ML20153A951
| ML20153A951 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 03/11/1988 |
| From: | NRC OFFICE OF SPECIAL PROJECTS |
| To: | |
| Shared Package | |
| ML20127A683 | List:
|
| References | |
| NUDOCS 8803210303 | |
| Download: ML20153A951 (12) | |
Text
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SAFE 1Y EVALUATION REPORT BY THE OFFICE OF SPECIAL PROJECTS EMPLOYEE CONCERN ELEMENT REPORT CO 11306-SON "TESTING OF ANCHORS"
, TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR POWER PLANT, UNITS 1 AND 2 DOCKET NOS 50-327 AND 50-328 I.
SUBJECT Category:
Construction (10000)
Sub Category: Anchorages (11300)
Element:
Testing of Anchors (11306)
Concerns:
IN-85-285-002, IN-85-347-007. IN-86-115-001, IN-86-190-003 and XX-85-023-001.
The basis for element report C0 11306, Revision 5, dated July 9, 1987 are five enployee concerns which question TVA's testing of anchor bolts.
II.
SUMMARY
The Employee Concerns Task Group (ECTG) report identified the following four issues from the employee ccreerns:
a.
Shell-Self-Drilling (SSD) type anchort were inproperly tested or the pull test was bypassed.
b.
550 type anchers are tested by sampling rethods instead of on an individual basis, c.
SSD type anchors are being overtorqued.
d.
There is no torquing requirerent for instrument panal bolts.
III. EVALUATION A technical review of Employee Concerns Elenent Report CO 11306-SON, Revision 5 was performed by an NRC centractor under NRC Contract No. 05-86-156.
The results of this review are sumarized in the attached contractor techrical evaluation report dated Jaruary 21, 1988, on Employee Concerns Element Repert C0 11306, Revision 5.
Element Report CO 11306-SQN, Revision 5 found that there was a proarannatic concern with all TVA sites concerning the issue of overtighteninn of anchor bolts.
The recomended corrective action was that Secuoyah initiate a progran BBP2;gggggggagp P
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, to address this issue.
In addition, the element report recomerded that TVA's procedures be revised to assure anchor shells are not in contact with the base plates when they are tested.
Resolution of this corrective action was not considerd a restart iten in the ECTG report.
The contractor review of Elemen*
Report CO 11306, Revision 5 fcund that TVA's evaluation of the employee concerns and the proposed correctiva actions were adaquate.
The staff concurs
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with the conclusioni prese'1ted in tha contractor technical evaluation report.
IV.
CONCLUS!0NJ Based on the review of Employee Concerns Element Report CO 11306, Revision 5 the staff concludes that Employee Concerns IN-85-285-002, IN-SS-347-C07, IN-86-115-001, It!-86-190-003 and XX-85-023-001 have been adequately addressed at Sequoyah.
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Report 4 CO 11 306
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SEQUOYAH NUCLEAR POWER PLANT, UNITS 41 & 2 TECHNICAL EVALUATION REPORT FOR EMPLOYEE CONDERNS ELEMENT REPORT CO 11306-SON "TESTING OF ANCHORS" 1.
Subject Category: Construction,(10000)
Sub category:
Anchorages (11300)
Element:
Testing of Anchors 11306 Shell - Self-Drilling (SSD) type anchors were improperly tested, not pull tested, tested by sampling method, overtorqued and were not torqued.
Concerns: IN-85-285-002, IN-85-347-007,IN-86-115-001, IN-86-190-003 and XX-85-023-001.
The basis for element report CO 11306 Rev 45 dated 7/9/87 are-the following concerns:
I4-85-285-002 "TVA INSPECTED AND PULL TESTED REDHEADS IMPORPERLY:
PULL TESTING WAS NOT 100%.
BASE PLATE OR HANGER WAS BOLTED IN PLACE.
EVEN REDHFADS THAT WERE LOOSI COULD HAVE PASSED BY BEARING AGAINST THE BACK OF THE PLATE.
BECAUSE THE HOLES WERE NOT INSPECTED BEFORE REDHEADS WERE SET QC COULD NOT TELL IF REBAR HAD BEEN CUT.
CI HAD NO MORE INFORMATION.
CONST. DEPT. CONCERN.
NO FOLLOW UP REQUIRED."
IN-85-347-007 "PRCCEDURE DOES NOT REQUIRE TORQUING OF INSTRUMENT PANEL BOLTS."
IN-86-115-001 "SELF DRILLING EXPANSION SHELL ANCHORS ARE BEING OVERTORQUED.
THIS IS DONE TO CORRECT EXCESSIVE GAP-BETWEEN BASEPLATE AND WALL, CRAFT PERSONNE' ARE NOT TRAINED TO THE REQUIREMENTS OF SPEC. G-3 2 PARAGRAPH 3. 2.
CONSTRUCTION DEPT. CONCERN -CI HAS NO FURTHER INFORMATION UNITS 1 &
2."
IN-86-190-003 "AN EMPLOYEE TOLD THE CI THAT THE SAFETY RELATED CONCRETE ANCHORS (REDHEADS), WERE TESTCD BY A SAMPLING PLAN RATHER THAN INDIVIDUALLY.
CI QUESTIONED THE ACCEPTABILITY OF THIS PRACTICE.
NUC POWER CONCERN.
UNIT A1.
CI HAS NO ADDITIONAL INFORMATION."
r XX-85-023-001 SEQUOYAH UNIT #2.
PULL TESTS WERE ROUTINELY BY-PASSED AND/OR INCORRECTLY DOCUMENTED ON HANGERS / ANCHORS INSTALLED IN THE ANNULUS AREA.
MID-1977.
NO NAMES OR SPECIFIC
/
LOCATIONS WERE PROVIDED.
CONSTRUCTION DEPT. CONCERN.
CI HAS NO MORE INFORMATION.
NO FOLLOW-UP REQUIRED.
Concerns IN-85-285-002, IN-86-115-001, IN-86-190-003 and XX-85-023-001 were evaluated by TVA to be potentially nuclear safety significant and Concern IN-85-347-007 was evaluated by TVA to be not nuclear safety related.
II.
Summary of Issues The probltr. as defined by TVA is as follows:
.\\.
Shell-Self-Drilling (SSD) type anchors were improperly tested or the pull test was bypassed.
B.
SSD type anchors are tested by sampling methods instead of on an individual basis.
C.
SSD type anchors are being overtorqued.
D.
There is no torquing requirement for instrument bolts.
III. Evaluation The TVA Employee concern task group (ECTG) evaluators reviewed applicable documents and interviewed cognizant sequoyah engineering personnel to determine if the above concerns are valid.
IN-85-5-285-002 addressed three concerns.
Pull testing of red head expansion anchors was not 100%
o Pull testing was performed improperly.
Red dead shell o
beared against back of base plate.
o Holes were not inspected before red deads were set.
QC could not tell if rebar had been cut.
100% pull testing of expansion anchors was never a requirement in SON procedures.
The SQN procedures specify that the testing be performed on a sampling bas s.
This is an acceptable method.
The TVA procedures do not specific that the red head expansion anchors be tested with the base plate not in place or that the base plate be shimmed to prevent the shell of the anchor
o contacting the base plate during testing.
This is a weakness in TVA's procedures.
However the inspections and testing performed for the 79-02 response did require that the base plates be shimmed when the anchors were tested with the base plate left in place.
The results of the reinspection programs performed since r
79-01 indicate the failure rates are within acceptable limits.
This evaluation is acceptable.
However the TVA procedures should be revised to ensure that the base plates will be shimmed when testing is performed through the base plate.
This element report does not address the concern that rebar could have been cut but it is addressed ir element report number CO 11305-SQN.
Concern IN-85-347-007 states that the procedure does not require torquing of instrument panel bolts" The TVA evaluation agrees that the procedures do not require SSD type expansion anchor be torqued.
Setting of SSD bolts is accomplished by using a method other than torquing and torquing is not a requirement of this type of bolt.
The procedures do require that the wedge type be torqued.
This reviewer agrees with the evaluation.
Concern IN 86-115-001 states that "self drilling expansion shell anchors are being overtorqued."
Interviews with responsible personnel indicated that this issue had rever been addressed by SON.
The evaluation recommends that this issue be evaluated by SQN personnel.
This reviewer agrees with this evaluation The responses to concerns IN-86-190-003 and XX-85-023-001 are included in response to IN-85-285-002.
IV.
Conclusion As part of my review of this element report CO 11306, I also reviewed, General Construction specification 4G-32 Rev 41, M&AI procedures 49 Rev 7 & 11 Rev. 15.
TVA's investigations of these concerns are adequate and TVA 4
conclusions resolutions and commitments to resolve the concerns addressed in this element report are satisfactory.
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'Jhe commitments are as follows:
The specifications and procedures will be revised.to o
ensure that testing procedures do not allow pull testing of the anchors when the shell is in contact with-the f
base plate.
The potential of overtightening of SSD type anchors will o
be evaluated by SON personnel.
T
Safety Evaluation Report Sequoyah Nuclear Plant Employee Concerns I.
SUBJECT CATEGORY:
Construction (15000)
SUBCATEGORY:
Damage (15100)
ELEMENT:
Damage to Electrical Penetrations (C015102)
CONCERN:
(IN-85-346-002) - Electrical penetrations entering the reactor building are not strong enough and have been damaged because of being walked on.
~
II.
SUMMARY
OF ISSUE A
A generic concern about the condition of the electrical penetrations was identified as result of a review conducted on Watts Bar specific item IN-85-346-002.
An additional concern over the adequacy of the penetration covers was.also identified.
III. EVALUATION The licensee determined that there was.no damage to containment electrical penetrations and that the damage identified by the concerned individual was actually to the penetration. covers..The electrical penetration-covers are constructed of sheetmetal and are plant specific.
In addition, the electrical covers have been included in the seismic calculations com'pleted for the penetrations.
The licensee determined that the penetration covers specific to Sequoyah were more sturdy than those used at Watts Bar and that there was no present damage to either the electrical penetrations or the electrical penetration covers.
IV.
CONCLUSION The licensee's review of potential reportable occurrences and field verification appear to be adequate to determine that there is no damage to the electrical penetrations at Sequoyah.
The licensee has also determined that, at Sequoyah, adequate penetration covers are in place and that no excessive penetration cover damage exists.
No actions are required to be completed prior to the startup of either unit.
This issue is closed.
i a
Safety Evaluation Report Sequoyah Nuclear Plant Employee Concerns I.
SUBJECT CATEGORY:
Construction (15000)
SUBCATEGORY:
Damage (15100)
ELEMENT:
Damage to Instrumentation Tubing (015109)
CONCERN:
(IN-86-200-006) - Copper and stainless instrumentation tubing throughout Units 1 and 2 is unprotected and span between hangers are bent.
(IN-85-119-002) - Instrumentation lines through out power block are damaged, bent, flattened and touching.
(IN-85-618-004) - Instrument tubing in the Unit 2 accumula-tor #4 area is being severly damaged; protection should be provided.
II.
SUMMARY
OF ISSUE A generic concern about damage to instrumentation tubing was identified as a result of three Watts Bar specific items (IN-86-200-006, IN-85-119-002, and IRt85-618-004).
III. EVALUATION The licensee reviewed a Watts Sar specific element report (C015109-WBN) and determined that the Division of Nuclear Construction had a program to limit instrument line damage at Watts Bar.
It was determined that the Division of Nuclear Power at both the Watts Bar site and the Sequoyah Nuclear Plant did not have a program expressly designed to limit instrument line damage.
The need for a program to limit instrument line damage was assessed by the licensee through a walkdown of approximately 800 instrument tubes.
No instrument tube damage was identified during the walkdown process.
In addition, interviews with knowledgeable personnel did not identify any instrument line damage issues specific to the Sequoyah Nuclear Plant.
IV.
CONCLUSION The licensee's actions to evaluate the extent of instrumeht line damage appear to be adequate.
In the absence of present damage, the cetermination by the licensee that a damage prevention program was not needed at the Sequoyah Nuclear Plant, appears to be appropriate.
There are no actions required by the 1teensee prior to the startup of either unit. This issue is closed.
_.7
o SEQUOYAH NUCLEAR POWER PLANT, UNITS 1 & 2 i
SAFETY EVALUATION REPORT FOR EMPLOYEE CONCERN ELEMENT REPORT CO 17101-SQN "VALVES" I.
Subject Category: Construction (10000)
Subcategory:
Mechanical (17100)
Element:
Valves (17101)
Employee Concerns:
IN-85-055-N04 and XX-85-094-007 i
The basis for Element Report CO 17101-SQN, Rev. 2, dated October 10, 1986 is Bellefonte Employee Concern XX-85-094-007 and Sequoyah-specific Employee Concern IN-85-055-N04, which state (respectively):
"Bellefonte:
Limitorque valves were stored and installed with wrong altitude (upside down) and were r.ot maintained (stroked, etc.).
Construction Dept. concern.
CI has no further information.
No follow up required."
"NRC identified the following concern from review of QTC file.
"Emergency hand valve incorrectly installed."
These concerns were evaluated by TVA as potentially nuclear safety-related, and XX-85-094-007 was evaluated by TVA as potentially applicable to Sequoyah (generic).
II.
Summary of Issue:
The problems as defined by TVA are that (a) Limitorque motor operators for valves were not stored, installed or maintained properly, and (b) an "emergency hand valve" was incorrectly installed at Sequoyah and thus could not be manually operated.
III. Evaluation:
The two issues were separately investigated and evaluated by the Employee Concern Task Group (ECTG) evaluators.
They reviewed applicable mainte-nance, inspection, storage and installation procedures for the Limitorque valve motor operators, and also reviewed contract and design documents and held discussions with cognizant Sequoyah personnel.
The ECTG evaluators found that Sequoyah procedures and instructions met all of the vendor's requirements for storage, maintenance and installation.
Valves have been installed in other than the manufacturer's preferred orientation; however, both Limitorque and TVA's engineering staff have stated that if proper lubricant and maintenance practices are implemented the performance of Limitorque motor operators for valves is not affected by the mounting position.
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y
The second issue, involving the "emergency hand valve," was determined by the ECTG evaluators to be invalid.
Procedures in effect at Sequoyah should result in the identification and correction of equipment inaccessibility problems.
IV.
==
Conclusions:==
The NRC agrees with the TVA findings and conclusions as described in Element Report No. CO 17101-SQN Rev. 2.
No further action by the NRC required.
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t SEQUOYAH NUCLEAR POWER PLANT, UNITS 1 AND 2 SAFETY EVALUATION REPORT FOR EMPLOYEE CONCERN ELEMENT REPORT CO 17105-SQN, "PIPE / FITTINGS AS RELATED TO CONSTRUCTION" I.
Subject C6tegory:
Construction (10000)
Subcategory: Mechanical (17100)
Element:
Pipe / Fittings as Related to Construction (17105)
Employee Concerns:
IN-85-211-002, IN-86-282-004, and WI-85-053-012.
The basis for Element Report CO 17105-SQN, Rev. 2 dated March 10, 1987 is Watts Bar Employee Concerns IN-85-211-002, IN-86-282-004, and WI-85-053-012, which state (respectively):
"ERCW line was designed to be stainless steel however it is not."
"Pressure tests were not applied on many vendor NPP-1 ASME code data forms for containment penetrations.
The penetrations were installed and hydro tests were never verified and documented.
Additional information known to QTC, withheld to maintain confidentiality.
No further information may be released.
Construction Dept. Concern.
CI has no further information."
Buried penetrations have vendor welds tnat were not inspected during hydro tests.
Construction Dept. concern.
CI has no further i nf orma tion.
No followup required."
These concerns were evaluated by TVA as potentially nuclear safety-related anc potentially applicable to Sequoyah (generic).
Sur. mary of Issue The problems as defined by TVA are that the ERC'n piping was installed using material other than the design specified stainless steel, and the hynrotest ir.spection cf penetration welds were not verified or documet.ted, l
III. Evaluation TVA performeo a cetailed investigation of Nuclear Safety Review Staff (NSRS) Investigation Reports IN-85-166-WBN and I-65-118 WBN and of appli-cable SON documents to determine if the concerns were applicable to SON.
It was found that the ERCW System was designed to carbon steel pipir.g. but that a portion of the sys. tem was being changed to stainless steel because of flow restrictions aue to accumulation of corrosion products.
TVA a
t
. implemented programs to monitor the performance of the ERCW System, Surveillance Instruction SI-566 ERCW Flow Verification Tests, and to replace carbon steel pipe with stainless steel as required, to ensure design flow rates at the safety-related components in the ERCW System under various postulated plant conditions and system configurations.
Most of the small bore (2 inch or less) lines have been already changed to stainless steel.
TVA determined that Sequoyah had installed some penetrations un which at least one internal process pipe weld was not hydrostatically tested by either the vendor, or by TVA as part of the system hydrotest.
- However, the ASME code applicable to Sequoyah allows substitution of 100 percent radiography of the weld for the hydrostatic test, and the vendur had performed 100 percent radiography of the subject welds.
IV.
Conclusion The NRC inspectors revi'ewed the Element Report and the related ECTG file, discussed the issues with the ECTG evoluetor and Sequoyah personnel, and reviewed related documentation available at Sequoyah which was not in the ECTG file.
The NRC staff agrees with the evaluation, conclusions, and resolution of the concerns covered in Element Report C0 17105-SQN, Rev. 2.
The correc-tive actions to be implemented by TVA are adequate.
The one corrective action defined by TVA as required prior to restart, which concerns seismic analyses for the change from carbon steel to stainless steel pipe, should be verified by NRC inspection prior to restart.
No other action by the NRC is required.
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