IR 05000266/1989005
| ML20235X337 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 03/02/1989 |
| From: | Hasses R, Lougheed V, Phillips M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20235X296 | List: |
| References | |
| 50-266-89-05, 50-266-89-5, 50-301-89-05, 50-301-89-5, NUDOCS 8903130419 | |
| Download: ML20235X337 (9) | |
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U. S. NUCLEAR REGULATORY COMMISSION
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REGION III
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Reports No. 50-266/89005(DRS);50-301/89005(DRS)
Docket Nos. 50-266; 50-301 Licenses No. DPR-24; DPR-27 Licensee: Wisconsin Electric Power Company 231 West Michigan Avenue - Room 308 Milwaukee, WI 53201 Facility Name:
Point Beach Nuclear Plant, Units 1 and 2 Inspection At: Two Rivers, Wisconsin Inspection Conducted: January 30-February 3, 1989 Rf)A(.$w& kt F/;V/H
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Hasse/
Inspector:
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Date '
[/Y.itw [t V.
. Lougheed NA//f Date '
?h?[ sus hu A. M. Congiofanni 3/2/r1 IIate Approved By: Monte P ill p,
3/2[
Operational P ograms Section Date'
Inspection Summary Inspection on January 30 through February 3, 1989 (Reports No. 50-266/870X(DRS);
No. 50-301/89005(DRS))
Areas Inspected: Routine, announced inspection of design changes and modifications.
This inspection was conducted in accordance with Inspection Module 37700.
Results: One violation of 10 CFR Part 50, Section 50.59 was identified during the inspection for which no Notice of Violation i.s being issued.
(Paragraph 2.
a.(5)). The licensee showed strength in recent efforts at self-assessment and corrective action. A historical weakness in design documentation and 10 CFR Part 50, Section 50.59 evaluations was being effectively addressed.
8903130419 890302 PDR ADOCK 0500
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DETAILS
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1.
Persons Contacted Wisconsin Electric Power Company
- T. Koehler, General Superintendent, Maintenance
- B. Fromm, Modifications Engineer
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J. Knorr, Regulatory Engineer i
M. Rinzel,-Nuclear Quality Assurance Department i
- F. Flentje, Administration
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US NRC
- R.'Leemon, Resident Inspector C. Vanderniet, Senior Resident Inspector Other personnel'were contacted as a matter.of routine during the inspection.
- Designates those personnel attending the exit interview held on February 3, 1989.
2.
Design Changes and Modifications The purpose of this inspection was to determine if design changes and modifications were conducted in accordance with the requirements of the licensee's QA program, 10 CFR 50.59, the Safety Analysis Report, and the Technical Specifications.
-a.
Permanent Plant Modifications The_ inspectors reviewed eight permanent plant modifications.
The results of these reviews are given below:
(1) Modification M-784-03 This modification revised the logic for fans W85 and W86 which j
provide ventilation to the battery and inverter rooms.
Prior to this modification, these fans were stripped from the bus upon safety injection (SI) initiation.
The logic was revised
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to switch these fans to low speed upon SI initiation coincident with a loss of offsite power.
The inspector reviewed the documentation for this modification and determined it to be minimal but adequate.
Post j
edification testing was appropriate and well documented.
The safety evaluation for this modification provided a good j
assessment relative to impact on plant safety ' 0 wever, it
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failed to explicitly address the 10 CFR 50.59 criteria for
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determining the existence of an unreviewed safety question'
(USQ). The inspector was satisfied that no.USQ existed. The
- licensee displayed a historical weakness in addressing the 10 CFR 50.59 criteria and has taken corrective-action. This is addressed further in Paragraph 3 " Quality Verification
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Effectiveness" of this report.
(2) Modification IC-218 This modification installed a redundant environmentally qualified wide range reactor coolant pressure monitoring system. This modification was implemented in response to NUREG-0578 and Regulatory Guide 1.97.
The documentation for the modification appeared complete and appropriate post modification testing had been performed. The inspector had no concerns.
(3) Modification 85-14 This modification installed a new water treatment facility.
The inspector reviewed this modification only in regard to the need.to do a 10 CFR 50.59 evaluation and as to the engineering adequacy of the tie-in to the Condensate Storage Tank (CST) in the Auxiliary Feedwater (AFW) suction line.
The inspector agreed with the licensee determination that no 10 CFR 50.59 evaluation was required. The water treatment systems are not discussed in the FSAR, while the CST is mentioned only as a water source for AFW, and is not described itself. Therefore, addition of a new water treatment facility does not affect any component described in the FSAR.
The tie-in from the water treatment facility to the secondary side is on a line from which the AFW pump suction from the CST is drawn. This tie-in does not affect the AFW system's performance for the following reasons:
(1)Thequalified unlimited source for AFW in the service water system. Service water is supplied to the pump suctions through remote manual actuation from the Control Room of two motor operated valves for each suction line; (2) There is a check valve in the new line to prevent emptying the CST should a break in the new line occur (additionally, there are check valves in the AFW system to prevent backflow should a break occur in the non-seismic portion of the CST line); and (3) The tie-in is in the i
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non-seismic portion of the CST discharge line. The inspector had no concerns.
(4) Modification 87-019 Originally there was an interlock between the feedwater bypass valve and the feedwater regulating valve (MFRV) such that if i
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the bypass valve reset button was depressed the MFRV would
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close and then reopen when the pushbutton was released.
If the reset button was pressed while the reactor was at power, a feedwater trarsient would result.
This modification removed the interlock by disconnecting a relay and permanently jumping around a contact for that relay.
The inspector reviewed the modification package (including the 10 CFR 50.59 evaluation and engineering design), the completed drawings, the installation and testing procedures and results, and the completed training procedures.
The inspector also discussed with the training coordinator _how training procedures were written and used.
The modification was completely
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incorporated into all drawings and procedures.
The inspector noted a problem with respect to the post modification testing results:
the MFRV opening and closing times exceeded the acceptance criteria.
Review of some earlier l
modifications to the MFRV (83-111,84-046, 84-129, and 86-008)
indicated that only the SI or high-high steam generator level initiated closing time for this valve is governed by the design basis accident analyses.
The licensee has a letter from the NSSS architect-engineer stating that a longer MFRV closing time was acceptable.
The closing time for any other case and the opening time following a. reactor trip have no impact on FSAR safety analyses.
As it presently stands, to exceed the acceptance criteria for
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the SI/hi-hi SG level closing time requires a management (I&C I
Supervisor) signature but no documentation of why or how he determined it was acceptable.
The remaining closing and opening times are given as " approximate" times and can be exceeded without any approval.
No guidance exists in the test to say by how much they can exceed the approximate acceptance criteria.
It is not an acceptable practice to specify acceptance criteria and then, basically, ignore them.
The basis may have been clear to the testing personnel; however, there was no documentation of the rationale for accepting the results.
Thic concern was discussed with the Modification engineer.
A discussion was also held with a representative of the Nuclear Quality Assurance Division (NQAD).
He indicated that the lack of setpoint bases has been identified as a generic issue as a result of the in-house vertical slice review of the
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Residual Heat Removal system.
The reconstruction and
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documentation of the design basis, including setpoints, was an
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issue under consideration by the utility and that work had already been started on reconstructing the electrical design bases.
There is no safety concern since the longer MFRV closing time for SI actuation has been reviewed and accepted, in terms of the accident analyses, by the NMSS architect engineer.
The licensee should review its administrative procedures for
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acceptance of modification or surveillance test results to ensure that adequate documentation showing why the result I
is acceptable, is required when acceptance criteria are exceeded. The inspector had no further concerns.
(5) Modification 84-169 Thi:; modification replaced the tube bundles in the #1 and #2 low pressure feedwater heaters from Unit 1.
Similar modifications replaced the tube bundles in the Unit 1 #3 low-pressure heaters and in the #5 high-pressure heaters, as well as the same heaters on Unit 2 (Modifications84-170,84-171,84-172,85-059,and85-060).
Because of the extensive plugging that had occurred, the decision was made to use a different material than was originally used. This resulted in changing a number of parameters for the tubes (e.g., the total number of tuber.
I and the design pressure). According to the modification supervisor, these changes were necessary to maintain the same heat transfer rate - and were only partially successful,
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as the heat transfer rate decreased slightly.
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No 10 CFR 50.59 evaluation was done for any of these changes.
At the time of the modifications, a 10 CFR 50.59 review was done only if it was felt (see Paragraph 3 for further discussion of this issue) an FSAR Chapter 14 accident analysis would be affected. Since replacement of the heaters was considered straightforward plant maintenance, no 10 CFR 50.59 was deemed necessary.
FSAR Section 14.1.6, " Reduction in Feedwater Enthalpy Incident," discusses opening of the bypass valve around the low-pressure feedwater heaters. since the design parameters of the heater tube bundles were changed, the results of this analysis could be impacted. This concern was discussed with the Modification engineer. He was knowledgeable of the FSAR section and agreed that a 10 CFR 50.59 evaluation should have been done. He agreed to perform the evaluation for these modifications to determine if an unreviewed safety question existed.
The failure to perform the evaluation is a violation of 10 CFR 50.59. Since this appeared to be an isolated violation and the licensee had taken effective generic action to improve the implementation of this regulation, no Notice of Violation will be issued pursuant to 10 CFR Part 2, Appendix C, l
Section V.A.
Completion of this evaluation will be tracked as Open Item (50-461/89005-01; 50-301/89005-01).
(6) Modification 88-124 This modification replaced the existing single Safety Injection (SI) block switch with two train specific block switches. A
discrepancy concerning the potential for blocking both trains
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by a single switch failure was identified by the licensee.
This modification separated the trains to prevent such'a failure.
The modification package was reviewed for technical content and adequacy.
The inspector reviewed selected drawings and procedures and noted that all appropriate changes were made.
The 10 CFR 50.59 evaluation was complete with enough detail to describe the impact of the modification.
The inspector noted that the format of the evaluation had improved and contained a more detailed review than in previous years.
The package appeared to be complete and no concerns were identified.
(7) Modification 86-006 This modification removed the containment isolation trip (T).
signal from each steam generator blowdown containment isolation valve, CV-2042 and CV-2045.
Removal of the T signals eliminated the requirement'to perform the 10.CFR 50, Appendix J, Type C leak rate test on these valves.
Prior to May 1980, each blowdown line was a closed system with-one manual isolation valve, CV-2042 or CV-2045, located outside containment.
In late 1979, the licensee was required to comply to NUREG 0538 which stated that the initiation of containment isolation shall be from diverse signals and that all non-essential systems shall be automatically isolated under containment isolation conditions.
In May 1980, the licensee committed to provide the double barrier isolation by diverse signals and restored the T signals to the valve.
Double barrier protection was provided by the closed system and automatic isolation valves.
In June 1981, the licensee installed an additional isolation valve inside containment on each blowdown line in response to IE Bulletin 80-11 concerning masonry wall design. Valves CV-2042 and CV-2045 were subject to falling debris during a seismic event and thus possibly would not be capable of their isolation function.
Further investigation by the licensee showed that the blowdown lines were inconsistent with the definition of a closed system as described in the Final Safety Analysis (FSAR).
To qualify as a closed system, the lines had to be seismically qualified and missile protected throughout its length to the penetration.
Subsequent analysis showed that the blowdown
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lines in Unit 1 were qualified and additional modifications were needed to qualify the lines in Unit 2.
Nonconformance Report, NCR 85-44, was written to track this item.
Discussions with the licensee and NRR showed that the removal of the T signals did not change the commitment made in May 1980.
Double barrier protection was provided by a seismically qualified closed system and an inside containment
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. isolation valve.
Removal of the T signals.was also cons.istent L
with 10 CFR 50, Appendix A, General Design Criterion 57, in that ' remote manual isolation valves,. CV-2042 and CV-2045, existed outside containment.
The, inspector had no further concerns regarding the. intent of the modification.
The' inspector reviewed the final documentation for completeness
.and. technical adeouacy.
The 10 CFR 50.59 evaluation accurately addressed the technical issues and was found to be adequate.
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l Tho inspector reviewed several procedures and controlled drawings and determined;that all necessary changes were completed.
Table 5.2-1 of the FSAR was revised to reflect
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the changes to penetrations 50 and 51..The inspector had no concerns.
(8) Modification 85-233 This modification installed an additional U-bolt type restraint.
on an one inch drain line connected to the A loop steam
~ generator blowdown line.
The modification was necessary to seismically qualify the blowdown lines in Unit 2.
The documentation appeared to be complete.
The inspector reviewed the calculations and had no concerns.
b.
Temporary Modifications Temporary _ modifications were controlled by Procedure PBNP 4.17,
" Temporary Modifications." Temporary modifications packages were maintained in the control room along with a log documenting implementation and closeout dates.
The packages contained the implementing form, tagging records, 10 CFR 50.59 evaluations, and supporting documentation including calculations if appropriate.
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temporary modifications were reviewed semiannually by the operations staff for continued appropriateness.
Field. tags were also inspected on a six month basis to assure'they were still in place and legible.
The inspector reviewed the open packages located in the control room.
and a sample of closed packages.
There were a total of 49 open packages (including Units 1 and 2).
Nineteen had been issued prior to 1988.
All had closure actions identified (e.g., permanent modification or completion of maintenance activities).
Supporting documentation including 10 CFR 50.59 evaluations was complete.
The evaluations performed pursuant to 10 CFR 50.59 displayed much better quality than some reviewed for permanent modifications.
The inspector' attributed this to the licensee's recent effort to improve performance in this area.
Most temporary modifications were recent and reflected licensee success in improving these evaluations (this was also reflected in the quality of recent evaluations performed for permanent modifications).
The inspector identified no concerns.
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3.
Quality Verification Effectiveness
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The inspectors reviewed recent licensee efforts in assessing and l
improving the quality of modifications and plant design basis documentation. These efforts included:
A Safety System Functional Inspection of the emergency diesel
generator system performed by a contractor (WESTEC).
A vertical slice audit of the Residual Heat Removal System (RHR)
performed by the Nuclear Quality Assurance Department (NQAD).
A Unit 1 outage modification surveillance performed by NQAD.
- A special audit of 10 CFR 50.59 evaluations performed by a
consultant.
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All findings from these activities were issued as Audit Finding Reports (AFRs) by fiQAD to obtain and track corrective actions.
The inspector reviewed the AFRs and corrective actions.
Generic actions prompted at least in part by the results of these audits included:
The development of a procedure to address FSAR revisions.
- A review of all modifications done in 1987 to determine if
10 CFR 50.59 evaluations were performed when necessary and if those performed were adequate.
An upgrade of the procedure covering the preparation of 10 CFR 50.59
evaluations with a commitment for further upgrade when the NUMARC guidelines for performing these evaluations were issued.
Initiation of an effort to upgrade or regenerate the plant design
basis.
The effectiveness of the efforts to upgrade design documentation and the quality of the 10 CFR 50.59 evaluations was apparent to the inspectors by the improved quality in these areas for the more recent modification packages reviewed.
The inspectors were satisfied that the licensee was implementing an effective quality verification program and taking appropriate actions as a result of the findings.
4.
Open items Open items are matters which have been discussed with the licensee which will be reviewed further by the inspector, and which involve some action on the part of the "I nr licensee or both.
An open item disclosed during the inspection is discussed in Paragraph 2.a.(5).
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5.
Exit Interview
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The-inspectors met'with licensee representatives (denoted in Paragraph 1)
at the conclusion of the inspection on February ~3, 1989, and summarized l
the purpose, scope and findings of the inspection. The licensee stated that the inspectors had no access to proprietary information.
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