ML20141E479
| ML20141E479 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 06/26/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20141E437 | List: |
| References | |
| 50-483-97-05, 50-483-97-5, NUDOCS 9707010071 | |
| Download: ML20141E479 (39) | |
See also: IR 05000483/1997005
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ENCLOSURE
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION IV
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Docket No.:
50-483
License No.:
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Report No.:
50-483/97-05
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Licensee:
Unicn Electric Company
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Facility:
Callaway Plant
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Location:
Junction Hwy. CC and Hwy. O
Fulton, Missouri
Dates:
February 10-14 and 24-28,1997
Inspectors:
T. Stetka, Team Leader, 8: .gineering Branch
P. Goldberg, Reactor lnspector, Engineering Branch
W. Wagner, Reactor inspector, Engineering Branch
K. Thomas, Project Manager, Office of Nuclear Reactor
Regulation
Approved By:
C. V::nDenburgh, Chief, Engineering Branch
Division of Reactor Safety
ATTACHMENT:
Supplemental Information
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9707010071 970626
ADOCK 05000483
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TABLE OF CONTENTS
EXECUTIVE SUMM ARY . . . . . . . . . . . .
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Report Details . . .
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1. Engineering . . . . . . .
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Conduct of Engineering
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E1.1
System Review
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E1.2 Temporary Plant Modification Review . . . . . . . . . . . . . . . . . .
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E1.3 Suggestion-Occurrence-Solution Report Review . . . . . . . . . . . . . 4
E2
Engineering Support of Facilities and Equipment . . .
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E2.1
Review of Facility and Equipment Conformance to the Final
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Safety Analysis Report Description . . . . . . . . . . . . . . . . .
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E2.2 Validation and Control of Design Basis Documents . . . . . . . . . . . 8
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E2.3 Engineering Backlog . . .
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E2.4
10 CFR 50.59 Implementation
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E2.5 System Walkdowns . . . . . . . . . . . .
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E4
Engineering Staff Knowledge and Performance . . . . . . . . . . . . . . . . . . 16
E5
Engineering Staff Training and Qualification . . . . . . . . . . . . . . . . . . . . 17
E6
Engineering Organization and Administration
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E7
Quality Assurance in Engineering Activities
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V. Management Meetings . . . .
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Exit Meeting Summary . . . . . . . . . . .
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ATTACHMENT: Supplemental Information
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EXECUTIVE SUMMARY
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NRC Inspection Report 50-483/97-05
Callaway Plant
This team inspection evaluated the current effectiveness of the licensee's plant and design
engineering organizations to respond to routine and reactive site activities, which included
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the identification and resolution of technical issues and problems. This inspection asse:: sed
engineering and technical support by focusing on the functional aspects of the component
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cooling water system. The inspect!on also reviewed 10 CFR 50.59 safety evaluations and
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screenings, engineering evaluations for design modifications, and general engineering
performance. The inspection covered a 4-week period with 2 of these weeks conducted
onsite,
Enaineerina
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The conduct of engineering activities was considered to be generally good. Aspects
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of good engineering practices included strong system engineers, a minimum
engineering backlog, effective control of plant modifications, good interfaces
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between engineering and other plant disciplines, a good design basis information
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process, and a very effective independent safety engineering group. However, the
inspection identified two examples involving Technical Specification interpretations
for the diesel generator building supply fans and the refueling machine, wherein the
10 CFR 50.59 safety evaluation screening process was ineffective. The inspection
also identified one instance wherein a safety evaluation was not performed for a
change to the method of operation of the post-accident sampling system. The
inspection also identified two instances where required reports were not made to
the NRC.
Modification packages for the component cooling water system were found to have
appropriate safety evaluations and post-modification testing reauirements to assure
component operability (Section E1.1).
The majority of the request for resolutions reviewed were of gor,d quality. The
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request for resolutions had proper engineering justification and proposed corrective
actions (Section E1.1).
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Work packages were found to be performed in accordance with their instructions
and no recurrent problems were noted. The team also concluded that no work
requests resulted in a modification to the system (Section E1.1).
The temporary plant modification program was found to be 'in conformance with
procedures and properly managed (Section E1.2),
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The majority of the suggestion-occurrence-solution reports had resolutions with
proper engineering justifications and proposed corrective actions. One violation was
identified for the failure to issue a licensee event report when main steam safety
valves had as-found setpoints in excess of the Technical Specification setpoint
tolerances (Section E1.3.b.1).
While some discrepancies between the actual plant con' guration and procedures
and the Final Safety Analysis Report were noted, an at.oon plan existed to correct
such deficiencies (Section E2.1).
Effective controls were implemented to ensure that c;esign basis documents were
available, were being adequately maintained, and were easily retrievable
(Section E2.2).
The backlog of engineering work was properly managed (Section E2.3).
Overall, procedural guidance for implementation of 10 CFR 50.59 was appropriate.
However, the inspection identified two examples in which the guidance was not in
accordance with the requirements of 10 CFR 50.59. Specifically, the licensee did
not report safety evaluations for temporary modifications and did not require safety
evaluations when a change was considered to be a plant improvement. The team
identified the first example as a violation of 10 CFR 50.59(b)(2) (Sections E2.4.1b.2
and E2.4.1.b.3(1)).
The implementation of the 10 CFR 60.59 program was adequate; however, the
licensee failed to perform a safety evaluation for a modification to the post-accident
sampling system. This modification changed the method of operation of the system
described in the Final Safety Analysis. Report. This was considered to be an
apparent violation (Section E2.4.1b.3(2)).
The implementation of the Technical Specification interpretation program was
adequate; however, the team identified two interpretations that provided guidance
that was contrary to the Technical Specification requirements and the Final Safety
Analysis Report. in the first example, the interpretation effectively changed the
setting of the trip setpoints for the refueling machine without performing a 10 CFR 50.59 safety evaluation. In the second example, the interpretation changed the
operation of the diesel generator building supply fans from automatic to manual
operation. This change may have increased the probability of occurrence of a
malfunction of equipment important to safety previously evaluated in the Final
Safety Analysis Report and increased the possibility for a malfunction of a different
type than any evaluated previously in the Final Safety Analysis Report. This change
was considered to potentially constitute an unreviewed safety question. Both
examples were cited as apparent violations of 10 CFR 50.59 (Sections E2.4.2b.1
and E2.4.2b2).
The plant material condition and housekeeping were good and some improvements
in material condition was noted. The boron control program was considered to be
effective in improving the plant's material condition (Section E2.5).
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Engineering management expectations were clear and understood by engineering
personnel. Communications between system engineering and other plant
departments were effective. System engineers were knowledgeable of their
assigned systems (Section E4).
The training program for the engineering staff was effectively supporting the role of
the system and design engineers (Section ES).
The independent safety engineering group was effective in providing an independent
assessment of plant operations, providing an independent assessment of the effect
of internal and external events on plant operations, and in providing
recommendations to improve plant safety. The use of an independent safety
engineering group engineer as a shift technical advisor was considered to be an
effective and notable application of the group's experience base (Section E6).
Self assessments of engineering activities were conducted through the use of
quality assurance audits and surveillances. The results of these audits and
surveillances were generally consistent with the team's findings (Section E7).
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Report Details
I. Enaineerina
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Conduct of Engineering (37550)
E1.1
System Review
The team reviewed the component cooling water system to verify the licensee's
ability to maintain this system in an operable status. The team reviewed the
adequacy of the licensee's plant modification process, engineering calculations,
problem evaluation requests, and technical evaluation requests. In addition, the
team interviewed the system engineer to determine the engineer's knowledge of the
system.
E1.1.1 Permanent Plant Modification Review
a.
Inspection Scone
The team reviewed two safety-related plant modification records to verify
conformance with applicable installation and testing requirements as prescribed by
procedures. Specific attributes reviewed and/or verified by the team included:
10 CFR 50.59 safety evaluations, post-modification testing requirements,
safety-related drawing updates, Final Safety Analysis Report updates, training
requirements, and field installation,
b.
Observations and Findinas
The team reviewed Plant Modification Records RFR 16981 A, " Replacement Gages
and Transmitters," and RFR 17402A, " Material Change for Component Cooling
Water Heat Exchanger End Cover Gasket." These modifications had a proper
10 CFR 50.59 screening or safety evaluation performed and neither represented an
unreviewed safety question. The team also found that the post-modification testing
requirements were adequate to assure component operability. The team verified
that affected drawings and procedures were updated for the plant modification
records. In addition, the team verified, through walkdowns, that the physical
installations of these plant design changes were consistent with the descriptions in
the modification packages.
E1.1.2 Reauests for Resolution Review
a.
Inspection Scoce
Requests for resolution were used to request technical evaluations, document the
evaluation, recommend action, and obtain management concurrence. The team
reviewed ten requests for resolution associated with the component cooling water
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system and other selected areas to determine whether proper engineering resolution
was performed and that issues requiring the use of the plant modification process
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were properly identified.
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b.
Observations and Findinas
The team reviewed Request for Resolution 16444, " Operability of Component
Cooling Water Pumps with Safety-Related Room Coolers inoperable," which
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requested an engineering evaluation to justify the component cooling water pump
operability while the pump room coolers were inoperable. This operation was
permitted by Technical Specification Interpretation 35. This request for resolution
was originally written for Technical Specification 3.7.12. This Technical
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Specification was subsequently deleted and the requirements incorporated into Final
Safety Analysis Report, Section 16.7.4.1. Technical Specification interpretation 35,
which was revised so that it applied to the new Final Safety Analysis Report
section, specified that the component cooling water pumps could withstand
ambient room temperatures of up to 119 F without pump room coolers. The team
also noted, however, that the licensee had calculations indicating that the pump
room temperature could increase to 128 F during accident conditions without the
pump room coolers. The team noted that the request for resolution only stated that
the pumps were qualified to 121 F and did not address the effect of the higher
128 F temperature on pump operation. However, following discussions with
engineering personnel, the team determined that the pumps were capable of
withstanding temperatures in excess of 128 F. T he team concluded that this
request for resolution was inadequate because it lacked the pertinent information
needed to determine that all aspects of the issue were addressed. The team
considered this to be an isolated occurrence.
E1.1.3 Review of Enaineerina Calculations
a.
inspection Scope
The team reviewed the adequacy of two design engineering calculations associated
with the component cooling water system to determine whether the calculation
assumptions were technically reasonable and properly supported.
b.
Observations and Findinas
The team found that the licensee's calculations were satisfactory. The calculations
reviewed provided sufficient information and assumptions to reach the conclusion
stated. The team concluded that the hcensee's calculations were acceptable.
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E1.1.4 Review of Work Reouests
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a.
Insoection Scope
The team reviewed 18 work requests associated with the component cooling water
system to determine if repetitive problems existed and to deterrnine the present
material condition of the system. This information was compared with the results
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of the system walkdown. In addition, the work requests were reviewed to
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determine if any unauthorized modifications were being performed using work
requests.
b.
Observations and Findinas
The team found that the work packages were performed in accordance with their
instructions and that the engineering staff was knowledgeable of the work
performed. No recurrent problems were noted. The team's walkdown results
indicated that the licensee was maintaining the system in good condition and that a
very low threshold for deficiency identification had been established. The team did
not find any recent work reqwsts that resulted in a system modification,
c.
System Review Conclusions
The team found that the modification packages reviewed included appropriate
safety evaluations and that post-modification testing was appropriate to assure
component operability.
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The team concluded that in all but one isolated instance, the request for resolutions
reviewed were of good quality. The request for resolutions had proper engineering
justifications and proposed corrective actions.
The team concluded that the work packages reviewed were found to be performed
in accordance with their instructions and no recurrent problems were noted. The
team also concluded that no work requests resulted in a modification to the system.
E1.2 Temocrarv Plant Modification Review
a.
Inspection Scope
The team reviewed temporary plant modifications to verify conformance with
applicable installation and testing requirements as prescribed by licensee
procedures. Specific attributes reviewed by the team included: 10 CFR 50.59
safety evaluation, license impact review, post-modification testing requirements,
plant installation, and the process for periodically reviewing the status of the
modifications.
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b.
Observations and Findinas
The team found that there were eight open temporary plant modifications. Seven of
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these modifications were nonsafety related and one was safety related. The team
reviewed Temporary Modification TM-960E010, " Removal of the SR Power Supply
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Source from D/P Gauges GKPDIS50028,39,100, and 103 on SGK04A, SGK048,
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SGK05A, and SKGOSB." and found that the modification had the proper safety
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evaluations, license impact review, and that the post-modification testing
requirements were properly specified. The team also verified that the control room
had a copy of the temporary modification and that the affected equipment in the
plant was properly tagged. The team found that the temporary plant modification
was being tracked for closure.
c.
Conclusions
Based on the review of this one temporary modification and the low number of open
temporary modifications, the team concluded that the temporary plant modification
program was in conformance with procedures and being properly managed.
E1.3 Suaaestion-Occurrence-Solution Reoort Review
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Insoection Scope
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The licensee issued suggestion-occurrence-solution reports as a means to identify
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problems with components and systems and to place these problems in their
corrective action system for resolution. The team reviewed 37 suggestion-
occurrence-solution reports to determine the adequacy of the resolution, whether
the component / system operability was properly determined, and that the proposed
corrective actions were adequate to preclude recurrence, in addition, the team
interviewed the applicable licensee personnel to discuss the resolution of the
suggestion-occurrence-solution reports,
b.
_ Observations and Findinas
During the review of these suggestion-occurrence-solution reports, the team
identified instances where main steam safety valves and a pressurizer safety valve
were found to have lift setpoints that were out of tolerance. The team found that
these findings were not reported to the NRC.
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The team reviewed Suggestion-Occurrence-Solution Report 95-0508, which
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reported that 14 out of the 20 main steam safety valves exceeded their Technical
Specification setpoint tolerance of i1 percent when surveillance tests were
performed during Refueling Outage 7. The team reviewed the test data and noted
that 3 of the 14 valves exceeded the setpoint tolerance by more than 3 percent
(+ 3.6, + 3.01, and + 3.4 percentL One of the four steam lines had all five valves
outside of the setpoint tolerance and included the two valves with tolerances
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greater than + 3 percent of the setpoint. The licensee's corrective actions were to
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adjust and re-test tha valves. The re-test indicated that the valves were set
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properly. In addition, the licensee directed their nuclear steam system supplier,
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Westinghouse, to perform an analysis of the safety related impacts of having the
main steam safety valve tolerance at + 3.6 percent and to provide information to
support a Technical Specification amendment submittal to the NRC to increase the
valves' setpoint tolerance to + 3, -1 percent.
When questioned, the licensee indicated that a licensee event report had not been
issued because they considered that exceeding the setpoint tolerance occurred at
time of discovery and not during the operating cycle. In addition, the licensee
stated that the three main steam safety valves that had opening setpoints greater
than the +3 percent tolerance, exceeded their safety analysis assumptions and the
one valve that had a setpoint less than -1 percent of the tolerance exceeded the
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component fatigue analysis assumptions. However, the licensee also stated that
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preliminary evaluations indicated that the excessive setpoints were enveloped by
the existing safety analyses.
The team concluded that due to the number of failures, it was unlikely that the main
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steam safety valves failed at the time of discovery. The team also concluded that
during Refueling Outage 7, two independent trains became inoperable in the main
steam system that were designed to mitigate the consequences of an accident.
The licensee disagreed that these valve failures needed to be reported. Their basis
for this disagreement was that they complied with the requirements of
10 CFR 50.73 and the guidance provided in NUREG 1022, Revision 0, " Event
Reporting Guidelines for 10 CFR 50.72 and 50.73," which specified that when
f ailures occur, the failures are assumed to have occurred at the time of discovery
and not during the operating cycle,
in a November 2,1993, memorandum issued by the Office of Nuclear Reactor
Regulation, which was sent to the existing eight Region IV power reactc,r licensees,
the staff stated that the guidance in NUREG 1022, Supplement No.1, " Licensee
Event Reporting System," was clear that if conditions were discovered during an
outage, but were believed to have existed during operation, they were reportable,
as long an applicable threshold for reporting was reached. Although the licensee
was not a Region IV licensee at the time the letter was sent, the licensee obtained a
copy of the letter from another plant and was aware of the NRC position.
10 CFR 50.73(a)(2)(vii) requires that an event be reported when a single cause or
condition caused at least one independent train or channel to become inoperable in
multiple systems designed to mitigate the consequences of an accident. The failure
to issue a licensee event report for this occurrence was considered to be a violation
of 10 CFR 50.73 (50-483/9705-01).
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b.2 Pressurizer Safety Valves
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The team reviewed Suggestion-Occurrence-Solution Report 96-1273, which
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reported that, while performing surveillance testing during Refueling Outage 8, one
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pressurizer safety valve opened at -2.31 percent. This opening pressure exceeded
the Technical Specification setpoint tolerance of t1 percent. The licensee's
corrective action was to adjust and re-test the valve. The adjustment was
completed and the valve had a satisfactory re-test.
While reviewing the pressurizer safety valve data, the team also determined that,
prior to Refueling Outage 8, the licensee upgraded their inservice testing program to
incorporate the 1989 edition of the ASME Code,Section XI. This ASME Code
edition does not require increasing the sample size unless the as-found valve
setpoint exceeds the setpoint criteria by 3 percent or greater, even though the
Technical Specification setpoint tolerance is il percent. Due to the change in
code years for inservice testing, the licensee did not test the other two pressurizer
safety valves. The team discussed this with the licensee and determined that the
Final Safety Analysis Report, Chapter 15, safety analysis was based on the
Technical Specification tolerance of
1 percent and not the ASME Code allowed
tolerance of i3 percent.
While reviewing Suggestion-Occurrence-Solution Report 96-1273 regarding the
failed pressurizer safety valve, the team noted that the plant could have
exceeded the Chapter 15 safety analysis. The licensee received a preliminary
analysis from their vendor (Westinghouse) in letter SCP-97-105, dated February 26,
1997. This letter concluded that the out-of-tolerance valve was enveloped by
the accident analysis. However, the letter also stated that a change in the
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pressurizer safety valve setpoint from -1 to -3 percent would require a change to
one of calculations before permanent implementation of an increased setpoint
tolerance. The final Westinghouse analysis to support an increased setpoint
tolerance for the pressurizer safety va!ve and the licensee's plant-specific
calculations will be reviewed when available. to verify that the increase to a -3
percent setpoint would still be enveloped by the accident analysis. This is
considered to be an inspection followup item (50-458/9705-02).
c.
Conclusions
The majority of the suggestion-occurrence-solution reports had resolutions with
proper engineering justification and the proposed corrective actions were adequate.
A 10 CFR 50.73 violation was identified for the failure to issue a licensee event
report when the as-found rnain steam safety valve setpoints exceeded the Technical
Specification setpoint tolerances.
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E2
Engineering Support of Facilities and Equipment (37550)
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E2.1
Review of Facility and Eauioment Conformance to the Final Safety Analysis Report
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Descriotion
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a.
Inspection Scope
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A recent discovery of a licensee operating its facility in a manner contrary to the
Final Safety Analysis Report description highlighted the need for a special focused
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review that compares plant practices, procedures and/or parameters to the Final
Safety Analysis Report descriptions. As the result of this discovery, the inspectors
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reviewed selected sections of the Final Safety Analysis Report.
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b.
Observations and Findinas
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The team identified three discrepancies between the Final Safety Analysis Report
and the actual plant configuration. These discrepancies are discussed in
Sections E2.4.1.b.3(2), E2.4.2.b.1 and E2.4.2.b.2 of this report.
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The team also noted that there was a previous NRC finding that identified that
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the component cooling water system temperature was below the lower limit
specified in the Final Safety Analysis Report (as described in NRC Inspection
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Report 50-483/96-11). As a result of this finding, the licensee reviewed the
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component cooling water system safety system functional assessment that was
previcusly performed. This review was initiated because the assessment was
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limited to a design basis review instead of a review of all areas of the Final Safety
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Analysis Report that pertained to component cooling water. This review was
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conducted prior to this inspection to determine if cdditional problems existed. As a
result of the review, the licensee identified several discrepancies between the Final
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Safety Analysis Report and the actual configuration and operation of the component
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cooling water system.
Since the licensee had conducted seven safety system functional assessments, the
licensee further expanded their review to include the safety system functional
assessments that were previously conducted on the essential service water system
and the auxiliary feedwater system. These reviews also identified several
discrepancies between the Final Safety Analysis Report and the actual configuration
and operation of these systems. Subsequent to this inspection, the licensee
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decided to review the remaining four safety system functional assessments as a
part of a task team formed in March 1996 to review all sections of the Final Safety
Analysis Report.
The purpose of this task team was to identify and prioritize sections of the Final
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Safety Analysis Report for a compliance review against plant hardware and
procedures. The task team completed its review in July 1996 and identified actions
and prioritized sections for further review. These additional reviews were scheduled
for completion prior to the end of 1998. This licensee effort was documented in
their letter ULNRC-03530 dated February 5,1997, to the NRC regarding an NRC
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enforcement policy revision. Enforcement Guidance Memorandum EGM-96-005,
dated October 21,1996, set forth a revised enforcement policy applicable to
voluntt.ry licensee efforts to correct inconsistencies in licensing documents,
including programs for licensee reviews of the Final Safety Analysis Report.
The licensee's efforts to determine the extent of Final Safety Analysis Report
discrepancies were ongoing during this inspection. Further review of this effort will
be conducted during future inspection efforts. This is considered to be an
inspection followup item (50-483/9705-03),
c.
Conclusions
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While some discrepancies between the actual plant configuration and procedures
and the Final Safety Analysis Report were noted, the team concluded that the
licensee had a process underway to identify and correct such deficiencies.
E2.2 Validation and Control of Desion Basis Documents
a.
inspection Scope
The team reviewed the licensee's controls of design basis documents to determine
if the documents were available, maintained, validated, and were easily retrievable,
b.
Observations and Findinas
The team found that the licensee's program for identification and control of design
basis documents was described in Procedure EDP-ZZ-04055, " Design Basis
Control." This procedure also described the sources of design basis information and
how this information was located, validated, and maintained for future use. The
team found that system design basis validation was accomplished through safety
system functional assessments performed by the quality assurance organization.
The licensee informed the team that seven safety system functional assessments
were performed between 1988 and 1995, which validated 17 of 45 safety-related
systems.
The team reviewed the license's response to the NRC request for information
pursuant to 10 CFR 50.54(f) regarding adequacy and availability of design basis
information. The licensee response was documented in letter ULNRC-3531, dated
February 6,1997. The team found this letter contained two licensee commitments
for future work intended to verify the adequacy and availability of design basis
information. Those commitments were: (1) initiate a review of the Callaway Plant
in accordance with the Nuclear Energy Institute initiative as described in Nuclear
Energy Institute 96-05; and (2) perform two safety system functional assessments
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by December 31,1998. The team found that the Nuclear Energy Institute review
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focused on licensing basis information, which included a sample review of Final
Safety Analysis Report information, whereas, the safety system functional
assessments focused on design basis information to support the as-huilt
configuration of the plant.
The licensee informed the team that they would evaluate the results of their reviews
to determine if they had reasonable assurance that the original design basis would
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be maintained for future use. The licensee stated that the need for additional
validation of design basis information would also be based on this evaluation.
The team observed implementation of the licensee's program for maintaining,
updating, and retrieving design basis information for the emergency service water
system, the residual heat removal system, and the auxiliary feedwater system. The
documentation was adequately maintained and easily retrieved.
c.
Conclusions
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The team concluded that the licensee implemented effective controls to ensure that
design basis documents were available, were being adequately maintained, and
were easily retrievable.
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E2.3 Enaineerina Backloa
a.
Insoection Scope
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The team evaluated the extent of backlogged engineering work to determine the
size of the backlog and to determine whether it was being properly managed,
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b.
Observations and Findinas
The team found that the engineering backlog consisted of 238 suggestion-
occurrence-solution reports,138 request for resolutions,153 central action
tracking items, and 151 modification packages ror a total of 680 items. The
team found that the licensee had a process to set priorities, such that, work
and resources were allocated first to the most significant items. This process
assigned a weighing factor to set the priority within a category. For example,
a weighing factor of five was assigned for category items involving nuclear,
industrial, or radiological safety. The lowest weighing factor was a two, which
was assigned to items involving management discretion. The team did not identify
any safety-significant issues that were not being properly resolved. The team
reviewed the trend of the backlog items and found that the number of backlog open
items had remained relatively constant over the last 12 months. The team noted
that there were no old open items that were safety significant.
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c.
Conclusions
1
The team concluded the licensee was effectively managing the backlog of
engineering work.
E2.4
10 CFR 50.59 Imolementation (37001)
E2.4.1 10 CFR 50.59 Program
a.
Insoection Scooe
l
The team reviewed the licensee's 10 CFR 50.59 program guidance, 20 screenings
that concluded that a safety evaluation was not required, and 16 safety evaluations.
The screenings and safety evaluations were associated with permanent and
temporary rnodifications to the plant and procedures, request for resolutions, and
Final Safety Analysis Report change notices.
1
b.
Observations and Findinas
b.1
Administrative Reauirements
The licensee's safety evaluation process for changes to the facility was controlled
by Procedure APA-ZZ-00140, " Safety, Environmental and Other Licensing
Evaluations." This procedure delineated the methods and responsibilities to
determine and document whether procedure and facility changes could be made
without prior NRC approval. The licensee's safety evaluation process began with a
safety evaluation screening that utilized specific screening criteria. This screening
was performed to determine whether the proposed activity needed additional review
to determine if an unreviewed safety question existed.
Procedure APA-ZZ-00140 provided this screening criteria in the form of questions
that were answered by a reviewer. Specifically, these questions were: (1) the
activity did not change the facility or a procedure as described in the Final Safety
Analysis Report; (2) the activity was not a test or experiment not described in the
Final Safety Analysis Report; and (3) the activity did not involve a change to the
Technical Specifications. If the results of this screening concluded that one or more
of the screening criteria was not satisfied, the process then required that a formal
safety evaluation be performed to assess the merits of the activity and to determine
whether an unreviewed safety question existed. The unreviewed safety question
determination was documented in this safety evaluation. If it was determined that
an unreviewed safety question existed, then NRC approval was required prior to
implernenting the proposed change.
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The team found the documentation contained in the safety evaluations that
were performed to be sufficiently detailed and the conclusions logically
supported. While the team determined that overall, the guidance contained
in Procedure APA-ZZ-00140 was adequate, the team identified problems with this
guidance and with the implementation of the 10 CFR 50.59 process.
b.2 Reoortina of Plant Chances
During a review of Procedure APA-ZZ-00140, the team noted that the procedure
stated that short-term modifications (e.g., temporary modifications) did not fall
within the periodic reporting requirements. A review by the team of the
l
10 CFR 50.59 safety evaluation reports submitted to the NRC, confirmed that the
safety evaluations for temporary modifications were not reported. As the result of
discussions with licensee personnel, the team determined that the licensee had not
reported these safety evaluations since June 14,1988. The licensee further stated
that the decision to not report safety evaluations for the temporary modifications
was an error. The team identified that Temporary Modification 95-MOO 2,
" Temporary Filter and Piping for BTRS Chill Water Loop," was an example of a
temporary modification that was not reported.
10 CFR 50.59(b)(2) requires licensees to submit a report containing a brief
description of any changes, tests, and experiments, including a summary of the
safety evaluation of each. This report was required to be submitted annually or
along with the Final Safety Analysis Report updates. Since 10 CFR 50.59(b)(2)
does not differentiate between long- and short-term modifications, all safety
i
evaluations for modifications are required to be reported. The f ailure to report
temporary modification safety evaluations is considered to be a violation of
10 CFR 50.59 (50-483/9705-04).
b.3 Performance of Safety Evaluations
During the review of documentation involving screening to deterr.iine if safety
evaluations were required, the team identified the following two examples where
safety evaluations were not performed:
(1)
The team noted that Procedure APA-ZZ-00140 stated that, if the design,
function, or method of performing the function of an associated system,
structure, or component was either unaffected or improved, there was no
change in the f acility as described in the Final Safety Analysis Report. Since
10 CFR 50.59 requires a safety evaluation for all changes to the facility
irrespective of whether or not the change is believed to be an improvement,
the team considered this procedure guidance to be inconsistent with the
10 CFR 50.59 requirement.
The team identified one instance where this guidance was used as
justification for not performing a safety evaluation. Modification
CMP 95-1027A upgraded the source of power for instrument cabinet fans
from nonsafety related to safety related. The licensee documented that this
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modification was an improvement; therefore, a safety evaluation was not
performed. However, a subsequent review of this modification by the team
determined that the modification did not change the racility as described in
the Final Safety Analysis Report. Therefore, a safety evaluation was not
required. Nevertheless, the team considered the guidance in Procedure APA-
ZZ-00140, regarding improvements, to be misleading.
>
(2)
Modification RMP 94-2005A was implemented to redesign the post-accident
sampling system by replacing the computer control of the sample panel with
j
manual control. Before the modification, the system was manually initiated
'
by selecting a particular analysis. The computer would then automatically
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position the valves, as necessary, to accomplish the analysis. Due to
problems with softwear, the computer controls did not work properly. After
the modification, the computer was removed and the necessary lineups to
perform specific analyses were performed by personnel following specific
analysis procedures.
The safety evaluation screening, performed on November 1,1995,
concluded that a safety evaluation was not required. The conclusion stated
.
that Final Safety Analysis Report Change Notice 94-05, issued in February
i
1994, and Change Notice 94-23, issued in July 1994, were already
approved and incorporated in the Final Safety Analysis Report and that the
modification involved nonsafety-related equipment. Therefore, the
modification did not require any additional Final Safety Analysis Report
changes.
The team noted that Change Notices 94-05 and 94-23 involved a change to
the chemical analyses that were being performed by the post-accident
sampling system. The analyses for atmosphereic oxygen, dissolved oxygen,
pH, conductivity and in-line chloride analysis were eliminated. However, the
'
team also noted that the change from a computer controlled analysis to a
manually controlled analysis was not identified in these change notices.
When this finding was discussed with the licensee, the licensee stated that
the change was not included because the description regarding the computer
controlled operation of the post-accident sampling system was only
described in a letter to the NRC, dated November 4,1983. Since this letter
was only referenced by the Final Safety Analysis Report, they did not
consider the operation described in this referenced letter to be a part of the
Final Safety Analysis Report.
The team did not agree with the licensee's position. The February 4,1983,
letter referenced by Section 18.2.3 of the Final Safety Analysis Report
provided additional details on the post-accident sampling system operation.
This information was used by the NRC to determine acceptability of the
design of the post-accident sampling system. Specifically, the information
contained in this letter was referenced in Supplement 3 to the NRC's Safety
'
Evaluation Report. Therefore, the inspectors concluded that the method of
operation of the system, as described in the Final Safety Analysis Report,
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was changed by the modification, and a safety evaluation was not
performed.
10 CFR 50.59(b)(1) requires the performance of a safety evaluation when
plant modifications change the plant as described in the Final Safety Analysis
Report. The failure to perform this safety evaluation is considered to be an
apparent violation of 10 CFR 50.59 (50-483/9705-05).
c.
Conclusions
While the team concluded that the licensees procedural guidance for implementation
of 10 CFR 50.59 was adequate, the team identified two areas where this guidance
,
was not in accordance with 10 CFR 50.59. These involved the failure to report
safety evaluations for temporary modifications and not performing a safety
,
evaluation when a change was considered to be a plant improvement. The team
also concluded that the implementation of the 10 CFR 50.59 requirements was
adequate; however, the team 'dentifica an apparent violation involving the f ailure to
3
perform a safety evaluation for the change in the automatic operation of the post-
J
accident sampling system.
E2.4.2 Technical Soecification Interpretations
a.
insoection Scone
,
As a result of a concern identified in October 1996 at another facility regarding
Technical Specification interpretations that were found to be in conflict with
Technical Specification requirements, the team reviewed 41 of the licensee's
Technical Specification interpretations to ensure that these interpretations did not
conflict with Technical Specification requir9ments.
b.
Observations and Findinas
The tec.m fourid that the Technical Specification Interpretation Program was
controlled by Procedure APA-ZZ-00104, " Technical Specification Interpretations and
Notice of Enfcrcement Discretion." Procedure APA-ZZ-00104 defined a Technical
Specification interpretation as a formalinterpretation that provided guidance for
both the Technical Specifications and Section 16 of the Final Safety Analysis
Report. This procedure also specified that all Technical Specification interpretations
were reviewed by the onsite review committee and approved by the plant manager.
The licensee informed the team that, as a result of the concern identified in October
1996, the onsite review committee performed an additional review of the Technical
Specification interpretations to ensure that the interpretations did not conflict wRh
Technical Specification requirements. Nevertheless, the team identified an example
in which an interpretation provided guidance that potentially violated the
requirements of the Technical Specifications, in addition, the team also identified
an example in which an interpretation p;ovided guidance that was inconsistent with
the Final Safety Analysis Report.
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Technical Soecification Interpretation 25
Section 16.9.2 of the Final Safety Analysis Report described the limits for
setting the overload and load reduction trip setpoints for the refueling machine-
at 250 pounds above and below the weight of the suspended loads, respectively.
Technical Specification Interpretation 25 interpreted Section 16.9.2 to mean that
4
these trip setpoints could be set to 250 pounds above the heaviest fuel assembly
'
load for the overload trip and 250 pounds below the lightest fuel assembly load for
the load reduction trip.
The team was concerned that this interpretation allowed these trip setpoints to be
set in excess of 250 pounds by approximately 150 pounds (the estimated weight of
a rodded assembly). This meant that, while the overload trip would be correct for a
rodded fuel assembly, it would be excessive for the unrodded fuel assembly and
would not trip until the weight of the suspended load was 400 pounds above the
suspended load weight. It also meant that, while the load reduction trip would be
correct for the unrodded fuel assembly, it would be excessive for the rodded fuel
assembly and would not occur until the insertion force was 400 pounds less than
the suspended load weight.
Through discussions with licensee personnel, the team determined that these were
the trip setpoints used during Refueling Outage 8. Additional review by the team
indicated that on October 20,1995, Technical Specification 3.9.6, which provided
the same trip setpoint setting requirements, was deleted and the requirements
incorporated into Final Safety Analysis Report 16.9.2. Therefore, during the period
of October 18,1984, through October 20,1995, Technical Specification 3.9.6 was
violated during seven refueling outages (Refueling Outages 1 through 7).
)
10 CFR 50.59 (b)(1) requires the performance of a safety evaluation when plant
modifications change the plant as described in the Final Safety Analysis Report.
The failure to perform this safety evaluation is considered to be an apparent
violation of 10 CFR 50.59 (50-483/9705-06).
b.2 Technical Soecification Interpretation 18
Technical Specification Interpretation 18 provided an interpretation regarding
the operation of the diesel generator building supply fans. Technical Specification 3.8.1.1b required the diesel generators to be operable. In
addition, Technical Specification 1.19 required that for a system to be operable,
all supporting subsystems must also be operable. The diesel generator building
supply fans are a subsystem of the diesel generators that the licensee determined
are required to be operable when outside ambient temperature is greater than 65 F.
Final Safety Analysis Report, Section 9.4.7.2.3, stated that the diesel generator
building supply fans automatically start when the room temperature exceeds 90 F
and automatically shut down when room temperature falls below 86 F. If the
building temperature exceeded 90oF, Final Safety Analysis Report, Section 16.7.4,
allowed the temperature to rise to a maximum of 119 F. At 119 F, Section 16.7.4
required the temperature to be lowered below 119oF within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or to perform
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an analysis to demonstrate that equipment was not affected by the elevated
temperatures. In addition, Section 16.7.4 also required that if the temperature
'
exceeded the 119 F limit by 30 F (149 F) for more than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the affected
equipment (i.e., the diesel generators) were to be considered inoperable.
4
i
The licensee developed Technical Specification Interpretation 18 to allow the
diesel generator building supply fans to be placed in manual operation (i.e.,
,
defeating the automatic starting function by placing the supply fan control switches
in a " pull-to-lock" position), without declaring the diesel generators inoperable when
the outside ambient temperature was greater than 65 F. The purpose for this
interpretation was to eliminate excessive cooling of the diesel generator jacket
!
water system that was causing system alarms. In addition. the licensee revised
Procedure OTN-NE-00002, " Standby Diesel Generator Auxiliary Systems," to add a
precaution and limitation (Step 2.6), which stated that each diesel generator
building supply fan was considered capable of performing its intended safety
function (i.e., the ability to supply air to the building if temperatures rise
above the fan start setpoint) if the fan was placed in pull-to-lock and was under
'
,
the control of the operator. The procedure also directed the operator to assign
!
the room temperature point to annunciate on Window 65F, " Optional Parameter
Setpoint," at or below 110 F. However, the licensee concluded, through the
10 CFR 50.59 screening for the procedure change, that a safety evaluation for the
change was not required. Therefore, a safety evaluation for the procedure change
was not performed.
Since the diesel generator building supoly fans were a subsystem of the diesel
generators and these fans were considered to be inoperable when they were in a
" pull to-lock" condition, the team concluded that the diesel generators were also
inoperable. Therefore, in effect, Technical Specification Interpretation 18 changed
the diesel generator Technical Specifications by allowing the diesel generators to be
,
declared operable while the diesel generator buildiqq supply fans were inoperable.
Based on this finding, the team requested that the licensee review the operations
.
logs to determine when the Technical Specification interpretation / procedure
guidance was implemented. Although licensee representatives stated that the
interpretation was used to place the fans in manual operation during the Fall and
Spring evenings from 1987 to 1990, the licensee's review of the operator logs from
1987 to 1990, revealed that there were no documented instances in the operator
logs in which the f ans were placed in pull-to-lock. In addition, since a plant
modification in 1990 eliminated the need to place the fans in manual operation, che
licensee interviewed five operators to determine if they recalled any instances of
placing the fans in manual operation since 1990. As a result of these interviews,
the licensee informed the team that no operator interviewed recalled placing the
f ans in manual operation.
Since the system was described in the Final Safety Analysis Report as operating
automatically, it appeared that a safety evaluation was required to substitute the
manual operator action for this automatic function. This substitution may have
increased the probability of occurrence of a malfunction of equipment importam to
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safety previously evaluated in the Final Safety Analysis Report and increased the
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possibility for a malfunction of a different type than any evaluated previously in the
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Final Safety Analysis Report, and therefore, potentially constitutes an unreviewed
safety question.
l
10 CFR 50.59 (b)(1) requires tne performance of a safety evaluation when plant
'
modifications change the plant as described in the Final Safety Analysis Report.
i
The failure to perform this safety evaluation is considered to be an apparent
violation of 10 CFR 50.59 (50-483/9705-07),
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c.
Conclusions
)
Overall, the team concluded that the implementation of the Technical Specification
'
interpretation. program was adequate. However,' the team identified two examples
in which interpretations provided guidance that was inconsistent with the Final
Safety Analysis Report and potentially violated the requirements of the Technical
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Specifications. One example involved an apparent violation for the failure to
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perform a safety evaluation for a change to the refueling machine load setpoints.
!
The second example involved an apparent violation for the substitution of manual
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operator action for the automatic operation of the diesel generator building supply
fans.
E2.5 Svetem Walkdowns
a
a.
Insoection Scope
At different times during the inspection, the team performed walkdowns of selected.
plant areas to determine the overall material condition of equipment and the
maintenance of housekeeping.
b.
Observations and Findinos
During the walkdowns, the team noted a number of tags on pumps and valves that
were called " boric acid" tags. The licensee stated that they started a boron control
program in January 1997. The purpose of the program was to hang such tags on
components to identify that there was some leakage that required occasional
cleaning, but the affected valves were not in need of immediate repair. The
licensee stated that the purpose of the program was to differentia'.e between
acceptable periodic residue removal on components and those ccmpoaents that
needed to be repaired. For components that needed repair, work request tags were
hung. The team noted that these boric acid tags were not limited to boric acid
problems and extended to other leakage problems as well. This explained the
existence of such tags on such systems as the component cooling water system,
which was not a borated water system. Based on these observations, the team
considered the boron control program to be innovative and effective toward
maintaining the material condition of the plant.
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Conclusions
The team's walkdown of the plant indicated that the material condition of the plant
was good and that some improvements were noted. Housekeeping was also noted
to be good. The boron control program was considered to be effective in
maintaining the plant's material condition.
E4
Engineering Staff Knowledge and Performance (37550)
a.
inspection Scope
The team interviewed the nuc' ear engineering department manager, one
engineering supervisor, five system engineers, and one design engineer. Interview
topics included management expectations for staff engineers, training regarding
system interrelations, and interface with other plant organizations. The team also
questioned staff engineers about knowledge of their assigned systems and
conducted system walkdowns with the system engineers. In addition, a detailed
walkdown of the component cooling water system was conducted with the
associated system engineer to determine the level of knowledge of this engineer.
,
b.
Observations and Findinas
The team found management expectations for staff engineers were clearly defined.
System engineers were effectively coordinating with design engineering to evaluate
and improve their specific systems. System engineering personnel indicated that
communication and cooperation with operations, maintenance, and design engineers
were effective in resolving work issues and also in assuring that modifications were
properly installed. Team interviews and plant walkdowns indicated that system
engineers were knowledgeable of their assigned systems.
The detailed walkdown of the component cooling water system with the system
engineer indicated that this system engineer spent approximately 25 percent of the
time in the plant performing system walkdowns, witnessing surveillance tests, and
witnessing maintenance activities. The system engineer trended flow versus
differential pressure for the component cooling water and essential service water
pumps to detect pump degradation. The system engineer indicated that their
identification of a degrading trend in one of the essential service water pumps will
result in the replacement of this pump during the next refueling outage. The team
determined that the system engineer was also trending for fouling in the heat
exchangers. This engineer also explained component deficiencies in detail and
discussed specific problems with system operation. This walkdown further
confirmed that system engineers were knowledgeable of their systems.
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c.
Conclusions
The team concluded that engineering management expectations were clear and
understood by enginee-ing personnel. Communications between system
engineering and other p ant departments were affective. System engineers were
knowledgeable of their assigned systems.
E5
Engineering Staff Training and Qualification (37550)
a.
Inspection Scope
.
The team reviewed the licensee's training and certification program requirements for
the engineering staff. This review included a review of training records fo-
-9
engineering staff.
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b.~
Observations and Findinas
The team found that all 35 design engineers had completed their qualification
j
modules to ensure that they possessed sufficient knowledge and skills to
{
independently perform their assigned tasks. The team found that 16 out of
23 system engineers completed their qualification modules. The team noted
that engineering management's goal was to have all system engineers qualified by
l
June 1998. The team's review of the training records and schedules indicated that
engir eering management was on track for completing the system engineer
qualifications.
c.
Conclusions
The team concluded that the training program for the engineering staff was
effectively supporting the role of the system and design engineers.
E6
Engineering Organization and Administration (37550)
a.
insoection Scoce
The team evaluated the overall effectiveness of the independent safety engineering
group by reviewing selected reports, interviewing independent safety engineering
group personnei, and by determining if issues identified by the independent safety
l
engineering group were corrected or in the process of being corrected.
b.
Observations and Findinas
in addition to their Technical Specification required activities, the independent safety
engineering group provided an independent review of plant operating activities to
]
detect potential operational problems. To accomplish this operational experience
mission, the group performed operating experience reviews and analyzed industry
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events for applicability to the Callawev Plant. The indecendent safety engineering
group also reviewed and analyzed internal plant events to determine the corrective
actions needed to prevent recurrence.
The independent safety engineering group consisted of eight engineers and a
supervising engineer. Of these eight engineers., six were also qualified shif t
technical advisors. Each week one of these six shift technical advisors stood a day
shift watch in the control room. The team considered the practice of having an
independent safety engineering group engineer performing shift technical advisor
duties to be an effective method for the independent safety engineering group to
keep abreast of ongoing shift operations.
As the result of their operating experience reviews, the independent safety
engineering group issued periodic, " Operating Experience Journals," to provide
information to plant personnel regarding events that have occurred at the Callaway
Plant and in the industry. The independent safety engineering group also performed
operating experience crew briefings to assure that plant personnel were aware of
operating experience issues. The team reviewed four operating experience journals
and six operating experience crew briefing sheets. As the result of these reviews,
the team considered these reports to be informative and effective at keeping plant
personnel informed of operating experience events.
As required by the Technical Specifications, the independent safety engineering
group issued monthly reports to the quality assurance manager and the plant
manager. The team reviewed the independent safety engineering group monthly
reports for the period of August 29,1996, through February 3,1997. The team
found these reports to meet the requirements of the Technical Specifications and to
be indicative that the independent safety engineering group provided independent
assessments of on-going plant activities.
The team also reviewed a listing of suggestion-occurrence-solution reports written
by independent safety engineering group engineers to determine the group's
involvement in identifying plant problems. The licensee's suggestion-occurrence-
solution reporting system is used to p ovide documentation of plant problems into
their cotrective action system and to assure that these problems are tracked for
resolution. The team was informed tnat the independent safety engineering group
wrote approxirnately 10 percent of all the suggestion-occurrence-solution reports
generated. The team confirmed this approximation. There were 229
suggestion-occurrence-solution reports written by the independent safety
engineering group engineers for the period of January 1996 through February 1997.
Of these 229 reports, approximately 87 were open, but not overdue, and were
being processed for closure. Five were Open and overdue. The team reviewed the
five overdue reports and an additional six other reports that were still open and did
not identify any problems. This review indicated that the independent safety
engineering group was identifying plant issues and that these issues were being
tracked to completion.
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The team also reviewed the licensee's listing of items entered in their centralized
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action tracking system by independent safe':y engineering group personnel to
determine the extent of the group's involverient with the system. The centralized
action tracking system was used to track istues identified in the industry. The team
noted that a total of 64 items were written t y the independent safety engineering
4
group over the past year. Of these items,311 were closed and 26 were still open.
]
Of the 26 open items,14 were overdue for closure. Thase 14 items were reviewed
with licensee personnel. Based on this review, the team determined that they were
1
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properly classified as low priority and were bring tracked for resolution,
i
The team's interviews of four independent saiety engineering group engineers and
the group's supervising engineer did not identify any issues. All personnel
interviewed indicated that they were qualified to perform their assigned activities
and that they were proactive toward resolving plant issues. In addition, they felt
'
that they had a low threshold for writing suggestion occurrence-solution reports and
that plant personnel were responsive to their fi1 dings,
i
c.
Conclusions
The team concluded that the independent safety engineering group was effective in
providing an independent assessment of plant o?erations, providing an independent
assessment of the effect of internal and externa; events on plant operations, and in
providing recommendations to improve plant safety. The team considered the use
of an independent safety engineering group engineer as a shift technical advisor to
be an effective and notable application of the group's experience base.
4
E7
Quality Assurance in Engineering Activities (37550)
,
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a.
Inspection Scooe
i
The team reviewed six quality assurance audit reports and eight quality assurance
surveillance reports of plant engineering that were performed from December 1994
a
through August 1996. These reports were reviewed to evaluate the effectiveness
of the licensee's process to self identify and resolve plant problems.
b.
Observations and Findinas
The team found that the licensee considers their self assessments to be the audits
and surveillances performed by the quality assurarle department. These audits and
surveillances were conducted at the request of the nuclear engineering department.
The team noted that 27 self assessments of engineering activities were performed
during the period of 1994 through 1996. The team found that the quality
assurance audit and surveillance findings were generally consistent with those
identified by the tearn. Specifically, the team noted that system engineers were
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knowledgeable and qualified in their systems and were effective in identifying plant
problems. However, the team also found that self-assessment findings were not
consistent with the team's findings in that the self-assessments did not identify the
failure to perform 10 CFR 50.59 safety evaluations as discussed in Section E2.4 of
this report.
The team found that the responses to the quality assurance audits and surveillances
were timely and acceptable. An example of this responsiveness was evidenced by
Surveillance Report SP96103. The team reviewed the discrepancies identified in
,
this report and found that they were properly evaluated and that necessary changes
to the Final Safety Analysis Report and/or plant procedures were being
implemented.
c.
Conclusions
The team concluded that the self assessmei.ts of engineering activities were
generally consistent with the team's findings except in the area the 10 CFR 50.59
safety evaluations. The team concluded that licensee responses to identified
discrepancies were timely and acceptable.
V. Manaaement Meetinas
X1
Exit Meeting Summary
The team presented the inspection results to members of licensee management at
the conclusion of the inspection on February 28,1997. In addition, a final exit
meeting was held on June 24,1997. The licensee acknowledged the findings
presented. During both meetings, the licensee stated the following objections with
regard to the inspection findings:
The licensee disagreed with the violations for a failure to report
the inoperability of main steam safety valves as discussed in Section
E1.3.b.1. The licensee's position was that they comply with the
requirements of 10 CFR 50.73 and the guidance provided in NUREG 1022,
Revision O. They also do not believe that they need to comply or are
committed to guidance prouded to other licensees. (This statement was
made in reference to a lett.sr dated December 8,1993, that was sent to the
existing Region IV power reactor licensees by Samuel J. Collins regarding the
interpretation of reporting requirements for setpoint drifting of main steam
and pressurizer safety valves.)
The licensee did not agree that the guidance in Procedure APA-ZZ-00140,
which states that a safety evaluation was not required for plant
" improvements," was misleading as discussed in Section E2.4.1.b.3.(1).
The licensee's position was that the statement would not result in a plant
change that affected the design, function, or method of performing the
function in an associated system.
21
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4
The inspectors asked the licensee whether any materials examined during tlie
inspection should be considered propriotary. The licensee identified some
information reviewed by the team that was considered to be propriety. The team
was aware of this information, which involved maximum flows through component
cooling water heat exchangers, and stated that this information had no bearing on
inspection results and would not be discussed in the report.
22
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.
ATTACHMENT
SUPPLEMENTAL INFORMATION -
PARTIAL LIST OF PERSONS CONTACTED
Licensee
R. Affolter, Plant Manager
D. Bono, Supervising Engineer, Site Licensing
B. Hampton, System Engineer
D. Hollabaugh, Supervising Engineer, Technical Support
G. Hughes, Supervising Engineer, independent Safety Engineering Group
L. Kanuckel, Supervisor Civil Design Group
K. Kuechenmeister, Superintendent, Design Engineering
J. Laux, Manager, Quality Assurance
J. McGraw, Superintendent, Technical Support Engineering
C. Naslund, Manager, Nuclear Engineering
A. Passwater, Manager, Licensing and Fuels
C. Pilkington, Outage Supervisor
G. Randolph, Vice President, Nuclear
M. Reidmeyer, Engineer, independent Safety Engineering Group
R. Rice, Design Engineer
T. Sharkey, Supervising Engineer, NESM
W. Witt, Superintendent, Systems Engineering
NRC
D. Passehl, Senior Resident inspector
LIST OF INSPECTION PROCEDURES USED
iP 37001
10 CFR 50.59 Safety Evaluation Program
Engineering
LIST OF ITEMS OPENED AND CLOSED
Ooened
50-483/9705-01
Failure to report the inoperability of main steam safety valves
as required by 10 CFR 50.73 (Section E1.3.b.1).
50-483/9705-02
IFl
Review the final analysis for the f ailed pressurizer safety valve
in Refueling Outage 8 to assure that the accident analysis is
still valid (Section E1.3.b.2).
1
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. _ . . - _ _ _
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50-483/9705 03
IFl
Review the licensee's efforts to determine the extent of Final
Safety Analysis Report discrepancies and the adequacy of
corrective actions (Section E2.1b).
50-483/9705-04
Failure to report the 10 CFR 50.59 safety evaluations
performed for temporary modifications as required by
10 CFR 50.59(b)(2) (Section E2.4.1.b.2).
50-483/9705-05
APV
Failure to perform a 10 CFR 50.59 safety evaluation for
changing the method of operation of the post-accident
sampling system (Section E2.4.1.b.3(2))
)
50-483/9705-06
APV
Failure to perform a 10 CFR 50.59 safety evaluation for
changing the setpoints on the refueling machine (Section
E 2.4. 2.b.1 ).
50-483/9705-07
APV
Failure to perform a 10 CFR 50.59 safety evaluation for the
substitution of manual operator action for the automatic
function for the diesel generator building supply fans. Based
on the increase in probability of failure, this change was an
unreviewed safety question as defined in 10 CFR 50.59
(Section E2.4.2.b.2).
LIST OF DOCUMENTS REVIEWED
Plant Procedures
Procedure
Revision
Title
APA-ZZ-00007
11
Quality Assurance Organization, Responsibility and
Conduct of Operations
JDP-ZZ 04100
8
Operating Experience Review Procedure
APA-ZZ-00107
3
Review of Current Industry Operating Experience
JDP-ZZ-04400
2
Callaway Plant Event Reduction Program
JDP-22-01100
5
ISEG Tracking Log -
JDP-ZZ-02000
5
STA Personnel Qualification and Training
JDP-ZZ-03C00
6
ISEG Engineer Control Room Watch
JDP-ZZ-04000
2
Document Reviews
l
JDP-ZZ-04200
4
Callaway Operating Experience input to Nuclear
Network
JDP-ZZ-04300
3
Review of Nuclear Safety Review Board Material
!
APA-ZZ-00140
20
Safety, Environmental and Other Licensing Evaluations
,
!
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Procedure
Revision
Title
1
APA-ZZ-00104
7
Technical Specification Interpretations and Notice of
APA-ZZ-304
11
Control of Callaway Equipment Lists
1
APA-ZZ-500
27
Corrective Action Program
APA-ZZ-604
16
Requests for Resolution
,
1
APA-ZZ-605
7
Temporary System Modifications
APA-ZZ-600
15
Design Change Control
APA-ZZ-325
4
Initiating, Authorizing, and Removing Condition
Reports
EDP-ZZ-4023
13
Calculations
EDP-ZZ-4055
2
Design Bases Control
MSM-BB-QV001
16
Pressure Safety Valve Testing
MSM-AB-OV001
10
Main Steam Safety Valve Set Pressure Test
'
OPS-EG-V001 A
17
CCW Train a Section XI Valve Surveillance
OPS-EG-V001 B
13
CCW Train B Section XI Valve Surveillance
TDP-ZZ-0065
0
Training and Qualification of Engineer Support
Personnel
Plant . Modifications
Modification
Title
FiFR 16981 A
Approved Replacement Gages and Transmitters
RFR 17402A
Change Heat Exchanger End Cover Gasket Material
MP 93-1055
Modification to Hangers EJ01-R502/134 and EJ02-R504/133
MP 93-1058
Move Close Torque Switch Bypass to Rotor 3 on EGHV62
MP 96-1003
Modification to Provide a Level Instrument on the Turbine Exhaust Line
of Turbine Driver (KFCO2) Associated With the TDAFP
MP 96-1014
Installs An isolation Valve In The B Train ESW To AFP Suction Line
Temocrary Modifications
Modification
Title
TM-960E010
This TM removes the SR power supply source from differential pressure
gauges GKPDIS50028, 39,100 and 103 on SGK04A, SGK04B,
SGK05A, and SKG05B, respectively.
3
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Suaaestion-Occurrence-Solution Reports
SOS
Title
96-0926
EO discovered NB01 room warmer than NB02 room. He found the A/C
unit would start and valve GKV0767 opened but would go immediately
shut.
96-0201
During the performance of OSP-EC V001 A, EC-HV-0011 was timed
stroked closed, but not open as required.
96-0646
Fuel and/or lube oil was discovered by the NRC in the 'B' diesel firepump
sump.
96-0815
During performance of OSP-SA-0017A, there were complaints that the
resultant lineup caused excessive HVAC pressure in the fuel builoing.
96-0178
Letdown HX outlet temp control valve (BGTV0130) experienced problems
again at controlling letdown temperature.
96-1621
During restoration of some WPA on the heater drain pump, a tag was in
place but the valve was out of the tagged position,
j
96-1775
Due to high vibration on the "B" RCP, the pump was secured and the
_
reactor shutdown per TS 3.4.1.1.
96-1945
While attempting to isolate ESW on WPA 21514, it was discovered that
the pointer for EFV0275 was 180 degrees out of position.
96-1801
Significant oil leak noted on the "A" MFP.
]
96-1981
Resolution of a level 4 violation
95-1952
A single active failure of the CCW train could result in loss of cooling to
the CCP miniflow
95-0230
Suspect valve seat leakby
95-1428-
Unexpected levelincrease noted in B CCW surge tank
96-0355
Review Velan valves for their applicability to OE 7640
95-2140
Equipment required for remote shutdown is not being tested on a periodic
basis
95-1593
CCW valve must be positioned open or throttled to prevent CCW low
flow to RCP motor cooler
95 2297
CCW valves have not been verified to be in their correct position
95-2094
Only one CCW penetration has an automatic isolation valve
95-0787
Pin hole leak discovered in weld upstream of CCW valve
96-1380
Pin hole leak in vent pipe of CCW heat exchanger
4
.
.
SOS
Title
96-1795
CCW system temperature below Final Safety Analysis Report minimum
95-2065
Several valves credited for operating following a single failure have been
designated as passive
,
95-0860
A potential operability concern with EGD02A and the A train Diesel
]
Generator
'
95-1792
Valves did not fully stroke per their indicators
96-1263
During testing a PSV exceeded its Technical Specification tolerance
96-1247
During testing a MSSV exceeded its Technical Specification tolerance
l
90-2908
During testing a PSV exceeded its Technical Specification tolerance
95-0508
During testing 14 MSSVs exceeded their Technical Specification
tolerance
90-1682
During testing 3 MSSVs exceeded their Technical Specification tolerance
l
95-0013
Failed surveillance of the CCW surge tank level transmitter
95-2219
Cases were identified reviewing CCW flow verification tests that did not
meet the testing requirements
95 2105
Resolve Callaway's definition of safe shutdown
95-2126
Discrepancies noted during review on auxiliary building flood calculations
95-1852
No CEL ECN was initiated when a hanger was rernoved
94-0775
During review it was found that some valves listed as active in the Final
j
Safety Analysis Report were listed as passive in the equipment list
i
i
94-1047
RFRs had been generated without correction of errors in the equipment
list or Final Safety Analysis Report
94-0702
The valve nozzle and guide ring settings could not be verified
96-0877
A separation violation exist in the vendor terminal box for GKPDIS0103
I
97-0255
Work done on Diesel Generator A consisted of replacing the cam cover
gasket on the left bank with new gaskets made of Bluegard 3000 which
leaked oil
j
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,,
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Reauests for Resolution
!
RFR
Title -
'
176726
Establishing range of design flows for CCW components
16449
Revise valve drawing and vendor manual to' include size of vent plug hole
l
16458
Evaluate seal water heat exchanger capacity.
16528
Evaluate safety function of excess letdown heat exchanger
16444
Operability of CCW pumps with pump room coolers inoperable
16457
Update CEL' Q-list reason fields for various components
16448
Raise CCW surge tank low level alarm setpoint
15246
Evaluate differences between CEL and Final Safety Analysis Report
Table 3.9(B)-16
17572
Update drawings for storage items
15489A
To provide positive indication of the status of the Turbine Driven Auxiliary,
17206A
Modification request to rewire GKPDIS0028,39,100 and 103
.
Calculation .
Revision
Title
-
J
EG-32
0
Calculation Determines Volume Contained in the CCW
l
Surge Tanks Versus Fluid Level
j
EG-34
0
Upper Recommended Flowrate Determination for
Components Using CCW
Work Reauests
Title
W1b3699
Change EG-LSHL-000224 setpoint -
G525907-181
Install and remove glove bag
G561893
Generic WR to perform troubleshooting
W163006
Replace lower wedge in valve internals
W169645
Remove PSA snubber and replace
W172312
Actuator air supply regulator is plugged
W173733
Replace vent cap
W174151
Adjust open limit switch on EGHV0016
W177243
Valve handwheel required to be locked in place
6
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Work Recuests Title
W178013
Perform ISI VT-3 exam on EF02-C003
W178273
Retorque bolting due to chemical leakage
W531533
Replace relays
W531535
Replace relays
W531564
Replace relays
j
W531568
Replace relay due to outgassing concern
W575857
Install DP gage and take readings
G579598
Generic work request for cleaning boric acid leaks
W174735
Drill hole in versa valve vent plug to 11/32 in, diameter
10 CFR 50.59 Screenina and Safety Evaluations For The Followina Documents:
Plant
Title
Modifications
RMP 94-2005A
Redesign of PASS System (screening)
CMP 95-1027A
Rewire Cabinet Cooling Fans to Receive Safety Grade Power
(screening)
CMP 95-1019
Change Cable Size for EJHV8701B (screening)
CMP 96-1008
Add Local Control Stations to S/G PORVs "B" & "C" (safety
evaluation)
CMP 95-1004
Modify SSPS Power Supply (safety evaluation)
CMP 95-1007
Change RHR Miniflow Valves to Limit Close (safety evaluation)
CMP 95-2015
Remove Flow Switches from Control Logic (safety evaluation)
Temocrary Modifications
_N_o ,.
Title
q
TM 95-E0006
Install Recorder to Monitor Battery Charger NK22 (screening)
TM 96-E0013
Add Manual Switch to Control Speed of Refueling Machine
(screening)
TM 94-M005
Provide Lube Water for the Cirewater Pump (screening)
TM 95-M019
Install Temporary Lube Oil Fill Line for RCP-C (screening)
7
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TM 96-E0005
Jumpering of Temperature Switches for UHS Sump Heaters
,
(screening)
TM 96-M004
Install Splash Shield on "B" CCP (screening)
i
TM 95-E0006
Remove Failed Signal from COPS (safety evaluation)
,
TM 95-M002
Temporary Filter and Piping for BTRS Chill Water Loop (safety
evaluation)
TM 96-M014
Installation of Blank Flanges on Cooler SGN01 A (safety evaluation)
New Procedures and Procedure Chances
Number
Title
ETP-AE-STOO8
AEFV0042 Repair and Retest Procedure (safety evaluation)
ETP-RJ-ST001
Test of Rod Drop Software (safety evaluation)
ETP-BG-ST015
Letdown Heat Exchanger Flow Test (safety evaluation)
ETP-MB-ST001
Main Generator Excitation Stability Test (safety evaluation)
ETP-ZZ-ST019
Plant Radio Testing (cafety evaluation)
ETP-ZZ-ST006
Bank Reactivity Worth Measurement (Rod Swap) (safety evaluation)
OTN-NE-0001 A
Standby Diesel Generator System - Train "A" (screening)
OTN-NE-00002
Standby Diesel Generator Auxiliary Systems (screening)
MSM-KJ-QT001
10 Year Emergency Diesel Generator Fuel Oil Storage Tank Cleaning
(screening)
Reauests for Resolution
RFR
Title
16337 A & B
Sediment in Diesel Generator Fuel Oil Storage Tank (safety evaluation)
17402 A
Change HX End Cover Gasket Material (screening)
13573B
Modify PORV Leakoff Line (screening)
13877 A
Cavity Return Temperature Alarm Setpoint (safety evaluation)
16805 A
Increase DP Capabilities of Valves (screening)
16981 A
Approve Replacement Gauges and Transmitters (screening)
14464 B
Pressurizer Safety Valve Drawing Changes (screening)
16311 B
Evaluate Containment Cooler Motor Acceptability (screening)
8
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. _ _ _ _ . . _ . . _ _ .
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Final Safety Analysis Report Chanae Notices
l
. _C N
Title
I
95-046
Correct Final Safety Analysis Report Descriptions of MStV Status on Loss-
of-Electrical Power (screening)
Remove Final Safety Analysis Report Table 3.11(8)-3 from Final Safety
,
Analysis Report - Put in CEL (screening)
. Delete Note on Figure 6.2.4-1 in Final Safety Analysis Report for Type A
CIV Test (safety evaluation)
i-
Technical Specification Interoretations
i
Rev.
TS/ Final Safety Analysis Report Section
'
1
1
TS 3/4.5.1
i
4
1
Final Safety Analysis Report 16.3.4
7
O
TS 3/4.3.1
.
11
4
TS 1.2, TS 3/4.9
_
Final Safety Analysis Report 16.0-2
13
2
TS 3/4.10.3
14
7
TS 3/4.3.3, TS 3/4.4.6
)
Final Safety Analysis Report 16.3.3.4
18
8
TS 3/4.8.1.1 -
19
9
TS 3/4.6.3
20
4
TS 3/4.0
l
23
1
TS 3/4.7.6e, 3/4.9.13d & 3/4.7.7d
25
2
Final Safety Analysis Report 16.9.2
32
2
TS 3/4.3.1
33
7
TS 3.7.1.5 & 3.7.1.6
36
2
Final Safety Analysis Report 16.4.5
42-
5
TS 3/4.9.2
l
44
1
TS 3/4.7.6, 3/4.7.7 & 3/4.9.13
46
6
TS 3/4.7.6, 3/4.7.7 & 3/4.9.13
49
6
TS 3/4.8.1.2, 3/4.4.1.4.2, 3/4.7.6 & 3/4.9.8
Final Safety Analysis Report 16.1.2.1 & 16.1.2.2
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Rev.
TS/ Final Safety Analysis Report Section
50
14
TS 3/4.7.1
51
1
TS 3/4.6.1.4 & 3/4.6.1.5
53
3
TS 3/4.4.6
54
0
. TS 3/4.3.3.5
56
6
TS 3/4.3.3
57
2
TS 3/4.6.1.1
Final Safety Analysis Report 16.6.1.1
58
1
TS 3/4.3.2
60.
4
TS 3/4 2.1
62
0
TS 3/4.3.1 & 3/4.3.2
63
3
TS 3/4.3.1
64
O
65
2
TS 3/4.4.9.3
66
O
TS 3/4.3.1
67
O
TS 3/4.9.12
68
1
TS 3/4.1
69
3
TS 5.1.3,' 6.8.4.e, 6.9.1.5, 6.12 & 6.14a
Final Safety Analysis Report 16.11.1
,
70
5
TS 3/4.1.3
72
1
Final Safety Analysis Report 16.0.3
73
4
TS 3/4.1, 3/4.1.1, 3/4.9 & 3/4.9.1
74
1
Final Safety Analysis Report 16,16.7.2.1 & 16.8.1
75
O
TS 3/4.5.2 & 3/4.5.3
Final Safety Analysis Report 16.1.2.1,16.1.2.2,16.1.2.3 16.1.2.4
76
O
TS 3/4.1.3.2
77
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TS 3/4.7.3
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Nuclear Enaineerina Quality Assurance Audits
4
l
AP94-019
Quality Assurance Audit of Materials and Nuclear Engineering Technical
'
Support (Materials Engineering)
$;
AP95-004
Quality Assurance Audit of Qualifications of Nuclear Division Personnel
4
i
AP96-001
Quality Assurance Audit of Qualifications of Nuclear Division Personnel
i
i
AP96-002
Quality Assurance Audit of Design Control
1
AP96 011
Quality Assurance Audit of Operator and Engineering Support Personnel
~
Training
l
AP96-010
Quality Assurance Audit of Corrective Action
-
Nuclear Enaineerino Quality Assurance Surveillances
l
SP94-107
Fire Barrier Penetration Seal Qualifications
!
SP95-013
10 CFR 50.59 Safety Evaluation for Final Safety Analysis Report Change
Notice 94-31
i
i
1
'
i
SP95-061
Plant Modification Configuration Control
SP95-073
Review of EDP-ZZ-04023, Calculations
1
SP95-101
Vendor Equipment Technical Information Program
SP96-027
Technical Adequacy and Configuration Management of the Request for
i
Resolution Program
I
.SP96-059
Cancellation of Modifications
SP96-103
Review of Component Cooling Water System Operating Parameters
Miscellaneous Documents
s
Technical Specifications
Final Safety Analysis Report
Operating Review Committee Meeting Minutes and Materials Review Log for the Operating
Review Committee Ineeting minutes for the period of December 15,1995 through
February 7,1997
SEGR 96-05-001, independent Safety Engineering Group 18 Month Summary, dated
May 23,1996
11
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<
SEGR 95-12-003, Independent Safety Engineering Group Review of NRC Region IV SALP
.
and Violations, dated February 20,1996
i
!
SEGR 96-10-003, Independent Safety Engineering Group Followup review of EF system
failures, dated October 10,1996
"
)
SEGR 9610-008, independent Safety Engineering Group 1995 and 1996 highlights, dated
November 4,1996
)
SEGR 96-11-007, independent Safety Engineering Group Review of SOS 96-1385, Gripper
Damage, dated November 13,1996
'
j
independent Safety Engineering Group Responsibility List, dated Jar.uary 13,1997
Reports to the industry via the INPO Nuclear Network submitted by Independent Safety
.
Engineering Group personnel, dated February 12,1997
,
l
Centralized Action Tracking System general information report providing the status of
j
Independent Safety Engineering Group entered items
'
Independent Safety Engineering Group 18 Month Summary, dated May 23,1996
Suggestion Occurrence Solution Report listing for 1996 and 1997 occurrences identified by
Independent Safety Engineering Group personnel
Independent Safety Engineering Group Action Tracking list for the past year
'
>
Operating Experience Crew Briefing Sheets, dated November 4,1996, September 10,
1996, October 1,1996, October 8,1996, October 10,1996, and January 13,1997
Operating Experience Journal on Excerpts from NRC Violations and Reports, dated
March 1996
^l
Operating Experience Journal on Shutdown Operations, dated May 1996
Operating Experience Journal on Drain Down and Midloop Operations, dated May 1996
Operating Experience Journal on Radiation Protection, dated May 1996
Operating Experience Journal on Shutdown and Startup Operating Experience, dated
June 1996
l
February 5,1997 Letter ULNRC-03530 to NRC, "NRC Enforcement Policy Revision"
February 6,1997 Letter ULNRC-3531 to NRC "Callaway Plant Response to Request for
Information Pursuant to 10 CFR 50.54(f) Regarding adequacy and Availability of Design
Basis Information"
Nuclear Energy Institute guidelines of NEl 96-05, " Guidelines for Assessing Programs for
Maintaining the Licensing Basis," Revision 0, dated October 7,1996
12
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