ML20205L220
| ML20205L220 | |
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|---|---|
| Site: | Maine Yankee |
| Issue date: | 03/18/1986 |
| From: | Maine Yankee |
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| PROC-860318, NUDOCS 8604030332 | |
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MAINE YANKEE EERGENCY OPERATING PROCEDURES TECHNICAL GUIDLINES O
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MAINE YANKEE PROCEDURES GENERATION PACKAGE.
PLANT TECHNICAL GUIDELINES I
i REVISION 1 05-06-85 Prepared by:
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I. L. Anderson THERE ARE 44 PAGES IN THIS DOCUMENT INCLUDING THE COVER SHEET.
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1
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CONTENTS
- 1. INTRODUCTION ...... .. ................. .. .I
- 2. COMPARISDN OF SYSTEM DESIGNS (DESIGN DIFFERENCES) ....... .2
- 3. ANALYSl5 APPLICABILITY . ... . .. ... .... ...... . .7 3.1 Pressurized Thermal Shock (PTS) . .... ... ...... . .7 3.2 Natural Circulation Cooldown . ..... .. .......... 13 3.3 Loss of Coolant Accident (LOCA) . . ......... ..... 14 3.4 Post LOCA Cooldown and Depressurization ... .... .... . 15 3.5 Steam Generator Tube Rupture (SGTR) ......... ..... 16 3.6 Loss of All AC Power . .. ................... 17 3.7 Secondary Side Break . ... .. . ... ............ 18 3.8 Inadequate Core Cooling (ICC) ..... ............ 18 3.9 Anticipated Transient Without Scram (ATWS) ........... 20 3.10 Loss of (Secondary) Heat Sink (LOHS) .. .... ... ... .. 20 3.11 Reactor Coolant Pump (RCP) Trip . ... ... ....... .. 21
- 4. BASIS FOR USING THE GENERIC WESTINGHOUSE ERGS ........ .. 22
- 5. METHOD FOR DEVELOPING EOPs FROM ERGS ... ....... .... . 23 5.1 General . . . ... . .. ....... ....... ... . . 23 5.2 Preparation . . . . . .. ....... ....... ... . . 23 5.3 Method . . . . . . . . . ........ ....... ... .. 23
- 6. CONCLUSION . . . . . . . . .................... 25
- 7. LIST OF REFERENCES . . . .. .... . ............. . 26 g3 APPENDIX A: COMPARISON OF SYSTEM DESIGNS (DESIGN DIFFERENCES) . . . . 27
\) APPENDlX B: LISTING OF APPLICABLE EMERGENCY RESPONSE GUIDELINES (REV.1) 38 APPLICABLE OPTIMAL RECOVERY GUIDELINES .. .. ...... .. . . . 38 APPLICABLE FUNCTION RESTORATION GUIDELINES ... ..... ... .. 39 4
APPENDlX C: MAINE YANKEE E0P DOCUMENTATION FORM .. . ... ... . . 40 4
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Contents I
1 MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1 O
LIST OF ILLUSTRATIONS Figure 1. EFFECT OF FLUX REDUCTION MEASURES ON ASSOLUTE RTNOT AT CRITICAL LONGITUDINAL WELD 2-203 at 270*
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(REG. GulOE 1.99 REV. I WITH UNCERTAINTIES) . .... ..... . .... . ~10 Figure 2. EFFECT OF FLUX REDUCTICN MEASURES ON ASSOLUTE RTNOT AT CIRCUMFERENTIAL WELD 9-203 PEAK AZlMUTHAL FLUENCE LOCATION 0*
(REG. GUIDE 1.99 REV. I WITH UNCERTAINTIES) . .... ....... ... . 11 T
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d MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1
- LIST OF REVISIONS I Revision 0 4/4/85 - Original issue i
Revision 1 5/6/85 - Revise upper support plate thickness Delete need for verification of fluence and list references Revise FW isolation description
, Add appropriate references i
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1
- 1. INTRODUCTION The development of plant-specific technical guidelines is one of the four elements of the Procedures Generation Package, which is required by NUREG-0899 and Supplement I to NUREG-0737. For the Maine Yankee plant, applicable generic Westinghouse Owners Group Emergency Response Guidelines (ERGS) Revision I listed in Appendix 8 will be used as the basis for writing upgraded Maine Yankee Emergency Operating Procedures (E0Ps) .
This document describes the method of developing plant specific E0Ps from applicable generic Westinghouse Owners Group (WOG) ERGS for the Maine Yankee plant. This document includes a summary of design dif f erences between the Maine Yankee plant and the HP reference plant design as described in the WOG ERGS Rev. 1 Executive Volume Reference Plant Description, dated September 1. 1983. In some areas additional detail beyond that in the reference plant description is included here to give added detail to the procedure writers. In addition. an analysis applicability summary is included describing which generic analyses performed for Revision 1 WOG ERGS apply to Maine Yankee and which plant specific analyses are required to justify the upgraded EOPs.
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1
() 2. COMPARISON OF SYSTEM DESIGNS (DESIGN OlFFERENCES)
During the development of the generic WOG' ERGS, a generic reference plant design configuration was assumed, and the technical content included in the ERGS is based upon the reference plant design. The following systems are included in the reference plant (Maine Yankee terminology is used):
Reactor Protective System Engineered Safeguards Features Actuation System Nuclear instrumentation System Control Element instrumentation System Containment instrumentation System Reactor Coolant System Chemical & Volume Control System Emergency Core Cooling System
- Residual Heat Removal System Radiation Monitoring System Containment Spray System Containment Air Recirculation System Component Cooling Water System Service Water System Main Feedwater and Condensate System Main Steam System Emergency and Auxiliary Feedwater Systems Steam Generator Slowdown System Sampling System
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Spent Fuel Storage and Cooling System Control Element Drive Mechanism Cooling System CEA Control System Turbine Control System Electric Power System Pneumatic Power System To aid the development of plant specific E0Ps for the Maine Yankee plant, a comparison of the 3bove systems from an emergency operations perspective for the Maine Yankee and reference plant was made. This comparison was done in a systematic and complete manner by reviewing all of the above systems. The purpose of the comparison was to identify areas of the Maine Yankee plant which are different from the r'eference plant f rom the standpoint of emergency system operations and thus these areas will be explicitly considered and included as appropriate during the development of the Maine Yankee E0Pt. The comparison for each system follows. Please see Section 7 for a listing of references used to compile this comparison. Appendix A was developed to provide a detailed comparison of each system based upon its use in the ERCS. Appendix A i should be referred to during the following comparison of each system. I I
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MAINE YANKEE PLANT TECHNICAL CUIDELINES REVISION 1 s
(m,) REACTOR PROTECTIVE SYSTEM The function of the Reactor Protective System (RPS) is to monitor specified process parameters and equipment status and to acteate reactor trip if conditions exceed specified limits. The Maine Yankee design has differences from the reference plant. However these differences only af fect the list of automatic reactor trip signals which are entry conditions to the E0Ps and do not af fect the body of the procedure set.
ENGINEERED SAFEGUARDS FEATURES ACTUATION SYSTEM
- The function of the Engineered Safeguards Features Actuation System (ESFAS) is to monitor specified process parementers and to actuate Engineered Safeguares Features (ESF) operation if conditions exceed specified limits.
The Maine Yankee design has differences with the reference plant, as shown in Appendix A and these differences should be incorporated in the writing of the E0Ps.
NUCLEAR INSTRUMENTATION SYSTEM The function of the Nuclear instrumentation System (N I S) is to monitor and display the reactivity state of the reactor core. The Maine Yankee design has differences with the reference plant, as shown in Appendix A and these differences should be incorporated in the writing of the E0Ps.
() CONTROL ELEMENT INSTRUMENTATION SYSTEM The function of the Control Element Instrumentation System (CEIS) is to monitor and display the position of the reactor core control rods. From the standpoint of emergency operations, the CEIS is the same for the Maine Yankee and reference plant.
P CONTAINMENT INSTRUMENTATION SYSTEM The function of the Containment instrumentation System (ColS) is to monitor the environmental condition and Isolation status of the containment. From the standpoint of emergency operations, the Col $ is the same for the Maine Yankee and reference plant.
REACTOR COOLANT SYSTEM The function of the Reactor Coolant System (RCS) is to transfer heat from the reactor core to the main steam system or residual heat removal system to provide a barrier against the release of reactor coolant or radioactive material to the containment environment. The Maine Yankee design has differences with the reference plant as shown in Appendix A and these differences should be incorporated during the writing of the E0Ps.
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I MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1 l l CHEMICAL ANO VOLUME CONTROL SYSTEM The function of the Chemical and Volume Control System (CVCS) system is to provide makeup to the reactor coolant system and to provide reactivity l control for normal operations and any event that does not require i engineered safeguards features operation. From the standpoint of emergency '
. operations, the CVCS is the same for the Maine Yankee and reference plant.
EMERGENCY CORE COOLING SYSTEM The function of the Emergency Core Cooling System (ECCS) is to provide makeup to the reactor coolant system and to introduce negative reactivity or restrict the addition of petitive reactivity for events that require engineered safeguards features operation. The Maine Yankee design has 2
differences with the reference plant as.shown in Appendix A and these differences should be incorporated during the writing of the E0Ps.
RESIDUAL HEAT REMOVAL SYSTEM The function of the Residual Heat Removal System (RHRS) is to remove residual heat from the reactor coolant system during plant shutdown *
, operations at low reactor coolant system pressures. The Maine Yankee design has dif f erences wi th the reference plant, as shown in Appendix A and these differences should be incorporated during the writing of the E0Ps.
RADIATION MONITORING SYSTEM The function of the Radiation Monitoring System (RMS) is to monitor the radiation levels in specified process systems and specified areas internal and external to the plant. From the standpoint of emergency operations, the RMS is the same for the Maine Yankee and reference plant.
CONTAINMENT SPRAY SYSTEst T'he function of the Containment Spray System (CSS) is to provide containment pressure suppression and airborne fission product removal for events that require engineered safeguards features actuation. The Maine Yankee design has differences with the reference plant as shown in Appendix A and these differences should be incorporated during the writing of the E0Ps.
CONTAINMENT AIR RECIRCULATION SYSTEM The function of the Cuntainment Air Recirculation System (CARS) is to provide containment heat removal and combustible gas mixture control. The Maine Yankee design has differences with the reference plant as shown in Appendix A and these differences should be incorporated during the writing of the E0Ps.
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1
() COMPONENT COOLING WATER SYSTEM The function of the Component Cooling Water System (CCWS) is to provide heat removal from system process and equipment via an intermediate closed-loop system. From the standpoint of emergency operations, the CCWS is the same for the Maine Yankte and reference plant.
SERVICE WATER SYSTEM The function of the Service Water System (SWS) is to provide heat removal from system processes and equipment to the ultimate heat sink via an open-loop system. From the standpoint of emergency operations. the SWS is the same for the Maine Yankee and reference plant.
MAIN FEEDWATER AND CONDENSATE SYSTEM The function of the Main Feedwater and Condensate System (MFCS) is to provide water to the secondary side of the steam generators during plant power operations. The Maine Yankee design has differences with the reference plant and these differences should be incorporated during the writing of the E0Ps.
EMERGENCY FEEDWATER SYSTEM & AUXILIARY FEEDWATER SYSTEM The functions of the Emergency Feedwater System (EFS) and the Auxiliary Feedwater System (AFS) are to provide coolant to the secondary side of
_) the steam generators during plant shutdown operations and fcr events that require engineered safeguards features actuation. The Maine Yankee design has differences with the reference plant as shown in Appendix A and these differences should be incorporated during the writing of the E0Ps.
MAIN STEAM SYSTEM Tha function of the Main Steam System (MSS) is to provide controlled heat removal from the reactor coolant system via the steam generators. From the standpoint of emergency operations, the Maine Yankee design has differences with reference plant as shown in Appendix A and these differences should be incorporated during the writing of the EOPs.
STEAM GENERATOR BLOWOOWN SYSTEM The function of the Steam Generator Blowdown System (SGBS) is to provide letdown from the secondary side of the steam generators. From the standpoint of emergency operations, the Maine Yankee design is the sanie as the reference plant.
SAMPLING SYSTEM The function of the Sampling System (55) is to provide a means for sampling process systems. From the standpoint of emergency operations. the Maine Yankee design is the same as the reference plant.
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t MAINE YANKEE PLANT TECHNICAL GUlD LINES REVISION 1 1 l
SPENT FUEL STORAGE ANO COOLING SYSTEM The function of the Spent Fuel Storage and Cooling System (SFSCS) is to j control fuel storage positions to ensure a subcritical geometric i configuration and to provide heat removal to maintain stored fuel within ;
specified temperature limits. From the standpoint of emergency l operations, the Maine Yankee design is the same as the reference plant. l l
CONTROL ELEMENT ORIVE MECHANISM COOLING SYSTEM The function of the Control Element Drive Mechanism Cooling System (CEMCS) ,
is to provide heat removal from the control element drive mechanisms. From the standpoint of emergency operations. the Maine Yankee design is the same as,the reference plant. .
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CONTROL ELEMENT ASSEMSLY CONTROL SYSTEM , l
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- The function of the control Element Assembly Control System (CEACS) is to l s
control.the position of the control' elements in the reactor core. From the l
'the standpoint of emergency oper ations, the Maine Yankee design is the same !
as' the ref erence plant. -
l TUR81NE CONTROL" SYSTEM The function of the Turbine Control System (TCS) i s to control the
% turtii ne-genera tor . From the standpoint of e.nergency operations, the Maine Yanket design has' differences with the reference plant and these differences i should be incorporated during the writing of the'! ops.
ELECTRICAL POWER SYSTEM 4
The function of the Electrical Power System (EPS) is to provide ac and de eledtrical power to equipment that require electrical power to accomplish their functions. The Maine Yankee design has differences with the reference plant, as shown in Appendix A and these dif f erences should be' incorporated in.the writing of the E0Ps.
PNEUMATIC POWER SYSTEM The function of the Pneumatic Power System (PPS) i s to supply pneumatic power (typically control air) to equipment that require pneumatic power to I accomplish their functions. From the standpoint of emergency operations.
. the Maine Yankee design is the same as the reference plant.
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1 x
~), ANALYSIS APPLICABillTY The Maine Yankee Plant design has been reviewed with respect to the )
reference plant analyses which were performed to support'the development of the generic WOG ERGS. Since the Maine Yankee Plant is similar to the reference plant, many of the analyses are directly applicable to Maine Yankee. A review of the generic analyses shows that although some generic analyses results do not bound Maine Yankee (therefore plant specific analyses are required), the generic analysis methodology does apply.
The following table summarizes the assessment of analysis applicability:
ASSESSMENT COMPARISON
SUMMARY
ADDITIONAL ANALYSIS ANALYSl$ ASSESSED RECOMMENDED 7 Pressurized Thermal Shock No Natural Circulation Cooldown No Loss of Coolant Accident (LOCA) No Post LOCA Cooldown & Depressurization Yes Steam Generator Tube Rupture (SGTR) Yes Loss of All AC Power No l Secondary Side High Energy Line Break No inadequate Core Cooling (I CC) Yes Anticipated Transient Without Scram (ATWS) No
[} Loss of (Secondary) Heat Sink (LOHS) Yes The following sections detail the assessments summarized above.
3.1 Pressurized Thermal Shock (PTS)
The generic analysis performed to establish the generic r essure-temperature limit curves for the WOG ERGS applies to the Maine Yankee plant. The reasons for this determination include:
- Applicability of Transients The Integrity Limit Analysis utilized step change transients that are l independent nf reactor coolant system (RCS) response during a pressurized thermal shoc( '?.TM etant. The resultant pressure-temperature limit curve (Integrity Limit) conservatively bounds transient cooldown rates.
The analysis is therefore applicable independent of the RCS transients and other ccerational functions. So long as'the measured system responses are indicative of the actual system conditions. the Integrity Limit Analysis apolies. Therefore, the integrity status of the plant during a potential PTS event can be determined.
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s MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1 e Applicability of Thermal. Stress, and Fracture Analysis The reactor vessel thermal response and stress response are applicable since the vessel design and the analytical boundary conditions are represented by the generic analysis. In particular, the thicknes's of the Maine Yankee vessel of 8.5 inches is within 1.5 percent of the vessel thickness of 8.625 inches used in the generic Integrity Limit Analysis.
The fracture analysis is also applicable since the same assumptions apply to the Maine Yankee vessel beltline as apply for the generic analysis vessel beltline region.
e Appilcability of Generic Categories The Generic Category designations are applicable for RTNDT values projected for the reactor vessel inner surface critical locations. The Generic Category ll Is applicable for RTNDT values between 200'F and 250'F for both circumferential and longitudinal welds and base plate materials. The Generic Category lilb is applicable for RTNOT values above 250'F and below 300*F for circumferential weld materials only.
The Maine Yankee reactor vessel limiting RTNCT values currently fall within these generic Ilmits. .
Three cases will be described in defining the Applicable Generic Category, based on Figures 1 and 2:
( CASE PATH DESCRIPTION I
I A No credit for flux reduction measures 2 8&C Credit for flux reduction measures l
3 D Credit for flux reduction measures -
HEDL NUREG/CR-2805, vol 3 3 E Credit for flux reduction measures -
HEDL NUREG/CR-3391 (to be issued)
Loncitudinal Weld The limiting longitudinal weld number 2-203 at 270' uses surveillance weld properties (Cu = .36 wt%, Ni = .99 wt%, initial
'RTNDT = -56'F). For Case I above, using the design basis flux, the
, longitudinal weld reaches the upper limit of Generic Category 11 (250
- F) by January 1990 using the Maine Yankee "RTNDT Versus Year of Operation" curve in Figure 1, which is Path A. For Case 2, using the Maine Yankee curve, Paths B and C, the weld reaches the upper limit of the Generic Category 11 by November 1992 based on a low leakage core installation in 1983. For Case 3 Paths 0 and E, the weld reaches the upper limit in 1996. I s- g i
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MAINE YANKEE PLANT TECHNICAL GulDELINES REVISION 1
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.' Circumfarential Weld For Case 1, the limiting circumferential weld num-ber 9-203 uses surveillance weld properties (Cu = .36 wtt, NI = .99 wt%,
initial 'RTNOT = -56'F) . For Case I above, using the design basis flux, ,
the circumferential weld reaches the upper limit of Generic Category ll l (250'F) in March 1987 using the Meine Yankee curve. Path A. In Figure 2.
For Case 2. using the Maine Yankee curves, Paths 8 and C, the weld reaches the upper limit of Generic Category ll In February 1988 based on !
a low leakage core installation in 1983. For Case 3. Path D, the weld 1 reaches the upper limit in June, 1990. 1 I l
I Aoolicable Generic Categories Based on the above assessment, the dates of applicability for the three cases are described-for each case and listed in Table 1. The dates are approximate based on the Maine Yankee projected usage factor. The evaluation is for the most limiting circum-farential weld, number 9-203, and longitudinal weld number 2-203 at 270*.
When the longitudinal weld exceeds the upper bound of the Generic Cate- l gory ll (250*F RTNDT) for any of the cases considered, a plant specific 1 analysis will be required to generate a plant specific P-T limit curve '
for the Emergency Procedure, Vessel integrity Limit guidelines.
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1 O
TABLE 1 .
DATES OF APPLICABLE CATEGORY FOR MAINE YANKEE REACTOR _ VESSEL (a)
Category 11 Category lilb Plant Specific Analysis CASE From To From To After No credit for Flux Reduction c Measures Present Mar 1987 Mar 1987 Jan 1990 Jan 1990 Credit for Flux Reduction Measures Present 'Feb 1988 Feb 1988 Nov 1992 Nov 1992 Credit for Flux Reduction Measures based on Re-evaluation .
of surveillance Capsule Fluence by HEDL (b) Present Jun 1990 Jun 1990 Jan 1996 Jan 1996 l
(a) Dates of applicability are approximate based on projected usage factor. l l
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MAINE YANKEE PLANT TICHNICAL GUIDELINES REVISION 1 O ,
Figure 1 EFFECT OF FLUX REDUCTION MEASURES ON ASSOLUTE RTNOT AT CRITICAL LONGlTUDINAL WELD 2-203 at 270*
(REG. GUIDE 1.99 REV.1 WITH UNCERTAINTIES) 4W LICte: i Re Creelt for Flus Reduetten Measures t&C Creelt for Flus Roemetten moesures O Creelt for Flus asengties Measures-NEBL 6-m5. tel. 3 398 - t Creest for' Flus Ameuttles Amesures . EE IBMC/CB.3319 (to be issuse)
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1 Figure 2 .
EFFECT OF FLUX REDUCTION MEASURES ON ABSOLUTE RTNDT AT CIRCUMFERENTIAL WELD 9-203 PEAK AZIMUTHAL FLUENCE LOCATION 0*
(REG. GUIDE 1.99 REV.1 WITH UNCERTAINTIES)
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me Cavett for Fles asesetten meeserve B 8 C Crevet for rios asemetten meeser,s 3 Creett for fles asemetten measures NEBL ase-
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1 3.2 Natural Circulation Cooldown The WOG ERG Rev. I natural circulation cooldown analysis is not directly applicable to Maine Yankee because of differencer in (1) the amount of flow communication between upper head fluid and the rest of the RCS. and (2) the thickness of the reactor vessel upper plate.
The amount of flow into'the upper head provided by the spray nozzles in the generic ERGS for a plant with the upper head at TH0T*(0.15% of total flow at full flow) is much greater than that provided by the alignment keyways at Maine Yankee (0.0k% of total flow at full flow) . .
The ERG generic analysis yielded an average upper head fluid cooldown l rate of 10*F/hr due to flow from the guide tubes into the upper head l l
region and out of the upper head region through the spray nozzles.
Any credit taken for upper head temperature reduction due to flow through the guide tube / upper head / alignment keyway path at Maine Yankee would be significantly less than 10*F/hr.
The second reason that the WOG ERG Rev. I natural circulation cooldown analysis is not directly applicable to Maine Yankee is that in the generic ERG analysis, if no CEDM fans are running, a wait period during which plant depressurization was stopped was required for plants with the upper head at THOT*. This wait period was to allow the upper head to cool off to a temperature corresponding to a saturation pressure of 400 psig (cut-in pressure of the Residual Heat Removal System) via heat conduction from the upper head fluid through the reactor vessel upper plate to the rest of the RCS. In the generic ERG Os analysis the thickness of the upper (suppor t) plate was either 12 inches or 5 inches. Maine Yankee has an upper plate that is about k inches thick. Thus the cool off rate of the upper head fluid at Maine Yankee due to conduction through the upper support plate would be i faster than that for the generic ERG analysis.
Although the WOG ERG Rev. 1 natural circulation analysis is not directly applicable to Maine Yankee, no plant specific analysis is required for Maine Yaakee. This is the case since (1) Maine Yankee has the Reactor Vessel Head Thermowell which allows the upper head fluid temperature to be monitored during a natural circulation cooldown. (2) Maine Yankee has the Primary Inventory Trend System (PITS) to monitor any upper head void growth during a natural circulation cooldown, and (3) the upper head fluid for Malne Yankee cools more rapidly via heat conduction through the upper plate than the ERG reference plant.
- THOT upper head plants are plants which have an amount of flow through the spray nozzles lnto the upper head region such that the fluid temperature of the upper head at full power operation is close to the fluid temperature of the hot leg. For conservative analyses the upper head fluid temperature for these plant is assumed to be equivalent to the hot leg fluid temperature.
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1 MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1
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in general the method of natural circulation cooldown for Maine Yankee would be as follows. Cool down the RCS as fast as possible while maintaining adequate RCS subcooling margin and monitoring RCS pressure to ensure that it is greater than the saturation pressure corresponding to the upper head fluid temperature. PITS would be used to ensure no upper head voiding occurs. As many CEDM fans as
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possible would be started early in the E0P and kept on (if possible) ,
until the entire RCS, including the upper head region, is below I 210'F.
3.3 Loss of Coolant Accident (LOCA)
The analyses bases for the WOG ERG E-1 are the LOCA analyses performed for Saf ety Analysis Reports and those in WCAP-9600, " Report on Small Break Accidents for Westinghouse NSSS Systems". To determine ERG Rev. I analyses' applicability, Maine Yankee LOCA analysis was was investigated to determine if any significant differences exist that would affect the ERG recovery method. Also, design differences between the ERG reference plant and Maine Yankee were analyzed to determine if a significant variation in plant response following a ,
LOCA would occur due to these differences.
Two design dif ferences between the ERG reference plant and Maine Yankee which could potentially affect the reference piant analysis applicablity are (1) the safety injection tanks for Maine Yankee start injecting O' at 230 psig while the accumulators on the ERG reference plant start injecting at 650 psig, and (2) Maine Yankee has less atmospheric steam dump capacity (5% of total steam flow) than the ERG reference plant (10% of total steam flow) . ERG E-1 is designed to recover the plant completely for larger sized LOCAs. For plant recovery from smaller sized LOCAs ERG ES-l.2, Post LOCA Cooldown & Depressurization, is also used (see section 3.4) . For larger-s zed LOCAs the RCS pressure decreases rapidly. Due primarily to this rapid RCS depressurization, the time of safety injection tank injection following a large LOCA for Maine Yankee (13.7 to 15.6 seconds) is essentially the same as that for the ERG reference plant accumulators (14 seconds to 16 seconds) . Also, the time it takes to empty the safety injection tanks and accumulators is essentially the same. Thus the variation in safety injection tank and accumulator injection pressure has a negligible ef fect in the ERG reference plant and Maine Yankee system responses following a larger sized LOCA.
Late in the recovery from a larger sized LPCA steam may have to be.
dumped to atmosphere, if the condenser.is not available, to reduce secondary pressure and aid in further cooldown and depressurization of the RCS. If this secondary side depressurization is required it would be performed when the RCS pressure is low and the plant is on cold leg recirculation. Due to the low atmospheric steam dump capacity for Maine Yankee a secondary side depressurization would take longer to accomplish, but with the plant in a relatively stable condition (on cold leg recirculation) there would be no problems as a result of this longer depressurization time and no change is required in tne method Gg of plant recovery given in ERG E-1.
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1 b)
N-- No significant differences between the ERG reference plant and Maine Yankee were found in the method of analyzing larger sized LOCAs. Both assume the same single f ailure (loss of one LHSI pump) both account for Si spilling through the break, both use 1.2 times ANS Infinite decay heat, etc.
Both the ERG reference plant and Maine Yankee have the capability to vent the reactor vessel upper head region. Both switch automatically to the recirculation phase when the RWST switchover setpoint is reached. Also, both required switch to het leg recirculation at the required time following the LOCA.
In summary, no variation in analysis presented for a larger sized LOCA
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for Safety Analysis Reports for the ERG reference plant and Maine Yankee was found which would affect the method of plant recovery given in ERG E-l. AIso, no plant design variation exists which would significant-ly alter the plant response to a larger sized LOCA at Maine Yankee f rom that for the ERG reference plant. Thus no plant specific analysis for a larger sized LOCA is required for Maine Yankee.
3.4 Post LOCA Cooldown and Depressurization
- The WOG Post LOCA Cooldown and Depressurization ERG. ES-1.2. is structured primarily to cool down and depressurize the RCS to cold shutdown following smaller sized LOCAs. This is done by establishing a steam generator cooldown and selectively reducing safety injection flow by stopping safety
\ injection pumps sequentially or establishing normal charging flow if specified requirements are met.
In the guideline for post LOCA cooldown and depressurization the safety injection pumps are stopped and normal charging is established in a sequential manner. Since Maine Yankee HPSI flow can be throttled from the control room while the reference plant Si flow can not, the method of reducing safety injection flow or establishing normal charging flow during post LOCA coolcown and depressurization for Maine Yankee can be significantly different from the ERG reference plant and is to be analyzed for Maine tannee.
If the condenser is unavailable, then steam dump to atmosphere would be required to cooldown during post LOCA cooldown and depressurization.
Since Maine Yankee has less atmospheric steam dump capacity than the ERG reference plant (5% of total steam flow for Maine Yankee versus 10% of total steam flow for the ERG reference plant), the cooldown of Maine Yankee would be at a slower rate.
For larger sized small LOCAs analyzed for ERG ES-l.2 the RCS pressure decreased sufficiently to result in accumulator injection for the ERG reference plant. The ERG reference plant accumulator injection was initiated when the RCS pressure decreased to approximately 650 psig. l The Maine Yankee safety injection tanks are set to start injecting at 1 an RCS pressure of about 230 psig. Thus, for certain break sizes the l safety injection tanks would not inject into the RCS while the ERG i O, I reference plant accumulators would inject.
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1 O Finally, the overall effect on attaining cold shutdown due to Maine Yankee having less flow communication between the upper head fluid and the rest of the RCS, and having a thinner reactor vessel upper support plate than the reference plant (see Section 3.2) should be accounted for.
Because of the variations between Maine Yankee and the ERG reference plant itemized above, a plant specific post LOCA cooldown and depressurization analysis is recommended for Maine Yankee to determine the ef fect of these variations on recovery methods and operator actions. Cases for a smaller sized LOCA should be run both with and without offsite power available to demonstrate a successful recovery strategy. These cases should be run to conditions for initiating RHR.
3.5 Steam Generator Tube Ruoture (SGTR) in the ERG reference plant each steamline has an atmospheric relief valve and main steamline isolation valve (MSIV) with the relief valve upstream of the MSIV, in case of a steam generator tube rupture (SGTR) wi th the condenser unavailable, the steam generator with the rupture can be isolated, from the intact steam generators by closing the MSIV, atmospheric relief valve and bypass valves on the steamline with the ruptured steam generator.
These valves can be remotely closed from the control room. Plant cooldown is then accomplished by using the intact steam generator atmospheric relief valves. This Ilmits the release of radioactive steam to the atmosphere.
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At Maine Yankee, if the condenser is unavailable, the atmospheric steam dump valve (A50V) is used to cool down the plant by dumping steam to the atmosphere. There are no valves in the lines from each steam generator to the A50V which can be remotely closed to isolate the ruptured steam generator from this atmosgrieric release. The valve used to isolate-the ruptured steam generator from the intact steam generators must be closed locally. Also, the ASDV relief capacity is less than that for the ERG reference plant atmospheric relief valves. Thus, with the condenser unavailable, the cooldown would be at a slower rate at Maine Yankee.
Since Maine Yankee has less flow communication between the upper head fluid and the rest of the RCS as well as a thinner reactor vessel upper plate than the ERG reference plant (see Section 3.2), the overall effect of these upper head variations should also be accounted for in the cooldown and depressurization following a SGTR. Finally, since HPSI pump flow can be throttled from the control room at Maine Yankee and ERG reference plant 51 flow can not be throttled the method of reducing saf ety injection flow following a SGTR needs to be analyzed for Maine Yankee.
A plant specific SGTR analysis for Maine Yankee is recommended due to the plant variations discussed above. Three cases have been ider.tifled which should be analyzed.
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1 The first case would attempt to simulate the system response of the plant to the operator actions included .in the present Maine Yankee SGTR E0P. An analysis basis for the existing E0P is not available. It is anticipated that I unless this case is performed, extensive ef fort would be expended during procedure writing attempting to resolve conflicts between the existing j Maine Yankee E0P and Revision 1 of the ERGS.
The remaining two cases would predict the system response to a full double ended SGTR with operator actions consistent with the ERGS Rev.1.
Cases would be analyzed both with and without offsite power available.
3.6 Loss of all aC Power A loss of all AC power will result in a loss of all pumped injection including component cooling water and charging flow for reactor coolant pump seal injection. The resulting high pressure thermal transient could result in RCP seal degradation accompanied by an increase in RCP seal leakage and loss of RCS inventory. WOG guideline ECA-0.0 attempts to minimize seal degradation and the loss of RCS inventory with time. AC power must be restored to replenish lost RCS inventory and to re-establish RCP cooling.
l For WOG ERG support, best estimate analyses were performed for standard Westinghouse 2, 3 and 4 loop plants assuming a total loss of AC power.
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i Initial leakage rates of 5. 50 and 300 gpm per RCP were analyzed to cover l the expected to severely degraded seal cases. To minimize the RCP seal leakage, an operator controlled cooldown/depressurization of the RCS was simulated. The analyses showed that RCS cooldown/depressurization is beneficial in minimizing RCS inventory loss through the seals during a loss of all AC power event. Leakage rates for the Maine Yankee seals are less than the 300 gpm maximum used in the WOG analysis (Ref erence 9) .
Therefore. the Maine Yankee RCP seals' performance during a loss of l all AC power event is bounded by WOG analysis.
l l The capabilities of Maine Yankee for secondary depressurization are significantly different from a Westinghouse plant. The Maine Yankee atmo-spheric steam dump valve (ASOV) can handle 5% of total steam flow while the Westinghouse SG PORVs can handle 10%. The difference does not indicate a need for a plant specific analysis because the WOG analyses, as applied to the ERGS, identify only that RCS depressurization is beneficial, and and ERG ECA-0.0 has the maximum rate of secondary steam dump f easible established manually or locally to depressurize the plant.
Because of this and because the RCP seal leakage rates at Maine Yankee are below the 300 gpm maximum used in the WOG analyses, the WOG Ioss of all AC power analyses are found to be applicable to Maine Yankee. Therefore, no p plant specific analyses are recommended.
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4 MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1 3.7 Secondary Side Break The RCS response to loss of secondary coolant is dependent on the size and location of the break. Small secondary breaks may be compensated by normal plant control systems. Larger breaks will cause secondary depressurization resulting in a reactor trip and subsequent RCS cooldown/depressurization.
The WOG ERG analyses.for this topic included an intermediate (0.6 sq. f t) steamline break and a double ended feedline break. Best estimate modeling
+
assumptions for a 3 loop Westinghouse PWR were used in the analyses. The results of these analyses represent typical thermal hydraulic behavior for these transients.
'The Maine Yankee plant systems involved in these transients were compared to the Westinghouse generic plant systems. This comparison included the followings reactor trip actuation system. ESF actuation system, auxiliary feedwater system, main steam system, and the main feedwater and condensate system. The most notable difference is when main feedwater is-isolated. For the reference plant design, main feedwater is isolated on SIAS but for Maine Yankee it is isolated on low SG pressure. This difference has negligible effect on the analyses. All of the other systems were found '
to be similar to the reference plant as applied to these analyses.
I Because the Maine Yankee systems are similar to those used in the WOG secondary side break analyses, the WOG analyses are found to be applicable to Maine Yankee and no plant specific analyses are recommended.
O 3.8 Inadeavate Core Coolina (ICC)
The indication of an inadequate core cooling condition requires prompt operater action. Without adequate heat removal, core decay energy will cause fuel temperatures to increase thus endangering clad integrity. The purpose of the Response to inadequate Core Cooling ERG is to restore core cooling before fuel temperature renains above an elevated level for a sustained period.
i Four different recovery techniques were analyzed for the WOG ERGS. The recovery techniques included $1 reinitiation, rapid steam generator depressurization, restarting the RCPs, and opening the pressurizer PORVs.
The analyses were based on best estimate data for a 4 loop Westinghouse plant. Separate LOCA analyses were performed for a one inch and a four inch equivalent diameter break size. In order to achieve ICC conditions.
, a loss of all high pressure safety injection was assumed. Additionally, the accumulators were locked out in the four inch break analyses.
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION I .
Prior to the analyses, a core exit thermocouple temperature of 1200'F was chosen as an indication of an inadequate core cooling situation. This was chosen based on the errors and limitations of the core exit thermocouples so that there would be significant margin between design basis small LOCA thermocouple indications and inadequate core cooling -situation thermo-couple Indications. During the NRC review of the ERGS BASIC version, 1200*F. was found to be an acceptable symptom. The ICC analyses were performed to verify that the operator could recover the core after the core exit thermocouples reached a temperature of 1200*F. The analyses were also performed to determine the ef f ectiveness of each of the recovery techniques so that they could be prioritized for use in the guideline.
The third purpose of the analyses was to verify that each of the techniques is beneficial to the system even if it can not recover the core by itself.
The WOG ERG analyses showed that the core could be recovered if SI was i reinitiated soon after the core exit thermocouples reached 1200*F. Since
! the $1 system capacity at Maine Yankee is comparable to that for a Westinghouse plant, no plant specific analysis is required for this case.
The WOG ERG analyses showed that the core could also be recovered by secondary depressurization using the steam dump system after the core exit i thermocouples reached 1200*F. The pressure of the RCS was lowered enough to actuate accumulator injection which recovered the core soon after initiation of accumulator injection. However, the set pressure (230 psig)
() of the Maine Yankee safety injection tanks is significantly different f rom the set pressure of the ref erence plant accumulators (650 PSIG) .
Because of this difference, the WOG analyses of core recovery using secondary depressurization do not apply to Maine Yankee. Plant specific analysis is required to verify-that Maine Yankee can recover the core by secondary depressurization after the core exit thermocouples reach 1200*F.
The third recovery technique analyzed for the reference plant'was restart of a reactor coolant pump. Restarting an RCP recovered the core quickly for a one inch break and a long term recovery was predicted when the accumulator injected because the system depressurized to the set l
point pressure before the core exit thermocouples reached 1200'F a second time for the four inch break, restart of an RCP did not recover the core because of the small amount of inventory remaining in the l system and the accumulators being locked out. The effectiveness of the l recovery technique depends on system conditions when actions are performed. Secondary depressurization may be required after the RCP is restarted to achieve a long term core recovery for Maine Yankee due to its lower safety injection tank set pressure.
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MAINE YANKEE PLANT TECHNICAL Gul0ELINES REVISION 1 b
o Because of the difference between the set pretsure of the reference plant accumulator tanks and the set pressure of the Maine Yankee safety inject-Ion tanks, plant specific analysis is recommendsd for the restart of an RCP. Since the reference plant analyses showed that a secondary depress-urization may be required for long term recovery after RCP restart, the recommended plant specific analysis is RCP restart followed by a secondary depressurization. This depressurization should be performed by locally opening the ASDV (the steam dump system is not working otherwise the operator would already be using it) . The time delay between the RCP restart and the ASDV opening should reflect the time required for plant personnel to locally open the ASDV. The analysis is required to verify that Maine Yankee can achieve a long term ccre recovery with a combination of RCP restart and secondary depressurization after the core exit thermocouples reach 1200*F.
The fourth technique analyzed was the opening of the pressurizer PORV's.
The technique was found to be the least desirable of the recovery techniques analyzed. For both the one inch and the four inch breaks studied, the generic analyses showed that core recovery cannot be achieved by opening the pressurizer PORV's alone. Long term recovery was achieved only af ter accumulator injection or LHSI operation. However, the analysis Indicated that this recovery technique is beneficial in speeding up the RCS depressurization to reach the accumulator injection set pressure or LHSI. $1nce other actions are required to recover the core and since the conclusion that opening the PORV's is, in general.
() beneficial to the system, no plant specific analysis is recommended for this action. Although analysis is not recommended, it can be performed to determine how much time will pass before the safety injection tanks inject into the RCS after the PORV's are opened to depressurize the system.
3.9 Anticipated Transient Without Scram (ATWS)
The analytical basis for the existing Maine Yankee ATWS related steps can be found in References 14, 15, and 16 as confirmed by Reference 10.
The existing Maine Yankee ATWS related steps will be used for the basis for the upgraded E0Ps. Thus, no further plant specific analyses for ATWS are required.
3.10 Loss of (Secondary) Heat sink (LOHS)
A loss of secondary heat sink is characterized by decreasing steam generator water inventory (l eve l) caused by reduced or loss of feedwater supply to the steam generators followed by increasing RCS pressure and temperature. A " bleed and feed" recovery technique is used if emergency or auxiliary feedwater is unavailable or ineffective in restoring the secondary heat sink. This technique involves initiating high pressure safety injection and manually opening the pressurizer PORV's to provide core cooling.
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1 7
Analyses were performed for generic Westinghouse plants to determine the time at which RCS " bleed and feed" must be initiated to prevent sustained core uncovery. For high pressure safety injection plants (plant designs for which the shutoff head of the highest discharge pressure saf ety injection pumps is greater- than nominal operating primary system pressure), analyses were performed for a f our-loop. 3411 MWt plant. A range of pressurizer PORV capacity-to-core power ratios were analyzed. These analyses showed that for plants with a pressurizer PORV capacity-to-core power ratio greater than 140 pounds per hour per MWt. " bleed and feed" can be initiated when the steam generators' secondary sides dry out. For siellar plants with ratios less than 140 pounds per hour per MWt. " bleed and feed" must be initiated before the steam generators dry out. The lowest possible steam generator water level allowable before " bleed and feed" is initiated must be determined on a plant specific basis.
Maine Yankee has a PORV capacity-to-core power ratio less than 140 pounds per hour per MWt. This would indicate for a Westinghouse plant that " bleed and feed" must be initiated before the steem generator secondary sides dry out. Plant specific analysis is recommended for Maine Yankee to see if this is the case. The analyses should first check if steam generator dryout can be used as a symptom for " bleed and feed" initiation. If steam generator dryout is not suitable, the plant specific minimum steam generator water level that is acceptable for use as a symptom must then be determined.
(l 3.11 Reactor Coolant Pump (RCP) Trio RCP trip analyses to respond to Generic Letter 83-10d were separately committed outside the scope of this E0P upgrade. The results of these analyses will be reported separately and will be incorporated into the E0Ps.
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MAINE YANKEE PLANT TECHNICAL Gul0ELINES REVISION I O
V 4. BASl5 FOR USING'THE GENERIC WESTINGHOUSE ERGS To the greatest practicable extent, the Westinghouse Emergency Response Guidelines (ERGS) have been constructed to be generic. It can be seen from the comparison made in Section 2 that the Maine Yanke's plant is very slallar to the reference plant. which was used as the basis for developing the ERGS. Also, as noted in the analyses discussion provided in Section 3. certain analysis performed to support the generic ERGS are applicable to Maine Yankee. Additional plant specific analyses have also been identified. Therefore, the Maine Yankee E0Ps will be based upon the generic Westinghouse ERGS HP-Revision I and plant specific analyses IIsted in Section 3. When writing the E0Ps. modifications to ERG steps must bs made to account for the Maine Yankee plant design differe'nces which are delineated in Appendix A.
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MAINE YANKEE PLANT TECHNICAL GUlOELINES REVISION 1
/T s s- 5. METH00 FOR DEVELOPING E0Ps iROM ERGS 5.1 General Generic Westinghouse Emergency Response. Guidelines (ERGS), Revisica I will be used as the basis for writing the E0Ps for the Maine Yankee plant. A list of the Revision 1 ERGS to be used to develop Maine Yankee E0Ps is included as Appendix B.
This section describes the method that will be used to convert the applicable generic guidelines into E0Ps.
5.2 Prenaration The FDP writing team will obtain and review the following source documents for Maine Yankees Applicable Westinghouse generic ERGS, Rev. I and background documents Maine Yankee Plant Specific Technical Guidelines Maine Yankee Writers Guide for E0Ps '
Technical Specifications Final Safety Analysis Report Engineering . low Diagrams Maine Yankee Systems Training Manual
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Existing E0Ps (See letter MN-83-84, 5/17/83)
Calculated Mathematical Values used in E0Ps (included in Background information Manual) 5.3 Method The E0P write s will follow the applicable ERGS step by step. The writer will research the source documents and then construct the EOP and an associated E0P Documentation Form (Appendix C) . This Maine Yankee Background Documentation Form will list how each generic guideline step is used in the E0P and also list any additional steps added to the E0P with its basis, if applicable. Any difference betwsen the ERG step and the Maine Yankee step will be explained.
This form along with the calculation for mathematical values used in the E0Ps will be kept in the Background Information Manual for the Maine Yankee EOPs.
The following additional instructions for writing the E0Ps and completing the EOP Documentation Form are provided.
- 1. If the generic step is compatible with the Maine Yankee plant design, then the step should be copied into the Maine Yankee E0P.
Since the technical basis for the step is explained in the ERG Background Document, there is no need to repeat this on the background documentation form.
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1 7
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- 2. When an ERG step specifies a numerical value to be calculated, the value will be determined and put into the Maine Yankee E0P. The documentation form should indicate where the method of derivation is located.
- 3. When an ERG step requests plant specific details or actions to be added to the procedure, add the information to the procedure.
However, if the operator actions are highly routine or well within the knowledge of the operator, the specific information should not be included. The reason for this should be explained on the documentation form.
- 4. If the ERG guideline falls to identify or address systems or actions that are unique to Maine Yankee (Refer to Appendix A),
then steps should be included to encompass the necessary actions. These should be explained on the documentation form.
- 5. If an ERG step specifies an action that cannot be performed at the Maine Yankee plant, the step will be deleted or modified and the reason explained on the documentation form.
- 6. .lf an ERG step is modified such that the intent of the ster is 4 changed, then the basis will be explained on the documentation form.
T 7. Minor modifications to ERGS steps are acceptable without-extensive justification provided that the change does not alter the intent of the guideline. Examples of these types of changes are as follows:
- a. Deletions of level of detail (see i tem #3) .
- b. Rewording of ERG steps to conform to standard Maine Yankee terminology, abbreviations and acronyms.
- c. Rearranging ERG steps to reflect Maine Yankee control room design and for operator convenience if it does not affect plant recovery via the ERG.
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1
- 6. CONCLUSION For the Maine Yankee plant, the generic WOG ERGS Revision 1 listed in Appendix 3 will be used as the basis for writing the plant specific E0PS.
This document provides a description of the planned method for developing the Maine Yankee E0Ps from the generic WOG guidelines. Also, deviations from an emergency operations perspective resulting from differences between ,the reference plant and. Maine Yankee designs have been identified.
it is intended that this document along with the Maine Yankee Writer's Guide for E0Ps will be used to aid in the preparation of the Maine Yankee E0Ps.
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MAINE YANKEE PLANT TECHNICAL GulOELINES REVISION 1
- 7. LIST OF REFERENCES
- 1. Maine Yankee Atomic Power Station Final Safety Analysis Report.
Volumes 1, it, and Ill.
- 2. Maine Yankee Cycle 8 Core Performance Analysis. January 1984.
- 3. Maine Yankee Atomic Power Station Technical Specifications.
. Amendment No. 78.
- 4. Maine Yankee Systems Training Manual (Volumes One through Ten)
- 5. Drawing Number 11550-FM-70A. Revision 20
- 6. Drawing Number 11550-FM-9A-S.C.. Pevision 17
- 7. Letter GNS-84-23. (Stowers to Fuoto) December 14, 1984
- 8. Westinghouse Owners Group Emergency Response Guidelines Revision 1 (HP Version)
- 9. Letter RPJ-85-07. (Jordan to Fuoto) March 29. 1985
- 10. Letter RPJ-85-08. (Jordan to Fuoto) March 29. 1985
- 11. CEN-152. Rev. 0 " Emergency Procedure Guidelines". CE Nuclear Power Systems Division June 1981
- 12. Letter NSS-A-81-021. (Burchill to CE Owners' Group) . "Draf t Reactor Protection System Failure Guidelines" January 28. 1981
- 13. Letter CE-ATWS-008. (DeLozier to distribution) . "ATWS Review of RPS Failure Guideline" September 29, 1980
- 16. CENPD-263-P, "ATWS Analysis: Response to NRC Letter of 02/15/79.
Alterrete 3 Analysis per NUREG-0460. Volume 3 for CE NS$$'s". CE Nuclear Power Systems Division October 1979
- 17. Letter FSD-RSA-84/2062. (Anderson to Jones) . " Supplementary Information for Neutron Flux Comparisons at Survalliance Capsules" May 11. 1984
- 18. Letter GNS-85-12. (Stowers to Fuoto), " Answers to Questions Raised 1 During Meeting of 4/23/85" April 26. 1985
- 19. Letter RPJ-85-13. (Jordan to Fuoto), " Transmittal of PTS Ref erences" s_ May 6. 1985 1
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1 APPENDIX A COMPARISON OF SYSTEM DESIGNS (OESIGN DIFFERENCES)
Westinghouse Reference Plant Maine Yankee (Maine Yankee terminology used)
REACTOR TRIP ACTUATION SYSTEM o Reactor Trip Signal Same (I) o Turbine Trip Signal Same ESF ACTUATION SYSTEM o Si Actuation from manual, high containment Si actuation from
. pressure, low pressurizer pressure, & manual, high contain-low steamline pressure signals. ment pressure, & low pressurizer pressure signals, o 51 Roset by using reset buttons. SIAS cleared using reset Block interlocked with pressurizer switches when above, pressure permissive setpoint. SIAS pressurizer pressure
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block automatically cleared.when SIAS AUTO UNBLOCK pressurizer pressure permissive setpoint with high setpoint exceeded. containment pressure signal not present. OR
() by manual block of low pressurizer pressure SlAS when below BLOCK ENABLE setpoint. SIAS BLOCK automatically cleared when pressurizer pressure greater than SIAS AUTO UNBLOCK setpoint.
o Feedwater isolation Signal on SIAS. Feedwater isolation does not occur on SIAS. 1 (See p.35. MFCS) o Main Steamline Isolation Signal on Main Steamline low SG pressure which also causes SlAS. Isolation Signal Block interlocked with pressurizer on low SG pressure pressure permissive setpoint. SlAS which trips main block automatically cleared when feed system and pressurizer pressure permissive isolates all feed setpoint exceeded. to affected SG.
Block and reset interlocked with SG pressure permis-sive. Automatic signal can be de-feated on component level.
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1 Westinghouse Reference Plant Maine Yankee (Meine Yankee terminology used)
ESF ACTUATION SYSTEM (Cont.)
o Cont. Isolation Signal and Roset Cl$ block stellar to SIAS block.
o Cont. Spray Actuation Signal on manual CSAS on manual or high and high containment pressure setpoint containment pressure with reset allowed below actuation coincident with SlAS.
setpoint pressure. Roset allowed when SIAS cleared and containment pressure below actuation setpoint.
NUCLEAR INSTRUMENTATION SYSTEM o Source Range Startup Rate Same o Neutron Flux Recorder Same o Separate ranges betwsen source. Wide range log scale Intermediate, and power ranges with that automatically Interlocks and permissives to change changes range.
range.
CONTROL ELEMENT INSTRUMENTATION SYSTEM o Control Element Position Some o Control Element Bottom Lights Same CONTAINMENT INSTRUMENTATION SYSTEM o Containment Pressure Same o Containment Temperature Same o Containment Recirculation Sump Level Same t
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_ _ _ _ _ _ _ . _ _ - . _ - . _ _ -- - u
3 MAINE
- YANKEE PLANT TECHNICAL GUIDELINES REVISION 1
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\-- Westinghouse Reference Plant Malne Yankee (Maine Yankee terminology used)
REACTOR COOLANT SYSTEM o 4-Loop 3-Loop o Hot & Cold Leg RTD Bypass Hot & Cold Leg RTD directly mounted in RC Piping on SG side of LSVs. I o Two PORVs & Associated Block Valves Two PORVs and Sized at a Nominal 240.000 lbm/hr Associated Block (saturated steam) each at a nominal Valves slaed at a actuation setpoint of 2335 PSIG. nominal 150.000 lbe/hr (saturated steam) each at a nominal setpoint of 2385 PslG.
o RV Head Vent to Containment RV Head Vent to PressurI2er Quench Tank-o RVLIS (3 ranges) Primary inventory Os Trending System (PITS) (3 ranges)
(2) o Nominal hot Zero power Nominal hot Zero conditions of 557'F. power conditions 2235 PSIG of 532'F.
2235 PSIG o N/A One hot leg LSV.
one cold leg LSV and one loop bypass valve (all motor operated) per reactor coolant loop.
o Flow no221es from reactor vessel inlet N/A (3) plenum to reactor vessel upper head o Reactor vessel upper support plate is RV upper plate of 12 inches thick or 5 inches thich the upper support structure is k inches thick. I g
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1 Westinghouse Reference Plant Maine Yankee (Maine Yankee terminology used)
CHEMICAL & VOLUME CONTROL SYSTEM o Two Centrifugal Charging Pumps which Two Centrifugal are also used for St Charging Pumps plus one Installed spare which are also used for Si o One PD Pump Same o Charging & RCP Seal injection using Same one Charging Pump o Letdown-Regenerative HX. Letdown HX Same to VCT '
o 4% Boric Acid System 6.25%
o Boric Acid Pumps Supply Charging Pumps Same through either normal Reactor Makeup Water System or Emergency Soration Path EMERGENCY CORE COOLING SYSTEM High Pressure Saf ety injection o Two Charging /SI Pumps take suction Two Charging /SI from RWST or Low-Head St Pumps Pumps plus one Installed spare take suction f rom RWST or the containment spray pumps.
o Charging /SI Pumps shutoff Head > RCS Same
. Design Pressure o 12% SIT is injected by Charging /$1 N/A Pumps to Cold Legs o 81T Contents are circulated by 2 Boron N/A injection Recirculation Pumps o N/A HPSI Pump flows can be throttled.
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MAINE YANKEE PLANT TECHNICAL Gul0ELINES REVISION 1 Weetinghouse Reference Plant Maine Yankee (Maine Yankee terminology used)
EMERGENCY CORE COOLING SYSTEM (Cont.)
Intermediate Pressure Safety injection o Two 51 Pumps separate from the charging / N/A Si pumps with shutoff head of approximately 1600 psig.
o $1 Pumps separate from the Charging /SI N/A take suction from the RWST or Low-Head
$1 Pumps.
o Suetions of Charging /SI and $1 Pumps sep- N/A arate from Charging /SI Pumps Connected.
o 51 Pumps separate from the Charging /SI N/A Pumps deliver to Cold Legs (through Accumulator lines) and all Hot Legs Low Pressure Safety i nj ection o Two LPSI Pumps Two LPSI/RHR Pumps
'h (N/ o Low-Head $1 Pumps take suction from RWST Same or Containment Sump o Low-Head $1 Pumps deliver to Cold Legs and Deliver to Cold Concurrently feed Charging /51 and High-Head Legs only 51 Pumps (Cold Leg Recirculation Mode) o Low-Head St Pumps deliver to Hot Legs and Hot Leg recircul-Concurrently feed charging /51 and High-Head ation via HPSI St Pumps (Hot Leg Recirculation Mode) pumps (which s imul-taneously deliver flow to the hot legs and cold legs) and containment spray pumps.
o Switchover initiation-Automatic Sump LPSis stopped.
Valve Opening RWST suction closed on RAS.
LPSis may be manually restarted on recirculation.
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1 Westinghouse Reference Plant Maine Yankee (Maine Yankee terminology used)
EMERGENCY CORE COOLING SYSTEM (Cont.)
Safety injection Tanks o One Accumulator Tank per loop with One Safety Nitrogen Cover Gas at a nominal Injection Tank 650 PSIG with a nominal 1400 FT3 per loop with volume (800 gas. 600 water) . Nitrogen Cover Gas at a nominal 230 PSIG with a nominal 3500 FT3 volume (2000 gas, 1500 water).
RESl0UAL HEAT REMOVAL SYSTEM o Two Low-Head Pumps. Same o Low-Head Pumps take suction from Two Hot Low Head Pumps Legs and return Flow to Four Cold Legs take suction from Loop 2 and can return flow to all three cold legs
( 4 (normally only two used).
o RHR initiated at nominal 400 PSIG, RHR Initiated at 350'F RCS conditions nominal 350 PSIG, 400*F RCS conditions.
RADIATION MONITORING SYSTEM o Condenser Air injector Monitor Same o SG Blowdown Monitor Same o Containment Atmosphere Monitor Same o Auxillary Building Monitor Same o SG Steamline Monitor Same O I t
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MAINE YANMEE PLANT TECHNICAL GUIDELINES REVISION 1 O
Westinghouse Reference Plant Maine Yankee (Maine Yankee terminology used)
CONTAINMENT SPRAY SYSTEM o Two Low-Head Containment Spray Pumps Two Low-Head Containment Spray Pumps out of a total of three (one is an installed spare) aligned to feed Charging /SI pumps suction and containment spray after RAS occurs.
CONTAINMENT AIR RECIRCULATION SYSTEM o Four Emergency Fan Coolers Six one-fifth (two speeds) capacity recirculation fans o Two Hydrogen Recombiners - N/A (5)
Manual Actuation
- O EMERGENCY AND AUXILIARY FEEDWATER SYSTEMS o Two Motor Driven Pumps (Emergency) Same o One Steam Driven Pump (Auxillary) Same o Domineralized Water Storage Tank for Sama preferred water supply o Alternate Water Supply Same o AFW Control Valves $ame o AFPs start on manual, low SG water MD AFPs start on manual level. Si actuation and loss of and low SG water level offsite power (T/D AFP) signals signals. TD AFP starts manual signal only. (6)
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MAINE YANKEE PLANT TECHNICAL Gul0ELINES REVISION 1 Westinghouse Reference Plant Maine Yankee (Maine Yankee terminology used)
MAIN STEAM SYSTiM o Steam Generator PORVs sized at Common header to 10% nominal loop flow rate each atmospheric steam (1 per loop) with a nominal setpoint dump valve sized 25 PSIG greater than hot zero power at a nominal 5%
steam pressure and clso capable of total steam flow.
being manually controlled for plant Manual control only.
cooldown. The PORVs have no interlock Interlocked with with safeguards signals. CIS/SIAS. (7) o Steam Generator Safety Valves Same o Condenser Steam Dump Valves Same o Main Steamline Isolation Valves Same o Main Steamline Bypass Valves Same (8) ,
COMPONENT COOLING WATER SYSTEM o CCW Pumps (PCC and SCC) Same o RCP Thermal Barrier Valves Same SERVICE WATER SYSTEM o Service Water Pumps Same o Service Water Valves Same MAIN FEEDWATER AND CONDENSATE SYSTEM o Main Feedwater Condensate, and Same Heater Orain Pumps o Feedwater Regulating Valves Same o Bypass Feedwater Regulating Valves Same o Feedwater Isolation Valves Same o All Main Feedwater Reg Valves closed on All FW reg valves low RCS temperature signal following close after reactor reactor trip trip then bypass 1 FW reg valves open to 354.
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MAINE YANKEE PLANT TECHNICAL Gul0ELINES REVISION 1 O' Westinghouse Reference Plant Maine Yankee (Maine Yankee terminology used)
MAIN FEEDWATER AND CONDENSATE SYSTEM (Cont.)
o Feedwater Isolation on SIAS and high SG Main FW & condensate water level signals pumps trip on low SG pressure coincident with SIAS. All FW l isolated to affected SG. FW isolation on high SG water level.
STEAM GENERATOR BLOWDOWN SYSTEM l o SG Blowdown Isolation Valves Same SAMPLING SYSTEM o SG 81owdown Sample Isolation Valves Same SPENT FUEL STORAGE AND COOLING SYSTEM o Spent Fuel Pit level Same
(T
\ms/
CONTROL ELEMENT ORIVE MECHANISM COOLING SYSTEM o Control Element Drive Mechanism Fans Same CONTROL ELEMENT ASSEMBLY CONTROL SYSTEM o Control Element Assemblies Same TURBINE CONTROL SYSTEM o Turbine Runback N/A ELECTRIC POWER SYSTEM o Two Diesel-Generators Same o Automatic start signals on loss of Automatic start signal offsite power or on SlAS on loss of offsite power.
PNEUMATIC POWER SYSTEM o instrument Air Compressor Same o Instrument Air Valves Same O I t
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MAINE YANKEE PLANT TECHNICAL GUl0ELINES REVISION 1 O' Notes on systems comparison informations (1) The differences in these signals only affects the list of reactor trip signals which are entry conditions to the 10Ps.
These dif ferences do not af fect the body of the procedure set.
(2) PITS has three ranges:
A. Top of the pressurizer to the bottom of the reactor vessel.
B. Top of the reactor vessel to the Loop 2 Hot leg pipe.
C. Top of the reactor vessel to the bottom of the reactor vessel.
I (3) Two RTDs (TE-145A. TE-1453) are located in the reactor vessel upper head to measure upper head fluid toeperature and are inputs to the Subcooling Margin Monitor (SMM) .
(4) HPSI can be manually throttled after use of key switches to bypass SI AS to valves HSI-M-kl and HSI-M-42.
l (5) Two taps exist for hooking up hydrogen recombiner.
(6) Main steam supply to TO AFP isolated on SIAS/CIS. -
(7) SG's must be locally Individually isolated from the atmospheric steam dump valve and the aux feed pump turbine steam supply by use of MS-59 (Loop 1) . MS-79 (Loop 2) , or MS-99 (Loop 3) .
(8) Main steam bypass valves (MS-51. MS-71. MS-91) are locally operated and are normally closed.
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1 O
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1 s_/ APPENDIX St LISTING OF APPLICABLE EMERGENCY RESPONSE GUIDELINES (REV.1)
APPLICABLE OPTIMAL RECOVERY GUlOELINES E-0 Reactor Trip or Safety injection ;
ES-0.0 Redlagnosis j E3-0.1 Reactor Trip Response 1 ES-0.2 Natural Circulation Cooldown l ES-0.3 Natural Circulation Cooldown with Steam Veld in Vessel (with RVLIS) l ES-0.4 Natural Circulation Cooldown with Steam Void in Vessel (wi thout RVLIS) l l
E-1 Loss of Reactor or Secondary Coolant ES-1.1 51 Termination ES-1.2 Post-LOCA Cooldown and Depressurization ES-1.3 Transfer to cold Leg Recirculation ES-1.4 Transfer to Hot Leg Recirculation E-2 Faulted Steam Generator Isolation E-3 Steam Generator Tube Rupture ES-3.3 Post-SGTR Cooldown Using Steam Dump ECA-0.0 Loss of All A.C. Power ECA-0.1 Loss of All A.C. Power Recovery Without 5.1. Required ECA-0.2 Loss of All A.C. Power Recovery With S.,l. Required ECA-l.2 LOCA Outside Containment l
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MAINE YANKEE PLANT TECHNICAL GUIDELINES REVISION 1 APPLICABLE FUNCTION RESTORATION GUlOELINES F-0 The Critical Safety Function Status Trees F-0.1 Suberiticality s
F-0.2 Core Cooling F-0.3 Heat Sink F-0.4 Integrity F-0.5 Containment F-0.6 Inventory FR-5.1 Response to Nuclear Power Generation /ATVS FR-5.2 Response to Loss of Core Shutdown FR-C.1 Response to inadequate Core Cooling FR-C.2 Response to Degraded Core Cooling FR-C.3 Response to Saturated Core Cooling FR-H.1 Response to Loss of Secondary Heat Sink FR-H.2 Response to Steam Generator Overpressure FR-H.3 Response to Steam Generator High Level FR-H.4 Response to Loss of Normal Steam Release Capabilities ,
FR-H.5 Response to Steen Generator Low Level FR-P.1 Response to Iseninent Pressurized Thermal Shock Condition FR-P.2 Response to Anticipated Pressurized Thermal Shock Condition
, FR-Z.1 Response to High Containment Pressure
-FR-Z.2 Response to Containment Flooding FR-Z.3 Response to High Containment Radiation Level FR-l.1 Response to High Pressurizer Level FR-l.2 Response to Low Pressurizer Level FR-l.3 Response to Volds in Reactor Vessel O1, 1
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MAINE YANKEE PLANT TECHNICAL GulOELINES REVISION 1 APPEN0lX C: MAINE YANKEE E0P OOCUMENTATION FORM Page __ of__
MAINE YANKEE E0P DOCUMENTATION FORM b
EOP No. Rev.
Title Prepared by: Date MAINE YANKEE ERG
() STEP N0. STEP NO. EXPLANATION OR BASIS FOR OlFFERENCE O
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MAltE YANKEE ,
i EERGENCY ORRATING PROCEDURES I- l l TASKANALYSIS oO I
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APPLICABILITY 0F TASK ANALYSIS
,. m WHICH StPPORT MAIPE YAtKEE'S E0PS l
During Maine Yankee's DCRDR program a Task Analysis was performed on Maine Yankee's plant, utilizing a multi-disciplinary team, including personnel
-knowledgeable in Maine Yankee's Transient Analysis and qualified Human Factors expertise. The methodology employed in the DCROR Task Analysis have been documented in Maine Yankee's DCROR submittal,' reviewed by the NRC and found to be acceptable. Westinghouse in the development of the generic ERG and FRG sets performed a Systems Review and Task Analysis (SRTA) down to the level of
-generating a list of required Instruments and Controls. This SRTA has also '
been reviewed by the NRC and accepted, in that the NRC accepted the generic ERG and FRG sets.
When Maine Yankee employed Westinghouse to develop Maine Yankee's Emergency Operating Procedures, we realized that a similiarity in plants existed and that any differences had to be documented and accounted for.
Three areas of' prime importance were then identified as follows:
- 1) Plant Differences on a systems level
- 2) Analysis' applicability
- 3) Instruments and Controls differences Sections 2 and 3 of the Maine Yankee Plant Technical Guidelines address the first two of the three areas. The third area consisted of the identification and evaluation of Instrumentation and Control differences.
0938A-TIC
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This program started by bouncing Westinghouse's SRTA derived list of required Instrunents and Controls against! Maine Yankee's Task Analysis derived list of required Instruments and Controls and documenting any differences. The differences then fell into several categories with each to be reviewed in further depth to determine whether the difference presented a problem that required correction.
The first category are Instruments and Controls that perform identical functions but have different names. The differences in plant terminology can be easily identified and essentially thrown out as not
/~') _being a problem.-
V The second category are those Instrument and Control differences due to different requirements of different systems. If Maine Yankee doesn't require a particular system or trip scheme due to plant analysis requirements then the Instrument and Control requirements would be different for those systems.
- The third category are those extra Instruments and Controls that Maine Yankee requires but that Westincflouse does not. Maine Yankee's list of required Instruments and Controls is longer then Westinghouse's and can be attributed to the conservativeness of Maine Yankee's task analysis.-
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- The fourth category consists of those Instruments and Controls required by Westinghouse which are not required by Maine Yankee but are available in the plant. For this category of equipment a task analysis would have been performed to enable a comparison of the required I&C characteristics with those on the actual plant equipment. Any deficiencies would have been assessed for importance to safety and resolved as necessary.
- Category five consisted of those _ Instruments and Controls that Westinghouse requires, that Maine Yankee doesn't require, and also that the actual plant doesn't have. For the EOP ' work and the
_( requirements for certain instruments and controls to be present on the Maine Control Board at Maine Yankee, Westincflouse's SRTA would take precedent. If the comparison of extra Westinghouse required equipment with the actual M.Y. plant inventory found the instrument missing, then Maine Yankee would have performed a task analysis on those areas in the Westinghouse E0P's where the instrument was required. This task analysis would have resulted in determining the characteristics required for the instrument to facilitate purchase.
The first three categories are differences that are acceptable with no further work required. The equipment in these categories will continue to receive inprovements as stipulated by Maine Yankee's DCRDR program.
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g This methodology assumes that if the plant's are identical, if ~they respond identically and if the instruments and controls' are identical then the characteristics for those instruments and controls should be identical. If the assumption is true, then Maine Yankee's task analysis applies and the results can be used to improve the existing I&C equipment and enhance the operators ability to rbitigate plant casualties using Westinghouse ERG based procedures.
The results obtained from implementing this methodology are on file as part of the DCRDR documentation. The bottom line is a total of 74
~'N differences, reviewed indepth, of which all are either differences in (d
terminology, category 1, or plant systems, category 2. Several category 3 differences were evaluated and set aside as not being a problem. No category 4 or 5's were found thus verifying the close similiarity between plants as well as the conservativeness of Maine Yankee's Task Analysis. Any of the category 1, 2 and 3 Instrument and Control differences will continue to be improved through the DCRDR Progrm and thus be usable in conjunction with the New Westinghouse E0Ps. Maine Yankee is confident that the comparison of Westinghouse's Generic Plant Systems, applicable analysis and Instrument and Control requirements with Maine Yankee's Plant Specific information, along with the resolution of differences, fully address and supports the applicability of Maine Yankee's Task Analysis to our Westinghouse ERG based emergency operating procedure.
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MAINE YANKEE EERGENCY OPERATING PROCEDURES SUPPLIENTAL TRANSIENT. ANALYSIS
- O~ .USING TREAT CODE-I
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N VERIFICATION OF THE E0Ps FOR MAINE YANKEE USING TREAT CODE By James D. Robichaud Nobuyuki Fujita Prepared By: h. /). 8 mal , //- /f- FI games D. Robichaud Engineer (Date)
Prepared By: ( NNM i (
Nobuyuki fujita, Nudiear Engineer (Date) pr y . // /S 8[
Paul A. Berbron,jfanager IDatd)
Transient Analysit Group Approved By:
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Bruce C. Slifer, Manag '( Da t'e )
Nuclear Engineering De tment Yankee Atomic Electric Company Nuclear Services Division 1671 Worcester Road Framingham, Massachusetts 01701 l O
DISCLAINER This document was prepared by Yankee Atomic Electric Company (Yankee) pursuant to a contract between Yankes and Maine Yankee Atomic Power Company (Maine Yankee). The use of information contained in this document by anyone other than Maine Yankee', or for any purpose other than for which it is i intended, is not authorized, and with respect to any unauthorized use, neither Yankee nor its officers, directors, agents, or employees assume any obligation, responsibility, or liability, or make any warranty or representation concerning the contents of this document or its accuracy or completeness.
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i ABSTRACT In order to provide a quantitative basis for improving the pnergemey Operating grocedures (EOPs) for Maine Yankee, four events were selected for analysis with a best-estimate model using the TREAT computer code. The events selected were recovery from: Steam generator lube Rupture (SGTR), Inadequate Gore Gooling (ICC) condition, koss gf Secondary Heat link (LOHS), and a small break Loss-af-goolant Accident (LOCA). The results of the analyres demonstrate that the revised Maine Yankee E0Ps are acceptable to enable plant recovery from the above accident scenarios. ;
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TABLE OF CONTENTS Pale
() DISCLAIMER........................................................ 11 ABSTRACT................ ........................................ iii TABLE OF CONTENTS................................................. iv LIST OF ACRONYMS.................................................. vi LIST OF TABLES.................................................... vil i
LIST OF FIGURES................................................... viii
1.0 INTRODUCTION
...................................................... 1 2.0 GENERAL CODE DESCRIPTION.......................................... 3 3.0 MODEL DEVELOPMENT................................................. 5 3.1 Model Descriptions.......................................... 5 3.1.1 Nodalization........................................ 6 3.1.2 Core Model.......................................... 8 3.1.3 Plant Control Systems............................... 9 3.1.4 Steady-State Full Power Mode 1....................... 10 N 3.2 Model Verifications......................................... 11 j
3.2.1 Reactor Scram From Full Power....................... 11 3.2.2 Pump Coastdown Test................................. 12 3.2.3 RELAPSYA SGTR Test.................................. 12 4.0 E0P ANALYSES...................................................... 21 4.1 SGTR Analysis............................................... 21
-4.1.1 Assumptions......................................... 23 4.1.2 Re su l t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 4.1.2.1 SGTR (Recovery Action - Auxiliary Spray /ADV)................................ 26 4.1.2.2 SGTR (Recovery Action - One PORV/ADV)..... 29 4.1.2.3 SGTR (Recovery Action - Normal Spray /
Steam Dump)............................... 30 4.1.3 Conclusion.......................................... 31 4.2 ICC Analysis................................................ 58
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TABLE OF CONTENTS (continued)
PAge 4.2.1 Assumptions......................................... 58 4.2.2 Results............................................. 59 4.2.2.1 ICC (Recovery Action'- Secondary Depressurization).......................... .
60 4.2.2.2 'ICC (Recovery Action - RCP Re start /
Delayed Secondary Depressuriration)........ 60 4.2.2.3 ICC (Recovery Action - Primaty Depressurization).......................... 61 4.2.3 Conclusion.......................................... 62 4.3 LONS Analysis............................................... 79 4.3.1 Assumptions......................................... 80.
4 . 3.2- Results............................................. 81 4.3.2.1 LOHS (Without Steam Dump and Bypass)....... 81 4.3.2.2 LOMS (With Delayed Feed and Bleed)......... 82 4.3.3 Conclusion.......................................... 83 4.4 Small Break LOCA Cooldown/Depressurization Analysis......... 94 4.4.1 Assumptions......................................... 95 4.4.2 Results............................................. 95 4.4.3' Conclusion.......................................... 96 5.0
SUMMARY
AND CONCLUSION............................................ 104
6.0 REFERENCES
........................................................ 106 4
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LIST OF ACRONYMS O ADV Atmospheric Dump Valve (Decay Heat Relief Valve)
V AFW Auxiliary Feedwater (Emergency Feedwater)
BE Best-Estimate l CE Combustion Engineering D Dimension.
DEGB Double-Ended Guillotine Break ECCS Emergency Core Cooling System EFCV Excess Flow Check Valve E0P Emergency Operating Procedure ERG Emergency Response Guideline FW Feedwater FWRBV Feedwater Regulator Bypass Valve HFP Hot Full Power NZP Hot Zero Power ICC Inadequate Core Cooling LOCA Loss-of-Coolant-Accident LOHS Loss of Heat Sink LOOP Loss of Off-Site Power LPSI Low-Pressure Safety Injection MFWRV Main Feedwater Regulator Valve NWt Mega-Watt Therinal
- NE Non-Equilibrium ,
NR Narrow Range NRV Nonceturn Valve PORV Pilot-Operated Relief Valve (Power-Operated Relief Valve)
PWR Pressurized Water Reactor RCP Reactor Coolant Pump RCS Reactor Coolant System RHR Residual Heat Removal RPS Reactor Protection System RTP Rated Thermal Power SG Steam Generator SGTR Steam Generator Tube Rupture SI Safety Injection (High Pressure Safety Injection)
TREAT Transient Real-Time Engineering Analysis Tool
, MAINE YANKEE Maine Yankee Atomic Power Company WOG Westinghouse Owners Group WR Wide Range YANKEE Yankee Atomic Electric Company l
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1 LIST OF TABLES I
3.1 TREAT Model Assumptions and Features 14 3.2 RPS Trips and Setpoints 16 3.3 Full Power Conditions 17 4.1.1 Sequence of Events for a SGTR (Recovery Action - Auxiliary Spray /ADV) 32 4.1.2 Sequence of Events for a SGTR (Recovery Action - One PORV/ADV) 34 4.1.3 Sequence of Events for a SGTR (Recovery Action - Normal Spray / Steam Dump) 35 4.2.1 Sequence of Events for ICC (Recovery Action - Secondary Depressurization) 63 4.2.2 Sequence of Events for ICC (Recovery Action - RCP
- Restart / Delayed Secondary Depressurization) 64 4.2.3 Sequence of Events for ICC (Recovery Action - Primary Depressurization) 66 4.3.1 Sequence of Events for a LOHS (Without Steam Dump and Bypass) 85 4.4.1 Sequence of Events for a Small Break LOCA (Without Off-Site Power) 97
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LIST OF FIGURES Wumber Title Paje 3-1 Maine Yank'ee TREAT Model (Primary Side) 18 3-2 Maine Yankee TREAT Model (Secondary Side) 18 3-3 Pump Coastdown Comparison 19 Benchmark - Break and SI Flow Rates 20 j 3-4 3-5 Benchmark - Pressurizer and SG Pressures 20 SGTR Results (Recovery Action - Auxiliary Spray /ADV) 4.1-1 SGTR - Pressurizer Pressure (psia) 36 4.1-2 SGTR - Pressurizer NR Level (U span) 36 4.1-3 SGTR - Faulted SG Pressure (psia) 37 4.1-4 SGTR - Intact SG Pressure (psia) 37 .
4.1-5 SGTR - Faulted SG NR Level (U span) 38 4.1-6 SGTR - Intact SG"NR Level (b span) 38 4.1-7 SGTR - SGTR Tube Side (lbm/sec) 39 4.1-8 SGTR - SGTR Plenum Side (lbm/sec) 39 4.1-9 SGTR - Atmospheric Dump Valve (lbm/sec) 40 4.1-10 SGTR - Safety Injection (lbm/sec) 40 4.1-11 SGTR - Intact Auxiliary Feedwater (lbm/sec) 41 4.1-12 SGTR - Core Exit Flow (lbm/sec) 41 4.1-13 SGTR - Corrected Faulted SG Pressure (psia) 42 4.1-14 SGTR - Corrected SG NR Level (U span) 42 SGTR Results (Recovery Action - One PORV/ADV) 4.1-15 SGTR - Pressurizer Pressure (psia) 43 4.1-16 SGTR - Pressurizer NR Level (k span) 43 4.1-17 SGTR - Faulted SG Pressure (psia) 44 4.1-18 SGTR - Intact SG Pressure (psia) 44
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LIST OF FIGURES (continued)
Wupber Tit'a Paae 4.1-19 SGTR - Faulted SG NR Level (k span) 45 4.1-20 SGTR - Intact SG NR Level (k span)' 45 4.1-21 SGTR - SGTR Tube Side (lba/sec) 46 ,
I 4.1-22 SGTR - SGTR Plenum Side (lba/sec) 46 4.1-23 SGTR - Intact Auxiliary Feedwater (lbm/sec) 47 4.1-24 SGRR - Safety Injection (lba/sec) 47 4.1-25 SGTR.- Atmospheric Dump Valve (1bm/sec) 48 4.1-26 SGTR - One Pressurizer PORY (lbm/sec) 48 4.1-27 SGTR - RCS Saturation Temperature (OF) 49 4.1-28 SGTR - Core Exit _ Temperature (OF) 49
- SGTR Results (Recovery Action - Normal Spray / Steam Dump) 4.1-29 SGTR - Pressurizer Pressure (psia) 50 0 4.1-30 SGTR - Pressurizer NR Level (b span) 50 4.1-31 SGTR - Faulted SG Pressure (psia) 51 4.1-32 SGTR - Intact SG Pressucc (psia) 51 4.1-33 SGTR - Faulted SG NR Level (b span) 52 4.1-34 SGTR - Intact SG NR Level (b span) 52 4.1-35 SGTR - SGTR Tube Side (1bm/sec) 53 4.1-36 SGTR - SGTR Plenum Side (lbm/sec) 53 4.1-37 SGTR - Intact Auxiliary Feedwater (lbm/sec) 54 4.1-38 SGTR - Safety Injection (1bm/sec) 54 4.1-39 SGTR - Steam Dump and Bypass (lbm/sec) 55 4.1-40 SGTR - Normal Spray (1bm/sec) 55 4.1-41 SGTR - RCS Saturation Temperature (OF) 56
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Wumber Title Paje 4.1-42 SCTR - Core Exit Temperature (OF) 56 SGTR (Conclusion) 4.1-43 SGTR - Fleshing Fractions (%) 57.
4.1-44 SCTR - Pressuriser Pressure Comparison (psia) 57 4
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ICC Kesults (Recovery Action - Secondary Depressurization) 4.2-1 ICC - Pressurizar Pressure (psia) 67 4.2-2 ICC - SG Pressure (psia)- 67 4.2-3 ICC - SG NR Level (Tespan) 68 4.2-4 ICC - Auxiliary Feedwater (lbm/sec) 68 4.2-5 ICC - Break Flow (Ibm /sec) 69 I
4.2-6 ICC - Core Exit Flow (lbm/sec) 69 4.2-7 ICC - Core Exit Temperature (OF) 70 4.2-8 ICC - Atmospheric Dump Valve (lbm/sec) 70 1
ICC Results (Recovery Action - RCP Restart / Delayed Secondary Depressurization) 4.2-9 ICC - Pressurizer Pressure (psia) 71 4.2-10 ICC - SG Pressure (psia) 71 4.2-11 ICC - SG NR Level (%-span) 72 4.2-12 ICC - Auxiliary Feedwater (lbm/sec) 72 4.2-13 ICC - Break Flow (lbm/sec) 73 4.2-14 ICC - Core Exit Flow (lbm/sec) 73 4.2-15 ICC - Core Exit Temperature (OF) 74 4.2-16 ICC - Atmospheric Dump Valve (lba/sec) 74 ICC Results (Recovery Action - Primary Depressurization) 4.2-17~ ICC - Pressurizer Pressure (psis) 75 i
v 4.2-18 ICC - SG Pressure (psia) 75 eXe 4
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' k . Wugber I.itle. P_ age 4.2-19 ICC - SG NR Level (%-span) 76 4.2-20 ICC - Auxiliary Feedwater (lbm/sec) 76 4.2-21 ICC - Break Flow (Iba/sec) 77 1
4.2-22 ICC - Core Exit Flow (Iba/sec) 77 4.2-23 ICC - Core Exit Temperature (OF) 78 4.2-24 ICC - Pressurizer PORY (ibm /sec) 78 i LONS Results (Without Steam Dump and Bypass) 4 4.3-1 LONS - Pressurizer Pressure (psia) 87 4.3-2 LONS - SG Pressure (psia) 87 ,
- 4.3-3 LOHS - Pressurizer NR Level (%-span) 88 4
- 4.3-4 LOHS'- Safety Injection (lbm/sec) 88 I 4.3-5 LONS - SG NR Level (%-span) 89 i T 4.3-6 LPHS - SG WR Level (%-span) 89 4.3-7 LONS - Auxiliary Feedwater (lba/sec) 90 4
4.3-8 LONS - Core Exit Flow (1bm/sec) 90 4.3-9 LOHS - Pressurizer PORY (1bm/sec) 91 4.3-10 LONS - Atmospheric Dump Valve (lbm/sec) 91 4.3-11 LONS - RCS Saturation Temperature (OF) 92 4.3-12 LOHS - Core Exit Temperature (OF) 92
. 4.3-13 LONS - Extrapolated Pressuriser Pressure (psia) 93 4.3-14 LONS - Extrapolated Core Exit Temperature (OF) '93
, Small Break LOCA Results (Without Off-Site Power)
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>- 4.4-1 LOCA - Pressurizer Pressure (psia) 98 4.4-2 LOCA - Pressurizer NR Level (%-span) 98
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LIST OF FIGURES (continued)
Number Title Page 4.4-3. LOCA - SG Pressure (psia) 99 4.4-4 LOCA - SG NR level (%-span) -99 4.4-5 LOCA - Break Flow (1tna/sec) 100 4.4-6 LOCA - Safety Injection (lbm/sec) 100 4.4-7 LOCA - Pressurizer PORY (lbm/sec) 101 24.4-8 LOCA - Atmospheric Dump Valve (lbm/sec) 101 4.4-9 LOCA - Auxiliary Feedwater (lbm/sec) 102 4.4-10 LOCA - Core Exit Flow (lbm/sec) 102 4.4-11 LOCA - RCS Saturation Temperature (OF) 103-4.4-12 LOCA - Core Exit Temperature (OF) 103 O.
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1.0 INTRODUCTION
The analyses described herein were perforined to provide a quantitative basis for supporting E0P development for Maine Yankee. Revised E0Ps at Maine Yankee are based on the Westinghouse Emergency Response _ Guidelines (ERGS) developed for the Mestinghouse _0wners Group (WDC). The Westinghouse ERGS were written for standard Westinghouse plants. The four events identified below
, were selected based on differences between Maine Yankee and standard Westinghouse plants. These analyses will demonstrate the acceptability of the revised Maine Yankee EOPs by examining the plant-specific capability of the Maine Yankee emergency equipment. The four events selected were:
o Steam Generator Tube Rupture o Inadequate Core Cooling o Loss of Secondary Heat Sink o Small Break LOCA Cooldown/Depressurization Maine Yankee had previously upgraded their EOPs in response to NUREG-0737 and implemented them in April of 1982. Following-a supplement to NUREG-0737, they agreed to integrate the EOPs with the remaining emergency response capability programs and revise them as necessary. This analysis supports the upgrade of the Maine Yankee EOPs as required by NUREG-0737.
The TREAT computer code (Reference 1) was used to analyze the above mentioned transients. This code was developed by Westinghouse for several uses, one of which is to aid in verifying the ERGS for standard Westinghouse plants (Reference 2). Numerical input data is required to describe the components and geometry of the system being analyzed. The input data includes fluid volume sizes, pump features, power generation, heat exchanger properties, material compositions and control system characteristics. Section i 2.0 describes the TREAT code in more detail.
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The Maina YtnkOo plcat 10 a throo-locp gombustien Engin2cring (CE)
Eressurized W!ater Reactor (PWR) rated at 2630 MWt. The TREAT code was designed and "hard-wired" for a generic Westinghouse PWR. Therefore, some assumptions for the plant systems had to be made. The necessary modeling assumptions, as well as, a general model overview, and model verification are l presented in Section 3.0. i The analysis results for the SGTR, ICC, LONS and small break post-LOCA cooldown are presented in Sections 4.1 through 4.4, respectively. Two additional SGTR analyses were evaluated in Section 4.1 to access alternative depressurization strategies beyond those contained in the Westinghouse ERGS.
For each of the transients analyzed the E0P recovery method, operator and equipment delay times, and equipment availability came from either Reference 3 or through discussions with the Maine Yankee staff. Section 5.0 contains a summary and conclusion of the analyses.
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2.0 GENERAL CODE DESCRIPTION
() TREAT (Transient Real-Time Engiceering Analysis Tool) is a general purpose interactive thermal-hydraulic network code. It was developed by Westinghouse for several uses, one of which is to aid in verifying the ERGS for standard Westinghouse plants (Reference 2).
The TREAT code requires input data equivalent to codes such as RELAP w
and RETRAN. TREAT input requires a node-junction type description of the system. Fluid properties in each node are solved independently using two energy and two mass conservation equations. Hence, non-equilibrium thermodynamic conditions are allcwed in any node. The mass equations are solved explicitly. The energy equations are solved by a predictor / corrector method. The predictor / corrector is used to find a global pressure that conserves total system energy. The overall system vol uut is conserved; however, the fluid in an individual node may not occupy the same volume as the physical node, thus, producing a local volume error. This local volume error is coupled to the momentum equation using the " dual variable" method to predict a volumetric flow rate through each flow link. If two-phase
() conditions exist in a node, phase separation is calculated by the drift flux method.
The TREAT PWR System includes models for:
o Neutronics D kinetics to compute the axial flux, power and fission production.
o Heat Transfer - computes core, steam generator and metal heat transfer.
o Controllers - for reactor power, pressurizer pressure, pressurizer level, steam / feed flow, steam generator pressure and steam generator level.
o Reactor Protection Systems (RPSs) - for reactor trip, Safety Injection (SI) actuation, turbine trip, steam line isolation,
() feedwater isolation and auxiliary feedwater actuation setpoints.
i o Boundtry Flows - fcr manusi edjustment of SI, charging, letdown, PORVs and pressurizer spray flows.
O o Reactor Coolant $mp (KCP) - computes Reactor Coolant gystem (RCS) flow by a four-quadrant pump homologous curve.
Reference 1 (the TREAT Users Manual) furnishes a more detailed description of the code.
1 i
O t
i e-t Note: The controllers and RPS are inherently designed for a genoric
.; Westinghouse plant.
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l 3.0 MODEL DEVELOPMENT i
The purpose of this section is to describe the Maine Yankee TREAT model
- developed for the support of the Maine Yankee E0Ps. The major areas discussed "are: l l
o Model Description and o Model Verification. ,
Plant geometry such as fluid volumes, elevations and areas were developed from the RELAP4 large break LOCA model used in Reference'4 because of the similarity in the nodalization of the large break model and that used for the TREAT model. Plant geometry such as resistance and inertia terms were 7
developed from the Maine Yankee RETRAN-02 model used in Reference 5 because it is a best-estimate, full power / full flow model. The plant control parameters were generated from plant design information. Westinghouse data for a-standard plant were used in some cases to complete the model, i.e., 1-D core kinetics data. The type of data used from the standard plant was appliad only to complete the model and does not play a significant role in-the transients analyzed.
Validation of the reactor scram decay heat versus time was accomplished l
by comparison with NUREG-75/087. Pump coastdown was checked against Maine Yankee plant test data (Reference 6). The qualitative plant thermodynamic response was benchmarked against a Maine Yankee SGTR analysis perfonned with RELAPSYA in Reference 7.
3.1 Model Descriptions
. The following sections highlight the Maine Yankee' TREAT model. The highlight includes a discussion on the primary and secondary nodalization, the
. core model, plant control systems and the steady-state full power model.
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-S--
f 3.1.1 Nodalisation TheLthree-loop Maine Yankee plant is explicitly represented by a
~
three-loop TREAT nodalization scheme as shown in Figures 3-1 and 3-2 for both l
the prir.ary and the secondary sides, respectively. This nodalization scheme was adapted from the Westinghouse model via their recommendation. However, I '
f because of the differences in plant geometry between a generic Westinghouse plant and the Maine Yankee plant, some modification was required. The most l important modification was the upper head flow path. This flow path is a major difference between a generic Westinghouse plant and a CE plant.
Vessel I
Five nodes were used to model the reactor vessel, reactor core, upper head, control rod guide tubes, upper head and.downcomer/ lower plenum as shown in Figure 3-1. Flow links are connected between the nodes to represent the coolant flow paths. Flow Link 4 connecting Nodes 4 and 2 is unique to CE's vessel. On a Westinghouse vessel, this flow link would connect Modes 4 and 5 (i.e., upper head injection).
O Information for nodes and flow links, such as volumes, flow areas,
[ elevations, etc. , were generated from Maine Yankee's RELAP4 model used in Reference 4. The TREAT nodalization scheme is less detailed compared to the RELAP4 model. However, the lack of detail is subdued by the utilization of subnodes and a more elaborate form of input data. Each node can be subdivided into finer horizontally stacked nodes, called "subnodes," each with their own horizontal area. For example.. node five has two subnodes specified to track the water level - one for the downcomer and one for the lower plenum.
Although no explicit momentum equation is solved by the TREAT code, fluid inertial effects are simulated by a so-called " volume error flow rate."
Local flow rates are calculated from a volume error and the local flow resistance and inertia. The local fluw resistance and inertia are input and were prepared from Maine Yankee's M RAM-02 model used ir. Reference 5. The RETRAN model was used because d r % sins a best-estimate, full power / full flow model. Thus, the Maine Yankee IAEAT model should yield results comparative to best-estimate, full power / full flow conditions.
l RCS Loops Each RCS loop is modeled by seven internal nodes: one hot leg including Steam Generator (SG) inlet plenum, two nodes for SG tubes, one SG outlet plenum with pump suction leg (downward), one pump suction leg (upward) with RCS pump and one cold leg. Internal flow links connect the internal nodes of the RCS and the reactor vessel to construct a closed system.
The pressurizer is attached to the hot leg of Loop 1 the same as the plant. Non-Equilibrium (NE) conditions (superheated steam above subcooled or saturated liquid) are modeled in the pressurizer as well as all the RCS nodes.
Charging, letdown, _ Emergency Core Cooling Eystem (ECCS), pressurizer spray, etc., are represented by the boundary nodes connected to the RCS by critical flow links as shown in Figure-3-1.
Secondary Side The main steam line and the three SGs were modeled by nodes and noncritical flow links as shown in Figure 3-2. The turbine, condenser and Feedwater (FW) System were not explicitly modeled. However, the flow paths to I the trebine and from the FW System were represented by boundary nodes and i
critical flow links.
Internals of the SG shell side were subdivided into three nodes to model internal circulation. For the SG of Loop 1, Mode 1 represents the downcomer node, Mode 2 - the boiling region. The steam dome was modeled by Mode 3. Subnodes were extensively utilized to express the complex internal features. Critical Flow Links 52 and 53 model the liquid return flow to the SG downcomer. Flow Link 13 allows steam to escape to the steam dome. Mode 4 represents an average main steam line between the SG and the main steam isolation valve. Hence, the asynunetry of the main steam lines was not modeled.
Nodes 1 through 4 are repeated for the other two SGs and steam lines.
Mode 13 represents the header region up to the turbine stop valve.
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Heat Modes
) Heat stored in the internal structures and walls was modeled via heat nodes as shown in Figures 3-1 and 3-2. All heat nodes were assumed to be passive (i.e., no internal heat generation). Hence. Heat Node 1 represents only the core region internal structures excluding the fuel. Heat nodes representing walls were assumed to be perfectly insulated on the outside surfaces. Steam generator tubes are a special kind of. heat node in the TREAT code. SG heat nodes allow heat transmission from one side to the other. The heat capacitance of the tube wall cannot be modeled by the code. This is a reasonable approach since the tube wall is relatively thin. Correct heat transfer resistance is accounted for via supplying tube wall conductance, including U-fouling on the tube surface, as well as code computed primary and secondary heat transfer coefficients.
The reactor core'is another special type of heat source described in Section 3.1.2.
3.1.2 Core Model O The reactor core is simulated by two-dimensional heat conduction with internal heat generation, convective heat transfer, a one-dimensional neutron kinetics model and a decay heat model.
The core is segmented into eighteen equally sized axial nodes, sixteen of the eighteen nodes represent the active fuel, the two additional nodes located at the top and bottom of the core represent the nonactive fuel regions. The built-in pellet-to-clad gap heat transfer model is selected rather than the Maine Yankee-specific model. However, the fuel temperature should not be affected because the long-time duration of cooling will remove most of the stored energy during the transient periods analyzed. The fuel geometry of Cycle 9 is modeled.
Cross-section data for Maine Yankee core could not be generated within the scope of this project. The data were substituted from a generic Westinghouse core for beginning-of-life. The cross-sections were adjusted by O
- u.
trici and cerse to match Mains Ycnkoo'o ested care power. Th3 cv:nts cnolyz:d are not reactivity induced, thus, the results are not impacted by.this modeling approach.
The decay heat level used was the ANS standard (NUREG-75/087) without uncertainty.
3.1.3 Plant Control Systems Since the us e selected variables for the plant control systems were developed for generic Westinghouse plants, numerous improvisations to the data were necessary. Sone af the changes required code modifications to accept the
' Waine Yankee design. The major modeling improvisations are listed in Table 3.1.
The TREAT code allows'only one charging pump connection to RCS, hence, Loop 2 and Loop 3 charging pump flows are combined into one connection at Loop 3. Furthermore TREAT does not allow regulating the letdown flow during normal charging, as in the Maine Yankee design. Hence, the not flow of charging and letdown is controlled via regulating the charging flow in this model.
The steam bypass valves in the TREAT code are designed to be automatically controlled by the RCS temperature, whereas Maine Yankee's are controlled by SG pressure. TREAT also allows the user to manually switch to automatic SG pressure control for RCS cooldown procedures. Maine Yankee's j model uses only the automatic SG pressure control. The relief capacity was set for relief from all twelve valves. This will have no effect on the 4 transient results.
l Steam Bypass for Westinghouse:
I
' o Pre-Trio - T control corresponding to turbine demand vs. core.
avg i
power. It uses six of the twelve available valves.
o Post-Trio - T control corresponding to T, vs. T .
It uses twelve of the twelve available valves.
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o Pressure Control - Used to cool down during plant recovery procedures. -
O , Steam (Dumo and) Evoass for Maine Yankee:
, o Pre Trio and Post-Trio - Pressure control maintains'SG pressure at 900 psia. It uses seven of the twelve available valves. '
- o Post-Trio - T, control corresponding to T, vs. T , ,.
It uses five of the twelve available valves. l The reactor protection logics are' programmed for generic Westinghouse t plants and no improvisation could be used to simulate Maine Yankee's 3'
Protection System. Therefore, a Maine Yankee-specific subroutine was written to override TREAT's protection logic. This subroutine controls: reactor scram, SI, AFW actuation, turbine trip, feedwater and steam isolation, and ,
letdown isolation. The reactor protection system models are shown in
. Table 3.2.
3.1.4 Steady-State Full Power Model l
4 The solution techniques used in TREAT are such that a steady-state full
. power flow condition'cannot be initialized directly through the input data, t t Instead, the model must be initialized for a stagnant (i.e., no flow). zero power condition. Then, coolant flow is induced by camping the RCPs to rated I speed. After reaching rated full flow in the RCS.'the control rods are slowly I removed until core thermal power reaches the rated power level. The secondary I side was then added to the primary side model.- Main FW and SG pressures were
- adjusted to obtain best-estimate full power operating conditions.
t
+ Maine Yankee's full power model had been obtained in a similar fashion, except the rated core thermal power was achieved by removing boron from the coolant after all rods were removed from the core.
4 LO
! _10
The basic maneuver to reach full p:wer from a stagntnt H t D ro Power (HZP) condition is similar to that used in a real plant. However, to'obtain the correct conditions with TREAT for such parameters as pressures, temperatures, flow and power level requires a number of iterations. Due.to system feedback, varying any parameter (during the course to achieve the correct full power conditions) varies many other parameters simultaneously.
A final set of data for Hot _ Full Power (HFP) conditions obtained by TREAT is compared to .B_est-Estimate (BE) plant data in Table 3.3.
3.2 Model Verifications TREAT code applications have been presented in recent publications, Reference 2; but the Maine Yankee E0P support analysis is the first application of the code to a non-Westinghouse plant. Verification of the code as well as its input model for this application was performed by benchmarking the TREAT predictions against plant data and RELAPSYA predictions. The following section describes the benchmark cases.
3.2.1 Reactor Scram From Full Power The steady-state full power Maine Yankee TREAT model (described in Section 3.1.4) was manually scrammed and the plant control system was kept in an automatic control mode. Due to an apparent inadequacy of the TREAT code, which was later resolved, steam dump to the condenser was regulated by the T,y controller rather than the pressure controller. Therefore, the post-reactor trip hot and cold leg temperatures correspond to 547 F (T,
)
rather than to 532 F-(T sat at 900 psia, i.e., the pressure controller setpoint). All other controllers were representative of Maine Yankee's automatic control systems. In this transient, the following parameters were
)
monitored:
o RCS pressure o Pressurizer level o RCS temperatures l
l r - --. --.. -- -,. , - -
O SG pr00sure o . Decay heat
-Following reactor scram, the RCS pressure decreased approximately 200 psi and gradually increased to the initial pressure after 500 seconds. The pressurizer level dropped from 57% to 34%, then slowly increased to 43%
level. The hot leg temperature decreased to about 549 F, this corresponds to T,y plus 2 F; the cold leg temperature decreased to 547 F. The SG pressure increased momentarily above 1000 psia and stabilized at approximately 940 psia.
For a normal Maine Yankee reactor trip, the RCS pressure level and temperatures would have decreased and stabilized at lower values. The SG pressure would have stabilized at the pressure controller-setpoint. However, since the pressure controller was not modeled in this scenario, the cooldown was limited to 547 F (T ) instead of 532 F (T no-load * " *
- adjusting for the pressure controlled impact,the thermodynamic response compares reasonably.well with the normal trip response.
O The decay heat level was compared to the values obtained from NUREG-75/087 and shows very good agreement.
3.2.2 Pump Coastdown Test Pump coastdown tests were performed from stabilized HZP conditions.
The results were compared against Maine Yankee's test data (Reference 6).
A three-pump coastdown test was analyzed to match the coastdown characteristics. The frictional torque parameters were varied to obtain good agreement with plant data, as shown in Figure 3-3.
3.2.3 RELAPSYA SGTR Test SG tube rupture calculation performed by RELAPSYA model used in Reference 7 was repeated using the TREAT code for benchmarking purposes.
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This. event was initiated from full power. At 60 seconds, a single tube rupture occurred as a double-ended guillotine near the cold and of a tube. No operator actions were assumed.
Figures 3-4'and 3-5 present the comparison of critical parameters.
Figure 3-4 shows break flow comparisons along with ECCS flows. The magnitude of the combinsd break flow rates were not exactly equal. However, the tube sheet (plenum side) break flow matches well. This demonstrates that the critical flow models in both codes are similar. In contrast to the tube sheet side, the tube side break flow does not match well. The tube side break flow calculated by TREAT is larger than that calculated by RELAP5YA. This mismatch can be attributed to differences in the nodalization of the tube region.
RELAP5YA models a single ruptured tube by two fluid volumes (nodes) separated from the other tubes. In the TREAT model, all tubes, including the ruptured
-one, are modeled as a combined node. Hence, the uniquely distinct conditions inside the ruptured tube cannot be explicitly modeled in the TREAT analysis.
However, the combined break flow calculated by TREAT is larger than that predicted by RELAPSYA, therefore, the TREAT break flows are conservative in respect to SG overfill.
The total break flow calculated by the TREAT model was slightly larger than the RELAPSYA model, however, the RCS depressurization rate was less than the TREAT model as shown in Figure 3-5. This may be expected because of the different modeling of the charging and letdown control system rather than the differences in the codes. As stated in Section 3.1.3, the letdown flow was not explicitly modeled in the Maine Yankee TREAT model.
The results of the-benchmarking effort described above provide confidence that the model developed is an accurate representation of the Maine Yankee plant. Therefore, the model can be used to depict plant response characteristics in support of EOP development.
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TABLE 3.1 TREAT Model Assumptions and Features O 1~. A nonequilibrium pressurizer model is used.
- 2. The core bypass flow is not separately modeled. The core is modeled as one node with eighteen subnodes, sixteen of which represent active fuel.
i The core bypass is-lumped into-the core node.
- 3. Beginning-of-life moderator and fuel feedback parameters are used.
- 4. . Charging and letdown is modeled as not flow.' charging fluid temperature is assumed to be 50 degrees below the cold leg fluid temperature.*
- 5. Charging on Loops 2 and 3 are lumped into only Loop 3.*
- 6. The proportional pressurizer heaters are modeled as on-off heaters that operate midway through their normal control range.*
- 7. Mass of pressurizer spray added to the RCS during each transient is considered to be small in comparison to the total RCS mass.
- 8. The core neutronics is represented by 1-D kinetics.
- 9. Pressurizer pressure control, level control and steam generator level control are in automatic mode.
() 10. Steam dump is considered the preferred way to cooldown, if available.
- 11. Normal pressurizer spray is considered the preferred way to depressurize, if available.
- 12. Loss of Off-Site Power (LOOP) results in:
o Tripping all RCPs o Closure of the Main Feedwater Regulator Valves (MFWRVs) o Closure of the Eeedwater Regulator Bypass Valves (FWRBVs) o Loss of steam dump valves o Loss of normal pressurizer spray o Loss of auxiliary pressurizer spray o Safety Injection (SI) returns on-line 20 seconds af ter LOOP (powered by diesel generators) o Auxiliary Feedwater (AFW) returns on-line 30 seconds after LOOP (powered by diesel generators)
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4.
TABT.E 3.1 (continued)
- 13. Operator action is assumed.
- 14. All cases start at HFP conditions.
- 15. Westinghouse core cross-sections were used. (Full power and decay heat levels are adjusted for Maine Yankee.)
- 16. RCS flow resistances are based on full power forced forward flow conditions.
l l
1 2
0
- Modeling improvisions that were necessary due to the differences between 1
Maine Yankee and standard Westinghouse plants.
O l !
TABLE 3.2 RPS Trips and Setpoints
( Reactor Tri2 RCS SEC MI FLUX = 1.065% LO SG LEVEL = 35%
LO PRES = 1850 psia LO SG PRES = 500 psia HI PRES.= 2400 psia TURBINE-TRIP LO FLOW = 93% of nominal l SIA LO PRES = 1600 psia AFWI LO SG LEVEL = 35% (Also at top of tubes)
() HI SG LEVEL = 91%
REACTOR TRIP MFWI REACTOR TRIP Feed ramps to 5% of HFP flow rate 4 Also on SIS w/ LO PRES 415 psia /SG, this is not modeled.
MSIV LO STEAM PRES = 415 psia (Any SG LO PRES isolates all EFCVs)
LETDOWN ISOLATION LO PRES = 1600 psia (SIS) l 0
TABLE 3.3 l
- Full Power Conditions V (3 TREAT Nodel BE Plant Data RCS Pressure (psia) 2253 2250 Temperature (F) Hot Leg 596.8 595 to 600 Cold Les 550.6 550 Total Loop Flow Rate (lb/sec) 41,124.0 41,012.1 Pressuriser Level (%) 58.0 60 Core Power (MWth) 2629 2630
- SG Pressure (psia) 881.5* 840*
FW Flow Rate (lb/sec) 1074 1,064.67 Total Steam Flow Rate (lb/sec) 3194 3194 4
SG Level (%) 67.5 67 Recirculation Ratio 5.9 5.9
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- The BE plant data is measured at the main steam isolation valve, the TREAT data for this corresponds to the pressure in the steam dome.
i
^^
Maina Ycnkoo TREAT Mod 31
- Figure 3-1 T'N. PRIMARY SYSTEM
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- 20-
4.0 50P ANAI,YSES The Maine Yankee TREAT model is described in Section 3.0 of this report. This model was used to analyze four events:
o Steam Generator Tube Rupture o Inadequate Core Cooling o Loss' of Secondary Heat Sink o small Break LOCA Cooldown/Depressurization
- The four events identified above were analyzed to provide a quantitative basis for developing revised EOPs for Maine Yankee. The revised Maine Yankee EOPs were based on the Westinghouse ERGS developed for the WOG. ,
The Westinghouse, ERGS were written for standard Westinghouse plants. These transients were selected based on differences between Maine Yankee and
^
1 standard Westinghouse plants. The analyses will demonstrate acceptability of the revised Maine Yankee E0Ps by examining the plant-specific capability of the Maine Yankee emergency equipment. For each of the events analyzed the E0P recovery method, operator and equipment delay times and equipment availability
- were developed from either Reference 3 or through discussions with the Maine Yankee staff. Performance characteristics of the plant systems that detet1nine the course of these transients were simulated as realistically as possible.
For all scenarios described, operator action did not include loop stop valve closure as part of the accident recovery schenie.
4.1 SGTR Analysis i
This transient is a SGTR from 100% R_ated }'hermal Power (RTP) with a coincident loss of off-site power with reactor trip. The magnitude of the break flow is that of a Double-Ended Guillotine Break (DEGB) of a single tube located at the SG outlet plenum. The purpose of this analysis is to O
o Determine if the revised SGTR E0P will be adequnto to cllow for plant recovery from a DEGB of a single tube.
% s/
o Determine if the RCS cooldown and depressurization will be limited by Maine Yankee's emergency equipment relief capacity.
The key operator recovery actions from the SGTH EOP, as denoted in Reference 3, were:
I o Diagnosis of the ruptured SG, o -Isolate the ruptured SG (on the secondary side),
o Cooldown the RCS, o Terminate cooldown at the predetermined subcooling margin, and then 1
o Depressurize the RCS to terminate the break flow.
() Cooldown and depressurization were also evaluated as simultaneous steps.
Three transients were analyzed for the SGTR. The recovery methods include:
o Cooi.'own with the ADV and depressurize with the pressurizer auxiliary spray, Section 4.1.2.1.
o Cooldown with the ADV and depressurize with one (or two) pressurizer PORV(s), Section 4.1.2.2.
o Cooldown with the steam dump system and depressurize with the pressurizer normal spray, Section 4.1.2.3.
The scenario in Section 4.1.2.1 was evaluated to validate the revised SGTR E0P developed by Westinghouse. The scenarios in Sections 4.1.2.2 and j 4.1.2.3 are additional analyses that were evaluated to assess alternative depressurization strategies.
(}
I
1 I
4.1.1 Assumptions
() The following assumptions apply to all the SGTR cases:
l
~
o The SGTR break flow area = 2.333 x 10 ft . This area corresponds to one DEGB. Based on results from Chapter 14 of the Maine Yankee FSAR, the initial flow from a double-ended SGTR will be greater than 50 lbm/sec. This size break will be high enough to drain the pressurizer and cause reactor trip and SI actuation within a few minutes after the break occurs.
o The SGTR occurred in Loop 1. The loop was chosen arbitrarily and is not assumed to affect the results, o The SGTR was located near the SG outlet plenum. The " cold leg side" of the SG was selected to maximize the break flow. .
o A time of five minutes from the diagnosis was assumed to isolate the faulted SG. SG isolation consists of:
O o Isolate blowdown, MFWRVs, FWRBVs and AFW valves, o Close gonreturn Valve (NRV), Excess Flow gheck Valve (EFCV) and atmospheric steam dump header valve in ruptured SG, and o close steam traps, drain lines, etc.
o once the RCS pressure reaches the faulted SG pressure, the system is considered to be under control.
The following assumptions apply to the SGTR case in Section 4.1.2.1:
o LOOP was assumed. This would minimize the RCS cooldown rate to the relief capacity of the atmospheric dump valve. The method RCS depressurization on a LOOP would be two pressurizer PORVs where pressurizer level and RCS subcooling would be maintained by O
4 throttling SI. Auxilitry cpery is ni,t cyniltblo on LOOP bectusa Valve CH-M-52 is not on the emergency buses. However, to limit the
(~S ~RCS depressurization, this valve was assumed to be operable. Thus, V SI was terminated (when the SI termination criterion were met) and the pumps where realigned to normal charging. This would enable the auxiliary pressurizer spray for RCS depressurization (i.e.,
instead of the PORVs) where pressurizer level and RCS subcooling would be maintained by throttling charging. It is possible to maintain injection through the SI lines and use auxiliary spray by resetting the SI signal. The SI reset criterion will be met at or prior to SI termination criterion because of their restrictions.
o SI Roset Criterion:
- 12 F subcooling
- 5 percent pressurizer level, and 5 percent SG NR level o SI Termination Criterion:
- 12 F subcooling 2 percent pressurizer level
- 35 percent SG NR level in at least one SG, and
- increasing RCS pressure o A time of five minutes after reactor trip was assumed for diagnosis of the ruptured SG.
o Throttle SI to maintain pressurizer level greater than 50% provided
~
that at least a 12 F subcooling margin exists. l I
o cooldown was performed following faulted SG isolation.
- 1) Cooldown was maintained until a 32 F subcooling margin exists corresponding to the ruptured SG pressure,
- 2) Depressurization was initiated when cooldown was complete.
i l
l The following assumptions apply to the SGTR case in Section 4.1.2.2: l 1
o LOOP occurs coincident with reactor trip.
1
- o. Diagnosis of the ruptured SG takes 8 minutes from the time of I 1
l o Throttle SI to maintain pressurizer' level greater than 5% provided i that at least a 12 F subcooling margin exists, o cooldown using the ADV as fast as possible as soon as isolation of the ruptured SG is cosplete. This assumes that the operator was previously aware of the LOOP and does not attempt to cooldown with steam dump to the condenser.
o Depressurize using one PORV as soon as subcooling > 32 F (i.e.,
20 F for a subcooling margin, plus 12 F for instrument uncertainty). This assumes that the operator was previously aware of the LOOP and does not attempt to depressurize with pressurizer nomal spray or auxiliary spray.
Note: Cooldown and depressurization will actually be initiated simultaneously since the SI should have enabled a subcooling.>_ 32 F long before the ruptured SG was isolated.
o If subcooling $ 12 F stop depressurizing, SI will be throttled to maintain this subcooling margin.
o operator maintains a close surveillance of the subcooling margin.
The following assumptions apply to the SGTR case in Section 4.1.2.3:
o Diagnosis of the ruptured SG takes 8 minutes from the time of reactor trip.
O o Throttlo SI to maintcin pre:surisce 10'o1 gre ter than 5%, provided that at least a 12 F subcooling margin exists. '
l O #
o Cooldown using the steam dump to the condenser as fast as possible as soon as the ruptured SG is isolated.
i o Depressurize using pressurizer normal sprays as soon as subcooling 2 32 F (i.e., 20 F for a subcooling margin, plus 12 F for instrument uncertainty).
Note: Cooldown and depressurization will actually be initiated simultaneously since the SI should have enabled a subcooling 2 32 F long before the ruptured SG was isolated, o If subcooling i 12 stop depressurizing, SI will be throttled to maintain this subcooling margin.
o Operator maintairis a close surveillance of the subcooling margin.
O All other assumptions, sumnarized in Table 3.1 and discussed in Section 3.1, apply as well.
4.1.2 Results 4.1.2.1 SGTR (Recovery Action - Auxiliary Spray /ADV)
The SGTR E0p transient consists of automatic and operator action. The sequence of events is provided in Table 4.1.1 and results are illustrated in Figures 4.1-1 through 4.1-12.
The SGTR causes an imbalance of primary to secondary pressure. The primary to secondary mass leakage resulted in draining the pressurizer. As the RCS pressure decreased to a new equilibrium pressure, the low pressurizer pressure setpoint was reached causing a reactor trip. The resulting RCS cooldown and emptying of the pressurizer caused a rapid decrease in pressure.
O ,
1 1
Du3 to LOOP coinsid:nt with re:cter trip, SI cctuated 20 saconds efter LOOP and AFW actuated 30 seconds after LOOP. The SI flow rate overcame the break O flow rate, resulting in an increase in RCS pressure and in restoration of t
C o pressuriser level. The addition of 80 F AFW resulted in a decrease in the l SG pressure below the safety valve setpoint and restoration of SG 1evel.
Diagnosis of the ruptured SG was completed five minutes after reactor trip. An additional five minutes was assumed for isolating the ruptured SG.
i Cooldown was achieved by opening the ADV. This resulted in a rapid pressure j drop in the intact SGs. The primary to secondary heat transfer increased in the intact loops, thus increasing the subcooling margin. The required subcooling margin to begin RCS depressurization was achieved at approximately 2000 seconds. Auxiliary spray was assumed to be the most limiting method for depressurizing the RCS under LOOP conditions. Condensing steam in the pressuriser by auxiliary spray decreased the RCS pressure below the ruptured SG pressure, thus ending the transient. Reverse heat transfer in the faulted loop enabled the faulted SG pressure to remain below the safety valve setpoint, j
Auxiliary Spray i
O The pressurizer auxiliary spray is considered to be a less effective means of RCS depressurization under LOOP conditions, as opposed to the pressurizer PORVs. Also, the Maine Yankee operators would be more apt to use the pressurizer auxiliary spray, if available. This was the agreed consensus at the time of the analysis. Therefore, if the pressurizer auxiliary spray I was available, it would be a more likely and more limiting way of f depressurizing.
j A constant rated volumetric flow rate of 185 spm was used for the RCS depressurization. Upon initiated RCS depressurization, SI was terminated and
-notinal charging was initiated. The maximum rated charging volumetric flow
! rate is also 185 spa. Thus, the combined rated volumetric flow rate of charging and auxiliary spray is 370 spm. From the charging pump (i.e., SI pump) flow characteristic curves of p-14A, P-14B or P-14S, a flow rate of 370 gym can be achieved from one pump as long as the RCS pressure is less than
- 1840 psia. The analysis assumed two pumps available and the RCS pressure O
- 1
t ranged between 800 psic cnd 1,800 psio during the depras:uris:ticn.
Therefore, the combined rated flow rate of charging and auxiliary spray is obtainable for this scenario as long as one pump is operable.
I If SI has been previously activated, it must be reset to make auxiliary j spray available. After resetting the SI, the operator may use auxiliary spray and still maintain flow through the SI lines (see Reference 8). If the SI i
termination criteria are met as well, the operator may use auxiliary spray and
]
realign flow through the normal charging lines (see Reference 9). ;
Faulted SG Overfill
)
4 <
j The faulted SG is approximately 92% full at approximately 3800 l seconds. This value was obtained by integrating the break mass over 3800 q.
seconds plus the mass contribution from the auxiliary feedwater. From 1600 seconds to the end of the transient, the SG NR level indication remains at 79% ;
l
- NR.(see Figure 4.1-14). This corresponds to the bottom of the separators, or I
approximately 65% full. . Physically, the 79% NR level should continue to
) increase to 100% NR. However, due to structural differences between a Westinghouse SG and Maine Yankee's SG, the NR level indication never exceeds f() 79%. Thus, extrapolating the level in Figure 2 gives a more realistic SG NR indication for the scenario given.
When the SG downcomer node became full (corresponding to the 79% SG NR level) a 50 psi drop occurred in the faulted SG. This was caused by steam condensing in the steam dome node.
i The phenomenon is not physically realistic; however, it is a familiar i
i problem with thermal-hydraulic codes. Figure 4.1-13 shows the pressure response with a 50 psid correction. With the additional pressure, the faulted SG safety may open a second time, thus making a total opening time of t =
- dit +At2 = 660 seconds. However, this does not pose a problem because the safety valve will not remain open since the faulted SG does not overfill.
O
- t
)
l _ _ . _ . _ , . . _ _ _ - . , _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ , _ _ . _ , _ _ . _ . _ _ _ , , . _ ,,_,_m_..
8I Throttlina SI was throttled after an indication of 50 percent level in the pressuriser. This level was the agreed consensus at the time of the analysis. However, the EOPs will use a lower value. Throttling early would lessen the primary to secondary leakage and, therefore, would be less conservative for SG overfill and allow recovery to occur sooner. If SI was throttled upon entry into the E0P E-3 (i.e., upon diagnosis of the ruptured SG) for the given scenario, the fanited SC would be approximately 90% full at approximately 3800 seconds (see Reference 8).
4.1.2.2 SGTR (Recovery Action - One PORV/ADV)
Usinz One PORV The sequence of events is provided in Table 4.1.2 and results are illustrated in Figures 4.1-15 through 4.1-28.
The plant response is similar to Section 4.1.2.1 until the initiation of cooldown.
The results using one PORV for depressurization show that:
o A 32 F subcooling exists at the on-sight of cooldown, therefore, cooldown and depressurization were initiated simultaneously,
. o The break flow from the ruptured tube was terminated at 35 minutes, and o The maximum percent level achieved in the ruptured SG was 67% full.
The key parameters controlling the discharge of activity to the environment are the flashing fraction and the actual mass release of contaminated steam. The calculated flashing fractions given in Figure 4.1-43 represent conservative estimates of the percentage of break flow that would flash to steam (for the given scenario) as a function of time. Figure 4.1-43 O
shows that the flashing fractions drop significantly following reactor trip.
Comparison of Figures 4.1-17, 4.1-21 and 4.1-22 shows that the actual mass release of contaminated steam will be significantly lower than the break flow and terminate much sooner.
Usinz Two PORVs .
The results using two PORVs for depressurization (see Reference 8) show that:
o A 32 F subcooling exists at the on-sight of cooldown, therefore, cooldown and depressurization were initiated simultaneously, o 'The break flow from the ruptured tube was terminated at 34 minutes, and o The maximum percent level achieved in the ruptured SG was 67% full.
4.1.2.3 SGTR (Recovery Action - Normal Spray / Steam Dump)
O The sequence of events is provided in Table 4.1.3 and results are illustrated in Figures 4.1-29 through 4.1-42.
The plant response is similar to section 4.1.2.1 until the initiation of cooldown. The analysis was terminated at 2400 seconds (40 minutes). At this time:
o RCS pressure = 1060 psia, and o Faulted SG pressure = 750 psia.
The method of cooldown was steam 4. imp to the condenser at a rate equivalent to the ADV not to exceed a cooldown limit of 100 F/hr. This rate was selected to cover the situation where the condenser was not available.
The method of depressurizing was normal spray (with a rated volumetric flow rate of 375 spm).
O
The precauce responsa becomes icds and lecs effectiva with time (i.o.,
psi /sec approaches 0). This implies that the RCS can not be depressurized enough to terminate the break flow solely by the pressurizer normal sprays.
If the condenser is available and a larger cooldown rate is used, it may be feasible to effectively depressurize using the pressurizer norinal spray (see Reference 10).
Soray Comparison j Thc sensitivity of the RCS-pressure response from colder spray is shown
- in Figure 4.1-44. Figure 4.1-44 compares the anticipated pressure responses using normal spray (T = Tcold leg) and auxiliary spray (T = 80 F). The auxiliary spray response was superimposed from Figure 4.1-1. The normal spray response is from Figure 4.1-29.
4.1.3 Conclusion i
Results of the SGTR analysis show that Maine Yankee can recover in a timely manner without SG overfill. The recovery methods that were demonstrated acceptable for Maine Yankee were:
o Cooldown with the Atmospheric _ Dump yalve (ADV) and depressurization with the pressurizer auxiliary spray, Section 4.1.2.1.
o Cooldown with the ADV and depressurization with one (or two)
! pressurizer pilot-Operated Relief Yalve(s) (PORV), Section 4.1.2.2.
Recovery could not be demonstrated cooling down with the steam dump system and depressurizing by pressurizer normal spray, see section 4.1.2.3. In this ,
scenario, the steam dump was throttled back to the flow capacity of the ADV.
For a larger cooldown rate, it may be possible to recover with normal pressurizer spray prior to SG overfill.
4
- O
-,~,,-,_-,,-..,-,--,----,...,,,,.._--.,,,.,...,,..+,---.-..--,,n,
TABLE 4.1.1 53qu:nc2 of Ev:nto '
For a SGTR y (Recovery Action-Auxiliary Spray /ADV )
TIME (SEC) EVENTS l ,
i 0.0 start of ansIysis. Steady State at 100% RTP. )
50.0 Begin SGTR in Loop-1 (1-Tube, DEGB).
s 834.0 Low Pressuriser Pressure setpoint is reached (P-RCS=1850
. PSIA). Reactor trip o curs with a coincident LOOP (i.e. RCPs, MFWRVs, FWRcVs and Steam Dump trip). SG safeties open (on all loops).
' s I
854.0 SI returns on line and actuated (P-RCS 1600 PSIA),
powered by Diesel Generator.
< 1 AFW returns tn line and actuated (L-SG 35% narrov
~
864.0 i range), powered by Diesel Generator.
1080.0 SG safeties close (on all loops).
t A Diagnosis of ruptured SG complete. (i.e. 5 min. for
() 1134.0 post trip recovery). Beginning steps to isolate faulted SG (Loop-1): '3 (1) Isolate Blowdown, MFWRVs, FWRBVs and AFWVS, and
~
(2) Close NRVs and (3) Close steam traps, drain lines, etc.
(Totaltimetoisolate-5 min.)(
\..
< 1434.0 Isolation of faulted SG complete. Beginning cooldown at maximum rate possible using the intact SGs. Method of cooldown is the ADV, rated at 54 full power steam 4-flow.
v\
- 1567.0 All-SG narrow range levels are above 50%. Throttle AFW in; intact loops to maintain SG 1evels.
1998.0 RCS cooldown complete 6 The RCS core exit gesperature was cooled down to 20 F subcooling plus 12 F measurement
, uncertainty,at the, ruptured SG pressure.
T-RCS = 497 F + 20 F + 12 F = 529 F.
T-SG (P-SG = 880 PSIA) = 52,3 F. i i
1
i 1
l l
TIME (SEC) EVENTS i Initiate RCS depressurization. Method of depressurization
[\-~
)# is Auxiliary spray. Auxiliary spray uses charging pumps. Must re-align charging pumps. SI termination
- criteria:
(1) Increasing RCS Pressure YES (2) A secondary heat sink is available (35% NR level in at least one intact SG). YES (3) RCS subcooling is greater than the allowanceforsubcooging uncertainty (e.g. 12 F) YES (4) Pressurizer Level is greater than 0%
plus an allowance for pressurizer level uncertainty (e.g. 2%) YES All SI termination criteria are met. Letdown and normal CVCS operations were established to control pressurizer level above 10% NR and supply water to Auxiliary spray.
3840.0 RCS Pressure matches Faulted SG pressure.
End-of-Transient. ,
e
/
O
.n .-, . - -~.,- . - , _ , __ . . . . , , - - _ , , _ _ . , , ,,._,..,,,,,.-,,,v.,,,,-.,.,,.-,-.-._n---- .-
TABLE 4.1.2 Sequ nc3 cf Ev:nt3 For a SGTR (Recovery' Action-One PORV/ADV)
A Time (Sec) ,
Event 0.0 Start of analysis. Steady State at 100% RTP.
50.0 Begin SGTR in loop-1 (1-Tube, DEGB).
840.0 Low Pressurizer Pressure setpoint is reached (P-RCS=1850 psia). Reactor trip occurs with coincident LOOP.
860.0 SI returns on line powered by the diesel generator.
P-RCS i 1600 psia, thus SI is actuated 870.0 AFW returns on line powered by the diesel generator.
L-SG f 35% NR (in any one of three), thus AFW is-actuated.
Operator throttles AFW to maintain SG 1evels between 50% and 70% to prevent SG overfill.
1320.0 Diagnosis of the ruptured SG is complete.
Operator may now throttle SI if pressurizer level is.
greater than 5% provided that at least a 12 F subcooling n margin existo.
( -
1620.0 Isolation of the ruptured SG is complete. Operator immediately booins to cooldown using the ADV, and to depressurize using one pressurizer PORV. This assumes no delay timo for the operator to note that a 32 F subcooling margin exists.
SI is throttled to maintain a subcooling > 12 F.
2100.0 The RCS pressure drops below the ruptured SG prossure, thus, terminating the break flow.
I l
O 4
i i
- t. .
_ )
TABLE 4.1.3 Sequ;nc3 of Ev:nta For a SGTR (Recovery Action-Normal Spray / Steam Dump) l d("'s Time (See) ,
Event 0.0 Start of analysis. Steady State at 100% RTP.
50.0 Begin SGTR in loop-1 (1-Tube, DEGB).
840.0 Low Pressurizer Pressure setpoint is reached (P-RCS=1850 psia). Reactor trip occurs.
860.0 P-RCS 4 1600 psia, SI actuates.
870.0 L-SG < 35% NR (in any one of threo), ATW actuates.
Operator throttles AFW to maintain SG 1evel between 50%
and 70% to prevent SG overfill.
1320.0 Diagnosis of the ruptured SG is complete. -
Operator may now throttle SI if pressurizer level is greater than 5% provided that at least a 12 F subcooling margin exists.
1620.0 Isolation of the ruptured SG is complete. Operator immediately begins to cooldown using Steam Dump to the condenser, and to depressurize using pressurizer normal spray.
This assumes no delay time for the operator to note that l
a 32 F subcooling margin exists.
SI is throttled to maintain a subcooling 2; 12 F.
2400.0 The RCS pressure = 1060 PSIA, the ruptured SG pressuru = 750 PSIA.
The transient was terminated because of the RCS prassure responso from the normal spray seems inadequate of terminating the break flow.
l Os U
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4.2 ICC Analysis
, This transient is an.ICC from 100% RTP with a coincident loss of
{x~')T off-site power with reactor trip. This analysis was recommended to determine
~
an appropriate temperature at the core exit thermocouples to indicate an ICC condition for Maine Yankee. When the given value is reached the operator will take the appropriate actions to reduce the RCS pressure and temperature.
Three recovery methods were analyzed (Reference 3):
o Secondary depressurization using the ADV; Section 4.2.2.1.
o RCP restart and delayed secondary depe ssurization via the ADV; Section 4.2.2.2.
o Primary depressurization using the pressurizer PORVs; Section 4.2.2.3.
The generic Westinghouse ERGS use an ICC indicating core exit thermocouple temperature reading of 1200 F, based on best-estimate large
(} ~
break LOCA analyses. The purpose of this plant-specific analysis was to verify that a value of 1200 F will be acceptable for Maine Yankee. This value will be acceptable for Maine Yankee if the plant can recover using the procedures in FR-C.1 (i.e., the Response to Inadequate Core Cooling EOP),
Section 4.2.2.1 and Section 4.2.2.2 identify these procedures.
4.2.1 Assumptions The following assumptions apply to all the ICC cases:
o A one-inch diameter break occurred in the cold leg. A cold leg break would have the most restrictive steam vent path, thus, less inventory would remain in the vessel at the time steam would be vented out the break. !
l l
o Failure of high pressure safety injection. This enables an ICC condition to develop.
M o LOOP coincident with reactor trip. Therefore, steam dump and bypass to the condenser is not available. This would limit the secondary depressurization to the relief capacity of the ADV.
o AFW was throttled to prevent SG overfill. No other operator action was assumed until (and so that) an ICC condition develops.
Normally, the operator would perform several actions based on the E0Ps to stabilize and recover the plant.
t o The assumed indict. ting parameter for ICC was a reading of 1200 F at the core exit thermocouples.
o once Low Pressure S_afety Injection (LPSI) is initiated, the system is considered under control.
All other assumptions, sununarized in Table 3.1 and discussed in Section
~
. 3.1, apply as well. ,
4.2.2 ICC Results O- The transient was identical for all recovery methods up to a reading of 1200 F at the core exit thermocouples. The results up to this point are -
sutunarized below.
The break flow caused the RCS pressure to decrease. At approximately 830 seconds, a reactor trip occurred on low pressure. LOOP occurred coincident with reactor trip, SI returned on line 20 seconds after LOOP and failed to start, AFW returned on line 30 seconds after LOOP and actuated. RCS '
, 1 pressure equilibrated corresponding to secondary conditions. Lack of SI caused the RCS mass inventory to be depleted. Core uncovery occurred after approximately three and one-half hours. This caused a rapid increase in core
~
exit temperature. At approximately 13,800 seconds (~ 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />), .the core exit temperature reached 1200 F. The remaining portion of the transient is
- discussed in Sections 4.2.2.1, 4.2.2.2 and 4.2.2.3 for each recovery method analyzed.
1 l
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1 r.~,--~,-. ..-....-,...-,_....,-,_._-_._,-_____..,____,-__.-,,,,.__-.--..._._,...-_-_-____--_..I
4.2.2.1' ICC Recovery Action - Secondary Depressurization thia case is the first of the two analyses that identify the recovery procedure outlined in FR-C.l. The method of recovery was secondary depressurization via the ADV. The purpose was to determine if the limited relief captcity of the ADV would impede the recovery. The sequence of events is shown'in Table 4.2.1 and results are illustrated in Figures 4.2-1 through 4.2-8. ,
, 1 ,
s 9 .
The first operator action for core recovery was secondary depressurization. A 30-second delay was added after reaching a value of 1200 F because the operator may attempt to dump to the condenser if possible before using the ADV. The results show that the ADV was adequate for depressurizing the system and cooling down the core exit temperature. The peak core exit temperature observed was approximately 1390 F. At approximately 18,100 seconds the RCS pressure was g brought below the LPSI .
shut-off pressure, thus actuating LPSI flow. Th'e plant is considered under control once LPSI is initiated, thus answering the concern about secondary depressurization using the ADV. \
4.2.2.2 ICC Recovery Action - RCP Restart / Delayed Secondary D49Pd3surization
(
This case is the second of the two analyses that identify the recovery procedure outlined in FR-C.1. The method of recovery was RCP restart with a delayed secondary depressurization via the ADV. The purpose was to demonstrate if the forced circulation from restarting one RCP is adequate to cool down the core exit temperature. The sequence of events is shown in
's Table 4.2.2 and results are illustrated in Figures 4.2-9 through 4.2-16.
The first operator action for core recovery was to attempt to depressurize the secondary system. The condenser was unavailable due to the LOOP. The ADV was assumed to be inoperable from the Control Room. A delay time of two minutes was added after reaching a value of 1200 F, this represents the time spent by the operator trying to depressurize the secondary system.
O v
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_ , _ _ . - ,. , . ,, ,.m,_,__.____m- _ , , _ _ _ _ . _ _ _ . , . . . _ _______.-_%, _ _ _ _ . _ ___
Off-cito power was cscumed to be r ster:d p:rmitting RCP r ctset, but the condenser was still assumed unavailable for steam dump. At this time, plant personnel were sent to manually open the ADV. (Ten minutes were assumed for this task.) Pump restart was performed in Loop 1, the broken loop.
Restart could be perforined in any loop; however, it would be less effective in the broken loop.
The results show that RCP restart in the broken loop was adequate for cooling down the core exit temperature. The peak core exit temperature observed was approximately 1340 F. After the ten-minute delay, the ADV was opened, thus depressurizing the RCS. The system response following the secondary depressurization was similar to that observed in Section 4.2.2.1.
Therefore, the TREAT anslysis was terminated at 16,615 seconds (~ 4.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />).
The remaining times in Table 4.2.2 were approximated from the results of Section 4.2.2.1. Thus, the RCS pressure can be brought below the LPSI shut-off pressure to initiated LPSI flow. The plant is considered under control once LPSI is initiated, thus answering the concern about delayed secondary depressurization.
4.2.2.3 ICC Recovery Action - Primary Depressurization In this case, the attempted method of recovery was primary depressurization by the two pressurizer PORVs. The purpose was to determine the effects of opening the pressurizer PORVs. WOG analyses have shown that for generic Westinghouse plants, core recovery cannot be achieved solely by depressurizing with the pressurizer PORVs (see Reference 12). The same results were expected for Maine Yankee. The sequence of events is shown in Table 4.2.3 and the results are illustrated in Figures 4.2-17 through 4.2-24.
The first operator action for core recovery was to attempt to depressurize the secondary system. The condenser was unavailable due to the LOOP. The ADV was assumed to be inoperable from the Control Room. An attempt was made to restart RCP. RCP was unavailable due to LOOP. A delay time of three minutes is assumed, this represents the time spent by the operator trying to restart a RCP and to depressurize the secondary system. The operator attempted to depressurize using the pressurizer PORVs. The PORVs were opened. At this time, plant personnel were sent to manually open the
l ADV. (Ten Cinutcs wera ccsumed fcr this tcsk.) The results show that depressurizing solely with pressuriser PORVs is not adequate for core
() recovery. The core exit temperature continued to. increase sharply. After approximately 1480 F, the core exit temperature began to have severe oscillations. This response is not physical, but a limitation of the code (see Reference 11). Thus, depressurizing with the PORVs could not be demonstrated a successful ICC recovery action.
4.2.3 Conclusion Results of the analysis show that Maine Yankee can recover from an ICC condition (defined as 1200 F at the core exit thermocouples). The recovery methods that were demonstrated acceptable for Maine Yankee were:
o Secondary depressurization using the ADV; Section 4.2.2.1.
o RCP restart and delayed secondary depressurization via the ADV; Section 4.2.2.2. .
() Recovery could not be demonstrated by depressurization with pressurizer PORVs, alone, in the absence' of high pressure safety injection, see Section 4.2.2.3.
I t
V r'
h
TABLE 4.2.1 Sequence of. Events p e For an ICC v) (Recovery Action-Secondary Depressurization)
Time (Sec) Events 0.0 Start of analysis. Study state at 100% RTP 50.0 A1 inch diameter break occurs in the cold leg (loop --1).
370.0 Low Pressure setpoint is reached (P-RCS=1850 PSIA). Reactor trip occurs with a coincident LOOP (i.e. RCP, MFWRVs, FWRBVs and Steam Dump
, Trip). SG safetics open.
390.0 SI returns on line (P-RCS=1600 PSIA), powered by Diesel _ Generator. SI pumps fail to start
- l. (i.e. no SI). .
400.0 AFW returns on line and actuated (L-SG=35% narrow range), powered by Diesel Generator.
AFW was throttled to prevent SG overfill
~'s,()% (maintain NR SG level between 35%.and 70%).
No other operator action, so that the ICC condition develops.
13800.0 The core exit thermalcouples reach a temperature of 1200*F (e.g. the indicating parameter for ICC).
13830.0 Begin secondary depressurization. Method of depressurization is the ADV. A 30 second delay is added after the core exit thermalcouples reach 1200*F because the operator may attempt to dump to the condenser if possible before using the ADV.
16950.0 RCS pressure decreases below 245 PSIA.
The accumulators discharge into the RCS (Nitrogen pressure, 245 PSIA). The RCS begins to recover.
18100.0 RCS pressure decreases below 196 PSIA. The LPSI starts to discharge into the RCS. LPSI shut-off pressure = 196 PSIA.
Once LPSI is initiated the system is considered f}N/ under control.
End-Of-Transient
. TABLE 4.2. 2 .
fr.g Sequence of Events t 1 For an ICC
\~'
(Recovery Action-RCP Restart / Delayed Secondary Depressurization)
Time-(Sec); Events 0.0 Start of analysis. Steady State at 1004 RTP.
50.0 A'1 inch diameter. break occurs in the cold leg >
(loop - 1).
1 370.0 Low Pressure setpoint is reached (P-RCS=1850 PSIA). Reactor trip occurs with a coinsident LOOP (i.e. RCP, MFWRVs, FWRBVs and Steam Dump Trip). SG safetics open.
390.0 SI returns on line (P-RCS = 1600 PSIA), powered by Diesel Generator. SI pumps fail to start -
(i.e. no SI). J.
400.0 AFW returns on line and actuated (L-SG=35% narrow range), powered by Diesel Generator.
AFW was throttled to prevent SG overfill (maintain NR SG 1evel between 35% and 70%).
t No other operator action, so that the ICC condition develops.
13800.0 The core exit thermalcouples reach a temperature of 1200*F (e.g. the indication parameter for ICC).
Attempt to depressurize the secondary system.
The condenser is unavailable due to the LOOP.
The ADV is assumed to be unoperable from the-control room. A delay time of two minutes is assumed, this represents the time spent by the operator trying to depressurize the secondary system.
L 13920.0 offsite power is restored,-permitting RCP restart, but the condenser is not available for steam dump. The RCP is started ,on loop-1. This is the loop with the break.
i At this time plant personnel are sent to manually open the ADV (10 minutes is assumed to open the
^ v>-
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1 4
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. . -_. - . . - .. .-- -~ . . . .
t d'
s18700.0 RCS pressure decreases below 245 PSIA. The accumulators l
'/ discharge into the RCS (Nitrogen pressure, 245
'k PSIA). The RCS begins to recover.
i' -%19800.0 RCS pressure decreases below 196 PSIA. The LPSI starts to discharge into the RCS. LPSI shut-off pressure = 196 PSIA.
Once LPSI is initiated the system is considered under control.
4 i End-Of-Transient t
i*
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4
)
t 2
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TABLE 4.2.3 Sequerce of Events For
/7 An ICC
( ,) '
(Recovery Action-Primary Depressurization)
Time (Sec) Events s.
~
0.0 Start of analysis. Steady State at 100% RTP.
50.0 A1 inch diameter break occurs in the cold leg (loop - 1).
370.0 Low Pressure setpoint is reached (P-RCS=1850 PSIA). Reactor trip occurs with a coinsident LOOP (i.e. RCP, MFWRVs, FWRBVs and Steam Dump Trip). SG safetics open.
'390.0 SI returns on line (P-RCS = 1600 PSIA), powered by Diesel Generator. SI pumps fail to' start *
(i.e. no SI).
400.0 AFW returns on line and actuated (L-SG=35% narrow range), powered by Diesel Generator.
(') AFW was throttled to prevent SG overfill (maintain NR SG 1evel between 35% and 70%).
No other operator action, so that the ICC condition develops.
13800.0 The core exit thermalcouples reach a temperature of 1200 F (e.g. the indication parameter for ICC).
Attempt to depressurize the secondary system.
The condenser is unavailable due to the LOOP.
The ADV is assumed to be unoperable from the control room. Attempt to restart RCP. RCP is unavailable due to LOOP. A delay time of three minutes is assumed, this represents the time spent by the operator trying to restart a RCP and to depressurize the secondary. system.
13980.0 The pressurizer PORVs are opened.
~
At this time plant personnel are sent to manually open the ADV (10 minutes is assumed to open the ADV).
g~g 14400.0 Unable to achieve core recovery solely
(~,) - with RCS depressurization using the pressurizer PORVs.
End-of-transient
A , _ =4 O O O 1
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4 ICC Results (Recovery Action-Secondary Depressurization)
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SG PRESSURE (PSIR) PRESSURIZER PRESSURE (PSIR) 0.0 500.0 1000.0 1500.0 0.0 500.0 i
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w, Delayed Secondary Depressurization) o Figure 4.2-11 d t i f 3
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ICC Results (Recovery Action-RCP Restart /
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(y a Figure 4.2-13 g i i i 9
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4.3 LOMS Analysis
() This transient is a LOHS from 100% RTP with failure of the steam dump and bypass system. Both the main feedwater and AFW were assumed to fail. The consensus of Reference 3 is that a reasonoble indicating parameter to initiate feed and bleed is zero indication on the narrow-range SG 1evel for two of three steam generators. The reasoning was based on the potential inadequacy of two systams. First, Maine Yankee has a relatively small pressurizer PORY l capacity to thermal power ratio as compared to generic Westinghouse plants.
Second, wide-range SG level instrumentation for Maine Yankee is not qualified and, therefore, may not be reliable for adverse containment conditions.
Yankee regards this indication as being premature in that the narrow-range SG level can be lost for other types of events shortly after reactor trip. For this reason, Yankee suggests that a more reasonable indicating parameter for Maine Yankee would be an increasing RCS pressure and temperature. The purpose of this analysis is to:
! o Determine if the revised LOHS EOP will be adequate to allow for plant recovery with an indicating parameter of increasing RCS
() pressure and temperature.
o Determine if maintaining or recovering from feed and bleed will be limited by Maine Yankee's emergency equipment.
o Determine the maximum operator action time to initiate feed and bleed.
I The key operator recovery actions from the LOHS EOP, as denoted in Reference 3, were:
I o Initiate feed and bleed. Feed and bleed is accomplished by two pressurizer PORVs and SI flow, respectively, o Obtain AFW flow in each loop and regain a SG KR reading of 35%, l 1
o Close one pressurizer PORV, and then
o At the onset of decreasing pre:sure eleco the remaining PORV cnd cooldown.
O) t v
Two cases were analyzed:
o The base case, Section 4.3.2.1.
o Delayed feed and bleed, Section 4.3.2.2.
The scenario in Section 4.3.2.1 was evaluated to validate the LOHS IOP developed by Westinghouse while using an indicating parameter of increasing RCS pressure and temperature for initiating feed and bleed. Section 4.3.2.2 was performed to assess the maximum operator action time to initiate feed and bleed.
4.3.1 Assumptions The following assumptions apply to the LOHS cases:
() o Steam dump is considered the preferred way to cooldown, if available. However, to construct a more limiting case, the steam dump was assumed to be unavailable and the ADV was used to cooldown. (See Reference 13.)
o LOHS consists of a system malfunction and/or operator error that results in the closure of the NFWRVs, FWRBVs and AFW valves in all loops and/or main feedwater and AFW pump failure.
o ,
Initiate feed and bleed when:
- a. Pressurizer pressure starts to increase, and
- b. Core exit temperature starts to increase, o Feed and bleed consisted of SI (all loops) and opening two pressurizer PORVs.
O I
1 o For reasons of SG metal integrity, in the recovery phase, each dried-out SG was fed at 150 gym for five minutes.- After the five minutes, the SG was fed one-third of the total AFW pump flow rate.
This is assumed to have no effect on the transient.
i o Maintaining feed and bleed for ten minutes will demonstrate the integrity of Maine Yankee's emergency ~ equipment.
] o A subcooling margin greater than 12 F (instrument uncertainty) from the core exit thermocouple will prevent voiding'in the t.pper head.
o SI Termination Criterion:
l o 12 F subcooling, o 2% pressurizer level, o 35% SG NR level in at least one SG, and
.O
, o increasing RCS pressure.
i
- o Once the RCS temperature is on a steady decrease, the system is considered under control.
All other assumptions, summarized in Table 3.1 and discussed in Section 3.1, apply as well.
4.3.2 Results i
4.3.2.1 LONS.(Without steam Dump and Bypass)
The sequence of events is provided in Table 4.3.1 and results are illustrated in Figures 4.3-1 through 4.3-12.
l0 i
?
The loss of all feedwater resulted in dryout of the SGs and cubsequent heatup of the RCS. Reactor trip was caused by low SG narrow-range level. The secondary pressure remained constant during the linear depletion of SG mass
- inventory. SG dryout occurred at approximately 1800 seconds. This caused an issnediate increase in RCS pressure and temperature.
Feed and bleed was initiated on indication of increasing RCS pressure and temperature. All RCPs were tripped to reduce the heat addition into the RCS. The pressuriser went solid shortly thereafter. Feed and bleed was i successful in removing decay heat. The feed and bleed process was maintained for 10 minutes to demonstrate system integrity. Following this time period i
AFW was restored (one loop at a time).
Once a SG narrow-range level of 35% was achieved in at least one loop, l
one of two pressurizer PORVs was closed. This caused a reduction in heat removal, causing an increase in RCS pressure and temperature. SI was t
throttled to decrease the RCS pressure and temperature. The remaining PORV was closed, terminating feed and bleed. The ADV was inunediately opened to remove the decay heat. SI was terminated (on the SI termination criteria) to stop the RCS pressure surge and to restore a pressurizer level indication.
The case was run out to near Residual Heat gemoval (RHR) conditions where a steadily decreasing temperature was observed. RHR entry criteria are:
o RCS pressure 1 365 psia o RCS tegerature 1 400 F
- 4.3.2.2 LOHS (With Delaved Feed and Bleed) i The analysis in Section 4.3.2.1 assumed a 50 psi increase in RCS pressure and a 5 F increase in RCS temperature as an indication for initiating feed and bleed. These indicating parameters are purely visual (i.e., no audio alarms). Therefore, additional calculations were made to simulate delayed operator action until an audio alarm spurs operator action.
The pressure response was extrapolated linearly up to the pressurizer PORV setpoint (2400 psia), see Figure 4.3-13, where an audio alarm would sound.
O
_ d
1.
The temperc.tura rcep:nso was exterp31cted lin:ccly until s turation
- . conditions, see Figure 4.3-14. A linear ramp is conservative since decay heat would be decreasing within this time. Approximately 400 seconds after the RCS j pressure begins to increase, the PORVs open to maintain RCS pressure at 2400 psia. The RCS temperature would not reach saturation until approximately 1840 i, seconds after.it begins to increase. The pressurizer should so solid shortly ,
after the PORVs open; thus venting liquid. After the pressuriser is packed i- (i.e., goes solid) another pseudo-equilibrium condition would be established.
' Since there is no SC heat transfer, the core decay heat will go into heating 4
up the RCS fluid. The pressuriser PORVs were shown to be sufficient to
{ accommodate the volumetric expansion up to saturation conditions (see Reference 8). Therefore, the PORVs are adequate until saturation conditions
{ exist. This implies the operator has approximately 30 minutes to initiate feed and bleed from the onset of increasing RCS pressure and temperature, with
] the RCPs running.
Natural Circulation i
If loss of off-site power occurred coincident with reactor trip, the T across the core would be greater, thus, shortening the operator action time to initiate feed and bleed. A reasonable AT across the core for Maine Yankee j' during natural circulation is 26 F (see Reference 8). Assuming this AT, the t time to reach saturation conditions after the onset of increasing RCS pressure and temperature is approximately 1430 seconds. This implies the operator has j approximately 20 minutes to initiate feed and bleed from the onset of increasing RCS pressure and temperature, assuming no RCPs running. A LOHS
. analysis assuming no operator action was performed on a plant that is of the i same class as Maine Yankee in Reference 14. The results of the Reference 14 j analysis agree well with the extrapolated results for operator action times l I
presented above.
i j 4.3.3 Conclusion i
i The results show that the revised LOHS E0k is adequate to recover from this event using an increasing RCS pressure and temperature as an indicating parameter for the operator to initiate feed and bleed. The results also show that the relief capacity of Maine Yankee's pressurizer PORVs and SI are
}:
L,---_-_---_.---,_. _ _ _ - _ . . . _- . _ - , . - - .__ ,-
adequate to maintain feed and bleed, and that recovery from feed and bleed can be accomplished by the ADV. Following the onset of increasing RCS pressure and temperature, the operator action time to initiate feed and bleed is:
o Approximately 30 minutes assuming forced flow, i
o Approximately 20 minutes assuming natural circulation.
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TABLE 4.3.1 Sequence of Events For a LOHS rN (Without Steam Dump and Bypass)
Time (sec) Events 0.0 Start of analysis. Steady State at 100% RTP.
-50.0 Close all MFWRVs and MTRSVs and ATWVs. Steam release by Steam Dump and Steam Bypass.
76.0 Reactor Trip from Low SG Level Narrow Range (L-SC=35%).
1800.0 . Initiate Feed and Bleed when:
- 1. Pressurizer pressure starts to increase, and
- 2. Core exit temperature starts to increase.
This should be just prior to SG Dryout. Feed and bleed consists -
of SIA (all loops) and opening two (2) Pressurizer PORVs. Trip all RCPs to reduce the heat addition into the RCS.
() 2400.0 Maintain core cooling by feed and bleed for 10 minutes.
Initiate AFW (to one SG, loop-1) at 150 GPM for 5 minutes.
2700.0 Set AFW in loop-1 to 177 GPM (one-third of AFW pump-530 GPM). Initiate AFW (to one SG, loop-2) at 150 GPM for 5 minutes.
3000.0 Set AFW in loop-2 to 177 GPM. Initiate AFW (to one SG, loop-3) at 150 GPM for 5 minutes.
3300.0 Set ATW in loop-3 to 177 GPM.
i 4550.0 Maintain the SG 1evel narrow range greater than 35% in at least one (1) SG. Close one (1) pressurizer PORV..
4850.0 Observe a decreasing pressurizer pressure. Close the remaining pressurizer PORV. Throttle 51 to increase RCS subcooling and limit the pressure response. Steam relieved by Atmospheric Dump Valves.
AFW was throttled to prevent SG overfill.
O
1 l
1 1
4880.0 SI. termination criteria: 1 (1) Increasing RCS Pressure ,
(2) A secondary heat sink is available (35% NR level in at least one intact SG). ,
(3) RCSsubcoolingisgreaterthangheallowancefor subcooling uncertainty (e.g. 12 F)-
(4) Pressurizer Level is greater than 0% plus an allowance for pressurizer level uncertatnty (e.g. 2%).
l Terminate SI (to all loops) and initiate normal charging and letdown.
5700.0 Temperature was on a steady decrease, thus all decay heat was being i removed by the SGs.
End-of-transient.
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O O- O SG PRESSURE (PSIA) PRESSURIZER PRESSURE (PSIA) 0.0 200.0 400.0 600.0 800.0 1000.0 i
0.0 500.0 1000.0 1s00.0 2000.0 2s00.0-i i e i e i i g
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LOHS Results ~
p (Without Steam Dump and Bypass)
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O O O (loop-1)
CORE EXIT FLOW (LB/SEC)x10' RUX. FEEDWATER (LB/SEC) 0.0 1.0 2.0 3.0 4.0 s.o 0.0 50.0 100.0 150.0 .
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LONS RESULTS (Without Steam Dcap and Bypass) .
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l
, 4.4 small Break LOCA Cooldown/Depressurization Analysis This transient is a small break LOCA cooldown/depressurization from 100% RTP with a coincident loss of off-site power with reactor trip. 'The LOCA is caused by a one-inch diameter break located on the cold leg of Loop 1. A plant-specific analysis for post-LOCA cooldown was recommended for three reasons:
o Maine Yankee has the capability rf throttling SI in addition to turning SI pumps off sequentially. This may provide a simple SI i reduction schosa ccapared to that used for the Westinghouse i Emergency Response guidelines (RRGs). The generic Westinghouse plants do not beve SI throttling capability.
o If steam duay to the condenser were not available, the Atmospheric Dump Valve (ADV) must be used. Its limited capacity may restrict the cooldown rate to less than 100 F/hr.
o Maine Yankee has restricted flow paths to the upper head as
() compared to a generic Westinghouse plant. A significant amount of voiding in the upper head may make pressurizer level difficult to control and complicate the recovery.
The key operator recovery actions from the small break post-LOCA cooldown EOP, as denoted by Reference 3, were:
o Cooldown the RCS, o Regain a 12 F subcooling margin, and then o Depressurize the RCS.
Results of this analysis are presented in Section 4.2.2. The analysis was performed to validate the small break post-LOCA cooldown E0P developed by Westinghouse.
l O
4.4.1 Assuustions The following assumptions apply to the LOCA cooldown/depressurization case:
o A one-inch diameter break was assumed in the cold leg. This size break would be large enough to meet the RCP trip criteria (trip criteria is less than 25 F subcooling in the RCS), but small enough that pressurizer level could be controlled.
o The one-inch LOCA occurred in Loop 1. The loop was chosen arbitrarily and is not assumed to affect the results.
o LOOP was assumed. This would minimize the RCS cooldown rate to the relief capacity of the ADV.
o ES-1.2 (i.e., the Post-LOCA Cooldown and Depressurization EOP) entry was assumed to be reached approximately twenty minutes after reactor trip.
O o once RHR entry conditions are met, the system is considered under control. RHR entry criteria are:
RCS pressure s 365 psia RCS temperature 1 400 F
. All other assumptions, sumarized in Table 3.1 and discussed in Section 3.1, apply as well.
4.4.2 Results The sequence of events is provided in Table 4.4.1 and results are illustrated in Figures 4.4-1 through 4.4-12.
l The break flow caused the RCS pressure to decrease. Reactor trip occurred on low pressurizer pressure. Due to LOOP coincident with reactor O
trip. SI actuated 20 seconds after LOOP and AFW actuated 30 seconds after LOOP. The cold water addition from the AFW resulted in a decrease in the SG pressure below the safety valve setpoint, and restoration of the SG 1evel.
Following entry into ES-1.2 (approximately 20 minutes after reactor trip), RCS cooldown was initiated via the ADV. The cooldown rate was 100 F/hr (i.e., cooldown limits). The ADV increased primary to secondary heat transfer. As a result, subcooling margin increased.
When a 12 F RCS subcooling margin was obtained RCS depressurization was initiated. The method of depressurization was the pressurizer PORVs. SI was throttled to maintain an adequate pressurizer level. The transient was terminated at RHR conditions.
Plant recovery was demonstrated by a TREAT analysis for a small break iSCA without off-site power. Based on the results and the similarity in the transient sequence, the case with off-site power available is expected to be bounded by these results (see Reference 15).
4.4.3 Conclusion The results show that the small break LOCA cooldown/depressurization EOP is adequate to recover from a one-inch diameter break in the cold leg.
The results also show that:
o Throttling leads to a simpler SI reduction scheme as compared to that used in the Westinghouse ERGS.
o The Maine Yankee ADVs are a capable means of RCS cooldown.
o Upper head voiding did not impede the recovery or the control of pressurizer level.
l l
0 4
TABLE 4.4.1 Sequence of Events For a Small Break LOCA O (Without Off-site Power)
Time (Seconds) Events 0.0 Start of analysis. Steady State at 1004 RTP.
50.0 A 1 inch diameter break occurs in the cold leg (loop-1).
370.0 Low Pressure setpoint is reached (P-RCS=1850 PSIA). Reactor trip occurs with a coincident LOOP (i.e. RCP, MFWRVs, TWRBVs and Steam Dump Trip). SG safeties open.
390.0 SI returns on line and actuated (P-RCS=1600 PSIA), powered by Diesel Generator.
400.0 AFW returns on line and actuated (L-SG=35% narrow range), powered by Diesel Generator.
AFW was throttled to prevent SG overfill (maintain NR SG 1evel between 35% and 70%).
No further operator action is modeled until af ter ES-1.2 entry (ES-1.2 entry is assumed to be approximately 20 minutes after reactor trip).
1570.0 Cooldown at the maximum rate possible (100*F/HR). Method of cooldown is the ADV, rated at 54 full power steam flow.
1600.0 A 12*F subcooling was maintained in the RCS. Initiate RCS depressurization Depressurize without pressurizer heaters. Method of depressurization is the PORV.
1800.0 End RCS depressuriaztion, pressurizer level ji 20% NR.
1840.0 Throttle SI to maintain a 12'F subcooling in the RCS and to maintain pressurizer level between 20% and 90%.
6000.0 RCS reaches RHR conditions: 1. RCS pressure gi 365 PSIA, and
- 2. RCS temperature g; 400'F.
End-Of-Transient O
i
1 Small Break LOCA Results (Without Off-site Power)
ON o Figure 4.4-1
- t i f I l C
O Q_ O.
- 8_ _
ua Z
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Figure 4.4-2 o,
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Small Break LOCA Results (Without Off-site Power) x g Figure 4.4-3 o' , , t 1 I 8
c
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en u> o -
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o 0.0 10$0.0 20$0.0 30$0.0 40b0.0 50$0.0 6000.0 TIME (SEC)
O Figure 4.4-4 o
t I t t 1 E o.
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O 0.0 1000.0 2000.0 3000.0 4000.0 5000.0 6000.0 TIME (SEC)
Small Break LOCA Results .,
-. (Without Off-site Power)
Figure 4.4-5 t t t 1 -I O
o t.n o N U* -
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_ a-I h3O "o
J fk U. ~
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v e o U U Z
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I i i i I 0.0 1000.0 2000.0 3000.0 4000.0 5000.0 6000.0 TIME (SEC)
O Figure 4.4-6 o
t i
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Od O 0.0 1060.0 20o0.0 3060.0 TIME (SEC) 40o0.0 5060.0 e000.0
-100-
Small Break LOCA Results (Without Off-site Power)
[D
\j Figuts 4.4-7 i
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Cr 9 l G- o 0.0 10$0.0 20$0.0 30$0.0 40$0.0 50$0.0 6000.0 TIME (SEC)
Figure 4.4-8 Oo Ud CD e
' t , , ,
N CD J
>o Ed e _.J o - _
$C 3 Q-e E s
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$ i g-* 0.0 1000.0 2000.0 3000.0 4000.0 5000.0 6000.0 TIME (SEC)
-ioi_
O O o (Loop-1)
CORE EXIT FLOW (LB/SEClx10' RUX. FEEDWATER (LB/SEC)-
o.o 1.0 2.0 3.0 4.0 s.o 0.0 50.0 100.0 iso.o i i i i i i g
b b s F a g_
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8 i g i i i i g i O
O
Small Break LOCA Results (Without Off-site Power)
O~ Figure 4.4-11 L. O. t ! I I O I (f")' o u"
u Go W d- ~
ag m__ -
a_
r i u 9 o.
g_ _
zm O
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eo_ _.
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in S o a i i i i 5000.0 S000.0 e 0.0 1000.0 2000.0 3000.0 4000.0 TIME (SEC)
Figure 4.4-12 o
' t f f f C) 1 ns E
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um 1
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TIME (SEC)
-103-
l l
5.0 SUIStARY AND CONCLUSION The purpose of this report was to demonstrate acceptability of the revised EOPs and to examine plant-specific capability of the Maine Yankee ;
equipment. Four events were selected for this task and analyzed with a best-estimate model using the TREAT code. The events selected were:
o Steam Generator Tube Rupture o Inadequate Core Cooling o Loss of Secondary Heat Sink o Small Break LOCA Cooldown/Depressurization Results of the SGTh analysis show that Maine Yankee can recover in a .
timely manner without SG overfill. The recovery methods include:
o Cooldown with the Atmospheric Dump Valve (ADV) and depressurization with the pressurizer auxiliary spray.
o Cooldown with the ADV and depressurization with one (or two) pressurizer P_ilot-operated Relief Valve (s) (PORV).
Recovery could not be demonstrated cooling down with the steam dump system and depressurizing by pressurizer normal spray. In this scenacio, the steam dump was throttled back to the flow capacity of the ADV. For a larger cooldown rate, it may be possible to recover with normal pressurizer spray prior to SG overfill.
Results of the ICC analysis show that Maine Yankee can recover from an ICC situation with an ICC entry condition reading of 1200 F at the core exit thermocouple. Upon reaching this value, the operator will begin plant recovery by reducing the RCS pressure and temperature. The recovery methods were:
o Depressurizing with the ADV only.
-104-
o Mixing with a RCP restart and d21cywd deprescurizaticn with ADV.
Recovery could not be demonstrated by depressurizing with pressurizer PORVs, alone, in the absence of high pressure safety injection.
Results of the LOHS analysis show increasing RCS temperature and 4
pressure are adequate indicating parameters for initiating feed and bleed.
Plant recovery was demonstrated from the LOHS event, assuming loss of steam dump and bypass. Following the onset of increasing RCS pressure and temperature, the operator action times to initiate feed and bleed are:
l o Approximately 30 minutes, assuming forced flow.
o Approximately 20 minutes, assuming natural circulation.
l Results of the small break LOCA cooldown/depressurization analysis demonstrate that:
o Throttling SI leads to a simpler SI reduction scheme as compared to that used in the Westinghouse ERGS. Plant recovery was demonstrated from a LOCA (one-inch diameter break) without off-site power.
o The ADV is adequate for RCS cooldown.
o Upper head voiding did not complicate the recovery.
l l
t O
-105-
--w-- . . _ . , , . _ . , , , - , . , _ _ . _ , , , ,c..my- ,..___,___,_,,,.,.,y ,_.__.m- , ,,,_,m-,._ ,-_.,,_m,_ - ,._._. , f
6.0 REFERENCES
- 1. " Transient Real-Time Engineering Analyses Tool Code (TREAT)," WCAP-10771 February 1985.
- 2. SED-0SA-0414. " Westinghouse Emergency Response Guideline Development for Small LOCA Recovery Using the TREAT Interactive Thermal Hydraulic Code,"
Letter from H. Julian (Westinghouse) to G. Stowers (Maine Yankee),
November 11, 1984.
- 3. " Maine Yankea Analysis Scenario Definition Document," April 4, 1985, Revision O.
- 4. YAEC-1160, " Application of Yankee - WREM - Based Generic PWR ECCS Evaluation Model to Maine Yankee (3-Loop Sample Problem)," July 1978.
- 5. YAEC-1447, " Application of RETRAN-02 MOD 02 and BIRP to the Analysis of the MSLB Accident at MYAPC," September 1984.
- 6. "MY Startup Test Report."
- 7. YAEC-1484, " Primary Fluid Subcooling as a Reactor Coolant Pump Trip Criterion for Maine Yankee," May 1985.
- 8. TAG 85-270, " Response to Westinghouse Comments of the EOP Analyses -
Cases I, III and IV," Memo from J. D. Robichaud to P. L. Anderson, September 1985.
& G. Stowers, June 26, 1985.
- 10. WCAP-10698, **SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill, December, 1984.
- 11. TAG 85-245, " Case II ICC - EOP Upgrade," Memo from J. D. Rob 3.chaud to P. L. Anderson, September 20, 1985.
- 12. WCAP-9753, " Inadequate Core Cooling Studies of Scenarios with Feedwater Available Using the NOTRUMP Computer Code," (June 1980).
- 13. TAG 85-171, " Case III LOHS - EOP Upgrade", Memo from J. D. Robichaud to G. Stowers (NY), June 27, 1985.
- 14. CEN-114-P, Amendment 1-P, " Review of Small Break Transients in Combustion Engineering Nuclear Steam Supply Systems," July 1979.
- 15. TAG 85-177, " Case IV LOCA - EOP Upgrade", Memo from J. D. Robichaud to G. Stowers (NY), July 5, 1985.
-106-
MY WRITERS GUIDE - REVISION 1
~\
(G LIST OF ILLUSTRATIONS I'
Figure -1. STAPOARD LEADER BOX FORMAT . . . . . . . . . . . . . . . . 10 Figure- 2. COVER PAGE FORMAT . . . . . . ... . . . . . . . . . . . . 11 Figure 3. ACTION PA T (2-COLUMN) FORMAT . . . . . . . . . . . . . . 11 Figure 4. COVER SHEET EXA WLE FOR PROCEDURE 2-70-0 . . . . . . . . . 15 Figure 5. COVER SHEET EXAWLE FOR PROCEDURE 2-81-3.1 . . . . . . . . 16 Figure. 6.,EXAWLE INSTRUCTION STEPS . . . . . . . . . . . . . . . . 19 Figure . 7. EXAWLE GRAPH . ......................28 Figure 8. EXAWLE TABLES ......................30 '
Figure 9. EXAWLE ATTACHENT PACE FORMAT , . . . . . . . . . . . . . 31 Figure 10. EXAWLE FOLDOUT PA E FORMAT ...............33 Figure 11. BLOCK FORMAT FOR STATUS TREES . . . . . . . . . . . . . . 35 Figure 12. STATUS TREE PRIORITY IDENTIFICATION . . . . . . . . . . . 36 O
List of Illustrations 11 0428A-GNS
MY WRITERS GUIDE - REVISION 1 PURPOSE Af0 SCCPE The purpose of this document is to provide administrative and technical guidance for the writing of Emergency Operating Procedures (EOPs) for the Maine Yankee nuclear power plant.
This writers guide' applies to the writing of all emergency operating procedures.
Revisions to the Writers Guide are denoted by a vertical line in the left - i margin.
O .
. PURPOSE AND SCOPE 1 0428A-GNS
MY WRITERS GUIDE - REVISION 1 EOP DESIGNATION abo NLM3ERING The EOPs specify operator actions to be taken during plant emergency situations to return the plant to a safe stable condition. Each procedure shall be uniquely identified to ~ facilitate preparation, review, use, and subsequent revision.
Procedure Identification a Procedure Number Each EOP will have a unique procedure number assigned to it in accordance - with Maine Yankee Procedure No. 0-06-1, " Procedure Preparation, Classification and Format." To assure consistency when numbering EOPs, discrete numerical ranges have t,een assigned to specific categories of E0P's. Table 1 summarizes the numerical. ranges assigned to the various ECP categories.
The procedure number will be used for document control purposes only.
o Procedure Index Number (PIN) q O Each EOP will have a unique alpha-numeric Procedure Index Number (PIN) assigned to it. To assure consistency, PINS will be assigned using the guidance presented in Table 2.
The PIN . number and procedure title will be used whenever an EOP is referenced from within procedures.
The purpose of the PIN is to provide the operator with a rapid, non-ambiguous method of moving between EOPs.
o Procedure Title Each emergency procedure will have a descriptive title which explains its function or the symptoms which that procedure is designed to mitigate. Table 3 presents a summary of the procedure numbers, PINS, and titles for the current MY ECP set.
EOP DESIGNATION AND NlNBERING 2 0428A-GNS
. - _ . . _ . ~ -- - - - ? -- - --
MY WRITERS GUIDE - REVISION 1
/~ N Y.] -
TAa.E 1 DEFINITIONS OF NUERICAL DESIGNATORS FOR EERGENCY PROCEDURES 2 - the leading designator for all emergency procedures 70 -
a procedure for diagnosis and recovery from design basis events 71 -- a procedure which supplements the recovery actions of a 70 procedure 72 - a procedure which supplements both the 70 and 71 procedures by providing recovery actions.for low-probability or unique event sequences which are not easily covered'in the 70 or 71 procedures or which may complicate or reduce the effectiveness of these procedures Following.a 70, 71, or 72 designator, individual procedures are further distinguished by integer numbers, separated by a decimal point to denote related procedures.
80 - a procedure for diagnosis of challenges to a Critical-(V~~} Safety Function - represented in tree format 81 - a procedure for restoration of a Critical Safety Function (CSF) to a satisfied condition 1 - designator for SUBCRITICALITY CSF 2 - designator for CORE COOLING CSF 3 - designator for HEAT SIW CSF 4 - designator for VESSEL INTEGRITY CSF 5 - designator for CONTAINENT CSF 6 - designator for RCS INVENTORY CSF Critical Safety Function Restoration Procedures shall be identified by their function designator, and by an additional integer to denote a related series Numerical procedure designators will be separated by hyphens.
Examples: 2-70-1 2-71-1.2 0 2-72-1 2 2-80-1 2-81-1.1 EOP DESIGNATION ANOP NLNBERING 3 0428A-GNS
i MY WRITERS GUIDE - REVISION 1 g
b TABLE 2 DEFINITIONS OF DESIGNATORS FOR PROCEDLRE ItOEX NUMBERS E - a procedure for diagnosis and recovery from design basis events ES - a procedure which supplements the recovery actions of an E procedure ECA - a procedure which supplements both the E and ES procedures by providing recovery actions for low probability or unique event sequences which are not easily covered in the E or ES procedures or which may ,
complicate or reduce the effectiveness of these procedures Following an E, ES, or ECA designator, individual procedures are further distinguished by ' integer numbers, separated by a decimal point to denote related procedures. Letter and number designators shall be separated by a hyphen:
Examples: E-0, E-1, E-2, ES-0.1, ES-1.1, ES-1.2 A ECA-0.0, ECA-1.1, ECA-2.1 b F - a procedure for diagnosis of challenges to a Critical Safety-Function - represented in tree format FR - a procedure for restoration of a Critical Safety Function (CSF) to a satisfied condition
's . designator for SU8 CRITICALITY CSF C - designator for CORE COOLING CSF H - designator for HEAT SItK CSF P - designator for VESSEL INTEGRITY CSF Z - designator for CONTAIt#ENT CSF I . designator for RCS INVENTORY CSF V
E0P DESIGNATION Ato NUMBERItO 4 0428A-GNS
MY WRITERS GUIDE - REVISION 1
. TABLE 2 (Continued)
DEFINITIONS OF DESIGNAGRS FOR PROCEDURE-ItOEX NUMBERS Critical Safety Function Status Trees shall be designated by the -
letter F plus a number designator. Number designators shall consist of the number zero plus a decimal integer which shall be assigned sequentially Letter and number designators shall be separated by a hyphen.
Example: -F-0.1 F-0.2 Function Restoration Guidelines shall be designated by the letters FR plus an additional letter which corresponds to the respective Critical Safety Function. All the separate procedures related to a particular Critical Safety Function are assigned decimal integers in increasing order. ,
The procedure-letter and decimal integers are separated from the FR designator by a hyphen.
' Examples: FR-S.1 FR-S.2 i
i I
l l
l D
EOP DESIGNATION AND NUMBERING 5 0428A-GNS
.MY WRITERS GUIDE - REVISION 1 s/
TABLE 3 LISTING OF EERGENCY PROCEDlRE CODES AND TITLES PROC.
NO. PIN EERGENCY PROCEDURE TITLE 2-70-0 E-0 Energency Shutdown from Power or Safety Injection 2-71-0.0 ES-0.0 Event Rediagnosis 2-71-0.1 ES-0.1 Reactor Trip Response 2-71-0.2 ES-0.2 Natural Circulation Cooldown
'2-71-0.3 ES-0.3 Natural Circulation Cooldown with Steam Void in Vessel (with PITS) 2-71-0.4 ES-0.4 Natural Circulation Cooldown with Steam Void in Vessel (without PITS) 2-70-1 E-1 Loss of Primary or Secondary Coolant 2-71-1.1 ES-1.1 SI Termination 2-71-1.2 ES-1.2 Post-LOCA Cooldown and Depressurization 2-71-1.3 ES-1.3 Transfer to Recirculation Cooling (RAS) 2-71-1.4 E S-1.4 Establishing Hot Leg Injection 2-70-2 E-2 Steam Line Break
() 2-70-3 2-70-3.1 E-3 ES-3.1 Steam Generator Tube Rupture Post-SGTR Cooldown Using SG Backfill 2-71-3.3 ES-3.3 Post-SGTR Cooldown Using Steam Dumps 72-0.0 ECA-0.0 Loss of All AC Power 2-72-0.1 ECA-0.1 Loss of All AC Power Recovery, SI Not Required 2-72-0.2 ECA-0.2 Loss of All AC. Power Recovery, SI Required 2-72-1.2 ECA-1.2 LOCA Outside Containment O
EOP DESIGNATION APO NUMBERING 6 0428A-GNS
l l
MY WRITERS GUIDE - REVISION 1 '
_k c)
TABLE 3 (Continued)
LISTING OF E ERGENCY PROCEDURE CODES AND TITLES PROC.
NO. PIN EERGENCY PROCEDURE TITLE 2-80-1 F-0.1 Subcriticality Status Tree 2-80-2 F-0.2 Core Cooling Status Tree 2-80-3 F-0.3 Heat Sink Status Tree 2-80-4 F-0.4 Integrity Status Tree 2-80-5 F-0.5 Containment Status . Tree 2-80-6 F-0.6 Inventory Status Tree 2-81-1.1 FR-S.1 Nuclear Power Generation /ATWS 2-81-1.2 FR-S.2 Loss of Shutdown Margin 2-81-2.1 FR-C.1 Inadequate Core Cooling 2-81-2.2 FR-C.2 Degraded Core Cooling ,
2-81-2.3 FR-C.3 Saturated Core Cooling 2-81-3.1 FR-H.1 Loss of Secondary Heat Sink 2-81-3.2 FR-H.2 Steam Generator Overpressure t'~N 2-81-3.3 FR-H.3 Steam Generator High level (s-) 2-81-3.4 FR-H.4 Loss of Normal Steam Release Capabilities 2-81-3.5 FR-H.5 Steam Generator Low level 2-81-4.1 FR-P.1 Imminent Pressurized Thermal Shock Condition 2-81-4.2 FR-P.2 Anticipated Pressurized Thernal Shock Condition 2-81-5.1 FR-Z.1 High Containment Pressure 2-81-5.2 FR-Z.2 Containment Flooding 2-81-5.3 FR-Z.3 High Containment Radiation Level 2-81-6.1 FR-I.1 High Pressurizer Level 2-81-6.2 .FR-I.2 Low Pressurizer Level 2-81-6.3 FR-I.3 Voids in Reactor Vessel E0P DESIGNATION AND NUMBERING 7 0428A-GNS
MY WRITERS GUIDE - REVISION 1 (m
d Revision Numbering Every separate procedure shall have an assigned Revision number to identify its date of origin. The initial Revision will be designated " Revision 0."
Any PEW procedure (s) added to the E0P set will follow this convention and be assigned " Revision 0" designators.
a Each procedure " Revision" number will have two dates associated _ with it.
The Issue Date (Day, Month, Year) of the Revision and the Review Date (Month, Year) which identifies the month and year during which. the procedure is due for review. These dates will appear on each E(P cover sheet in the document control block.
Page Numbering and Identification Each page of a procedure will be identified by the procedure number, procedure Revision number and page number at the upper right of each page.
Each page number will be specified as "Page N1 of N2," where N1 is the T current page number and N2 is the total number of pages in the procedure.
(d O
E0P DESIGNATION AND NLNBERING 8 0428A-GNS
-MY WRITERS GUIDE - REVISION 1-p v
FORMAT This section describes the format that is to be applied consistently to all Emergency Operating Procedures.
Procedure Organization i
All procedures in the EOP set are to employ a common structure consisting of five elements as shown in Table 4. Any indivicijal procedure might contain only the two required elements, or additional elements as necessary
- to present the intent of the procedure.
Thel sequence of procedure elements is always in the order shown in Table 4.
Page numbering is sequential through all the elements comprising any procedure.
/~ Page Formats D}
All pages of the Emergency Operating Procedures will use the same page structure except the Foldout Page which is discussed later. This page structure employs a boxed-in header across. the top of each .page and a one-inch margin on both sides and at the bottom of the page. The boxed area contains all required procedure identification, including the full title (with PIN), the full procedure number, the revision number, and the page number. Layout of these separate elements is shown in Figure 1.
Each cover sheet will provide space for the approval signatures of the Dept. Head, PORC Secretary and Manager of Operations, indicating acceptance of that procedure for implementation. These personnel will be designated only by their position title on the typed procedure. In addition the cover sheet will identify the Class designation of the particular procedure. The layout of these items, along with the standard procedure identifiers, is shown on Figure 2.
The pages for presentation of operator action steps will use a two-column format below the page header. The left-hand column is designated for
. expected operator action and response, and the right-hand column is designated for contingency actions when the expected response is not obtained. These pages will use underscored headers above the separate columns (including the " step" column) for uniformity (see Figure 3).
O FORMAT 9 0428A-GNS
1 MY WRITERS GUIDE - REVISION 1 y
- V TABLE 4 EMERGENCY OPERATING PROCEDURE ELEENTS COVER SEET (all procedures) . summarizes procedure intent and either entry symptoms or transitions l' INSTRUCTION STEPS (all procedures) - presents the stepwise ,.
operator instructions FIGURES (as required) - presents usually .
graphical data to supplement action steps ATTACHENTS (as required) - presents non-graphical information to supplement action steps FOLDOUT PAGE (as required) - presents information ,
which is applicable throughout the procedure (s) that it follows O I I l- 1 I l Procedure Title l Proc. No. PNNNNNN l l l l PIN l Rev. No. NN I l l l PROCEDURE TITLE I Page ' N1 of N2 l l l1 I I I I I l l 1 l l l l l l l l Figure 1. STAPOARD EADER BOX FORMAT l l l O
FORMAT. 10 0428A-GNS
MY WRITERS GUIDE - REVISION 1 L(
l I l- 1 I l Procedure
Title:
l Proc. No. NPNNNNN I l l l PIN I Class LLL I I I l. PROCEDURE TITLE I Rev. No. NN l l l ,1 1 Issue Date DD/MM/YY I I l l- l Review Date MM/YY' l l l 1 l Page 1 of N2 I l l l Dept. Head PORC M00 I .I I l i I I I
.I I I l
-l A. PURPOSE l i l l l l B. SYWTOMS OR ENTRY C0rOITIONS I I I l 1) {
l l l 2) l l l-l C..Il44EDIATE ACTIONS I '
I I I I '
I Figure 2. COVER PAGE FORMAT s s
l I I I I l Procedure
Title:
l Proc. No. NNNNNNN l l l l PIN I Rev. No. NN l l l -l PROCEDURE TITLE i Page N1 of N2 l l l l l 1I l l l I I STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED l i e i I I I
) l l l Figure 3. ACTION PAGE (2-COLLNN) FORMAT FORMAT 11 0428A-GNS
W WRITERS GUIDE - REVISION 1 The Foldout Page is intended to summarize only the information which an operator should have continuously available, so page content will vary by procedure. Each FOLDOUT PA E shall be titled at.the top in upper case type
" FOLDOUT FOR xxx SERIES PROCEDLRES".
Instructional Step Numbering Procedure steps will be individually numbered. Substeps are lettered sequentially according to expected order of performance. . If the order of substep ' performance' is not important, the substeps are designated by.
bullets (o). If the logical OR, is used, both choices must be designated by bullets with the - OR - on a separate line between the choices offered.' The -
same numbering sche E is to be used in both the right-hand and left-hand columns of the procedures.
Example:
1 = Hicfi-level step
- a. Substep (if necessary)
- 1) Detailed instructions (if necessary)
- 2) . Detailed instructions (if necessary)
- b. Substep (if necessary) o Detailed instructions (if necessary) o Detailed instructions (if necessary)
- c. Substep (if necessary) o . Detailed instructions (if necessary)
E o Detailed Jnstructions (if necessary)
Immediate Action Steps For. those procedures which are the entry procedures into the EOP set, certain initial steps may be. designated "immediate actions". This
. designation implies that those steps are required to be memorized in sufficient detail that the operator can perform the intent of the step from memory, and. therefore without reference to the written procedure. These-steps should be limited to verifications, if possible. Immediate action steps'(Wiere present) are identified by a . dedicated section on the procedure cover sheet.
FORMAT 12 0428A-GNS
- . . . . _ . . . . _ . - . . - . =______._ . . - . - . .
4 MY WRITERS GUIDE - REVISION 1
- O
- Continuous Steps Many of the operator actions provided in a procedure imply continous
. performance throughout the remainder of the procedure. This intent is
- conveyed by the use of appropriate action verbs such as monitor or ,
maintain.
l l
i a
e i~
i Y
O I
i .
i i
t t
i 4
i O
i FORMAT 13 1
'e 0428A-GNS
_ . _ - . _ . _ _ - = - . - - _ - _ _ _ _ _ _ _ _ ._ . _ - _ . . . _ _ -._-. _ _ . . _ . - _ . - _
MY WRITERS GUIDE - REVISION 1 A
b' WRITING EERGENCY OPERATING PROCEDutES The following format is to be applied consistently when writing Emergency Operating Procedures.
Cover Sheet Each cover sheet will contrin three sections in addition to procedure and page designators. The first section will be titled PURPOSE and will briefly describe what the procedure is intended to do. The second.section is a summary of those conditions which require entry into the procedure.
This section will be titled SYM3 TOMS OR ENTRY C0tOITIONS. For procedures that are entry procedures into the E0P set, a symptom summary is sufficient (see Figure 4). For other procedures, which can only be entered by ,
transition from previous procedures, a summary of the entry cunditions (and procedure / step) should be provided (see Figure 5 for the preferred format).
The third cover sheet section will be labeled IMEDIATE ACTIONS and will O state which of the procedure steps, if any, qualify as immediate actions.
'b Figures 4 and 5 should be used as examples for wording on all cover sheets.
WRITING EKRGENCY OPERATING PROCEDURES 14 0428A-GNS
MV WRITERS GUIDE'- REVISION 1 -
O L) l l l l Procedure
Title:
l Proc. No. 2-70-0 l 'l l l E-0 l Class A l l l -l .
l Rev. No. 0 l l l l EERGENCY SHUTDOWN FROM POWER OR l Issue Date DD/ W/YY l l l l SAFETY INJECTION l Review Date MM/YY l l l l l Page 1 of N2 l l l l Dept. Head PORC M00 l l l l l l l l l A. Pt.RPOSE l l l l This procedure provides actions to ensure proper response of the I.
I automatic protection systems following manual or automatic l l emergency shutdown from power or-safety injection, to assess l l plant conditions,and to identify the appropriate recovery procedure. l l l l B. SYMPTOMS OR ENTRY CONDITIONS l l l l 1) The following are synptoms that require a reactor trip, if l l one has not occurred: l O l [ Enter plant specific setpoints and requirements] l V I l l 2) The following are symptoms of a reactor trip: l l l l a. Any First Out annunciator lit. l l b. Rapid decrease in neutron level indicated by nuclear l l instrumentation. l l c. All Trippable CEAs are fully inserted. Rod bottom lights l
.l are lit. l l l l 3) The following are symptoms that require a reactor trip and l l safety injection, if one has not occured: l l [ Enter plant specific setpoints and requirements] l l l l 4) The following are symptoms of a safety injection: l l 1 l a. Any SI annunciator lit. l l b. SI pumps running: l l 1 l C. I WEDIATE ACTIONS l l l l Steps 1 through 14 are INDIATE ACTION steps. l l l l 1 l l
, Figure 4. COVER SHEET EXA WLE FOR PROCEDURE 2-70-0 WRITING EERGENCY OPERATING PROCEDURES 15 0428A-GNS
MY WRITERS GUIDE - REVISION 1
- A N_,l l l l l Procedure
Title:
l Proc. No. 2-81-3.1 l l l l FR-H.1 l Class A l l l l l Rev. No. 0 l l l l EERGENCY SHUTDOWN FROM POWER OR l Issue Date '00/ W/YY l' l l l SAFETY INJECTION l Review Date MWYY l l l l l Page 1 of N2 l -l l l Dept. Head PORC M00 l l l l l l l l l A. . PURPOSE l l .
l
.I This procedure provides actions to respond to a loss of I l secondary heat sink in all steam generators. l 1 l 1 B. SYMPTOMS OR ENTRY C0tOITIONS l l .
l l This procedure is entered from: l l l l 1) E-0, EMERGENCY SHUTDOWN FROM POWER OR SAFETY INJECTION, l l Step 16, when minimum feed water flow is not ensured. l l l p1 l 2) F-0.3, EAT SINK Critical Safety Function Status Tree on a l v l RED condition. l I I l C. IMMEDIATE ACTIONS l l- l l None 1 l l l 1 l l l 1 Figure 5. COVER SHEET EXA WLE.FOR PROCEDURE 2-81-3.1 Operator Actions All steps which direct operator action should address the capabilities of the minimum control room complement of operators. For the purposes of emergency procedures, the following crew composition should be assumed:
o Control room SRO - will physically use the written procedures and direct the actions of the other operators.
o' Control Room Operator 1 (CRO1) - will normally perform required actions and verifications on primary system and safeguards equipment. In addition, the CR01 will perform the duties of the CR02 when the CR02 is absent from the control room, d
WRITING E ERGENCY.0PERATING PROCEDURES 16 0428A-GNS'
MY WRITERS GUIDE - REVISION 1 o Control Room Operator- 2 (CR02)- when present in the control room, ' the f) CR02 will perform required actions and verifications on Balance of
'd Plant. equipment.
Additional control room personnel will be responsible for implementing the plant Emergency Plan (Plant Shift Superintendent, PSS) and Critical Safety Function Status Tree monitoring (Nuclear Safety Engineer, NSE). These functions are assumed to be performed in parallel with the performance of emergency operating procedures.
Steps directing operator action should be written in. short and precise language. The statement should present exactly the task which the operator is to perform. The. equipment to be operated should be specifically identified, and only those plant parameters which are presented by instrumentation available in the control room should be specified. Tasks which require local (outside the control room) actions should explicitly state which local operator or plant area is to be paged. It is not necessary to state expected results of tasks for which the expected ~ result is implicit in the action statement.
Example: Trip All RCPs Align Charging Pump Suction to VCT Continue RCS Cooldown to Cold Shutdown
'O All steps are assumed to be performed in sequence unless stated otherwise d in a preceding NOTE. To keep the individual steps limited to a single action, or a small number of related actions, any complex evolution should be broken down into sequential steps.
If a particular tek MUST EE COMPLETED prior to continuation, this requirement MUST BE STATED CLEARLY using standard wording.
Example:
CAUTION Isolation of ruptured SG(s) shall be complete before continuing to Step 13 unless a ruptured SG is needed for RCS cooldown.
Actions required in a particular step should not be expected to be COH)LETE before the next step is begun. If assigned tasks are short, then the expected action will probably be completed prior to continuing. However, if an assigned task is very lengthy, additional steps may be performed prior to completion of the initial task.
Because of the redundant nature of the reference ERGS, and the repetitive nature of many recovery sequences, checkoff provisions should NOT be incorporated into the writing format.
The wording should be (q/ and construction of operator action steps standardized as much as possible. Therefore, if an action step with a particular intent occurs in several locations within the procedure set, each instance should be worded in the same way, if possible.
WRITING E ERGENCY OPERATING PROCEDURES 17
d 4
MY WRITERS GUIDE.- REVISION 1 '
L?O ~
3
- Refer to' Figure 6 as an example of the format 'for presenting operator actions in the following subsections.
i i
)
1
.' /
I a
e i
e i
T
't
~
, 'f i
}
I 3
1 1
I-i i
O t WRITING E ERGENCY OPERATING PROCEDURES 18
$ 0428A-GNS i .
-Y'W NPW-t"a- 7,19, y " yqw - M %en- yw w + g -w rq e . -mgyv -im- g.+gg
MY WRITERS GUIDE - REVISION 1 O
I- l l l Procedure
Title:
l Proc. No. 2-70-2 l l l l E-2 l Rev.'No. 0 l l l l STEAM LIFE BREAK I Page 2 'o f 4 l l l l 1 I I I l l STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAIPED I i l l l l l l CAUTION I I I I At least one SG shall be maintained available l l for RCS cooldown. 1 I I l Any faulted SG or secondary break should remain l I isolated during subsequent recovery actions 1
-l unless needed for R S cooldown. I I I I . l l 1 Check EFCV(s) On Manually close valves. l l Affected SG(s) - CLOSED l
_p v
l I
(MS-ll, MS-22, MS-33) I I
I .
l l 2 Isolate Faulted SG(s) Manually close valves. l l IF valves CAN NOT l l a. Isolate main feedline E closed, THEN dispatch l l . operator to locally close l l b. Isolate EFW flow block valves valves or block valves. I l l I c. Isolate other secondary l l piping: l I I I 1) i I I I 2) l I I I I l 7 Go T4. E-1, LOSS OF PRIMARY OR I l SEC0tCARY COOLANT, Step 1 -
1 I I I I l -Em- 1 I I I I Figure 6. EXAlfLE INSTRUCTION STEPS WRITING EMERGENCY OPERATING PROCEDURES 19 0428A-GNS
MY WRITERS GUIDE - REVISION 1 r%
Instruction Steps, Left-Hand Column The left-hand column of the two-column format will be used for operator instruction steps and expected responses. The following rules of construction apply:
o High Level Action steps should begin with an appropriate verb, or verb with modifier. Appendix A to this guide presents a list of recommended verbs, along with the their understood meanings.
o Expected responses to operator actions are shown in ALL i CAPITAL LETTERS.
o If a step requires nultiple substeps, the'n each substep will have its own expe ted response.
l Example:
l Check SIT Isolation Valve Status:
- a. Power to isolation valves - AVAILABLE p b.- Isolation valves - OPEN LJ o If only a single task is required by the step, the high-level step contains its own expected response.
Example: Check RCP Status - AT LEAST ONE RUNNING a left-hand column tasks should be specified in the sequence that they are expected to be performed. The user would normally move down the left hand column when the expected response to a particular step is obtained, o Plant-specific information and data are included as necessary to assure correct performance. If an instrument setting is referenced, and that instrument is subject to harsh containment conditions and is environmentally qualified, then two separate values will be provided.
Example: Check RCS pressure - GREATER THAN 2000 PSIG
[1850 PSIG FOR HARSH CONTAIMENT]
o When a step or substep is applicable to three or more objects, list the items separately (in parenthesis or tabular form as appropriate for clarity).
Example: Ensure Reserve Breakers - CLOSED Ato RED FLAGGED (lR, 2R, 3R, .4R)
WRITING EtERGENCY OPERATING PROCEDURES 20 nA?An nm
MY WRITERS GUIDE - REVISION 1 (v '
Example: Align EFPs and AFP to PWST by locally opening their suction valves:
1 PUW l SUCTION VALVE l l l l 1 P-25A l EFW 43 l l 'l I.
I P-25B l AFW-4 l l l l l P-25C I EFW-315 -l l l 1 o Put system alignment checklists in tabular format (see the section on TABLES in this guide). If any checklist is of such length to interfere with procedure usage, it should be placed in an ATTACHENT, and only referenced within the action step. ;['
o When the ACTION / EXPECTED RESPONSE _is not obtained, the user is expected to move to the right-hand column for contingency instructions.
q o All procedures should end with a transition to another
(/ emergency procedure, a transition to a plant procedure or with direction to consult the plant technical staff for guidance.
Instruction Steps, Right-Hand Column The right-hand column (RESPONSE NOT 08TAIED) is used to present
, contingency actions which are to be taken in the event that a stated condition, event, or task in the left-hand column does not represent or achieve the expected result. Contingency actions will be specified for steps or substeps for which useful alternatives are available. The following rules apply to the right-hand column:
o Typically, contingency actions should tell the operator-how to override automatic controls or to initiate manually what is normally initiated automatically.
O Contingency actions should be numbered consistently with the ACTION / EXPECTED RESPONSE for substeps only. A contingency for a single-task high-level step will not be separately numbered but will appear on the same line as its related. step.
o Unlike the left-hand column, contingency instructions are O t be ritteo 1" se"te"ce e r et-o If the rirpt-hand column contains multiple contingency WRITING E E RGENCY OPERATING PROCEDURES 21 0478A-r.NS _
MY WRITERS GUIDE - REVISION 1
/
U actions for a single high-level action in the left-hand column, the phrase " Perform the following:" should be used as the introductory high-level statement. Sufficient space should be allowed between sequential action instructions in the left-hand column to insure that multiple contingency actions in the right-hand column do not overlap subsequent left-hand column steps.
o .If the ricfit-hand column contains multiple contingency actions which do not correspond to multiple substeps in the left-hand column, then different designators should be used in the two columns. Also, the two columns will be separated by a vertical demarcation, using asterisks, extending from the high level step to the last substep.
Example:
Establish Letdown:
- Establish alternate letdown:
- a.
- 1)
- b.
- 2)
- c.
- 3)
The user is expected to proceed to the next step or (q.s o
substep in the left-hand column after taking contingency action in the right-hand column.
o As a general rule, all CONTINGENT transitions to other procedures take place out of the right-hand column.
(Deliberate transitions may be made from the left-hand column. )
o If a contingency action MUST be completed prior to continuing, that instruction MUST appear explicitly in the right-hand column step or substep.
o If a contingency action CANNOT BE COWLETED, the user is expected to proceed to the next step or substep in the left-hand column UNLESS specifically instructed otherwise. When writing the procedure, this rule of usage should be considered in wording subsequent left-hand column instructions.
Use of Logic Terms The logic terms IF, AND, OR NOT, IF NOT, WHEN, CAN NOT, and TWN are to be used to desWibe preHs,ely a set of conditions or a sequence of actions. Logic terms will be highlighted for emphasis by capitalizing and underlining (see Figure 6).
[v]
The two-column formt equates to the following logic: " IF NOT the expected response in the left-hand column, THEN perform the contingency WRITING EERGENCY OPERATING PROCEDURES 22 nA7AA Gm ___. _ _ _ __ _ _ _ __ _
MY WRITERS GUIDE - REVISION 1 q
,/
action in the right-hand column." The logic terms should not be repeated in the right-hand column contingency. However, the logic terms may be used to introduce a SEC0tOARY contingency in the
.ricpt-hand column.
When action steps are contingent upon certain conditions, the step shall begin with the words E or WHEN followed by a description of those conditions, a comma, the word THEN ,and the action to be taken.
E is used for an unexpected, but possible condition.
WHEN is used for an expected condition.
AND calls attention to combinations of conditions and shall be placed between conditions. If more than two conditions are to be combined, a list format is preferred.
OR implies alternative combinations or conditions. OR means either Ene, or the other, or both (inclusive). The use o T OR should be-avoided whenever possible.
IF ... NOT or IF ... CAN NOT should be used when an operator must respond to the second of two possible conditions. IF should always be
('N; used to specify the first condition. (The right Fa'nd column of the
\. ~
two-column format contains an implicit E NOT).
Notes and Cautions Because the present action step wording is reduced to the minimum essential, certain additional information is sometimes desired, or necessary, and cannot be merely included in a background document.
This non-action information is presented as either a NOTE or a CAUTION (see Figure 6).
To distinguish this information from action steps, it will extend across the entire page and will immediately PRECEDE the step to which it applies. Each category (NOTE or CAUTION) will be preceded by its descriptor in uppercase letters centered on the page. Multiple statements included under a single heading shall be separated by one blank line. The text of a NOTE or CAUTION will be presented in a block of lines not more than 5 inches in length.
The entire statement (s), including the descriptor, will be separated from the surrounding action steps by solid, horizontal lines running across the page. Adjacent CAUTIONS and NOTES will be grouped together within the same solid line separators. The separator lines will be separated from any adjacent text by one blank line.
(n)
CAUTION denotes some potential hazard to personnel or equipment associated with the following instructional step. NOTE is used to present advisoty or administrative information necessary to support the WRITING E ERGENCY OPEilATING PROCEDURES 23 0428A-GNS
MY WRITERS GUIDE - REVISION 1 a
r i .
following action instruction.
A CAUTION or NOTE may also be used to provide a contingent transition based on changes in plant conditions.
CAUTION statements shall not contain instructions which direct operator
. actions.
NOTE statements,'in general, should not contain instructions whidi direct operator actions.
CAUTIONS precede NOTES when they occur. together unless the NOTE-contains information which clarifies the CAUTION.
Either a NOTE or a CAUTION may make reference to actions which are expected to be in progress, or to procedural transitions required as a result of changing plant conditions.
.The wording of NOTES and CAUTIONS should be standardized as much as possible. Therefore, if a NOTE or CAUTION with the same intent occurs in several locations within the procedure set, each occurrance should
- be worded in the. same way, if possible.
I Transitions to Other Procedures or Steps
~J Certain conditions require the use of a different procedure or step sequence. Transitions are specified by using the words "go to" followed by the procedure title (in ALL CAPITAL LETTERS), step number 4
for procedures within the ECP 2-70 and 2-80 set, and an arrow pointing to the right. This arrow shall consist at least four "=" followed by a "l". If there is not sufficient space, then the arrow should be put on the next line after the "go to" statement.
Examples: Go To ES-1.1, SI TERMINATION, Step 1 1
========_____=-s-=================>.
Go To OP l-7, PLANT C00LDOWN =====c-Transitions shall NOT contain a " return" feature (e.g., perform steps X through Y in some other procedure and then return).
Transitions to a different step later in the same procedure are specified in a similar manner but without the arrow.
Example: Go to Step 20 Transitions to an earlier step in a procedure are specified by using y the words " return to" without the arrow.
Example: Return to Step 2 WRITING EERGENCY OPERATING PROCEDURES 24 0428A-GNS
MY WRITERS GUIDE - REVISION 1 f).
G .
Transitions to a step which is preceded by a CAUTION or NOTE may include special wording (in ALL CAPITAL - LETTERS) to emphasize that the .
CAUTION or NOTE is to be observed.
Example: IF conditions are-MIT satisfied, TIEN go to Step 22.
OBSERVE CAUTION PRIOR TO STEP 22.
Component Identification Equipment, controls and displays will be identified in " operator la v age" terms. Standard abbreviations and acronyms which may be used throughout the procedures are listed alphabetically in Table 5. Where similar components are used in both primary and secondary systems, it ,
is always necessary to clarify the location, even if the wcrding appears redundant.
Level of Detail To allow an operator to efficiently execute the action steps in a procedure, all unnecessary detail must be removed. Any information which an operator can reasonably be expected to know and remember under stress (based on his training and experience) should not be included.
When a specific device is intended, the exact wording used on the control identification tag and eauipment ID number for that device will be provided. If several equiv n ent devices are available, the common name should be provided.
Location information (for equipment and - controls) is generally to be avoided in the procedures. However, for infrequently-used controls or equipment, limited location information may be included to assist the operator. In addition, any instrument or control not visible from the "at the controls position", and all PAN ALARMS specified shall have location information provided. This information should immediately follow its corresponding equipment or control.
Example:
LOCATION: Back of M'8 Section C O OR V
LOCATION: RH-1-2 WRITING E ERGENCY OPERATING PROCEDURES 25 n/ ann nw:
MY WRITERS GUIDE - REVISION 1
,/}
V Whenever keys are required to bypass interlocks, the key number will be listed as it is numbered in the " Control Room' Key Inventory Index".
This information should be immediately follow its corresponding equipment or control.
Example:
KEYS: B.-13,B-14 Many actuation devices (switches) in the control room are similar, even though the remotely performed functions are not, so certain action verbs listed here are recommended for general usage.
o Use " start /stop" for power-driven rotating equipment.
o Use "open/close/ throttle" for valves, o Use " adjust" to describe a manually maintained process variable (flow, level, temperature, pressure). ,
o Use "open/close" for electrical breakers.
Whenever possible, operator instructions to change the position of any q control device should refer to the actual control device labeling. If a V switch has positions labeled PULL TO LOCK-AUTO-STOP and the intended action is to place the pump in standby, the procedural step should be worded " Place the pump in AUT0" rather than " Place the pump in stand)y."
Figures If needed to clarify operator action instructions, figures shall be added to a procedure. Any figure used will be constructed to fit on the standard operator action page (see Figure 6). Certain rules of construction will apply:
o All wording on the figure shall be at least as legible (type size and spacing) as the instruction ~ steps in the procedures.
t o Each figure will occupy a complete page and will be uniquely identified by a figure number and title. The figure number will consist of an integer number only. Multiple figures will be assigned sequential integers.
o Figure titles will explain the intent or content of the figure.
o The figure number and title will be placed at the TOP of the figure, centered three blank lines below the header block.
WRITING EKRGENCY OPERATING PROCEDURES 26 ?
0428A-GNS
! W WRITERS GUIDE - REVISION 1
/O O o If the figure is a graph, all the numbers and wording will be horizontal. If space is a problem for long names alongside the y axis, the name should be placed above the coordinate line.
By convention, the independent variable is plotted on the horizontal (X) axis. Grid line density should be consistent' with the resolution expected from the graph, but line spacing should not be finer than 1/8 inch. Any labeling required on the grarn will have a white (not graph) background. Figure 7 is an exainple figure showing presentation of a graph.
o All figures for a procedure are numbered sequentially and are located immediately after the instruction step pages and prior to any ATTACHENTS which might follow. Figure pages are numbered as pages of that procedure. Any figures required for an ATTACH E NT are assigned numbers in sequence with the procedure figures.
o References to a figure from an action step should use the
/ figure number, page number, and title.
/
i
/
O i
i
/
/
J O
WRITING EERGENCY OPERATING PROCEDURES 27
< 0428A-GNS
MY WRITERS GUIDE - REVISION 1 O
I I I I Procedure
Title:
l Proc. No. 2-81-4.1 l l l l FR-P.1 1 Rev. No. O I I I IIMMIENT PRESSURIZED THERMAL SHOCK CONDITION I Page 9 of 9 l l 1 I I I I I I I I I I <
l FIGURE 1 1 I POST-SOAK C00LDOWN LIMIT I I I I I l RCS PRESSURE (PSIG) l I .
I I 3000 l l l l l l l l l l l l 2500 l l 1 .
I I l- 1 I I I I I 2000 I I I I I I I l 1500 l l l l l l l l l l l 1000 1 I I I I I I I I I I i 500 l l l
'l I I I I I I I I O I I I l 0 100 200 300 400 500 l l l l RCS COLD LEG TEWERATURE (oF) l I l l I I I I I I I l
I l l l 1 I I I I I I I Figure 7. EXAMPLE GRAPH WRITING E ERGENCY OPERATING PROCEDURES 28 0428A-GNS
MY WRITERS GUIDE - REVISION 1 Tables Tables may be used within the text- of a procedure to clearly present a large number of separate options. A table will immediately follow the step or substep which makes use of it. Therefore, it does not require a unique number and title. Any table will be completely enclosed by a distinct outline; if necessary, it may extend into the adjacent column because of this delineation.
All information presented in a table shall be at least as legible (type size and spacing) as the instruction steps in the procedure.
All columns and rows of information in a table will be defined by solid lines.
All column and row headings shall be presented in ALL CAPITAL LETTERS.
If the same table is needed in several steps, it may be more efficient ,
to place that table in an ATTACHENT.
Absence of a table element will be indicated by the letters NA.
~
' Figure 8 presents two separate types of example tables.
O WRITING E ERGENCY OPERATING PROCEDURES 29 0428A-GNS
MY WRITERS GUIDE - REVISION 1 I I i' l l l l Procedure
Title:
l Proc. No. 2-70-3 ll ll E-3 l Rev. No. 0 ll l l STEAM GEERATOR TUBE RlPTURE l Page 7 of 20 ll II I ll
^
I I l- l l l 1l RWTWED S1 PRESSWE (PSIG) l CORE EXIT TEWERATW E ('F) ll l1 ll ll l ll l l l [ HARSH CONTAIf4ENT OF] ll l l 1200 l 519'F [ 530 ] ll l l l 1100 1 508'F [ 519 ] ll
- ll 1000 l 496*F [ 508 ]- ll ll 900 l 4840F [ 496 ) lI ll 800 l 470af [ 484 ] ll ,
ll 700 l 456*F [ 470 ] ll ll 600 l 439'F [4%] I1 l1 I II
- I l
'O
- l ll l RLPTURED SG LEVEL Il l
l ll ll l ll Pzr. LEVEL INCREASING l DECREASING l OFFSCALE ll 1
-l l l l l l l 1
1I l- o Increase RCS l Increase RCS lo Increase RCS l l l l l Makeup Flow l Makeup Flow I Makeup Flaw ll l l LESS THAN 6% l l l ll l l [18% FOR HARSH l o Depressurize I la Maintain RCS l l l l CONTAINENT] I RCS Using i I And Ruptured l l 2
ll l Step 6b l l SG pressures l l l
ll l l l Equal ll II I I II l l E TWEEN 64 l Depressurize l Turn On Maintain RCS I l
! I I [18% FOR HARSH l RCS Using l Pzr. l And Ruptured l l
) l l CONTAINENT] l Step 6b i Heaters l SG Pressures l l I
l l Af0 50% l l l Equal ll l ll l l l ll
, I l Figure 8. EXA WLE TABLES
- Attachments Supplementary information or detailed instructions which would unnecessarily
( e complicate the flow of a procedure may be placed in an attachment to that
! procedure. An attachment is referenced by the words: Refer to ATTACHPENT N
{ (where N is the attachment letter designator).
l I WRITING EERGENCY OPERATING PROCEDURES 30 0428A-GNS l
_ _ _ . . - _ - _ . _ , _ __ _ _.-.,_ _-_.- ___.-_~-_____._ - __ ._-_
MY WRITERS GUIDE - REVISION 1 O Attachments are identified by the title "ATTACHENT" followed by a single letter designator. This title is centered at the top of a standard format page three blank lines below the procedure identification block. Attachments will use a single-column, full-page-width format.
Any table included in an attachment will be identified by a title centered on the page and located one blank line below the ATTACHENT designator.
Any figure included in an attachment will use the same format as a figure included in the procedure (sequential figure number and title located below the figure). The ATTACHENT heading will be located above the figure on the page.
Physically, Attachments will be located after any Figures belonging to the procedure. Attachment pages are numbered in sequence with normal procedure pages. Figure 9 is an example ATTACHENT page.
I I Il Procedure
Title:
l Proc. No. 2-71-1.1 ll i ll ES-1.1 l Rev. No. 0 l l t II SI TERMINATION l Page 11 of 11 II II I II I I I I
+
1 1 I ATTACHENT A l l l 1 l l The following conditions support or indicate natural circulation l I flow: l I I I o RCS subcooling based on core exit Temp. - GREATER THAN 280F I I I l o SG pressures - CONSTANT OR DECREASING l l l l 0 WR T-HOTS - CONSTANT OR DECREASING I I I I o Core exit Temp. - CONSTANT OR DECREASING l l l l l l l l 1 1 I I I I I O ' '
Figure 9. EXATLE ATTACHENT PAT FORMAT WRITING EERGENCY OPERATING PROCEDURES 31 0428A-GNS 4
- . ~ - ._ , . ,- - - - - - _ -- --,-r--_.m .,e.,_ y. - _.- ,-- . -,.e .m, - . _ - - . - . . , - , . , ,. ---- y .----r --
MY WRITERS GUIDE - REVISION 1 O Foldout Page
, - A foldout page will be supplied for each emergency procedure series. The sheet will be located at the end of each procedure series and will be treated as the last page of the last non-ECA procedure in its related series. The title for the foldout page will be put in the procedure title section of the
- - page header block in place of the normal procedure title and will be in the 1 form of " FOLD 00T FOR XXX SERIES PROCEDURES".
Foldout pages will use a single-column, full-page-width, format (see Figure 10).
Each set of operator information on the foldout page will be numbered sequentially and have an explanatory title. The title will be capitalized and underlined for emphasis. This page contains those important actions which can be performed at any step in the procedure of which it is a part.
O 4
b WRITING E!ERGENCY OPERATING PROCEDURES 32 0428A-GNS
MY WRITERS GUIDE - REVISION 1 D
(V .
l 1 l l Procedure
Title:
l Proc. No. 2-70-1 l l 1 l l Rev. No. 0 l l l l FOLDOUT FOR E-1 SERIES PROCEDURES l Page 20 of 20 l l 1 l l I l l 1 l 1. SI REINITIATION CRITERIA l l l l Initiate SIAS if EITER of the following conditions listed l l below occur: l l l 1 o RCS subcooling based on core exit Temp. - LESS THAN 280F l l l l o Pzr. level - CAtNOT BE MINTAINED GREATER THAN 6% [24% l l FOR HARSH CONTAINE NT] l l l l 2. SECO@ARY INTEGRITY CRITERIA l l l '
l Go to E-2, STEAM LIE BREAK, Step 1, if any SG pressure is l I decreasing in an uncontrolled manner or has completely l 1 depressurized, and has not been isolated. l l l Q
V l
l
- 3. E-3 TRANSITION CRITERIA l l
l Go to E-3, STEAM GENERATOR TUBE RUPTURE, Step 1, if any SG level l I increases in an uncontrolled manner or any SG has abnormal l l radiation. l l l 1 4. RECIRCULATION COOLING (RAS) SWITCHOVER CRITERIA l l l l Go to ES-1.3, TRANSFER TO RECIRCULATION COOLING (RAS), Step 1 if I l RWST level decreases to less than 100,000 gallons. l l l l S. EFW SUPPLY SWITCHOVER CRITERIA l l .I I DWST level - LESS THAN 40,000 GALLONS l l l l l l l l l Figure 10. EXA WLE FOLDOUT PAGE FORMAT
\
v WRITING EERGENCY OPERATING PROCEDURES 33 0428A-GNS
MY WRITERS GUIDE - REVISION 1 O STATUS TREE FORmT Critical Safety Function Status Trees will be presented in the " block" version (see Figure 11). All trees in the set must use the same format and will be oriented vertically on the page.
Color-coding and/or line-pattern-coding shall be used from each last branch point to its terminus (see Figure 12).
All text on 'the Status Trees shall be at least as legible (type size and spacing) as the instruction steps in the-procedures.
Each Status Tree shall have at the top of the page, a designator block identical to that used in the standard procedure format, and containing the same information. Each tree shall also be labelled with an individual title, as indicated in TAELE 3 (LISTING OF EERGENCY PROCEDLRE CODES AND TITLES).
Statements shall be worded so that the favorable response is the downward (usually the "yes") exit from the decision block.. Termini shall be ordered so that REDS are uniformly at the top and GREENS at the bottom. Termini order should be RED - ORANGE - YELLOW - GREEN if possible.
O O
STATUS TREE FORMAT 34 04284-GNS
MY WRITERS CUIDE - REVISION 1 O
I I I I ProcedJre
Title:
l Proc. No. 2-80-1 1 I I I F-0.1 1 Rev. No. O I I i 1 SUBCRITICALITY STATUS TREE l Page 1 of 1 I I I I I I I I l l l -
1 l s . . GO TD g
- FR-S.1
' , . . CD TO FR-S.1 1 ND l
VR POVER
' l "
- l LESS THAN -STARTUP NO I 2% !
RATE I IEI ~
ZERO CR l l NEGATIVE ygg l ,
I I I I I '
GO TD l
[ FR-S.2 l
'l VR POVER I 4
l CHANNELS l
l ANG: TARMP NO I YES l RATE MORE l NEGATIVE g l THAN '
i 1 -0.2 DPM YES I 1 I
.I I l CSr I
' ' SATISFIEll
, l l 4
GO TD I '
FR-S.2 !
I I I I
! NC 1 STARTUP 1 1 RATE I l ZERO CR I NEGATIVE ygg l l
- I I i i i I I cSr I l SATISFIES l I I i I i i I i Figure 11. BLOCK FOR*T FOR STATUS TREES STATUS TREE FORmi 35 0428A-GNS
MY WRITERS GUIDE - REVISION 1 1 I I I I gTATUS TREE PRIORITY CODING I I- l 1 I I l COLOR CODING STATUS / RESPONSE I I I I A. I I RED The Critical Safety Function is under l l w extreme challenge; IMEDIATE operator action is required.
i I I I i 1 l _A. l 4
I ORANGE The Critical Safety Function is under l I
I h severe challenge; PROPPT operator action is required.
i I
I I I I .
l YELLOW The Critical Safety Function condition i I
I k is OFF-NORMAL; operator action MAY be taken.
I I
I I I GREEN The Critical Safety Function is l I satisfied. N0 operator action is I I needed. l I I I I I I I I I I I I Figure 12. STATUS TREE PRIORITY IDENTIFICATION 1
9 L)
STATUS TREE FORMAT 36 0428A-GNS
MY WRITERS GUIDE - REVISION 1 h,m ECHANICS OF STYLE Spelling All spelling should be consistent with modern usage. If more than one spelling is given in standard dictionaries, the first form listed, indicating the most common usage, will be used.
Punctuation Punctuation should be used only as necessary to aid reading and prevent misunderstanding. Word order should be selected to require a minimum of punctuation. The following rules apply:
o Use a COLON to indicate that an item or a list of items is to follow.
Example: Stop the following equipment:
o Use a COMMA after conditional phrases for ease of reading. .
(N Example: E level exceeds 50%, TEN ...
'%)
o Use PARENTHESES to indicate added information, o Use a PERIOD to indicate the end of complete sentences in right-hand column actions statements and in NOTES and CAUTIONS and for indicating the decimal place in numbers.
No periods are to be used in left-hand column action steps.
o Use a DASH to separate a required action and its expected response.
Example: Ensure SI Pumps - RUNNING o Use square ERACKETS to designate [ HARSH CONTAINENT SETPOINTS]
Capitalization Capitalization 'shall be used in the procedures for emphasis in the following cases:
o Logic terms will be capitalized and underlined, o Expected responses (left-hand column of instructions) are capitalized.
ECHANICS OF STYLE 37 0428A-GNS
MY WRITERS GUIDE - REVISION 1
,a V
o Titles of procedures will be completely capitalized whenever referenced within any procedure, o Operator action steps may be capitalized for emphasis o Acrenyms (see Table 5) are capitalized. Abbreviations (see Table 5) may be capitalized when they are labels for controls or displays the operator is to use when performing a procedure step.
o Section headings on foldout pages are capitalized and underlined.
o The phrase "[FOR HARSH CONTAI?NENT]" *ere ever used.
Vocabulary
~
- Words used in the proceidures should convey precise meaning to the operator.
Simple words having few syllables are preferred. These are typical of words in common usage.
Verbs with specific meaning should be used.
The verb should exactly define V]
the task expected to be performed by the operator. A list of frecuently used verbs is included as Appendix A (located at the end of this document).
. Some wcrds have unique meanings as listed below:
manual (manually) - an action performed by the operator in the control room. (The word is used in contrast to an automatic action, which takes place without operator intervention.)
local (locally) - an action performed by an operator outside the control room.
Example: " Locally close valve" means directly turning a handwhsel to close a valve.
Certain other words are to -be avoided simply because they are not adequately defined when used without modification.
~
These include:
approximately, rapidly and slowly. The same words become acceptable when some clarification is provided; clarification normally is part of a lower-level substep.
Example: Rapidly (up to 200 F/HR) cool down the RCS.
Inequalities are to be expressed in words rather than symbols: 1.e.,
"gieater than, less than". These words are always appropriate for comparing pressures, temperatures, levels and flowrates. The words "above" and "below" should not be used in this context.
ECHANICS OF STYLE 38 0428A-CNS
MY WRITERS GUIDE - REVISION 1 p
V Numerical Values All numerical values presented in the procedures should be consistent with s at can be read on instruments in the control room (i.e., consistent with instrument scale and range).
The number of significant digits presented should be equal to the required reading precision of the operator.
Acceptance values should be stated in such a way that any addition and subtraction operations are avoided, if possible. This is done by stating acceptance values as limits.
Examples: 2500 psig maximum 350'F minimum between 450*F and 500*F Tolerances can be expressed by stating the normal value followed by the acceptable range in parenthesis. .
Example: 5500F (5400F to 5600F) -
Correct 550*F +/- 100F -
Incorrect
~N, Engineering units should always be specified when presenting numerical (V values for process parameters. They should be the same as those used on the control room displays.
Abbreviations and Acronyms Abbreviations and acronyms should be limited to those commonly used by operators. Table 5 lists those approved for use at Maine Yankee. These abbreviations and acronyms are consistent with the Maine Yankee Standards for Control Panel Labeling.
Abbreviations and acronyms should be used whenever possible to simplify the procedures. However, care must be taken to ensure that readability is not impaired.
{
Acronyms from Table 5 will be uniformly capitalized whenever they are used to refer to Maine Yankee plant design features. The first letter of each ;
abbreviation will be capitalized and the abbreviation will be followed by a period.
Underlining The use of underlining is restricted to certain specified ' applications to maintain its effectiveness as a highlighting device:
O ECHANICS OF STYLE 39 l
l 0428A-GNS l l
MY WRITERS GUIDE - REVISION 1
- O o The column headings on operator action pages are underlined '
o The headings CAUTION and NOTE are underlined, o Logic terms are underlined whenever they appear. ,
- o Section headings on the foldout pages are underlined.
Steam generator status descriptions are underlined in o
procedure steps for clarity. (i.e. Faulted SG, Intact SG) >
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O ECHANICS OF STYLE 40 0428A-GNS
MY WRITERS GUIDE - RCVISION 1 TABLE 5 MIE YAM (EE ACRONYMS & A8REVIATIONS
~ Abnormal Operating Procedure AOP
~
Addition Add.
~
Adjust Adj.
Aerated drain tank Aer. Drn. Tk.
- Air circuit breaker Air. Ckt. Bk.
Air conditioning Air. Cond.
~
Air operated valve A0V Air particulate detector
~ Air. Partic. Det.
~
Alternate Alt.
~ Alternate control room operator ACR0 Alternate shutdown panel Alt. Shtdn. Pnl.
Alternate shutdown system Alt. Shtdn. Sys.
Alternating current AC And &
Anticipated Transient Without Scram ATWS
~
Automatic Auto.
- Automatic bus transfer Auto, bus Trnsfr.
Auxiliary Aux.
~ Auxiliary condensate system Aux. condensate Sys.
(~~ s Auxiliary feed water AFW s Auxiliary feed (water) pump
[ AFP Auxiliary operator ,A0
[ Auxiliary steam Aux. Stm.
Backup Bckp.
~
Bearing Brg.
~
Blowdown BD Boric acid Bor, acid Boric acid mix tank BAMT Boric acid storage tank BAST Baron concentration Bor. Conc.
Baron waste storage tank BWST Breaker Bkr.
Building flow diagram FB
~
Calculation . Calc.
Cathode ray tube CRT Cavity purification Cav. Purif.
~
Central alarm system CAS
Channel Chan.
Charging Chrg.
~
Chemical Chem.
~
Chemical addition Chem. Add.
Chemical volume control system CVCS O
ECHANICS OF STYLE 41 0428A-GNS
MY WRITERS GUIDE - REVISION 1 TABLE 5 (Continued)
MIE YAPEEE ACRONYMS & ABREVIATIONS Circuit breaker Ckt. Bkr.
Circulating water Cire. Wtr.
Circulating water pump Circ. Wtr. Pmp Circulation Circ.
Circulation water Cire. Wtr.
Coefficient Coef.
Concentration Conc.
Condensate Condensate Condensate water storage tank CWST Condenser Condsr.
Containment Ctmt.
Containment control air Ctmt. Cntrl. Air Containment isolation signal CIS
. Containment spray Ctmt. Spry.
Containment spray actuation signal CSAS Containment spray pump Ctmt. Spry. Pmp.
Containment ventilation and purge Ctmt. Vnt. & Prg.
Control Cntrl.
Control element assembly ~
CEA Control element drive CED O- Control element drive mechanism 4~ CEDM Control t ment drive system CEDS Control room operator CR0 Controller Cntrlr.
Coolant Coolnt.
Cooling Coolng.
Counts per second CPS Critical Safety Function CSF Decades per minute OPM Demineralized water storage tank DWST Demineralizer Demin.
Department Dept.
Departure from nucleate boiling DtB Departure from nucleate boiling ratio DNBR Detector Det.
Diesel generator DG Differential pressure d/p Differential temperature d/t Diffuser vacuum priming Dif. Vac. Priming Direct current DC Dischage Disch.
Disconnect Discon.
Display Dsply.
Disposal Disp.
Domestic water Domestic Wtr.-
O oreia ore-ECHANICS OF STYLE 42 0428A-GNS
MY WRITERS GUIDE - REVISION 1 TAELE 5 (Continued)
MIE YA*EE ACRONYMS & ABREVIATIONS Driven Orvn.
- Duty call officer DC0 Electrical schematic ESK Electrahydraulic (turbine) control system EHC Emergency Emerg.
Emergency Contingency Action (Procedure) ECA Emergency core cooling system ECCS Emergency feed water EFW Emergency feed (mter) pump EFP Emergency on-site facility EOF Emergency Operating Procedure E0P Emergency shutdown from power Emerg. Shtdn. from Pwr, Emergency Supplemental Procedure ES Engineered safety feature ESF Excess flow check valve EFCV -
Exchanger Exch.
Extraction Xtractn.
Extraction steam Xtractn. Stm. .
Feed water FW Filter Fltr. *
(*]' Fire suppression Fire Suprsn.
Flow Flo.
Fuel oil F0 Fuel pool' purification Fuel Pool Purif.
Fuel temperature coefficient Fuel Temp. Coef.
Functional Recovery (Procedure) FR Generator Gen.
Gland Glnd.
Gland seal G1nd. Sl.
Governor Gov.
Ground Gnd.
Header Hdr.
Health physics Hlth. Phys.
Heat Heat Heat exchanger Ht. Exch.
Heater Htr.
Heater drain pump HDP Heating Htg.
- Heating and ventilation Htg. & Vntln.
High Hl.*
High pressure HP*
High pressure and reheater drain Hl. Press. & Rehtr. Drn.
High pressure drain hPD ,
High pressure safety injection WSI '
N High pressure setpoint relief FPSR ;
Hydrogen H2 l Inadequate core cooling ICC Inches of mercury Inches Hg.
ECHANICS OF STYLE 43 0428A-GNS l
L . -
MY WRITERS GUIDE ~- REVISION 1 O TABLE 5 (Continued)
MITE YAPKEE ACRONYMS & ABREVIATIONS Indicator Ind.
Instrument Instr.
Instrument air Instr. Air Inverse boron worth IBW Isolation Isol.
Kilo-volt KV Kilo-volt-ampere KVA Leakage monitoring Lkg. Monit.
Letdown LD Level Lvl.
Level control valve LCV Logarithmic Log.
Loss of coolant accident LOCA Low pressure LP*
Low pressure safety injection LPSI ,
Low temperature over pressurization LTOP J
Main control board MCB Main feed (water) pump WP I Main feed water regulator valve MFRV Main steam
~
MS
(,_x) Manager of Operations M00 Margin Marg.
Mechanical flow diagram- FM Megawatt electric MWe Megawatt thermal MWth Minimum pressurization temperature MPT Miscellaneous Misc.
Moisture .
Moisture Moisture separator reheater Moisture Sep. Rehtr.
Monitor, monitoring Monit.
Motor Mtr.
- Motor control center Mtr. Cntrl. Cntr.
Motor driven feed water pump MDFP Motor generator Mtr. Gen.
Motor operated valve MOV Multi-point trend- from computer WX Narrow range NR
- Net positive suction head N3SH Nitrogen N2 Non-Nuclear Safety NNS Non-return valve NRV Nuclear Regulatory Commission NRC Nuclear safety engineer NSE Offset Offset s Operated Oper.
Operating Procedure OP Panel Pnl.
ECHANICS OF STYLE 44 0428A-GNS
MY WRITERS GUIDE - REVISION 1 I
TABLE 5 (Continued)
MITE YAtKEE ACRONYMS & A8REVIATIONS Particulate Partic.
Plant shift superintendent PSS Plant Operations Review Committee PORC Points Points Post accident purge PAP Pounds per square inch absolute PSIA Pounds per square inch differential PSID
. Pounds per square inch gage PSIG Power Pwr.
- Power dependent insertion limit PDIL Power operated relief valve PORV Pressure Press.'
Pressure and pressure relief Press. & Press. Rlf.
Pressure control valve PCV Pressure recorder Press. Rec. .
Pressurizer Pzr.
Primary Prim.
Primary auxiliary building PAB Primary auxiliary building emergency panel PAB Emerg. Pnl.
,Q Primary component cooling PCC V Primary drain tank FOT Primary . inventory trending system PITS Primary sampling Prim. Samp.
Primary vent and. drain PD Primary vent stack Prim. Vnt. Stk.
Primary water PW Primary water storage tank PWST Priming Primng.
Procedure Proc.
Procedure Index Number PIN Proportional Prop.
- Proportional heater Prop. Htr.
Pump Pmp.
- Purge Prg.
- Purification Purif.
Quench Qnch.
Quench tank Qnch. Tk.
Radiation Rad.
Radiation monitoring panel RMS Raw water RW RCS average temperature T-AVE RCS cold leg temperature T-COLD RCS hot leg temperature T-HOT Reactor Rx Reactor coolant RC 0 Reactor coolant pump RCP MECHANICS OF STYLE 45 0428A-GNS
l
~ MY WRITERS GUIDE - REVISION 1 I TABLE 5 (Continued)
MIE YA*EE ACRONYMS'& ABREVIATIONS Reactor coolant pump seal water RCP Sl. Wtr.
Reactor coolant purification .RC Purif.
Reactor coolant system RCS Reactor operator R0 Reactor protective system R'S Reactor regulating system RRS Reactor trip Scram Reactor vessel Rx Vsl.
Receiver Revr.
j Recirculation Recirc.
Recirculation actuation signal RAS r Recorder Rec.
- Reference Ref.
- Refueling Rfing.
Refueling water storage tank RWST ,
{. Regenerative Regen.
Regulation Reg.
Reheater Rehtr.
' Relay Rly.
! -Relief Rif. *
. Removal Removal t
Residual heat removal RHR i l Resin Resn. ,
Resistance Resist.
Resistance temperature detector RTD Return Rtrn.
Safety injection tank SIT Safety parameter display system SPOS Sample, sampling Samp.
- Saturation Sat.
- Screen Scrn.
Seal Seal i Secondary Sec.
- Secondary component cooling SCC Secondary seal and leakage Sec. Sl. & Lkg.
Section Sect. '
Selector- Sel.
Senior reactor operator SR0 i
Separator Sep.
- Service Serv.
. Service water SW Service water pump SW Pmp.
4 ECHANICS OF STYLE 46 4
0428A-GNS
MY WRITERS GUIDE - REVISION 1 TABLE 5 (Continued)
MIE YAM (EE ACRONYMS & A8REVIATIONS Shift operations supervisor SOS Shutdown Shtdn.
Solenoid operated valve SOV Solidification Solidif.
Spray Spry.
Spray chemical addition tank SCAT Stack Stack.
Start-up rate Start-up rate Steam Stm.
Steam driven Stm. Drvn.
Steam generator SG Steam generator blowdown demineralizer ~SG BO'Demin.
Steam generator emergency panel SCEP Storage Star.
Suction Suct.
Supply Sply.
- Suppression Suprsn.
/l kJ
. Switch Swch.
Symmetric offset Sym. Offst.
Symmetric offset trip calculator SOTC System Sys.
Tank Tk.
Technical data book TDB Technical support center TSC Temperature Temp.
Temperature control valve TCV Thermal Thrm.
- Thermal margin / low pressure TM/LP Thermocouple T-couple Throttleable Throt.
. Transfer Trnsfr.
- Transformer Trsns.
Travelling water screen Travelling Wtr. Scrn.
Treatment Trtmnt.
- Trip Ckt. Bkr.
Trip circuit breaker Turbine Turb.
Turbine driven feed (water) pump Turb. Drvn. Feed Pmp.
Undervoltage ' UV Vacuum Vac. j Valve Vlv.
Variable pressure setpoint relief VPSR Vent Vent
. Ventilation Vntitn.
ECHANICS OF STYLE 47 0428A-GNS
MY. WRITERS GUIDE - REVISION 1 TABLE 5 (Continued)
MIE. YAM (EE ACRONYMS & AMEVIATIONS ;
- Volume Vol. i Volume control tank VCT l I
Waste Waste i Waste disposal Wst. Disp. !
Waste gas disposal Wst. Gas Disp.
Waste solidification Wst. Solidif.
Water Wtr.*
Well water Well.Wtr.
Wide range WR *
- To be used only~ in-r.ontext with other. words or abbreviations.
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ECHANICS OF STYLE 48 l
-0428A-GNS
MY WRITERS GUIDE - REVISION 1 r
~ PRINTED FORMAT General' Printing Instructions
- For all emergency operating procedures, the following general instructions apply:
o Paper size should be 8-1/2 by 11 inches for all pages except the foldout page, o White bond paper should be used.
o Type size and style are to be consistent across all emergency procedures. A 12 point block style font with'6 lines per inch will be used to assure unifor:nity and legibility. (WANG Tile 10/12 12-pitch and. IBM Selectric 12-pitch Letter Gothic are examples ' of acceptable type fonts for printing) o Figure labeling may be done on standard electric typewriting equipment, using 12 pitch Letter-Gothic type, provided the requirements of this c guide are followed. .
- b. o All text material should be maintained on current word-processing.
equipment.
Page Layout The following instructions apply to the placement of text on the procedure pages:
o The-boxr i hes iing will begin one-half inch from the top of the page, and will . ^ wain one inch margins at both sides.
o All text will maintain 1-inch margins at both sides of the page.
o The bottom page margin will always be one inch, o Two blank lines will follow the header box prior to any text.
o Two blank lines will separate operator action steps.
o A single blank line will separate substeps within any action step.
. o Adjacent lines of text will be separated by normal spacing for the type fonts.
PRINTED FORMAT 49 0428A-GNS
I a
E i -
s O !
I MAINE YANKEE'
- EERGENCY OPERATING PROCEDURES !
i f
WRITERS GUIDE i O :
s 6
I 4
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,,-wer-, -. - . , .-.
T MY WRITERS GUIDE - REVISION 1 p
CONTENTS PURPOSE AND SCOPE ........................... 1-E0P DESIGNATION AND NUMBERING ..................... 2 Procedure Identification ...................... 2 Revision Numbering ......................... 8 Page Numbering and Identification .................. 8 FORMAT ................................. 9 Procedure' Organization ....................... 9 Page Formats ............................ 9 Instructional Step Numbering .................... 12 Immediate Action Steps ....................... 12 Continuous Steps ........................... 13 WRITING EERGENCY OPERATING PROCEDlRES ................ 14 Cover Sheet ............................. 14 Operator Actions .......................... 16 Instruction Steps, Left-Hand Column ................. .20 Instruction Steps, Right-Hand Coluan ................ 21 Use of Logic Terms . . . . . . . . . . . . . . . . . . . . . . . . . . 22
(s Notes and Cautions ......................... 23
\s,) Transitions to Other Procedures or Steps .............. 24 Component Identification ...................... 25 Level of Detail ........................... 25 Figures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... 26 Tables ...............-................ 29 Attachments . . . . . . . . . . . . . . . . . . . . . . . . . . . . .- 30 Foldout Page ............................ 31 STATUS TREE FORMAT .......................... 34 MECHANICS OF STYLE .......................... 37 Spelling .............................. 37 Punctuation .............................. 37 Capitalization ........................... 37 Vocabulary ............................. 38 Numerical Values .......................... 39 Abbreviations and Acronyms .....................
39 Underlining ............................. 39 PRINTED FORMAT ............................ 49 General Printing Instructions . . . . . . . . . . . . . . . . . . . . . 49 Page Layout ............................. 49 REPRODUCTION ............................. 51 b
v REFERENCES .............................. 52 Appendix A. ACTION VERBS ....................... 53 Appendix B. COMMON TERMS '....................... 59 ;
Contents i i
MV WRITERS GUIDE - REVISION 1 O o Every effort should be made to avoid splitting action steps across pages. If a lengthy step requires a full page, 'then it may be necessary to leave much of a preceding page blank. To prevent any confusion, the words " PROCEED TO EXT PAGE" will be centered in the
-white space if margin at the botton of the page is greater than 3 inches, o If an action step (including its preceding NOTES and/or CAUTIONS) does not fit onto a single page, then the words " STEP xx CONTINUED ON EXT PAGE" should be centered on the page, two blank lines below the last line of text.
o Text following any underlined header (column headings on action step pages or section headings on the foldout pages) will be separated from the underline by two blank lines.
O Page rotation is to be avoided for emergency operating procedures.
O O
PRINTED FORMAT 50 0428A-GNS
t MY WRITERS GUIDE - REVISION 1 REPRODUCTION Procedure reproduction will be done from original type written copies only on a standard copier set for full size reproduction. All pages shall be copied i single-sided only.
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1-O l REPRODUCTION 51 !
i l 0428A-GNS l
MY WRITERS GUIDE - REVISION 1 b,
REFERENCES
- 1. NUREG-0899, Guidelines for the Preparatio.1 of Emergency Doerating '
-Procedures, August, 1982.
- 2. IP O 82-017, Emergency Operating Procedures Writing Guideline,~ July, 1982.
- 3. Maine Yankee Procedure 0-06-1, Procedure Preparation, Classification, and Format.
- 4. W4-83-84, Emergency Operating Procedures Generation Package (Letter from Garrity, Maine Yankee, to Clark, tEC) May 17, 1983.
- 5. REQlEST FOR ADDITIONAL IWORMATION ON PROCEDURES TfERATION PACKAT, (Letter from Miller, tac, to Randazza, Maine Yankee), June,1984.
O 1 REFERENCES 52 0428A-GNS
REVISION 1
.(3 V ACTION VERBS APPEM)IX A. ACTION VERBS adjust to manually operate equipment as necessary to satisfy procedure requirements on process parameters: pressure, temperature, level, flow, etc.
Example: Adjust Pzr. level align to arrange components into a desired configuration Examples: Align the system for normal charging, Align valves as appropriate
(] allow to permit a stated condition to be t/ - achieved prior to proceeding Example: Allow discharge pressure to stabilize block to inhibit an automatic actuation Example: Block SIAS check to note a condition and compare with some procedure requirement i
Example: Check Pzr. level -
GREATER THAN 20%
close to change the physical position of a mechanical device. Closing 1
a valve prevents fluid flow. l Closing a breaker allows electrical current flow.
complete to accomplish specified procedure requirements O
Appendix A. . ACTION VERBS 53 0428A-GNS
REVISION 1
(^,
V ACTION VERBS continue to~go on with a.particular process Example: Continue with this procedure determine to evaluate using formulae or graphs Example: Determine maximum venting time dump to release steam by use of a means which is manually or automatically controlled Example: Dump steam to condenser energize To provide electrical power to Example: Energize Pzr. heaters. ,
ensure to determine if an expected condition exists and if not take action to cause it to exist Examples: Ensure Reactor Trip, Ensure HPSI Pumps - RUNNING enter to insert into or add to; commonly used in reference to plant-specific additions Example: Enter-plant specific means
. equalize to make the value of a given parameter equal-to the value of another parameter Example: Equalize charging and letdown flow establish to make arrangements for a stated condition Example: Establish normal Pzr. pressure and level control Appendix A. ACTION VERBS 54 0428A-GNS
REVISION 1
(,-.s s ACTION VER8S evaluate to examine and decide; commonly used in reference to plant conditions and operations Example: Evaluate plant conditions implement to commence the performance of a procedure or adrninistrative guideline-Example: Implement Site Emergency Plan initiate to put into action or motion; commonly used to refer to automated, multi-faceted operations ,
Examples: Initiate SI, Initiate CIS load to connect an electrical component or unit to a source of bq ~ electrical energy, may involve a
" start" in certain cases Example: Load H3SI pump on ac emergency bus
, maintain' to control a given plant parameter to some procedure requirement continuously Example: Maintain SG 1evel in the narrow range minimize to ma'<e as small as possible Example: Minimize secondary system contamination monitor similar to " check", except implies a continuous activ.ity open to change the physical position of a mechanical device to the unobstructed position. Opening a valve permits fluid flow.
Opening an electrical breaker prevents current flow.
Appendix A. ACTION VER8S 55 0428A-GNS
_ _ _ _ _ . _ - . _ - _ _ _ _ ~ - . _
REVISION 1 b,m. ACTION VERBS place to move a control to a stated position Example: Place controls in MAMJAL purge To flush an undesirable gas or fluid from a system.
Example: Purge the Main Generator with Carbon Dioxide record to document specified characteristics Example: Record T-AVE reduce to lessen in extent or amount Examples: Reduce charging flow to 65 cym, .
Reduce to one WP reset to remove an active output signal 4 .from a retentive logic device even with the input sicpal still Os . present; commonly used in reference to protection and/or safeguards logics in which the actuating signal is " locked-in".
The reset allows equipment energized by the initial signal
,to be deenergized Example: Reset SI actuation signal
. restore to cause a condition which 4
existed in the past to be regained Example: Restore Pzr. water level to not less than 28%
sample to take a representative portion for the purpose of examination; commonly used to refer to chemical or radiological examination Examples: Sample RCS boron concentration, Sample secondary side O radioactivity Appendix A. ACTION VERBS 56 0428A-GNS
REVISION 1 d,- ACTION VEPBS set to physically adjust an adjustable feature to some specified value:
Example: Set diesel speed to 525 rpm.
start to originate motion of an electrical or mechanical device, either directly or by remote control Example: Start one RCP stop to terminate motion of an electrical or mechanical device Example: Stop both diesel-generators 9
^
throttle to operate a valve in an intermediate position to obtain a certain flow rate
-p V Example: Throttle charging flow control valve to establish ~ desired flow trip to manually actuate a semiautomatic feature. Commonly,
" trip" is used to refer to component deactuation.
Examples: Trip reactor, trip turbine, trip Mtr. Gen. set try to make a continued effort when success may not be immediately obtainable Example: Try to restore station service power turn off to remove electrical energy from (something) l l
Examples: Turn off Pzr. heaters, 1 Turn off LPSI pumps O
Appendix A. ACTION VERBS 57 0428A-GNS
i REVISION 1 O. ACTION VERBS turn on to supply electrical energy to a non-mechanical component Example: Turn on Pzr. heaters use to employ as necessary to achieve
.the stated objective Example: Use Pzr. heaters to i maintain pressure ,
vent .to " . - liquid i u . .ao unow ssure to escape j at a vent
]
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i Appendix A. ACTION VERBS 58 i
0428A-GNS
REVISION 1 COM TERMS APPEtOIX B. Co m TERMS affected Refers to a SG which is either faulted or ruptured auto Refers to a switch position or component condition in which that component is available for starting by a control or protection system input (without manual action) constant Refers to maintaining a parameter at or about a defined value, either -
manually or automatically faulted Refers to SGs whose secondary side .
pressure boundaries are not intact, in stan&y Refers to a piece of equipment in p an inactive status but ready for L/ start on demand. (see auto) intact Refers to pressure boundaries that are functioning as designed.(i.e., no leaks or breaks) local (locally) An action performed by an operator outside the control room.
Example: " Locally close valve" means directly turning a handwheel to close a valve.
manual (manually) - An action performed by the operator in the control room.
(The word is used in contrast to an automatic action, which takes place without operator intervention.)
may When used to denote permission, tne action is neither a requirement nor a recommendation.
O Appendix 8. COMMON TERMS 59 0428A-GNS
d REVISION 1 0
I i COMON TERMS ruptured Refers to SGs whose primary-to-secondary pressure boundaries (i.e., SG tube bundle) are not intact. ,
shali - Denotes a REQUIREENT, deviation i is not allowed.
should Denotes a recommendation.
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! Appendix B. COW ON TERMS 60 0428A-GNS
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MAINE YANKEE EERGENCY OPERATING PROCEDURES VERIFICATION PROGRN1 O
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TABLE OF CONTENTS
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SECTION
- 1.0 Program Overview 2.0 Program Description 2.1 Program Objectives 2.2 Program Evaluation Criteria 2.3 Program Process 2.3.1 Preparation 2.3.2 Assessment 2.3.3 Resolution 2.3.4 Documentation 2.3.5 Results 3.0 Responsibilities 4.0 Verification Team f% 5.0 Verification Process d
6.0 Program Results 7.0 Verification Principles Written Correctness (pgs.15 thru 24) 8.0 Verification Principles Technical Accuracy (pgs. 25 thru 30)
'9.0 Verification Guideline (pgs. 30 thru 43) l ' Attachment #1 E0P Verification Forms Form #1, Part I Comparative Evaluation (pg. 39)
Form #1, Part II E03 Verification Close Out (pg. 39)
Form #2 ' E0P-ERG Differences (pg. 40)
Form #3 Part "A" - Discrepancy (pg. 41)
Part "B" - Resolution (pg. 41)
Form #4 Sequence Step Comparison Sheet (pg. 42)
Attachment #2 Sample Format for E0P's O
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EERGENCY PERATING PROCEDURE
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Verification Program i'
1.0 PROGRAM OVERVIEW Verification is a process developed as a result of the requirements of NUREG 0799 Guidelines for the preparation of Emergency Operating Procedures. These procedures written to Owners Group Guidelines are more complex in nature. The level of detail that these procedures provide ensure that all aspects of accidents are addressed within the framework of the Emergency Operating Procedures.
These procedures address multiple failures and ensure Functional Recovery
/ Procedures are developed to restore Critical Safety Functions. The restoration of the CSF ensures the " Barrier Concept" is maintained, thereby minimizing the potential for the release or radioactive material to the environment.
NUREG-0737, Item I.C.1 and NUREG-0899 Item 3.3.5.1 require that Emergency Operating Procedures be developed to enable licensed operators to interface with the procedures and the machine with the least amount of conflict. Human Factors review of the Control Room Design Review (CRDR) identifias the instruments and mechanical equipment that must be available for the operator to manage the accident. The EP's provide the operator with the direction and information he requires when mitigating an accident.
Information presented to the operator in the procedure set must be consistent and easy to use under stressful conditions.
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The verification process ensures that the procedure set is written correctly as O,
prescribed in the plant specific Writers Guide. Technical accuracy is required and provided thru analysis and updated source documents. The blending of these features provides the bases for workable E0P's. The Verification process encompasses the efforts necessary to ' support a comparative evaluation of ET's to the source documents.
2.0 PROGRAM DESCRIPTION The E0P Verification Program consists of the follcaing areas:
o Program Objectives Evaluation Criteria used to determine that the Program Objectives are satisfied.
The process to be followed.
In this program, the evaluation criteria are applied to determine if the program objectives are satisfied. The evaluation criterit are applied each time the EOP Verification Process is implemenced.
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2.1 Program Objective 3
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l- The objective of the E0P verification program is to determine that
, consistency has been maintained between ECP source documents and the E0Ps.
l This determination can be 'made by evaluating the characteristics of EOP
- written correctness and E0P technical accuracy.
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E0P Written Correctness - addresses proper-incorporation of information from the Maine _ Yankee plant specific Writers Guide for EOPs and other appropriate administrative policies.
E0P Tecnnical Accuracy - address proper incorporation of plant-specific technical information from EOP source documents and plant hardware into the EOPs.
2.2 Program Evaluation Criteria f
Program evaluation criteria for use during the comparative evaluation are developed. This comparative evaluation will determine whether or not the E0P verification program objective has'been satisfied. - These criteria are developed by addressing ECP verification principles that are based on the characteristics of E0P written correctness and EOP technical accuracy.
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2.3 . Program Process The Program Process consists of the following:
Preparation Assessment Resolution Documentation 2.3.1 Preparation Preparation includes identifying the information needed to perform the
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comparative evaluation between the E0Ps and the applicable source documents, determining how to apply the developed evaluation criteria, designating the personnel to perform the comparative evaluation, and scheduling the assessment.
2.3.2 Assessment Assessment includes the performance of the comparative evaluation between designated source documents and the E0Ps using the evaluation criteria, identifying any discrepancies between them.
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t 2.3.3 Resolution Resolution includes the determination of the appropriate solution for any discrepancy, the incorporation of the solutions into the E0Ps, and the approval of the modified EOPs. A discrepancy detected during the comparative evaluation does not always necessitate an EOP change.
2.3.4 Documentation Documentation provides a record of the Program Process.
2.3.5 Program Results
,bove Upon completion of the EOP verification process described ja thorough f review of the EOPs for written correctness and technical accuracy will have been conducted. Discrepancies between the E0Ps and the E0P ' source documents will have been identified evaluated and resolved.
Documentation will exist that records the performance of the comparative evaluation and that confirms all discrepancies were satisfactorily resolved or identified and presented to management for further evaluation.
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3.0 RESPONSIBILITIES o Manager, Operational Support (Corporate)
The Manager, Operational Support will maintain corporate oversite of E0P development.
Resources available to the Manager, Operational Support are contracted MVAPCo., consultants (Yankee Nuclear Services Division), the Site and Corporate Technical ~ Staffs.
o EOP Project Co-ordinator
/ The EOP Project Co-ordinator is responsible for overall coordination of E03's effort.
- E0P Verification Co-ordinator The E0P Verification Co-ordinator is responsible for the successful completion of the verification phase of the EOP effort.
l 4.0 VERIFICATION TEAM RESPONSIBILITIES
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The management of the Verification Program is the responsibility of the Verification Co-ordinator. To ensure the expertise is available with background in the required disciplines the Verification Co-ordinator shall assign specific corporate and' site personnel to the verification team thereby minimizing the impact on.the '0perations personnel at the site.
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Final' resolution of discrepancies will be reviewed and concurred by the ECP n
U Project Co-ordinator o_r_ a panel of three with disciplines in plant. operations,-
hardware famallarity, technical analysis and Human Factors.
5.0 VERIFICATION PROCESS The process of E0P verification consist nf four phases:
c Preparation
- Assessment Resol'ution o Documentation 5.1 Preparation The preparation phase consist of the. following activities; o
Identifying the information required to perform the comparative ,
evaluation between the E0P's and their applicable source documents.
- Application of the evaluation criteria to the EOP's.
Designating personnel to perform' the comparative evaluation.
- Scheduling the assessment
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, 5.1.1 EOP Source Documents Selections V
The bases for the EOPs are found in applicable source documents representing actual plant design and engineering analysis. These source documents may include, but are not restricted to, the following:
Westinghouse emergency operating procedJres guidelines (EPGs) o Setpoints document relating to the derivation of mathematical values that are to be included in the EOPs o
Design Differences Documents identifying the selection of plant-specific equipment to be included in the E0Ps o
Licensing commitments relating to E0Ps such as Technical Specifications, Final Safety Analysis Report (FSAR) statements
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- WY plant-specific writers guide for EOPs a WY plant-specific written bases for EOPs 4 5.1.2 Designated Maine Yankee E0P Source Documents:
o WY Plant-specific writers guide o WY Plant-specific design differences document o
WY Plant-specific setpoint document WY Plant specific analysis documents 0
Westir.ghouse Emergency-Response Guidelines (ERG's) Rev. #1 H.P.
o FSAR MYAPCo.
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o 1984 Refueling Modifications Summary Rev.1 b
V o 1985 Refueling Modifications Summary 5.2 ASSESSENT PHASE 5.2.1 Apolication of Evaluation Criteria to the E0P's The comparative evaluation will determine.whether or not the E0P verification program objectives are satisfied.
Evaluation criteria is needed to Control and Direct the assessment. The
. applicability of the evaluation criteria _is, for WY Plant Specific
'EOP's, developed from the Westinghouse Emergency Response Guidelines
-(ERG's) and the Design Difference Documents. The verification program will apply the evaluation criteria to the individual steps of the E0P's to identify. differences.
In the assessment phase the evaluator shall:
Make a general review of the ECP using the procedure-specific portion of the evaluation criteria.
Indicate on Form #1 (Part I) of the ECP Verification Forms (Attachment 1) that the evaluation was performed, either by checking the acceptable column or by designating the appropriate discrepancy sheet for any discrepancies identified.
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Make a step-by-step review of the newly developed ECP using.the step, caution, note-specific portion of the evaluation criteria -
and-source documents (EPG's).
- _ Indicate for each step on Form #2 of the EOP Verification Forms (Attachment 1) that the comparative. evaluation was performed, l either by checking the acceptable column or by designating the 4
appropriatie discrepancy sheet for any discrepancies identified.
Complete Form #3, Discrepancy Part "A" of the EOP Verification Forms (Attachment 1) and forward the verification forms with the
-discrepancy sheets to appropriate personnel.
5.3 Resolution (Form #3' Part "B")
. After any discrepancies have been identified, potential-resolutions.will be developed. These resolutions are developed by the ECP writer or by the !
personnel conducting the verification assessment.
5.3.1 Resolution of discrepancies will be determined by the verification team and presented to the Verifica' tion Co-ordinator to resolve any conflicts between the writers and evaluator comments. ,
'5.3.2- Forward potential s_olutions to the verification team for review and . l approval.
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Once the potentici resolutions are developed, solutions for each k- 'I discrepancy will be selected, approved, and incorporated. An approved solution could be to accept the discrepancy. The' approval activity could be accomplished in the existing review and approval cycle for procedures.
t 5.3.3 Implement changes to the applicable source document and procedures with approved resolutions as directed by the Manager, Operational Support.
The verification process uses the EOP writer to provide justifying remarks for discrepancies. The Verification Co-ordinator assigns ,
someone to resolve the writers' and evaluators' conflicts and tnen reviews and approves a resolutions. The resolution is then incorporated in applicable EOPs.
, 5.4 Documentation Documentation will record the activities during the program process. Of
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specific historic interest are the 1dentification of discrepancies and the resolution for each discrepancy.
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. Documentation facilitates future verification needs by identifying the source documents used and the resolution of discrepancies and their bases.
If future changes occur in the EP source haants, the appropriate EP can
'be reviewed for impact. Likewise, if a similar discrepancy appears in a future assessment, a previous resolution can be evaluated for applicability or for change. - Without documentation to record the pertinent ' facts, required future verification could destroy or overlook what has been done before.
Close out of the Verification Process will be documented on Attachment 1, Form #1 Part II. This ensures all documentation anc' the verified EP are in one package for filing.
Retention of the E0P Deficiency Documentation will be for the life of the E P 's developed from the Westinghouse ERG set.
6.0 PROGRAM RESLLTS The Emergency Operating Procedures have been verified per the requirements of this program. All deficiencies have been resolved, the EP changes have been made. Documentation for aJditability is completed and available thru plant 1 files.
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SECTION 7.0 g
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VERIFICATION PRINICPLES FOR WIE YAtKEE PLANT SPECIFIC EG"S 7.0 VERIFICATION PRINCIPLE OF WRITTEN CORRECTESS The verification principle of written correctness; is the information in the ECPs consistent with the plant-specific Writers Guide for EOPs.
3 The evaluation criteria to address written correctness is to divide the principle into the following areas:
7.1 Legibility
.7.2 E0P Format Consistency 7.3 - Identification Information
-7.4 Information Presentation 7.5 Procedure Referencing and Branching .
The evaluation criteria that are developed are based on the Maine Yankee plant-specific Writers Guide and applicable administrative policies addressing each area. The evaluation criteria provided under each area are based on the plant-specific Writers Guide (cr ECPs.
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7.1 Legibility Evaluation Goal To determine if the E0P's are legible.
To address the problem of inconsistent printing of E0P documents, the General Printing Instructions are followed:
7.1.1 ASSESS >ENT - are the general printing instructions follo'wecf?
General Printing Instructions
'() For all emergency operating procedures, the following General Printing Instructions apply:
(a) Papef size should be 81/2 by 11 inches for all pages except the foldout page.
(b) White bond paper should be used (except for the fold-out page).
(c) Type size and style are to be consistent across all emergency procedures. A 10 point block style font with 6 lines per inch will be used to assure uniformity and legibility. (WANG Tile 10 and IBM Selectric Letter Gothic are examples of acceptable type fonts for printing).
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(d) Figure labeling may be done on s'tandard electric typewriting equipment, using 12 pitch Letter-Gothic type, provided the requirements of this guide are followed.
(e) All text material should be maintained on current word-processing equipment.
Evaluation Goal o- To address the potential for inconsistency in the page layout format the Page Layout Instructions are followed:
- 7.1.2 ASSESSFENT - are the Page Layout Instructions followccf? To address the potential for inconsistency in the reproduction of E0P's;
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- (a) are the text, tables, graphs, figures and charts clearly readable?
(b) All pages copied single-sided only, except for FR-P.1 Status Tree?
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7.2 E0P Format Consistency tO
-V Evaluation Goal To determine if the format used for all E0P's.is consistent.
To address the problem of inconsistent format in E0P's which create delays in information location:
.7.2.1 ASSESSENT - are the Page Layouts of the E0P set consistent in the format and information presentation?
Page Layout O
The following instructions apply to the placement of text on the proceciJre pages:
-(a) The boxed heading will begin one-half inch from the top of the page, and will maintain one inch margins at both sides.
(b) All text will maintain one-inch margins at both sides of the page.
(c) The bottom page margin will always be one inch.
(d) Two blank lines will follow the header box prior to any text.
(e) Two blank lines will separate each operator action steps.
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(f) A single blank line will separate substeps within any action (m
b step.
(g) Adjacent lines of text will be separated by normal spacing for the type fonts.
(h) Every effort should be made to avoid splitting action steps across pages. If a lengthy step requires a full page, then it may be necessary to leave much of a preceding page blank. To prevent any confusion, the words " PROCEED TO PEXT PAGE" may be centered in the white space.
(i) If any action step (including its preceeding NOTES and/or CAUTIONS) cannot be fitted onto a single page, then the words
" CONTINUED ON PEXT PATH chnuld be Centered at the botton of t .c page.
f) (j) Text following any underlined header (column headings on action a
step pages or section headings on the foldout pages) will be separated from the underline by two blank lines.
(k) Page rotation is to be avoided for emergency operating procedures.
7.2.2 ASSESSMENT - is the Operator Action Section presented in a dual column r
format? s the page layout consistent with ti.e sample page format?
Reference Sample Page Layout Attachment #2 O
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7.3 Identification Information Evaluation Goal
- - To determine if the EOP identification is complete.and correct.
17.3.1 ASSESSENT - is the procedure title descriptive for the purpose of the procedure?
- 7.3.2 ASSESSENT - does the cover sheet correctly provice the following
(a) procedure title (b) procedure number (c) revision number (d) number of pages 4
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EVALUATION GOAL To provide individual page identification information, the writers guide specifies what is required on each page. .The evaluation criteria, based on the writers guide, are as follows:
7.3.3 ASSESSENT - does each page correctly provide the following:
(a) procedure designator '
(b) revision number j (c) Page of numbers 4
/.>.4 ASSESSMENT - does the procedure have all its pages in the correct order?
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7.4 Information Presentation n
Evaluation Coal To determine if the instruction steps, cautions, and notes are clearly and consistently presented, understandable, and distinguished from each other.
7.4.1 ASSESSENT (a) Are instruction steps numbered correctly?
(b) Are operator-optional sequence steps identifie@
(c) Are instruction steps constructed to comply with the following:
- 1) Steps deal with only one idea.
- 2) Sentences are short and simple.
- 3) Operator actions are specifically stated.
- 4) Objects of operator actions are specifically stated.
(d) Do instruction steps make proper use of logic structure per Writers Guide (e) When an action instruction is based on receipt of an annunciator alarm, is the setpoint of the alarm identifies (f) Are cautions used appropriately?
(g) Are cautions placed properly?
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(h) Are cautions constructed to comply with the following?
A V 1) They do not contain operator actions.
- 2) They do not use extensive punctuation for clarity.-
. 3) They make proper use of emphasis.
(1) Are notes properly use@
-i (j) Are notes properly place @
(k) Are notes worded so that they do not contain operator actions?
.(1) Are numerical values properly written?
(m) Are values specified in such a way that mathematical operations are i
not required of the user?
I . ( n) Is a chart or graph provided in the procedure for necessary operator calculations?
(o) Are units of measurement in the E0P the same as those used on equipment?
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7.5 Procedure Referencing and Branching' 2
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Evaluation Goal To determine if the transitions within the E0Ps are consistent and compatible with the established rules of referencing and branching.
1 Note: Emergency Operating Procedure transition flow chart may be used to determine procedure exit point and entry points.
7.5.1 ASSESSENT (a) do the referenced and branched procedures identified in the E0P's exist for operator use?
(b) is the use of referencing minimize @
(c) are referencing and branching instructions correctly worde@
(C.1) are arrows ======== used for each "go to" instruction.
"go to" (branching)
" refer to" (referencing)
(d) do the instructions avoid routing users past important information such as cautions preceeding steps?
(e) are the exit conditions compatible with the entry conditions of the referencted or branched procedure?
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SECTION 8.0 C 'g O
VERIFICATION PRINCIPLES FOR MIE YAFKEE PLANT SPECIFIC ECP's 8.0 VERIFICATION PRINCIPLE OF TECWICAL ACCURACY The verification principle of technical accuracy is; the technical information contained in the E0Ps is consistent with and adequately incorporates the technical information contained in the E0P source documents.
The evaluation criteria to address technical accuracy is to divide the principle into the following areas:
8.1 entry conditions or symptoms information 8.2 instruction step, caution, or note information 8.3 quantitative information 8.4 plant hardware information 8.1 Entry Conditions or Symptoms Information Evaluation Goal To determine if the entry conditions or symptoms used for identifying when to use tha E0P are correct.
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8.1.1 ' ASSESSMENT O
(a) are the entry conditions of the E0P's listed correctly?
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(b) if additional entry conditions have been added, do they comply with the following:
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- 1) appropriate conditions for which the EOP should be used i
- 2) not excessive n ;
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.8.2- -Instructional Step Caution, and Note Information O
Evaluation Goal To determine if the content and arrangement of E0P instructional. steps, cautions, a.id notes are supported by information from E0P source documents.
To determine if the E0P's accurately incorporates appropriate 'information from the EPG Source Documents.
8.2.1 ASSESSENT - Are E0P/EPG differences:
(a) documented 1.e., Design Differences Document (b) explained 1.e., EOP Bases Document 8.2.2 ASSESSENT Is the EPG technical foundation (strategy) changed by the following changes in E0P steps, cautions, or notes:
(a) elimination (b) addition (c) sequence (d) alterations O
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8.2.3 ASSESSMENT O
o Are correct plant-specific adaptations incorporated per EPG:
(a) systems (b) instrumentation (c) limits (d) controls (e) indications 8.2.4 ASSESSENT (a) Have licensing commitments applicable to EOPs been addressecrl O
(b) Are differences between the licensing commitments and the EOPs or EPGs documented?
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8.3 Quantitative Information (a3 Evaluation Goal
- To determine if calculated or translated quantitative values used in the E0Ps are confirmed and correct.
8.3.1 ASSESSENT (a) Do the quantitative values, including tolerance bands, used in
! the ECP comply with the applicable ECP source document?
(b) Where EPG values are not used in the E0P, are the EOP values computed accurately?
4 (c) When calculations are required by the EOP, are equations
- presented with sufficient information for operator use?
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-8.4 Plant Harchtare Information Evaluation Goal 4
To determine if the plant hardware identified in the E0P is available for operator use that does not exist. ,
8.4.1 ASSESSENT -- Is the following plant hardware specified in the EOP available for operator use:
(a) equipment (b) controls
, (c) indicator (d) instrumentation 4
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MAIE YAM (EE ATOMIC POWER COWANY VERIFICATION GUIDELIE
9.0 INTRODUCTION
The purpose of this Guideline is to guide the administrative process used in the Verification of the emergency operating procetres (EOPs) and to assign responsibilities for carrying out the activities of the process.
9.1 Scope This process identifies and directs the phases of the verification process.
9.2 Applicability 9.2.1 This Verification Guideline applies to the initial Emergency Operating ProceWres being developed by Westinghouse for use at the Maine Yankee Atoimc Power Company.
9.3 REFERENCES
9.3.1 IW O 83-004 9.3.2 WAPCo. Writers Guide 9.3.3 IW O 82-017 31 of 43 8541N-REA
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- 9.4 IREENTING The E0P~ source documents to be used during the assessment phase are presented on Form #1 of the ECP Verification Forms (Attachment 1). For each E0P to be verified, additions, deletions, or revisions of the source documents to be used in the comparative evaluation should be noted in the spaces provided.
Comparative evaluation requires the evaluator to be famallar with the source doct.nents specific to the procedure set. The comparative evaluation of the E0P's for written correctness 'may be performed after the initial review of' the procedure set by operators. Written correctness should be addressed prior to the Validation process to ensure the transfer of information from
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the procedure set to the operator does not impact on the use of the procedures.
9.5 DEFINITIONS To establish uniformity in the meaning of key words used in this program, the following definitions are provided.
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id Emergency Operating Procedures (EOPs) - Plant procedures directing operator actions necessary.to mitigate the consequences of transients and accidents that cause plant parameters to exceed reactor protection system setpoints, engineered safety features setpoints, or other appropriate technical limits.
Emergency Operating ProcedJre Guidelines (EPGs) - Guidelines that provide technical bases for the development of E0Ps.
. EOP' Operational Correctness - A characteristic of E0Ps that indicates the
! degree to sich the E0Ps are compatible with plant responses, plant L hardware, and the shift manpower to manage emergency conditions in the plant.
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EOP Source Documente - Documents or records upon m ich E0Ps are based.
E0P Technical Accuracy - A characteristic of E0Ps that indicates the degree to which proper incorporation of generic and/or plant-specific technical information from EOP source documents and plant hardware has been made.
EOP Usability - A characterisitc of E0Ps that indicates the degree to which the E0Ps provide sufficient and understandable operator information to manage emergency conditions in the plant.
E0P Validation - The evaluation performed to determine that the actions specified in the E0Ps can be followed by control room operators to manage emergency conditions in the plant.
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i k/ EP Verification -- The evaluation performed to confirm the written correctness of the EOPs and to ensure that the generic and/or plant-specific technical aspects have been properly incorporated.
EOP Written Correctness - A characteristic of EPs that indicates the dgreee to which proper incorporation of information from the plant-specific Writers Guide for EOPs and other appropriate administrative policies has been made.
Symptoms - Plant characteristics that directly or indirectly indicate plant status.
Writers Guide for E0Ps - A document that provides instructions for writing f) EOPs, using good writing principles, o
Design Differences Document - A document that identifies system differences between the generic plant systems identified in the ERG Set and Maine Yankee Plant Specific Systems.
9.6 RESPONSI8ILITIES o Manager, Operational Support (Corporate)
The Manager, Operational Support will maintain corporate oversite of E0P development.
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f'h O Resources available to the Manager, Operational Support are contracted MYAPCo., consultants (Yankee Nuclear Services Division), the Site and Corporate Technical Staffs.
- EOP Project Co-ordinator The EOP Project Co-ordinator is responsible. for overall coordination of E03's effort.
EOP Verification Co-ordinator The EOP Verification Co-ordinator is responsible for the successful
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v completion of all aspects of the verification phase of the E0P effort.
9.7 Verification Team Responsibilities The management of the Verification Pr : gram is the responsibility of the Verification Co-ordinator. To ensur me expertise is available with background in the required disciplines the verification Co-ordinator shall assign specific corporate and site personnel to the verification team thereby minimizing the impact on the Operations personnel at the site.
Final resolutions of deficiencies will be reviewed and concurred by the Verification Co-ordinator or, a panel of three with disciplines in plant operations, hardware familiarization, technical analysis and Human Factor s.
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n 9.8 ECP VERIFICATION PROCESS The process of E0P verification consists of four phases: preparation, assessment, resolution, and documentation.
9.8.1 Preparation Phase
! The preparation phase consists of the following activities:
a designate personnel to conduct the comparative evaluation obtain and review the EOP source documents i 9.8.2 Designate Personnel The Verification Co-ordinator shall appoint the necessary personnel as evaluators to conduct the comparative evaluation. Personnel should be appointed based on operating experience and understanding of plant hardware, Design Differences Document, the EPGs, and the M.Y. plant specific Writers Guide and Human Factors background.
9.8.3 Obtain and Review the E0P Source Documents The listing of E00 source documents is provided on Form #1 of the E0P Verification Forms (Attachment 1) and shall be available for review by the personnel conducting the assessment phase. These documents shall be
. reviewed to ensure they are current. Any additional applicable source documents shall be listed if required.
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NJ 9.9 - ASSESSENT PHASE In the assessment phase the evaluator shall: ,
e 9.9.1 Make a general review of the E0P using the procedure-specific portion of the evaluation criteria (Section 7.0).
9.9.2 Indicate on Form #1 of. the EP Verification Forms (Attachment 1) that the evaluation was performed, either by checking the acceptable column or by designating the appropriate discrepancy sheet for any discrepancies identified.
, 9.9.3 Transfer the identified deficiencies to Form #2 for tracking & resolution.
9.9.4 Make a step-by-step review of the EP using the step, caution, i
note-specific portion of the evaluation criteria (Section 8.0). This review identifies the differences between the EPGS and the Plant Specific E0P's. List the EPG/EOP differences on Attachment #1 Form 4 (sequence step comparison sheet).
9.9.5 Indicate for each step on Form #2 of the E0P Verification Forms j (Attachment 1) that the comprative evaluation was performed, either by l checking the acceptable column or by designating the appropriate
! discrepancy sheet for any discrepancies identified.
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9.9.6 Transfer the identified ' discrepancy from Form #2 to Form #3 in written terminology (Part "A") of Form #3.
9.9.7 Complete Form #3 of the E(P Verification Forms (Attachment 1) and forward the verification forms with the discrepancy sheets to the appropriate personnel.
9.9.8 Resolution Phase (Reference Form #3, Part "B")
In the resolution phase, the appropriate department shall:
9.9.9 Review the evaluator's comments and resolve any conflicts between the
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writers' and evaluators' comments.-
9.9.10 Forward potential solutions to the Verification Co-ordinator for review and approval.
9.9.11 Update applicable source documents and procecijres with approved resolutions as directed by the. Verification Co-ordinator.
9.10 DOCLM NTATION PHASE i
The documentation developed throughout the process will be maintained in accordance with Administrative Procedures.
O 38 of 43 1 8541N-REA
ATTACHPENT #1 l3 C) l FORM #1 l l PMItE YANKEE ATOMIC POWER COWANY l l ECP VERIFICATION l lI COWARATIVE EVALUATION l l (a) E0P TITLE: l l E0P: OPAFT: PIN: l l REASON FOR VERIFICATION: l l l 1 (b) E0P SOURCE DOCLENTS USED: EVALUATORS l l 1. W EPG 's, Rev. #1 1. l l 2. M.Y FSAR 2. l l 3. M/Y EOP WRITERS GUIDE 3. l l 4. FC8 Series Orawings 4. .
l l 5. 5. l l (c) PROCEDWE-GENERAL VERIFICATION COMPLETED DATE l
(~l l 1. WRITTEN CORRECTNESS l l AREAS ACCEPTAa.E DISCREPANCY SFEET #(s) l l LEGIBILITY l l EOP FORMAT CONSISTENCY l l IDENTIFICATION IW ORMATION l l IWORMATION PRESENTATION l' l PROCEDWE REFERENCES & BRANCHING l l l l 2. TECHNICAL ACCURACY l l l lII EOP VERIFICATION CLOSE OUT l l INITIALS /DATE l l (a) All Comparative Discrepancies / Resolutions Resolved. / l l (b) All Technical Accuracy Discrepar.cles/ Resolutions Resolved. / l l (c) Final Procedure Copy Draft # Reviewed for l l Incorporation of Discrepancies / Resolutions. / l l (d) Forms #2, 3 & 4 attadled. / l Ol (e) Final Verified Procedure Attached Rev # . / l l (f) Verification Coordinator Approval: l l Date: Sign: l l l-8541N-REA
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ATTACHENT #1 LO. i F- n i E0P TITLE: ,
, E0P: ORAFT: PIN:
i i WLIE YA*EE ATOMIC POWER CCWANY l I EOP VERIFICATION I I E(P. ERG. DIFFERENCES I
. I PAGE OF I I I 1 STEP. CAUTION, WTE-SPECIFIC VERIFICATION l ;
I I j' I STEP NLMEER, I WRITTEN CORRECTNESS l TECHNICAL ACCIRACY .l l 3 l CAUTION, (R I ACCEPTABLE I DISCREPANCY I ACCEPTABLE l OISCREPANCY I I NOTE I I SEET # 1 l SHEET # 1 l i l l 'l 'I I l- 1 I I l i
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ATTACHENT #1 d('" l FORM #3 I Part A & B E0P TITLE:
EOP: DRAFT: PIN:
1 MIE YAPKEE ATOMIC POWER COWANY I I E0' VERIFICATION I I DISCREPANCY SHEET NUMBER I l PAT of I I STEP NUMBER (S): l I I I DISCREP/INCY: (Patt "A") l I I I ,I I I I I I I I I I I I I I I I EVALUATOR: DATE: l l FORWARDED TO FOR RESOLUTION: DATE: l l__________________________________________ l l RESOLUTION: (Part "B") l I I i 1 I I l 'l l I I I I I I l I I Verification Co-ordinator: Date:
Approved Yes No (Check one)
Resolution Incorporated by: Date:
41 of 43 8541N_REA
ATTACHENT #2 EP Rev Page _ of I- l l l l l Procedure
Title:
l Proc. No. 2-70-2 I l l l . E-2 ' - l Rev. No. l l
'l l ~ STEAM LIE WEAK l Page 2 of 4 l l l 1 l l l STEP ACTION / EXPECTED REPONSE RESPONSE NOT OBTAIED l l l l l
'l l l l l CAUTION l l l l At lease one SG rmJst be maintained available for l l RCS cooldown. I
- 1. l l Any faulted SG or secondary break should remain l
'l isolated during subsequent recovery actions 'l l unless needed for FCS cooldown. I l' l I .
I A i 1. Check Main Steamline Isolation Manually close valves. l l And Bypass Valves of Affected l l SG(s) - CLOSED I I I l 4 Isolate Faulted SG(s): Manually close valves. IF l l valves CAN NOT be closef I l a. Isolate main feedline TEN dispatch operator to l l locally close valves or l l b. Isolate AFW flow block valves block valves. l l l l c. Isolate other secondary l l piping: ,
I I l 1) l l 1 l 2) l l l l 7 Go To E-1, LOSS OF REACTOR OR l l SEC0tOARY COOLANT, Step 1 1 I I l -EW- l l l l l l l l l FicyJre 6. EXAWLE INSTRUCTION STEPS --- l O li i 42 of 43 8541N-REA
- ATTACHENT #1 v' l FORM #4 l MIE YAtKEE ATOMIC POWER COWANY SEQUENCE STEP COW ARISON SHEET EOP TITLE:
E0P: PIN: ORAFT: DATE:
1 I ERG I l l ERG l l MYEOP l EQUIL.l REMlHKS I MYEOP l EQUIL.I I I I I I I I I I I I I I I I I I I l l I I I I l l l l l l l l l l l l l l 1 I I I I I I I I I l l l l l l l l 1 1 I l l l l l l l
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I I I I I l v i l I I I l 1 1 I I I I I I I I I I I I I I l l I I I I I I I I I I I l l l l l l l l 1 1 I I I I I I I I I I I I I I l l l l l l l l l l l l l l l l l 1 I I I I I I I I I I I I I I I I I I I I I I I I I ~
I I I I I l i I I I I I l l I I I I I l i I I I I I I I I I I I I i i i I
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l l I I I l l l l l l l WPS: Maine Yankee Plant Specific Step NA: ERG Step is not applicable to Maine Yankee 43 of 43 8541N-REA
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- s. 1 l r EERGENCY OPERATING PROCEDURES !
! VALIDATION PROGRAM .
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TA8LE OF CONTENTS
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v PWAPCo. Validation Program PAGE 1.0 PROGRAM 0VERVIEW...............................................
2.0 PROGRAM DESCRIPTION............................................
3.0 PROGRAM 0BICT IVE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
4.0 PROGRAM EVALUATION CRITERIA....................................
5.0 PROGRAM PR0 CESS................................................
5.1 Preparation..................................................
5.2 Assessment...................................................
5.3 R e so lu t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
5.4 Docunenta t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
6.0 E0P VALIDATION PRIOR TO EJP IWLEENTATION. . . . . . . . . . . . . . . . . . . . . -
6.1 ECF Validation Subsequent to E0P Implementation. . . . . . . . . . . . . .
7.0 ~ VALIDATION PRINCIPLES APO EVALUATION CRITERIA. . . . . . . . . . . . . . . . . .
(S 7.1 Validation Principle of Usability............................
7.2 L ev e l o f De ta i l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
7.2.1 Goa l o f Evaluation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
7.2.2 E xpla na t ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
7.2.3 G u i da nce . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
7.3 U nder s ta nda b ili ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
7.3.1 Goal o f Evaluation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
7.3.2 E xpl a na tio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
7.3.3 G u i da nce . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
8.0 VALIDATION PRINCIPLE OF CPERATIONAL C0RRECitESS. . . . . . . . . . . . . . . .
8.1 Plant C ompa tab 111ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
8.1.1 Goal o f Ev a lua tion. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
8.1.2 E x p l a na t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
8.1.3 Guidance.................................................
8.2 Opera tor Compa tability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
8.2.1 Goal o f Evalua tion. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
8.2.2 Explanation..............................................
8.2.3 G u i da nce . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
9.0 E0P VSLIDATION GUIDELINE.......................................
9817N-REA
EOP VALIDATION PROGRAM
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im 1.0 PROGRAM OVERVIEW The E0P validation program and the EOP verification program are interdependent. The overall combination of both programs evaluates four E0P characteristics: usability, operational correctness, written correctness, and tednical accuracy.
The validation program evaluates the ECP for usability and operational correctness. The evaluation of written correctness and temnical accuracy is addressed in an EOP verification program. This division of the evaluation for the four ECP daracteristics is based on tre type of evaluation to be used. The EOP validation program encompasses the em efforts necessary to support a performance evaluation of the EOPs, D
whereas the EOP verification program encompasses the efforts necessary to support a comparative evaluation of the EOPs.
2.0 PROGRAM DESCRIPTION The EOP validation program consists of the following:
0 the objective of the program o
the evaluation criteria to be used a the process to be followed In this program the evaluation criteria are applied to determine if the
( program objective are satisfied. The evaluation criteria are applied each time the EOP Validation process is used.
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3.0 PROGRAM OBECTIVE iO v i The objective of E0P validation program is to determine if the control room operators can manage emergency conditions in the plant using the i
E0Ps. This determination can be made by evaluating the E0Ps with regard to the validation principles of usability and operational correctness.
Usability - The _EOPs provide sufficient information that is understandable to the operator.
' Operational Correctness - The E0Ps are compatible with plant [
responses, plant hardware, and the shift manpower.
4.0 PROGRAM EVALUATION CRITERIA Program evaluation criteria to be used during the assessment of the E0Ps to determine that the ECP validation program objective has been satisfied. These criteria are developed by addressing E0P validation principles which support the objective of the E0P validation program.
5.0 PROGRAM PROCESS The program process consist 'of the following:
preparation, assessment, O reso1etioo.
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5.1 Preparation l
l Preparation consists of selecting the validation method, planning for the validation method, developing the validation scenarios, and determining the use of the evaluation criteria.
5.1.1 Selecting the Validation Method l
The selection of one or more validation methods for the assessment is dependent on the degree to ditch the E0Ps are to be evaluated for compliance to the validation l principles. Additionally, the resources available at the time of E0P validation nust be considered.
j The method's for WY E0P validation will be the Simulator Method. E the simulator is unable to simulate a desired l
condition, the EOP will be " walked-thru" using the same l
! assessment criteria.
I Simulator Method - A validation method whereby control room operators perform control functions on simulated equipment during a scenario for an observer / reviewer.
This validation method will yield useful information about the E0Ps. This method permits more comprehensive testing of the E0Ps. The relative effectiveness of the validation method for evaluating EPs at WY with regard to the EOP validation principles is:
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- Simulator method provides for assessment of the compatibility of the plant process and operator actions. This is the most comprehensive challenge to the E0Ps.
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5.1.2 Plaming for the Validation Method l
l Plaming for the simulator method is crucial. Personnel and l
l equipment resources required for the method must be scheduled. The following is a list to consider dien checking the schedule of available resources.
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a scenarios (selection and development) control room equipped with adequate table surface to lay out E0Ps and related documentation
- support equipment (simulator methods)
- audiovisual equipment (if required)
- required persomel j - operating crew l - nonoperating specialists
- observer / reviewers l
- support equipment operators documents to be used l
l - required EOPs
- data recording forms
- other plant specific procedures O
! 2-4 l 9817N-REA
t a training
- operators using the EPs during E0P validation
, should be trained to a level #11ch ensures they are familiar with the " Rules of Usage". and how the EP 's work tio mitigate accidents-
- observer / reviewers should be trained to support their assigned function in the assessment ;
I Planning for the simulator validation method is more complex
- ciae to the need to integrate hardware, software, and personnel during relatively short periods of simulator l -
availability. Therefore, appropriate personnel should be organized, documents should be researched, and appropriate O
.t) data collection methods should be prepared.
l The Maine Yankee Plant Specific Control Room simulator will be used for EP validation. Where differences in the characteristics exist between the actual control room l
setting and the simulator setting, it is important to understand that the evaluation of the EOPs in the validation process is beina made for the actual plant setting. The simulator is simply being used as the setting for making an evaluation of the E0Ps in the validation process. A comparison of the two settings with respect to the actual control room's design will aid the observer / reviewers later in determining whether or not operator performance '
deviations occurring in the simulator setting could also occur in the actual plant.
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Conversely,: the observer / reviewers might be able to identify potential problem areas that could arise at the plant but not at the simulator.
Operational differences between the plant and simulator can be minimized by adapting the initial symptons designated in the scenario on the plant to equivalent simulator syrsptoms through the programming of malfunctions. Differences that exist between actual plant equipment and simulator equipment may be resolved by rewording the references to plant-
! specific equipment in the proce&res. However, the strict
(- intention of amending the proce&res is only to reword references to plant-specific equipment sud1 that the procedures' characteristics are not altered and the operators will not be faced with equipment substitution i
decisions as they use the procedures on the simulator.
Planning the techniques for detecting and recording operator performance data on the simulator during the assessment is i
necessary. These data collection tediniques will fall into l two major categories - automatic and/or manual. Automatic data collection techniques would require the use of l computer-related software, whereas manual techniques would require the use of personnel to observe operator actions using observation forms and/or videotape, i O 2-6 9817N-REA
- A combination of these measc;es can be employed to record i
J operator performance and provide a reliable data base for evaluating the E0Ps.
5.1.3 Developing the Validation Scenarios A scenario is a structural plan of parameter and plant sympton changes that provide operating cues for the conduct of assessment. It is a written description of a transient and resulting parameter changes used to exercise the EOP.
It provides the background, prerequisite conditions, and the 4
proper sequence of realistic plant symptoms and responses.
The scenario is designed to guide the E02 user through (o) v anticipated procedure steps so that-the evaluation criteria can be addressed. It will vary in length, complexity, and presentation style depending on the validation method used.
5.1.4 Determining the Use of Evaluation Criteria Evaluation criteria are needed to control and direct the assessment. Once developed, the evaluation criteria can be used in many ways by the personnel conducting the performance evaluation. The evaluation criteria developed can be tailored to discussion, observation, or debriefing forms depending on the validation method.
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5.2 Assessment g
V i During assessment, performance deviations (actual versus expected responses) are detected as a validation method. The performance deviations are then used to identify potential problems with the E0Ps. Assessment only identifies deviations and does not classify l them. Classification of the deviations occurs during resolution.
Assessment for the simulator mett.od is the detection of performance deviations by observing plant responses and operator actions. Table l 1 is a sample list of behavioral symptons that may aid in this detection by observers. Failure or success in ad11eving expected plant parameter response will determine the technical adequacy of i
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the E0P instructions.
j Certain general actions are applicable to the assessment regardless i
of the method used. The observer / reviewer leader should review the responsibilities of team members. Some individuals may be observers and data recorders, while other may be support equipment operators.
l All observer / reviewers should be familiar with the overall format and sequence for conducting the assessment. If reasibic, they 1-should perform a " dry-run" of the assessment and may use an abbreviated form of the scenario (s) for this purpose. If an observer / reviewer has questions concerning i
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9817N-REA !-
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I-L I TABLE 1 D BEHAVORIAL SYr TOMS TO AID IN OBSERVATION
- OF PERF0fMNCE DEVIATIONS l-i l
l OPERATOR COMMISSION DEVIATION IPOICATORS does not walk to anticipated area of control room does not look at anticipated display
- does not touch anticipated control l
does not set control to anticipated value performs an action not in the procedire ,
a does not select anticipated procedures O
- a tror ctio" oet or ae oc-has difficults determining values from charts, graphs, tables, etc.
l OPERATOR OMISSION DEVIATION INDICATORS
- does not perform an action or step
- allows a limit to be exceeded l
l l
- fails to detect sympton a
fails to perform task within specified time (dien required) does not use procedures (dien procedures are available) o cannot find information in procedures O
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y use of the data collection forms, the questions should be i
V). clarified. If videotape is used, coordination with the video operators should be finalized at this time.
Before starting the assessment, operators taking part will be '
briefed. Answers should be provided to questions such as the following:
. What are the objectives of the assessment?
- How will the assessment be conchjetecf?
O What will the operators be required to do?
The observer / reviewers will be introduced to tr.a operators, and their roles during the assessment will be explained. The use of i observation and debriefing forms should be reviewed with the operators. This briefing should acquaint the operators wi,th the observer / reviewers and snould familiarize the operators with the overall purpose and technique of the assessment. Operators must be told that the assessment is of the E0Ps and not the operators.
Although their actions and responses are important to the validation program, the operators themselves are not being critiqued.
For the simulator method of validation. Observer / reviewers should i observe operator performance during each scenario run on the simulator, recording their observations on the previously prepared forms and checklists or the actual E0P being validated. A combination of any of these is acceptable.
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To supplement their information collection capabilities, they may v use videotape and/or computer recordings as desired. However, the impact that videotape recording or other intrusions in the control room might have on the operators' performance should be considered.
5.3 Resolution In resolution, the deviations identified during assessment are evaluated. For some deviations, no resolution is required.
Deviations that are determined to require resolution should be evaluated for their apparent cause with relationship to the personnel, training, plant design, or E0P. Solutions can then be determined for each deviation and a resolution from the possible
[) solutions is selected to remedy the deviation. For example, if the operating crew felt that the E0Ps were not detallad enough, possible solutions might be either to increase the level of detail in the E0P or to increase the training emphasis on the indicated Steps with a follow-up evaluation to ensure tne operator does not need additional procedure detail. A resolution would be chosen from these two possible solutions.
Depending on M/Y constraints, resolutions for any given deviation will vary. If a like deviation is noted during a subsequent EOP validation after a resolution has been enacted, a different resolution may be required. If the resolution results in a diange in the EOP, Ole change should be verified and if warranted, p
( ) revalidated.
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For the simulator methods an operator debriefing session can be (V')
conducted for resolutions immediately-after the running of each scenario during the assessment. In this way, the events of the scenario will be fresh in the operator's minds and their comments to the observer / reviewers will be more comprehensive. It is suggested that the debriefing session be recorded, either by tape recorder or manually by a secretary or stenographer. Recording the session will aid in later review of the data.
A sequence for conducting a debriefing that provides for operator comment without initial influence by the observer / reviewers is as 4
follows:
O w/
Debriefing sequence:
o Observer / reviewer leader briefs the participants on the purpose and objective for debriefing, o
Operators present problems and discrepancies that they had identified during assessment. These verbal explanations may be augnented by videotape displays of problems.
- Operators provide possible reasons for problems.
- Operators present potential solutions to problems.
- Observer / reviewer team presents other problems and discrepancies identified during assessment. Verbal explanations may be augmented by videotape displays.
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- Operators describe possible reasons for the other problems, a
Observer / reviewers summarize the findings of the debriefing.
Another medianism for uncovering information about the E0Ps is
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through a survey given to the operators. The results of the survey could then be discussed with the operators.
5.4 Documentation Documentation provides a traceable history of the validation program that WY can reference. Because it is intended to aid WY, .
documentation can be as brief or as exhaustive as desired. The following is a proposed list of items that could be saved as part of
(^T the documentation:
%.)
o scope of the validation a list of participants
- description of validation method (s) used a copy of any validation procedures used results of any plant / simulator. compatibility checks amendment of E0Ps for simulator method 8
scenarios used o evaluation criteria used
- observation forms used debriefing forms used a possible solutions for each deviation O - rese1etiee fer eech eeveitie, 2-13 9817N-REA
6.0 EP VALIDATION PRIOR TO E0P IhPLEENTATION
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(3 E0P validation prior to E0P implementation will demonstrate that the operators' can' manage the emergency conditions through the use of the procedures and that the EOPs are operationally correct.
I It is the responsiblity of Maine Yankee to determine the point at #11ch '
it considers its EOP validation effort to be adequate. This determination should be made based on diat WY has accomplished in addressing the validation principles of usability and operational correctness. The validation methods should be applied selectively to yield an efficient and effective validation program. The simulator method provides for assessment of the time-dependent interactions of
( plant processes and operator. interactions.
6.'l EP Validation Subsequent to E0P Implementation After conducting EOP validation prior to implementation, WY may choose to perform subsequent E0P validation as necessary. This could be conducted in conjunction with existing training programs and could provide periodic feeWack on the E0Ps in a manner which effectively uses material and personnel resources. For example, the
- regular classroom or simulator instructor for an operator requalification training session could also serve as the EOP validai. ion observer / reviewer. The instructor and operators could complete Forms identifying any problems with the EOPs found.during in Q- operator requalification training.
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4 7.0 VALIDATION PRINCIPLES APO EVALUATION CRITERIA 7.1 Validation Principle of Usability The validation principle of usability is stated as follows: The E0Ps provide sufficient and understandable information to the operator.
In developing evaluation criteria to address this validation principle, it is helpful to divide the principle into two procedure characteristics:
o level of detail
- understandability 7.2 Level of Detail 7.2.1 Goal of Evaluation The goal of evaluation is to determine that the EOPS contain sufficient information, consistent with' training, for the operator to properly execute the EOP instructions.
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7.2.2 ' Explanation The level of detail in the EOPs is a balance between providing all the possible operating information and '
providing the minimum amount of needed operator
-information. 'The plant-specific writers guide addresses the desired level of detail, and during EOP validation, _the E0P '
user's judgement and that of the observer / reviewer on sufficiency is obtained.
4 7.2.3 Guidance Specific evaluation criteria for this procedure Il characteristic should be developed to identify any U
discrepancies. Sample evaluation criteria are as follows:
0 Was there sufficient information to perform the specified actions at each step?
o Was there sufficient information for decisions at each transition in the EOPs?
- Were the alternatives adequately described at each decision point?
- Did the operator use labeling, abbreviations,
'I symbols, and location information as provided in the E0Ps to find the needed equipment?
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a Were the E0Ps missing information needed to manage t 1 V the emergency condition?
- Did the operator use the title and numbers to find referenced and branched' procedures?
7.3 Understandability
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7.3.1 Goal of Evaluation The goal of evaluation is to determine that the user comprehends the information presented in the EOPs.
7.3.2 Explanation O
.The understandability of the E0Ps results from proper presentation and consistency of information (such as readable. print, standard terminology, standard method of format'and emphasis). The E0Ps have been written in accordance with the WY plant-specific writers guide to facilitate the understanding of the written inforamation.
During EOP validation,' the EOP user is asked to evaluate how well he could comprehend the EOP under emergency conditions.
7.2.2 -Guidance Specific evaluation criteria for this procedure (Ov characteristic should be developed to identify any discrepancies.. Sample evaluation criteria are as follows:
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o Was the E(P easy to read?
g o' Were the figures and tab 1'es identified, understood, and accurately rea d a
Were the values on figures and charts easily determined? ~ '
o were cautions and notes identified and understood?
o Old the operators follow branching and referencing instructions?
o Were the EOP steps complied with?
o Did the operator find the particular step or set of steps when required?
Did the operator return to the procedure exit point without omitting steps when required?
o Did the operator enter the branched procedure at the correct point?
-o Did the operator exit from a given EOP at the correct branch?
8.0 VALIDATION PRINCIPLE OF OPERATIONAL CORRECTNESS The validation principle of operational correctness is stated as follows: The EOPs are compatible with plant responses, hardware, and the.
shift manpower.
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,. In developing evaluation criteria to address this validation principle, it is helpful to divide the principle into two procedure characteristics:
a plant compatability
- operator compatability 8.1 Plant Compatability 8.2.1 Goal of Evaluation The goal of the evaluation is to determine that the EOPs are compatible with plant hardware and plant responses.
. () 8.2.2 Explanation The operator must be able to carry out the actions specified
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in the EOPs with existing plant hardware. In addition, if the plant responses during the transient are different than those which the EOPs address, the operator could lose confidence in the procedures' correctness. Therefore, this procedure diaracteristic addresses whether the information in the EOPs is correct for operating the plant hardware during emergency conditions and whether the plant response is addressed by the EOPs.
3-5 9817N-REA
ys 8.2.3 Guidance N
Specific evaluation criteria for this procedure characteristic should be developed to identify any discrepancies. Sample evaluation criteria are as follows:
o Could the actions specified in the procedure be performed in the designated sequence?
o Were alternate success paths found that were not included in tne EOPs?
8 As specified in the procedure, could the operator obtain the necessary information from the plant inst:umentation that is providecf?
[
- Did the plant symptoms provide adequate information for the operator to select the applicable E0P7 o
Were the E0P entry conditions appropriate for the
. plant symptoms seen by the operator?-
- Did the operator have to use information or equipment not specified-in the EOPs to accomplish his task.
Did the plant responses agree with the EOP basis?
8 -Were the instrument readings and tolerances consistent with the instrument values stated in the EP?
o Were the EOPs physically compatible with the work situation (too bulky to hold, binding would not
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(mj allow them to lay flat on work space, no place to lay ~the EOPs down to use)?
3-6 9817N-REA
Were the instrument readings and tolerances specified by the E0P for remotely located o
instruments accurate?
'8.2 Operator Compatability 8.2.1 . Goal of Evaluation The goal of the evaluation is to determine that the E0Ps are compatible with the shift manning levels and policies.
8.2.2 Explanation The shift manning level must be adequate to comply with the actions specified in the EOPs. In addition, the plant's policies for individual operator's responsibilities and duties must not conflict with the actions specified in the EOPs. Therefore,' this area addresses the ability of the shift manpower to perform the actions specified in the EOPs.
8.2.3 Guidance Specific evaluation criteria for this procedure characteristic should be developed to identify any discrepancies. Sample evaluation criteria are as follows:
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3-7 9817N-REA 1
- If time intervals are specified, could the procedure actions steps be performed on the plant within or at the designated time intervals?
I Could the procedure action steps be-performed by the operating shift?
If specific actions are assigned to individual shift personnel, did the E0Ps help coordinate the
[ actions where necessary?
o Was the operating shift able to ' follow the designated action step sequences?
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1 3-8 9817N-REA 3
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! D 9.0 EOP VALIDATION GUIDELItES i
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~: 1 TAER E OF CONTENTS l l l l Section 9.0 .I.
I I I
9.1 INTRODUCTION
................................................. I I .
I I 9.1.1 Purpose.................................................. l l 1 l 9.1.2 Scope.................................................... I I .
I I 9.1.3 A ppl icab ility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I I.. I I 9.2 R EFER ENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . l 1 I I 9.2.1 General.................................................. I i l l 9.2.2 EOP Source............................................... I I I I 9.3 DEFINITIONS.................................................. I I I I 9.4 RESP 0NSIBILITIES............................................. I 1 I I 9.4.1 Manager, Operational Support............................. l I I 9.4.2 Valida tion Co-ordina tor. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I l 9.4.3 ~ E0P Project Co-ordinator . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . l I I l 9.5 PROCESS FOR EOP VALIDATION................................... I
-l l I 9.5.1 Preparation............................'..................
I I I
- I 9.5.2 A s se s smen t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I I I l 9.5.3 R esolu tion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I I I l 9.6 0 0Cl)ENTAT I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I i l l l l 9.7 ATTACH E NTS.................................................. I 1 I i FORMS l l l l Checklist for Simulator Method of Validation l I I l I l' I I I I I I I l - I f I l
I l
1 Page 2 of 22 I I I 9817N-REA
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9.1 INTRODUCTION
l l I I i
l 9.1.1 Purpose l l l l The purpose of this guideline is to guide the adninistrative process l l used in validation of the WY emergency operating procedures (EOPs) l l ard to assign responsibilities for the process. l l l l 9.1.2 Scope l l l
] This guideline identifies the aspects of the validation program processl I and provides guidance that encompasses the simulator validation method.l O i i l 9.1.3 ' Applicability l I i l This guideline applies to EOP validation prior to implementation l l as well as to validation subsequent to implementation for the Maine l l Yankee Atomic Power Cortpany. l 1 l l 9.2 REFERENTS l l 1 l 9.2.l' General l I I l
Emergency Operating Procedures Validation Guideline l l (ItF0 83-006) l l l 0 WCAP 10599 l l l Page 3 of 22 l 9817N-REA
b,O l 9.2.2. E(P Source Documents l l I l
. Westinghouse Owners Group Emergency Response Guidelines, H.P. l l Revision - Rev. #1 l 1 l-o FSAR, Unit 1 1 l 1 l
.l 0 Results of EOP verification l 1 l l
8 Analysis Differences Document l I I l
0 MY Plant Specific Writers Guide l n
k_)
l 9.3 DEFINITIONS l l l l Control Room Simulator - A device. that dynamically models the plant l l functions as presented in the control room. l l l l ' Emergency Operating Procedures (EOPs) - Plant procedures directing l l operator actions necessary to mitigate consequences of transients and I i accidents that cause plant parameters to exceed reactor protection l l setpoints, engineered safety feature setpoints, or other appropriate l l temnical limits. l 1 l l Emergency Operating Procedure Guidelines (EPGs) - Guidelines that l l provide tecnical bases for the development of EOPs. I I i l Page 4 of 22 l 9817N-REA
l E0P Source Doctsnents - Documents or records upon which EOPs are based. l l 1 ,
l EOP Validation - The evaluation performed to determine that the actions l l specified in the EOP can be followed by trained operators to manage l l the emergency conditions in the plant.' l l l l Mock-Up - Static device (e.g. , models, photos, drawings) that p'ortrays l l control room hardware and configuration. l 1 l l Reference Validation - Method of validation whereby data developed in l l common ECP validation program is referenced by similar plants. l l l
( l Scenario - A structural plan of parameter and plant symptom changes l
,l that provide operating cues for the conduct of assessment. l l l l Simulator Validation - Method of validation whereby control room l l operators perform actual functions on simulated equipment during a l I scenario for' an observer / review team. l l l l Table-Top Validation - Method of validation whereby personnel explain l l and/or discuss procedure action steps for an observer / reviewer in re- l l sponse to a scenario or as part of an actual industry operating l l experience review. l I I l l l l l l l Page 5 of 22 l 9317is-REA
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l Walk-Through Validation - Method of validation diereby control room l l operators conduct a. step-by-step enactment of their actions during a l l scenario for an observer / review team without carrying out the actual l l control functions. l I l l 9.4 RESPONSIBILITIES l 1 l l 9.4.1 Manager, Operational Suoport (Corporate)- l 1- l 1 The Manager, Operational Support will maintain corporate oversite of l l the ECP development. I I I I 9.4.2 ECF Validation Co-ordinator l I I l The ECP validation co-ordinator is responsible for the successful l l completion of the validation phase of the E0P effort. l 1 l o
l managing the validation program and ensurjng its smooth l l coordination with the Simulator Group l l 1 l
o determining if validation is needed and its scope l l l l selecting the validation method i I I I l
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- appointing and training an observer / reviewer team I l l - three persons 'for the simulator validation method l I I
- I a cupleting applicable portions of the EP Validation Form's l l l l
arranging for rotating operating crews through the training / -l l validation sessions l l- l l
schecialing simulator training time for validation purposes as I l appropriate l l l I
- reviewing discrepancies and resolutions forwarded to him by
, I f I observer / review personnel I l- 1 o
I forwarding recommended resolutions and procedure dianges to the 1 I E0P Project Co-ordinator for approval and incorporation into l I EP's l l l l 9.4.3 ET Project Co-ordinator I I I 2
i The EOP Project Co-ordinator is responsible for overall co-ordination l I of the EOP effort, l I l~
l 1 1 I O i i f
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I 9.5 PROCESS FOR E(P VALIDATION l l l-I Regardless of the validation method, the EOP validation process can l I be described by the three phases of preparation, assessment, and l l resolution. I I I I 9.5.1 Preparation l l l l Each validation method will reference the applicable evaluation i I criteria presented in Table 1, and the scenario to be used will be l l recorded on the appropriate scemrio form: I 1 l I
- Simulator Scenario Form #2 l l l l 9.5.2 Assessment (Evaluation Criteria) I
.I I l Specific guidance for assessment is presented on the checklist l l for the simulator validation method Attachment 2. l I I I I I I I I I I I I I I bI I I I l Page 8 of 22 1 9817N-REA
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N l 9.5.3' Resolution i I I l Resolution will be accomplished by reviewing discrepancies and com- l l ments presented on the Discrepancy Sheet, Form 3. The observer / l l reviewer will propose solutions, if needed, and forward to the - l i l Validation Co-ordinator for approval, with the other designated l l documentation. l 1 1 I 9.5.6 Docunentation l ,
l l l The documented items needed to provide a history of the validation l l l program are specified on cach validation method checklist. These l
(] l l
Items wi]l be maintained as a validation package in the document control storage area of the WAPCo. corporate area.
l I
I I I I I l i I
-l I i I
.I I l l l I I I I I O i i l I l Page 9 of 22 I 9817N-REA
l ATTACHENT 2 l
-l CHECKLIST FOR SIMULATOR l l ETH00 0F VALIDATION l l l
['1 PWPOSE l l l l The purpose of this checklist is to provide guidance for. the simulator l l method of validating E0Ps. l l l l 2 VALIDATION PROCESS l l l l EOP validation will be conducted in three parts: preparation, assessment, l l and resolution. l l l 1 2.1 Preparaticn, l 1 l l The designated observer / reviewer will be responsible for the following: l l l l
8 using and completing the ECP Validation Form (Form 1) l l l
- l. o reviewing the scope of the validation as directed by the l l Validation Co-ordinator or his representatives. l l l o developing or modifying scenario runs to support the scope of l I l validation as necessary. I O i i l l l Page 10 of 22 l 9817N-REA
f s, U l completing the upper portion of the Simulator Scenario Form I l (Form 2) and forwarding to the Validation Co-ordinator. I l l l o develop data collection technqiues. l l l
- l.
- evaluating plant-to-simulator characteristics. l
.I l l
'o Recommend the required adjustments to the ECP set to use on the l l simulator. I
~l l l
. modify / select the evaluation criteria to support the scope l l of validation. l I I l
ensure the E0Ps and supporting procedures are available. l I l l 2.2 Assessment !
I l l The designated observer / reviewer will perform the follow duties: l l 'l
.l
- brief the operating crew on the scope of the validation and how I
- ~l the assessment will'be conducted. I I I l i I I I I (O I I l l l
l Page 11 of 22 l l 9817N-REA l l
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,7-U l ensure the observer / reviewer does not interfere or interact with l 1
l the operating crew.. 1
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- - brief the operating crew on initial plant conditions for each l l scenario run. l
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! I I a conckJct a debriefing with the operators as soon as possible after
-l l l l each scenario run using the following sequence: l l
l l l
l - brief the participants on the purpose and objectives for l
~l debriefing l l_
1 - have operators present problems and discrepancies which they had l
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\_s l identified during assessment l l' - present other problems and discrepancies identified during l L
I assessment l l - have operators describe possible reasons for the other problems I L
l' - summarize the findings of the debriefing for the operators l l l j
l l
- record discrepancies and comments on Form 3 Part "A". l l l l 2.3. Resolution l I I
! l The designated observer / reviewer will perform the following duties: l l l l
- review comments and discrepancies on Form 3 Part "A". l f I l l v i l l l Page 12 of 22 l l 9817N-REA i
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- propose resolutions on Form 3 to the Verification Co-ordinator. l I l l
submit the validation package to the E0P Co-ordinator. l l l l The E0P Co-ordinator will perform the following tiJties: l l l l
- review proposed resolutions with appropriate staff. l l
- . select resolutions for incorporation in the ECPs. l l
0 present the revised E0Ps to PORC for approval. l l l l 3. DOCtMENTATION .l 1 I
'l The following documentation will be submitted with the validation package: l
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- completed Form 1 cover sheet l 0
l Form 2 (Simulator Scenario Form) l I a l~ Form 3 Discrepancy / Resolution l I o EOPs used for the validation l l
0 data on simulator test scenario run l I I I I I I I I I I I I O i i
! l l Page 13 of 22 l
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l FORM #1 l l MYAPCo. l l 1
-l E0P VALIDATION l l Page 1 of 3 l I I I -l l .I E0P
Title:
l [
l EOP NO._ DRAFT /REV NO. PIN l l Scope of V111dation: l I l I I ,
-l II Validation Method or Methods to be Used: (Simulator) l .
$ l l lIII Designated Observer / Reviewer (s): Title l ol l l
C l
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1 D l l E l l l l Operator (s) Involved: Qualification: (SRO, RO, Other) l
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I l 1 I I I I I I I I I I- I.
I l i- l Resolution Completed / / By: l l l Documentation Package Completed / / By: I
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.l SIHJLATOR SCENARIO FORM l
.I E0P VALIDATION l l Page 2 of 3- l l l l ECP: PIN: l l .I I PROCEDLRE NO.: REV./0 RAFT NO.: l l' I l' DATE: l l l .
I PURPOSE: .l
.; I O l l l SCENARIO DESCRIPTION: l l 1 1 I I I I I I I I I I I
-l ~ INITIAL PLANT C0tOITIONS: l l- l 1 1 I I I I I I I l O- l l l ATTACHENTS YES -
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a O i FORM #3 i
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- l DISCREPANCY SEET NlbeER: -l I I l ,
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l E(P:- ORAFT/REV. NO.: -
STEP NL)6ER: I !
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'I I. DISCEPANCY: (PART A) i I I I I I I I I I i ol l l
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< l EVALUATOR: DATE: I 1 I l II. RESOLUTION: (PART B) l
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- I I I I l l APPROVED
- YES NO (circle one) l 1 RESOLUTION INCORPORATED BY: DATE:
1 I 4 I VALIDATION CO-ORDINATOR: DATE: I j l I 9817N-REA
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EVALUATION CRITERIA FOR EP NO. l I I l - VALIDATION - l l Page 1 of ' 4 l l I. USABILITY l l l l l l Step i l lS U l- No. A. LEVEL OF DETAIL l l l l 1 l
l l l l 1. Is there sufficient information to perform the specified I i l l l actions at each step? I I l- ,
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I l l l l 2. Are the alternatives adequately described ht each l l l l l decision point? I I I l l l l l l l .
I l l l l 3. Are the labeling, abbreviations, and location informationi l l l l as provided in the EOP sufficient to enable the operator l l l l l to find tne needed equipment? l l l l l l l l l l 1 l l l l4 Is the E0P missing information needed to manage the l l l l l emergency condition? l O lll l l l l
l 5. Are the contingency actions sufficient to address the l
l l l l. I symptoms? l 1 1 I I I l l l l .
l l l l l 6. Are the titles and numbers sufficiently descriptive to l I l l l enable the operator to fino referenced and Dranched l l l l l procedures? I l l l l l l l l Step i 'l iS Ul No. I B. UPOERSTAPOABILITY l
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i l l l l l l 1. Is the EP easy to read? l l l l l l l ,
I I I l l l l 2. Are the figures and tables easy to read with accuracy? l l l l l l l l l l l l l l l 3. Can the values on figures and charts be easily l l l l l determinecf/ l l l l l l l l l l l l l l l4 Are caution and note statements readily understandable? I l l l
O l
i i i is are t"e eoa stea reeeitv ""eer te"eee e2 i l i I I I 9817N-REA
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- b l l EVALUATION CRITERIA FOR ECP l
l l l 1 - VALIDATION - l l Page 2 of _,A _
l l I. USABILITY l l .. I l l l Step i .1 lS Ul No. B. UPOERSTAPOABILITY l l l .
I I l l l 6. Can the particular steps or sets of steps be readily l l l l l located when required? I I I l l l l l 1 I l l 7. Can procedure exit point be returned to without omitting l l l l l l l l steps when required? I I I I I I I I I I l l l l l 8. Can procedure branches be entered at the correct point? l l l l l 1 l l l 1 l l l l l 9. Are EOP exit points specified adequately? I
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EVALUATION CRITERIA FOR E0P l l
I - VALIDATION - l l Page 3 of 4 l l II. .PERATIONAL MRRECTNESS I l 1 l l l Step i l
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No. I A. PLANT COWATIBILITY l I I l l .
l l 1. Can the actions specified in the procec1Jre be preformed I l l l '
l l l l in the desicpated sequence? I I l l l l l 1 1 I i l l l l 2. Are there alternate success paths that are not included l l l l l in the EOPs? l l 1 l l l j i i l l l l, l 3. Can information from plant instrumentation be obtained, I l l l l as specified by the EP? l I I I I I .
l l l l l I I l l4 Are the plant symptoms specified by the EP adequate to I l l l l enable the operator to select the applicable EOP7 l 1 l l I i l l l l S. Are the EP entry conditions appropriate for the plant l l .l l l symptoms displayed to the operator? I I l 1 l l l l l l l 1 l l l 6. Is information or equipment not specified in the EP l l l l l required to accomplished the task? l l -l l l l l l l l l l l l l 7. Do the plant responses agree with the E0P basis?
l l l l 1 1 I I I I I l 8. Are the instrument readings and tolerances stated in the l
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l l l I .l I l E0P consistent with the instrument values displayed on l l l l l the instruments? l l l l l l l l l l .
I l l l l9 Is the' EOP physically compatible with the work situation l l l l l (too bulky to hold, binding would not allow them to lay l l' l l l flat in work space, no place to lay the E0Ps down to l l I l l use)? l'
.I l l I I I I I I l l l .I 110. Are the instrument readings and tolerances specified by 1 l l l l the E0P for remotely located instruments accurate? l
-l l l l l 9817N-REA
EVALUATION CRITERIA FOR EOP ,.
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l - VALIDATION - *l
'l Page 4 of 4 l l II. (PERATIONAL CORRECTNESS. l l l 1 l l Step i l
lSIU No. l B. OPERATOR COWATIBILITY l l 1 1 I
I l l l 1. If time intervals are specified, can the procedJre actioni l ! l l steps be performed on the plant within or at the - 1 i l l I designated time intervals? l l l 1 l l l l l l l l 1 I 2. Can the procedure action steps be performed by the l l l l l operating shift? l l 1 I l l l l l ! l l l l l 3. Specific actions are assigned to individual shift l l l l l personnel, does the E0P adequately aid in the i ,
l .I I I coordination of actions among shift personnel where l l l l 1 necessary? l l l 1 1 l l l l l 1 l l l l 4. Can the operating shift follow the designated action l l l l l step sequences? l l l l l l l l l l l l l EOP OBSERVER / REVIEWER: DATE: 1 1 l l 1 l ADDITIONAL REMARKS SECTION I I I I I I I I I I I I l I i 1 I I I I I i 1 l 2
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MAINE YANKEE 1,
EERGENCY OPERATING PROCEDURES 4
h l TRAINING PROGRAM-O l 4
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PROGRAM DESCRIPTION m
Program
Title:
Emergency Operating' Procedure Training Program Overview:
Maine Yankee integrates Emergency Operating Procedure (EOP) Training into its Operations Qualification and Requalification Courses to provide operations personnel with a coherent performance based program. This document briefly describes the program elements, how they interrelate with each other, and how they fit within their host course.
~ Objective:
The'overall objectives for the E0P Training Program follow:
- 1. To provide M.Y. Operations personnel with the knowledge and skill needed to use the E0Ps effectively to react to off normal and emergency situations.
- 2. To reinforce and_upgrace on~a periodic basis the knowledge and skill needed by M.Y. operations personnel to use the E0Ps effectively to react to off normal and emergency situations.
q Subjects Addressed:
U l. Development Process
- 2. Structure
- 3. Technical Basis
- 4. Technical Content
- 5. Usage Rules
- 6. Critical Safety Function Criteria
- 7. Individual Procedure Strategies
- 8. Use of Safety Parameter Display System (S'DS)
- 9. Team Member Roles and Responsibilities
- 10. Simulator Exercises Program Elements:
A. E0P training consists of elements from the following courses:
- 1. Licensed Operator Requalification Course:
- a. -Licensed Operator EOP knowledge and skills are routinely reinforced and ertanced through:
(1) Annual required document review of EOP's.
(2) Biennial lectures using ECP related lesson materials from Operator Qualification Courses (Appendix 8)
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1284A-ARS
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(3) Simulator exercises using EP related scenarios from Operator Qualification Courses (Appendix B)
(4) Required control manipulations (Appendix A)
- b. Training for the transition from the old to the new E0P set was provided during two blocks of the 1985-1986 Requalification Course. During these blocks licensed -
operators and Nuclear Safety Engineers participated in lectures and simulator training using EOP related materials listed in Appendix B.-
- c. Individual and team performance is evaluated using written examinations and simulator evaluation. t
- 2. Reactor Operator and Senior Reactor Operator Qualification Courses:
Candidates for operator licenses participate in comprehensive
. training courses which include an Emergency and Abnormal Operating Procedures Mo t le..'This mo Wle provides the-candidates with the necessary EOP knowledge and skill through the ECP related lectures '
and simulator exercises listed in Appendix B. Candidate n performance is_ evaluated and recorded using written, . simulator, L) and oral examiantions.
- 3. Auxiliary Operator Qualification Course:
Candidates for non-licen.ad operator positions participate in a comprehensive training course which includes an Emergency and Abnormal Operating Procedure Module. This module provides the candidates with the necessary EP knowledge and skill through the E P related classroom seminars listed in Appendix C. IndiviWal performance is evaluated by written examinations.
- 4. Auxiliary Operator Requalification Course:
- Auxiliary Operator E0P knowledge is reinforced and enhanced on a selective basis using the EP related seminars listed in Appendix C. Individual performance is evaluated by written
, exanination.
- 5. Nuclear Safety Engineer Qualification Course:
Nuclear Safety Engineer Qualification Candidates participate in a comprehensive training course which includes E0P's as part of a Plant Specific Training Component. Candidates gain knowledge and skill through self study and participation in lectures and simulator exercises when available from Licensed Operator ,
i O-V Qualification and Requalification Courses. Candidates are evaluated by written or oral examination and simulator observation.
4 1284A-ARS
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6.- Nuclear Safety Engineer Requalification Course:
' Nuclear Safety Engineer E0P knowledge and skill are routinely reinforced and enhanced through participation in the Licensed Operator Requalification Course lectures and Simulator Training.
Evaluation is provided through written examination and simulator observations.
B. The Maine Yankee plant specific simulator is used extensively in providing E P training. The fidelity of the simulator provides
. sufficient. realism and flexibility that training in-the plant control i- room is unnecesary. - Consequently the benefits of procedure walkthrough training can be gained through simulator activities .
' without risk to actual plant operation. Appendix D summarizes the instructional methods used in the above courses.
I Program Revisions:
~
When icsued, revisions to the E0Ps and their basis documents are A.-
reviewed by the-Operations Training Section Head for impact on training materials. - Training material changes are tracked, documented, and approved through a formal training department revision control process.
B. Operator feedback resulting from EP validation and training is fed
- back to the EP development task force. Changes to the E0Ps or their l
basis documents resulting from this feedback -is incorporated into training as outlined in paragraph A above.
L Program Evaluation:
l I
The effectiveness of each contributing course in this program is evaluated by procedure as required by the Maine Yankee Systematic Approach to Training. The specific methods for each course are listed in their current course description. At a minimum this includes student evaluations, Instructor Course Summary, and an independent post course effectiveness assessment.
Documentation:
Documentation is accomplished in accordance with the M.Y. Training Documentation System. ' At a minimum this includes content, attendance, dates, and results.
L L
O 128AA-ARS L
Page 1 of 2 APPEPOIX A CONTROL MANIPULATIONS The following control manipulations and plant evolutions are acceptable for meeting the reactivity control manipulations required by Appendix A, Paragraph 3.a. of 10 CFR Part 55. The starred (*) items shall be performed on an annual
. basis; all other items shall be performed on.a two-year cycle. 'Those control j manipulations which are not performed at the plant may be performed on a simulator. The use.of the Technical Specifications should be maximized during the simulator-control manipulations. Personnel with senior licenses are credited with these activities if they direct or evaluate control
. manipulations as they are performed.
- (l) Plant or reactor startups to include a range that reactivity feetback from nuclear heat addition is noticeable and heatup rate is established.
(2) Plant shutdown.
- (3) Manual control of steam generator feedwater during startup or shutdown.
(4) Boration and/or dilution during power operation.
- (5) Any significant (greater than 10%) power changes in. manual. rod control.
(6) Any reactor power change of 10% or greater when load change is performed with load limit control.
- (7) Loss of coolant including:
significant steam generator tube leaks a.
- b. inside and outside primary containment
- c. large and small, including leak-rate determination
- d. saturated Reactor Coolant response.
(8) Loss of instrument air.
(9) Loss of electrical power (and/or degraded power sources).
- (10) Loss of core coolant flow / natural circulation.
(11) Loss of condenser vacuum.
(12) Loss of service water.
(13) Loss of shutdown cooling.
1284A-ARS
-Page 2 of 2 O
. C' APPENDIX A
' Continued (14) Loss of component cooling system or cooling to an individual component.
(15) Loss of normal feedwater or normal feedwater system failure.
- (16) Loss of all feedwater (normal and emergency).
(17) Loss of protective system channel.
(18) Mispositioned control rod or rods (or rod drops).
(19) Ir. ability to drive control rods.
(20) Conditions requiring use of emergency boration.
(21) Fuel cladding failure or high activity in reactor coolant.
(22) Turbine or generator trip.
(23) Malfunction of automatic control system (s) sich affect reactivity.
(24) Malfunction of reactor coolant pressure / volume control system.
(25) Reactor trip.
(26) Main steam line break (inside or outside containment).
(27) Nuclear instrumentation failure (s).
O.
U 1284A-ARS
Page 1 of 2 G
Q) APPEPOIX B MODULE 8 EP's/AP's Lessons R0-L-8.1 A0P 2-2 Loss of Vacuum Seminar R0-L-8.2 AP 2-3 HELB Seminar R0-L-8.3 AOP 2-7 Excess Steam Demand Seminar R0-L-8.4 AOP 2-10,17 RCS Leak, Post Accident H2 Control Seminar R0-L-8.5 AOP 2-11 Loss of Instrument Bus Seminar R0-L-8.6 AOP.2-12 Bus 5 or Bus 6 Under Voltage Seminar R0-L-8.7 AOP Loss of Offsite Power Seminar R0-L-8.8 AOP . Emergency Loop Snutdown Seminar R0-L-8.9 A0P 2-16 Partial Loss of Load Seminar R0-L-8.10 AOP. 2-20, 21, 22, 23 RRS Failure, Dropped / Stuck Rod / Reactivity Addition Seminar R0-L-8.11 AOP 2-25 High Radiation Levels Seminar R0-L-8.12 AOP 2-27 High SG Chlorides Seminar q R0-L-8.13 .AOP 2-28, 29 Loss of Compressed Air Seminar Q R0-L-8.14 R0-L-8.15 AOP 2-30 Loss of Containment Integrity Seminar AOP 2-31, 32, 33 Service Water / Component Cooling Failure. Sainar R0-L-8.16 AOP 2-34, 35 Loss of RHR/ Refueling Accidents Seminar R0-L-8.17 AP 2-36, 40, 41, 42, 43 Natural Disasters / Safeguards Ann /CR Isolation Seminar R0-L-8.18 EOP 2-90-0 Plant Fire Assessment Seminar R0-L-8.19 EOP 2-90-1 Plant 50 for fire in CR (station blackout) Seminar R0-L-8.20 AP Control Room Evacuation Seminar R0-L-8.21 E0P 2-90-2 through 7 Plant 50 for various fires Seminar R0-L-8.22- EOP 2-90-0 Plant Cooldown from ASP Seminar R0-L-8.23 Introduction to E0P's R0-L-8.24 E-0 ESOP/SI, ES-0.0 - Event Redlagnosis R0-L-8.25 ES-0.1 - Reactor Trip Response R0-L-8.26 ~ES - 0.2, 0.3, 0.4 - Natural Circulation Cooldown R0-L-8.27 E-1, ECA-1.2 - Loss of Primary or Secondary Coolant R0-L-8.28 ES-1.1 - SI Termination R0-L-8.29 ES-1.2 - Post LOCA Cooldown and Depressurization R0-L-8.30 ES-1.3,1.4, - Recirculation Cooling and H.I R0-L-8.31 E Steam Line Break R0-L-8.32 E-3, ES-3.1, ES-3.3 - SGTR and Cooloown R0-L-8.33 ECA - 0.0, 0.1, 0.2 - Loss of all AC R0-L-8.34 FR-S.1, S.2 - Subcriticality Challenge Recovery R0-L-8.35 FR - C.1, C.2, C.3 - Core Cooling Challenge Recovery R0-L-8.36 FR - H.1, H.2, H.3, H.4, H.S - Heat Sink Challenge Recovery R0-L-8.37 FR - P.1, P.2 - RCS Integrity Challenge Recovery
-O V
R0-L-8.38 R0-L-8.39 FR - I.1,1.2, I.3 - FCS Inventory Challenge Recovery FR - Z.1, Z.2, Z.3 - Containment Challenge Recovery 1284A-ARS
Page 2 of 2 7%
APPEPOIX B ,
Continued MODULE 8 E0P's/AOP's SimJ1ator RO-S-8.1 ESOP or SI and Response (E-0, ES-0.0, ES-0.1, ES-1.1)
R0-S8.2- Natural Circulation Cooldown (ES-0.2, 0.3, 0.4)
R0-S-8.3 Loss of all AC Power and Recovery (ECA-0.0, 0.1, 0.2)
RO-S-8.4 Loss of Primary Coolant (E-1, ECA-1.2, ES-1.2,1.3,1.4)
R0-S-8,5 Loss of Secondary Coolant (E-1, E-2)
RO-S8.6 SGTR and Cooldown (E-3, ES-3.1, 3.3)
R0- S8.7 Subcriticality Challenge Recovery (FR-S.1, S.2)
R0-S-8.8 Core Cooling Challenge Recovery (FR-C.1, C.2, C.3)
R0-S-8.9 Heat Sink Challenge Recovery (FR-H.1, H.2, H.3, H.4, H.5)
RO-S-8.10 RCS Integrity Challenge Recovery (FR-P.1, P.2)
R0-S-8.ll- RCS Inventory Challenge Recovery (FR-I.1, I.2, I.3)
,cg R0-S-8.12 Containment Challenge Recovery (FR-Z.1, Z.2, Z.3)
V RO-S-8.13 R0-S-8.14 Loss of Condenser Vacuum (AOP 2-2)
Excess Steam Demand ( AOP 2-7)
R0-S-8.15 Loss of OP/ PAC or OP/IAC (AOP 2-11)
R0-S-8.16 Misaligned (Oropped) CEA (AOP 2-21)
R0-S-8.17 Inoperable CEA ( AOP 2-23)
R0-S-8.18 High Chlorides (AOP 2-27)
R0-S-8.19 Loss of Control Air (AOP 2-28, 29)
R0-S-8.20 Service Water Header Rupture (ACP 2-31)
R0-S-8.21 Loss of SCC /PCC (AOP 2-32, 33)
R0-S-8.22 Emergency Loop Shutdown O
1284A-ARS
APPEtOIX C
! Auxiliary Operator Qualification Course MODULE 8 Emergency and Abnormal Procedures Lessons A0-L-8.1 AOP 2-2 Seminar (Loss of Vacuum)
A(M.-8.2 AOP 2-3/2-7 Seminar (HELB, Excess Steam Demand)
Ao-L-8.3 AOP 2-10/2-17 Seminar (RCS Leak, H2 Control)
A0-L-8.4 AOP 2-11/2-12 Seminar (Electrical Distribution Abnormalities)
A0-L-8.5 AOP 2-25 Seminar (High Radiation Levels)
A0-L-8.6 AOP 2-27 Seminar (High Chlorides)
Ao-L-8.7 AOP 2-28/2-29 Seminar (Loss of Air)
A0-L-8.8 AOP 2-30 Seminar (Containment Integrity)
A0-L-8.9 AOP 2-31/2-32-2-33/2-34 Seminar (SW/PCC/ SCC /RHR Abnormalities)
A0-L-8.10 A03 2-35 Seminar (Refueling Accidents)
Ao-L-8. ll AOP 2-40/2-41/2-42 Seminar (Natural Disasters)
A0-L-8.12 E-0 Series Emergency Procedure Seminar AO-L-8.13 E-1 Series Errergency ProceWre Seminar 73 Ao-L-8.14 E-2 Series Emergency Procedure Seminar
(
l A0-L-8.15 E-3 Series Emergency Proce ire Seminar Ao-L-8.16 Critical Safety Function Restoration Procedures Seminar A0-L-8.17 Control Room Evacuation Seminar A0-L-8.18 EOP 2-90 Series Emergency Procedures Seminar i
v 1284A-ARS
I
- APPEFOIX D t
Control Simulator Manipu- Self Document Exam./ Record Lecture Training lations Study Review Eval. Keeping LORC X X X X X X SROQC X X X -X t ROQC X X X X !
I l AOFC X X X A0QC X X X NSERC X X X X NSEQC 'X X X X X Abbreviations:
p LORC Licensed Operator Requalification Course Senior Reactor Operator Qualification Course
~
SROQC P0QC. Reactor Operator Qualification Course A0QC Auxiliary Operator Qualification Course AORC . Auxiliary Operator Requalification Course NSERC Nuclear Safety Engineer Requalification Course NSEQC Nuclear Safety Engineer Qualification Course O
1284A-ARS
l s-TECHNICAL DIFFERENCES BEREEN MAINE YANKEE E0P's
&~ AND WESTINGHOUSE ERG's O
MAINE YANKEE E0P STEP DOCUMENTATION FORM INTRODUCTION ,
The E0P step documentation form was developed for the purpose of documenting the technical differences between the WOG ERGS and the Maine Yankee E0Ps.
Recorded on these forms are the technical differences, and the explanations or bases for them. The documentation forms are compiled in order of procedure number, and can be used ef fectively with the following information:
- 1. When there are technical differences between a given E0P and its corresponding ERG, the E0P step number will be listed in the left column of the form. The step number of the respective ERG step will be listed in the center column, and the explanation or basis of the difference will be given in the right column.
- 2. A copy of the ERGS and E0Ps must be used in conjunction with step documentation forms as a step text is not presented on the forms.
- 3. For E0P steps not listed on the forms, there are no technical differences from the ERG steps. Therefore, no explanation or bases are needed.
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'a l 4292e:1d/021086 1
l MAINE YANKEE E0P STEP DOCUMENTATION FORM E-0 EMERGENCY SHUTDOWN FROM POWER OR t SAFETY INJECTION Maine Yankee ERG ,
, Sten No. SteD No. Explanation or Basis for Difference 1st NOTE before -
A note was added to implement the emer-Step 1 gency plan as necessary with respect to the E0Ps.
1 -
Step was added to manually trip the reactor and turbine to ensure that all i reactor and turbine trip functions have been initiated.
- o- 5 Step deleted since M.Y. does not have an automatic FW isolation on 51 actuation.
13 & 14 9 The ERG step was broken down into two different steps to address the PCC &
SCC pumps separate'y.
CAUTION before -
A caution was added to alert the opera-Step 17 tor that a rapid restoration of feed-water may cause waterhammer.
17 16 The step was changed since main FW could be available to provide an adequate heat sink since main FW is not isolated on 51.
O '
V 4292e:1d/021086 2 !
l l
.g E-0 (Continued)
Sten No. SteD No. Explanation or 8 asis for Difference 18 7 & 17 The 2 ERG steps were cortined to address the specific-M. Y. des,1gn.
NOTE before The note to maintain seal injection was Step 21 deleted since for Maine Yankee, seal water supply flow is isolated on SIAS and PCC flow is maintained.
CAUTION before Maine Yankee utilizes a DWST and a CST, Step 28 each with a large' volume of water.
Alternate water sources are not a concern.
CAUTION before Caution was deleted because SI will Step 32 actuate upon AC power restoration if required.
34 -
A step was added to reset CSAS if required. However, it is unlikely that the signal would be generated unless containment pressure reached 20 psig.
36 The step to stop the diesel generators was deleted since for Maine Yankee, the DGs do not receive an AUTO start signal on an SIAS.
O U
4292e:1d/021186 3
N l ?
. l MAINE YANKEE E0P STEP DOCUMENTATION FORM 4
ES-0.0 EVENT REDIAGNOSIS s
Maine Yankee ERG i Sten No. Steo No. Explanation or Basis for Dif ference l
1
- l j General. - General A flow chart f ormat was used to direct 1 ~the operator through rediagnosis rather than the ERG procedure format.-
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- 4292e:1d/021086 4 i
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MAINE. YANKEE E0P STEP DOCUMENTATION FORM ES-0.1 REACTOR TRIP RESPONSE ,
Maine Yankee ERG Sten No. Steo No. Explanation or Basis for Difference CAUTION before -
A caution was added to alert the opera-Step 2 tor that rapid restoration of feedwater may cause water hammer.
NOTE before -
A note was added to remind the operator Step 4 that makeup to the VCT may be necessary in subsequent actions to maintain level greater than 20%.
CAUTION before -
A caution was added to guard against Step 5 possible thennal stresses on the spray nozzle if Pzr. Aux. spray is established too rapidly.
NOTE before -
A note was added to remind the operator Step 5 that Pzr. pressure will be between 1800 PSIG and 2260 PSIG following a reactor trip.
CAUTION before -
A caution was added to prevent opening Step 7 the main generator output breakers prior to the required time for dutomatic transfer to occur.
8 The ERG step was deleted because the M. Y. steam dump system is always in an AUTOMATIC control mode.
4292e:ld/021186 5
. . . _ ~
b ES-0.1 (Continued) j O
1 i
Sten No. Step No. Explanation or Basis for Difference i
NOTE before The RTDs are mounted directly in RC [
Step 9 piping on the SG side of the LSVs and [
not in the hot of cold leg RTO bypass manifolds, i
10 ERG step deleted because wide range log i scale automatically transfers to the source range on decreasing flux.
11 -
A step was added to check Pzr. level as an indication that the PLCS is functioning properly to restore RCS inventory.
12 -
A step was added to align secondary plant equipment following a reactor trip.
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O 4292e:1d/021086 6 1
J
MAINE YANKEE E0P STEP DOCUMENTATION FORM O
ES-0.2 NATURAL CIRCULATION C00LOOWN Maine Yankee ERG SteD No. Step No. Explanation or Basis for Difference CAUTION before Maine Yankee has a DWST and a CST, each Step 6 with a large volume. Alternate water sources are not a concern.
6 -
Maine Yankee wants to have loop stop valves available for loop maintenance af ter a safe shutdown condition is established.
7 -
Maine Yanke,s stops chem feed to SGs L prior to cooldown to prevent corrosion conditions.
8 - Maine Yankee starts warmup of aux.
boilers to minimize thermal stresses.
7 The ERG step was deleted since Maine Yankee maintains RCS temperature and Pzr. pressure within restrictive bounds of Technical Data Book Fig. 1.2.3.1 when blocking SI.
NOTE before -
If required, charging flow is throttled Step-10 to increase flow to spray system to provide adequate spray flow.
O 4292e:1d/021086 7
ES-0.2 (Continued)
. f~h)
' Maine Yankee ERG Step No. Sten No. Explanation or Basis foi Difference 12 -
Maine Yankee has containment spray actuation (CSAS) in addition to SIAS and CIS. CSAS is blocked to prevent spurious CSAS actuation during cooldown/depressurization.
13 -
Maine Yankee has automatic recirc.
actuation (RAS). RAS is blocked to prevent spurious RAS actuation during cooldown/depressurization.
CAUTION before -
A caution was added to warn of possible q Step 16 spray nozzle damage if Pzr. Aux. spray C' is initiated too rapidly.
16 12 Maine Yankee uses the Reactor Vessel l
Head Thermowell to ensure sufficient subcooling to prevent upper head voiding.
17 -
A step was added to bypass the SG low pressure trip to prevent isolation of l
the SGs during depressurization.
18, 24 - Maine Yankee puts 1 of 2 HPS1 pumps in l
PULL-TO-LOCK to lessen severity of cold l overpressure event due to spurious SI.
Flow restrictor used on second HPS!
pump.
l l 4292e:1d/021086 8 i
t,
ES-0.2 (Continued)
Sten No. SteD No. Explanation or Basis for Dif ference 24, 18 15 The ERG step was deleted because the SITS at Maine Yankee inject at a pressure which will not be reached in this procedure. The SITS will be locked out in OP 1-7.
Caution before -
Maine Yankee has VPSR setpoints which Step 23 must be reset to prevent a low tempera-ture overpressure event.
23 - A step was added in order to meet VPSR setpoint requirements during a cooldown/depressurization to prevent low temperature overpressure event.
25 -
A step was added to reduce the VPSR setpoint in order to prevent an LTOP condition.
26 -
A step was added to prevent
" fluttering" of main steam NRV with very low steam flow rates.
21 Maine Yankee combined this step with the caution before step 27.
Note before - Note added to inform operator that ini-Step 28 tial conditions for RHR recirculation have been satisfied by this procedure before he is transferred to OP l-7.
4292e:1d/021086 9
MAINE YANKEE E0P STEP DOCUMENTATION FORM ES-0.3 NATURAL CIRCULATION C00LDOWN -
WITH STEAM VOIDS IN VESSEL (WITH PITS)
Maine Yankee ERG Sten No. Step No. Explanation or Basis for Difference 1 -
Maine Yankee has E0Ps for nat. circu-lation cooldown with void in vessel -
both with and without PITS. If PITS is not available, this step transfers the operator to the E0P without PITS.
4 3 Maine Yankee uses Reactor Vessel Head Thermowell to monitor upper head subcooling.
6 5 Maine Yankee PITS performs same function as RVLIS in this step (i.e.,
ensures any upper head void is above top of hot legs).
12, 8 6 The ERG step was deleted because the SITS at Maine Yankee inject at a pressure which will not be reached in this procedure. The SITS will be l locked out in OP 1-7.
7 - A step was added to bypass the SG low
{ pressure trip to prevent isolation of SGs during depressurization.
O i l
4292e:1d/021186 10 l l
ES-0.3 (Continued)
O Maine Yankee ERG Step No. SteD No. ExDianation or Basis for Dif ference 8, 12 -
Maine Yankee puts 1 of 2 HPSI pumps in PULL-TO-LOCK to lessen severity of overpressure event due to spurious SI.
Flow restrictor used on second HPSI pump.
CAUTION before -
Maine Yankee has VPSR setpoints which Step 11 must be reset to prevent a low temperature overpressure event.
11 -
A step was added in order to meet VPSR setpoint requirements during a cooldown/depressurization to prevent O
low temperature overpressure event.
13 -
Maine Yankee must meet VPSR setpoints to prevent a low temperature overpressure event.
14 -
A step was added to prevent
" fluttering" of main steam NRV with very low steam flow rates.
12 Maine Yankee combined this step with the caution before step 15.
Note before -
Note added to inform operator that ini-Step 16 tial conditions for RHR recirculation have been satisfied by this procedure before he is transferred to OP 1-7.
4292e:1d/021086 11
MAINE YANKEE E0P STEP DOCUMENTATION FORM ES-0.4 NATURAL CIRCULATION C00LDOWN WITH STEAM VOIDS IN VESSEL (WITHOUT PITS)
Maine Yankee ERG Sten No. SteD No. ExDlanation or Basis for Difference 3, 11, 17 3, 5, 11, 15 Temperatures in step-wise cooldown/
depressurization for Maine Yankee are set to meet VPSR setpoint requirements.
4 Initial depressurization step covered in ES-0.2 prior to transfer to ES-0.3 (or ES-0.4) at Maine Yankee.
O 6, 13. 19, 22 8, 13, 17 Pressures in step-wise cooldown/
depressurization for Maine Yankee are set to meet VPSR setpoint requirements and to place HPSI flow restrictor in service.
8, 21 -
Maine Yankee puts 1 of 2 HPSI pumps in PULL-TO-LOCK to lessen severity of overpressure event due to spurious SI.
Flow restrictor used on second HPSI pump.
l l 9 -
A step was added to bypass the SG low pressure trip to prevent isolation of l
SGs during depressurization.
O 4292e:1d/021086 12
--. ._ . -- _. ~
ES-0.4 (Continued)
O Maine Yankee ERG Sten No. Sten No. ExplanationorBasisforhifference CAUTION before -
Maine Yankee has VPSR setpoints which Step 10,10,15 must be met to prevent a low pressure overtemperature event.
16 -
Maine Yankee added step to prevent
" fluttering" of main steam NRV with very low steam flow rates.
- 21, 8 10 The ERG step was deleted because the SITS at Maine Yankee inject at a pressure which will not be reached in this procedure. The SITS will be locked out in OP 1-7.
22 Maine Yankee combined this step with the caution before step 23.
Note before 24 - Note added to inform operator that initial conditions for RHR recircula-tion have been satisfied by this procedure before he is transferred to OP 1-7.
O 4292e:1d/021186 13
MAINE YANKEE E0P STEP DOCUMENTATION FORM
[ 't V ECA-0.0
^
LOSS OF All. AC POWER w
Maine Yankee ERS Step No. Stec No. Explanation or Basis for Difference IST NOTE before -
A note was added to implement the Emer-Step 1 gency Plan as necessary with respect to the E0Ps.
1 -
Step was added to manually trip the reactor and turbine to ensure that all reactor and turbine trip functions have been 'mitiated.
3 2 Fo11 wing a loss of AC power, the only k indication of a turbine trip is to ens'ure the EFCVs are closed. The step was rewritten accordingly.
2nd CAUTION Tnc caution was deleted because Maine ,
before Step 6 Yankee blocks the SI signal prior to actua'tlon.
N0lt before - >
A note was addtid to provide information Step 9 '
to the operator regarding the possibility of thermal / hydraulic shock upon restoration of control air.
16 -
A step was added to stop gland steam since the EFCVs are closed and to break condenser vacuum.
!O 4292e:1d/021086 14
ECA-0.0 (Continued)
O O
Maine Yankee ERG SteD No. SteD No. ExDlanation or Basis for Difference 18 -
Step was added to prevent isolation of !
the intack SG(s) on a low pressure signal during depressurization~.
21 18 Maine Yankee blocks the SI signal prior to actuation rather than allowing the signal to come in and reset it.
27 -
A step was added to ensure BUS 7 and l BUS 8 are energized because there are no voltage indications on these two busses.
2nd CAUTION -
A caution was added to isolate PCC before Step 29 return flow prior to starting a PCC pump to prevent damage to the RCP seals.
26 Step deleted because the diesel generator cooling concern was covered in the previous step by loading one PCC and SCC pump.
4292e:Id/021086 15
MAINE YANKEE E0P STEP DOCUMENTATION FORM (h
ECA-0.1 LOSS OF ALL AC POWER RECOVERY,
~
SI NOT REQUIRED Maine Yankee ERG Sten No. _
Step No. ExDlanation or Basis for Difference 2 2 The strategy to block SI (ECA-0.0, Step
- 21) should also prevent a CIS signal.
The step was reworded to ensure the CIS j switches are reset. '
3 -
The check on containment control air was perfortned in the ERGS in Step 2b RNO. Due to the restructuring of Step
~ A 2 in the E0Ps, the containment control
-- air warranted it own step to verify an adequate air supply is available.
1st CAUTION -
A caution was added to prevent starting before Step 4 a charging /HPSI pump without adequate cooling water.
2nd CAUTION -
A caution was added to prevent over-before Step 4 loading the diesel generators.
1st CAUTION Maine Yankee uses a DWST and a CST, before Step 7 each with a large volume of water.
Alternate water sources are not a concern.
O 4292e:1d/021086 16 4
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ECA-0.1 (Continued) :
O Maine Yankee ERG Sten No. Sten No. Explanation or Basis for Dif ference
~2nd CAUTION and The caution and note were deleted NOTE before because Maine Yankee prefers to run the
$tep 7 motor-driven EFW pumps.
CAUTION before -
A caution was added to alert the opera-Step 7 tor that a rapid restoration of feedwater flow may cause waterhammer.
t CAUTION before The RTDs are :nounted directly in RC Step 15 piping on the SG side of the LSVs and not in the hot or cold leg RTD bypass manifolds.
I \
16 The ERG step was deleted because wide range log scales automatically transfer ,
to the source range or decreasing flux.
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4292e:1d/021086 17
MAINE YANKEE E0P STEP DOCUMENTATION FORM ex ECA-0.2 LOSS OF ALL AC POWER RECOVERY, SI REQUIRED Maine Yankee ERG
' Step No. SteD 40. ExDlanation or Basis for Difference 1 -
A step was added to ensure that an air compressor was loaded on the restored ac bus for control of air operated equipment.
3 ERG step was deleted because the PCC pump was started in ECA-0.0.
4 -
A caution was added to prevent overloading the diesel generators.
5 -
A check on EFW status was made into a step because, regardless of whether the plant is in the nonnal injection alignment or the RAS alignment, an EFW pump running is required.
6 -
Regardless of plant alignment, the containment recire, fans are checked running to restore.the containment. ;
environmental conditions.
1st CAUTION Maine Yankee utilizes a DWST and a CST, before Step 6 each with a large volume of water.
Alternate water sources are not a concern.
4292e:ld /021086 18
ECA-0.2 (Continued)
O Maine Yankee ERG Steo No. Sten No. Explanation or B' asis for Difference 2nd CAUTION The caution and note were deleted and NOTE before because Maine Yankee prefers to run the Step 6 motor driven EFW pumps.
CAUTION before -
A caution was added to' alert the Step 8 operator that a rapid restoration of feedwater flow may cause waterhansner.
7 The ERG step was deleted because for Maine Yankee, the containment spray pumps will be running, however, the spray header isolation valves will be closed.
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MAINE YANKEE E0P STEP DOCUMENTATION FORM p .
t]
E-1 LOSS OF PRIMARY OR SECONDARY COOLANT .
Maine Yankee ERG SteD No. SteD No. ExDlanation or Basis for Dif ference 2nd NOTE before The note to maintain seal injection was Step 1 deleted since for Maine Yankee, seal water supply flow is isolated on SI and PCC flow is maintained.
1 1 1 The logic of the step was rever:.ed per utility request.
CAUTION before -
A caution was added to alert the opera-Step 3 tor that rapid restoration of feedwater may cause waterhansner.
CAUTION before -Maine Yankee utilizes a DWST~and a CST, Step 3 each with a large volume of water.
Alternate water sources are not a concern.
CAUTION before -
If containment spray is not required, Step 8 the valves are closed in Step 9. The note states that the pumps should remain running to provide water to the-suction of the HPSI pumps if necessary.
1st CAUTION Caution was deleted because SI will before Step 9 actuate upon AC power restoration if
~
required.
4292e:ld/021086 20
,.. . . . . - - . . - . . _ - . . . -- , - . . - . - - . - = - - . - - . - - - . __~
4 I
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E-1 (Continued) 2
Step No. Step No. Explanation or Basis for Difference i
4 11 The step to stop the diesel generators !
! was deleted since for Maine Yankee, the i DGs do not receive an AUTO start signal ;
l on SI. :
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4292e:1d/021086 21
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MAINE YANKEE E0P STEP DOCUMENTATION FORM
. ,-~
ES-1.1 SI TERMINATION Maine Yankee ERG SteD No. Sten No. Explanation or Basis for Dif ference None Caution.before Caution was deleted because SI will Step 1 actuate upon AC power restoration if required.
1 -
Maine Yankee has the capability of throttling HPSI flow whereas the ERG reference plant can not. This step replaces SI reduction sequence steps.
S 2 -
Necessary Controlability Check After HPSI Flow has been throttled.
4' 2 Maine Yankee Step is equivalent to Phase A and Phase B isolation.
4 5 -
CSAS is coincident with SIAS. Reset allowed when SIAS is cleared in Haine~
Yankee Step 3.
5 This is a controlability check that was performed in Maine Yankee Step 2.
9 7 Maine Yankee does not have a BIT, so stopping HPSI flow effectively stops boration. i
(
4292e:ld/021086 22
ES-1.1 (Continued)
Maine Yankee ERG SteD No. SteD No. ExDlanation or Basis for Difference 10 8 The Pzr. Level is controlled automatically for Maine Yankee as opposed to manually in the ERGS by controlling charging flow.
9 Maine Yankee has no high-head SI pumps.
15 -
A step was added to reestablish seal water supply flow which was isolated on SIAS.
18 .- Containment spray pumps automatically
'~
started on SIAS. If no continued requirement to run them, they should be stopped.
24 16 Restructuring of steps. Intent remains ths same as the ERGS.
17 The ERG step was deleted because the Maine Yankee steam dump system is always in an AUTOMATIC control mode.
Caution before -
To alert operator that rapid actuation Step 20 of the Pzr. Aux. Sprays may cause thermal stresses on the nozzle.
l v
i 4292e:ld/021086 23 !
1
ES-1.1 (Continued)
N Maine Yankee ERG Sten No. Sten No. Explanation or Basis for Difference Caution before -
A caution was added to alert the opera-Step 21 tor that rapid restoration of feedwater may cause water hansner.
l Caution before Maine Yankee has a DWST and a CST, each Step 20 with a .large volume of water. Alter-nate water sources are not a concern.
22, 23 21 The ERG step was broken down into two steps for RCP cooling. The intent l
remains the same.
'"N -
2nd Caution The RTDs are mounted directly in RC before Step 23 Piping on the SG side of LSVs and not in Hot or Cold Leg RTD Bypass.
24 ERG step deleted because wide range log scale automatically transfers to the source range on decreasing flux.
25 The step to stop the diesel generators was deleted since for Maine Yankee, the DGs do not receive an AUTO start signal on an SI.
30 -
To properly align the secondary equipment for further recovery actions.
A U
' 4292e:ld/021086 24
MAINE YANKEE E0P STEP DOCUMENTATION FORM
_(-
ES-1.2 '
l POST LOCA C00LDOWN AND DEPRESSURIZATION l Maine Yankee ERG Step No. SteD No. Explanation or Basis for Difference 2nd CAUTION Caution was deleted because SI will before Step 1 actuate upon AC power restoration if required.
3 -
A step was added to reset CSAS since the signal is coincident with SIAS.
The reset is allowed when SIAS is cleared in Step 1.
5, 6 4, 5 The order of the two steps was reversed because the LPSI pumps can be stopped quickly, thus reducing the load on the energized busses.
CAUTION before -
A caution was added to alert the opera-Step 7 tor that a rapid restoration of feed-water may cause waterhansner.
CAUTION before Maine Yankee utilizes a DWST and a CST, Step 6 each with a large volume of water.
Alternate water sources are not a concern.
2nd NOTE before Note was deleted because Maine Yankee Step 7 does not have a low steamline pressure SI signal.
O 4292e:ld/021086 25
ES-1.2 (Continued) ,
. Maine Yankee ERG Sten No. Sten No. Explanation or Basis for Difference 2nd CAUTION The RTDs are mounted directly in the RC before S'ep 12 piping on the SG side of the LSVs and not in the hot or cold leg RTD bypass manifolds.
14, 15 13, 14, 15 Maine Yankee has the capability to throttle HPSI flow until makeup can be supplied by normal charging. Thus the SI' reduction sequence of the ERGS does not apply.
17 17 Maine Yankee does not have a BIT.
Stopping HPSI flow effectively stops boration.
18 18 The Pzr. level is controlled automati-cally for Maine Yankee, as opposed to manually in the ERGS, by controlling charging flow.
2nd CAUTION The RTDs are mounted directly in the RC before Step 19 piping on the SG side of the LSVs and not in the hot or cold leg RTD bypass manifolds.
J 20, 21 20, 21 ERG step sequencing allows steps to be interchanged.
O 4292e:ld/021086 26
ES-1.2 (Continued)
A U
Maine Yankee ERG Sten No. Step No. Explanation or Basis for Difference 24 The step to stop the diesel generators was deleted since for Maine. Yankee, the.
DGs do not receive an AUTO start signal on an SIAS.
24, 25 25 The ERG step was broken down into two steps for RCP cooling. The intent remains the same. '
27 ERG step deleted because wide range log scale automatically transfers to the source range on decreasing flux.
~30 -
A step was added to properly align the secondary equipment for further recovery actions.
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V 4292e:ld/021086 27
MAINE YANKEE E0P STEP DOCUMENTATION FORM !
1 0_ ES-1.3 TRANSFER TO RECIRCULATION COOLING (RAS) ,
Maine Yankee ERG Sten No. SteD No. Explanation or Basis for Difference General General The ERG guidance showing typical tasks pertaining to the transfer to cnid leg recirculation does not follow the Maine Yankee sequence. The transfer to RAS was written accordingly.
O r
O 4292e:1d/021086 28
MAINE YANKEE E0P STEP DOCUMENTATION FORM O ES-1.4 ESTABLISHING HOT LEG. INJECTION 1
__ Sten No. Sten No. Explanation or Basis for Difference
- General General The ERG guidance showing typical tasks pertaining to the transfer to hot leg injection does not apply to the Maine
' Yankee design. The sequence for establishing hot leg injection is given.
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4292e:ld/021086 29
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. MAINE YANKEE E0P STEP DOCUMENTATION FORM l
ECA-1.2 LOCA OUTSIDE CONTAI'NMENT f
1 - ,
! - Step No. SteD No. ExDlanation or Basis for Difference ,
1
{
No deviations from ERGS. .
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i 4292e:1d/021086 30 .
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i MAINE YANKEE E0P STEP DOCUMENTATION FORM
, i 4 E-2
< . STEAM LINE BREAK-r 4
~
l Sten No. Step No. Explanation or Basis for Difference .
l !
2 The ERG step was deleted since it ;
4 served as a transition to ECA-2.1.
- This was not part of the E0P procedure package.
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4292e:1d/021096 31 i
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l MAINE YANKEE E0P STEP DOCUMENTATION FORM t
E-3 STEAM GENERATOR TUBE RUPTURE .
Maine Yankee ERG Sten No. Sten No. Explanation or Basis for Difference 3rd NOTE before The note to maintain seal injection was Step 1 deleted because, for Maine Yankee, seal water supply flow is isolated on SIAS and PCC flow is maintained.
3rd NOTE before -
A note was added to show Maine Yankee Step.3 preference to stop RCP in the ruptured loop if another RCP is running that provides spray. Analyses shows that there is not significant difference if RCP running is in the ruptured loop or non-ruptured loop.
CAUTION before Maine Yankee utilizes a DWST and a CST, Step 7 each with a'large volume of water.
Alternate water sources are not a concern.
7 - A step was added since main feedwater is not isolated on SIAs for Maine Yankee.
CAUTION before - A caution was added to alert the opera-Step 8 tor that a rapid restoration of feed-water may cause waterhammer.
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4292e:ld/021086 32
E-3 (Continued)
(
~3 V
Maine Yankee ERG Sten Nos Step No. Explanation or Basis fof Difference NOTE.before -
A note was added to inform operator that Step 8 EFW or AFW is the preferred feed flow source to minimize radioactive contaminants.
CAUTION before Caution was deleted because SI will Step 8 actuate upon AC power restoration if required.
12, 13 11, 12 The order of the two steps was reversed because the LPSI pumps can be stopped quickly, thus reducing the load on the i energized busses.
14 -
Step was added to prevent isolation of the intact SG(s) on a' low pressure signal during depressurization.
13 The ERG step was deleted since it served as a transition to ECA-3.1.
This was not part of the E0P procedure package.
NOTE before The note was deleted because Maine Step 14 Yankee does not have the low steamline pressure SI logic.
4292e:1d/021086 33
,s E-3 (Continued)
/ \
SteD No. SteD No. ExDlanation or Basis for Difference 16 -
A step was added because Maine Yankee has the capability to throttle HPSI flow to aid in the prevention of Pzr.
overfill during the cooldown.
15, 16 THe ERG steps were deleted because they served as a transition to ECA-3.1.
These were not part of the E0P procedure package.
2nd CAUTION -
A caution was added to alert the opera-before Step 17 tor that the bounds of Technical Data Book Figure 1.2.3.1 must be maintained during the depressurization.
19 The ERG transition to ECA-3.1 was 1 deleted because ECA-3.1 was not part of 1 the E0P procedure package.
23 23 Maine Yankee does not have a BIT.
Stopping HPSI flow effectively stops 1 boration.
24 24 The Pzr. level is controlled automati-cally for Maine Yankee, as opposed to l manually in the ERGS, by controlling l d
charging flow.
4292e:ld/021186 34 1
E-3 (Continued)
C)
SteD No. SteD No. Explanation or Basis for Difference 29 -
. Step was added to check containment spray pumps because pumps are started on SIAS. If there is no continued requirement to run them, they should be stopped.
30 The step to stop the diesel generators ,
was deleted, since for Maine Yankee, l
the DGs do not receive an AUTO start signal on SIAS.
32 -
A step was added to warmup the Aux.
boilers to minimize thermal stresses.
34,-35 33 The ERG step for. RCP cooling was broken down into two separate steps. The
, intent remains the same.
2nd CAUTION The RTDs are mounted directly in the RC before Step 35 piping on the SG side of the LSVs and not in the hot or cold leg RTD bypass manifolds.
36 ERG step deleted because wide range log scale automatically transfers to the source range on decreasing flux.
39 38 'The transition to ES-3.2 was deleted since it is not part of the E0P procedure package.
4292e:ld/021186 35
1 1
MAINE YANKEE E0P STEP DOCUMENTATION FORM
/N
\n. 1 ES-3.3 l POST-SGTR COOLDDWN USING STEAM DUMP ,
Maine Yankee ERG Sten No. Sten No. Explanation or Basis for Difference 2 Step deleted because the SIT injection pressure is 230 PSIG. This procedure only depressurizes to 350 PSIG.
f
, CAUTION before -
A caution was added to alert the opera-Step 2 tor that additional boration may be required due to backfill from the ruptured SG.
CAUTION before -
A caution was added to alert the opera-
/L.
)T Step 3 tor that a rspid restoration of feed-water may cause waterhammer.
CAUTION before Maine Yankee utilizes a DWST and a CST, Step 4 each with a large volume of water.
Alternate water sources are not a Concern.
6 -
A step was added to place one HPSI pump in PULL-TO-LOCK to reduce the potential for low temperature overpressure condition.
CAUTION before -
Maine Yankee has VPSR setpoints which Step 7 must be reset to prevent a low tempera-ture overpressure condition during
() cooldown.
'4292e:ld/021086 36 1
ES-3.3 (Continued) v Maine Yankee ERG Steo No. SteD No. ExDlanation or Basis for Difference 7 -
A step was added in order to meet VPSR setpoint requirements during a cooldown/depressurization to prevent an L10P condition.
10 -
A step was added to check if the SG low pressure trip should be bypassed to prevent SG isolation.
NOTE before The ERG note was placed before Step 24 Step 9 since it pertains to placing the RHR system in service.
14 -
A step was added to reduce the VPSR setpoint in order to prevent an LTOP condition.
15 -
A step was added to prevent "flut-tering" of the main steam NRV with very low steam flow rates.
16-22 -
Steps were added to further cooldown and depressurize to RHR entry condi-tions while maintaining the limits of the TDB curve. These steps follow the sequence of the ERG cooldown/
depressurization steps.
4292e:1d/021086 37 l
l
. . _ . _ _ - _ = - _ _ - _ . . .. .. ._ - .. - - -
ES-3.3 (Continued)
Maine Yankee ERG SteD No. Sten No. Explanation or Basis fof Difference CAUTION before 15 The ERG step was made into a caution Step 24 prior to exiting procedure.
2nd NOTE before -
A note was added to inform the operator Step 24 that initial conditions for RHR recir-culation have been satisfied by this procedure before the transition to OP~
1-7 is made.
14, 16 The Maine Yankee procedure did not~
provide guidance down to cold shutdown condition as the ERGS did. Once RHR entry conditions are established, Os OP1-7isusedtoreachcoldshutdowg{.
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4292e:1d/021086 38 e
MAINE YANKEE E0P STEP DOCUMENTATION FORM s,
FR-S.1 NUCLEAR POWER GENERATION /ATWS Maine Yankee ERG Sten No. SteD No. Explanation or Basis for Difference 1 -
Step was added to manually trip the reactor and turbine to ensure that all reactor and turbine trip functions have been initiated.
CAUTION before - A caution was added to alert-the opera-
. Step ~4 tor that a rapid restoration of feed-water may cause~waterhanner.
4 3 The check on adequate feedwater supply was changed to include main feed since main feedwater is not isolated on SIAS.
5 - A check for suberiticality was added before emergency boration is initiated. This is also an ERG maintenance item.
8 6 The step was changed to read total FW flow because main feedwater is not isolated on SIAS.
CAUTION before 12 The RNO column of the ERGS was reworded Step 14 and formed into a caution per utility request.
O 4292e:ld/021086 39
I.
i j FR-S.1 (Continue 1) t Maine Yankee ERG '
t- Sten No. Sten No. ,_fxplanation or Basis for Difference
!~ 2nd NOTE before -
A note was added to ensure that the
, Step 15 charging pumps are operating properly for boration.
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. . - . - - .-.. . ._ . .. - . = _ . . . . _ . _ _ _ . -
MAINE YANKEE E0P STEP DOCUMENTATION FORM j T FR-S.2 LOSS OF SHUTDOWN MARGIN Naine Yankee ERG
__ SteD No. Step No. Explanation or Basis for Difference
]
i General General Maine Yankee desired that this 1 procedure be restructured for clarity. ,
i However, the overall intent of respondhj to an abnormal flux condition following l
a trip or an SIAS remains the same.
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O l 4292e:ld/021086 41
,:,) ,
s MAINE YANKEE E0P STEP DOCUMENTATION FORM
~
FR-C.1 INADEQUATE CORE COOLING ,
Step No. Step No. Explanation or Basis for Difference 2nd NOTE before 3 The step to establish RCP support condi-Step i tions was moved to the note at the front of the procedure so the operator was aware of the importance of the step.
6 Maine Yankee does not have a RVLIS full range indication.
7 ERG step deleted because the PITS
() '
s system does not give a reliable level indication in checking if core cooling
- is restored. '
ist CAUTION .
A caution was added to alert the opera-before Step 6 tor that a rapid restoration of feed-water may cause waterhammer.
1st CAUTION Alternate sources for AFW pumps is not before Step 9 a concern because of the large volume of the CST and the DWST as backup.
13 16 The RVLIS check for core cooling recovery was deleted because PITS only shows a level trend.
20 23 Maine Yankee does not have a RVLIS full range indication.
~ '
4292e:ld/021086 4?
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i
,_ s MAINE YANKEE E0P STEP DOCUMENTATION FORM
/'%)\
FR-C.2 DEGRADED CORE COOLING .
Maine Yankee ERG SteD No. Sten No. Explanation or Basis for Difference 4 NOTE before -
A note was added to monitor LPSI pump Step 1 bearing temperatures to prevent pump damage due to lack of component cooling.
'S Maine Yankee does not have a dynamic RVLIS indication.
5 6 .The "Go to Step B" was deleted because if an RCP is stopped, the affect on core cooling is checked since there is no PITS indication.
6 7 The RVLIS full range indication was deleted.
1st CAUTION -
A caution was added to alert the before Step 8 operator that a rapid restoration of feedwater may cause waterhammer.
1st CAUTION Alternate sources for AFW pumps is not before Step 9 a concern because of the large volume of the CST and the DWST as backup.
17 18 Maine Yankee does not have a RVLIS full range indication.
O 4292e:1d/021186 43
. _ . _ _ - . . . - . . . . - . _ . _ . . . . . _ _ _ . _ _ _ _ _ __._..-._---_...-.___..m_..
MAINE YANKEE E0P STEP DOCUMENTATION FORM O FR-C.3 SATURATED CORE COOLING ,
Maine Yankee- ERG SteD No. Step No. Explanation or Basis for Difference 1
NOTE before The note was deleted since it served as Step 1 a transition to ECA-3.2. ECA-3.2 was 4 -
not part of the E0P package.
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4292e:1d/021086 44 :
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(. . _ . , , , _ . . . . . . _ - , . _ . . , _ . _ _ . . . _ . .
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l MAINE YANKEE E0P STEP DOCUMENTATION FORM l
FR-H.1 LOSS OF SECONDARY HEAT SINK Maine Yankee ERG
-SteD No. SteD No. ExDlanation or Basis for Difference 1st CAUTION The note was deleted since it served as before Step 1 a transition to ECA-2.1. ECA-2.1 is not part of the E0P package.
1 2nd CAUTION The check on the requirement of bleed before Step 1 and feed was rewritten as a step, rather than a note.
s -
4 This ERG step only applies to plants ~
, 'with intermediate high-head SI pumps.
CAUTION before Caution was deleted because SI will Step 5 actuate upon AC power restoration if required.
22, 23, 24 12, 13, 14 Bleed and feed should be done as quickly as possible. The three steps were moved to the end of the procedure because the Pzr. PORVs are solenoid operated and do not need instrument air established.
18 21, 22, 23 and The SI stop sequence of the ERGS does NOTE before not apply because Maine Yankee has the Step 24 capability to throttle HPSI flow.
4292e:ld/021186 45
FR-H.1 (Continued)
Steo No. Sten No. Explanation or Basis for Difference 19 -
Following bleed and feed,'.the operator
, ensures all bleed paths are isolated.
Here, the RCS vent paths are closed.
This is also an ERG maintenance item.
21 -
A step was added to determine if throttling HPSI flow was successful in order for normal charging flow to be restored.
26 26 Maine Yankee does not have a BIT.
Stopping HPSI flow effectively stops-boration.
27 This ERG step was included in Step 20 for Maine Yankee. A loop was~ built in to the procedure to ensure all PORVs were closad prior to exiting the bleed and feed ioop.
27 28 Pzr. level h controlled automatically for Maine Yankee, as opposed to manually in the ERGS by controlling charging flow.
O 4292e:ld/021086 46
MAINE. YANKEE E0P STEP DOCUMENTATION FORM ,
FR-H.2 STEAM GENERATOR OVERPRESSURE Maine Yankee ERG Step No. Step No. Explanation or Basis for Difference
-2 2 FW isolation is not automatic on an SIAS for Maine Yankee. The step was changed to ensure manual' isolation.
4, 8 4, 8 Per Maine Yankee request, the condenser steam dump was added as an option to dump steam even though it is unlikely that it will be available.
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4292e:1d/021086 47
MAINE YANKEE E0P STEP DOCUMENTATION. FORM FR-H.3 STEAM GENERATOR HIGH LEVEL Maine Yankee ERG Sten No. SteD No. Explanation or Basis for Difference 2 2, 3 Main FW is not isolated on the SI-signal for Maine Yankee. The two ERG steps were combined to manually isolate both main feedwater and AFW.
4, 5 5, 6 The order of the two ERG steps was switched because closing the atmos-pheric steam supply valves is a local action which requires-dispatch of an operator. For personal safety, the NRVs and NRV bypass valves are closed first.
CAUTION before -
For personal safety, a caution was Step S added warning against dispatching an
- operator before the NRVs are closed.
O 4292e:ld/021086 48 l
MAINE-YANKEE EOF STEP DOCUMENTATION FORM
\
i FR-H.4 LOSS OF NORMAL STEAM RELEASE CAPABILITIES i
~ Sten No. ' Step No. Explanation or Basis for Difference No deviations. 3 i
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i...._..__.._____.....~__._..._... _ _ _ _ . . _ _ . _ . . . _ _ . , _ _ _ . . . , _ . , , , , _ , , _ , . _ _ , _ _ _ . , , _ , _ _ _ _ , . . _
n MAINE YANKEE E0P STEP DOCUMENTATION FORM O FR-H.S STEAM GENERATOR LOW LEVEL ,
Maine Yankee ERG Sten No. SteD No. Explanation or Basis for Difference 2nd CAUTION -
A caution was added to alert the opera-before Step 1 tor that a rapid restoration of feed-water may cause waterhansner.
CAUTION before -
A caution was added to alert the opera-Step 4 tor of the restrictive flow rate for feeding a boil dry SG to prevent large thermal stresses.
O 4292e:ld/021086 50 i _ . . _ . . _ _ _ _ _ . _ . _ _ _ . _ _ . _ . . . .
e MAINE YANKEE E0P STEP DOCUMENTATION FORM
/s t i V FR-P.1-IMMINENT PRESSURIZED THERMAL SHOCK CONDITION Maine Yankee ERG
' Sten No. Step No. Explanation c' lasis for Difference 1st CAUTION -
A caution was added to alert the opera-before Step 1 tor that a transition to FR-H.1 is only required if minimum FW flow capability is lost. This is also an ERG '
maintenance item.
2nd CAUTION -
A caution was added to alert the opera-before Step 1 tor of the potential of waterhammer damage if EFW flow is transferred to
-s the first point heaters.
\b CAUTION before Caution was deleted because SI will Step 6 actuate upon AC power restoration if required.
11 11 Maine Yankee does not have a BIT, so stopping HPSI flow effectively stops boration.
CAUTION before -
A caution was added to guard against Step 15 thermal stresses on the spray nozzle if Pzr. Aux. spray is established rapidly.
20 -
A step was added to determine if the containment spray pumps should remain running. For Maine Yankee the pumps O were started automatically on SIAS.
4292e:1d/021086 51 l
MAINE YANKEE E0P STEP DOCUMENTATION FORM FR-P.2 ANTICIPATED PRESSURIZED THERMAL SH0CK CONDITION Maine Yankee ERG SteD No. Step No. Explanation or Basis for Difference 1st CAUTION -
A caution ~was added to alert the opera-before Step i tor that a transition to FR-H.1 is only required if minimum FW flow capability is lost. This is also an ERG maintenance item.
2nd CAUTION -
A caution was added to alert the opera-before Step i tor of the potential of waterhanner damage if EFW flow is transferred to the first point heaters.
CAUTION before -
A caution was added to guard against Step 3 thermal stresses on the spray nozzle if Pzr. Aux. spray is established rapidly.
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4292e:ld/021086 52 L
MAINE YANKEE E0P STEP DOCUMENTATION FORM FR-Z.1 HIGH CONTAINMENT PRESSURE Maine Yankee ERG Sten No. Sten No. Explanation or Basis for Difference CAUTION before The ERG caution to transition to ECA-1.1 Step 3 was deleted because ECA-1.1 was not part of the E0P procedure package.
S The ERG step to close the MSIVs was included in the Maine Yankee step to isolate the SGs.
- 8, 9, 10 These ERG steps were deleted because Ov Maine Yankee does not have a hydrogen recombiner system and the. notification ,
of the technical support center was ,
included in Step 6.
b 4292e:1d/021086 53
o MAINE YANKEE E0P STEP. DOCUMENTATION FORM O FR-Z.2 l
. CONTAINMENT FLOODING
~
l Maine Yankee ERG Sten No. Sten No. Explanation or Basis for Difference 1 -
A step was added to determine if there is an adequate water supply in the RWST or if recirculation alignment is required.
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O 4292e:1d/021086 54 L
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MAINE YANKEE E0P STEP DOCUMENTATION FORM FR-Z.3 HIGH CONTAINMENT RADIATION LEVEL Maine Yankee ERG SteD No. SteD No. ExDianation or Basis for Difference 2 Step deleted because Maine Yankee does not have a Containment Atmosphere Filtration System similar to the ERGS.
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4292e:1d/021086 55 L
1 i
l MAINE YANKEE E0P STEP DOCUMENTATION FORM FR-I.1 l i- HIGH PRESSURIZER LEVEL
- 4 1-4 i
-Sten No. Steo No. Explanation or Basis for Difference __
i f No deviations.
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t, 4292e:1d/021086 56
MAINE YANKEE E0P STEP DOCUMENTATION'FORN FR-I.2 LOW PRESSURIZER LEVEL .
Maine Yankee ERG Sten No. Sten No. Explanation or Basis for Dif ference No deviations.
O-O 4292e:ld/021086 57 L
MAINE YANKEE E0P STEP DOCUMENTATION FORM FR-I.3 VOIDS IN REACTOR VESSEL -
Maine Yankee ERG Step No. Step No. Explanation or Basis for Difference 8 8 A check on head region subcooling was added to the step s ,ce PITS does not give a reliable level indication.
10 10 Head region subcooling was used to determine the offects of starting an RCP had on condensing the void.
. 21 20 Since Maine Yankee does not have a
) RVLIS system to determine if venting has eliminated the void, the RCS is depressurized by 25 PSIG to determine '
the response in Pzr. level. An unexpected increasing Pzr. level indicates that the void in the vessel has not been eliminated.
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1 4292e:1d/021086 58
. . ]