IR 05000341/1989017
| ML20006A998 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 01/18/1990 |
| From: | Defayette R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20006A993 | List: |
| References | |
| 50-341-89-17, NUDOCS 9001310144 | |
| Preceding documents: |
|
| Download: ML20006A998 (24) | |
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U. S. NUCLEAR REGULATORY COMMISSION
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REGION III
Report No. 50-341/89017(DRP)
Docket No. 50-341 Operating License No. NPF-43 Licensee: Detroit Edison Company 2000 Second Avenue Detroit, MI 48226 Facility Name: Fermi 2 Inspection At: Fermi Site Newport, Michigan Inspection Conducted: May 30 through November 1, 1989 Inspector:
W. G. Rogers
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Approved By:
R.DeFayette,dief
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Reactor Projects Section 2B Date Inspection Summary Inspection on May 30 throuah November 1.1989 (Report No. 50-341/89017(ORp))
Areas Inspected:
Special ~ safety inspection associated with safety reviews
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required under 10 CFR 50.59.
Result: Most of the deficiencies identified in the safety review process
l dealt with the preliminary safety evaluation / safety evaluation interface, One matter, the RHR minimum flow valve, was considered an unreviewed safety l
question that was not identified by the licensee's safety review process.
In almost every instance the personnel involved in the defittent safety reviews had not been trained to the most current 50.59 training program due to the failure of the licensee to require requalification-training. Also, there has been a breakdown in the licensee's deviation event report review / approval-of corrective action extensions.
Two violations were identified (Paragraphs 4.g, 6.g and 8.g.4 for instances of inadequate
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10 CFR 50.59 reviews and Paragraphs 8.g.1, 8.g.2 and 9.c for instances'of I
inadequate implementation of the deviation event report and temporary modification programs).
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DETAILS
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Persons Contacted a.
Detroit Edison company
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- P. Anthony, Licensing
- T. Bradish, Production, Quality Assurance i
- R. Ballis, Superintendent, Nuclear Engineering /I&C
- J. Contoni, Superintendent, Mechanical / Fluids
- S. Catola, Vice President, Nuclear Engineering and Services
- G. Cranston, General Director, Nuclear Engineering
- D Drotar, Supervisor, Nuclear Training
- D. Gipson, Plant Manager
- L. Goodman, Director, Nuclear Licensing
- K Howard, Principle Engir.eer, Plant Systems
- A. Kowalczuk, Superintendent, Maintenance
- R. McKeon, Superintendent, Operations
- R. May, Director, Nuclear Materials Management
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'G. Ohlemacher, Principal Engineer, Licensing
- W. Miller, General Supervisor, Plant Safety
- W. Orser, Vice President, Nuclear Operations
- L. Schuerman, General Supervisor, Nuclear Engineering
- A. Settles, Superintendent; Technical Engineering
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- G. Smith, Principle Reactor Engineer
- R. Stafford, Director, Quality Assurance
- B. R. Sylvia, Senior Vice President, Nuclear Operations
- R. Thorson, Outage Manager
- J. Tibai, NSRG, Staff Engineer b.
U.S. Nuclear Regulatory Commission
- W. Rogers, Senior Resident Inspector
- Denotes those attending the exit meeting on November 1, 1989.
The inspectors also interviewed others of the licensee's staff during this inspection.
2.
Background a.
Regulatory Requirements 10 CFR 50.59 is a requirement applicable to all license holders of production and utilization facilities. This section of the Code of Federal Regulations establishes review criteria for proposed changes, tests and experiments to assure that the licensing bases of the facility are not changed without NRC review and approval.
If the change, test or experiment does effect the licensing bases it is considered an unreviewed safety question or a change to the Technical Specifications.
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The review criteria set forth in 10 CFR 50.59 to determine whether the licensing bases are affected are:
May the change, test or experiment increase the probability of occurrence or the consequences of a previously evaluated s
accident / transient?
May the change, test or experiment create the possibility of a different type accident / transient than previously evaluated?
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Is the margin of safety as defined in the bases of the Technical Specifications reduced?
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Does the change, test or experiment involve a change to the Technical Specifications?
If the answer to any of these questions is yes and the licensee wishes to continue to pursue the change then an application for amendment to the license must be sought and granted prior to implementation.
b, previous Enforcement History During the first quarter of 1987, the licensee isolated the l
moisture separator reheaters from service while greater than 25%
power and substantially reduced feedwater heating while operating up to 50% power. Both of these situations required safety evaluations in accordance with 10 CFR 50.59 and the licensee failed to perform the evaluations prior to establishing these l
conditions.
Violation 341/87021-01 was issued against 10 CFR 50.59 on May 12, 1988.
The licensee responded to the violation on June 13, 1988.
In that response the licensee stated that training requirements for personnel performing preliminary and full safety evaluations had been instituted as part of Procedure NOIP 11.000.53, Revision 5, approved June 3, 1988. Also, the licensee stated that a new procedure, FIO-FMP-01, " Safety Review Group Organizations," was issued requiring the Onsite Safety Review Organization (OSRO) to ensure that a preliminary evaluation is performed, and if required, a full safety evaluation is performed for any proposed off-normal plant operation, 3.
Recent Issues l
In recent months a number of issues associated with the 10 CFR 50,59 l
process have been identified by NRC inspectors, Most of these issues were embodied in unresolved item 341/88012-12, apparent violation 341/88037-05, apparent violation 341/89006-02 and unresolved item 341/88035-04 Presented in this report are all the 10 CFR 50.59 issues and the inspector's evaluation of whether the corrective actions for violation 341/87021-01 were effective.
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4.
Residual Heat Removal (RHR) Minimum Flow Valve, Unresolved Item I
341/88012-12 j
a.
Regulatory Requirements Technical Specification 3.5.2 requires at least two subsystems of emergency core cooling be operable while in operational
conditions 4 and 5.
Two subsystems of core spray or two subsystems of low pressure coolant injection (LPCI) in any
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combination can be utilized to meet this requirement.
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LPCI subsystem is utilized then both pumps and an operable flow
path capable of taking suction from the suppression chamber and -
i transferring the water to the reactor vessel for that subsystem
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must be operable.
Technical Specification 3.4.9.2 requires two operable shutdown cooling loops of RHR. An operable loop consists of only one RHR pump and the heat exchanger,
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Regulatory Guide 1.139, May.1978, Section C, Item 4 on RHR system pump protection states, "The design and operating procedures for the RHR systems should include provisions to prevent damage to the RHR system pumps due to overheating, cavitation, or loss of adequate pump suction head." Detroit Edison's Updated Final Safety Analysis Report, Appendix A', Item 1.139, states, "The~ Fermi 2
plant is in compliance with Regulatory' Guide 1.139."
The Safety i
Evaluation Report for Fermi, NUREG 0798, states in Section 5.4.2,-
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Residual Heat Removal System, "The Residual Heat Removal System design has been compared with the functional, isolation, pressure
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relief, pump protection, and test requirements of Branch Technical
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Position RSB 5-1, " Design Requirements of the Residual Heat Removal System," and was found to comply with the implementation criteria for FERMI-2." Section D of Branch Technical Position RSB 5-1
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reiterates Regulatory Guide 1.139, Section C, Item 4.
b.
Deadheading_ Incident During the LLRT outage in the spring of 1988 the reactor was placed in cold shutdown, operational condition 4, and the reactor vessel head removed. On March 28, 1988, Core Spray division I was taken out of service for maintenance and on March 31, 1988 RHR/LPCI division I was taken out of service for the same reason.
On March 30, 1988, Safety Evaluation 88-0074 was prepared and approved. The safety evaluation allowed the division 2 RHR/LPCI minimum flow valve to be removed from service in the closed position and considered that division to still be operable.
Early in April, the minimum flow valve was taken out of service as authorized by the safety evaluation.
Therefore, on April 9, 1988, division 2 LPCI and division 2 Core Spray were considered to meet Technical Specification 3.5.2 requirements.
Division 2 RHR was considered to meet Technical Specification 3.4.9.2 requirements. At 0330 on
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April 9, 1988, the RHR Division 2 discharge valve unexpectedly closed.
For approximately the next half-hour the operating RHR pump ran deadheaded. After the half-hour, operators observed the condition, shut down the pump and ascertained whether any damage had occurred. The reactor coolant temperature did not increase duo
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to the small decay heat load that was present and no pump damage
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was incurred, c.
Inspector and Licensee Review of Safety Evaluation 88-0074
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Subsequent to the deadheading incident a confirmatory action letter was issued and the inspector began to review the safety evaluation.
l During the review the inspector could not ascertain how the
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licensee assured that provisions to prevent damage to the division 2 RHR pump had been maintained from overheating since the minimum flow valve was designed to prevent overheating.
Regulatory Guide 1.239 was not mentioned or referenced in the safety evaluation. Also, the safety evaluation stated in part,
"In operational condition 4 the reactor pressure is O psig and the minimum flow bypass function of the RHR is not required...." Reactor pressure is not always 0 in operational condition 4.
Hydrostatic testing is done in operational condition 4 and at much higher pressures than the discharge pressure of the RHR pumps.
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On April 15, 1988, these concerns were conveyed to the engineering
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authori ty.
There were a number of meetings between the inspector and engineering authority personnel discussing these matters and why the inspector felt the safety evaluation was questionable.
I The licens.s responded that the intent of the safety evaluation
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was only for when reactor pressure was at 0 psig and that at l
0 psig there was no need for the minimum flow valve since an injection path would always be available.
However, the licensee did not establish any administrative or physical controls to assure O psig reactor pressure was maintained. Also, as the deadheading event revealed, a flow path was not always available.
On April 29, 1988, the. Independent Safety Engineering Group (ISEG) reviewed safety evaluation 88-0074 and questioned two aspects of the evaluation.
Under a licensee initiated program, all safety evaluations are reviewed by ISEG and their findings
presented to the offsite review committee entitled the Nuclear Safety Review Group (NSRG). This review effort partially grew out of the events associated with Violation 87021-01 e,u though this was not mentioned in the Notice of Violation response.
The *irst aspect of the ISEG concern dealt with the safety evaluation not addressing the potential for RHR pump damage
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resulting from the possible inadvertent closure of a discharge valve.
The second aspect dealt with the statement in the Updated Final Safety Analysis Report that the minimum flow line is used for flushing upon starting the RHR system and the safety evaluation did not state whether the flushing was done.
On May 4, 1988, ISEG wrote DER 88-1008 against safety evaluation 88-0074 ISEG findings on safety evaluation 88-0074 were presented to the NSRG on May 19, 1988, as documented in NSRG minutes 88-03.
DER 88-2008 was forwarded to nuclear engineering for resolution.
The resolutiun to the first concern was a discussion of the intent of the safety evaluation.
The resolution to the second concern was an explanation that the j
flushing function was accomplished by another valve in the i
system. These responses were acceptable to the ISEG personnel who concurred with nuclear engineering that an unreviewed safety
question did not exist as documented on the DER on July 11, 1988.
In NSRG meeting 88-04 on July 21, 1988, safety evaluation 88-0074 j
was documented closed in the NSRG minutes.
l During the inspector's exit for Inspection Report 341/88012 on August 15, 1988, the inspector reiterated his concern with safety
evaluation 88-0074 and his intent to submit this matter for NRR review.
In response to these statements, memorandum NE-PJ-88-0454 dated August 19, 1988, was generated from the mechanical and fluid.
systems supervisor to the director of licensing. This memorandum
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discussed the chronology of events associated with this valve and that engineering, after reviewing the history surrounding the safety evaluation, believed that the issues posed by the resident inspector had been addressed.
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d.
NRR Review of Safety Evaluation 88-0074
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On October 27, 1988, a memorandum from E. G. Greenman, Region III to M. J. Virgilio, NRR, was issued requesting NRR technical assistance in evaluating issues raised by the inspector.
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response was issued on May 25, 1989, in a memorandum from M. J. Virgilio for E. G. Greenman.
The first question was,
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"Does an operable flow path in the context of Technical Specifications 4.5.2 include the minimum flow valve line?" The response was yes. The second question was, "Is the overheating
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l protection an inherent condition required under the Technical l
Specifications to assume operability of the shutdown cooling function of RHR in Mode 4?" The answer was yes.
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Effectiveness of Violation 341/87021-01 Corrective Action
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The violation response indicated that personnel performing
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.10 CFR 50.59 screening would be required to be qualified through
training as of June 13, 1988. However, in 1986 licensee personnel
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began receiving safety evaluation training. Also, personnel l
performing 10 CFR 50.59 screening were required to be qualified P'
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through training on November 11, 1987, when Revision 4 of Procedure NOIP 11.000.53 was issued.
In Revision 4, the preparer, reviewer and approver were required to have successfully passed a course (one day) on 10 CFR 50.59.
Therefore, the actual corrective action to improving 10 CFR 50.59 screening went into effect in November 1987 not June 1988.
Since Safety Evaluation 88-0074 was performed on March 30, 1988, the administrative controls of NOIP 11.000.053, Revision 4 were in effect. The preparer and reviewer for safety evaluation 88-0074 had completed the long course on 10 CFR 50.59 on October 13, 1987, and October 8, 1987. The approver had only received a short course (half day) on November 7, 1986.
Four of the five CSRO members who reviewed and approved the safety evaluation later on March 30, 1988, had received 10 CFR 50.59 training, ihree of the OSR0 members had received the long course and one had received the short course. The two ISEG members involved in the DER challenge and subsequent acceptance of safety evaluation 88-0074 had received the long course on May 12, 1987, and September 5, 1986. Therefore, nine personnel trained to some degree on 10 CFR-50.59 reviewed and determined that safety evaluation 88-0074 did not constitute an unreviewed safety question or a change to the Technical Specifications.
With regards to the approver of safety evaluation 88-0074 not being long course qualified', Revision 5 to NOIP 11.000.053, the one referenced in the Notice of Violation response, reduced the qualification requirements of the approver to the short course.
This revision was issued on April 4,1988, not June 3,1988, as stated in the Notice of Violation response, f.
Inadvertent Reactor Vessel Drainage Consequences of an Open Minimum Flow Valve During discussions with the licensee it became apparent that certain licensee personnel considered closure of the minimum flow valve as an additional measure to assure an inadvertent drainage of the reactor vessel did not occur. The licensee postulated drain path was from the reactor recirculation system suction piping through the RHR suction piping and the RHR minimum flow line to the suppression pool.
The licensee supported their concern about this drain path with a report from the Nuclear Safety Analysis Center..
The report was NSAC 88, " Residual Heat Removal Experience Review and Safety Analysis for Boiling Water Reactors," published in March 1986. The report reviewed 480 BWR RHR events from 1977 through 1983. Ninety of these events posed an actual loss or degradation of RHR.
Thirteen of the ninety events were from the loss of coolant from the reactor vessel.
The drain path for the loss of reactor coolant from the reactor vessel was via the
minimum flow line in four of the thirteen events.
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Recognizing that this drain path existed, the inspector evaluated the plant response to such an event. Without any operator or engineered safety features (ESF) actions the top 1/3 of the active fuel would be uncovered.
However, there is an engineered safety feature, low pressure emergency core cooling, that is required to be in service in cold shutdown.
Technical Specifications require a low reactor vessel water level 1 (31.8 inches above the top of the active fuel) actuation of core spray and/or low pressure core injection which would keep all of the active fuel covered.
There is another ESF that could automatically isolate the minimum flow drain path. The RHR suction valve isolation on low reactor vessel water level 3 (173.4 inches above the top of the active fuel) would isolate this pathway prior to ECCS actuation.
In addition this isolates any other drain paths associated with the RHR system. This particular ESF is what secured an inadvertent
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drainage of the reactor vessel during cold shutdown at Fermi in July 1987.
However Technical Specifications do not require this ESF to be operable in cold shutdown / refueling.
The inspector's evaluation was consistent with NSAC 88 which states on pages 2-3, "BWR procedures for operation of the RHR system should require that the automatic shutdown cooling isolation feature be maintained fully operational throughout cold shutdown. Operating experience has demonstrated that thir. automatic protective feature is a very effective means of terminating loss of inventory events promptly at a safe water level." Later on the same page the report states, " Essentially, every credible loss of inventory path which has the potential for uncovering fuel would be terminated by this
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automatic isolation feature."
In this regard, the inspector contacted the licensee to determine whether or not they intended to deactivate this Level 3 isolation at any time during the forthcoming refueling outage.
The licensee stated that there were no plans to deactivate the isolation during the outage.
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Results (1) Based upon the NRR response, the inspector determined that the safety evaluation performed by the licensee in concluding that a division of RHR for shutdown cooling and LPCI mode was operable with the RHR minimum flow valve disabled was incorrect and an unreviewed safety question did exist. A license amendment was not sought for the unreviewed safety question. Also, the
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Technical Specification Limiting Condition for Operation 3.5.2 and 3.4.9.2 were not complied with when the minimum flow valve was taken out of service.
This is an apparent violation (341/89017-01A(DRP)) of 10 CFR 50.59.
(2) The corrective actions of Violation 341/87021-01 were implemented when safety evaluation 88-0074 was performed but the unreviewed safety question was not identified.
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5.
Diaanostic Evaluation Team (DET) Safety Evaluation Findings a.
Preliminary Safety Evaluation Weaknesses and Licensee Response t
Sections 3.6.5.2 and 3.6.6 of the DET report dated November 16, 1988, identified five changes to the facility which required safety evaluations (SE) to be performed.
Instead preliminary safety evaluations (PSE) were performed and concluded that safety evaluations were not required.
The changes involved two temporary modifications and three engineering design packages (EDP). As a result of these findings, the licensee performed a review of six randomly selected preliminary safety evaluations.
The result was
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two of the changes were misclassified and should have required L
safety evaluations but neither change constituted an unreviewed safety question.
DER 89-0198 documented the results of the review.
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The final disposition on the DER was that procedure FIP-SR2-01,
" Preliminary Evaluations and 10 CFR 50.59 Safety Evaluations,"
issued July 15, 1988, assured that the same decisions would not i
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be made after issuance of that procedure which provided enhanced
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guidance on safety evaluations. Though it is not discussed in
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the DER, the Vice President of Engineering and Services in June
1989, tasked ISEG to review PSEs in the same manner as SEs were reviewed, b.
Inspector Review
The inspector reviewed the time frame in which the two temporary modifications and the three EDP PSEs occurred and noted that they all occurred before the licensee required qualification through training for personnel performing PSEs or SEs (November 11,1987).
The three EDP PSEs were done on August 7,1985, February 18, 1986, I
and March 1, 1986. The two temporary modifications were associated l
with the feedwater temperature reduction event which was one of the i
l examples of Violation 341/87021-01.
Therefore, these PSEs are only additional examples of the 10 CFR 50.59 evaluation weakness identified in Inspection Report 341/87021, The inspector reviewed the two PSEs which the licensee identified in DER 89-0198 as misclassified.
The two PSEs were performed on September 30, 1985, and December 2, 1986.
Once again these PSEs t
were performed prior to the licensee required qualification through training for personnel performing PSEs oe SEs (November 11,1987).
t The other four EDP PSEs reviewed were also performed prior to November 11, 1987.
Therefore, the random selection / review of PSEs
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provided no insight into the current acceptability of PSEs but reinforced what was already known about the 10 CFR 50.59 review i
process prior to qualification through training.
c.
Conclusion
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(1) The changes that required SEs but only received PSEs did not constitute unreviewed safety questions and are additional examples of Violation 341/87021-01 and no Notice of Violation should be issued.
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(2) The inspection observations and the licensee's response associated with the DET findings do not establish the effectiveness of the qualification through training requirement established in NOIP 11.000.053, Revision 4, on November 11,.1987.
6.
HPCI Discharge Valve Design Change a.
The Design Change Following MOVATS testing during the LLRT outage the licentee determined that a larger horsepower (H.P.) motor was required on HPCI discharge valve E4150F006.
Therefore, Engineering Design Package (EDP) 7964 was generated to authorize installation of a 10.83 H.P. motor to replace the 7.22 H.P. one.
On February 4, 1988, a preliminary safety evaluation was performed to support the EDP and concluded that a safety evaluation was not required, b.
Original Inspector Review During the inspection period for Inspection Report 341/88035, December 3 through December 31, 1988, the preliminary safety evaluation was reviewed by a NRC inspector and questioned as to whether a safety evaluation should have been performed.
Section 8.3.2.1.2 of the UFSAR discusses the 260/130 VDC Class 1E power system.
Included in that section was the statement, "The loads supplied from the ESF batteries and their running and starting amperage are shown in Table 8.3-15."
The table lists all the division 1 battery loads but division 2 was omitted.
Originally, the inspector considered the omission a clerical oversight. Clearly, if the direct current loading increase from the addition of the larger motor were to exceed the battery capacity the consequences of an evaluated accident would be increased. An increase in the consequences of an evaluated accident is one of the definitions of an unreviewed safety question.
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Licensee Response l
During the exit for Inspection Report 341/88035 on January 24,
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l 1989, the inspector expressed concern regarding the lack of a safety evaluation and an unresolved item was identified on this matter.
In a meeting on February 3,1989, the inspector inquired if the licensee had any further information regarding the preliminary safety evaluation.
On February 10, 1989, the licensing compliance
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supervisor issued a memorandum, NL-89-0013, to the nuclear engineering general director requesting any additional input, On February 28, 1989, nuclear engineering responded.
.The response stated:
"The design change package in question is EDP 7964 Rev. A.
This EDP was prepared to change out the motor on valve E4150F006 from 100 ft.lb. to 150 ft.lb. This EDP was reviewed in detail by both INPO and the NRC. Refer to NRC
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Inspection Report no. 50-341/88027(DRS).
No comment was
made by either group concerning the adequacy of the preliminary safety evaluation.
As a portion of the design change it was necessary to make revisions to the calculation that supports the sizing of the 260/130 VDC batteries.
It was because of the fact that these calculation changes had i
already been performed and the fact that this valve previously had a 150 ft.lb. motor that the conclusion that no unresolved safety concern was involved in the design.
The NRC has concluded that engineering was unaware that the FSAR did not contain any Div. II battery loads. Quite to the contrary, it is because of this that it was concluded that the FSAR was not affected.
The FSAR does contain a tabulation of Div. I battery loads but these serve little purpose and cause a configuration control probler:.
Engineering prefers that thi3 information be deleted from the UFSAR and has initiated a licensing change request (LCR)
to accomplish the same."
d.
Inspector Evaluation The inspector's evaluation of the licensee response as to why no safety evaluation was performed is as follows:
(1) INPO and NRC reviewed the EDP and had no comment - The
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. inspector did not contact any INPO personnel about the scope of their review.
The inspector did contact the NRC personnel involved in Inspection Report 341/88027.
They stated that the safety evaluation was not the focus of the inspection effort but the technical competency of the engineering design calculations / package. As documented in the inspection report the calculations for this design change were incomplete.
Load calculations were performed to assure that the battery would not be overloaded but calculations to assure that the valve motor would receive enough motive force (voltage / current)
were not performed as required at the time of EDP issuance.
(2) The valve had previously been 150 ft.lb. and design calculations were performed assuring an unreviewed safety question did not exist - the original calculations / design
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effort may have been adequate for the new motor not to be considered as requiring a safety evaluation from a mechanical and system response perspective.
However, additional loads had been added to the battery since the 150 ft.lb. operator had been utilized. Therefore, battery load calculation revisions were necessary..Those calculations should have been the basis of the safety evaluation to prove that an unreviewed safety question did not exist. Under the licensee's rationale, the only time a safety evaluation would be required would be when an unreviewed safety question was identified, i.e., the batteries being overloaded.
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(3) The division 2 battery loads were not listed in a UFSAR table
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This is consistent with the guidelines, the Standard Review Plan, utilized by the NRC in the licensing review of Fermi 2.
The NRC Standard Review plan, NUREG-0800, states in part:
"The descriptive information, analyses, and referenced documents, including electrical single line diagrams, electrical schematics, functional piping and instrument diagrams, logic diagrams, tables, and physical arrangement drawings for the d-c onsite power system, presented in the applicant's safety analysis report, are reviewed."
" Design information and analyses demonstrating the suitability of batteries...are reviewed to assure that have sufficient capacity and capability to perform their intended functions. This will require an examination of the characteristics and design requirements of each load...."
To examine the characteristics and design requirements of each load requires a delineation or reference to all the loads for the batteries.
Section 8.3.2 of the Safety Evaluation Report, NUREG 0798, reflected that a load review was performed since the second paragraph of that section identified that nonsafety related loads were powered from the Class IE batteries.
A central po1nt of the licensee's response was that the addition of a load on a safety related battery does not always constitute a change to the facility as described in the FSAR. This is flawed. When the NRC reviewed the design of the direct current power system each load on that system was encompassed.
Finally, as discussed in the licensee's LCR change request, the design calculation is composed of "Approximately 300 pages with approximately 50 of these pages devoted to defining the loading under various scenarios.
In condensing this loading to one page, it is difficult to avoid misrepresenting the information presented." With battery load control of such complexity it only heightens the necessity for a safety evaluation and the ensuing collegial body reviews (OSRO & NSRG).
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e.
Design Calculation Review The capacity and load determination of the safety-related batteries was performed in Design Calculation (DC) 213. The inspector reviewed the calculation to ascertain whether adding this load constituted an unreviewed safety question. The inspector began with the calculation revision of DC 213 in effect at initial licensing, Revision A,
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associated with battery 2PB against the load profile of Technical Specification surveillance requirement 4.8.2.1.d.
These two documents should have been consistent since the Technical Specification load profile has not been revised since original issuance.
The two documents were not consistent.
Page 5 of 15 of DC 213 indicated a larger load profile than is portrayed in the Technical Specifications. The same condition existed in DC 213 Revisions B and C.
Revision D reflected a calculated load profile more conservative than the Technical Specification load profile.
It was in Revision D that the higher horsepower motor was added into the battery calculations. However, there was a significant change in the assumptions reducing the continuous duty loads in conjunction with the small load addition from the HPCI valve.
Further review revealed that problems with DC 213 had been identified by a licensee initiated SSFI and the NRC in 1988.
Subsequently, this calculation was upgraded and revised in Revisions 0 and E.
f.
Effectiveness of 341/87021-01 Violation Corrective Action The inspector reviewed the licensce's 50.59 training records to determine whether the personnel involved in the evaluation of EDP 7964, Revision A, had been trained. The preparer of the PSE was qualified through training on October 29, 1987. The approver was not qualified by the training records until May 9,1988.
However, in subsequent discussion with the licensee the inspector ascertained that the approver had received 50.59 training prior to reviewing the PSE for EDP 7964 but did not take the test until May 9, 1988.
R_e sults g.
e (1) A safety evaluation should have been performed to support EDP 7964, Revision A, and failure to do so is an apparent violation (341/89017-01B(DRP)) of 10 CFR 50.59.
(2) The facility change accomplished under EDP 7964 Revision A, was not an unreviewed safety question.
(3) The licensee identified poor design practices associated with battery load control in Design Calculation 213, Revisions A, B and C, and corrected those practices in Revisions D and E.
(4) The corrective actions of Violation 341/87021-01 were implemented when the PSE for EDP 7964, Revision A, was performed but the necessity for performing a safety evaluation was not identified.
7.
Railcar Door Flood Protection Deficiency Evaluation Paragraph 5 of Inspection Report 341/89006 discussed how the licensee evaluated the flood protection aspects of a deficiency to the reactor building railcar door. The evaluation was documented in DER 89-0219
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initiated on February 8, 1989.
As discussed in Inspection Report 341/89006 the licensee inapproprlately concluded that the as-found condition was acceptable and in effect made a change to the facility that affected the licensing bases.
The technical evaluation that supported the DER conclusion was done in a memo to file performed on February 4,1989.
The two individuals associated with the memo were trained on 10 CFR 50.59 in the 1986 and 1987 time frame.
Therefore, the inspector determined that the personnel involved in this review had been trained under the corrective actions to Violation 341/87021-01 and those personnel did not identify an unreviewed safety question.
5.
Installation of Digital Fluke Meters in the Congrol Room a.
The Event Paragraph 3.b. of Inspection Report 341/89002 discussed a situation where the licensee installed three digital Fluke meters in main panels of the control room control boards. These meters were installed to replace broken recorders until the recorders were repaired. Two of the Fluke meters were installed on January 16, 1989, to monitor circulating water temperature and the third meter was installed on January 21, 1989, to monitor torus water temperature.
All the Fluke meters were installed under the work request system using work requests 001C890116 and 016C890113.
However, installation of the Flukes constituted a change to the facility and installation of the meters should have been ~ performed under the temporary modification system.
Procedure FIP-OPI-02, " Temporary Modification,"
was the controlling document.
Section 1.0, "The Purpose," stated,
"To prescribe administrative controls for temporary minor alterations made to plant equipment that do not conform with approved drawings and design documents. These alterations'are temporary in nature l
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and are expected to be installed for a short duration."
Section 6.1.5 of FIP-OPI-02 had mandatory provisions for a safety review, either a PSE and/or SE, when installing equipment / components.
The work request system did not contain such provisions. Consequently, a safety review was not performed when the change to the facility occurred.
Evidently, the four I&C. personnel involved in the three installations misinterpreted the administrative controls asscciated with interim alteration of electrical circuits while troubleshooting the recorders.
Procedure NPP-mal-03, " Interim Alteration of Electrical Circuitry,"
Section 5.1, directed the use of this document to document the lifting and landing of electrical ' leads when repairing or replacing components.
However, Section 2.2 of NPP-mal-03, stated, "This procedure does not apply to temporary or permanent modifications that are made to operable structures, systems, and components...." Also, the operating authority personnel who were cognizant of the Cluke meter installations did not recognize the need for temporary modification controls.
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On January 23, 1989, the inspector observed the installed Fluke meters, reviewed the Temporary Modification (TM) records and noted that no TM existed for these meters. The inspector immediately identified this situation to the licensee, b.
Licensee Response The licensee immediately began to initiate TM reviews but the recorders were repaired and installed before completion of the TM reviews. This occurred on January 23, 1989. Also, deviation event report (DER)89-108 was written on this situation that same day.
The DER system is the licensee's internal corrective action system utilized to meet 10 CFR 50, Appendix B, Criterion XVI, " Corrective Action." The system is established such that the in-line organizations perform the corrective action reviews and implement the corrective actions in accordance with time tables established by the in-line organization. The plant manager assigns a responsible department to perform the investigation and corrective action determination.
This department is then responsible to coordinate the corrective actions with any other departmer.ts as necessary. However, the department of plant safety, an independent organization from the in-line organization, coordinates the DER effort, reviews the corrective actions for adequacy, approves the corrective action time tables and overviews / trends the DERs.
T5e I&C department was assigned as the responsible department on DER 89-108. The corrective action to the DER was established on February 7,1989.
The remedial correct action (RCA) was:
Reinstall the recorders (completed),
Perform safety evaluations on the Fluke meter installations by February 17, 1989.
The comprehensive corrective action to prevent recurrence (CATPR)
was to train all the I&C foreman on safety evaluations with training to be scheduled by February 17, 1989.
Over the next seven months portions of the corrective actions stated above were performed. The safety evaluation training was completed on June 13, 1989.
A safety review on the circulating water Fluke meters was performed under a PSE on July 17, 1989, concluding that the installation was acceptable. As of September 13, 1989, no safety review had been performed on the torus water temperature installation.
Clearly, these completion dates were not consistent with the original time tables. With regards to performing the safety reviews of the
. Fluke meters, I&C determined that Technical Engineering (TE) should perform the reviews. However, the TE personnel continually provided I&C with dates they could not support. With regards to the safety evaluation training, two classes were necessary (May 30 and June 13)
for all 6 foremen to attend.
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The time tables changed during in-line/ plant safety correspondence on these matters.
I&C requested extensions on the corrective actions via memos dated March 8. May 2. July 11, August 23 and September 12, 1989. On February 10, April 24 June 22 and July 31, 1989, the designated plant safety reviewer reviewed and accepted revised corrective action time tables on DER Review Forms for certain corrective actions.
The DER Review Form is a plant safety internally generated form used to document the reviewer's review and communicate the reviewer's outstanding issues to the responsible department.
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Inspector Followup of DER 89-108 Dn September 13, 1989, the inspector began a follow-up to determine
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the status of DER 89-108. The inspector reviewed plant safety records on the DER and noted that all the corrective actions had not been performed.
The last memo from I&C dated July 11, 1989, stated that the safety reviews on the Fluke meters-had not been performed and requested extension to July 20, 1989.
The memo stated that the safety evaluation training had been completed.
The last DER Review Form dated July 31, 1989, stated that the documentation of the I&C safety evaluation training had not been received and an extension to August 15, 1989,- had been granted.
The form was mute to the Fluke meter reviews.
A complete review of the DER package revealed no safety reviews or safety evaluation training attendance records.
The inspector contacted the plant safety evaluator to determine what was outstanding and why.
From this discussion the inspector i
ascertained that the safety evaluations on the Fluke installations had not been completed and documentation on the I&C training was needed.
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The inspector contacted the responsible I&C individual for the DER and determined that one of the safety evaluations, the one on circulating water temperature, had been performed and the DER was
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being sent back to plant safety for DER reassignment to Technical Engineering to complete the evaluation.
Also, there were two
memos dated August 23 and September 12, 1989, from I&C to plant safety on this DER not in plant safety files.
The August 23rd memo requested an extension to August 31, 1989, for the meter safety reviews and the latter memo explained the conversations with I&C and TE on transferring the DER to TE with a completion date targeted at September 26, 1989.
Discussions with plant safety personnel revealed that the August 23rd memo had been received but not reviewed.
Discussions with TE personnel revealed that the torus water temperature safety review had not been performed.
Upon gathering this information the
. inspector requested all information on this DER and the licensee began a follow-up into this matter.
Upon review of this information, the inspector noted that four extensions had been granted by the plant safety reviewer on this DER. None of the extensions were escalated within the plant safety 16 *
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department or I&C for signature approval. The last extension for the Fluke meter safety review was to July 12, 1989.
Some of the other previous extension dates had been exceeded without plant safety approval.
No reason as to why TE could not support the safety reviews was ever provided in any of the correspondence.
Include in this information was a memorandum from the Plant Manager /Ouality Assurance Director dated September ll,1989, stating that any DER with corrective action in excess of six months will require the responsible director / superintendent approval.
From the September memo, the status of DER 89-108 and discussion with plant safety personnel, it was apparent to the inspector that DER closure was a concern.
The QA Director and the inspector requested information on how many DERs were older than six months and how many had been granted extensions.
The inspector noted that nowhere in the DER corrective actions was the operating authority discussed.
In discussion with I&C personnel and plant safety they stated that no corrective action was assigned that department.
The inspector pointed out that the operating authority should have been aware of I&C activities since they authorized the work requests and the Fluke meters were installed in plain sight in the control boards of the main control room, d.
The Circulatina Water Temperature Safety Review The inspector reviewed the July 17, 1989, safety review on installation of the Fluke meter for circulating water temperature.
The inspector noted that the TE personnel had concluded that the installation only required a PSE.
Installation of the Fluke meters constituted a' change to the facility as described in the UFSAR.
The meters were mounted within the control center panels approximately 8 inches above the controls of the division 1 RHR mechanical draft cooling tower fans.
Installation of these meters would have to meet the category II/I seismic design criteria of Section 3.7, " Seismic Design." Also, Table 3.2-1 of Section 3.2, " Classification of Structures, Systems and Components," identified the control center panels as seismic class I.
Finally, Figure 10.4-8 depicted the circulating water temperature recorder in the control center panels.
The inspector requested the head of TE to review the PSE.
Subsequently, a safety evaluation, SE 89-0164, was generated on the installation.
The inspector verified that the preparer and reviewer of PSE 89-108 had received safety evaluation training.
The preparer received training on May 13, 1987, and the reviewer received training on September 5, 1986, d.
Licensee Corrective Actions to DER 89-108-17
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Following these discussions, TE completely reviewed DER 89-108 again to determine root cause and corrective actions. This was documented on September 26, 1989, in the DER as the RCA. The results of the review and status of the corrective actions were documented as the CATPR on October 16, 1989.
The corrective actions and status were:
Operations to improve their knowledge of FMP-DP-1 (Operations to respond to this action by October 27,1989);
Plant safety staff to review DER 89-108 (complete);
Revise the DER Review Form to include signature place for Plant Safety Director to sign for acy extensions (complete);
Modification of computer system for DERs to include tracking of extension status (complete);
Counseling of maintenance and TE personnel involved in DER 89-108 (complete); and
TE to implement a tracking sy tem for informal requests.
The systems engineers were ir,tructed to use the Artemis work tracking program (comp'ete).
On October 11, 1989, the inspector interviewed the acting Head-of Plant Safety to determine what the licensee review / corrective actions were to date with DERs.
The results of the discussion were similar to those described above from-the DER.
In addition the inspector determined that draft guidance on acceptable /
unacceptable reasons for extensions was being circulated to the plant safety staff for comment for formalization at a later date.
Also, there were 1239 DERs outstanding on or about September 15, 1989.
Six hundred and ninety were greater than six months old.
Two hundred seventy nine had been granted more than one extension.
There was no current schedule for the review of these DERs to assure that the extension granted was acceptable other than these DERs needed to be reviewed after completion of the refueling outage, e.
Further Inspector Review (1) Review of the administrative procedure associated with DERs The administrative procedure governing DERs is FIP-CA1-01,
" Deviation and Corrective Action Reporting." Section 6.1.18 of FIP-CA1-01 stated:
" Submit the following for review to QA and PS.
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Results of investigation and corrective action.
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If DER is not ready for closeout retain original and include a plan with assigned responsibilities and due dates for all remaining actions.
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If action cannot be completed by the due date, an extension r quest _may be submitted with suitable i
justificetion and a revised schedule.
This will be reviewed by Director, Plant Safety and an
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extension granted, if acceptable."
l Section 5.9 of FIP-cal-01 stated:
"The responsible organization is responsible for:
5.9.1 Completing required investigations.
5.9.2 Recommending remedial and corrective actions
'q to prevent recurrence.
5.9.3 Assigning ~ actions to other organizations as required, mutually agreeing on due dates."
The inspector noted these weaknesses in the procedure:
(a) Section 6.1.18.3 used "may" instead of "shall" for submittal of a revised schedule.
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(b)' Nowhere was the term " suitable justification" in
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Section-6.1.18.3 defined.
(c) The format for submittal.and review cf extensions was not clearly-defined.
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The inspector interviewed personnel involved in the DER process and ascertained that it was generally understood
that extension requests were needed and:necessary for
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modifying completion dates.- Also, the' term suitable justification was subject to individual interpretation and j
no-formal or informal guidance had been provided to the in-line organization or the plant safety' reviewers.
i (2) Review of newly established corrective actions to DER 89-108
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The inspector noted a substantially' improved corrective
action to the DER.- However, the corrective actions were-mute to the inadequate' PSE' performed in July, the j
ramifications of the.large number of DERs that'were already~
older than six months before the Plant Manager /QA Director
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memorandum of September 11, 1989, and the weaknesses in the administrative procedure for DERs discussed above.
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Additional Licensee Actions
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On October 31, 1989, the -QA Director met with the inspector
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to discuss additional corrective actions to a numberEof the
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inspector's concerns.
The actions were in the form of a comprehensive corrective action plan. A DER, 89-1263, had been
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generated earlier that day to document the deficiencies in the DER program and action plan for those deficiencies.
Beyond those actions already mentioned previously in this report the licensee:
(1) reviewed all the DERs with greater than one extension, excluding operating experience reports, as to whether a safety review was warranted on those DERs, was an operability determination necessary, does the DER effect important to safety equipment and had the extension been approved by the Director of Plant Safety.
Some DERs were identified as needing additional safety review but once the reviews were performed no unreviewed safety questions were identified.
One hundred thirty six DERs were identified as improperly extended under the administrative procedure.
(2) required completion of root cause training of all DER. evaluators by January 1990, including revision of the root cause training to encompass the criteria for extending DERs.
(3) intended to revise procedure FIP-CA1-01 by November 15, 1989.-
(4) will systematically review' older DERs during the schedule review meeting after the refueling outage ends.
(5) will periodically have a sample of problem DERs reviewed by senior management (Vice President level) with the QA Director.
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Results i
(1) Personnel failed to implement the requirements of FIP-0P1-02-when the Flukes were installed into the control room panels.
i These actions are an apparent violation'(341/89017-02A(DRP))
of 10 CFR 50, Appendix B, Criterion V, " Instructions, Procedures and Drawings."
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(2) Complete, timely corrective action did not occur on DER 89-108 because:
(a)
In-line personnel did not adhere to the administrative controls of FIP-cal-01 for completion of corrective actions
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and did not work as a team in implementing the corrective I
actions.
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(b) The Plant Safety reviewer did not adhere to administrative controls of Section 6.1.18.3 of FIP-cal-01 for granting extensions which only allowed the Director of Plant Safety to grant extensions.
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(c) Incomplete corrective actions were initially established in j
that the operating authority was excluded.
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(d) Weak administrative procedure direction on what constituted
" adequate justification" and what was the format for extension requests, review and approval / rejection.
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Items-2a, 2b, 2c and 2d above are additional examples of an apparent violation (341/89017-02B, C, D and E) of 10 CFR 50, Appendix B, Criterion V, " Instructions, Procedures and Drawings.'
(3) There was a breakdown in the review process for granting-extensions for DER corrective actions. The proposed corrective actions to this breakdown appear comprehensive.
(4) The inadequate conc 1m ion of the July 17, 1989, PSE that installation of the rculating water temperature Fluke meters did not constitute a change to the facility is another example of an apparent violation (341/89017-01C(DRP)) of 10 CFR 50,59.
However, installation of the Fluke meters did not appear to have caused an unreviewed safety question.
(5) The corrective actions of Violation 341/87021-01 were implemented when the PSE for the circulating water temperature Fluke meter installation was performed but the necessity for performing a safety evaluation was not identified.
9.
RHR Mechanical Draft Cooling Tower Fan Brake Temporary Modification a.
The Event
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On July 11, 1989, the licensee declared the C mechanical draft cooling tower (MDCT) inoperable due to no pressure in the nitrogen bottles for the brake system (The reasons-as to why.the
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MDCT. fan was declared inoperable on July lith will be the subject
of a future special safety inspection).
The nitrogen is used as motive force for engaging the brake to the fan shaft with a solenoid valve used as the controller for controlling the nitrogen to the brake..The brake system prevents mechanical damage of the fan in the event of a tornado overspeeding the fan. At 1545', a Limiting Condition for Operation (LCO) log sheet was filled out by shift-personnel in the operating authority. The log sheet was to identify that the licensee was in a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement.
Troubleshooting activities commenced on the swing shift.
These activities determined that there was a leaking flexible hose
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between one of the nitrogen bottles and the solenoid valve controlling the nitrogen to the brake. Also, there was no i
replacement hose available in the warehouse to effect a repair.
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Therefore, later in the shift, a temporary modification, 89-0021,
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was prepared to install rigid tubing between the bottle and the
valve replacing the flexible hose.
The on-duty systems engineer and the Shift Technical Advisor (STA)
reviewed drawing M721N-2099 to determine the quality classification of the hose and determined that no seismic requirements were applicable.
Therefore, the tubing installation was not subject to seismic considerations.
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At 2100 on July 11, 1989, Technical Engineering Supervision approved the TM per telecon, and during a teleconference, OSR0
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approved the TM at 2200.
Implementation of the TM began after swing / grave shift change at 0115 on' July 12, 1989. TM implementation
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was completed at 0400.
At 0512, the fan was declared operable and LC0 89-571 was cleared.
Upon clearing the operability restraints on
the Division 1 MDCT fan, the graveyard NSS released Division 2 EDG 14
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for a planned maintenance outage.
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On dayshift, Technical Engineering personnel began
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preparation /research of the Temporary Modification for an in person i
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presentation to OSR0 later that day. As a matter of policy, all
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telecon OSR0s meetings are followed'up by an in person OSR0 meeting the next day.- This action was taken by the licensee'following recommendations of the NRC OSTI.
During the preparation, the Technical Engineering person was confused as to the quality classification of the flexible hose-after review of drawing 6M721N-2045.
He contacted cognizant personnel in Nuclear Engineering who responded to him just before the OSRO meeting.that there were seismic requirements'on the component connecting the N2 bottles with the solenoid valve.
At 1430, the'0SR0 meeting commenced. TE personnel presented.the new information on flexible hose qualification.
OSR0 considered
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the MDCT fan LC0 still in effect but the LCO had been-cleared i
10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> earlier. OSRO requested a qualified NE engineer be dispatched to the RHR complex to observe the installation at.
1530.
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The NE engineer observed the installation and considered the arrangement as unacceptable.
He reported his findings at the OSR0 meeting still in session.
Subsequently, 0SRO rejected the temporary modification. At 1800, the NSS declared the MDCT fan inoperable and the appropriate action statements were entered.
b.
Inspector Follow-up On October ^7, 1989, the inspector tried to obtain the DER on the improper :: :orary modification. After discussion with plant safety personnel and cognizant TE personnel the inspector ascertained that no DER was written on this matter.
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Returning systems / components to service without them being operable is a condition adverse to quality and a DER should have been written.
The licensee's program for identifying conditions adverse to quality required this matter to be documented on a DER.
Procedure FMD CA1,
" Evaluation and Corrective Action," Section 2.1.1, defined.
"Conditions Adverse to Quality," in part as, "... equipment deviations from approved specifications, codes, regulations, orders, drawings, standards...."
Procedure FIP-CA1-01, " Deviation and Corrective Action Reporting," Sections 2.1 and 6.1.1, require the initiation of a DER for a condition adverse to quality.
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Licensee personnel showed the inspector drawing 6M721N-2045.
Note 2 of this drawing stated, "UNLESS NOTED OTHERWISE ALL EQUIPMENT SHOWN WILL BE Q. A. LEVEL 1 SEISMIC."
However, a quality code break was designated between the solenoid valve and the nitrogen cylinders. The code break note was "QA LEVEL II SEISMIC CLASS I FOR CYLINDER MOUNTING ONLY."
From this information the personnel concluded that the code break note applied to the flexible hose, and the hose did not have seismic or QA 1 requirements.
The inspector reviewed the safety evaluation training records to determine whether the preparer and reviewer of the temporary modification were qualified.
The personnel were properly trained in 1986 and 1987, c.
Results (1) The licensee did not initiate a DER as required by FIP-CA1-01 on the improper temporary modification and subsequent improper return to service of the MDCT. fan.
This is another example of
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an apparent violation (341/89017-02F(DRP)) of 10 CFR 50, Appendix B, Criterion V.
(2) The MDCT fan temporary modification safety review was inadequate
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due to incorrect / incomplete information on drawing 6M721N-2045.
The inadequate design document material will be evaluated under i
unresolved item 341/89018-03 dealing with potential design deficiencies associated with the MDCT' fan brake system.
(3) The additional management control action of having an OSR0 '
the next day on telecon OSRO meetings was effective at identifying an improper design / installation. This reduced the window of vulnerability on the improper configuration.
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10.
Present Procedures for 10 CFR 50.59 Review The inspector reviewed the present procedure, FIP-SR1-01, the licensee had for conducting 50.59 reviews.
The inspector did not note any-significant deficiencies associated with the procede*e.
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11. Training Content on 10 CFR 50.59 a.
Inspector Review
.The inspector contacted the licensee's training organization to determine how 50.59 training was conducted.
The inspector
ascertained through interviews that training had been performed l
since 1986 by a contractor.
The inspector noted that the l
contractor's services were purchased under a non-Q purchase order
and there were no licensee approved lesson plans.
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The licensee did provide the inspector with the current training participant's manual provided by the contractor.
The inspector also reviewed the manual from 1988.
Both the 1988 and 1989 versions appeared to provide good information and insight into
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how to determine what required only a PSE and when a SE was warranted. Also the manual provided a good discussion on what the phrase "as described in the FSAR" meant. The manual included all the current approved guidance on 10 CFR 50.59.
The inspector did not note any significant deficiencies associated with the manual. Unfortunately, the inspector was not able to attend l
the actual classroom training to ascertain how the manual was
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The inspector inquired as to whether any requalification training was required to be considered trained to participate in the safety review process.
The licensee. responded that no requalification had been established, b.
Licensee Actions Subsequently, at the end of the inspection period licensee personnel informed the inspector that procedure FIP-SR1-01 was being revised to include a two year requalification period.
Retraining of all personnel on the current training manual was l
targeted for completion by March 1990. Also, the contractor training would utilize lesson plans which would be reviewed by a licensee subject matter expert.
12.
Final Observations
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The inspector noted that most of the safety review inadequacies dealt with identifying when a SE was required versus a PSE.
The present:
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- quality of training material, if properly presented, appeared adequate in this area. However, except for one individual, all the personnel
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involved in the safety review process had not been trained to the current contractor manual (1988 or 1989).
Finally, in the instance where the inspector determined that an
unreviewed safety. question existed, the licensee still did not arrive at that conclusion by the end of the inspection and the' licensee so stated at the exit.
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13.
Exit Interview (30703)
The inspectors met with licensee representatives (denoted in Paragraph 1) on November 1, 1989, and informally throughout the inspection period and summarized the scope and findings of the inspection activities.
The inspectors also discussed the likely informational content of the inspection report with regard to
. documents or processes reviewed by the inspectors during the inspection.
The licensee did not identify any such documents /
processes as proprietary. The licensee acknowledged the findings of the inspection.
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