ML20198L524
| ML20198L524 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 09/23/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20198L449 | List: |
| References | |
| 50-440-97-08, 50-440-97-8, NUDOCS 9710270062 | |
| Download: ML20198L524 (27) | |
See also: IR 05000440/1997008
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. U. S, NUCLEAR REGULATORY COMMISSION
REGION lli
Docket No: 50-440
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License No: NPF 58
Report No: 50 44C/97008(DRS)
Licensee: Centerict Service Company
Facility: Perry Nuclear Power Plant
Location: P. O. Box 97. A200
. Perry, OH 44081
Dates: July 21 through August 27,1997
Inspectors: M. J. Miller, Reactor Engineer
Approved by: Mark Ring, Chief, Lead Engineers Branch
Division of Reactor Safety
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9710270062 970923
PDR ADOCK 05000440
0 PDR
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EXECUTIVE SUMMARY
Perry Nuclear Power Plant, Unit 1
NRC Inspection Repart 50-440/97008
This inspection reviewed the unresolved items and inspection follow-up iterr. identified by
the Design Inspection conducted from February 17 through March 27,1997.
Enaineerin.q
- The licensee's assessment of the collective significance of issues identified in the
Design Inspection Report was proactive and thorough. The conclusions
demonstrated that the issues had been identified to the licensee through other
means and that previous corrective actions did not resolve the issues.
- The safety evaluation which reviewed the changing of the emergency closed cooling
surge tank from a 7-day supply to a 30 minute supply failed to identify an
unreviewed safety question (USQ). Although engineering had the lead with the
evaluation and failed to identify the USQ, several additional barriers failed as well.
This is an apparent violation and is being considered for escalated enforcement
actions.
- Design control problems which resulted in violations were identified in:
- not incorporating the design oasis into the actual plant conceming tornado
missile protection of safety related equipment
- not taking into account worst case conditions for the condensate storage tank
swap-over point
a using conservatisms from the wrong perspective when performing a flooding
analysis for emergency closed cooling
e accepting open assumptions in calculations for long periods of time
- not updating calculations to reflect modifications to various systems
- an inadequate analysis concerning high pressure core spray (HPCS) over
pressure protection
- Corrective action problems, which resulted in a violation, were identified in -
- not recognizing the discrepancy between the HPCS keep-full pump performance
and the stated performance in the Updated Safety Analysis Report (USAR)
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finding a problem with the HPCS over pressure protection relay and then not
taking thorough corrective actions
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- not taking timely corrective action to repeated failures of the emergency' diesel
generator testable rupture discs -
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- . Inadequate leak testing of the emergency closed cooling boundary valves .
- Changes in methods of operating the plant were not evaluated appropriately when-
the suppression pool cleanup system was switched from occasional use to nearly -
continuous use and resulted in a violation.
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- Actions were taken that were assumed to be equivalent to the commitments made to
the NRC and resulted in deviations.
- Many examples of USAR discrepancies were identified which resulted in a violation.
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Report Details
This inspection was to follow up on the items identified in the Perry Unit 1, Design
Inspection (NRC nspection Report No. 50-440/97-201) conducted from February 17
through March 27,1997. The 22 unresolved items and 3 inspection follow-up items
identified in the report are discussed below.
111. Enaineerinn
E1 Conduct of Engineering
E1.1 Collective Sianificance of the identified issues
a. Inspection Scoce (37550)
The licensee issued Potential issue Form (PlF) 97-1024 to address the collective
significance of the issues identified in the Design Inspection Report. In addition, the
licensee reviewed previous inspections and audits to determine common issues.
The inspectors reviewod the PlF to determine the level of effort, assessment quality,
and final conclusions.
b. Observations and Findinas
The licensee's assessment team reviewed the Design inspection Report in detaii for
findings, conclusions, and identified corrective actions to be taken by the licensee.
Ten other assessment documents were then reviewed in a similar manner. A matrix
was created based on the Design Inspection Report data. The data gleamed from
the other documents was then incorporated into the matrix to make the comparison.
The team identified rine general conclusions from the review.
1. Design basis not readily retrievable, design basis documentation inconsistencies
2. Weakness with respect to Criterion Ill/ Calculation development and control
weakness
3. Untimely / ineffective Corrective Actions
4. Indication of a lack of rigor in documentation and understanding of the design
bases, and maintenance of the design and licensing basis
5. Facility being operated or maintained differently than described in the USAR
and/or vendor design input information/ Undocumented modification / Change to
the plant from that described in the USAR
6. Inconsistency in the plant's design and hcensing basis /USAR
7. Deviation for licensing commitments
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8. Raquirements and acceptance limits not in accordance with design documents.
Test control weakness
9. Weaknesses in the design control prograr.
The assessment also identified that,' corrective actions are typically not globalin
scope; they tend to be focused on correctiot. of the specific issues at hand, rather
it' n broader based extent-of condition type improvements initiatives.... Corrective
actions if more global in nature , tend to be more ' future fit" (for new work / products)
than for 'backfit." The assessment further identified that future-fit cotiective actions
would not have a substantial impact to upgrade ovarall product quality since
relatively few design documents would be captured in the corrective action.
The licensee's team concluded that the issues identified by the Design Inspection
had been previously identified by other assessments and that previous corrective
actions affecting old products tended to be ineffective.
c. Qonclusions
The licensee's assessment was proactive, thorough, and straight forwwd. The
licensee's conclusions paralleled concerns of the Design inspection team. Since it is
clear that these issues had been previously identified to the licensee and that
previous corrective actions had been ineffective in improving the older products,
future corrective actions specifically need to address the older design engineering
products.
E8 Miscellaneous Engineering issues (37550)
E8.1 LQig}d)J)nresolved item 50-440/97201-0t the c alculation for fue condensate
storage tank (CST) low level swap-over set point for high pressure core spray
(HPCS) suction from the CST to the suppression pool did not address worst case
conditicns. The calculation was to ensure that the HPCS system had adequate net
positive suction head (NPSH) and that no vortex would occur before the suction
valve swap-over to the suppression pool. Two specific conditions were not
addressed:
1. The licensee based the calculation on a flow rate of 700 gpm for reactor core
isolation cooling (RCIC) and 1550 gpm for HPCS. The licensee combined the
HPCS and RCIC flows since both would be taking a suction from the CST
through a common line. However, operating data specifies a maximum HPCS
flow rate of 6110 gpm at 200 psi back pressure in the reactor vessel and 7800
gpm at run out flow
2. The CST water level set points did not address continued draw down of the
CST as HPCS suction transferred from the CST to the suppression pool. HPCS
suction from the CST continues as the suppression pool suction isolation vaive
first strokes open, and the CST suction isolation valve then strokes closeo.
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The licensee had previously evaluated the adequacy of the swap-over set point as
part of the system based instrumentation and control Inspection (SBlCl) in 1995, and ,
found it to be acceptable. In resolving the issue in 1995, the licensee stated that the
primary function of HPCS was to alleviate the consequences of a smallline break,
when the reactor was at pressure and the required HPCS flow was 1550 gpm. For a
large break loss of cooling accident (LOCA) during which the HPCS would deliver
full flow, the licensee contended that suppression pool swell caused by the LOCA
would lead to a transfer of suction to the suppression pool as a result of the
suppression pool high water level swap-over set point. However, the licensee could
not identify design basis documentation that would substantiate the assertion that
pool swell negated the need for the CST level to cause the suction swap during
high flow conditions, as specified in the GE design documentation.
Regarding Doth items: The licensee revised the calculation to consider valve stroke
time and the worst case pump run out flow of 7200 gpm. To support the current set
point, the licensee had to use a less conservative methodology, which considers
operation in the region where vortex formation was possible. The licensee was
continuing to investigate the feasibility of changing the set point to provide additional
margin.
The licensee's use of non-conservative flow rates and not considering the impact of
valv6 timing within the original calculation to resolve the issue in 1995 were
inappropriate. 'Ihese issues represent an example of a violation of 10 CFR 50,
Appendix B, Critorion lil, ' Design Control,' (VIO 50-440/97008 01a).
E8.2 [ Closed) Unresolved item 50-440/97201 02: the HPCS keep-full pump was not
capable of delivering the pressure and flow specified in the Updated Safety Analysis
Raport (USAR). Two issues had been identified:
1. PlF 961609 written March 20,1996, identified that surveillance data for the
HPCS keep-full pump did not meet the USAR requirements for the pump. The
team identified that this condition had existed since a surveillance test
conducted on July 24,1993.
2. Concern was raised over contingency measures found in the annunciator
response inst uction ARI H13 P60116. In the event that operators were
unable to raise HPCS keep full system pressure after receiving a low pressure
annunciator, operators were directed to confirm that the system was filled by
checking its fill status every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or by performing the HPCS high point
vent. The licensee had not determined the rate of discharge line pressure
decay when the keep-full pump was not operating. Consequentiy, the arbitrary
time period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> may exceed the time at which voids were introduced in
the system.
Regarding item (1) The licensee indicated that , even though the keep full pump
was degraded, it was capable of maintaining system pressure above the alarm set
point. The licensee further Indicated that if the alarm was received, operators would
attempt to raise system pressure in accordance with procedures. Following the team :
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questioning the adequacy of the HPCS pump, the licensee parformed calculations
for the HPCS, RCIC, and residual heat remuval (RHR) keep full pumps. The
licensee determined that the pumps were adecuate for the systems.
10 CFR Part 50, Appendix 8. Criterion XVI, requires that conditions adverse to i
quality (such as failures, malfunctions, deficiencies) must be promptly identified and
corrected. The condition existed since July of 1993 based on surveillance date but
was not recognized. The licensee documented that the pump was degraded in PlF
901000, on March 20,1990. In a memo to the corrective action review board l
(CARB) dated June 25,1996, the failure to update the USAR was again identified. !
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The memo states: 'However, the USAR should have been revised to reflect the new
pump performance parameters generated in calculation E22 24, and it has not
been." The issue was still unresolved when the team started the design inspection
in February of 1997. The failure to take timely corrective actions is an example of a
violation of 10 CFR 50, Appendix B Criterion XVI, * Corrective Action," (VIO 50-
440/97008 02a).
Regarding item (2), PlF 97 0513 addressed this issue. The alarm response
instruction ARl H13 P60110 was changed to direct operators to use the HPCS
alternate keep-full method if the HPCS waterleg pump and keep-full instrumentation
were inoperable. The licensee indicated that upon a loss of both keep full methods,
the system would be declared inoperable and tne appropriate limiting condition for
operation would be entered. The licensee was not taking credit for system
operability based on checking the fill status every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without a keep-full
system aligned. This item is closed.
E8.3 (Closed) Upresolved item 50-440/97201 03: the licensee identified that HPCS
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should be declared inoperable during a specific surveillance; however, the team
questioned if technical specification (TS) violations occurred prior to the identification
of the issue.
This issue was identified through a Quality Assurance (QA) audit and a standing
instruction was issued on March 8,1995, to declared HPCS inoperable during
operation in the test mode. PlF 95 570 was issued and QA reviewed the logs back
to July of 1994 and confirmed that no violation of TS occurred, in addition, PIF 97-
0500 reviewed operating practices which were in place since initial operations.
These practices would not have allowed for removal of a safety system's redundant
train or the parallel safety function before or during secondary mode testing and
secondary mode testing would be postponed if the redundant train or the parallel
safety function was inoperable when the testing was scheduled to be performed.
Based on these two facts the licensee concluded that no violation of TS
requirements regarding this issue. This issue is closed.
E8.4 LQ1gsed) Unresolved item 50--440/97201 04: the over frequency protection relay for
HPCS discharge piping over pressure protection was never installed during
construction.
During construction of Perry Nuclear Power Plant Unit 1 (PPNP 1), the Architect /
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Engineers (AE) decided not to install the HPCS pump motor over frequency
protection relay. The licensee's safety system functional inspection (SSFI) of the
HPC9 system in 1992 recognized that the basis for not installing the relay was not
well founded. Consequently, the licensee performed Calculation E2219,
' Justification for Elimination of HPCS Over frequency Relay,' Revision 1 dated July
23,1992, to evaluate the effect of not installing the relay. The licensee's corrective
action decision in 1992 was improper for the following reasons:
in order to justify exceeding the system design pressure in the event that
the Division lli emergency diesel generator frequency goes above 60 Hz.
The licensee did not identify the specific eoition or addenda of the Code;
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however, Design Specification (DSP) E2214549-00, Revision 3, dated
April 18,1980, specifies the 1974 ASME B&PV Code with addenda up to
and including the winter 1975 issue, Section Ill, Division 1. This code
edition and addenda did not provide adequate basis to enable the licensee
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to justify the allowances for exceeding the design pressure in accordance
with rG36541.
- The licensee's calculation methodology provided a relief path to limit
pressure using the minimum flow valve and its actuation circuitry as
overpressure protection devices. However, the licensee was unable to
demonstrate compliance with the requirements for this valve and its
actuation circuitry, as specified in ASME Code Section Ill, Article NC 7000
- Protection Against Overpressure."
The team determined that the calculation methodology, code application, and
review / approval process used by the licensee did not ensure design quality as
specified in the USAR Section 17.2, Quality Assurance (QA) program and 10 CFR
Part 50, Appendix B, Criterion Ill,' Design Control." The licensee stated that they
would reevaluate the possible need to install the overfrequency relay as part of the
effort to resolve PIF 97-0575. Additionally, the team determined that the licensee's
improper disposition of the 1992 discovery of this inue constitutes ineffective
corrective actions This is an example of a violation of 10 CFR 50, Appendix B
Criterion XVI, (VIO 50-440/97008 02b).
E8.5 (Qgie.dLUnresolved item 50-440/9720105: multiple failures of the testable rupture
disc (TRD) for the emergency diesci generator (EDG) exhaust system. Two issues
had been identified:
(1) The TRD on the safety-related exhaust of the EDG was designed to
provide pressure relief in case the nonsafety related portion of the exhaust
or silencer was blocked, restricted, or inoperable. The team considered
that the licensee's corrective actions were deficient, since TRD reliability
problems appeared repetitive. There had been more than 12 failures to
date associated with the three divisions of EDGs, more than 6 years after
the first failure to open.
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(2) The team questioned whether the EDG would be able to start with a higher !
back pressure should the nonsafety-related exhaust be blocked before
starting the EDG, l
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Regarding item (1) In 1990, a modification was considered but never implemented. !
In 1995, a lifting lug was installed on all three TRDs which eliminated one of the :
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failure mechanisms associated with the testing methodology, However, other failure
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mechanisms associated with the TRD design were not addressed. The most recent
failure occurred on February 19,1997. j
10 CFR Part 50, Appendix B, Criterion XVI, requires that conditions adverse to ,
quality (such as failures, malfunctions, deficiencies) must be promptly identified and i
corrected. The failure to correct the failures of the TRDs in a prompt manner is an ;
example of a violation of 10 CFR 50, Appendix B Criterion XVI,' Corrective Action,' !
(VIO 50-440/97008 02c). !
Regrading item (2) The licensee contacted the vendor and confirmed that the EDG
would be able to start with the nonsafety related exhaust blocked and the associated
back pressure until the TRD lifted (documented in telecon memoranda PES NS$S-
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97 000/5 dated March 25,1997). This portion of the unresolved item is closed.
E8.06 (Closed) Unresolved item 50-440/97201 06: the droop setting for the Division lil
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EDG was procedurally set at 20 without documented design input.
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During initial plant startup, the Division til EDG govemor droop setting was kept at
zero when in the standby mode. When the diesel was tested, it was paralleled to
the grid after adjusting its droop setting to acccmmodate operation in the parallel ,
mode. After the diesel was shut down and returned to the standby mode, the droop
setting was returned to zero. The licensee was not able to determine when the !
change in droop setpoint occurred but present practice at PNPP 1 was to maintain .
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the DG3 droop setting at a value of 20 on the dial face at all times, The licensee ;
could not produce any existing documentation to support the current setpoint or to
define the impact of isochronous mode operation at a droop setting other than zero.
The inability of the licensee to produce documentation for this change is considered ;
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a weakness. This issue fits with a common theme throughout the Design inspection
Report (50 440/97201), a lack of retrievable documentation or failure to perform
certain analyses. For this specific issue, the vendor manual recommendation to set :
the governor droop at zero for the isochronous mode did not constitute a design
basis and therefore no violation of NRC requirements existed. This issue is closed.
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E8.07 LQpen) Unresolved item 50-440/97201-07: the methodology of addressing the
HPCS diesel generator govemor droop setting in an analysis appeared
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Inappropriate. The licensee has revised the calculation; however, this issue will
remain open for NRC review.
E0.08. LQ!osed) Unresolved item 50-440/97201-06; tomado missile protection for plant .
equipment was not as described in the USAR. The team klentified two methods by
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which safety related equipment located by the CST could be damaged by missiles:
1, During a seismic event, two non-seismic stacks on the top of the auxiliary
building had the potential to fall on and damage the CST water level instrument
piping and RCIC/HPCS suction piping.
2. Tornado generated missiles could damage the CST water level instrument
piping and RC!C/HPCS suction piping.
Regarding item (1) Although the stacks were not built for seismic considerations, the !
design requirements to address wind speeds bounded the seismic concerns. The
licensee documented this in calculation 1:05.7, ' CST Missiles & A.B. Stack," Rev.1,
completed April 4,1997. This item is closed,
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Regarding item (2) The plant safety evaluation report was issued in 1982. At that
time the licensee assumed that the probability of tomado missile strikes to exposed
safety related equipment vis one, Since the licensee elected to use a probability of
one, the safety evaluatior. riid not consider any lower probability in the review. In
1984, the keensee startet u Ng probability to determine where tornado missile
protection was
needed but the licensee did not recognize that they had changed their methodology
and the licensee did not submit a change The plant was licensed in 1986.
When the design inspection team raised questions with missile protection, the
licensee approach the issue using probability. The team was concerned with the
references being used with the probability approach. While the licensee was trying
to determine the correct references, they determined that Perry was not licensed to
use any probability approach for tornado missile protection. Once the licensee
recognized the licensing problem and that a potential existed for other missile
targets, immediate actions were taken to resolve the concern. During the licensee
actions, additional tornado missile targets were identified including the emergency
service water discharge piping.
As compensatory measures, the licensee modified the severe weather procedure
and later erected temporary protective barriers for the items identified by the team
and the items identified by the licensee. The licensee's actions were documented in
two letters to the NRC: PY-CEl/NRR 2180L dated June 13,1997, and PY-CEl/NRR-
2194L dated August 11,1997. The licensee has cubmitted a license change request
to allow for using probability in the tomado missile protection calculations.
The licensee failed to correctly translate the licensing basis into the design of the
plant, wh;ch is a violation of 10 CFR 50, Appendix B Criterion Ill, ' Design Control,"
(vio
50-440/97008-03).
E8.09 (Closed) Unresolved item 50-440/97201-09: suppression pool cleanup (SPCU) was
being operated continuously, which was not consistent with the USAR and required
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HPCS to be aligned to the suppression pool.
As part of a response to two residual heat removal (RHR) suction strainers being
deformed due to clogging in January 1993, (see Inspection Reports 92026,93004,
93005, and 93011) the licerme staried running the SPCU continuously to improve
the quality of the suppression pool water and reduce the potential of further strainer
clogging.
Continuous operation of the SPCU conflicted with the USAR since it stated that
containment isolation requirements for the SPCU return line were satisfied, in part,
on the basis that the line was normally closed, in addition, due to the piping
configuration, whenever the SPCU was placed in operation, the HPCS suction had to
be switched from the condensate storage tank (CST) to the suppression pool.
Having HPCS aligned to the suppression pool on a continual basis was also
inconsistent with the facility operation as described in the USAR.
Following identification of the issue the licensee issued a memorandum dated March
31,1997, that minimized the operation of the SPCU. In addition, the licensee
plaaned to install a new ECCS strainer system during refueling outage six,
scheduled to start September 12,1997, to address the stainer clogging issue.
In the process of addressing the strainer cloggi,,g issue, the licensee did not
adequately review the ramifications of the compensatory measures. The failure to
develop a safety evaluation as required in 10 CFR 50.59 for continuous operation of
the SPCU system, which was different than that described in the USAR, is a
violation of 10 CFR 50.59 (VIO 50-440/97008 04).
E8.10 (Open) Unresolved item 50-440/9720110: the team questioned an assumption
made for the suppression pool cleanup system. The assumption relaxed an original
PNPP assumption of full circumferential breaks in moderate-energy, nonsafety-
related, non seismic, Category I piping outside containment to permit consideration
of leakage cracks only. This item will remain open for NRC technical staff review.
E8.11 (Open) Unresolved item 50-440/97201-11: the design of the SPCU/HPCS Interface,
which would require HPCS isolation in the event of a SPCU system leak represented
an apparent oversight in the design. This condition had existed since the initial
licea. sing. This configuration was the subject of Engineering Design Deficiency
Report (EDDR) 10, dated February 13,1984, which was reported to the commission
via letters dated April 30 and June 8,1984. This item will remain open for NRC
technical statf review.
E8.12 (Closed) Unresolved item 50-440/97201-12.: a commitment to clean and inspect the
HPCS toom cooler was not met. The licensee had committed to clean and inspect
the HPCS room cooler at each refueling outage. The licensee inspected the cooler
each refueling, determined that cleaning was not necessary, and did not perform the
cleanings. .The licensee failed to change the commitment and in fact documented, in
a letter to the NRC (Implementation of GL 89-13, ' Service Water Problems Affecting
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_ Safety Related Equipment,' PY-CEl/NRR 1734L, dated April 8,1994), that the
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coolers had been cleaned and inspected.
In letter PY-CEl/NRR 2194L dated August 11,1997, the licensee changed the
commitment with respect to this unresolved item. The new commitment states, "The
Perry plant staff will continue to inspect and clean as necessary, as originally
intended in a commitment made in CEI letter (PY-CEl/NRR 1121), dated January 28,
1990.
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It appeared the licensee assumed their actions were equivalent to the commitment
made to the NRC. However, the missed cleaning and the licensee's subseqtunt l
report that the commitments had been satisfied constitutes a deviation from the
licensing commitments (DEV 50-440/97008-05).
E8.13 (Closed) Unresolved item 50 440/97201 13: a change in commitment was not met.
In 1994, the commitment to clean and inspect the HPCS room coolers was changed
from cleaning and inspecting to testing the coolers once per cycle until such time as
the testing demonstrated that a reduced frequency was warranted. The coolers were
tested once in June of 1995, which provided inconclusive results. Since 1995, no
other operability test was conducted. The team noted the licensee had not
established a performance test program for the HPCS room cooler and had reverted
to the insoection program, but had extended the frequency beyond each cycle
without a history of testing to demonstrate that a reduced frequency was warranted.
In letter PY-CEl/NRR 2194L dated August 11,1997, the licensee changed the
commitment with respect this unresolved item. The new commitment states, 'The
Perry plant staff will continue to use altemate me itoring methods as described in
EPRI NP 7552 whi!e developing a performance based test, as originally intended in
a comrnitment made in CEl letter (PY CEl/NRR 1734L), dated April 8,1994.
The failure to maintain test frequencies of once per cycle until such time as testing
demonstrated that a reduced frequency was warranted as stated in PY-CEl/NRR-
1734L, dated Ar/il 8,1994, is a deviation from licensing commitments (DEV 50-
440/97008-06).
E8.14 (Closed) Inspection Follow up Item 50-440/97201-14: the top and bottom rows of
batteries on Division til racks did not have the same number of vertical clamp-down
supports. Field variance authorization (FVA) 5847-33 1204 dated November 18,
1983, authorized the brackets not to be installed but did not provide justification.
The licensee performed calculation 40:71 Rev. O,'Div. 3 Battery Rack," dated
March 17,1997, which verified the adequacy of the as build clamp-down support
configuration. The applicable drawing was revised. This item is closed.
E8.15 LQigsed) Unresolved item 50-440/97201-15: several discrepancies were identified
with statements made in the USAR.
- The team identified the following administrative discrepancies between
Calculation PSTG-0014, * Diesel Loading Division I, ll, and 111,* Revision 3, and
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USAR Table 8.31: The licensee had previously identified similar discrepancies
with USAR Table 8.31 and Calculation PSTG-0014 as documented in PIF 96- ,
2780.
a. USAR Table 8.31 listed fuel oil tonsfer pumps 1R45C001C and 20 as 0- ,
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second loads.1R45C001C and 2C were 40-minute automatic cyclic loads
b. USAR Table 8.31 identified a 9 kW load for 1E22C004B and did not agree
with the 8 kW load in Calculation PSTG 0014. ,
- The team identified discrepancies desenbed below indicated that deshn and 7
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calculation changes were not accurately reflected in the USAR:
a. USAR Table 8.31 listed the inrush currents for HPCS fuel oil transfer
pumps 1R45 C001C and 2C as 109A, whereas Calculation PRMV-0017
listed the inrush current as 130A.
b. USAR Table 8.31 listed the inrush currents for HPCS diesel generator
room fans OM43 C001C and 2C as 362A, whereas Calculation PRMV-0017
listed the inrush current as 376A.
c. USAR Table 8.31 listed the FLA of HPCS diesel generator starting air
compressor 1E22 C004B as 13A, whereas Calculation PRMV-0017 listed
the FLA as 11 A.
d. USAR Table 8.31 listed the HPCS emergency service water (ESW) pump
1P45-C002 as 75 hp,88.5 FLA, and 557A inrush, whereas Calculation
PMRV-0017 listed the same load as 75 hp,85.4 FLA, and 543A inrush.
e. USAR Table 8.31 listed the rating of HPCS diesel generator space heater
1E22 D011 as 2 kW, with a load current of 3 amp. Calculation PRMV-
0017 listed the same space heater as 1.6 kW, with the load current of 2.01
amp. Drawing D 206-029/BB, " Electrical One Line Diagram, Class lE,480
V Bus EF10," listed the same space heater as 2.4 kW.
- USAR Table 3.9-30 listed active valves not associated with the nuclear steam
supply system (NSSS). This table had not been updated to reflect several
emergency closed cooling (ECC) system modifications. Valves P42-F315A, B,
"
C should have been deleted from the table, since they were converted from
automatic to manual valves by DCP 92-0060. Valves P42 F550 and P42 F551
should_have been added to the table, since they were converted from manual to
automatic valves by DCP 90-0012.
two different values for ECC system operating flow rate (1860 versus 1820
gpm). Since all pump flow is delivered to the heat exchanger, the flow values
should be the same.
14
. - . . _ .-.- - - . .- -_. . . - - . - - - - - -,
. __._ _ _ _ _ _ . - _ _ . _ _ _ _ - - --- _ - _ . _ . _ _
The licensee issued potential issue forms (PlF) to address the discrepancies listed
above and was in the process of evaluating the issues. The failure of the licensee to
'
correct the above discrepancies or update the USAR to ensure that the USAR
contained the latest information is a violation of 10 CFR 50.71 (VIO 50-440/97008- !
07),
{
E8.16 (Closed) Inspection Follow uo item 50-440/97201 16: within different sections of the
.
USAR, the definition of passive failure was inconsistent.
The GE design basis defines that passive failure design is only applied to electrical !
equipment. However, during the design of the emergency closed cooling (ECC) !
system the AE also included valve stem and pump packing failures. The licensee
agreed that the USAR was not consistent concoming this issue and that the USAR !
would be updated by December 12,1997, to clarify that the original design of the !
ECC system also included pump seals and valve stem packing failures, in addition, :
a review was planned for other USAR sections to identify and clarify similar
problems. This item is closed.
i
'
E8.17 Mosed) Unresolved item 50 440/9720117: a safety evaluation analyzed the
change of the ECC surge tank from a 7 day supply to a 30 minute supply without
recognizing an unreviewed safety question. .
,
in compliance with 10 CFR 50.59, the licensee prepared Safety Evaluation 96-128,
dated October 10,1996 to evaluate the USAR and procedural revisions associated
with the change in the surge tank sizing basis from a 7-day supply without necessary
makeup to a 30 minute supply. The safety evaluation was also used as a basis for
the use as is disposition of PlF 96 2846, associated drawing change notice (DCN) '
5541, and USAR change request (CR) 96150. The safety evaluation concluded that
the change did not constitute an unreviewed safety question.
The inspector concluded that the change to the ECC surge tank sizing basis from a
7 day supply to a 30 minute supply, with operator actions required outside of the
control room to initiate makeup from the ESW system, constituted an unreviewed
safety question, as defined in 10 CFR 50,59, because it- t
- Increases the probability of an occurrence of a malfunction of equipment
- important to safety. Reliance on operator action at 30 minutes after the
accident, under stressful and hazardous working conditions, increases the
probability that the operator would not correctly perform the required actions.
- Increases the consequences of an accident. Total cumulative operator
exposure had increased by 12 rem,: and the potential existed that an individual
operator's exposure could exceed the General Design Criteria (GDC) 19 limits
specified by NUREG 0737, item 11,B.2. ,
The licensee retested the valves using a revised procedure. Based on total system
leakage, the available surge tank supply was greater than 7 days. The USAR and l
all associated procedures involved in this issue were revised to reflect the 7 day
15
.
1
nn-- ,,-.m-,,nwr, m--r-v-v,+-,-..., wm -o -w-wn n m nn , -- a wn,r. -an,---, ..m,n,-,-,--n---,nn.r-n,- - , ro w , , -- rwem, 4
.. .. ._ - - _ - _ _ - -
surge tank supply. ,
,
The licensee's second review of the safety evaluation (documented in the Plant
Operations Review Committee meeting minutes dated August 11,1997) determined
that the original change to the USAR did involve a USO. However, during the first
review multiple barriers failed to detect the USQ. The safety evaluation was first
reviewed by engineering and then by management from multiple departments. The
failure to recognize the unreviewed safety question during the safety evaluation and
failure to submit the USQ for NRC review is an apparent violation of 10 CFR 50.59 i
(eel
50-440/97008 08).
E8.18 (Closed) Unresolved item 50-440/97201 18: flooding analysis for ECC surge tank
makeup used a non conservative flooding rate. Engineering had been focused on j
the minimum flows to the tank to address other concerns and failed to recognize that !
'
for flooding considerations the maximum flow to the tanks should have been used.
This was a lack of attention to detail on the part of the engineering staff and on the
part of the reviewer. The failure to use conservative flooding rates in Safety
_
Evaluation 96-128 is an example of a violation of 10 CFR 50, Appendix B Criterion
111, ' Design Control," (VIO 50-440/97008-01b).
E8.19 (Closed) Unresolved item 50-440/9/2011% the test procedure and acceptance
criteria, for testing ECC boundary valves leakage, did not adjust the leakage
measured under test conditions to expected leakage under post accident differential
pressures.
The leakage test was originally designed to determine gross leakage through the
butterfly valves, which would indicate the disc was not properly positioned by the
motor operator limit switch. The valve leakage was the subject of escalated
enforcement action in late 1996 (see inspection report 50-440/90008(DRS)). In the
licensee's response to the Notice of Violation (PY-CEl/NRR 2118L dated December
6,1996)it stated: "These valves were re categorized as ASME Code,Section XI.
Category 'A' valves on October 8,1996. As such, they will be periodically leak
tested as part of the In service Testing program (ISTP) against specific leakage
acceptance criteria." The letter further stated that full compliance had been
achieved.
The leak test procedure originally written to determine gross leakage was not revised
to demonstrate that the equipment could perform satisfactorily under accident
conditions. Specifically the differential test pressure specified for system boundary
valve seat leak testing was only approximately % of the pressure that the valves
would be subjected to under accident conditions, No extrapolation of test data to
compensate for this difference in test condition was included in the procee e.
However, the Ucensee took credit for previous leakage tests performed with the
original procedure to achieve compliance. This test procedure was part of the
licensee corrective actions for the escalated enforcement item stated above. The
inadequate test procedure was an ineffective corrective action and represents an
example of a violation of 10 CFR 50, Appendix B Criterion XVI, (VIO 50-440/97008-
10
02d).
E8.20 (C101f4)JntipEtion Follow up Item 50-440/9720b2Q; due to conflicting
information, the team was concerned that the ECC pump may have been
allowed to operate below the minimum flow rate allowed by the pump vendor.
The design inspection team noted that the Ingersoll-Rand certified pump curve
indicated that the minimum required continuous flow for the ECC pumps was 800
gpm although the system operating instruction (SOI) stated a minimum flow value of
500 gpm. The licensee's investigation determined that the pump vendor's technical
manual specified a minimum flow equal to 25% of the best efficiency point or 575
gpm. The change from 800 to 575 gpm was previously evaluated in 1990 in CR 90
100; '3 wever, the curve had not been revised.
Through the corrective actions of PlF 97 0470, the vendor pump curve was revised
to delete the note and the sol value was corrected to 575 gpm. The licensee
determined that the sol did not have any operating mode where flow would have
been below the 575 gpm. PIF 97-0470 attributed the wrong value listed in the sol
as a calculation error on the part of a system engineer.
The licensee had identified a problem in 1990, and took actions to resolve the issue.
The concern identified by the team was raised due to incomplete closure, on the part
of the licensee, of an identified issue. This was a minor example of a lack of
attention to detail in completely closing issues. This item is closed.
E8.21 LQ1gitd) Unresolved item 50-440/9720141: a calculation, ' Evaluation of Heat
Transfer Coefficient and Minimum Required Wall Thickness for ECC Heat
Exchangers 1P42 80001 A/B," Revision 0, dated May 1,1996, referenced ASME
Section Vill criteria instead of ASME Section ill criteria. The licensee review of the
c.alculation Indicated that Sections lll and Vill use the identical code methodology
and the calculation results were unchanged. The error was in the reference and not
in the calculation. This issue is minor in nature and representr insufficient attention
to detail on the part of engineering. This item is closed.
E8 22 LClosed) Unresolvedjlem 50-440/0720122: calculations were found with open
assumptions that had not been verified or confirmed in a timely manner. The
concem was that these open assumptions may potentially have some impact on the
plant,
The team identified a post-test calibration, which was specified to confirm a
calculation (Calculation P42 31, 'ECC A Heat Exchanger Test Results-1995,"
Revision 0 dated September 15, 1995), had been outstanding for 18 months. The
calculation had been used as the basis for equipment operability evaluations. During
follow-up to the teams concerns, the lice,'see identified ht 1 of the 8 temperature
measurements on both the ECC intet and outlet were out of . Nh.*ation. However,
engineering was not informed of the failure. In this case, the failure did not affect the
conclusion of the operability evaluations.
17
)
l
The licensee's follow up on the team's concem included a partial list of 44
calculations with unconfirmed assumptions, in the past, the license had no formal
computerized (or other) tracking system that was user friendly. Until the recent
initiation of the Calculation Database Project started in late 1996, the only method of
checking for unconfirmed assumptions was to do a manual, periodic search of
voluminous calculation logs.
The failure of licensee personnel who performed the post-test calibration to alert
engineering to instrumentation that was out of calibration demonstrated poor
communications. However, the failure of engineering to follow up on an open
assumption for over 18 inonths is an example of a violation of 10 CFR 50, Appendix
B Criterion Ill, * Design Control," (VIO 50-440/97008-01c)
E8.23 { Closed) Unresolved item 50-440/97201 21: the post accident operator action to
cross tie Unit i emergency service water (ESW) to the common fuel pool cooling i
and cleanup (FPCC) was not evaluated for the radiological exposure to the operators l
in accordance with NUREG-0737, item II.B.2. Prior to the inspection, the licensee
identified this as a generic issue in PlF 97-0248. This issue did not appear to be
willful, was not reasonably preventable by previous corrective actions, and should be
corrected in a reasonable time frame commensurate with the requirements of the
licensee's corrective action program. This item is closed. ;
I
E8.24 (Closed) Unresolved.ifem 50-440/97201-24: the licensee has modified various
systems as reflected in design drawings and did not update or revise the !
calculations. The licensee's control of calculations was questioned due to the
following discrepancies identified by the team:
- Electrical drawing D 206-029, " Electrical One. Line Diagram, Class IE,480 V
Bus EF1D," Revision BB, identified the installation of a 10-hp electnc motor for
compressor 1E22 C004 A Calculation PRMV 0017, *EHF 1 E Transformer Breaker
EH1305 " Revision 0, did not list the compressor motor. Following the team's
identification of the issue, engineering performed an operability evaluation and
concluded that sufficient margin existed between the estimated values and the actual
50/51 relay settings. Therefore, operability or breaker EH1305 was not a concern.
Calculation PRMV-0017 was last updated on March 11,1985 (12 years ago), and
did not reflect the current plant loads and settings.
- Calculation PSTG-0003 '480-V Safety-Related Motor Starting Voltage Drop,"
Revisiun 2, dated June 29,1995, contained an open assumption that required
confirmation. Calculation PSTG 0001, *PNPP Auxiliary System Voltage Study,"
Revision 2, approved on August 24,1995, provided the information to resolve the
open assumption. As of March 27,1997, calculation PSTG-0030 was not upJaled to
close the open assumption. Although, the information to close the assumption was
available on August 24,1995.
- Calculation PRDC-0006, * Load Evaluation and Battery Sizing of Division lll Class IE
DC System," Revision 0, dated April 8,1991, did not address Division ll1 HPCS
pump 1E22C001 breaker EH1304 spring charging motor load at t=0 second, the load
18
_ _ . _ . _ _
profile for 0-1 min for continuous (L2) load, and the DC control circuit loads (L2 ;
'
loads) of the breakers.
dated May 30,1995, did not address switch 412 added to drawing D206 051,
15,1992, in accordance with DCP 90-0012. The Drawing D206-051 was at current
revision WW, dated April 7,1996.
1
- Calculation PRLV-0004, ~480 V Breaker Coordination,' Revision 2, dated April 30,
1996, was reviewed against associated electrical D 206 series drawings for 480 V
motor control centers (MCCs). Various discrepancies and typographical errors were
found between the calculations. and the drawings as noted below:
l
MPL# Calculation PRLV-0004 Drawing D-206
series
P42 F551 MISSING 0.13 HP
P42F550 MIS 3tNG G.13 HP
._
PIF 97-0494 was issued by tne licensee to resolve the deficiencies and typographical
errore listed in the table. Engrneering verified that the calculation was still valid for
the over current protective devices of the 480 V switchgear breakers, and adequate
protection of the downstream equipment was still provided without premature tripping
on short time demand.
The licensee had not performed a review to determine the extent of the above
conditions as they relate to other (similar) calculations. The licensee generated
Potential issue Forms to address the issues above. These calculation control
deficiencies did not meet the !icensee's 10 CFR Part 50, Appendix B, Criterion 111,
Design Control Program as described in USAR Section 17.2 and are examples of a
violation of 10 CFR 50, Apper: dix B Criterion ill,' Design Control," (VIO
50-440/97008-09).
E8.25 (Closed) Unr_qsolved item 50-440/97201 25: an analysis for over pressure protection
for portions of the HPCS system were inadequate (Calculation E22 2, 'Over
pressure Protection Analysis," Revision 0, dated February 23, 1983). Three issues
had been identified:
19
l
l
- -. -
1. The analysis did not identify operating conditions under which pressure relief
devices were required to function including the relief capacity required to
prevent system components from being subjected to pressures exceeding
code allowable values. Specifically the maximum pressure considered for the
suction piping was 31.25 psig whereas the suction side relief was set at 100
psig. The maximum discharge pressure considered was 1130 psig whereas
the discharge side thermal relief nive was set at 1560.
2. The maximum pressure to which the suction piping could be subjected was
contingent on the static head from the normal water level of the CST, The
calculation did not consider the static head from the maximum overflow level
within the CST, it did not consider alignment to the suppression pool with
- consideration for containment overpressure, and did not consider back
leakage from the reactor pressure vessel, which may result in pressurization
of the suction line.
3. The licensee considered a pump total discharge head (TDH) at the shutoff of
2630 feet and did not consider coincident suction pressure. In addition, no
consideration was made for pump overspeed conditions with the associated
increase in pump TDH.
The calculation was not used to determine the system relief valves setpoints. It had
been performed at the request of the State of Ohio Enforcement Authorities to
ensure that no components within the system were subjected to pressures and
temperatures which were beyond the design parameters of the components.
However the methodology and system modeling concerning this calculation were
inadequate and this is considered an example of a violation of 10 CFR 50, Appendix
B Criterion lil,' Design Control," (VIO 50-440/97006-01d).
LMananoment Meetings. <
X1 Exit Meeting Summary
The inspector presented the inspection results to members of licensee management at the
conclusion of the inspection on August 27,1997. The licensee acknowledged the findings
presented.
The inspectors asked the licensee whether any material examined during the inspection
should be considered proprietary. No proprietary informatiors was identified.
20
- _ . _ _ _ _ _ _ _ _ _ _ . _ . _ . - _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _
- _ . - .-
!
PARTIAL LIST OF PERSONS CONTACTED l
klGADifA .
C. Angstadt, Engineering Assurance Lead 1
H. Bergendahl, Director Nuclear Services Department !
J. Grabner, Projects Unit Supervisor t
D. Gudger, Compliance Engineer ,
'
W. Kanda, General Manager Plant Department
L. Myers, Vice President Nuclear :
J. Powers, Design Engineering Manager
R. Schrauder, Perry Nuclear Engineering Director
- J. Stetz,- Senior Vice President, Nuclear
t
'
INSPECTION PROCEDURES USED
'
lP 37550: Engineering
i
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
50 440/97008-01 VIO The calculation for the condensate storage tank (CST) low level
swap over set point for high pressure core spray (HPCb)
suction from the CST to the suppression pool did not address
worst case conditions (from URI 50-440/97201-01)
Flooding analysis for emergency closed cooling (ECC) surge
tarik makeup used a non-conservative flooding rate (from URI
50-440/97201 18) ,
Calculations were found with open assumptions that had not
been verified or confirmed in a timely manner (from URI 50-
440/97201 22)
An analysis for over pressure protection for portions of the
HPCS system were inadequate (from URI 50-440/97201-25)
50-440/97008-02 VIO The HPCS keep full pump was not capable of delivering the
pressure and flow specified in the USAR (from URI 50-
440/97201 02)-
i
The over frequency protection relay for HPCS discharge piping
over pressure protection was never installed during construction
(from URI 50 440/97201-04)
Multiple failures of the testable rupture disc for the emergency
21
'
i
'
.- . - . . - . . .- . . _ - - -- -
. . . _ _ __- _ - = . _ _ - ._- - ----_ .
. - .__
diesel generator exhaust systems (from URI CD 440/97201-05)
The test procedure and acceptance criteria, for testing ECC
boundary valves leakage, did not adjust the leakage nisasured
under test conditions to expected leakage under post accident
differential pressures (from URI 50-440/97201 19)
50-440/97008 03 VIO Torriado missile protection for plant equipment was not as
described in the Updated Safety Analysis Report (USAR) (from
URI 50-440/97201-08)
50-440/97008 04 VIO Suppression pool cleanup (SPCU) was being operated
continuously, which was not consistent with the USAR and
required HPCS to be aligned to the suppression pool (from URI
50 440/97201 09)
50-440/97008-05 DEV A commitment to clean and inspect the HPCS room cooler was
not met (from URI S0-440/97201-12)
50-440/97008 06 DEV A change in commitment to test HPCS cooler once per cycle
was not met (from URI 50-440/97201-13)
50-440/97008-07 VIO Several discrepancies were identified with statements made in
the USAR (from URI 50-440/97201 15)
50-440/97008 08 eel Apparent violation - a safety evaluation analyzed the change of
the ECC surge tank from a 7-day supply to a 30 minute supply
without recognizit g an unreviewed safety question (from URI
50-440/97201-Q
50-440/91008-09 VIO The licensee had modified various systems as reflected in
design drawings and did not update or revise the calculations
(from URI 50 440/97201 24)
Closed
50-440/97201 01 URI The calculation for the CST low level swap-over set point for
high pressure core spray suction from the CST to the
suppression pool did not address worst case conditions (closed
to VIO 50-440/97008 01)
50 440/97201-02 URI The HPCS keep-full pump was not capable of delivering the
pressure and flow specified in the USAR (closed to VIO 50-
440/97008-02)
50-440/97201 03 URI The licensee identified that HPCS should be declared
inoperable during a specific surveillance; however, the team
22
.
- . . _ . _ . . _ _ _ _ -._ ___ . . _ . . . . _ . _ _-_ . _ _ _
questioned if technical specification violations occurred prior to ;
'
the identification of the issue.
'
50-440/97201 04 URI The over frequency protection relay for HPCS dischaige piping
over pressure protection was never installed during construction :
(closed to VIO 50-440/97008-02)
50-440/97201 05 URI Multiple failures of the testable rupture disc for the emergency
diesel generator exhaust systems (closed to VIO 50-440/97008-
02)
50-440/97201-06 URI The droop setting for the Division lll emergency diesel
generator was procedurally set at 20 without documented
design input
50 440/97201 08 URI Tornado missile protection for plant equipment was not as
described in the USAR (closed to VIO 50-440/97008 03)
50-440/97201 09 URI SPCU was being operated continuously, which was not
consistent with the USAR and required HPCS to be aligned to
the suppression pool (closed to VIO 50-440/97008 04)
50 440/97201 12 URI A commitment to clean and inspect the HPnS room cooler was
not met (closed to DEV 50-440/97008-05)
50-440/97201 13 URI A change in commitment to test HPCS coolers once per cycle
was not met (closed to DEV 50-440/97008 06)
50-440/97201 14 IFl The top and bottom rows of batteries on Division lll racks did
not have the same number of vertical clamp-down supports
50-440/97201-15 URI Several discrepancies were identified with statements made in
the USAR (closed to VIO 50 440/97008-07)
50-440/97201 16 IFl Within different sections of the USAR, the definition of passive
failure was inconsistent
50 440/97201 17 URI A safety evaluation analyzed the change of the ECC surge tank
from a 7 day supply to a 30 minute supply without recognizing
an unreviewed safety question (closed to apparent VIO
50-440/97008-08)
50-440/97201 18 URI Flooding analysis for ECC surge tank makeup used a non-
conservative flooding rate (closed to VIO 50-440/97008-01)
50-440/97201 19 URI The test procedure and acceptance criteria, for testing ECC
boundary valves leakage, did not adjust the leakage measured
23 ,
- . . .-- - . - ._ _
_ . . __ _ . _ . _ _ _ _ _ _ _ . _ _ _ _ __ _. ___ _ . . .-_ _
under test conditions to expected leakage under post accident i
differential pressures (closed to VIO 50-440/97008-02)
50-440/97201 20 IFl Due to conflicting information, the ECC pump may have been
allowed to operate below the minimum flow rate allowed by the
pump vendor i
50-440/97201-21 URI A calculation * Evaluation of Heat Transfer Coefficient and
Minimum Required Wall Thickness for ECC Heat Exchangers
1P42 80001 A/Bf referenced ASME Section Vill criteria
instead of ASME Section ill criteria.
50 440/97201 22 URI Calculations were found with open assumptions that had not
been verified or confirmed in a timely manner (closed to VIO
50-440/97008 01)
50-440/97201 23 URI The post accident operator action to cross tie Unit i emergency
service water to the common fuel pool cooling and cleanup was
not evaluated for the radiological exposure to the operators
50-440/97201 24 URI The licensee had modified various systems as reflected in
design drawings and did not update or revise the calculations
(closed to VIO 50-440/97008 09)
50-440/97201 25 URI An analysis for over pressure protection for portions of the
HPCS system were inadequate (closed to VIO 50-440/97008-
01)
Qlscussed
50-440/97201-07 URI The methodology of addressing the HPCS diesel generator
governor droop setting in an analysis appeared inappropriate
50-440/97201-10 URI An assumption made for the SPCU system relaxed an original
assumption of full circumferential breaks in moderate-energy,
nonsafety-related, non seismic, Cate0ery I piping outside
containment to permit consideration of leakage cracks only.
50-440/97201-11 URI The design of the SPCU/HPCS interface, which would require
HPCS isolation in the event of a SPCU system leak
represented an apparent oversight in the design
24
- _ . - . . - - -
. -. .. . . .- - . - - . _ - . - - - _ - . . - . . . - - - - - . . - . = . - . . - .
i
h
i
LIST OF ACRONYMS USED i
AE Architect / Engineers i
ASME American Society of Mechanical Engineers lt
CARB Corrective Action Review Board
CST Condensate Storage Tank
-DCN Drawing Change Notice
DEV Deviation '
ECC Emergency Closed Cooling ,
ECCS- Emergency Core Cooling System i
EDDR Engineering Design Deficiency Report !
EDG Emergency Diesel Generator !
EPRI Electric Power Research Institute l
ESW Emergency Service Water
'
!
FLA Full Load Amperage
FPCC Fuel Pool Cooling and Cleanup ,
.FVA Field Variance Authorization f
GDC General Design Criteria
IFl Inspection Follow up Item .
ISTP. In service Testir.g Program
LOCA Loss of Cooling Accident
MCCs Motor Control Centers >
NPSH Net Positive Suction Head
NSSS Nuclear Steam Supply System
PIF Potential Issue Forms
- PPNP 1 Perry Nuclear Power Plant Unit 1
QA. Quality Assurance
RCIC - Reactor Core Isolation Coolmg
SBICI System-Based Instrumentation and ControiInspection
sol System Operating Instruction
SPCU Suppression Pool Cleanup
SSFl Safety System Functional Inspection
TDH Total Discharge Head
TRD Testable Rupture Disc .
TS Technical Specification :
URI Unresolved item
'
Violation ,
._
t
25
-- , . . , - _ - - , ~. - - . - . .
-.. -.. .. - - . - - - - . . - - . - _ _ _ . - - - . - _-. - ..-..- - - . .
I
LIST OF DOCUMENTS REVIEWED
Document No. Qpjg Document Title / Descriotion
ARl-H13 P60116 Rev 4 Revised annunciator response instruction for HPCS
water let pump discharge pressure low
'
Calc 1:05.7 4/04/97 Review of aux boiler stacks to determine if they are
tornado / seismic hazards for the HPCS pip!ng and i
instrumentation within the CST dike
Calc 40:71 3/17/97 Division 3 battery rack
Calc E22 29 7/14/97 HPCS Pump Penfm. 1ce Acceptance C.iteria '
Calc E22 36 5/5/97 Minimum acceptance r. ,uirements for the HPCS
waterleg pump
Cale P42 33 8/11/97 Evaluation of Heat Transfer Coefficient and min.
required Wall Thickness for ECC Heat Exchanger 1P42-
B0001A/B
DCN 05638 4/21/97 Revised drawing to reflect one hold down clamp on a
number of batteries
PlF 961609 3/20/96 HPCS waterleg pump performance
PlF 97-0165 1/28/97 Operation of EDG with 2% droop ,
PlF 97-0248 2/06/97 No method to track dose to operators for postulated
post accident radiological conditions
PlF 97-0325 2/19/97 EDG testable rupture disc failure
PlF 97-0343 2/19/97 llSAR discrepancies
PIP 97-0350 2/20/97 USAR discrepancies
PIF 97-0395 2/27/97 USAR discrepancies
- PlF 97-0406 2/27/97 Non-conservative ass amption in flooding evaluation
PlF 97-0407 2/28/97 Missing hold down brackets on battery rack
PlF 97-0416 2/25/97 CST vortexing question due to HPCS suction flow rates
PlF 97-0425 3/04/97 USAR discrepancies
26
. . _ . _, _ ___
._ _ _
- . _ _ _ . _ .__ _ _ _ _ _ _ _ _ _ _ - _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ . . - _ . _
PlF 97 0426 3/04/97 Over pressure protection analysis for HPCS
i
PIF 97-0463 3/08/97 Missed commitments on the HPCS room cooler :
PlF 97-0469 3/07/97 USAR clarity concerning control room Indication and
annunciation
!
PlF 97-0470 3/10/97 Minimum flow concerns for ECC pump
PlF 97 0494 3/12/97 Discrepancies between drawing and calculations ;
PIF 97-0496 3/12/97 Discrepancies between drawing and calculations
PlF 97-0497 3/12/97 Open assumption not confirmed
,
PlF 97-0500 3/11/97 USAR discrepancies
PIF 97 0511 3/13/97 Battery load profile calculation did not account for spring
charging motor
PlF 97 0512 3/12/97 USAR discrepancies
PlF 97-0513 3/13/97 Alarm response instruction allowed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> between
checks of HPCS water fill
PlF 97 0517 3/13/97 Corrective actions for safety related electrical protection
calculations
PlF 97-0526 ?/17/97 Design questions conceming the interface of the HPCS
and the SPCU systems
PlF 97 0531 3/17/97 Wrong reference to ASME section - tube thickness
calculation for heat exchanger
PIF 97 0543 N/A Open assumptions had not been verified in a timely
manner
PlF 97-0500 3/24/97 Reportability of past operability of other systems prior to
HPCS being declared inoperable during testing
PlF 97-0561 3/24/97 Tornado and seismic missile protection of the
HPCS/RCIC system suction pipe near the CST
PIF 97 0575 3/26/97 Elimination of the HPCI overfrequency relay
PlF 97-0578 3/26/97 ECC leak rate testing non conservative
PlF 97-0815 5/15/97. Review of the flow through two ECC valves - USAR
27
. --- -. . - - - . . _ - . . -- - . _ - . - . . . .-
- . . . _ _ - - . . - _ - . . - . . - . - . - - _ - _ .
discrepancies
Memorandum 3/31/97 Suppression Pool Demineralizer Service Life
Optimization Plan
Memorandum 9/04/97 Operation of Suppression Pool Clean up and
High Pressure Core Spray System
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