IR 05000266/1998009

Insp Repts 50-266/98-09 & 50-301/98-09 on 980414-0523. Violation Noted.Major Areas Inspected:Operations,Maint, Engineering & Plant Support

ML20249C286
Person / Time
Site:	Point Beach
Issue date:	06/19/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20249C284	List: IR 05000266/1998009 ML20249C283 ML20249C285
References
50-266-98-09, 50-266-98-9, 50-301-98-09, 50-301-98-9, NUDOCS 9806290004
Download: ML20249C286 (29)

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U.S. NUCLEAR REGULATORY COMMISSION REGION lli Docket Nos: 50-266; 50-301 License Nos: DPR 24; DPR-27 Report No: - 50-266/98009(DRP); 50-301/98009(DRP)

Licensee: Wisconsin Electric Power Company Facility: Point Beach Nuclear Plant, Units 1 & 2 Location: 6612 Nuclear Road Two Rivers, WI 54241-9516 Dates; April 14 through May 23,1998 i

inspectors: F. Brown, Senior Resident inspector P. Louden, Resident inspector P Simpson, Resident inspector M. Kunowski, Project Engineer Approved by: J. W. McCormick-Barger, Chief Reactor Projects Branch 7 9806290004 980619 "

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e EXECUTIVE SUMMARY Point Beach Nuclear Plant, Units 1 & 2 NRC Inspection Report 50-266/98009(DRP); 50-301/98009(DRP)

This inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a 6-week inspection period by the resident inspector Operations

. The Unit 2 licensed operator was not aware of a Unit 1 abnormal chemical and volume control system (CVCS) lineup and the impact this lineup had on the Unit 2 CVCS due to an incomplete turn-over and problems with control room activity coordinatio (Section O1.2)

+ A minor, non-cited violation of Technical Specifications occurred when a radioactive effluent discharge was made without first source checking an in-line radiation monito This failure could have been prevented by better control room command and control of shift activities, operator turnover, procedural adherence, and operator self-verificatio (Section 08.3)

- The licensee performed an effective self-assessment of the conduct of operations. The use of industry peers was highly beneficial. The need for improvement to bring the operations department up to current industry standards was identified. All self-assessment findings were consistent with the inspectors' observations. (Section 01.3)

. The licensee's root cause evaluation for a February 7,1998, waterhammer event attributed the event to a poor operating procedure and poor decisions by the operating crew. The corrective actions were appropriate and thorough, with the exception that the licensee did not address a poor drain trap configuration associated with main steam  !

piping on the 8-foot level of the turbine building, in addition, the inspectors identified a (

failed, nonsafety-related pipe support which licensee staff had overlooked during the {

event assessment and routine operator rounds. (Section O2.2) /

. The Manager's Supervisory Staff (MSS), which compose the station's onsite review i

committee, appropriately identified weaknesses with a proposed safety evaluation (SE)

for a new fuelload. Because of the incomplete staff work associated with the SE, the MSS actively participated in development of the final product. Although no problems were identified with the SE reviewed by the inspectors, the practice of having MSS l members provide so much input into SEs had the potential to affect the objectivity of the MSS. (Section 07.2)

. Senior plant and corporate management continued to demonstrate a otrong commitment to improving performance. Examples of this commitment included increases in plant staffing, additional management changes, and extending the U1R24 outage to allow

! installation of many modifications. Significant progress was made during this outage in i addressing main control board wire separation issues, electrical separation issues

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associated with safe shutdown equipment, removal of partiallength control rod drive

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housings, and inspection arid repair of service water system piping and pipe supports in containment. (Section 07.3)

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. The relatively high level of management attention which was required to ensure safe and effective completion of outage activities, coupled with the large backlog of corrective action and modification work, prevented resource allocation for necessary program improvement initiatives, effectively ~ prevented immediate resolution of all degraded and nonconforming plant conditions, and diluted plant management's ability to drive committed improvements to completion. As a result, the licensee was evaluating a significant restructuring of commitment dates for program and hardware reviews and upgrades. Senior licensee management indicated that any change in docketed commitment dates would be formally submitted to the NRC in writing. (Section O7.3)

Maintenance

. A senior reactor operator with no concurrent duties was assigned to directly supervise the reactor operators who performed the Unit 2 bi-weekly rod exercise test. The operators appropriately exercised self-checking techniques, independent verification, and three-way communications. The senior reactor operator maintained a good overview and ensured control room distractions were minimized. The operators utilized the surveillance procedure, and signed off each step as it was performed, consistent with licensee procedural controls. (Section M1.1)

. The inspectors identified that minor damage had occurred to instrument tubing ard cables located near work areas within the Unit 1 reactor coolant system loop cubicle The licensee initiated corrective actions for the specific damage and committed to review the process controls for preventing such damage. No other conduct of maintenance problems were identified in the refueling area during the inspection perio (Section M1.2)

. Annunciator alarms for flow spiking of the seal water return from the reactor coolant pumps were a distraction to the operators. The alarms began when the gas stripper system was taken out-of-service to address material condition problems; however, the licensee was unsure of why the out-of-service gas stripper caused the spikin (Section M2.1)

. Inadequate maintenance for containment upper personnel hatch latch and interlock mechanism components was determined to be the cause of repetitive surveillance test failures. (Section M2.2)

. The licensee responded promptly to two 10 CFR Part 21 vendor notifications, resulting in the removal of defective or degraded components from the auxiliary feedwater system and tl'c emergency dieses generators. (Sections M8.6 and M8.7)

Enaineerina

. A draft SE, presented to the MSS for approval, for a slightly longer than normal operating cycle core load was conceptually adequate, but lacked appropriate reference to allowable plant conditions. The design engineers who developed the SE had not proposed appropriate administrative controls for ensuring that the plant conditions covered (allowed) by the proposed SE were maintained while the SE was in effec (Section 07.2)

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• A licensee quality assurance report was effective in that it identified programmatic '

weaknesses in the licensee's response to Generic Letter 96-01," Testing of Safety-Related Logic Circuits," specific concerns with the licensee's current position on some Generic Letter 96-01 issues, and a failure to perform some aspects of a required surveillance. (Section E4.1)

. The licensee engineering and regulatory affairs staff who responded to a licensee quality assurance finding of inadequate surveillance testing for refueling system interlocks failed -

to apply sufficient rigor in their assessment of the issue. Inspector intervention was required to ensure that the Technical Specification-required test was performed appropriately One violation of Technical Specifications was identified. (Section E4.1)

Plant Support

+ . A self-assessment of the chemistry program identified opportunities for improvement (assessment findings)in the areas of chemistry system material condition and design control, chemistry procedures, and chemistry management's effectiveness in reinforcing standards and expectations. The recommendations (corrective actions) contained in the self assessment were appropriate to address these assessment findings. The findings were consistent with observations made by the inspectors during chemistry system walkdowns and routine review of condition reports. (Section R7.1)

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Report Details Summarv of Plant Status During this inspection period, Unit 1 remained in the Cycle 24 refueling (U1R24) outage. Unit 2 operated at 100 percent power throughout the period.

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l. Operations l 01 Conduct of Operations 0 General Comments (Inspection Procedure (IP) 71707)

The inspectors conducted frequent reviews of ongoing plant operations. During this inspection period, the inspectors observed Unit 1 and Unit 2 control room shift turnovers and control room operation .2 Operator Not Aware of System Configuration (IP 71707)

An expected volume control tank (VCT) high-level automatic divert signal was generated during a planned reactor coolant system (RCS) dilution for Unit 2 on May 12,1998. The signal caused the level divert valve to realign, resulting in the letdown return flow path being changed from the VCT to the chemical and volume control system (CVCS) hold-up tank. However, the Unit 2 reactor (control) operator noted abnormalletdown flow and pressure indications following the repositioning of the divert valve. The operator then realigned the level divert valve back to the VCT, and the letdown return flow and pressure indications returned to norma While investigating the abnormal letdown flow and pressure indications, the operating crew determined that the Unit 2 letdown divert flow path was isolated from the hold-up tank. This had been done by the previous shift as part of the Unit 1 CVCS restoration valve lineup for filling and venting following refueling outage maintenance activities. The facts that the Unit 1 CVCS recovery was ongoing and that a partial valve alignment had been performed were discussed during the pre-shift turnover; however, the Unit 2 control operator was not aware of the impact of the Unit 1 CVCS fill-and-vent valve alignment on the Unit 2 CVCS. The licensee documented this event in Condition Report (CR)98-192 This event was of minor safety significance; however, it indicated an inadequate turnover between operating crews. Improved supervisory oversight could also have identified the effect that the evolution on the shutdown Unit 1 would have on the operating Unit 2. The procedure used to restore the Unit 1 CVCS was also weak in that it did not identify the cross-unit impact of the fill-and-vent valve line-u .3 Areas for improvement in the Conduct of Operations Inspection Scope (IP 71707 & IP 40500)

The inspectors attended the exit meeting for a licensee self-assessment of the operations )

department, on May 15,199 .

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I Observations and Findinas The self-assessment team consisted of three licensed senior reactor operators (SROs)

from other nuclear facilities, and several plant employees, including a licensed reactor operator (RO). The bases for the team's assessment were industry guidelines for the conduct of operations, plant configuration control, and requalification training. The team spent one week interviewing operations personnel, observing control room and non-licensed operator activities, and reviewing operations procedures and program The self assessment team reached the overall conclusion that the operations department was adequate; but, that it needed to make improvements to reach current industry standards. The team stated that the previous rate of improvement had slowed dramatically and attributed this to the lack of a self-perceived need for further improvement among the operators. In addition, operations personnel were unable to focus clearly on improvements because of the work load from the long and frequent plant shutdowns and outage The team recognized several areas of strength, including good control room communications, good use of procedure feedback systems, and strong leadership from the operations manager. The areas identified for improvement consisted of the following:

operations staffing, quality of procedures, management standards and expectations, operations department facilities, plant configuration control, and operations personnel sense of plant ownership and authority. These observations were consistent with those being made by the inspector Plant management at the exit meeting acknowledged and concurred with the self-assessment team findings. The plant manager and operations department manager stated that they recognized that improvement was needed to reach industry standards, and emphasized that numerous improvement initiatives were ongoing to reach those standard Conclusions The licensee performed an effective self-assessment of the conduct of operations. The use of industry peers was highly beneficial. The need for improvement to bring the operations department up to current industry standards was identified. The self-assessment findings were consistent with the inspectors' observation O2 Operational Status of Facilities and Equipment O Independent Verification of Safety System Operability (IP 71707) inspection Scope (IP 71707)

The inspectors performed a safety system walkdown using the guidance of IP 7170 _ - _ _ _ _ _ _ _ _ _ _ _ _

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b. Observations and Findinas On May 12,1998, the inspectors performed a walkdown of the Unit 2 containment spray system. The containment spray system consists of two trains of pumps and valves. The pump for each train takes suction from the refueling water storage tank (RWST) and discharges to the containment spray ring header. A small portion of the pump discharge is recirculated through an eductor, back to the pump inlet. The eductor draws a sodium hydroxide solution from the spray additive tank into the containment spray system. The flow path from the spray additive tank to the train specific eductors includes a common manual outlet valve from the tank (SI-831 A), parallel air-operated valves which open on a containment spray initiation signal, another common manual isolation valve (SI-8318),

and then separation for the eductors. The sodium hydroxide performs the safe *y-related function of removing radioactive iodine from the post-accident containment atmospher The Final Safety Analysis Report (FSAR) stated that approximately 14 gallons per minute of sodium hydroxide solution would be drawn into the containment spray system. Overall, the Unit 2 containment spray system appeared to be properly configured and maintaine The inspectors observed that the Unit 2 spray additive tank was at a slight vacuu Section 6.4 of the FSAR stated that a nitrogen blanket was maintained in the spray additive tank, and implied that the tank was normally slightly pressurized. The inspectors questioned whether maintaining the tank at a slight vacuum was consistent with the conditions under which flow from the tank was tested. The inspectors then identified that surveillance testing of flow from the spray additive tank was not performed as part of the system operability surveillance test The system engineer wrote CR 98-1940 to evaluate the fact that the spray additive tank was not being maintained at a slightly positive pressure, as implied by the FSAR. The system engineer subsequently informed the inspectors that the spray additive tank was maintained within a pressure range which was based on ensuring adequate flow from the sealed tank during the injection phase of containment spray, and that a slightly negative '

pressure was acceptable. During system flow tests, water from the RWST was used to verify that a check valve at the train specific eductors opened to pass full flow. A test line supplied water from the RWST to the spray additive tank side of the eductors downstream of SI-831B. The containment spray system engineer stated that the inservice testic,9 (IST) system engineer relied upon this test to track operability of the eductors, in reviewing the information provided by the system engineer, the inspectors noted that I the IST surveillance procedure and test acceptance criteria handbook both stated that the test using RWST water through the eductors was performed to validate operation of the check valve, and neither document mentioned the need to monitor eductor operation, in addition, the RWST pressure was significantly higher than the spray additive tank pressure, so the test line-up did not simulate accident conditions for purposes of

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evaluating the operation of the eductors or for monitoring flow resistance through the spray additive lines. The inspectors noted the following uncertainties associated with this test configuration:

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• acceptance test results and acceptance criteria reported to be for the eductors was not labeled as such in the controlled test procedure, and a the test configurations did not verify that an operable flow path existed through two manual valves and two air-operated isclation valves between the spray additive tank and the RWST (for blockage or increased resistance to flow).

Section 6.4.1 of the FSAR contained the following statement:

Testina of Containment Spray Systems

"A capability shall be provided to the extent practical to test periodically the delivery capability of the containment spray systems at a position as close to the spray nozzles as is practical GDC [ General Design Criterion) 60."

The inspectors concluded that the FSAR indicated that the licensee's test program was expected to demonstrate the ability of the containment spray system to perform its safety function by use of appropriate flow tests. The inspectors discussed the testing concerns with the licensee. The licensee was unable to provide any documentation which exempted the spray additive outlet pipe flow path from testing, but subsequently completed an acceptable operability determinatio This issue is being treated as an Unresolved item (URI 50-266/98009-01(DRP);

50-301/98009-01(DRP)) at the request of the Office of Nuclear Reactor Regulation (NRR). This URI will remain open until NRR has reviewed the following:

• whether the current licensing basis and NRC regulations require that the licensee test the spray additive portion of the containment spray system, and

• whether there are minimum acceptable test configuration requirements that the licensee is required to satisfy while performing required testin Conclusions The inspectors determined that the containment spray system was aligned correctly during a safety system walkdown. The inspectors noted many discrepancies regarding the manner in which the containment spray system is tested. This issue is considered a URI pending a review by NR .2 Follow-up to a Unit 2 Waterhammer Event Inspection Scope (IP 71707)

The inspectors reviewed the operations department root cause evaluation (RCE) for a waterharnmer event documented in inspection Report (IR) No. 50-266/98003(DRP);

50-301/98003(DRP).

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A Observations and Findinas The waterhammer event occurred on February 7,1998, during a Unit 2 start-up. The level"A(highest in the licensee's corrective action program) RCE, RCE 98-026, concluded that the wcterhammer was caused by a combination of an inadequate drain path and excessive steam line warm-up rates. These conditions were the result of a poor procedure and poor decisions by the operating cre The inspectors concurred with most of the findings of the RCE and noted that the recommendations thoroughly addressed the issues identified in the RCE. However, the inspectors were concerned that the RCE did not address the specifics of the main steam system pipe displacement on the 8-foot level of the Unit 2 turbine building. Based on the damage to two energy absorbers associated with the steam line to the south condenser steam dump header, the inspectors concluded that the western end of an east-west section of pipe had been displaced to the north. The inspectors walked this section of main steam piping down and noted that the drain traps for significant lengths of the 16-inch diameter pipe were configured so that drained condensate had to be forced upward to the atmospheric relief tank during system heat-up. A similar design condition for the drain traps associated with main steam piping between the main steam isolation valves and the main steam non-return check valves had been addressed by adding connections to floor drains via Valves MS-120A and MS-123A. While the RCE adequately addressed problems with the operation of MS-120A and MS-123A, it failed to note that these two valves could not provide a drain path for condensation downstream, and lower than piping near the non-return check valves. The inspectors were, therefore, concerned that a significant impact load had been applied to main steam piping on the 8-foot level of the turbine building because of an inadequate drain configuration that was not addressed by the licensee's RCE. The licensee was evaluating the inspectors'

concern at the end of the inspection period. The results of that evaluation will be reviewed by the inspectors. This item is considered an Inspection Followup Item (IFl 50-266/98009-07(DRP); 50-301/98009-07(DRP)).

While performing a routine walkdown of the Unit 2 turbine building on May 4,1998, the inspectors observed that a pipe hanger which was supposed to support a 1%-inch stainless steal pipe had broken free from the wall and was hanging from the pipe. The pipe is a Seismic Class til, nonsafety-related, common drain line to the atmospheric blowoff tank from condensate traps for the main steam lines upstream of the main steam isolation valves. This observation was documented in CR 98-1806. Although it was not clear that the failure of the pipe support was caused by the February 7,1998, waterhammer event, the damage should have been identified during the RCE or during routine operator round Conclusions The licensee's RCE for a February 7,1998, waterhammer event attributed the event to a poor operating procedure and poor decisions by the operating crew. The corrective actions were appropriate and thorough, with the exception that the licensee did not address a poor drain trap configuration associated with main steam piping on the 8-foot level of the turbine building. In addition, the inspectors identified a failed, nonsafety-related pipe support which licensee staff had overlooked during event assessment and routine operator round _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ - - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -

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w 07' Quality Assurance in Operations 07.1 Off-Site Review Committee (OSRC)

The inspectors attended portions of the OSRC meeting held on May 5 and 6,199 The OSRC meeting was conducted in accordance with the requirements of Technical Specification (T/S) 15.6.5.2, "Off-Site Review Committee." The safety evaluation (SE)

review subcommittee of the OSRC reported a total of 78 SEs that had been reviewed in the last quarter. The subcommittee verified that the proposed changes, evaluated in the 78 SEs, did not constitute unreviewed safety questions (USOs). Three were_ determined to have ins"fficient discussion of the proposed change to allow the committee to independe ./ determine that a USO did not exist. These were subsequently rewritten to provide the necessary detail. Eight SEs contained various administrative flaws. During the portions of the OSRC meeting when plant staff were giving presentations, the OSRC members were actively engaged and involved in the presentation .2 Manaaer's Supervisory Staff (MSS) Inspection Scope (IP 71707)

The inspectors attended a meeting of the MSS (the onsite review committee) to assess performance of the MSS and to review the analysis associated with a slightly longer than normat operating cycle core load, Observations and Findinas The inspectors attended portions of the MSS meeting held on May 13,1998. The MSS meeting was conducted in accordance with the requirements of T/S 15.6.5.1, " Manager's Supervisory Staff." The inspectors independently verified that a quorum was present, as required by T/S 15.6.5.1.2 and T/S 15.6.5. The purpose of the MSS meeting was to review a draft SE for core reload during the U1R24 refueling. The draft SE was presented by engineers from the corporate design engineering group. The proposed U1R24 core reload differed from previous core loads in that a higher fuel enrichment was being utilized so as to allow a longer operating cycl Based upon the data presented, the inspectors concluded that reloading the core using the proposed fuel mix would not be a USQ. However, the scope of the SE was limited to certain plant conditions associated with refueling the core. Additional calculations were being performed to support placing the unit in cold shutdown, and then hot shutdown through power operation, with the new fuel load. Although the design engineers were aware of these limitations, the correct method of defining limited (allowed) plant conditions had not been determined prior to bringing the draft SE to the MS Additionally, the proper administrative controls to ensure that the plant did not leave the allowed plant conditions covered by the SE prior to completion of additional SEs had not been determined prior to the MSS meeting. As a result of the incomplete staff work, the MSS members established the definition of allowed plant conditions and established the administrative controls for ensuring that the plant did not exit the allowed plant conditions until such time as the subsequent SEs were completed. The inspectors considered the

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quality of the review performed by the MSS to be good, but noted that the objectivity of the MSS members could be affected by the fact that thcy essentially developed

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a significant portions of the SE. The inspectors were also concerned that the incomp?ete staff work could have been at least partially attributable to a perceived schedule pressure to have an SE in place to support core reloa Conclusions A draft SE for a slightly longer than normal operating cycle core load was conceptually adequate, but lacked appropriate reference to allowable plant conditions. The design engince., who developed the SE did not propose appropriate administrative controls for ensuring that the plant conditions covered (allowed) by the proposed SE were maintained while the SE was in effec The MSS appropriately identified weaknesses with the proposed SE for the new fuel loa Because of the incomplete staff work associated with the SE, the MSS actively participated in development of the final product. Although no problems were identified with the SE reviewed by the inspectors, the practice of having MSS members provide so much input into SEs had the potential to affect the objectivity of the MSS revie .3 Plant Performance in Implementing Lona-Term improvement Initiatives Senior plant and corporate management continued to demonstrate a strong commitment to improving performance. Examples of this commitment included increases in plant staffing, additional management changes, and extending the U1R24 outage to allow installation of many modifications. Significant progress was made during this outage in addressing main control board wire separation issues, electrical separation issues associated with safe shutdown equipment, removal of partiallength control rod drive housings, and inspection and repair of service water (SW) system piping and pipe supports in containment. However, the relatively high management attention which was required to ensure safe and effective completion of outage activities, coupled with the large backlog of corrective action and modification work, prevented resource allocation for necessary program improvement initiatives, effectively prevented immediate resolution of all degraded and nonconforming plant conditions, and diluted plant management's ability to drive committed improvements to completion. As a result, the licensee was evaluating a significant restructuring of commitment dates for program and hardware reviews and upgrades. Senior licensee management indicated that any change in docketed commitment dates would be formally submitted to the NRC in writin Miscellaneous Operations issues 0 (Closed) Violations (VIOs) 01013. 01023. and 01033 from Enforcement Action (EA)

96-273: irs No. 50-266/96006(DRP): 50-301/96006(DRP) and No. 50-266/96007(DRP):

50-301/96007(DRP): Violations of Criterion V," Instructions, Procedures, and Drawings,"

of 10 CFR Part 50, Appendix B, for watching training videotapes in the control room (VIO 01013), an RO briefly leaving the control panels without a relief (VIO 01023), and an RO not responding promptly to an alarm (VIO 01033). The licensee committed to extensive corrective actions in its response to the Notice of Violation (NOV) in a Jetter dated January 31,1997, including a revision and re-emphasis of the procedural standards for control room conduct. Over the past year, NRC inspectors have reviewed the implementation of these standards, described in Operations Manual (OM) 1.1, " Conduct of Operations." This review has indicated that the conduct of control room activities has l

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greatly improved (for example, see irs 50-266/97006(DRP); 50-301/97006(DRP),

50-266/97010(DRS); 50-301/97010(DRS), 50-266/9702C(DRP); 50-301/97020(DRP),

50-266/97023(DRS); 50-301/97023(DRS), and 50-266/97026(DRP);

50-301/97026(DRP)). A recent violation for an RO leaving the main control panel area (of the defueled Unit 1) was documented in IR 50-266/98006(DRP); SU-366/98006(DRP), but was considered by the inspectors to be an isolated problem, not indicative of the failure of corsctive actions for VIO 01023, which occurred about 2 years ago in July 1996, and involved a reactor that was at full powe .2 (Closed) VIO 01043 from EA 96-273: irs 50-266/96006(DRP): 50-301/96006(DRP) and 50-266/96007(DRP): 50-301/96007(DRP): Violation of Criterion V," Instructions, Procedures, and Drawings," of 10 CFR Part 50, Appendix B, for the duty technical advisor (DTA) leaving the site while on duty. The licensee committed to extensive corrective actions in its response to the NOV in a letter dated January 31,1997, including a revision of the procedural position description of the DTA and a relocation of the DTA'

sleeping quarters from outside the protected area to inside. These corrective actions have been effective in preventing a reoccurrence of the violation. Recently, the inspectors identified an unrelated concern with the working hours of some DTAs. This issue is discussed in IR 50-266/98006(DRP); 50-301/98006(DRP).

08.3 (Closed) Licensee Event Report (LER) 50-266/98011: 50-301/98011: Missed T/S Surveillance of a Radiation Monitor Prior to Discharge. This LER described the discovery that the required T/S surveillance test of a radiation monitor had not been completed prior to initiation of a radioactive waste discharge on February 12,1998. The safety significance of this missed surveillance test was minor, but illustrated the breakdown of several barriers within the operations department intended to prevent instances like this from occurring including control room command and control of shift activities, operator turnover, procedural adherence, and operator self-verification. This licensee-identified and corrected failure to verify the radiation monitor's operability prior to the radioacti waste discharge as required by T/S 15.7.4 is considered a non-cited violation (NCV 50-266/98009-02(DRP); 50-301/98009-02(DRP)), consistent with Section Vll.B.1 of the NRC Enforcement Polic .4 (Closed) LER 50-266/98002: 50-301/98002: Failure of the High Voltage Station Auxiliary Transformer. This issue was fully described in IR 50-266/97022(DRP);

50-301/97022(DRP). The LER was ieviewed using the guidance of IP 92700 and determined to be accurate and complet . Maintenance I

M1 Conduct of Maintenance M1.1 Performance of Surveillance Test (IP 61726)

On May 1,1998, the inspectors witnessed the operations department's performance of J Technical Specification Test (TS) 6, "Bi-Weekly Rod Exercise Test Unit 2," Revision 2 An SRO with no concurrent duties was assigned to directly supervise the ROs who performed the reactivity manipulations. The operators appropriately exercised self-checking techniques, independent verification, and three-way communications. The SRO I

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maintained a good overview and ensured control room distractions were minimized. The operators utilized the surveillance procedure, and signed off each step as it was perforrned, consistent with licensee procedural control The inspectors noted that TS 6 did not require independent verification that the rod control selector switch was restored to the " automatic" position. The licensee's normal practice was to specify independent verificat?on for "as left" switch positions when recovering from surveillance test alignments. The operations department's lead person for procedure upgrades indicated that independent verification would be included in the next revision of TS M1.2 Conduct of Outaae-Related Maintenance Activities a. < Inspection Scope The inspectors observed maintenance activities associated with the U1R24 outage using the guidance of IP 6270 Observations and Findinas

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The licensee's new work control process contained improved mechanisms for ensuring that suitable plant conditions existed prior to initiating maintenance or modification wor A list of " protected equipment" was maintained to ensure that safety functions were not lost by taking redundant system trains or equipment out-of-service at the same time. No examples of work being performed during inappropriate or non-conservative plant conditions were identifie The inspectors observed parts of the maintenance activities and reviewed portions of the procedures associated with the modification of SW pump P-32E. Most of the specific performance issues identified in IR 50-266/98003(DRP); 50-301/98003(DRP) appeared to have been addressed. No significant problems were identified. However, late in the inspection period, the licensee initiated CR 981999 to address some attention-to-detail problems with the close-out of the P-32E work packag The inspectors also performed walkdowns of the maintenance activities in the Unit 1 RCS loops. No significant problems were identified; however, the inspectors noted that some instrumentation tubing and cables located in the vicinity of maintenance activities .

appeared to be deformed or frayed, respectively. Additionally, some small-diameter CVCS system and component cooling water (CCW) system pipe associated with removed portions of a reactor coolant pump (RCP) were supported by ropes. This resulted in a significant amount of " play"in the tied-off CVCS and CCW pipe, a condition which the inspectors considered to be a poor work practice. The inspectors discussed the specific observations with the RCS system engineer. The inspectors and the RCS system engineer also discussed whether the processes for planning modification and

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repair work included adequate provisions for planning the removal of interferences in the l- areas adjacent to the work. The RCS system engineer initiated CR 98-1891 to document operability determinations for the specific conditions observed, and to initiate corrective actions for the failure of existing work planning and control processes to ensure that plant equipment was not damaged while maintenance activities were performed in adjacent area !

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Licensee personnelidentified that poor work practices, such as the use of wire brushes on steam piping above the steam generators, had resulted in the spread of foreign materialinto the refueling cavity and within the RCS cubicles. These foreign material exclusion (FME) concerns were documented in CRs 98-1921 and 98-1941, and corrective actions were initiated to address the specific issue Conclusions The inspectors identified that minor damage had occurred to instrument tubing and cables located near work areas within the Unit 1 RCS loop cubicles. The licensee initiated corrective actions for the specific damage and committed to review the process controls for preventing such damage. No other conduct of maintenance problems were identified in the refueling area during the perio M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Unit 2 RCP "A" Seal Flow Spikina On May 7,1998, the Unit 2 gas stripper (GS) system was removed from service due to increased circulating pump sealleakage. The Unit 2 "A" RCP,2P-1 A, subsequently experienced periodic seal flow perturbations, as indicated by a control room annunciator alarm. The annunciator alarm came in about every 45 minutes for the next 6 days until the GS was returned to service on May 13,1998. Based upon the observed operator response, the inspectors concluded that this alarm had become a distraction to the control room operator Corrective maintenance was delayed because GS system valves for the pump did not provide adequate isolation during an initial valve line-up. The leakage delayed the start of repalis for two days while alternatives to the planned protective boundary were contemplated. Operators subsequently established adequate isolation by cycling the GS system isolation valves several times. Following the repair of the GS system circulating pump, the mechanical maintenance personnel spent two days attempting to get proper shaft bearing alignment because of excessive wear in the shaft bearing housings. A temporary modification was eventually made to the bearing housings because of the non-availability of replacement parts. The return-to-service of the GS was delayed an additional day by equipment related problems with the cryogenic portion of the GS syste A previous example of repetitive RCP seal flow annunciator alarms causing a distraction to the operators was documented in IR 50-266/98003(DRP); 50-301/98003(DRP). The licensee technical staff had yet to determine the reason why RCP seat flow spiking occurred with the GS system out-of-servic M2.2 Material Condition of Containment Personnel Hatch interlocks and Seals Inspection Scope UP 62707)

The inspectors monitored the licensee's response to continued problems with the operation of the containment personnel hatches.

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b. Observations and Findinas Recurring examples of the Unit 1 containment personnel hatch inner door failing its leakage tests were documented in irs 50-266/96012(DRP); 50-301/96012(DRP),

50-266/96015(DRP); 50-301/96015(DRP), and 50-266/96019(DRP); 50-301/96019(DRP).

Several personnel hatch problems also occurred during 1997, and were tracked in the licensee's corrective action program. During this inspection period, the Unit 2 upper personnel hatch inner door failed a leak test, and the interlock for the inner and outer doors was inoperable. The system engineer determined that the inner door seal was lost when the interlock mechanism was adjusted, because of wear in the keyways on the door operating levers. The condition of the keyways had not been addressed in the existing maintenance procedures, and previous corrective action reviews had not identified this wear issue. The system engineer informed the inspectors that maintenance procedures were being revised to address this problem. The Unit 1 hatch system was being rebuilt during the U1R24 outag The licensee recognized and appropriately resolved the containment interlock keyway wear issue; however, the inspectors were concerned that the problem had existed for an extersded period of time before the apparent root cause was identified. The inspectors initiated a review of the licensee's handling of the containment hatches with regard to the requirements of 10 CFR 50.65, " Maintenance Rule," at the conclusion of the inspection period. The inspectors identified the following issues which were not resolved during this period:

1) The system engineer had not classified any of the hatch surveillance test failures as Maintenance Rule " functional failures" or " maintenance preventable functional failures (MPFF;," because a safety-related function had not been los The inspectors were concerned that a loss of operability for a T/S-required component was not considered a functional failure under the licensee's maintenance rule progra ) The inspectors obtained a list of all CRs which identified MPFFs for 199 Only two records were found, both for the CVCS system, which were flagged as MPFFS, and the text for one of these records indicated that the condition had been evaluated as not being an MPFF. The inspectors noted that a single MPFF in a 5-month period for a plant of Point Beach's vintage was lower than expecte ) The inspectors verified that the system engineer for the CCW system recognized the failure of Valve 2CC-745 (see IR 50-266/98006(DRP);

50-301/98006(DRP)) as being a MPFF; however, this MPFF was not on the list of MPFF CRs for 1998. Licensee procedures appeared to require that all MPFFs be documented on CR ) A system engineer for the SW system told the inspectors that MPFFs were not  ;

considered when determining "(a)1" systems (as defined in 10 CFR 50.65). The licensee procedures required that systems with repetitive MPFFs be moved to the

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Followup inspection effort to determine the validity and regulatory significance of these concerns will be tracked as an inspection Follow-up Item (IFl 50-266/98009-03(DRS);

50-301/98009-03(DRS)) pending a review by an NRC Division of Reactor Safety, Maintenance Rule specialist inspecto Conclusion,_s inadequate maintenance for containment upper personnel hatch latch and interlock mechanism components was determined to be the cause of repetitive surveillance test failures. While the response to this long-standing material condition problem was adequate during this report period, the inspectors were concerned with how the surveillance test failures were handled relative to the licensee's Maintenance Rule progra M4 Maintenance Staff Knowledge and Performance M FME Controls flP 62707)

Inconsistent application of FME controls had been a recurring issue during previous inspection periods. During this period, the inspectors noted improvement in the application cf FME standards, but also noted additional examples of inconsistent application and maintenance worker attention-to-detail. The licensee set up a display on proper and improper FME techniques near the main cafeteria. The posting of FME areas improved, and the delineation of FME control boundaries improved during the inspection period. Improvements in the use of practical FME area boundaries were also noted. For example, an FME boundary for maintenance work on Valve 2CC-745 during a previous inspection period included e.e entire mezzanine area in which the valve was locate During this period, the FME boundary for wc* on Valve 1CC-745 was limited to an appropriately sized work area. One example of a poor application of FME standards observed by the inspectors involved the use of a frayed FME cover for the open internals of Valve 1CC-745. No material was observed to enter the valve. A second example involved loose bolts associated with electrical work on the 1P-1 A RCP inside the missile shield. These bolts were observed in a location where they might have been able to fall into the RCP oil collection sys, tem. The workers were on a break at the time of this observatio M8 . Miscellaneous Maintenance issues M8.1 (Closed) LER 50-266/96002: Auxiliary Feedwater (AFW) Pump Discharge Valves Found Shut When (Reactor Was) Critical. This issue involved inappropriate aspects of the completion of post-maintenance and surveillance testing of a turbine-driven AFW pum These problems were discussed in detailin NRC irs 50-266/96006(DRP);

50-301/96006(DRP) and 50-266/96007(DRP); 50-301/96007(DRP). On December 3,1996, an NOV was issued for these problems (EA 96-273) and included, for l the AFW pump issue, a violation of T/S 15.3.4.A.2 (VIO No. 02013) and four violations of Criterion V," Instructions, Drawings, and Procedures"(VIOs No. 02023,02033,02043,

and 02053). The licensee committed to extensive corrective actions in response to the NOV, including counseling and training of personnel, a detailed review of the post-maintenance testing and equipment return-to-service processes, and revision of procedures and checklist l

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Additional programmatic problems, including a significant breakdown of the corrective actions program, were identified during an NRC inspection in late 1996 (IR 50-266/96018(DRP); 50-301/96018(DRP)). In view of the problems identified in the three 1996 inspections, the licensee undertook an extensive improvement effort. In mid-1997, NRC inspectors reviewed the results of these efforts and determined that committed corrective actions had been implemented (irs 50-266/97010(DRS);

50-301/97010(DRS) and 50-266/97023(DRS); 50-301/97023(DRS)). The inspectors reviewed corrective actions specific to the AFW issue during the current inspection period. The specific actions and the, broad-scope actions taken for the programmatic issues appeared adequate to preclude a repeat of both the original violation of T/S 15.3.4.A.2 for the AFW pump valves and the four associated Criterion V violations. This completes NRC review of LER 50-266/96002 and the five violations associated with the AFW system as cited in the NOV issued January 31,1997, for EA 96-27 M8.2 (Closed) VIOs 02013. 02023. 02033. 02043. and 02053 from EA 96-273:

irs 50-266/96006(DRP): 50-301/96006(DRP) and 50-266/96007(DRP):

50-301/96007(DRP): AFW Pump Discharge Valves Found Shut When [ Reactor Was)

Critical. These issues are discussed in Section M M8.3 (Closed) LER 50-266/96012: 50-301/96012: EDG [ Emergency Diesel Generator) Fuel Oil System Tests Not Performed in Accordance With T/S. This issue was discussed in IR 50-266/96018(DRP); 50-301/96018(DRP) and considered with other problems for escalated enforcement action. On August 8,1997, an NOV was issued (EA 97-075)

which included a violation of T/S 15.4.6.A.5 for the failure to test the automatic operation feature of the fuel oil transfer systems of the four EDGs. The licensee responded to the NOV in a letter dated October 10,1997. Corrective actions for programmatic aspects of EA 97-075 were reviewed in irs 50-266/97010(DRS); 50-301/97010(DRS) and 50-266/97023(DRS); 50-301/97023(DRS). During this inspection period, the inspectors reviewed the procedure revisions completed as corrective actions specific to the EDG fuel oil system testing problem. These reviews indicated that adequate corrective actions had been taken to address the issues described in the LER and the violation conceming the testing of the EDG fuel oil syste M8.4 (Closed) Escalated Enforcement item (EEI) 50-266/96018-11(DRP):

50-301/9601811(DRP): EDG Fuel Oil System Tests Not Performed in Accordance With T/S. This issue is discussed in Section M M8.5 (Closed) eel 50-266/96018-10(DRP): 50-301/96018-10(DRP): Failure to start all associated safety-related loads during annual EDG testing initiated by a loss of altemating current followed by a simulated safety injection signal. On August 8,1997, an NOV was issued (EA 97-075) which included a violation of T/S 15.4.6.A.2 for the failure to test the EDGs with all associated safety-related loads. The licensee revised the procedures for the annual tests to include the appropriate safety-related loads and conducted the tests as required. The procedure revisions and the conduct of the tests using the revised procedures were reviewed by NRC inspectors and were cor Jidered to be acceptable as documented in irs 50-266/97010(DRS); 50-301/97010(DRS) and l 50-266/97013(DRP); 50-301/97013(DRP).

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M8.6 (Closed) LER 50-266/98007: Defective Turbine-Driven AFW Pump Low Suction Trip Time Delay Relay. This LER described the licensee's response to a 10 CFR Part 21 notification concerning potentially degraded or defective relays. The licensee responded promptly to the Part 21 report and determined that potentially defective relays were installed in the AFW system. Allinstalled relays were inspected and a single defective relay was found installed on the Unit 1 turbine-driven AFW pump. The defective relay was replaced within the applicable T/S limiting condition for operation allowed outage time. The licensee response to this issue was considered to be goo M8.7 (Closed) LER 50-266/98008:50-301/98008: EDG Air Start Motor Solenoid Valves Found with Springs Not in Accordance with Design. This LER described the licensee's response to a 10 CFR Part 21 notification concerning defective air system solenoid valves. The licensee responded promptly to the Part 21 report and determined that three of the four EDGs were affected by the use of these solenoid valves. The three EDGs were subsequently declared inoperable. The licensee then initiated action to return a second EDG to an operable status by replacing the solenoid. The completion of this maintenance activity was delayed by problems with the performance of work by maintenance personnel, but the T/S limiting condition for operation allowed outage time was not exceede M8.8 (Closed) IFl 50-266/96019-04(DRP): Containment inner PersonneI Hatch Test Failure The cause and corrective actions for the test failures are discussed in Section M2.2 of this repor Ill. Ennineerina E4 Engineering Staff Knowledge and Performance E Inadeouate Response to Quality Assurance (QA) Department Findina Inspection Scope (IP 37551 & IP 40500)

The inspectors reviewed QA Surveillance Report S-P-98-02, "T/S Line items," dated April 13,1998, and the documented follow-up corrective action Observations and Findinas l Quality Assurance Report S-P-98-02 documented the review of the implementation of T/S line item surveillance requirements. Fifty-nine refueling outage frequency line items were

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reviewed, and two " deficiencies"(licensee terminology) were identified. One of the two deficiencies was administrative in nature, and the second dealt with an apparent failure to test three of the seven required refueling system interlocks. These deficiencies were documented in Quality Condition Report (OCR) 98-0099 and QCR 98-0110, respectivel The inspectors reviewed the corrective action records for QCR 98-0110 to ensure that adequate tests of fuel handling equipment would be completed prior to core reload during the U1R24 outage. Line item 14 of T/S Table 15.4.1-2, specified that refueling system interlocks be tested each refueling outage to verify functionality. On April 22,1998, in describing completed corrective actions, QCR 98-0110 stated that three new tests needed to be performed to satisfy Line item 14 of T/S Table 15.4.1-2, but that the as-

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found condition was not reportable. The corrective action record stated that procedure changes to incorporate the three new tests were a U1R24 core reload restraint. Based q upon an independent review of the applicable test procedures, the inspectors identified the following concems with the corrective actions descnbed above:

a) Only two of six required manipulator bridge, trolley, and winch drive interlock l com'oinations were tested. This was not identified by the license )

i b) The test for the manipulator winch gripper safety interlock only verified that a relay !

could be heard actuating, it did not verify the associated contact function, and was therefore inadequate. This was not identified by the license c) Some required tests were being performed in accordance with Instrumentation &

Control Procedure 10.9, " Routine Maintenance Procedure - Dillon Load Cell Installation," yet this procedure was marked as being "not T/S related." This was not identified by the license d) The tests associated with T/S Table 15.4.12, Line item 14 had not been comprehensive, yet licensee staff had concluded that this was not a violation of the T/S, and therefore not reportable under 10 CFR 50.73(a)(2)(B). This appeared to be an inappropriate conclusio e) Corrective actions were appropriately identified as being core reload restraints; however, the inspectors determined that these restraints were not reflected in the outage planning schedule or in the corrective action / work control process list of core load restraint The inspectors discussed these issues with appropriate plant staff. Each of the inspectors' concerns were acknowledged as being accurate, and plant staff initiated additional corrective actions to address these issues. The failure to test all required refueling system interlocks was a violation (50-266/98009-04(DRP);

50-301/98009-04(DRP)) of T/S 15.4.1.B., which requires that equipment and sampling tests be conducted as detailed in T/S Table 15.4.12. This violation was applicable to both units, and had existed since each unit's initial refuelin The QA report also evaluated some aspects of the licensee's program for completing actions required by Generic Letter (GL) 96-01, " Testing of Safety-Related Logic Circuits."

Based upon concerns identified during this QA surveillance, a QA Program Significant issue was opened to address weaknesses in the execution of the licensee's GL 96-01 review. In addition to the concerns with management of the program, the auditors wrote several OCRs to identify specific issues that the auditors felt might not have been appropriately addressed. The inspectors considered the concerns raised by the auditors to be good issues which should have been addressed much earlier in the licensee's GL 96-01 proces Conclusions A licensee QA report was effective, in that it identified programmatic weaknesses in the licensee's response to GL 96-01, specific concerns with the licensee's current position on some GL 96-01 issues, and a failure to perform .ome aspects of a required surveillanc .

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The licensee engineering and regulatory affairs staff who responded to a licensee QA finding of inadequate surveillance testing for refueling system interlocks failed to apply sufficient rigor in their assessment of the issue. Inspector intervention was required to ensure that the T/S-required test was performed appropriately. One violation of T/Ss was identifie E8 Miscellaneous Engineering issues E (Closed) VIO 02063 from EA 96-273: irs 50-266/96006(DRP): 50-301/96006(DRP) and 50-266/96007(DRP): 50-301/96007(DRP): Violation of Criterion XI, " Test Control," of 10 CFR Part 50 Appendix B, for not incorporating the requirements and acceptance limits of applicable design documents in the IST program for the safety injection (SI)

pumps. The licensee committed to revise the IST program for the Si pumps to ensure that the appropriate requirements and acceptance criteria were included. The licensee also committed to review the IST program to ensure that appropriate acceptance limits were incorporated for other pumps and valve The licensee's implementation of its corrective actions for this violation has been reviewed over the past 1% years by NRC inspectors and is considered adequate overal These reviews were documented in irs 50-266/96013(DRP); 50-301/96013(DRP),

50 266/97009(DRP); 50-301/97009(DRP),50-266/97010(DRS); 50-301/97010(DRS), and 50-266/98006(DRP); 50-301/98006(DRP). However, occasional problems with the IST program have been identified by the inspectors. For a recent example involving the charging pumps and station battery, corrective actions are being tracked as VIO 50-266/97016-04(DRP); 50-301/97016-04(DRP), and for a recent example involving the component cooling water system, corrective actions are being tracked as VIO 50-266/97021-04(DRP); 50-301/97021-04(DRP). As discussed in Section E8.3 of IR 50-266/98006(DRP); 50-301/98006(DRP), the licensee was in the midst of an extensive rebaselining of the IST program undertaken in response to the number of concerns identified in the past 2 years. Because previous NRC inspections have indicated that overall the IST program is being adequately implemented, and because tracking of corrective actions for the recent problems involving the IST program is being performed under separate follow-up items, VIO 02063 is considered close E8.2 (Closed) VIO 03013 from EA 96-273: irs 50-266/96006(DRP): 50-301/96006(DRP) and 50-266/96007(DRP): 50-301/9600NDRP): Violation of Criterion XVI," Corrective Action,"

of 10 CFR Part 50, Appendix B, for the failure of the licensee to promptly correct a self-identified condition adverse to quality, in that a license amendment was not requested when the licensee determined that T/S 15.3.3.D did not accurately specify the lowest functional capability or performance level of the SW system. The SW T/S was appropriately revised on July 9,1997, (Amendment No.174 for Unit 1 and Amendment No.178 for Unit 2). However, during NRC followup of the licensee's review of the other T/Ss, the inspectors identified two additional T/Ss that were not conservative and for which license changes had not been requested. These issues were discussed in IR 50-266/96018(DRP); 50-301/96018(DRP) and considered with other problems for escalated enforcement action. An NOV was issued for these problems (EA 97-075) on August 8,1997, and cited the two more recently identified problems as examples of a violation of Criterion XVI. The licensee responded with proposed corrective actions in a letter dated October 10,1997. Because the original problem with the adequacy of the SW T/S has been corrected and corrective actions for the two more recent problems with l 1

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nonconservative T/Ss are being tracked as eel 50-266/96018-07c(DRP);

50-301/96018-07c(DRP) and eel 50-266/96018-07m(DRP); 50-301/96018-07m(DRP), i VIO 03013 is considered close f l

E (Closed) IFl 50-266/96006-04(DRS): 50-301/96006-04(DRS): Inspectors to review the results of an analysis by the licensee of inappropriate separation of wires in the main q

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control board panels. The NRC completed the review and discussed with the licensee the proposed method and schedule for resolution of the identified discrepancies. One aspect of the wire separation issue involving AFW pump control circuits was included in EA 97-505 (IR 50-266/97022(DRP); 50-301/97022(DRP)), which comprised numerous licensee-identified and corrected old design issues. Enforcement discretion was applied to this EA and no NOV or civil penalty was issued. Additional reviews of the resolution of the wire separation issue were documented in irs 50-266/96010(DRS);

50-301/96010(DRS), 50-266/97013(DRP); 50-301/97013(DRP) and 50-266/98006(DRP);

50-301/98006(DRP). The status and schedule of the licensee's corrective actions were most recently documented in a letter dated January 27,1998, to the NRC from the licensee. Because sufficient progress has been made by the licensee in resolving this issue, the IFl is considered close E (Closed) eel 50-266/97022-14(DRP): 50-301/97022-14(DRP): RCP Rotor Stand Support Not Seismically Adequate. This issue was documented in LER 50-266/97-035 and dispositioned along with numerous other licensee-identified and corrected old design issues as EA 97-505 in IR 50-266/97022(DRP); 50-301/97022(DRP). Enforcement discretion was applied to EA 97-505 and no NOV or civil penalty was issued. The inspectors verified that containment close-out procedures had been revised to address the identification and assessment of equipment left in containment during power operatio E (Closed) LER 50-266/98012: 50-301/98012: Missed Section XI Pressure Test Program Surveillance. This LER documented that 27 containment penetrations were tested in accordance with the licensee's 10 CFR Part 50, Appendix J, test program, but not the American Society of Mechanical Engineers (ASME)Section XI test program. The licensee subsequently determined that the intent of the ASME Section XI program requirements were met by the Appendix J program, declared the penetrations operable, and submitted a request on December 16,1997, to NRR for relief from the Section XI requirements. The safety significance of this issue was minor. This licensee-identified and corrected failure to test the containment penetrations under the Section XI test program is considered a non-cited violation (NCV 50-266/98009-05(DRP);

50-301/98009-05(DRP)) of T/S 15.4.2.B.1, consistent with Section Vll.B.1 of the NRC Enforcement Polic E8.6 { Closed) LER 50-266/97040: Potential Overstress of Seal Water Return Lines During Loss of Seal Coolant Events Such as Station Blackout or Appendix R Fires. This LER described a condition where nonsafety-related pipe was non-conforming but operabl The LER accurately described the condition, and corrective actions were completed during the U1R24 outag E (Closed) LER 50-266/97002: 50-301/97002: Potential to Overpressurize Piping Between Containment Isolation Valves. This LER described containment penetration overpressurization issues associated with NRC GL 96-06 (" Assurance of Equipment

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Operability and Containment integrity During Design-Basis Accident Conditions") review In one case, a containment penetration was determined to be inoperable pending temporary modification. All penetrations were returned to an operable status through the installation of pressure relief valves. Technical Specification 15.3.6.A.1.b requires that each containment penetration be operable to satisfy containment integrity. The failure of the original design of these penetrations to provide overpressure protection to accommodate the thermally-induced overpressurization that may occur during a design-basis accident is a non-repetitive, licensee-identified and corrected design issue and is considered to be an NCV (50-266/98009-06(DRP); 50-301/98009-06(DRP)) of T/S 15.3.6.A.1.b, consistent with Section Vll.B.1 of the NRC Enforcement Polic E (Closed) LER 50-266/96004: 50-301/96004: Operation of the SW System Outside the Design Basis. This issue was reviewed in detailin IR 50-266/96006(DRP);

50-301/96006(DRP). The condition was addressed as VIO 03013 of EA 96-273 and was closed in Section E E (Closed) LER 50-266/96005: 50-301/96005: Potential SW Flashing in Containment Fan Coolers. This issue is closely associated with the issue described in Sections E8.2 and E IV. Plant Support R1 Radiological Protection and Chemistry (RP&C) Controls R General Comments (IP 71750)

The inspectors performed tours of the radiologically controlled areas in the plant on a routine basis. During a tour on April 24,1998, the inspectors noted that a set of keys was laying on the floor within, and near the exit from, a contaminated area. The keys were adjacent to the access for a locked high radiation area. The inspectors pointed the keys out to a health physics supervisor who was nearby. The supervisor confirmed that they were controlled keys for access to locked high radiation areas. The supervisor reestablished control of the keys. This observation was evaluated during a concurrent NRC Division of Reactor Safety inspection, and will be dispositioned in IR 50-266/98012(DRS); 50-301/98012(DRS).

R7 Quality Assurance in RP&C Activities R Chemistry Self-Assessment (IP 71750 & IP 40500)

The inspectors reviewed Chemistry Self Assessment Report S-A-98-01, dated February 23,1998. This self-assessment concluded that the chemistry program was being effectively implemented. The assessment identified opportunities for improvement (assessment findings)in the areas of chemistry system material condition and design control, chemistry procedures, and chemistry management's effectiveness in reinforcing standards and expectations. The recommendations (corrective actions) contained in the self-assessment were appropriate to address these assessment findings. The assessment findings were consistent with the observations made during the inspector's chemistry system walkdowns and routine review of CRs. The inspectors venfied that all

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recommendations contained in the self-assessment were entered into the licensee's corrective action tracking syste P2 Status of Emergency Planning Facilities, Equipment, and Resources The inspectors performed a routine tour of the licensee's emergency response facilities on April 27,1998, and noted that the technical support center (TSC) did not contain a complete set of plant documents for use by the licensee during TSC activation. The plant drawings and system descriptions were located in the TSC building, but not the TSC proper. The discrepancy in the location of the system descriptions and drawings was of minor safety significance because the reference material would be available for use in an event. Emergency Plan Manual 7.0, " Emergency Facilities and Equipment," Revision 40, Section 2.3, stated that the TSC was to contain appropriate plant drawings and system description This failure of the TSC to contain appropriate plant drawings constitutes a violation of minor significance and is not subject to formal enforcement action. The EP supervisor stated that the emergency plan was already undergoing review, and that this and other inconsistencies would be corrected in an emergency plan revision this yea V. Manaaement Meetinas X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee managernent at the conclusion of the inspection on May 22,1998. The licensee acknowledged the findings presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified. The inspection period ended on May 23,1998. There were no significant inspection findings that were made between the exit meeting on May 22 and the end of the inspection period on May 2 ,

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PARTIAL LIST OF PERSONS CONTACTED Licensee Wisconsin Electric Power Company S. A. Patuiski, Site Vice President M. E. Reddemann, Plant Manager R. G. Mende, Operations Manager W. B. Fromm, Maintenance Manager J. G. Schweitzer, Site Engineerir.g Manager R. P. Farrell, Health Physics Manager D. F. Johnson, Regulatory Services and Licensing Manager i

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INSPECTION PROCEDURES USED IP 37551: Onsite Engineering IP 40500: Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems IP 61726: Surveillance Observations IP 62707: Maintenance Observation IP 71707: Plant Operations IP 71750: Plant Support Activities ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-266/98009-01(DRP) URI NRR to Evaluate Adequacy of Containment Spray 50-301/98009-01(DRP) System Testing 50-266/98009-02(DRP) NCV Missed T/S Surveillance of a Radiation Monitor ,

50-301/98009-02(DRP) Prior to Discharge l l

50-266/98009-03(DRS) IFl Containment Hatch Requirement, Maintenance l 50-301/98009-03(DRS) Rule,10 CFR 50.65 l 50-266/98009-04(DRP) VIO Failure to Test All Required Refueling Systems 50-301/98009-04(DRP)

50-266/98009-05(DRP) NCV Failure to Test the Containment Penetrations 50-301/98009-05(DRP) Under the ASME Section XI Test Program 50-266/98009-06(DRP) NCV Containment Penetration Determined to be 50-301/98009-06(DRP) Inoperable Pending Temporary Modification 50-266/98009-07(DRP) IFl Inspectors to Review Evaluation of Load to 50-301/98009-07(DRP) Main Steam Piping from inadequate Drains Closed 50-266/VIO 01013(DRP) EA 96-273 Watching Training Videotapes in the Control 50-301/VIO 01013(DRP) Room 50-266/VIO 01023(DRP) EA 96-273 RO Briefly Leaving the Control Room Front Panels 53-301/VIO 01023(DRP) Without a R lief 50-266/VIO 01033(DRP) EA 96-273 RO Not Responding Promptly to an Alarm 50-301/VIO 01033(DRP)

50-266/VIO 01043(DRP) EA 96-273 DTA Leaving Site While on Duty 50-301/VIO 01043(DRP)

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50-266/98011- LER Missed T/S Surveillance t , . "Nn Monitor 50-301/98011 Prior to Discharge 50-266/98002 LER Failure of the High Voltage Station Auxiliary 50-301/98002 Transformer 50-266/ % 002 LER AFW Discharge Valves Found Shut When Critical 50-266NIO 02013(DRP) EA 96-273 AFW Pump Issue, Violation of T/S 15.3.4. NIO 02013(DRP)

50-266NIO 02023(DRP) EA 96-273 Criterion V, instructions, Drawings and Procedures 50-301NIO 02023(DRP)

50-266NIO 02033(DRP) EA 96-273 Criterion V, instructions, Drawings and Procedures 50-301NIO 02033(DRP)

50-266Nto 02043(DRP) EA 96-273 Criterion V. Instructions, Drawings and Procedures 50-301NIO 02043(DRP)

50-266NIO 02053(DRP) EA 96-273 Criterion V, instructions, Drawings and Procedures 50-301NIO 02053(DRP)

50-266/96012 LER 96012 EDG Fuel Oil System Tests Not Performed in 50-301/96012 Accordance with T/S 50-266/%018-11(DRP) eel Violation of Fuel Oil Pump Start Requirements 50-301/96018-11(DRP)

50-266/96018-10(DRP) eel Violation on Load Testing of EDGs 50-301/96018-10(DRP)

50-266/98007 LER Steam Driven AFW Low Suction Pressure Trip 50-266/98008 LER EDG Air Start Motor Solenoid Valves Found With 50-301/98008 Springs Not in Accordance with Design 50-266/96019-04(DRP) IFl Containment inner Personnel Hatch Test Failures

50-266NIO 02063(DRP) EA %-273 Criterion XI Test Control, IST l 50-301NIO 02063(DRP)

l 50-266NIO 03013(DRP) EA 96-273 Criterion XVI, Corrective Action, SW System l.

! 50-301NIO 03013(DRP)

50-266/96006-04(DRS) IFl Cable Separation 'in the Control Room 50-301/%006-04(DRS)

50-266/97022-14(DRP) eel '

RCP Rotor Stand Support Not 50-301/97022-14(DRP) Seismically Adequate i

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50-266/98012 LER Missed ASME Section XI Pressure Test Program 50-301/98012 Surveillance 50-266/97040 LER Overstress SW Return Lines During Loss of Seal Coolant Events 50-266/97002 LER Potential to Overpressure Piping Between Containment isolation Valves 50-266/96004 LER Operation of the SW System Outside the Design 50-301/96004 Basis 50-266/96005 LER Potential SW Flashing in Containment Fan Coolers 50-301/96005 Discussed 50-266/97016-04(DRP) VIO Acceptance Criteria within IST Test ano 50-301/97016-04(DRP) Maintenance / Surveillance Procedures 50-266/97021-04(DRP) VIO Inadequate Testing of CCW System 50-301/97021-04(DRP)

50-266/97035 LER RCP Rotor Stand Support Not Seismically Adequate 50-266/96018-07c(DRP) eel T/S Did Not Specify Lowest Function Capability 50-301/96018-07c(DRP) or Performance :.evel of Crossover Steam Dump System 50 266/96018-07m(DRP) eel T/S for Setpoints for the 480-Volt 50-301/96018-07m(DRP) and 4160-Volt Loss-of-Voltage Relays were not Conservative i

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LIST OF ACRONYMS USED AC Attemating Current AFW Auxiliary Feedwater ASME American Society of Mechanical Engineers CCW Component Cooling Water CFR Code of Federal Regulations CLB Current Licensing Basis CR Condition Report CVC Chemical and Volume Control System DRP Division of Reactor Projects DRS- Division of Reactor Safety DSS Duty Shift Superintendent l

DTA Duty Technical Advisor L EA Enforcement Action

.ECCS- Emergency Core Cooling System EDG Emergency Diesel Generator eel Escalated Enforcement item lESF . Engineered Safety Feature EP Emergency Planning FME Foreign Material Exclusion FSAR Final Safety Analysis Report GL Generic Letter GS Gas Stripper IFl Inspection Follow-up item IP inspection Procedure IPE Individual Plant Evaluation IR Inspection Report IST Inservice Testing ILRT Integrated L.eak Rate Test IT Inservice Test Procedure LCO Limiting Condition for Operation LER- Licensee Event Report MPFF Maintenance Preventable Function Failures MSS Manager's Supervisory Staff NCV Non-Cited Violation NDE Non-Destructive Examination NOV Notice of Violation NP Nuclear Power Department Procedure NRC Nuclear Regulatory Commission OF Operating Instruction OM Operations Manual OOS Out-of-Service OP Operating Procedure OR Operations Refueling Test OSRC Off-Site Review Committee PASS Post-accident sampling System POD Prompt Operability Determination QA_ Quality Assurance QCR .- Quality Condition Report

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RCS Reactor Coolant System RCE . Root Cause Evaluation RCP Reactor Coolant Pump RHR Residual Heat Removal RO . Reactor Operator RP Radiation Protection RWST Refueling Water Storage Tank l SE Safety Evaluation i SER Safety Evaluation Report l SPF' Spent Fuel Pool l SRO Senior Reactor Operator l SW- Service Water TDAFW Turbine Driven Auxiliary Feedwater T/ Technical Specification TS Technical Specification Test TSC Technical Support Center URI - Unresolved item -

'USQ Unreviewed Safety Question

.VCT Volume Control Tank

- VIO Violation VNCR Control Room Ventilation

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