
,
I s
.
/
(
-
.
i
.
,
.
.
'
e
. -
.....
...
- - - - - -
--
- -
-
y
.._, _
w m
--4-.,
,
,
e.,
,,
-
,.-ms w
w yr
.m._--
_ _.
.. _
_
_
_ _ _ _. - - - _
._
_
_
.
-
,
%
.
v IE Rpt. No. 50-302/76-22 II-1 l
l DETAILS II Prepared by:
7(,
G. L. Troup', Radidion Specialist
'Date Radiation Support Section Fuel Facility and Materials Safety Branch Dates of Inspect N ember 9-10, 1976 ll[7T !M Reviewed by:
//
e'
i N
A. F. Gibson, Chief Date Radiation Support Section Fuel Facility and Materials Safety Branch 1.
Individuals Contacted J. Alberdi - Project Manager G. P. Beatty, Jr. - Nuclear Plant Superintendent P. F. McKee - Assistant Nuclear Plant Superintendent J. L. Harrison - Assistant Chemical and Radiation Protection Engineer G. D. Perkins - Health Physics Supervisor D. W. Pedrick, IV - Compliance Engineer P. E. Griffith - Training Coordinator W. A. Cross - Plant Engineer R. E. Fuller - Plant Engineer E. E. Froats - Manager, Site Surveillance
<
R. S. Dorrie - Quality Engineer J. C. Hobbs, Jr. - Manager, Generation Testing G. H. Ruszala - Test Engineer 2.
Testing of Radioactive Waste Sample Lines (75-19/4)
This item was originally disc'ssed in IE Report No. 50-302/75-19, u
'
Details III, paragraph 2 and dealt with the need to test waste sample lines to verify representative sampling and assess the amount of plate out in sample piping, as required by Regulatory Guide 1.68.
The inspector discussed this item with licensee representatives regarding status, schedule and test methods. A
-
,
licensee management representative informed the inspector that a test procedure had been drafted but was not yet approved. A licensee representative' stated that the actual testing would be performed by a contractor in accordance with the test procedure but had not yet been accomplished. This item remains open.
o
.
.
4 m,e6M*
y-y
_
,,-
-
_
-- -
e--
y
_ _ _ _ - _ _ _ _ _ _
-
.
,
-s
'
IE Rpt. No. 50-302/76-22 II-2
.
3.
Evaluation of Sampling Media Collection Efficiencies (75-19/6)
This item was originally discussed in IE Report 50-302/75-19, Details III, paragraph 4 and dealt with the need to establish the collection efficiencies for sample collection media, especially charcoal cartridges. In IE Report No. 50-302/76-20, Details II, 4.b it was noted that an engineering change notice had been issued to modify the air samplers to increase the residence time to an acceptable value. A licensee representative informed the inspector that the modification would be accomplished on the four samplers at environmental release points but the process monitors will not be modified at this time. Once the monicors are modified, the adsorber efficiency will be selected based on the flow rate and media; the efficiency will then be factored into the appropriate chemistry procedures. This item remains open pending the final selection of the adsorber efficiency.
4.
Conduct of Preoperational Radiochemistry Testing (76-20/1)
a.
This item was originally discussed in IE Report 50-302/76-20, Details II, paragraph 2 and dealt with incomplete records for test procedure 7 2 500 3, Initial Radiochemistry Test, relating to the sign-offs for prerequisites and sample collectisn for test phases which have been completed.
In some cases, samples had been collected but the procedure was not signed, but in others, the samples were not collected and no deviations were recorded in the procedure.
b.
The inspector reviewed the official record copy of the procedure and verified that procedure had been signed off for samples previously collected and deviations noted for samples which had not be collected. A licensee representative advised the inspector that a review had been made of the sampling and counting records for the,r<ent fuel pool and it was determined that the samples had been collected and some analyses performed, as required. These were signed off in the procedure. When samples were not collected due to equipment or system problems, a deviation was entered and signed.
i c.
A licensee management representative informed the inspector
'
'
that the test procedure will be periodically reviewed by a Chem Rad supervisor during the power ascension test program to verify that the required samples are being collected and the test procedure is being completed as required. The inspector
,
informed licensee mangement that he had no further questions and that this item was closed.
-
e S
e
m-e e
ew e-
- 6 b
.
_
.
_
- - -.
.
.
.
-.
.
.
.
IE Rpt. No. 50-302/76
-
II-3
>
.
i
~
5.
Control of High Radiation-Areas a.
10 CFR 20, Section 20.203.(c)(2) specifies the requirements for-the control of the entrance or access points to a high radiation area. Section 20.203. Cc)(3) requires that such
controls "shall be established in such a way that no indivi-
dual will be prevented from leaving a high radiation area."
l b.
The inspector observed the barriers being installed at the
!
entrances to potential high radiation areas and noted that some areas (such as the hold-up tank room) were provided with gates secured with padlocks on the outside.
Such locks, if in place, would prevent on individual in the area from leaving
.
the area. This situation was discussed with licensee manage-
'
ment, who stated that this condition would be reviewed and action taken to comply with 10 CFR 20.
- 6.
Radiation and Respiratory Protection Training a.
FSAR Section 12.2.2.4 states, in part, "All personnel assigned to the nuclear plant will receive specialized training in radiation saferf..." As discussed in IE Report No. 50-302/76-17, Deca 11s I, paragraph 6, during the inspector's review of training records, the test indicating satisfactory completion for one individual could not be located. A licensee
representative informed the inspector that during a review of
'
the training records, several instances were found where the test could not be located although other records indicated satisfactory completion of the training; in these cases the j
individuals attended the training again and took the test.
The licensee representative also discussed changes which were being made in the storage of training records. The inspector had no further questions.
i b.
Plant Radiological Control Procedure RP-102, Respiratory Equipment Manual, Section 4.0 specifies the training require-ments for plant personnel who are assigned to the respiratory protection program. The inspector reviewed the training and equipment fitting records for the training done so far and discussed the training program with licensee representatives.
,
The inspector provided comments relative to the records to licensee management who acknowledged them. The inspector had
no further questions.
l
i
.
.
.
.. _..
.
..
.
...,-.
..
, -
--
- -
-
..
-..
_.
.
.
-- L-..
- - - -. - -
.. - -.... -
- -
.. - - -, -
.
.
,
.
-s
,
.
,
.
,
IE Rpt. No. 50-302/76-16 II-4-
.
7.
Charcoal Cannister Test Results a.
This item was originally discussed in IE Report No. 50-302/76-
'
17, Details I, paragraph 3.c and dealt with possible degradation in the efficiency of charcoal in the ventilation systems adsorbers. During performance testing of the adsorbers, the testing was delayed due to high background readings which may have been the result of organic cleaning compounds.
Charcoal test cannisters were removed from the filter housing in three systems and sent off-site for laboratory analysis to determine any degradation in the charcoal efficiency.
b.
A licensee representative showed the inspector copies of the laboratory results performed on the test cannister charcoal specimens. The laboratory test results showed each cannister had a removal efficiency for methyl iodide greater than 99.7%
under the test conditions of Regulatory Guide 1.52, Rev. 1.
The licensee ecpresentative also informed the inspector that the in-place testing of the adsorbers had been satisfactorily completed af ter purging to lower the background. After review-(g ing the laboratory test reports the inspector had no further questions on this matter.
8.
Confirmatory Measurements a.
Scope The licensee is required to measure the quantities. and concen-trations of radioactive material in effluents from his facility to assure that they are within the limits specified in his license and the NRC Regulations. The inspection consisted of testing the licensee's measurements of radioactivity in prepared test standards by comparing his measurements with those of the NRC's reference laboratory. The result from the licensee is compared to its corresponding NRC value to determine if the result fell within established ranges for agreement or possible agreement. These ranges vary with resolution, which is defined i
as the ratio of the NRC result to the NRC statistical uncer-
'
tainty. As resolution increases, the ratio of the licensee result to the NRC result must more closely approach unity.
,
The measurements made by the NRC's laboratory are referenced to the National Bureau of Standard-radioactivity measurements system by' laboratory intercomparisuus.
j
-
,
.
$
.
_e-w---
-
,n_ _ _ _
y
.p_,
e-r-
- -
y-wy-----w
--,vvwyeg-3-*i,-vyvw-,-+waw-yw----:
-
V-v----
.
.
.m
-
'
-
.
'IE Rpt. No. 50-302/76-16 II-5
'
.
b.
Comparison of Analytical Results The results reported by the licensee were in agreement with the NRC results except for one analysis, which was in possible agreement. Based on these results, no further action is required at this time.
9.
Status of Previously Identified Items i
The followip; facilities, systems and founctional areas were reviewed by the inspector. These items were previously discussed in IE reports as incomplete.
a.
Administrative instructions for the conduct of chemistry and
radiation protection - a revision to AI-700 has been reviewed and approved; the inspector has no further questions.
j l
l b.
Discharges from the laundry sump - all review has been completed j
of the sources of water into the sump but no procedures have been issued to establish adminstrative controls over the sump during discharges to meet the batch release criterion.
s
\\s_s c.
Waste solidification - procedures for the solidification of waste with no free liquid have not been prepared.
d.
Decontamination room - the decontimination room has not been completed nor have temporary decontamination facilities been
'
provided.
e.
Make-up System letdown line - the shielding has not been installed.
f.
Chemical fume hoods performance - contractor personnel will be working on the fans and on the system balance to c.orrect this problem but it is incomplete.
.
/
-
.
e
.
+...... _.... -.
,..... ~... - -.....
.... -. ~.,. **. - -. 9 - -. -
~
'
'
- - -
,.
-.-
.
. -. -,
.
.-.-
-
,
,.
___
.
.. ~,
Q.
-
\\
IE Rpt. No. 50-302/76-22 -
III-1
_
s'/j ('
)/
.4-
/4-
/d!8 /7/-
DETAILS III Prepared by: 7'/-
'<
,
S. D. Ebneter, Reactor Inspector Date Engineering Support Section No. 2 Reactor Construction and Engineering Support Branch Dates of Inspect n November 10-12, 1976 d
Reviewed by:
~
A. R. Herdt, Chief Date Engineering Support Section No. 2 Reactor Construction and Engineering Support Branch 1.
Persons Contacted Florida Power Corporation (FPC)
J. Alberdi - Project Manager g
G. Beatty - Nuclear Plant Superintendent j
S. Johnson - Plant Engineer D. Ruzic - Results Engineer 2.
Scope This inspection consisted of audits of the preservice inspection (PSI) final report, reactor internals inspection and decay heat valve modifications. Primary emphasis was on the PSI report and the evaluation of code rejectable indications.
3.
Preservice Inspection Data and Final Report Preservice inspection of the Crystal River 3 reactor coolant system and associated components was performed by Southwest Research Institute (SwRI), Babcock and Wilcox Construction Company (B&W),
and Zetec, Inc. Additional data related to resolution of indica-
,
tions observed in Class 2 piping and visual inspection of reactor internals has been developed by FPC personnel and will be incorporated or referenced in the final report compilation.
,
The bulk of the PSI data is contained in the ten volume report prepared and approved by B&W. The report consists of the following volumes:
O
- meme W- - d
-.
A
%
'sp)'
~
-
s_,
IE Rpt. No. 50-302/76-22 III-2
.
.
Volume Title
Summary Report
.
Preoperational Inspection Manual
Reactor Vessel, Pressurizer, and Steam Generator Data
Class 1 Piping
Class 1 Piping
6 Class 1 Piping
Class 1 Pumps and Valves, Class 2 Vessels and Piping
Class 2 Piping, Pumps and Valves
Reference Documentation N
,
Ultrasonic Calibration Sheets These are supplemented by the Zetec, Inc. report, and revisions thereto; Eddy Current Inspection of Crystal River #3 Steam Gen-erator.*ubing; SwRI report and data for CRDM tube to head welds and pump visuals; and evaluations of Class 2 indications performed by FPC.
The inspector selected several weld areas for audit to determine compliance with ASNE Code and regulatory requirements. The data and records for Circle Seam MK Al ta MK A2 were audited. The examinationdatafortheweld.wascontainedinVogume2,gection4, Figure 1.1.4.
The UT examinations consisted of 0 and 45 scans with calibration performed using calibration block 40704. In reviewing the records it was noted that this calibra. tion block and three others listed in Volume 2 were not cross referenced to drawings. The licensee has detected this and transmitted the information to B&W for inclusion in Revision 1.
Calibration sheets
,
111015,111018,111019 and 111020 detailed parameters of the examination and included dates, times, and equipment serial numbers.
/
I J
-
.
G
&
G
-
-
,
._
... -..
,... _...,
.
-.
m
.
.
._
.
~
.
<
-,
IE Rp t. No. 50-302/76-22
~
III-3
,
Cross checking of times and dates of data and calibration sheets did not reveal any areas of concern. Personnel certifications were contained in Volume 9, Section 2 and all appeared to be in conform-ance with SNT-TC-1A and B&W procedures. B&W conducted a reexamina-tion of indications that exceeded 100 percent DAC.
The examination data for the pipe-to-elbow safe end NK49 to MK53 shown on drawing 141586E-5 was audited. The PT data is recorded en Figure 4.1.77 which cross referenced materials and personnel.
Certifications attesting to halogens and sulfur content less than 1% for developer 4LO21, Penetrant 4K029 and cleaner SD091 were in Volume 9 Section 3 of the report. Personnel performing the test vers certified to level II per SNT-TC-1A as evidenced by records in Volume 9 Section 2.
Ultrasonicexaminationdatawasrecorgedon Figure 4.1.j8andreferencedgalibrationsheets 1200131 (0 ),
1800124 (45 ) and 1800130 (60 ) which were in Volume 10.
Calibra-tion dates, dates of examination, recalibration times, personnel certifications, couplant certifications, and equipment calibration records were available in Volume 9.
A typographical error in-
!
volving an apparent transposition of digits occurred in Volume 9, (
Section 3 where the batch number for Item 3.2 in the index was
,
specified as 835 and the actual certification shows it to be 853.
This did not affect the ability to trace the certifiestion and/or verify materials on the data sheets.
The disposition of Code rejectable indications was discussed in Volume 1 and Volume 9 of the report. These were reviewed to assure compliance with code and to verify disposition of indications referenced in previous IE reports. The disposition of these appear-ed to be in conformance with code requirements.
One reportable indication discussed in Volume 9 was investigated further to determine if the recommended disposition had been accomplished. Figure 4.9.14 contained data which exceeded code acceptance criteria. The recommended disposition of this was detailed in a letter dated August 17, 1976, and signed by the B&W Level III examiner. The Level III evaluation was that the indica-tions represented an area of probable lack of penetration of a lug-to-pipe weld. It was recommended that the design be checked to
'
determine if a repair was necessary. FPC took action on this as documented on work request 7-01002 dated August 24, 1976, repaired it per W-97-1627, 1626 and 1680, and reinspected it on September 16, 1976.
'
1-
,
e
9
- - -.,,
,-
e
-
, - -
, - - - -
r
-
-ep-sem - -- --
_
_
-
_
.
.
.
.
-
.
I s_,)
IE Rpt. No. 50-302/76-22 III-4
.
'
Seven additional' indications were evaluated by the Level III examiner utilizing examination data, reexamination data and data from complementary NDE examinations. However, the inspector in i
reviewing data discovered some indications which apparently exceed code acceptance criteria which had not been adequately evaluated as listed below:
o a.
Figure 4.4.995.2 for a 0.61" thick material records a 360 re-
'
flector at 125% DAC and an indication 0.5 inches long from 100% to 100% DAC. There is a graphical sketch attached which was signed by a Level II examiner. The sketch indicates the
'
reflection is from the root area. There are no notes with regard to damping, no counter bore is shown and an evaluation
!
by a Level III examiner apparently was not performed.
b.
Figure 4.4.790 data lists two arc strikes and references only the diameter. There is no other assessment of these or
'
definition of depth of metal removal.
c.
Figure 4.4.951 references an indication of 100% DAC for 360 with no further evaluation.
d.
Figure 4.4.997 lists an indication in 0.52 inch thick material i
with a maximum amplitude of 125% DAC with 100% to 100% of one
'
inch.
e.
The data recorded on Figure 4.4.775 contains several reflectors which exceed 100% DAC. The material is 0.28 inches thick and a reflector is listed as being 0.45 inches from 100% to 100%
DAC. There is a graphical sketch attached and a note on the i
data sheet indicates one is dampable. An evaluation by a Level II examiner is shown on the sketch but no evaluation by a Level III examiner.
The licensee is committed to 'he ASME Section XI, 1971 issue t
through Winter of 1972 addenda.
ASME Code Section II,1971 issue, Article 221 states in part that i
all nondestructive examination results shall be evaluated by
'
qualified NDE personnel. Article IS-311 states that evaluation
'
shall be made of any indications which exceed acceptance standards
! -
specified in ASME Section III. For the welds in questions, ASME Section III, Article IX-3470 requires that all indications that
'
produce a response greater than 20% of the reference level shall be investigated to the extent that the operator can determine the shape, identity and location of all.such reflectors and evaluate-them in terms of the acceptance standards of NB-5330.
,
i
-
.
-
,
.
.
-
- --
~n
-
.......
.. _.. - -. ~...
..
.
. -. -...---
. -..
. -
-
~
.
.
.
.
.
_ _, _ _ -
-
,.
-, _. _
.
.. _.... ~
_
-
_
_ _
. - _.
.
.
.
.
,
s.
IE Rpt. No. 50-302/76-22 III-5
.
.
The Summer 1971 Addenda to Section XI modifies Article IS-311 to-read that evaluation shall be made of any indications detected i
during any of the examinations including the preoperational ex-aminations which exceed acceptance standards specified in Sec-
,
!
tion III, 1971 Edition. This is further modified by Errata to j-Summer 1971 Addenda, issued in Winter 1971 Addenda to Section XI,
to clarify that evaluation shall be to the Section III edition applicable to ths crastruction of the component.
Article IS-312.2 is modified by Summer 1971 Addenda to Section II
}-
to state that all indications which produce a response greater than l
100% of the reference level shall be investigated and evaluated in l
accordance with IS-311. Article NB-5330 of Section III states that j-discontinuities are unacceptable if the amplitude exceeds the i
reference level and have lengths which exceed:
1/4 inch for e up to 3/4 inch j
1/3 t for t from 3/4 inch to 2 1/4 inches 3/4 inch for t over 2 1/4 inch
!
("t" is the thicknesc of the weld being examined)
The indications of Figures 4.4.775, 4.4.995.2, and 4.4.997 exceed the above acceptance criteria. The report does not reference any l
evaluation of the reflectors in some cases, and in others only contains a geometric sketch with no text explaining the final-
'
determination and disposition. This is an apparent noncompliance
to 10 CFR 50.55a, item g, and represents inadequate evaluation of
'
data per ASME Sections XI and III, as referenced therein. This is
,,
identified as noncompliance 76-22-Al(II).
l The licensee contacted B&W personnel by telephone and discussed j
some of the data. No reason for the apparent lack of evaluation could be given.
i
!
.
The evaluation of Class 2 systems indications was performed by FPC
,
personnel. The results have been transmitted to B&W for inclusion
in the Revision 1 to the final PSI report. As a result of the
,
evaluations, FPC has committed to perform a limited number of ultrasonic reexaminations on a periodic basis to attempt to ascer-tain if changes in the system's integrity can be distinguished.
The inspector requested that the licensee specifically identify which welds would be reexamined by UT and for how many outages.
z
'
The licensee agreed to incorporate specifics in the final PSI report. With regard to the remainder of the Class 2 components, RT
,
!
will be utilized as the volumetric technique.
-
.
t
.
.
.
.
_ _
.
.....
.
.
_
..
--
-
-
-
f e --
--m
.m
.-
e
,
,-,-a ea-ea,-
w
-,-m-e w,ey>.--ee,,---,w---e-anr--w-ee--s-,
w p e-w w, m.m,
-
-w
,e
-, ---
J
-
-
.
.
,
'
.~
'
'~N
'
.
i IE Rpt. No. 50-302/76-22 III-6
.
Other data related co the PSI have also been fowarded to B&W for inclusion in Revision 1.
Includ-d in this is the Revised Zetec report on steam generator tube data, the SwRI data, and calibration block data. The Zetec report was reviewed for conformance to commitments and RG 1.83.
FPC personnel had reviewed the initial report and requested several revisions which were incorporated in l
the final report.
4.
Reactor Internals Inspections
The licensee performed visual inspections of the reactor internals before and af ter hot functionals to determine effects of flows and other stresses.on the components.
Surveillance Procedure SP-298 was used to conduct the inspection after the hot functional and Procedure SP-306 was used prior to hot functionals. The procedures provided guida.nce and check.*ists for visual inspections of the core support assembly, plenum as: embly, vessel, closure head and nozzle sealt. The initial inspectfan was conducted in September 1975 and the post-hot functional inspection in September / October 1976. The latter inspection revealed:
a.
no flow induced flaws; b.
some fabrication flaws; and c.
some slight damage due to handling.
Section XI of the ASME Code, Item 1.15 of Table IS-261 requires a visual inspection of interior surfaces and internals and integrally welded internal supports. These inspections were conducted in September 1975 per B&W procedure ISI-350 and the results recorded as " clear" on Figures 1.15.1 and 1.15.2 of the preoperational inspection report. The licensee will cross reference the pre-operational inspection final report to reflect those indications that are not presently recorded in the PSI report.
.
5.
Decay Heat Valves DHV-110 and DHV-lli
'
Decar *.leat valves DHV-14 and DHV-25 are being replaced by DHV-110
,
and DRV-111, respectively. The valve replacement is necessary to provide a throttling capability on the discharge side of the decay heat pumps.
\\
l
.
.
W
,
_... _ _
-
!
,,
_.
__
_
__
-
.
_
_ _.
...
_
_
_.
..
_ _
O
-
,.
.,
.
I IE Rpt. No. 50-302/76-22
' III-7
.
The inspector observed valve-to-pipe welding and actuator installa-tion on the two decay heat lines. The valve locations and nomen-
'
clatures were verified per FD-302-641. The inspector noted that the welds are subject to in-service inspection and the replacement invalidates the preservice inspection data. The licensee stated they would examine the replacement welds and incorporate the data in the baseline. On November 29, 1976, the 'icensee informed the inspector that B&W personnel had visited the site and ultrasonically examined the applicable welds.
.
,
i.
I
!
,
t
i O
s
!
!
-
.
e G
,
,,
.,,,,
.,
.
.
- -.. - =..
. = = = = -. _..
.-
_
.., _.,... _ -.
_ _ _,.... _ _ _ -. _ _
.m,
. _ _. _. _,,,.
. _. -.., _, - -.
-
-
..
,w
IE Rpt. No. 50-302/76-22
-
IV-1
.
DETAILS IV Prepared by:
424L
/*
F. Jape, Reactor Inf(pector Date Reactor Projects Section No. 2
.
Reactor Operations and Nuclear Support Branch Dates of Inspection: November 15-19, 1976 Reviewed by: E.6,
/2.[/3M R. C. Lewis, Chief Date Reactor Projects Section'No. 2 Reactor Operations and Nuclear
,
Support Branch 1.
Individuals contacted Florida Power Corporation (FPC)
J. Alberdi - Project Manager G. P. Beatty, Jr. - Nuclear Plant Superintendent G. J. Walker - Manager, Field Testing J. C. Hobbs, Jr. - Manager Generation Testing J. E. Barrett - Plant Engineer D. E. Olson - Quality Auditor B. L. Chastain - Data Collection Supervisor B. Black - Test Engineer R. R. Henderson - Chief Nuclear Operator K. O. Vogel - Computer and Controls Engineer j
D. A. Morrison - Nuclear Operator M. P. Holmes - Assistant Nuclear Operator J. F. Heilman - Assistant Nuclear Operator T. N. Mount - Nuclear Operator H. M. Embach - Nuclear Shif t Supervisor Gilbert Associats Inc. (CAI)
J. P. Babb - Engineer
,
Babcock and Wilcox Company (B&W)
J. M. Putnam - Technical Support Engineer i
O
.
.
.
.
...
r-
,,
, -
~
v v
iw-m--
--.
.
.
.
-A f).
(,)
IE Rpt. No. 50-302/76-22 IV-2
_
_
2.
Plant Operational Status Followup on the operational status of the plant was continued during this inspection.
The master punch list, maintained by the licensee, was examined to determine the status.of preoperational testing and construction activity.
In summary, a list of eight items was compiled and identified as restraints on initial criti-cality.
There were no items identified as a restraint on initial fuel loading.
Followup on the initial criticality restraints will be conducted on
subsequent inspections.
3.
Hydraulic Snubbcr Sight Class Leakage Followup on the replacement program of leaky sight glasses on hydraulic snubbers revealed that approximately 45% of the snubbers in the reactor building have been repaired, tested and re-installed.
All snubbers have been removed from the reactor building for repair of the sight glass and testing prior to reinstallation. The licensee estimates this program will be completed by December 15, 1976.
V The repair and test work are being conducted as prescribed in HP-116, "PPC Hydraulic Snubber Installation Inspection," and SP-201,
"In-Service Inspection of Hydraulic Snubber.
These procedures were reviewed and the work-in-progress was witnessed by the inspector.
There were no questions or comments concerning this activity. The inspector examined a random sample of the repaired and re-installed snubbers. None of the snubbers had evidence of oil leakage. A spot check of tcat data revealed that the tested snubbers are within the established test criteria. Followup on this item will be conducted during future inspections.
4.
Annunciator Alarm Procedures
.
.
.
During inspection 50-302/76-16, unresolved item 76-16/1 was identi-fied due to differences in setpoints, probable causes, symptoms, automatic actions and operator action between the same annunciator alarm on A and B panels. In a response on this item,' dated October 18, 1976, licensee management described the problem and corrective
,
actions. Followup, by the inspector, revealed that the alarm
-
procedures have been completely rewritten and the problem has been corrected.
.
.
<-
.
.
..
,,
4 -=
.m....
..
.
.-.e
_
-
.-_
._____. __. _ _..__.__ -_
-
,
'
--
-
.
-
.
_
+ - ~ <.
)
IE Rpt. No. 50-302/76-22 IV-3
_
.
.
l The new annunciator alarm procedures have been organized in tabular form to improve the operator's ability to locate an alarm when needed. A review of randomly selected procedures revealed that
,
they are complete and concise. This unresolved item is closed.
5.
Preoperational Test Results Evaluation
,
i
'
Completed preoperational tests were reviewed to assure that the licensee had performed an adequate evaluation of test results and that the test results were within established acceptance criteria.
The review also verified the adequacy of the methods used for identifying and correcting test deficiencies and assured that administrative control practices required by Section 13.2 of the FSAR and the Test Program Guide (TPG) were being followed in test results documentation. The inspector tracked the incomplete. test steps and found that open items were identified on the " Exception and Deficiency" page and also were identified on the " Punch List"
of open test items. The following preoperational tests were reviesed:
TP 207-1, Heat Tracing TP 600-25, Integrated ES Actuation Test TP 600-3, Soluble Poison Concentration Control TP 172-2, Control Complex Ventilation Functional Test TP 150-8, Reactor Building Isolation Valve Leakage Rate Test At the completion of the inspection period, the open items related to the above tests were resolved. There were no discrepancies
,
~
identified.
- 6.
Initial Fuel Loading Procedure During inpection 50-302/76-6, a comment concerning radiation control practir.es required to support initial fuel loading was identified.
'
Follesup revealed that the item.is covered in AI 700, " Conduct of Chemistry and Radiation Protection" and RP-106, " Radiation Work Permit Procedure." The inspector reviewed these two procedures and had no questions. The comment is considered resolved..
!
7.
Plant Tour I
i A inspection tour of the plant was conducted wi h emphasis on the t
housekeeping status of the reactor building.
Extensive housekeeping activity in the reactor building was noted.
The reactor building
!
was essentially ready for initial operation. During inspection
.
.
.
.
'
.
.
.
.
e
.
-
m.e-e e--
s-
-
g
-.
+.,.
--
,.,*.%.g,
- -,,.
y e
-
+,.--g qm,9 o-g
_,,e
--w.y-
-p
.-p,yyy,
- -
-
.
.
.
.N
.
i
-
IE Rpt. No.. 50-302/76-22
-
IV-4-
'
l
,
50-302/76-13, it was noted that the reactor building sump screens were not in place.
Followup on this consnent revealed that the
,
'
screens are now in place, as described in Section 6.2.2.1 of the FSAR.
]
The condition of the auxiliary building has noticeably improved over that observed in past inspections.
j-8.
Safeguards System Pump Runout The status of safeguards system pump runout was reviewed and
,
unresolved item 75-8/2 is now closed. Engineering change notice
3489 and modification approval record 76-11-13 have been completed.
i Testing and appropriate re-testing have been c spleted with satis-factory results. A flow control method has been installed in the low pressure injection system and the reactor building spray system to control pump runout. Results from TP 203-3, " Decay Heat System Functional Test," and TP 20-5, " Reactor Building Spray System Functional" were reviewed by the inspector and found acceptable.
9.
Licensee Evaluation of Preoperational Test Result The status records of all preoperational tests were examined ta
.
determine if the licensee has completed their evaluation and awep -
!
ance of test results.
Six tests were identified as incomplete.
Satisfactory completion of these six tests, listed below, have been identified as requiring resolution prior to initial criticality:
,
a.
TP 150-3, " Post Accident H Purge Test"
b.
TP 160-3 and 600-36 " Reactor Building Cooling"
.
TP 160-7, " Reactor Cavity Cooling" c.
d.
TP 267-5, "CRD Cooling Functional" e.
TP 302-2 and 3, "Incore System" f.
TP 600-22, " Liquid Waste Operational"
Followup inspection will be conducted on the listed tests.
10.
Faergency Facilities and Equipment
,
During inspection 50-302/76-7, comments were made concerning implementation of the emergency plan.
Followup on these items
,
.
e
.
.
e--v-wee tL--y--r,9 y-g*g.e--
m-w-
-g
--s+.,
a
--.y--e+--
g ww.i--w, y-rw.,-
rwr,w.9
-
ww g-ww
.gwr+e-w-w
.ww-w,,,y++.,grv,y wwy---.-.,w',
_
~
.
-
.
-
'N '
.
IE Rpt. No. 50-302/76-32 IV-5
.
revealed that all have been resolved.
The comments and disposition of each are summarized below, a.
Installation of Filters for Control Complex Ventilation and Issuance of Surveillance Procedures The inspector verified that filters have been installed and testing of the control complex ventilation has been completed and results accepted. Surveillance Procedure - 186 has been issued to cover the requirements of Technical Specifi-cation 4.7.7.1.
b. ' Area and Process Radiation Monitoring Equipment The inspector verified that the system has been installed and calibrated.
TP-360-2, " Atmospheric Radiation Monitoring Functional" has been completed and results accepted, c.
Plant Communication Systems
"N Preoperational test TP 452-2, " Communications System Functional,"
has been completed and the results accepted. The Power Radio Service Base has been installed and is functioning in the control room.
'
d.
Fire Training for Plant Personnel Resolution of this item was accomplished by ammendment 48 of the FSAR. Section 7.2.2, Appendix 12B of the FSAR now states that key fire brigade personnel will be trained in the use of fire equipment, rather than all plant personnel. Previous inspections verified that key fire brigade personnel were trained.
.
11.
Zero Power Physics Testing Comments were noted in inspection report 50-302/76-7 concerning the zero power physics testing program.
Followup revealed that the issues have been res61ved.
- Previously the inspector noted that the limiting values for measured moderator temperature coefficient and control rod group withdrawal-rate appeared to be less conservative than the values in Table 14-3 of the FSAR.
The inspector verified that the procedure has been revised and is now consistent with the FSAR.
t
.
i l
-
~.
-
-
-
.
.
-s
-
/
.
$
.
,
IE Rpt. No. 50-302/76-22 IV-6
,
..
12.
Initial criticality Procedure Review The initial approach to criticality is prescribed in TP 710-1,
"Zero Power Physics Test."
This procedure was reviewed using Sections 13.1.1 and 13.4.7 of the FSAR and Appendix C, part C of Regulatory Guide 1.68 as guidance.
The review verified the ade-quacy of dhe method to achieve initial criticality.
The test procedure is in agreement with the format described in ANSI N18.7 and has been reviewed by the Plant Review Committee as required by Section ~ 5.1.6 of the licensee's proposed Technical Specifications. No discrepancies were identified regarding this procedure.
13.
Power Ascension Test Procedure ~s Review Power ascension test procedures were reviewed to determine if the tests conformed to Regulatory Guide 1.68, Appendix C, "Preparati:n of Procedures" and Section 13.2.5.2 of the FSAR. Within the test
'N procedures reviewed, there were no discrepancies identified.
The following tests were reviewed:
a.
TP800-11. " Core Power Distribution Test" b.
TP 800-19, " Pseudo Ejected Control Rod Test" c.
TP 800-27, " Shutdown From Outside Control Room Test" d.
TP 800-24, "Incore Detector Testing" PP 800-25, " Pipe and Component Hanger Hot Inspection at Power" e.
.
e e
,
'
.
e
-,
-
-
_
.
. -
__ -
_. -. - - - - - -,,,