ML18066A314

Provides Update to Design Insp Action Items Re Insp Rept 50-255/97-201 Conducted on 970916-1114.Util Recommends That NRC Consider Scheduling Efforts Early in 1999 to Review Insp Items for Closure Based on Completion Dates for Items

ML18066A314
Person / Time
Site:	Palisades
Issue date:	10/01/1998
From:	Haskell N CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
50-255-97-201, NUDOCS 9810070265
Download: ML18066A314 (55)

Text

A CMS Energy Company October 1, 1998 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington D.C. 20555 Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert. Ml 49043 DOCKET 50-255 -LICENSE DPR-20 -PALISADES PLANT

• Tel. 616 764 2276 Fax.* 616 764 2490 Nathan L. Ha.ks/I Director.

Licensing OCTOBER 1, 1998 UPDATE TO DESIGN INSPECTION ACTION ITEMS During the period from September 16 through November 14, 1997, the NRC conducted a design inspection at the Palisades Nuclear Plant. By letter dated December 30, 1997, the NRC issued Inspection Report No. 50-255/97-201, and requested a response within 60 days detailing our plans to complete the corrective actions required to resolve the open items listed in Attachment A of the inspection report. Contained within our March 2, 1998 response was a single commitment to provide the NRC a status of our progress in completing actions associated with each open inspection item. The purpose of this commitment, in part, was to assist the NRC in planning for follow-up review and closeout of these items. Attachment A of this letter contains the text of each open inspection item from the December 30, 1997 inspection report, followed by our 60 day response as submitted in our March 2, 1998 letter, followed by the status of associated action as of October 1, 1998. This status includes the results of our investigations and corrective actions, along with planned completion dates for ongoing actions. Attachment B contains similar information for programmatic issues related to inspection findings.

_J Based on completion dates for the remaining open items, we recommend that NRC consider scheduling efforts early in 1999 to review inspection items for closure. A review of completion dates for open items indicates that a majority of actions will be completed by the end of 1998. 9810070265 981001 PDR ADOCK 05000255 G PDR

-. . .:.; * * -.. -Sl:JMMAR¥-'-8F COMMITMENTS This letter closes the March 2, 1998 commitment as .restated below, and contains no new commitments. "By October 1, 1998, Consumers Energy will provide NRC with a status of our progress in completing all actions identified in the attachments to this letter.

• Nathan L. Haskell . Director, Licensing CC Administrator, Region Ill, USNRC Project Manager, NRR, USNRC NRC Resident Inspector

-Palisades Attachments 2

ATTACHMENT A CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255 STATUS OF PLANS FOR CORRECTIVE ACTIONS TO RESOLVE NRC DESIGN INSPECTION OPEN ITEMS 45 Pages

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Unresolved Item 50-255/97-201-01 The team questioned whether the CCW system design met the vendor-recommended minimum flow of 2000 gpm for the CCW pumps under all operating conditions.

The team was concerned that small differences in the pump operating characteristics could cause significant differences in flow through each pump during parallel pump operation due to the flatness of the pump operating

• curves at low flows. The licensee had no analysis available to demonstrate that the CCW pumps met the minimum flow requirements.

During the inspection, the licensee developed a preliminary system flow model, which showed that, when all three pumps were started upon receiving a safety injection system (SIS) signal, the minimum pump flow was through CCW pump P-52A at 1768 gpm. The licensee received a revised minimum flow requirement of 1600 gpm from the pump manufacturer.

The team's review of the licensee's completed flow model calculation will be an Inspection Fol/owup Item 50-255197-201-01.

• Palisades 60 Day Response:

As a result of CCW system balancing, scheduled for the 1998 refueling outage, a reanalysis of minimum predicted CCW system flow rates will be performed.

This reanalysis will verify that minimum flow rate requirements will be met under a worst case scenario with appropriate pump IST degradation input. This action will be completed by September 1, 1998. 10/1/98 Update: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-01) was identified as open. Pump performance data was obtained during the 98 refueling outage. The completion for the reanalysis has been rescheduled for August 1, 1999 to accommodate emerging higher priority analytical work. Unresolved Item 50-255/97-201-02 The team verified the heat removal capability of the CCW heat exchangers by reviewing the results of various accident analyses.

The licensee had performed the following LOCA analyses:

• EA-D-PAL-93-207-01, "LOCA Containment Response Analysis With Reduced LPSI Flow Using CONTEMPT El-28 Code," Revision 0,

• EA-D-PAL-93-272-03, "LOCA Containment Response Analysis With Degraded Heat Removal System Using CONTEMPT El-28A Computer Code," Revision 0, *and

• EA-GEJ-96-01, "A-PAL-94-324 Containment Spray System (CSS) Sensitivity on the Containment Heat Removal During Recirculation (Post-RAS)," Revision 1. The team verified that the input assumptions relating to the CCW system for the above analyses were correct. The above LOCA analyses demonstrated that the heat exchangers could remove sufficient heat from containment following a LOCA to keep the containment pressure and 1

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ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS temperature within the design limits. In each case, the analysis documented a CCW temperature exiting the shutdown coolers exceeding the system design temperature of 140 degrees Fahrenheit (140 °F) as stated in FSAR table 9-6 and DBD 1.01, "Component Cooling Water," Revision 3. The team noted that the licensee accepted the maximum CCW temperature that resulted from the scenarios analyzed in EA-D-PAL-207-01 and EA-D-PAL-93-272-03 by Corrective Action D-PAL-93-272G, based primarily on an evaluation of the effects on pipe stress. However, the licensee had not considered the other negative effects, such as any detrimental effects from elevated CCW temperature on pump seals. Also, the licensee had not determined the maximum possible CCW temperature under worst case conditions and had not identified that a change to the FSAR could be required.

The team reviewed the latest LOCA analysis, EA-GEJ-96-01, and determined that it documented a CCW temperature exiting the shutdown cooling heat exchanger was 184 °F. The licensee determined the system was operable under this condition and issued Condition Report (CR) C-PAL-97-1363F to determine the most limiting CCWtemperature for any condition and to evaluate all the effects resulting from that limiting temperature on the CCW system. ' It appeared that the requirements of 10 CFR 50, Appendix B, Criterion 111, "Design Control," were not met in this case in that the design basis for the CCW system, as defined in 10 CFR 50.2, did not encompass the entire range of bounding temperatures.

The team identified this item as Unresolved Item 50-255197-201-02.

Palisades 60 Day Response:

Prior to the Design lnspection;.we determined that the CCW system is operable at a predicted maximum system temperature of 184°F. The CCW system will be analyzed to confirm the most limiting temperature for any design basis condition, and to determine the effects of this temperature on system components by October 1; 1998. The FSAR will be updated as appropriate.

The programmatic design control aspects related to this issue will be addressed as identified in Attachment B, Item 1. 10/1/98 Update: In June of 1998, Engineering Analysis EA-LOCA-98-01 was performed to determine the limiting condition CCW temperature.

The results show a maximum 180°F CCW temperature out of the CCW heat exchanger.

The effects of this temperature on system components was then evaluated.

It was determined that the CCW heat exchanger outlet temperature indication range was too narrow and needed to be expanded to meet RG 1.91 requirements.

By December 15, 1998, these temperature indicators will be replaced and full compliance with RG 1.97 requirements will be achieved.

All other evaluated CCW system component peak temperature ratings fall within the predicted 180°F temperature.

The FSAR was changed to clarify CCW system design temperature and LOCA maximum temperatures.

The temperature indicator range issue (50-255/97201-02) was identified as open, and was the subject of a NOTICE OF DEVIATION (50-255/98003-02), in NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION." Palisades responded with additional information to the NRC under correspondence dated June 24, 1998, entitled "RESPONSE TO NOTICE OF VIOLATION AND NOTICE OF DEVIATION FROM INSPECTION REPORT 50-255/98003." 2

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Refer to Attachment B, Item 1 for the programmatic "design control" aspects associated with this issue. Unresolved Item 50-255/97-201-03 The team reviewed C-PAL-96-1-63-01, "120 day response to GL 96-06, Assurance of Equipment Operability and Containment Integrity during Design Basis Accident Condition," Revision 0, which was the licensee's response to Nuclear Regulatory Commission (NRG) Generic Letter 96-06, "Assurance of Containment Operability and Containment Integrity During Design-Basis Accident Conditions," and observed that the licensee took credit for relief valve RV-0939 to protect the CCW piping inside containment from overpressurization in the event of a LOCA. RV-0939 was not included in the /ST program. The team questioned whether RV-0939 performed a safety function and if it should have been included in the /ST program. The licensee issued CR C-PAL-97-1686 to evaluate this discrepancy.

10 CFR 50.55a requires /ST in accordance with ASME Section XI of valves that perform a safety function.

It appeared that the licensee did not fully implement these requirements for RV-0939. The team identified this item as part of Unresolved Item 50-255197-201-03.

Palisades 60 Day Response:

During the Design Inspection, it was determined that sufficient overpressure protection is provided for the CCW system without taking credit for relief valve RV-0939, and the CCW system is therefore operable.

The CCW piping in containment is not required during an accident and is classified non-Q, safety related. As a result, the ISl/IST programs have classified the CCW piping and related components, including RV-0939, as non-class and excluded the same from inspection/test requirements of the Code. The Palisades response to GL 96-06 determined acceptability of systems by generally taking credit for 1) steam/gas service, 2) available expansion paths, or 3) relief valves as a means to provide *sufficient protection against thermally induced over pressurization.

In the case of the CCW system, "available relief valves" serves as the basis for acceptability.

Relief valve operation is considered important but not a safety related function, and therefore, the classification of the CCW system and its components such as RV-0939 were not changed. Although RV-0939 is not in the IST program, it, along with RV-2108 and RV-0956, is inspected, maintained and set point verified via maintenance activity PPAC CCS043 on a 10-year interval.

These are essentially the same as the requirements of the Code (ASME/ANSI OM-1987, Part 1 ). Based on this evaluation, no further action is required.

RV-0939 is appropriately classified, maintained and tested. Our existing GL 96-06 submittal is accurate.

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• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS 10/1/98 Update: This response has not changed since the submittal of our original 60-day inspection report response.

Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-03) was identified as closed. No further actions on this item are planned. Unresolved Item 50-255/97-201-04 FSAR Section 9.3.2.3 stated that the CCW pipingwithin containment was not vulnerable to failure caused by a high energy line break (HELB) and referred to Deviation Report (DR) D-PAL-89-061, "Post Accident Operation of CCW System, 11 dated March 23, 1989, for the evaluation.

This DR referred to Engineering Analysis (EA) EA GW0-7793-01, "CCW Piping Inside Containment HELBA," Revision 0. This EA was reviewed by the team, and it concluded that the CCW piping inside containment was not affected by HELBs, but did not contain the analysis performed or a reference to the analysis.

The EA contained an outline of the methodology, listed the drawings and walkdowns used, and referenced the source of the postulated HELBs. Palisades Administrative Procedure No. 9.11, "Engineering Analysis, 11 Revision 9, stated that an EA shall present an argument which substantiates the conclusion of the EA. The EA also contained an error in the identification of the Systematic Evaluation Program (SEP) topic number for evaluation of the effects of internally generated missiles.

The licensee initiated Engineering Assistance Request (EAR) EAR-97-0632 to revise EA-GW0-7793-01.

During the inspection, the licensee issued Revision 1 of EA-GW0-7793-01, which included a discussion of the walkdown analysis used and corrected the SEP references.

This revised EA was acceptable to the team. It appeared that the requirements of 10 CFR Part 50, Appendix B, Criterion . Ill, "Design Control," regarding verifying the adequacy of designs were not adhered to in this case. Also, the requirements of the licensee's Administrative Procedure

9. 11 were not fully met in that EA-GW0-7793-01, Revision 0, did not contain full substantiation of the conclusion.

The team identified this item as Unresolved Item 50-255197-201-04.

Palisades 60 Day Response:

As a remedial action, EA-GW0-7793-01 was revised to provide justification for its conclusion and to correct references to related NRC corresponqence.

The related programmatic design control and calculation control aspects will be addressed as identified in Attachment B, Items 1 and 2. 10/1/98 Update: This response has not changed since the submittal of our original 60-day inspection report response.

Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-04) was identified as closed. No further actions are planned for this item . Refer to Attachment B, Item 1 for the programmatic "design control" aspects associated with this issu.e. 4 ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Unresolved Item 50-255/97-201-05 The team reviewed the implementation of the licensee's commitment to NRG Regulatory Guide (RG) 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following an Accident," Revision 3, as described in FSAR Appendix 7C. The RG stated a range for CCW flow instrumentation of 0-110 percent. Since there was no instrument to directly measure CCW flow, the licensee used a combination of instruments, including TE-0912 and TE-0913, which measure shutdown cooling heat exchanger outlet temperature, to indicate flow. Use of instruments (other than flow indicators) to monitor for CCW flow was determined as acceptable by the NRG (a letter from NRG to Consumers Power Company, dated July 19, 1988, entitled "Palisades Plant-Response to Generic Letter 82-33 Conformance to Regulatory Guide 1.97 "Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following an Accident).

The required range for these TEs in FSAR Appendix 7C was 0-180 °F. This range did not encompass the temperature determined in EA-GEJ-96-01, "A-PAL-94-324 Containment Spray System (CSS) Sensitivity on Containment Heat Removal During Recirculation (Post-RAS)," Revision 1. This analysis determined an outlet temperature of the CCW from the shutdown cooling heat exchanger of 184 °F. The licensee issued CR C-PAL-97-1363E to evaluate the process instrumentation and controls associated with the CCW system for the effects of the higher temperature predicted by the analysis.

The licensee did not appear to meet their commitment to NRG RG 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following an Accident," in that the installed CCW temperature indicators were not capable of monitoring the full temperature range expected to be observed in the CCW system. The team identified this item as part of Unresolved Item 50-255197-201-05.

Palisades 60 Day Response:

Prior to the Design Inspection, we determined that the COW system is operable at a predicted maximum system temperature of 184°F. The CCW system will be analyzed to confirm the most limiting temperature for any design basis condition, and the effects of this temperature on system components.

In response to this specific issue, process instrumentation and controls associated with the CCW system will be reviewed to identify the impact of the maximum predicted temperature.

This action will be completed by October 1, 1998. 10/1/98 Update: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-05) was identified as closed. This item was also the subject of a NOTICE OF DEVIATION (50-255/98003-02) from the same letter. Palisades responded with additional information to the NRC under correspondence dated June 24, 1998, entitled "RESPONSE TO NOTICE OF VIOLATION AND NOTICE OF DEVIATION FROM INSPECTION REPORT 50-255/98003." In summary, the range of the CCW heat exchanger outlet temperature indicators will be changed to meet RG 1.97 requirements by December 15, 1998. 5

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• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Unresolved Item 50-255/97-201-06 The team identified a lack of closure verification testing on SI system check valves that could potentially result in an overpressure condition affecting the low-pressure piping on the suction of the HPSI pumps. The minimum flow recirculation lines associated with the two HPSI pumps and the two LPSI 'pumps were interconnected upstream of the air-operated minimum flow recirculation isolation valves. In the event that only one HPSI pump was operating under post-accident conditions with the minimum flow recirculation isolation valves closed, back leakage through the minimum flow piping associated with the idle HPS/ pump could over pressurize the idle HPS/pump suction piping. Backflow between the HPS/ minimum flow lines should be prevented by check valves CK-ES3339 or CK-ES3331, and CK-ES3340 or CK-ES3332.

However, EGAD-EP-01, "lnservice Testing Program-Valve Test Program," Revision 10 indicated that closure verification testing of these check valves was not included in the /ST program. *The team asked the licensee if closure of these check valves was considered a safety function requiring

/ST. The licensee initiated CR C-PAL-97-1660 to evaluate the testing requirements of these check valves. On November 10, 1997, the operability determination concluded that these system check valves had not been subject to closure verification testing as required, and both HPSI pumps were declared inoperable.

In accordance with TS Section 3.0.3, 3.3, and 4.0.3, the licensee entered a Limiting Condition for Operation (LCO) action statement, performed closure verification testing of check valves CK-ES3339 and CK-ES3340, and verified the operability of these valves. The licensee stated that closure verification testing of these check valves would be added to the /ST program. The team also identified a lack of closure verification testing on SI system valves that could potentially result in a Safety Injection Tank (SIT) being degraded under post-accident conditions.

The normally closed SIT vent valves, CV-3051, 3063, 3065, and 3067, could be opened in accordance with SOP-3, "Safety Injection and Shutdown Cooling System," Revision 28, to reduce SIT pressure.

SOP-3 did not require the affected SIT to be declared inoperable when a vent was opened. When a vent valve was opened the SIT pressure boundary (250 psig design pressure) was exposed to the SIT vent header piping (100 psig design pressure).

SOP-3 did not include d(rections to isolate an open vent valve in the event of an accident.

EGAD-EP-01, lnservice Testing Program -Valve Test Program," Revision 10, indicated that closure verification testing of these valves was not included in the /ST program. The team asked the licensee if the failure of a valve to close could result in a SIT being degraded under accident conditions, and if closure of these valves was considered a safety function requiring

/ST testing. The licensee initiated CR C-PAL-97-1592 to evaluate this item and placed caution tags on the control room switches for vent valves CV-3051, 3063, 3065, and 3067 to prevent the valves from being opened without entering an LCO for the SITs. The licensee also stated that these valves had been opened rarely during plant operation.

1 O CFR 50. 55a requires in-service inspection in accordance with Section XI of the ASME Boiler and Pressure Vessel Code. This code requires testing of valves which perform a safety function.

It appeared that the licensee did not implement these requirements with regard to valves CK-ES3339, CK-ES3340, CV-3051, CV-3063, CV-3065, and CV-3067. The team identified this item as part of Unresolved Item 50-=255197-201-06.

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ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Palisades 60 Day Response:

During the Design Inspection, high pressure safety injection pump minimum flow recirculation line check valves CK-ES3339 and CK-ES3340 were tested and the HPSI system was declared operable.

Action to include check valves CK-ES3339 and CK-ES3340 in the IST Program will be completed by July 15, 1998. lri the interim, the check valves are tested to meet quarterly testing requirements.

During the Design Inspection, the Safety Injection Tank (SIT) vent valves CV-3051, CV-3063, CV-3065 and CV-3067 were closed and cautioned tagged with the tanks declared operable.

Action to revise operating procedures to address opening the SIT vent valves will be completed prior to removal of the caution tags. Prior to March 15, 1998, a representative sample of check valves, AOVs and MOVs will be reviewed and verified to be incorporated in the IST program as required.

10/1/98 Update: Check valves CK-ES3339 and CK-ES3340 have been included in the IST Program. Operating procedures have been revised to address opening of the SIT vent valves CV-3051, CV-3063, CV-3065 and CV-3067 and caution tags have been removed. A representative sample of check valves, AOVs and MOVs have been sampled to determine if they are included in the IST Program as required.

The sampling identified additional AOVs and one check valve that required inclusion into the IST Program. These valves have been incorporated into the IST Program and have been tested to confirm their safety related function.

In addition, several other actions associated with the IST Program are underway to enhance databases, review ISi Program bases for IST Program impact, and revise IST Program and bases to enhance purpose, scope and program descriptions.

These actions are projected to be complete by . May 1, 1999. Presently, Palisades is in full c_ompliance with the ISi and IST program requirements.

Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-06) was identified as closed. .This item was also the subject of a NOTICE OF VIOLATION (50-255/98003-03) from the same letter. Palisades responded with additional information to the NRC under correspondence dated June 24, 1998, entitled "RESPONSE TO NOTICE OF VIOLATION AND NOTICE OF DEVIATION FROM INSPECTION REP.ORT 50-255/98003." Unresolved Item 50-255/97-201-07 The team reviewed the HVAC system serving the cable spreading room. The team observed that DR F-CG-91-072 was prepared in May 1991 when it was discovered that the assumptions in calculation EA-FC-573-2, "Calculated Required Air Flow for Inverter/Charger Cabinet Cooling Fan," dated October 3, 1982, used an ambient temperature of 94 °F instead of the correct design basis temperature of 104 °F. The Safety System Design Confirmation (SSDC) Team that found this discrepancy recommended that the EA be updated. Procedure*9.11, "Engineering Analysis," Revision 9, required all EAs to be revised if analytical inputs or major assumptions change. The 7 ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS licensee aec1dedtiotl6 reVisetfie EA-; and ffie alscrepaiicy was recorded in DBD 4.02 (125-V de system) and DBD 4.03 (preferred ac system). The fans were installed in 1983 and were not safety related. DR F-CG-91-072 was closed in October 1994, when the decision was made not to revise the calculation.

The licensee stated that specifications were being developed for replacing the inverters and chargers during the time the discrepancy was being evaluated and that this knowledge contributed to the decision not to update the EA. The inverters and chargers were scheduled to be replaced in the near future by Specification Change (SC) SC-96-033.

The new equipment would have internal cooling fans designed for a 104 °F maximum ambient and SC-96-033 would supersede EA-FC-573-2 upon installation.

The team had no other concerns about the cable spreading room HVAC system. It appeared that the requirements of 10 CFR Part 50, Appendix B, Criterion/I, "Quality Assurance Program," were not followed in this case in that the requirements of Procedure

9. 11 regarding revising EAs were not fully implemented.

The team identified this item as part of Unresolved Item 50-255197-201-07.

Palisades 60 Day Response:

Prior to the Design Inspection, Design Basis Documents were revised to address this discrepancy.

Analysis EA-FC-573-2 will be revised or superseded by December 1, 1998. The calculation control aspects related to this issue (in this case, the revision of all analyses whenever analytical inputs or major assumptions change) will be addressed by the action described in Attachment B, Item 2. 10/1/98 Update: The schedule for resolving remains as stated above. Per NRG correspondence dated May 18, 1998, titled "NRG INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-07) was identified as closed. This item was also the subject of a NOTICE OF VIOLATION (50-255/98003-04) from the same letter. Palisades responded with additional information to the NRC under correspondence dated June 24, 1998, entitled "RESPONSE TO NOTICE OF VIOLATION AND NOTICE OF DEVIATION FROM INSPECTION REPORT 50-255/98003." Unresolved Item 50-255/97-201-08 The team identified the following discrepancies in SJ system mechanical calculations:

• EA-DBD-2.01-004, "Electrical and Mechanical Failure Analysis for the Low Pressure Safety Injection System," Revision 0, pages 10 and 25, identified a situation in which a Joss of an emergency diesel generator (EOG) during a large-break LOCA would result in only one LPSI pump and two LPS/ injection valves being operable.

The EA stated: "The acceptability of this situation could not be verified." The team asked if this statement was correct. The licensee replied that the statement was not current, and that the statement appeared to be based on superseded calculation ANF-88-107, "Palisades Large Break LOCA/ECCS Analysis With Increased Radial Peaking," Revision 1. Calculation ANF-88-107 was superseded by Seimens calculation EMF-96-172, "Palisades Large Break LOCA/ECCS Analysis," Revision 0. The licensee initiated Engineering Assistance Request (EAR) 97-0635 to revise EA-DBD-2.01-004.

8 ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS

• EA-A-NL-92-185-01, "Worst Case Operating Conditions for the LPSllSDC System MOVs," Revision 1, addressed the most limiting conditions under which the system motor-operated valves (MOVs) were required to open and close. This analysis included MOVs M0-3015 and M0-3016. These valves were the isolation valves installed in the shutdown cooling inlet . piping from primary coolant system (PCS) loop 2. For all normal operations

-other than shutdown cooling being in service, -the valves were electrically locked closed. Page 19 of EA-A-NL-92-185-01 stated that the scenario that could produce the most limiting differential pressure was that these valves would be required to close in the event of a downstream pipe break. The EA addressed a potential 12-in. downstream pipe break and determined that complete depressurization and blowdown of the PCS to the hot-leg elevation would occur before operators could enter the EOPs and attempt to isolate the break. Therefore, the analysis then established a maximum flow rate of 4120 gpm through valves M0-3015 and M0-3016, based on a normal system flow rate of 3000 gpm and a calculated leakage of 1120 gpm through a break of a 1-112-inch branch line downstream of the valves. The team asked the licensee to provide the basis of the postulated 1-112-inch branch line failure, since it did not appear to be consistent with the postulated pipe crack used in the internal flooding analysis of the safeguards areas (EA-C-PAL-95-1526-01, "Internal Flooding Evaluation for Plant Areas Outside of Containment," Revision 0). The licensee verified that the flooding analysis break flow was different and that this difference would not affect the conclusions of EA-A-NL-92-185-01.

Assumptions 5.9 and 5.10 of EA-A-NL-92-185-01 stated that the HPS/ and LPSI injection flows to the loops were approximately equal under post-accident conditions.

These assumptions did not appear consistent with the flow values calculated in EA-SDW-95-001, "Generation of Minimum and Maximum HPSllLPSI System Performance Curves Using Pipe-Flo," Revision 2. The team asked the licensee to provide the bases of these values. The licensee stated that the values were not current and verified that the difference between these values and the current values would not affect the EA results. The licensee initiated CR C-PAL-97-1670 to resolve the discrepancies in EA-A-NL-92-185-01.

• EA-E-PAL-93-004E-01, "/ST Check Valve Minimum Flow Rate Requirements to Support Chapter 14 Events," Revision 0, identified 1601 gpm as the required test flow for the LPS/ injection check valves. The team observed that this value appeared to be less limiting than the values calculated in EA-SDW-95-001, "Generation of Minimum and Maximum HPS/ILPSI System Performance Curves Using Pipe-Flo," Revision 2. The licensee initiated CR C-PAL-97-1603 to address this discrepancy.

The licensee determined that the LPSI test flow presented in EA-E-PAL-93-004E-01 was less than the current calculated requirement.

However, the actual LPSI check valve flow acceptance criterion in /ST Procedure Q0-88, "ESS Check Valve Operability Test (Cold Shutdown)," Revision 17, was verified to be 1690 gpm, which was greater than the current calculated requirement.

The licensee stated that the affected documentation will be corrected.

Administrative Procedure

9. 11, "Engineering Analysis," Revision 9, Section 6. 1. 5. c stated that an analysis shall be revised if analytical inputs changed. In the above instances, engineering analyses were not updated to reflect analytical input change. The licensee initiated C-PAL-97-1636 to evaluate the overall issue of calculation control. The team identified this item as part of Unresolved Item 50-255197-201-08.

9 ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Palisades 60 Day Response:

During the Design Inspection, it was determined that the LPSI check valves are operable since IST acceptance criteria and actual test flow rates exceeded the minimum required flow rates in analysis EMF-96-72 which had superseded EA-E-PAL-93-004E-01.

By June 1, 1998, engineering guideline EGAD-EP-09 and IST procedure Q0-8B basis document will be revised to assure that the increased minimum design flow requirement is met, and that design bases agree with IST acceptance criteria.

Remedial actions to revise EA-DBD-2.01-004 to accurately reflect electrical system response to events will be completed by August 15, 1998. EA-A-NL-92-185-01 and EA-SDW-95-001 are bounding analyses which will not be required to be revised or superseded.

Specifically,

• the calculation control process will be revised to allow bounding analyses to remain unchanged when revisions to inputs or assumptions do not affect the analysis conclusions.

The calculation control aspects related to this issue will be addressed by the action described in Attachment B, Item 2. 10/1/98 Update: Engineering guideline EGAD-EP-09, IST procedure Q0-8B Basis Document, and engineering analysis EA-DBD-2.01-004 were revised as stated above. Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-08) was identified as closed. This item was also the subject of a NOTICE OF VIOLATION (50-255/98003-04) from the same letter. Palisades responded with additional information to the NRC under correspondence dated June 24, 1998, entitled "RESPONSE TO NOTICE OF VIOLATION AND NOTICE OF DEVIATION FROM INSPECTION REPORT 50-255/98003." Unresolved Item 50-255/97-201-09 During an SI system walkdown on October 6, 1997, the team observed scaffolding installed adjacent to the SIRWT on the roof of the auxiliary building.

The team questioned how the installation of scaffolding in the vicinity of safety-related equipment was controlled to prevent damage to the safety-related equipment during a seismic event. The licensee provided Procedure MSM-M-43, "Scaffolding," Revision 2, for the team's review. Section 5. 3 of this procedure required an engineering review of scaffolding installed in the vicinity of safety related equipment.

However, the licensee determined that the scaffolding observed during the walkdown had not received engineering review in accordance with the procedure.

The licensee initiated CR C-PAL-97-1417 to address the scaffolding installation, and the scaffolding was removed on October 8, 1997. EA-C-PAL-97-1417A-01, "Operability Reassessment of SIRWT Scaffolding," Revision 0, was completed during the inspection.

Based on a structural analysis of the maximum loading on the SIRWT due to seismic interaction with the scaffolding during a safe shutdown earthquake, this analysis concluded that the SIRWT was not inoperable due to this nonconforming condition.

10

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS During another SI system walkdown on October 30, 1997, the team observed additional scaffolding installed in the east ESG room adjacent to safety-related piping. An evaluation by the licensee determined that this scaffolding had not been installed in accordance with Procedure MSM-M-43, "Scaffolding," Revision 2. The licensee initiated CR C-PAL-97-1585 to address this scaffolding installation and, based on a visual inspection, concluded that this nonconforming scaffolding would not render any safety-related piping or components inoperable.

The licensee removed the scaffolding.

In addition, the licensee performed a walkdown of all plant scaffolding during the inspection and verified that there were no additional nonconforming conditions.

The licensee stated that all scaffolding erections would cease until appropriate personnel underwent remedial training.

The team observed the following three separate conditions in the west ESG room involving potential seismic interactions with safety-related equipment.

The team noted that, during a seismic event, unrestrained items could potentially damage safety-related piping and equipment.

The safety-related piping and equipment in the west ESG room were required for operation of the HPSI, LPSI, and containment spray systems in the event of an accident.

• The team observed an unsecured operations storage cabinet located adjacent to safety-related piping and valves. The team asked the licensee if the condition was in accordance with plant procedures.

The licensee initiated CR C-PAL-97-1587, which determined that the cabinet was not placed in accordance with the spacing requirements of Administrative Procedure 1.01, "Material Condition Standards and Housekeeping Responsibilities," Revision 11. The operability evaluation concluded that the nonconforming condition did not result in any safety-related equipment being inoperable.

The cabinet was laid on its side to eliminate the toppling concern. The licensee stated that the cabinet would be removed from the area.

• The team observed an* unsecured chainfall located adjacent to and above the shutdown cooling heat exchangers.

A similar chainfall in the east ESG room was secured. The team asked the licensee if the condition was in accordance with plant procedures.

The licensee determined that the chainfall location was not in accordance with Administrative Procedure 1.01, and initiated CR C-PAL 97-1586. The operability evaluation concluded that the nonconforming condition did not result in any safety-related equipment being inoperable.

The licensee stated that the chainfall chains would be moved away from the heat exchanger.

• The team observed a ladder in the west ESG room that appeared to be improperly stored. The ladder was lying on the floor under the installed ladder rack. The team asked the licensee if the condition was in accordance with plant procedures.

The licensee initiated CR C-PAL-97-1601 and determined that the ladder location was not in accordance with the "Palisades Ladder Control Policy for Operating Spaces," dated May 14, 1997. The CR concluded that, although the ladder storage did not meet the ladder control policy, the nonconforming condition did not result in any safety-related equipment being inoperable.

The licensee stated that the ladder was removed from the area. Procedure MSM-M-43 required an engineering review of scaffolding installed in the vicinity of safety-related equipment.

Procedure

1. 01 and the "Palisades Ladder Control Policy for Operating Spaces," dated May 14, 1997, contain requirements for storing items in the vicinity of safety-related equipment.

In these cases, the licensee did not comply with the procedural requirements for activities affecting quality as required by 1 O CFR Part 50, Appendix B, Criterion V, "Instructions, 11 ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Procedures, and Drawings." The team identified this item as Unresolved Item 50-255197-201-09.

Palisades 60 Day Response:

Remedial actions consisted of dispositioning all scaffolding and unrestrained items near the SIRW Tank and in the East and West Safeguards Rooms to assure operability of safety-related equipment.*

Subsequently, walkdowns were conducted in other areas containing safety-related equipment and no conditions similar to the scaffolding conditions identified in this open item were observed.

Maintenance and construction crews were briefed on the lessons learned pertaining to scaffolding erection.

By July 15, 1998, we will revise procedures, provide training and reinforce management expectations as necessary to maintain compliance with seismic interaction requirements for related equipment.

10/1/98 Update: Specific actions to revise procedures, provide training and reinforce management expectations as necessary to maintain compliance with seismic interaction requirements for safety-related equipment have been completed.

Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003

• (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-09) was identified as closed. This item was also the subject of NOTICES OF VIOLATION (50-255/98003-05 and 50-255/98003-06) from the same letter. Palisades responded with additional information to the NRG under correspondence dated June 24, 1998, entitled "RESPONSE TO NOTICE OF VIOLATION AND NOTICE OF DEVIATION FROM INSPECTION REPORT 50-255/98003." This response is associated with plans to enhance maintenance personnel scaffolding training, and provide training for Auxiliary Operators to recognize unrestrained items for prompt identification.

Training will be completed by March 1, 1999. Unresolved Item 50-255/97-201-10 During the surrogate tour, the team obseNed the ends of two vent pipes that connected the containment sump to the 590-ft elevation of the containment.

The team asked the licensee to explain the design of these vent lines. During a review of the vent lines, the licensee determined that the top of the vents were located inside the containment at an elevation of approximately 595-ft. The maximum calculated post-accident water elevation was at elevation 597-ft. The vent pipes did not have screens on their inlets. The licensee also determined that the two vent lines entered the containment sump inside the sump screens, creating a potential path for debris to enter the EGGS pump suction piping under post-accident conditions.

The licensee initiated CR C-PAL-97-1571, on October 29, 1997, to evaluate this condition and determined that the postulated type and quantity of debris that could enter the vent pipes under post-accident conditions would not prevent the SI and containment spray systems from performing their safety function, and that these systems were operable under this condition.

The licensee also installed Temporary Modification TM-97-046, on October 29, 1997, to add screens to the top of the vent pipes during the inspection.

These screens would prevent debris from entering the EGGS pump suctions in the event of an accident.

12

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS It appeared that the requirements of 10 CFR Part 50, Appendix B, Criterion Ill, "Design Control," were not met in this instance in that the design basis of the containment sump to exclude debris from the EGGS pump suction piping was not fully implemented.

The team identified this item as part of Unresolved Item 50-255197-201-10.

Palisades 60 Day Response:

As stated above, an operability determination concluded the Engineered Safeguards Systems were operable in the as-found condition.

As additional assurance for continued operability, temporary screens were placed over the vent pipes. These screens will be permanently installed in the 1998 refueling outage. The programmatic design control aspects related to this issue will be addressed as identified in Attachment B, Item 1. 10/1/98 Update: Containment sump vent screens were permanently installed during the 1998 refueling outage. Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-10) was identified as closed. This item was also the subject of a NOTICE OF VIOLATION (50-255/98003-0?a) from the same letter. Palisades responded with additional information to the NRC under correspondence dated June 24, 1998, entitled "RESPONSE TO NOTICE OF VIOLATION AND.NOTICE OF DEVIATION FROM INSPECTION REPORT 50-255/98003." As part of our annual design basis document update projected for June 1999, the Containment Spray Design Basis Document DBD-2.03 will be revised to address issues vital to the function of the Engineering Safety Features following a LOCA. Refer to Attachment B, Item 1 for the programmatic "design control" aspects associated with this issue. Inspection Followup Item 50-255/97-201-11 The team also observed several piping penetrations between the east and west ESG rooms which included rubber piping expansion joints used as penetration seals. The team questioned the design of these piping penetration seals. The licensee stated that the engineering analyses that demonstrated that these penetrations met the design basis did not-specifically address the use of rubber piping expansion joints in the penetration seals. The team reviewed EA-RJC-92-0508, * Analysis of the Effect of a Fire on the Fire Barrier Penetration Seal Number FZ-0508," Revision 0, and verified that the rubber piping expansion joints were not addressed.

The licensee initiated CR C-PAL-97-1627 and determined that the failure to specifically justify the presence of rubber expansion joints did not invalidate the conclusions of the original engineering analyses and that the penetration seals were adequate.

The licensee also stated that the affected documentation would be corrected, and that an "extent of condition" review would be performed.

The team identified this item as Inspection Fo/lowup Item 50-255197-201-11.

13

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Palisades 60 Day Response:

An operability determination during the Design Inspection concluded that the safety function provided by the fire barriers separating the East and West Safeguards Rooms is not affected by the use of rubber expansion pipe joints. By August 1, 1998, we will revise the design basis engineering analysis to formally justify the installed rubber expansion pipe joints, and perform an investigation of other area fire barriers for potential unanalyzed designs. 1011198 Update: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-11) was identified as closed. The revision to the design basis engineering analysis for rubber expansion pipe joints is complete along with investigations for other fire barriers for potential unanalyzed designs. No other unanalyzed fire barrier design issues were discovered.

No further actions are planned for this inspection item. Inspection Followup Item 50-255197-201-12 The team reviewed 10 SI system calculations and 1 pressurizer pressure uncertainty calculation; these were identified as "basis documents." Basis Document Rl-38, "SIRW Tank Level Instrument Calibration," Revision 6, was reviewed for adequacy.

It provided the basis for calibration of SIRWT level indicators LT-0332A *and LT-0332B to enable their use to monitor the TS requirement that the tank contain at least 250, 000 gallons of borated water. Rl-38 used a tank boron concentration of 1720 parts per million (ppm) and did not consider the range of 1720 to 2500 ppm allowed by TS Section 3.3. Rl-38 was the basis document for the calibration of the level indicator that supported manual actuation of post-accident recirculation operation.

The team was concerned that the increased density of the tank water at higher boron concentrations would increase the instrument uncertainty.

The calculation also did not account for variation in boron concentration density caused by temperature changes; an effect which could also affect the total uncertainty.

The licensee recalculated the total instrument uncertainty using the most conservative boron concentrations and temperature, and the *resulting change to the total uncertainty remained bounded by the original uncertainty value. Bases Document Rl-69, "Subcooled Margin Monitor Surveillance," Revision 6, was reviewed for adequacy.

The subcooled margin monitor (SMM) provided the operator indication of the PCS margin to .saturation conditions.

Rl-69 evaluated possible errors induced in the SMM. The team found that Rl-69 did not account for seismic uncertainty.

This was inconsistent with RG 1.97 "Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following an Accident," May 1983. This RG identifies subcooled margin as a Category/, Type A variable, which must continue to read within the required accuracy following, but not necessarily during, a safe-shutdown earthquake event. The team was concerned that the calculated error was nonconservative because it did not consider seismic uncertainty, and could provide misleading information to the operators.

The licensee reanalyzed the potential error in the SMM, including seismic uncertainty, and the resulting total uncertainty remained bounded by the original uncertainty value. The licensee assigned Procedure Change Request (PCR) 5569 to revise Rl-69. 14

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS EA-RSW-94-001, "F/-0404 Instrumentation Uncertainty Calculation," Revision 2, was also reviewed for adequacy.

The analysis established the recommended uncertainties of Fl-0404, which was used in flow testing of the SJ pumps. The instrument was installed in 1989, and has been calibrated five times since then. Drift error was determined using historical calibration data. For the first 4 years, the instrument was calibrated once a year. The team found that 24 months had transpired between the fourth and fifth calibrations.

The licensee stated that the interval was in 1993 from 11 months to 24 months. The team asked if the drift analysis was revised to account for this change in the calibration interval.

The* team was concerned that increasing the calibration interval to 24 months would increase the drift error and consequently increase the total uncertainty of the instrument.

The licensee reanalyzed the Fl-0404 uncertainty using appropriate drift performance data for the longer calibration interval, and the resulting change to the total uncertainty remained bounded by the original uncertainty value. The licensee issued EAR-97-0658 to revise EA-RSW-94-001.

The team also reviewed Basis Document Rl-15A, "Safety Injection Tank Pressure Channel Calibration," Revision 7, for adequacy.

Rl-15A formed the bases for the pressure channel setpoints for PIA-0363, 0367, 0369, and 0371, which defined low-and high-pressure alarms for the S/Ts. The /ow-pressure alarms warned the operators of decreasing nitrogen pressure in the tanks. The channel alarms were set to annunciate earlier than the pressure limits of TS Section 3.3. 1 (b) so appropriate action could be taken before pressure reached the setpoints of pressure switches PS-03408, 03448, 03738, and 30508, which were set to alarm at the TS limits. The team was concerned that Rl-15A did not consider uncertainties such as stability and temperature effects and that the current total uncertainty was not adequate.

Considering the low alarm point of 207 psig, the calculated uncertainty allowance of +/-6.85 psig could result in an alarm at close to 200 psig, which was the TS limit. If additional uncertainties were added, the channel pressure switches could alarm after the TS pressure switches.

The licensee reanalyzed the setpoint for P/A-0363, 0367, 0369, and 0371 using additional appropriate uncertainty inputs and determined that the resulting instrument uncertainty was bounded by Rl-15A. The team observed that the results of these basis documents were determined to encompass specific additional uncertainties due to the assumed margins used in the documents to account for unquantified effects. The licensee had a guide entitled "Design & Maintenance Guide on Instrument Setpoint Methodology," EGAD-PROJ-16, Revision 0, and the team concluded that it provided a satisfactory methodology for setpoint calculations and was consistent with industry standard S67-04, Part I, "Setpoints for Nuclear Safety-Related Instrumentation." The licensee stated that EGAD-PROJ-16 provided identical guidance as EGAD-PROJ-08, Revision 0, of the same title, which was the current designation of the guide. The instruments that were re-analyzed during the inspection used the guidance of EGAD-PROJ-08.

This methodology affirmed that margins remained bounded. The licensee stated that use of this guide was not required by plant procedures.

However, the licensee has previously recognized from past assessments that its basis documents were not as rigorous as required by the current /SA standards.

The licensee stated that EGAD-PROJ-08 was being revised and that the appropriate procedures would be revised to require its use. The team identified this item as Inspection Fol/owup Item 50-255197-201-12.

Palisades 60 Day Response:

None of the above calculational deficiencies identified during the Design Inspection affected the operability of any safety-related equipment.

During the inspection, EGAD-ELEC-08 Rev 1 was approved and issued to provide for instrument setpoint methodology.

Our engineering staff 15

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS has been briefed as to the need to utilize this guidance.

Plant procedures will be revised by August 15, 1998, to incorporate EGAD-ELEC-08 for use when setpoint calculations are required.

10/1/98 Update: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-12) was identified as closed. Applicable plant administrative procedures have been changed to reference guidance document EGAD-ELEC-08 for use when performing setpoint calculations, and enhanced to more clearly . describe the applicability of EGAD-ELEC-08.

No further actions are planned for this inspection item. Unresolved Item 50-255/97-201-13 During a walkdown of the SI system, the team observed that transmitters for containment spray flow, FT-0301 and FT-0302, and shutdown cooling heat exchanger flow, FT-0306, were properly mounted below their flow elements, but the process tubing was observed to be inadequately sloped back to the transmitters.

Additionally, a walkdown performed by the licensee at the team's request during an

• in-containment inspection revealed that the process lines to the HPSI cold-leg flow transmitters FT-0308, FT-0310, FT-0312, and FT-0313, and the LPSI flow transmitters, FT-0307, FT-0309, FT-0311, and FT-0314, were also installed with inadequate slope. The team was concerned that inadequate slope in instrument tubing could contribute to significant instrument uncertainty by entraining unequal amounts of air in either leg of the transmitter, causing erroneous readings.

This was shown to be a valid concern when an operator observed an erroneous reading in the left channel containment spray loop indicator, Fl-0301A.

The "below zero" reading was caused by air trapped in one of the process iines. The licensee issued CR C-PAL-97-1561 to vent the line. The lack of tubing slope was inconsistent with original plant installation specification J-F020, Revision 0. This specification stated: "Flow instruments (differential tyP.e) in liquid and condensable vapor service shall preferably be mounted below the main line connection so that the impulse lines will slope down to the instrument." The specification also stated: "Impulse lines to flow instruments shall slope (up or down) a minimum of one inch per foot." Plant drawings J-F133, Revision 1;

• J-F134, Revision O; J-F140, Revision O; and J-F141, Revision 0, depict various acceptable installation configurations for a differential transmitter.

The current installations of the flow instruments identified above were not consistent with these drawings.

A later specification, J-465 (Q), "The Technical Specification for Installation of Instrumentation For Nuclear Service for CPCo Palisades," Revision 0, dated 1981 stated: "The installation shall be neat in appearance, properly supported, and shall provide for proper slope for adequate drainage or venting of the instrument lines." This specification has since been incorporated into specification 20557-J-59 (Q) under the same title, which requires that a "horizontal tubing run is continually sloped in accordance with design drawings." The licensee issued CR C-PAL-97-1561 to evaluate these instrument tubing sloping discrepancies.

According to the operability determination of the CR, the instruments have never shown any adverse effects of trapped air during the last 20 years of operation.

The HPSI and LPSI flow transmitters were mounted as much as 8 ft above their flow elements.

To accommodate instruments mounted above flow elements, specification J-F020 stated: "5 foot minimum "drop legs (equivalent of a loop seal)" may be required before the tubing is sloped up the I 16

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS meter." Plant drawings J-F152, Revision 1, and J-F153, Revision 0, depict these mounting configurations.

The licensee stated that the bottom and side tap locations for the tubing would tend to limit the amount of air getting into the transmitters and that air entrainment would be minimal due to the ratio of the volume of the HPSI and LPSI pump suction piping to the tubing volume. EA-C-PAL-95-0877D, "Evaluation of the Potential for Excessive Air Entrainment Caused by Vortexing SIRWT During a LOCA," Revision 0, evaluated the potential for excessive air entrainment in the lines of the pumps caused by vortexing in the SIRWT during a LOCA, and determined that the air f]ntrainment would be a small percentage of the flow volume. The licensee also stated that technicians are required to vent the transmitters during every 18 month surveillance.

However, the team was concerned that, since the transmitters sense low static pressure during normal standby operation, air may accumulate between calibration intervals and between system tests. Additionally, the water circulated through the SI lines from the containment sump could contain significant amounts of dissolved gasses, which could enter the tubing up to the flow transmitters.

The team was concerned that the effect of air entrapped in the instrument tubing could cause large and unquantifiable errors in the flow indications.

EOP Supplement 4, "Loss of Coolant Accident Recovery Safety Function Status Check Sheet," contained curves presenting total SI flow ranges intended to help ensure that the minimum values utilized in the accident analyses (LOCA, MSLB, Steam Generator Tube Rupture (SGTR)) were met. There was also a minimum total flow criterion for the operators to meet, which ensured the containment sump check valves remained in a stable condition in EOP-4. 0, "Loss of Coolant Accident Recovery," Revision 9. The operators would use the HPSI and LPSI flow indication from FT-0308, 0310, 0312, 0313, 0307, 0309, 0311, and 0314 to compare SI system performance against the EOP requirements.

The team was concerned that the potentially large errors could confuse the operator and impair decision making. The licensee stated that the opetators are trained to use all available indications and that alternate/additional instrumentation could be used to confirm trending of PCS conditions such as that for pressurizer level, subcooling margin, reactor vessel level, and charging pump flows. The licensee issued EAR-97-0699 to evaluate this item. It appeared that the design basis for instrument tubing installation was not implemented in the plant installation as required by 10 CFR Part 50, Appendix B, Criterion Ill, "Design Control." The team identified this item as Unresolved Item 50-255197-0201-13.

Palisades 60 Day Response:

During the Design Inspection, an operability determination was made concluding the HPSI and LPSI flow indication is operable based on plant operating experience.

Since the inspection, a plant walkdown was conducted which revealed that the HPSI and LPSI tubing configuration met design requirements but did not conform to associated design drawings.

The existing tubing configurations

• were observed, and the tubing was determined not to be susceptible to air entrainment.

The

• conclusions reached from this walkdown review further justify the reliability of the HPSI and LPSI flow indication, although configuration discrepancies exist. By August 15, 1998, we will resolve the HPSl/LPSI flow indication tubing discrepancies and compare our design requirements to additional samples of safety related instrument tubing to identify any additional nonconformances with design criteria.

The programmatic design control aspects related to this issue will be addressed as identified in Attachment 8, Item 1. 17

• *

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS 10/1/98 Update: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-13) was identified as closed. This item was also the subject of a NOTICE OF VIOLATION (50-255/98003-07b) from the same letter. Palisades responded with additional information to the NRC under correspondence dated June 24, 1998, entitled "RESPONSE TO NOTICE OF VIOLATION AND NOTICE OF DEVIATION FROM INSPECTION REPORT 50-255/98003." Subsequent to the Design Inspection, Palisades walked down these installations during the 98 refueling outage and confirmed that the sensing lines for HPSI and LPSI flow transmitters FT-0308, 0310, 0312, 0313, 0307, 0309, 0311 , and 0314 are appropriately sloped -thus no deviations from design requirements exist. A sampling of other sensing lines associated with safety-related equipment were also walked down and confirmed to meet design requirements for sensing line slope. NRC correspondence dated August 3, 1998 rescinded this cited potential violation.

No further actions are planned for this inspection item. Unresolved Item 50-255/97-201-14 The team reviewed EA-ELEC-LDTAB-005, "Emergency Diesel Generator 1-1 & 1-2 Steady State Loading," Revision 4, and verified that the analysis was consistent with the design basis information in the FSAR. All required accident loads for a LOCA and a LOOP were identified and tabulated.

The electrical loads exceeded the continuous rating of the EOG during the first 32 minutes of operation but were below the EOG maximum 2-hour rating. One of the inputs to this analysis was the electrical toad estimate for LPSI pumps P-67 A and P-678. These electrical load estimates were based on the minimum hydraulic LPS/ pump performance used in EA-A-PAL-92-037, "Emergency Diesel Generator Loadings-First Two.Hours," Revision 1, which determined that LPSI pump flow would be* 3600 gpm. Although the LPS/ pump flow was conservative for evaluating LOCA mitigation, it was not conservative for determining the maximum load the EOG could experience during a LOCA. The team determined that the LPS/ pumps could pump 4500 gpm with one LPS/ pump discharging into all four injection loops as identified in EA-SDW-95-001, "Generation of Minimum and Maximum HPSllLPSI System Performance Curves Using Pipe-Flo," Revision

2. The team was concerned that the licensee had not analyzed for the worst-case electrical load demand on the EDGs. Preliminary evaluations by the_ licensee using the correct maximum loads indicated that the electrical loading on one EOG could be higher than that determined in EA-ELEC-LDTAB-005.

The licensee issued CR C-. PAL-97-1650 to review and correct all necessary electrical analyses and determined the EDGs to be operable.

The team reviewed EA-ELEC-VOL T-13, "Palisades Loss of Coolant Accident With Off$ite Power Available," Revision 0, which evaluated the ac voltage available during normal operating, refueling, and accident conditions.

The team noted that the calculation had not been revised since 1993 and . that the load magnitudes identified in EA-ELEC-LDTAB-005, Revision 4, and EA-SDW-95-001, Revision 2, had not been included.

The licensee reviewed the impact of the revised loads on EA-ELEC-VOL T-13 and determined that the changes had minimal effect on the analysis.

The team also noted that FSAR Section 8.3 stated that backfeeding via the main and station power transformers could be utilized; however, EA-ELEC-VOL T-13 had not analyzed this particular operating mode. The licensee stated that it had recognized that an analysis for backfeeding needed to be performed in 1994 and had issued AIR A-PAL-94-223 to create an analysis in order to bound 18

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS this condition of operation.

The licensee initiated C-PAL-97-1619 to review and update EA-ELEC-VOLT-13 for load changes. It appeared that the requirements of10 CFR Part 50, Appendix B, Criterion Ill, "Design Control," had not been met for EA-ELEC-LDTAB-005 an*d EA-ELEC-VOLT-13 in that the design basis had not been updated to document the actual plant parameters.

The team identified this item as part of Unresolved Item 50-255197-201-14.

Palisades 60 Day Response:

During the Design Inspection, an operability determination was made which concluded, based on an evaluation which bounded recent load changes, that the electrical system is operable.

Mechanical flow model analyses, which serve as input to the electrical load flow analyses, will be completed by December 15, 1998. The electrical load flow analyses, which will assure plant loads are accounted for and applicable operating scenarios are addressed, will be completed by August 15, 1999. A specific backfeed analysis will be completed by Januar}t 15, 1999. The programmatic design control aspects related to this issue will be addressed as identified in Attachment 8, Item 1. 1011/98 Update: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", th.is item (50-255/97201-14) was identified as closed. The mechanical and subsequent electrical flow model analyses are on target for completion by December 15, 1998 and August 15, 1999, respectively, as stated above. Backfeed analysis EA-ELEC-FL T-009, "GSU Short Circuit Analysis" was completed with design attributes captured in the applicable Design Basis Document.

Refer to Attachment 8, Item 1 for the programmatic "design control" aspects associated with this issue.

• Inspection Followup Item 50-255/97-201-15 FSAR Section 8.5.2 stated that cables would be sized in accordance with the National Electric Code (NEC) or Insulated Power Cable Engineers Association

(/PCEAllCEA) ampacity values and the cable ampacities would be adjusted on the basis of actual field conditions when possible.

The adjustments included conductor operating temperature, ambient temperature, cable overall diameter, raceway fill, and fire stops. The licensee had recently initiated a program to verify the adequacy of its cable ampacity sizing. EA-ELEC-AMP-032, "Ampacity Evaluation for Open Air Cable Trays With a Percent Fill Greater Than 30% of the Usable Cross Sectional Area," Revision 1, was issued in 1997 to address cable sizing. While reviewing the EA, the team noted the absence of fire stop derating and increased cable temperatures due to thermal radiation from hot pipes. The licensee had initiated AIR A-PAL-97-062 to evaluate the effects of local heat sources on fire stops; however, evaluation of the effects on cable degradation due to the close proximity of hot piping systems had not been included.

The licensee stated that evaluation of the effects of hot piping would be included under A-PAL-97-062.

The team identified this item as Inspection Followup Item 50-255197-201-15 . 19

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Palisades 60 Day Response:

We will complete our Cable Ampacity Sizing Program by September 15, 1998 which will identify any cable degradation due to the close proximity of hot piping, and any degradation of fire stops due to local heat sources. 10/1/98 Update: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-15) was identified as open. Cable

• degradation due to the close proximity of hot piping, and any degradation of fire stops due to local heat sources has been evaluated.

Results confirm that the cable design is acceptable.

No further actions are planned for this inspection item. Unresolved Item 50-255/97-201-16 The 120-V ac safety-related and non-safety-related loads were powered from instrument ac bus Y-01. Bus Y-01 was powered from either motor control center (MCC) 1or2 via automatic transfer switch Y-50. MCCs 1 and 2 were redundant safety-related busses. The licensee stated in a January 24, 1978, letter to the NRG that it would. implement the recommendation of RG 1. 6 in that no . provision would exist for automatically transferring loads between redundant power sources. The NRG issued a safety evaluation report, dated April 7, 1978, confirming the licensee's commitment.

FC-364, "Feeder Change for Instrument Bus Y-01," Revision 0, implemented this commitment and powered bus Y-01 from MCC 1 and non-safety-related MCC 3. However, FC-854, "Y-01 Power Supply Feed Modification," Re.vision 0, moved the backup power source from MCC 3 to the safety-related MCC 2, and resulted in a departure from the plant's licensing basis. The modification installed fuses in series with the existing breakers, which provided an additional level of protection for the two safety-related busses. The team observed that the safety evaluation performed for FC-854 did not identify that prior NRC approval was required.

The licensee issued CR C-PAL-97-1678 to document this deviation from the licensing basis. It appeared that this modification was a USO in that the possibility of a common-mode failure of the redundant safety-related busses was created, which was not previously evaluated in the FSAR and, thus, the criterion of 10 CFR 50.59(a)(2)(ii) was satisfied.

The team identified this item as Unresolved Item 50-255197-201-16.

Palisades 60 Day Response:

During the Design Inspection, an operability determination was completed which concluded that the implemented design meets the intent of RG 1.6 and provides a single failure proof method of preventir:ig the transfer of a fault between redundant load sources. The current configuration was implemented under FC-854 with the modification safety evaluation concluding that an unreviewed safety question does not exist. Prior NRC approval of the change was not required.

A description of the implemented modification was transmitted to the NRC in our Annual Report of Facility Changes, Tests and Experiments dated April 2, 1991. This 1989 modification resulted in a change to a prior NRC commitment.

In accordance with NEI guidelines, we will submit by November 1, 1998, a revised commitment which reflects the existing plant configuration and governing design basis. 20

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN iNSPECTION OPEN ITEMS 10/1/98 Update: Per NRG correspondence dated May 18, 1998, titled "NRG INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-16) was identified as closed. This item was also the subject of a NOTICE OF DEVIATION (50-255/98003-08) from the same letter. Palisades responded with additional information to the NRG \ . .mder correspondence dated June 24, 1998, entitled "RESPONSE TO NOTICE OF VIOLATION AND NOTICE OF DEVIATION FROM INSPECTION REPORT 50-255/98003." In summary, Palisades concludes that our commitment to assure that redundant safety related power sources cannot be both affected by a fault on the instrument bus has been maintained.

NRG correspondence dated August 3, 1998 concluded that a USQ does not exist, and that Consumers appropriately notified the NRG of past design changes, and rescinded this cited potential deviation.

No further actions are planned for this inspection item. Inspection Followup Item 50-255/97-201-17 The team observed that no system analysis existed to show that all the Class 1 E 120-V ac loads had *adequate voltages.

The licensee demonstrated during the inspection that adequate voltages did exist for selected loads. For example, EA-ELEC-VOLT-24, "Voltage Drop From Preferred AC Power Source Y10 Breaker 2 and Y40 Breaker 2 Out to the 5U12 Relays," Revision 0, showed that adequate ac voltage for those selected components was available at the minimum.inverter voltage. The licensee initiated CR C-PAL-97-1621 to evaluate and resolve this concern. The team identified this item as part of Inspection Fol/owup Item 50-255197-201-17.

Palisades 60 Day Response:.

During the Design Inspection, an operability determination was made concluding the Class 1 E 120 V

• ac loads are operable based on past plant operating experience and the expected minimal change in supplied voltage between normal and accident plant conditions.

By August 15, 1998,. we will perform a bounding analysis to confirm that Class 1 E 120 V ac loads have adequate voltage during accident conditions.

10/1/98 Update: Per NRG correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-17) is identified as open. A bounding calculation was performed under EA-C-PAL-97-1621A-01 that developed worst case voltage levels for the Preferred AC System and confirmed adequate available voltage during accident conditions . These analysis results will be incorporated into Design Basis Document DBD-4.03, "Preferred AC System" and tracked under change request number 4.03-12-R3-0728.

No further actions are planned for this inspection item . 21

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Unresolved Item 50-255/97-201-18 The team reviewed relay settings for protective relays associated with LPSI pump P-67 A, HPSI pump P-66A, SW pump P-7A, CCW pump P-52A, EOG 1-1 differential protection, bus 1C undervoltage protection, and Bus 1 C second-level undervoltage protection.

The settings were consistent with the design parameters of the devices being protected.

However, during the review, the licensee determined that the overcurrent relays for supply breakers 152-105 and 152-106 to bus 1C had not been calibration tested during the last refueling outage (1995) as required by Periodic and Predetermined Activity (PPAC) SPS025, "Bus 1 C Relay Testing." The licensee stated that these relays would be calibrated during the 1998 refueling outage. The licensee reviewed past calibration data for this type of relay and determined that negligible drift had previously been documented.

The licensee initiated CR C-PAL-97-1568 to resolve this discrepancy.

It appeared that the requirements of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," had not been implemented in this case in that certain relays had not been tested as required by the test program. The team identified this item as Unresolved Item 50-255197-201-18.

• Palisades 60 Day Response:

During the Design Inspection, an operability determination was made concluding that past calibrations of overcurrent relays for breakers 152-105 and 152-106 revealed insignificant drift and the relays are operable.

We will perform maintenance activity PPAC SPS025 to calibrate the overcurrent relays during the 1998 refueling outage. Our corrective action history identified no other examples of failure to perform scheduled relay testing. 10/1/98 Update: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-18) was identified-as closed. This item was also the subject of a NOTICE OF VIOLATION (50-255/98003-09) from the same letter. Palisades responded with additional information to the NRC under correspondence dated June 24, 1998, entitled "RESPONSE TO NOTICE OF VIOLATION AND NOTICE OF DEVIATION FROM INSPECTION REPORT 50-255/98003." In summary, the overcurrent relays for breakers 152-105 and 152-106 will be tested/calibrated by December 31, 1998. The requirements for PPAC SPS025 have been revised to allow performance of the testing and calibration while the plant is at power operation.

Unresolved Item 50-255/97-201-19 The team questioned the replacement schedule for Agastat E7000 series relays. The team was aware that the manufacturer, in correspondence to other utilities, had recommended a 10-year replacement schedule for these relays. The licensee stated that 52 E7000 series relays were installed and that 7000 series Agastats were also installed in Class 1 E applications.

Some circuits containing 7000 series relays included the 2400-V bus 1C and*1D supply breakers, time delay relays associated with charging pumps. P-55A, B, and C, and auto transfer failure alarms for 2400-V busses 1C and 10. The manufacturer's stated qualified life forthe E7000 relays was 10 years. The licensee stated that the qualified life applied if the relays were located in a harsh environment and, 22

• *

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS since the E7000 relays were located in a mild environment, no qualified life determination was required.

Based upon this justification, the licensee issued PPAC Deletion Form MSE 034, dated March 3, 1995, which stated that the relays would not require replacement at 10-year intervals.

The team believed that the qualified life stated by the manufacturer applied to any environment.

The team verified with the manufacturer that the projected qualified life of 10 years was the operating life of the E7000 series relay as long as the device did not exceed the equipment ratings, and that the life of 10 years was applicable to either a mild or harsh environment.

The licensee had not evaluated the qualified life ofthe 7000 series relays. The manufacturer of Agastat relays issued a 10 CFR Part 21 notification concerning the inability of the E7000 series relays to switch a 1-amp load at rated voltage. The licensee evaluated the installed E7000 series relays and identified no concerns.

The team observed that this evaluation did not review those 7000 series relays dedicated by the licensee to safety-related use. The licensee issued CR C-PAL-97-1663 to resolve the issues concerning Agastat relays and determined that all the relays were operable.

It appeared that the requirements of 10 CFR Part 50, Appendix B, Criterion Ill, "Design Control," had not been met in this instance in that the design basis lifetime for Agastat relays as stated by the manufacturer had not been correctly implemented in the facility.

The team identified this item as Unresolved Item 50-255197-201-19.

Palisades 60 Day Response:

During the Design Inspection, an operability determination was made concluding that the 7000 series relays are operable based on their similarity in application and design to E7000 relays. By July 15, 1998, we will complete our analysis of both 7000 and E7000 series relays dedicated for safety related use to confirm their ability to perform safety-related functions during their installed life and their conformance with applicable design requirements.

10/1/98 Update: Per NRG correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-19) was identified as open. A review of both 7000 and E7000 relay age-sensitive components was performed that indicates that all relay materials will last for greater than 40 years without significant degradation when installed in mild environments.

Based on this review, a 10 year replacement interval is not justified and the relays can be expected to perform their design function for greater than 40 years. No further actions are planned for this inspection item. Unresolved Item 50-255/97-201-20 The 125-V de system was divided into two independent systems. Each system consisted of a battery, switchgear, distribution panel, and two chargers.

Station battery 1, battery charger 1, and battery charger 3 supplied 125-V de bus 1. Battery charger 1 was supplied from MCC 1 and battery charger 3 was supplied from MCC 2. Administrative controls limited the operation so that only one charger per battery was in service. This prevented a common-mode failure from affecting both

• 23 ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS emergency busses. The supply to 125-V de bus 2 was similar, with battery charger 2 fed from MCC 2 and battery charger 4 fed from MCC 1. Operating Procedure SOP-30, "Station Power," Revision 20, required the battery chargers to be operated in pairs (1 and 2 or 3 and 4). The licensee stated that the battery chargers were swapped monthly to provide equal operating time for each battery charger. During swapping of the battery chargers in accordance with Section 7. 7. 2 of SOP-30, the 125-V de breaker on the in-service battery charger was opened and then the 125-V de breaker for the battery charger to be placed in service was closed. During this evolution, both battery chargers were disconnected from the station battery and 125-V de switchgear bus. Although temporary disconnecting the battery charger from the de bus had minimal safety impact on the plant, the team observed that TS 3. 7. 1 h required two station batteries and the de systems (including at least one battery charger on each bus) to be operable when the PCS was above 300 °F. The licensee stated that an LCO was not entered when no battery chargers were connected to the de busses. The licensee initiated CR C-PAL-97-1537 to resolve this discrepancy.

The team identified the licensee's failure to enter an LCO during battery charger switching evolution as Unresolved Item 50-255197-201-20.

Palisades 60 Day Response:

Prior to the Design Inspection, we concluded that our design bases were met and an LCO would not entered when realigning battery chargers.

This conclusion was based on no appreciable battery discharge occurring during the short realignment period when neither charger was connected to the 125 Vdc bus. In response to this Design Inspection item, however, operating procedure SOP-30 was revised in anticipation of an amendment approving our December 27, 1995 technical specifications change request. Although the requested change does not require a connected charger, the change defines 125 Vdc bus operability in terms of applied bus voltage. SOP-30 now requires entry into an LCO whenever performing charger realignment.

On January 26, 1998, a technical specification change request was resubmitted as part of the Improved Technical Specifications Program. An amendment in response to this latest change request will resolve this open item. 10/1/98 Update: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-20) was identified as closed. In July 1998, Amendment 180 of the Palisades Electrical Technical Specifications was implemented that clarifies the 125 Vdc system operational requirements.

With the issuance and implementation of Amendment 180, no further actions are planned for this inspection item. Inspection Followi.Jp Item 50-255197-201-21 The team reviewed the 125-V de battery loading during the normal and alternate battery charger alignment.

During the normal battery charger alignment, battery charger 1 was powered from EOG 1-1 and battery charger 2 was powered from EOG 1-2. During a LOCA combined with a LOOP in this normal alignment, the batteries would be without ac power for approximately 1 O seconds until the EDGs restored power. The team reviewed EA-ELEC-LDTAB-009, "Battery Sizing for the Palisades Class 1 E Station Batteries ED-01 and ED-02," Revision 2, which verified that the battery was sized to provide adequate power during the 10 second interval until the EDGs provided ac power to battery 24

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGNINSPECTION OPEN ITEMS chargers 1 and 2. During the alternate battery charger alignment with battery charger 3 powered from EOG 1-2 and battery charger 4 powered from EOG 1-1, the station batteries would be required to carry the de loads for more than 10 seconds in the event of a LOCA combined with a LOOP and a single failure of ac power. EA-ELEC-LDTAB-009 did not analyze the battery loading for station batteries ED-01 and ED-02 during this condition.

When questioned by the team the licensee stated that the de loading during this scenario would be greater than the worst-case loading assumed in ELEC-LDTAB-009.

The licensee issued CR C-PAL-97-1596 to resolve this discrepancy.

Additionally the team had concerns on whether the licensee met the single failure criterion when the alternate battery charger alignment was in effect. The team identified the question with respect to the single failure criterion and the additional loading on the battery as an Inspection Followup Item 50-255197-201-21.

Palisades 60 Day Response:

During the Design Inspection, an operability determination was made concluding that the station batteries are operable.

Operability was based on a preliminary analysis where additional

• conservative loads were included in the battery load analysis showing that the battery terminal voltage would be greater than the required minimum output of 105 Vdc throughout the exp.ected load duration until an operable charger would be connected to the bus. Operating procedures control alternate charger alignment but do not restrict this practice which is allowed by technical specifications.

By January 15; 1999, we will complete a formal analysis of battery loading considering the battery chargers are in their alternate alignment, and a combined event of a LOCA, LOOP and single failure of ac power occurs. 10/1/98 Update: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-21) was identified as open. As stated above, by January 15, 1999, the formal battery loading analysis will be completed.

Inspection Followup Item 50-255/97-201-22 The team identified that TS Section 4. 7.2c required that each station battery be demonstrated operable by verifying that the battery capacity was adequate to supply and maintain in an operable status all of the actual emergency loads for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when the battery was subjected to a battery service test. The battery service tests performed on station batteries ED-01 and ED-02 were performed for a duration of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The 4-hour duration and loading was based on the design basis station blackout (SBO) coping time. The team noted that the 2-hour requirement of TS 4. 7.2c was non-conservative with respect to the design basis, which required the station batteries to be available for4 hours. The design basis duration of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> was included in FSAR Section 8.4.2; DBD 4.01, "Station Batteries," Revision 3; RE-83A, "Service/Modified Performance Test-Battery No. ED-01," Revision 9, and RE-838, "Service/Modified Performance Test-Battery No. ED-02," Revision 9. Testing the batteries in accordance with RE-83A and B has ensured that batteries ED-01and02 have met the 4-hour design basis requirement.

The licensee has submitted TS changes to correct the non-conservative TS Section 4. 7.2c and issued CR C-PAL-97-1551 to resolve this discrepancy.

The team identified this item as Inspection Followup Item 50-255197-201-22.

25

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Palisades 60 Day Response:

During the Design Inspection, an operability determination was made concluding that the 4-hour SBO station battery load profile envelops the 2-hour OBA load profile. By January 15, 1999, we will complete a formal analysis of battery loading considering the battery chargers are in their allowed alignment configurations with a combined event of a LOCA, LOOP and.single failure of ac power. We submitted a technical specification change request on December 27, 1995 to describe the test profile as the design basis profile without stipulating a specific period for the profile. On January 26, 1998, a technical specification change request was resubmitted as part of the Improved Technical Specifications Program which identifies a four hour load profile for the service test. An amendment in response to this latest technical specifications request will resolve this open item. 10/1/98 Update: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-22) was identified as open. As identified above, by January 15, 1999, the formal battery loading analysis will be completed.

In July 1998, Amendment 180 of the Palisades Electrical.Technical Specifications was implemented.

Amendment 180 does not specify a duty cycle (profile) duration in units of time. Therefore, the design basis requirements found in the FSAR can be used. Inspection Followup Item 50-255/97-201-23 EA-ELEC-FL T-005, "Short-Circuit for the Palisades Class 1 E Station Batteries ED-01 and ED-02," . Revision 0, was submitted to the team as the short-circuit analysis for the Class 1E 125-V de system. The following discrepancies with the assumptions, methodology, and conclusions were identified:

• Section 4. 4 and 4. 5 assumed various breaker and fuse impedances, which had not been verified against the installed facility.

• Section 5. 2 utilized the battery charger current limit of 220 amps as the maximum short-circuit contribution without supporting documentation.

• Section 5.2 stated that the open-circuit voltage was 2.06 V per cell, whereas the EA utilized an open-circuit voltage of 2. 0 V per cell.

• Section 8. 0 stated that the results were to be further reviewed by the licensee; however, the team found no evidence of this review. Section 8. O also contained no conclusion about the de system acceptability.

The licensee issued A/Rs A-PAL-97-108, 109, and 110 to resolve these discrepancies.

The licensee stated that the* analyses would be reviewed and the conclusions revised. During the 1995 refueling outage, FES-95-206 replaced existing batteries ED-01 and ED-02. The team questioned if the sh9rt-circuit current provided by the new battery was analyzed and if there were any effects on the de distribution panel breakers, since the team noted that EA-ELEC-FL T-005 26

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS . had not been updated since 1994. The team also noted that the design basis for the evaluation of fault current contributions on de circuits was in FSAR Section 8.5.2, which stated "The 125 volt de protection design considers the fault current available at the source side of the feeder protective device." However, the licensee stated that the short-circuit contribution value for de circuits was taken at the electrical load terminals and not at the breaker load terminals (de short-circuit current value would be less when calculated at the load terminal vice the source side of the feeder protection device because voltage available at the load terminal would be less than at the source breaker).

The licensee determined that the short-circuit contribution at 8 breakers (breakers 72-101, 72-105, 72-106, 72-121, 72-127, 72-133, and 72-135) on distribution panels 011-1and011-2 could exceed the short-circuit interrupting ratings when evaluated in accordance with the design basis method in the FSAR. Also, when the team questioned the assumed breaker fault ratings on de busses 010, 020, 011-1, and 011-2of13,000 amps in EA-ELEC-FLT-005, the licensee was unable to show manufacturer or testing documentation to support this assumption.

The team believed that this assumption was inconsistent with its experience.

The licensee performed an operability review and issued CR C-PAL-97-1652 to resolve these discrepancies.

The maximum short-circuit current of the battery installed by FES-95-:206, as provided by the manufacturer, was 17094 amps. Calculation EA-ELEC-FL T-005 did not reflect this new short:..circuit current. Upon questioning by th.e team, the licensee stated that an evaluation was performed to ensure that the system short-circuits were acceptable.

During the team's review of this evaluation it was determined that the maximum battery short-circuit current was not utilized.

The.licensee stated that the short-circuit current utilized, 12,821 amps, was provided by the manufacturer as a more realistic value than 17,094 amps. However, the licensee could not document a basis for the 12,821 amps and stated that they would verify it with the manufacturer.

The team identified these discrepancies concerning EA-ELEC-FL T-005 as part of Inspection Followup Item 50-255197-201-23.

Palisades 60 Day Response:

During the Design Inspection, an operability determination was made concluding that a fault would more likely occur at the load rather than at the breaker terminals.

A fault at the load (esults in a reduced value of fault current which falls within the breaker interrupting rating. We have since obtained vendor specifications which envelop our calculated peak short circuit currents assumed to occur at the breaker terminals.

These specifications confirm our earlier conclusion that the breakers are suitable for their intended service, and resolve any concerns with respect to breaker short circuit interrupting capability.

Revisions to analysis EA-ELEC-FL T-005, to correct the plant-identified deficiencies described in the Design Inspection report, will be complete by January 15, 1999. 10/1/98 Update: Per NRG correspondence dated May 18, 1998, titled "NRG INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-23) was identified as open. Revisions to analysis EA-ELEC-FL T-005, to correct the plant-identified deficiencies described in the Design Inspection report, remains scheduled for completion by January 15, 1999 . 27

• **

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Inspection Followup Item 50-255197-201-24 FSAR Section 8.4.3.3 stated that the batteries were designed to furnish their maximum load down to an operating temperature of 70 °F without dropping below 105 V de, and that the equipment supplied by the batteries was capable of operating satisfactorily at this voltage rating. EA-ELEC-VOL T-026, "Voltage Drop Model of the Palisades Class 1 E Station Batteries D01 and D02," Revision 0, evaluated the de voltages at the distribution panels based upon a battery voltage of 105 V de, but did not evaluate the voltages that would be available at the load device terminals.

The team was concerned that the additional voltage drop from the distribution panel to the loads could result in voltages less than the design basis of the loads, and that no analysis was performed to evaluate this situation.

For example, the deign-basis minimum input voltage for the inverters was 105 V de and the licensee could not show any vendor documentation to support operating at a value Jess than 105 V de. The team noted that the inverters could be subjected to an input voltage of approximately 102 V de if the battery voltage were 105 V de. The licensee stated that battery surveillance testing has shown that battery voltage, when subjected to an SBO duty cycle, did not decrease below 108 V de. During the inspection, the licensee evaluated several safety-related loads and verified that adequate voltages would exist at 105 V de battery voltage. The licensee issued CR C-PAL-97-1620 to evaluate the lack of an EA to ensure that adequate voltages would exist at the load terminals.

The team identified this item as part of Inspection Followup Item 50-255197-201-24.

Palisades 60 Day Response:

During the Design Inspection, an operability determination was made concluding that the 125 Vdc system is operable based on an evaluation of several safety related loads, in which adequate load voltage was found to exist with a 105 Vdc battery terminal voltage. By November 15, 1998, we will perform a bounding analysis to identify the worst-case minimum voltage levels at the load assuring that minimum load voltage req.uirements are met.

• 1011198 Update: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-24) was identified as open. As stated above, this issue is scheduled for completion by November 15, 1998. Unresolved Item 50-255197-201-25 The team also questioned the capability of solenoid valves to operate at voltages of 87 V de as stated in DBD 1. 01, Cooling Water System," Revision 4. The licensee determined that the DBD was incorrectly worded and that the correct solenoid capability was90-140 V de. Upon further review, the licensee identified that improperly rated coils, rated 102-126 V de, were installed in solenoid valves SV-0918 and SV-09778.

The licensee initiated Engineering Assistance Request (EAR) 97-0652 to replace the coils. It appeared that the requirements of 10 CFR Part 50, Appendix B, Criterion Ill, "Design Control," were not followed in that the design basis for the solenoid valve coils was not implemented in the plant. The team identified this item as Unresolved Item 50-255197-201-25 . 28

• • ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Palisades 60 Day Response:

Since the design inspection, further evaluation identified that there is no impact on the mitigation of an accident if solenoid valves SV-0918 and SV-09778 fail to open due to low voltage since the close position is both the failed position and the required safety position.

Based on this review, the design basis is met by the existing solenoid valve installation.

The actions in response to Inspection Followup Item 50-255/97-201-24 will identify any other minimum voltage problems.

10/1/98 Update: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-25) was identified as closed. No further actions are planned for this inspection item. Inspection Followup Item 50-255/97-201-26 The team identified other discrepancies in calculations as follows:

• Assumptions

4. 6 and 4. 7 of EA-ELEC-VOL T-26, Revision 0, and assumptions

4. 8 and 4. 9 of EA-ELEC-M/SC-022, "Electrical Systems Model of the Palisades Class 1 E Safety Re/a.fed 125 V de System," Revision 1, assumed various fuse and breaker impedances which had not been verified against the installed equipment.

• Section 7. 0 of EA-ELEC-VOL T-26, Revision 0, "Conclusion," stated that the results were to be further reviewed by the licensee; however, the team found no indication that this review had been performed.

The "Conclusion" section also contained no statement concerning the de system acceptability.

• EA-ELEC-VOL T-26, Revision 0, utilized a correction factor for battery temperature of 77 °F instead of the correction factor for 70 °F, which was the minimum design basis temperature for the battery. The number utilized is less conservative and the licensee evaluated that the overall effect on voltages in the calculation would be less than 0. 5 percent.

• EA-ELEC-LDTAB-029, Revision 2, stated the type of battery constant as 1.0 in Attachment A and 1.4 on Sheet 4. The constant to be utilized depended on the type of battery. 1. 0 referred to a lead acid battery; 1.4 referred to a nickel-cadmium battery. The licensee reviewed the EA and determined that the correct constant was utilized in the EA and that the reference to 1. 4 was an editorial error. The licensee issued CR C-PAL-97-1656 to address the battery temperature correction factor and stated that the other discrepancies would be corrected in future revisions to the calculations.

The team identified this item as part of Inspection Fo//owup Item 50-255197-201-26.

29

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Palisades 60 Day Response:

During the Design Inspection, an operability determination was made concluding that the calculation deficiencies identified had no affect on the analyses conclusions; ie, supplied voltages remain within equipment ratings and the station batteries are not affected.

By January 15, 1999, EA-ELEC-VOLT-26, EA-ELEC-MISC-022 and EA-ELEC-LDTAB-029 will be revised to resolve the deficiencies noted above. 10/1/98 Update: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-26) was identified a:s closed. Analyses EA-ELEC-VOLT-26, EA-ELEC-MISC-022 and EA-ELEC-LDTAB-029 will be revised by January 15, 1999 as projected above. Inspection Followup Item 50-255/97-201-27 The team noted that TS Section 4. 7.1.b required testing to be performed at every. refueling to demonstrate the overall automatic operation of the emergency power system. Proper operation was verified by bus load shedding and automatic starting of selected motors and equipment to establish that emergency power had been restored within 30 seconds. FSAR Tables 8-6 and 8-:-7 stated that sequencing would occur in 65 seconds. Technical Surveillance Procedure RT-BC, "Engineered Safeguards System -Left Channel," Revision 8, and RT-8D, "Engineered Safeguards System -Right Channel," Revision 8, required performance testing to be within the 65-second requirement.

The team questioned the use of a 30-second test duration in the TS instead of a 65-second duration, which would demonstrate that all required equipment would start. The licensee stated that the TS did not specifically require full testing of the entire diesel load sequence but only required testing of selected loads. The team noted that the licensee was testing the diesel loading to the full accident loading sequence and has submitted a proposed TS change which would be more consistent with the current design. The team reviewed Test Procedures R0-128-1, "Diesel Generator 1-1 24 Hour Load Run," Revision 2, and R0-128-2, "Diesel GeneratOr 1-2 24 Hour Load Run," Revision 2. The team noted that Section 3. O of the Acceptance Criteria and Operability Sheet for Procedure R0-128-2 referred to TS Section 3. 7. 1 and 4. 7. 1. 11, and that these references would only be correct when the proposed improved TS, which have been submitted to NRG for approval, became effective.

The licensee issued CR C-PAL-97-1566 to resolve these discrepancies.

The team identified this item as Inspection Followup Item 50-255197-201-27.

Palisades 60 Day Response:

Several issues identified in the Design Inspection are associated with interpretation of existing Technical Specifications.

On December 27, 1995 we submitted an electrical technical specifications change* request which served to resolve the discrepancy noted above pertaining to the Emergency Diesel Generator (EOG) load sequence test. On January 26, 1998, we submitted a request for improved technical specifications which specifies testing the EOG to the load 30 ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS intervals programmed by the sequencer; eliminating any specific reference to the sequence time. It is expected that the amendment resulting from the most recent .technical specification change request will serve to resolve this and other technical specification related open items. 1011198 Update: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-27) was identified as closed. In July 1998, Amendment 180 of the Palisades Electrical Technical Specifications was implemented.

Amendment 180 specifies testing the EOG to the load intervals programmed by the sequencer; eliminating specific reference to the sequence time. No further actions are planned for this item. Inspection Followup Item 50-255197-201-28 The team identified the following discrepancies when reviewing station battery Test Procedures RE-83A, "Service/Modified Performance Test-Battery No. ED-01," Revision 9, and RE-83B, "Service/Modified Performance Test-Battery No. ED-02," Revision 9:

• The tests evaluated whether the final discharge voltage (105 V de) of station batteries ED-01and02 was met at the end of the test (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />). Load parameters (amps) at 1 and 239 minutes were not verified during the test. These load parameters were design requirements of EA-ELEC-LDTAB-009, Revision 2. The licensee demonstrated that the 1-and 239-minute data were recorded elsewhere and that the duty cycle was* tested in accordance with the design requirements.

The licensee stated that the battery testing procedures would be revised to include verification of these design parameters.

• The procedures did not require any calibration tolerances for the discharge testing shunt and control unit. The licensee stated that the tolerance was removed from the procedure before testing during the 1996 refueling outage and issued PCRs 5422 and 5423 to change the procedures to include these tolerances.

• The battery charging data in Procedure RE-83B for the 1996 refueling outage did not meet Step 5. 2. 2, which required the battery charging rate to be decreasing and to remain within 5 percent over the last 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> before stopping the equalization process, in that the process was stopped before the end of the 8-hour period. The licensee stated that the nearly steady state voltage operation of the charger gave adequate assurance that the battery was operable before exiting the test and issued CR C-PAL-97-1460 to resolve this discrepancy.

• During the performance of procedure RE-83B at the 1996 refueling outage, the elapsed time recorded manually did not agree with the testing control unit time. The licensee stated that because the testing unit did not have the capability to record the time, the test start and stop times were recorded manually.

The inconsistencies were minor and had no effect on the test results. The licensee issued C-PAL-97-1460 to evaluated this discrepancy.

The team identified this item as Inspection Followup Item 50-255197-201-28.

31

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Palisades 60 Day Response:

Note: Inspection Followup Item 50-255/97-201-28, Unresolved Item 97-201-30 bullets 7, 8, 9, 10, 11 and 12, and Unresolved Item 97-201-31 bullets 6 and 13 are completed under this action due to their subject similarity.

Surveillance tests RE-83A and RE-838 will be revised as appropriate to eliminate the identified deficiencies to support 1998 refueling outage performance.

By December 15, 1998, we will review DC system requirements, FSAR Chapter 8 and surveillance tests RE-83A and RE-838 for consistency, and resolve the deficiencies identified in this open item and the following:

• Reconcile FSAR section 8.2.3 concerning the battery supplying safe shutdown loads for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> with the requirement to strip loads. (Inspection report item #30-7.) *

• Disposition battery shunt and de tie breakers which are not consistent with FSAR section 8.3.5.2. (Inspection report item #30-8.)

• Reconcile one battery charger capability to supply normal loads and recharge battery in less than 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with FSAR section 8.3.5.3. (Inspection report item #30-9.)

• Reconcile alternate alignment of battery chargers with FSAR section 8.4 .. 2.2. (Inspection report item #30-10.)

• Reconcile battery chargers cross connection with FSAR section 8.5.2. (Inspection report item #30-11.)

• Reconcile design of system 1, 2, 3, 4 circuits and their separation requirements with FSAR section 8.5.3.2. (Inspection report item #30-12.)

• Add battery discharge restriction to the D8D. (Inspection report item #31-6.)

• Disposition battery cell specific gravities. (Inspection report item #31-13.) 10/1/98 Update: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION, this item (50-255/97201-28) was identified as open. Surveillance tests RE-83A and RE-838 were revised and satisfactorily performed during the 1998 refueling outage. The June 30, 1998 FSAR revision resolved inspection report items #30-8, #30-9, and #30-12. The above remaining items are scheduled to be complete by December 15, 1998 .. Inspection Followup Item 50-255/97-201-29 The team reviewed the following electrical modification packages and found them consistent with the plant design basis: 32

• * * * * * *

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Temporary Modification TM-96-027, "lnsta/1152-Spare

1. 5 Breaker in 152-113 Cubicle," dated April 10, 1996 FES-95-206, "ED-01 and ED-02 Station Battery Replacement," Revision O FC-364, "Feeder Change for Instrument Bus Y-01," Revision O FC-854, "Y-01 Power Supply Feed Modification," Revision 0 FC-638, Add Component Cooling Water Pumps to the Normal Shutdown Sequencer," Revision 0 FC-798, "Battery Room Temperature Indication and Alarm," Revision O FC-683, "Removal of Pressurizer Heaters from SIS Trip," Revision O Except as previously discussed, all these modifications were adequately prepared, provided the necessary technical basis for the changes, and contained adequate installation instructions and testing requirements.

The 10 CFR 50. 59 safety evaluations were adequate, except for the two listed below: = Safety Reviews 95-1431and95-1432, dated July 7, 1995, for FES-95-206 stated that the battery duty cycle service test duration for station .batteries ED-01 and ED-02 was changed from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The licensee noted that TS Section 4. 7.2.c was affected by this design change. However, the USQ evaluation, Question 2 of Section II, was not checked "Yes" for a TS change. TS 4. 7.2.c required that a 2-hour battery test be performed; while design analysis ELEC-LDTAB-009 and FSAR Section 8.4.2 required a 4-hour battery duty cycle. The licensee has submitted a proposed TS change to reflect the proper battery test duration and issued CR C-PAL-97-1551 to address this discrepancy.

• The safety review documentation for TM-96-027 stated that the FSAR was not reviewed.

Administrative Procedure

3. 07, "Safety Evaluations," page 12, required that the FSAR be reviewed and that thos*e sections reviewed be noted on the safety review sheet. The licensee initiated .C-PAL-97-1493 to evaluate this discrepancy.

The team identified these safety review discrepancies as Inspection Fol/owup Item 50-255197-201-29. Palisades 60 Day Response:

It was not documented in the safety evaluation for FES-95-206 that a technical specification change would be required to change the battery duty cycle service test duration from 2 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. An FES-95-206-specific technical specifications change was not considered necessary by the preparer of the safety evaluation since a technical specifications change request eliminating reference to a specific duty cycle time was to be submitted under the Improved Technical Specifications Program in the near term. Since completion of the FES-95-206 safety evaluation, Palisades has implemented a Safety & Design Review Group which reviews and approves all design changes and safety evaluations.

The purpose for forming and employing this group is to provide consistent oversight The quality of safety evaluations and their reviews has significantly improved over the recent years. It is unlikely that a safety evaluation deficiency, similar to that associated with FES-95-206, would have occurred since deployment of the Safety & Design Review Group. 33

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS The original safety review for TM-96-027 inappropriately indicated that FSAR sections had not been reviewed.

In reality, the FSAR was reviewed during safety review preparation and the FSAR was found to contain description at a level of detail that the TM would not affect. The review of the TM-96-027 safety review was performed by telecon (an infrequent practice) with no follow-up review performed by the Safety & Design Review telecon reviewer.

By April 15, 1998, design control procedures will be revised to require a follow-up review whenever a review is performed by telecon. 1011198 Update: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND.NOTICE OF VIOLATION", this item (50-255/97201-29) was identified as closed. Administrative Procedure AP 3.07, "SAFETY EVALUATIONS" was revised to require follow-up reviews as stated above. No further actions are planned for this inspection item. Note: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-30) was identified as open. FSAR changes identified in Unresolved Item 50-255/97201-30 are identified below. Some of these bullets are grouped and evaluated with other URl's or IFl's. For clarity, each bullet's actions will be separately addressed.

Unresolved Item 50-255197-201-30 The team identified the following discrepancies in the FSAR:

• Page 6. 7-4 stated that 'containment isolation valves fail closed with loss of voltage or control air except for the CCW return isolation valves. However, the CCW supply isolation valve (CV-0910) is also a fail-open valve and should have *been noted as an exception to fail-closed containment isolation valves. The licensee issued FSAR Change Request 6-142-R20-1426 to correct the FSAR. Palisades 60 Day Response:

The next FSAR annual update revision will incorporate this change. 1011198 Update: Annual FSAR update issued June 30, 1998, included this change.

• Section 6. 7 classified the CCW penetrations as Class C-2, which was defined as penetrations with lines not missile protected.

However, EA-GW0-7793-01 stated that the entire CCW system (both inside and outside containment) was missile protected.

The licensee issued FSAR Change Request 6-143-R20-1427 to state that the CCW penetrations were not vulnerable to internally generated missiles . 34

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Palisades 60 Day Response:

The next FSAR annual update revision will incorporate this change. 10/1/98 Update: Annual FSAR update issued June 30, 1998, included this change.

• Table 9-10 stated that valves 3029 and 3030, containment sump suction valves, failed closed upon loss of air and were equipped with an accumulator.

The valves actually failed as is and had no accumulator.

The licensee issued FSAR Change Request 9-293-R20-1431 to correct *the FSAR and CR C-PAL-97-1559 to evaluate and trend the FSAR discrepancies being identified at the plant. Palisades 60 Day Response:

The next FSAR annual update revision will incorporate this change. 10/1/98 Update: Annual FSAR update issued in June 30, 1998, included this change.

• Table 9-9 correctly stated that the high-pressure air piping was seismic Class I from the receivers to the valve operators.

However, FSAR Table 5.2-3 stated that the entire system was seismic Class I. The licensee issued FSAR Change Request 5-155-R20-1432 to correct the FSAR 5. 2-3. Palisades 60 Day Response:

The next FSAR annual update revision will incorporate this change. 10/1/98 Update: Annual FSAR update issued in June 30, 1998, included this change.

• Section 8.4.2.2 stated that the station batteries would be tested to Institute of Electrical and Electronics Engineers (IEEE) 450-1975.

However, battery testing procedures RE-83A, Revision 9, and RE-838, Revision 9, referred to IEEE 450-1995.

FSAR Change Request 8-126-R20-1249 had been initiated, but the licensee did not intend to act on this change until approval was received from NRG of a related proposed TS change. Palisades 60 Day Response:

This FSAR change is on hold until the license amendment responding to our improved electrical technical speeification change request, submitted January 26, 1998, is received.

This change cites IEEE 450-1995 for the battery testing . 35

• ** ATTACHMENT A

• STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS 1011198 Update: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-30) was identified as open. In July 1998, Amendment 180 of the Palisades Electrical Technical Specifications was implemented with IEEE 450-1995 as a reference.

FSAR change 8-126-R21-1249 will be implemented as part of the next annual FSAR update. to reflect the use of this IEEE standard.

• Table 5. 7-8 listed the seismic design value for the station batteries and racks as "later" instead of including the actual values of the batteries installed by FES-95-206.

The licensee issued EAR-97-0636 to evaluate this discrepancy and revise the FSAR. Palisades 60 Day Response:

The table in the FSAR is designated as containing the original seismic design values for the plant. The term "later" was an original FSAR description which acknowledged that an impending upgrade to install a second redundant electrical train would be made and the applicable seismic criteria would not be available until then. Since we have chosen to keep this table for historical record, the word "later" will be removed and the table maintained as original seismic criteria.

The next FSAR annual update will incorporate this change requested by FSAR Change Request 5-157-R20-1456.

1011198 Update: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this portion of unresolved item 50-255/97201-30 was identified as closed. The annual FSAR update issued June 30, 1998, included this change. No further actions are planned for this inspection item.

• Section 8.2.3 stated the "The de battery system is designed to supply the required shutdown loads, with a total loss of ac power, for at ieast 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />." This statement did not reflect the fact that load stripping was required during the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for the battery to perform its intended function during a loss of ac power.

• Palisades 60 Day Response:

Refer to our response to Inspector Followup Item 50-255/97-201-28.

1011198 Update: The resolution of this issue is addressed in Inspection Followup Item 50-255/97201-28 due to subject similarity.

This item is projected to be complete by December 15, 1998. Section 8. 3. 5. 2 stated that "Operation of all circuit breakers in the de and the preferred ac systems is manual with automatic trip for fault isolation." The battery shunt trip breakers and the de bus tie breakers do not comply with this statement.

36

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Palisades 60 Day Response:

Refer to our response to Inspector Followup Item 50-255/97-201-28.

10/1/98 Update: Revision 20 of FSAR Chapter 8 incorporates the exclusion of the battery isolation shunt trip breakers and tie breakers between the left and right sections of each switchgear bus that do not have an automatic trip for fault isolation.

Our June 30, 1998, annual FSAR update includes this change.

• Section 8. 3. 5. 3 stated that "Each of the two battery chargers provided on the. de bus is capable of supplying the normal de loads on the bus and simultaneously recharging the battery in a reasonable time. A fully discharged battery can be recharged in less than nine hours." Contrary to the statement, one battery charger could not supply the normal loads and recharge a fully discharged battery in less than 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />. Palisades 60 Day Response:

Refer to our response to Inspector Followup Item 50-255/97-201-28.

10/1/98 Update: Revision 20 of FSAR Chapter 8 now states that two battery chargers are needed to recharge a fully discharged battery in less than nine hours. Our June 30, 1998, annual FSAR update includes this change.

• Section 8.4.2.2 stated that "Emergencv Operation

-.On loss of normal and standby ac power, the batteries will supply power to all preferred ac and de loads, until one of the (diesel generators)

DGs has started and can supply power for the chargers." This statement was not correct if the battery chargers were in their alternate alignment and did not reflect load shedding during the 4-hour duration.

Palisades 60 Day Response:

Refer to our response to Inspector Followup Item 50-255/97-201-28.

10/1/98 Update: The resolution of this issue is addressed in Inspection Followup Item 50-255/97201-28 due to subject similarity.

We plan to complete this item by December 15, 1998.

• Section 8.5.2 stated that The power source for the driven equipment and the control power for that system are supplied from the sources in one channel." This statement would not be correct if the battery chargers were cross-connected . 37

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Palisades 60 Day Response:

Refer to our response to Inspector Followup Item 50-255/97-201-28.

10/1/98 Update: The resolution of this issue is addressed in Inspection Followup Item 50-255/97201-28 due to subject similarity.

We plan to complete this item by December 15, 1998. *

• Section 8.5.3.2 referred to "System 1, 2, 3, 4 Circuits" and separation requirements for those circuits.

The licensee was not able to identify these circuits.

• Palisades 60 Day Response:

Refer to our response to Inspector Followup Item 50-255/97-201-28.

10/1/98 Update: Revision 20 of FSAR Chapter 8 expands the definition along with providing routing and isolation requirements for 'left', 'right' and channel '1 ', '2', '3', and '4' circuits.

Our June 30, 1998, annual FSAR update includes this change.

• Section 8.4.1.3 required clarification as to whether the reserve capability margin referred to the capability of the overall EDG and engine or if it referred to the capability of the EOG to handle an increase loading due to a control circuit ma/function during the loading sequence.

The licensee issued C-PAL-97-1309 to resolve this discrepancy.

Palisades 60 Day Response:

Prior to the Design Inspection, an operability determination was made concluding that the EDGs are operable.

This conclusion was reached based on the capability of the EDGs to provide the required design function in the event of a control. circuit malfunction or delayed containment high pressure signal; but not both concurrently.

The design basis accident analysis does not require that these two events occur simultaneously.

Due to the change being descriptive in nature, rather than licensing basis information, we have elected to use the Design Basis Documents rather than the FSAR to make the clarification.

Design Basis Document Change 5.03-11-R3-0617 was initiated and the revision will be made by December 15, 1998. 10/1/98 Update: Revision 4 of DBD 5.03 incorporates the requested change which evaluated the system functional requirements of the EOG starting and carrying the largest load due to a control circuit malfunction.

Revision 4 also includes discussion regarding the EOG control circuit malfunction and starting a containment spray pump during a delayed containment high pressure scenario;

• concluding that the malfunction and the pump start are mutually exclusive.

No further actions are planned for this item. 38

• .ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS

• Section 6.1.2.3 stated that The RAS ... provides a permissive to manually close the valves in the pump minimum flow lines." EOP-4. 0, "Loss of Coolant Accident Recovery," Revision 9, Step 23, directed the operators to place the hand switches for these valves in the pump minimum flow lines (CV-3027 and CV-3056) to CLOSE when SIRWT level lowered to between 25 percent and 15 percent. Per EOP-4.0, Step 51, the RAS occurred when the SIRWT level reached 2 percent. The FSAR appeared to conflict with EOP-4.0. The licensee initiated FSAR Change Request 6-141-R20-1425 to update the FSAR. Palisades 60 Day Response:

The next FSAR annual update revision will incorporate this change. 10/1/98 Update: Annual FSAR update issued in June 30, 1998, included this change.

• The footnote for Table 14.17.1-1 implied that a containment building temperature of 90 °F was used as input to the large-break LOCA analysis because it is the limiting temperature during normal operation.

The 90 °F value did not appear to be limiting.

The licensee stated that the 90 °F value was the nominal containment building temperature, not the limiting temperature, and was used in the accident analysis in accordance with Seimens Power Corporation's large-break LOCA methodology guidelines.

The licensee initiated FSAR Change Request 14-95-R20-1441 to update the FSAR.

• Palisades 60 Day Response:

The next. FSAR annual update revision will incorporate this change. 10/1/98 Update: Annual FSAR update issued in June 30, 1998, included this change. The above discrepancies had not been corrected and the FSAR had not been updated to ensure that the material in the FSAR contained the latest material as required by 10 CFR 50. 71(e). The team identified this item as Unresolved Item 50-255197-201-30.

Palisades 60 Day Response:

10 CFR 50.71(e) requires that the FSAR be updated to contain the latest material developed and that it includes the effects of all changes made in the facility or procedures described in the FSAR. Although several of the identified FSAR discrepancies were clear errors, most were cases where statements in the FSAR were misleading or unclear and not cases where the FSAR was not updated per 10 CFR 50.71 (e). Our ongoing FSAR verification and validation effort should provide identification and correction of similar conditions which may exist in the FSAR. Our processes were also changed a few years ago to require a safety review (1 O CFR 50.59 screening) for 39 ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS all analyses, modifications, etc which have the potential to affect the design basis of the facility.

This widespread 10 CFR 50.59 screening will prevent failures to update the FSAR in accordance with 10 CFR 50.71(e).

In addition, a license basis self assessment performed in accordance with NEI 96-05, "Guidelines for Assessing Programs for Maintaining the Licensing Basis," found few discrepancies in the FSAR sections sampled which had not been previously identified for correction by other plant processes.

Therefore, we feel that the current efforts underway will correct other errors which may exist in the FSAR and the current plant processes will ensure that the FSAR is updated properly.

10/1/98 Update: The above response remains unchanged from our 60-day response.

Note: Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-31) was identified as open. DBD changes identified in Unresolved Item 50-255/97-201-31 are identified below. Some of these bullets are grouped and evaluated with other UR l's or IFl's. For clarity, each bullet's actions will be separately addressed.

Unresolved Item 50-255/97-201-31 The team identified the following discrepancies in the DBDs:

• DBD 1.07, Auxiliary Building HVAC Systems," Revision 1, Table 3.2.1, incorrectly stated that the design basis temperature for Room 123, which contains the CCW pumps, was 125 °F. The correct temperature was 104 °F as stated in 080 7.01, "Electrical Equipment Qualification Program," Revision 1, Appendix A. The 125 °F temperature was a conservative assumption used to size the outside air supply fans. Table 3.2.1 also contained a typographical error in a reference number. The licensee issued 080 Change Requests 1.07-71-R1-0512 and 1.07-72-R1-0532 to correct the 080. Palisades 60 Day Response:

The identified Design Basis Document Change Request will be incorporated into the DBD by July 1, 1998. 10/1/98 Update: Revision 2 of DBD 1.07, "Auxiliary Building HVAC Systems" incorporated the above changes. The basis for the 125 ° F CCW room temperature was clarified and references were corrected.

• 080 1.07, Revision 1, Section 3.2.1.3, listed maximum room temperatures for the west ESF room from an outdated analysis.

The latest analysis, EA-O-PAL-93-272F-01, "Engineering 40

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Safeguards Room Heatup Following LOCA in Conjunction With a Loop," Revision 0, determined lower maximum room temperatures for various SW flows through the air coolers. The 080 also required clarification of the normal design temperature of the ESG room. The licensee issued 080 Change Request 1.07-73-R1-0543 to correct the 080. Palisades 60 Day Response:

The identified Design Basis Document Change Request will be incorporated into the DBD by July 1, 1998. 10/1/98 Update: Revision 2 of DBD 1.07, "Auxiliary Building HVAC Systems" incorporated the above change. The basis for the 135°F Engineering Safeguards Room temperature was clarified.

• 080 7. 08, "Plant Protection Against Flooding, 77 Revision 1, incorrectly stated that the EOG would be inoperable before a flood reached the EOG windings because the lube oil heaters were located below the windings at 7 inches above the floor. EA-C-PAL-95-1526-01, "Internal Flooding Evaluation for Plant Areas Outside of Containment, 77 Revision 0, stated that the minimum flood level at which the EOG could become inoperable was 10 inches due to the exciter cubicle bus bars and that the lube oil heaters were not needed for EOG

• operability.

The licensee issued CR C-PAL-97-1557 to initiate a 080 change and evaluate the item. Palisades 60 Day Response:

During the Design Inspection, an operability determination concluded that the EDGs

• are operable based on other indications available to inform operations that water level in the rooms is increasing.

DBD change request 7.08-40-R1-0561 was initiated to state that the limiting component is not lube oil heaters but the exciter cubicle bus bars located ten inches above the EOG room floor. The identified Design Basis Document Change Request will be incorporated into the DBD by December 15, 1998. 10/1/98 Update: This DBD change is on target for completion by December 15, 1998 as identified above.

• 080 2. 03, "Containment Spray System, 77 Revision 2, stated that the air supply to the sump outlet valves, CV-3029 and 3030, was backed by an accumulator.

There were no accumulators for these valves. The licensee identified this error while evaluating an FSAR statement that these valves had an accumulator backup that was questioned by the team, and issued 080 Change Request 2.03-22-R2-0531 to correct the 080 . 41

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Palisades 60 Day Response:

The identified Design Basis Document Change Request will be incorporated into the DBD by July 1, 1998. 10/1/98 Update: Revision 3 of DBD 2.03, "Containment Spray System" corrected the terminology from "accumulator" to "high pressure air receivers".

No further action is planned.

• DBD 1.01, "Component Cooling Water System," Revision 3, Section 3.3. 7, incorrectly indicated that Class 1 E and non-Class 1 E breakers were installed in the same distribution panels. The licensee initiated DBD Change Request 1.01-14-R3-0518 to correct the DBD. Section 3. 3. 7 of this DBD also stated that solenoid valves had been tested to operate at 87 V de instead of 90 V de. The licensee stated that the DBD would be corrected.

Palisades 60 Day Response:

The identified Design Basis Document Change Request will be incorporated into the DBD by July 1, 1998. 10/1/98 Update: Due to competing priorities, this DBD change has been rescheduled to be completed by December 15, 1998. * *

• During the teain's review of FES-95-206, it was noted that the battery manufacturer had imposed a limit of 40 battery discharges for the 20-year life of the battery. This restriction had not been identified in any DBD. The licensee stated that the requirement would be added to DBD4.01. . Palisades 60 Day Response:

A Design Basis Document Request will be incorporated into the DBD by December 15, 1998. Refer to our response to Inspector Followup Item 50-255/97-201-28.

10/1/98 Update: This DBD change is on target for completion by December 15, 1998, as above.

• Appendix A of DBD 7. 02, "Palisades Design Basis Document EQ Master Equipment List," Revision 2, incorrectly listed the location for L T-0383; referred to EIP 0343 instead of E/P 0346; and did not include SV-32138 in Table A-1. The licensee issued DBD Change Requests 7. 02-4-R2-0522, 7. 02-6-R2-0527, and 7.D2-4-R2-0523 to correct the DBD . 42

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• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS Palisades 60 Day Response:

The identified Design Basis Document Change Request will be incorporated into the DBD by December 15, 1998. 10/1/98 Update: These DBD changes are on target for completion by December 15, 1998.

• DBD 2.01, "Low Pressure Safety Injection System," Revision 3, and DBD 2.02, "High Pressure Safety Injection System," Revision 3, both contained references to ANF-88-107, "Palisades Large Break LOCNECCS Analysis With Increased Radial Peaking," Revision 1. ANF-88-107 was superseded by Seimens Calculation EMF-96-172, "Palisades Large Break LOCNECCS Analysis," Revision 0. The licensee Initiated DBD Change Requests 2. 01-30-R3-0519 and 2.02-27-R3-0520 to update the DBDs.

• Palisades 60 Day Response:

The identified Design Basis Document Change Request will be incorporated into the DBD by July 1, 1998. 10/1/98 Update: Revision 4 of DBD 2.01, "Low Pressure Safety Injection System," and Revision 4 of DBD 2.02, "High Pressure Safety Injection System," incorporated reference to the most current LOCA analysis.

No further action is planned for this item. DBD 2.01, "Low Pressure Safety Injection System," Revision 3, Section 3.3.1.3, stated that the SIRWT must maintain a minimum of 20,000 gallons at the time of a RAS to limit the radiological consequences of an accident.

The DBD reference for this statement was TAM-95-05, "Radiological Consequences for the Palisades Maximum Hypothetical Accident & Loss of Coolant Accident," Revision 0. A review of EA-TAM-95-05 indicated that this analysis did not take credit for the 20,000 gallons at the time of RAS to limit the radiological consequences of an accident.

The licensee issued DBD Change Request 2.01-31-R3-0524 to update the DBD. Palisades 60 Day Response:

The identified Design Basis Document Change Request wlll be incorporated into the DBD by July 1, 1998. 10/1/98 Update: Revision 4 of DBD 2.01, "Low Pressure Safety Injection System," clarifies the SIRW tank minimum volume design requirements.

No further action is planned for this item . 43

• *

• ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS The team also identified the following discrepancies in other documentation:

• P&ID M-232, Sheet 2A, incorrectly identified L T-0383 as connected to penetration

1. 54 instead of#56. The licensee issued Document Change Request (OCR) 97-0856lo correct the drawing.

Palisades 60 Day Response:

P&ID M-232, Sheet 2A has been reviseo to incorporate OCR 97-0856. 10/1/98 Update: No further update necessary.

• Documents E-33, Revision 46, and E-37, Revision 46, were not revised to reflect the installed condi(ion of the battery charger cabling that was rerouted by SC-89-284.

The licensee issued CR C-PAL-97-1495 to resolve this discrepancy.

Palisades 60 Day Response:

E-33, Rev 46 and E-37, Rev 46 have been revised to reflect the correct battery charger cable routing installed by SC-89-284 . 10/1/98 Update:

• No further necessary.

• P&ID M-209, Sheet 3 (Revision 34), incorrectly depicted valves SV-0918 and SV-09778 as normally deenergized.

The licensee issued EAR 97-0652 to revise the drawing.

• Palisades 60 Day Response:

P&ID M-209, Sheet 3, Revision 35 has been issued to depict SV-09778 as normally energized.

Further evaluation of SV-0918 identified that the normally deenergized state as depicted on M-209 Sheet 3 is appropriate per FSAR Table 9-10. 10/1/98 Update: No further update necessary.

• Vendor drawing E-12A, Sheet 39, Revision 0, indicated that the battery discharge characteristics were based upon battery cell specific gravities of 1.215 +/- 0.005. However, the batteries were being maintained to a criterion of 1.215 +/- 0.010. The licensee issued EAR 97-0669 to update the drawing. Palisades 60 Day Response:

E-12 A, Sheet 39, Rev O will be updated by December 15, 1998. Refer to our response 44 ATTACHMENT A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION OPEN ITEMS to Inspector Followup Item 50-255/97-201-28.

10/1/98 Update: This item is on target for completion by December 15, 1998. These documentation discrepancies were not consistent with 1 O CFR Part 50, Appendix B, Criterion Ill, "Design Control," which requires that the design basis be correctly translated into drawings.

The team identified this item as Unresolved Item 50-255197-201-31.

The programmatic design control aspects related to this issue will be addressed as identified in Attachment B, Item 1. 'i 45

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• ATTACHMENT B CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255 STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE 6 Pages

• ATTACHMENT 8 STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE Per NRC correspondence dated May 18, 1998, titled "NRC INSPECTION REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", inspection item 50-255/98003-01 was identified as open. As stated in the report, this item will remain open pending NRC review of the results of the collective significance of individual inspection items and planned programmatic improvements.

The following summarizes of our programmatic improvements.

1. DESIGN CONTROL ISSUES: The following issues were identified in the Design Inspection report as potentially not meeting requirements of 10 CFR 50, Appendix B, Criterion Ill, "Design Control." Our design control program provides assurance that the plant as-built configuration conforms to design requirements, and the configuration is operated, tested and maintained within required design parameters.

The deficiencies identified during the Design Inspection relate to these design control program objectives.

Design Objective For Operating Systems Within Design Parameters:

• Loss-Of-Coolant Accident analysis identified the maximum CCW temperature of 184°F yet the effects of this temperature on CCW system components was not performed. (Unresolved Item 50-255/97-201-02.)

• Incomplete analysis (inadequate justification for conclusion and incorrect references to related NRC correspondence) for CCW piping for High Energy Line Break. (Unresolved Item 50-255/97-201-04.)

• Some AC Load calculations have not been updated to reflect current design. (Unresolved Item 50-255/97-201-14.)

Design Objective For As-Built Conditions Conforming To Design Requirements:

• *

• Unscreened Emergency Core Cooling System Suction piping vent. (Unresolved Item 50-255/97-201-10.)

Some instrument tubing is not sloped consistent with design requirements . (Unresolved Item 50-255/97-201-13.)

Design Basis Document I design documentation discrepancies. (Unresolved Item 50-255/97-201-31.)

1

• ATTACHMENT B STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE Palisades 60 Day Response:

Elements comprising and supporting our.design control program consist of our calculation control program, instrument setpoint program, FSAR verification and validation (V&V), design basis documents (DBDs) with associated safety system design confirmations, and as-built confirmation through drawing review or field walkdown.

These elements will be revised as appropriate by December 15, 1998 to prevent the recurrence of conditions similar to those identified in the Design Inspection and cited above. Resolution of any nonconforming conditions identified will be implemented through our corrective action program. 10/1/98 Update: Programs exist at Palisades that ensure proper station design attributes are considered, evaluated, changed and documented.

These programs makeup our overall "Design Control" program. In past months, several programs have been reviewed in various inspections and routine assessments such as:

• NRC INFORMATION NOTICE 98-22:"DEFICIENCIES IDENTIFIED DURING NRC DESIGN INSPECTIONS" was evaluated by comparing the adequacy of our program design controls against other station Design Inspection identified concerns.

• Self assessments were performed in areas such as design document control and modification programs.

• NRC inspections and internal NPAD audits in the areas of Engineering and Technical Support were performed in mid 1998 that evaluated several Palisades design and configuration program attributes.

As a result of these and other efforts, "Design Control" Program enhancements have been identified and incorporated into the appropriate programs.

For example, several changes have been made to design change processes to better define the applicability of each distinct process, and to ensure that design change inpuUoutput requirements are adequately addressed.

2

• ATTACHMENT B STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE No major programmatic weaknesses were identified in these reviews and program enhancements are now complete.

To conclude, the Palisades "Design Control" Program is considered effective.

2. CALCULATION CONTROL ISSUES: The Design Inspection issues identified below reflect weaknesses in our calculation control program. Improvements in our calculation control program will serve to prevent recurrence of these conditions.

Inspection Report Issues:

• Required justification for conclusion and correct references to related NRC correspondence not provided in analysis. (Unresolved Item 50-255/97-201-04.)

• Not all analyses revised whenever analytical inputs or major assumptions change. (Unresolved Item 50-255/97-201-07.)

• Analyses not reflecting accurate as-built configuration and system operation, not all interdependent analyses have been revised together in response to changes, and analytical design bases do nofagreewith test acceptance criteria. (Unresolved Item 50-255/97-201-08.)

Palisades 60 Day Response:

Prior to the Design Inspection, calculation control weaknesses were recognized and an improvement plan was implemented.

Over 19,000 calculations have .been indexed to provide for improved retrievability.

A cross-index between selected calculations of record and the documents that use the results of the calculations is being developed.

These and other improvements to our calculation program serving to prevent recurrence of the deficiencies cited above will be made by December 15, 1998. 10/1/98 Update: The identification of calculations referenced in the major design documents has been completed.

The Calculation Control Improvement Project is on target for 3

• *

• ATTACHMENT 8 STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE completion of the detailed calculation cross-index by December 15, 1998. Development of the computerized calculation retrieval application and completion of associated engineer training will follow in early 1999. 3. SETPOINT CONTROL ISSUES: Station procedures and guidance to require the use of established uncertainty methodology need to be implemented.

The plan for implementation should be validated against weaknesses identified in* Unresolved Item 50-255/97-201-12.

Palisades 60 Day Response:

An instrument uncertainty evaluation methodology manual has been developed.

Uncertainty calculations for Reactor Protection System and Engineered Safety Features Actuation System setpoints have been performed Ul?ing .the methodology manual. Incorporation of instrument uncertainty evaluation requirements in procedures, and training select engineers to perform uncertainty calculations, will be completed by December 15, 1998. 10/1/98 Update: As stated in Inspector Follow-up Item 50-255/97201-12, station procedures have been revised to consider use of established instrument uncertainty guidance when developing test acceptance criteria and determining errors for operating instrument loops. In addition, a self assessment of the Setpoint Control Process was performed with potential areas for improvement being evaluated.

4. 10 CFR 50.54(F} RESPONSE:

Evaluate inspection findings, both specific and programmatic, against the Palisades response to NRC's October 9, 1996 request for information pursuant to 1 O CFR 50.54(f) regarding adequacy and availability of design bases information.

Palisades 60 Day Response:

After review of the inspection findings and comparison to our response to the 1 O CFR 50.54(f) letter regarding the adequacy and availability of design basis .4

..

• ATTACHMENT B STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE information, we have determined that our response to the 10 CFR 50.54 (f) letter remains complete and accurate.

Improvements to our design programs, initiated through our response, will be directly responsible for resolution of issues

• identified within the Design Inspection report. The programs and projects being improved include our Calculation Control Program, Setpoint Methodology and Control Program, FSAR Verification and. Validation Project, and our Fuse Control Program.

• Beyond programmatic improvements, design basis knowledge will be further enhanced by the development of 1 O additional DB Os and the performance of. three safety system design confirmations similar to the NRC's safety system functional inspections.

To date, four of the new DBDs have been issued and one design confirmation has been completed.

No additional programmatic improvement efforts have initiated as a result of actions being taken to satisfy our 10 CFR 50.54(f) response.

A final review of the adequacy of our response will be completed by December 15, 1998. 10/1/98 Update: Some of the initiatives noted in our 60-day response to the Des_ign Inspection were not part of Palisades formal response to the NRC's October 9, 1996 request for information pursuant to 10 CFR 50.54(f) regarding adequacy and availability of design bases information.

Our February 6, 1997, 50.54(f) response coneluded that the Palisades' design bases information was adequate, and reasonabie assurance exists that: 1) design bases information has been translated into operating, maintenance, and testing procedures, and 2) system, structures, and component configuration and performance are consistent with the design bases. Our 50.54(f) response also referred to specific initiatives to further strengthen plant processes and design basis documentation.

Specifically noted as

• commitments in the 50.54(f) response were: 1) performing an FSAR Verification Project, 2) completing ten new Design Basis Documents, 3) conducting one Safety System Functional Type inspection per fuel cycle, and 4) updating and re-instituting use of a Quality Assurance Requirements Matrix database.

Other initiatives to strengthen plant processes and design basis documentation were also undertaken that were not specifically included ln the 50.54(f) response 5

• ATTACHMENT B STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE such as: 1) implementing a calculation control improvement project, 2) implementing improvements in instrument setpoint uncertainty methodology, 3) performing an assessment of instrument setpoint control, and 4) performing an assessment of the fuse control program. The 50.54(f) response remains complete and accurate.

The response to Attachment B Item 1 relates to and supports this position.

It should be noted, however, that the 50.54(f) response and its committed programmatic initiatives, along with other initiatives noted above, will not resolve all issues identified within the Design Inspection since it is more effective to resolve certain issues on an individual, basis. A formal review that evaluates the Design Inspection findings against the 50.54(f) response is on target for completion by December 15, 1998. 6