ML18029A601
| ML18029A601 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry, Sequoyah, 05000000 |
| Issue date: | 04/18/1985 |
| From: | Garg H, Marsh L, Mckee P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE), Office of Nuclear Reactor Regulation |
| To: | Harold Denton, Taylor J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE), Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18029A600 | List: |
| References | |
| NUDOCS 8506260752 | |
| Download: ML18029A601 (19) | |
Text
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+y*yk UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 APR 18 ss5 MEMORANDUM FOR:
Harold R.
Denton, Director Office of Nuclear Reactor Regulation James R. Taylor, Director Office of Inspection and Enforcement FROM:
P.
McKee, Branch Chief Operating Reactors Program Branch Division of Inspection
- Programs, IE L.
B. Marsh, Section Leader Reactor Systems Branch Division of Systems Integration, NRR H. Garg, Electrical Engineer Equipment gualification Branch Division of Engineering, NRR
SUBJECT:
SE(UOYAH UNITS 1 and 2 CONTAINMENT PRESSURE TRANSMITTERS The enclosed report presents our review and findings of the TVA handling and technical adequacy of two Non-conformance Reports regarding the Sequoyah Units 1 and 2 containment pressure transmitters.
It. should be noted that the immediate safety function of the pressure trans-mitters was never an issue.
It is the possibility of reduced accuracy of these instruments during the post accident phase when operator actions are required that is the overall subject of the TVA report(s).
Our review leads us to believe that there were managerial and process break-downs in two major areas.
Each is briefly described below.
- First, we believe the licensee improperly delayed initiating the review of the Sequoyah containment pressure transmitters after becoming aware of an identical problem on Watt's Bar in October, 1984.
The review of the Sequoyah Units 1 and 2 was not begun until mid January, 1984, about 10 weeks later.
According to those inter viewed by the staff, this delay was a result of working on Watts Bar to avoid licensing delays.
- Second, we believe the licensee's systems and actions relative to the containment pressure transmitters were not consistent with the significance of the issue described in Revision 0 of the Non-conformance Report (NCR) and Failure Evaluation/Engineering Report (FE/ER).
Although low supervisory 8506260752 85062i PDR ADOCK 05000259 P
level personnel believed the NCR to contain technical errors, the potential for immediate threat to health and safety for operating was not escalated to responsible managers for resolution and appropriate disposition.
This is supported by the facts that the Office of Engineering determined the existence of a potential safety problem on March 5, 1985; management resolution of the issue was not completed until March 14, 1985.
Good safety practice would dictate a more expeditious resolution.
At this time, we are reviewing all Sequoyah 1 & 2 category III NCRs and FE/ERs from the last 12 months to determine the technical adequacy of the review performed and actions taken.
Staff from NRR's technical disciplines have been provided to assist in this review.
-We are also reviewing these NCRs and FE/ERs with respoect to the adequacy of the TVA process.
With the exception of one NCR, to date we have not determined any inadequacies similar or in addition to those described above.
Region II has taken enforcement action regarding licensee actions with respect to one of the NCRs.
We anticipate our review to be complete by April 22, 1985.
We were informed Thursday, April 4, 1985, by the Sequoyah project manager that the four PAM containment pressure transmitters for Sequoyah Units 1 and 2
have been replaced by fully qualified instruments.
Despite this action, we believe enforcement actions are appropriate based on the inadequate actions described above.
Region II has reviewed this report and they generally agree with the findings of the report and the above recommendation.
L.
B. Marsh, Section Leader Reactor Systems Branch Division of Systems Integration, NRR Operating Reactors Program Branch Division of Inspection
- Programs, IE 8/S~~ ~f H.
G a
ngineer Equipment gualification Branch Division of Engineering, NRR cc: J. Nelson Grace D. Verelli E. Jordan J.
Partlow J.
Axelrad J. Olshinski H.
Thompson R. Bernero
ENCLOSURE EVALUATION OF TVA'S HANDLING AND TECHNICAL ADE(UACY OF CONTAINMENT PRESSURE TRANSMITTERS NCRs I.
INTRODUCTION This report documents a review of the handling by TVA of Nonconformance Report (NCR)
SgNNEB 8501 Revisions 0 and 1.
This review was conducted to assess the TVA process and handling of these
Representatives of NRR and OIE, accompanied by Region II staff, interviewed TVA individuals involved in this NCR within the Office of Engineering and the Office of Nuclear Power.
These inter-views were conducted from March 27 to 29, 1985.
The staff interviewed not only the originators and recipients of the NCR and Failure Evaluation/
Engineering Reports (FE/ER), but also their management to assess their cognizance and decisions in this matter.
The report herein is divided into several parts.
Section II presents background information including TVA organization and locations.
Section III describes the TVA procedures for handling and processing NCRs and FE/ERs.
Section IV presents the chronology of events developed by the staff after interviewing all those involved.
The technical aspects of Revisions 0 and 1 are presented in Section V.
Based on the proceeding information, we assessed the adequacy of the technical evaluations conducted and the processing of the NCRs.
Section VI presents our findings.
ii.
BACKGROUND TVA Office of Engineering (OE) and Office of Nuclear Power (NUC PR) have recently undergone organizational and policy changes that bear on the handling of the NCRs (and associated FE/ERs)
S(NNEB 8501 Revisions 0
and 1.
Under the current organization, OE (previously designated Engineering and Design-ES DES) maintain the responsibility for NCR generation, review and processing for TVA reactor sites, both those under construction and those that are operational.
As discussed
- later, OE is the organizational unit charged with determining reportability for issues associated with CP and OL stage reactors.
- However, OE has been assigned no responsibility regar ding reportabi lity determination for the operating reactors (OR).
For these operating plants, NCRs and FE/ERs are only inputs to the OR site, under the organization unit of NUC PR, where the ultimate decisions regarding plant operations, equipment
- changes, and reportability are to be made.
Recently, TVA underwent changes in. organizational responsibility and working relationships.
According to those interviewed by the staff, until recently, technical judgements and evaluations conducted by OE were considered by site representatives in NUC PR as authoritative and more-or-less, the final assessment.
- However, under the current arrangement, which has been referred to as the "OWNER-OPERATOR" concept, OE is considered as the architect engineer for the NUC PR Office.
As
- such, OE competitively "bids" on jobs NUC PR desires to have done.
Input by OE is viewed similar to that provided by an outside contractor or architect engineer.
In an attempt to facilitate the new "OWNER-OPERATOR" concept, represen-tatives of OE were relocated to the site.
Although these individuals are in the OE organization, they work (in a matrix organization format) for the division of NUC PR.
They have been on the Sequoyah site for about six months, and their role in the NCR and FE/ER process is new and, in the opinion of one OE manager interviewed, evolving.
II
I. PROCEDURE
S The staff obtained and reviewed the Office of Engineering (OE) and Sequoyah plant site procedures covering the generation, processing and site actions regarding
Pertinent procedures are described below.
A.
Non-Conformance Re orts (NCR)
TVA procedure EN DES EP 1.26, "Nonconformances
- Reporting and Handling by EN DES" delineates requirements for Office of Engineering personnel to generate and process NCRs.
Any OE person who determines that there is a condition judged to be adverse to quality may generate a
NCR.
Once the NCR is generated, a determination of significance is made by the appropriate OE Branch Chief.
Normally, the originator's immediate supervisor makes a recommendation regarding significance, based on the guidance contained in procedure 1.26.
The procedure gives the following examples of significance.
a.
b.
e.
Any item or condition requiring extensive
- redesign, repair or rework.
Any item or condition which, if uncorrected, could adversely affect the safety of operations of the plant, or could have generic implications at other TVA nuclear plants.
A condition adverse to quality which has not been determined to be significant or nonsignificant within 8 calendar days of the identification of the condition.
Any falsification of records.
Repetitions of a particular condition adverse to quality.
Note, however, that a large number of repetitions may be required to define a significant condition, such as in:
f.
g.
(1)
A special manufacturing process such as welding, where a small rejection rate is unavoidable, or (2)
Administrative matters such as late arrival of records from a supplier.
Any deliberate failure to follow procedure.
Nonconformance reports which are sent to EN DES marked nonsignificant by CONST which are reviewed by EN DES and which cannot be verified to be nonsignificant within 2 weeks (14 calendar days) of receipt by EN DES.
Failure to meet a specific commitment to the Nuclear Regulatory Commission (NRC).
When the construction or operation of a facility activity, or a basic component supplied for such facility or activity (1)
Fails to comply with the Atomic Energy Act of
- 1954, as
- amended, or any applicable rule, regu-lation,'rder, or license of the NRC relating to a substantial safety hazard, or (2)
Contains a defect which could cause a substantial safety hazard.
If there is a doubt as to whether a condition is significant, it shall be classified as significant.
The procedure states that a determination of significance is to be made within 3 working days of the NCR preparation date.
The procedure further requires that the time for a determination of significance must not exceed 8 calendar days.
If the time exceeds 8 calendar
- days, the NCR is automatically designated as significant.
For significant NCRs, the procedure requires preparation of a Failure Evaluation/Engineering Report by OE.
B.
Failure Evaluation/En ineerin Re ort TVA Procedure EN DES EP 1.48 "Preparation of Failure Evaluations/
Engineering Reports of Deficient Conditions for Operating Nuclear Plants" describes the process to be followed once an NCR is determined to be significant.
The procedure applies only to operating nuclear plants, and specifies the preparer's reviewer's and approver's responsibilities and the allowed time for the preparation the FE/ER.
In general, the FE/ER is Chief within 15 calendar solely to provide NUC PR operating decisions such LCOs and for reporting to the NRC.
to be issued by the appropriate OE Branch days".
The stated purpose of the FE/ER is with engineering information to be used for as compliance with technical specification, As specified in Section2.4 of the procedure, there are certain situations in which this time may be exceeded,
- however, none of these situations seem to apply in this case.
The FE/ER, when completed by OE, results in a categorization of the deficient condition as either a Category I, II or III.
No guidance is provided within the instruction itself as to how to make this determination or of the safety significance of this determination.
The only guidance provided is on the FE/ER coversheet where the preparer is to check the appropriate block.
Procedure 1.48 requires that the design project manager (an onsite OE position) concur in the FE/ER, at which time the document is an approved document ready for issuance.
Once the Nuclear Engineering Branch Chief (Chief Nuclear Engineer) signs the FE/ER, it is sent to the site manager for the unit affected.
Once the FE/ER is formally transmitted to the site, the NUC PR organization processes the FE/ER in accordance with procedure SQA 118.
C.
Handlin of Non-Conformance Re orts Sequoyah site procedure SQA 118 "Handling of Non-Conformance Reports" specifies the system to handle incoming NCRs and FE/ERs from OE.
The procedure requires the Regulatory Engineering Section (RES) which is an organization unit under NUC PR, upon formal receipt of the NCR and FE/ER, to determine the adequqcy of the documents.
Also the procedure states that OE will be immediately notified if the FE/ER is judged to be inadequate.
If necessary, the FE/ER is to be returned to OE for further evaluation which is to be completed as soon as possible.
If RES accepts the FE/ER, three working days are allowed for a safety evaluation to be prepared.
(As noted in EN DES EP 1.48, NUC PR may request OE to perform the safety evaluation).
Once RES has completed the safety evaluation (SE), the FE/ER is forwarded.to the Compliance Staff, another organizational unit under NUC PR.
The Compliance Staff has an additional three days to determine the reportability of the issue.
A NCR Disposition Record Sheet is provided to track the
There is no guidance or specific steps given in this procedure for informing site management should a Category III FE/ER be received and rejected.
The procedures do not distinguish, in terms of notifying plant management or operators, between the three different categories assigned in FE/ER, The determination of reportability, made by the Compliance Staff, is performed in accordance with procedure SQA 84.
D.
Re ortabl e Occurrences Sequoyah site procedure SgA 84 "Reportable Occurrences" gives the standard practice for the initial evaluation, notification and documentation of occurrences that are potentially reportable to the NRC.
The instruction states that any plant employee who identifies an event or occurrence deemed to meet the criteria, should complete the Potentially Reportable Occurrence (PRO) form, and informs his immediate supervisor who in tur n will inform the shift engineer(SE) or shift technical advisor (STA).
The SE or STA will determine if immediate notification (10 CFR 50.72) is required and will assign a number to PRO.
The Compliance Staff will process the PRO and is responsible for the final evaluation of the completed PRO and the preparation of any required report to NRC.
Plant 0 erations Review Committee The Plant Operations Review Committee (PORC) serves in an advisory capacity to the Plant Manager and as an investigating and reporting body to the Nuclear Safety Staff in matters related to safety in plant operations.
- However, PORC is not required to be informed anytime a Category III designation is assigned to a FE/ER, or even in the case of a disagreement between the OE and NUC PR staff regarding a Category III FE/ER.
IV.
CHRONOLOGY OF EVENTS The following is a chronology of the development of Revision 0 and Revision 1 of the licensee Failure Evaluation/Engineering Reports (FE/ER) and Nonconformance Reports (NCR) for the containment pressure transmitters.
This chronology was based on interviews with TVA staff.
Subsequently, TVA provided a more detailed chronology.
A.
Revision 0
Three NCRs were written for Watts Bar in October 1984 on containment pressure transmitters that were similar to the containment pressure transmitters used at Sequoyah.
The engineer that prepared the Watts Bar NCRs did not begin the evaluation of the Sequoyah pressure transmitters until January 1985.
Rev.
0 of the NCR for Sequoyah was signed by the preparer on January 16, 1985.
Between January 16 and January 31, four supervisors initialed the NCR. It was stated that there were delays in getting one initial (almost two weeks) because that individual had been out of the office for personal reasons.
The NCR was designated by the Nuclear Engineering Branch Chief (Chief Nuclear Engineer) as having a "Significant Condition Adverse to guality" on January 31, 1985.*
Thus, the time requirement of EN DES 1.26 was exceeded.
According to OE, the NCR still would have been designated as having a "Significant Condition Adverse to guality" even if the time limit had not been exceeded.
A determination of "Significant Conditions Adverse to guality" on an NCR, by procedure, requires the preparation of a FE/ER (form TVA 10826).
In the case of Revision 0 of the subject NCR, engineering staff in OE other than the preparer of the NCR were asigned to prepare the FE/ER.
The FE/ER was completed by the preparers on February 15, 1985 and categorized as a Level III deficient condition (unable to perform its required design function(s) unless corrective modifications are made).
During preparation of the FE/ER, the preparers stated that they had several discussions with site engineering staff and some site operations engineers.
Originally the preparers had proposed that the classification of the deficient condition would be a
Level II (not acceptable for some design loading combination or design condition.
- However, a functional impairment of component(s) is not likely).
But the NCR was upgraded from a Level II to Level III based on recommendations from supervisors in the engineering organization.
The FE/ER was concurred in by site engineering staff on March 4, 1985 and signed out by the Nuclear Engineering Branch Chief on March 5, 1985.
With respect to reporting requirements, the engineering organization asserted that they had no responsibility as to reportabi lity of items gener ated in thei r organization for operting facilities.
That responsibility is (by procedure) delegated to the onsite Office of Nuclear Power (NUC PR) organization.
Based on interviews with OE staff, no one stated that they were aware of the implications of Rev.
0 on reportability or Technical Specification applicability.
By procedure, when the FE/ER is concurred in by the onsite organization, it is an approved document (by OE) ready for issuance and the FE/ER is informally sent to the appropriate NUC PR contact.
In the case of Rev.
0 to the FE/ER, a telecopy from
.the Knoxville Office to OE, on site, was hand carried to the NUC PR, Regulatory Engineering Staff, on March 5, 1985.
- However, RES did not consider this copy as a formal document that required immediate action per SgA 118.
RES contacted site engineering, and based on several technical considerations, it was informally agreed that Revision 0
of the NCR and FE/ER may be inaccurate and further information was needed.
Office of Engineering (OE) procedure 1.26 requires the NCR to be categorized as significant if the time between initiation of the NCR (in this case January 16, 1985) and the time of the disposition (in this case January 31, 1985) is greater than 8 days.
During the interviews, the OE'anagers were surprised when they were informed that NUC PR did not consider the copy hand delivered on
.March 5, 1985 by OE onsite engineering staff as a formal document.
The formal copy of Revision 0 of the NCR and FE/ER (signed by the Nuclear Engineering Branch Chief) was not sent out of the Engineering office in Knoxville until March 7, 1985 and was received on site on March 8, 1985.
The supervisor of RES received his formal copy of the NCR on March ll, 1985 at which time he brought it to the attention of the Compliance Supervisor.
The Compliance Supervisor stated he considered that if Rev.
0 of the NCR and FE/ER were accurate, both Sequoyah Units, which were then operating, would be required to shut down.
The Plant Superintendent, Plant Manager, Office of Engineering Management, and other NUC PR representatives were called to a meeting by the Compliance Supervisor to discuss actions to be taken.
It was agreed that Revision 0 of the NCR and FE/ER were inaccurate and that the pressure transmitters were operable.
Work'as initiated to expedite revisions to both the NCR and FE/ER and to prepare a justification for continued operation.
On March 13 and 14, 1985, staff from OE and NUC PR discussed the changes needed to the FE/ER and NCR.
These changes included review of a footnote in WCAP 8541 (See Section V) which imposed an additional 7X inaccuracy which NUC PR felt had been inappropriately applied and consideration that the similarity comparison was being made to an instrument tested in a LOCA environment when the transmitters in question were located in the annulus building.
The March 14, 1985 discussion is referenced by note on Revision 0 of the FE/ER as documentation of the agreement that Revision 0 was incorrect and needed revision.
B:
Revision 1
Revision 1 of the NCR was completed and signed by the preparer on March 20, 1985.
As with Revision 0 ofthe NCR, Revision 1 was designated as having a "significant condition adverse to quality" and was signed out by the Engineering Branch Chief on March 21, 1985.
Revision 1 of the FE/ER was completed and signed by the preparer on March 21, 1985.
All parties responsible for concurrence in the revision also signed the form on March 22, 1985.
The Chief Nuclear Engineer signed Revision 1 of the FE/ER on March 22, 1985.
The primary changes made to the FE/ER were:
(1) the deficient condition category was changed from Level III to Level II, and (2) information for justification for continued operation was included.
The FE/ER was not clear as to the qualification status of the pressure transmitters in accordance with 10 CFR 50.49.
Based on interviews with OE and NUC PR staff, there appeared to be disagreement on this issue.
Revision 1 to the FE/ER and NCR was formally received on site on March 25, 1985.
The Supervisor of
Regulatory Engineering completed and forwarded Revision 1, with their safety evaluation, to the Supervisor of Compliance on March 27, 1985.
The safety evaluation essentially agreed with the justification for continued operation contained in Revision 1
to the FE/ER and proposed replacement of the transmitters.
The Supervisor of Compliance is responsible for making determinations of reportabi lity.
V.
TECHNICAL ARGUMENTS/JUDGEMENTS IN NCR/FEER Sequoyah Units 1 and 2 and Watts Bar Units 1 and 2 are both Westinghouse (W) four loop PWRs with ice condensor containments.
For all four units, containment pressure is monitored and is an Engineered Safety Feature (ESF) actuation parameter.
The containment pressure is used for the following automatic functions":
1.
2.
3.
4.
5.
6.
ESF (Safety Injection) (from Cont.
Pres.
HI)
Containment spray (from Containment Pressure HI-HI)
Containment Isolation (Phase A) (from ESF-SI)
Containment Isolation (Phase B) (from Cont.
Pres.
HI-HI)
Containment Ventilation Isolation (from ESF-SI)
Steam Line Isolation (from Cont.
Pres.
HI-HI)
Containment Pressure HI Containment Pressure HI-HI 1.54 psig 2.81 psig These functions are assumed in the licensing basis calculations for the following design basis accidents:
Main Steamline Break (inside containment),
Main Feedline Break (inside containment),
and LOCA (certain break sizes).
Additionally, based on indication of containment
- pressure, a number of operator actions following these accidents are to be taken.
According to the licensee, these manual operator actions include:
1.
Termination of Contairiment spray (after automatic initiation) 2.
Initiation of Containment Spray (from RHR system) 3.
Termination of Containment Spray (from RHR system) 4.
Open Containment Vacuum Breaker isolation valves 5.
Reset phase B Containment isolation Because of their post-accident requirements, two pressure transmitters are designed, in accordance with Regulatory Guide 1.97, as Post Accident Monitoring (PAM) instruments.
"Sequoyah Units 1 and 2 Technical Specifications
A.
Revision 0
This NCR was based on the Watts Bar NUREG-0588 review (about October 1984), which had determined that these pressure trans-mitters are not environmentally qualified.
- Hence, a review on mitters were not qualified for their application at the Sequoyah plants.
701 feet.
During a design basis accident (OBA), they are exposed
$o a temperature of 150 F for approximately 27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br />, that decays to 105 F in 30 days and remains there for the rest of the accident.
The total6integrated radiation dose based on 7 year qualified life is 9.5 X 10 rads.
The FSAR states that these transmitters should operate within + lOX accuracy.
The installed transmitters are Foxboro Model E11GM (Non-MCA) with a range of -1 psig to + 15 psig.
In performing the review for
- Sequoyah, the licensee used qualification data from WCAP-8541.
In Table 2 on Page 12 of the WCAP, qualification data is presented for Foxboro transmitter model E11GM (MCA).
The table states that these transmitters are accurate to within 2 14.75K when exposed to a design basis accident evironment inside containment.
The licensee stated that this data was used since it was virtually all that was available for their transmitters, even though the data is not for the same transmitter (MCA versus non-MCA), nor for the same location (inside containment versus annulus).
Based on this data, the licensee prepared Revision 0 of the Failure Evaluation/Engineering Report (FE/ER).
This FE/ER analyzed the overall post-accident consequences assuming the transmitters will operate with the maximum inaccuracy of 14.75X of the span.
These consequences could affect the operator action required by Emergency
Response
Guideline (ERG) E-1 as well as Functional Restoragion Guideline (FRG) Z-1.
Based on this evaluation the Office of Engineering
- recommended that these transmitters be replaced or relocated to an environment to which the present transmitter are qualified, (i.e.,
completely outside containment).
B.
Revision 1
At the request of NUC PR, Revision 1 was prepared to correct what NUC PR believed were technical errors in Revision 0, and to add a
JCO.
The errors were related to the fact that there are only two transmitters required for Sequoyah post-accident monitoring compared to the four identified in Revision 0.
- Also, NUC PR believed that a footnote to Table 2 in the WCAP used by OE was not properly considered.
This footnote states that 27K of accuracy was added to the total accuracy figure of 14.75K to account for the increase in pressure of the containment during the DBA.
This is based on the fact that the transmitter is calibrated to operate at the normal ambient condition of atmospheric pressure.
However, during the DBA the ambient pressure inside containment is much greater than normal ambient pressure.
This footnote applied to the pressurizer pressure instrument in the WCAP.
The transmitters at Sequoyah do not see any significant pressure change during DBA conditions.
Rev.
1 of the NCR and FE/ER corrected the inaccuracy regarding the number of transmitters required for Sequoyah and also provided the justification for continued operation (JCO) in accordance with paragraph (i) of 10 CFR 50.49.
Revision 1 does not explicitly address the NUC PR concern regarding the footnote on Table 2.
OE evaluated the footnote on the table in the WCAP, but did not believe that the note could be used to justify the accuracy requirement for the Sequoyah application.
Because of the difference-between the tested equipment (MCA, Large span and environmental conditions) and the transmitters at Sequoyah, it was difficult to extrapolate the data from the WCAP report.
C.
Evaluation We have reviewed the technical judgements made by the originators of Revision 0 and believed that based on the available technical information, the use of 14.75K as an accuracy figure may not have been a true indication of the instrument's
- accuracy, but was conservative in that it led to the conclusion that the instruments should be replaced with qualified ones.
Whether the installed instrument's accuracy is better or worse than the value used cannot be determined, as nq documentation demonstrating the instrument's accuracy is available Several of the statements in Revision 0 imply that the instruments wi 11, without any doubt, experience reduced accuracy.
While these are conservative statements, they are not necessarily
- accurate, since there is no data available regarding the performance of the actual Sequoyah transmitters when located in the annulus.
On balance, we believe Revision 0 is a conservatively based document that clearly supports the conclusion regarding instrument replace-ment, but was not based on information applicable to the actual instruments installed.
Revision 1 on the other hand, is a confusing and contradictory document.
It contains the following:
a.
Revised NCR (Rev. 1) b.
Revised FE/ER (Rev.
1) c.
Modified Equipment gualification Sheets (EgS) d.
Description of equipment location and environment e.
Differences between MCA and non-MCA instruments f.
Evaluation of Non-MCA transmitters g.
Overall conclusion (JCO)
In reviewing this document and its parts, we noted that the Revision 1 FE/ER, paragraph 1 (Summary of Non-Conformance) states the transmitters are located in the annulus and are subject to the environmental condition for a LOCA/MSLB inside containment.
This conflicts with the description of equipment location and environment, which implies that the environment in the annulus following a MSLB/LOCA is not as severe as the environment inside containment.
Paragraph 8 of Revision 1, Conclusions, states that based on the JCO, the existing transmitters will be capable of providing their intended long term safety function.
However, paragraph 9
on the same page gives two possible correction actions (replacement or relocation).
Thus, there is a contradiction as to whether the equipment is qualified or not.
The EgS has two blocks checked that are contradictory.
One ualification b similarit and two, un uglified com onent.
It should be noted that Revision 1 does not contain or cite any test data that would document the qualification of the non-MCA transmitter in the annulus environment.
The evaluation of the non-MCA transmitters also contains inadequacies.
Under paragraph (4), the licensee states that the accuracy of the MCA transmitter is similar to the non-MCA transmitter, however there are design differences that the licensee pointed out (top of page 2),
but apparently ignored when concluding that the instruments have similar accuracies.
Also, the licensee
- asserts, in paragraph 4 of the conclusions, that the instruments would be able to perform their long-term safety function, but goes on to say in paragraph 5 that the instruments will be replaced with qualified transmitters.
On balance, Revision 1 is a less conservative document that contains contradictions.
The JCO contains the essential elements necessary to provide an adequate basis for interim operation, but there are inadequacies and technical items not properly addressed.
Based on the information in the NCR and FE/ER, we do not consider that these transmitters were qualified for their application.
Also after reviewing the JCO, we believe the requirements for JCOs specified in paragraph (i} of 50.49 were generally met.
However, the following information was lacking:
(a)
Confirmation that the subject transmitters maintain an accuracy of i 0.5X of span at the accident temperature for the period of time to cover the instrument's operability requirement.
(b)
Confirmation that the radiation testing of the amplifier for the MCA transmitter (WCAP 8541) was done as it is installed in non-MCA transmitters, and that no additional shielding was installed during this test.
(c)
Since 0-rings are different in non-MCA transmitters compared to MCA transmitters and no test data is available to establish the life of the O-rings, confirmation that these 0-rings are replaced any time the transmitters are opened or at least once every refueling outage.
We have been recently informed that the subject transmitters (PAM instruments) have been replaced with qualified transmitters.
VI.,
FINDINGS The following are the findings of this review effort.
These findings were preliminary presented at an exit meeting to TVA staff on March 29, 1985.
1.
There were problems in the process for dealing with the issues raised in Revision 0 to the NCR over the timing between March 5 to March ll, and the actions therein (or lack thereof).
There appears to have been delays in (1) getting Rev.
0 to the site, and (2) getting management apprised of the substance of Rev.
0 and the implications of it.
2 Confusion/ambiguity/lack of definitions existed regarding the deficiency categories of a FE/ER (Level I, II and III).
There was no universally accepted definition of a Category III FE/ER, other than that it was important.
3 There was no method to assure expeditious management appraisal of Category III FE/ER status, especially if, as in this case, there was a technical disa reement betweem two organizations (OE L NUC PR) and there was a pending safety issue.
For example, when there was technical concern by a member of the NUC PR, RES staff at the site, regarding the Rev.
0 assessment, he went to the Nuclear Services
- Manager, who considered himself as merely a "conduit" for information flow between NUC PR and OE.
The Nuclear Services Manager did not ask nor did he seem concerned over the the nature of the disagreement, impact on plant safety or the extent to which upper management had been apprised.
Also, the Nuclear Services Manager did not ensure that his OE counterpart, the Design Project Manager, who was the next level in the communication chain, was informed or consulted.
13-In general, there was no separate, distinct, positive means of
. ensuring all significant NCRs and FE/ERs get the proper management attention.
The procedures in place had various time limits and organization contacts, many of which were exceeded but plant management (i.e., the actual operating personnel) and other operating management were not required to be informed until far in the process.
5.
At least two supervisors who have signature responsibility in the FE/ER process, believed their responsibilities when concurring in FE/ERs is only an administrative task.
Their managers believed their responsibilities were far more - to read, understand and technically concur in the FE/ER approach and conclusions.
In general NCRs and FE/ERs contain no assessment of safety significance, impact on plant operations, or reportability impact.
(OE perceives their responsibility to end once the FE/ER is generated and transmitted to site staff.)
7.
As of March 29, 1985 (the day of the exit) there was confusion and, in at least one case, disagreement, regarding what Rev.
1 is actually concluding.
OE believed Rev.
1 says the PTs are not environmentally qualified, are operable for an interim (not specified) period based on a JCO, but needed to be replaced with qualified instruments (non-specified time).
NUC PR (at least 2 individuals),
on the other hand, believed Rev.
1 says the instruments are qualified, are operable, but should be replaced.
With respect to the technical merits of Revision 0, we believe Revision 0 was a conservatively based document in that it concludes that the pressure transmitters should be replaced.
Revision 0 is clear in this conclusion.
However, Revision 0
'used accuracy data (taken from WCAP 8541) that was not appropriate for the actual Sequoyah pressure transmitters located in the annulus.
9.
Revision 1 is a contradictory and somewhat confusing document.
It implies the pressure transmitters are both qualified and unqualified.
It contains a
JCO that is, for the most part, acceptable, but still lacks proper assessments of all aspects.
Revision 1 is less conservatively based than Revision 0 in that it cites maniufacturer s design specifications (i.e., a.
5X) as a representation of the instrument's post accident performance.
The a.
5X figure was never verified by test.
We believe, the above findings are indicative of managerial and process breakdowns in two major areas.
- First, we believe the licensee improperly delayed initiating the review of the Sequoyah containment pressure transmitters after becoming aware of an identical problem on Watt's Bar in October, 1984.
The review of the Sequoyah Units 1 and 2 was not begun until mid
- January, 1984, about 10 weeks later.
According to those interviewed by the staff, this delay was a result of working on Watts Bar to avoid licensing delays.
- Second, we believe the licensee's systems and actions relative to the containment pressure transmitters, were not consistent with the significance of the issue described in Revision 0 of the Non-conformance Report (NCR) and Failure Evaluation/Engineering Report (FE/ER).
Although low supervisory level personnel believed the NCR to contain technical
- errors, the potential for immediate threat to health and safety for operating was not escalated to responsible managers for resolution and appropriate disposition.
This is supported by the facts that the Office-of Engineering determined the existence of a potential safety problem on March 5, 1985; management resolution of the issue was not completed until March 14, 1985.
Good safety practice would dictate a more expeditious resolution
Distribution DWI LPDR SECY ACRS CA
- ABBeach, IE JLieberman, ELD Enforcement Coordinators RI, RII 0 R III,
- RIV, RV
- JNGrace, RII
- BHayes, OI SConnelly, OIA
- FIngram, PA VStel.lo, DED/ROGR
- JCrooks, AEOD
- HDenton, NRR
- RStark, NRR
- EJordan, IE
- JPartlow, IE
E. Fredrickson, Reactor Project Inspector State of Tennessee IE:ES ABBe 6/
/85 e rad 6'/S5 IE RV mer
~o 6Q/85 4~