ML18106A643

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LER 98-006-01:on 980227,determined Incorrect Scaling Error of First Stage Pressure Transmitter Existed.Caused by Human Error.Revised Setpoint Calculation SC-MS002-01 & Revised Associated Instrument Calibr Database info.W/980529 Ltr
ML18106A643
Person / Time
Site: Salem PSEG icon.png
Issue date: 05/29/1998
From: Bakken A, Bernard Thomas
Public Service Enterprise Group
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-98-006, LER-98-6, LR-N980254, NUDOCS 9806080289
Download: ML18106A643 (10)
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Text

PS~G Pl;Jblic Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit MAY 2 9 1998 LR-N980254 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 LER 311/98-006-01 SALEM GENERATING STATION - UNIT 2 FACILITY OPERATING LICENSE NO. DPR-75 DOCKET NO. 50-311 Gentlemen:

This Supplemental Licensee Event Report entitled "Incorrect Scaling of the First Stage Turbine Impulse Pressure Transmitters" is being submitted pursuant-to the requirements of the Code of Federal Regulations 10CFR50.73(a)(2)(i)(B).

Sincerely, Ja-LP2at~

(j ~*r /i.e. $:;Jie->r ///

A. C. Bakken Ill General Manager -

Salem Operations Attachment BJT )

C Distribution LER File 3.7 I,

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NRCFORM366 U.S.N R REGULATORY COMMISSION OVED BY OMB NO. 3150-0104 (4-95) EXPIRES 04/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS.

REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSEE EVENT REPORT (LER) LICENSING PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (T-6 F33J, U.S. NUCLEAR (See reverse for required number of REGULATORY COMMISSION, WASHINGTON, DC 205 5-0001, AND TO THE PAPERWORK REDUCTION PROJECT ~150-0104), OFFICE OF digits/characters for each block) . MANAGEMENT AND BUDGET, WASHINGTON, C 20503.

FACILITY NAME (1) uu<..r.- ,._,,, __ , (2) PAGE(3)

SALEM GENERATING STATION UNIT 2 05000311

  • 1OF9 TITLE (4)

Incorrect Scaling *of the First S~age Turbine Impulse Pressure Transmitters EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR I SEQUENTIAL NUMBER IREVISION NUMBER MONTH DAY YEAR FACILITY NAME SALEM UNIT 1 ~ .

DOCKET NUMBER 05000272 02 27 98 98 -- 006 -- 01 05 29 98 FACILITY NAME DOCKET NUMBER OPERATING 5 ITHIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more) (11)

I MODE (9)

POWER I 20.2201 (b)

  • 20.2203(a){1) 20.2203(a)(2)(v) 20.2203(a)(3)(i)

X 50.73(a)(2)(i)

50. 73(a)(2)(ii) 50.73(a)(2)(viii)
50. 73(a)(2)(x) 000 LEVEL (10) 20.2203(a)(2)(i)
  • 20.2203(a)(3)(ii) 50. 73(a)(2)(iii) 73.71
/,°': Y<": 20.2203(a)(2)(ii) 20.2203(a)(4) 50. 73(a)(2)(iv) OTHER 20.2203(a){2)(iii) 50.36(c)(1) 50. 73(a)(2)(v) Speci~ in Abstract below or in NR Form 366A

',o, 20.2203(a){2)(iv) 50.36(c)(2) 50.73(a)(2)(vii)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER (Include Area Code)

Brian J. Thomas, Licensing Engineer 609-339-2022 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TONPRDS TO NPRDS SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR

!YES (If yes, complete EXPECTED SUBMISSION DATE). xi NO SUBMISSION DATE (15)

ABSTRAl; 1 (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

On February 27, 1998, Engineering personnel determined that a scaling error of the first stage pressure transmitter existed when Salem Unit 2 operated from August 1997 to February 1998. The P-13 permissive (1 of 2 Turbine impulse chamber pressure channels ;::: a pressur~ equivalent to 11 % of rated thermal power) is an "or" input along with permissive P-10 (2 of 4 Power Range Neutron Channels;::: 11 % of rated thermal power) for the P-7 permissive as defined in Technical Specification (TS) Table 3.3-1. Therefore only one of either of the P-13 or P-10 permissives js necessary to actuate the P-7 permissive. The P10 permissive was not affected by the scaling problem and therefore would have actuated the P-7 permissive at the proper value. However with the first stage pressure transmitters scaled at a higher pressure for 100% thermal power, the P-13 permissive setpoint was actually higher than the value required by TS Table 3.3-1 (above 11 % thermal power).

The cause of occurrence for the incorrect scaling of the Unit 2 first stage pressure transmitters is attributed to human error This event is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B), any condition prohibited by the plant's Technical Specifications.

NRC FORM 366 (4-95)

NRC FORM 366A .S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

SALEM GENERATING STATION UNIT 2 05000311 YEAR I SEQUENTIAL NUMBER I REVISION NUMBER 2 OF 9 98 - 006 - 01 TEXT (If more space is required, use additional copies of NRG Form 366A) (17)

PLANT AND SYSTEM IDENTIFICATION Westinghouse - Pressurized Water Reactor Reactor Control and Protection (RCP) {JC/-}*

  • Energy Industry Identification System (EllS) codes and component function identifier codes appear as {SS/CCC}.

CONDITIONS PRIOR TO OCCURRENCE At the time of discovery, Salem Unit 2 was in Mode 5 and Salem Unit 1 was in Mode 4.

DESCRIPTION OF OCCURRENCE On February 27, 1998, Engineering personnel determined that a scaling error of the first stage pressure transmitter existed when Salem Unit 2 operated from August 1997 to February 1998. The first stage pressure transmitters, 2PT505 and 2PT506, were incorrectly scaled for the turbine pressure expected at 100% power. During the period of operation of Unit 2 above, the first stage turbine pressure was identified as reading approximately 555 psia with the Unit at 100% thermal power. 2PT505 and 2PT506 were calibrated in accordance with setpoint calculation SC-MS-002-01, Revision 1, dated May 29, 1997. In accordance with SC-MS-002-01, Revision 1, pressure transmitters 2PT505 and 2PT506 were scaled with a span of 0 to 690 psia corresponding to 0 to 120% power. The equivalent pressure at 100% thermal power based on this scaling is 575 psia.

Therefore the first stage pressure transmitters were indicating a lower first stage turbine pressure when Unit 2 was at 100% thermal power.

The basis of the 690 psia is derived from the original thermal performance (heat balance) data developed by Westinghouse for the Unit 2 Turbine-Generator and documented in Public Service Blue Print (PSBP) 131382 dated February 15, 1972. Reactor Engineering also confirmed on May 16, 1995, that the Unit 2 first stage* turbine pressure value at 100% rated thermal power would be 573 psia. Based on this documentation, setpoint calculation SC-MS-002-01 was developed and issued on August 8, 1996, for the scaling of the first stage pressure transmitters for a span of 0 to 690 psia corresponding to 0 to 120% reactor power. However, in June of 1995 just prior to the extended shutdown of Salem Unit 2, thermal performance data* indicated that the first stage turbine impulse pressure at 100% reactor power was approximately 552 psia.

NRC FORM 366A (4-95)

NRC FORM 366A .* NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

SALEM GENERATING STATION UNIT 2 05000311 YEAR I SEQUENTIAL NUMBER IREVISION NUMBER 3 OF 9 98 - 006 01 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

DESCRIPTION OF OCCURR~NCE (cont'd)

During the Salem Unit 2 outage prior to September of 1997, the Unit 2 low pressure turbine blades were replaced. As a result of the new low turbine pressure blades, Westinghou~e provided new thermal performance (heat balance) data. The new thermal performance data for the low pressure turbine replacement indicated that the first stage turbine pressure corresponding to 100% thermal power was lower than the value in the 1972 thermal performance heat kit (heat balance). However, discussion with Westinghouse indicated that the replacement of the low pressure turbine should not have had an effect on the first stage turbine pressure. Westinghouse has stated that the change in first stage pressure was due to a new heat balance calculation method. PSE&G is continuing to pursue further explanation of the change in first stage pressure with Westinghouse.

The first stage turbine pressure transmitters provide inputs to the Reactor Protection and Control system. The first stage turbine pressure transmitters provide input for the following safety and non-safety related functions:

Safety Related

  • The high steam line flow variable setpoint is adjusted based on the first stage turbine pressure. High steam line flow coincident with either low Tavg or low steam line pressure generates a Safety Injection (SI) signal and a Main Steam Isolation (MSI) signal.
  • The P-13 permissive (1 of 2 Turbine impulse chamber pressure channels ~ a pressure equivalent to 11 % of rated thermal power) is an "or" input along with permissive P-10 (2 of 4 Power Range Neutron Channels ~ 11 % of rated thermal power) for the P-7 permissive as defined in Technical Specification (TS) Table 3.3-1. Therefore only one of either of the P-13 or P-10 permissives is necessary to actuate the P-7 permissive. The P-7 permissive prevents or defeats the automatic block of the reactor trip on low flow in more than one primary coolant loop, reactor coolant pump undervoltage and under-frequency, pressurizer low pressure, pressurizer high level, and the opening of more than one reactor coolant pump breaker.
  • Permissive P-2 inhibits the automatic rod withdrawal on low turbine impulse pressure below a setpoint equivalent to 15% of full power.

NRC FORM 366A (4-95)

NRC FORM 366A .S. NUCLEAR REGULA TORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

SALEM GENERATING STATION UNIT 2 05000311 YEAR I SEQUENTIAL NUMBER I REVISION NUMBER 4 OF 9 98 - 006 01 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

DESCRIPTION OF OCCURRENCE (cont'd)

Non-Safety Related

  • The AMSAC circuitry is armed at a first stage turbine impulse pressure equivalent to 40%

reactor power.

  • AT-ref (reactor coolant temperature reference value) signal based on first stage impulse pressure provides an interlock for the condenser steam dumps.
  • The rod control system receives a T-ref signal based on first stage turbine pressure for comparison with Nuclear Instrumentation System (NIS) power level to determine direction and speed of control rod motion. Also the T-ref signal based on first stage turbine pressure is compared with Tavg in the rod control system.
  • First stage turbine impulse pressure also provides a T-ref input to the Main Control Room indicator for reactor coolant temperature.

With the actual first stage turbine impulse pressure corresponding to 100% Reactor Power being at a lower value than the pressure transmitters were scaled, for a given pressure of turbine load the actual reactor thermal power level would be at a higher value (i.e., 10% turbine load based on impulse pressure would be greater than 10% reactor thermal power).

Technical Specification (TS) Table 3.3-1 requires a P-7 setpoint of "... 1 of 2 Turbine Impulse chamber pressure channels~ 11% of RATED THERMAL POWER". With the first stage pressure transmitters scaled at a higher pressure for 100% thermal power, the P-7 (P-13 turbine load input) permissive setpoint was actually higher than the value required by TS Table 3.3-1 (above 11 % thermal power).

Since the actual turbine impulse pressure for the P-7 permissive correlated to a Rated Thermal Power above 11 %, the setpoint for the P-13 permissive input to the P-7 permissive was non-conservative.- There was no impact to the P-10 permissive input for the P-7 permissive. Therefore for the period of September 1997 to February 1998, Salem Unit 2 operated with a non-conservative TS setpoint. The sensor calibration procedures for the Unit 2 first stage turbine impulse pressure channels I (2PT505) and II (2PT506) were revised on February 28, 1998. The first stage turbine pressure channels 2PT505 and 2PT505 were calibrated using the revised procedures prior to Unit 2 entering Mode 2. This event is reportable in accordance with 10CFR50.73(a)(2)(i)(B), any condition prohibited by the plant's Technical Specifications.

NRC FORM 366A (4-95)

NRC FORM 366A .S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

SALEM GENERATING STATION UNIT 2 05000311 YEAR I SEQUENTIAL NUMBER I REVISION NUMBER 5 OF 9 98 - 006 01 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

DESCRIPTION OF OCCURR.ENCE (cont'd)

Also during the review of the scaling of the first stage pressure transmitters, a region of non-linearity of turbine load (first stage pressure) with respect to reactor thermal power was i.dentified. This area of non-linearity existed at the lower end of the first stage turbine impulse transmitters. The non-linearity of the turbine pressure with respeCt to the reactor thermal power also caused the setpoint for the P-7 (P-13 input for turbine pressure) permissive to be set at a non-conservative value for Salem Units 1 and 2. In the past, the P-7 permissive was set at a value of 10% turbine load based on the turbine impulse pressure, however based on the review of the non-linearity of the transmitter a 1.6%

offset for Unit 1 and 1.9 % offset for Unit 2 existed. Therefore the actual Rated Thermal Power level for the P-7 permissive for the turbine impulse transmitters would correspond to 11.6% for Unit 1 and 11.9% for Unit 2. These setpoints did not comply with the requirements of TS Table 3.3-1.

The channel calibration procedures for the Unit 2 first stage turbine impulse pressure channels I and II were revised on March 3, 1998. These procedures were revised to reflect the re-scaling of the Unit 2 first stage pressure transmitters and to ensure the proper setting of permissive P-13. However during a review of the corrective actions associated with this LER, no work order was identified as having performed the re-calibration of the P-13 permissive prior to Unit 2 entering Mode 2. Upon discovery that the revised channel calibration had not been performed, on March 26, the Unit 2 bistables associated with permissive P-13 were placed in the tripped condition. Placing the P-13 bistables in the tripped condition ensures that all the reactor trips associated with permissive P-7 are enabled at all reactor power levels. Subsequently, the P-13 permissive was re-calibrated using the revised channel calibration procedure.

CAUSE OF OCCURRENCE The cause of occurrence for the incorrect scaling of the Unit 2 first stage pressure transmitters is attributed to human error. During the review and issuance of setpoint calculation SC-MS002-01, the.

most current data available to verify the scaling of the pressure transmitters was not reviewed. There was no communication between the design engineering organization and the thermal performance engineer during the development of the setpoint calculation to verify the scaling information with empirical data.

During the development and review of the modification for the replacement of the Unit 2 low pressure turbine, the modification preparer (contractor) did not identify any impact to the instrumentation setpoint calculations although the revised thermal performance data provided by Westinghouse indicated a lower first stage turbine pressure at 100% reactor power. Also, the review of the modification package by the Instrumentation and Controls (I &C) Engineering group did not question the fact that no setpoint calculations were affected by the turbine replacement.

NRC FORM 366A (4-95)

NRC FORM 366A .S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE(3)

SALEM GENERATING STATION UNIT 2 05000311 YEAR I SEQUENTIAL NUMBER I REVISION NUMBER 6 OF 9 98 - 006 - . 01 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

The cause of occurrence for the failure to re-calibrate the Unit 2 P-13 permissive prior to the Unit 2 start-up is attributed to inadequate communication. Setpoint calculation SC-MS002-01 was revised to correct the P-13 interlock setpoint and channel calibration procedures were revised on March 3, 1998 to reflect the new P-13 setpoint. However, due to inadequate communication between the personnel involved in the revision of the P-13 setpoint, no corrective maintenance work orders were generated to perform the re-scaling of the Unit 2 P-13 setpoint in the field prior to the Unit 2 start-up.

PRIOR SIMILAR OCCURRENCES A review of LERs for Salem Units 1 and 2 for the past two years identified several LERs associated with personnel errors and inadequate communication, however, the corrective actions were specific to the events identified and would not have prevented this event from occurring.

SAFETY CONSEQUENCES AND IMPLICATIONS As discussed previously, the first stage turbine impulse pressure transmitters provide inputs to both safety related and non-safety related reactor protection and control systems.

The impact of the improper scaling for the safety related functions associated with the first stage turbine pressure transmitters is as follows:

  • High Steam Flow Reference Setpoint:

With the actual fi.rst stage turbine pressure lower than the scaling for 2PT505 and 2PT506, the setpoint for the high steam flow trip was conservative. The steam flow trip setpoint would be at a lower value for the actual reactor power level and therefore would be conservative.

  • P-7 Permissive The P-7 permissive is an "or" combination of the P-13 (turbine impulse pressure) and P-10 (N IS power) inputs. Although the P-13 portion of the P-7 permissive was set in the non-conservative direction, the P-10 permissive, 2 of 4 Power Range Neutron Channels~ 11 % of Rated Thermal Power, would ensure that the P-7 permissive operated at the proper value.

NRC FORM 366A (4-95)

NRC FORM 366A .* NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE(3)

SALEM GENERATING STATION UNIT 2 05000311 YEAR I SEQUENTIAL NUMBER I REVISION NUMBER 7 OF 9 98 - 006 01 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

SAFETY CONSEQUENCES AND IMPLICATIONS (cont'd)

  • P-2 Permissive The P-2 Permissive inhibits the automatic withdrawal of control rods below a turbine impulse pressure setpoint equivalent to 15% rated thermal power. The actual value that the P-2 permissive was set at was approximately 18.5% rated thermal power. Therefore, automatic control rod withdrawal was blocked up to 18.5% rated thermal power which is conservative for this function.

Blocking the automatic withdrawal of the control rods at low power levels requires the control room operators to perform control rod movement and oversee any changes of reactivity at low power levels.

Based on the above, although the first stage turbine pressure transmitters were scaled incorrectly, the reactor protection system would have operated when required to protect the health and safety of the public.

Although the first stage turbine pressure transmitters provide input to the Rod Control System, Steam Generator Level control, ADFCS, steam dump controls, and the T-ref recorder in the control room, these non-safety related control systems do not affect the ability of the reactor protection system to generate a reactor trip or initiate the required Engineered Safety Feature (ESF) equipment.

In accordance with the Salem UFSAR and the NRC safety evaluation report for AMSAC, the AMSAC system is armed at a value of 40% rated thermal power. The AMSAC interlock (C-20) for arming the system is set at 40% turbine impulse power. Setting the C-20 interlock at 40% turbine impulse power does not ensure that the AMSAC system is armed prior to 40% rated thermal power. However, in accordance with WCAP-11293-A the AMSAC system is designed to prevent overpressurization of the reactor coolant system. The Westinghouse analysis demonstrates that the AMSAC system is not required to actuate below 70% rated thermal power in order to limit the peak pressure in the reactor coolant system. Since AMSAC would continue to be armed before 70% rated thermal power, there is no safety consequences associated _with the incorrect scaling of the first stage turbine pressure transmitters.

CORRECTIVE ACTIONS

1. Setpoint calculation SC-MS002-01, "Turbine lmpul_se Pressure Scaling/Uncertainty Calculation" was revised on February 28, 1998. This calculation revised the scaling of the first stage turbine pressure transmitters for Salem Unit 2 to equate a turbine pressure of 555.69 psia for 100%

reactor power.

NRC FORM 366A (4-95)

NRC FORM 366A .S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

SALEM GENERATING STATION UNIT 2 05000311 YEAR I SEQUENTIAL NUMBER I REVISION NUMBER 8 OF 9 98 - 006 01 TEXT (If more space is required, use additional copies of NRG Form 366A) (17)

CORRECTIVE ACTIONS (cqnt'd)

2. Setpoint calculation SC-MS002-01 was revised on March 3, 1998, to ensure that Permissive P-13 is set at a value not to exceed 11 % Reactor Thermal Power. ~
3. The associated instrument calibration database information was revised on February 28, 1998, to reflect the re-scaling of the Unit 2 first stage turbine pressure transmitters.
4. The sensor calibration procedures for the Unit 2 first stage turbine impulse pressure channels I (2PT505) and II (2PT506) were revised on February 28, 1998. The first stage turbine pressure channels 2PT505 and 2PT506 were calibrated using the revised procedures prior to Unit 2
  • entering Mode 2.
5. The channel calibration procedures for the Unit 2 first stage turbine impulse pressure channels I and II were revised on March 3, 1998. These procedures were revised to reflect the re-scaling of the Unit 2 first stage pressure transmitters and to ensure the proper setting of permissive P-13.

The re-calibration of the P-13 permissive was completed on March 26, 1998.

6. The channel calibration procedures for the Unit 1 first stage pressure transmitter channels 1PT505 and 1PT506 were revised on March 27, 1998, to reflect the proper setting of permissive P-13. Calibration of these channels using the revised procedures was completed on March 27, 1998.
7. Procedure SC.IC-PT.SSP-0014(Q), "AMSAC Functional Test," was revised to reflect the re-scaling of the Unit 2 first stage turbine impulse pressure transmitters. This procedure adjusted the setpoint of the C-20 interlock for the arming of the AMSAC system. The revised AMSAC functional procedure was performed prior to Salem Unit 2 entering Mode 1. The revised AMSAC procedure was completed on Unit 1 on March 19, 1998.
8. l&C personnel involved with the review of the design modification have been held accountable for their actions in accordance with PSE&G policies.
9. This LER will be reviewed with the l&C Design Engineers to discuss the importance of verifying setpoint scaling information with current plant parameters by June 12, 1998. Also, an assessment of the technical standard for instrument setpoint calculations will be performed to determine if any enhancements are necessary.

NRC FORM 366A (4-95)

" NRC FORM 366A .S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

. TEXT CONTINUATION 1

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

SALEM GENERATING STATION UNIT 2 05000311 YEAR I SEQUENTIAL NUMBER I REVISION NUMBER 9 OF 9 98 - 006 01 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

CORRECTIVE ACTIONS (corit'd) 10.A lessons learned discussion was conducted with the l&C Design and System Engineers to ensure that the engineering review of modification packages assesses the overall impact to plant

. systems and to stress the necessity to ensure work orders are. generated when field work is*

required.

11.An assessment of the current procedures has determined that the processes in place for the control and oversight of plant modifications prepared by contract personnel provide adequate guidance and expectations for development of plant modifications.

. NRG FORM 366A (4-95)

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