ML19269D178
| ML19269D178 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 12/31/1978 |
| From: | NORTHEAST UTILITIES |
| To: | |
| Shared Package | |
| ML19269D177 | List: |
| References | |
| NUDOCS 7903070235 | |
| Download: ML19269D178 (50) | |
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O f1ILLST0flE UtiITS 1 AND 2 AfifiUAL REPORT This Annual Report has been prepared pursuant to the requirements of Title 10, Code of Federal Regulations, Section 50.59, and Sections 6.9.1.4 and 6.9.1.5 of the units' Appendix A Technical Specifications.
There are two tabs separating the reporting information by unit.
Connon site reporting requirenents are addressed in this section.
Connon site facility changes and tests are administered under the control of only one unit. Hence, their evaluations are provided in the section applicable to that unit assigned the responsibility.
Connon site procedure changes are addressed here.
The only procedure changed requiring a 10 CFR 50.59 review during 1978 common to both units' FSAR was the f1illstone Energency Plan.
fione of the changes involved an unreviewed safety question since they did not affect the safety analysis or Technical Specifications.
Changes to the Energency Plan in 1978 involved the following areas:
1.
General organizational changes to reflect the latest organizations, both company and off site support groups, and to improve communications.
2.
There were various changes of an editorial or general update nature.
3.
Changes were incorporated to reflect additions made to energency equipnent inventories.
4.
Additional training requirements were imposed, particularly in the area of fire crew training.
5.
Provisions for off site agency participation in emergency drills were incorporated.
6.
Energency actions were provided in greater detail and responsibilities were clarified.
7 9 3o7 D "s
lilLLSTONE UNIT 1 CONTEilTS Page CHANGES Design Changes l/l
- l/4 Technical Specification Changes 1/5 - 1/7 Environmental Technical Specification Changes 1/8 TESTS Feedwater Temperature Reduction 1/9 - 1/13 P,ADIATION EXPOSURE 1/14 - 1/15
Page 1/1 PLANT DESIGN CHANGES THAT WERE C0fiPLETED DURING 1978 4 0-71 Renove Alarn Card on N flakeup Low Pressure Alarn 2
On the N2 nakeup systen, there are two nodes of operation; a purging and trickle feed node.
Currently, we use the purging node and in this mode, a continuous N2 low pressure alarn is received.
Thus, the alarn is not representative of an abnormal condition.
30-75 Service Water Extraction Evaporator Basket Strainer A basket strainer was installed in the service water supply line to the extraction evaporator.
This prevents plugging in the deaerator tank sparger.
11-76 Torus Support System - Inner Row of Colunns The inner row of tcrus support colunns, their botton connections and the botton connections of the outer row of colunns were modified to provide extra safety margin to withstand postulated dynamic loads.
22-76 Torus Support System - Colunn Supports The colunn connections in the outer row were nodified to increase the capability of the existing pin and lug connections to support compressive loads.
Support blocks were added to provide extra safety margin for the botton connections of the columns to withstand postulated loads.
38-76 Stator Cooling Pump Reservoir Coolers Cooling water was piped to coolers for "A" and "B" stator cooling pump coupling reservoirs.
This was done to renove heat, thereby reducing bearing failures.
63-76 Condenser Vacuun Prining Systen The condenser vacuum priming system was nodified to elininate float valves at the inlet and outlet condenser water boxes. The nodification also prevented water seal in the U-loop and allowed venting of the discharge tunnel.
Page 1/2 22-77 Fuel Pool Blowdown Systen A pneunatic booster and filter were installed in the fuel pool filter blowdown system.
This elininated the need for using canpressed nitrogen in filter blowdown.
35-77 Shutdown Cooling Instrument Line Renoval Two 3/4" instrument valves and a nipple off of a shutdown cooling line, that was used only during the startup test progran, were renoved and capped.
The lines were no longer used and were continual naintenance problems.
54-77 Condenser Bay - Moisture Separator Shield Walls Two shield walls were constructed between the noisture separators and the condenser water boxes to decrease radiation exposure of individuals involved in condenser maintenance work during power operations.
57-77 Loss of DC Power Alarns Loss of DC power alarn circuits were standardized for five generic annunciator alarms.
This provided a standard alarn circuit and short circuit protection.
62-77 Control Rod Drive Area Vent Controls The CRD area vent controls, TS 26-1 & TS 26-2, were relocated from the recirculation pump notor areas to the under vessel area of the drywell.
They were relocated to more effectively monitor tenperature in the CRD area.
This will provide additional cooling to the CRD area when
=
temperatures are high.
3-78 Solid Radwaste Compactor Roon Heater A stean heater was installed in the solid radwaste compactor roaa to provide heat for personnel working in the compactor roal during the winter nonths.
4-78 Health Physics Office Area Shield Wall A shield wall was constructed along the existing wall between the condensate punp area and the Health Physics office passageway.
Page 1/3 This was done to decrease the radiation exposure to individuals working in the Health Physics office and personnel walking through the passageway.
7-78 Denineralized Hater Supply to Turbine Area A new line was connected between a discontinued demineralized water makeup line and the Unit 1 turbine area.
This will supply denineralized water to the turbine area to be used for washing turbine parts during overhaul periods.
25-78 Recirculation Flow Control Systems The existing nanual/ automatic flow control stations were replaced with new bias nanual/ automatic stations per vendor reconnendation.
This provided isolation between the two control loops to eliminate the source of spurious power fluctuations and provide capability for balancing loop flows.
26-78 CR0 Punp Suction Fra, CST A new line was installed between the condensate reject line from the main condenser and the control rod drive pump suction from the condensate storage tank.
This provided the CPD pumps with a lower oxygen content water which could prevent potential danage to the control rod drive collet retainer.
27-78 Connection for ffobile Air Compressor A connection, adjacent to Unit 1 maintenance area, was installed to pennit a mobile air compressor to be connected to the station air systen. This would provide supplemental air, as required, during periods of high station air consumption and/or station air canpressor overhauls.
34-78 Stop Valves for Radwaste Flush Water Stop valves were installed in the flush water lines for the spent resin pump and the clean up filter sludge pump.
This was done to preclude the possibility of leaking downstream solenoid valves while centrifuging and to allow for isolation of the solenoid valves for maintenance.
Page 1/4 53-78 Turbine Thrust Bearing Modification The old turbine thrust bearing wear detectors were replaced with new detectors per vendor reconnendation. The old detectors had generic problems.
73-78 Containment Isolation Equipment Modifications Certain cables, splices and teminations were replaced with environmentally qualified materials to assure operation of safety related electrical equipment inside the drywell requ ired for containment isolation.
Page 1/5 TECHNICAL SPECIFICATIONS APPROVED IN 1978
~
Date Description Amendment #
3/3/78 Fire Protection 44 Fire protection systens were added to Technical Specifications and administra-tive controls were changed accordingly.
3/10/78 LER Fonnat 45 Section 6.9.1 was incorporated in the Routine Reports and Reportable Occurrences section per NRC letter.
3/10/78 Core Spray Conditionally Inoperable 46 This change allows the inoperability of the core spray systen during control rod drive work and/or fuel novement.
This can be acconplished because of redundant ECCS systens having capacities to adequately deliver more flow than core spray.
4/12/78 MCPR - MAPLHGR 47 This amendment involves actually four separate Technical Specification changes.
First, the tenn nuclear engineer is changed to reactor engineer to be consistent with station technology.
Secondly, the Mininum Critical Power Ratio is altered to reflect the results of transient analysis for Cycle 6, which were found to be technically acceptable.
The third portion adds Figure 3.ll.l.d, the single 7D250 assembly inserted at B0C 4.
Lastly, the Maxinum Average Planar Linear Heat Generation Rate limits for all previous fuel types at all exposures were increased based on new ECCS analysis. Additionally, the MAPLHGR limit is shown for the Cycle 6 8D274H assemblies.
Page 1/6 Date Description Amendment #
6/16/78 MAPLHGR 49 These changes to Maximum Average Planar Linear Heat Generation Rate limits are based on detailed analysis perforned by the vendor contained in the LOCA topical report ~ for Millstone 1.
Additional documentation to support the applic-ability of this MAPLHGR multiplier at reduced flow was referenced.
7/6/78 Containment Suppression Pool Operation 51 These changes were incorporated to reduce the containment suppression pool water volume (and level) and reduce suppression pool temperature. This will reduce dynamic loading on the torus during LOCA transients.
The temperature reduction will insure that adequate heat sink properties are maintained.
7/11/78 Power / Flow 52 Figure 3.3.1 was revised to allow operation to a higher power / flow curve.
This provides more flexible operation during startup and other reduced power /
flow naneuvers within approved transient response analysis.
9/26/78 Fire Protection 53 Fire protection systens were added to Technical Specifications and changes made to the Administrative Controls section of Technical Specifications. This will assure the availability of fire protection and detection systens for safe shutdown areas and comply with regulatory requirements.
Page 1/7 Date Description Arendnent #
12/4/78 Standby Gas Treatment System 54 This change eliminates the Technical Specification requirement to perforn the in-place leak test at normal design flows.
The design flow rate can only be obtained if all other reactor building ventilation is tenninated. This is not practical during nonnal operation with personnel in the building.
12/4/78 Instrumentation (IRM) 34 The calibration frequency of the Intennediate Range Monitor (IRM) rod blocks was changed to agree with IRM scram function. The calibration of IRM's during each startup for scran function was previously deleted from the STS.
It was felt that calibration of the IRM rod blocks should be the same.
12/8/78 Administrative Changes - High Radiation 56 This change involves administrative controls to reflect organizational structure concerning entry into high radiation areas.
Page 1/8 ENVIR0t.NENTAL TECHNICAL SPECIFICATION CHANGES
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Date Description Amendment #
4/21/78 Radioactive Effluents 48 The Environmental Technical Specification regarding radioactive effluents was changed to allow discharges fron Condensate Polishing waste neutralizing sumps to Long Island Sound.
6/19/78 Gaseous Waste Effluents 50 Specification 2.4.2.2.K was added to limit the noble gas in-process activity rate.
This is provided to linit off site doses due to the seismic failure of the new stean dilution recanbiner and augmented off gas systen.
12/4/78 Radioactive Effluent Sanpling 55 On Table 2.4-1, primary coolant sampling and analysis for I-131 and I-133 were d ele ted.
Also, footnote #4 on page 2.4-16 was deleted. The changes were made because prinary coolant sampling and analysis for iodines are covered by the Unit 1 Safety Technical Specifications (Section 3.6.C and 4.6.C.1).
Page 1/9 FEEDWATER TEMPFRATURE REDUCTION Preliminary Analys3 The nuclear response of the liillstone Unit 1 reactor to feedwater te1perature reductions at the end of Cycle 5 was analyzed with the same standard design procedures and techniques used in the original licensing work for Cycle 5.
A new heat balance was generated consistent with operation with two feedwater heaters valved out of service at full rated power. These operating conditions were used to extrapolate from the previously predicted end of cycle point on reactivity depletion to the new, longer exposure capability achieved with the decrease in the inlet water temperature. Then, at the new end of cycle point (defined with all control blades withdrawn),
standard analyses were conducted to evaluate the scran response, void coefficient and doppler coefficient values.
The scran response was slower during the first half of the transient; the core average void fraction was less and the dynamic void coefficient was significantly better.
Analyses indicate Technical Specification operating linits for Reload 4 would not be exceeded for E0C5 conditions with feedwater tenperature reduced by a naxinum of 75 F.
The !!CPR for the feedwater derate condition is less than the Technical Specification f!CPR limits.
Therefore, the nargin from transition boiling is greater for the feedwater derate condition than that analyzed in NED0-21364 for Reload 4.
(See Table 1)
The peak neutron flux and peak vessel botton pressure for both Load Rejection without Bypass and MSIV Closure with Flux Scram are less than the feedwater derate condition analyzed in NED0-21364.
Therefore, a greater nargin fra, the AStiE Vessel Code pressure limit is obtained for the feedwater derate condition.
Although the scran worths were slightly less for the feedwater derate condition, other differences (e.g., decrease in void coefficient) nore than ca,pensated for the decrease in scram worths.
Description On February 17, 1978, reactor power was reduced to 50% power and the extraction steam to both high pressure feedwater heaters was re,oved.
The object of this action was to increase the gross electric production of this unit.
Page 1/10 Prior to valving out the extraction stean to the high pressure
. eaters, transient and thennal hydraulic analyses were perfonned, i
as nentioned above, for reactor operation with a feedwater temperature reduction of 75 F.
(The 75 F decrease in feedwater temperature assumed the isolation of the high pressure and intennediate high pressure feedwater heaters.)
The isolation of the high pressure feedwater heaters was perforned in accordance with Operating Procedure 346, Revision 2, Change 1, and the data during the feedwater telperature reduction and the ascension to. power was acca1plished in accordance with Special Procedure 78-1-05.
Tables 2 and 3 sunmarize the results of reducing the feedwater tenperature by renoving the extraction stean to the high pressure heaters.
Conclusions Although the feedwater tenperature reduction data was sonewhat affected by the renoval of the 74 control rod notches fran the core after the heaters were isolated (resulting in a loss of 12 MWE), it can be concluded the unit gained gross generation of 21 f9WE/hr. and lost 0.55% efficiency.
Therefore, for the 19 days the unit perforned before the refueling, an additional 9,576 MWE hours were produced.
Also, it can be concluded that the 40 F tenperature reduction did not have any substantial impact on our thennal limits at this power l evel.
However, if the plant were operating at full power, feedwater temperature reduction would be possible if the fuel thernal limit margins were greater than 5% prior to reducing the feedwater te,pe ra tu re.
The sane 2 f1WE per day coast down that was seen prior to feedwater tenperature reduction was also witnessed af ter feedwater tenperature reduction through the end of cycle.
- - --mens
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Page 1/11 Table 1 MILLSTONE CYCLE 5, TRAtlSIEllT DATA SUtiMARY 100% Power /100% Flow, E0C-5 Scran Core
^
^
^
^
Power Flow 4
Q/A Psl Pv
, (%)
(%)
(%)
(%)
(psig)
(psia)_
7x7 8x8 Load Rejection w/o 100 100 298 117.2 1191 1225 0.23 0.30 Bypass MSIV Closure, Flux 100 100 534 1238 1268 Scran 100% Power /100". Flow, E0C-5 + 277 with 75 F Feedwater Temperature Reduction Load Rejection w/o 100 100 284 113.5 1165 1198
<0.18
<0.24 Bypass MSIV Closure, Flux 100 100 467 1208 1204 Scran Synbols:
- Peak neutron flux Q/A - Peak heat flux Psl - Peak steam line pressure Pv - Peak vessel botton pressure
Page 1/12 Table 2 FEEDWATER TEMPERATURE REDUCTION At 50% Power with 74 Notches in the Core Before Isolation After Isolation Change Core Themal Power 1044 MHTH 1063 fiWTH
+ 19 f MTH Gross Generation 335.4 MUE 346.1 MWE
+ 10.7 ffWE Feedwater Temperature 317.2 F 286.9
- 30.3 F 11267.75fWHR B
Unit Heat Rate 11244.57
+
'l Kt HR Kil-HR Unit Efficiency 30.29%
30.35%
+.06% (1)
Turbine 1st Stage Pressure 443 psia 439 psia
- 4 psia (2)
Core Average Void Fraction 33%
31.7%
- 1.3%
+ 3.7 fB Subcooling 33.33 37.03 L
B MCPR 0.580(7x7) 0.591 (7x7)
+ 0.011 0.662 (8x8) 0.664(8x8)
+ 0.002 MAPLHGR 0.467 0.482
+ 0.015 Total Peaking Factor 2.243 2.279
- 0.036 (1) & (2) Circulator Punp Removed from Service a
Page 1/13 Table 3 FEEDUATER TEMPERATURE REDUCTION At 100% Flow with all Rods Out Before Isolation After Isolation Change Core Themal Power 1809.34 Mkm1 1883.29 ffWTH
+ 73.95 MWTH, 3.7%
Gross Generation 613.7 MWE 628.12 MWE
+ 14.42 t'WE (3)
Feedwater Temperature 361.03 321.66
+ 39.37 F BTU 0
0 Unit Heat Rate 10540.9 10723.14
+ 182.33 gp HR K HR Unit Efficiency 32.38%
31.83%
0.55%
Turbine 1st Stage Pressu re 841 psia 837 psia
- 4 psia (3)
Core Coverage Void Fraction 33.9%
33.9% U BU B"
Subcooling 21.35 25.09
+ 3.74 B
B f1CPR 0.733 (7x7) 0.784(7x7)
+ 0.051 (4) 0.775(8x8) 0.825 (8x8)
+.05 (4)
MAPLHGR 0.664
.0764
+.01 (4)
Total Peaking Factor 2.240 2.065
- 0.175 (5)
(3) Reactor core was coasting down for several weeks prior to tenperature feedwater reduction.
Figure 1 illustrates the change in electric production.
(4) After the feedwater temperature reduction, the location of the three limiting fuel bundles changed location.
(5) Total Peaking Factor reduced because of all rods out.
MILLSTONE 1 116 REPORT
- 1978***
NUMBER OF PERSONNEL OVER 100 MREM TOTAL fMN-REM STATION UTILITY CONTRACT WORKERS STATION UTILITY CONTRACT WORKERS EMPLOYEES EMPLOYEES AND OTHERS EMPLOYEES EMPLOYEES AND OTHERS REACTOR OPERATIONS &
SURVEILLANCE MAINTENANCE PERSONNEL 18 5
31 1GO 2.000 7.243 OPERATING PERSONNEL 35 1
0 56.810
.105
.000 HEALTH PHUSICS PERSONNEL 7
0 37 13.690
.000 27.005 SUPERVISOR PERSONNEL 10 0
1 7.880
.010
.230 ENGINEERING PERSONNEL 4
1 2
1.690
.230
.833 ROUTINE fMINTENANCE MAINTENANCE PERSONNEL 5
0 4
1.970
.545 2.381 OPERATING PERSONNEL 0
0 0
.070
.000
.000 HEALTH PHYSICS PERSONNEL 0
0 0
.080
.000
.000 SUPERVISORY PERSONNEL 0
0 0
.020
.000
.015 ENGINEERING PERSONNEL 0
0 0
.030
.025
.055 INSERVICE INSPECTION fMINTENANCE PERSONNEL 3
0 9
.445
.055 2.215 OPERATING PERSONNEL 1
0 0
.160
.000
.000 HEALTH PHYSICS PERSONNEL 0
0 0
.070
.000
.095 SUPERVISORY PERSONNEL 0
0 1
.065
.055
.405 ENGINEERING PERSONNEL 2
2 13 1.025
.545 4.370 SPECIAL fMINTENANCE
[
MAINTENANCE PERSONNEL 44 74 642 76.595 55.054 629.136 o
OPERATING PERSONNEL 28 0
0 9.240
.000
.000 g
HEALTH PHYSICS PERSONNEL 5
0 30 3.600
.000 8.300
-^
SUPERVISORY PERSONNEL 10 2
34 3.635
.440 44.348 ENGINEERING PERSONNEL 12 24 61 10.987 13.050 51.207
_ MILLSTONE 1 116 REPORT
- * *1978 * *
- flUMBER OF PERSONNEL OVER 100 MREM TOTAL Mall-REM STATION UTILITY CONTRACT WORKERS STATION UTILITY CONTRACT WORKERS EMPLOYEES EMPLOYEES AND OTHERS EMPLOYEES EMPLOYEES AND OTHERS WASTE PROCESSING IWINTENANCE PERSONNEL 18 1
23 11.950 1.215 42.260 OPERATING PERSONNEL 4
0 0
4.435
.000
.000 HEALTH PHYSICS PERSONNEL 2
0 1
1.560
.000
.825 SUPERVISOP,Y PERSONNEL 1
0 1
.995
.000 2.775 ENGINEERIrlG PERSONNEL 2
1 0
.735
.145
.290 REFUELING MAINTENANCE PERSONNEL 12 17 18 6.115 4.355 11.165 OPERATING PERSONNEL 4
0 0
2.465
.000
.000 HEALTH PHYSICS PERSONNEL 0
0 1
.040
.000
.470 SUPERVISORY PERSONNEL 0
1 0
.145
.190
.100 ENGINEERIflG PERSONNEL 1
3 7
.360
.985 2.755 TOTALS MAINTENANCE PERSONNEL 100 97 727 108.745 63.224 694.400 OPERATIllG PERSONNEL 73 1
0 73.180
.105
.000 HEALTH PHYSICS PERSONflEL 14 0
69 19.040
.000 36.695 SUPERVISORY PERSONNEL 21 3
37 12.740
.695 47.873 ENGINEERING PERSONNEL 21 31 83 14.827 14.980 59.510 TOTAL 229 132 91 6 228.532 79.004 838.478 C
G
MILLSTONE UNIT 2 CONTENTS Page CHANGES Design Changes 2/1
- 2/9 Procedure Channes 2/10 TESTS 2/11 - 2/16 STEAtt GENERATOR TUBE ISI 2/17 - 2/23 RADIATI0tl EXPOSURE 2/24 - 2/25
Page 2/1 PM NT DESIGN CHANGES The following list by design change number sunmarizes those design changes completed in 1978, relating to safety related equipment, which could have a potential impact on safety related systens, could potentially impact the environment, or required a change to the FSAR.
PDCR 2-45-75 Installation of drain pipe and isolation valve on CVCS strainer L-5 was rade to permit controlled drainage of hot fluid prior to renoving the horic acid punp discharge strainers.
This change does not affect the design function or operability of the system.
PDCR 2-51-75 Installation of a bypass switch to allow charging pump operation when pressurizer level deviates from the progran by greater than
+13 inches.
This change does not affect the operability or functional requirenents of the charging pump.
PDCR 2-58-75 Installation of RCP seal leakage and pressure recorders.
This change inproves nonitoring capability of operating personnel and does not result in changes to any safety systems.
PDCR 2-63-75 Installation of a hard piped venting system with a collection tank for the reactor coolant pumps to reduce the possibility of a radio-logical spill occurring.
This change reduces radiological hazards and does not affect any safety systems.
PDCR 2-35-76 Replacement of the diesel air start solenoid valves, deleting the strainers, hypass valves and water traps from the starting air tanks.
This change improved diesel reliability.
Page 2/2 PDCR 2-95-76 Addition of key lock switches for control of ESF punps' mininun flow recirculation valves, 2-SI-659 and 660, such that the valves will be locked open and an amber light will indicate if the valves are out of the open position.
This was to ensure that the minimun flow bypass line for safety injection pumps is not isolated without positive operator action.
The change ensures ESF pump operability by allowing positive means to maintain pump recirculation flow.
PDCR 2-222-76 Installation of supports to seisnically support the safety injection test header as corrective action for LER 76-38.
The change was to ensure the availability of the test header to conform to the small break analysis.
PDCR 2-230-76 Modification of the low suction pressure trip for the steam generator feed pumps to trip on a time delay at 280 psig and immediately at 250 psig.
This change prevents tripping a steam generator feed pump during a feedwater transient of short duration, does not adversely affect any safety related systems and improves system rel iabil i ty.
PDCR 2-281-76 Installation of improved position indication gear trains for eight HPSI isolation valves.
This change improves reliability and does not affect valve operation.
PDCR_2-286-76 Replacenent of CEDS 15 volt power supplies with new dual output units.
This change provides a nore reliable power supply and will not adversely affect any safety related functions.
PDCR 2-293-76 Addition of a static de-excitation circuit to the Alterrex excitation system to autonatically remove the excitation from the main generator field if the exciter field breaker did not open on a generator trip.
This change will not affect any safety related systems and provides additional generator protection.
-s.
.im men-.
.. i i-m
Page 2/3 PDCR 2-45-77 Provision for a long term low pressure / low tenperature reactor coolant systen relief capability which in turn provides nitigating capability against potential RCS overpressurization transients while the RCS is in a solid water condition.
This change meets the margins of safety as defined in the basis for the Technical Specifications and is within the safety criteria as defined in the Safety Analysis Report.
PDCR 2-50-77 Installacion of nitrogen lines for purging and blowdown of the steam generators.
This change will improve reliability and will not adversely affect any safety related systens.
PDCR 2-66-77 Installation of a shutdown cooling sample point on the LPSI punp discharge to provide sufficient notive force to obtain samples when the RCS is on shutdown cooling. This change will not adversely affe::t any safety related systen.
PDCR 2-85-7/
Provision for an alarm on C08 for diesel feeder breaker unavailability.
This change will inprove reliability and does not affect diesel operabil i ty.
PDCR 2-87-77 Addition of two globe valves with a vent between them in the nitrogen header inside the containment to facilitate the performance of local leak rate testing.
This change is to reduce personnel radiation exposure and to increase the accuracy of the test.
There will be no adverse affect on any safety related systen.
PDCR 2-151-77 Modifications to the design of the hangers for the extraction steam pip:ng to reduce system vibrations.
This change will decrease the probability of a pipe break event and will have ro adverse effect on any safety related systems.
Page 2/4 PDCR 2-194-77 Recoating of A and 8 service water strainer internals with carboglas 1678 to improve the corrosion resistance.
This change is not an unreviewed safety question because it will improve the reliability of the strainers.
PDCR 2-203-77 Installation of a one inch fiarinite board fire barrier between filters L29A and L29B and between filters L30A and L30B.
This change does not adversely affect any safety related system and will inprove the safety of the Enclosure Building filtration and Control Roan air conditioning systems.
PDCR 2-204-77 Cutting a twenty inch slot in the concrete foundation directly opposite the discharge of the A circulating water pump to facilitate pump renoval for maintenance and repair.
This change has no adverse effects on any safety related syste, or intake structure integrity.
PDCR 2-205-77 Installation of solenoid operated pneumatic operators to two existing hand operated valves in the chemical and volume ccntrol system.
This change conforms to the design analysis for the engineered safety systems and enables making rapid changes in reactor coolant system baron concentration for power control.
PDCR 2-209-77 Elimination of the diesel auto start function on a reserve station transformer (RSST) lockout (primary and secondary) and an RSST audio trip operate (primary and backup) to reduce unnecessary starts.
This change has no adverse effect on any safety related system and the starts were not credited in the FSAR.
PDCR 2-211-77 Replacement of the refueling machine hoist, bridge and trolley motor drive panels with programmed and Remote Systens Corporation's newest design. This change will greatly increase the operability and reliability of the refueling machine.
Page 2/5 PDCR 2-212-77 Replacenent of approximately three feet of conduit to the wide range nuclear instrumentation detectors with a junction box to prevent damage to the detector's cable when the detector is removed.
This change does not affect any safety functions and improves rel iabil i ty.
PDCR 2-217-77 This change connected reactor protection systen connon to plant ground to reduce RPS noise. The change resulted in increased RPS reliability and does not adversely affect any RPS functions.
PDCR 2-218-77 Installation of exhaust and inlet air capability in the Fire Punp House to reduce high hunidity conditions.
This change has no effect on any safety systens and improves reliability.
PDCR 2-221-77 Construction of enclosures for instrumentation on the RWST, PWST, CST and condensate surge tank.
This change is to prevent the instrumentation from freezing, ensuring operability of the protected c omponents.
PDCR 2-223-77 Addition of a reed switch in parallel to every other reed switch on the group 7 rods reed switch assemblies.
This change will increase the reliability of the reed switches used in group 7 and does not adversely affect any rod safety functions.
PDCR 2-225-77 Installation of two brackets in the vicinity of the dodge taper lock coupling on the refueling machine drive shaft to prevent the possibility of the coupling falling into the refueling cavity.
This change improves the reliability of the refueling machine and does not adversely affect any safety related system.
Page 2/6 PDCR 2-246-77 Replacement of diesel generator mufflers.
This change will not affect the diesel's design performance and will reduce exhaust noise in the mid and low ranges of the sound spectrum.
PDCR 2-255-77 Installation of an additional pipe hanger on the safety injection recirculation line. This change will prevent any further stress failures of this line as reported in LER 77-10.
PDCR 2-256-77 Addition of two trisodium phosphate baskets in the containment.
This change will improve the ability to maintain the design pH following the design LOCA with maximum RWST boron concentrations.
PDCR 2-270-77 Replacement of four existing low voltage containment penetration nodules with new Conax nodules.
This is to correct for shorted wires in existing penetrations and meets the plant safety and design criteria.
PDCR 2-284-77 Replacement of the RWST level transmitters with an improved type.
This change will improve level accuracy.
PDCR 2-285-77 Modification to the low steam generator pressure bypass bistable, to increase the bistable impedance to eliminate a voltage error.
This change will improve operations.
PDCR 2-8-78 Decrease the histable deadband for all RSST sensor assembly bistables.
This change does not adversely affect any safety related systen and inproves systen performance.
' ?s.
Page 2/7 PDCR 2-9-78 Installation of high radiation area barriers across the steam generators' secondary side manways to control access during naintenance periods. This change does not affect any safety related sys t em.
PDCR 2-14-78 Add provision for a turbine trip on steam generator high water level to protect the turbine. This change does not affect any RPS safety related trip functions.
PDCR 2-17-78 Addition of stiffeners to the 12U diesel generator skid base to reduce vibration. This change improves diesel reliability and does not affect the engine's operability.
PDCR 2-18-78 Renoval of the part length control elenent assemblies.
The part length CEA's were removed since their use was prohibited by Technical Specifications and they were not required for plant operations.
No credit was taken in the safety analysis for the part length CEA's.
However, deletion of these CEA's with no other changes decreases the shutdown margin discussed in the Technical Specification Bases.
The shutdown margin requirements were increased for Cycle 2 operation and approved by the NRC through the Cycle 2 reload Technical Specification submittal.
PDCR 2-22-78 Replacement of conductors in electrical penetration assemblies that exhibit a low insulation resistance. This change inproved system rel iabil i ty.
PDCR 2-27-78 Removal of the dynamic response function from the delta-T calculation in the reactor protection system.
This change was analyzed and meets SAR and Technical Specification requirements and is in conformance to the NSSS vendor setpoint study.
a 9-
Page 2/8 PDCR 2-34-78 The allowance of power operation with CEA guide tubes sleeved.
Guide tube sleeving was performed to ensure adequate fuel assembly integrity due to CEA guide tube wear.
This condition wis approved in the License Anendnent for Cycle 2 operation dated April 19, 1978.
PDCR 2-37-78 Podification to the tube support plates in steam generators one and two. The structural integrity of the primary and secondary steam generator systems is not compromised.
PDCR 2-39-78 Installation of seismic supports for Enclosure Building auxiliary steam and condensate.
This change will upgrade the piping system's supports and improve Enclosure Building integrity.
PDCR 2-44-78 Reconnection of the General Electric solid state negative sequence relay for the nain generator.
This change is to provide a trip vice an alarm for the main generator and improves generator protection.
PDCR 2-45-78 Replacement of the diesel service water TCV bypass flow orifices with blanks to inprove reliability by preventing mussel fouling.
PDCR 2-48-78 Elimination of the high SUR and high steam generator level reactor trips as called for by the Cycle 2 Technical Specifications.
This change is not an unreviewed safety question because the trips are not taken credit for in the safety analysis.
PDCR 2-54-78 Modification to the bearing plate for tendon ID21 to correct for a deformed plate during tendon surveillance.
This change does not affect the tendon integrity.
Page 2/9 PDCR 2-61-78 Installation of an upper end fitting on fuel assembly A066 capable of lifting and. holding 3,000 pounds and naintaining assembly integrity for storage in a spent fuel pool rack location.
This change is to replace the upper end fitting that was removed for a CEA guide tube study and restored the fuel assembly to a condition acceptable for storage and handling.
PDCR 2-68-78 Reduction of the T cold high annunciator setpoint to 543 F from 545 F to comply with a Technical Specification limit change.
This change does not affect any safety related systen.
Page 2/10 PROCEDURE CHANGES The following sunmarizes safety analyses for procedure changes as listed in the FSAR in accordance with the provisions of Title 10, Code of Federal Regulations, Section 50.59.
Procedure Change 2401D - RPS Matrix Logic and These changes modified the RPS Trip Path Relay Test calibration procedures to delete reference to the high stean generator 240lG - RPS Bistable Trip Test level and high startup rate reactor trips, deleted in PDCR 2-18-78 The safety analysis for this change is as listed in the design change section.
2617A - Radioactive Liquid The procedure was changed concerning Waste Discharges use of a new tank nixing systen prior to discharging tanks.
The previous tank nixing nethod used tank recircul-ation with installed radwaste systen discharge pumps.
The mixing method provided equivalent or better results than recirculation.
2211 A - Spent Fuel Inspection The procedure for spent fuel inspection was changed to specify the new inspection method utilizing the new fuel elevator, nodified under PDCR 2-219-77.
The safety analysis for this change is contained in the design change section.
2202 - Reactor Startup These procedures were changed to reflect deletion of the part length 2302A - Control Element Drive CEA's.
The evaluation of operation without these CEA's is contained with PDCR 2-18-78.
Page 2/11 TESTS The following list by test number summarizes those tests performed under the provisions of Title 10, Code of Federal Regulations, Section 50.59.
A summary of the safety analysis is included for each test.
None of the tests were evaluated as an unreviewed safety question.
T7 7-36 Trend of Shorted Penetration Wires - This test was performed to establish a trend of resistance readings based on time and temperature of various low voltage containment electrical penetrations. The testing involved data taking of unused penetration conductors and did not affect any safety systems.
T77-41 Rhodiun Detector Sensitivity Depletion Test - This test installed four incore test rhodium detectors for the purpose of obtaining accurate incore detector sensitivity correction factors.
The use of the test incore detectors is in conformance to all existing operations with the normal incore detector system.
T77-43 Special fionitoring of Insulation Resistance of Penet. ation Wires -
This test was for the purpose of nonitoring the insulation resistance values of a representative sample of equipment with connections through containment low voltage electrical penetrations.
The test involved megger data taking only and did not adversely affect the tested circuits' integrity.
T77-47 Retest of Low Tenperature/0verpressurization Circuit - This test procedure was conducted to verify operability of the RCS power operated relief valves following completion of a design change to incorporate low taaperature/ overpressure protection setpoints.
The test involved routine oper ation of the power operated relief valves.
The valves were isolated tron the RCS and the plant was in Mode 6.
Page 2/12 T78-2 RPS Grounding - The purpose of this test was to collect data on the results of modifying the reactor protection system fran a floating ground to a plant grounded scheme.
The test was performed with the reactor shutdown and on only one RPS channel.
T78-3 "A" Diesel Run-In After Overhaul - This test procedure was perforned to control the run-in of the "A" emergency diesel engine following periodic overhaul.
The test was conducted in accordance with maaufacturer's reconnendations.
T78-4 "A" Diesel Generator Operational Run - This test was conducted following T78-3 to verify proper engine performance under load.
The test was conducted in accordance with normal operating procedures modified to include additional data taking.
T78-6 "A" Diesel Generator Starting Logic Test - This test was conducted as a retest to verify proper operation of the diesel control schene following design changes deleting unnecessary start signals. The test was conducted in accordance with test methods established and used for Technical Specification surveillance testing.
T78-7 Spent Fuel Pool Platform Crane and Long Spent Fuel liandling Tool Test - This test was conducted to verify proper functioning of the SFP platform crane and long handling tool following modifications.
The test was conducted using normal fuel handling procedures with a dumny fuel bundle in an area of the SFP not containing fuel.
T78-9 Criticality / Low Power Physics Test - Cycle 2 - This test was part of the Cycle 2 reload startup test progran performed to verify proper core load and nuclear characteristics.
The testing was conducted in accordance with normal operating procedures and the Technical Specifications, including applicable special test exceptions.
Page 2/13 T78-10 Power Ascension Test - Cycle 2 - The power ascension test was perforned as part of the Cycle 2 reload startup test program.
The testing verified proper power range plant perfornance, including power dependent nuclear characteristics.
The power ascension test was conducted in accordance with normal plant operating procedures and Technical Specification requirements.
T78-11 "B" Diesel Generator Starting Logic Test - Sane test and evaluation following modifications as performed for the "A" diesel in Test T78-6.
T78-12 Steam Generator Leak Test - This test was conducted by pressurizing the stean generator secondary sides to verify zero leakage following extensive repairs and tube plugging during the refuel outage.
The testing was conducted within steam generator pressure / temperature limitations and did not affect steam generator integrity.
T78-13 Post-LOCA Boron Precipitation Control Flow Path Verification - This test was conducted to measure flow rates of the system alignments used during post-LOCA boron precipitation control to obtain data for use in the operating procedure.
The testing involved normal system flow paths and was conducted during reactor cavity fill with the core of f-loaded.
T78-14 Radwaste Tank Mixer - Eastern Model RSG This test was conducted following the installation of tank nixers in various radwaste tanks.
The test deten,ined appropriate mixing times to meet Environmental Technical Specification requirements and did not affect safety.
T78-15 CEA Pulldown Coil Voltage Test - This test obtained data on the operating electrical characteristics of the CEA pulldown coils and was conducted with the reactor shutdown using normal control element drive systen operating procedures.
Page 2/14 T78-16 RWST Level Transmitter Calibration Check - Tne purpose of this test was to provide more accurate data for use in the refueling water storage tank level transnitters' calibration procedure.
The test was conducted in Mode 6, when the transmitters were not required.
T78-17 Scavenging Steam Supply to Moisture Separator / Reheater A & B -
This test verified motor operated valve operability following design nodifications.
The test did not affect any safety related systems.
T78-18 RCS Low Temperature Overpressure Alarms Test - This test was conducted in addition to Test T77-47 to verify the operability of the new low tenperature overpressure protection control schene, including alarms. The test was conducted in Mode 6 with no overpressure protection required.
T78-19 Motor Operated Valves - Steam to Moisture Separator / Reheater -
This test verified valve operability following nodifications and did not affect any safety related systems.
T78-22 Diesel Generator SIAS Start Test - This test was performed to verify automatic start of the energency diesel generators on receipt of a safety injection actuation signal to complete a surveillance testing requirement. The test was performed in accordance with normal surveillance methods and did not affect diesel operability.
T78-23 Pressurizer Spray Valve Hot Functional Test - This test was performed to demonstrate proper spray perfonnance following replacement of the pressurizer spray valves.
The test was conducted within the bounds of nornal operating procedures and system parameters.
Page 2/15 T78-24 Steam Generator Flush - This test was conducted to reduce residual steam generator chloride levels following plant startup af ter refuel ing. The procedure was conducted in hot standby by drain and fill evolutions within normal limits and normal blowdown at power.
T78-27 Waste Gas Flow Test - This test was for data taking through the normal waste gas discharge path and did not affect any safety related systens.
T78-29 CEA Motion Inhibit Verification - The purpose of this test was to verify acceptability of using CEA Group 5 for perfonnance of periodic surveillance testing. The test was conducted within the linits provided by Technical Specifications and did not affect any safety related functions of the CEA system.
T78-30 Diesel Fuel Oil Pressure Drop - The test objective was to neasure diesel fuel pressure drop on supply valve closure to gain data for a design change. The test was conducted with the diesel out of service for routine preventive maintenance, and thus did not affect engine operability.
T78-34 CEA 42 Position Indication System Test - This test was a retest following modification to CEA 42 reed switch position indication systen.
The test did not affect any safety related components.
T78-37 Feedwater Venturi Flow Data - This test was performed using a Na-24 tracer injection into the feed train to gather data on feed flow Venturi calibration. The test did not affect any safety related systens and precautions were taken to prevent any unmonitored Na-24 releases.
Page 2/16 T78-38 Feedwater Heater Tube Leak Test - This test involved the injection of a fla-24 tracer into the condensate header in order to provide data for the evaluation of feedwater heater tube integrity. The testing did not affect any safety related systems and precautions were taken to prevent any unmonitored releases.
T78-39 floisture Separator Drain Flow fieasurement - This test involved the injection of a fla-24 tracer into the moisture separator drain lines in order to determine drain flow rates.
The testing did not affect any safety related systens and precautions were taken to prevent any unmonitored releases.
T78-40 Set Fire Pump Bypass Relief and Throttle Valve - This test was an operability test following a design change to install a fire pump bypass relief valve and throttle valve.
The test confirmed proper design completion and did not affect any safety systens.
Page 2/17 STEAM GENERATOR TUBE INSERVICE INSPECTION RESULTS This section provides a summary of the steam generator inservice inspection results for the steam generator tubes, in accordance with Technical Specification 4.4.5.1.5.b.
In response to the specific requests for data per Technical Specification 4.4.5.1.5.b:
1.
The number and extent of tubes inspected are included in the summa ry.
Inspection results for tube denting inspections, in addition to the inservice inspection requirements, are also sunnarized.
2.
There were no tubes identified with an indication of an inperfection affecting wall-thickness penetration.
3.
There were no tubes plugged as a result of the inservice inspection results.
The initial stean generator tube inservice inspection was conducted in 1977.
However, additional testing was required due to the evidence of tube denting as reported in the sunnary. The results of the tube denting measurenents and corrective actions progran, performed during the 1977-1978 refueling outage, were presented to the NRC in letter W. G. Counsil to R. Reid, dated 8/28/78.
SUMMARY
The first inservice eddy current examination was perforned on steam generators #1 and #2 at Millstone Point-II for Northeast Utilities during the period of May 23, 1977 through June 4, 1977.
The exanin-ation was performed by Combustion Engineering Power Systens, Systens Integrity Services Group personnel. The inspection progran exceeded the requirements of the U. S. Nuclear Regulatory Connission Regulatory Guide 1.83, Revision 1 (July 1975), " Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes", and the plant's Technical Specifications.
These required that a minimun of 3% of the 8,500 tubes (256 tubes) in each of the two generators be inspected for wall degradation.
The stean generator tubing is nickel-chromium-iron alloy 600 (Inconel) with noninal dimensions of 3/4 inch OD and 0.050 inch wall.
The calibration tube standards for the defect detection examination were supplied by Millstone Point-II Station.
Page 2/18 The comparison standard tube for the dent assessment exanination i
along with pertinent documentation was supplied by Combustion Engineering and retained by Station personnel for future use.
The liillstone Point-II steam generator design has 8,519 U-tuhes.
There are a handful of manufacturing plugs in each generator to bring the net number of tubes to 8,508 and 8,500 for steam generators
- 1 and #2, respectively.
The tube bundle is supported at eleven elevations.
From the botton up, the first seven elevations have carbon steel egg crate supports for the full bundle, elevations 8 and 9 are carbon steel egg crates over a portion of the bundle and elevations 10 and 11 are drilled hole carbon steel support plates over a portion of the bundle.
The egg crate structure contacts two inches of the tube and the drilled hole supports are 3/4 inch thick.
The number of U-tubes passing through the number 10 and 11 support plates are 2,225 (26.1% of total) and 771 (9.1% of total) respectively.
An elevation view of the secondary side support structures is shown in Figure 1.
The NRC Regulatory Guide mandatory inspection defect detection prograns were concentrated in one quadrant of each hot leg and included tubes in the outer periphery, in the center of the bundle, around the support stays and near the divider lane.
In particular, the section of the tube bundle passing through drilled hole carbon steel support plates was exanined for service induced denting.
When denting was found in this region, a supplemental dent assessment examination was added to the inspection program.
A cold leg defect detection inspection was done in one generator as well as routine sludge height neasurements.
An inspection of peripheral tubes in the hot leg of steam generator #2 was done to detect if any tube damage had been caused by loose parts found in the secondary side adjacent to the tube bundle.
RESULTS OF INSPECTION Upon completion of each reel of tape, the strip chart and magnetic tape recorded data were reviewed by the data analyst to evaluate any indications. A sanple of these inspection results was later compared with the previous baseline data to detect any changes in the tubes. The results of both the on-site analysis of the 1977 inspection data and the comparison with the 1973 baseline data for both generators are discussed below. Tables I and II summarize the inspections of steam generators #1 and #2, respectively.
Page 2/19 Stean Generator #2 Steam generator #2 was the first one to be inspected. A total of 347 tubes (4.1% of bundle) met the plant's Technical Specifications for a complete defect detection inspection. Of this group, 79 were inspected from both the hot side and cold side so that the entire tube length was inspected.
In addition, a number of tubes were "short runs" for which an incomplete inspection for the horizontal run of the U-bend was made.
However, for these tubes, the hot leg data is valid and was analyzed.
The defect detection inspection for cold side included 79 tubes.
The hot side and cold side defect detection inspection revealed no tube wall degradation indications in any area of the stean generator.
Denting of the tubes at the drilled hole supports was observed at every tube / support plate intersection inspected.
Based on this sampling inspection, it was concluded that each tube / support plate intersection is dented, involving approximately 26% of the total number of tubes.
In order to assess the approximate size of the dents at the drilled hole supports, an additional eddy current test was done on 104 tubes to conpare dent signal amplitudes to those from known dents made by uniformly reducing the tube diameter around the circumference.
The dents on the cold side appeared larger than those of the hot side and support #10 had larger dents than support #11.
This dent assessnent data will serve as a baseline for comparison with future inspections tr. conitor any potential increase in denting.
Around the tube bundle periphery in the hot side,138 tubes were inspected, typically only to the first egg crate.
This inspection was done to detect damage that may have been caused by loose parts in the secondary system. This inspection complemented a visual inspection done for the same purpose.
No indications of damage were observed.
Sludge height testing was done in 61 and 70 tubes in the hot and cold side, respectively.
The sludge profiling was done in one quadrant of each side.
It is assuned that the adjacent quadrant has essentially the sane sludge profile.
The maxinun sludge depth both the hot and cold side was approxinately 6 inches.
A summary of the inspection program, including the results of the dent assessment program for steam generator #2, is given in Table II.
Page 2/20 Stean Generator #1 In steam generator #1, the inspection prograns were done on the hot side only.
The defect detection inspection to satisfy the plant's Technical Specifications included 271 tubes (3.2% of total).
Again, the tubes that are "short runs" are not included in the total, since the entire U-bend was not inspected, however, the hot leg data is valid and was analyzed.
No tube degradation was observed in any region of the-generator. As in the other steam generator, denting was observed at the drilled hole supports on both the hot and cold legs and it was concluded that every tube / support plate intersection was dented (affecting 26% of the totai number of tubes).
A dent assessment program was done on 99 tubes in the hot leg in steam generator #1. As before, the dents at support plate #10 are slightly larger than at support plate #11.
The size of the dents in steam generator #1, however, are slightly less than those in s team generator #2.
A sludge height inspection was done on 74 tubes in stean generator El hot side. The sludge profile was done in one quadrant and it is assumed that the adjacent quadrant is similar. The maximun sludge height was 10.5 inches.
A summary of the inspections, including the dent assessment results, is given in Table I.
Page 2/21 TABLE I
SUMMARY
OF STEAM GENERATOR #1 EDDY CURRENT INSPECTION Test Type Tubes Examined Results
% of total in S/G Hot Side Defect Detection 271 3.2 No tube wall degradation; denting at drilled hole supports Hot Side Sludge Measurement 74 0.9 Maximum depth - 10.5 inches Hot Side Dent Assessment 99 1.2 See Below.
DENT ASSESSMENT RESULTS Tubes Examined Dent Size (mils off nominal radius)
- % of total
- Mean i 1 7 Range
@ #11 Support Hot Side 63 8.2 3.8 1 1.0 2-6
@ #10 Support Hot Side 99 4.4 4.9 i 1.5 2-9
- Based on number of tubes passing through partial support plate.
Page 2/22 TABLE II
SUMMARY
OF STEAM GENERATOR #2 EDDY CURRENT INSPECTION Test Type Tubes Examined Results
% of total in S/G Hot Side Defect Detection 347 4.1 No tube wall degradation; dents at drilled hole supports Cold Side Defect Detection 79 0.9 No tube wall degradation; dents at drilled hole supports Hot Side Sludge Measurement 61 0.7 Maximum depth 6 inches Cold Side Sludge Measurement 70 0.8 Maximum depth 6 inches Dent Assessment (Hot and Cold) 104 1.2 See Below.
Peripheral Tubes 138 No damage observed.
DENT ASSESSMENT RESULTS Tubes Examined Dent Size (mils off radius)
% of total
- Mean i 10-Range
@ #10 Support Cold Side 95 4.3 8.6 1 2.0 6 - 14
@ #11 Support Cold Side 59 7.7 7.9 1.5 5 - 13
@ #11 Support Hot Side 65 8.4 4.9 1 1.0 2-8
@ #10 Support Cold Side 104 4.7 5.6 1 2.0 3 - 14
- Based on number passing through partial support plates.
u STEAM s
OUTLET DRYERS SEPARATORS SECONDARY
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MANWAY N qwgD00g ll
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AN
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PARTIALDRILLEDHOLE y
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- IT 4
FEEDWATER INLET STEAM GENERATOR LIQUID LEVEL 13' 9" OD -
> TUBES 62' 5"
= _.
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EGGCPATE
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N TUBE SUPPORT (TYP)
I.
4 HANDH0LE L
M-i- -eN BOTTOM 2
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REACTOR COOLANT N REACTOR COOLANT INLET OUTLET e
FIGURE 1 - ELEVATI0ri VIEW 0F MILLSTONE P0lf4T-II STEAM GENERATORS
MILLSTONE 2 116 REPORT
- * *1978* *
- NUMBER OF PERSONNEL OVER 100 MREM TOTAL MAN-REM STATION UTILITY CONTRACT WORKERS STATION UTILITY CONTRACT WORKERS EMPLOYEES EMPLOYEES AND OTHERS EMPLOYEES EMPLOYEES AND OTHERS REACTOR OPERATIONS &
SURVEILLANCE MAINTENANCE PERSONNEL 3
0 25
.940
.115 13.100 OPERATING PERSONNEL 12 0
0 S.353
.000
.000 HEALTH PHYSICS PERSONNEL 5
0 42 5.330
.000 36.852 SUPERVISORY PERSONNEL 1
0 4
.665
.010 1.415 ENGINEERING PERSONNEL 5
0 5
1.512
.295 1.405 ROUTINE MAINTENANCE MAINTENANCE PERSONNEL 2
0 1
1.335
.195
.520 OPERATING PERSONNEL 1
0 0
.390
.000
.000 HEALTH PHYSICS PERSONNEL 0
0 0
.030
.000
.045 SUPERVISORY PERSONNEL 0
0 0
.000
.000
.000 ENGINEERING PERSONNEL 0
0 0
.005
.005
.01 5 INSERVICE INSPECTION MAINTENANCE PERSONNEL 3
3 378 1.145
.665 524.922 OPERATING PERSONNEL 1
0 0
.195
.000
.000 HEALTH PHYSICS PERSONNEL 2
0 3
.425
.000
.970 SUPERVISORY PERSONNEL 0
0 24
.005
.000 29.760 ENGINEERING PERSONNEL 3
0 39 2.640
.125 37.212 SPECIAL MAINTENANCE MAINTENANCE PERSONNEL 39 20 497 31.218 8.100 647.616 OPERATING PERSONNEL 6
0 0
1.630
.000
.020 y
HEALTH PHYSICS PERSONNEL 3
0 11 1.015
.000 3.380 g
SUPERVISORY PERSONNEL 2
0 38 1.170
.000 46.190 y
ENGINEERING PERSONNEL 4
3 50
.940 1.550 57.507 g
WASTE PROCESSING MAINTENANCE PERSONNEL 0
0 1
.095
.025
.325 OPERATING PERSONNEL 1
0 0
.330
.000
.000 HEALTH PHYSICS PERSONNEL 0
0 0
.170
.000
.000 SUPERVISORY PERSONNEL 0
0 0
.050
.000
.005 ENGINEERING PERSONNEL 0
0 0
.000
>000
.000 MILLSTONE 2 116 REPORT
- 1978***
NUMBER OF PERSONNEL OVER 100 MREM TOTAL MAN-REM STATION UTILITY CONTRACT WORKERS STATION UTILITY CONTRACT WORKERS EMPLOYEES EMPLOYEES AND OTHERS EMPLOYEES EMPLOYEES AND OTHERS REFUELING MAINTENANCE PERSONNEL 24 4
30 10.490 1.330 15.065 OPERATING PERSOUNEL 1
0 0
.545
.000
.000 HEALTH PHYSICS PERSONNEL 0
0 0
.000
.000
.025 SUPERVISORY PERSONNEL 1
0 2
.380
.000
.975 ENGINEERING PERSONNEL i
0 3
.450
.025 1.195
--TOTALS MAINTENANCE PERSONNEL 71 27 932 45.223 10.430 1201.548 OPERATING PERSONNEL 22 0
0 8.443
.000
.020 HEALTH PHYSICS PERSONNEL 10 0
56 6.970
.000 41.272 SUPERVISORY PERSONNEL 4
0 68 2.270
.010 78.345 ENGINEERING PERSONNEL 13 3
97 5.547 2.000 97.334 TOTAL 120 30 1153 68.453 12.440 1418.519 2
c N