ML20093E145
| ML20093E145 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/05/1984 |
| From: | Hukill H GENERAL PUBLIC UTILITIES CORP. |
| To: | Haynes R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| 5211-84-2127, NUDOCS 8407170387 | |
| Download: ML20093E145 (18) | |
Text
_
ENuclear tsu",n";ffra" "
Route 441 South Middletown, Pennsylvania 17057 0191 717 944 7621 TELEX 84 2386 Writer's Direct Dial Number:
July 5, 1984 5211-84-2127 Mr.
R. C. Ilaynes Region I, Regional Administrator U. S. Nuclear Regulatory Omnission 631 Park Avenue King of Prussia, PA.19406
Dear Mr. Ilaynes:
Three Mile Island Nuclear Station, Unit I (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 10 CFR 50.59 Report In accordance with the requirments of 10 CFR 50.59, enclosed please find two copies of changes to TMI-l systes and procedures as described in the IEAR.
Sincerely,
- 1. D. Iukill, Director, 'IMI-l IIDil/ RAS /mle Enclosures cc: Director, Office of Inspection and Enforcement (40 copies)
U. S. Nuclear Regulatory Cm1 mission Washington, D.C. 20555 Director, Office of Management Information and Program Control U. S. Nuclear Regulatory Ormission Washington, D.C. 20555 John F. Stolz, Office of Nuclear Reactor Regulations U. S. Nuclear Regulatory Omnission Washington, D.C. 20555 kD 0
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q GPU Nuclear Corporation is a subsidiary of the General Pubile Utilities Corporation
B/A 412021 - Reactor Coolant System High Point Venting System Reactor Vessel Head Vent DESCRIPTION OF PROJECT:
The Reactor Vessel Head Vent has been installed in order to improve the plant's ability to vent a mixture of reactor coolant liquid / steam and/or non-condensable gases from the Reactor Coolant system, without having an adverse impact on core cooling. This safety grade modification satisfies Seismic Class I criteria and is supplied with Class IE electrical and instrumentation power.
The Reactor Vessel Head Vent is controlled by olenoid-operated isolation valves RC-V 42 & 43 which are mounted in series and vents directly to the Reactor Building atmosphere. Flow element FE 1081 and differential pressure transmitter DPT-1081 provide input signals to the " flow /no flow" lamps installed on panel "PC".
The vent valves are activated frein the Main Control Room "PC" panel which includes open/ closed key lock switches and indicating lights. Annunciation of inadvertent valve opening has been provided in light box "G" in the Main Control Room. The Reactor Vessel Hea0 Vent will be maintained and operated via strict administrative controls'.
SAFETY EVALUATION
SUMMARY
The purpose of this modification is to provide a remote power operation of the vent line from the Reactor Vessel Head.
Administrative procedures will be implemented for controlling the operation of the Reactor Vessel Head Vent from remote controls in the Main Control Room. This modification does not create the possibility of an accident or malfunction different from any previously evaluated in the SAR, i.e., failure of the PORV which is already an evaluated accident. No safety margins have been reduced as a result of this modification.
The probability of an inadvertent venting cf the Reactor Vessel Head Vent is slightly increased, since remote power operation of the double isolation valves is controlled from the Main Control Room; however, strict administrative controls and key lock switches will govern actuation of the Head Vent line.
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A GANGE FDDIFICATION:
Screenhouse Wall / Emergency Lighting Upgrade B/A 412347 DESCRIPTION OF GANGE:
This nodification adds a UIeA-Iabeled 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire rated roll-up door, 3-hour fire rated penetration seals in the Screenhouse wall, an 8-hour emergenef light and a acumunication handset in the intake screen and pumphouse between fire zones ISPH-FZ-1 and ISPH-F2-Z containing redundant equipTent needed to bring the plant to safe shutdown. In addition, an existing 8-hour energency light and handset will be relocated so as not to interfere with the operations of the door.
211s nodification also includes the installation of an 8-hour emergenef light in the Intermediate Building floor elev. 295'-0", near valve RRV-2.
These nodifications were identified in the Safe ShutdcEn Evaluation Report, "TMI-l Fire Hazards Analysis and Appendix R,Section III, G" dated June 28, 1982 as needed to elcminate non-empliance with 10 CPR 50 Appendix R.
SAFE 1"I EVALUATION:
The important Safety Functions enhanced by these modifications are that, within the intake screen and pumphouse a single fire in one fire zone cx:uld darrage a nuclear service river water putp and the redundant nuclear service pmp. The addition of a 3-hour fire rated roll-up door, and the 3-hour penetration seals will separate these two purrps and will have no impact on the functional operation of the pmps and associated acrrponents.
The installation of the emergency lights in the Internediate Building and Intake Screen and Pumphouse will have no impact on the functional operation of the plant.
The addition of the roll-up door and the 3-hour penetration seals between fire zones ISPH-FZ-1 and ISTH-FZ-2 will enhance the safety function by separating redundant equipmnt needed to bring the plant to safe shutdown. This door will remain open at all times. It will have an electro thermal link which will be interlocked with the ionization detection system of both fire zones. In case of a fire, the roll-up door will close and separate fire zones ISPH-FZ1 and ISPH-FZ-2.
It is, therefore, concluded that the subject nodification does not involve any unreviewed safety concern.
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i CHANGE. MODIFICATION:
B/A-412375 Turbine Building Contamination Detection and Control DESCRIPTION OF CHANGE:
1 This modification upgraded existing systems / components and enhanced utilization methods to provide more accurate and timely OTSG leak detection l
' techniques and monitoring (for accountability). New instruments were added to provide redundancy and relief from manual sampling which was utilized to I
achieve these functions and included portable steam line (NaI), area and airborne radiation monitors, turbine building sump liquid effluent monitor and composite sampler proportional to flow.
i SAFETY EVALUATION:
i The Liquid Radiation Monitoring System (LRM) for the Turbine Building sump a
1 functions continuously during all plant operating modes once the equipment is energized.
If a high radiation setpoint is detected the sump pump is tripped 4
off and an alarm is annunciated in the control room. A plant operator must go to the LRM control panel to investigate and resolve the problem.
If high i-radiation is confirmed by the control panel meters and/or analog recorder, operations will be required to either re-align the sump discharge valves from the IWTS to the Unit #2 Condensate Storage tank *, or increase the dilution j
factor (by increasing MDCT flow and decreasing the release rate from the T.B.
j Sump) such that the activity level will be below 10 CFR 20 MPC limits at the discharge point.
The keylock normal / override switch on the LRM control panel must be placed in ' override' to remove the trip function and restart the sump 4
i pumps after valves re-alignment.
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The LRM system performs a monitoring function of alarm annunciation and pump trip and no other control functions. The equipment can not cause a casualty event.
It assists operations to assure that plant discharge liquid and i
filtercake will not exceed 10 CFR 20 maximum permissible concentrations, j
j The LRM is classified as Important to Safety, non-seismic and non-1E.
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Conclusion:
1.
The probability of occurrence or consequence of an accident previously evaluated have not been increased.
j 2.
No accident other than those previously considered will be introduced.
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No safety margins have been reduced.
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B/A 412021 - Reactor Coolant System High Point Venting System RCS Loop A & B Hot Leo Vents DESCRIPTION OF PROJECT:
The RCS Loop A & B Hot Leg Vents have been installed in order to improve the plant's ability to vent a mixture of reactor coolant liquid / steam and/or non-condensible gases from the Reactor Coolant System.
This safety grade
- modification satisfies Seismic Class I criteria and is supplied with Class 1E electrical and instrumentation power.
The RCS Loop' A & B Hot Leg Vents are controlled by solenoid-operated isolation valves RC-V 40A and 41A and RC-V 40B and 41B which are mounted in series and vent directly to the Reactor Building atmosphere.
Flow elements FE 1080 and 1082 and differential pressure transmitters DPT-1080 and 1082 provide an input signal to the " flow /no flow" lamps installed on panel "PC".
The vent valves are activated from the Main Control Room "PC" panel which includes open/ closed key lock switches and indicating lights. Annunciation of inadvertent valve opening has been provided on light box "G" in the Main Control Room. The RCS Loop A & B Hot Leg Vents will be maintained and operated via strict administrative controls.
SAFETY EVALUATION
SUMMARY
The purpose of this modification is to provide remote power operation of the vent lines from the high points in the RCS Loop A & B Hot Leg piping.
Administrative procedures will be implemented for controlling the operation of the RCS Loop A & B Hot Leg Vents from remote controls in the Main Control Room. This modification does not create the possibility of an accident or malfunction different from any previously evaluated in the SAR, i.e.,- failure of the PORV which is already an evaluated accident. No safety margins have been reduced as a result of this modification. The probability of an inadvertent venting of these vents is slightly increased, since remote power operation of the double isolation valves is controlled from the Main Control Room; however, strict administrative controls and key lock switches will govern actuation of the isolation valves.
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CHANGE MODIFICATION:
H02 Waste Gas Analyzer B A 412215 DESCRIPTION OF CHANGE:
The Hays Gas Analyzer was the only hydrogen / oxygen monitor installed in the WDG System. This project added the second analyzer (Beckman).
The Beckman Analyzer added by this project will alleviate the maintenance and cperability problems with the existing Hays Gas Analyzer.
SAFETY EVALUATION:
The addition of a second H 022 Analyzer will allow the Waste Gas Holdup System to remain operational when existing Hays Gas Analyzer is out of service for maintenance.
Therefore, the addition of the Beckman Analyzer does not increase the probability of occurrence or the consequence of an accident or malfunction of equipment and increase the overall plant availability.
GANGE FDDIFICATION:
B/A 412052 - Containment Isolation - Line Break Detection (IM-5D)
DESCRIPTICN OF CHANGE:
The 4 psig reactor building pressure signal was deleted from several valves to maintain services to the BC Punps and notors. The systens involved are closed systems thus preventing a leak path to outside containment. Autcmatic isolation in the event of line break has been added to preclude loss of containment integrity in the event the closed system is damaged.
(bntairroent Isolation on NSCC and ICC system pipe line break is provided by the addition of redundant IE safety grade actuation trains. The trains are cmprised of level transmitters which monitor NSCC and ICC surge tank levels, signal conditioning for indication, detection and alarming tank leakage (pipe line break) and pmvide an isolation signal to valves IC-V2, IC-V3, IC-V4, IC-V6, NS-V4, NS-V15 and NS-V35.
Ica surge tank level ooincident with an HPI actuation signal closes the valves on the affected system.
Service Lines Isolated by Line Break with HPI Isolation Signal l
Intermediate Cboling Water Outlet Line Intermediate (boling Water Supply Line l
Intermediate Cooling to CRD! Cooling (bils Reactor Coolant Pump bbtor Cooling Water Supply Reactor Coolant Pump Motor Cooling Water Return SAFLTI EVALUATICN:
1.
The syetem is designed as safety grade and single failure proof. Thus, the system will perform its safety function when required. The probability of containment isolation occuring on demand is increased.
2.
Spurious. initiation of an isolation signal will not introduce new accidents into the plant design. Because the system is designed to be single failure proof, spurious initiation of any of the above signals will not isolate any conponents.
3.
No safety margins have been reduced, only oquignent added to insure that radiation is not released frcm containment.
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B/ Ail 23004 - Replaosnent of (bntainnent Isolation Valve Mh44 ~
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aw WDG-V4 Sas a solid wedge Hahoock 950W gate valve with an air-to-open/ spring-to-close actuator. %e valve ntem was sealed by a conventional packing gland.
s An autcznatic fluid blockirt eater supply was connected to the valve bonnet area.
- gi' Se new. valve is'a 2" Target hock Model' 80Z-14-008 giche valve with a silicone rubber "O" ring in the disc seating surface,'and with a spring-to-close/ solenoid-
%c)s/alve/bannet-/ stun assenbly is aaaled for zero leakage to-open actuator.
(no stem packing).
Mat rlal is 315 stainless steel and the, ends are sockec-s welded. %e solenoid actuator ga low DC current of G mps with 90-145 volts. Nameplate rating is' 300 F/55 psig/ ASESection III Class s
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A detailed purchasing spec 61caticm,was uk d,to assure that the new' valve would b
l ~3' meet the requirements of WOG s/otaservice in the reactor building. %e reactor building envirconent is nerefseverepan that inche auxiliary. building where this valve will be installect.s Q, the Equipnent' Qualification Greg of EP & I has reviewed a11'qualifica onf etanentation.
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NO me valve design and installation d5sure that it functions properly for both containment imlation and WDG ' istem beurylary.,.It,is expected to perform these, s
functions mohe reliably,than the former systemtvalve? We necessary elimination p,.'c of the fluid blocking provigicm.is fully Mified because of the denonstrated i aliability v.d low leakage of other qxisting containment isolation globe valves.
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'ItElehkege.@ expected %be fpr belch the ma'.cimtun allowable per 10 CFR Appendix J.
We fluidyack,systern will.be more reliable for other renaining gate valves when it no longer nee'ds to supply the high ledcage to WDG-V4.
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me nodif' cation will not therefore increase tne probability,of occurrence or the consa m of an a m h. N N
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me valve; during nonnal plant operat. ion, will function as did the former valve, i
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3 to allow flow of waste gas from the R. C. ' Drain Tank to the waste gas treatment
,systen.
(his function prcNents over-presshn of the R. C. Drain Tank whic5 would lead to thvfa to the drain thk rupture disc 'and resultant contamination of containment.) ne proposed. modification will notl therefore ' increase the probability of occurrence or,cLwmas'of a malfunction of equiptent I. T. S.
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Likatise, it wikno't create a posslibility[for'an ' accident or malfunction of a.
g different type than any previously ' identified 'in the FSN. since'it replaces,3 valve which already performs the sans norad1 and emarwney function.
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baeradiol'gidal'exposur IIthatitwil pNEtde Use cf this h'sys'vgwill r o
tighter MXi ten and Redctor Building boundaries. %ere will be no change in predicted relbases beyond license application. Therefore, the deiligned
.systen will nd,p edversely affqct the safetyR and is consisten,t wit;h AIARA
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BA 412052 - Centalment Isolation en 'digh Padiatica (Fy-5B)
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DFSCRIPTICN OF CRWGE::
p A s:rall break IDCA can result in high reactor building radicactivity kithout i
generating the 4 psig reactor building isolatica signal which would no= rally isolate the reactor building.. Therefore, the actuaticn of the reacter building isolation is assured by partial isolation and diversity of signals fran hich I
reactor building radiation. Piping which could transfer high levels of radiation fzun either the reactor coolant system er the reactor building are individually nonitored to detect high levels of radioactivity and initiate ',
closure of the appropriate valves, or almm.
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Follcwing services lines are detected using existing or new detectors and isolated on high radiation.
Scu'rce Radiation Detector Incation Wre of Moniter Steam Generator Sanple Outside the R.B. near the sangling Area Ga:rra Lines line deanstream cf the contai:Inent Cetectors (Kex);
isolation valve and upstream of connection for Turb. Plant Saapling Letdcun Line to Purification Outside R.B.
In line' Demineralizers (existing)
Pressurizer-and Peactor Outside the R.B. hetween the Area Gam lra
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Coolant Sa::ple Lines isolation valve and the sample Datectors (Nes);
cooler Reactor Ccolant Ptnps '
Oatside the F3 downstream of the Area Ga. a.
Seal Return containment isolation Valves Detectors (New)
(Ala=n cnly, operator actica required to close valves)
Reactor Ccolant Drain Tank Oatside of the tank Area Ga:ra Vent Datector (New)
Reactor Coclant Drain Tank-Pc:p Discha.rge -
Reactor Building Oatlet Outside the R.B.
In line and Inlet Purge Lines (Existing)
Pasetor Building Strrp F3 Step, mounted ir. side a Stnp Area Drain seismically supported pipe l'e. iter (New) 4 a
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New and. existing ram atiati :nonitors will be utilized to detect high levels of radioactivity and to initiate the closure of the associated
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isolation valves in piping which could transfer high level of radiation frcrn either the reactor coolant system or the reactor building to the outside. This nodification therefore provides timely isolation of con-tainment on high radiation, reducing the radiological safety concerns without impacting the designed functions of the lines being nonitored.
s All of the lines which receive the high radiation isolation signal y
currently receive the 4 psig reactor building isolation signal. There-fore, the spurious initiation of any ona of these signals does not I
introduce a new accident or transient into the plant design.
er This modification does not increase the probability of occurrence or
., ~, the consequences of an accident or malfunction of equipnent inportant to safety. This modification does not create the possibility for an accident or malfunction of a different type than any evaluated previously fin the safety' analysis report. No safety margins defined in the basis for any technical specification have been reduced by this nodification.
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CHANGE MODIFICATION:
RB Spray System Sodium Thiosulfate Tank Deletion - B/A 412073 - LM-7A DESCRIPTION OF CHANGE:
This modification removes a section of the 4 inch Sodium Thiosulfate Tank supply line to the Reactor Building Spray Pump suction headers and closes the remaining piping by installing weld neck and blind flanges.
SAFETY EVALUATION:
This is an Important to Safety modification only from a pressure boundary integrity standpoint. The DHRS cnd RBSS pressure boundary integrity is assured by the 'Important to Safety" classification and quality assurance pertaining to this modification.
This modification will not increase the probability of occurrence of an accident because the deleted sodium thiosulfate tank had no normal operation function nor did it support the reactor normal ar safe shutdown. Also, the modification will not increase the consequences of an accident because the iodine removal function, originally performed by the deleted sodium thiosulfate, will be successfully peformed by the sodium hydroxide.
Off-site radiation doses will remain far below 10CFR100 limits.
This modification will not increase the probability of a malfunction of ITS equipment because (1) corrosion effects of the spray solution on ITS components are considered minimal (per requirements of ANS-56.5 and SRP 6.5.2) if the system sprays solution with pH between 8.5 and 11, and (2) where the sodium thiosulfate piping shares space with ITS equipment, a seismic Category I support was added to the sodium thiosulfate piping to assure its integrity during a SSE. Also, the modification will not increase the consequence of a malfunction of ITS equipment because the fodine removal function will be successfully perfonned by the sodium hydroxide.
A possibility for an accident or nalfunction of a different type than any previously f aentified in the SAR was not created by this modification because (1) the iodine removal function of the deleted portion of the RBSS was transferred to the portion of RSS that existed before and functioned in the same manner, and (2) there were no components or equipment added to the disconnected sodium thiosulfate piping except a seismic Category I support to assure pipe integrity during a SSE (continued)
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' SAFETY EVALUATION (cont'd.)
This modification will not decrease the margin of safety as defined in the basis of any technical specification because the performance of RBSS remains the same as before the modification. Offsite radiation doses are kept below those.regulatedLin 10CFR100 by keeping spray solution pH between 8.5 to 11 and 8.5 to 9.5 during the injection and recirculation phases, respectively, as specified in ANS-56.5 and SRP 6.5.2.
.The modification does not violate any license requirements or regulations because the system performance remains within the requirements of 10CFR100 and
.the limits delineated in TMI-1 FSAR Paragraph 1.4.70 (Criterion 70).
1 Compliance with these regulations is assured by keeping spray solution pH between 8.5 and 11 as required by ANS-56.5 and SRP 6.5.2.
The modification does not involve a radiological safety concern since the modified RBSS will perform the iodine removal function under LOCA conditions by keeping the radiation doses at the site boundary below 189 Rom for the thyroid and 7.Il Rem for the whole body.
These are well within the 10CFR100 guidelines (300 Rem for thyroid and 25 Rem whole body dose).
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Fire Detection System in the Intake Screen and Punp Ilouse (B/A 412388, Phase I)
DESCRIPTION OF OING:
B/A 412388, Phase I, installed six (6) ionization (snoke) detectors in eadi of the Intake Screen and Ptmp Ilouse (ISPII) punp roans (i.e., fire zones ISPil-FZ-1 and 2). % e detectors are monitored by a local control panel located in the
. adjacent traveling screen area (i.e., fire zone ISPII-FZ-3). W e local control panel provides local alarm and relay contacts for a~ renote alarm in the 'IMI-l control Room. %e local panel ja equipped with a four (4) hour self contained energency power supply.
SAFITI EVALUATIN:
We ionization detection system installed was manufactured in accordance with the QA requirenents of the USNIC Brandi Tedinical Position (BTP) ASB-9.5-1
The balance of the engineering and installation was performed as Inportant to Safety (ITS). All equipnent'is seismically mounted so as not to create a missile hazard. Installation of the ionization detection system meets a conmit-ment made to the NIC in GPW Report "TMI-l Fire llazards Analysis Report and Appendix R, dated June 28, 1982,Section III.G., Safe Shutdown Evaluation".
'Ihis modification, along with other modifications discussed in the above report will result in the ISPl! being in ampliance with-10 GR 50,. Appendix R.
The probability of occurrence or the consequence of an accident or malfunction of equignent important to safety previously evaluated in.the Safety Analysis Report has not been increased. It is, therefore, concluded that the subject nodification
- does not involve an unreviewed: safety question per the criteria of 10 CPR 50.59.
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OWE !ODIFICATICN:
Emergency Feedwater Differential Pressure Flow thasurement System (412398)
DESCRIPTICN OF OIANGE:
. As a part of the Emergency. Feedwater, System. (EFS;7) _ Upgrade, GPUN ommitted to install by " Restart" safety grade flow devims, which were to indicate in the control rocm the system operating flow conditions.
(Restart Report Section 2.1.1. 7). To meet this acumitment safety grade differential pressure transmitters system utilizing annubars were installed.
SAFETY EVAIARTICN:
The EFWS functions continue to operate as designed and are not degraded by the additions of Emergency Feedwater Differential Pressure Flow bhasure-
, ment System. The plant control performance during normal and abnormal operations remain unchanged.
The changeout of flow sonic transmitters to delta pressure transmitters with associated equipnent and cable routing increase reliability of readout information for operators use.
This nodification does not in any way increase the probability of an accident' or malfunction or decrease the margin of safety.
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Procedures for TDR 400 Guidelines for Plant Operation with Steam Generator Tube Leakage t
PCR. No.
Procedure Procedure Title 1-OS-83-0266 1203-24 Steam Line Rupture Detection System Actuation 1-05-83-0268 1102-11 Plant Cooldown 1-05-83-0269.
1101-1 Plant Limits and Precautions 1-OS-83-0270 1102-10 Plant Shutdown 1-OS-83-0271 1202-12 Excessive Radiation Levels 1-OS-83-0272 1301-1 Shift and Daily Checks 1-0S-83-0274 1106-13 Powdex System 1-EG-83-0051 1303-1.1 Reactor Coolant System Leak Rate
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The procedural changes incorporating the TDR-400 recmnendations provide the necessaryoperatingrestrictions to mntrol radioactive contamination of the secondary plant. These restrictions in'clude crSG Tube leakage determination and nonitoring requirements; disposal as radioactive waste of the condensate
polishing" Dowdex resins; Turbine Building sump monitoring _and contamination limits; Industrial Waste operating restrictions and ad:ninistrative plant shutdown instructions based on OTSG tube leakage.
Incorporation of.these reaminandations along with operator training on the new procedure guidance will control secondary plant contamination and offsite doses to within the acceptable levels developed by GPLNC. Because this - level of contamination is well below NBC guidelines and 10 TR 20 regulations, there is no Environmental or Nuclear Safety imimet.
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Procedures for TDR 406 Steam Generator Tube Rupture Procedure Guidelines PCR No.
Procedure Procedure Title 1-0S-83-0230 H-1-8 T Sat Margin A/f, Low 1-0S-83-0231 1202-2 Station Blackout 1-OS-83-0232 1202 2A Station Blackout with Loss of Both Diesel Generators 1-0S-83-0233 1202-4 Reactor Trip 1-0S-83-0234 1202-6B Small Break LOCA 1-0S-83-0235 1202-6A Loss of Reactor Coolant within Makeup Capability 1-05-83-0236 1202-6C Large Break LOCA i
1-OS-83-0237 1202-14 Loss of Reactor Coolant Flow 1
1-OS-83-0238 1202-36A Loss of Instrument Air-Backup Air Available 1-0S-83-0239 1202-36B Loss of Instrument Air-No Eackup Air Available 1-OS-83-0240 1202-39 Inadequate Core Cooling 1-0S-83-0241 1102-16 RCS Natural Circulation Cooling 1-0S-83-0242-1202-5 OTSG Tube Leak / Rupture 1-OS-83-0270 1102-10 Plant Shutdown The procedural changes incorporating the TDR-406 recommendations provide instructions to the operator for the best method of handling an OTSG tube leak casualty to min'imize off-site dose. These instructions are based on up-to-date industry experience in addition to analytical work on multiple OTSG Tube Failures. These instructions include subcooling margin limits, emergency RCP NPSH limits, OTSG Tube to Shell temperature difference limits, OTSG isolation criteria, and RCP trip criteria based on subcooling margin. These new in-structions along with operator classroom and simulator training will ensure that an OTSG tube rupture casualty is controlled and te;minated to minimize off-site exposure t.
to.within NRC accepted limits. The NRC has endorsed the technical content of TDR-406 with thbir SER concerning our tube rupture emergency procedure.
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' Reactor Ccclant System Cleaning Reactor Coolant system cleaning was initia_ted for the rer: oval of residual sulfur contaminants, to the extent possible, from the surfaces of primary systaa equiprent and piping. Reducing the gaantity of, residual sulfur was necessary to preclude the possibility of reactivation of the corrosion
- nedanism whid had damaged the Steam Generator tubes. Sulfur which was present as deposits in the pressurizer vapor space has been removed by hydrolasing.
The cleaning procedure rapidly converts the insoluble reduced sulfur to an oxidized, soluble fom under protective high pH conditions. The sulfur is then removed using installed plant purificatica system.
7ne selected cleaning process utilizes a 1c xncentration (-20 ppn) of hydrogen peroxide at an elevated pH (8.0) and at a slightly elevated temperature (~130 F). Hydrogen peroxide is norrally forred in slightly lcwer concentrations (5-10 ppa) in the reactor ecolant at every shut-dcwn due to radiation effects on the coolant. Therefore,theuseofthib additive will not adversely affect the normal syste:n materials.
In
-l' addition, many PWPs add more peroxide at shutdcun to quickly solubilize scme o
nuclides and then remove them to avoid interference in the refueling process. Peroxide concentrations of up to the 15-20 ppn level have been used in this process with no adverse effects.
Stress corrosion tests carried out with highly stressed C-rings fabricated frcrn tubing renoved from Tm-1 have shcun no detrimental effects frem the. cleaning process. Ionger tem 1 cop tests with realistically stressed-tubing are also underway.
The TMI-l primary system was chemically cleaned without adverse effects on the prizrary system, the remainder. of the plant or the public. Tne Steam Generators were not required for decay heat remcrial at any tine curing the cleaning procedure. No unanalyzed accident was introduced and the prcbability or consequences of any analysed ac?ident was not -
increased.
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%e revised 'IMI-l Fire Protection Plan was submitted to the NRC on January 1, 1984 by GPUN letter 5211-83-359.
The Plan was updated to reflect danges in organization and fire brigade manning and training. The most significant dange in the Plan (which is also the only deviation frczn the basis of the Fire Protection SER for License Amencinent #44) is in the assignmentiof the Shift Maintenance Foreman (non-licensed) ~ to the brigade leader position. mis reassignment relieved the SIO-licensed Shift Foreman frczn the responsibility of fire fighting.
To empensate for this change and to ensure nuclear safety, a shift foreman or CR0 is required to respond to all fires whi& could affect plant safety.
Being frm of fire fighting duties, nuclear safety is enhanced since with this organization, the SRO is allowed to concentrate on the impact of the fire on the plant and its operation. The ability to control the plant will be demonstrated during fire drills and emergency exercises as outlined in AP 1038, the inplementing procedure for the Plan.
Other danges to the Plan reaffirmed the SER requirments on training sdedules and attendance. The Plan now clearly requires all fire brigade menbers to attend well defined initial training, all quarterly training Iraking up the two-year program, drills, and nonthly meetings.
The revised Plan exceeds the SER requirments in most areas and in addition now defines requirenents for annual firewatch training and annual training of personnel providing a support role to the fire brigade. A final signi-ficant dange is the removal of TMI-2 from this Plan which was previously covered under a comnon site program.
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