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#REDIRECT [[B11638, Forwards Reviews of Integrated Safety Assessment Program Topics,Including 1.01 Re Gas Turbine Generator Start Logic mods,1.04 Re Reactor Water Cleanup Sys Pressure Interlock, 1.05 Re Ventilation Sys & 1.15 Re FSAR Update]]
| number = ML20137B883
| issue date = 08/13/1985
| title = Forwards Reviews of Integrated Safety Assessment Program Topics,Including 1.01 Re Gas Turbine Generator Start Logic mods,1.04 Re Reactor Water Cleanup Sys Pressure Interlock, 1.05 Re Ventilation Sys & 1.15 Re FSAR Update
| author name = Opeka J
| author affiliation = NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
| addressee name = Grimes C
| addressee affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR)
| docket = 05000245
| license number =
| contact person =
| document report number = B11638, NUDOCS 8508220164
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, UTILITY TO NRC
| page count = 24
}}
 
=Text=
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August 13,1985                                        l Docket No. 50-245                                      ,
B11638                                    I Director of Nuclear Reactor Regulation Attn:    Mr. Christopher I. Grimes, Chief Systematic Evaluation Program Branch U.S. Nuclear Regulatory Commission Washington, D.C. 20555
 
==References:==
(1)  3. F. Opeka letter to C. I. Grimes, dated May 17, 1985.
(2)  H. L. Thompson letter to J. F. Opeka, dated July 31, 1985.
Gentlemen:
Millstone Nuclear Power Station, Unit No.1 Integrated Safety Assessment Program In Reference (1), Northeast Nuclear Energy Company (NNECO) provided a proposed scope for the Integrated Safety Assessment Program (ISAP) review of Millstone Unit No.1. In Reference (2), the Staff formally issued the results of the ISAP screening review process, establishing the scope of ISAP for Millstone Unit No. I and initiating issue-specific evaluations. Reference (1) also indicated that for each issue or topic included in ISAP, NNECO would provide a discussion of the safety objective and an evaluation of the plant design with respect to the issue being addressed to identify specific items to be considered in the integrated assessment. In accordance with this commitment, reviews for the following ISAP topics are attached:
o  ISAP Topic 1.01 " Gas Turbine Generator Start Logic Modifications" o ISAP Topic 1.04 "RWCU System Pressure Interlock" o ISAP Topic 1.05 " Ventilation Systems" o ISAP Topic 1.15 "FSAR Update" o ISAP Topic 1.18 " Anticipated Transients Without Scram" o ISAP Topic 1.20 " Motor Operated Valve (MOV) Interlocks" o ISAP Topic 1.24 " Emergency Power" If you have any questions concerning the attsched reviews, please contact us.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY 8508220164ByBh45 ADOCK O o gDR                          PDR                            {(g  U J. F. Qpeka Senior Vice President cc:    J. A. Zwolinski
                                                                              \k*
 
ISAP TOPIC NO.1.01 GAS TURBINE GENERATOR S~ nRT LOGIC MODIFICATIONS l
 
r ISAP Topic No.1.01 Gas Turbine Generator Start Logic Modifications I. Introduction This topic resulted from the review of SEP. Topic VIII-2, On-Site Emergency Power Systems, for Millstone Unit No.1.          The SEP- review included an evaluation of the on-site emergency power system with respect to diesel generator loading and protective trips.          The adequacy of emergency AC power system status indications (i.e., annunciators) was also reviewed. At Millstone Unit No.1, the review included both the diesel generator and the emergency gas turbine generator.
II. Review Criteria
: 1. 10CFR50, General Design Criterion 17
: 2. Standard Review Plan Section 8.3.1
: 3. Branch Technical Position ICSB 17
: 4. IEEE Standard 279 - 1971
: 5. Regulatory Guide 1.9 III. Related Projects / Interfaces The following ISAP topics are related to the overall reliability of the emergency power system:
Topic 1.21 Fault Transfers Topic 1.23 Grid Separation Procedures Topic 1.24 Emergency Power Topic 1.25 Degraded Grid Voltage Procedures Topic 2.18 4.16 kV,480V and 125V DC Plant Distribution Protection Topic 2.25 Off-Site Power Systems IV. Evaluation The on-site emergency AC power system for Millstone Unit No. I consists of one diesel generator and one gas turbine generator with associated distribution equipment. The review criteria in Section II are intended specifically for diesel generators; there are no regulatory criteria related to use of gas turbine generators as emergency power sources. Therefore, the SEP review of Topic VIII-2 for Millstone Unit No. I applied diesel generator criteria for the evaluation of the gas turbine generator.
The SEP review concluded the following:
o The continuous and emergency loadings of the diesel and gas turbine complied with current licensing criteria.                  ,
o  The use of protective trips for the diesel generator complied with current licensing criteria.
o  The gas turbine and diesel generator annunciators either met the guidelines of IEEE Standard 279 - 1971 or had been modified to meet IEEE criteria.
 
Additionally, as a result of the _ operating history of the gas turbine generator, the Staff recommended that Northeast Nuclear Energy Company (NNECO) review the gas turbine preventative maintenance program and revise it where warranted. This issue is being reviewed separately under ISAP Topic 1.24. .
The- SEP review concluded that the use of protective trips on the gas turbine generator did not rrt "t Jrrent licensing criteria. Specifically, the following protective-trips do wt use coincident. logic and are not bypassed under emergency cenu.pns (current criteria allow protective trips provided - they are eifr.er '2ypassed under accident conditions or utilize coincident logic):
: 1) Light-off speed no . reached in 20 seconds.
: 2) Light-off temperature not reached 15 seconds af ter light-off.
: 3) Starting air ignition cut-off speed not reached 60 seconds after start.
: 4)    Generator excitation speed not reached in 60 seconds.
: 5)  High Exhaust Gas Temperature
: 6)  High Lube Oil Temperature
: 7)  High Gas Generator Speed
: 8)  High Turbine Overspeed
: 9)  High Vibration 3et
: 10)  Low Lube Oil Pressure
: 11)  Loss of Excitation
: 12) Opening of Exciter Breaker
: 13)  Generator Differential-
: 14)  Negative Sequence
: 15)  Reverse Power
: 16)  Generator Underspeed
: 17)  Voltage Restrained Overcurrent                                                      ,
Trips (1) through (4), above, are associated with the start-up sequence- of the gas turbine engine. Trips (5) through (10) protect the turbine during startup and steady-state operation, and (11) through (17) are associated with the generator itself.
                                                                    ,,w,w-
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As a result of this review, NNECO proposed to modify the following trips so that they would be bypassed during accident conditions:
(1)    Light-off speed not reached in 20 seconds.
(2)- Generator excitation speed not reached in 60 seconds.
(3) ' High Lube Oil Temperature (4)    Loss of Excitation (5)    Opening of Exciter Breaker (6)    Negative Sequence (7)    Reverse Power (8)    Generator Underspeed For the trips for which no modifications were proposed, NNECO provided justification for maintaining those trips <in their existing configuration.
The proposed modifications and the justification for retaining certain trips were accepted by the Staff.
V. Conclusions The use of protective trips for the. Millstone Unit No. I gas turbine generator does not meet current licensing criteria. Specifically, there are nine (9) trips associated with start-up and steady-state operation of the gas turbine generator which do not use coincident logic or are not bypassed in emergency conditions. The expected safety impact and the need to implement the modifications to the eight trips identified in Section IV, and a priority for implementation, should. be determined in the Integrated Assessment.
VI. - References (1)    D. M. Crutchfield letter to W. G. Counsil, dated September 30,1981.
(2) NUREG-0824 letter, Section 4.28.
(3)    W. G. Counsil letter to D. M. Crutchfield, dated September 27,1982.
(4)    W. G. Counsil letter to D. G. Eisenhut, dated February 4,1985 (Response to Generic Letter 84-15).
 
I 1
l l
i ISAP TOPIC NO.1.04 RWCU SYSTEM PRESSURE INTERLOCK
 
ISAP Topic No.1.04 RWCU System Pressure Interlock
    ' I. Introduction The review of SEP Topic V-II.A, Requirements for Isolation of High and Low Pressure Systems, evaluated the independence and diversity of pressure interlocks for motor-operated valves which isolate the reactor coolant system from other systems that have lower design pressure ratings.
Current criteria specify that interlocks should prevent the opening of MOVs until the reactor coolant system pressure is below the system design pressure, and close them automatically when RCS pressure increases above the system design pressure.
II. Review Criteria Standard Review Plan Section 7.6 Branch Technical Position ICSB3 III. Related Topics / Interfaces None.
IV. Evaluation The SEP review included an evaluation of the Reactor Water Clean-up System (RWCU), the Low Pressure Coolant Injection System (LPCI) and the l          Core Spray System (CS). This review concluded that the LPCI and CS systems meet current criteria regarding isolation requirements.
The RWCU system was determined not to meet current criteria. Isolation l
of the RWCU system on the suction side is provided by a motor-operated valve inside the drywell, and two motor-operated valves in parallel outside of containment (auxiliary pump suction valve and pump bypass valve).
Isolation on the discharge ' side is provided by a motor-operated valve outside containment and three check valves.
l          The RWCU system takes suction on the reactor coolant system, cools the water by directing        flow through three regenerative and two nonregenerative heat exchangers, and lowers the pressure to below the
;          design pressure of the low pressure portion of the RWCU system by use of l          a pressure control valve. Operational experience has shown that adequate l          purification is obtained even without use of the 'RWCU system filters.
Thus, flow bypasses the filters to the demineralizers and is then pumped at high pressure through the regenerative heat exchangers and back to the reactor coolant system via the feedwater system.
None of the MOVs will open if pressure in the low pressure portion of the system is higher than its designed pressure. The system will automatically isolate on the following signals:
o Low Vessel Water Level o High Temperature af ter Nonregenerative Heat Exchangers
 
                                            .                o High Pressure Downstream of Pressure Control Valve                    .
o Standby Liquid Control System Actuation                              l o High Auxiliary Pump Seal Water Temperature                            l l
Of concern is the fact that the high pressure isolation signal for all of the MOVs is from a single sensor. This does not meet current criteria, which      -
requires diversity in signals used for isolation purposes.
V. Conclusions The concern expressed by the Staff during the SEP review was that failure of the pressure control valve followed by a single failure ~of the one pressure interlock would result in a LOCA which bypasses the containment.
As a result, the Staff recommended that a redundant pressure interlock be installed for the inboard isolation valve, or that NNECO demonstrate the adequacy of the RWCU relief valve. The capacity of the relief valve and the need for an independent pressure interlock should be evaluated in the integrated assessment.
VI. References
: 1. NUREG-0824, Section 4.18.
: 2. D. M. Crutchfield letter to W. G. Counsil, dated July 8,1981.
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I ISAP TOPIC NO.1.05 VENTILATION SYSTEMS
 
n ISAP Topic No.1.05 Ventilation Systems I. Introduction To assure that ventilation ~ systems have the capability to provide a safe environment for plant personnel and for engineered safety features it is necessary to review the design and operation of these systems. For example, proper functioning of a ventilation system may be necessary in order to provide a controlled environment for engineered safety features following design basis accidents 'and anticipated transients to prevent safety-related equipment from exceeding its design thermal' limits. The Millstone Unit i ventilation systems were reviewed previously under SEP Topic IX-5, Ventilation Systems.
II. Criteria The following regulatory criteria were utilized in the SEP treview of ventilation systems for Millstone Unit 1:
: 1. Standard Review Plan, Sections 9.4.1, 9.4.2, 9.4.3, 9.4.4, 9.4.5 III. Related Topics / Interfaces ISAP Topic 1.12, Control Room Habitability ISAP Topic 2.11, Drywell Ventilation System l          ISAP Topic 2.24, Instrument, Service and Breathing Air Systems j          Environmental Qualification of Electrical Equipment (10CFR50.49)
IV. Evaluation The design and operation of the control room area ventilation system is not included in the scope of this review. The control room. ventilation system has been reviewed under NUREG-0737 Item III.D.3.4, Control Room Habitability, and is included in the ISAP review as Topic No.1.12.
The SEP review of the Millstone Unit I ventilation systems concluded that the following plant area ventilation systems satisfied the intent of the SEP topic:
o  Reactor Building Ventilation System o  Spent Fuel Area Ventilation System o  Radwaste Area Ventilation System o  Gas Turbine Building Ventilation System The Staff concluded that the following area ventilation systems did not meet all current criteria for design and operation:
o  Turbine Building Ventilation System The turbine building ventilation system does not automatically start following a loss of off-site power, and operator action is required to reinitiate ventilation in the turbine building, which includes safety-related equipment. In addition, a concern was raised regarding hydrogen generation in the battery rooms.
 
o    Low Pressure Coolant Injection and Core Spray Systems Space coolers' which service the areas of the low pressure coolant injection (LPCI) and core spray pumps are subject'to single failures.
: o. Feedwater Coolant Injection and Diesel Generator Space Coolers The. space coolers in the areas of FWCI components and in the diesel generator area do not automatically. start following a loss of off-site power.
o    Intake Structure Ventilation System The ventilation system for the intake structure, which houses the service water and emergency service water pumps, does not automatically restart following a loss of off-site power.
In References (1) and-(2), Northeast Nuclear Energy Company (NNECO) provided information regarding the ventilation systems identified by the .
Staff as being of concern.
In Reference (1), NNECO provided the Staff with documentation of testing that had been performed.on the~ space coolers in the areas of the LPCI and core spray pumps. This testing demonstrated that the space coolers were not required to prevent this equipment from exceeding design limits. This was accepted by the Staff in Reference (3).
In Reference (2), NNECO demonstrated that active ventilation was not required to prevent hydrogen in the battery rooms from reaching flammable concentrations, and also provided information to show that the-space coolers for the diesel generator automatically receive emergency power. In Reference (3), the Staff concluded that the concerns in these areas had been adequately resolved.
With respect to the turbine building ventilation system, NNECO provided an analysis of heat loads in the turbine building and the expected heat-up rates in the vicinity of safety-related equipment. This evaluation concluded that active ventilation would be requircJ only when the gas turbine starts and runs after a loss of off-site power. If the gas turbine does not start, the heat loads in the turbine building would be significantly reduced.      This analysis also demonstrated that sufficient time was available, even in the worst case, for operator action to manually restart the turbine building ventilation system. As a result, NNECO committed to modify operating procedures to instruct the operator to manually start the turbine building ventilation if the gas turbine starts and runs following a loss of off-site power. These procedure changes have been completed.
NNECO also evaluated heat loads in the area of FWCl components and in the intake structure. This evaluation concluded that the space coolers in the area of the FWCl components (HVH 3,3A,4,4A,5, and SA) are needed in order to maintain these areas within the 1040F design temperature of safety-related equipment. As a result, NNECO committed to modify these
 
space coolers so that they would be automatically sequenced onto the gas turbine generator. The single power source is acceptable since the FWCI system can be powered only by the gas turbine. If the gas turbine does not start, no cooling of the FWCI components is required. Regarding the intake structure, NNECO concluded that the worst case heat-up rate in the structure due to operation of the service water and emergency service water pumps indicates that starting of one of two ventilation exhaust fans is necessary. This would provide one complete air change in the building every four minutes. NNECO committed to modify the exhaust fans to automatically start one fan on the gas turbine and one on the diesel generator.
V. Conclusions NNECO has proposed physical modifications to the emergency power capabilities for the FWCI area space coolers and for the intake structure exhaust fans. All other plant ventilation systems have been determined to meet current criteria or have been demonstrated to have no effect on the ability to achieve and maintain safe shutdown. The proposed modifications should be evaluated with respect to the conservatism in equipment design limits and risk insights and best estimate system success criteria obtained from the probabilistic safety study to determine the need to implement these modifications and, if determined to be necessary, a priority for implementation.
VI. References
: 1. W. G. Counsil letter to D. M. Crutchfield, dated December 3,1982.
: 2. W. G. Counsit letter to D. M. Crutchfield, dateo April 18,1983.
: 3. 3. 3. Shea letter to W. G. Counsil, dated July 5,1983.
: 4. 3. 3. Shea letter to W. G. Counsil, dated September 14, 1982.
        - 5. NUREG-0824, Section 4.32, Ventilation Systems.
 
O ISAP TOPIC NO.1.15 FSAR UPDATE l
 
l ISAP Topic No.1.15 FSAR Update I. Introduction A Final Safety Analysis Report (FSAR) for a nuclear power plant is submitted at the time the application for an operating license is filed with l
the NRC. The FSAR provides the NRC with the information which the Staff requires in order to judge whether or not to issue an operating license l
for the plant. Licensees are required to update the FSAR on an annual basis to reflect changes in the plant design.
II. Review Criteria
: 1. 10CFR50.71 III. Related Topics / Interfaces
: 1. ISAP Topic 2.10, Upgrading of P&lDs
: 2. ISAP Topic 2.29, Long Term Cooling Study
: 3. ISAP Topic 2.30, FWCl Assessment Study IV. Evaluation The FSAR for Millstone Unit No. I which was submitted as part of the operating license application has not been updated and submitted to the NRC. Although 10CFR50.71(e)(3)(i) requires that an updated FSAR be submitted by July 22,1982, Millstone Unit No. I participated in the NRC's Systematic Evaluation Program (SEP). 10CFR50.71(e)(3)(ii) exempted plants participating in the SEP review trom the requirement for updating of the FSAR until 24 months after the licensee receives notification that the SEP review has been completed for that facility. For Millstone Unit No.1, this notification was received by.the licensee on March 21,1983; the first annual FSAR update was thus required to be submitted by March 21, 1985.
On February 4,1985, NNECO requested an exemption from the schedular requirement of 10CFR50.71(e) in order to permit this requirement to be subjected to an integrated evaluation in the context of ISAP. On April 11, 1985 the NRC issued an exemption until October 11, 1985 for NNECO to submit a plan and schedule for complying with the FSAR update rule.
V. Conclusions The FSAR for Millstone Unit No. I has not been updated since the time of original submittal. The licensee is to provide a plan and schedule for performing the update to respond to Reference (2). The optimal plan for updating the FSAR, including consideration of other configuration control options, and its associated resource burden will be considered in the integrated assessment.
VI. References
: 1. W. G. Counsil letter to 3. A. Zwolinski, dated February 4,1985.
: 2. 3. A. Zwolinski letter to W. G. Counsil, dated April 11, 1985.              I
 
f t
i ISAP TOPIC NO.1.18 ANTICIPATED TRANSIENTS WITHOUT SCRAM (ATWS) i f
t
 
ISAP Topic No.1.18 Anticipated Transients Without Scram (ATWS)
: 1. Introduction 1
General Design Criterion 20 of Appendix A to 10CFR50 requires that nuclear power plants be provided with a protection system which will:
(1)                                        initiate automatically the operation of the appropriate systems, including the reactivity control systems, to ensure that specified acceptable fuel design limits are not exceeded as .a result of anticipated operational occurrences, and (2)                                        sense accident conditions and initiate the operation of systems and components important to safety.
An anticipated operational occurrence (as defined in Appendix A to                                                              -
10CFR50) followed by a failure of the protection system to shut down the reactor, could result in exceeding plant design limits.                                                                          As a result, 10CFR50.62 was promulgated to require additional' protection against an anticipated transient without scram (ATWS). The objective of this topic is to review plant design against the requirements of 10CFR50.62.
II. Review Criteria I
: 1.                10CFR50.62, " Requirements for Reduction of Risk from Anticipated i
Transients without Scram (ATWS) Events for Light Water Cooled Nuclear Power Plants."
;                                                      III. Related Topics / Interfaces l
l                                                          ISAP Topic 1.26, Equipment Classification / Vendor Interface ISAP Topic 1.27, Post-Maintenance Testing Procedures t                                                          ISAP Topic 1.28, Post-Maintenance Testing Tech. Spec. Changes ISAP Topic 1.30, Post-Trip Review Data and Information ISAP Topic 1.31, Equipment Classification / Vendor Interface ISAP Topic 1.32, Items 3.2.1 and 3.2.2, Post-Maintenance Testing Procedures ISAP Topic 1.33, item '3.2.3, Post-Maintenance Testing Tech. Spec.
Changes ISAP Topic 1.34, items 4.5.2 and 4.5.3, Reactor Trip System Testing
!                                                          ISAP Topic 1.36, item 4.5.1, Reactor System Functional Testing ISAP Topic 1.43, Tech. Spec. for Anticipatory Trips ISAP Topic 2.17, Reactor Protection Trip System IV. Evaluation 10CFR50.62 specifies three separate requirements that are applicable to boiling water reactors. These are:
(1)                                          Each boiling water reactor must have an alternate rod injection (ARI) system that is diverse from the reactor trip system from sensor output to the final actuation device. The ARI system must
 
have redundant scram air header exhaust valves. The ARI must be designed to perform its function- in a reliable manner and be independent from the existing reactor trip system from sensor output to the final actuation device.
(2)    Each boiling water reactor must have a standby liquid control system (SLCS) with a minimum flow capacity and boron content equivalent in control capacity to 86 gallons per minute of 13 weight percent sodium pentaborate solution. The SLCS and its injection location must be designed to perform its function in a reliable manner. The SLCS initiation must be automatic and must be designed to perform its function in a reliable manner for plants granted a construction permit after July 26, 1984, and for plants granted a construction permit prior to July 26, 1984, that have already been designed and built to include this feature.
(3)    Each boiling water reactor must have equipment to trip the reactor coolant recirculation pumps automatically under conditions indicative of an ATWS. This equipment must be designed to perform its function in a reliable manner.
Millstone Unit No. I has an alternate rod insertion system that satisfies the requirements of item (1), above. Millstone Unit No. I also has equipment to trip the recirculation pumps under conditions indicative of an ATWS that satisfies the requirements of item (3), above. Millstone Unit No.1 is equipped with a Standby Liquid Control System (SLCS), however the capacity of the SLCS does not meet the requirements of item (2) above.
The Millstone 1 SLCS has an equivalent control capacity of 43 gpm of 13 weight percent sodium pentaborate solution, which is half of the effective capacity required by 10CFR50.62.
V. Conclusions Millstone Unit No.1 is currently in compliance wit 7 the requirements of 10CFR50.62 concerning alternate rod insertion and recirculation pump trip systems.      Millstone Unit No. I does not meet the SLCS capacity requirement of 10CFR50.62. The significance of this difference and an assessment of the need to increase the SLCS capacity will be evaluated in the fntegrated assessment.
VI. References
: 1. 10CFR50.62
: 2. 49FR26044, dated June 26,1985
: 3. 49FR27736, dated July 6,1984
 
st 9  e ISAP TOPIC NO.1.20 MOTOR-OPERATED VALVE (MOV) INTERLOCKS u-_------_____
 
ISAP Topic No.1.20 Motor-Operated Valve (MOV) Interlocks I. Introduction The SEP review for Millstone Unit 1 included a review of Topic III-10.A, Thermal Overload Protection for Motors of Motor-Operated Valves. The primary objective of thermal overload relays is to protect motor windings of motor-operated valves against excessive heating. Excessive heating could result from, among other causes, too frequent exercising of the valve motor or by excessive current draw due to encountering an obstruction during valve travel. Thermal overload devices detect this motor heat-up and operate to stop the valve motor before overheating damage occurs.
This feature of thermal overload relays could, however, interfere with the successful functioning of a safety-related system. In nuclear power plant safety system application, the ultimate criterion is to drive the valve to its proper position to mitigate the consequences of an accident, rather than to be concerned with degradation or failure of the motor due to excessive heating. However, thermal overload devices can provide useful protection provided that they are sized properly.
II. Criteria
: 1. IEEE Standard 279-1971.
: 2. Regulatory Guide 1.106.
: 3. The Dangers of Bypassing Thermal Overload Relays in Nuclear Power Plant Motor-Operating Valve Circuits, IEEE Transactions on Power Apparatus and Systems, Volume PAS-99, No. 6, November / December 1980.
III. Related Topics / Interfaces ISAP Topic 1.17, Replacement of Motor-Operated Valves ISAP Topic 2.16, Torque Switch Evaluation for MOVs IV. Evaluation The SEP review of Topic III-10.A for Millstone Unit 1 identified 12 motor-operated valves that used thermal overload devices in their control circuits and would be required to change position during an accident.
In Reference (1), Northeast Nuclear Energy Company (NNECO) provided its philosophy on protection for motor-operated valves, including the use of thermal overloads, torque switches and limit switches. While NRC criteria require that thermal overload devices be bypassed during emergency or accident conditions, NNECO's position was that thermal overload protection should be retained; however, analyses should be performed to confirm that these devices are conservatively set. That is, they should be set so as to preclude inadvertent actuation and thus prevent fulfillment of the valves required safety function. Reference (1) provided a complete discussion and justification for this position. NNECO also proposed criteria by which to evaluate the thermal overload setpoints. These criteria had been published by IEEE and are specifically oriented to applications in
 
nuclear power plants (see Reference (3) in Section II, above).            In Reference (2), the Staff accepted NNECO's position and requested that NNECO demonstrate that the thermal overload devices for the 12 MOVs noted above are properly set.
In reviewing the subject MOVs and their associated safety functions, NNECO determined that several of the valves identified by the Staff should not be considered in the thermal overload analysis for the following reasons:
o The designation 1-IA-14 had been reused on a check valve and the motor operated had becn previously identified as 1-IA-68 (Instrument Air to Drywell).          However, due to recent modifications, this MOV has now been relabeled as 1-AC-50 and serves a different function (Nitrogen to Drywell).      It is not a safety-related valve, and thus does not need to be evaluated.
o    Valve 1-MS-7 is not a Category lE MOV and does not receive an isolation signal. Therefore, this valve ryerator does not need to be evaluated.
o  The designation 1-SA-6 had been used on a hand operated valve and the MOV had been labeled 1-SA-35 (Station Air to Drywell).
This is not a Category 1E MOV and does not receive an isolation signal and thus does not need to be evaluated.
Additionally, the designation 1-MW-97 is now used for a hand operated valve and the MOV has been relabeled 1-MW-96A.
Therefore, the remaining 9 MOVs which were evaluated are as follows:
: 1. 1-CU-2
: 2. 1-CU-3
: 3. 1-CU-5
: 4. 1-CU-28
: 5. 1-M W-96A
: 6. 1-R R-2A
: 7. 1-R R-2B
: 8. 1-5U9
: 9. 1-IC-10 The 9 MOVs of concern have been analyzed using the method described in Reference (1).      This analysis confirmed that the thermal overload protective devices have been conservatively and properly set. Calculations to support this conclusion are attached. It was not necessary to account for variations in Motor Control Center ambient temperature since the thermal overload relays are ambient compensated.
As requested by the Staff, NNECO also evaluated the operating experience of these valves. Review of operating experience identified no Instances where valves failed to operate as a result of thermal overload protection devices. One valve motor (1-MW-96A) failed on February 6,1980 as a result of a contactor stuck in the closed position. One other valve (l.CU-
 
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: 3) has experienced three motor winding failures, the most recent occurring on December 4,1981. No other failures involving these valves have been identified.
V. Conclusions There are 9 safety-related motor-operated valves at Millstone Unit I that use thermal overload protective devices that are not bypassed in an emergency and which are required to change position. This does not meet current criteria, however, the use of thermal overloads has been accepted by the Staff provided they are properly set. NNECO has demonstrated that the thermal overloads for the 9 valves in question are properly set and thus the safety objective of this topic has been satisfied.
VI. References
: 1. W. G. Counsil letter to D. M. Crutchfield, dated March 22,1982.
: 2. 3. 3. Shea letter to W. G. Counsil, dated April 12,1982.
: 3. W. G. Counsil letter to 3. A. Zwolinski, dated March 22,1985.
: 4. NUREG-0824, Section 4.14.
 
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l ISAP TOPIC NO.1.24 EMERGENCY POWER
 
ISAP Topic No.1.24 Emergency Power I. Introduction Millstone Unit No. I utilizes one diesel generator and one gas turbine generator as emergency power sources. During the Systematic Evaluation Program review, the emergency power system was reviewed with respect to emergency generator loading (30 minutes and continuous loading) and      l the use of protective trips. The adequacy of the emergency generator        ,
status indications (i.e., annunciators) was reviewed to ensure that        '
conditions which would render the generator inoperable were clearly L              indicated.
The issue of diesel generator (or gas turbine generator) testing was not included in this review. This issue is being addressed generically by the NRC via Generic Letter 84-15.
During the SEP review, the NRC concluded that failure of the gas turbine l
generator appeared in approximately one quarter of the dominant accident sequences. Based on a review of 31 reported failures of the gas turbine generator over a 12-year period (see Table 4.2 in NUREG-0824), the NRC concluded that some of these failures could have been prtvented by a more effective preventative maintenance program. .This prompted the NRC to request that NNECO review its preventative maintenance program for the gas turbine to identify areas for improvement. The scope, of this ISAP
(              topic is a review of the adequacy of this program.
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II. Criteria l              There are no regulatory criteria related to preventative maintenance programs for gas turbine generators.
l        III. Related Topics / Interfaces l              ISAP Topic 1.01, Gas Turbine Generator Start Logic Modifications l              ISAP Topic 1.21, Fault Transfers ISAP Topic 1.23, Grid Separation Procedures ISAP Topic 1.25, Degraded Grid Voltage Procedures
(              ISAP Topic 2.18,4.16 KY,480 V and 125 VDC Plant Distribution Protection
!              ISAP Topic 2.25, Offsite Power Systems IV. Evaluation in response to Section 4.28.3 of NUREG-0824, NNECO performed a review of the preventative maintenance program for the gas turbine generator.
The results of this review were forwarded to the Staff in Reference (1).
NNECO's review of the preventative maintenance program identified three vital components for which there were no records of maintenance being performed. These were the engine-mounted fuel pump, the fuel shutoff valve and tim air start regulating valve. NNECO committed to disassemble and inspect these components when their cumulative running time met the
 
surveillance interval recommended by the manufacturer. Subsequently, as described in Reference (2), it was determined that both the engine-mounted fuel pump and the fuel shutoff valve had been replaced along with the gas generator when the engine was damaged by stones which entered the compressor inlet in 1976. Since that time, these components have logged only a small fraction of the recommended running hours between inspection intervals and thus disassembly and inspection of these components was not warranted.            The engine manufacturer, General Electric, has advised against more frequent inspections due to the extremely close tolerances required when the parts are reassembled and the extensive bench testing necessary to ensure proper operation prior to reinstallation.
Following review of the preventative maintenance program submitted in Reference (1), the Staff indicated that they believed that more frequent calibration of the control circuit timers should be performed. This concern was supported by past gas turbine failures resulting from improper timer calibration. In Reference (3), NNECO committed to revise the preventative maintenance program to require calibration of control circuit timers at each refueling outage. This is an increase in surveillance from the frequency of every third refueling outage noted in Reference (1). This change in surveillance frequency has already been implemented.
The Staff's concern over the adequacy of the preventative maintenance program was the result of several failures of the gas turbine generator during 1981 and 1982 that were attributed to causes which could have been prevented by better preventative maintenance. The Staff was also concerned by what was perceived as a relatively high failure rate for the gas turbine generator.
NNEC7 has recognized that some degradation has occurred in the gas turbine generator, and has taken a number of actions to restore components to high quality conditions and to prevent further degradation.
As a result of a series of failures in 1982 that were attributed to rust contamination, NNECO installed new stainless steel piping in the air start system, installed filters in the regulator sensing lines and cleaned and coated the inside of the air receiver tank. Additionally, it has been noted that continued exposure to a salty-air environment has caused degradation (e.g., oxidation and falling open of resistors) that has resulted in failures of the gas turbine generator. Examples are described in Licensee Event Reports (LER) 81-28 and 81-31. To address this, NNECO has installed equipment in the area of the gas turbine generator which will provide a i                      controlled environment for critical, environment-sensitive components. It l                      Is expected that this will effectively limit the rate of degradation of the machine. Since the 1981-1982 failures, there have been no failures of the i                      gas turbine generator that are attributed to either of these causes.
As part of its response to Generic Letter 84-15, NNECO performed an evaluation of the reliability of both the gas turbine generator and the diesel generator at Millstone Unit 1. The results of this evaluation were provided in Reference (4).
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  . a l                                                V. Conclusions NNECO has conducted a thorough review of the gas turbine generator preventative maintenance program and identified several areas where l        improvements were warranted.        Recent experience with gas turbine generator surveillance indicates that the problems which have arisen in the past have been alleviated to the point where overall reliability is high.
Based on the above, NNECO considers this issue to be resolved.
VI. References
: 1. W. G. Counsil letter to D. M. Crutchfield, dated May 10,1983.
: 2. W. G. Counsil letter to D. M. Crutchfield, dated June 12,1984.
: 3. W. G. Counsil letter to D. M. Crutchfield, dated July 20,1983.
: 4. W. G. Counsil letter to D. G. Eisenhut, dated February 4,1985.
: 5. 3. Shea letter to W. G. Counsil, dated August 16,1984.
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