IR 05000302/1977004
| ML19308D191 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 04/12/1977 |
| From: | Jape F, Robert Lewis, Troup G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19308D185 | List: |
| References | |
| 50-302-77-04, 50-302-77-4, NUDOCS 8002270657 | |
| Download: ML19308D191 (18) | |
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II Inspection Report No. 50-302/77-4 Licensee: Florida Power Corporation 3201 34th Street, South P. O. Box 14042 S t.
Petersburg, Florida 33733 Facility Name: Crystal River 3 Docket No.:
50-302 License No.:
DPR-72 Category:
BP_
Location: Crystal River, Florida Type of License:
B&W, PWR, 2452, Mut Type of Inspection: Routine, Unannounced (March 7-11, 1977)
Routine, Announced (March 10-11, 1977)
Dates of Inspection: March 7-11, 1977 Dates of Previous Inspection: February 22-25, 1977 Inspectors:
G. L. Troup, Radiation Specialist (March 7-11, 1977)
S. D. Ebneter, Reactor Inspector (March 9-11, 1977)
J. D. Martin, Reactor Inspector (March 10-11, 1977)
Other Accompanying Personnel: None Principal Inspector:
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F. Jape, Reactor Insp @ tof Date Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch Reviewed by: [. 6,
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R. C. Lesis, Chief
' Date Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch
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I IE Rpt. No. 50-302/77-4-2-
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SUMMARY OF FINDINGS I
I.
Enforcement Matters i
Deficiencies 1.
Lack of Procedure Contrary to Technical Specification 6.8.1.c, surveillance
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tests were conducted in December 1976 and February 1977 on the i
filters and absorbers in three ventilation systems without approved procedures.
(Details I, paragraph 2)
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2.
Loss of Test Record
Contrary to the requirements of license condition 13.B of by-product material license 09-07682-02, the licensee did not retain records of leak tests performed on sealed radioactive sources.
(Details I, paragraph 3)
3.
Failure to Certify or Approve Procedure CP-108 Contrary to 10 CFR 50, Appendix B, Criterion V and the q ration QA Program as delineated in Section 1.7.6.7.1.e of the FSAR, procedure CP-108, " Nondestructive Examination," was not certified or approved by a Level III examiner prior to implementation.
(Details II, paragraph 5)
II.
Licensee Action on Previously Identified Enforcement Matters Infraction: Failure to Evaluate Examination Data The licensee has evaluated the PSI NDE data in question relative to ASME Section XI requirements and have taken corrective action as stated in FPC's letter to NRC, dated January 4,1977. This item is closed.
(Details II, paragraph 3)
III. New Unresolved Items
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l UN 77-4/1 Loss of Off-site Power Test
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The plant configuration used to conduct the loss of l
off-site power test does not appear to meet the FSAR
or R.C. 1.68.
Item referred to IE:HQ for evaluatic'.
(Details III, paragraph 21
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IV.
Status of Previously Identified Unresolved Items UN 77-2/4 Survey Records for Radioactive Sources This item has been elevated to a deficiency since the licensee has not retained the required survey records.
The unresolved item is therefore closed.
(Details I, paragraph 3)
V.
Unusual Occurrences
,.None VI.
Other Significant Findings Personnel Changes The following changes were announced, effective March 7, 1977:
J. Alberdi, formerly Project Manager, Crystal River Unit 3 was appointed Manager, Fossil Operations.
R. C. Bonner, formerly Electrical Construction Supervisor was appointed Manager, Crystal River Unit 3 Project Transition.
VII. Management Interview A management interview was held on March 11, 1977, with G. P. Beatty, Jr.,
and staff. The inspection findings were discussed and licensee personnel were informed of the three deficiencies identified during this inspection.
(Details I and II) The closeout of the previous infraction regarding failure to evaluate examination data was dis cussed.
(Details II) The inspectors findin6s regarding the loss of off-site power test was discussed and the item remained unresolved.
(Details III)
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IE Rpt. No. 50-302/77-4 I-l s
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As DETAILS I Prepared by:
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G.L.Trob,RafationSpecialist Date Radiation Support Section Fuel Facility and Materials Safety Branch Dates of Inspection: ? reh 7-11, 1977 Reviewed by:
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4/4 /~7 7 A. F. Gibson, Chief Date Radiation Support Section Fuel Facility and Materials
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Safety Branch 1.
Individuals Contacted G. P. Beatty, Jr. - Nuclear Plant Superintendent P. F. McKee - Assistant Nuclear Plant Superintendent J. R. Wright - Chemical and Radiation Protection Engineer J. L. Harrison - Assistant Chemical and Radiation Protection Engineer g
G. D. Perkins - Health Physics Supervisor D. H. Ruzie - Results Engineer D. W. Pedrick, IV - Complainee Engineer W. A. Cress - Plant Engineer R. E. Fuller - Plant Engineer M. E. Collins - Plant Engineer E. E. Froats - Manager, Site Surveillance R. S. Dorrie - Quality Engineer j
D. E. Olson - Quality Engineer l
4 Chem Rad technicians 2.
In-place Testing of AEPA Filters and Charcoal Absorbers
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a.
Technical Specifications sections 4.6.4.2, 4.7.7.1 and 4.7.8.1 specify the surveillance requirements to demonstrate operability of the hydrogen purge (Reactor Building exhaust), Control Room emergency ventilation and Auxiliary Building ventilation exhaust systems, respectively. These sections require that the system satisfy the in-place testing acceptance criteria of Regulatory Guide 1.52, Revision 1 when tested in accordance with ANSI N5.10-1975, " Testing of Nuclear Air-Cleaning Systems."
Testing is required on a periodic basis as well as af ter replacement of the filters or absorbers in the systems.
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IE Rpt. No. 50-302/77-4 I-2
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b.
Technical Specifications section 6.8.1.c requires that " written procedures shall be established, implemented and maintained
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covering... surveillance and test activities of safety related equipment." The filters and absorbers in the systems stated in paragraph 2.a above are specified as safety related equipment in plant documents. ANSI N510-1975, paragraph 4.2 states, in part, " Detailed procedures shall be prepared for i
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each required test, based on the requirements of this standard."
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Plant surveillance procedures SP-185, -186 and -187 specify that the requirements fc
- he various tests specified in the Technical Specifications bu state the actual testing shall be performed by a contractor. Each of these surveillance proce-
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dures contains the statement "all procedures used by the contractor shall be approved by Florida Power Corporation CR-3 staff prior to their use."
c.
During the period December 15-23, 1976 a contractor conducted tests to verify that the systems were operable prior to initial criticality and conducted tests on February 22, 1977 af ter the Reactor Building exhaust filters and absorbers were replaced. The inspector discussed the tests conducted by the contractor with the cognizant engineer and reviewed the test j
reports submitted by the contractor. The engineer informed the inspector that the contractor hcd performed the tests using his calibration procedures and ANSI N510-1975 as a guideline but did not have " detailed written procedures" for
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each test. The inspector informed licensee management that this constituted an item of noncompliance based on Technical
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Specifications section 6.8.1.c as surveillance tests were performed without the use of written procedures as required by the Technical Specifications as well as by approved plant procedures and the referenced ANSI standard. This was acknowl-edged by management; a responsible licensee representative stated that preparation of detailed test procedures by the
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plant staff or requiring such procedures to be provided by the
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contractor would be considered as a solution of this apparent item of noncompliance.
d.
As the charcoal absorbers were replaced in the Reactor Building exhaust system in February 1977, the inspector reviewed the
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certifications for the new charcoal and compared the values against the requirements of Regulatory Guide 1.52, Rev.1, Table 2.
The charcoal certifications conformed to the specified acceptable results; the inspector had no further questions on the replacement charcoal. The inspector also reviewed the contractor's test reports for the in-place testing of the filters and absorbers and noted that the results exceeded the
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IE Rpt. No. 50-302/77-4 1-3
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minimum removal efficiencies specified in the Technical Speci-fications. As the test methods conformed to ANSI N510-1975 the inspector had no questions on the efficiency testing.
3.
Survey Records for Sealed Radioactive Sources (77-2/4)
a.
This item was originally discussed in IE Report No. 50-302/77-4, Details II, paragraph 2 and dealt with the retention of the survey records for the leak tests for sealed radioactive
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sources. Retention of the records was required by license condition 13.B of by-product material license 09-07683-02, which was applicable to the radioactive sources in the posses-
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sion of the licansee prior to the issuance of operating license DPR-72.
b.
During the inspection a licensee representative informed the inspector that after conducting a search of the files and records, the survey records could not be located and it was concluded that the records were not retained. The inspector informed licensee management that this is an item of noncom-pliance as the license required that the retards be retained.
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This. action closed the unresolved item. This was acknowledged
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by licensee management.
c.
The inspector reviewed the records of the source leak tests conducted in accordance with Technical Specifications sections 3.7.10 4.7.10 and verified that the tests had been performed at the required frequency, results were within the specified limits, and the records were on file. After reviewing the most recent tests (February 1977) and discussing the probable causes for failure to retain previous records with licensee representatives, the inspector informed licensee management that corrective action appeared to have been taken to correct the situation and no response would be required for the item of noncompliance. This was acknowledged by licensee management.
4.
Power Ascension and Operations Chemistry Tests a.
Technical Specifications sections 3.4.7 and 3.4.8 specify the
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limits for reactor coolant chemistry and specific activity, respectively. Surveillance requirements and frequencies are specified in Table 4.4-3 for chemistry and Table 4.4.4 for specific activity. The inspector reviewed the reactor coolant chemistry records for the period February 7-24, 1977 to deter-mine that the required analyses were performed at the specified frequencies. The inspector noted that during this period one analysis (fluoride) was performed outside the required frequency O
on one ocassion but no other problems were noted.
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IE Rpt. No. 50-302/77-4 I-4 I
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i b.
Technical Specifications Table 4.4-4 requires that an isotopic analysis for dose equivalent iodine-131 per once per 14 days during power operations and an isotopic analysis for iodine be performed between two and six hours following a power change exceeding 15% of rated power with a one hour period. The
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inspector reviewed the chemistry records and verified that the dose equivalent iodine-131 determination was done at the required frequency during the period February 7-24, 1977 and that the iodine analyses were done as required during the power changes which occurred during March 1-9, 1977. The inspector had no further questions on the chemistry or specific activity analyses.
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c.
The inspector reviewed the following power ascension test procedures and ascertained that the required sampling and analyses were performed at the specified power levels, that the data were being recorded as specified and that water quality was being controlled as specified:
TP 7 2 500 1 Reactor Coolant Chemistry Test TP 7 2 500 2 Steam Generator Chemical Test TP 7 2 500 3 Initial Radiochemistry Test Afte. reviewing these test procedures the inspector had no questions on the conduct of the tests or the results but made several minor comments on the recording of data from the laboratory in the record copy of the test procedure. These comments were acknowledged by the cognizant supervisor, who
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initiated corrective action. The inspector had no further coaments.
5.
Licensee Event Report 77-13 a.
Licensee Event Report 77-13 describes an event in which a licensee employee inadvertently activated the deluge system in the Reactor Building exhaust system while cleaning the actua-tion station. Actuation of the deluge system resulted in water being sprayed onto the charcoal absorbers, rendering them inoperable.
b.
The inspector reviewed the subject event ret. ort and verified that it was schmitted in accordance with the requirements of Regulatory Guide 1.16, including timing of submittal. The inspector also discussed the event with licensee representatives and verified that the event was reviewed and evaluated, that appropriate corrective action was taken and that generic implications were considered. The inspector verified by
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observation that protective covers were installed over the s.
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i actuation station for the deluge system and that covers were also installed on other stations where similar situations were possible (e.g. the fire suppressant system in the cable spread-ing room). Replacement of the effected filters and absorbers is discussed in Details I, paragraph 2.
The inspector had no further questions.
6.
Radiological Surveys a.
10 CFR 20.201(b) requires that "each licensee shall make or cause to be made such surveys as may be necessary for him to comply with the regulations in this part."
Plant procedure
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RP-202, " Radiological Surveys" specifies the requirements for routine radiation and airborne activity surveys. Plant proce-dure RP-106, " Radiation Work Permit Procedure" specifies the requirements for special surveys 1n conjunction with work to be performed with the permit. 10 CFR 20.401.(b) requires that the licensee maintain the records of surveys performed in accordance with 10 CFR 20.201.(b).
b.
The inspector reviewed the survey records for February 7-15,
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1977, February 21-28, 1977 and March 6-9, 1977 and verified that the required surveys were performed as required by proce-dure RP-202 for daily, twice weekly, weekly and monthly surveys.
The inspector also verified that areas which exceeded the limits were identified. and that the surveys were reviewed by
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supervision. The inspector also reviewed thirty radiation
work permits and verified that the special surveys required by procedure RP-106 were performed and the results recorded on the permits. The inspector had no further questions on radi-ological procedures.
c.
Test procedure TP 7 2 850 1, " Biological Shield Survey" speci-
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fies the special survey to be performed to assess the adequacy of the biological shield. The inspector reviewed the procedure
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and verified that the measurements were taken at the required
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power levels and that measurements of the specified type were
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being taken at specified locations. The inspector had no
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questions ccacerning the performance of the test for the power levels completed. The inspector also discussed the results of the completed surveys with licensee representatives, who informed the inspector that evaluations were being made of the
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areas which were above design levels and the corrective action would be initiated based on the evaluation and further measure-ments at higher power levels.
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IE Rpt. No. 50-302/77-4 I-6
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7.
Posting and Control _
a.
10 CFR 20.203 specifies the requirements for caution signs, j
labels and controls for radiation areas, high radiation areas and airborne radioactivity areas. Technical Specification Section 6.13 modifies the requirements of 10 CFR 20.203.(c)(2)
for the control of high radiation areas.
b.
On March 7 and 9, 1977 the inspector toured plant areas and checked the posting and controls of areas to see that they
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were consistent with current survey results. The inspector l
l also took radiation measurements using a licensee's instrument
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to verify that areas were properly classified as radiation areas and to identify any improperly posted areas. On March 7, 1977 the inspector identified one area around make-up pump 3A which was not posted as a radiation area but where the dose rate exceeded 5 mrem / hour for the whole body. A licensee
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representative who was accompanying the inspector acknowledged that this area should have been posted as a radiation area and initiated action to have it posted. The licensee representative explained that the radiation area was present when the pump was running but was not present when the pump was shut down, and that the pump had been running only part of the day. On March 9 the inspector again took measurements in the area when the pump was shut down; no radiation area was present. No other areas were identified which were improperly posted.
c.
The inspector reviewed the control of high radiation areas and contamination areas, and noted that they were barricaded and posted. The inspector had no questions on these areas.
d.
The inspector reviewed thirty radiation work permits (RWP's)
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for. work $n the period January-March 1977 and verified that they were being used consistent with procedure RP-106. The inspector also observed the posting of contamination and high radiation areas and verified that these areas were posted with i
the notation "RWP required." The inspector had no further
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questions on the use of radiation work permits.
8.
External Exposure Control The inspector reviewed the licensee's external exposure control program by examining exposure control records, reviewing relevant procedures, observing licensee practices concerning issuance and use of personnel monitoring devices and discussing practices with licensee representatives. Based on this review, the licensee appeared to be in compliance with 10 CFR 20.202(c) with respect to external exposure monitoring.
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9.
Radiological Protection Procedures The inspector reviewed the changes made to ten plant radiological protection procedures (RP-100 and RP-200 series) during 1976 and 1977 and verified that the changes had been reviewed and approved in accordance with Technical Specification 6.8.2.
The inspector also reviewed the contents of the changes and verified that they were consistent with 10 CFR 20 and the Technical Specifications.
The inspector had no further questions.
10.
Exposure of Minors
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, The inspector discussed the exposure of minors to external radiation l
and airborne radioactivity with licensee representatives. A licensee representative stated that, at the present time, no minors are employed at or assigned to the p) ant, but should the situation arise, controls would be established to restrict their exposure.
The inspector also reviewed plant procedure RP-101, " Radiation Protection Manual" and noted that the procedure restricts external exposures to minors in accordance with 10 CFR 20.104a but does not set limits on the exposures to airborne radioactivity. This was discussed with a licensee representative who acknowledged the need to define the Itaits for airborne exposures and stated that a revision would be made to RP-101 to clarify the limits for minors.
The inspector had no further questions.
11.
Reports to the NRC The inspector discussed 10 CFR 20 reporting requirements for the loss or theft of licensed material and incidents with the cognizant supervisor. Based on these discussions the inspector determined that the licensee has had no activities or occurrences which would require reporting 10 CFR 20.402 or 10 CFR 20.403. The inspector
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had no further questions.
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12.
Changes in Procedures or Facility l
The inspector discussed changes in plant procedures or in the j
facility design which might have resulted from problems encountered
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during power ascension. A licensee representative stated that-while some changes had been made to radiological protection proce-
dures since start-up, these were the result in changes in the regulations or to improve the performance of the procedures and were not the result of deficiencies in the procedures. One change in the facility currently in progress is the installation of shield-
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ing on the make-up system letdown line; this was identified prior j
to start of power ascension although the need for shielding was
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IE Rpt. No. 50-302/77-4 I-8
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i verified during power ascension. Additional changes (additional i
shielding) are being evaluated as the result of the power ascension l'
shield survey (see paragraph 6.c).
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IE Rpt. No. 50-302/77-4 II-1
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DETAILS 11 Prepared by:
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If77 S. D. Ebneter, Nondesductive Date Examination Engineer Engineering Support Section No. 2 Reactor Construction and Engineering Support Branch Dates of Inspection: March 9-11, 1977 Reviewed by:
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A.R.Hardt{SupportSectionNo.2 ChMiW
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Engineering Reactor Construction and Engineering
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Support Branch
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1.
Persons Contacted Florida Power Corporation (FPC)
G. P. Beatty - Nuclear Plant Manager D. W. Pedrick IV - Manager, Compliance S. W. Johnson - Plant Engineer 2.
Scope This inspection consisted of:
(1) a review of corrective action taken by the licensee to comply with ASME Section II requirements; (2) an examination of actions taken in response to IE Bulletin 76-06; and (3) an audit of the licensee's control of the nondestructive examinations (NDE) program.
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3.
Preservice Inspection (PSI)
II: Report No. 50-302/76-22 identified Noncompliance 76-22-Al(II),
" Failure to Evaluate Examination Data," as a failure of the licensee to fully evaluate ultrasonic test (UT) data as required by ASME Section XI.
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The PSI contractor instituted additional data reviews by the Level II examiner and further evaluation by a Level III exuiner to assure compliance with requirements and preclude recurrence. In a letter from Babcock and Wilcox Company (B&W) to FPC, twenty (20)
additional indications were evaluated by the Level III examiner
including those identified in the above IE Report. The evaluation l
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concluded that the majority were geometric in nature and the remaining were within code acceptance limits. The evaluation will be incorpo-rated in the final PSI report. The inspector reviewed the data for the re-evaluated indication and compared the data with the Level III's evaluation to the extent possible. The corrective action appears to be adequate and this item is closed.
The inspector reviewed the licensee's program for inservice inspection of pumps and valves. The licensee has a well documented program relative to those pumps and valves for which relief (waivers) have been requested in accordance with 10 CFR 50.55a.
The request for relief was submitted to NRR for comment on March 2, 1977.
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4.
IE Circular 76-06 As a result of previous problems encountered at other FWR facilities
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an information copy of IE Circular 76-06 was forwarded to FPC.
In January 1977, FPC discovered a through wall crack in an 8-inch j
Schedule 40, 304 stainless steel reactor building spray pipe.
Since FPC had been issued an operating license, the actions recommended in IE Circular 76-06 were initiated. This event was reported by
telephone and confirmed by mailgram on January 3,1977. Subsequently, LER 77-1 was submitted and additional investigations were performed.
The additional investigation included nondestructive examination of 49 (17 percent) additional welds in the decay heat system and 22, (26 percent) additional welds in the reactor building spray system.
The ultrasonic examina : ions were performed by Babcock and Wilcox (B&W) personnel in accordance with procedures ISI - 101 and ISI - 102.
The examiners were certified per SNT-TC-1A and all calibration and data sheets were filled out in accordance with procedural requirements. Two additional defective welds were found during the additional investigations, a thru-vall crack in pipe (BS-82) of 8-inch Schedule 40, 304 stainless steel reactor building spray pipe and a lack of penetration in weld BS-8.
The section of pipe containing Weld BS-82, reported in LER 77-1, was cut out and sent to Battelle Memorfs?. Institute for analysis.
This weld in located inside the conta?>.aeut at penetration 341 and
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utilized a consumable insert at the weld of 8-inch Schedule 80, 304 l
stainless steel penetration pipe to an 8-inch Schedule 40, 304 stainless steel process pipe.
This particular veld had undergone four repair cycles during erection and several flushing generations.
The system was not drained af ter final flush and therefore contained stagnant borated water.
Batte11e's analysis included optical x
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IE Rpt. No. 50-302/77-4 II-3
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scanning electron microscope and electroprobe analyses.
Their conclusion was that the through wall crack was due to intergranular stress corrosion cracking of the sensitized pipe. Electroprobe analysis confirmed the presence of chlorides at the crack surface.
High residual stress was probable due to the maltiple weld repairs.
Wald BS-81 was located on the outside of containment penetration 341 and exhibited cracking similar to that of BS-82.
It also had a similar history with regard to flushing and containing stagnant borated vatar.
Both welds were repaired by removal of the existing pipe and replacement with a new section.
The inspector reviewed the documentation related to these repairs including inspection results, RT film, and process records. The inspector noted in the e' review of records that the plant engineer had initialed / signed some NDE records.
It was explained that the plant engineer was a certified Level II visual examiner as well as the primary contract administrator for B&W and PTL contract NDE services; thus his sign off did not necessarily represent an inspection approval.
During ultrasonic examination of Weld BS-8-B, a large reflector was noted which was diagnosed as lack of penetration at the root.
This weld was repaired and re-examined in accordance with documented proceduru.
All records and data appeared to be adequate.
This defect was documented internally but was not due to stress corrosion cracking and/or relt.ced to the conditions identified in IE Circular 76-06.
The licensee has not submited the final report detailing the results of actions taken per IE Circular 76-06.
The licensee has proposed in the draf t report the following actions to provide future surveil-lance and prevent recurrence:
Additional volumetric examination of 5 percent of the welds in a.
reactor building spray and decay heat systems during each ISI interval; b.
Drain lines, where possible, to prevent accumulation of liquids 5.
NDE Program The licensee's Quality Manual addresses the control of nondestruc-tive examination procedures and personnel through the following documents:
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Number Title QP 9.50 Control of Special Processes QP 9.11 Certification of Special Process Procedures and Personnel QP 2.51 Training for Operations QP 9.10 Special Process Control Policy 9.1 Special Process Certification and Control The procedural requirements appear to be adequate to assure special process control and certification of NDE personnel. Procedures are
/ required to be reviewed and approved by a Level III examiner and NDE laboratories providing NDE services are required to be certified by a Level III examiner. Personnel qualifications and certifications are specified to be in accordance with SNT-TC-1A and administered by the Level III examiner.
FPC is utilizing B&W NDE procedures and personnel to accomplish PSI examinations. Quality Program Policy 9.1, paragraph 4.5 requires certification of NDE laboratories providing contract NDE services.
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The licensee stated that this provision is not applicable to B&W
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which is to be certified by "grandfathering" as a result of contractual obligations incurred prior to implementation of the procedure.
FPC uses Compliance Procedure CP-108, " Nondestructive Examination,"
an internal procedure, to perform NDE.
This procedure was reviewed by the plant review committe and approved by the Nuclear Plant Manager. Licensee personnel stated that they have used the procedure to perform penetrant examinations detailed in Section 8.
The procedure has not been approved for use by the Level III examiner as specified in QP 9.50.
Paragraph 4.1.2 of QP 9.50 states that the Level III examiner witnesses the demonstration of the NDE Procedure and if satisfied that all requirements are met, he approves the procedure for use. This requirement is restated in QAP No.17, i
paragraph 4.9.
In addition to procedure CP-108, it appears that procedure CP-109, " Nondestructive Examination Personnel Qualification,"
should have been reviewed and approved by the Level III examiner.
Failure to follow the above procedural requirements is an apparent noncompliance with FSAR Section 1.7.6.7.1.e and 10 CFR 50, Appendix B,
Criterion V.
This noncompliance is identified as 77-4-1A(III).
Based on discussions with the licensee, only Section 3 of procedure CP-109 has been implemented. A review of this section revealed several areas which are of concern:
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IE Rpt. No. 50-292/77-4 II-5
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a.
The procedure contains no scope. With regard to PT examination of welds what is the extent of the examination coverage? Does it extend to one wall thickness beyond the weld edge into the base metal?
l b.
The procedures states that it is for nonswashable dye penetrants which essentially means solvent removable. Paragraph 8.7.2
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allows swabbing of the developer. ASMI Code practice (Section V; 1974) permits solvent suspensions to be applied by spray only.
c.
Paragraph 8.8.2 defines linear indications as those in which the length is more than twice the width. ASME Code practice is to define linear indications as those in which the length is greater than three times the width. This is the definition used in PSI and other code related requirements. Inconsistency of this definition could lead to, problems in evaluation of indications.
d.
Paragraph 8.2.2 allows grinding and machining to remove indications.
There is no provision to perform volumetric examination to verify that minimum wall thickness has not been violated.
Similar examples of procedural inadequacies can be cited for other sections of this document. In all cases, no scope, or coverage of the examinations has been defined.
In many cases, calibrations or checks are specified but no acceptance limits are defined. Examples are the requirement for the UT couplant to be free of sulfur and halogen and the statement to check amplitude and horizontal linearity of the UT equipment with no reference as to how and what constitutes acceptable conditions.
All of the above itens and others should be identified by adequate review of the procedure.
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The inspector audited NDE examination data, calibration records, personnel qualification records, and specific contractor prepared procedures to determine compliance with commitments and requirements.
The inspector had no further questions.
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IE Rpt. No. 50-302/77-4 III-1
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DETAILS III Prepared
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~nartirg Reactor Inspector
[Da[a Nu ar Support Section Reactor Operations and Nuclear Support Branch Dates of Inspectio.
March 10-11, 1977
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Reviewed by:
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R. D' Martin, Chief Date Nuclear Support Section
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Reactor Operations and Nuclear l
Support Branch
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Individuals Contacted Florida Power Corporation (FPC)
G. P. Beatty, Jr. - Nuclear Plant Superintendent P. F. McKee - Assistant Nuclear Plant Superintendent W. R. Nichols - Operations Supervisor D. W. Pedrick, IV - Compliance Engineer J. C. Hobbs, Jr. - Manager Generation Testing l
Babcock and Wilcox Company (B&W)
J. Kelly - Testing Superintendent
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l J. Bohart - Startup Test Engineer i
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Power Ascension Test Witnessing Portions of the power ascension testing that were in progress during the dates of this inspection were witnessed by the inspector.
In addition, an evaluation was made of previously conducted tests.
l The performance of each test was evaluated against the requirements j
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of Regulatory Guide 1.68 - November,1973 and FSAR Chapter 13, I
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" Initial Tests and Operation."
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The following documents were reviewed:
TP-71-800-26 (Rev.,1), Loss of Offsite Power Test TP-71-800-14 (Rev. 2), Reactor / Turbine Trip Test TP-71-800-0 (Rev. 2), Power Escalation i
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IE Rpt. No. 50-302/77-4 III-2
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Within the areas inspected one new unresolved item resulted.
(UN 77-4/1) The plant configuration used during the conduct of the
" Loss of Offsite Power Test" (TP-71-800-26) on March 9, 1977, does not appear to meet the intent of Regulatory Guide 1.68 November, 1973 Section D.l.k.
The FSAR Chapter 14.1.2.8.1 states that the facility is designed to withstand a complete loss of all system and unit power except the station batteries. Also item 11 of FSAR Table 13-4, titled " Loss of Offsite Power" states that the purpose of the test is to demonstrate operation of the diesel generators (DG), turbine driven FW pumps and engineered safeguard (ES) buses in the event of a total loss of offsite power.
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When the test was conducted the unit main generator was separated
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from the grid and lef t in service supplying facility auxiliary loads (house loads) through the Unit Auxiliary transformer. An emergency DG was then manually started and placed in service carrying a 2000 KW load on ES bus 3A.
Loss of offsite power was simulated by opening the breaker from the startup transformer to ES bus 3B.
The remaining DC automatically started and reenergized ES bus 3B.
ES bus 3A was similarly tested by energizing ES bus 3B with a DG and interrupting the offsite power supply to ES, bus 3A.
No inter-ruption of reactor or turbine - generator operation was experienced since the main unit generator continued to supply the house loads and one ES bus was being energized by a DG, Since the intent of Regulatory Guide 1.68 Section D.l.k is to demonstrate the response of the plant for a simulated condition of loss of the unit turbine - generator coincident with a loss of offsite power, the validity of the as-run test has been referred to g
IE:RQ for evaluation.
(UN 77-4/l)
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