IR 05000282/1988010

Insp Repts 50-282/88-10 & 50-306/88-10 on 880606-21. Violations Noted.Major Areas Inspected:Verification That Emergency Operating Procedures Technically Correct & Specified Actions Accomplished Using Existing Equipment

ML20151E980
Person / Time
Site:	Prairie Island
Issue date:	07/11/1988
From:	Love R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20151E957	List: IR 05000282/1988010 ML20151E956 ML20151E971
References
50-282-88-10, 50-306-88-10, NUDOCS 8807260283
Download: ML20151E980 (18)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Reports No. 50-282/88010(DRS);50-306/88010(DRS)

Docket Nos. C0-282; 50-306 Licenses No. DPR-42; DPR-60 Licensee: Northern States Power Canpany 414 Nicollet Mall Minneapolis, MN 55401 Facility Name: Prairie Island Nuclear Generating Plants Units 1 and 2

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Inspection At: Prairie Island Plant Site, Welch, Minnesota Inspection Conducted: June 6-21, 1988 Y W Inspectors: R. S. Love Team Leader 9"/kG/ fBP Reactor Inspector, Region III Date J. Hard, Senior Resident Inspector Prairie Island J. Hopkins, License Examiner, Region III J. Lennartz, License Examiner, Region III B. Glickstein, Consultant (SAIC)

J. Sears, Consultant (Comex) ,

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Approved By: T. M. Burdick, Chief Operator Licensing Section 2 (DRS) Date Region III Inspection Summary Inspection on June 6-21, 1988 (Reports No. 50-282/88010(DRS);50-306/88010(DRS) Areas .'.nspected: Special announced safety inspection to verify that the Priirie Island Emergency Operating Procedures (E0/s) are technically correct; that ;;eir specified actions can be meaningfully accomplished using existing equipment, 88072o0283 880712 PDR ADOCK 05000282 Q PDC

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controls, and instrumentation; and that the available procedures have the ;

usability necessary to provide the operator with an effective operating too The inspection was conducted in accordance wf th Temporary Instruction (TI) 2515/92. (SIMS No. HF 4.1)  ;

Results: One violation was identified against 10 CFR 50, Appendix B, Criterion XVIII - Failure to perform planned and periodic audits of the Prairie Island Emergency Operating Procedures between April 1984 and April 14,

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1988 (Paragraph 4).

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DETAILS

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- Persons' Contacted ,

-Northern States Power Company (NSP)

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• L. Eliason, General Manager, Nuclear Power Plants

• M. Sellman, General Superintendent, Plant Operations

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• E. Watzl, Plant Manager

,. *M.Wadley,ShiftManager(SRO) a

• J. Goldsmith, Superintendent, Nuclear Technical Services

'*7. Bacon, QC Specialist-

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' *H. Julian, E0P Writer (Volian Enterprises)

• D. Schuelke, Superintendent, Radiation Protection

• D. Reynolds, Operations Training Supervisor

• Waldron, Senior 0perations Specialist . .

• "mdele, Superintendent, Engineering and Radiation Protection

• 5 a -spr' .3, Lead Production Engineer

<cGillie, Operations Training Supervisor (Monticello)

% Goranson, Senior Production Engineer -(Monticello) ,

• M. Werner, Training Instructor * i D. Smith, Onerations Instructor (Westinghouse)

M. Gardzin.Ki, Instructor .(SRO)

J. Sorenson, Shift Manager (SR0)

S. Rogers, Reactor Operator (RO)

W. Eppen, Reactor Operator-(RO)

G. Dammann, lead Reactor Operator (RO)

L. Henry, lead Reactor Operator (SRO)

H. Pemble, Shift Supervisor (SRO)

R. Thorkelson, Lead Reactor Operator (RO)

S. Chezick, Reactor Operator (RO)

W. Irvin, Lead Plant Equipment Operator (RO)

G. Woodhouse, Shift Supervisor (SRO)

W. Mather, Lead Plant Equipment Operator (SR0)

S. Groh, Assistant Plant Equipment Operator  ;

P. Kramer, Apprentice Plant Attendant

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D. Page, Assistant Plant Equipment' Operator J. Gosman, Lead Plant Equipment Operator i U.S. Nuclear Regulatory Commission (US NRC)

• H. Regan, Jr. , Chief, Human Factors Assessment Branch, NRR

• J. Miller, Director, Division of Reactor Safety, Region III

• Moser, Resident Inspector, Prairie Island SR0 denotes a licensed Senior Reactor Operator I R0 denotes a licensed Reactor Operator Other licensee personnel were contacted / interviewed during the inspectio l

• Denotes those personnel +1n attendance at the exit interview on June 21, 198 !

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2. Emergency 0perating Procedures (25592) Background Emergency Operating Procedures (E0Ps) have u_ndergone significant changes due to the 1979 accident at the Three Mile Island (TMI)

facility. The post-TMI procedures are symptom-oriented rather than '

event-oriented. Symptom-oriented E0Ps provide the operator guidance on how to verify the adequacy of critical safety functions-and how to restore a'nd maintain these functions-when they are degraded.-

Symptom-oriented E0Ps are written in a manner that the cperator- '

need not diagnose an event to maintain-the plant in a_ safe shutdown condition for all accidents that are within the scope of the E0P The purpose of this inspection was to verify that the Prairie Islan E0Ps are technically correct; prepared in accordance with the writer's guide; that their specified actions can be accomplished using existing

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equipment, controls, and instrumentation; and that the available procedures have the usability. necessary to provide the operator with an effective operating too *

This was accomplished by performing: a desk-top review of 25 Optimal Recovery Procedures, six Critical Safety Function Status Trees,18 Function Restoration Procedurcs, and two Abnormal Procedures; system walkdowns of eight Recovery Procedures, two Restoration Procedures, and one Abnormal Procedure; eight scenarios on your plant specific '

simulator that exercised 19 procedures and the six status trees; and a human factors review during the desk-too review and walkdown of the-procedures, and during the simulator scenarios. In addition, 11 users and developers of the E0Ps were interviewed. For a detailed listing ,

of the procedures and status trees reviewed,-walked down a..d exercised on the simulator, see Appendix ,

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This inspection report provides examples of observations noted during l the inspection. The licensee was provided detailed debriefings in which all of the inspection team's observations were discussed. In addition, a detailed listing of all observations will be orovided tc -

the Prairie Island NRC Resident Inspector's office for followup and -,

. closur l Desk-Top Review The desk-top review was accomplished by comparing the Prairie Island l (PI) procedures and status trees identified in Appendix A with the l Westinghouse Owners Group (WOG) Emergency Response Guidelines (ERGS), l PI Procedure Generating Package (PGP), Writer's Guide and the Plant '

Specific Setpoint Document. The inspection team also reviewed ~the ERG and E0P background documents, the ERG Executive document, ar.d the

. PI Design Differences Document. When deviations between the various documents were identified, the inspectors verified that the deviations-

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were identified, documented, and justified in the Deviation Documen l When required, the inspectors also verified that a safety analysis I report had been prepared in accordance with 10 CFR 50.5 In addition,. l the inspectors reciewed the licensee's verification and validation  !

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(V & V) of the Prairie Island E0P !

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Results of Desk-Top Review Generic Technical _ Guidelines were prepared 'or all of the ERGS. These-generic guidelines provide a complete and documented analytical basis for each of the procedures. The Generic Technical Guidelines have-been verified by the WOG. The PGP and E0Ps were developed from the WOG_ Low' Pressure ERGS, Revision 1A.. A review was conducted by the licensee and it was concluded that in~ general, the ERG reference plant analysis was applicable and that no additional ' analysis wEs required to support the use of the_ ERGS to develop the Prairie Island PGP and

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E0Ps. A review of technical adequacy due-to-several design differences between the reference plant and Prairie Island was performed-by NSP engineering personnel. Any differences between the E0Ps and the WOG ERG, with exceptions, were identified, documented, and justified in-the Deviation Documen ~

In general, the Prairie Island.(PI) E0Ps_were found acceptable, however, the following concerns were identified:

E0P E-0 (See Appendix A),_ Step 7. Response Not Obtained column (RN0). Substeps b and c from the comparable ERG step were not included in the E0P and there was inadequate justification in the Deviation Document for this deviation. Due to design differences, PI does not have Phase.8 containment isolation valves. The licensee has consnitted to add additional i information to the Deviation Document from the Design Difference Documen E0P E-0, Step 16a and b, RN The licensee.has committed to expanding the justification in the Deviation Document to explain the differences between the ERG and E0P transition '

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E0P ES-0.0, Steps 3, RNO and 4, Action / Expected Response column (A/ER). After evaluation by NSP and NRC, the licensee has committed to word these steps in accordance with the ER E0P ES-0.1, Step 12.c, A/ER. The licensee has committed to I expand the justification in the Deviation ~ Document to explain the differences between the ERG and E0P steam generator (S/G)

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E0P ES-0.2, Attachment The licensee has comitted to define an "L" and "S" signal in the writer's guide or -)

in the attachmen E0P ES-0.?B, Step 11, A/E The licensee has committed to revise the Deviation Document to reflect deletion of pressurizer level versus RCS pressur E0P ES-0.4, Steps 1.a and b, A/E The licensee has comitted !

to expand the justification in the Deviation Document on ~l subscep sequence deviation ,

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E0P ES-0.4, Step 2.b,-A/ER. .The licensee has-comitted to provide additional information to the Deviation Document on pressurizer level control methods at PI to justify the deviation between the ERG and E0 *

E0P ES-1.1, NOTE before Step 1 (1N). The Deviation Document indicates that a NOTE had been added before Step 1, however, the NOTE does not appear in the E0P. 'he licensee has comitted . s' '

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to correct the Deviation Document by celeting the reference -to .

the NOT .

E0P ES-1.1, 100. The licensee deleted ERG CAUTIO "On natural

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e circulation, RTD bypass temperatures and associated interlocks will be inaccurate," with justification. The inspection team recomended that this CAUTION be reinstated. The licensee has committed to reevaluate the seed for this CAUTION in this procedure and throughout the E0Ps (Generic Issue).

E0P ES-1.1, Step 12, RNO. : At the Inspection Team's recomendation, the licensee committed to evaluate the need to add a contingency transition step for a case of both RHR pumps and no SI pumps runnin '

E0P ES-1.2, General. When a temporary change is made to a procedure, the licensee enters the change in the body of the procedure as though it was a revision to the procedure. If '

the temporary change cover letter was to be detacheu, there would be two different controlled procedure revisions on file, however, both revisions would carry the same revision number The same situation occurred with E0P ES-1.3. The-licensee committed to revising ES-1.2 and ES-1.3 and their associated background documents, and to delete the temporary change memo E0P ES-1.2, Steps 10.d and e, RN There was some confusion on the part of the inspectors as to when the transitions should be

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made in these two substeps. The licensee committed to clarify these transitions during the next procedure revisio *

E0P ES-1.3, Step 9.d and e, RNO. Same problem and resolution as noted in ES-1.2 Step 10.d ard e abov *

E0P E-3, Step 13. The adverse containment temperature values provided in E-3 do not match the temperature valves provided

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in the PI Setpoint Document. An investigation by the licensee indicated that the values contained in the Setpoint Document were correct. The licensee committed to revise E0P E-3 to correct the adverse containment temperature value E0P E-3, Steps 14 and 25. The licensee comttted to add ad11tional justification in the Deviation Document for step seeuerce deviations between the ERG and the E0 '

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' Generic Issue. There are instances throughout the E0PS where-the operator is directed to be in two E0Ps at the same tim Thi., is contrary to the general usage guidelines for E0P Paragraph 4.2.5 of the ERG Writer's Guide states "Transition shall not contain a ' return' feature-(e.g., performed Steps X through Y in some other procedure and then-return)."

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This type of deviation from good practice was identified in the following E0Ps:

ES-0.2, Step 13c, RNO ES-1,1,' Step 200, RNO E-3, Step 33c, RNO ECA-0.2, Step 1, RNO ECA-2.1, Step 25c, RNO ECA-3.1, Step 26c, RNO ECA-3.2, Step 19c, RNO ECA-3.3, Step 17c, RNO The above concerns are typical of the type identified during the-desk-top review. After an evaluation was performed, none of the concerns were identified as being safety-significan An E0P validation had been performed by the licensee to verify that the procedures were usable, i.e., they.can be understood and followed without confusion, delays, and error In addition, the validation program verified that the E0Ps guided the operator-in mitigating transients and accidents. The validation of the E0Ps were performed by a multi-discipline team. One or more of the following methods were used in the validation program: (1) desk-top review; (2) control  :

room plant walkdowns; (3) exercising the E0Ps on the plant specific simulato ,

During the review of Prairie Island's verification and validatio program, no safety-significant concerns were identifie c. Plant Walkdown Plant walkdowns of select E0Ps were performed during the inspection to verify that the specified actions could be accomplished by the operators using existing equipment, controls, and instrumentatio See Appendix A for listing of procedures walked down by the three

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inspection teams. Each team consisted of'two NRC personnel and a licensed Senior Reactor Operator (SRO), a Reactor Operator (RO), or a non licensed Plant Equipment Operator. During the walkdowns, ,

the inspectors specifically looked at component accessibility and

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identification (labeling / tagging), tools and protective equipment j needed for local equipment operations, emergency lighting, comuni-cations, and environmental conditions (radiation and temperatures)

during a Design Basis Accident (DBA). j (1) Prior to this inspection, the licensee updated their E0Ps to the WOG ERGS, Revision 1A. As part of their V&V i licensee walked down (February through May 1988) program, their E0P to the -I ensure usability. To supplement their V&V program, the licensee l generated a "Local Action Checklist" (see Appendix B) to identify potential problem areas such as: accessibility, environmental l

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conditions, inadequate lighting, and shift staffing. During their walkdowns, the licensee identified the need for: dedicated ladders / platforms, additio cl/ replacement components labels / tags, additional emergency lighting, and the installation of additional sound powered phone jack The inspectors were informed that corrective action on the identified items included the issuance of work orders or the item has been sent to engineering for evaluatio (2) During the inspection, the inspectors identified additional examples of the type of deficiencies identified by the license Examples follow:

E0P ES-1.2, Step 6, RNO. The operator is required to locally close Valve MV-32084 or MV-32085. The operator needed to climb on pipes and/or hangers to reach the valve The licensee took corrective action to install dedicated ladders in both RHR pit areas so that the operators can locally close these two valve *

E0P ES-1.2, Step 12, RN The operator is required to locally close Valve MV-32162 or MV-32163. At present, the operator needs to climb on pipes and/or hangers to reach the valves. Ladder or platform is neede E0P ES-1.2, Step 13, RNO. The operator is required to locally close Valve MV 32206 or MV 32207. At present, the operator needs to climb on pipes and/or hangers to reah the valves. Ladder or platform is neede During a walkdown of the containment isolation valves, it was noted that additional emergency lighting is needed in the steam generator blowdown (SGB) flash tank are The licensee has committed to submit to Plant Engineering Staff ;

the need for a more detailed evaluation for additional Emergency ,

Lighting in areas determined to be inadequate and the need for '

permanently installed catwalks and access ladders in the areas identified. Pending the review of Plant Engineering Staff's ,

evaluation on the need for permanently installed catwalks and i access ladders, this item is open (50-282/88010-01; )

50-306/88010-01).

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(3) As part of the licensee's environmental study, floor plans were l developed to show the Design Basis Accident (DBA) dose rates (R/hr) throughout the plant. The DBA dose rates are a result of safety system failure leading to major core damage with release to the containment atmosphere. This study is contained in Emergency Plant Implementing Procedure No. F3-25, Revision 4 Attachment A. "Dose Rate Calculation Description."

During plant walkdowns, the inspectors noted the areas entered to perform actions required by the E0Ps. Using this information, the inspectors were able to determine the calculated dose rates,

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(buringaDBA)fortheareasenteredfromth'elicensee's *

environmental study. 'It was.noted that during a DBA with '

recirculation mode in progress, the operator would have to

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enter a Radiation Field of approximately 100 R/hr to locally operate' Containment Isolation Valves in the SGB flash tank are i (4) In. general, the tools and protective equipment needed to .

i perform local operations as_ described in the E0Ps was adequate l .'

and readily available. Also, the labeling throughout the plant was very good. Several minor labeling concerns were identified and are discussed in Paragraph 3 of this repor Simulator Scenarios l . .

Eight scenarios were conducted on the PI plant specific simulator to verify that the PI E0Ps' provide the operator with an effective operating tool to place the plant in a safe shutdown condition for accidents and transients that-are within the scope of the E0P Nineteen E0Ps and the six Critical Safety Function Status Trees,.

as identified in Appendix A, were exercised during the scenario The scenarios were conducted in two four hour sessions utilizing the licensed operators from Crew 4, that were in Requalification Training. The simulator operating crew consisted of r Shift Supervisor (SS),LeadReactorOperator(LRO),ReactorOperator (RO),andaShiftTechnicalAdvisor(STA). This crew size meet the PI minimum Technical Specification requirement '

In general, the specific actions detailed in the E0Ps were:

technically correct; could be accomplished using the existing-- i equipment, controls, and instrumentation; and.provided the-  !

operators with an effective tool to place the' plant in a safe :'

i shutdown condition. The E0Ps led the operators through the correct transition points without much confusion. The Inspection Team observed one incorrect transition during the execution of E0P ECA-3.1 l

due to incorrect wording of Step 13, A/ER.

As worded, ECA-3.1, Step 13, directs the operator to check the '

status of the Safety Injection (SI) and Residual Heat Removal (RHR)

l pumps. If both pumps were running, the operator is to continue with !

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Step 14. However, if both pumps are not running, the operator is '

directed to tr;.sition to Step 1 In accordance with the WOG ERG, the intent of Step 13 is to continue with Step 14 if either the SI _or the RHR pump is running and to transition to Step 19 if neither pump is runnin During the scenario, the SI pumps were running and the RHR pumps were secured. The correct operator action was to continue with Step 14. However, the "0R" was omitted in Step 13 which resulted in the SG making an incorrect transition to Step 19. The safety significance of this transition error was evaluated by the inspectors.-

It was determined that subsequent procedural steps would have transitioned the operator back to Step 15 and automatic functions

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would'have deenergized the pressurized' heaters as required by Step 1 Therefere, the incorrect transition to Step 19 is not considered.to be safety significan It was also noted that the same generic step (Step 7) appears in E0P ECA-3.2 and the required "0R" was also omitted. During the evaluation of this procedural error it was also determined that subsequent procedural steps would have transitioned the operator back to Step 9 and automatic functions would have deenergized,the pressurizer heaters as required by Step 8. .Therefore, an incorrect transil. ion at Step 7 is not considered to be safety significan The licensee took prompt corrective action by revising-EOPs E'CA- and ECA-3.2 to incorporate an "0R" into the appropriate steps. The ,

revised procedures were placed in the Control Room' E0P binders .

before the Inspection Team departed the PI sit . Human Factors Review

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The objectives of the human factors review was to ensure that the E0Ps followed the guidance provided in the Writers Guide for Prairie Island ,

Nuclear Plant Emergency Operating Procedures, Revision 1, date May 1, 1988 and to ensure that the E0Ps can by physically and' effectively carried out. To achieve these objectives, the human factors evaluator performed a desk-to) review of the E0Ps, observed simulator. scenarios,

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participated in wal(downs of the E0Ps, and interviewed select users and developers of the E0P ' .esk-Top Review The E0Ps reviewed are listed in Appendix A and in general, they comply with the PI Writers Guide. The following exceptions were noted:

, Attachment E is not labeled to indicate that this graph is to be used for adverse containment conditions only. The attachment is not identified as to which E0P it pertains, nor-is there a revision number on the attachment (examples: E0Ps

E-0, ES-3.3 and FR-H.2).

Paragraph 4.3 of the Writers Guide states that NOTES and CAUTIONS shall not contain action steps. In E0P E-0, the CAUTION before Step 31 contains an action statement and in E0P ECA-3.1, the NOTE before-Step 16 contains an action statement (Generic).  ;

Paragraph 4.3 of the Writers Guide states that a description of ,

the consequence shall be included in the CAUTION so that the

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operator will know the concern. In E0P ECA-0.0, the CAUTION before Step 6 does not provide a consequenc E0P FP-P.2, Step 4 b.1, A/ER. Figure FRP2-1 is incorrectly referenced. The correct reference is Figure FRP2-2.

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The same step' is worded' differently in different procedures, e.g., EOPs E-3, Step 7 versus FR-C.1 Step 9 and ECA-1,1,

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Step 18.b, RNO versus ECA-3.1,. Step 31.b, RNO. Consistent wording should be maintained when the steps 1are the sam E0P ES-0.3B, Figures ES038-1 and ES03B-2. These figures.are '

labeled the same and contain different graphs. 0ne graph is used to determine the required condensate while the other graph is used to determine the available condensat ,

The' licensee has committed to correct the above listed deficiencies as well as the other. deficiencies provided during the debriefing session h. Walkdown of E0Ps The Inspection Team noted that~ the operators knew the plant and were able to simulate implementation of the E0Ps. .The location and labeling of E0Ps was very good. Durin and an Abnormal Procedure (see Appendix theA)g walkdowns following of the E0Ps findings were identified:

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(1) During walkdowns of Abnonnal Procedure C1.8, Step 9, it was noted that large yellow tags were placed on valves that need to be locally operated following a turbine / reactor tri This was considered to be an outstanding operator ai (2) Figure Cl-10, used for calculating boron addition, was missing from Abnormal Procedure C1.8. This figu're is required to perform Step 8 of the procedure at the Hot Shutdown Panel. The licensee took innediate action to place Figure C1-10 in all the controlled copies of Procedure C (3) A labeling inconsistency was identified in E0P FR-S.1, Step Sa, RNO, which directs the operator to open the MG set input and output breakers. Locally, the breakers are listed as Motor and ,

Generato (4) In E0P E-0, Step 5, RNO, the operator was directed to manually or locally align safeguard components. At present, there is no listing of containment isolation valves in the procedure nor is ,

a list available to the Equipment Operator in the Control Room or in the Auxiliary Building work station. However, the licensee was in the process of. generating a listing of Containment Isolation Valves and committed to place this list in the Control Room and Auxiliary Building work statio (5) Meters were identified with no unit designators ( F or psig).

Examples include: .Wide Range Steam Generator Level, ILR-460; RCS Cold Leg Temperature, ITR-450 and ITR-451; and Pressurizer i Level, 1RP-42 i (6) In several areas of the plant, labels are made with dyno-tape and has the potential for falling off. This concern was

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previouslyl identified'by the licensee and' corrective action was in progres The licensee has committed to correct the above listed deficiencie Simulator Scenarios \

lDuring observation of simulator scenarios, it was noted that staffing levels were adequate, the roles-and responsibilities of the crew were clearly defined, a team approach was utilized, and the operators were able to carry out their assigned tasks ,without difficulty'or' conflict The SR0 followed the procedures and. demonstrated an understanding of the intent of the E0P In general, transitions within procedures and between procedures was s.mooth. See Paragraph 2.d of this report for an identified transition erro With regs"d to place-keeping, each procedure is in a separate binder and the binders are clearly labeled. The operators, using special pens, check-off the steps in the procedures as the steps were completed, and used pens, pencils,'and paperclips as' place-keeping aids. Although no specific instances were noted where the operators lost their place, some difficulties were obscrved when the' operator was directed to return to a procedure and step in effect. The. licensee committed to evaluate place-keeping "methods .and revise practicesnas appropriat '

During the simulator scenarios, it was noted that emergency lighting in the Control Room was adequate to~ implement the E0Ps during a station blackout. See Paragraph 2.c(1) of this report for inplant emergency lighting deficiencies note Personnel Interviews ,

During the inspection,11 users / developers of the Prairie Island E0Ps were interviewed. Interviewees included: one Shift Supervisor (SRO); six Lead Plant Equipment and Reactor Operators (two SR0s and four R0s); two Assistant Plant Operators; an Apprentice Plant Attendant; and one E0P write In ger.eral, the operators interviewed had a positive attitude toward i the E0Ps and supported their use. The' interviewees identified two !

concerns which had also been 16 it ied-by the Inspection Tea )

The first concern pertaineo e of negatives in the action /

expected response column. T h t ., arn was discussed with the licensee and with the NRR Reviewe. for WOG ERG, Revision 1A'.

The second concern pertaineo to component location. At present, the location of an item is listed by column and elevation, however,'the columns are not always clearly identified. This concern was discussed'

with the licensee and they committed to review the column markings and to evaluate the need to supply additional locating aid _ ']

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The operators felt the comunications systems (pager, dial phones,

.. -sound-powered phones, etc.) throughout the plant were good. However, the licensee and the NRC identified areas ~ where addit.ional sound-powered phones and jacks would be desirable. See Paragraph 2.c o this report for. additional detail : QA Audit-During the inspection, the inspectors requested copies _of the last two Q audits of the Prairie Island PI) E0Ps for revie The inspectors were informed that only one audit was performed in this area and was provided with a copy of QA Audit Report Number AG-88 20-13, dated May 23, 1988. A review of the audit report indicated that it was an indepth revie Following are examples of the audit findings: In procedure E-1, there were four transition steps that referenced-the wrong procedure. Procedure E-1 was revised during the-audit to reference the correct procedure Identified 20 E0Ps that had attachments as part of the procedure but were not listed on the procedure's introduction page. Thes deficiencies were corrected during the audi Identified potential step sequence errors in E0P E-3. As of June 10, 1988, the licensee is evaluating the need to revise E-3 so as to be consistent with the ER During walkdown of various E0Ps, it was identified that the SI test line to the RWST valves were missing the proper labels. As of June 10, 1988, these valves had proper tags applie Nine potential deficiencies were identified and the auditor recommended corrective action on six of the deficiencies that were not corrected during the audit. As of June 10, 1988, all but two of i.he potential deficiencies have been corrected. These two items are still being ,

evaluated by the licensee. One item under evaluation is discussed in as l

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item 4.c abov The other item pertains to the need to environmentally qualify (EQ) the steam generator (SG) narrow range level indicatio Under normal containment conditions, the operator uses SG narrow range i level indication. Under adverse containment conditions, the operator must use wide range instrumentation because the SG narrow range level instrumentation is not qualified for harsh environmen This item is being evaluated by NSP Technical Engineering staf During personnel interviews, the inspectors were infonned that the QA

, audit discussed above had been requested by the PI Operations Department ,

because the audit of PI E0Ps was not on the QA Audit Schedule. It was  !

also learned that the PI E0Ps were first approved for use in April 198 l All evidence obtained from review of documents and personnel interviews  !

confirms that the PI E0Ps were not audited by QA from April 1984 until l the first audit was started on April 14, 1988. The licensee was informed !

that failure to perform planned and periodic audits of the PI E0Ps is a violation of 10 CFR 50 Appendix B, Criterion XVII :

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Further investigation'into this matter revealed that as of June 20, 1988, the QA Audit Schedule contained the requirements for_ planned and periodic *

audits of the PI E0P In summary, the inspection showed that action has been taken to correct 1 the identified violation (faiure to perform planned and periodic.audtis '

of the PI E0Ps)' and to> prevent recurrence. An indepth audit of the PI E0Ps was conducted'on April 14 - May 11,1988,-and the ~QA Audit Schedule now contains the requirements for planned and periodic audits of the PI E0Ps. Consequently, no reply- to the violation is required and-pe have no further questions regarding this matte . E0P Summary

. As noted in Paragraphs 2, 3 and 4 of this report, various concerns were identified by the Inspection Team. For the concerns identified, the '

licensee has either:

Completed the corrective action before the Inspection Team left the sit *

Comrc.itted to specific corrective action to resolve the deficienc *

Committed to perform an evaluation to determine corrective action, if require During the inspection, the following positive attributes were ,

identified: *

Prairie Island (PI) has upgraded their E0Ps to meet Revision 1A

, of the WOG Emergency Response Guideline PI has an effective Verification and Validation (V&V) program 1 in place. The feedback from the operators / engineers and the VAV program is also very goo #

To supplement the V&V program, P1 has developed a Local Action Checklist which is included as Appendix B.to this repor *

Background informatior has been added to individual procedure l steps to clarify the 1 tent of the ste In addition, all of the background documentation for a procedure is filed in the procedure ;

binder. Each E0P is in a separate binder and is well identifie . Open Items Open items are matters which have been discussed with the licensee which will be reviewed further by the inspectors and which involves some actions on the part of the NRC or licensee or both. O ,

this inspection is discussed in Paragraph 2.c(pen item disclosed during 2)ofthisrepor l 4 I 14 ,

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7~. - Exit Interview Theinspectorsmetwithlicenseerepresentatives-(denotedinParagraph1)

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on June.21, 198 The inspectors summarized the: purpose, scope, and ~<.

findings of the inspection and the likely-informational . content of the!

inspection report. The licensee acknowledged this information and-did not identify any proprietary informatio r (J-l i

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APPENDIX A PRAIRIE ISLAND EMERGENCY OPERATING PROCEDURES

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, OPTIMAL REC 0VERY PROCEDURES

• fE-0, Revision 4, Reactor Trip or Safety Injection - ,

ES-0.0, Revision 2, Rediagnosis

1. ES-0.1, Revision 4, Reactor Trip Response fES-0.2, Revision 3, SI Termination

• fES-0.3A, Revision 0, Natural Circulation Cocidown w/CRDM Fans ES-0.3B, Revision 0, Natural Circulation Cooldown w/o CRDM Fans ES-0.4, Revision 1, Nataral Circulation Cooldown w/ Steam Void in Vessel

• kE-1, Revision 5. Loss of Reactor or Secondary Coolant ES-1.1, Revision 4, Post LOCA Cooldown and Depressurization ,

1. ES-1.2, Revision 3, Transfer to Recirculation  !

• fES-1.3, Revision 0, Transfer-to Recirculation With One Safeguard Train Out of Service

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1. E-2, Revision 1, Faulted Staam Generator Isolation

• fE-3, Revision'3, Steam Generator Tube Rupture (SGTR)  !

ES-3.1, Revision 2, Post-SGTR Cooldown Using Backfill ES-3.2, Revision 2, Post-SGTR Cooldown Using Blowdown ES-3.3, Revision 2, Post-SGTR Cooldown Using Stean Dump

• fECA-0.0, Revision 2, Loss of All AC Power ECA-0.1, Revision V, Loss of All AC Power Recovery Without SI Required

1. ECA-0.2, Revision 0, Loss of All AC Power Recovery With SI Required

• fECA-1.1, Revision 1, Loss.of Emergency Cociant Recirculation

. ECA-1.2, Revision 0, LOCA Outside Containae't

• fECA-2.1, Revision 2. Uncontrolled Depressurization of Both Steam Generators

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1. ECA-3.1, Revision 4, SGTR With Loss of Reactor Coolant - Subcooled Recovery

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! ECA-3.2, Revision 5, SGTR With Loss of Reactor Coolant - Saturated Recovery ECA-3.3, Revision 2, SGTR Without Pressurizer Pressure Control

, CRITICAL SAFETY FUNCTION STATUS TREES

1. F-0.1, Revision 1, Subcriticality l

1. F-0.2, Revision 2, Core Cooling

1. F-0.3, Revision 2, Heat Sink

1. F-0.4, Revision 2, Integrity

1. F-0.5, Revision 1, Containment

1. F-0.6, Revision 2. Inventory

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FUNCTION RESTORATION PROCEDURES ,

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• 1. FR-S.1~. Revision 5, Response to Nuclear Power Generation /ATWS FR-S.2, Revision 2 Response to loss of Core Shutdown

• FR-C.1, Revision 2, Response to Inadequate Core Cooling

1. FR-C.2, Revision 2, Response to' Degraded Core Cooling FR-C.3, Revision 1, Response to Saturated Core Cooling '

#FR-H.1, Revision 2, Response to Loss of Secondary Heat Sink FR-H.2, Revision 1, Responseoto Steam Generator Overpressure FR-H.3, Revision 1, Response to Steam Generator High Level FR-H.4, Revision 1, Response to-Loss of Normal Steam Release Capabilities FR-H.5, Revision 1, Response to Steam Generator Low Level

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1. FR-P.1, Revision 2, Response to Imminent Pressurized Thermal Shock Condition FR-P.2, Revision 1 Response to Anticipated Pressurized Thermal Shock Condition .

1. FR-Z.1, Revision 1, Response to High Containment Pressure FR-Z.2, Revision 1, Response to High Sump B Level  :

FR-Z.3, Revision 1, Response to High Containment Radiation FR-I.1, Revision 1, Response to Pressurizer Flooding ,

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FR-I.2, Revision 1, Response to Low System Inventory FR-I.3, Revision 3, Response to Voids in Reactor Vessel

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i *C1.8, Revision 4, Shutdown from Outside the Control Room FS, Appendix B, Control Room Evacuation (Fire) Safe Shutdown Procedure

A desk-top review of all these procedures was perfermed as noted in Paragraphs 2.b and 3.b of this repor *A walkdown of these procedures was performed as noted in Paragraphs 2.c and

3.c of this report.

J #These procedures were exercised during the simulator scenarios as discusscd l l in Paragraphs 2.d and 3.d of this repor j i i J

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APPENDIX B LOCAL ACTION CHECKLIST PROCEDURE STEP IS THE COMPONENT SPECIFIED IN THE LOCAL ACTION ACCESSIBLE?

YES NO IF NO LIST WHAT NEEDS TO BE DONE (LADDER, PLATFORM, ETC.)

- ENVIRONMENTAL CONDITIONS DO EXPECTED RADIATION LEVELS FOR DBA OR LONG TERM RECIRCULATION PREVENT LOCATION ACTION?

YES NO DBA/RECIRC EXPECTED RADIATION LEVEL FROM F3-25 / DOES THE AREA IN WHICH THE COMPONENT IS LOCATED HAVE ADEQUATE LIGHTING (NORMALANDEMERGENCY)?

YES N0 l IS THE COMPONENT LOCATED IN A CONTAMINATED AREA?

YES N0 CAN MINIMUM SHIFT STAFFING SUPPORT THIS l0 CAL ACTION?

YES N0 APPR0XIMATE LOCATION OF COMP 0NENT (USE THE PLANT C0 ORDINATES) ANY RECOMMENDATIONS REGARDING THE ABILITY TO PERFORM LOCAL ACTIO EVALUATION PERFORMED BY DATE OF EVALUATION