TMI-85-084, Forwards Addl Info Re Procedures Generation Package,Per Bj Youngblood 850830 Request for Info on Functional & Task Analysis

Forwards Addl Info Re Procedures Generation Package,Per Bj Youngblood 850830 Request for Info on Functional & Task Analysis

ML20198C348
Person / Time
Site:	Wolf Creek
Issue date:	11/04/1985
From:	Koester G KANSAS GAS & ELECTRIC CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
CON-NRC-TMI-85-084, CON-NRC-TMI-85-84 TAC-57368, NUDOCS 8511120061
Download: ML20198C348 (14)

Text

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KANSAS GAS AND ELECTRIC COMPANY THE ELECTFMC COMPANY GLENN L NOESTER vocs pngstossev.asucigae Novenber 4, 1985 Mr. Harold Denton, Director Office of mclear Reactor Regulation U.S. lAlclear Regulatory Conmission mshington, D.C. 20555 KMLNRC 85-245 RE: Docket No. SfN 50-482 SUBJ: Response to Request for Additional Information on the Functional and Task Analysis REF: NRC Istter Dated August 30, 1985 from BJYoungblood to G Moester, KG&E

Dear Mr. Denton:

The referenced letter requested additional information be provided concerning the Wolf Creek Procedures Generation Package. The enclosure contains the responses to these requests.

Yours Very Truly, Glenn L. Voester Vice President - Iliclear Gm:see Enclosure xc: PO'Connor (2)

JCunmins h*

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1 1

201 N. Market -Wictita, Kansas - Mail Address: PO. Box 208 I Wichita, Kansas 67201 - Te'ephone: Area Code (316) 261-6451

Enclosura to page 1 Novenber 4,1985 KMLNRC 85-245 RESPONSES 1D REQlESP FOR AIDITIONAL INFORMATION REGARDING THE NOIF CREEK GE1ERATING SfATION FUNCTIONAL A}D TASK ANALYSIS A technical deviation as defined in letter dated February 29, 1984 to G.L. Koester, Kansas Gas and Electric Conpany, from B.J. Youngblood, NBC, is any modification, addition, or deletion of guideline actions necessary because of plant design differences, equignent operating charactedstics, or plant operating philosophy.

Nblf Creek Generating Station, (NCGS) Emergency Procedure Generation Package (PGP) , ADM 01-052, is the administrative procedure for the development and revision of the Emergency Operating Procedures (EMGs) . The review process conducted to identify potentially safety significant deviations from the Westinghouse Owners Group (NOG)

Emergency Response Guidelines (ERG) (as described in our response to Request 3) concluded that there are no safety-significant deviations.

Therefore, the plant-specific technical guidelines portion of administrative procedure ADM 01-052 need not be revised as asked in Requests 1, 1, and 3.

RE00 ESP 1 Describe and justify, as part of the plant-specific technical guidelines portion of the PGP, Findings 1, 6, 8, 9, and 10 contained in the Task Analysis Final Report and clarification letter dated May 24, 1985, as potentially safety-significant deviations from the generic technical guidelines.

RESPONSE

Findina 1 BCA 0.0, step 23 requires BIT tenperature indication in the control room. None is provided.

JUSPIFICATION Per ERG background document, monitoring BIT tenperature for solubility limitations is only a concern for systems having Boron concerntrations greater than 7000 ppm. The reduction of BIT Boron concentration from 21,000 ppm at the reference plant to 2000 ppm at NCGS is addressed in SINRC 84-0070 dated April 17, 1984. It is also reflected in FSAR Sections 6.3.2.2 and 15.3.

ERG ECA-0.0, Loss of all AC Power, Step 23a which required the operator to check BIT tenperature was deleted in EMG C-0 due to the reduced Boron concentration.

CONCLUSION BIT Boron solubility is not a concern at NCGS, therefore, the need to monitor BIT tenperature and the need for control room indication is not necessary. It does not constitute a safety-significant deviation.

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EnclosurQ to page 2 Ncr.wier 4,1985 10DBC 85-245 Findina 6 KC4E procedure E-1, step 13A requires operator action at greater than 535 GPM. Control room indicators are graduated in increments of 100 GPM. Therefore, this value of 535 GPM cannot be read accurately.

JUS 1'IFICATION EMG E-1, Rev. O used RIR punp recirculation valve automatic closure at 535 GPM to be indication of .RIR flow to RCS. The procedure was 4

charged to use a calculated value of 500 GPM (which includes instrument errors) as read on EJ FI-618 or EJ FI-619. This change results in positive indication of mininum RIR injection flow to the RCS.

The Westinghouse Owners Group (NOG) Emergency Response Guidelines (ERG) Procedure ERG E-1 step 13A directs operator action after verifying RIR injection into the RCS. NCGS Emergency (ENG) procedure EMG E-1, step 13A Revision 1 requires operator action after verifying RIR pung flow to the RCS to be greater than 500 gpn, which indicates flow to RCS. The operator action is the same so there is no deviation from NOG ERG guideline actions.

CONCLUSION This remains within the NOG ERG and is not a deviation.

i Findina 8 The ERG background h=ntation for FR-C.1, Step C-lb, lists CCW to RIR heat exchanger flow as an instrumentation requirement. No instrument for this exists in the control room.

JUSTIFICATION I The NOG ERG FR-C.1 step C-lb cautions the operator to verify that the RIR pungs are not operated longer than a specified time without CCW flow to RIE heat exchanger to prevent punp damage. NCGS EMG FR-Cl step C-lb contains the same caution. Acceptable alternatives exist for the indication of CCW flow. Control room annunciators 51A and 53A alert the operator to III/ID CCW flow conditions. RIR inlet / outlet tenperature indication across the heat exchangers is an acceptable indication of CCW flow to the heat exchangers, and is available in the control room. In addition, OCW to RIR heat exchanger flow indication is available locally and on the BOP CRT located in the control room.

Since the reactor operator has adequate OCW flow information available, the actions of EMG FR-Cl step C-lb remain within the NOG

. ERGS.

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Enclosura to page 3 Novenber 4, 1985 4

100mC 85-245 OCW to RHR_ heat exchanger flow indication is classified as backup plant instrumentation per the Instrumentation Section of the Generic Issues portion of the Executive Volume. Backup plant instrumentation, as defined in Generic Issues, is not required to meet the stringent design, qualification and display requirements of key plant instrumentation. For exanple, backup instrumentation is not required

to be redundant, powered from a highly reliable source, and is not i

needed to be either accessible on demand or recorded. Therefore, the instrumentation used in the ENG to verify OCW flow to the RHR heat exchanger meets the ERG criteria and is not an instrument and control deviation.

C0fCLUSION 1

Since OCW flow informaton is available to the operator and the actions and instrumentation remain within the NOG ERGS, this is not a deviation.

. Findina 9 ERG background documents for eight of the ERGB list CCW Flow to Seal Water Heat Exchanger as an information

. requirement. No instrumentation for this information is provided in the control room.

JUSTIFICATION The eight EMGs that were referenced in the finding are listed below with the NOG equivalent procedure and step cross-referenced.

1. EMG FR-I.1, Response to Pressurizer Flooding, Step 2 & 4 (ERG FR-I.1, Step 2)

2. EMG E-3, Steam Generator Tube Rupture, Step 34 (ERG E-3, Step 34)

3. EMG ES-ll, Post-IOCA Cooldown and Depressurization, Step 26 (ERG ES-1.2, Step 26) f

4. EMG ES-03, SI Termination, Step 16 (ERG ES-1.1, Step 16)

5. ENG C-21 Uncontrolled Depressurization of ALL Steam Generators, Step 27 (ERG ECA-2.1, Step 27)

6. . ENG C-31, SGTR with Ioss of Reactor Coolant-Subcooled Recovery Desired, Step (ERG ECA-3.1, Step 31)

7. ENG C-32, SGTR with Ioss of Reactor Coolant-Saturated Reovery Desired, Step 25 (ERG ECA-3.2, Step 25)

8. EMG C-33, SGTR Without Pressurizer Pressure Control, Step 18 (ERG ECA-3.3, Step 18) '

OCW flow to Seal Water Heat Exchanger is also classifed as backup plant instrumentation. Acceptable alternatives exist in the control l room for indication of proper CCW flow. A check of service loop flow on EG-FI-55A (of which the seal return heat exchanger is a part), OCW to and from service loop valve positions, and the absence of

Enclosura to page 4 Novenber 4, 1985-10DEC 85-245 annunciator 54F "CCW Seal HX Flow HIIO" are sufficient to assure that proper CCW flow thru the seal return heat exchanger exists.

CONCLUSION The ERG criteria has been met and this finding is not a deviation.

Findirx 10 The background h=arit for ERG E-3, step 2, lists steamline radiation monitors as an instrument requirement. None is provided in the SNUPPS control room.

JUSTIFICATION The intent of the Emergency Response GL11delines is to utilize steamline monitors as one possible means to. identify which steam generator (s) have ruptured tubes (this is the " purpose" for E-3, step 2). The Executive Volume and Background Documents allow for alternate instruments, such as the steam generator level indication (for larger leaks) or the sanpling system (effective for smaller leaks). Per the Generic Instrumentation Section of Generic Issues portion of Executive Volume, only two channels of secondary radiation detection are necessary. NCGS procedures EMG E-3, step 2, utilizes turbine driven AEW pung steam exhaust monitor,- SG POEW plume monitors, SG blowdown monitors and SG sanple monitor.

Therefore, the two channel criteria is met, and two backup methods of determining SG radiation are provided.

CONCLUSION Because the ERG lists the steamline radiation monitors as one of several options, and because we meet the two channel criteria, steamline radiation monitors are not an instrument requirement.

Therefore,a safety-significant deviation from NOG guidelines does not exist.

REO(EST 2 Describe and justify, as part of the plant-specific technical cuidelines portion of the PGP, the eight EMGs (the licensee's identification for ECPs) related to Finding 9 of Task Analysis Final Report as potentially safety-significant deviations.

RESPONSE

Refer to the response to Request 1. The operator action is the same in each of the eight DIGS. Therefore these items need not be

diam aaad in the ADM 01-052, PGP.

Enclosura to page 5 Novenber 4, 1985 KMI2mc 85-245 REOLEST 3 Review its methods for identifying potentially safety-significant plant-specific technical deviations from the generic technical guidelines and deviations from the generic instrumentation and control characteristics to determine whether or not additional potentially safety-significant deviations exist. Describe and justify, in the plant-specific technical guidelines portion of the PGP, any

' additional, potentially safety-significant deviations or indicate that none exist as a result of its review.

RESPONSE

The method for identifying deviations from generic instrumentation and control characteristics was submitted via SIhaC 85-11 dated April 1, 1985. The actions and information requirements were developed independent of existing control room instrumentation and utilized NOG ERGS Revision 1. The results of this review are M'n=nted in SINRC 85-012 dated April 26, 1985 and clarified by SINRC 85-016 dated May 24, 1985. The findings that resulted consisted of human factor findings which address instrumentation and control characteristics but are not necessarily deviations from the generic guidelines.

A review of the EMG's and background material was conducted to identify potentially safety-significant plant-specific technical deviations from the NOG ERGB. The following criteria was utilized during the review:

j 1. Plant-specific steps which differ from the NOG HP Rev.

, 1 reference plant procedure steps were not considered to be deviations if they agreed with step conversion guidance of the background document or generic issues section of the administrative volume.

2. Control and instrumentation criteria was reviewed only for those cases where differences existed. This control and instrumentation criteria was then reviewed to ensure that the ERG j criteria was still adhered to.

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3. The following were not considered as deviations because ADM 01-052, Emergency Procedure Generation Package NCGS, Section 2.2.1.2 specifically exenpts them:

a) Ievel of detail.

l b) Rewording to conform to standard NCGS Terminology.

l c) Rearranging steps to armmv4te the NCGS control room design for

operator convienence when allowed by the ERG step sequence table. ]

l (The procedure ADM 01-052 has previously been submitted in letter l KMINRC 84-048 dated April 4, 1984 to H.R. Denton, NRC, from G.L. l Koester,KG&E) i

Enclosur3 to page 6 Novenber 4, 1985 10 D BC 85-245

4. Setpoints were not reviewed because they underwent independent review during the plant specific procedure generation process.

The potertial safety-significant deviations that resulted from this review are described below. After reviewing these, it was determined that no safety-significant deviations exist.

ITim 1 DESCRIPTION Actions in two (2) EMG's which address entry into DtG ES-02, Reactor Trip Response, have been modified from the ERG HP Revision 1 actions due to plant operating philosophy.

- JUSPIFICATION The decision was made by the Plant Safety Review Comittee to allow direct entry into ERG ES-02 following a reactor trip only if SI has not occurred and is not required.

• ES-02, Reactor Trip Response, Synptons or Entry Conditions and Steps 1-4

'Ihe phrase "or based on operator jt*=, aartt" was added to the section Syn #<== or Entry Conditions. The section now reads:

"This procedure is entered from EMG E-0, SAFE 1Y INJECTION, Step 4, or based on operator jt*=n=rit, when SI is neither actuated nor required."

The following steps, Steps 1-4, were added:

1. Verify Reactor Trip

2. Verify Turbine Trip

3. Verify Power to AC Eraergency Busses

4. Check if Safety Injection is actuated i In Step 4, if SI has actuated, the reactor operator is directed to l go to EMG E-0, Safety Injection step 5. IF SI has not actuated, the i operator is to check if SI is required. If SI is required the i reactor operator will manually actuate SI and go to DIG E-0, Safety l Injection Step 5. IF SI is not required, the reactor operator

! continues onto Step 5 in DtG ES-02.

!_ *DtG FR-S1, Rempaidie to Itaclear Power Generation, Synytoms or Entry l Conditions l

l A statement was added to the Sv=I*twa or Entry Conditions Section to allow an entry from DIG ES-02, Reactor Trip, Step 1. This allows the

Enclosur2 to page 7 Novenber 4, 1985 KMMRC 85-245 ,

operator to respond to al; anticipated tNansient without trip condition. The section now reads:

s.,

s This procedure is entered from:

1) EMG E-0, SAFE 1Y INJECTION, Step 1, when reactor trip is not verified and manual trip is not effective. , ,

4

2) EMG ES-02, REACIOR TRIP, Step 1, when >

reactor trip is not verified and manual trip is not effective.

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3) SUBCRITICALITY Critial Safety Function Status Tree on either a RED or ORAI E condition.

The addition of the 4 steps to EMG ES-02 h5s proven to be of benefit to the operator through sinulator use. He is able to direct more attention to plant activities rather than spend the initial critical time in procedure transition.

CONCLUSION ,

The entry into EMG ES-02 modification to the EMGs does not constitute a safety-significant deviation.

ITEM 2 DESCRIPTION ,

Actions in seven (7) f:MG procedures were modifie from the ERG actions due to a PORV block valve design difference. The affected procedures / steps are listed below: N s

. EMG E-1, Ioss of Reactor or Secondary Coolant, Ete Sc (ERG E-1)

DIG E-3, Steam Generator Tube Rupture, Step Sc (ERG E-3)

DtG C-21, Uncontrolled Depressurization of All iteam Generators, Step Sc and RNO (ERG ECA-2.1) 2 DIG FR-Cl, Response to Inadequate Core Cooling, Sted llc and RNO

~~ '

(ERG FR-C.1) .

EMG FR-C2, Response to Degraded Core Cooling, Step 3C and RNO (ERG FR-C.2)

DtG FR-C3, Response to Saturated Core Cooling Condition, Step 3c and RNO (ERG FR-C.3) ,

DIG FR-P1, Response to Inninent Pressurized Thermal Ghock Coridition, Step 2b ard RNO (ERG FR-P.1) ,

JUSI'IFICATION ,

The NOG guideline procedures direct the reactor operator to verify that at least one PORY block valve is open. The Wolf Creek emergency procedures direct the reactor operator to verify that at least one PORV block valve is capable of being opened.

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Enclosurm to page 8 Novenber 4, 1985 KMDIRC 85-245 The step was modified due to auto closure of PORV Block Valves at 2185 psig decreasing. Therefore, wording of this step was revised to assure relief capability, yet it also recognizes that the block valves may not remain open. It should be noted that by placing the system in

" ARM" that the block valves will remain open. This option was rejected since vapor space leakage protection could be reduced. This design difference was mandated by NUREG-0737 II.K.3.1 as described in FSAR Section 5.1.3.f. Note: Block Valves automatically re-open when RCS Pressure increases above 2135 psig.

CONCLUSION The modification due to PORV Block Valve design does not constitute a safety-significant deviation.

ITEN 3 DESCRIFfION The option to use the procedure without RVLIS was added to ten (10)

DIG's. The affected procedures are listed below.

ENG ES-05, Natural Circulation Cooldown with Steam Void in Vessel (without RVLIS), Synptom (ERG ES-0.4) dig ES-06, Natural Circulation Cooldown with Steam Void in Vessel (with RVLIS), Step 1 (ERG ES-0.3) dig C-11, Loss of Emergency Coolant Recirculation, Step 18a & c (ERG ECA-1.1) dig C-31, SGTR With Ioss of Reactor Coolant-Subcooled Recovery Desired, Step 2b of Foldout (ERG ECA-3.1)

ENG C-32, SGPR With Ioss of Reactor Coolant-Saturated Recovery Desired, Step 20b and RNO, Step 1 & 2b of Foldout (ERG ECA-3.2)

EMG C-33, SGrR Without Pressurizer Pressure Control, Step 7c, Step 11, Steps 1 & 2b of Foldout (ERG ECA-3.3)

ENG FR-Cl, Response to Inadequate Core Cooling, Step 6 (ERG FR-C.1) dig FR-C2, Response to Degraded Core Cooling, Steps 5, 7a, 18a (ERG FR-C.2)

EMG FR-P1, Response to Inminent Pressurized Thermal Shock Condition, Steps 5 & 12 (ERG FR-P.1)

ENG FR-I3, Response to Void in Reactor Vessel, Steps 8a & RNO, Steps 10a & RNO, Step 20, Attachment 'A' (ERG FR-I.3)

Each procedure was modified to allow the reactor operator to perform the procedure step with or without RVLIS.

JUSTIFICATION The ERG's provide guidance for developing procedures for using RVLIS or procedures if RVLIS is not installed. Since the ERG's contained guidance for plants without RVLIS, Plant Management made the decision to develop Contingency Actions for the case when RVLIS may not be operable. These contingency Actions were developed using the ERG's.

Enclosura to page 9 Novenber 4, 1985 KMLNRC 85-245 nis effort provides additonal information to the operator and does not detract from the plant specific EMG's.

CONCLUSION Providing guidance for the condition when RVLIS is not operable does not constitute a safety-significant deviation.

HEL i DESCRIPTION Se ' reset of SI has been added to DIG ES-05, Natural Circulation With Steam Void In Vessel (Without RVLIS), Step 1. This is a potentially safety significant deviation from the generic guidelines due to a plant design difference.

JUSTIFICATION The step was added to ensure undervoltage relays do not actuate when

' the reactor coolant punp(s) are started. This' reflects a connitment regarding Confirnatory Issue #18 in SDIRC 83-006 dated February 2, 1983. Resetting the SI signal prior to an attenpt to start RCPs C or D will reset the SI output relays and the inmediate undervoltage trip is removed from the offsite power breaker control circuits.

C0tCLUSION W is is not a safety-significant deviation.

I'IBt 5 DESCRIPTION A foldout in the generic guidelines which addressed RHST switchover criteria has been deleted in EMG ES-13, Transfer to Cold Iag Recirculation Following Ioss of Reactor Coolant.

JUSTIFICATION The foldout was deleted to avoid confusion since RHST switchover to cold leg recirculation nust occur prior to initiating hot leg recirculation.

CONCLUSION The deletion of the foldout step that addressed RHSP switchover criteria is not a safety-significant deviation.

Enclosura to- page 10 Novenber 4, 1985 IGENC 85-245 ITDI 6 DESCRIPTION The~ phrase "High Radiation from any SG Steamline" was deleted in ENG E-3, steam Generator 'nabe Rupture, Step 2. "High Radiation from any SG Relief or Safety Plume Monitor" was inserted since Nt,GS has plume monitors but does not have steamline radiation monitors.

JUSTIFICATION i The use of steamline radiation monitors is not mandatory based on guidelines in the Generic Issues, Instrumentation Section which states: '"At least two channels of a measurement system for detecting secondary radiation are assumed to be available to the operator.

Several means of inplementing this monitoring function are available.

These may include dedicated steamline radiation monitors, condenser air ejector radiation monitors and steam generator blowdown radiation monitors."

CONCLUSION This modification in EMG E-3, Step 2 does not constitute a safety-significant deviation.

ITEM.7 F DESCRIPTION 4

The phrases "EoS bottom light-lit" and " Rod position indicators-at zero" in ERG ECA-0.0, Ioss of all AC Power, Step 1 were deleted in EMG C-0.

JUSTIFICATION The NOG Guidelines instruct the operator in ECA-0.0, Step 1 to verify reactor trip by the following:

. Rod bottom lights-lit

. Reactor trip and bypass breakers open i

. Rod position indicators-at zero

. Neutron flux-decreasing In the SNUPPS design, the rod bottom lights indicator and rod position indicator is the same indication. Also, upon loss of AC power, this

- indication is deenergized. Reactor trip is adequately verified by in EMG C-0, Step 1 by verifying the reactor trip and bypass breakers are open and decreasing neutron flux.

CONCLUSION ,

This is not a safety-significant deviation.

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Enclosura to page 11 Novenber 4, 1985 1001RC 85-245 ITEN 8 This item is discussed in the response to Request 3 under Finding 1.

ITEM 9 DESCRIFfION The method of establishing mininum EOCS flow for decay heat removal in ENG C-ll, Loss of Emergency Coolant Precirculation, Step 11 was modified from the guidelines due to plant design differences.

JUSTIFICATION The step in which the operator is to establish ECCS flow to remove decay heat was modified to allow use of BIT inlet Dypass Valves. These valves are able to be throttled from the Control Room and the reactor operator can establish mininum EOCS flow effectively. This was determined preferable to starting / stopping ECCS punps to achieve the same flow rate as reconmended in the guidelines. The ERG Executive Volume Background document addresses throttling EOCS flow for SI reduction allowing it as a refined option.

CONCLUSION This modification does not constitute a safety-significant deviation.

ITEN 10 DESCRIPTION Additions were made to ERG ECA-2.1 in EMG C-21, Uncontrolled  !

Depressurization of All Steam Generators. These additions address cold overpressure protection. The additions are listed below.

Hgtg M-RCS pressure should be less than 650 psig prior to reducing I RCS cold leg tenperature(s) below 368 P and arming cold overpressure protection.

Sten R-Check if cold overpressure protection system nust be placed in service. If RCS cold leg is less than 368 F, RCS cooldown continues. If not, the cold overprotection system is placed in service.

Sten R-Check if RCS hot leg tenperature is less than 350 F. If it is BCS cooldown continues. If not, the SI punps, one (non-operating) CCP, and the PDP are placed in " Pull to Iock" position and the punp breakers are opened prior to RCS cold leg tenperatures decreasing below 325 F.

Enclosur3 to page 12 Novenber 4, 1985 10 D EC 85-245 JUSTIFICATION

'these additions ensure coupliance with the Wolf Creek Generating Station Technical Specification.

CONCLUSION This does not constitute a safety-significant deviation.

REOWST 4 The PGP should be modified to state that the task analysis for the EOP Upgrade Program was described as part of the DCBDR.

RESPONSE

The PGP has been changed to reflect the option of using the task analysis of the Detailed Control Room Design Review as one of the means of " validation".

REOUESP 5 Lescribe, as part of Finding 10 of the Task Analysis Final Report, the operator actions that will be taken at Wolf Creek, other than observation of SG water level, to identify the SG having a ruptured tube. Limiting operator action times should be included. Also, the radiation detectors and their location in the main steam system, the i emergency feedwater turbine exhaust and, in the SG blowdown system, should be described.

RESPONSE

The reactor operator at hCGS, in addition to observing SG water level, will utilize the following to identify the faulted steam generator (s):

1. Check for abnormal radiation from any of the following:

l l a. Turbine Driven AFW Pung Exhaust l If radiation is high then probability of ruptured tube in 'B' or 'C' steam generator.

b. SG Relief Plume Monitor i

If radiation is high two itens are indicated

1) Release is in progress

2) High tenperature high pressure reactor coolant via a ruptured tube causing overpressure condition.

c. Inline Steam Generator Blowdown or Sanple Radiation Monitor Monitor for abnormal radiation levels by utilizing one generator at a time.

l

__ _ _ _ _ _ _ - - . _ _ J

Enclosure to page 13 Novenber 4, 1985 KMLNRC 85-245

2. If actions and indications of Item 1 do not positively identify the ruptured steam generator, then the operator is directed to re-establish SG sanple and to request chemistry to abtain a grab sanple of the most suspect steam generator, followed by grab sanples of all other steam generators.

Operator action times have previously been provided in SI2mc 84 044 -

dated March 16, 1984 and SINRC 84-129 dated Decenber 3,1984.

Radiation detectors and location:

Main Steam Rad MonitgI,g.

AB RE-114 'A' SG PNV plume monitor (M-12AB01/1)

AB RE-113 'B' SG PORV plume monitor (M-12AB01/1)

AB RE-ll2 'C' SG PCRV plume monitor (M-12AB01/1)

AB RE-lll 'D' SG PORV plume monitor (M-12AB01/1)

SG Blowdown M RE-25 SG Blowdown non-regenerative heat exchanger outlet (M02m02/11)

BM RE-52 SG blowdown surge 'IK outlet to liquid radwaste discharge header (M02 m04/5)

SG Rannle SJ RE-2 SG sanple downstream of the sanple isolation valves (solenoid operated) and sanple flow indicators. Note: Sanple can be 4

individaully restored to determine ruptured SG. Also, grab sanple may be drawn for analysis.

A. revision to the Task Analysis Final Report is not required.

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