ML20202G556

Amends Response in Util Re Functional & Task Analysis,Per 860505 & 23 Telcons.Addl Changes Made to More Clearly Differentiate Between Callaway & Wolf Creek Emergency Response Guidelines & Operating Procedures

ML20202G556
Person / Time
Site:	Callaway
Issue date:	07/09/1986
From:	Schnell D UNION ELECTRIC CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
ULNRC-1323, NUDOCS 8607160001
Download: ML20202G556 (13)

Text

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- Union Etscraic a

1901 Gratiot Street. St. Louis Donald F. Schnell Vce President JUIY 9: 1986 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Denton:

ULNRC-1323 DOCKET NUMBER 50-483 CALLAWAY PLANT REQUEST FOR ADDITIONAL INFORMATION REGARDING FUNCTIONAL AND TASK ANALYSIS

Reference:

ULNRC-1242 dated January 14, 1986 The purpose of this letter is to amend the response provided in the referenced letter as a result of phone conversations held on May 5 and 23, 1986. The amended portions have been marked by revision bars. Additional changes have also been made to more clearly differentiate between Callaway and Wolf Creek procedure numbers and Emergency Response Guidelines (ERG) and Emergency Operating Procedures (EOP).

If there are any questions please contact us.

Very truly yours, Donald F. Schnell DS/llb Attachment i

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8607160001 860709

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PDR ADOCK 05000483 '

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Mading Address: P.O. Box 149, St. Louis, MO 63166

n STATE OF MISSOURI )

) SS CITY OF ST. LOUIS )

t Donald F. Schnell, of lawful age, being first duly sworn upon oath says that he is Vice President-Nuclear and an officer of Union Electric Company; that he has read the foregoing document and knows the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and that the facts therein stated are true and correct to the best.of Fis knowledge, information and belief.

By Donald F. Schnell Vice President Nuclear SUBSCRIBED and sworn to before me this f day of - T dd- , 1986.

&nh lg'&

BARBARA JNFAFF f NOTARY PU2UC, STATE OF Miss0URI WY COMMISSION EXPIRES APRIL 22,198'J ST. LOUIS COUNTY l

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r cc: Gerald Charnoff, Esq.

Shaw, Pittman, Potts & Trowbridge 1800 M. Street, N.W.

Washington, D.C. 20036 Nicholas A. Petrick f

Executive Director SNUPPS

5 Choke Cherry Road 20850 Rockville, Maryland

't C. W. Hehl Division of Projects and ,

Resident Programs, Chief, Section lA 4 U.S. Nuclear Regulatory Commission 1 Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 l Bruce Little

) Callaway Resident Office U.S. Nuclear Regulatory Commission RRil i Steedman, Missouri 65077 I L Paul O'Connor Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 316 7920 Norfolk Avenue Bethesda, MD 20014 Manager, Electric Department Missouri Public Service Commission P.O. Box 360 ,

Jefferson City, MO 65102 i

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Page 1 of 10 Attachment to ULNRC-1323 l UNION ELECTRIC RESPONSE TO THE NRC CONCERNING FUNCTIONAL AND TASK ANALYSIS REQUEST 1 Demonstrate that the task analysis based on Revision 1 of the Emergency Response Guidelines (ERGS) is applicable to Callaway.

RESPONSE

The Callaway Emergency Operating Procedures (EOPs) follow the generic ERGS in format and identification. The Callaway EOPs have been based on Revision 1 of the ERGS since January 1, 1986.

(See Response to Request 5) .

REQUEST 2 Modify the Procedures Generation Package (PGP) to state that the task analysis which supported the Emergency Operating Procedures (EOP) Upgrade Program was described as part of the Detailed Control Room Design Review (DCRDR). ,.

RESPONSE

The PGP (APA-ZZ-0 010 2) has been modified to indicate the task analysis, which was done as part of the DCRDR, is used in the EOP Update Program.

REQUEST 3 Describe and justify the deviations from Revision 1 of the ERGS indicated in the Task Analysis Final Report, Findings 1, 6, 8, 9, and 10.

RESPONSE

Note: These justifications will be included in the Callaway Procedures Generation Package (APA-ZZ-00102) by reference to this letter, ULNRC-1323.

JUSTIFICATION Per ERG background document, monitoring BIT temperature for solubility limitations is only a concern for systems having Boron concentrations greater than 7000 ppm. The reduction of BIT Boron concentration from greater than 20,000 ppm at the reference plant to 2000 ppm at Callaway is addressed in SLNRC 84-0070 dated April

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, Page 2 of 10 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 temperature was deleted in EOP ECA-0.0 l due to the reduced Boron concentration.

CONCLUSION BIT Boron solubility is not a concern at Callaway, therefore, the need to monitor BIT temperature and the need for control room indication is not necessary. It does not constitute a safety-significant deviation from Revision 1 of the ERGS.

Finding 6 UE 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.

JUSTIFICATION EOP E-1, Rev. 2 used RHR pump recirculation valve automatic l closure at 535 GPM to be indication of RHR flow to RCS. The procedure was changed to use a calculated value of 550 GPM (which includes instrument errors) as read on EJ FI-618 or EJ FI-619.

This change results in positive indication of minimum RHR injection flow to the RCS.

The Westinghouse Owners Group (WOG) Emergency Response Guidelines (ERG) Procedure ERG E-1 step 13A directs operator action after verifying RHR injection into the RCS. Callaway Emergency (EOP) procedure E-1, step 13A Revision 1 requires operator action after verifying RHR pump ficw to the RCS to be greater than 550 gpm, which indicates flow to RCS. The operator action is the same so there is no deviation from WOG ERG guideline actions.

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

Finding 8 The ERG background documentation for FR-C.1, Step C-lb, lists CCW to RHR heat exchanger flow as an instrumentation requirement. No instrument for this exists in the control room.

JUSTIFICATION The WOG ERG FR-C.1 cautions the operator to verify that the RHR pumps are not operated longer than a specified time without CCW flow to RHR heat exchanger to prevent pump damage. Callaway EOP FR-C.1 directs the operator to check for RHR flow and if there is none, directs him to initiate CCW flow. Acceptable alterna-tives exist for the indication of CCW flow. Control room annunciators 51A and 53A alert the operator to HI/LO CCW flow conditions. RHR inlet / outlet temperature indication across the

Page 3 of 10 heat exchangers is an acceptable indication of CCW flow to the heat exchangers, and is available in the control room. EOP FR-C.1 provides guidance to establish CCW flow to the RHR heat exchanger. In addition,CCW to RHR heat exchanger flow indication is available locally and on the BOP CRT located in the control room. Since the reactor operator has adequate CCW flow information available, the actions of the FR-C.1 caution remain within the WOG ERGS.

CCW 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 example, backup instrumen-tation is not required to be redundant, powered from a highly reliable source, and is not needed to be either accessible on demand or recorded. Therefore, the instrumentation used in the EOP to verify CCW flow to the RHR heat exchanger meets the ERG criteria and is not an instrument and control deviation.

CONCLUSION Since CCW flow information is available to the operator and the actions and instrumentation remain within the WOG ERGS, this is not a deviation.

Finding 9 ERG background documents for eight of the ERGS list CCW Flow to Seal Water Heat Exchanger as an infor-mation requirement. No instrumentation for this information is provided in the control room.

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

1. Callaway (Cal) EOP FR-I.1 (Wolf Creek (WC) FR-I .1) , l Response to High Pressurizer Level, Step 4 (ERG FR-I.1, Stop 2)

2. Cal EOP E-3 (WC E-3), Steam Generator Tube Rupture, Stop l 34 (ERG E-3, Step 34)

3. Cal EOP ES-1.2 (WC ES-ll) , Post-LOCA Cooldown and l Depressurization, Step 26 (ERG ES-1.2, Step 26)

4. Cal EOP ES-1.1 (WC ES-03), SI Termination, Step 16 (ERG l ES-1.1, Step 16)

5. Cal EOP ECA-2.1 (WC C-21), Uncontrolled Depresaurization l of ALL Steam Generators, Step 27 (ERG ECA-2.1, Step 27)

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6. Cal EOP ECA-3.1 (WC C-31), SGTR with Loss of Reactor l Coolant-Subcooled Recovery, Desired, Step 31 (ERG ECA-3.1, Step 31)

7. Cal EOP ECA-3.2 (WC C-32), SGTR with Loss of Reactor l Coolant-Saturated Recovery Desired, Step 25 (ERG ECA-3.2, Step 25)

8. Cal EOP ECA-3.3 (WC C-33), SGTR Without Pressurizer l Pressure Control, Step 19 (ERG ECA-3.3, Step 18)

These steps reference CCW flow to the Seal Water Heat Exchanger.

This indication is also classified as backup plant instrumentation. Acceptable alternatives exist in the Control Room for indication of proper CCW flow. A check of service loop flow on EC-FI-55A (of which the seal return heat exchanger is a part), CCW to and from service loop valve positions, and the absence of annunciator 54F "CCW Seal HX Flow HILO" are sufficient to assure that proper CCW flow thru the seal return heat exchanger exists. These procedures have been modified to include a specific reference to annunciator 54F to verify sufficient CCW flow.

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

Finding 10 The background document for ERG E-3, Step 2, lists steamline radiation monitors as an instrument requirement. None is provided in the SNUPPS control 1 room.

JUSTIFICATION The intent of the Emergency Response Guidelines 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 sampling system (e f fective r for smaller leaks). Per the Generic Instrumentation Section of Generic Issues portion of Executive Volume, only two channels of secondary radiation detection are necessary. Callaway procedure EOP E-3, step 2, utilizes SG blowdown monitors and SG sample monitor, and high radiation from any steamline.

Therefore, the two channel criteria is met, and one backup method of determining SG radiation is 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.

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1 Therefore, a safety-significant deviation from WOG guidelines i does not exist.

REQUEST 4 Review the method used for identifying deviations and describe and justify all potentially safety-significant deviations not identified in Items 1 and 3 above.

RESPONSE

The method for identifying deviations from generic instrumen-tation and control characteristics was submitted via SLNRC 85-11 dated April 1, 1985. The actions and information requirements were developed independent of existing control room instrumen-tation and utilized WOG ERGS Revision 1. The results of this review are documented in SLNRC 85-012 dated April 26, 1985 and clarified by SLNRC 85-016 dated May 24, 1985. The findings that resulted consisted of human factor findings which address instru-mentation and control characteristics but are not necessarily

deviations from the generic guidelines.

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

1. Plant-specific steps which differ from the WOG.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 criteria was still adhered to.

3. The following were not considered as deviations because APA-ZZ-00102 specifically exempts them:

a) Level of detail.

b) Rewording to conform to-standard Callaway Procedure Terminology.

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

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

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ITEM 1 l DESCRIPTION

The option to use the procedure with or without RVLIS was added to all E-0, E-1, and E-3 series procedures on their respective foldout page. .The specific items dealt with the red path summary for core cooling and SI initiation criteria, J

f Specific "in-body" changes to incorporate RVLIS were added to several Callaway EOP's. These specifically are as follows:

q l EOP ECA-1.1, Loss of Emergency Coolant Recirculation, Step

' 18a (ERG ECA-1.1)

! EOP ECA-3.2, SGTR With Loss of Reactor Coolant-Saturated j Recovery Desired, Step 20a and RNO, (ERG ECA-3. 2) 4 EOP.ECA-3.3, SGTR Without Pressurizer Pressure Control, Step 8c, Step 12, and respective RNO columns, Steps 1 & 2b

! of Foldout (ERG ECA-3.3, Step 7c and Step 11)

EOP FR-C.1, Response to Inadequate Core Cooling, Step 6 and RNO column (ERG FR-C.1) l EOP FR-C.2, Response to Degraded Core Cooling, Steps 5 and 7 and respective RNO column (ERG FR-C. 2)

EOP FR-P.1, Response to Imminent Pressurized Thermal Shock

Condition, Steps 5 & 12 (ERG FR-P.1) j EOP FR-I.3, Response to_ Void in Reactor Vessel, Steps 8a, 10a j

and respective RNO (ERG FR-I.3)

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

JUSTIFICATION J The ERG's provids guidance for developing procedures for using i i RVLIS or procedures if RVLIS is not installed. .Since the ERG's l

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.

This effort provides additional information to the operator and 1 does not detract from the plant specific EOP's.

4 CONCLUSION ..

i Providing guidance for the condition when RVLIS is not operable

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does not constitute a safety-significant deviation.

l ITEM 2 i

i DESCRIPTION I

The reset of SI has been added to EOP ES-0.4, Natural Circula- l i

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Page 7 of 10 tion 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 i

The step was added to ensure undervoltage relays do not actuate when the reactor coolant pump (s) are started. This reflects a commitment regarding Confirmatory Issue #18 in SLNRC'83-006 dated February 2, 1983. Resetting the SI signal prior to an attempt to start RCPs C or D will reset the SI output relays and the immediate undervoltage trip is removed from the offsite power breaker control circuits.

CONCLUSION 1

This is not a safety-significant deviation.

ITEM 3 DESCRIPTION A foldout in the generic guidelines which addressed RWST switchover criteria has been deleted in EOP ES-1.3, Transfer to Cold Leg Recirculation Following Loss of Reactor Coolant.

! JUSTIFICATION This is to clarify that only one step of the foldout was deleted versus the whole foldout. The Cold Leg Switchover criteria directs the Reactor Operator to Step 1 of EOP ES-1.3, Transfer to Cold Leg Recirculation, if RWST level drops below 36%. The procedure from which this was deleted was in fact EOP ES-1.3. It is not necessary to direct an operator to a procedure he is already in.

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

. ITEM 4 DESCRIPTION l

1 The phrases " Rod bottom light-lit" and " Rod position indica-tors-at zero" in ERG ECA-0.0, Loss of all AC Power, Step 1 were deleted in EOP ECA-0.0.

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Page 8 of 10 JUSTIFICATION The WOG 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

. 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 EOP ECA-0.0, Step 1 by verifying the reactor trip and bypass breakers are open and decreasing neutron flux.

CONCLUSION This is not a safety-significant deviation.

ITEM 5 This item is discussed in the response to Request 4 under Finding 1.

REQUEST 5 Provide a cross-reference of the Callaway EOPs to Revision 0 and Revision 1 of the ERGS and identify each step of the eight EOPs given in Enclosure 2 that lists the Component Coolant Water Flow to the Seal Water Heat Exchanger as an instrumentation requirement.

RESPONSE

Comparison of Emergency Procedures GENERIC ERG Wolf Creek Callaway

a. FR-I.1 FR-I.1 FR-I.1

b. E-3 E-3 E-3

c. ES-1.2 ES-ll ES-1.2

d. ES-1.1 ES-03 ES-1.1

e. ECA-2.1 C-21 ECA-2.1

f. ECA-3.1 C-31 ECA-3.1

g. ECA-3.2 C-32 ECA-3.2

h. ECA-3.3 C-33 ECA-3.3 l

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Page 9 of 10 All the referenced procedural steps have been reviewed in both the Callaway EOP's and generic ERG's. All steps that are ques-tioned by the NRC involve plant specific steps for system resto-ration. The generic ERG considers these items to be plant specific, thus the instrumentation requirement for CCW flow to the seal water heat exchanger is not identified in the generic ERG step.

REQUEST 6 Describe the indications, other than steam generator water level, that the operator will use to identify the steam generator with a ruptured tube.  ;

RESPONSE

Operators at Callaway, 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:

a. High Steamline Radiation

b. Inline Steam Generator Blowdown Monitor for abnormal radiation levels by utilizing one generator at a time.

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

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 sample and to request chemistry to obtain a grab sample of the most suspect steam generator, followed by grab samples of all other steam generators.

Operator action times have previously been provided in SLNRC 84-044 dated March 16, 1984 and SLNRC 84-129 dated December 3, 1984.

Main Steam Rad Monitors AB RE-ll4 'A' SG PORV plume monitor (M-12AD01/1)

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

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

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

SG Blowdown BM RE-25 SG Blowdown non-regenerative heat exchanger outlet (M02BM02/11)

BM RE-52 SG Blowdown surge TK outlet to liquid radwaste discharge header (M02BM04/5)

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> SG Sample SJ RE-2 SG sample downstream of the sample isolation. valves (solenoid operated) and sample flow indicators.

4 Note: Sample can be individually restored to

determine ruptured SG. Also, grab sample may be drawn for analysis.

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

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