ML20099L402
| ML20099L402 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 09/22/1995 |
| From: | Ajluni M GEORGIA POWER CO. |
| To: | |
| References | |
| OLA-3-I-MOSB-95, NUDOCS 9512280229 | |
| Download: ML20099L402 (44) | |
Text
- /
Z-Moss n -95 IS/nL Intcr0ffice C:rr sp:ndence Georgia Poner d 000,ETED p
rf DATE:
August 23, 1990
'95 gi 20 Pa :58 RE:
Voatle Electric Generatina Plant Y..-~
F,page !
of NRC Areas of Concern OF F '
Exhibit Log: SR8S-00044 CC c.
Security Code:
NC FROM:
M. J. Ajiuni TO:
Distribution Attached for your information is a copy of the recent NRC Operational Safety Inspection areas of concern and VEGP's position.
s' M. J. Ajluni MJA/ghj cc:
W. B. Shipman G. Bockhold, Jr.
S. C. Ewald L. K. Mathew C. C. Miller /s W. C. Ramsey P. D. Rushton NORMS SRB File 9512280229 950922 fg D EA/B PDR ADOCK 05000424 G
PDR YM.gh6~
NUCLEAR REGULATORY COMMISSION Dock.t No. 50424/425-OLA-3 EXHIBITNO. T
- iI in en matur of o. orate Power co..t at.. voatt. Units 1 a 2 92 PRO.TECT O staff O Applicant Effntervonor O Oewr 045529 O identifi.d p'.ived O R $cmd Reporter O m3 Dem $ /M/Tf witn
n AUGUSl 17. 1990 AREAS OF CONCE9HS MRC VEGP CONTACT CORPORATE CONTACT o
D/G Records Starts / Failures Pete Taylor G. Frederick 0
l 3/1/90 S R Monitor Inop Mode Change Heal Hunemuller JES/D. Carter
- Missed Sury. Cont. Isol.
Neal Hunemuller JES/S. Swanson
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o March 15 RHR Train 8 Ron Aiello JES/J. Gasser P. 9'. Rushton o
Temp. Change Notice to AOP Robert Carrol JES/J. Cash 18028-C-7-90-1 ESFA Sequencer Out of Service Robert Carrol JES/Horton J. A. Bailey o
o Alternate Radwaste Building Ron Aiello Ron LeGrand/JES P. D. Rushton
- Snubber Reduction /
Larry Garner Gus Williams Ward /Stringfellow
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LCO Action Statement Cont. Integrity Hydrogen Monitor Morris Branch Dean Gustcfson Ward /Stringfellow Valve Opened o Precision Heat Balance Morris Branch Gus Williams B. Florian o Personnel Accountability
'C. VanDenburgh JES/GB Methodology for Reporting o Tech. Spec. 3.0.3 Philosophy J. D. Wilcox J. E. Swartzwelder J. Stringfellow o ESFAS Reportability J. D. Wilcox R. M. Odom J. A. Bailey Plant Review Board (PRB) Composition C. VanDenburgh G. Bockhold 0
Ya g,o Tech. Specs. Interpretation Morris Branch J. E. Swartzwelder J. Stringfellow u :0
- Overtime / Training & Qualification Larry Garner J. E. Swartzwelder O
- Electrical Separation Zone 80 Larry Garner M. Horton P. D. Rushton T. S. 3.4.7.3 CCW J. D. Wilcox J. E. Swartzwelder
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- Training Department Comments on OSTI
- Shift Experience
- Plant Equipment Operator Morale
- Shtft Communications
- Analyzer Operation Following St
- Chilling Affect / Intimidation of PRB Members
- Quality concern Program
- Conflicting Statements
- While Containment Cooler is Inoperable DGIB is Rendered Inoperable
- Exit of Diesel Generator LCO h
i t
3 2,
38 2M 4
l L.
i DIESEL STARTS AND FAILURE REPORTING 8/22/90 Time:
13:00 l
3 I
l Page 1 of 2
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NRC Concern I
1.
The NRC is concerned about the incorrect number of diesel starts reported 3
in LER 1-90-06 and the number of starts presented to the NRC on April 9, l
1990 and in the confirmation response letter of April 9,1990. The major issue remaining. is to try and determine through personal interviews, how the number of 19 for diesel 18 was arrived at in the April 9 letter to the
^
NRC.
The NRC believes the intent of the April 9 letter and the presentation discussed consecutive successful starts.
The revised
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4 j
response to LER 90-06 did not clarify the number of starts reported to the
^
NRC April 9, and did not clarify that the 19 starts were not consecutive.
i 2.
The inspector noted that documentation provided by Operations to support i
diesel trending (14980-C and 13145-C data sheets) does not contain an i
adequate description of what happens during the start attempt.
The plant -
is not interpreting Reg Guide 1.108 properly with regard to reporting l
valid and non-valid failures.
There may be valid and non-valid failures l
that were not reported.
The NRC does not consider the current status of reporting diesel failures to be in compliance with connitments made to the j
NRC in Violation 50-424/87-57.
i NRC Documentation The NRC has reviewed the diesel start log and supporting documentation (14980-C and 13145-C data sheets).
The NRC currently believes some problems i
identified on 14980's and 13145's should be classified as non-valid failures i
and reported to the NRC.
The NRC has requested and received written analysis to explain the disposition of the following IB diesel starts:
- 's 123, 124, 132, 133, 134, 136, 160, 161, 162, 164, 165, and 190.
LER 1-90-06, revision l
1; QA Audit Report OP26-90/33; QA Audit Report OP09-90/31; and Special Report 1
1-90-05, dated August 7, 1990; GPC confirmatory action letter dated April 9, l
1990.
VEGP Position l
1.
The error made in the number of diesel starts reported to the NRC on April 9, 1990, and in LER 1-90-06 is attributed to two factors:
a.
The testing as described in LER 90-06, revision 0,
was in the i
" context of" and "in reference to" the diesel control systems.
The first two sentences of the 5th paragraph explain actions taken with regard to sensor calibrations and control system testing.
In this i
context, the test nrocram correlates to testing discussed with the NRC on April 9,1990, and reported in the April 9,1990, confirmatory letter.
The LER 90-06 comment of " subsequent to the test program" was not intended to exclude successful diesel starts before declaring the diesel operable. As a result, diesel starts after testing of the control
- systems, but before a
declaration of operability were l
counted. The transmittal letter for LER 90-06, revision 1, describes the confusion and attempts to clarify the concern by redefining the types of starts and the point of counting.
92 PROJECT 045532 l
l DIESEL STARTS AND FAILURE REPORTING
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Page 2 of 2 b.
LER 90-06, revision 1,
was intended to clarify any inadvertent l
l
" misleading" of the NRC on successful operation of the diesel control i
systems.
Mhen Vogtle Management was aware of the problem in LER j
90-06, revision 0, management notifled the NRC Residents.
Also at 4
the corporate office on 6/11/90, M. Shipman contacted Ken Brockman and on about 6/11/90, H. G. Hairston, III, contacted Mr. S. Ebneter of NRC Region II. The revised LER was submitted on 6/29/90.
The 19 starts discussed on April 9 were based on operator assessments of the starts as successful using VEGP procedures. Additional review
~
of these starts by both the NRC and Vogtle personnel indicates start
- 134, performed on March 23, 1990, could be counted as unsuccessful.
If start #134 is not counted, only 14 successful starts occurred before April 9,1990.
This start will be reviewed in detail and an appropriate report to clarify the number of starts reported April 9 j
1990 will be made.
i i
i 2.
After a thorough review of Reg Guide 1.108 Engineering Support (Mike Norton) agreed that all diesel start nroblems have not been reported as failures.
GPC's response to NRC Violation 424/87-57 committed to report t
such equipment problems as failures;
- however, due to internal administrative problems, the commitment was not implemented.
Engineering Support intends to review diesel start records for any unreported failures.
i VEGP Documentation o
LER 1-90-06, revision 1; OA Audit Report OP26-90/33; QA Audit Report l
OP09-90/31; and Special Report 1-90-05, dated August 7,
1990; GPC confirmatory action letter dated April 9,1990.
o 18 diesel start analysis available 8/15/90 and Reg Guide 1.108 position i
from Engineering Support.
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92 PROJECT 045533
8/22/90 Response to NRC Questicn Concerning Time:
13:00 Olesel Starts Reported on April 9,1990 j
and in LER 90-06, Revisions 0 and 1 t
]-
Question #1 1.
Who prepared the slide for the 4/9/90 presentation?
Answer:
G. Bockhold, Jr., J. P. Cash, and K. Burr working as a group.
i 1
2.
Who approved use of the slide?
i.
Answer:
G. Bockhold, Jr.
1 Ouestion #2 i
j 1.
Who prepared the confirmatory letter of April 9, 19907 j
Answer:
C. K. McCoy, J. A. Bailey, H. G. Hairston, III as a group.
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2.-
Who approved the letter?
Answer:
H. G. Hairston, III j
Ouestion #3 (with regard to LER 90-06, revision 0, dated 4/19/90) 1.
Who prepared the LER?
Answer:
Several draft revisions of the LER were prepared by Tom Webb and others of the NSAC group of the Vogtle Site Technical Support.
These drafts were reviewed and commented on by the Plant Review l
Board.
The final revision of LER 90-06, revision 0 was prepared by a phonecon between site management and corporate management.
Those participating are believed to be G. Bockhold, Jr., A. L.
l Nosbaugh, J. G. Aufdenkampe, H. Shipman.
2 i
2.
Who reviewed the LER?
Answer:
All revisions of the LER were reviewed by the PR8 and the j
General Manager-Plant Vogtle.
L l
3.
Who approved the LER?
Answer:
The LER was approved by H. G. Hairston, III a
i Ouestion #4 i
l 1.
Who prepared the cover letter for LER 90-06, revision 17 Answer:
The cover letter was prepared by H. H. Hajors of the corporate l
staff.
This letter was prepared under the guidance of H. G.
Hairston.
J.
2.
What was the purpose (intent) in the wording of the cover letter with regard to the number of diesel starts?
Answer:
The cover letter was intended to document discussions with NRC Region II to clarify the starts documented in LER 90-06, l
revision 0.
By picking a well defined. point to specify
" subsequent to the test program" it was possible to identify a i
substantial number of successful diesel starts.
This was intended to remove any additional ambiguity.
Ouestion #5 1.
Who in corporate added the words " subsequent to the test program" in LER l
93-06, revision of Answer:
Corporate Licensing personnel in conjunction with the phone conversation described above made editorial
- changes as F
directed.
Those present during the phone conversation are thought to be H. Shipman, G. Bockhold, Jr., A.
L. Hosbaugh, J. G. Aufdenkampe, and J. Stringfellow.
92 PROJECT l
045534 4
i
Page 1 of 2 LER 1-90-004 i
Original NRC Concem To determine if Technical Specification 3.0.4.
was violated, when Unit 1 entered Moe's 6 from Mode 5 while Source Range IN31 was out of service for an 18 month calibration, solely for the purpose of progress on the critical path schedule; if the shift was subsequently congratulated for making that progress; and if-the Shift Superintendent demonstrated a willingness to violate Technical Specifications for the sake of schedule.
Found to be unsubstantiated NRC Concern The inadequacy of the root cause determination and corrective actions of LER 1-90-004, in that, hisnan factor problems involving the LC0 sheet may have contributed to the Shift Superintendent's failure to note the LCO mode change restriction.
NRC Doctanentation Technical Specification 3.9.2 Deficiency Card 1-90-0050 LER 1-90-004 12007-C Refueling Entry (Mode 5 to Mode 6)
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Unit 1 Shif t Supervisor Log (2/28/90 and 3/1/90)
Unit 1 Control Log i
LC0 Status Sheet 1-90-152 LCO Log (10008-C P.8 of 11 dated 2/28/90) i 14000-1, Operations Shift and Daily Surveillance Loos, dated 2/28/90 i
1R2 Outage Schedule (actual vs. schedule)
Turnover Checklist (11870-C dated 2/28/90 j
Completed Procedures, dated 2/28/90 - 12007-C, 14000-1, 11871-C and 11872-C I
VEGP Position 2
YEGP's position i s that hinan factor problems with the LC0 sheet was not a significant contributing causal factor in this event.
- However, due to a number of human factor concerns noted during the IR2 refueling outage, VEGP has revised procedure 10008-C twice, to enhance usability 2
and htman factoring.
Furthermore, VEGP will review Procedure 10008-C to determine if further enhancements are warranted.
4 r
4 92 PROJECT 045535
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,.,m.
..,...-.4
.,... ~, -
Pag 3 2 of 2 i.
LER 1-90-004 1
i 4
VEGP Documentation i
LER 1-90-004 12007-C, fiefueling Entry (Mode 5 to Mode 6)
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Unit 1 Control Log, 2/27/90 to 3/2/90 Unit 1 shift Supervisor Log, 2/27/90 to 3/2/90 Unit SS Relief Check 11'sts, 2/28/90 and 3/1/90 Support SS Relief Checklists 2/28/90 and 3/1/90 Operations Supervisor Relief Checklist, 2/28/90 and 3/1/90 R0 Relief Checklists, 2/28/90 and 3/1/90 B0P Relief Checklists, 2/28/90 and 3/1/90 14000-1, opentions Shift and Daily Surveillance Loos, 2/28/90 4
1R2 Outage Schedule, 2/23/90 thru 3/3/90 10008-C Recording Limitine Conditions for Operation Rev.12 i
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A e
92 PROJECT 045536 i
1 UNIT 2 LER 90-001 08/22/90 PAGE 1 0F 2 08:45 MISSED SURVEILLANCE ON
" CONTAINMENT INTEGRITY VERIFICATION VALVES OUTSIDE CONTAINMENT" NRC CONCERN:
Required Tech Spec actions may have been delayed by initiating an investigation.
Was management pressure a contributing factor?
Concern:
Potential concealment of correct Tech Spec LCD entry time to prevent a forced shutdown and immediate notification of the NRC.
I Finding:
The correct T.S. LCO entry was not concealed.
Concern:
Cause for confusion over the Surveillance Task Sheet.
Finding:
The cause for the confusion was an inconsistent use of equipment identification numbers on these sheets. Corrective actions adequate.
l Concern:
Extent 6f emphasis on keeping the plant in operation and ifmiting NRC notifications.
l Finding:
There was no indication of unreasonable emphasis on keeping the plant in operation or limiting NRC l
notifications.
i NRC DOCUMENTATION:
D.C. 2-90-022 Surveillance 14475-201 Jan 3,1990 Feb 1,1990, Feb 28, 1990 Unit II LER 90-001 Control Room Logs from Feb 27 and Feb 28, 1990 NRC Inspection Report 90-10 92 PROJECT 045537
I.
s."
RHR PLW 18 VIBRATION Original NRC Concern A non-conservative decision was made concerning the operability of the 18 RHR pump in order to avoid substantial impact to the outage critical path schedule Found to be unsubstantiated NRC Concern A Deficiency Canf was not generated in a timely fashion concerning the IB RHR pimp cooler leak and elevated vibration levels.
NRC Doctmentation Unknown VEGP Position VEGP concurs that a Deficiency Card was not generated in a timely fashion.
Since the occurrence of this event, VEGP management has taken positive action to improve the effectiveness of the Deficiency Ca rd Program.
These improvements include:
l 1.
Revision of Reactor Trip Review Procedure 10006-C to specifically require a sign-off indicating a Deficiency Card has been written.
2.
Address by General Manager to the PRB stressing the necessity for timely Deficiency Card generation, and memo to all managers from the Technical Support Manager stressing the requirements for timely submittal of Deficiency Cards.
In addition, the Deficiency Card Program has received increased management attention and oversight to ensure Deficiency Cards are generated in a timely fashion.
This will ensure that operability and reportability determinations and appropriate engineering evaluations are performed.
VEGP Doctmentation Letter, Manager Technical Support to Department Managers dated 6/22/90.
92 PROJECT 045539
~.
1.
l RHR Ptw 18 VIBRATION Original NRC Concern 4
i A non-conservative decision was made concerning the operability of the 18 RHR pump in order to avoid substantial impact to the outage critical path schedule j
Found to be unsubstantiated 1
NRC Concern A Deficiency Card was not generated in a timely fashion concerning the 18 RHR pump cooler leak and elevated vibration levels.
i l
NRC Docisnentation 1
Unknown VEGP Position VEGP concurs that a Deficiency Card was not generated in a timely fashion.
Since the occurrence of this event, VEGP management has taken positive action to improve the effectiveness of the Deficiency Ca rd Program.
These improvements include:
i 1.
Revision of Reactor Trip Review Procedure,10006-C to specifically require a sign-off indicating a Deficiency Card has been written, j
2.
Address by General Manager to the PR8 stressing the necessity for timely Deficiency Card generation, and memo to all managers 4'
from the Technical Support Manager stressing the requirements for timely submittal of Deficiency Cards.
1 In addition, the Deficiency Cand Program has received incrtased management attention and oversight to ensure Deficiency Cards are generated in a timely fashion.
This will ensure that operability and reportability determinations and appropriate engineering evaluations are performed.
4 VEGP Doctamentation l
Letter. Manager. Technical Support to Department Managers dated 6/22/90.
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92 PROJECT 045539
.... o. w-,o Page 1 of 2 IMPROPER TCP PROCESSING 3
j i
i NRC Concern i
j TCP 18028-C-7-90-1 was "back-dated" to avoid violating section 6.7.3.c of Technical Specifications.
NRC Doctmentation
-l TCP 18028-C-7-90-1 DC 1-90-282 j
DC.1-90-283 i
PR8 Minutes 90-81 and 90-82 i
Procedure 00052-C 1
VEGP Position A violation of Section 6.7.3.c of the VEGP Technical Specifications occurred.
Howeve r, the cover sheet of TCP 18028-C-7-90-1 was not dated 6-12-90 to avoid this violation.
i TCP 18028-C-7-90-1 (written against Rev. 7 of the permanent plant procedure), Loss of Instrunent Air, was approved by the Operations Manager on 5 90.
On 6-8-90 the PR8 tabled this TCP to allow the Operations 3
Depa rtment to determine if additional instructions for Modes 3,
4, 5,
j and 6 should be added to the revi sion to strengthen the AOP.
Revi sion 8 of the permanent procedure was prepared by Operations and approved by the PR8 on 6-12-90. This revision addressed both TCP 18026-C-7-90-1 and additional instructions for a Loss of Instrument Air in Modes 3, 4 5,
and 6.
The Acting Operations Manager understood that the TCP would not be used in the field once Rev. 8 was issued.
Upon approval of Rev. 8 of the permanent procedurt by the PRB, verbal instructions were given by the Acting Operations Manager to the procedure coortlinator to void TCP 18028-C 90-1.
The TCP was next in the procedure coordinator's possession on 6-15-90.
On that date the acting Operations Manager signed the TCP cover sheet and dated it 6-12-90 to reflect his understanding, i
based on discussions with the procedure coordinator and his verbal instructions of 6-12-90, that the TCP was voided on 6-12-90.
The Acting Operations Manager assisned that the approval of Rev. 8 of the permanent procedure (which he assumed occurred on 6-12-90) resulted in the voiding of the TCP, and that his vertal instruction to the Operations staff was adequate to close-out required paperwork.
T'is was an error n
and risulted in a failure to comply with Procedure 00052-C, Section 4.6.2 i
in a timely manner.
GPC notes that the minutes of PR8 meeting 90-82 indicate the TCP "was voided" on 6-12-90 which reflected the understanding 4
of the Acting Operations Manager.
A 92 PROJECT 045540 I
1600 8-1 6-90 Page 2 of 2 l
i l
IMPROPER TCP PROCESSING i
l On 6-18-90, the Operations Manager instructed the Acting Operations Manager to write a DC on the inaccurate dating of the TCP close-out sheet and a failum to meet the 14 day period set under Procedure 00052-C Section i
3.2.4.
This instruction was a result of normal Technical Support Group i
i, riview and verification close-out of TCPs.
On 6-22-90 two DCs were written by Technical Support.
On the same i
date the (former) Acting Operations Manager, in preparing a DC on the inaccurate dating of the TCP cover sheet, determined that the permanent procedurw. Rev. 8, was not issued until 6-13-90, that the TCP was pulled from the Control Room on 6-13-90 and that the 14 day limit under Procedure l
00052-C had been exceeded.
i 92 PROJECT 045541
c.
t SEQUENCER IMOPERABILITY J
NRC Concern Inadequate infonnation exists for shift personnel to determine which Technical Specification to apply if the sequencer is inoperable.
In addition, previous verbal guidance was inadequate..
NRC Doctanentation 1.
Sequencer related work orders 4
2.
Previous sequencer LCO sheets 3.
Control room narrative logs 4.
Sequencer related surveillances VEGP Position The NRC position is accurate in that no Tech. Spec. interpretation exists.
Previous guidance connected sequencer inoperability to diesel generator inoperability. Recent infonnation has demonstrated that sequencer inoperability should also be tied to " actuation logic and actuation relays",
as found in the instrisnentation specifications.
VEGP will further review i
and eval ua te this issue to ensure an adequate interpretation exists for the shift personnel.
i Ne intend to develop a clear interpretation with input from Operations, Engineering, and Maintenance.
Based on that development a Tech Spec i
Interpretition will be written and a Tech Spec change requested to clarify this issue.
VEGP Doctsnentation As above 9
92 PROJECT 045542
8/17/90 i
(
10:00 Paga 1 of 2
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ALTERNATE RADWASTE BUILDING NRC Concern Concerned that the FAVA system was installed without performing adequate engineering l
and safety evaluation (50.59), because the fabrication and quality of the system did not meet the RG-1.143 and ASME code requirements.
Concluded that the FAVA system was originally installed without an adequate j
safety evaluation. As a result of a VEGP QA finding in early 1989 involving a breakdown in the procurement and failure to meet FSAR coseitments, the system was removed from service.
Subsequently the system was returned to service following two SEs (dated 11/89 and 2/90) which adequately addressed the use of PVC piping with respect to radiation degradation and pipe rupture. Although these SEs did not address the effects of a break in the hoses (which could j
result in wall spray down or leakage), the use of hoses and effects of hose 4
breaks (i.e., airborne activity and puddling) were addressed in SER Supplements 3 and 4.
Although these SEs did not address high temperature effects our interview indicated that these effects were considered in performance of the SE.
i Concluded that the SE performed on 6/90 at the request of RII to evaluate j
the effects of a FAVA system wall spray down and wall leakage to an unrestricted area have been adequately addressed for the use of the FAVA system, because j
the FAVA System has a protective cover.
However, the June 90 safety evaluation inadequately addresses the potential effects of wall spray down from any other j
source in the ARB due to erroneous assumptions concerning the release path and the dilution volumes.
This is a potentially unreviewed safety question concerning the use of the alternate radwaste building. And as such will be j
followed as an unresolved item pending further review and evaluation.
(Unresolved i
item concerning unreviewed safety question).
NRC Documentation IEC 80-18 SSER 3 Section 11.4 SSER 4 Section 11.4 l
SSER 8 Section 11.4 l
VEGP Position The safety evaluation for the FAVA microfiltration system was adequate for use of the system.
The calculation performed to evaluate the " spray accident" in the ARB was flawed due to erroneous assumptions regarding the release path.
These flawed assumptions do not affect the 50.59 evaluation made for the FAVA unit.
The analysis of the " spray accident" in the AR8 should not have been included as a revision to the safety evaluation in the FAVA unit.
Doing so confused the:;e two separate issues and was not appropriate.
4 92 PROJECT 045543
8 /17/90 10:00 Page 2 of 2 During the plant licensing process, details of the construction of the AR8 were provided to the NRC. NRC personnel visually inspected the as-built condition 1
of the AR8 and associated solid waste processing steams and interface connections that tie other support systems to the equipment in the AR8.
Flexible hoses and couplings were in use in the facility at the time of this inspection.
The NRC found the facility acceptable for use based on our submittal and their visual inspection with one exception which was subsequently corrected and had to do with exhaust air filtration.
This information is well documented in the SER Supplements 3, 4, and 8.
NRC concerns with wall spray-down from the FAYA System have been adequately addressed by the installation of a protective cover.
VEGP considers previous analysis and NRC SER Supplements adequate with respect to the AR8.
Since the configuration of the ARB and the hoses in question is the same as that addressed by such previous analysis and SER Supplements, no unreviewed safety question exists.
VEGP Documentation Safety Evaluation dated 2/26/90 Other documentation same as NRC 4
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1 92 PROJECT 045544
j D:tes August 22, 1990
' i Time:
1300 t
i Snubber Reduction and Use of the Anoropriate jl Technical Specification and LC0 Action Statement Entry to Imolement Desian Chance i
4 NRC Concern j
Voluntary entry into a LCO action statement is acceptable for the purpose of surveillance testing but is discouraged for modification work.
(See NRC internal j
correspondence Murley to Martin dated May 18,1990).
NRC has determined that applying the action statement associated with Specification 3.7.8 and then applying i
the action statement of the applicable system is a correct interpretation of 1
Technical Specification requirements. Specification 3.7.8 is intended for broken j
snubbers or functional testing and not for other purposes. With respect to snubber i
reduction you must have a valid safety evaluation which considers the ramifications l
of performing the modification at power.
If the modification renders the system inoperable during the installation, as determined by analysis, then the applicable i
system action statement must be applied.
Entering a LC0 action statement should represent a net safety benefit and be i
warranted by operational necessity, not just by convenience. The practice should not be abused by repeated entry into and exit from the LCO.
Implementation of i
snubber reduction during power operation is non-conservative.
(
l NRC Documentation 1
LCO's Safety Evaluations for DCP's 88-VIN 0ll4, 89-VIN 0047 Copies of MWO's for NSCW "A" Train Letter from Murley to Martin, dated 5/18/90 i
Letter from W. C. Ramsey to C. C. Miller, dated 8/15/90 j
Letter from Denton to Norelius, dated 5/27/90 j
VEGP Position j
Voluntary entry into a LC0 action statement for the purpose of implementing a design
{
change is acceptable provided the activity is accomplished within the provisions of 1
the Technical Specification and proper consideration has been given to the impact on j
plant safety.
This position is supported by NRC Standard Technical Specification i
interpretation which actually endorses voluntary entry into an action statement condition on the basis that the NRC "has structured the Technical Specifications to permit the licensee to exercise judgement within the latitude permitted by the 4
Action Statement language in the Technical Specification".
4 i
VEGP implements design changes on safety related systems for the purpose of improving system reliability and thereby enhancing plant safety.
VEGP maintains that voluntary entry into a LC0 action statement to implement a design change is acceptable and desirable in specific cases.
VEGP considers this consistent with industry practice.
a 92 PROJECT 045545 i
i
l VEGP agrees t:1th the NRC position that applying the action statement assreinted with specification 3.7.8 then applying the associated system action statement is the appropriate way to implement Technical Specification requirements.
In addition, VEGP agrees that when performing a modification a Safety Evaluation must be l
performed.
If the evaluation determines the system is rendered inoperable during l
the installation process then the action statement associated with the system must be followed and 3.7.8 cannot be applied.
Since snubber reduction increases system reliability by eliminating potential failure modes, implementation of snubber reduction during power operation is j
conservative.
VEGP Doc eentation
)
LC0's Safety Evaluations for DCP's 88-VIN 0ll4, 89-VIN 0047 Copies of MWO's for NSCW "A" Train STS, Section 3.0 Voluntary Entry into Action Statements, dated 1/1/82 Letter from Murley to Martin, dated 5/18/90
. Letter from W. C. Ramsey to C. C. Miller, dated 8/15/90 i
Letter from Denton to Norelius, dated 5/27/90 l
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1 92 PROJECT 045546 4
d 8/16/97 1600 Page 1 of 4 CONTAINMENT HYDROGEN MONITORS CONCERNS:
I.
OPENING VALVES AT POWER i
II.
ANALYZER OPERATION FOLLOWING SI j
I.
OPENING H2 MONITOR VALVES AT POWER
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NRC QUEST!0N Are the following valves considered containment isolation valves?
HV-2792A i
HV-27928 HV-27938 HV-2791B NRC CONCERN i
NRC. feels they are based on:
FSAR Table 6.2.4-1 i
FSAR Table 16.3-4 YEGP POSITION ies, the above identified valves are containment isolation valves for the A-Train Containment H2 Monitor.
NRC QUESTION If they are containment isolation valves does Tech Spec 3.6.3 apply to the operation of these valves?
NRC CONCERN NRC feels Tech Spec 3.6.3 applies for the following reasons:
)
i
- Tech Spec 3.6.3 applies to containment isolation valves.
i
- We say they are containment isolation valves in Tech Spec j
i interpretation to 3.6.3 (1-18-90).
- Operations Procedure 13130-2 page 4 " Caution" statement.
- Maintenance Procedure 24932-2 step 3.2 " Prerequisites or Initial Conditions"
- FSAR 6.2.4.2.3 J
i VEGP POSITION 1
Tech Spec 3.6.3 applies to containment isolation valve operability.
Opening these valves to perform channel calibration of the H2 analyzer does not render the valve inoperable and therefore Tech Spec 3.6.3 is not entered.
If one of these valves became inoperable (e.g. would not close leaked excessively, etc.), then Tech Spec 3.6.3 would apply to that valve and the associated containment penetration.
Tech Spec 3.6.3 interpretation (1-18-90) applies to hydrogen monitor valves as described above.
Procedure 13130-2 confirms the way we want to operate these valves.
i 92 PROJECT 045547
i.
U/lb/90 1600 Page 2 of, Procedure 24932-2 was only recently revised to i clude the reference to the LCO condition.
The LC0 condition was with reference to breaching j
the piping boundary outside containment.
FSAR 6.2.4.2.3 states these essential Ifnes are normally closed and remain closed during power operation.
The configuration of these valves are normally closed during power operation. Opening these valves to perform calfbration does not conflict with the system description in the FSAR.
CALIBRATION PROCEDURE 24551-2 NRC QUESTION j
Do we feel step 2.9 still valid, precautions / limitations "may be performed
~
j in any plant mode"?
VEGP POSITION 1
Yes, precaution 2.9 is valid.
The procedure for H2 monitor calibration is required to be performed every 92 days on a staggered test basis (i.e.
one channel must be tested approximately every 46 days), per Tech Spec 4.6.4.1.
Thus Tech Specs recognize this surveillance as one that can j
be and should be perforined at power.
NRC QUESTION Do we feel it is necessary to open the isolation valves to perform calibration?
4 VEGP POSITION Yes, for the following reasons:
- 1) By estab11shing a flow path to and from containment we are verifying an open flow path extsts.
- 2) Verification that pump will operate in the normal flow path configuration i
is confirmed.
- 3) ALARA concerns associated with positioning the vent valves.
i
- 4) Risk associated with vent valve manipulations.
NRC QUESTION l
Was operations involved in review and approval of the procedure?
i VEGP POSITION
{
No t
NRC QUESTION f
Was the 50.59 safety evaluation performed adequately?
VEGP POSITION Yes e
8 i
(
92 PROJECT 045548
---4---
m
- - ~--
~ - - - - -
1600 Page 3 of
.a NRC QUESTION When this issue was pointed out 8/7/90 on Unit 2, why was test performed on Unit 1 the next day?
VEGP POSITION When the issue was brought up with the Operations Manager, he began gathering i
information on the issue. At the time he felt our procedures were correct j
and did not review the next day's activities.
When shift personnel were
)
made aware of the NRC concern the test was terminated imediately.
1 4
NRC QUESTION What Tech Spec requires LLRT testing of the system?
24910 24930 24931 24932
/
24933 VEGP POSITION Pmcedures 24930, 24931, 24932, 24933 satisfy the requirements of Tech Spec 4.6.1.2.d for components defined in FSAR Table 6.2.4-1.
Procedures 24910 and 24932 satisfy the requirements of leakage assessment of Tech j
Spec 6.7.4.A.
NRC QUESTION Is leak rate testing performed on these containment isolation valves
}
added to overall containment leak rate?
VEGP POSITION Leak rate for these isolation valves is added to the total type B and C leakage.
It is not added to type A results.
j NRC QUESTION j
Evaluate applicability of Tech Spec 6.7.4.A to this system VEGP POSITION The piping outside containment is covered under the leakage assessment program as addressed in Tech Spec 6.7.4. A.
NRC QUESTION i
Do we feel we violated Tech Spec 3.6.3 on the following two occasions?
1 Unit 2 0411 8/6/90 to 0122 8/7/90 21 hrs. 11 min.
Unit 1 2053 8/7/90 8/8/90 18 hrs. 47 min.
VEGP POSITION No, the containment isolation valves were not inoperable on these dates.
Further, in the past calibrations have been scheduled in accordance with 4
Tech. Spec. requirements (approximately every 90 days).
92 PROJECT 4
045549 4-
l'0h'"Page4of 6
ANALYZEhi DPERATION FOLLOWING SI NRC QUESTION Are the analyzers placed in service 30 minutes after a safety injection?
Requirement: NUREG 0737 IIF1 Attachment 6 TMI requirement Provide station position relative to NUREG 0737 also provide proof of implementation.
NRC has looked the following places:
19000-C 19251-C Loss of primary & secondary coolant Reference SER 6-4 i
VEGP POSITION Procedure 19010-C (Loss of Primary or Secondary Coolant) step 12 currently addresses obtaining containment H2 samples following an SI.
VEGP intends to enhance this procedure relative to placing the H2 analyzers in service for this purpose.
i a
p 4
I h
M i
92 PROJECT 045550
Date: August 16, 1990 Time:
1600 Precision Heat Balance j
NRC Concern The corrective actions associated with LER 90-015 (Failure to Calibrate Computer Points Prior to Precision Heat Balance Flow Measurements) are technically correct, the decision not to re-perform the surveillance test was non-conservative.
4 NRC Documentation LER 90-015-00 DC cards, RCM's Reactor Engineering Calculations Completed test procedures t
VEGP Position The decisioi: not to reperform the surveillance was conservative, based on l
Engineering eveluation of available data. Additionally, all associated data was reanalyzed assuming potential calibration errors.
This reanalysis j
verified the initial engineering analysis.
l When Unit I final feedwater temperature instrumentation was determined to be out of calibration, examination of the data indicated sufficient margin to address the out of calibration condition.
Reanalysis of the data considering the out of calibration condition confirmed the conclusion.
For Unit 2 the calibration of final feedwater temperature was never suspect. Again, examination of calibration data taken after the test indicated sufficient margin to address potential miscalibration problems.
Reanalysis of the data also confirmed this conclusion.
VEGP Documentation None 1
4 92 PROJECT 045551 4
=_-
8/16/90 1600 1
PERS0MIEL ACCOINTABILITY NRC Concern Holding shift supervision accountable for the number of reactor 4
trips. LER's, and ESFAS actuations has a potential negative influence on plant safety because personnel might not be open about reporting these types of plant problems.
NRC Documentation Typical Shift Superintendent Accountabilities VEGP Position i
These accountabilities enhance reactor safety because they focus personnel attention on safety and compliance issues.
Reporting problems is required to achieve good SALP ratings and is part of shift supervision accountabilities.
VEGP Documentation 4
Typical Shift Superintendent Accountabilities 1990 Organizational Goals Perfonnance Appraisal Forms l
4
}
92 PROJECT 045552
a x
-a a
a 4
3.0.3 1 HR. ACTIONS NRC Concern Inadequate documentation exists to demonstrate all actions taken during the first hour af ter entry into Technical Specification 3.0.3.
NRC Doctmentation Control Rocci Narrative Logs
.YEGP Position Doctmentation of all actions taker: during the first hour after entry into Technical Specification 3.0.3 does not exist as stated by the NRC.
However, this information (doctmentation) is not procedurally required, nor is it a regulatory compliance issue.
Appropriate actions have been taker. in the past 3.0.3 entries to meet the time table cf the action statement.
VEGP Doctmentation As Above b
92 PROJECT 045553 i
= _... _ _ _ _
Date: August 16, 1990 i
Time:
1600
[EAS REPORTABILITY ISSUE 4
l NRC Concerns In regard to reportability of ESF actuation NRC has developed a position that "If, for any reason (except expected responses to testing) the ESF components are caused to operate, then an ESF actuation did occur".
This position was formulated in response to GPC Corporate internal meno (June 11,1987) which provides guidance concerning ESF actuation reportability.
l-5 NRC Documentatim i
1.
Internr.1 NRC memo dated July 12, 1990 from Charles E. Rossi to Gua C. Lainas.
I 2.
Internal GPC memo dated June 11, 1987 from R. Baker to
}
L. T. Gucwa.
VEGP Position The June 11, 1987 letter was written by a member of the Corporate Nuclear Safety and Licensing Department to his Supervisor.
The information contained i
in the attachment to this letter was intended to be used as guidance when i
determining ESF actuation reportability.
This information was never adopted l
by the Vogtle Electric Generating Plant (VEGP) to be used for reportability guidance.
The Vogtle practice has been to report the ESF actuation regardless l.
of "what caused the actuation' or "how the actuation occurred".
Based on i
discussions with individuals who review deficiencies for reportability and a j
review of past deficiency eval M ions identified no instances where the position described in the June 11, 1987 letter was utilized in ESFAS L
reportability determination at VEGP.
VEGP Documentation 1.
W. F. Kitchens memo to 0505 dated June 9, 1987 RE: ESF Actuation 2.
Sort of DC's by keyword "ESFAS" 4
3.
List of ESFAS LER's for Vogtle k
4 92 PROJECT 045554 1
6
~
l'
\\
i 8/16/90 1600 PUWT REVIEW BOARD (PRB) C(NOSITICII George Bockhold Jr.
NRC Concern The Tech Specs may not allow the Manager Technical Support to be the PRB member for both Quality Control (QC) and Nuclear Safety &
Compliance (NSAC).
NRC Documentation Tech Spec 6.4.1.2 - PRB Composition Tech Spec Interpretation - 6.4.1.2 VEGP Position When tb Company restructured the PRB, we raised the Level of the PRB members to be line managers to enhance board activities.
Our plant i
organization had been restructured, and the Manager Tech Support was responsible for both QC and NSAC, and should be their representative on the PRB.
VEGP Documentation Tech Spec 6.4.1.2 - PRB Composition Tech Support Interpretation - 6.4.1.2 Nuclear Operations Organization - FSAR Fig.13.1.2-1 Manager Technical Support - Details Resolution:
C. VanDenburgh dropped the issue on 8/14/90.
We should consider a Tech Spech revision at the next update.
)
PROJECT 92 t
045555
4 8/17/90' 1000 TECl5(ICAL SPECIFICATION INTERPRETATI0ftS i
NRC Concern 1.
The control of the generation, approval and distribution is not fonnal enough.
2.
The level of review and approval of Tech. Spec. interpretations is not a high enough level (i.e. PR8 review and concurrence should be requi red).
In fact. Tech. Spec. 6.4.1.6 a or d may apply.
~
3.
If NRC guidance i s used, author of guidance should be sent a
" Info Copy *.
NRC Docisnentation 1
None 4
VEGP Position Since VEGP Tech. Spec. interpretations a re not designed to modify the intent or breadth of the Technical Specification but merely to clarify 4
the specification for the on-shift opera tions crews, a fonnal process is not required and T.S. 6.4.1.6 a and d are not applicable.
The Operations Manager, being the senior member of plant management required to maintain a Senior Reactor Operators License, is the appropriate approval authority l
for the Technical Specification interpretations generated.
As required by subject matter, input from other sections including Nuclear Safety and Compliance, is utilized in the development of the interpretations.
1 VEGP Documentation None 4
e i
92 PROJECT 04s556
l 1
OVERTIE i
[
NRC Concern No provision exists in our procedures that would prevent operators from working _ an -excessive amount of overtime during an ' extended period i,
of time, i.e. monthly or yearly.
Our restriction of no more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> worked in a seven day period would not prevent excessive overtime on a j
monthly or yearly basis, i
NRC Doctanentation l'
None i-l VEGP Position I
Operations department personnel use established procedurts and i
guidelines, based on existing regulatory
- guidance, that limit the j_
possibility of this situation occurring and. ensure compliance with the i
j regulations governing overtime.
r Procedure 00005-C gives the cvertime guidelines and requi res the
)
department head to evaluate and approve the consistent use of overtime.
j.
The GMNP, or designee is also required to review excess overtime assigned to individuals each month to ensure proper authorization per Figure 1 of this procedure. Also, he reviews the overtime to ensure that assignment 4
j of excess overtime does not become routine.
l For operators working under the union contract, additional guidance is provided for overtime assignment and equalization in the Memorandum of Agreement, paragraph 49.
)
Based on review of overtime records, LER's, Reactor Trips, and ESFAS Actuations, no conclusions can be drawn that indicate excessive overtime
~:
has caused operator fatigue or an increased frequency of operator errors.
Howeve r,-
VEGP intends to review this ites for potential enhancements.
The timely. completion of required documentation was identified as
]
a problem by a Quality Assurance audit.
Corrective action from.that audit has improved performance in documentation.
i VEGP Doctamentation '
i
-Week at a Glance Paragraph 40, Memorandian of Agreement Procedure 00005-C 92 PROJECT 045557
~i w
i a
NON-LICENSED OPERATOR TRAINING N
I
}
l l
NRC Concern i
i 1.
The PE0 training program does not include under-instruction watches j
for building qualification.
j 2.
The training prograr. may not be adequate to train and evaluate l
the ability to make routine rounds; some operators may not have completed l-their actua! rounds task properly.
i NRC Documentation In NRC interviews with new building operators some said they did
}
not actually perform their rounds in trainino.
1 i
V__ EGP Position L
i j
1.
Qualification is based on the successful completion of required knowledges and skills, which are arrived at through the analysis phase l
4 of a systematic approach to training (SAT) process.
The INP0 accredited program does not rely on any arbitrary ntaber of under-instruction shifts for qualification.
However, due to, requests from Plant Equipment Operators j
surveyed. VEGP will re-evaluate the addition of under-instruction watches to the building cperator qualification checklists.
2.
The routine conduct of rounds is an iv;ntified task with associated supporting knowledges in the PE0 training and qualification program.
i A comg rehensive instructional unit provides sufficient information and l
guidance to assure that operators have the ability to perform routine rounds.
A specific weakness has been identified in the implementation of the evaluation process for the task conducting rounds.
This weakness l
will be handled through the SAT feedback process.
l l
VEGP Doctmentation Procedure 11958-C. " Auxiliary Building Operator Training Qualification
. Checklist" Qualification Signoff Criteria Cluster 51 - NLO Administrative Duties YEGP Instructional Unit NL-IU-51401-001 C.
Conduct Auxiliary Building Rounds VEGP Instructional Unit NL-IU-51401-0C2-C Conduct Control Building Rounds
[
Management Observation Report (MORE) - TQ.3, "On-The-Job-Training" i
?
i l
92 PROJECT I
o45558 i
i OPERATOR ROUNDS GENERAL INSPECTION d
NRC Concern There are differences in the depth of general inspections performed by operators during their rounds.
1 NRC Doctanentation NRC observation of rounds by new Auxiliary Building Operators.
i VEGP Position, The general inspection is intended to identify acy type cf abnormal 1
condition which may develop.
The procedural guidelines are accordingly very broad.
It is not our intent to detail every possible check which the operator could make in our procedures, The guidance in Operations j
Procedure 10001-C "Logkeeping", describes the overall areas of inspection required of the Plant Equipment Operators.
4 We expe::t there will be differences in the focus of different operators based on their personal experience and shift supervi sion instructions.
l This diversity is a plus to increase the breadth of the general inspection.
For the rest of 1990, VEGP will increase the number of supervisors and managers doing Management Obse rvation Reports on operator rounds, during both day and night shifts.
These observations will be reviewed
]
to establish a baseline performance standar11 and any needed ccrrective actions wf11 be implemented in procedures, training and practice.
4 VEGP Doctanentation Procedure 10001-C, Logkeeping Training Cluster 51-ML0 Administrative Duties i
1 92 PROJECT 045559
8/16/90
' l 16:00 l
ELECTRICAL SEPARATION ZONE 80
)
NRC Concern l
j Upon an LOSP a postulated fire in Zone 80 would render Train A inoperable, and may trip the Train 8 Diesel Generator output breaker.
VEGP should insure that no equipment required to cope with this condition would be damaged by the fire while the diesel generator output breaker is being reclosed, j
NRC Documentation a
2 DC 1-90-299 and 2-90-080 l
VEGP Position 4
1.
The design requirement for a fire in this area is to be able to shut down the plant using Train "B" equipment (FSAR 9A.1.40.L.la).
This scenario does not affect our ability to safely shut down the plant, is not a condition outside our design basis, and is therefore not a reportable condition.
A.
The postulated fire scenario would not damage the Train "B" safety related equipment necessary for safe shutdown of the plant.
The 0/G would continue to run and an annunciator would indicate D/G trouble.
4 8.
The operator would be required to observe a loss of power on the "B" Train safety related bus, recognize there was a fire in the room where the attached non-safety related bus is located, separate the nonsafety related bus from the safety related bus, and reclose the D/G output breaker.
There is adequate time for the operator to take these actions.
2.
As a conservative measure, the feature which could cause the D/G c9tput breaker to open in this scenario is being eliminated so as not; to rely on operator action to reclose the breaker.
3.
VEGP intends to provide additional infonnation to support these a
conclusions.
VEGP Documentation FSAR 9A.1.40.L.la Letter #5G-9471 Letter #5G-9510 DC 1-90-299 DC 2-90-080 92 PROJECT 045560
h CCW val.YE SURVEILLANCE t
5 k
NRC Concern Tech Spec surveillance of CCW Valves is not conducted monthly as 4
required by T.S. 4.7.3.a 4
NRC Doctmentation j
Tech Spec 4.7.3.a CCW Yalve Surveillance VEGP Position Vogtle Technical Specifications Section
- 4. 7. 3.a requi res that all valves in the Component Cooling Water System that are not locked, sealed, i
or otherwise secured in position be verified in the correct position once every 31 days.
A surveillance is not required for any CCW flowpath valves at Vogtle because all CCW flowpath valves are included in the Vogtle locked
,41ve program.
This program ensures control of the valves locked status through the Locked Valve Verifica tion Checklist 11867-1,2 and Shift Supervisors administrative control of locking and unlocking of the valves pe r Ontrol of Safety Related Locked Valves,10019-C.
i All valves required for system operability are locked, no surveillance is requimd.
The valves not locked are minor system valves which do not effect operability.
l VEGP Docunentation Procedure 11715-1/2 CCW Alignment Procedure t
92 PROJECT 045561
4 e
i J
Training Department Comments on OSTI NRC Concern 4
Repeat backs and communications were different between sitifts.
i (In one instance the BOP had to request a repeat back from the USS to establish closed loop communications)
VEGP Position l
)
Communications is an area that is streased in both simulator training and evaluation. During this sequent of PEO continuing i
1 training a listening course is being taught.
There is a Management Observation on the control room that addresses communications. Communications is an area that requires ' constant i
i attention and feedback.
2 (Comment on observation of differences between shifts and adequacy of " lowest level" of performance)
QEg 1
1 l
1 i
Additional facts about PEO Rounds Trainino:
4 Contract training instructors rather than qualified Ops Dept 1
SRos were doing evaluation due to SRO shortage.
2 The PEOs were given direction to conduct the rounds with a qualified PEO and get the qualified PEO to sign the rounds sheet. This was done due bectuse it was felt that a qualified PEO had a better feel for conducting the rounds than the instructor. Few PEOs fully complied with this direction.
3 All other evaluations were directly observed by the instructor.
4 The instructors observed that each rounds sheet had been signed off and asked the trainee if they had any questions or problems. The instructors thought the trainees had gone with a qualified PEO.
i
}
1 4
l 1
f l
l 1
i 92 PROJECT 045562
SHIFT EXPERIENCE J. E. Swartzwelder NRC Concern Backshift operators are relatively inexperienced.
NRC Doctmentation Unknown VEGP Position The collective bargaining agreement covering union employees does result in an imbalance in experience on shift with less experienced personnel on night shifts.
However, the personnel a re fully qualified for their assignments.
VEGP Docunentation Union contract - shift picks 1
I 92 PROJECT 045563
PLANT EQUIPMENT OPERATOR MORALE NRC Concern Plant Equipment Operators (PEOs) complained to the NRC Inspector of low morale due to lack of advancement opportunity.
I NRC Documentation Unknown i
VEGP Position The Operations organization is filled with ex-operators at all levels up to Shift Superintendent.
In addition, two Reactor Operators recently a
accepted promotions / transfers to other departments and Assistant Plant Operators art interviewing with other departments.
The opportunities are there, but may not appear fast enough to the PE0's.
Promotion is expected to slow down as the plant staff stabilizes.
i We indent to promote personnel to maintain a mixture of ex-operators and engineers in supervision.
We intend to encourage participation in the ATI degree program to provide a path for ex-operators to progress into mana gement.
i VEGP Docisnentation None l
f 4
1 l
92 PROJECT 045564
~
SHIFT COM M ICATIONS NRC Concern 1.
Shift briefings are not hea rd by all personnel due to room j
crowding / overflow and noise.
S
~
2.
Safety meetings are not covered in Operations shift briefs as planned.
'R NRC Doctanentation Unknown VEGP Position 1.
We intend to increase management observation of shif t briefings and consider modifications of the facility to improve briefing audibility.
2.
We intend to increase safety meetings in shift briefs.
92 PROJECT 045565
ANALYZER OPERATION FOLLOWING SI NRC QUESTION Are the analyzers placed in service 30 minutes after a safety injection?
Requirement: NUREG 0737 IIF1 Attachment 6 TMI requirement Provide station position relative to NUREG 0737 also provide proof of implementation.
NRC has looked the following places:
19000-C 19251-C Loss of primary & secondary coolant 19010-C Reference SER 6-4 YEGP POSITION Procedurt 19010-C (Loss of Prima ry or Seconda ry Coolant) step 12 currently addresses obtaining containment H2 samples following an SI.
Procedure 19251-C (Response to High Containment Pressure) step 7
currently addresses obtaining containment H2 samples with elevated containment pressure (Red or orange paths on CSFSTs).
VEGP intends to comply with NUREG 0737 relative to placing the H2 analyzers in service for these punoses.
However, the current language and procedure ortler may not meet the 30 minute in service requi rement of NUREG 0737.
We intend to revise our Emergency Operating Procedure to improve the timeliness of this action.
92 PROJECT 045566
.4 s
t Chilling Affect / Intimidation of PRB Members i
j NRC Concern
~
Intimidation of PRB members.
~
NRC Position 4
4 l
The presence and comments of the Project General Manager in
~
the PRB meeting has affected the voting some of members.
4 This was a new item identified during the 08/17/90 NRC exit.
l VEGP Position While this may be a recently identified item with the special NRC Operations Safety Team Inspection, it was previously identified to the Quality Concern Program.
It's documented as part of 90V0015.
The investigation corasisted of interviewing the voting members of PRB meeting 90-15.
During the interviews one individual did expressed some, i
hesitancy of being "true and candid" because of Bockholds presence.
This was later addressed by Bockhold to all PRB members in one of their meetings.
He reinforced the commitment of their independence and said that neither he j
or anyone else should ever influence someone's vote...that
)
if any of the members ever felt they were incapable of 4
i functioning freely and independently, they should be excused and assign someone else this responsibility.
As a result of renewed interest in this subject a new inquiry was made.
This time members (25 of 27 were l
contacted, two were unavailable), both voting and i
non-voting, were questioned on their opinion of l
intimidation.
Each was afforded the protection of confidentiality and briefed on the history of the i
allegation.
This investigation found no evidence of persuasion by presence, intimidation or coercion by the Plant Manager j
toward PRB members.
It's believed that any evidence of this i
allegation pertained strictly to the issue previously l
identified in quality concern 90V0015.
d 4
4 4
92 PROJECT i
045567 i
r d
1 Quality Concern Program NRC CONCERN I
- 1. At the exit the NRC expressed a concern about the independence of investigations performed on issues reported to the Quality Concern Program.
1
- 2. The NRC felt there was a compromise of confidentiality within the Quality concern Program by having employees other than its own performing exit interviews.
1 NRC Position 1
- 1. After reviewing procedure 00015 and discussions with" QCP members, the NRC questioned our method of investigating concerns.
They disagree with assigning investigations to department managers who are directly associated with the allegation.
This method of business l
is perceived by them as conflicting and unethical.
2.
In discussions with the NRC they inquired into the confidentiality of exit interviews when conducted by individuals other than representatives of the Quality Concern Program.
They felt that a potential exists for a breach and that business conducted in this fashion did not afford the submitter total confidentiality.
e VEGP Position
- 1. We agree that the assignment of investigations to involved parties is not ethical.
To protect against this practice the QCP Coordinator considers this when assigning an investigator.
However, it has always been our practice to use site resources in these investigations.
This arrangement was conceived primarily because of the knowledge and understanding that a department manager would have in the subject matter.
Unless directly involved in the allegation, these managers are not perceived as having biased opinions or conflict.
i 92 PROJECT l
045568 4
b i,
ie 4
f As recognized in the program's developmental stages, i
weaknens may exist if the sole resolution of concerns a
l was dependent upon departmental input.
Because of this i
one of the primary responsibilities of the QCP 1
coordinator.is to assure independence.
So that he can l
function in this capacity he reports to no department other than Plant Manager (reference procedure, 00015, section 4.4.2).
Evidence of the QCP coordinators responsibility toward independence is found in procedure 00015, sections 3.2.d & J, 4.5.4.1.
+
If the QCP coordinator is doubtful of independence, as an j
alternative investigations can be assigned to someone from off site.
Many examples of this can be found in past concerns.
As a final step in the assurance of independence, submitters of quality concerns are contacted for closure.
s Steps for closure with the submitter is found in section i
4.6 of the QCP procedure.
This is done not as a convenience to the submitter, but primarily as a check to i
the program for concern coverage (i.e.,
thoroughness and i
accuracy).
i
- 2. To eliminate this issue the concern program would have to employ personnel around the clock.
As an alternative l
we have elected to train those conducting QCP exits on the subject of confidentiality.
Currently there is only a very small select group of individuals assisting in this capacity.
This has been the method of business for several years and to date has not posed any problems.
i Aside from this upon entry on site each individual is introduced to the Quality Concern Program and the various i
ways of which they may submit concerns.
Should there be j
a problem of confidentiality upon exiting the site, they i
can always call the QCP representative at their leisure.
i l
l i
1 1
1 i,
~
t 92 PROJECT 045569
-m--
,~
i
)
Conflicting Statements NRC Concern
" Conflicting information from cognizant managers and operators on several occasions."
i NRC Position 4
In the exit the NRC reported several instances of conflicting l
information.
Areas of conflict were identified as being in:
l 1
Cont. Integrity Hydrogen Monitor Valve Opened 2
Snubber Reduction / LCO Action Statement 3
Reportability Requirements 4
Required or Anticipated Actions With Tech. Spec. 3.0.3.
t i
i Details of Research i
In an attempt to understand the NRC's position the Quality Concern Coordinator was requested to question involved individuals.
The following qg"the facts as determined through these interviews.
j 3
i 1
To assess this issue Dean Gustafson and Jim Swartzwelder were interviewed.
Gustafson reported that his involvement in this matter did not begin until the second week.
Prior to this Swartzwelder was speaking with the NRC.
His (Gustafson) only involvement during this time was the supplying of information to Swartzwelder.
Swartzwelder wanted to know:
1 \\
How do we test these monitors?
2 Were they included in the ILRT test?
I In the second week of their review he began speaking to NRC inspector, Morris Branch.
Branch asked very specific j
questions relative to the hydrogen monitors (ref. previous response).
At this time Gustafson said there was not an exchange of information, he merely received Branch's questions and made sure he understood his issues. These issues and our response was later reviewed with Branch in -
1 detail by both Gustafson and Swartzwalder two days prior to the exit.
During this time several questions were asked by Branch of which Jeff Davis was brought in and answered.
This summarizes Gustafsons involvement, at no time was he aware of any, controversial statements between himself and I(
Swartzwelder.
He felt if there was a problem that it was before his involvement.
92 PROJECT 045570 5
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i o.
In the interview with Swartzwelder he agreed with Gustafson in that he was not involved and that any comment by the NRC was in reference to their initial investigation.
He believes that the confusion was started when a shift supervisor during the initial tour of the control room informed Morris Branch that the containment hydrogen monitor 1
isolation valve received a containment isolation phase A signal.
Later on there was a discussion between a shift l
superintendent and Branch on this same issue.
At this time i
the superintendent said that regardless of whether they get i
a signal or not the system had been LLRTed to design j
accident pressure.
It's Swartzwelders belief that either the shift superintendent accidentally said ILRT or that Branch understood ILRT.
Another confusing issue was that whether the hydrogen monitors are or are not, containment isolation valves.
3 Swartzwelder told them that they were in the list of containment isolation valves because they are valves that are physically located in the given system nearest the containment.
However, he was also told that they did not receive containment isolation signal and that the system was designed to withstand accident pressure.
swartzwelder is not sure were the confusion came but its thought that it was sometime during this discussion.
4 h
d 1
1 4
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t 92 PROJECT 045571
Dates Anguet 22, 1990 Time.s 1300 While contalement cooler in T=== arable DGiB is mondered Inonerable MRC concerns The NRC reviewed the event that made the Diesel Generator 1A inoperable while a containment cooler on "B" Train was inoperable for application to Tech spec 3.0.3.
Their review concluded that it was appropriate not to enter Tech Spec 3.0.3 when the DGIA was discovered inoperable on 6-20-90 at 0121 CDT since the I40 for the "B" Train containment cooler had been exited at 1415 CDP on 6/19/90 even though the diesel had been inoperable since the installation of the tape on 6/18/90.
Nac Doemmentations LER 90-014 VEGP Positions we concur that we are not required to back date entry into Tech Spec 3.0.3 for this or similiar situations.
VEGP Documentatient t
f LER 90-014 i
i i
I 92 PROJECT I
045572
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