ML20133L527

Forwards Listed Documents Re Onsite Emergency Planning & Safety Issues in Response to Board 851004 Order.Responses to Specific Questions Raised by Board Re Environ Qualification of Equipment Also Provided.Related Correspondence

ML20133L527
Person / Time
Site:	Seabrook
Issue date:	10/21/1985
From:	Perlis R NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD)
To:	Harbour J, Luebke E, Wolfe S Atomic Safety and Licensing Board Panel
References
CON-#485-884 OL, NUDOCS 8510240204
Download: ML20133L527 (50)

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October 21, 1985 Sg h !

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i Sheldon J. Wolfe, Esq., Chairman Dr. Jerry Harbour Administrative Judge Administrative Judge Atomic Safety and Licensing Atomic Safety and Licensing Board Panel Board Panel U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Washington, D.C.

20555 Commission Washington, D.C.

20555 Dr. Emmeth A. Luebke Administrative Judge Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C.

20555 In the Matter of PUBLIC SERVICE COMPANY OF NEW HAMPSHIRE, et al.

(Seabrook Station, Units 1 and 2) --

Docket Nos. 50-443 OL and 50-444 OL On-site Emergency Planning and Safety Issues

Dear Administrative Judges:

In an Order dated October 4, 1985, the Board requested the Staff to provide information on various matters that were incomplete at the time hearings were held in August of 1983. The Staff response to the Board's request follows. We have attached the following documents to this response:

1.

Section 4.1.4.A of Applicants' EQ Report 2.

Section 5.0 of Applicants' Emergency Plan 3.

Emergency Procedure ER-1.1 4.

Emergency Procedure ER-5.4 5.

Staff's Preliminary Review of Westinghouse Emergency Response Guidelines, Rev. 1 In response to the specific questions asked by the Board, the Staff submits the following:

1.

The Applicants submitted to the NRC Staff a document entitled

" Environmental Qualification of Electrical Equipment Important to Safety" on 8510240204 851021 PDR ADOCK 05000443 Q

PDR DSo

7.

August 13, 1983.

(Enclosure to SBN-549, Letter from DeVincentis to Knighton1/). The Applicants' environmental qualification (EQ) submittal was approximately three inches thick; Section 4.1.4.A of that document addresses the operating time component of the accident environment. A copy of that Section is attached.

The Applicants' submittal of August 13th is not their final EQ submittal. On May 7, 1984, the NRC Staff sent a Request for Additional Information to the Applicants (Letter from Knighton to Harrison); the Applicants submitted a response on September 7, 1984 (SBN-710, Letter from DeVincentis to Knighton). The Staff has been subsequently informed that the Applicants will be submitting a revised version of their August 13, 1983 submittal near the end of the year. The Staff will perform its SER review on j

the revised submittal; the Staff has not published (and because of Applicants' intention to submit a revised report, will not publish) any SER review of the Applicants' August 13th EQ submittal.

2.

On July 26, 1985, the Applicants submitted to the NRC Amendment 55 to the Seabrook FSAR (SBN-845, enclosure '.o Letter from Johnson to Knighton).

Amendment 55 contains extensive revisions to the onsite Radiological Emergency Response Plan, including a new Section 5.0.

The Staff has enclosed a copy of the new Section 5.0; the Staff review of this document should be contained in the next Supplemental Safety Evaluation Report for Seabrook, which is currently scheduled for publication in January of 1986. Applicants have also submitted Emergency Plan Implementing Procedures to the Staff.

(SBN-844,LetterofJuly 25, 1985, from Johnson to Knighton). The Staff has attached Procedures ER-1.1 (" Classification of Emergencies") and ER-5.4

(" Protective Action Recommendations"). The Staff review of the onsite plan will include a review of the procedures.

Information bearing on the l

-1/

The documents submitted by the Applicants to the NRC are identified by date and "SBN" number. According to the service list appended to these documents, copies of all SBN documents are sent to all the l

parties in the Seabrook operating license proceeding.

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'I questions asked by the Board can be found in the attached documents in the following areas:

(a) Set Points - Set points are set out in the status trees that are documented in Figures 5.1 through 5.5 at the end of the new Section 5.0.

(b) Additional Radiation Level Monitors - ER 1.1 litts initiating conditions and emergency acticn levels for miscellaneous emergency conditions. Figure 1 includes references in various locations to radiation levels as indicators of emergency cor.ditions. 2/

(c) Correlation with NUREG-0654 - The Staff has not yet seen a correlation of Applicants' assessment / classification scheme with Appendix 1 to NUREG-0654 (d) Recommended Protective Measures - This area is addressed in ER-5.4.

3.

The Staff has perfonned a preliminary review of Revision 1 to the Westinghouse Owners Group Emergency Response Guidelines; a copy of the review is attached. The Staff review of Revision 1 is continuing. Revision 1 itself is being provided to the Public Document Room in Washington, D.C.

Sincerely,

/

Robert G. Perlis Counsel for NRC Staff Attachments: As stated cc w/att:

Service List

-2/

flease note that the copy of ER-1.1 provided to the Board does not include Attachment 7.2, the Emergency Classification F;Ji Chart.

The flow chart is provided as Figure 5.6 to Section 5.0 of the 1

I onsite plan.

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-wJ ATTACHMENT 1 PUBLIC SERVICE COMPANY OF NEW HAMPSHIRE SEABROOK STATION

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ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT IMPORTANT TO SAFETY ddition, Seabrook will have in pifce a rig equipment preventativ intenance program will detect equipment deteri-oration in a*ta m

to ensure equipment performance duri normal and accident conditions. -

aging beyond that predict models a nisms utilized by rook EQ Program wi 1 ffect plant c' The Seabrook Maintenance /Skrt-lance s-

,ram is discussed further in Sectio 3N44 4.1.4 Accident Environments Each piece of equipment entered into the Seabrook Program was evaluated to determine if it would function as required during exposure to postulated accident conditions.

The specific environmental parameters evaluated are discussed in detail below.

A.

Operating Time

(

All equipment in this program which is required to function in a harsh environment will be qualified for the postulated post-accident duration of one year.

%In order to meet this one year operating time, Seabrook equipment must be qualified to the forty year normal plus one year accident total i

integrated ra'diation dose; and must demonstrate qualification in the harsh environment, it could be exposed to for at least the amount 'of time required for said environment to return to maximum normal service conditions.

This is considered sufficient to demonstrate qualification since after the environmental i

conditions return to normal the qualification of the device for the remainder of the one year operating time will be enveloped by the qualified life of the device.

As an example, a piece of equipment with a qualified life of 20 years would be replaced after 19 years because the remainder of its qualified life would not cover the postulated post-accident duration of one year.

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ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT IMPORTANT TO SAFETY i

Should it be determined that specific pieces of equipment cannot be qualified for one year, the required operating time for that component will

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be determined and the component will be qualified for at least that duration plus margin.

DBE Testing - Temperature / Pressure / Humidity equipment which could be subjected t a

ste (i.e.,

100% relative humidity) ironment durin a postulated accident han be tested to demonst te that the equipment vi functicn as requir when exposed to the ccident tem-perature, p essure and humidi conditions.

To ensure quali ' cation, the te time-temperature profile must e elope the stulated accident profile.

The te t time-essure profile should show that the tes was onducted at saturated steam conditions as minimum, in order to y/

qualify the equipme for 100% relative humidity, and tha the eak test pressure evelopes the pe postu ted accident pressure.

Comparison of time depende pressure and humidity pr iles was not co sidered necessary since the are no recognized ime dependent effects f those parameters.

was also taken i

into.c sideration that..ifethe t t temperature

. prof e envnibpes the. postulated a cident tem- ("/g

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per ures profile,1 pressure and hum ity condi-t ns would also be enveloped for sat rated eam conditions.

C Radiation Accident radiation exposure is accounted fo during equipment pre-aging as discussed in Section 4.1.3.D above.

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ATTACHMENT 2 SB 1&2 Amendment 55 FSAR July 1985 i

5.0 EMERGENCY CLASSIFICATION SYSTEM i

5.I Summary j

The Seabrook Station Emergency Classification System categorizes a wide spectrum of component or system failures and other occurrences that would reduce station safety margins. One of four emergency classifications is made upon the recognition of an initiating condition which indicates a degraded station status. Many of these initiating conditions are defined by five Critical Safety Function (CSF) color coded status trees which indicate the 4

severity of an of f-normal condition and are available to operators on the Safety Parameter Display System. Other initiating conditions are defined by quantitative or observable indications of station conditions called Emergency Action Levels (EAL's).

5.2 Emergency Classifications Seabrook Station utilizes the four emergency classifications as specified

• .n NUREG-0654/ FEMA-REP-1 (November, 1980). In order of increasing severity ::hese are: UNUSUAL EVENT, ALERT, SITE AREA EMERGENCY, and GENERAL EMERGENCY. The following definitions and descriptions of these emergency classes are used at Seabrook Station.

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UNUSUAL EVENT -

AN UNUSUAL EVENT INDICATES A POTENTIAL DEGRADATION OF STATION SAFETY MARGINS. f!O RELEASE OF RADIOACTIVE MATERIAL REQUIRING OFF-SITE RESPONSE OR j

MONITORING ARE EXPECTED.

i ALERT -

AN ALERT INDICATES AN ACTUAL OR POTENTIAL SUBSTANTIAL DEGRADATION OF STATION SAFETY MARGINS. ANY RELEASES ARE EXPECTED TO BE LIMITED TO SMALL FRACTIONS OF THE EPA PROTECTIVE ACTION GUIDELINE EXPOSURE LEVELS.

SITE AREA EMERGENCY -

A SITE AREA EMERGENCY INDICATES AN EVENT WHICH INVOLVES LIKELY OR ACTUAL MAJOR 4

l FAILURES OF STATION FUNCTIONS NEEDED FOR THE PROTECTION OF THE PUBLIC. ANY RELEASES ARE NOT EXPECTED TO EXCEED EPA PROTECTIVE ACTION GUIDELINE EXPOSURE LEVELS EXCEPT NEAR THE SITE BOUNDARY.

i GENERAL EMERGENCY -

4 A GENERAL EMERGENCY INVOLVES ACTUAL OR IMMINENT SUBSTANTIAL CORE DEGRADATION OR MELTING WITH THE POTENTIAL FOR THE LOSS OF CONTAINMENT INTEGRITY. RELEASES CAN BE REASONABLY EXPECTED TO EXCEED EPA PROTECTIVE ACTION GUIDELINE EXPOSURE LEVELS OFF-SITE FOR MORE THAN TIIE IMMEDIATE AREA.

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.fuly 1985 5.3 Scope of Classification System The classification system for Seabrook Station provides the ability to classify approximately sixty discrete symptom-based or miscellaneous events.

The system considers and classifies events specified in Appendix 1 of NUREG-0654/ FEMA-REP-1; upset conditions defined by the Critical Safety Functions, and the discrete acciients contained in the Seabrook Station Final Safety Analysis Report, Chapter 15, Accident Analysis.

5.4 Symptomatic Approach to Classification A symptomatic approach has been developed to assist operators in emergency reengnition and classification. Critical station process data are condensed on color-coded status trees which allow the operator to recognize an off-normal condition and take appropriate actions. Symptomatic status trees are available to the operator and at the emergency response facilities on the plant process computer displays and on hardcopy.

A wide spectrum of events that represent varying degrees of safety margin reduction are illustrated on the color-coded status trees. The status trees (Figures 5.1 - 5.5) are based on the following five Critical Safety Functions:

S - Subcriticality C - Core Cooling 11 - Heat Sink P - RCS integrity Z - Containment Integrity Color coding is used to identify event priorities for the individual branches of the status trees as follows:

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CREEN - The Critical Safety Function is satisfied - No operator action is

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called for.

YELLOW - The Critical Safety Function is not fully satisfied - Operator action may eventually be needed.

ORANGE - The Critical Safety Function is under. severe challenge - Prompt operator action is necessary, i

RED -

The Critical Safety Function is in jeopardy - Immediate operator action is required.

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If a status tree is coded in a color other than green, the control room operators will take corrective action consistent with the Emergency Operating Procedures. In addition, if a status tree (or combination of status trees) is in a condition other than green, the Shift Superintendent will use the Emergency Classification Flowchart (Figure 5.6) to determine whether.an be declared.

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SB 1 & 2 Amendment 55 FSAR July 1985 Figure 5.6 presents the critical safety functions in descending order of importance as one reads down the figure.

If more than one classification is reached, the emergency will be classified according to the most severe.

3 5.5 Hiscellaneous Station Conditions The capability also exists for the classification of emergencies based on conditions that do not challenge a Critical Safety Function. Based on the guidance of Appendix 1 of NUREG-0654/ FEMA-REP-1, niscellaneous emergency conditions (e.g., fire, electrical, security, natural events) have been evaluated, initiating conditions identified and Emergency Action Levels developed. The specific miscellaneous initiating conditions are indicated on Figure 5.6.

In some cases a combination of miscellaneous conditions or a complication of a miscellaneous condition with a critical safety function are an indication that an emergency classification has been reached. These combinations and compli-cations are also on Figure 5.6.

5.6 Classification of Emergencies Classification of an emergency at Seabrook Station is made based on one or more of the conditions listed in Figure 5.6.

Specific EAL's (color status t ree s, meter indications, alarms, or limits) for initiating conditions are provided in an emergency response procedure and in operator training.

In all cases, if several emergency classifications are indicated, the most severe emergency classification will be made whether based upon status trees or miscellaneous initiating conditions.

5.7 Sample Classifications To ensure understanding of the emergency classification system, the following sample classifications are presented. These examples explain t' process by which the operators would come to the decision to classify an

rgency.

EXAMPLE 1 - Condition - Critical Safety Function Core Cooling (Figure 5.2) indicates orange.

First locate C, Core Cooling under the Critical Safety Function column on the left of Figure 5.6.

Then moving to the right, find C Orange under the 1

appropriate, emergency class, Site Area Emergency.

EXAMPLE 2 - Condition - Critical Safety Functions, Heat Sink (Figure 5.3) indicates red, and Core Cooling (Figure 5.2) indicates orange.

i Combinations of separate Critical Safety Function indicators sometimes warrant i

a higher level emergency classification. First locate C, Core Cooling under the Critical Safety Function column on Figure 5.6.

Moving to the right, find C Orange (Site Area Emergency), then C Orange plus H Red (General Emer-gency). Then locate H, Heat Sink. Moving to the right, find H Red (Site Area Eme rgency). Using the most severe classification, select General Emergency.

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-Cr SB 1 & 2 Amendment 55 FSAR July 1985 EXAMPLE 3 - Condition - Critical Safety Function Heat Sink (Figure 5.3) indicates red and emergency power is not restored to at least one train of operable safeguards equipment within 15 minutes.

i Complications of other miscellaneous emergency conditions along with Critical l

Safety Function indicators may also warrant increased levels of emergency classification. From Example 2 recall that H Red indicated a Site Area

'j Emergency.

To consider the electrical problem, locate category 6.

Electrical Failures under the Miscellaneous Emergency Conditions column on the lef t of Figure 5.6.

Moving to the right, locate condition 6e (Site Area Emergency). To consider the complication, follow the Heat Sink line to the right and find H Red plus 6e (General Emergency). Using the most severe classification, select General Emergency.

EXAMPLE 4 - Condition - Indication of a steam generator tube rupture by procedure E-3.

First locate the category of condition, Steam Generator Tube Leakage / Rupture under the column labeled Miscellaneous Emergency Conditions on Figure 5.6.

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Moving to the right, locate condition 7b (Alert). The condition is classified as an Alert.

EXAMPLE 5 - Condition - Indication of a steam generator tube rupture by procedure E-3 and bus E-5 and E-6 cannot be powered f rom an of f-site source.

First locate the classification for the steam generator tube rupture, 7b, as in Example 4 (Alert). Then locate the category, Electrical Failures, and move to the right to condition 6a (Unusual Event). Following either category 6 or 7, find the combination 6a plus 7b (Site Area Emergency). Using the most severe classification, select Site Area Emergency.

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GREEN PUBLIC SERVICE COMPANY OF NEW HAMPSHIRE STATUS TREE FOR CRITICAL SAFETY FUNCTION l

SEABROOK STATION - UNITS 1 & 2 NUMBER S - SUBCRITICALITY I

FINAL SAFETY ANALYSIS REPOR T RADIOLOGICAL EMERGENCY PLAN l

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Amendment 55 July 1985 RED RED N

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13% 1 RCP YELLOW GREEN PUBLIC SERVICE COMPANY OF NEW HAMPSHIRE STATUS TREE FOR CRITICAL SAFETY FUNCTION SEABROOK STATION - UNITS 1 & 2 NUMBER C - CORE COOLING FINAL SAFETY ANALYSIS REPORT RADIOLOGICAL EMERGENCY PLAN l

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GREEN PUBLIC SERVICE COMPANY OF NEW HAMPSHIRE STATUS TREE FOR CRITICAL SAFETY FUNCTION SEABROOK STATION - UNITS 1 & 2 NUMBER 4 - HEAT SINK FINAL SAFETY ANALYSIS REPORT RADIOLOGICAL EMERGENCY PLAN l

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GREEN PUBLIC SERVICE COMPANY OF NEW HAMPSHIRE STATUS TREE FOR CRITICAL SAFETY FUNCTION SEABROOK STATION - UNITS 1 & 2 NUMBER P - RCS INTEGRITY FINAL SAFETY ANALYSIS REPORT l

RADIOLOGICAL. EMERGENCY PLAN l

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CLASSIFICATION OF EM12GENCIES

~ l 1.0 OBJECTIVES This procedure specifies the classification of emergencies in accordance with the Seabrook Station Radiological Eeergency Plan."

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2.0 RESPONSIBILITIES 2.1 Unit Shif t Supervisor 1

Responsible for recognizing potential esecgency conditions and notifying the Shif t Superintendent. Should the Shif t Superintendent be unable to respond to the affected' control room in a prompt manner (5 minutes), the Unit Shif t Supervisor will assume the duties and responsibilities of the Shif t Superintendent.

2.2 Shif t Superintendent Responsible for classifying observed station conditions in accordance with the emergency.glassification system specified in this procedure.

t 2.3 Emergency Director Responsible for analyzing chaNss$ station conditions and in reclassifying the emergency c tfication in accordance with this procedure.

3.0 PRECAUTIONS

' j 3.1 When two or more critical safety functions (CSF) or energency con-ditions exist and differer.t emergency classes result, the higher emergency class will be used.

3. 2 Final emergency classifications and emergency plan procedures imple-mentation is contingent upon the evaluation and discretion of the Shif t Superintendenc.

4.0 PREREQUISITIES A CSF has been challenged and/or one of the miscellaneous emergency con-ditions has occurred.

5.0 ACTIONS Shif t Superintendent / Emergency Director f

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m _. ; __ _ a._ ____;n.

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ER-1.1 i

Prg3 3 of 43 Esv. 00

^-

5.1 Critical Safety Function Classification 1.

Determine and verify using hardwired information whethee any of the following critical safety functions (CSF) are challenged:

S.

Suberiticality, C.

Core cooling, H.

Heat sink, yl F.

RCS integrity, and

,j Z.

Containment integrity

^

NOTE Following initial verification of a cr'itical safety function, j

subsequent SPDS indications need not be verified.

i 2.

If no CSF is challenged, proceed to Step 5.

1 3.

Identify the color coded event for the challenged CSF(s') from either:

}

.a.

The plant computer SPDS, or 1

b.

If the SPDS is unavailable, refer to Emergency Operating i

Procedures hard copy Status Trees.

d 4.

Circle the letter and color of each CSF event or combination of i

i events identified in step 3. on Form ER-1.1A. " Emergency 1

Classification Flow Chart".

NOTE Only CSF combinations which require emergency classification are listed on Form ER-1.1A.

5.

Circle the Miscellaneous Emergency Conditions and combinations of Miscellaneous Emergency Conditions which correspond to actual station conditions on Form ER-1.1A.

NOTE Only siscellaneous combinations which require emergency classifi-cation are listed on Form ER-1.1A.

NOTE Emergency Action Levels pertaining to specific initiating con-dicions are described in Figure 1, " Miscellaneous Emergency Condition EALs".

~ i 6.

Circle any complications of Miscellaneous Emergency Conditions and Critical Safety Functions which correspond to actual station conditions on Form ER-1.1A.

l 7.

Identify the most severe emergency class which corresponds to the events circled on Form ER-1.1A.

, W,.

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m___.._.__.

d' ER-1.1 Peg 2 4 of 43

~ -

Rev. 00 8.

Evaluate, and if appropriate, immediatelyimplementeneofthe(

f ollowing Seabrook Station Radiological Emergency Plan Implementing Procedures in accordance with the identified emergency classification.

UNUSUAL EVENT

. ER-1.2 j

ALERT ER-1.3 SITE AREA EMERGENCY ER-1.4 i

GENERAL EMERGENCY ER-1.5

^

J m;

d 6.0 _ REFERENCES

- 4 4

6.1 Seabrook Station Radiological Emergency Response Plan.

i 7.0 ATTACHMENTS 7.1 Figure 1, " Miscellaneous Emergency Conditien and Emergency Action Lr.<els ".

7.2 l'orm ER-1.1 A, " Emergency Classification Flow Chart".

I

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ER-1.1 Pegs 5 of 43 Rev. 00 Page 1 of 39

(. :.. <

Figure 1 MISCELLANEOUS EMERGENCY CONDITIONS AND EMERGENCY ACTION ',EVELS Initiating Condition i

Busses E5 and E6 cannot be powered from an offsite source.

6a.

'1 LC EMERGENCY ACTION LEVELS Offsite power is not being supplied to both busses E5 and E6 and cannot be made availab le.

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Figure L

.,'ls:

Initiating Condition 6b.

No emergency diesel generator can power operable required safeguards equipment.-

EM EGENCY ACTION M YELS 1.

Soth indicating 1is hts t

MCB-HF UL-19 "A"

diesel no t available UL-20 "B" diesel not available OR

.J 2.

Bo th VAS alarms, D or F Point V.,

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f ca*:.4 F6587 Train A emergency power inop F6637 Train B emergency power inop a

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Pago 7 of 43 Rev. 00 Page 3 of 39 Figure L Initiating Condition I

i 6c.

14ss of both trains of emergency AC power VITH power restored to one train of operable safeguards equipment within fif teen minutias.

j i

EMERGENCY ACTION LEVETJ l

L.

Both Hardwire Alaras MCB-HF UA 54 4160V Bus 5 Volts Lo UA 55 4160V Bus 6 Volts Im

_OR 2.

Both VAS Alarms D or F Point Me ssag e F7300 Bus E5 Loss of Power

..,g 77310 Bus Z6 Loss of Power AND 3.

One alarm has cleared within fif teen minutes.

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ER-L.1 Pego 8 of 43 Rev. 00 Psge 4 of 39

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Figure 1 t

l Initiating Condition I

6d.

Slaultaneous loss of vital DC Busses 11A, 11B, llc and LlD for less than 15 sinute s.

d "0 l )

EMERGENCY ACTION LEVELS l.

All he Following Hardwire Alarms fo'r less than 15 minutes MCB-HF UA 54 DC Bus llA Volts Lo DC Bus 11C Vo

• t.y to UA 55 DC Bus llB Vol ts la DC. Bus 11D Volta Lo 3

2.

All he Following VAS Alar:ss for less than 15 minutes D Poin t Me ssag e D6094 125V DC Bus 11A Volts 14-Lo D6095 125V DC Bus 11C Volts Lo-Lo D6096 125V DC Bus 115 Volts Lo-Lo D6097 125V DC Bus llD Volts Io-Lo SK l

3.

A Ioss Of Voltage Condition As Indicated By All he Following Voltmeters For Iass han 15 Minutes MCB-HR 1-EDE-VM-9750 Ba ttery Bus llA Vol tag e 1-EDE-VM-9752 Ba ttery Bus 11B Vol tag e 1-EDE-VM-9754 Battery Bus LLC Vo l tag e 1-EDE-VM-9756 Ba ttery Bus llD Vol tag e I

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i DL-L. L' Pmgo 9 of 43 13v. 00 Page 5 of 39 Figure 1 Initiating Condition 6e.

Faergency power g restored to at least one train of operable safeguards equipment within L5 minutes.

EMERGENCY ACTION MYEM L. Both Hardwire Alares MCB-HF UA'54 4160V Bus 5 Volts to UA 55 4160V Bus 6 Volts Lo NOrt:

If one of these has cleared (light out), a bus is reenergized.

OR 2.

Bo th VAS Alarms D or F Point Me ssat e F7300 Bus ES Loss of Power F7310 Bus E6 Loss of Power AND 3.

No alars has been cleared within fif teen sinutes.

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Figure L Initia ting Co ndition

~

6f. ' Simultaneous loss of vital DC Busses LLA, ILB, LLC and 11D for more than 15

'sinute s.-

EMERGENCY ACTION LEVELS 1.

All The Following Hardwire Alarms For More Than 15 Minutes MC3-HF UA 54 DC Bus LLA Volts Lo DC Bus llc Volts Lo UA 55 DC Bus LLB VoltsLo DC. Bus llD Volts Lo OR 2.

All The Following VAS Alarms For More Than 15 Minutes j((f D Poin t Me s sag e D6094 125V DC Bus LLA Volts Lo-Lo D6095 125V DC Bus Llc Volts Lo-Lo D6096 tzav pu aus LLs voit3 to-Lo D6097 125V DC Bus LLD Volts Lo-Lo g

3.

A Loss Of Voltage Condition As Indicated By All of The Following Voltme ters For More Than LS Minutes MC3-HR l-EDE-VM-9750 Ba ttery Bus LLA Vol tag e 1-EDE-VM-9752 Ba ttery Bus 113 Vol tag e L-EDE-VM-97 54 Ba ttery Bus 11C Vol tag e

~

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w. = c ::w

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G9 ER-1.1 Pag 3 11 of 43 Rev. 00 Page 7 of 39 k-A Figure 1 Initiating Condition 7s.

Primary to secondary leakage greater than 500 gal per day E steam genera-tor specific activity greater than 0.1 pCi/cc dose equivalent I-131.

EMERGENCY ACTION LEVELS 1.

a) Any one of the following alarms:

RDMS CRT Monitor Indication Equipment Tag #

Channel Displav #

Steam Generator Hi Alarm 1-RM-RM-6510 ILM211 Blo-aown 1 Stee.a Generator Hi Alarm 1-RM-RM-6511 LLM212 Blowdown 2 Steam Generator Hi Alarm 1-RM-RM-6512 ILM213 Blowdown 3 Steam Generator Hi Alarm 1-RM-RM-6513 IL4214 Blowdo'wn 4 Steam Generator Ei Alarm 1-RM-RM-6519 IL4215 Blowdown Flash Tank Drain Condenser Air Hi Alarm 1-RM-RM-6505 1GM810 Evac O,,R b) Any other indication of primary to secondary leakage.

AND 2.

Analysis of a blowdown liquid sample by the chemistry department indi-cates activity greater than 0.1 uCL/cc dos,e equivalent I-131.

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Page 12 of 43 Rev. 00 Page 8 of 39 Figure L Initiating Condition 7b.

Identification of steam generator tube rupture by Emergency Procedure E-3.

EMERGENCY ACTION LEVELS Indication of a Steam Generator tube rupture may be determined by any of the following:

1.

Condenser ef fluent radiation above normal.

_OR 2.

Steam line radiation above normal.

OR 3.

Steam Generator blowdown line high radiation.

..:~~.

OR s.y 4.

High radiation in a Steam Generator sample.

OR 5.

An unexpected rise in Steam Generator level.

S

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ER-1.1 Pcgo 13 of 43 Rev. 00 Page 9 of 39

.'^ '

Figure 1 Initiating Condition 8a.

Coolant activity sample greater than 20 uC1/cc gross activity (y).

DIERGENCY ACTION LEVELS 1.

a) Latdown Monitor Im Alarm Equipment Tag #

RDMS CRT Display Channel #

l-RM-RM-6520-3 LLM 240

_OR b) Any indication of potential fuel clad damage.

g 2.

Analysis of a reactor coolant sample by the ebenistry department indi-cates gross activity (T) greater than 20 pC1/cc.

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ER-L.1 Page 14 of 43 Rev. 00 Page 10 of 39 Figure 1 Initiating Condition 1

8b.

Coolant activity sample greater than 200 uC1/cc groes activity (y).

LMERGENCY ACTION LEVELS 1.

a) Latdown Monitor Hi Alarm kquipment Tag #

RDMS CRT Display Channel i 1-RM-RM-6520-3 ILM 240 OR b)

Any indication of potential fuel or clad damage.

OR 2.

Analysis of a reactor coolant sample by the chemistry departasent indi-

' Ja catas gross activity (Y) greater than 200 pCi/cc.

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ER-L. L Pego 15 cf 43 Rev. 00 Page 11 of 39 i

r Figure L Initiating Condition 9a.

Loss of VAS and hardwired alarm systems.

EMERGENCY ACTION LEVELS 1.

VAS loss indicated by:

a.,

Sta tic time on CRT screen DR b.

Blank VAS display screen ANL 2.

The annunciator power failure lights on each of the following annunciators indicaJe loss of 120V AC power supply.

UA-50 MCB-BF UA-51 MCB-BF UW UA-52 MCB-DF 2;.1+ #

UA-53 MCB-FF UA-54 MCB-HF UA-55 MCB-HF A

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ER-1.1 i

Pcgo 16 of 43 Rev. 00 Page 12 of 39 Figure 1 Initia ting Condition 9b.

Failure of qualified RMIS panel (CP-lS0).

EMERGENCY ACTION LEVELS Both lE panel' Trains A and B represented on the RDNS CRT indicata Blank or erroneous LED displays ee

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IA-1.1 Pcgo 17 of 43 I

Rev. 00 Page 13 of 39 Figure L

)

Initiating Condition

..J

'7 9c.

VAS _and hardwired alarm system not functional for more than two hours.

Stable plant conditions existing throughout the two-hour restoration M

period.

s 1 l

EMERGENCY ACTION LEVEfJ l'.

VA3 Loss indicated by:

a.

Static time.

OR b.

Blank VAS display screen.

AND 2.

The annunciator power failure lights on each of the following annunciators indicate loss of 120V AC power supply.

UA-50 MCB-BF UA-51 MCB-BF UA-52 MCB-DF UA-53 MCB-FF UA MCB-HF UA-55 MCB-HF t

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AND d

1 3,

Two hours have elapsed.

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'i Pego 18 cf 43 Rev. 00 Page 14 of 39 Figure 1 i

Initiating Condition 3

-1a 9d.

Loss of VAS and hardwired alarm system with an abnormal traustent in progress.

'i h

EMERGENCY ACTION LEVELS q

1.

VAS Loss indicated by:

a.

Sta tic time.

OR b.

Blank VAS display screen.

AND 2.

The annunciator power failure lights on each of the following annunciators indica _te loss of 120V AC power supply.

UA-50 MCB-BF UA-51 MCB-BF UA-51 MCB-DF 1,f.,.

UA-53 MCB-FF UA-54 MCB-HF UA-55 MCB-HF AND i

3.

An abnormal transient in progress.

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Pcgo 19 of 43 i

',j Rev. 00 Page 15 of 39 j

i J.

Figure 1 1

Initia ting Condition

.l 10.

Loss '*of all communications capability.

q INERGENCY ACTION LEVELS I

Completa loss of all the following control room communications.

a. Telephones

b. Radios

c. Gaitronics I

d. Sound Powered Phones

];

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-l Pcgo 20 of 43 Rev. 00 Page 16 of 39

)

i Figure 1 Initiating Condition i

j

/;

i 11.

Shutdown technical specifications surpassed.

IMERGENCY ACTION LEVELS

' Initiation of shutdown to the cold condition as required by technical specifications.

N(rEE Initiation of shutdown does g include preparatory actions such as load reduction preceeding the decision that a shutdown aust be performed.

9 a

a I

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ER-1.1 Pego 21 cf 43 Rev. 00 i

Page 17 of 39

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Figure 1 Initiating Condition

.a 1

f lla. Radiological effluents exceed T'achnical Specification 3.11.2, Gaseous Effluents Instantaneous Limit.

EtERGENCY ACTION LEVELS

- :)

The results of chemistry department analysis of gaseous effluent release s.

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Ptgo 22 of 43 Rsv. 00 Pag e 18 of '39 Figure i s.,

Initiating Condition 12b. Radiological effluents greater than 10 times Technical Specification 3.11.2, Gaseous Effluents Instantaneous Limit.

. 1 DIERGENCY ACTION LEVELS I

The results of chemistry departinent analysis of gaseous effluent releases.

I6 2

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IR-1.1 Pcgo 23 of 43 Rty. 00 Page 19 of 39 Figure 1 k,;.j*

s.u.

Initiating Condition 12c. Indications tha t Environmental Protection Agency Pro tective ' Ac tion Guides (EPA PACS) are projected to be exceeded at the site boundary.

EMERGENCY ACTION LEIELS 1.

a) Any one of the following alarms:

RDNS CRT LE Panel Monitor Indication Equionent Tag 4 Channel Display #

(CP-180)

Recorder Plant Vent Hi-Hi Alarm 1-RM-RM-6533-3 IlMI226 1-RM-RK-6533 1-RM-RR-6 53 3 Air Radiogas (Channel 3)

Main Steam Line Hi Alarm 1-EM-RM-6481-1 LGM801 Monitor Imop 2 Main Steam Line Hi Alarm 1-RM-EM-6481-2 LGM802 Monitor Loop 3

('M Tiain Steam Line

.. f..

Hi Alarm 1-RM-RM-6482-1 1GM803 Monitor Loop.1 Main Steam Line Hi Alarm 1-EM-RM-6482-2 IGM 804 Monitor Loop 4 Cont Hi Range Hi Alarm 1-RM-EM-6 5 76A '

1AM106 1-RM-RK-6576A 1-RM-RR vj7t Post LOCA 1-EM-EM-6576B 1AM107 1-RM-RK-6576B l-RM-RR-65 7t AND b) Results of ER-5.3, "Off-site Dose Es timatas" which proj ec t doses a t site boundary to exceed 1 rem whole body.

E 2.

Offsite monitoring results indicate doses at the site boundary which exceed 1 rem whole body.

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=d-ER-t.t Pego 24 of 63 Rev. 00 Page 20 of 39 f-Figure 1 l ~..

x. ;7 Initia ting Condition j

L2d. Indications that projectad' or measured does rates at. site boundary exceed 1 res/hr.

s 1

IMERGENCY ACTION LEVELS 11 l.- a) Any one of the following alarms:

RDMS CP.T LE Panel Monitor Indica tion Equipment Tag # Channel Display 4 (CP-180 )

Recorder Plant Vent Hi-Hi Alarm 1-RM-RM-6 533 -3 IlMI226 l-RM-RK-6533 jl Air Radiogas (Channel 3)

.1i Main Steam Line Hi Alarm 1 -RM-RM-6481-1 LGM801

/lj Monitor Loop 2 ej Main Steam Line Hi Alaru l-RM-RM-6481-2 LGM302 l

Monitor Loop 3

,3

. c.;;.

Main Stesa Line Hi Alarm 1-EM-RM-6482 -1 1CM803 Monitor ' Loop 1 Main Steam Line Hi Alars 1-EM-RM-6482-1 LGM804 Monitor loop 4 Cont Hi Range Hi Alarm 1-95 =w A5764 L" 10' 1 a w "-6 5 7' A 1-93-? n -6 5 7 Post IDCA l-RM-EM-65768 1AM107 1-RM-RK-65768 1-RM-RR-657 l

MD 2.

Results of ER-5.3,. "Of f-site Dose Estimates" which proj ect dose rates at site boundary to exceed 1 ram /hr whole body.

1 OR

~

t 3.

Of fsite monitoring results Indicate dose rates at tha sLte bNn-dary which exceed 1 res/hr whole body.

i 3

Qy/

l l

1

(

+

ei.

Peg 2 25 of 43 ER-1.1 3

Rsv. 00 Page 21 of 39 Figure 1 Initiating Condition

13. Fuel handling accident with release of radioactivity.

-i

^ Ij EMERGENCT ACTION LEVELS "i

1.

Sh'ift Superintendent has received notification from the fuel handling j

building that the following has occurred, resulting in fuel damage.

-i a) Dropping, bumping or otherwise rough handling of an i

irradiated fuel bundle.

E

,'i 1

b) Dropping of a heavy object onto irradiated fuel.

';{

)

AND aj 2.

a) Containment Manipulator Crane Radiation Detectors and Monitor Hi-Hi alarms I

- fa Equipment Tag i RDMS CRT Display Channel #

lE Panel (CP-180) 1-RM-RM-6535A 1AM102 RK-5535A 1-RM-RM-65355 1AM103 RK-65358 9.R l

4 b) Spent Fuel Pool Area Monitor Hi-Hi alarms Equipment Tag #

RDMS CRT Display Channel #

IE Panel (CP-180) 1-RM-RM-6549 1AM402 Local alarm j

t 9

y a

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.'j ER-1.1 Pego 26 ef 43 Rev. 00 Page 22 of 39 e

f Fig,ure L

(

Initiating Condition I

14.

Rasctor trip or safety injection without a return to normal plant proco-si dures.

1 1

MERGENCY ACTION LEVELS A determination by the Shif t Superintendent that there has been:

1.

Reactor trip response (ES-0.1) WITHOUT return to normal plant procedures.

_Olt 2.

SI termination (ES-L.1) VITHOUT return to normal plant procedures.

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ER-1.1 Pege 27 of 43 Rev. 00

' ' Paga 23 of 39

/

k. ~

Figure 1 Initia ting Condition 15.

Loss of emergency coolant recirculation (ECA-1.1) or IDCA outside of con-tainment (ECA-1.2) with fuel desage imminent.

EMERGENCY ACTION LEVELS 1.

Initiation of one of the following emergency procedures:

a)

ECA-1.1 Emergency Coolant Recirculation

_OR b)

ECA-1.2 LOCA Outside of Containment AND 2.

Indications of imminent fuel damage.

/

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e le gh e g og v

4

-.-m o.--

a Qtij' -

ER-1.1

)

.1 Page 28 of 43

)

Rev. 00

~

Page 24 of 39

. Figure 1 Initiating Condition 16a. Fire in the protected area lasting more than 10 minutes WITH a safe shut-down capability maintained.

5 EMERGENCY ACTION LEVELS 1.

A fire is observed in the protected area burning for more than 10 minutes.

_OR 2.

a)

Fire detection alarm on Main Control Board Section HR.

AND b) Direct confirmation that the fire is still out of control 10 minutes af ter the alarm.

.M.-

AND

3. " A determination by the Shif t Superintendent tha t sait shutdown capability is maintained.

i i

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ER-l.1 Pego 29 of 43

^ '

Rev. 00 Page 25 of 39 I

Figure 1 Initiating Condition 16b. Controlled, contained fire, potentially affecting safe ty systees 91TM safe shutdown capability maintained at either train associated safe shutdown j

Panel.

DIERGENCY ACTION LEVEL Observation of fire through fire detection equipesnt, operation of automa-tic fire fighting systems and fire fighting teams which the Shif t Superintendent deter: mines has the po tential to affect safety systems.

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ER-1.1 Pago 30 of 43 Rev. 00 Page 26 of 39 Figure 1 Initiating Condition 16c. Uncontrolled, uncontaiaed fire affecting safety related equisent WITH safe shutdown capability maintained at one train, BUT may be affected by the J}

fire.

EMERGENCY ACTION LEVELS 1.

Observation of a major fire by fire detection equipment, operation of automatic fire fighting systems or fire fighting team s.

AND 2.

A deter:sination by the Shif t Superintendent that the safe shutdown capability is maintained at one train but may be affected by the fire.

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Initiating Condition

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16d. Safe shutdown capability degraded or lost due ta fire. ~

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EMERGENCY ACTION LEVEIS i

A determination by the Shift Superintendent that safe shutdown capability is degraded or lost due to fire.

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Initia ting Condition 17a. Control room evacuation anticipated or required with safe shutdown cap-ability established.

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s ENERGENCY ACTION LEVELS A determination'by the Shif t Superintendent tha t:

1.

Evacua tion of the control room is anticipa ted or required.

AND 2.

Safe shutdown capability has been established.

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El1-1.1 Pege 33 of 43 Rev. 00 i

Page 29 of 39 q-Figure 1 Initia ting Condition 17b. Control r'oom evacuation with safe shutdown capability established BUT station conditions are unstable or uncontrolled.

1 DIERGENCY ACTION LEVELS A determination by the Shif t Superintendent that:

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The control room evacua tion has been initiated or completad.

AND 2.

Control of systems and equipment needed for safe shutdown has been established.

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Cascable or uncontrolled station conditions warrant additional pre-

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R-L. L Ptgo 3$ of 43 Rev. 00 Page 30 of 39 Figure L Initiating Condition 17c. Control room evacuation with safe shutdown capability NOT established within 15 minutes.

MERGENCY ACTION LEVELS A dataraination by the Shif t Superintendent, that the following conditions exist:

1.

Control room evacuation has been completed.

AND 1.

Control of systems and equipment needed for safe shutdown has not been established within 15 minutes of the control room evacuation.

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ER-1.1 Pass 35 of 43 Rav. 00 Page 32 of 39

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Figure 1 Initiatin: Condition 18a. Hazards experienced or projected NOT affecting safety systems.

EMER0ENCY ACTION LEVEI.S

~

Naturai Phenomena 1.

Response spectrum seismic unit triggered.

Station computer alarm indi-cated on VAS.

D Point Message D5452 Seismic Instrumentation Unit Trigger 9E j

2.

Tornado observed do strike the site.

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OR C.~m 3.

A hurricane that strikes the site.

Man Made Events Notification of the Shift Superintendent that there has been:

1.

a) An aircraft crash onsite, OE b) Unusual aircraft activity over the site.

OR 2.

Train derailment onsite, OR 3.

Near site or onsite explosion, 9.E 4.

Near or onsite toxic or flammable gas release.

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DL-L. L Prgo 36 of 43 Rev. 00 Page 31 of 39 ligure 1 Initiating Condition 18a. (Con't)

Security Events A determination by the Shif t Superintendent or notification by the Security Supervisor that there is or wast 1.

A security threat in a vital area, OR i

2.

h attempted entry in a vital area,

_OR 3.

An attempted sa5otage in a vital area.

Discretionary Events Shif t Superintendent discretion that an event is in progress or has occurred which indicates a potantial degradation of the level of safe ty of the sta tion.

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.. IR-1.1 Page 37 of 43 Rev. 00 Page 33 of 39

.. :n Figure 1 Initia ting Condition 18b. Hazards experienced or projected which may affect safe ty systems.

EMERGENCY ACTION LEVELS Na tural Phenomena 1.

Response spectrum seismic unit indicates earthquake greater than con-tainment CBE levels on:

VAS Alarm D Point Ma s sag e D5451 Containment foundation OBE

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2.

Tornado striking and danaging safa ty aeruc tures.

_OR 3.

Severe weather (sustained winds exceeding 90 aph) which may affect safety systens as indicated by:

VAS Indicators A Po in t Ma s sag e A1626 Me t Tower Upper Wind Speed A1628 Met Tower Lower Wind Speed Man Made Events 1.

Shif t Superintendent observes or is notified of an aircraf t crash or missile inpact on plant structures or components.

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2.

Shif t Superintendent observes or receives notification of station l

damage caused by explosion.

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a) Shif t Superintendent receives notification of gasses in con-

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centrations which exceed the limits of toxicity within the station environs.

1 og b) Portable sampling equipment measures dangerous concentrations of toxic gas within the sta tion.

OR i.

c) Gaseous concentrations exceed the limits of flamability within the station environs.

i OR

.I d)

Flammable or explosive gases are detected by portable sampling

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methods in any plant structure containing safety related equipment.

Security Event An on-going security compromise in a vital area.

Discretionary Events Shif t Superintendent discretion that an event is in progress or has occurred *which involves an actual or potential substantial degradation of the level of safety of the station.

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Figure 1 N

Initia ting Condition 18c. Hazards experienced or proj ec ted which compromise safe ty systems.

IMERGENCY ACTION LEVELS

-j Matural Phenomenon i;4

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1.

Earthquake with potantial impact on SSE, aa indicatad by; 4

VAS AIARM s

D Poin e Mes sag e a) D5451 Containment Foundation OBE.

AND

.i b) Verification of SSE on field response recorders.

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3

. c..g a) Sustainad wind speeds of 100 mph as indicated by; 7AS l

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Ma s sag e A1626 Met Tower Upper Wind Speed A1628 Met Tower Iower Wind Speed OR b) Winds of short duration with estimated speeds in excess of 360 mph.

Man Made Events No tification to the shif t superintendent tha t there has been:

1.

An aircraf t crash causing damage or fire in any vital structures.

~ OR 2.

Severe damage to safe shutdown equipment from missile or explosion.

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OR

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Entry of uncontrolled flansable gases into vital areas.

Ea t:y o f uncontrolled toxic gases into viral areas where lack of access to the.

t area constitutes a safe ty problem.

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-t li3L-1.1 Page 40 of 43 Rev. 00 Page 36 of 39 Figure L Initia ting Condition 18c. (Con' t) b Security Event i

A determination by the Shif t Superintendent or notification by the Security Supervisor that a physical attack on the station is in progress which will result in imminent occupation of vital areas.

Discretionary Events Shif t Superintendent discretion that an event is 'in progress or has occurred which involves actual or likely major failures of station func-tions needed for protection of the public.

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Rev. 00 Page 37 of 39 Figure 1 Initiating Condition 18d. Hazards experienced or projected which result in loss of physical control cf the facility.

'l EMERGENCY ACTION LEVELS Na tural Phenomenon A detarzination by the Shif t Superintendent tha t there has been or will be major intarsal or external events which could cause sassive connon damage to plant sys tems resulting in any of the General Energency initiating con-ditions.

Security Threats A determination by the 5hif t Superintendent or notification by the Se.cu,rit'y Supervisor tha t a physical attack on the plant has resulted in unauthorised personnel occupying a vital area.

Discretionary Events t

Shif t Superintandent discretion that an event is in progress or has occurred which involves actual or imminent substantial core degradation or zelting with potential for loss of containment integrity.

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ER-L.L Page 42 of 43 Rev. 00 Page 38 of 39 I

Fig ure L

- ~ l Initiating Condition 19.

Energency transport of contaminated and injured person to local support

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9 EMERGENCY ACTION LEVELS 4

j No tification of the Shif t Superintendent tha t a contaminated and injured persea has been transportad to a local hospital.

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P:ge as of as a:v. oo Page 39 of 39

.y; I Figure 1 Initiating Condition I

2o. Failure to isolate containment requiring shutdown by Technical Specifications.

'1 MERGENCY ACTION LZYELS

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1.

Failure of Phase A valves to isolata af ter receiving a 'T' signal as indi-

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cated by Phase A safegt.ard status panel and valve (s) can not be closed by i

alternata means.

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Failure of Phase 3 valves to isolate af ter receiving a 'P' signal as indi-cated by Pb.sre B safeguard status panel and valve (s) can not be closed by alternata means.

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ATTACHMENT 4 ER-5.4 Pcge 1 of 10 Rev. 00 6

EMERGENCY PLMI IMPLEMENTING 1ROCEDURE COVER FORM I

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IDERIFICATION EmmER n-5.4 REVISION 00 t

TITLE PROTICTIVE ACTION RECOMMENDATIONS CRIGINATOR A. M. Callendrello 3.

INDEPENDENT REVIE*J

• ?. TITLE SIGNAIURE DATE

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RADIOLOGICAL ASSESSMENT MANAGER AP? ROV.U.

SIGNATURE DATE h) //

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SORC APPROVAL SCRC MEETING NO..

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APPROVAL AND IMPLEME WATION il W CL/25/85 x%

e STATIV4MddgKGER APPROVID DATE ETTECTIVE DATE 9

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ER-5.4 Paga 2 cf 10 Env. 00 PROTEC"I7E' ACTION RECOMMENDATIONS 1.0 OEJECTIVES This procedure provides guidance for determining protective action recommeh-I dations to be made to state asthorities.

2.0 RESPONSIBILITIES 2.1 Short Tern Emerzeney Director Responsible for initial protective action recommendations made to state asthorities. Responsible for all subsequent protective action recommendations until relieved by the emergency director.

2.2 Emergency Director Ralleves the short term emergency director of the responsibility for recommending protective actions to state authorities.

2.3 EOF Coordinator Responsible for the evaluation of radiological data, determination of protective action recommendations, and provision of protective action recommendations to the emergency director.

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3.0 PRECAUTIONS 3.1 The dose-saving effectiveness of protective actions can be influenced by many variable factors such as expected duration of releases, affected population, weather conditions, projected evacuation times,

,and station conditions. Na possible, the appropriate factors should all be considered prior to the recommendation of protective actions.

3.2 For an emergency that begins as a General Emergency, detailed protec-tive action recommendation calculations will not be performed for the initial recommendation.

3.3 For an emergency that begins as a General Emergency, protective

• action recounsendations sust be made within 15 minutes.

4.0 PREREQUISITES 4.1 A General E=ergency has been declared, C@t A Site Area Emergency has been declared and dose projections have 4.2 been completed in acecrdance with ER-5.3, "Off-site Dose Estimates".

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ER-5.4 Paga 3 cf 10 Rcv. 00

'.5.0 ACTICNS 3.1 Short-Term E=ergenev Director or EOF Coordinator 5.1.1 Site Area Emergency 1.

Obtain a copy of the Protective Action Recommendation Worksheet, Form ER-5.4A.

I 2.

Complete Part I of the worksheet obtaining information I

from the Follow-Up Information Form. Form ER-2.2C.

3.

For initial calculation, at Item 5, Distance to Receptor, use exclusion area boundary (0.6 miles), 2, 5, and 10 miles.

4.

If information is not available, enter N/A in appropriate spacts. Exa=ple:

If monitoring teams have not yet reported data, enter N/A in items 14, 15, 17, and 18.

5.

When worksheet has been completed and a protective action recommendation determined, complete the Follow-Up Information Form, Form ER-2.2C.

l 6.

Check worksheet for completeness and submit to the Emergency Director or, if completed by the Short-Term

,,'j Emergency Director, notify the states in accordance with ER-2.2, " Notification of Offsite Authorities".

5.1.2 General Emergency

~~

CAUTION PROTECTIVE ACTION RECOMMENDATIONS MUST BE TRANSMITTED TO l

STATE AUTHORITIES WITHIN 15 MINUTES OF EMERGENCY DECI.A-o RATION l

1.

Obtain a copy of the Protective Action Recommendation Worksheet, Form ER-5.4A.

2.

Complete Part II of the Protective Action Recommendation Worksheet.

3.

Check worksheet for completeness and submit to the Emergency Director or, if completed by the Short-Term Emergency Director, notify the states in accordance with ER-2.2, " Notification of Off site Authorities".

6.0 REFERENCES

6.1 USNRC IE Information Notice No. 83-23, Criteria for Protective Action Recom=endations for General Emergencies; May 4, 1983.

6.2 USE?A Protective Action Guides for Exposure to Airborne Radioactive Materials.

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RR-5.4 Pcg2 4 cf 10 R;v. 00 AM AC. fENTS T

7.0 7.1 Figure 1, Emergency Planning Zone With Sub-Areas and Sectors 7.2 Figure 2, Total Evacuation Clear Time 7.3 Figure 3. Protective Action Recommendation Guidance Charts 7.4 Figure 4. Post LOCA Containment Monitor Response / Personnel Hatch Area Monitor vs % of Core Activity Released 7.5 Figure 5 Predetermined Protective Action Recommendations for General.

Emergencies 7.6 Figure 6, Protective Action Recommendations by Sub-Area for General E=ergency Classifications 7.7 Form ER-5.4A, Protective Action Recommendation Worksheet A

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FICURE 2 t

TOTAL EVACUATION Ct. EAR TIMES (INCLUDING NOTIFICATION) BY WIND DIRECTION 1.

-i NORMAL WEATNER (1) (2) it i

0-2 Miles 0-5 Miles 0-2 Miles

j Plus 2-5 Miles Downwind Flus 5-EFZ Boundary Downwind

{f TIME (30988)

TIME (IBOURS)

TIME (HOURS)

FROH (DECREES)

SUB-AREAS WINTER (1)

SUISIER (2)

SUB-AREAS WIgrTER(1)

SUMMER (2)

SUB-AREAR WINTER (1)

SUMMER (2) unun A, B, C, i

NNU. N 326 to AB 2.92 5.75 D. E 3.25 6.08 y, NE 56 A, B, C, 56 to A, B, C 2.92' 5.75 D. E. F 3.25 6.08 ENE3_E 101 A, B, C, 101 to A, C 2.58 5.08 D. F 3.25 6.08 ESE 124 A

C, D 2.58 5.25 D. F 3.25 6.08 SE 146 146 to A, B, C, A, C, D 2.58 5.25 D. F. C 3.25 6.08 SSE, S 191 A, B, C, i

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2.58 5.75 D

3.08 6.08 U, unu 303 A, B, C, l

303 to A, B 2.92 5.75 D

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,3,26 Any Direction A

2.58 5.75 (1) For winter adverse weather conditions (heavy snow) add 2.5 kamre Not e s: (2) For suemmer adverse weather conditions (heavy rain and fog) add 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.-

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e 1 v. 00 Page 1 of 1 FIGURZ 3 PROTECTIVE ACTION RICOMMENDATION GUIDANCE CHARTS UBOLE BODY GUIDANCE CHART I

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,l evacuation dose (Iten 20) is l tess than shelter dose I

,,,,d Shelter is to be with ventilation control. Ventilation control means turning off air conditioners or fans, closing doors and windows, t'hus pre-venting ac:ess of outside air. Proceed to a basement if available.

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Page 9 cf 10 R:v. 00 Page 1 of 1 FIGURE 5 PREDITIRMINID ?RCTIC IVI ACTION RECCMMINDATIONS FOR GENERAL EMERGENCIES ACTUAL 01 FCTENTIAL DGENE2rr ColfrAINMENT RECOt0 ENDED MACNITUDE OF FAIIERE Q3 Roots)

PROTECTIVE CONDITION RADIATION SOUR,CE OR RELEASE UNDERWAY ACTION

• SEELTER:

I Iow 0-2 miles 360*

2-5 miles downwind II Medium '

No EVACUATE:

0-2 miles 360*

2-5 miles downwind SHELTER:

Remaining areas within EPZ i

III Medium or Eigh Yes EVACUATE:

0-5 miles 360*

af ter plume passes

' 5 miles - EPZ boundary

't downwind SHELTER:

0-5 miles any area that cannot be eva-cuated before plume arrival.

17 Righ No EVACUATE:

0-5 miles 360*

5 miles - EPZ boundary downwind 4

SHELTER:

Remaining areas,within EP:

• See figure 6 for specific sub-areas affected.

m. w,,, y _

ER-5.4 Pcgo 10 cf 10 Rev. 00 Page 1 of 1 FIGURE 6 PRCTICTIVE ACTION RECO?NENDATIONS SY SU3-AREA FOR

(

CENERAL EMERCENCY CI.ASSIFICATIONS CONDITION III (1)

WIND (DEGREES)

CONDITION I CONDITION II CONDITION IV FRCM Shelter Evacuate Shelter Evacuate Shelter !

Evacuate NNW, N 326 to C,D,E, A, B, C,

NNE. NI 56 A. B F, C A, 3 F, G D, E D. E. F, A, 3, C,

36 to ENE, !

101 A, B, C C

L A, B, C C

D, E, F 101 to 5 D E,

A, 5, C,

ESE 124 A. C F. C A, C E, C D. F 124 to 3, E F,

A, 3, C,

SE 146 A, C,-D C

A,C,D E. C D, F 146 to 3, E, F,

A, 5, C,

SSE, S 191 A.C,D C

A,C,D E

D, F. C 191 to 3, C, E A, B, C, SSV, SW 236 A, D F. C AD E, F DC D

i 226 to 3, C, E,

A, 3, C, j

'45W 25 3 A, D F. C A, D E,F,ClD 25d to B,C,D, A

5, C, V. V54 303 A

E, F. C A

E,F,C D

303 to C,D,E, A, 5, C, NW 326 A, B F, C A, 5 E. F, G D

NOTE:

(1)

If any sub-areas within five alles cannot be evacuated prior to plume arrival, sheltering should be reconnended.

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Page 1 cf 6 PROTECTIVE ACTION RECOMMENDATION WORKSHEET CAUTION FOR AN EMERGENCY THAT BEGIN5 AS A GENERAL EMERGENCT, PROCEZD DIRECTLT TO PART II 0F TIIS WORKSEEET.

y PART I - AIE30RNE RELEASE 1.

Time of calculation hours (see 24-hour elock) 2.

Time of release start hours (use 24-hour clock) i 3.

Release duration hours 4.

a.

Wind speed sph (43 foot data) b.

Wind direction from degrees (43 foot data) miles 5.

Distance to receptor 6.-

Affected evacuation subareas (Use items 4b

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and 5 and either page 1

/

or 2 of Figure 2) hours 7.

Flume travel time (Ices 5/ Item 4a) 8.

Time until esposure l

beslas (choose a or 4) a.

If release has benunt b.

Difference hours (Item 1 - Ites 2) hours c.

Time (Item 7 -~ Item 8b) l d.

If release will benin 1stgr e.

Difference hours (Ites 2 - Item 1) hours f.

Time

(!tes 7 + Item Se)

ER-5.4A Rev. 00 i

i I

l l.

Page 2 cf 6 PROTECTIVE ACTION RECCMENDATIDN fdORESHEET (continued) miles (Distance (from Ites 5) 9.

Evacuation Conditions Season (cf.rcle one) a.

1) Sammer: Memorial Day - I. abor Day

2) Winters b.

Weather (circle one)

Mild weather or light rain or snow

1) leoraal:

Summer /Reavy rain and fog, Winter /Reavy snow (see notes on

2) Adverse:

Figure 2)

10. Evacuation time (Use information recorded in Itama 6 and 9, along with Figure 2 to hours determine evacuation time.

11. Exposure time teen 10 - (Ices 8e or hours

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CADTION IF ITEM 11 IS & " Sin ?E NUMBER. INTER EERO 50UR$.

12. Evacuation exposure period Saalier of Ites 3 or hours Iten 11 res

13. Projected whole body dose (E1-5.3) res/hr

14. Monitoring team whole body dose rate El-$. t. A Rev. C0

...s.;.

m..

s Page 3 cf 6

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PROTICTIVE ACTION RECOMMINDATION WORKSHEET (continued) miles (Distance (from Ices 5) res

15. Monitoring team whole body dose (Item 14 x Item 3)

16. Most reliable whole body dose rea (Ites 13 or Ites 15)

17. Monitoring team thyroid res/hr dose rate

18. Monitoring team thyroid dose (Ices 17 x Item 3) rea

19. Whole body evaeustion

[

dose (Ites 12 x Itea 16/Iten 3) rea x 20. Thyroid evacuation dose (Ices 12 x Ites

• 18/ Item 3) res i

21. Whole body shelter dose (Ices 16 x 0.9*)

res

• Average shelter protection factor for wood frame home without basement

22. Thyroid shelter dose (pick a or b) a.

For release duration

+

less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (Ices 18 x 0.5) res b.

For release duration equal to or greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Ites 18 x T. 0.5

_. Item 3__,

res ER-5.4A Rev. 00 i,

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e Pago 4 cf 6 PROTECTIVE AC* ION RECOMMENDATION WORESHEET (continued) mile s (Distance (from Item 5) 23. Whole body indicated action - refer to whole body guidance chart (Figure 3)

(Indicate so estion, shelter, or evacuation

24. Thyroid indicated action -

refer to thyroid guidance chart (Figure 3)

(Indicate ao action, shelter, or evacuation) 25.

Recommended Protective Action (More severe action from Ites 23 or Ites 24 and affected subareas from Ites 6)

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CADTION

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IY EVACUATION IS RECCMMENDED FCE 2-5 MIuS 015-10 MIuS IN N DOWNWIND DIRECTION, RECCMMEND SEZLTE11NG FOR THE REMAINDER OF M SUBARIAS WITHIN THAT RADIAL SCUNDARY.

0-2 2-5 5-10 milea milea allea Evacuate - circle affected subarea (s)

A BCD EFG Shelter - circle affected subarea (s)

A BCD EFG Approved by Emergency Director o

Time Date IR-3.l.A Rev. 00

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o Page 5 cf 6 PROTECTIVE ACTION RECOMMENDATION WORKSHEET (continued)

PART II - GENERAI, EMERGENCY

26. Time Date General Emergency Declared.

27. Wind direction (YAS D point A1630) 43 foot level (from)

_ degrees i

28. Mas there bee's a loss of physical control of the facility to intruders?

Yes No If yes, recommend:

Evacuate - Subares : A

29. Has there been a release of fission products into containment?

Indicated by:

,1.

Containment Eigh Range Post-LOCA Monitor RDMS 1AM106,1AM107 i

. sad..,,

2.

Personnel Hatch Monitor RDMS 11M-EM-6536 Yes _

No

30. What is the estimated nagnitude. o'f the radiation source (from containment high range post-LCCA monitor reading, personnel hatch monitor reading, and Figure 4)?

If the radiation source is low, so to step 33 and make minissa protective actions in accordance with Candition I of Figure 6.

31.

Is a breach of containment anticipated within the next 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (As deter-mined by 1) status tree:

containment integrity - red, with pressure increasing or 2) containment isolation not indicated)?

Yes No If yes, anticipated time ER-$.4A Rev. 00 l

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_____._._______.___________________________A_____.'

Page 6 cf 6 PROTICTIVE ACTION RICCMMENDATION WORISREIT (continued) 3 2..

Condition:

Use information from Items 30 and 31 to determine the appropriate condition i

from Figure 5.

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33. Protective Action Recommendations Using wind direction (Item 27) and condition (Item 32), determine pro-testive action from Figure 6.

If Item 28 also indicates a protective action, enter the more severe of the protective actions.

0-2 2-5 5-10 milea mile miles Evacuate - subares(s) (circle):

A 3CD EFG Shelter - subarea (s) (circle):

A.

3CD EFG Approved by Emergency Director Time Date

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8 ER-5.4A Rev. 00 8

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UNITED STATES

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, n [,j NUCLEAR REGULATORY COMMISSION

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.E WASHIN GTO N, C. C. 20555 December 27, 1984

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Mr. J. J. Sheppard, Chairman Westinghouse Owners Group ATTACHMENT 5 Carolina Power & Light Company 411 Fayetteville Street Post Office Box 1551 Raleigh, North Carolini 27602

Dear Mr. Sheppard:

We have received your letter of August 15, 1984, transmitting the comparison of Revision 1 of the Emergency Response Guidelines (ERGS) with the approved BASIC (Revision 0) version.

This information was helpful in expediting a preliminary staff assessment of Revision 1 pending a more detailed, long term review.

Our imediate effort was limited to a determination of whether technical differences of Revision 1 from the approved BASIC version could influence plant licensing decisions. Accordingly, we conducted a limited review of the significant differences.

Based on the enclosed discussion, we believe that reasonable assurance exists that the identified emergency guideline changes would not result in a violation of approved plant licensing design bases and may be implemented.

This" approval does not preclude the need to pursue resolution of the open issues defined in our generic safety evaluation of the approved BASIC version and the reccmendations of the ERG Validation Report.

The outstanding comments and recomendations for enhancement together with further coments resulting from the completion of our more detailed review are to be resolved within the continuous program for maintaining ERGS.

Sincerely, V

~! IM L

Darrell. G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation

Enclosure:

Prelininary Review of Westinghouse Erergency Response Guidelines, Revision 1

PFELIMINARY REVIEW OF k'ESTINGHOUSE EMERGENCY RESPONSE GUIDELINES REVISION 1 INTRODUCTION The NRC staff has reviewed the August 15, 1984 transmittal from the Westinghouse Owners Group providing a comparison of Revision 1 of the Emergency Response Guidelines (ERG) with the approved BASIC version of the ERGS. The staff also has considered the findings of the ERG Validation Program (WCAP10599, June 1984). The purpose of this eview was to determine whether the Revision 1 ERG's contained sufficient differences from the approved BASIC version to influence plant licensing decisions prior to completion of a detailed, long-term review.

The staff',s limited review of major changes incorporated in the Revision 1 guidelines included consideration of changes to Safety Injection (SI)

Termination Criteria, SI Reduction Method, Loss of Heat Sink Guidelines and Reactor Coolant Pump Trip Criteria. The staff's evaluation of each of these changes follows.

OISCUSSION Safety injection Termination Criteria The SI Termination Criteria in BASIC included, as a necessary condition for

$1 termination, that Reactor Coolant (RC) pressure achieve either an established value or a 200 psi increase.

In Revision 1, this criterion was reduced to " stable or increasing" pressure to avoid unnecessary RC pressure increases during a stean generator tube ruoture and thus avoid the carecessary flow of PC into the r, team generator.

Since the modified critoria A

still require adequate RC subcooling, RC inventory and secondary heat sink, the modified criteria provide adequate assurance of core cooling under accident and transient conditions, and are therefore acceptable.

I SI Reduction tiethod In response to transients which require the reduction of SI flow in combination with cooldown and depressurization of the RCS, an SI Reduction Sequence is accomplished in Revision 1 by a series of pump trips.

The BASIC version of the guidelines required the " throttling" of SI flow.

This change to the guidelines is desirable because the throttling would be typically performed by an operator stationed at the valve location which could become a high radiation area under accident conditions.

Tripping pumps will accomplish the desired flow reduction and therefore is the preferred method..

Revision 1 presents the methodology for calculating appropriate subcooling criteria, and for evaluating on a plant specific basis, a sequence for reducing safety injection flow in coincidence with RCS cooldown.

The RE0VCE computer model was developed by Westinghouse for the utilities to use in the determination of the amount of subcooling required prior to the reduction of i

SI flow by shutting off an SI pump.

The methodology is presented for j

optimizing the series of $1 reductions during a cooldown while maintaining a minimum value of subcooling.

This refined cooldown technique is used in 1

Emorgency Response Guidelines ES-1.2, POST LOCA COOLDOWN AND DEPRESSURIZATION; ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DES!RE0; ECA-3.2, SGTR WITH LOSS OF REACTOR COOLANT - SATUPATED RECOVERY DESIRED; and FR H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK.

1

We have not completed the review of the REDUCE computer model, but we have considered the changes to the guidelines that are based on these analyses.

We note that even if the numerical values of subcooling were in error for the pump shutoff sequence, the guidelines contain sufficient overriding criteria in the foldouts and Functional Recovery Guidelines such that the resulting operator actions would not result in a significant adverse effect on the course of an accident or transient. On the basis of this limited review, we find this method of SI reduction acceptable for implementation pending completion of our longer-term review.

Response to loss of Secondary Heat Sink This guideline has been supplemented by the inclusion of a SI flow reduction and PORY closure sequence to restore the plant to controlled conditions after bleed and feed has been initiated. Greater attention is given in the backgrcund document to the need for prompt entry into bleed and feed cooling.

The background document in Revision 1 contains the methodology for determining the plant-specific operational response to loss of al.1 feedwater that would be necessary to avoid drying out the steam generators before the bleed and feed operations could be effective in preventing a condition of inadequatecorecooling(ICC).

The required operational response is sensitive to the capacity of the pressurizer relief valves and the mass of water in the steam generators.

This guidance also includes Reactor Coolant Pumps (RCP) trip criteria specifically for this transient.

Supportive ar,alyses are presented in the background material as well as in referenced documentation (WCAP9914:"PORVSensitivityStudyforLOFW-LOCAAnalyses").

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While we have not completed our review of these analyses, our review of the related guidelines indicates generally that Re' vision I requires more prompt response and greater attention to symptoms of a loss of secondary heat sink.

l Based on this limited review, we find the Revision 1 guidelines acceptable for implementation pending completion of our longer term review.

.l The' background material in Revision 1 also indicates that for plants with low l

capacity pressurizer relief valves, the loss of heat sink could lead to a condition of ICC.

Because of this close relationship between the loss of I

heat sink and ICC, the importance of the Heat Sink Critical Safety Function was increased. Accordingly, in Revision 1 the order of priority of operator i

response to alarm conditions is as follows:

Subcriticality, Core Cooling, i

Heat Sink, Integrity, Containment, and Inventory, whereas in the BASIC version, the Integrity and Heat Sink functions were third and fourth respectively.

l The staff considered the effect that this interchange might have en the

]

Integrity critical Safety Function.

The relative risk from thermal stresses i

j resulting from various transients is given in the background material i

regarding stagnant RC loops. The data presented shows a much lower risk of flaw extension resulting from a loss of heat sink than that which would result from a LOCA or SGTR.

This indicates that the actions taken to restore the steam generator heat sink or feplement bleed and feed operation should not significantly affect the parameters that signal the need for integrity

+

function actions.

Therefore, we find this change in priorities acceptable for implementation pending completion of our longer term review of the

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Emergency Response Guidelines.

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Reactor Coolant Pump Trip Criteria i

Revision I background material also contains an evaluation of alternate RCP trip parameters to establish a variable which will reduce the probability of unnecessary RCP trip for SGTRs and non-LOCAs, while still providing for timely RCP trip for small break LOCAs.

The results of this evaluation can be used by utilities to establish the appropriate RCP trip parameter and setpoints for use in Plant Specific Emergency Operating Procedures based on the Emergency Response Guidelines.

The relevant parameters evaluated are RCS pressure, RC subcooling and RC-to-Steam Generator (SG) pressure differential.

The resulting trip parameter selection is expected to result in improved operation of the RC pumps over a wide range of accident conditions.

This background material is, relevant n,the NRC staff's continuing review of this generic issue pursuant,

~

to NRC Generic Letter No. 83-10d:" Automatic Trip of Reactor Coolant Pumps"(ResolutionofTMIActionItemII.K.3.5). Accordingly, we conclude that, pending completion of our more detailed review of this matter, the RC Pump Trip Criteria methodology presented in Revision 1, together with the guidance provided in NRC Generic Letter No 83-10d, represents an acceptable basis for the implementation of plant-specific pump trip criteria.

Validation Program In addition to reviewing the Westinghouse Owners Group comparison of the Revision 1 guidelines with the approved BASIC version, the staff considered the findings of the ERG Validation Program (WCAP 10599, June 1984).

This j

documentation indicates that the actual validation of the Revision 1 ERGS was conducted on a plant-specific, full-scale control room simulator and that a 1

i 1

.1 normal operating crew complement used simulator-specific E0Ps to guide their actions in response to control rocm (plant /sidiulator) indications during major plant casualties.

The report further states that the E0ps used were based on, and closely resembled, the ERG Revision 1 set, and.that the training involved was developed for the same Revision 1 set. Thirty-five casualties were imposed on the plant including all major events and many events with multiple failures. Observations of operator response resulted in 39 recomendations for enhancement of the generic guidelines, to be addressed i

by the Owner Group as part of the continuous Ek3 maintenance program.

The report states that no Safety-related technical deficiencies in the ERGS were b

observed during the test program and concludes that the Revision 1 version of the ERGS represents a substantial improvement over the BASIC version.

The staff concludes that the results of the ERG Validation Program support implementation of the Revision 1 ERGS while the recomendations for enhancement are resolved as part of the long-term program for maintaining-ERGS.

CONCLUSION Based on our review of the August 15, 1984 " Emergency Response Guidelines Comparison of Revision I with BASIC", our audit of strategies related to this i

comparison, and our examination of the results of the ERG Validation Program i

(MCAp 10599), we conclude that reasonable assurance exists that the identified emergency guideline changes would not result in a violation of approved plant licensing design bases and implementation may proceed.

l l

l l

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-7 The outstanding comments and recommendations for enhancement together with further comments resulting from the completion of our more detailed review are to be resolved within the continuous ERG maintenance program for maintaining ERGS.

This approval does not preclude the need to pursue resolution of the open issues defined in our generic safety evaluation of the approved BASIC version.

Based nn our preliminary review, implementation of Revision 1 pending completion of a detailed evaluation is believed to be in the best interest of overall plant operational safety.

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