ML20213C830
| ML20213C830 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 11/04/1986 |
| From: | Mcdonald J TENNESSEE VALLEY AUTHORITY |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8611100315 | |
| Download: ML20213C830 (77) | |
Text
- _ _ _ _ -
TENNESSEE VALLEY AUTHORITY Watts Bar Nuclear Plant P. O. Box 800 Spring City, Tennessee 37381 November 4,1986 l
1 Nr. Harold Denton Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Dathesda, Maryland 20814
Dear Mr. Denton:
In the Matter of the Application of
)
Docket Nos. 50-327 Tennessee Valley Authority
)
50-328
SUBJECT:
EMPLOYEE CONCERNS TASK GROUP (ECTG) to this letter transmits for NRC's review seven initial element reports addressing Sequoyah Nuclear Plant employee concerns in the operations category.
(These reports have been reviewed and approved by the Senior Review Panel). Please note that corrective actions indicated in some of these reports are preliminary and do not represent the final version of these actions.
To assist you in your review of these reports, Enclosure 2 provides a comparison between the ECTG Writer's Guide recommended report format and the transmitted reports. Enclosure 3 provides a listing of the report numbers and-the related employee concerns.
Please telephone Martha Martin at 615-365-3587 (Watts Bar) if you have any questions.
Very truly yours, TENNESSEE VALLEY AUTHORITY
&AQx.A J. A. Mcdonald Watts Bar Nuclear Plant Site Licensing Manager Enclosures 8611100315 861104 PDR ADOCK 05000327 P
PDR 3 02,6 An Equal Opportunity Employer l
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Nr. Denton cc (Enclosures):
Mr. James H. Taylor. Director Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, D. C.
20555 Nr. B. J. Youngblood, Project Dirceter PWR Project Directorate #4 Division of PWR Licensing - A U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Room 440, 5 Story Phillips Bethesda, Maryland 20814 U.S. Nuclear Regulatory Commission Region II Attn:
Mr. Gary Zech 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323
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'a TENNESSEE VALLEY AUTHORITY I
SEQUOYAH NUCLEAR PLANT EMPLOYEE CONCERNS TASK GROUP OPERATIONS CEG Subcategory: Operations / Operational (310)
Element: Operations Procedures Need Clarification, Rewritten, and Used Report Number:
310.03 SQN Revision 1~
IN-86-055-003 I
Evaluator:
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D. E.' mit Date Reviewed by:
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Rsvisica 1 I.
Operations Procedures Need Clarification, Rewritten, and Used This report contains only one Sequoyah generic concern that addresses a procedure revision / rewriting.
The scope of this investigation was to review the Nuclear Safety Review Staff report I-85-415-WBN (attached) to verify its adequacy and completeness.
t II.
Specific Evaluation Methodology One concern was reviewed and included in this element evaluation.
IN-86-055-003 Hydrazine spill of 300 gallons in containment building.
Implies concern with inadequacies in plant
' operations / procedure adherence / control of valve and systems operation.
The methodology used for this evaluation was to review the NSRS report conducted on this WBN concern. The report recommendations were evaluated for their adequacy and completeness. The report recommendations were compared to the approved SQN procedures. SQN procedures were reviewed to ensure they incorporate the same' type procedural or programmatic steps as the WBN recommendations.
A review of the WBN generically related investigation and procedures was accomplished. Items reviewed are listed in the reference section of this report. Interviews with cognizant personnel were conducted to address the above listed concern.
This report was completed in accordance with the Evaluation Plan for the Operations Concern Evaluation Group and the Operations / Operational Subcategory Evaluation Plan.
i III. Findings Evaluation Results The spill occurred due to improper control of Steam Generator upper and lR1 lower tap-root valves to which temporary tygon tubing was attached.
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The tubing was installed for steam generator level indication while i
normal level indication was not available. The root valves were ll constantly open, instead of being opened only during the time a level reading was obtained. This set the conditions which allowed the tygon tubing to blow off the fitting when a leaky valve pressurized the steam generators. This occurrence was the second of this type within four days.
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Page 1 of 4
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Rsvision 1 III. Findings (Continued)
There were two recommendations identified in NSRS Report I-85-415-WBN:
- 1.
Delayed Recurrence Control Execution, which states that management should emphasize to the plant staff that a recurrence. control program is in place (CAR /DR system) that should be promptly used without hesitation to analyze events to determine root cause and generic applicability and to assure that decisive corrective action is taken to prevent recurrence.
2.
Inadequate Procedural Controls, which states requirements should be clearly established and delineated in writing which provide criteria for the selection, installation, and use of tygon tubing in abnormal configurations for water level measurement.
Specifically, a caution order should be issued to control the root valves to which the tubing is attached.
The above corrective action recommendations have been only partially addressed at SQN, as stated below:
L.
1.
There currently are programs and procedures which allow for assessment of corrective action, root cause, and generic applicability after a problem has been identified. These procedures are AI-12, Adverse Conditions and Corrective Actions; AI-18, Plant Reporting Requirements; SQN-84, Reportable u
jl Occurrences; and SQN-94, 10 CFR 21 Evaluation and Reporting Requirements. The plant reporting requirements instruction has recently been presented to all licensed operators in requalification training, j
I 2.
SQN presently has procedures (FHI-08, SOI-68.18, SOI-74.1C, and j
SI-673) stipulating the use of tygon tubing for reactor coolant j
system level monitoring during RCS filling and draining in Mode 6 e
operation. However, there are no procedures or instructions to control the root valves associated with the tygon tubing configuration allowed by the above procedures. There are no SQN administrative controls that address any other abnormal tygon tubing configuration.
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Conclusion This concern, as addressed by NSRS Report I-85-415-WBN, generically involves root cause assessment of plant problems and procedural controls to assure proper selection, installation, and use of tygon e
tubing.
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Revision 1 Conclusion (continued)
It is important to note that the incident occurred at WBN, not'SQN, and the l
NSRS report was issued to WBN and not SQN.
On that basis, this concern is substantiated for SQN in that there are no procedural. controls for the proper selection, installation, and use of tygon tubing outside of those procedures listed above.
SQN has adequate procedures for root cause assessment and the plant reporting requirement instruction has recently been presented to licensed operators in classroom training and will be presented in the future during requalification training.
SQN Operations Group is planning on performing the corrective actions (R1 identified :in Section VII of this report prior to plant startup.
l However, it is not a restart requirement.
l IV.
Root Cause There is no identifiable SQN root cause because the event addressed by lR1 this concern has not occurred at SQN. This-report has been written i
because of generic applications to SQN from the WBN element report.
l
.V.
Generic Applicability This concern (IN-86-055-003) was identified as generically applicable to SQN from the WBN concern. Because of the fact that there is an identified procedural deficiency at both Watts Bar and Sequoyah Nuclear Plants, this concern is deemed generically applicable to Browns Ferry and Bellefonte Nuclear Plants.
i VI.
References l
1.
NSRS Investigation Report I-85-415-WBN, dated September 19. -
October 9, 1985, subject - Hydrazine Spill (Attached) 2.
AI-12, Adverse Conditions and Corrective Actions, Revision 22 dated April 2, 1986 3.
AI-18 Plant Reporting Requirements, Revision 43, dated ~ March.14, 1986 i
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Page 3 of 4 1
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,, 4-1 Revision 1 VI.
References (Continued) i 4.
SQN-84, Reportable Occurrences, Revision 3, dated January 28, 1984 5.
FHI-08, Recovery from Refueling, Revision 8, dated January 9, 1986 6.
SOI-68.1B, Reactor Coolant System Filling and Venting, Revision 36, dated June 24, 1986 7.
SOI-74.1C, RHR Filling Reactor Cavity for Refueling, Revision 33.' dated i
June 18, 1986 8.
SI-673, ' Reactor Coolant System Level Verification UsEng Sight Glass or Tygon Hose, Revision 1, dated October 12, 1984 VII.
Immediate or Long-Term Corrective Action Sequoyah Nuclear Plant Corrective Action Response:
lR1 4
l 1.
SOI-68.18 and SOI-74.1C and SI-673, concerning use of tygon I
tubing on RCS system during NODE S or 6 operation, is adequate l
since the level is monitored at all times.
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2.
SOI-67-1, ERCW system concerning laying up the CS heat exchanger.
l will be revised to add a note or caution to isolate the tygon l
.i tubing when it is not being monitored.
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1 3.
A CAUTION ORDER will be added to EHC tank to isolate the tygon l
tubing when the level is not being locally monitored.
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4.
A memorandum will be written to Plant Maintenance to cover any l
use of tygon tubing not in Operations instructions such as WRs l
and their instructions.
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TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT EMPLOYEE CONCERNS TASK GROUP OPERATIONS CEG l
Subcategory: Naintenance 4
Element: Clan Control Program Report Number:
308.07 - SQN Revision 1 IN-85-948-001 IN-85-948-002 IN-85-948-003 h
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Evaluator:
F. R. Swearingen
/0////EI F. R Swe.ingen
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Reviewed by:
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/ Date
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Approved by:
W. R. Lagekdren
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'Rsvisicn 1 I.
Clam Control Program Three' concerns, IN-85-948-001, IN-85-948-002, and IN-85-948-003 were evaluated. All three concerns regard clan infestation in the following l
plant water systems: Essential Raw Cooling Water (ERCW), High Pressure lR1 Fire Protection (HPFP), Raw Cooling Water (RCW)..and Raw Service Water' l
(RSW). These concerns are specific to Watts Bar Nuclear Plant (WBN) and l this evaluation is performed at Sequoyah Nuclear Plant (SQN) because of the generic applicability.
The three concerns exact wording is " mussels." Mussels do not cause I
system problems unless salt water is the growth medium as documented by l
I. E.Bulletin 81-03, p. 3. It is thought that the concerned individuals lR1 have mistakenly called asiatic class, mussels, and the evaluation will I
cover the clan control program.
I II.
Specific Evaluation Nethodology l
Concern IN-85-948-001 states:
" Intake pumping station cannot or does not screen out mussels. The mussels found in lines are very small and y
perhaps are hatching. The ERCW line is also clogged with concrete debris. An 8 inch line may have a one and a half inch opening for water flow. The fire protection system will not operate properly due to this. clogging. Example: Six inch F. P. line in Unit 1 "Not Shop" was cut 2-4 years ago, and a one foot length of pipe had enough debris i
to fill a hard hat (713' Elevation behind security).
C.~I. had no further information.
Concern IN-85-948-002 states:
" Pipes to the sprinkler heads in the switch yard are filled with mussels and debris. Examples of past I
clogging are where the four inch diameter header joins the one inch ~
j diameter around every transformer.
C. I. Had no further information."
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i Concern IN-85-948-003 states:
"The flush hose was stopped up with mussels and identified while flushing the system two years ago.
Auxiliary Building, Unit 1, 692' Elevation. This system was F. P. and j
was supposed to be " dry."
C. I. had no further information."
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Reviewed the following documents:
I a.
Above listed employee concerns l
b.
Two SQN Generic Concern Task Force reports; the original and i
Revision 1 i
c.
WBN Employee Concern Task Group (ECTG) draft report
(
d.
e.
Three TVA memorandums f.
The applicable section of the SQN FSAR i
g.
SQN SQR32 l
h.
Technical Standard number TS 08.01.01.14.03 l
1.
Thirteen various SQN sis
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MI 6.24 l
k.
The completed data sheets from three SQN sis Page 1 of 6
Ravision 1 Informally interviewed individuals from: The Site Quality Assurance Organization Planning and Scheduling, Chemical Engineering, Mechanical Maintenance, Engineering Test, and Codes and Standards.
III. Findings Evaluation Results Concern IN-85-948-001 stated that there is concrete debris in the essential raw cooling water (ERCW) pipe. Concrete-lined piping is specific to WBN, and the concern about debris is to be evaluated at WBN lR1 by the Engineering CEG under subcategory 233. SON does not have I
concrete-lined RRCW pipes; therefore, this part of the concern is not generic to SQN and was not evaluated.
The following commitments to NRC were found when reviewing three TVA memorandums (reference 5):
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- 1.
Continuous chlorination of ERCW from May through October j
(reference Sa).
- 2.
RCW, RSW, and HPFP chlorinated continuously for two 3-week periods per year at beginning and end of clan spawning season (reference Sa).
- 3.
Chlorine equipment failures will be evaluated for added flushing j
or shock chlorination (reference Sa).
4.
Annual tests on centrifugal charging pump and safety injection lR1 pumps to verify bearing temperature difference of less than 72 degrees Fahrenheit. These bearing temperatures are indicative of lR1 aclan fouling of the bearing oil cooler (reference 5b).
I 5.
A heat exchanger inspection program will be instituted to prevent lR1 j
blockage by cisms (reference Sc).
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- Primary clam control commitments.
TVA memorandum (reference 5b) stated that class do exist near SQN. The SQN FSAR (Section 9.2.2.6) contains information about organic fouling by lR1 class, and that class will be controlled by chlorination. Chlorination l
during clan spawning season is the clan control method committed to by SQN in the FSAR (Section 9.2.2.6).
lR1 SQM32, Asiatic Clam Control, is a non-PORC (Plant Operations Review lR1 Committee) approved SQN standard practice (not mandatory) which contains l
l most of the commitments; however, there is no discussion of commitments l
3.
SQM32 does not generate any document data sheets or provide IR1 specific detailed instructions. SQM32 contains a requirement to drain and clean the Condenser Circuating Water (CCW) conduits if the water has i
been stagnant for a period of time since chlorination would not be taking
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place.
Page 2 of 6 L..
Rsvision 1 h.'
i Technical Standard TS 08.01.01.14.3? provides the details of chlorination for clam control and appears to be the basis for SI-712 Clam Control.
IR1 However, the technical standard is not a mandatory document that the plant must perform. Thir technical standard contains the same requirement-(paragraph 6.1.3) for CCW conduits that SQM32 contains as noted above.
r SQN sis 712, 692, and 171 were found to contain the control and documentation requirements for clam control (i.e. chlorination).
Problems found during performance of sis 122, 185 and 148 (hydraulic 4
performance and flushing instructions for High Pressure Fire Protection i
[HPFP) yard and building systems) could indicate infestation of clan /:
therefore, these sis should be supplemental sis to the overall clam
' control program.
SQN SI-712 is the document that controls clam infestation by chlorination at the required times and is written to document the control actions.
The SI meets items one and two of the caLa control commitments listed IR1 above; however, there were no requirements in the SI to meet commitment numbers 3, 4, and 5.
The SI was implemented in July 1985, and the lR1 performance records (reference 8) were reviewed. The SI data package reviews indicate SI-712 meets the present requirements that are written IR1 in SI-712.
l The evaluation requirement for chlorine equipment failures (commitment 3) is discussed in SI-712, but there were no details or data sheets to document the comunitment.
SI-692 is to be performed to verify chlorination levels in RRCW dead legs. No completed data sheets could be found in the Document Control Unit, and interviews with appropriate personnel verified that this SI has not been performed.
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IR1 i
SI-171 is performed to flush HPFP systems and to document clam control lR1 (chlorine levels in the HPFP systems) for the 3-week periods (commitment l 2).
Data sheets 171.2 from the SI performance (reference 8) were l
reviewed and found fully adequate for the purpose of documenting l
chlorination in the HPFP.
l SI-679 satisfies commitment 5, which is a heat exchanger inspection lR1 l-program. Reviews of a random semple of completed data sheets from l
ST-679 (reference 8) found no documentation of clam infestation in the heat exchangers. This review found data sheets from this SI to be inadequate QA documents, because there are no signatures and the data is missing. However, the data sheets do contain a space to be completed that documents the existence of class in a heat exchanger.
SI-40 and SI-129 contain requirements for bearing temperature readings on the centrifugal charging pump and the safety injection pump and s
appear to meet commitment 4.
However, there was no discussion, IR1 reference, or note about clam control.
Page 3 of 6
-,-- - - -. ~__--__ -_,-_____--._ __
Ravisicn 1 Reference Sc provides documentation of class at SQN that caused a flow blockage of the containment sp sy heat exchanger (CSHX) 1A during a test of the ERCW system in 1982.
This is the only documentethat t
discussed class existing in the plant systems that this evaluation found. Reference 5b disclosed a discussion about class at SQN (page 2) that strongly documents the fact that there have been virtually no clam infestation in plant safety systems.
Informal interviews with individuals in mechanical maintenance, chemical ensineering, safety and fire protection supported the above finding of only minor problems with class in the fire protection systems. The interview also revealed that numerous pipes have been cut open during modifications with virtue 11y i no class found, and that strainers and baskets removed per MI-6.24 and applicable sis (reference 7) have not revealed major clan infestation.
Conclusions Reviews of all pertinent documents and interviews with personnel indicate that there are class in the vicinity of SQN and clas growths have occurred in some of the water systems; however, the evaluation found no indication of any massive or recurring problem with class in plant water systems. Therefore, concerns IN-85-948-001, IN-85-948-002, and IN-85-948-003 generically evaluated at SQN are found to be not valid.
The plant sis need enhancements bat are adequate to prevent clan infestation. However, no SI or document was found that completely met lR1 commitment 3 that chlorine equipment failures will be evaluated.
SI-712 is adequate for its intended purpose; however, it needs:
(1)
References to all the other sis involved with clan control.
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(2)
Information regarding manual inspection and cleaning of CCW conduits when chlorination has not been possible as presented in l
SQM32 and T.S.08.01.01.14.03.
(3)
.Nore detail end documentation for incidents of chlorine equipment failure.
(4)
Data sheets to track the occurrences of clam infestation during 1
performance of all related sis. SI-679 is an adequate document to record clan infestation in the heat exchangers, but the data sheets reviewed are not adequate QA records (i.e., no sig' natures and all lines are not completed).
SQN is deficient in that SI-692 has not been performed. The performance is important f0r the documentation of commitment 1 to NRC.
IV.
Root Cause Since these concerns were evaluated generically for SQN and no clam infestation problems exist at SQN, no root cause was determined.
Page 4 of 6
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Ravisicn 1 V.
Generic Applicability In the documents that were reviewed, it was found that clan infestation is a potential problem at all TVA nuclear plants. However, the clam l
control program at SQN is specific to SQN and the findings cannot be applied to other TVA plants.
{
4 VI.
References 1.
Employee Concerns IN-85-948-001, IN-85-948-002, and IN-85-948-003 i
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'2.
Element Report: SQN Generic Concern Task Force; April 25, 1986, R1; " Mussels and Construction Debris Clogging ERCW and Fire Protection Flow Paths" 3.
Element Report: WBN Employee Concerns Task Group - No. 308.07;
" Clan Control Program"; Draft dated April 21, 1986 4.
IE Bulletin 81-03 (enclosed with letter fram James P. O'Reilly to H. G. Parris dated April 10, 1981)
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5.
a.
Letter to Mr. O'Reilly, USNRC, dated March 21, 1983, from D. S. Easumer, TVA, " Flow Blockage of Cooling Water to Safety Components by Corbicula" (A27 830321 019) b.
Letter to Mr. O'Reilly from L. N. Mills dated Nay 26, 1981, "I. E. Bulletin 81 Sequoyah Nuclear Plant" (A27 810526 028) c.
' Letter to E. Adenson, USNRC, from Mr. Mills dated September 7, 1982, " Docket Nos. 50-327, and 50-328" (A27 820907 034) 6.
SQN EIs: 171 RO, 692 RO, 712 R2, 122 R4, 185 RS, 148 R4, 40 R35, 129 R27', 704 R2, 679 R2, 181,3 RO, 181.4 RO, and 181.5 RO 8
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SI-171 SI-712 SI -679 f
171.2 A - 05/09/85 Data Sheet 1 03/84 171.2 F - 05/09/85 08/05/85 to 09/29/85 10/84 171.2 J - 05/09/85 10/21/85 to 11/18/85 02/85 171.3 04/17/86 to 05/26/86 09/27/85 171.1 y 10/20/83 10/01/85 a
10/10/85 01/86 i
s T
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R2 vision 1 9.
SQN NI-6.24 R1 10.
SQN 32 R1 11.
T. S. 08.01.01.14.03 RO VII.
Immediate or Long-Term Corrective Action The Clan Control Program summarized in SQN 32 and implemented by numerous surveillance test instructions appears to need numerous enhancements and controls.
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TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT EHPLOYEE CONCERNS TASK GROUP OPERATIONS CEG Subcategory: Health Physics Element: Health Pl.ysics Policies. Practices, and Management Control Report Number: 311.04 - SQN Revision 1 Concerns: SQP-86-009-001 X1-85-063-001 SQP-86-009-002 X1-85-028-102 11-85-084-001 II-85-028-103, 11-85-066-001 11-85-098-002 II-85-009-002 I-86-238-SQN WI-85-038-001 JLH-86-003 II-85-015-001 JNA-85-001 11-85-026-001 RII-85-A-0064-Evaluator:
D. C. Hall Jr.
/c 26 D. C. Hall Jr.
Date T. L. Reese
/c-/1-f6 T. L. Reese Date R. L. Huskin
/#-/7-f6 R. L. Huskin Data
/
D. L. Lovett
/c-/t -Ek_
D. L. Lovett Date Reviewed by: Audd&rneln/did) le))1l%
OPS CEG ember Dat'e Approved by:
. 4% ' M 10-//-$d W. R. Lagergreji Date 1872T
k Revision 1
-L
Title:
Health Physics Policies, Practices, and Management Control -
311.04-SQN The scope of the 311.04-SQN evaluation consisted of the investigation of 16 concerns. The concerns involved the following areas of the Health Physics (HP) program:
1.
Personnel contamination (Concern SQP-85-009-001) i f
2.
HP response to radiatics/ contamination alarms or indications of abnormal radiological conditions (Concerns 11-85-084-001 and XI-85-066-001) 3.
Distribution of personnel radiation doses (Concern 11-85-009-002) 4.
Containment "at power" entries (Concerns SQP-86-009-002, WI-85-038-001 and 11-85-015-001) 3 5.
Management support of HP programs (Concern 11-85-026-001) 6.
Verification of system contents (Concern 11-85-063-001) 7.
Radiation Work Permit (RWP) procedures (Concerns II-85-028-IO2 and II-85-028-IO3) l 8.
_ Radiological Survey frequency (Concern 11-85-098-002) 9.
C-Zone Emergency Procedures (Concern I-86-238-SQN) 10.
Auxiliary Building Secondary Containment Enclosure (ABSCE) breaches (airborne radioactivity concern) (concern JNA-85-001) 11.
Frisker Locations (Concern JLH-86-003) l 12.
Adequacy of the SQN HP program in general (Concern RII-85-A-0064) l l
II.
Specific Evaluation Nethodology i
1.
Concern SQP-86-009-001 states: An incident at Sequoyah Nuclear Plant which resulted in employees being radioactively contaminated I
could have been prevented and reflects managements attitude toward radiation safety and personal safety of the employees.
lR1 i
1 2.
. Concern SQP-86-009-002 states: The transfer of responsibility for HP from Muscle Shoals to Sequoyah places the individual responsible for HP in a position where much pressure from plant management can be exerted and has caused compromises of previously established HP policy regarding personnel access during unit operation.
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Revision 1 3.
Concern XX-85-084-001 states: Questionable practices by HP' at Sequoyah in 1982 led to possible overexposures.
HP would respond to radiation alarms and unplug units.
lR1 4.
Concern XX-85-066-001 states: Sequoyah: Three years ago HP at Sequoyah was notified of higher-than-expected radiation levels ~in the Reactor Building. When notified by telephone, HP personnel speculated on the reasons for the high radiation level, and did not respond immediately to investigate. CI feels that wasting time speculating on cause and not responding immediately is a concern for safety.
lR1 5.
Concern XX-85-009-002 states: Sequoyah: There is no regard for personnel safety at operating plants. Management (known) directed that the oldest employees be assigned to " hot" work in order for them to reach their radiation levels first. A supervisor (known) made the statement that " older folks won't be long around."
lR1 6.
Concern IX-85-028-X02 states: Sequoyah:
RWP 02-2-00214 (sign-in sheet) contains falsified signatures.
IR1 7.
Concern IX-85-028-X03 states: Sequoyah:
RWPs are not being completed according to procedure requirements.
RWP 02-2-00214 is an example.
IRl 8.
Concern 11-85-098-002 states: Sequoyah:
Radiation areas are not monitored often enough.
9.
Concern I-86-238-SQN states: An anonymous individual mailed in a safety concern to (NSRS) requesting that emergency procedures be written to encompass all aspects of possible emergency situations in a C-Zone. Procedures should cover specific areas such as spread of contamination, possibility of injury, possibility of a fire, possibility of poor breathing ataosphere, etc.
lR1 10.
Concern JLH-86-003 states: According to TVA's General Employee Training (GET) classes and plant procedures, employees are to be frisked as soon as exiting a "C-Zone."
Currently, an employee has to search for a frisker. In the process of looking for a l
frisker, an employee can contaminate doors and/or the floor.
One of TVA's objectives is to keep down contamination, and the current process does not adequately control the spreading of contamination.
Example: When exiting pipa chase on elevations 690 and 669, one has to pass through closed doors to get to a frisker. On elevation 669 an employee has to hunt for a frisker as evidenced on i
December 12, 1985.
lR1 Page 2 of 36 d
Revision 1 11.
Cor, ern JMA-85-001 states: A high risk possibility of not securing ABSCE type breaches if a valid high-radiation condition occurs in the Auxiliary Building or during an announced evacuation or evacuation alarm sounded may cause persons to leave the Auxiliary Building before sealing penetration.
IR1 12.
Concern WI-85-038-001 states: Watts Bar Nuclear Planti The-practice of persons entering the lower contaminated area of the reactor containment for nonemergency repairs while the reactor is operating should be reevaluated.
Recent studies indicate the biological effects of personnel exposure to neutron flux are more serious than previously believed. This practice is in effect at Sequoyah and resulted in an accident around 1983/1984
~
and-is planned to be implemented at Watts Bar.
IR1 13.
Concern XI-85-015-001 states: Sequoyah: The practice of
-personnel entering the lower containment area of the reactor containment for nonemergency repairs while the reactor is operating should be reevaluated since recent studies indicate the biological effects of personnel exposure to neutron flux are more serious than previously believed. This practice caused an accident in the incore instrument probe room at Sequoyah in 1985 and is still continued.
lR1 14.
Concern 11-85-026-001 states: Sequoyah: Inadequate upper management support provided the HP department to enforce an effective radiological safety program. No disciplinary action is taken when employees intentionally bypass monitors.
lR1 15.
Concern 1K-85-063-001 states: Sequoyah Operators and Health
. Physics: Failure to know and verify the contents of a system.
Example: HP gave go ahead to open a line in the unit 2 Turbine Building, saying everything was okay and clean.
After opening the line the next night, the~ entire area was roped L.
off for contamination. This occurred in January / February 1984.
IR1 16.
Concern RII-85-A-0064 states: This allegation expressed concerns 1
about the Sequoyah HP program. The concerns are summarized below:
1.
TVA does not have the ability to run an HP operation.
2.
An individual lost a radioactive source at the site and never reported the loss to management.-
3.
The location of radiation monitors are not as indicated on the ASIL-3 procedure.
4.
Smears are taken.into the HP office to count and are then thrown into the trash.
Page 3 of 36
Revision 1 5.
The smear counting area in the HP office was contaminated.
This " contaminated area" was used as an eating area.
6.
Air samples are taken improperly, e.
g.,
floor level.
Respirators were not worn by workers in high contamination areas (areas with surface contamination greater than ten thousand dpm).
7.
The individual claims he was dismissed from employment as a result of a conspiracy and that he was not treated fairly i
i during his training period.
(This item is being handled solely by the Intimidation and Harassment Category.)
f 8.
HP technician did not cover'the head and filters of air sampling monitors before and after exiting areas to be monitored.
Closure of this matter should involve an evaluation of the HP 4
program and practices to include air sampling program, respiratory protection program, and training program.
Implementation and compliance with written procedures should be assessed.
lR1 This report was prepared in accordance.with the Operations lR1 Concern Evaluation Group (Ops. CEG) evaluation plan and the l
}
Health-Physics subcategory evaluation plan.
l l
All-K-forms, previous NSRS line management, and ERT reports assigne J to element 311.04-SQN were evaluated. The evaluations were performed by IR1 four evaluators and consisted of investigations of all open1 item l
l concerns, evaluations and verifications of previous reports, responnes, I
l and investigations of closed 2 item concerns, interviews with l
cognizant personnel, and reviews of applicable regulations and j'
governing procedures. The specific items reviewed for each element are i
identified in the findings of that concern. All previous investigations and reports were assessed for the adequacy of the r
methodology, findings, and recommendations. Also, all respective corrective actions are verified completed or working.
L l
With the exception of X1-85-028-102 and item 7 of RII-85-A-0064, all
!R1 of the concerns are assigned solely to the Operations CEG.
l Note: 1 "open" item denotes-no previous investigation (s) were performed.
IR1 2 " closed" item denotes previous investigations were performed.
l l
Page 4 of 36 i
Revision 1 Item 7 of concern RII-85-A-0064 raises a question of potential intimidation and harassment in that the CI states he was terminated as the result of a conspiracy and treated unfairly during his in-plant HP training. This item will be evaluated solely by the Office of lR1 Inspector General. Concern XX-85-028-X02 raises allegations of I
document falsification and is, therefore, also a shared concern with I
the Office of Inspector General.
l III. Findings 1.
SQP-86-009-001 raises a concern that personnel at Sequoyah were contaminated and that the incident, which was preventable, reflected poor management attitudes regarding radiological health and safety. No information detailing the incident was available; therefore it is not known when the incident occurred, the area of the plant in which the incident occurred, the activity in progress which caused the incident, the number of persons contaminated, whether or not internal contamination was involved, nor the extent of the contamination. The evaluation consisted of 2 parts. Part 1 is an evaluation of plant procedures intended to prevent both internal and external radioactive contamination of personnel.
Part 2 is an evaluation of plant procedures regarding action taken when plant personnel.become contaminated, including corrective action taken to prevent recurrence.
IR1 A.
Part 1 - Prevention of Personnel Contamination lR1 10 CFR 20 establishes general requirements for protection of personnel in restricted areas against exposure to licensed radioactive materials. These requirements include limits on concentrations of radionuclides in air with regard to internal exposure, requirements for handling radioactive materials with regard to external exposure, and survey requirements pertinent to both internal and external exposure. In addition, U.S. NRC Regulatory Guides 8.15 (Acceptable Programs for Respiratory Protection) and 8.8 (Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Reasonably Achievable) establish guidelines for protecting personnel from both internal and external contamination hazards. Additional guidance and/or requirements are provided by 30 CFR Part 11. " Respiratory Protective Devices," and NUREG-0041, Manual of Respiratory Protection Against Airborne Radioactive Materials.
I l
Page 5 of 36
~
Revision 1 The evaluation included a review of both TVA-wide and Sequoyah-specific implementing procedures. TVA CODE VIII, OCCUPATIONAL RADIATION PROTECTION, establishes the general requirements for the radiation protection program.
The TVA Radiation Protection Plan defines more specific requirements applicable to all TVA nuclear facilities, including requirements for airborne radiological assessment and protection programs, protective clothing requicements, survey requirements, and radiolor'tal incident and personnel contamination reporting requiremerits.
At Sequoyah, the primary radiological control program implementing procedures are the Radiological Control Instructions (RCIs). The RCIs establish Seneral limits and guidelines governing the radiological protection program.
Detailed instructions which implement the RCIs within the HP Section are the ASILs, DSILs and HPSILs (section instruction letters). All TVA and Sequoyah procedures daaling with personnel contamination were reviewed and determined to be in compliance with regulatory requirements. Personnel contamination control programs are described by way of SQN RCI-1, RCI-3, RCI-4, and RCI-11.
In addition, RCI-14 describes the RWP program with regard to prescribing protective requirements for workers. Sequoyah HP-SILs 2, 3, 5, 7, 8, and 10 provide detailed instructions regarding both internal and external personnel contamination control programs, including respiratory protection and bicassay programs.
B.
Part 2 - Personnel Contamination Incidents lR1 The evaluation of HP practices following incidents of personnel contamination were examined. HP-SIL 10 establishes procedures to brs followed in the event of personnel contamination, both external and internal. This includes procedures for decontamination, reporting, and corrective action. -In addition, actual records of personnel contamination were examined.
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I i
i Page 6 of 36 i
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Revision 1 Sequoyah HP divides the incidents into 2 categories, reportable and nonreportable. Nonceportable incidents require-a Personnel Contamination Report, form TVA'17093, and are considered incidents which occurred because of unforseeable circumstances such as a punctured glove or torn protective 4
clothing. Reportable incidents require, in addition to a Personnel Contamination Report, a Radiological Incident Report, (RIR) form TVA 17143, and are considered incidents which were preventable and caused by a failure to follow prescribed procedures. Examinations of the reportable and nonceportable summary files revealed that since 1984 there i
have been 180 reportable incidents of personnel contamination and approximately 400-500 nonceportable incidents (not counted).
Both the Personnel Contamination Reports and RIRs require review by applicable HP and plant management, and they require recommended corrective action. It was noted that the number.
j of reportable incidents has declined, year to year, since 1984.
2.
Concern SQP-86-009-002 was evaluated with regard to the technical I
aspects and potential consequences of the alleged circumventing of I
HP personnel access requirements. Since the concern referred to I
access requirement for personnel during plant operation, it was I
determined that this reference pertained only to containment l
entries. The investigation, therefore, centered on containment lR1 entries, practices, and governing procedures, both past and present. l to. determine if indeed HP requirements had been detrimentally l
altered as a result of the referenced reorganization.
l
. A review of several TVA forms 9880 Employee Status and Information Record, for employees involved in the transfer of HP responsibilities from Muscle Shoals to_the Division of Nuclear Power identified June 1, 1982, as the effective date of transfer.
I Interviews with several members of Sequoyah HP management revealed that plant-level, PORC-approved, instructions for Reactor Building entry are contained in SQN AI-8, " Access to Containment." No i
specific HP instruction exists covering the same topic; however, certain hazards and/or conditions typically found inside the Reactor Building are addressed in several HP instructions.
A review of SQN-AI-8 (revision 17) and all of its prior revisions (revisicn 0 first approved January 26, 1977) revealed no significant changes in entry limitations or requirements during or i
after the transfer of authority in question.
4 4
+
4 Page 7 of 36
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Revision 1 Interviews with several members of Sequoyah HP management indicated specific guidelines for Reactor Building entry had not been changed to any great extent during the past 4 years with the possible exception that past practice had been to lower reactor.
level to approximately 30 percent of full reactor power before entry. Radiation surveys taken at 30 percent and 100 percent indicated no significant increase in man-rem if the scope of work was limited. Based on these findings, subsequent necessary entries have been made at power levels greater than 50-percent power. No plant instructions could be found supporting either the 30-percent or the greater-than-50-percent guidelines.
3.
Concern XX-85-084-001 was previously investigated by NSRS Report I-85-806-SQN. Findings of the NSRS report are as follows:
(Designations for individuals have been extracted directly from the NSRS report.)
A.
Based upon interviews with Public Safety Officers (individuals B, C, D, E, and F), no information was obtained that HPs failed to properly respond to radiation alarms (portal monitors, hand / foot monitors, or friskers).
B.
Individuals B. C, D and E stated that they had observed an RN-14 frisker alarming at the 690-foot elevation containment 4
air lock because of noble gases or other causes of high background. At one time, the frisker had read as high as 5,000 dpa. When the HP arrived and confirmed the radiation level, the public safety officer post and frisker would normally be moved to an area of lower background. When the radiation levels were not confirmed, the frisker was replaced if it continued to alarm.
C.
Individuals B, C, D, E, and F stated that the hand / foot monitors at the 690-foot elevation access point from the Turbine Building to the Auxiliary Building frequently went off. Both the hand / foot monitors and the portal monitor would i
alarm because of high background from trash, tools, or laundry in the area. The HPs would respond to these alarms and move the material causing the high background away from the
{
monitors. These individuals could not recall any cases where l
the monitors were unplugged or turned off when alarming to true radiation levels; if one hand / foot monitor was unplugged 1
or turned off because of instrument malfunction, the adjacent hand / foot monitor remained operative.
D.
Individuals B, C, D, E, and F could recall no instances where the hand / foot monitor or portal monitor from the refuel floor to the Control Building had been turned off or unplugged when alarming to a confirmed radiation level.
Page 8 of 36
Revision 1 E.
Individuals B, C, D. E, F, and G could recall no instances when both the hand / foot monitor and the portal monitor were out of service and a frisker was not then ured to check for I
i personnel contamination. No instances were recalled when the exit from the regulated area was left unmonitored.
F.
Individual H stated that entries into the Containment Building during plant operation allowed the transfer of small amounts of noble gas through the airlock. With the sensitivity'of the RN-14 frisker to very small increases in background, the noble gaser would frequently cause the frisker to alarm, thus t
requiring the relocation of the frisker station.
G.
No one interviewed stated that HP had zeroed their pocket chambers without recording the dose. However, an HP technician from the time period of concern (individual A) stated that on occasion he had zeroed a pocket chamber without recording the dose in the presence of the individual. Based upon the work an HP was doing when requested to read and zero a pocket chamber, past practices had included an occasional delay in recording the information.
Reading the dose and recognizing the individual would allow the HP to defer recording this information (SSN information was available in the HP laboratory). However, the current requirements of DSILs (reference 8) make this practice unlikely in that more information, including pocket chamber serial number, is now requ! red to be recorded. Regardless of any delays in recording pocket chamber dose or failure to record that dose, the official record of exposure would be unaffected since it is based upon thermoluminescent dosimetry (TLD).
Conclusions of the report are stated below, as well as the results j
of the document search.
Concern XX-85-084-001 was not validated. Based upon the lR1 statements of the CI, the concern involved multiple events that would have represented general HP practices that should have l
been readily observed by other individuals. However, NSRS could find no evidence from the randomly selected individuals interviewed that such practices existed.
A review of applicable documentation supports the findings of the-NSRS report I-85-806-SQN.
It was noted that Area Plan 3 (references 2 and 3 of the NSRS report) has been cancelled and superseded by the Radiation Protection Plan. Since all copies of the Area Plan (Radiation Protection Manual, Area Plan 3) were l
returned to the Distribution Center Clerk, LP 45164 D-C, it was not available for eeview; however, this did not affect the NSRS findings and conclusions.
Page 9 of 36
~. -
i
l Revision'l 4.
Concern XX-85-066-001 was previously investigated by Sequoyah line management in report XX-85-066-001 (reference 27), and involved the perception by the CI that because HP did not respond immediately to radiation alarms or unknown situations, the radiological safety of plant personnel could be compromised.
The Sequoyah Line Response report was reviewed for adequacy and determined to fully address the scope of the concern. Therefore, no follow-up was determined to be necessary.
Findings of the line report are as follows:
A.
Sequoyah has not exporienced abnormal radiation levels during periods of operation.
B.
The only event that resulted in unanticipated radiation levels in the Reactor Building was the thimble tube ejection in April 1984. HP was present at the beginning of the event and maintained control throughout the recovery process.
C.
Follow-up conversations with Quality-Technology Corporation (QTC) regarding additional information yielded only that unit 1 was operational and the alarm was in upper containment. No specific dates or persons contacted could be provided.
D.
HP supervisors cannot recall any instance that would coincide with the employee concern.
5.
Concern II-85-009-002 was previously investigated by NSRS
.(reference 51). It should be noted that the NSRS report also addresses concern II-85-009-001. II-85-009-001 was a concern which was retracted'by QTC when the CI indicated it contained inaccurate information as worded. The concern was reworded and reissued as XX-85-009-002. The concern involves an allegation by the CI that " hot" (high radiation area) work was assigned to older employees first, as directed by plant management. The NSRS investigation and review of radiation exposure records found no evidence that older individuals working at Sequoyah had received j'
disproportionately high levels of exposure when compared to other workers in their sections or organizations. The NSRS report was reviewed and determined to fully address the scope of the concern; therefore, it was determined that no additional investigative action or follow-ups were required. The NSRS findings are stated as follows:
i Page 10 of 36
+
Revision 1 A.
A review of radiation exposure records of 179 craft workers and foremen assigned to Sequoyah during the period from October 1979 to March 1981 revealed that none of them had received a dose which would have prevented or restricted their work in regulated areas. A review of doses for subsequent periods for these same individuals indicated that one individual had received a quarterly exposure above the currently imposed 70-percent administrative limit, thus influencing the work assignments made by the supervisor but not limiting the employment'of the individual.
B.
Sequoyah exposure records were reviewed for the period of January 1980 to June 1985 to determine if any personnel had exceeded 70 percent of either quarterly limits or annual limits. Thirty-six individuals exceeded a quarterly dose of 2.1 ren or an annual dose of 2.8 rem. Of the 20 TVA lR1 employees, 10 were craft engineers / technicians and 10 were craft personnel. Of the 10 craft personnel, 6 were currently employed at Sequoyah. A comparison of the employment records and exposure records of the other 4 individuals who had exceeded the 70-percent administrative limit revealed the following:
1.
One craft employee exceeded 70 percent of his quarterly lR1 i
exposure limit in the period January through March 1984.
He was terminated at the end of his temporary appointment on April 13, 1984--into the next quarter for exposure limits. There was no indication that the employee's termination was affected by his exposure at Sequoyah.
2..
Another craft employee exceeded 90 percent of his annual lR1 limit in 1983. However, his temporary appointment at Sequoyah was terminated in February 1983, with a first quarter dose at Sequoyah less than 70 percent of the quarterly limit. There was no indication that the employee's termination was affected by his exposure lR1 at Sequoyah.
l l
3.
A third employee exceeded 90 percent of his annual limit lR1 in 1984 and resigned at Sequoyah to accept other employment. The employee had been previously employed in 1984 at Browns Ferry Nuclear Plant (BFN) and subsequently returned to BFN during 1984. He remained a TVA employee into the second calendar quarter of 1985.
Almost all of his 1984 dose was received at BFN. There was no indication that this employee's resignation from Sequoyah was affected by his radiation exposure.
l l
Page 11 of 36
Revision 1 4.
A fourth craft employee exceeded 90 percent of his annual lR1 limit in 1983 and resigned at Sequoyah to accept other employment. The employee left Sequoyah during the first quarter of 1983 and had received less than 70 percent of the quarterly dose at that time. Although-the employee subsequently received radiation exposure in 1983, there was no indication that the employee's resignation was affected by his exposure.
1 C.
Based upon the exposure record of 179 craft personnel for the l
period October 1979 to March 1981, no pattern of selection of l
personnel for hot work based upon age was found in any of.the craft sections.
D.
Based upon an interview with the first craft employee, plant IR1 management had discussed, in the 1979-1980 time period, options that could be taken if employees approached the quarterly or annual dose limits established by RCI-1.
No information was received from the employee or the craft supervisor (the second lR1 craft employee) of that timeframe that any direction was l
provided to preferentially expose older workers.
E.
The supervisor who was alleged to have made the statement that
" older folks won't be long around" is no longer a TVA employee, could not be located from his last known address, and thus could not be interviewed.
F.
An individual who was craft foreman from the 1980 time period lR1 was unaware of any " management direction" regarding the assignment of personnel to " hot work" based upon age.
Conclusions from the NSRS report are as follows:
Concern (IX-85-009-002) was not validated. NSRS could find no lR1 objective evidence that Sequoyah management told supervisors in the 1980 timeframe to assign older personnel to work in high radiation areas (" hot work"). There is no evidence that older personnel ~were l
preferentially assigned " hot work." During the period in question, no individual received a dose high enough to require any consideration of work restrictions, even using the more conservative
[
TVA~ policy exposure limits.
Page 12 of 36
Revision 1 6.
Concerns WI-85-038-001 and XX-85-015-001 raise questions about lR1 personnel exposure-to neutron radiation during containment entries, specifically lower containment, while the reactor is at power (critical). Concern XX-85-015-001 was previously investigated at Sequoyah in a Sequoyah Line Response report (reference No. 54).
lR1 Concern WI-85-038-001 is an identical restatement of.XX-85-015-001 except that it is directed at Watts Bar. A review of the two concerns and the Sequoyah Line Response resulted in the determination that the line response adequately addresses both IR1 concerns; therefore, both concerns are eddressed as a single I
concern.
l Findings from the Sequoyah Line Response are as follows:
A.
Nazimum neutron dose (arem) for an individual was 190 and 210 in 1983 and 1984, respectively.
B.
Nazimum gamma dose (aren) for an individual was 3,110 and 3,360 in 1983 and 1984, respectively.
C.
Average neutron dose (ares) was 21 and 24 in 1983 and 1984, respectively, as compared with average gamma dose (arem) of 259 and 451.
D.
Neutron dose is typically a factor of 10 less than gamma dose.
E.
Quality factor (factor used to convert an exposure to radiation into dose to humans) of 10 for neutrons is accepted 4
by all scientific and rulemaking bodies.
.F.
Some recent literature publications suggest that quality factor be increased by about a factor of 2.
G.
Nearly all utilities enter containment for repairs and maintenance at power.
H. -Entry into' containment at power was not the direct cause of the thimble tube ejection incident.
Conclusions from the report are summarized below:
1 A.
Even if quality factor increased by a factor of 5, the effect from neutrons would still be of less concern than gamma i
radiation.
B.
Entry into containment at power is acceptable from a dose i
standpoint.
Recommendations from the Sequoyah Line Response are as follows:
Sequoyah HP and Site Services Branch will continue to monitor quality factor discussions and recommend changes accordingly.
i l
Page 13 of 36 i
Revision 1 This evaluation concurred with the findings of the Sequoyah Line Response report. A review of supporting documents justified the findings of the line response report, specifically in the area of neutron exposure quality factors. It was found in one journal report of recent publication (reference 30) that quality factors for neutrons range from 3.43 to 13.4 depending upon neutron energies.
ItLwas also found that a quality factor increase of a factor of 5, reference Sequoyah Line Response, based upon the 1983 and 1984 average neutron exposures reported, would not exceed the average gamma exposures and that the total gamma component of the overall exposure would still be the most limiting criterion for lR1 exposure.
It should also be noted that Sequoyah, as well as all TVA nuclear facilities, use the quality factor required by 10 CFR 20.4(c)(3) in determining neutron dose.
A review of the NSRS report I-84-012-SQN (reference 31) did not indicate that the thimble tube ejection, the accident at Sequoyah referred to in the concerns, was a direct result of entry into containment while at power.
7.
Concern XX-85-026-001 alleges that Sequoyah HP receives inadequate upper management support in enforcing the radiological safety program. Also, the CI states that no disciplinary action is taken when employees intentionally bypassed monitors. The concern was previously evaluated at Sequoyah by line management _in an Sequoyah Line Management Response report (reference 32). The report was reviewed for adequacy and determined to fully address the scope of the concern.
A follow-up interview was conducted to determine the status of the
. reports corrective. action recommendations.
Findings and recommendations of the line management report are summarized below:
No actual incidents were identified in the investigation where lR1 employees did not receive disciplinary action for deliberately I
bypassing radiation monitors.
I Interviews with HP personnel and reviews of plant procedures and records did not indicate inadequate upper management support to enforce an effective radiological safety program. The plant superintendent is immediately notified of all RIRs that have been designated as major by HP.
RIRs are then sent to the employee's supervisor for appropriate corrective action. Afterwards the plant superintendent or designee reviews the action taken. If he perceives the action to be inappropriate, he sends the RIR back to the supervisor for appropriate action.
Page 14 of 36
Revision 1 There were some instances where processing the RIRs took too long.
This is very ineffective when the employee is a temporary hire and has left by the time the RIR is processed.
In some cases, the person initiating the RIR did not receive feedback as to the disposition of the RIR.
lR1 The recommendation from the report is that a summary of RIRs will-be sent to all HP technicians for their information and those RIRs still active will be discussed with plant managers at the managers' meetings to ensure prompt action.
An interview was conducted with a supervisor in the HP Section to determine the status of the corrective action recommendation.
Based upon the results of the interview, it was determined that the corrective action has not yet been implemented. The HP representative stated that summaries of RIRs were not distributed to HP technicians nor were they currently being discussed in plant managers meetings.- The individual also stated that a procedure revision will be implemented that will specify disciplinary actions to be taken with a RIR.
3.
Concern 11-85-063-001 involves the perception by the CI that HP and Operations personnel may fail to know and verify system contents before authorizing the breaching of the system. The concern was previously investigated by NSRS (reference 52). A review of the investigation and report determined that the scope of the concern ~was fully addressed by NSRS and that further evaluation was unnecessary.
Findings of the NSRS report are as follows:
(Designations for individuals have been extracted directly from the NSRS repert.)
A.
Modifications personnel (individuals A and B) and HP personnel (individuals C and D) provided suggestions that any contamination in the Turbine Building, elevation 662.5 (under the condenser), would probably have been from work in the steam generator blowdown (SGBD) system. However, individual B could find no record of any unit 2 blowdown lines that had i
been breached with water in them during the months noted in the employee concern.
B.
Individual E atated that work had been done on the SGBD system (time period not remembered) involving the installation of two 4-inch valves which had required the draining of the associated piping up to a boundary valve.
He stated that there had been some leakage pcst the boundary valve and that the area had been roped off as a contamination zone as a precaution.
l Page 15 of 36 l
e Revision 1 C.
Individual E stated that when the SGBD system was cut into on the 685-feet level (adjacent to the flash tank), the workers had been dressed out as a precautionary measure. Once HP had surveyed the inside of the pipe, the area was declared clean and protective clothing requirements were removed.
D.
Based on HP surveys of the Turbine Building, elevation 662.5, unit 2, the only contamination area identified during the January-February 1984 period was on the SGBD pumps.
RWP 02-2-00925 timesheets 0001 and 0002 indicated general cleanup / decontamination of these areas at a time before 1400 on two days. This contamination area did not coincide with the concern of record because:
1.
These contamination areas were not established coincident with any work on the nearby SGBD piping.
2.
The timing of the decontamination on the RWPs was such that the CI would not have observed the decontamination process when he reported to work the "next night."
E.
Surveys of the unit 2 Turbine Building area during the January-February 1984 period showed that some areas around the SGBD system had been zoned as a regulated area because of radioactive material in the piping system as a result of primary-to-secondary leaks.
Two modif' cations to the SGBD system in the 1983-1984 period i
F.
were identified by RWPs in which radioactive /potentially radioactive piping was breached. However, as detailed below, neither of the cases fit the description provided by the CI.
1.
Work Plan 10476 required the draining and flushing of the steam generator blowdown lines to accomplish the tie-in of 4-inch lines. Although the work was performed in September 1983, details were compared with the event described by the CI to provide an indication of how HP imposed protective requirements and general practices.
In this work, the following sequence occurred:
a.
The drain valve on each SGBD pump was used as a sample point before draining. A lab coat, gloves, booties and shoe covers, and surgeon's cap were required.
Page 16 of 36
Revision 1 b.
HP coverage was required when draining the system.
Based upon the survey referenced in the RWP, the drain and flush operation was conducted in the immediate area of the SGBD pumps. The area around the SGBD pumps had previously been zoned as contaminated.
Coveralls, taped gloves, taped bootles and shoe-covers, and a surgeon's cap were required,
- c. -No evidence was found that the draining operation increased the level of contamination in the work area.
d.
The SGBD piping was subsequently cut, welding in 4-inch lines and associated valves. Protective requirements included coveralls, plastic suit, gloves, booties and overshoes, canvas hood, and full face
. mask. The plastic suit, hood, and facemask were 4
required only while breaching the system.
l 2.
WP 11021 cut into the SGBD system piping on the 685-foot level, This work was done in August of 1984. The following sequence indicates HP practices in that timeframe.
a.
Special instructions required continuous HP coverage and a requirement to contain all water.
b.
Protective requirements included continuous HP coverage and a requirement to contain all water.
G.
Modifications personnel (individuals A, B. E, and F) had no negative statements about the adequacy of HP personnel knowledge of plant systems. Individuals A, E, and F stated that the HP technicians establish conservative protective requirements; at times, they believed excessive protection was required.
if. A Modifications supervisor (individual A) stated that he considered Modifications personnel responsible for determining the contamination sample points before breaching a system and for understanding what contamination may be in the system and the potential leakage paths. He considered HP to be responsible only for performing surveys and setting protective requirements.
Page 17 of 36
Revision 1 An HP supervisor (individual G) considered HP personnel responsible for identifying potential contamination problem areas. Neither modifications nor HP personnel considered Operations personnel responsible for informing craft personnel of the contents of a system before breaching that system.
Conclusions of the NSRS report are stated below:
4
-Concern XX-85-063-001 was not validated. No evidence was found lR1 that an event occurred.as described by the CI.
Potentially contaminated systems in the Turbine Building had been breached on other occasions leading to scenarios similar to that described by the CI.
In these cases, the HP personnel treated'these systems as potentially contaminated conducting surveys, and ;;equiring protective clothing until the areas were declared' clean. No evidence was found to corroborate the opinion thrit Operations and HP personnel do not provide adequate information or verify system
]
contents.
9.
Concerns II-85-028-X02 and 11-85-028-X03 relate to the CI's l
perception that RWPs are not maintained in accordance with I
procedures and RWP timesheets contain falsified signatures.
IR1
~
A similar concern, I1-85-028-001, was evaluated in the
.I Operations CEG report 311.03-SQN. This report contains an l
evaluation of a QTC report regarding RWP timesheets and is considered pertinent to this report. The concerns were previously evaluated in NSRS report I-85-514-SQN.
+
The findings and recommendations of the NSRS rpport are summarized as follows:
I-85-514-SQN Revision to HPSIL-7 to Defit.e Worker Signature Transfer Requirements The RWPs provide a unique opportunity for Jncorrect entries which may not be discovered until after the worker is no longer available to correct his documentation. Although the NQAM and AI-7 provide overall guidance on the correction of quality assurance-records HPSIL-7 provides no additional guidance on correction of RWP entries. Corrections have been made to the RWPs-l without any traceability to the original documentation. Thus, it cannot be conclusively demonstrated that the employees had made the data entries as required by HPSIL-7.
t i
l-l 1
Page 18 of 36 i
l
Revision 1-
' Recommendation HPSIL-7 should be revised to clearly define the requirements for transcription of information between RWPs I-85-514-SQN Traceability for Transcribed RWPs 02-2-00214 and IR1 02-2-00250 l
RWP 02-2-00214, Timesheet 0002 (1984), and RWP 02-2-00250, Timesheet 0030 (1984), sign-in sheets were transcribed without traceability to the criCinal sign-in sheets.
Recommendation The Quality Assurance records for RWPs 02-2-00214 Timesheet 0002, 02-2-00250, and Timesheet 0030 should be supplemented with information providing traceability to the original worker sign-in sheets.
I-85-514-SQN-03, RWP Changes to Reflect Current Airborne radiological Information The need to transcribe data to a new timesheet due to " piggy backed" air data is indicative of programmatic problems with the RWP Timesheets. The Sequoyah HP-proposed changes to the RWP and RWP Timesheet should resolve the problem of individuals making entries on the timesheet for days beyond those covered by the airborne data.
Recommendation No action required beyond incorporation of the proposed changes to the RWP and RWP timesheet.
The Sequoyah line management response to the NSRS report I
(reference memorandum from Abercrombie to Whitt, dated l
January 16, 1986) is as.follows:
l Sequoyah Nuclear Plant Response to I-85-514-SON-01 Health Physics Section Instruction Letter (HPSIL)-7 will be revised to clearly define the requirements for transcription of information between RWPs. The revision will be completed by February 28, 1986.
RWP Timesheets 02-2-0214 Timesheet 0002, 02-2-0250, and Timesheet 0030 were reviewed to determine whether or not the I
recommended supplements had been made according to the NSRS I
recommendation. These timesheets were determined not to have been supplemented with the appropriate information as recommended by NSRS.
Page 19 of 36
Revision 1 A review of HPSIL-7 was conducted to determine whether the recommended revision to the section instruction letter had been affected.
It was found that ASIL-4 was revised to meet the recommendations of NSRS report I-85-514-SQN instead of HPSIL-7, IR1 as it was determined by Sequoyah HP that the revision was more I
appropriate there. This revision addresses the methodology for providing transcription copies of HP records.
An interview was conducted with an individual from HP to ascertain whether or not the revision to ASIL-4 addressed the handling of RWP timesheets. The revision has addressed the problem of transcriptions. Revisions to the RWP program have resulted in a decreased frequency of timesheet revisions.
In addition,' report 311.03-SQN identified QA record deficiencies lR1 in Sequoyah RWP timesheets and identified a need for appropriate l
corrective action.' These findings are applicable to this report.
l 10.
Concern 11-85-098-002 questions the frequency of radiological surveys and implies that they are not conducted often enough.
This concern was evaluated previously in NSRS report I-85-615-SQN (reference 33). A review of the NSRS report and applicable regulations, procedures and documents was conducted to verify the adequacy of the NSES report which was found to fully address the scope of the concern. The NSRS findings are as follows:
A.
The frequency of surveys required by Radiological Control Instruction RCI-1,Section I (reference 7), was found to satisfy the requirements and commitments. RCI-1 states:
Surveys shall be performed on a routine basis to assess radiation exposure rates, contamination, and airborne radioactivity levels. Additional surveys shall be performed whenever required by plant' conditions or work requirements to assure the protection of personnel and to monitor plant conditions.
B.
The specific frequency of radiological surveys required in acess with an active Radiation Work Permit (RWP) is established in RCI-14 (reference 8) and was found to meet the requirements of RCI-1.
RCI-14,Section III, requires that:
Periodic radiological surveys will be performed in all areas covered by an active RWP. The survey period will vary, depending upon radiological conditions, but will not exceed seven days....
I Page 20 of 36 J
_c_,,
_. _.. -., _ -, _,, ~. _ _ _ _, _ _, _,.
Revision 1 Provisions are made for m' ore frequent surveys if system changes occur to change the radiation dose rate.
If the job location is in an area where significant changes in dose rate are likely to occur, a radiological survey should be performed just before the start of work.
C.
The RPM requirement that a persoa should not unnecessarily expose himself to radiation while performing radiation surveys i.e., maintain exposure of HP technicians as low as reasonably achievable (ALARA) has been satisfied by an exception in RCI-14 that:
At the discretion of the plant health physicist or his assistant, the survey period may be extended for ALARA purposes, in increments of 7 days, by making the extension in writing to the responsible shift supervisors.
Additionally, according to HPSIL-7 (reference 9), routine surveys (a survey once every seven days) may be deleted for an individual area if an RWP is not in effect in the particular area or if radiation levels exceed 1000 millirem per hour and no work is scheduled in that area. Thus, radiation exposure of health physics personnel will be maintained ALARA if no surveys are required to support ongoing work.
D.
For many areas of the plant which are routinely accessible, surveys are documented on propriated survey sheets which establish the weekly survey routine to ensure that a survey is conducted once every seven days.
E.
Surveys are scheduled on these preprinted sheets for specific shifts throughout the week. A review of these proprinted sheets found that numerous areas outside the regulated area (i.e., the cafeteria and hallway by the electrical shop) were surveyed more frequently than once a week to check for the presence of transferable contamination.
F.
Routine surveys of the Containment Building and various rooms in the Auxiliary Building are scheduled based upon work planned during operation or for a particular outage. A survey status list and/or a monthly schedule of routine surveys are maintained at the HP lab / control point to ensure that the frequency of surveys meet the requirements of RCI-14.
A~
review of the monthly schedule at unit 1 containment control point (marked-up calendar) indicated that containment surveys were currently being conducted on a five-day schedule.
Page 21 of 36
' Revision 1 G.
Surveys for the Auxiliary and Containment Buildings were reviewed for the period of July through September 1985. The frequency of radiation surveys of 15 locations for the duration of this period indicated that these locations had received a routine survey on a seven-day schedule.
4 l
H.
RWP timesheets from 1984 demonstrated that surveys had been.
conducted on at least a seven-day schedule in accordance with RCI-14.
Because of the' nature of the work, one of the timesheets had radioactivity / contamination surveys performed on five days in an eight-day period.
I.
Based on interviews with individuals C and D (designated by NSRS report), few personnel (less than 25 percent) review the
-survey sheets at this time in the outage (two to three months into the outage) before entry into containment on an RWP.
Personnel were observed at the control points for unit 1 for a period during which approximately 20sindividuals processed through the control point, with none reviewing surveys. A check of the associated RWP timesheets showed that these individuals had previously worked in containment on those timesheets. Individual D stated that when an RWP timesheet is first opened, all radiation hazards are discussed by the HP i
with the associated foreman, using the survey map. The HP at the control point reiterates this information when the work crew enters the RWP for the first time. Additional instructions to workers on subsequent entries are provided to the workers only on a case-by-case basis. A control point HP Tochtician (individual C) was observed giving instructions to workers on special dosimetry requirements on a reentry on one job because of the nature of the work on reactor coolant pumps. Radiation levels were not reiterated to these individuals since it wa's
/
unchanged from their last entry.
Conclusions of "he report are as follows:
Concern 1X-85-098-002 was not validated. The frequency of lR1 radiation surveys, with the flexibility to have more surveys
~ when changes in radiation levels are anticipated, was judged to. adequately meet the requirements.
After a review of site procedures, it was determined that the conclusions are valid.
I i
Page 22 of 36 l
, l Revision 1 i
11.
Concern I-86-238-SQN consists of a request to implement a procedure encompassing al1~ aspects of possible emergency situations in a C-Zone.
1 No previous investigations of this concern have been conducted.
The evaluation of this concern consisted of a review of current HP procedures governing radiological safety in contaminated areas and i l Sequoyah emergency procedures, policies and guidelines to determine the adequacy of each to mitigate C-Zone emergency situations. The following general programmatic areas were examined:
A.
Training of plant employees in their responsibilities during i
emergencies B.
Scope of responsibilities for different classificationslof
~
employees.
C.
Training of those employees permitted access to radiologically i
controlled areas.
An interview with a supervisor identified plant instructions (listed 'in x
the reference section) issued to provide guidance to employees'in the event of situations described in the concern. The supervisor explained ~
how plant practice is to provide intensive training to those selected
?
groups of employees who will be responsible for handling specific problems such as fire, medical, or the release of radioactive material. Nonspecific training is provided to the general plant staff, and is designed to explain the responsibilities. The raployee has to identify and report the emergency and then to evacuate the area while
[
the selected groups handle the situation.
An interview with technicians and operations personnel reiterated the safety supervisor's position that specific groups such as Operations' l
and Radiological Control are responsible for handling emergencies
+
dealing with fires and injuries in contaminated areas. Other plant employees are expected to report such event and then evacuate the area.
.An interview with a supervisor identified those GET courses provided to all plant employees that explain each employee's responsibility. The supervisor also identified specialized courses provided to employees i
who frequent the plant's radioactively contaminated areas.
l These specialized courses provide additional information concerning now the employee should react to fire and/or medical situations when j
radioactive materials are involved.
l 1
Page 23 of 36 l
k\\
Revision 1 Attendance of the GET class on Fire Protection (GET-7) verified that objectives as listed in the training plan (SGET-GET-7) were covered during video presentation and by classroom discussion.
The Standard Practice (SQS-25) provides guidance in how to select a protective breathing apparatus, how to use the plant Hazard Control Nanual (LGA-181, SQS-7 and SQS-21) and how.to recover from a spill of radioactively contaminated liquid (SQA-131).
The Hazard Control Instructions (HCIs) deals with general responsibilities of super;isors (G-2) and employees (G-3).
4 Additional HCIs cover specific problems such as fire and medical emergencies (G-15, G-21, and G-23), the release of plant gases (HM-20) and respiratory protection (PPE-20).
Abnormal Operating Instructions (AOIs) provides guidance for fires (AOI-30), abnormal releases of radioactive material (AOI-31) and chlorine releases (AOI-33).
Site Radiological Emergency Plan and its Implementing Procedures Document (SQN-REP and SQN-IPPs) cover medical emergencies (IPD-10), and HP practices (IPD-14).
I Site Physical Security Instruction (PHSI-13) provides for the
' s correct response to plant fires.
A site Employee) Handbook is given to each employee and provides a brief overview of safety, security, and personnel procedures and steps.
.NRC Inspection Reports-50-327/85-07 and 50-328/85-07 reviewed lR1 TVA's actions during the radiological emergency preparedness drill held at Sequoyah between February 5 and February 7, 1985. No violations or deviations were identified.
General and Specific Training Plans (GET-7, GET-3.1, HP Level 0, I, and II) are designed to inform employees of their responsibilities and available procedures.
12.
Concern JLH-86-003 raises concerns about the location of friskers with regard to their proximity to contaminated area exits. This concern has not been previously investigated. The evaluation described in this report consisted of the review of applicable regulations and procedures, interviews with HP technicians and training supervisors, and field walkdowns to verify placement of friskers.
Page 24 of 36
\\
1 f
Revision 1 Sequoyah Nuclear Plant, RCI-1, revision 30', " Radiological
~
Program,"Section III, paragraph E, states that " frisking stations e
are located throughout the regulated area. These friskers are to, be used when personnel contamination is suspected, and upon' f
leaving a C-Zone."
In addition. HPSIL-10, revision 8. "Personnea Decontamination and Confiscation of Contaminated Articles," states personnel should frisk immediately after or as soon as practical
~
upon exiting a C-Zone.. Background readings can not exceed 200 dpa, in accordance with RCI-1, and this.means that there will be instances when a frisker will be a distance from the zone.
Because of this, it is possible that contamination could be tracked to a frisker.
Current HP procedures account for the possibility of spreading contamination on the way to a frisker.
RCI-1 states a person t
should contact HP immediately if contamination is detected, and.
stay there.. An HP technician will respond to the location fori assistance. The technician will also survey the pathway the employee took and any items they may have touched, such as3 phone, frisker probe, or door knob. If contaminatien is determined to have been spread, the area and items will be decontaminated' immediately, if possible, or zoned off until it can be deconned.
Instructors for Sequoyah's GET inform personnel that a frisker will not always be readily available because of reasons such as background being excessively high.
The example was substantiated concerning the fact that exiting elevation 690 and 669 pipe chases requires passing through closed doors; however, an independent survey revealed that background levels in both pipe chases exceeded 200 dpa, therefore a frisker had to be placed elsewhere. On elevation 669, the frisker had
~
been removed from the frisking booth near the elevator because of high background and placed near the
'A' holdup tank room.
Consequently, personnel may not have been aEare it had been moved 3
l and would have had to look for the frisker'.
l 13.
JKA-85-001 expresses a concern that.In the event of a radiation or evacuation alarm or notice, the operator in charge of an Auxiliary lR1 Building Secondary Containment Enclosure (ABSCE) type breach may l
d leave the area without sealing the breach. This concern was evaluated by a review of the governing procedures and interviews with Sequoyah Operations Section personnel. Sequoyah Technical Instruction 77 (TI-77) establishes the responsibilities and lR1 procedures governing the breaching of the ABSCE. Section 4.2.1 (note) on breaches requires an Unresolved Safety Question Determination (USQD) evaluation of the ability to isolate the breach within 4 minutes of receiving an Auxiliary Building Isolation (ABI) or high radiation signal. TI-77 requirements were confirmed in an interview with the Sequoyah Operations Supervisor who further stated that operators are instructed in this and are knowledgeable of their responsibility to seal any ABSCE type breaches before evacuating or leaving the area.
Page 25 of 36
'q' l'y
]
Revision 1 l
>\\
14.
Concor:t RII-85-A-0064 raises 8 items of concern.
IRI With the, exception of one item, which involved charges of lR1 intimidation and harassment and was referred to the Office of l
the Inspector General, the items were evaluated as follows:
A.
TVA Lacks Ability to Run an HP Operation lR1 3'
The evaluation included the review of NRC, INPO, TVA-QAB, and American Nuclear Insurers (ANI) audits / evaluations of the IRl Seguoyah HP program from 1984 to the present. Applicable j
.Section Instructions and Radiologic ~al Control Instructions were reviewed and implementation of the' instructions observed. Program documentation was reviewed and randomly verified by field walkdowns.
Interviewed personnel included
,E both HP technician and supervisory personnel.
The 1985 NRC-SALP Report gave radiological control at SQN a 2 rating. The 1984 SALP Report gave Sequoyah radiological controls a 1 rating. These ratings indicate a " satisfactory r
performance" (2 rating) to a "high level of performance" (1 rating). Since 1984 Sequoyah HP has had only one Severity
' Level III NRC violation (however, no civil penalty was involved and the violation involved a radiation waste shipment, not radiological protection). During this period, there were eight MRC inspections, and Sequoyah HP had eight y
Level IV and two Level V violations. The 1984 INPO evaluation l
listed three findings in the radiation protection area. The 1
1985 INPO evaluation identified three findings and one Good Fractice. Five'QAB audits were conducted during 1984 and 1985. A total of nine deviations were identified in the QAB Audit Reports.
The HP program at Sequoyah is currently under the direction of the Superintendent, Radiological Controls. This position was created in 1986 and reports directly to the Plant Manager.
The Superintendent, Radiological Controls is designated as the
" Radiation Protection Manager (RPN) as defined by NRC in a
The individual in this position meets the qualification criteria for the po Gr. ion of RPM according to Regulatory Guide 1.8.
B.
Unreported Loss of Radioactive Source IR1 HP SIL-11. " Leak Testing of Radioactive Sources," provides the guidelines for source inventory and control. Sources are routinely inventoried on a weekly basis.
In addition, these sources must be signed for by qualified personnel before and after use.
Interviews with HP technicians from different shifts demonstrated the procedure Page 26 of 36
Revision 1 was understood. None of the technicians could recall any instance of a lost or missing source. An independent s.urvey of the~ source locker verified that all sources were accountable. Random source inventories from 1985 and early
-1986 were reviewed with no discrepancies being found.
C.
Radiation Monitors Not Located According to ASIL-3 lR1 HP ASIL-3, revision 10. " Orienting of Health Physics Technicians for Inplant Work at Sequoyah," contains attachment C-6, which is a listing of radiation monitors-and their locations. This attachment is used by HP technician trainees as an aid in learning the location of these monitors.
Two HP technicians who had completed their Performance Verification Sheets within the last year stated that all.
monitors'are in the locations listed in Attachment C-6.
They i
I did say that some were difficult to locate because of their location, e.g., behind pipes, hangers. A random verification was performed by walkdown, and all monitors checked were in proper location according to attachment C-6.
D & E.
Smears Thrown into Trash / Smear Counting Area Used as an Eating lR1 Area l
Both of these items deal with the handling of smears in the HP y
lab counting room. The evaluation of these items consisted of L
interviewing HP field operations personnel and examination of applicable HP procedures. The findings of this evaluation are 1
as follows:
1.
Smears are handled and counted on a designated counter top in'the counting room. This area is posted as a regulated area; therefore, eating, drinking, and use of tobacco lR1
'~
products are not allowed in this area.
2.
The remainder of the count room and HP field-facilities is not a regulated area; therefore, eating, drinking, and use a
of tobacco products are allowed in these areas.
i 3.
The HP lab, counting room, and regulated counter top are
{
required to be routinely surveyed at least daily. Any contamination detected is required to be immediately deconned.
(
Reference:
SQN HP-SIL-4) 4.
After counting, all smears whether contaminated or not, are placed in a "conteminated material" designated container and never in the clean trash receptacles.
f Page 27 of 36 r
I o
~
Revision 1
-a 5.
HP technicians do not normally eat in the counting room even though it is not prohibited except on the regulated area counter top.
F.
Air Samples Improperly Taken/ Respirators Not Worn in.High lR1
(>10,000 dpm) Contamination Areas l
Interviews with HP trainees and training supervisors indicated that technicians are taught to avoid locating an air sampler on a contaminated surface since a possibility exists that the j
sempler might collect loose surface contamination. This could j
result in a higher calculated airborne activity that would not
)
be truly representative of the airborne activity. This would result in the recording of higher airborne radioactivity concentration levels on applicable survey forms and RWPs might lead to a requirement for respiratory protection. These measures would, however, be conservative and would not lead to an increased risk to the workers.
It is also understood by those interviewed that situations can develop where there may
~
be no alternate location to place an air sampler in order to obtain a representative semple of workers breathing zone. In j!
this case, technicians are instructed to exercise caution such that the air sample would not become contaminated because of loose surface contamination.
Random observations of HP technicians pulling air samples revealed proper sampling practices. All those observed est up the air sampler as close to breathing zone as possible, considering location of work and available equipment. All were knowledgable of their task.
NUREG 0041 establishes guidelines by which respirators should i
be utilized.
It states " Personnel who are responsible for establishing... and maintaining respiratory protection programs must exercise' sound judgment by providing and using
(
engineering controls, where feasible. and by avoiding i
. unwarranted use of respirators." RCI-14, revision 5, provides guidelines for use of protective clothing. Attachment 3.
states that except for (1)-breaching a radioactive or potentially radioactive system or (2) welding, scinding or burning a contaminated component, respiratory protection is not recommended until contamination levels exceed 10,000 dpa,
(
or 10 times the level expressed in the concern.. HP, according to TVA RPP, does have authority to prescribe respiratory protective devices when deemed necessary.
A review of randomly selected RWPs was performed, and in the cases reviewed, the initiating technician of the RWP followed the guidelines set forth in RCI-14, attachment 3.
1 I
Page 28 of 36
Revision 1 G.
(Not applicable to this report)
]
H.
Air Sample Heads Not Covered Prior to or After Sampling lR1 i
.This item expressed concern over HP technicians not covering
)
the air sampler heads before and after taking air samples.
The evaluation of this item consisted of an interview with a l
Seg'oyah HP-shift supervisor and review of HP procedures. HP u
4 technicians are taught to avoid cross-contamination of air r
sample. filters; however, the means by which they accomplish
~
this is up to their discretion. There are no requirements for covering air campler heads'before or after sampling.
It should be noted also that if an air sample filter should become cross-contaminated, the resulting air data would indicate higher airborne activity than that which actually existed resulting in more conservative protective measures being required than necessary and in no way compromising worker safety.
Conclusion t
1.
SQP-86-009-001 - The concern was not validated. No evidence of personnel contamination as a result of poor management attitudes toward radiological safety was found. Reviews of Sequoyah procedures indicated that the programs in place governing both internal and external personnel contamination control and safety j
adequately implement and comply with regulatory requirements.
Personnel contamination is documented and investigated by way of RIRs. This evaluation did not identify any deficiencies in the Sequoyah personnel contamination control program. The concern
.does not affect the safe operation of the plant.
2.
SQP-86-009-002 - The concern was not validated. ' Examinations of l
applicable procedures and interviews with cognizant personnel l
Indicated that changes made to containment access procedures were l
made prior to the transfer of HP to the DNP and that those IR1 changes did not compromise the health and safety of workers.-
l The concern does not affect the safe opert. tion of the plant.
l 3.
K1-85-084-001 - The concern was not validated. The NSRS l
investigation could find no evidence that HP personnel did not lR1 properly respond to radiation monitor alarms. This report concurs l
fully with the NSRS findings and conclusions. The concern does l
not affect the safe operation of the plant.
l i
Page 29 of 36
-. ~ -.
Revision 1 4.
XX-85-066-001 - The concern was not validated. This report concurs with the Sequoyah Line Management Report findings and conclusions that HP, or any other safety organization, responds to an alarm or unknown situation with deliberateness and caution to prevent possible hazard and ensure personnel safety. The concern does not affect the safe operation of the plant.
5.
XX-85-009-002 - (XX-85-009-001) The concern was not validated.
NSRS found no evidence indicating that older persons are assigned to the " hottest" (high radiation) work. This report concurs with the NSRS findings and conclusions. The concern does not affect the safe operationHof the plant.
6.
WI-85-038-001 and XI-85-015-001. These concerns were not-validated.
As~ stated in the Sequoyah line response report, "Even if [the] quality factor increased by a factor of 5, the effect from neutrons would still be of less concern than gamma radiation." Therefore, the practice of entering containment while at power'for nonemergency repairs does not need to be reevaluated. The investigation documented in the Sequoyah line i
report indicates compliance with 10 CFR 20 requirements regarding l
neutron dose assessment. The policy of allowing "at power" I
containment entries had no direct bearing on the thimble tube ejection accident at Sequoyah. This report concurs with the Sequoyah line response. The concern does not affect the safe i
operation of the plant.
l 7.
XX-85-026-001 - The concern was not validated in that HP does receive adequate upper management support to enforce the
-radiological safety program. No evidence was found by Sequoyah line management to support the allegation that employees who intentionally bypass monitors were not disciplined. Some needed improvements in the present RIR program were noted and corrective action recommended to upgrade the program. This-report concurs fully with the'Sequoyah line report. The concern does not affect the safe' operation of the plant.
i 8.
11-85-063-001 - The concern was not validated. NSRS found no evidence that the incident occurred as described by the CI or to corroborate the opinion that Operations and HP personnel do not i
provide adequate information or verify system contents. This report concurs with the findings and conclusions of the NSRS report. The concern does not affect the safe operation of the plant.
Page 30 of 36 i
Revision 1 9.
XX-85-028-X02 and XX-85-028-X03 - Concern XX-85-028-X02 was found to be indeterminate and XX-85-028-X03 was validated.
Both concerns were evaluated by NSRS in report I-85-514-SQN and were subsequently evaluated by QTC. NSRS subsequently referred this concern to the Office of General Counsel (OGC) for further investigation. OGC completed its evaluation and issued report OGC-86-021 on March 20, 1986.
HP committed to revise their procedures concerning transcription of QA records. The revision to ASIL-4 is considered to meet this commitment. A recommendation.to clarify QA record requirements i
for RWP timesheets and enhance worker awareness of their responsibility to properly handle QA records was made by Operations CEG report 311.03-SQN, and appropriate corrective action is being considered at this time by SQN personnel.
Official dose records are derived from TLD data and not RWP timesheets; therefore, these concerns do not affect the safe operation of the plant.
10.
Concern XX-85-098-002 - The concern was not validated. NSRS findings verified that radiological surveys are carried out according to procedural requirements, are sufficient to maintain an adequate assessment of plant radiological conditions, and comply with regulations. This evaluation concurs with the findings and conclusions of the NSRS report. The concern does not affect the safe operation of the plant.
11.
I-86-238-SQN - The concern was_not validated. The evaluation of the concern concludes that existing radiological protection procedures, emergency procedures, and personnel training programs
. address the handling and mitigation of any potential C-Zone emergency situations. No programmatic deficiencies were found.
The concern does not affect the safe operation of the plant.
I 12.
JLH-86-003 - The concern wac not validated. The review of i
applicable plant procedures, personnel training, and plant l
walkdowns indicated that an adequate number of friskers are placed throughout the plant in locations as convenient as possible to existing C-Zones with regard to background radiation requirements and that personnel training regarding knowing frisker locationa, using friskers properly, and knowing what action to take when contamination is indicated is in compliance with regulatory and plant procedural requirements. No programmatic deficiencies were found. The concern does not affect the safe operation of the plant.
Page 31 of 36
Revision 1 13.
JMA-85-001 - The concern was not validated. SQN TI-77 adequately addresses the securing of ABSCE breaches and it was determined that Sequoyah operators are properly instructed and aware of their responsibilities regarding this. The concern does not affect the safe operation of the plant since it was not validated.
i 14.
RII-85-A-0064 - The concern was not validated. None of the I
deficiencies expressed in the concern were found to exist and lR1 the concern does not affect the safe operation of the plant.
l IV.
Root Cause The following concerns were not validated; therefore, no root cause evaluation was necessary.
1.
SQP-86-009-001 8.
11-85-026-001 2.
SQP-86-009-002 9.
11-85-063-001 3.
11-85-084-001 10.
11-85-098-002 4.
11-85-066-001 11.
I-86-238-SQN 5.
11-85-009-002 12.
JLH-86-003 6.
WI-85-038-001 13.
JMA-85-001 7.
11-85-015-001 14.
RII-85-A-0064 Concern 11-85-028-102 was indeterminate.
Concern 11-85-028-103: The root cause of the concern, as stated, is l.
determined to be a programmatic deficiency in a plant procedure which IR1 has been corrected by the revision to ASIL-4.
l Y.
Generic Applicability Concern 11-85-028-103 is considered generically applicable to all other TVA Nuclear Plants that employ RWP timesheets because of the scope and nature of the programmatic deficiencies noted in the HP's QA records disposition and management system.
Concerns WI-85-038-001 and 11-85-015-001 are generically applicable to both Watts Bar rud Sequoyah but are not validated for either plant.
All other concerns evaluated in this report pertain to l
Sequoyah-specific incidents, were not validated, and are therefore l
not generically applicable to any other TVA facility. No evidence of lR1 similar incidents or situations existing at other TVA nuclear plants l
was found.
l Page 32 of 36 E'
Revision 1 VI.
References 1.
Title 10 Code of Federal Regulations, Part 20 2.
Title 30 Code of Federal Regulations Part 11 3.
U.S. NRC Regulatory Guide 8.8 - ALARA 4.
U.S. NRC Regulatory Guide 8.15 - Respiratory Protection 5.
NUREG 0041 - Manual of Respiratory Protection Against Airborne Radioactive Materials.
6.
TVA Code VIII, " Occupational Radiation Protection" 7.
TVA Radiation Protection Program (RPP) 8.
Sequoyah Nuclear Plant Technical Specifications (STS) 9.
Sequoyah Nuclear Plant Final Safety Analysis Review (FSAR) 10.
U.S. NRC Regulatory Guide 1.8, Revision 1 11.
Sequoyah Nuclear Plant Technical Instruction 77 (TI)77, " Breaching the Shield Building, ABSCE, or Control Building Boundaries" 12.
Sequoyah Nuclear. Plant Radiological Control Instructions (RCIs) 1-14 13.
Sequoyah Health Physics Section Instruction Letters (SILs),
HPSIL 1-37, ASIL 1-15, DSIL 1-24 14.
NRC Fifth Systematic Assessment of Licensee Performance (SALP) for March 1, 1984 through March 31, 1985 dated September 17, 1985 15.
NRC Fourth Systematic Assessment of Licensee Performance (SALP) for January 1, 1983 through February 29, 1984 4
16.
INPO Evaluation of S9quoyah Nuclear Plant - April 1985 17.
INPO Evaluation of Sequoyah Nuclear Plant - April 1984.
18.
NRC Inspection Reports, Sequoyah Health Physics Program a.
50-327/86-04, 50-328/86-04, 03/27/86 t
b.
50-327/85-20, 50-328/85-20, 06/20/85 c.
50-327/85-26, 50-328/85-26, 09/06/85 d.
50-327/84-34.
50-328/84-34, 11/21/84 i
e.
50-327/84-21.22 50-328/84-21,22 09/17/84 f.
50-327/84-14, 50-328/84-14, 07/27/84 g.
50-327/84-12, 50-328/84-12 03/29/84 h.
50-327/84-04, 50-328/84-04, 03/12/84
. i j!
Page 33 of 36
~
Revision 1
- 19. 'SQN-NRC-0IE Inspection Report Nos. 50-327/84-34 and 50-328/84,
Response to Violations Abercrombie to Hufham, dated January 9, 1985 (S53-841218-913) 20.
SQN-NRC-0IE Report 50-327/85-20 and 50-328/85-20 Response to Violations, Abercrombie to Hufham, dated July 15, 1985 (SS3-850712-964) 21.
SQN-NRC-0IE Report 50-327/85-26 and 50-328/85-26, Response to Violations, Abercrombie to Hufham, dated December 30, 1985 (S53-851230-981) 22.
SQN-NRC-0IE Report 50-327/86-04 and 50-328/86-04, Supplemental Response to Violations Gridley (TVA) to Grace (NRC), date July 3 s
1986 (L44-860703-800) 23.
QAB Audit Reports a.
QSS-A-85-0009 (L17-850308-801) b.
QSS-A-85-0010 (L17-850510-801 c.
QSS-A-850012 (L17-850905-800) d.
QSS-A-85-0016 (L17-860225-803) e.
CH-8400-14-01 24.
NSRS report I-85-514-SQN " Radiation Work Permits" dated December 27, 1985 25.
Memorandum from K. W. Whitt to W. T. Cottle, " Corrective Action Response Evaluation," dated January 30, 1986 26.
Memorandum O. L. There to N. A. Harrison, " Response to NSRS report I-85-514-SQN." dated February 3, 1986 27.
" Investigation / Evaluation of.NSRS Referred Employee Concern 11-85-066-001," (S01-851205-982) 28.
"Sequoyah Nuclear Plant (SQN) - Request For Evaluation of Employee Concern XI-85-066-001" (S01.851025 870) 29.
Investigation / Evaluation of NSRS Referred Employee Concern XI-85-015 "Sequoyah/ Personnel in Containment While Operating,"
dated August 28, 1985 I
30.
Radiation Protection Dosimetry, " Kerma Equivalent Factor for Photons and Neutrons Up to 20 MeV," Volume 14, Number 4, pp 289-298, (1986), Nuclear Technology Publishing 31.
"Sequoyah Nuclear Plant (SQN) - NSRS Investigation of Unit 1 Incore Instrumentation Thimble Tube Ejection Accident on j
April 19, 1984 - NSRS Report I-84-012-SQN," (LOD 840830 516)
Page 34 of 36
+4 Revision 1
-32.
Investigation / Evaluation Report, " Employee. Safety Concern - QTC Concern :
XX-85-026-001," dated February 4, 1986, (L61-860204-800) 33.
NSRS Report I-85-615-SQN, " Frequency of Radiation Surveys," dated December 10, 1985 34.
Sequoyah Nuclear Plant Engineering Section Instruction Letter ESSIL-C5, revision 0, "By product Material-Radiation Sources" 35.
Health Physics Technician Training Lesson Plant HPT-LP-14 36.
Sequoyah Nuclear Plant Radiological Survey, Form TVA 17069, Survey Number 0-85-2247 37.
Sequoyah Nuclear Plant HP Shift Coordinators Shift Daily Journal (Log), December 12, 1985 entries 38.
Title 10, Code of Federal Regulations, Part 50 39.
U.S. NRC Regulatory Guide 1.101 " Emergency Planning..."
40.
NUREG 0654, revision.1, " Criteria for Preparation and Evaluation of Rudiological Emergency Responses..."
41.
Sequoyah Nuclear Plant Hazard Control Instructions 42.
Sequoyah Nuclear Plant Standard Practices Manual I
a.
SQA - 131 " Recovery From a Spill..."
b.
SQA - 181 "Nazardous Material Control"
.c.
SQS - 7
" Hazard Control Plan" d.
SQS - 21 "SQN Hazard Contrc1' Instruction Manual" e.
SQS - 25
" Breath Apparatus" f.
SQS - 41
" Emergency Medical Treatment..."
g.
SQS Employee Complaints Concerning Safety and Health" l
43.
Sequoyah Nuclear Plant SOI-26.2 " Fire Interaction Manual,"
l revision 3, dated June 30, 1986 44.
Sequoyah Nuclear Plant A0I-30, AOI-31, and AOI-33 a
45.
Sequoyah Nuclear Plant Administrative Instruction. AI-14. " Plant Training Program" l
46.
Sequoyah Nuclear Plant Physical Security Instruction, PHYSI-13 l
" Fire" l
47.
Sequoyah Nuclear Plant Radiological Emergency Plan 48.
Memorandum NRC to TVA dated February 27, 1985 "SQN REP Exercise l
l Evaluation," 50-327/85-07 and 50-328/85-07 (A02 850304 020)
Page 35 of 36
Revision 1 49.
Sequoyah Nuclear Plant General Employee Training (GET) Lesson Plans a.
GET-2.1 "HP Level I" b.
GET-2.2 "HP Level II" c.
GET-2.4 "HP Level 0" d.
GET-3.1 " Security and Emergency Plans" e.
GET-7~" Fire Protection" 50.
Sequoyah Nuclear Plant Administrative Instruction AI-8 " Access to Containment," revision 17 51.
NSRS Report I-85-513-SQN*, " Radiation Exposure of Older Personnel," dated December 27, 1985 (Concern 11-85-009-001 and 11-85-009-002) 52.
NSRS Report I-85-513-SQN*, " Work Areas Contaminated / Lack of Knowledge of System Contents." (concern 11-85-063-001) 53.
Sequoyah Nuclear Plant, R'3P, Implementing Procedure, (IP)-15 54.
Memorandum from H. L. Abercronble to W. H. Thompson dated September 9, 1985, S01 850830 802 VII. Immediate or Lont-Term Corrective Action XX-85-028-IO3: Pertinent Procedures ** have been revised to reflect l
the current status of determining / classifying RWP-timesheets as QA or lR1 non-QA; however, all RWP-timesheets are retained as lifetime records.
l 11-85-026-001 - Recommendation to distribute RIR summaries to HP staff l
has been incorporated (first communications mailed for review 9/10/86) l and will be issued each quarter. In the future the summary sheet will be IR1 mailed to the Plant Manager as a possible agenda item for his weekly 1
meeting.
l The Corrective actions for these two concerns are being tracked on CATD lR1 Number 31104-SQN-01.
l
- Both NSRS reports are transmitted under the same NSRS report number.
Pertinent reports: AI-7 Rev 39, RCI-14 Rev 5, ASIL-4 Ret 11, HPSIL-7 Rev 15.
I Page 36 of 36 i
i
-. - m...
TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT EMPLOYEE CONCERNS TASK GROUP OPERATIONS CEG 1
Subcategory:
Instrumentation and Radiation Monitoring Employee Concerns: Location of Cold-leg Accumulator and RWST Level Transmitters Report Number:
303.02 - SQN Revision 2 IN-85-281-003 IN-85-142-006
- 4. Sta10 d<nAAA Evaluator:
10-?3-A4 G Dar 1 Gar ner Date Reviewed by:
~ 4//' M'
///h5//4 OPS C Maju6er
' Date m#
SkJ/b Approved by:
W. R. Lag (egk'en IDate 1522T l
{
Revision 2 I.
Location of' Cold-Leg Accumulator and RWST Level Transmitters This report evaluates two employee concerns for Sequoyah Nuclear i
Plant (SQN) regarding the calibration of level transmitters.
lR1 These concerns were determined potentially safety-related by the Employee Concerns Task Group (ECTG) Technical Assistance Staff (TAS).
II.
Specific Evaluation Methodology The employee concerns identified to Quality Technology Company (QTC) for Watts Bar Nuclear Plant (WBN) are as follows:
IN-85-281-003 "The two level transmitter on each of the cold-leg accumulators will not read the same level due to difference in elevation. Transmitters have been calibrated as per scaling data sheet. When put in service transmitters will read a 5% difference in level. When the' engineers were told about the problem they said don't worry about it adjust one until both level' indicators read the same.
(Unit 1)"
IN-85-142-006 "RWST in Unit 1 narrow range 1-LT-63-46 & 1-LT-63-49 reading were 6% off inst eng (name known) directed instrumentation mechanic to adjust to zero. This would make reading match the control room. Similar on SIS accumulators 1-4, elev 716. Two transmitters on each accumulator. This practice causes false readings in control room."
Since these concerns were identified specifically for WBN, this evaluation was limited to a review of the calibration methods and historical performance at SQN for the same instruments and i
did not consider allegations of improper adjustment instructions lR1 from engineers.
l SQN Technical Specifications (TS) were reviewed for requirements on the l
cold leg accumulator (CLA) and the refueling water storage tank (RWST) i levels. Surveillance Instruction (SI) data and scaling sheets were I
reviewed for historical data on the CLA level transmitters. sis, IR1 Maintenance Requests (MRs), and SI data were reviewed for historical l
performance of the RWST level transmitters. The historical data was I
supported by informal interviews with the cognizant instrument maintenance engineer and the assistant section supervisor.
l Page 1 of 6
Revision 2 III. Findings IN-85-281-003 i
The SQN TSs (reference 4) require a borated volume between 7,857 and 8,071' gallons in each CLA in modes 1, 2, at.d 3 which must be verified at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by the absence of alarms or measurements of levels and pressures. SQN i
SI-2 (reference 1) verifles these requirements by checking for
]
absence of alarms and verifying levels. The SI requires that the two independent level channels agree with less than 4.5-percent I
deviation between channels.
IR1 i
The cognizant engineer in Instrument Maintenance Section stated that problems with CLA level transmitters existed at SQN. These problems consisted of deviations between level channels on the same CLA and were attributed to elevation differences on the sense line taps. The 4
engineer stated ther.e problems were corrected by adjusting the scaling K
for the transmitters to the narrowest span using SI-161 (reference 2);
however, instrument drift remains a maintenance problem. Design Change 2l Request (DCR) 1848 (reference 6) was written to replace the obsolete 4'
Barton transmitters and will be implemented by Engineering Change Notice i[ 1 (ECN) L6358 during the next outage. According to the engineer, this change should reduce the maintenance caused by instrument drift on this I
y system. The assistant section supervisor confirmed that the CLA level IR1 1;
transmitters were being replaced.
l 1
1 A random review of SI-2 data packages for units 1 and 2 between b
Apell 1984 and April 1985 indicated that the CLA levels were indicating
/
within the allowable maximum channel deviation, y<
l
)
IN-85-142-006 n..
pg s The SQN TSs (reference 9) require that the borated water volume in the RWST be verified at least once every seven days in modes 1 through 6.
(Note: A lesser water volume is allowed in modes 5 and 6 than in modes
[
l 1 through 4).
SI-3 verifies water volumes meet the TS requirements by verifying the two independent RWST level channels indicate the required 4
level within a maximum 525-gallon deviation in modes 1 through 4 and a l
j 5-percent deviation in modes 5 and 6.
H SI-3 records only one narrow range level channel for verifying compliance
[
with TS RWST volume requireaants and the specific channel used is not l
designated during performance of the SI.
In addition, the acceptance l
criteria is specified as notes provided on the data sheet, and state that I operability is verified by having less than 525 gallons lR2
)
.(modes 1 through 4) or 5-percent (modes 5 and 6) deviation between I
i channels. As a result, channel calibration out-of-tolerance reports l
l cannot.be evaluated for impact on SI results nor can an audit be l
I l performed to verify past compliance.
I i
1 k
l Page 2 of 6
Revision 2 A review of MRs on RWST level transmitters and indicators for units 1 and 2 revealed approximately 50 MRs related to calibration problems.
.These MRs were generated from 1980 to 1985 and indicate a recurring
. problem with instrument drift. The cognizant instrument engineer stated that the scaling had been corrected for elevation differences l
on these transmitters in a similar manner to the CLA transmitters.
IR1 The assistant section supervisor stated tha* a lot of the problems may have been because of freezing of instrument lines which had been improved in the las'. few years.
It was noted during the search for calibration related MRs, several MRs had been written'for heat tracing lR1 on_these instruments and that the frequency of MRs had reduced in 1984 and 1985 compared to earlier years.
The supervisor stated that a new maintenance trending program was in place at SQN which will identify both old and new maintenance problems in the future. He noted that, while no plans are currently underway to address problems with RWST level transmitters, increased maintenance I
will continue to enture their operability in accordance with TS lR2 requirements.
I Conclusions The issues presented by employee concerns IN-85-281-003 and l
IN-85-142-006 were validated for SQN in that problems were encountered i
in the past with RWST and CLA level channel deviation because of sense IR2 line tap elevation differences. These problems were compensated by I
rescaling the transmitters to the most conservative span based on l
field measurements.
i Recurring maintenance problems exist with these level transmitters in IR1 the form of instrument drift. The CLA level transmitters tre scheduled to be' replaced during the next outage by RCN L6358 and should correct this problem. No plans have been made to address drift problems with l
RWST level transmitters, however, maintenance will continue to be lR1 i
performed to ensure their operability in accordancs with TS I
requirements. The RWST level transmitter surveillance IR2 data and acceptance criteria in SI-3 was determined insufficient based l
on recording only one channel and having its acceptability included as I
a note within the SI.
l IV.
Perceived Root Cause Excessive maintenance is required to meet TS requirements. The root IR2 cause is perceived as the lack of an adequate trending program in the past to identify and resolve problems for high maintenance equipment.
lR1 No root'cause could be readily identified for the lack of documentation required by SI-3.
Page 3 of 6
Revision 2 V.
Generic Applicability These issues were evaluated generically for SQN and specifically for WBN with siellar findings. These specific issues are not considered generically applicable to BFN or BLN; however, it is believed that BFN and BLN should address the maintenance trending issue, i
4 Page 4 of 6
_ _ ~ _
Revision 2 VI.
References 1.
SQN Surveillance Instruction (SI)-2, " Shift Log," Revision 40, July 3, 1986, and Data Packages April 1984-1985 2.
SQN SI-161, " Channel Calibration of SIS Accumulator Tank Water Level and Pressure Instrumentation," Units 1 and 2. Revision 11, February 20, 1986, and Revision 9. September 11,~1983 3.
SQN SI-165, " Channel Functional Test of SIS Accumulator Tank Water Level and Pressure Instrumentation," Units 1 and 2 Revision-6 August 16, 1984 4.
SQN Technical Specifications, LCO 3.5.1.1, SR 4.5.1.1.1, Bases 3/4.5.1, Unit 2 5.
Technical Instruction (II)-41-63, " Scaling and Setpoint Document "
Revision 4. July 10, 1985 6.
Design Change Request (DCR) SQ-DCR-1848, January 13, 1983 7.
Memorandum from H. B. Rankin to J. P. Yineyard on SQ-DCR-1848, dated January 3, 1985 (S58 841217 906) 8.
Maintenance Requests (MRs) for LT/LI-63-46 and -49 on Unit 1 and Unit 2, search conducted August 4, 1986 9.
SQN Technical Specifications, LCO 3.1.2.5, 3.1.2.6, 3.5.5, SR 4.1.2.5, 4.1.2.6, 4.5.5, Bases 3/4.5.5, Unit 2 10.
SQN SI-3, " Daily Weekly, and Monthly Logs," Revision 49, June 27, 1986 11.
Memorandum from H. L. Abercrombie to W. R. Brown, Element Report l
303.02 SQN Corrective Action Plan, dated September 22, 1986 (S03 l
860918 808) l 1
12.
Memorandum from W. R. Brown to H. L. Abercrombie, Transmittal of ECSP lR2 l
Report and Corrective Action Tracking Documents, dated i
September 29, 1986 (T25 860929 851) l i
l l
13.
Memorandum from W. R. Brown to H. L. Abercrombie, concurrence with l
Corrective Action Plan, dated October 9, 1986 (T25 861009 929) l l
1 I
Page 5 of 6 l
1
,--,-,c,
.m---
Revision 2 i
VII.
Immediate or Long-Term Corrective Actions SQN has provided the following acceptable Corrective Action Plan for i
Employee Concern Task Group Number 303.02 (IN-85-142-006) " Refueling Water i Storage Tank (RWST) Upper Level Transmitters" (references 11 and 13).
l l
Investigation of this employee concern at SQN has revealed a significant l
number of MRs related to problems on the RWST upper level transmitters.
l l
However, a large majority of these NRs were related to frozen sense lines l
or heat tracing problems on the sense lines'. Surveillance Instructions l
SI-706.1, and 706.2 were initiated in February 1985 to verify operability-1 of sense line heat trace circuits. The performance of these surveillance l
instructions on a monthly basis should reduce the number of maintenance i
problems relating to heat tracing.
I I
SQN has recently implemented a new maintenance trending program to l
Identify potential maintenance problems. However, sufficient data has l
not been collected since correcting sense line problems, to make any l
conclusive determination. SQN Instrument Maintenance Group has plans to l
l initiate a new preventive maintenance procedure to monitor the transmitter I output signal for 1,2-LT-63-46 & 49 on a more frequent schedule (most l
likely monthly). The performance of this PN will not only provide
_l additional trending information but will also ensure operability is I
maintained. The PN will be implemented before unit 2 start up.
l lR2 j
SQN Instrument Maintenance Section personnel have also taken field l
measurements of the transmitters and sense line taps to compensate for i
different head pressures at the transmitters. This was done to ensure I
an accurate redundant level indication between channels. This was 1
completed at the time the heat trace and freeze protection was l
[
upgraded.
Therefore, since corrective actions have already been taken to remedy l-maintenance problems associated with frozen sense lines and to correct I
head pressure values causing inaccurate level indications, and l
insufficient trending data has been collected to make conclusive i
determinations. SQN has no plans to replace the RWST upper level l
transmitters at present. As mentioned previously. SQN will continue to l
monitor the transmitters operation with the new treading program and the l
.new preventive maintenance instruction which will be implemented before l
start up.
l l
l An investigation of the RWST wide range level transmitters
(
(1,2-LT-63-50, 51, 52, 53) has shown that no significant maintenance or l
operability problem exists on these transmitters.
I l
The following deficiency remains open and is being tracked by corrective I
action tracking document (CATD) OP 30302-001-SQN (Reference 12):
l l
The RWST narrow range level transmitter surveillance data and I
acceptance criteria in SI-3 was determined insufficient based on I
recording only one level channel and having its acceptability I
included as a note within the SI.
This condition may represent a l
potential reportable occurrence (PRO).
l Page 6 of 6
- +
i TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT EMPLOYEE CONCERNS TASK GROUP OPERATIONS CEG Subcategory: Cable and Conduit Element: Procedure Problems Report Number: 304.01 SQN, Revision 1 TAK 86-005 IN-85-112-001 P
Evaluator:
/OM/[Jlo P. L. S 'ph6cd
'Da t'e Reviewed by:
/
8 IV
// "/ [M
/ Dite 40.A __
io/ze/ex Approved by:
- 9. R. Lagdeden
' Date j.
1770T
I Procedure Problems This report covers procedure problems addressed by two concerns. One concern addressed the misapplication of RTV 3140 on junction box terminal blocks. The other concern addressed cable pulling and. bending requirments.
II Specific Evaluation Methodology This element consists of two concerns. The first concern is specific to Sequoyah (SQN) and does not apply to any other plant.
TAK-86-005 -
"RIV applied to junction boxes on SMI-0-360-0 may not have been properly applied on all junction boxes.
RTV used was not QA level and has exceeded it's shelf life was not a concern as he has verified that the RTV in question will in fact set up properly.
The onsite employee concern file was reviewed to determine the specific areas of concern.
In addition, the supervisor that documented the employee concern was interviewed to further clarify the specific areas of i~
concern. Cognizant engineers in the Electrical Maintenance Section at SQN were interviewed.
The second concern is specific to Watts Bar Nuclear (WBN) Plant and i
applies to all other nuclear plants.
i IN-85-112-001 - "TVA Nuclear Power Department is not working in accordance with construction specifications when making
~
i modifications and additions. ~EG no maximum pull tension l
is specified when Nuclear Power Department pulls a new i
cable'and minimum bend radius is also not specified",
r f
The Construction Category Evaluation Group (CRG) has issued an element l
report No. C010900 SQN, dated August 7, 1986, for SQN that completely l
j addresses this concern. This report evaluated current ONP cable IR1
[.
installation practices and procedures. Nazimum pull tension, sidewall l
l pressure, and minimum bend radius cable issues are being actively l
l evaluated by DNE and final conclusions will not be available until I
they complete their evaluations and responses to C010900 SQN. No l
i r
further evaluation of this concern is required.
l I
i III Findings i
4 Naintenance Request (NR) No. A554518 was written on August 18, 1985 (Reference 1) to coat junction box terminal blocks with 3140 RTV. The
~
junction boxes to be coated were listed in Special Maintenance i
. Instruction SMI-0-360-3 Revisica 0 (Reference 2).
The contract number I
that purchase the RTV 3140 used on this NR cannot be determined.
(Discrepancy Report No. SQ-DR-85-08-111R, Reference 3 and Nonconformlog Item (NCI) tag No. N2-85-294, Reference 4).
I i
f f
Page 1 of 4 i
-(
III Findings (centinusd)
Only one junction box, 3190, is required by the Eevironmentd1 Qualification (EQ) Binder TB001 to be coated with QA RIV314(: but was coated with questionable coating on MR A-554518. The EQ Birider TB001 shows Reference 6 as documentation of the existence of termi',nal block coating in junction box 3190.
ThreeotherMR'swhichcoatedterminalblockswithRTV3140bnSMIO-360-3 were reviewed. All three had proper matertsl traceability.
All Junction Box terminal blocks listed on SHI 0-360-4 (Refel'ence 9)-have been inspected on SS3 860103 810 J(Reference 6) or replaced or! S02 860728 898 (Reference 10).
i The material traceability practices of the Electrical Maintentnce Tool
.I Room at the time the subject work, MR A-554518, was performed was not l
'consistant with the intent of SQN 45 Revision 17 dated September 27, 1985, lR1 (reference 7).
This material traceability problem was later identified I
and investigated by NSRS. NSRS Report No. I-86-165-SQN (refer ence 11) l addresses their investigation, conclusions, and recommendations.
l Corrective actions to clarify material traceability requirements were I
started with the issue of Revision 18 of SQA 45 on November 21, 1985 l
(reference 12) and continued through the issue of Revision 21 af SQA 45 l
on June 3, 1986 (reference 13). Corrective action was taken b;r lR1 Electrical Maintenance on March 6, 1986 (Attachment A to reference 14).-
l H. L. Abercrambie's responses (references 14 and 15) to the i
recommendations in the NSRS Report along with completed Discretancy l
. Reports SQ-DR-86-04-088R and SQ-DR-86-04-089R (references 16 as,d 17)
I and completed CATS No. 86171 (reference 18) document completion, of all l
recommended corrective actions.
l l
i Conclusions Based on the findings, concern TAK-86-003 was determined to be valid.
Since the RTV 3140 that was used to coat the terminal blocks iti junction box 3190 has now been marked "No-QA" (Reference 3), the qualifi' cation of the terminal block coating is questionable and should be requalified or I
reapplied. Thisopenitemdoesconstituteapotentialsafety-1l elated lR1
- issue, i
The additional issue of material traceability was completely alldressed l
byNSRSandthecorrectiveactionstakenwillresolvethematetlal lR1 traceability problems in the Electrical Maintenance Tool Room that l
caused this concern.
l 4
IV' Root Cause i
The traceability of the RTV 3140 used on MR A-554518 was not carrectly verified prior to use as required by SQA45, section 25.2 (Refegence 7) and AI-11 Section 6.4.3.6.a (Reference 8).
Page 2 of 4 4
m_ _ _,,
V Generic Applicability
-Concern TAK-86-005 is an isolated incident and is not generically.
~
applicable =to any other plant.
VI References 1.
Maintenance Request.(MR) Form no. A554518. " Coat Junction Box Terminal Blocks with 3140 RIV " dated August 18, 1985 (SMI-0-360-3 data sheets attached).
2.
Special Maintenance Instruction SMI-0-360-3 " Coating of Junction Box Terminal Blocks", Revision 0, dated August 18, 1985.
3.
Discrepancy. Report No. SQ-DR-85-08-111R " Electrical _ Maintenance installed "No-QA"-RTV-3140 on CSSC junction boxes and terminal strips on MR A-5544518," dated 8-28-85 reported by C. R. Stutz.
4.
Nonconforming Item (NCI) Tag No. N2-85-294 issued August 27, 1985 by Steve Campbell.
- 5. 0 AI-11. " Recommended Disposition of Nonconforming Items" submitted on August-26,'1985 by Steve L. West.-
6.
B. W. Hooper's memorandum to M. A. Skarsinski, "Sequoyah Nuclear Plant-NURRG 0588-Terminal Block Field Verification Sheets," dated January 3, 1986 (S$3 860103 810) (533 data sheets attached).
7.
Standard Practice SQA45, " Quality Control of Material and Parts and Services," Revision 17 dated September 27, 1985, paragraph 25.2
" Material Traceability for MR/WR materials."
8.
Administrative Instruction AI-11. " Receipt Inspection, Nonconforming Items, QA Level / Description Changes and Substitutions," Revision 33, dated August 6, 1985, paragraph 6.4.3.6.a. Attachments 1 and 3, and Section 7.
9.
Special Maintenance Instruction SMI 0-360-4 " Inspection of Junction Box Terminal Blocks," Revision 0, dated August 19, 1985.
10.
R. W. 01 son's memorandum to D. W. Wilson, "Sequoyah Nuclear Plant -
Significant Condition Report SQNRQP8521 R -RQ Binder No.
2 SQNEQ-TB-001-Open Item Sheet 1 of SA " (S02 860728 898) dated July 28, 1986.
11.
NSRS Report No. I-86-165-SQN, " Craftsmen Directed to Violate l
Procedures, Use Tools Other Than Those Specified, and Use Questionable QA Materials", dated March 17, 1986.
I i
12.
Standard Practice SQA 45, " Quality Control of Material and Parts and l
Services," Revision 18 dated November 21, 1985,'Section 25.0 l
" Material Traceability for MR Materials".
l lR1 Page 3 of 4
.a, VI References (continued) 13.
Standard Practice SQA 45, " Quality Control of Material and Parts and l
Services", Revision 21 dated June 3, 1986, Section 25.0 " Material l
Traceability," Subsections 25.1 " Material Traceability After Issue l
25.2, " Material Traceability for MR/WR Materials", and 25.3 l
" Material Traceability for Workplan Materials".
l l
14.
H. L. Abercrombie's memorandum to R. P. Denise, " Nuclear Safety l
Review Staff (NSRS) Investigation Report No. I-86-165-SQN, l
" Craftsmen Directed to Violate Procedures. Use Tools Other Than I
Those Specified, and Use Questionable QA Material"", dated l
May 21, 1986 1
1 15.
H. L. Abercrombie's memorandum to R. P. Denise. " Nuclear Safety l
l Review Staff (NSRS) Investigation Report No. I-86-165-SQN, IR1
" Craftsmen Directed to Violate Procedures, Use Tools Other Than l
.Those Specified, and Use Questionable QA Material"", dated l
July 01, 1986 l
l 16.
Discrepancy Report No. SQ-DR-86-04-088R dated Apell 16, 1986 l
required a revision of AI-11 to adequately implement the NQAM.
l Closed on June 16, 1986.
l l
17.
Discrepancy Report No. SQ-DR-86-04-089R dated April 16, 1986, l
required the revision of SQA 45 to adequately implement the NQAM.
l Closed on July 01, 1986.
I i
18.
Commitment Action Tracking System (CATS) Item No. 86171 l
Responselto NSRS Recommendation No. I-86-165-SQN-05 completed l
September 8, 1986.
I VII Immediate or Long-Term Corrective Action Junction Box 3190 was recoated with RTV 3140 by WR B200741.
IR1 i
t i
l Page 4 of 4 I
+j TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT EMPLOYEE CONCERNS TASK GROUP OPERATIONS CEG l
Subcategory: Cable and Conduit Element: Electrical Penetration Breached i
Report Number:
304.02 - SQN, Revision 1 IN-85-207-002 h!J/![4 Evaluator:-
des P. L. 3 he
~ Date 3eviewed by:
Yf
/f/7/) f4
' ' DPs C NjMiber
/ Da(e Approved by:
b-
/0/2 I/f 4
,-~-
W. R. Lage(gfen
/Date 1604T
L o
7 I.
Electrical Penetratien Breached This report evaluates the concern that a steel fish-tape is being used to breach penetration seals in lieu of fiberglass or wooden rods, i
II.
Specific Evaluation Methodology
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This element consists of one concern identified at WBN and determined to be generic to Sequoyah.
IN-85-207-002
" Crafts are using steel fish-tape in lieu of fiberglass or wood rods.to breach penetration seals. This may cause damage to existing cables in the breached penetration. This practice of using i
steel fish-taper, violates procedure NAI-and Construction Management (known) memorandum of approximately July 1984. This is a construction department concern, and additional information is on file, withheld due to confidentiality."
The E-form was reviewed to determine the specific areas of concern.
Referenced documents were researched to determine the requirements, and i
other historical documents including NSRS and Quality Technology Company'(QTC) (ERT) reports were reviewed. Cognizant engineers and i
craft in Unit A of the Modifications Branch of the Division of Nuclear Construction (NU CON) at Sequoyah (SQN) were interviewed to determine current practices being used at SQN. Cognizant engineers in the Division of Nuclear Engineering (DNE) were interviewed to clarify current requirements along with any pertinent history.
III. Findings NSR1: Report I-85-702-WBN, " Breaching Electrical Penetrations,"
j (Reference 4) documents the results of the investigation of concern IN-65-207-002 at WBN.
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The SQN procedure that applies to the subject concern is Nodifications and Additions Ir.struction N&AI-13 (Reference 1).
The review of this I
instruction revealed different requirements for breaching equipment being used on cable tray penetrations and conduit seals. According to l
the lustruction, the use of a steel fish-tape is acceptable for
[
breaching cable tray penetrations but a nonconductive probe must be used to breach conduit seals.
l A review of all previous revisions of M&AI-13 showed this:
I i
1.
A 1/2" conduit with a wooden conical nose was used as a breaching l
tool from December 4, 1979 until January 21, 1982 (Revisions 0, 1, lR1 and 2) l t
2.
Revision 3 was effective from January 21, 1982 until l
f February 22, 1982, and required that the breaching tool be l
nonmetallic.
l 3.
From February 22, 1982 until the current revision (Revision 4, 5, l
and 6) metallic breaching tools were authorized for use under i
special conditions.
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III.
Findings (continued)
The author of M&AI-13. Revision 3 used SQA 1/9 Attachment 1, dated l
t August 18, 1981 to initiate the procedure revision. During an l
3-interview with this author he was asked his reasons for his use of the llRI(N words "any tool made of nonmetallic material" snould be used as a breaching tool. His response was he was trying to change from the too1~
l described in Revisions 0, 1, and 2 to "any" tool. He stated that he did l
not have a specific reason (such as cable insulation damage) for i
stating nonmetallic other than good practice.
l The change request that resulted in Revision 4 of M&AI-13 dated v
l February 22, 1986, requested the "... use of metallie breaching tools I
under special conditions.
The current. Revision 6 of M&AI-13 lR1 uses the words "At the discretion of the responsible engineer a 1
metallic breaching tool may be used"..
p 1
I According to interviews with Electrical Modifications personnel, i
I 1/2" conduit, fish-tape and various nonmetallic breaching tools have l
been successfully used to breach penetrations since 1979 in accordance lR1 with M&AI-13 without any known cable insulation damage. This same fact l
was stated in ECTG's C010900 SQN Report.
(
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g It was determined from the interviews with the site modifications personnel that they prefer to use a nonnotallic probe to breach all.
penetrations / seals where possible and only use a steel fish-tape where space restrictions prohibit the use of a longer, unflexible probe, No violations of the current site instructions were discovered from these 5
interviews.
However, during this evaluation it was discovered that Modifications >
and Additions Instruction M&AI-4 " Control, Power, and Signal Cables."x Revision 8, dated December 31, 1985 (Reference 2) does not reference 2
M&&I-13 and does not show any responsibility for the craft to be aware-
?,
of the requirements of M&AI-13. Also Electrical Design Standard j
DS-E13.5.1, " Raceways - Electrical Penetration Fire Stops and Pressure
?
Seals " Revision 0, dated December 5, 1977 gives guidelines on 4
breaching penetrations on nuclear plants built after WBN.
Conclusions a
t n
Based on the findings, this concern was determined to be not, valid at l
H.
SQN. The use of a nonnotallic probe does not preclude cable in.shlation 1
damage resulting from the misuse of the probe by the craft. A metallic IR1
?j tool can be used to breach penetrations safely, without damaging' cable 3 l' d
insulation as indicated by six years of modification work at SQN in
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j accordance with M&AI-13.
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i IV.
Root Cause l3N No root cause could be determined by this evaluation.
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Page 2 of 3 Jl
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V.
G3nzeic Applicc5111ty This concern is specific to Watts Bar and could apply to Browns Ferry and Bellefonte Nuclear Plants depending upon the requirements established in their procedures. There are no upper-tier document I
requirements governing the use of breaching tools. The quality of IR1 workmanship at SQN does not indicate the need fo'r more specific l
requirements than those included in site procedures.
l
'p-r' VI. References s
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1.
Modifications and Additions Instruction M&AI-13. " Electrical Pressure Seal. Fire Stop Barrier and Flame Retardant Cable Coating, " Revision 6, dated January 28, 1985 s
g 2.
Modifications and Additions Instruction M&AI " Control, Power, and Signal Cables, " Revision 8, dated December 31, 1985 1
s3.
TVA General Construction Specification Number G-38, " Installing Insulated Cables Rated up to 15,000 Volts," Revision 7 dated January 15,- 1986
'4.
NSRS Investigation Report I-85-702-WBN, " Employee Concern IN-85-207-002 - Breaching Electrical Penetrations," dated November 22, 1985
' Si.
Electrical Design Standard DS-E13.5.1, " Raceways - Electrical Penetration Fire Stops and Pressure Saals," Revision 0, dated December 5, 1977
. N
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VII. Immediate or Long-Term Corrective Action M&AI-4 was revised on August 13, 1986 to incorpo ate the reference to i
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M&AI-13.
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,1 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT EMPLOYEE, CONCERNS TASK GROUP OPERATIONS CEG g
l Subcategory: Cable and Conduit Element: Cable Problems in Nanholes l
Report Number: 304.03 - SQN Revision 1 l
IN-85-145-001 l
BNP-QCP-10.35-8-13 I
1 l
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/ 8/![h Evaluater:
P. L. She ord
'Dat's Reviewed by:
A b
'Dat6
/#/L/!f[i Approved by:
xev W. R. Lagedsden
/Dat4 1722T
.o I.
CABLE PROBLEN IN MANHOLES
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This report evaluates the concern that electrical manholes (NH) are in a disorganized state and have water in them.
II.
SPECIFIC EVALUATION NETHODOLOGY This element consists of two concerns identified at Watts Bar Nuclear (WBN)'and Bellefonte Nuclose (BLN) and determined to be generic to Sequoyah (SQN). These concerns were expanded at SQN to include water in the manholes by the Category Evaluation Group Head (CEG-H) from a discussion with SQN QA.
IN-85-945-001
" Electrical manholes are in a very disorganized state.
Cables are laying out of trays with several feet of slack due to cables being spliced and not laced down properly. Examples may be found in the manhole next to the " Fab" shop or manholes in front of the Turbine Building and Auxiliary Building entrance. CI has no additional information. No follow-up required."
I BNb QCP-10;35-8-13 "CI expressed concern about condition of manholes in electrical cable trench to IPS."
The K-forms were reviewed to determine the specific areas of concern..
Referenced documents were researched to determine the requirements, and other historical documents ~ including an NSRS report were reviewed.
Cognizant engineers and craft in electrical maintenance were interviewed and a sample af six manholes was inspected.
This report was completed in accordance with the " Evaluation Plan for the Operations Category Evaluation Group" (CEG).
III.
FINDINGS EVALUATION RESULTS NSRS Report I-85-362-WBN, " Electrical Nanholes," (Reference 1) documents the results of the investigation of concern IN-85-945-001 at WBN.
i The SQN procedure that covers the installation of cables is Mt.AI-4 (Reference 2).
No specific documents were found that control the housekeeping state in electrical manholes.
Cognizant DNE managers and engineers in the Electrical Engineering Branch l
and in the Sequoyah Engineering Project Electrical Unit E2 were l
interviewed with the following results. There do not appear to be any l
specific engineering' requirements that prohibit electrical cables from lR1 being continuously submerged in water. DNE said that manholes that l
contain Class IE cables, especially Voltage Level II (6900 volt), should l
have operational sump pumps. This " requirement" is indicated by sump l
pumps showing on design drawings. As experienced at BFN, Voltage Level V l Page 1 of 5 l
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III. Findings (continued) cable with cross-linked polyethelene insulation (as installed at SQW) l have increased potential for failure after being subjected to excessive 1
moisture for several years due to " water treeing" (reference 3).
Voltage lEl Level Y, Class IE cables at SQN that are run through the manholes that I
were found to have water are Diesel Generator Emergency Raw Cooling l
Water and the Fire Pumps.
l The following seven manholes were inspected by the evaluator:
NH #
DRAWING REVISION DETAIL USE MH 9B 15W810- 5 37 D5 to Diesel Gen. Building.
(DGB) l NH10A 15N810- 5 37 D5 to DGB MH 12 15N810-12 16 A12 to Condenser Cooling Water Intake Pumping Station (IPS)
MH 22 15N810-14 12 B14 to Cooling Tower Pumping Station
{
NH 44 15NB10-40 5
A40 to Office and Power Stores (OkPS)
NH 43 15N810-40 5
.A40 to Const. Whse. & O&PS 1 HH 81 15N810-41 5
C41 to 5th DGB
[
The results of these inspections are as follows:
I HH95 - The cover was about a foot below the surrounding grade and l
had gravel washed across part of one cover. This atnhole i
was not opensd for internal inspection.
j 1
NH10A - This is a double manhole with a common sump and pump, which was in the north section. There was about a foot of water standing in the bottom of both sections which was apparently 5
as low as the sump pump would clear the water. This pump appeared to be functional. There were no cable trays under water. Cables in the south section were r'elatively nest and orderly.
j I
The cables in the north section were loose in some trays and the ends of several abandoned cables were evident in this
+
section.
NH12 - There are four independent taanholes side by side that have I
this number. Two are in a roadway and two are in the The manhole in the roadway closest to the adjacent grass.
curb was inspected. The manhole floor was dry and there was a small amount of water dripping out of a conduit into a small sump. There was no sump pump installed. Individual 3, said during an interview, that at least one of these manholes had recently been pumped out. The cable arrangement in this manhole was neat and the manhole was clean.
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o III. Findings (continued)
The top of this manhole was below the surrounding grade level and water had been draining under the cover into-the MH72 -
There was only about a foot of water in this manhole so the water must have been draining down to the manhole.
There was no permanent access ladder and no next manhole.The cables were neat and the bottom tray was There was scaffolding and ropes in sump pump.
just above the water.
the floor of the manhole.
The top of the manhole was initially found under about three After this water was removed and the MH43 -
manhole was opened the water inside was about six feat deep inches of water.
After removing the (about three feet below the MH cover).
water inside the manhole the cable arrangement wasNone of the cable trays and there were scaffold boards laying across inspected.
This was a non-CSSC manhole.
the track.
This manhole is adjacent to MH43, but it only had about two No further inspections were made in MH44 -
feet of water in it.
this manhole.
This manhole was about six feet square and about five feet It had about two feet of water.in~it with a sump pump HH81 -
Not only did the sump pump not run deep.
that was out of service.
There were but the discharge piping had been disconnected.
This only a couple of cables running through this manhole.
i manhole also had some debris lying on the floor.
There are several The cables in all manholes are coated with mud / dirt.
locations where sump pump controls / junction boxes are located below s These~ controls have been flooded due to power failu level.
f' pump would f ail to restart when the power was restored.
- Examples o pump failures.
ling..
these are the electrical manholes / sumps in the Essential Raw Coo Water (ERCW) Pumping Station; and the valve vault at the have been relocated above the flood area according to cognizant craft Pumping Station.
personnel.
SQN There is no preventative maintenance (PM) or housekeeping program at on electrical manholes and their associated sump pumps.
j Craft personnel interviewed, were asked why someone had not initiat l
The corrective action on the water problems in electrical h of from taking
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time required to effect a change was given as discouragement C
the initiative to request corrective actions.
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..' e CONCLUSIONS Only six manholes and one ERCW pumping station electrical sump were inspected. Based on the findings from these sample inspections and interviews with cognizant Electrical Maintenance personnel, there is a wide-spread problem at SQN with water in electrical manholes and cable sumps.
Cable arrangement in manholes inspected was generally good with isolated areas that could be improved. Several manholes need to be cleaned and debris removed.
Based on the findings, these concerns were determined to be valid at SQN and requires corrective actions. Raising selected manhole covers above ground level and establishing a PM and housekeeping program for electrical manholes and sump pumps should resolve this concern. This concern has potential safaty-ra.ated implications at SQN.
IV.
R0pT CAUSE Ths lack of procedures governing housekeeping and preventative maintenance on sump pumps in electrical manholes has led to the validity of this concern.
Corporate and sito procedures failed to incorporate. requirements to specify and maintain the environment for Class 1E cables'in manholes.
Management inattentiveness to trends of having to pump water out of manholes prior to performing work has prevented these conditions from being reported earlier.
Y.
GENERIC APPLICABILITY, This concern was specific to WBN and could apply to other nuclear plants. These concerns were substantiated at SQN and WBN. Both concerns have generic applicability to all TVA plants.
Page 4 of 5
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VI.
References i.
=
1.
NSES Report I-85-362-WBN 2.
M&A7-4, Installation of Control, Power, and Signal Cable, Revision 9.
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August 13, 1986.
3.
Memorandum from J. A. Teague to H. B. Bounds, subject: Class IE I
Cables Susceptible to Electro Chemical Treeing, July 26, 1982 111 (L23 820727 829).
l VII. IMMEDIATE OR LONG-TERM CORRECTIVE ACTION Electrical Maintenance will assign an engineer to:
l l
(1) Identify manholes to be FM ed, 111 (2) Prioritize manhole list, CSSC first, I
(3), Develop a PM program to address problems, and I
(4): Schedule and perform PMs- (initiate corrective action as necessary).
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OPERATIONS CATEGORY ECTG WRITER'S GUIDE OPERATIONS (OP) ELEMENT REPORTS NOTES /CORNENTS 1.0 Issue I.
Title The brief introduction Characterization Very brief introduction touches on the source of the concerns.
2.0 Summary I.
Title
- 1. The Conclusions are Very brief introduction towards the end of the findings.
III. Finding (Conclusions)
- 2. The Conclusions services as a summary of the findings.
3.0 Evaluators Cover _ Sheet
- 1. The printed name of the original evaluator (s) appears on the cover sheet.
- 2. The peer reviewer and CEG-H approval signa-tures are also on the cover sheet.
4.0 Evaluation II. Specific Evaluation Process Nethodology.
5.0 Findings
III. Findings l
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ECTG WRITER'S GUIDE OPERATIONS (OP) ELNENT REPORTS NOTES / COMMENTS I
6.0 Root Cause IV. Root Cause
- 1. At the element level (Collective Significance) the Root Cause is often not identifible.
- 2. Collective significance is not addressed at the element report level.
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7.0 Attachments /
II. Specific Eva1uation
- 1. A list of the concerns List of Concerns Nethodology with text are included within this section.
- 2. Other reports are sometimes attached.
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l Page 1 of 2 SEQUOYAH EMPLOYEE INITIAL CONCERN ELEMENT REPORTS NUMBER ISSUE Operations 31003 IN-86-055-003 Operations Procedures Need Clarification, Rewritten, and Used Operations 30807 IN-85-948-001 Clam Control Program IN-85-948-002 IN-85-948-003 Operations 31104 I-86-238-SQN Health Physics JLH-86-003 Policies, Practices, JNA-85-001' and Nanagement RII-85-A0064 Control SQP-86-009-001 SQP-86-009-002 WI-85-038-001*
XX-85-009-002 XX-85-015-001 XX-85-026-001 XX-85-028-X02
-XX-85-028-ZO3 XX-85-063-001 XX-85-066-001 XX-85-084-001 XX-85-098-002
- This concern was added.
after 9/26/86.
Operations 30302 IN-85-142-006 Location of Cold-(initial)
IN-85-281-003 leg Accumulator and RWST Level Transmitters Operations 30401 IN-85-112-001 Procedure Problems (initial)
TAK-86-005
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Page 2 of 2 SEQUOYAH EMPLOYEE INITIAL CONCERN ELEMENT REPORTS NUMBER ISSUE Operations 30402 IN-85-207-002 Electrical Penetration (initial)
Breached Operations 30403 BNP-QCP-10.35-8-13 Cable Problems (initial)
IN-85-945-001 in Manholes 4
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