ML20236B924

Forwards marked-up Parts of Abnormal Transient Operating Guidelines Manual (Atog) Used in Telcon Re Verification & Validation of Cooldown Procedures.Util Feels That Atog & Plant Documents Define Emergency Operating Procedures

ML20236B924
Person / Time
Site:	Rancho Seco
Issue date:	10/16/1987
From:	Andognini G SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
GCA-87-673, NUDOCS 8710260410
Download: ML20236B924 (19)

Text

i . . ., .

huo SACRAMENTO MUNICIPAL UTILITY DISTRICT O P. O. Box 15830, Sacramento CA

., 4p 95852 1830,(9 [4 n 2 3'2 @

RNIA i

October 16, 1987 AN ELECTRIC SYSTEM SERVING g lT GCA 87-673 k l j

I U. S. Nuclear Regulatory Commission Es l

Attn: J. B. Martin, Regional Administrator l Region V Office of Inspection and Enforcement

@C5 2c  ;

I 1450 Maria Lane, Suite 210 Halnut Creek, CA 94596

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,0 L DOCKET N0. 50-312 e i RANCHO SEC0 NUCLEAR GENERATING STATION N l

LICENSE NO. DPR-54 VERIFICATION AND VALIDATION OF C00LDOHN PROCEDURES

Dear Mr. Martin:

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Por the telephone conversation between Mr. Chris Caldwell of Nuclear Regulatory  !

Commission, Region V, and our Dennis Tipton and Steve Redeker, I am enclosing marked up copies of the portions of the Abnormal Transient Operating Guidelines  !

(ATOG) manual used in the discussion about the Verification and Validation of l Cooldown Procedures (cps). The ATOG manual makes a distinction between l

" Guidelines" and "Cooldown procedures." Also, Rancho Seco Administrative i' Procedure AP.47, " Emergency Procedures Writer's Guide," states in Step 3.2 that Emergency Operating Procedures (EOPs) are designated by the letter "E".

Based on these excerpts from ATOG and AP.47, which were originally submitted to NRR as part of the District's original Procedures Generation Package, the District has defined the E0Ps. Therefore, your concern about the V & V of cps l should be resolved. Also enclosed is a copy of the original ATOG manual, l Revision 0 for yc.ur information. l If you have any further questions on the matter, please contact Dennis Tipton j of my staff at Extension 4548. l l

Sincerely, h(

G. Carl Andognini l Chief Executive Officer, Nuclear 8710260410 871016' l PDR ADOCK 05000312 l Attachment P PDR l cc: G. Kalman, NRC, Bethesda (w/o atch)

A. D'Angelo, NRC, Rancho Seco (w/o atch) .

F. J. Miraglia, NRR,'Bethesda (w/o atch) j

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i RANCHO SECO NUCLEAR GENERATING STATloN O 1444o Twin Cities Road, Herald, CA 95638-9799;(209) 333 2935 J

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I Objectives: a) minimize offsite releases b) t e rmina t e leakage before BWST depletion I c) maintain core cooling / expedient cooldown and I depressurization Key Points: a) transient (leakage) not terminated until RCS cooled ,

depressurized, and drained below tube leak elevation 1

b) LOOP / natural circulation cooldown will probably result l l

in highe r offsite releases due to greater need to steam j the affected SG to the atmosphere 1

c) natural circulation cooldown necessitates use of EMOV l x ,~ m ,~ v ~ ~ ,.

to reduce RCS pressure , e, 1

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')h Cooldown Procedures / Inadequate Core Cooling ~) %Jw-

, ,.- - s The objective of these guidelines is to maintain adequate core cooling by  !

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t e rmina t ing t rans ie nt s and stabilizing the plants with controlled decay y/ 4 heat removal. Once stable conditions are achieved, further plant cooldown can be accceplished by existing plant proc ed ure s .

tions at stabilization following the execution of the guidelines will not However, the end condi-nec es s aril y coincide with the entry conditions for plant cooldown proced-m F gff ures. There fo re , procedures are provided in Part I to accomplish the p 4

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[ transit ion fecu the guidelines to the plant pec c ed ur e s . Five cooldown f k (*N }

p i procedures are provided to cover five possible end ccr.ditiona $

1 1) Cooldown following a large LOCA 4

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N 2) Normal cooldom

, 3 4; 3) Saturated cooldovn with primary to secondary heat tr ans f e r i

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4) HPI cooling

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jj 5) Solid plant cooldown/ recovery from solid plant. j e '

3 A sixth procedure is provided for the special case of Inadequate Core Cool-L; j ing (ICC). The philosophy and the objectives of the actions for ICC are 0 I

discussed in detail in the " Backup Cooling Methods" chapter in Volume 1 of I Part II. Y l

warwann At the end of the section containing the five cooldown proc ed ures (immedi-  ;

i ately be fore the ICC section) are specific rules. These rules are pro- j vided in a separate section to avoid repetition throughout Section III of the guidelines. These rules apply wherever they are referenced in Section

/] j III. Specific rules are provided to cover:

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1) Initiation of RPI l)

2) ELDI flow control i 1

3) Feedwater throttling methods i l

4) SC level setpoint.

In addition, three fig ur es are provided at the back of Part I fo r easy  ;

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' reference during the use of the guidelines. These figures provide: l

1) HP I flow vs . RCS pressure r l

2) RCS pres sure-temperature limits for brit tle fr ac t ur e / tiDT

3) Incore exit thermocouple temperature for ICC  ;

i Goerator Aids '

In addition to the guidelines in Part I and the training material in Part DATE: PAGE 10-8-82 255

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Y$ fu$j} 01 Nor CHAPTER C ('00(20m C/064t bl5ttMcN ABNORMAL TRANSIENT DIAGNOSIS AND MITIGATION  !

Introduction This chapter shows how an abnormal translent can be diagnosed and I mitigated using the information provided by the P-T diagram and the concepts on heat trans fer discussed in the previous chapters. A simplified flow chart of the approach to be used to diagnose and mitigate an ab no rmal transient is provided in Figure 20A " General Plant Transient 1

Mitigation". It is broken down into a few separate steps . although these steps will blend together into one con t inuous process in actual practice.

The Abnormal Transient Operating Guidelines are implemented whenever an m

1. automatic or manual reactor trip occurs or a' forced shutdown is necessary.

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The guidelines are provided in Part I. They list the appropriate operator l

actions necessary to mitigate an abnormal transient. They follow the approach outlined in Figure 20A. The guidelines incorporate the following features.

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1. Use of the P-T diagram which provides a constant feedback to the operator on his success or failure after taking each step in Part I.

This diagram should be checked frequently to make sure th ing s are progressing as expected. It will thus give the operator esrly indications of s ub s eque n t f ailure s that are delayed after the initial event, or mult iple failures that were masked by the predominant event and thus didn' t appear until that one was corrected.

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2. The guidelines are constructed such that the c.perator makes an attempt I

to correct the problem with a given piece of equipment 'or system (e.g.,

AFW to correct loss of main feedwater). If that' fails he is instructed.

l to go on to the next available syste'n (e.g., HP1 cooling). The failure o? a particular syatem (e.g., the AFW system) is. not given priority attention in Part I, protection of the core is.

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3. The operator is given fdquent present plant status ( STATUS) aids throughout the procedure to help him maictaiu proper orientation.

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4. If new symptoms appear he is instru :ed a recycle (go to the . section l

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that treats that symptom) to the 'appropiate part of the procedure . )

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A detailed discussion or. the use of the guidelines in Part I is provided i in Chapter H of this volume.

Immediate Actions ,

The first block in Figure 20A is the "Immediate Actions" block. The i:amediate actions should be completed in the first 2-3-minutes. The first action to be made is to determine' if a reactor trip has occurred or pla nt.

conditiens requiring a forced shutdown exist. If a reacter trip has occ urred the operator should manually trip the reactor and turbine, then ,

proceed to the next post trip step of the ATOG procedure which is. Vital Systems Status Ve r i f ic a t ions " . However, if plant cond i t io ns ' ' war r ant a forced shutdown, the operator should ini t ia t e the appropriate shutdown

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1 procedure. When the reactor is tripped during the forced shutdown I 7perations, the post-trip ATOG procedures should be implemented. If the

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I forced shutd own is due to a steam generator tube rupture (SGTR) the operator should. proceed directly to III D of the guidelines. )

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Vital Systems Status Verification l j

The next major block on Figure 20A is the Vit al Systems Status Veri-1 fication. This section requires reviewing specific plant status items including the P-T diagram to determine if they are behaving as they should )

i for' a normal reactor trip. If the s pecific plant status items cannot be ]

1 1 l ve ri fied as performing as expected, the operator should perform the l

l A specified remedial actions. The proced ures provide specific remedial

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d actions for each plant status item which cannot be verified. The plant status items which are checked first are the normal automatic post trip.

f unct ions which control core re ac t iv i ty , primary and seconda ry inventory,

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l and primary and secondary pressure. ~

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l j Next the operator must verify the operability of certain power supplies to

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l assure the important plant parameters can be monitored and the power is l

l available to the im po r t ant control devices. . These include pumps, valves, etc., which are needed to safely control the plant and to mitigate abnor-1 mal transients. The plant cannot be controlled or abnormal transients mitigated if the control devices do not work. Nor can the plant be safely controlled if the instrumentation parameters to be controlled are not available for diagnosing the plant status.

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1 Next the operator must check to see if any safety systems have been actuated and if so that they are operating properly. For example if the operator notes the presence of a high reactor building pressure alarm of :l l

30 psig he will be directed to verify actuation of the re ac t o r building _.

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spray system. )

I Next, the operator needs to check the P-T diagram fo r , a los s of subcool-ing margin, " overcooling" or " overheating" conditions. T able 2 summarizes generally the plant status items to be checked and the actions to be taken I

if the items are not as they should be.  !

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)i Finally the operator stust check to see if there are any steam generator 9/

tube failures.

The P-T diagram is the foundat ion for transient diagnosis and fo r the actions to correct abnormal transients.

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When the PT diag ram is checked, the reactor P-T should stay within the ,

" post trip" window and steam pressure should stay above the steam pressure i

i limit of 960 psig. If the plant res pond s so th a t these limits are not 4 l

exceeded then the transient is going "as expected". If the plant does not fe i "go as expected" the P-T characteristic should be checked to find out the i

" type" of abnormal transient so that proper corrective act'ons i can be made j i

to restore the reactor-steam generator heat transfer. )

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NUCLEAR power oENERATION OlVISION TECHNICAL DOCUMENT If the Vital Systems Status Verification shows everything is alright , then the pl ant is in a s t ab le subcooled condition with pr ope r primary to secondary heat transfer and no major primary boundary failures. Further j l

action will be at the discretion of the station management. ,

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I If one or more items are not alright, the ATOG procedure will direct the k l

operator to make a remedial action. The remedial action may be a finite action such as closing a valve or a " Followup Action".

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i Followup Action )

As shown on Figure 20A when the P-T diagram indicates a " loss of subcool-

[  %, ing margin, " overheating" or " overcooling", exists the operator must d e te rmine wh ich one of the three conditions exist, and then start the

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appropriate followup actions procedure. These followup actions are first 1

di rec t ed at rees t ab li shing the correct amount of steam generator cooling. )

l If it cannot be reestablished, then backup cooling methods are to be I implemented. (The backup cooling methods are di sc us sed in Chapter D of Part II).

Once s t e an generator heat t rans fer or backup cooling has been established ,

i the plant should be brought to a stable condition for plant cooldown. The condi tion may be het or ccid depencing on the circums t ances ; a return to the " post trip" window is not required for plant stability.

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Cooldown Procedures i Once stable plant conditions are reached, the plant is cooled down using

.kl i one of several cooldown procedures depending on the existing stable plant ( ,

e N I conditions such as a solid water system or saturated RCS.  ; 4 r; .T _ heat 00 _

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] g proc ed ures state which cooldown procedure~~~. to use. This is discussed in V i

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f j the Post-Accident Stability Chapter of AT00.

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=;wommwl Inadequate Core Cooline In the eve nt that neither ateam generato r cooling nor backup cooling is  ;

l e s t ab li sh ed , " Inadequate Core Cooling: will occur. Th is topic is i discussed in the Backup Cooling Chapter.

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Abnormal Transient Diagnosis and Treatment

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Although the type of transient may have become evident during 'the first 2 l

l l or 3 minutes after trip, plant monitoring is required to make sure that 1

the transient is going as expected. Generally, after 2 or 3 minutes the i plant will begin to stabilize within the " Post Trip Window" (examples of l 8

i this were given in the P-T Diagram Chapter). Actions have already been  !

t ake n to identify and handle the " fast" excessive main feedwater transient and the systems which should be operating have been checked to make sure ,

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thac they are working correctly. Further plant mo n i to ring should ' begin, l i

At this stsge the- e f for t should new be to make sure chat the plant i i

sesbilizes as it should. )i 1

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To do this the P-T diagram is kept under surveillance. If reactor coolant 3

. pressure and tem pe ra tur e stabilize within the P-T post trip wind o w, and i steam pressure is above the- low steam' pressure limit, the transient is-probably not abnormal and a quick check of the following should be made ' to c' ' '

ensure system and equipment parameters are within expected values: .l Heat Transfer Balance Indicators

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P-T diagram ( for RC pressure and temperature and subcooling and secondary saturation temperaute)' -l 1

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- Pressurizer Level .

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- Steam generator level and pressure

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Equipment Status and. Operation (depending on what was started); i

- Makeup /HPI flow rates and pump status-

- Main or auxiliary feedwater flow rates and pump status l

- RC pump operation including cooling water and seal injection service

- Position of important valves (letdown, EMOV, feedwater isolation s and control valves, pressurizer spray valve)

- Reactor Building isolation and cooling systems I

- Power supplies (AC and DC) l Once these reviews are completed a more thorough check can be cond uc t ed and a decision made to determine if the plant is stable. (Refer to the Chapter F on "Pos t Trans ient Stability Determination" .)

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i But if the first review of the P-T indic at es that the reactor coolant pressure and tem pe r a tur e are not going to remain within the pos t-trip window (or return to it), or that steam pressure is below the steam pressure limit, then some thing is wrong with heat transfer and corrective actions are required to bring the heat transfer into balance, i

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. POST TRANSIENT STABILITY DETERMINATION To determine if the transient has been brought under control- four general 1

areas must be checked.  ;

1. Reactivity Control - The reactor must have a suberitical margin of j at least 1% ak/k.

2. Core Heat Removal Control - The core must be c"vered and cooled; i i

the heat removal rate is equal to or slightly greater than the core heat generation rate.

3. Radiation Release Control - Release to of fsite is terminated .

4. Plant Equipment is Operating Correctly - Equipment to maintain the

, ) plant

/ safe and stable is operating and within design duty;  !

equipment. f ailures have been bypassed , isolated or repaired .

Several things around the plant must be chec ked to make sure these four  !

l general rules are being met. The following basic check list defines the more impo r t ant items. The list is divided into two cases. Case I applies l to LOCA's which can be stopped by complete isolation of the leak and to all other t rans ient s. Case II applies to LOCA's which cannot be isolated.

i The difference be tween the two parts is simple:

a reactor leak that can-not be s top ped is a transient that cannot be po s it iv ely t e rm ina t ed . How-eve:, a leak can be reduc ed to the smallest amotnt pos s ible and becoma i I

stable fo r "l o ng -t e rm c oo li ng" .

Steam generator heat removal can be u s ed for some small leaks but HPI or MU mu s t be kept r unning to maintain the i i

reactor coolant inve n to ry . Subcooling can be regained for some very small i

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break sizes at a time when the decay heat decreases and HPI is able to refill the RCS loops and add water to the pressurizer.

Case 1 - All transients (including LOCA's which can be isolated) =

1. Reactor coolant pressure and temperature are pre fe rably within the "po s t-t r i p window" of the P-T diagram;. however, pressure and temperature may be anywhe re on the P-T diagram within a region bounded by; a) NDT limits, b) the subcooling ' margin, c) an RC l

pr e s s ure upper limit equal to the EMOV setpoint, d) fuel pin j compression limits and e) RCP NPSH requirements, if ap pl ic ab le , j 1

Subcooling will exist in the. hot and cold legs of both loops. i

2. The "long term" trend of reactor coolant pressure and tempe r a tur e  ;

is constant or slowly decreasing with time. "Short-term" l l

fl uc tua t ions of temperature and pres sur e are small and can be' j i

attributed to periodic operations of other equipment. (pressurizer l l

heat ers , spray, or feedwater),

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3. Pressurizer level is within the indicated range.  !

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4. If fo rc ed circulation exists (RC pumps on) then reactor coolant T ave is abou t equal to the saturation temperature of the water in l

the steam generator (or generators) that is removing the heat. l l

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5. If natural circulation exists, Tc old will be about equal to the saturation t em pe r a t ure of the water in the steam generator (or i

generators) that is removing the heat. The dif ference between l incore th e rmo c ou pl e s and h T ot in the operating loop (or loops) will l track within 10F. ]

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6. Steam generator level will be at the correct se tpoint (either i natural or forced circulation setpoint) and will be steady. )

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7. Steam gene ra to r pressure is s t e ad y and is below the safety valve l l opening setpoint.

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8. The core is at least 1% A k/k sube ritic al on rods and boron. If I

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more than one rod did not fully insert the core is at 1%  ;

7-l l \ A k/k subcritical on boron alone, i

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l 9. If the transient caused water to enter the reactor building and the reactor building env iror. men t was increased, it will now be l l

reduced to near normal levels. Pressure will be close to j atmos pheric pressure; average re ac to r building temperature will be i

near prior operating temperature; relative humidity will be about  ;

l l 100%. i l

l 10. If radioactive water leaks oc cu rr ed in auxiliary buildings those l

areas will be sealed and the spillage either trapped or drained to s to rage t ank s .

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11. The component- failure (or f ailures) which caused the transient is l l

known. It has been bypassed, isolated,- repaired, or otherwise handled so that it no longer compromises plant safety, t

12. Components which suppo rt plant s afe ty are operatire, within their l

design limits (examples: pumps are operating away from the minimum l I

shutoff flow and have adequate NPSH, throt tl e valves are near the ]

. J proper o pening , electric motors are in the norma'l service range, .

elec t ro nic equipment is environmentally protected). If a component is operating off design and future f ailure is possible, then I i

redundant or alt ernate . equipment is on standby and ready to replace  !

the equipment which might fail.

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13. Stored water (condensate storage tank, BWSTI is adequate for long i term use or alternates are readily available.

14 Ins trume nt at ion to moni tor plant performance is operating l

- correctly. Potential failures of c rit ic al instrumentation have l been identified and alternate instrumentation is available.

Case II - LOCA's which cannot be isolated l

NOTE: With the exception of steam generator tube leaks, all reactor coolant I l

1eaks outside the reactor b uild.ing can be isolated, Although a l tube le ak is " ins id e" the reac te r building a direct path ou t s id e l the reactor building exists through the steam lines.

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TECHNICAL. DOCUMENT 74-1127469-00 Many of the criteria of Case I apply to this part except that the reactor coolant will not always regain the subcooled margin and operating condit ions that depend on subcooling will not apply. The very smallest reactor coolant leaks may allow the reactor coolant system to repressurize <

l (because of continued High Pressure Injection) .and some amount- of e s ub cooling may be regained, but it is not likely that the subc ooling  ;

margin will be restored. Consequently, the criteria ' for LOCA stability l does not include the subcooling margin. Also because subcooling may not exist the hot legs may have steam binding and natural circulation may not l i

exist; therefore, the criteria do not include natural circulation [

requirements (however, it can exist for very small breaks and should be f

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. checked). A re ac t o r-s t e am generator heat trans fer balance cannot usually j

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be accomplished because of saturated (or near saturated) conditions which j may not oermit the reactor coolant to move the heat from the core to the steam generacor, but some heat trans fer to the steam generator is pos s ible 1

fo r small b re ak s . The steam generator operating level should be at the 95% level for small breaks to permit condensation of primary side steam.

Pressurizer level cannot be relied upon if saturation exists.

The most impo rt ant criterion for LOCA is to keep the core covered. This condition is confirmed by readings of the incore thermocouple and the hot leg RTD 's; both should show that the reactor coolant is s a tur a t ed (or even subcooled) but not su pe rhe a t ed .

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The continued loss of coolant . from a LOCA will not permit the . transient to be truly terminated , .but the leak rate can be ' minimized . Lowering RCS pressure is the best 'way .to lower the leak rate. This can be done by' loss l

through the le ak , by opening the EMOV, or by lowering ~ secondary side

.s ,

pressure. Long term loss of coolant when the RCS is depressurized occurs in two ways: 1) steaming out of the leak because of continued boiling ,

and 2) water loss' bec ause the nead of water is above the b reak and water l l will "run" out- of it. The rate of leakage m depend on the system pressure, the decay heat level (which causes bo ili.n;;) , and the elevation 1

of the le ak (a leak high in the system will have a lower flow rate than a leak low in the system). The leak rate will also depend on the hole size.

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~--/ i1 The criteria for stability is that the leak rate is as low as possible and )

i that the flow into the core keeps it covered. It may take a very long j I time to recove r from some LOCA's and during that time there will be two 1

general stages shen the leak rate diminishes. The first s tage 'is when the reactor coolant system is depr es sur ized to the t e ac t o r building pr es s ur e ]

I (big breaks will depressurize rapidly, smaller breaks will take longer); q the second s t ag e is when the core heat drops so th a t it cannot boil the )

I water in the reactor vessel. Steaming will stop at that time (which may be

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as long as several months after the transient). .Until the water in the  !

l vessel becomes s ub c o oled (incore thermocouple r e ad less than 212F), the  !

plant must be o pe r a t ed by injecting reactor building sump water in the ')

recirculation mode or by continuing to inj ec t fresh borated water from

.):

1 i

\

i DATE: PAGE 10-8-82 171 , j j

,. .. 1 BWNP-20007 (6-76).

BABCOCK & WILCOX nomen NUCLEAA POWER otNERAiloN DIYl5!ON

.m TECilNICAl. DOCUMENT 74-1127469-00 other sources. When the vessel water becomes subcooled the operator has  !

the option to trans fer one train . of LPI to the decay heat removal mode and i keeping the other train on sump recirculation. The reason one train is I

left on recirculation is th a t it will keep water above the hot leg suction l for decay heat removal. Decay heat removal has the advantage of rapid RCS cooldown, but it must be carefully monitored to make sure the decay heat pump does not lose suction (or it will fail), and to make sure the decay I

heat pump does not run at shut-o f f head ,

i Because the leak may continue a long time until the decay heat system is placed in se rv ice , an arbitrary definition of s t ability is given. The following criteria define post-LOCA long term stability:

/ 1. The core is covered. Incore thermocouple readings show saturated or subcooled reactor coolant.

2. ECCS inj ec t ion is in the "long term cooling" mode. Long term cooling exists when the ECCS is operating with recirculation from the reactor building emergency sump. (NOTE: A decision may have been made not to transfer but to bring in backup water to refill th e BWST. Nevertheless, if recirculation could have been started , "long term cooling" is considered to have started).

3. The reactor coolant system is depressurized to near atmospheric  !

l pressure so that the leak rate is as low as po s s i'o le . The LPI f system is used to cool th e core. (NOTE: If the break size did  !

l 1

l 1

i DATE: 10-8-82 PAGE

e' ' '

BWNP-20C W (6-76) .

s- ,

BABCOCK-& WILCOX NUM8(R s

NUCllAR POWER oENERAflON olVl$ ION TECMilC Al.' DOCUMENT ~ 74-1127469-00 not permit depressurization before .the BWST was empty, and'HPI

" pig gyb ack" recirculation had to be .used while further depressurization took place the plant is not- considered to be l stable until the pressure 'and leak rate are as low as possible).

~^

4. Steam generator level is at 9'5% on the ' operate range and is s t e ad y .

l l

5. Reactor coolant pumps are off (operation of RC pumps could move 1

water past the break and increase the leak rate).

6. The following criteria from the previous part also apply:

Numbers 7, d, 9, 10, 11, 12, 13, 14

7. For the special case of steam generator tube leaks (LOCA's):

- I a) Fe ed wa t er (main and auxiliary) . has been s topped to the I bad generator. ."'

b) Steau created by boiling the RCS leakage is directed to the condenser (if it is operating). i l

c) The plant is on decay heat ' removal. or standby backup borated water sources are available to replenish BWST inventory, l

l l

l

)

DATE: PAGE 10-8-82 .

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