ML19249B290
| ML19249B290 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 06/25/1979 |
| From: | Goodwin C PORTLAND GENERAL ELECTRIC CO. |
| To: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| References | |
| NUDOCS 7909040236 | |
| Download: ML19249B290 (25) | |
Text
{{#Wiki_filter:' ,n n =, r;d.Z Q Portland General E!echic Ccmpany ,s n ,1 i %.) A June 25, 1979 Cm 4 Trojan Nuclear Plant Docket 50-344 1.icense NPF-1 [V k Mr. R. II. Engelken, Director .,9
- v.,
h(l ~ ?S Nuclear Regulatory Commission .., h. ), Y I Region V a Suite 202, Walnut Creek Plaza y -/ 1990 h. Ca} ifornia Blvd. p Walnut Cree.k, CA 94596 [' 'DlM
Dear Sir:
Subsequent to the Three Mile Island accident, a series of IE Bulletins were issued by the NRC: IE Bulletin 79-06 dated 4/11/79, IE Bulle-tin 79-06A dated 4/14/79, and Revision 1 to IE Bulletin 79-06A dated 4/18/79. Portland General Electric Compan, had reviewed these Bulletins and responded accordingly. Our original response was submitted on April 24, 1979 and updated on liay 4, 1979. The 30-day response to Eulletin Itea 13 was subnitted on May 18, 1979 A May 22 letter froa A. Schwencer of the NRC to C. Goodwin, Jr. of PGP notified us of a completion of a preliminary review of licensee responses to IE Eulletin 79-06A, including Revision 1, and of a meet-ing to be held in Bethesda, Maryland, on May 30, 1979. At that meeting, it was decided that the draf t SER prepared by the NRC Staf f and any requests for specific information would be transmitted to each licensee prior to fo rnal publication. We have reviewed the working paper transmitted by the NRC Staff, dated June 7, 1979, and the Request for Additional Information. Enclosed is our response to these docuaents, identified by the applicable Bulletin itea number. Sincerely, / C. Goodwin, Jr. / Assistant Vice President Thernal Plant Operation and !!aintenanc e CG/EM/4sb5A7 Attachments p UM,o.A '.; c: Mr. Lynn Frank, Director J State of Oregon Department of Energy <3 q 7909040 /'O 3 D Office of Management In fo rma tion f and Program Control ,1 C3 - l
Bulletin Iten 2 Revise your response after a thorough review of all transient and acci-dent conditions based on insight gained f rom TMI-2 to (a) assure that the action steps specifically warn of potential for voiding, with a description of all instrumentation which night provide indication of potential or actual voiding, (b) specifically address operator actions, based on operational nodes and instrument indications discussed above, for terminating conditions tending to lead to void formation, and (c) provide operators with guidance for enhancing core cooling, given the unexpected condition of actual voiding in the primary systen. Sunnarize the results of this review, including the revisions to pro-cedures. Identify all instrumentation which night be utilized in void recognition and to sunnarize the review results and actions taken with regard to the natural circulation node of operation. PCE Response a. Energency Operating Instruction EI-1 " Loss of Reactor Coolant" was codified as shown in Attachment 1, and the " CAUTION" state-nent in Innediate Ac lon Step 6 specifically warns operators of the potential for coiding as follows: "n. Maintaining pressurizer level alone may not prevent excessive boiling in the RCS and resultant voids that nay compromise the core cooling capability and natural circulation. Keeping pressure within the linits of the pressure tenperature curves of Figure 3.2 of the CROCTRM [ Control Roon Operating Curves and Tables Reference Manual] ensures that saturation is not reached." Figure 3.2 referred to in the " CAUTION" statement has been revised to ensure that 50 degrees sub-cooling is naintained, per the requirements of Eulletin 79-06A (see Attachment 2). Further-are preparing a procedure which provides a backup neans nore, we of indication for some primary Plant parameters for long-tern nonitoring. Attachment 3 is a list of backup neans of indica-tion and will be a part of the new p ocedure. The discussion associated with pressurizer level describes Plcnt parameter behavior under voiding conditions. This procedure, which will be in effect by June 30, 1979, along with Figure 3.2, will be valuable aids to the operator to guide hin in naintaining safe Plant conditions at all times, including adverse conditions.. b. A specific operator action is to ensure that the ECCS systens are naintaining RCS pressure above saturation, per Inmediate Operation Action Step 6 of the attached procedere, EI-l (Attach-nent 1), preventing void formation. T: n r()~3 r, I" "6. Verify the ECCS is keeping the IlCS pressure above saturation. The incore ther:rocouples nay provide better indication of core temperatures than the RTD bypass Icops after the RCP's are turned off. If it is not possible to keep RCS pressure above saturation, see subsequent action for response." A subsequent action mentioned in Step 6 above is shown in EI-l ( At t achment 1). c. Guidance for enhancing core cooling in the unlikely event of actual voiding in the primary system has been provided in the Innediate Operation Action Step 6 of the attached procedure, EI-1. Any nteaa voids would be eliminated by naintaining or relieving pressure above raturation by at least 50 degrees. These measures are to be taken when the potential for voiding exists, and enhance core cooling by providing a flow of as much cold injection water as necessary to maintain pressure above saturation, or as high as possibl.e if the leak is large. The discussion in Attachment 3, under Steam C'nerator Level (Loop h and T ), describes how to ensure that natural circulation T C loop flow is teing naintained. 'h i f ),*i t.- fev t.) t r- - - v_. es.
Bulletin Item 3 to maintain all pressurizer low-level histables in Provide a commitment a tripped condition whenever a saf ety injection signal from coincident pressurizer preseure and level is required to be operable. PGE Response Trojan plant was temporarily nodified by placing the pressurizer The water level bistables in the tripped position on April 25, 1979, while in Mode 1. This action removed the water level coincidence requirement for safety injection actuation, thus providing the temporary protection required. Since that tine, the Plant has remained in 'iode 5 for annual naintenance. Included in the schedule of codifications and repairs for the Plant during this time is a change to pressurizer safety injection logic. This change removes the pressurizer water level coincidence signals and substitutes a two-out-of-three low pressurizer pressure logic to initiate safety injection. These modifications are the subject of License Change Application 54 submitted en June 6, 1979 to the Director of Nuclear Reactor Regulation. This License Change Application will be approved and returns to the necessary codifications will be made before the Plant power operation. After these codifications, the pressurizer water level will no longer be involved in the developnent of a safety injection signal. The pressurizer water level signals will provide control and indication only. Thus the requirement of this Dulletin item is no longer applicable to Trojan. O(}20;p
Bulletin Iten 4 Discuss the procedures that would be necessary in order to allow reactor coolant pumps to be operated unde r con tainment isolation conditions. Discuss whether or not these procedures have been written and incorporated into Plant procedures, or your schedule for doing so. PGE Response As discussed in our original response to Bulletin Item 7.c, in the event of a Safety Injection Signal (SIS) or Containment Isolatioa Signal (CIS), the reactor coolant pump cooling water supplies, Conponent Cocling Water Systen (CCWS), and seal water return lines, Chemical Volume ar. Control System (CVCS), would automatically be isolated. This would preclude continued operation of the reactor coolant pumps. Therefore, new pro-cedures te allow reactor coolant puaps to be operated under containment isolation conditions are not possible at this time. A design review of the scope of the codifications necessary to operate reactor coolant pumps under safety injection and/or containment isolation conditions is underway and is e:.pected to be completed by July 31, 1979. There are eight valves involved in the cooling water and seal water lines for which the logic must be changed to allow them to remain open or be opened after an SIS or CIS. These valves are as follows: CCW Supply to RCP Motors !!O 3296, MO 3294 CCW Return from RCP Motors Mo 3320, FO 3300 CCW Supply to Seal Wat.>r Heat Exchanger MO 3295 CCW Return from Seal Uater Heat Exchanger MO 3319 CVCS Seal Water Return to Volume Control Tank MO 8100, MO 8112 A comprei.ensive safety evaluation still has to be perforced to determine the radiological, seismic, and emergency power ef fects of this modifi-cation. This safety analysis is expected to be comple ted by July 31, 1979. At that time, we plan to provide a description af the design modifications, a description of the necessary operating procedures, and a schedule for implementation of modifications to allow the reactor coolant pumps to continue to operate after SIS and CIS. - wm
Bulletin Iten 7.a your review and revision of training documents has You do indicate that been perforced, but you do not indicate that you have completed the In addition, you do not required review of operating instructions. the Bulletin guidelines are being incorporated, since you indicate that appear to address only reactor vessel integrity and safety injecticn. it is requested that you provide assurance that operating Therefore, procedures and training instructions will be reviewed to ensure that override automatic actions of engineered safety operators will not unless continued operation of engineered safety features will
- features, result in unsafe Plant conditions.
Provide a schedule for completion incor-of the review of operating procedures and training instructions, porating such nodifications as are necessary to comply with iten 7.a of the Bulletin. PCE Response reviewed and the Bulletin guidelines The Operating Instructions were implemented by a codification to Subsequent Action Instuctions in EI-l (Attachment 1). The following " CAUTION" statement was added to EI-1: Do not override automatic actions of engineered " CAUTION: safety features without careful review of plant conditions and only then if continued ESF operation will result in unsafe plant conditions. Do not nake operational decisions based on a single plant parameter or indication when a confirmatory indication is available, for example, pressuriner level without confirming with pressurizer pressure." The basis for this " CAUTION" statement is that " continued ESF operation" may result in unsafe conditions if allowed to continue unchecked in the event of a small leak which can be overcome by our centrifugal charging 200 The shutoff head of the centrifugal charging pumps is about pumps. and a continual popping of these psi higher than safety valve settings, safety valves could be expected to result in valve seat leakage which cannot be isolated. . ?50& O ., _ ~.
Bulletin Item 7.b Your response includes criteria for IIPI combinations which do not comply with those specified in the Bulletin. Therefore, you nust provide assurance that operating procedures will be nodified to keep high-pressure injection and charging pumps in operation in accordance with the criteria specified in Iten 7.h of the Bulletin. Provide a schedule for completion of the review of operating procedures incorporating such todifications a; are necessary to comply with Item 7.b of the Bulletin. PCE Response Plant procedures have been revised to caution the operators not to override the automatic function of Engineered Safety Features (ESP) as ^ dircussed in above response to Bulletin Iten 7.a. At Trojan, the ESF Systems consist of Centrifugal Chargind Pumps (IIPI), Safety Injection Pumps (LPI) and Residual llent Removal Pumps (LPI). The ternination of either liigh Pressure Injection (!!PI) or Low Pressure Injection (LPI) pumps is considered to involve the overriding of the automatic function of ESF. The llPI pumps at Trujan have a shutoff head approximately 200 psig higher than the pressurizer safety valve uct point. In the event of a small break or an inadvertent Safety Injection Signal (SIS), the Reactor Coolant System (RCS) pressure nay reach the safety valve setpoint in less than 20 minutes. Therefore, Plant procedures are being revised to allow the llPI pumps to be shut off when RCS pressure can he naintained above the saturation linit as shown in Figure 3.2 (Attachment 2). The LPI pumps will be kept running for at least 20 ninutes unless continued operation is likely to result in unsafe Plant conditions. If a safety injection signal has been generated and its initiation is known to be inadvertent, and if Plant parameters have reached stable conditions with Reactor Coolant Systen temperature being naintained at least 50 degrees below the saturation teaperature, both IIPl and LPI pumps may be shut off. The Energency Instructions must be naintained free of confusion and ambiguity in order for the operators to take adequate and necessary corrective actions. Although we realize the seriousness and consequences of terminating the llPI and LPI punps prematurely without careful review of Plant conditions, the IE Bulletin guidance for allowing the pumps to operate for more than 20 ninutes or until unsafe Plant conditions exist is subject to different interpretations by different operators and inspectors. Our revisions to the Emergency Instructions regarding operation and overriding of IIPI and LPI pumps have attempted to take this concern into consideration. OOI.35NAC ...c s.
Bulletin Item 7.c Your response stated that a design review of the necessary modifica-tions required to operate reactor coolant pumps (RCP) under safety injection conditions would be complete as of July 31, 1979. At that tine, we require a schedule for imple~tentation of any design nodifi-cations required to keep the RCP's operating. In addition, we will require a schedule for completion of the review of operating procedures incorporating such r'odifications as are necessary to comply with Item 7.c of the Bulletin. PGE Response Please refer to PCE's response to Bulletin Item 4 which covers both Bulletin Iteme 4 and 7.c. <.. ~,.,v' g j,n,_
- / L' A.*
_p
Bulletin Iten 7.d Identify those specific paraaeters other than pressurizer level identi-fled for operator use in evaluating Flant conditions, and verify that these paraneters have been included in appropriate operating procedures. PCE Fesponse Specific parameters for evaluatiag Plant conditions are identified under " Sy mp t om s *' at the beginning of each applicable caergency procedure, as is the case in the attached EI-1. The bodies of our procedures contain discussions in nore detail of how to use these parameters as a troubl2-shooting guide once the necessary immediate actions have been carried out. The ionediate actions ensure that the ECCS subsystems are func-tiening properly and contain the precautions necessary to prevent voiding. Please refer to the attached EI-l for identification of those specific parameters other than pressurizer water level used by the operators in evaluating Plant conditions. . c., ; >,r).~ n J v A.4 .F. -. v,;<, n w.
Bulletin Item 8 a summary of the results of the reviews of alignment Please submit requirements and procedures controlling manipulation of safety-related valves and any revisions necessary within 2 weeks after completion of Determine whether the Technical Specifications the review. (LTM-If not, add the follow-require reriodic surveillance of locked valves. ing request for information.) that Also review Plant procedures and revise them as necessary to ensure locked safety-related valves are subjected to periodic surveillance. sunmary of the results of the review. Subnit a PGE Response All ESF flow paths were identified on the Plant Piping and Instrument Drawings. All valves in these flow paths were checked for correct posi-tioning requirement. Eacii valve was checked against Plant administrative Plant control procedNes (Plant operating tests, locked valve lists, engineering iests, and valve line-upc) for proper positioning and controls. affect ESF Further, valves not directly in the ESF flow path that might allow bypass performnce were similarly checked, i.e., valves which night ESF systems that were checked fics or dilution of borated water sources. were: Containment Isolation System FSAR Fig. 6.2 6.2-47 6.3-1 Safety Injection System 6.3-1 Centrifugal Charging Pumps 6.3-1 Eoron Injection Systen 5.5-7 Residual Heat Removal System 6.4-1 Containment Spray System 6.2-48 Ilydrogen Vent System 6.2-48 Hydrogen Saupling System 9.2-4 Component Cooling Water System 9.2-1 Service Water System 9.5-3 Emergency Diesel Cenerators 9.2-4 Containment Coolers All valve position requirements were found to be correct. Valves i:. vital flow paths are either locked in position or verified to be in proper position monthly per Technical Specification requirements. Two valves were added to the locked valve list. These were: Valve 8735 - RllR Common Discharge to RWST (FSAR Fig. 5.5-7) Valve MD050 - Auxiliary Feedwater Pumps Suction from Condensate Storage Tank (FSAR Fig. 10.4-2). <. e rc e n .j G ' s v..
Control of locked valves is described in Administrative Order A0-3-13. Locked valves.ust be controlled in Modes 1, 2, 3 6 4. Position changes of locked valvet must be approved by the Shift Su pe rv i so r. All posi-tion chanf;es are raintained on a special list by the Control Operator (the locked valve list). tih e n the Plant is in Modes 5 and 6, permissio1 of the Shift Supervisor cust still be obtained to reposition a locked valve. Ilowe ve r, the locked valve lict does not have to be naintained in Modes 5 and 6. Prior to Plant heatup to Mode 4, the locked valve list is required to be updated. O - .> U.<, f ;a s.. s 3.yyy- '
hulletin Iten 10.c Please identify the Icvel of authority required for renoving and return-ing systeas to service, and describe the ne. hod used for *ransferring infornation about the status of safety-related systens of shift change. PCE Response Tn general, the Shift Supa visor is the level of authority required for removing and returning systens to service. Removal and return of safety-related equipment from/to service for periodic surveillance in accordance with test procedures reviewed and approved by the Plant Review o >ard require authorization by the Control Operator. For the removal of equipment from rcrvice for caintenance, the Shift Supervisor's approval with the Operations Supervisor's concurrence nust be received under the following circunstances: a. When renoval iron service is not by PRB's previously approved procedure, and b. When operability is required in the node that the Plant is in or nay be in by the tine the equipment is returned to service. Authorization for removal of safety-related equipment f rom service for naintenance is documented on outage worksheets. At shift change, the oncoming Control Operator and Shift Supervisor are inf onned of tests in progress, and ou_ age worksheets are reviewed at the beginning of each shift. cv*.,n; r-s.t, - -. y_.wv.
Bulletin Iten 12_ Confirm your review of operating nodes and procedures for removal of hydrogen from ihe Containment to assure that such removal can he accom-plished. Briefly discuss various methods for dealing with hydrogen in for completing procedures the Containment and provide your schedul incorporating these nothods. PCE Response We have reviewed codes and procedures for renoval of hydrogen from the Containment. We have revised applicable procedures in the nanner shown li. This in Subsequent Action Step 6 of the attached El-1 (Attachment recognizes the insight we have gained based on TMI-2 that hy ,an can be _..nce s. If e v o l '. e c earlier in post-1.CCA condition, under certain circur these circumstances have existed, hydrogen removal is commenced with the equipment which is already installed and operable during power operations. r the various _thods for dealing with hydrogen in A detailed discussion o Containment (sampling, nixing, recombiners, and venting) is provided in FSAR Section 6.2.5. The Trojan hydrogen reconbiners are tested monthly per Technical Specificatious to ve-ify operability. KM/sb/4sh6A6 c.. ,i ? s)yJ nr dise. -, n.. m.
. t. t a c une n t PORTLAND Gt!NERAL ELECTRIC CO?tPANY TROJAN NUCLEAR PLANT April 30, 1979 "" i " i "6 SAFETY-RELATED EMERGENCY INSTRUCTION EI-1 LOSS OF REACTOR COOLANT APPROVED BY L DATE 30 79 / / ~ A. SYhP TG>lS Listed below are the symptoms which may indicate a large leak in the reactor coolant system which will result in a loss of reactor coolar.t: 1. Pressurizer low pressure. 2. Pressurizer low level. 3. High containment pressure. 4. liigh containment humidity. 5. liigh containment recirculation sump level. 6. liigh containment radiation alarm. B. .AUTC5fATIC ACTIONS 1. Reactor trip. 2. Turbine trip. 3. Safety injection is initiated. 4. Containment spray may be initiated. C. DDIEDIATE OPERATOR ACTIONS 1. Verify reactor trip, turbine trip, and safety inj ection has occurred. CAUTION: If pressurizer pressure drops to 1765 psig and there is no automatic safety inj ection, n. - 11y start safety injection. 2. STOP reactor coolant pumps. 3. Verify all engineered safeguards valves and equipment are aligned and operating with status lamp panel. 4. 3rify safety injection flow when pressure is below pump's shutoff head. EI-1 Page 1 of 6 0;'el.'.T)l[ Revision 6 ~ v.m
5. Verify spray initiated if the containment pressure reaches the high-high set point. 6. Verify the ECCS is keeping the RCS pressure above saturation. The incore thermocouples may provide better indication of core temperatures than the RTD bypass loops after the RCP's are turned off. If it is not pos-sible to keep RCS pressure above saturation, see subsequent action for response. CAUTION: A. If it becomes necessary to reset containment isolation the attached list nust be verified before resetting. B. Maintaining pressurizer level alone r ay not prevent ex-cessive boiling in the RCS and resultant voids that nay connronise the core cooling capability and natural circulation. Keeping pressure within the limits of the pressure temperature curves of figure 3.2 of the CRCCTRM ensures saturation is not reached. 7. If both RHR pumps are running, manually isolate trains by closing RHR cross-connect valves MO-S716A/B. D. SUBSEQUENT OPERATOR ACTIONS 1. If there is an increasing pressuriner relief tank level, pressure and/or tecperature, along with a high relief line tecperature and the pressuriner pressure is below 1765 psig, isolate the POR's to see if one is stuck open. 2. If pressuriner pressure and/or level are decreasing and Tave is remaining constant, a loss of coolant accident is indicated. It may further be distinguished from a loss of secondary coolant or S/G tube rupture as follows: a. An increase in containment presnure, a containment high radiation alarm, and rising sump water level indicates a loss of coolant accident. b. An increasing pressurizer relief tank level, pressure, and/or temperature with possibly a high relief line temperature after both POR's are isolated indicates a loss of coolant accident due to a stuck open safety valve. c. A condenser air renoval equipment radiation alarn or a steam generator blowdown radiation alara indicates a steam generator tube rupture, d. Abnormally low pressure in one or more stean generators, coincident with low pressurizer pressure and level, and decreasing Tave indicate a nain steam line break or feed line break. EI-1 Page 2 of 6 Revision 6 .n o., s 73 s> U m.;J u g; ,.w
CAUTION: Do not override automatic actions of engineered safety features without careful review of plant conditions and only then if continued ESF,peration will result in unsafe plant conditions. Do not make operational decisions based on a single plant paranater or indication when a confirmatory indication is available, for example, pressurizer level without confirn-ing with pressurizer pressure. 3. If it is determined by the above descriptions that the accident is a loss of reactor coolant, proceed to step 4. If the accident is not a loss of reactor coolant, proceed to the appropriate Emergency Instruction. 4. If plant conditions require a planned evolution, i.e. stopping unneeded RHR puups, the safety injection signal may be reset. CAUT10h: In the event of a loss of off-site power following nanual blocking an automatic SI, the only loads that will re-sequence onto the diesel generator are those initiated by the shut-down sequencer. All other ESF loads required to be in operation as a result of the initial safety injection, must be nanually re-started by the operator. 5. Implement the Energency Plan. 6. If the RCS has spent a period of time below saturatio: or RCS samples show cladding damage or a buildup of hydrogen gas in the RCS, start the containment hydrogen recombiners and nixing fans and periodically vent the pressurizer to the pressurizer relief tank. If WGDT's are full, it nay be necessary to allow the PRT rupture disc to blow, venting gases to the containment. If RCP's are availabic, naxinize pressurizer sprays to aid in degassing RCS and dissolving any void.s which may now exist in the vessel head area. 7. When the RWST LO LEVEL annunciator actuates, start aligning the safety injection systen to take suction from the containment recirculation surp as follows: NOTE: Ensure the residual heat renoval (RHR) pumps tripped automatically on RNST LO LEVEL signal, a. Open RHR heat exchanger (Hx) component cooling water (CCW) inlet valves MO-3210A and MO-3210B. b. Close RHR pump suction valves MO-8700A, MO-87003 and MO-SS12 from the RWST. c. Open RHR pun.p suction valves MO-3811A and MO-SS11B from recir-culation sump. NOTE: These valves are interlocked such that MO-8700A/B nust be closed before MO-8711A/B can be opened. d. Verify the RHR Rx outlet cross-connect valves MO-8716A and MO-S716B are closed to provide train separation if both RHR pumps are operable. If both pumps are not operable leave the valves open. EI-1 Page 3 of 6 Revision 6 cc. e u das,; u u,3 t w.ev
c. START west /e:.st RHR pumps A/B. f. Flow to the vessel through the two cold leg injection lines can be checked by using west RIIR lix "A" outlet flow FI-9 71A and FI-9 71B, and cast RllR Hx "B" outlet flow FI-970A and FI-970B. g. Close safety injection pump miniflow line block valves MO-8833 and MO-8814. h. Open RilR pump discharge valve isolation valve MO-8804B to the safety injection pump suction. NOTE: Valve MO-SSO4B is interlocked such that the reactor coolant systen to RHR systen isolation valves MO-8701 or MO-8702 nust be closed, safety injection pump miniflow block valves MO-5813 or ':0-5814 must be closed, and recirculation sunp isolation valve MO-S811B nust be open before MO-8804B can be opened. i. Open RIIR pump discharge isolation valve MO-8804A to the charging pump suction. NOTE: Valve MO-SS0lA is interlocked such that the reactor coolant systen to RHR systen isolation valves ?!O-8701 or MO-8702 cust be closed, safety injection pump miniflow block valves MO-8S13 or MO-8814 cust be closed, and recirculation sump isolation valve "0-8811A nust be open before MO-SSO4A can be opened, j. Verify that east RHR pump "B" is supplying the safety injection pumps (increased safety inj ection pumps discharge pressure, PI-919, PI-923). k. Open RHR discharge to safety injection pump suction valves MO-SS07A and MO-8807B. 1. Close safety inj ection suction valve MO-8S06 from RWST. Close charging pump suction valves MO-112D and MO-112E fron RWST. a. 8. Shift the suction on both spray pumps one at a time as follows: a. STOP west / cast containnent spray pump A/B. b. Close containrent spray pump suction valve h0-2050 A/B from the RWST. c. Open containment spray punp suction valve MO-2052 A/B-from the recirculation sunp. d. START west / cast containment spray pump A/B. e. Repeat steps 1 through 5 and shift the other pump. EI-1 Page 4 of 6 Revision 6 0 050 - o m -s.
f. When the Na0!! TANK LO LO LEVEL annunciator actuates, close spray additive valves F10-2056A and 2056B to prevent air binding of the spray pumps. 9. If the hydrogen recombiners and mixing fans have not already been started per 6 above, start them now. 10. Approximately 17 hours after the accident, depending on the recirculation sump boron concentration, align the safety injection system for hot leg / cold Icg recirculation as follows: a. Open RHR to RCS hot legs isolation valve MO-8703. b. Open cross tie isolation valve M0-8716A. Close RiiR Cald Lag Inj ection valves MO-SSO9A/B. c. d. Verify hot leg recirculation flow on FI-600, c. Open hot leg isol:1 tion valve MO-SSO2A. f. Verify flow to the reactor coolant system through the hot leg header on FI-913. g. Open hot leg isolation valve MO-8802il, h. Verify flow to the reactor coolant system through the hot leg header on FI-022, i. Close Cold Leg Safety injection valves MO-8821 A/B and ?t0-8835. 11. Sample the recirculated coolant to determine boren concentration as follows: a. Every 15 minutes for the first hour. b. Every hour for the next 3 hours. c. Every 4 hours after the first 4 hours. d. Maintain boron concentration greater than 2,000 ppm B. e. Use emergency borate mode to increase the boren concentration, as required. EJC El-1 Page 5 of 6 Revision 6 s> U na,* p a
- . 37e O_
vvn
EMERGENCY INSTRUCTION EI-1 LOSS OF REACTOR COOLANT VALVE VERIFICATION BEFORE RESETTING CIS Before resetting Containment Isolation, you must verify the following valves are in the indicated positions: Valve Description Panel Position Verification MO-4180 Containment Sump Discharge C19 Pull to Lock CV-4181 Containment Sump Discharge C19 Pull to Lock CV-5652 Accum Sample Isol C17 Auto After Close CV-5661 Reactor Coolant Drain Tk Sample C17 Auto After Close CV-4000 Reactor Coolant Drain Tk N2 Supply C17 Auto After Close CV-4006 Reactor Coolant Drain Tk Outlet C17 Auto After Close CV-4301 Gas Collection Header Valve C17 Auto After Close CV-4471 Instrument Air to Containment C17 Auto After Close CV-4470 Service Air to Containment C17 Auto After Close CV-10001 Containment Purge Supply C17 Auto After Close CV-10004 Containment Exhaust C17 Auto After Close MO-10002 Containment Purge Isol C17 Auto After Close MO-10003 Containment Exhaust Isol C17 Auto After Close CV-10014 Chilled Water Return C17 Auto After Close CV-10015 Chilled Water Supply C17 Auto After Close MO-2310 A Steam Generator Blowdown Isol C15 Auto After Close MO-2813 B Steam Generator Blowdown Isol C15 Auto After Close MO-2812 C Steam Generator Blowdown Isol C15 Auto After Close MO-2808 D Steam Generator Blowdown Isol C15 Auto After Close CV-2811 A Steam Generator Blowdown Sample C15 Auto After Close CV-28SO B Steam Generator Blowdown Sample C15 Auto After Close CV-2814 C Stean Generator Blowdown Sample C15 Auto After Close CV-2809 D Steam Generator Blowdown Sample C15 Auto After Close EI-1 Page 6 of 6 Revision 6 bd1[5 $C,2,. I (1 I I Ei >s o C U .l. _--..-_.,-----..{O .l-
- O p
b c .;i...l.! [ .j t. 21 4 o c,7 o q-34 '.. j.
- N mU l.
- ' l.
I R 0
- d
...... M. __ - - H . p; E .:j t. p uu l l .. F. = "a.,u ..,i. ,t. uH. = y gg m lHuJ N L U I l z oa .I. [j.f'E 1 O O 3 .l. .i
- . t. : *L:.'
I. .l. I cn tn [.. ' j .I I.i.. C-* Ei e l- .1 l g: ....-t.l --I--. 1 .r-- '.l.----~ rd .W O
- g...t' i
'O e4 .l ! '.,... : : f. ' '. l': .-' ; ': l u' _,t. g to .) to O .'i. g l' .l-O +H @ Q j' ..,j f O L
- t. a b
{. 'j N' - b'- i r ~" o 't--'*- . ] 'i '. t t t. ]!: ' .H . f,., . b .j .. g;;. -. ! ;... l: CO . _d ! '. . !..I .; q .C u q g .,.4 m g- .. ~ l, I_: [- Cf: . '.' ; (.
- ...... _p} _..a_'
- ")' : -
-,l. O'g H_-_.-- -4 CO . - ff .I F. - O u O C -l ug.7 g*. l m - - - _.., g 7 4, .y -{-- .l. 9_ 1. a,. l.. - w ucus 3 o ej 4 f [*; gQ U li-p m- [-4 --9' 7 f. I I
- .._ ;.. _.,. [. f.,
.! W. H .! N H,.
- 7...
.._.l'_.. .f.
- g. _ _; _... _l.-_.
jLll:,H f. .__.pi.. -. _ Ql,.. y_ q r f. ~ N .____.4 p i_._ 4 _ _.E H
- t:
'!..!...l !; (/ N W l g y,?
- 1.
j e ..-o - fd - N H .4.. N ..my. .g .p
- r:
-1 : L oZ Z ..;J.! m 2" *... o o [ o M mN- .a i ^
- p. J. _
.n:1. 'l Cf a uN- ' K : 'l 4 O! I j M H: 1. H [ C.',. a -.t::'-l.-.t'
- 1:.
I'- p. Zf 'l ca H . 1 y cr. j i --- T7 pr -1 : - - - - -j:1 1 y { - g-g f: ! [-4 t--* - l-
- l,
.l-t A N i - t-~ I- -~~ r q. ; { : .-.\\ [l h p w !L-' 2._. ._f' .H.. s H N . r: - w ---&x -Y i . t- ._ t .c i + F-- r-I rd N
- -l; M
_'J-4- I .s. ..I: ' [: i D'o, .k p M '.]_ _U,__.__. g pa % m j _1._ b . h _ /C... % N .d.. : L .l 1
- =.
N g{e ..' 2. l _ g y ........d _. 8 g, ..g .I a 4 ._.,.. g j ;. -...J.._' 'I:::..'. " : ~ l :: :: o M
- t f
{
- j-A
__.-_g-- 'l. a,._;_tn _' '..[.. 'l-O U C9 f4 y g ..... -
- q4 f.
H
- t :
cd - l ,4 g -'~*"tr~! i m. ] [. j' N9._._'_'.'.. . f. '.._ _ _ '.. _ l
- }.
..l. p. .. _ 1. _ _ _ _ ' : j. ' .p ...a._. M m _...;.g% g-~ -N I - ll p ...:1 ' 7. f4 H g N Z j' .[; ; ;t - _.I C C: H ..; : s w I:.l. m d_;m V i. -t y ..N. d ,_q m L__ _ 'I.; :h 'l [h ~ j: $.:l:: ^b 1 O t-fu O . D.+ C -g-. l _ __ _. _ _.. ! ! : : i ',.. ". U i j 'f ! s{ ' i:. _y 1s .t.. .AN cd . I-L in f l' CM OH -'i
- p n g :-
.-f.. j- : -+--; .{:, }: Z ho I: . l. <- 4 m {-.---------l--------3.
L--
- -. y -- : g ---. - -[> -t-
- I i.
o 'G-O ,- M - '0 ___J.___....L. N t .l. / ,N N I l t/ M _ ' - l t-< . L. c' 1.1 - M; .H l N f: n; M .t l ' m '}: W. .. D: ,?,*. l
- ^....[ : : r-{ n.
-;* m fr' At .r t.: pH' ' j.: n,. ..._;'.I.m W ..H' 3-- 4{. M i -l :,
- ).
.w_..a._.. .h... ..;_. a. b .-_..._..?"._:... -.1 2....l.: - O. .l..
- .l P-- C I.
i .l..l N r { ' - [' : gH'- r-Z "t*: La1-n_____ re ~ - .i. L_u. a . __n_ ;.._2 .- :P i - ;- - = -er a i . i t .l ( W,_. qs c4 I, 'I: ;'I t.' : ~.:j, ^ N I, l '(:' m i ( Q g y '[ - I 'I 1' ':(I .I : i:
- ; ', l g'd*
t b h j I-g ..T--'- - ~ ~ ~
- --L-*
't! 'l*' l O ...{ 1 w Lo l. -4 c a. . : t :: ~. L. t .,...'. f. _. -< t2 i 't. cr y .... I _. i .-.t.. 3,. p i : + i j-r: g...., o y .l- ~l..u 4 l
- ,p :
. [: n o I t _ se 4ameuowneserweew.uawereir%tw.nesma.w =wrususcum aara :memin 4.Li o o o o o o o o o' m~ o o o o o o o D D = o e N w o m m a s a OD (3 sd) Rev.tston 26 I n'O"i l f 3HflSS3&l S3M Qril 1979 v .( O Figure 3.2 w &; wb e s.p.m.
Page 1 of 4 _ATTACIL'IENT 3 TROJAN NUCLEAR PIET POST-ACCIDENT BACKUP MONITORING CAPABILITY Nornal Instrument Backup Instrument Which nay be Lost Outside Containment Use of Backup Instrument (A) Pressurizer (1) Differential Read directly from remote Water Level Pressure indication. PGE is pre-Measured Outside sently in the process of Containment adding a differential pres-sure transmitter utilizing existing pressurizer sample lines located outside Containment. If operation of the instrumentation is tested and found successful, specific test procedures (and operating procedures required in the evert of an accident requiring long-term monitoring) can be made available before commercial operation is resumed, but are not available at this date (6/15/79). (2) Pressurizer By use of pressurizer temp-Steam Tempera-erature indication from ture and Water existing pressurizer RTDs, Temperature RTDs a comparison of pressurizer steam and water temperatures can be made by the operator. If the water temperature is approaching the steam temp-erature, the pressurizer may be empty. Ilowe ve r, if the steam temperature is approaching the water temp-erature, then the pres-surizer is going solid. [ NOTE: The existing pres-surizer RTDs contr.in teflon-insulated lead wires and may not be satisfactorily nois-ture tight for the post-LOCA environnent. The opera-bility of these RTDs in such an environment will be enhanced by a modification W )1%y p 4 m.J L e s t 2 _, -, w c= ~n;
Page 2 of 4 ATTACIDIENT 3 Normal Instrument Backup Instrument Outside Containment [a] Use of Backup Instrument Which may be Lost to the RTD wiring and encap-sulation of the terminals with a high temperature sealing compound. This work will be completed prior to Plant operation.] (3) RCS Pressure and Read RCS pressure directly. ^ Charging Pump The Reactor Coolant System Flow is going solid if RCS pres-sure is rapidly increasing at the same time that the charging flow decreases. (4) RCP Ammeter Read ammeter directly. A fluctuating or low reactor coolant pump ammeter reading indicates approach of two-phase (steam / water mixture) flow. (5) I;i. clear Read directly. An int _eas-lustrumentation ing count rate indicates a decrease in shielding as water level approaches core. (B) Steam Generator (1) Dif ferential Read directly from remote Water Level Pressure indication. PGE is pre-Measured Outside sently in the process of Containment adding signal transmitters for each steam generator utilizing the existing steam generator blowdown lines and main steam lines located outside Containment. If operation of the instru-centation is tested and found successful, specific test procedures (and oper-ating procedures required in the event of an accident requiring long-term raonitor-ing) can be made available before commercial operation is resumed, but are not available at this date (6/15/79). cu }., m) \\. ?)q,. 7 ? gg G
- mmvg
Page 3 of 4 _ATTACliME';r 3 Nornal Instrument Backup lustrument Containmentf"} Use of Backup Instrument Which may be Lost Outside (2) Steam Generator Read pressure directly. Pressure (a) The atmospheric relief valve can be periodi-cally opened. If the steam generator is nearly dry, it will rapidly depressurize. (b) While using the con-denser steam dump valves, the operator can listen for flow through the dump valve and con-firm that the steam generator pressure is stable near the no load value. Safe shutdown conditions (3) RCS Loop Thot and Tcold should show an RCS loop AT (i.e., Thot -Tcold) less than full load AT. Steam generator level has predict-able effects on RCS loop AT characteristics. (4) Fill Steam This method would be con-Generator Solid sidered a last resort. Indication is (a) water flow in steam lines and (b) pos-sible discharge of two phase water through atmospheric dump valves. (C) Reactor Coolant (1) Centrifugal Charging pump discharge System Pressure Charging Pump pressure is available from (Illgh Pressure) Panel C12 in the control Discharge room and from the Plant Pressure computer. (2) Centrifugal Charging pump flow is avail-Charging Pump able at Panel C12 in the Flow control room and from the Plant computer. With the charging pump running, the pump curve can be examined to estimate discharge pressure. n...,n. a) l*
- .n.>;.p
Page 4 of 4 ATTA' '4ENT _3 Normal Instrument backup Instrument Outside Containment [a] Use of Backup Instrument Which nay La Lost (3) Safety Injection Safety injection pump dis-Pump (Intermedi-charge pressure is available ate Pressure) at Panel C19 in the control Discharge room. This pressure is only Pressure accurate after the flow starts. If there is no safety injection flow, this indicates that the RCS pres-sure is greater than the safety injection pump dis-charge pressure. (4) RilR Pump (Low RllR pump discharge pressure Pressure) Dis-is available at Panels C12 charge Pressurc and C13 in the control room and from the Plant computer. This pressure is only accur-ate after flow starts. If there is no flow, this Indi-cates that the RCS pressure is greater than the RilR pump discharge pressure. (5) Pressure Indica-Manipulate sample valves as tors in Various necessary and read locally RCS Sample Lines at the sample stations. (D) RCS Loops Steam Generator Calculate RCS temperature by Pressure use of steam table.
- Tcold, Thot>
Incore Thernocouples [a] Listed in order of preference. b N.'. E.'[' KM/4kk3A5 .nn yr- , ~,. c r.:}}