ML20198L524

Insp Rept 50-440/97-08 on 970721-0827.Violations Noted. Major Areas Inspected:Unresolved Items & Insp follow-up Items Identified by Design Insp Conducted from 970127-0327

ML20198L524
Person / Time
Site:	Perry
Issue date:	09/23/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20198L449	List: ML20198L444 ML20198L459 ML20198L512 ML20198L524
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

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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:

• The calculation referenced Section Ill, NB36541, of the ASME B&PV Code

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

for both LOOP and LOCA.

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

'

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.

• ' USAR Tables 9.218 (ECC Pumps) and 9.2-19 (ECC Heat Exchangers) listed

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.

• Calculation PRDC-0004, ' Class IE DC Control Circuit Coordination," Revision 2,

dated May 30,1995, did not address switch 412 added to drawing D206 051,

• Electrical Main One Line Diagram, Class IE DC System" Revision RR, dated May

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

1821 F065A 6.6 HP 6.4 HP

P42 F551 MISSING 0.13 HP

P45 D004A 7 HP 1 HP

P42F550 MIS 3tNG G.13 HP

M25-C001B 100 HP 60 HP

1G33 F001 3.0 HP 3.9 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

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- _ . - .-

!

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

HPCS High Pressure Core Spray

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

RHR Residual Heat Removal

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

VIO

'

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

. . _ . _, _ ___

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- . _ _ _ . _ .__ _ _ _ _ _ _ _ _ _ _ - _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ . . - _ . _

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

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- . . . _ _ - - . . - _ - . . - . . - . - . - - _ - _ .

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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