ML19260B070
| ML19260B070 | |
| Person / Time | |
|---|---|
| Site: | Washington Public Power Supply System |
| Issue date: | 12/03/1979 |
| From: | Renberger D WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| GO1-79-580, NUDOCS 7912060476 | |
| Download: ML19260B070 (200) | |
Text
PP W shington Public Power Supply System A sCINT OPERATING AGENCY S
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.ro-..ssa e o. rso.) avs.sooo Docket Nos: 50-460 December 3, 1979 50-513 G01-79-580 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
2055E
Dear Mr. Denton:
Subject:
WPPSS Nuclear Projects Nos. 1 & 4 Response to NRC 10 CFR 50.54 Letter of October 25, 1979
Reference:
Letter, Mr. H. R. Denton, NRC to Mr. N. O. Strand, WPPSS, "10 CFR 50.54 Request Regarding the Design Adequacy of Babcock & Wilcox Nuclear Steam Supply Systems Utilizing Once Through Steam Generators (WPPSS 1 & 4)", October 25, 1979.
In the reference letter the NRC requested that WPPSS provide information on the sensitivity of the primary conlant system behavior to feedwater transients and the status of various systems. Six specific requests for information were made, identified as a through f.
Our detailed replies to these requests are attached, similarly identified.
Attachments a and b provide responser to the NRC requests to identify the most severe overcooling events and identify whether ECCS or RPS action is required for these events. We received input on these issues last week from Babcock & Wilcox and have initiated review of the material.
We will inform the staff if any changes to these attachments are appro-priate for WNP-1/4 in view of the B&W material.
The recommendations for plant changes are included in attachment f.
These recommendations involve changes in electrical systems, contrul.
instrumentation and valves.
They do not involve major changes to large pieces of equipment such as vessels, heat exchangers or pumps.
Like-wise, no major changes in pioing are considered necessary.
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7912060 d
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O Mr. H. R. Denton e
Page Two As discussed in attachment d, the ability to n.ake major changes to any of the WNP-1/4 systems identified in enclosure 3 of the refer.:nce is generally preserved because major access openings will be maintained in the Containment and General Services Building for at least ten months.
O In this light, and considering the design, fabricaticn and construction status of the various systems as discussed in attachment c, it is feas-ible and prudent to continue construction and integrate the reouired changes into the normal design and construction process.
6 Very truly yours, D. L. Renberger Assistant Director - Technology 9
DLR:AGH:oe Attachments cc:
A.
Bournia, NRC g
N. S. Reynolds, D&L C. R. Bryant, BPA Eng. Files - 1/4 O
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Docket Nos: 50-460 50-513 Sub: Response to NRC lo CFR 50.54 Letter of October 25, 1979 STATE OF WASHINGTON)) ss COUNTY OF BENTON
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D. L. RENBERGER, Beino first duly sworn, deposes ar.d rays: That he is the Assistant Director, Technology, for the WASHINGTON PUBLIC POWEA SUPPLY SYSTEM, the applicant herein; that he is authorized to submit the tore-going on behalf of said applicant; that he has read the foregoing and knows the coritents thereof; and believes the same to be true to the best of.his knowledge.
DATED
[et 9
, 1979
&/dnIny
- b. L. RENBERGER' On this day personally appeared before me D. L. RENEFRGER to me known to be the individual who executed the foregoing instrument and acknowledged that he signed the same as his free act and deed for the uses and purposes therein mentioned.
GIVEN under my hand and seal this & # day of M,.,,,mA<,~
, 1979.
(
ka-b idec e pttary Public in and for one State of Washington t
f Residing at e c l/h u._ V
/
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- 521 02a
ATTACHMENTS a and b Idettify.t/tc mest sevete ovetcooling evatts (costsideting botit aittielpated.ttastsiatts and accidatts) testicit could occwt a.t yowt factllty. Titese sitould be die evatts telticli causes tite greatest inva&ty sittinkage. Undet ute guidelines titat no opetator action occurs befo.te 10 etinates, and antij safety sys.cems can be used to utLtigate.tlie evatt, cacit Licatsee sitould sitoto Biat die core rema.ists adequately coated.
Idattiftj telletltet aeticsi of tite ECCS or RPS (or operator action) is stecessa.ty to protect. tite core follotcing tite mos.t seveste ovc<tcooling ttastolertt idottified.
If Diese stjstens are requited, you sitould sitolo titat its design etitetion for ble number of actuation cycles is adequate, costsideting artivat.tates fo.1 excessive cooling ttastsietts.
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ATTACHMENTS a and b OVERCOOLING EVENT CONSEQUENCE REVIEW l.0 INTRODUCTION AND CONCLUSIONS
1.1 Background
1.2 Scope 1.3 Applicability of Results 2.0
SUMMARY
2.1 Limiting Overcooling Event Confirmation 2.2 Shrinkage Effects 2.3 Adequacy of Core Cooling 2.4 Adequacy of Core Protective Measures 3.0 ANALYTICAL TECHNIQUES 3.1 Computer Codes 3.2 Transient Selection 3.3 Basic Assumptions 4.0 RESULTS OF CORE COOLING STUDIES 4.1 Anticipated Transients 4.1.1 Scope of Evaluation 4.1.2 Main Feedwater Overfeed Analysis 4.1.3 Conclusions 4.2 Accidents 4.2.1 Scope of Evaluation 4.2.2 Steam Line Break Analysis 4.2.3 Conclusions 5.0 DESIGN BASIS FOR CORE PROTECTION
6.0 REFERENCES
152i 025 ab-1
9
1.0 INTRODUCTION
AND CO'GU3 IONS
1.1 Background
On October 25, 1979, the NRC issued a letter to utilities (D
holding Construction Permits for B&W NSSS's; the utilities were requested to assess overcooling even*.s on their plants, accounting for balance-of-plant features. Two specific requests of the NRC letter were:
a)
" Identify the most revere overcooling events (considering gp both anticipated transients and accidents) which could occur at your facility. These should be the events which causes the greatest inventory shrinkage. Under the guide-lines that no operator action occurs before 10 minutes, and only safety systems can be used to mitigate the event, each licensee should show that the core remains adequately gg cooled."
b)
" Identify whether action of the ECCS or RPS (or operator action) is necessary to protect the core following the most severe overcooling transient identified.
If these systems are required, you should show that its design gp criterion for the number of actuation cycles is adequate, considering arrival rates for excessive cooling transj ents."
1.2 Scope This report responds to the specific NRC requests identified gg above. More than one transient type is analyzed to address different frequency of occurrence classifications and to assure the most severe cases are indeed included in the evaluation.
A qualitative assessment of possible non-mitigative operator actions in the O to 10 minute time frame is also provided.
This assessment provides indication of what operator action I
i.s anticipated during the intitial phases of an cvercouling cransient.
The analyses identify the frequency of the RPS, ESFAS, and operator action for mitigation of the transient.
II A summary of the results is given in Section 2.0.
Section 3.0 provides the details of the initial conditions, computer codes, and basic assumptions used in the analysis. The transient response data is given in Section 4.0.
Section 5.0 demonstrates the adequacy of the design criteria for tsch system.
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1.3 Conclusions Based on the analyses performed in this report, the following conclusions can be drawn:
a)
The overcooling accident (Main Steam Line Break) and the overcooling transient (Main Feedwater Overfill) analyzed herein, retain adequate core cooling even when analyzed with no operator action before 10 minutes and with only safety systems used to mitigate the event.
b)
RPS and ECCS actuation are required to mitigate the most severe overcooling transient; hcwever, operating data imply the arrival rate of transients requiring RPS or ECCS sctuation is within the design basis.
It should be noted that this report could not exhaustively determine the most severe overcooling transient in the allotted time; the reasons for selecting Main Feedwater Overfill are discussed in Section 4.1.1.
1.4 Applicability of Results The results presented in this report are applicable specifically to this NSS with the parameters tabulated in Section 3.
Specific attention has buen paid to the balance-of-plant equipment in the mitigative functions performed.
2.0
SUMMARY
This section provides a detailed summary including:
(a) identifi-cation of the safety concern and basis for selection of the tran-sients to resolve the concern, and (b) prf.ncipal results of the analysis. By reviewing this section, supported by the details given in Sections 3.0, 4.0, and 5.0, a concise overview can be obtained of the completed resolution of this concern.
Section 2.1 addresses the selection of anticipated transient and accident conditions causing greatest core shrinkage, and Section 2.2 discusses the phimmena of void formation under inventory shrinkage conditions. Section 2.3 summarizes the analyses.
Seccion 2.4 summarizes Section 5.0 in demonstrating that the design criteria for the number of actuation cycles of the RPS and ESFAS is adequate.
2.1 Limitina Overcooling Event Confirmation Maximum RCS coolant inventory shrinkage results frcm a decrease in the pressure and temperature of the coolant at a maximum rate, without a compensating coolant makeup addition. The double-ended steam line break (SLB) provides maximum cooldown rates and is analyzed in Section 4.2 as the limiting accident.
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O Several sensitivities and differing conditions vere analyzed to provide greater insight into the steam void formation and collapse which would occur and its subsequent effect on core cooling. These additional studies were performed on the SLB since this accident was expected to result in RCS voiding, GD whereas, it will be shown the limiting moderate frequency event analyzed does not produce voiding as a result of RCS cooling inventory shrinkage.
In selecting the limiting anticipated transient, SAR and oper-ating plant overcooling events were reviewed. The most severe (D
moderate frequency event in the SAR is the steam pressure regulator malfunction. In review of plant transient data (see Section 4.1.1), the overfeed by main feedwater after reactor trip has produced the most severe overcooline transients.
Hence, based on arrival rates for operating plants and cool-down rate associated with this transient, the main feedwater gp overfeed following a reactor trip / turbine trip is considered the limiting anticipated transient and is analyzad in Section 4.1.
2.2 Shrinkage Effects Shrinkaga of the RCS coolant liquid volume occurs as tempera-ture decreaces duging an overcooling event. The pressurizer volume of 2250 ft contains 1050 ft of mostly saturated water during normal operation. This liquid volume flows out of the pressurizer into the system as the systam inventory volume decreases.
IftheRgScoolantinventoryvolumedecreaseis gg greater than 1050 ft and continues to decrease, then the pressurizer steam space can be transferred into the RCS.
This type of steam voiding is limited by the inventory volume difference between hot, full pa*er and the final pressure /
temperature achieved during the transient.
Its effect is further mitigated by actuation of emergency core cooling system I
(ECCS) such as high pressure injection (HPI) low pressure injection (LPI), or core flood tank (CFT) systems.
The other mechanism which produces steam voids in the F'JS is flashing of RCS water. As the pressure rupidly decreases in the RCS, the liquid in the hotter portions of the system can become saturated and steam can form. This effect car. be II aggravated by the hot metal in this area flashing additional water to steam. This process, in a non-LOCA situation, how-ever, is self-regulating. Ar the steam separates, or addi-tional flashing occurs, the pressure decrease in the system lessens as the overcooling continues. The steam void forna-tion is then reduced and the steam void will tend to collapse GD as a subcooled state is again established.
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Examination of the SLB analysis indicates a small amount of steam formation occurs in the upper hot leg region prior to the pressurizer emptying, occuring almost exclusively on the side with the affected steam generator.
If the a"fected steam generator is on the loop with the pressurizer, emptying the pressurizer contributes to the steam void formation.
If the affected steam generator is on the opposite loop from the pressurizer, emptying the pressurizer has little effect on the steam voids on that side and they are quickly quenched.
Therefore, the limiting accident, in terms of void volume formation, occurs for the SLB in the same loop that has the pressurizer.
2.3 Adecuacy of Core Cooling In this section, a summary of the results presented in Section 4.0 is given and analyzed for determination of adequate core cooling. The anticipated transient analyzed is the overfeed of the steam generators by the main feedwater (Section 4.1).
This overcooling transient, with no mitigative operator action for 10 minutes, resulted in the pressurizer emptying briefly.
However, th3 HPI actuation and flow rate is sufficier.; to pre-vent ara steam voiding in the RCS.
The design basis steamline break (SLB) accident (Section 4.2) produces steam voiding in the upper hot leg regions of the RCS.
Several sensitivity studies were performed to assess impact on steam void formation and subsequent core cooling flow. The sensitivity studies included: a) varying the time of loss-of-offsite power (LOOP) from time of trip to time of ESFAS, b) with and without core decay heat, c) single failure assump-tions of stuck open relief valve on unaffected steam generator or loss of one HPI pump, and d) moving the break from the steam generator with the pressurizer in its loop to the side without the pressurizer.
In all analyzed cases, core flow continued, or if interrupted, resumed immediately upon collapse of the void in the unaffected steam generator side of the RCS.
The core region remained subcooled throughout the transient for a'i cases analyzed. The results of SLB cases are presented in Section 4.2.
Mitigative operator action was not assumed in the analysis in the first 10 minutes. From a review of potential operator actions during this time it is concluded that only two actions are of major importance. Operator control of the steam generator level would have reduced the extent of RCS inventory shrinkage for both MFW overfeed and SLB transients. A non-mitigative operator action would reault from the premature cut-off of the HPI flow.
Indication to the operator during steam voiding situations such as occurred during the SLB accident analyzed are to maintain HPI flow since pressurizer level and subcooled margin both indicate the necessity of HPI.
Adequate core cooling would necessitate that HPI makeup to the RCS be avail-able at some point during the course of the overcooling transients.
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O 2.4 Adequacy of Core Protective Measures Section 5.0 provides the details of the design basis for operating transient cycles. Operating plant data has shown gp the 40 cycles of actuation of HPI to be sufficient design basis to cover automatic initiation arrival rate for this system. The analysis presented in Section 4.0 confirms that the most severe overcooling events require ECCS actuation.
The operating plant data shows that ESFAS automatic actuations occur <1/ year and, therefore 40 cycler / lifetime is an adequate gg design for transients not expected to occur >40/ lifetime time of the plant.
3.0 ANALYTICAL TECHNIQUES 3.1 Computer Codes gg The B&W certified computer code TRAP 2 (Reference 1) has been used in the analyses presented in the following sections.
This computer code is a nodal type, digital simulation (similar to CRAFT 2, Reference) capable of handling rapid overcooling transients that may result in two-phase fluid conditions in II the reactor coolant system.
The noding flow path networks used in the TRAP 2 analysis of the plant are given in Figures 3-1 and 3-2.
A description of each node and the important flow paths are given in Tables 3-1 and 3-2.
The more detailed noding shown in Figure 3-1 (description in Table 3-1) is referred to as maxi-TRAP, the II less detailed model in Figure 3-2 (description in Table 3-2) is referred to as mini-TRAP. The more detailed maxi-TRAP model is used during the initial phase of the transient while the primary and secondary variables are rapidly changing.
In the interest of computer calculational time saving, the mini-TRAP model is used in the long-term solution where system (D
variables are more slowly varying and the additional noding is not required.
3.2 Transient Selection The types of overcor?ing events considered include (a) those gp which constitute the initiating event, (b) those which result from single failures following any initiating event, and (c) those which are made more severe from single failures following the initiating overcooling event. The specific systems whose malfunction or failure are considered either as initiating events or single failures which enhance overcooling are:
gg A.
Feedwater heater failure which causes a decrease in feedwater temperature, B.
Feedwater flow control malfunction that causes an increase in feedwater flow, gg ab-6 J2}
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C.
Steam pressure regulator malfunction which causes increased steam flow, D.
Inadvertent opening or stuck o*en steam relief valve c
which causes increased steam flow and/or depressuriza-tion of a steam generator, and E.
Steam system piping failure which causes excessive steam flow and depressurization of a steam generator.
The SAR analyses are referred to in order to narrow the most severe type overcooling events for consideration. Mors specifically:
a)
Events which constitute initiating events A through D are moderate frequency of which steam regulator malfunction is most limiting according to the SAR analyses. E is a design basis event for which the double-ended rupture (DER)
MSLB is limiting.
b)
Events which result from single failure following any initiating event: This infrequent occurrence is a com-bination of a moderate frequency event plus one of A through D occurring as a single failure. The event chosen to be analyzed in this category is an imnediate reactor trip on turbine trip signal (decrease the heat source), combined with a feedwater flow control malfunc-tion that allows continued main feedwater flow (increase the heat aink).
c)
Events which are more severe frem single failures following the initiating overcooling event:
The limiting design basis overcooling transient to be a double-ended SLB.
The single failure chosen to maximize continued long-term cooling is a stuck open relief valve on the unaffected steam generator.
The limiting or potentially limiting overecoling cases to be analyzed as discussed above are summarized in Table 3-3.
3.3 Basic Assumptions Key input parameters used in the plant analysis are given in Table 3-4.
These values represent as-built information, realistic setpoints, actuation times, flow rates, and valve closures. Other system parameters not listed are those applicable to the plant design. The assumption of stuck rod was removed from the shutdown rod worth, thereby being more realistic in a conservative direction for the overcooling type events concerned with maximum RCS coolant shrinkage.
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O Single failures of active components assumed in the analysis are given in Table 3-3.
Some parameterization of the single failure assumption is done for the limiting overcooling case.
Since only safety grade equipment is assumed to function, the GD singic failures of mitigative equipment is limited. Table 3-5 lists the equipment assumed to function for each transient analyzed.
No mitigative operator action is assumed for 10 minutes in the analysis.
dp 4.O RESULTS OF CORE COOLING STUDIES 4.1 Anticipated Transients 4.1.1 Scope of Evaluation
()
The anticipated transients analyzed in the SAR's were reviewed for cooldown rates and consequences in order to select the most limiting case for shrinkage. Oper-ating plant data was also reviewed. From this review, the transient with highest frequency of occurrence and gg the potential for greatest overcooling was due to mal-functions resulting in overfeed of the steam generators by main feedwater.
Operating plant data shows that overcooling of the RCS has occurred from primarily two types of events: 1) gg Failure of a relief valve to reseat at the proper pres-sure, which limits the overcooling to the saturation temperature of the pressure at which the valve does reseat. 2)
Overfeed of the steam generators following a reactor trip, which has caused the greatest primary cooldown observed. Steam pressure regultor malfunctions II that allow increased steam flow would represent over-cooling by depressurizing the secondary system. Its effect is very similar to a small SLB analysis. The arrival rate for this transient has been zero at oper-ating B&W plants; therefore, in the limited time frame for the preparation of this report, the MFW overfeed transient is presented. The MFW overfeed represents O
the maximum cooling that can be achieved by feeding the OTSG's first with uncontrolled main feedwater and then, after ESFAS, with colder auxiliary feedwater.
4.1.2 Main Feedwater Overfeed Analysis O
The initiating event is a turbine trip with simultaneous reactor trip and a control failure such that main feed-water continues to feed both steam generators at full capacity.
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The sequence of events for this transient is given in Table 4-1.
The analysis was performed using the models and assumptions given in Section 3.0.
Time constraints dictated that this analysi,7 be ccmpleted entirely with maxi-TRAP. Neither credit for ICS nor operator action was assumed.
Figures 4-1 through 4-8 present the system parameters.
The pressurizer does not empty during the 10 minutes period and no voiding occurs. The cooldown rate was slow enough that HPI flow adequately compensated for system shrinkage.
Without additional means to control OTSG levels, the steam generators will overfill with main feedwater.
Subsequently, the steam lines will become filled with water. This was found to occur prior to ESFAS isolation of the main feedwater.
It was assumed for the analysis that no immediate operator action was taken and water relief out of the safety valves was permitted. After ESFAS isolated main feedwater, auxiliary feedwater flow was then continued to the steam generators to maximize the cooling rate.
Frcm the system response observed, two probable operator actions during the course of the transient are suggested.
First, operator action would be needed to terminate the OTSG overfill by main feedwater early in the transient, which would stop the overcooling of the RCS.
Also, since sufficient subcooled margin exists throughout most of the transient, the operator would regulate HPI flow to maintain pressurizer inventory. However, this particular action is not required for the first 10 minutes of the transient.
4.1.3 Conclusior.,
The RCS coolant inventory remains subcooled throu;hout the transient, thus assuring core cooling. HPI act?ation and flow rate was sufficient to prevent the pressurizar from emptying throughout the 10 minute transient time.
Only additional failures, such as bypass relief valves stuck open, could increase the cooldown rate experienced during the transient. ESFAS terminates the excessiva feadwater flow. With the fill rates of main feedwater assumed, the steam generators will overfill in about 90 seconds.
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O The RC pumps running cast presented represents the maximum cooling rate. Therefore, no voiding for this case assures that the RC pump trip case also would not provide voids in the RC system.
O 4.2 Accidents 4.2.1 S_ cope of Evaluation Maximum overcooling of the RCS results from an uncon-trolled blowdown of the secondary plants, i.e.,
steam GD line break accident (SLB). The double-ended rupture from full power has been demonstrated in the SAR to result in maximum overcooling. Selection of the worst coolant inventory shrinkage case for this event has been studied by analyzing a spectrum of different con-ditions. Table 4-2 shows the various conditions and gp identifies these different analyses by case number for further reference in the discussion of results provided in following sections.
4.2.2 SLB Analysis O
In double-ened guillotine break is assumed to occure in the 28-inch ID eteam line. The location of the break is outside of the reactor building. Other system para-meters, models and assumptions are as presented in Section 3.0.
The sequence of events is given in Tables 4-3 through 4-10 for each case analyzed. The figures for each case are listed on the table for that case. The figures for Case 1 RC Pumps Running also include a comparison of maxi-TRAP and mini-TRAP results, showing reasonably good agreement between the two models. Subsequent SLB O
analyses were performed using the mini-TRAP model.
The SLB accident was anaP, zed for 10 minutes assuming no mitigative oper: tor action and only safety grade equipment for transient mitigation. The overcooling rate as would be anticipated is much higber for the SLB cases than that obtained for the MFW overfeed cases II presented in the previous section.
The case resulting in the most severe consequences of RCS shrinkage occurs with LOOP at time of ESFAS actua-tion. The assumption of no decay heat aggravates this shrinkage effect. A bubble rise velocity of 5 ft/sec GD was used in the hot leg piping nodes.
It is important to note that the void fonnation data presented includes O
ab-10 1 5 '2 1 0 3 4'
O entrained, as well as sepsrated bulble masses. Therefore, with RC pumps running and during ths start of flow coastdown, the bubble mass will be a 'rost totally entrained. Comparing the single fail.tre assumption of II a stuck-open relief valve on the una'.fected steam generator versus failure of one HPI pump, the stuck-open relief valve (Case 4) results in the maximum steady steam void formation. However, for the one HP' failure (Case 6), the steam void remains in the RCS longer.
The maximum steam void occurs in the hgt leg attached II to the pressurizer and is about 500 ft for the cases analyzed.
The first steam void formation that appears during the SLB accident is due to flashing, i.e.,
reaching satura-tion, in both hot legs. This occurs prior to the pree-4D surizer emptying. On the loop side opposite the pres-surizer and analyzed with the unaffected steam generator, this effect is small and returns to a solid, subcooled state about the time the pressurizer empties. On the loop with the pressurizer and the affected steam gen-erator, this steam void can'.inues to increase as the
()
pressurizer empties. ESFAS initiation also occurs at about this time and HPI injection, as well as isolation of the affected steam generato- 'ain steam and feedwater, tend to limit the size of the Lceam void formed. HPI flow is sufficient to overcome the shrinkage that is still ocurring from the heat removal through auxiliary feed-gg water to the unaffected steam generator. As refill and repressurization of the RCS continue by the HPI, the steam void is quenche? and collapsed. Core flow is main-tained throughout the transient. During LOOP cases, natural circulation is maintained by the cooling from the unaffected steam generator side of the RCS.
e Without additional means of steam generator level control, the auxiliary feedwater fills the unaffected steam generator in 6 to 7 minutes. The pressuri:er is filling, but has not complctely filled in the first 10 minutes of the accident. Thus, adequate time is available for operator action to prevent pressurizer overfill. Level g,
control on the unaffected steam generator would allow earlier repressurization of the RCS; thereby leading to earlier collapse of the void.
It was assumed for Fie analysis that no immediate operator action was taken and water relief out of the safety valve was permitted.
The most probable operator action during a steam line break would be to control auxiliary feedwater to the intact steam generator to maintain level and secondary pressure.
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4.2.3 Conclusions Steam void formation in the upper hot leg regions was found to occur during the steam line break accident.
The magnitude and duration of the steam void formation 45 varied with conditions under whic's the analysis was performed.
In all cases, core flow was maintained or, if interrupted, resumed upon col'. apse of the void in the unaffected steam generator side of the RCS.
In all cases, the core remained subcooled.
O Some of the specific phenomena noted for the various cases analyzed are:
- 1. The LOOP assumption at ESFAS produces slightly worse consequences then at an earliet time. This is because the pumps running maximize the overcooling, such (p
that the later the LOOP (up to ESFAS) the morc shrinkage that has occurred.
LOOP after ESFAS should not con-tinue to increase the severity, since isolation of the affected steam generator main feedwater supply occurs at ESFAS and greatly reduces the overcooling rate.
gp
- 2. The assumption of no decay heat aggravates the steam voiding situation. However, as decay heat level decreases, the need for additional core flou decreases.
In the extreme, no decay heat implies no core cooling is necessary.
gg
- 3. Single failure of a relief valve en the unaffected steau s3nerator to maximize cooling rate and a single failure of 1 HPI pump to maximice the refill-repressurization effects were examined. The larger magnitude of steam void occurred for the stuck-open gg relief valve case; whereas the steam void formation was of longer duration for the EPI failure case.
- 4. The void formation in a given loop was large enough to create temporary flow blockage in taat loop.
However, the net core flow remains positive through most of the analysis, and is never interrupted to II the point that saturation occurs in the core region.
No mitigative operator action was assumed for 10 minutes in the analysis. With the fill rates of auxiliary feedwater assumed, the unaffected steam generator will overfill in 6 to 7 minutes. Core GD cooling appears adequate for all cases analyzed, since subcooled conditions are maintained in the core region.
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5.0 DESIGN BASIS FOR CORE PROTECTION Required ECCS and RPS actions ne:essary to protect the core have been summarized in Table 3-5 and discussed for each transient in Section 4.0.
No operator action has been assumed within 10 minutes for mitigation in the analysis. The purpose of this section is to demonstrate that the design criteria for the number of actuation cycles is adequate.
Twenty-five different types of transient cycles ('everal are SAR analyses) are used in evaluating the acceptable number of design cycles. These operating transients are listed in Table 5-1 along with the number of design cycles for each transient type.
This data is the basis on which the stress evaluation is performed for the plant and will be contained in Section 5.7 of the Standard Technical Specifications for the plant (per NUREG 0103, Rev. 3).
The number of cycles for transient types listed in Table 5-1 is not meant to be an absolute limit but were chosen on the basis of expected frequency (plus margin) and shown to be acceptable in the stress evaluation. Special transient analyses can be performed based on any actual transient data, thereby allowing categoriza-tion of the special case into one of the allowable transient design cycles.
The adequacy of the number of design cycles can be inferred from operating plant data. Table 5-2 compares the actual arrival rate for RPS and ESFAS actuation to date on plants of B&W design to the rates allowed by the design basis (Table 5-2).
The operating data is less than the allowable actuation rate for both systems, thereby supporting the adequacy of design.
6.0 REFERENCES
1.
J. J. Cudlin, P. W. Dagett, " TRAP 2-FORTRAN Program for Digital Simulation of the Transient Behavior of the Once-Through Steam Gererator and Associated Reactor Coolant System", BAW-10128, Babcock & Wilcox, Lynchburg, Virgina, August 1976.
2.
R. A. nedrick, J. J. Cudlin, and R. C. Foltz, " CRAFT 2-FORTRAN Program for Digital Simulation of a Multinode Reacter Plant During Loss of Coolant," BAW-10092, Revision 2, Babcock and Wilcox, Lynchburg, Virginia, April 1975, ab-13 5 J c] *l 037 C
TA1LE 3-1: MAXI-TI:AP ? 0DE A!!D PATil DESCRIPTIO::S NODE I.1'53ER sDESCRIPTIO:1 1
RV lower plenum up thru the bottom of the core support. sheets.
2 Reactor core, core bypass, upper plenum, and outlet nozzles.
3 Hot Icg A, includes SG A upper plenum and upper tube support sheets.
4-13
,SG A tube region between tube support sheets.
14 A cold legs, including lower tubesheet, SG A lower plenum, and RCS pumps Al&A2.
15 RV annulus, includes inlet nozzles.
^
16 Hot leg B, includes SG B upper plenum and tube support sheets (upper).
17-26 SG B tube regions (primary) between the tube support sheets.
27 B cold legs, including lower tubesheet, SG B lower plenum, and RCS pumps Bl&L2.
28-30 Pressurizer.
31-40 SG A secondary side heat transfer region.
41-50 SG B secondary side heat transfer region.
51 SG A steam riscr.
' ~ '
52 SG B steam riser.
53 Main Steam Line 1 to the MSIV (SG A).
54,57 lbin Steam Line #2 to the MSIV (SG A) - Split to model a DESLB.
55,58 lbin Steam Line #3 to the IGIV (SG B) - Split to model a DESLB.
56 Main Steam Line #4 to the MSIV (SG B) 59 MSL #153 from'the MSlV's to the cross pairing (denoted here as "X").
60
..SL #2&4 from the MSIV's to the cross pairing (denoted as "Y"T.
61 From cr:ss pairing X to the turbine stop valves on X - includes h cross connection.
62 From cross pairing Y to the turbine stop valves on Y - includes cross connection.
63 Turbine.
64 From the !!FW pumps to llTR bank #6 65 FW litt banks #657, and associated piping.
1521 038
O T/ini.E 3-1 : MAX 1-TRAP ::0DE N:D PATl! DESCRIPTIO::S CO::T'D
?;cor ::' mr.R DESCRIPTIO.';
O f
66 From W litr bank #7 to the W fork (does not include any fork piping).
. 67 From the WL fork to MFIV "B".
O 68 From the WL fork to MFIV "A".
69 From MFIV B to SG B inlet.
70 From MFIV A to SG A inlet.
71
. SG B downcomer.
O 72 SG A downcomer.
O O
O O
O O
.21 039 O
Table 3-1 (cont'd.)
PATit DESCRIPTIONS 1
i PATH NC4BER DESCRIPTION 1
Core path 2
Core bypass 3,17 Paths from the re, or to the hotlegs 4,18 Paths from the hotlegs to the steam generators 5-13, 10-27 Steam generator primary paths 14, 28 Paths from the steam generators to the cold legs, including the RCS pumps 15, 29 Paths from the cold legs to the RV downcomer 16 Path from the downconer to the RV lower plenum 30 Pressurizer surge lines 51-54 Pressurizer flow paths 31-39, 41-49 Steam generator secondary flow paths (0,50 Paths from the SG heat transfer regions to the risers 55-58 Paths from the SG risers to main steam lines (MSL)
- 1-#4 60,61 MSL #2 & #3 59, 62-64 MSLV #1-#4 65,66 Paths from the junction of MSL #1,3 and 2,4 to the turbine stop valves (TSV's) 67-70 TSV #1-#4 71 Cross connection 72,73 FUTI pumps 74 Path from the MFW pumps to MFW HTR banks #6 & #7 75 Path from MFW banks #6 & #7 to the MFW branch 76,77 Path from the MFW brach to the MSLV's 78,79 MFIV "A"
& "B" 80,81 Paths from the MFIV's to the SG downconer 82,83 I.ths from the SG downconer t the SG heat transfer region 84,85 AFW flow paths 86 HPI 87 LPI 88,89 SLB paths 1521 04u
Table 3-2 (Cont'l.)
PATil DESCRIPTIO.M O
PATil NUMilER DESCRIPTIE 1
Core 2
Core bypass 3,11 llot leg 4,12,38,39 Ilot leg candy canc 5,13 I!ot leg piping (candy cane to SG upper plenum) 6,7,14,15 Steam generator primary tubes 8,16 RC pumps 9,17 Cold leg piping 10 Downconer, reactor vessel 18 Pressurizer surge line 19 Steam piping crossover 20,27 Main feedwater pumps O
21,28 reedwater piping 22,23,29,30 Secondary heat transfer region steam generator 24,31 Steam riser, SG 25,26,32,33
} bin steam piping 34,35 HPI pumps O
36,37,43,44 Auxiliary feedwater paths 40,41 Breaks 42 LPI 45 Stuck-open relief 9
'\\
1521 041
TABLE 3-2: MINI-TRAP NODE AND PATt! DESC:IPTIONS NODE NUM3ER Dy:SCRIPTION 1
RV lower plenum to bottom of support plates.
2 Core, core bypars, upper plenum, outlet nozzles.
3 llot Leg.
4 Candy Cane 5,6,7 Steam Generator prirary tube region 8
Hot Leg 9
Candy Canc 10,11,12 Steam Ceacrator primary tube region.
13 Cold Leg 14 Cold Leg 15 Recctor vessel downcomer 16 Pressurizer 17,18,19 Steam Cencrator secondary tube region, 20,21,22 Steam Cencrator secondary tube region.
23 Feedwater piping 24 Fe'edwater piping 25 Steam riser 26 Steam riser 27 Steam piping 28 Steam piping 29 Turbine 30 Atmosphere f
"J i i U N L'
+6 E
W6#%4'A*O a
W M2'e9 9 eng=egaMg g
ae""'** * *rw 4 M=Mw we 64e6es*Oe M-es we e -
eA ee ee.e se y%.
ma w a=m hM
- gg
TABLE 3-3: SUmfARY OF E'v*ENTS ANALYZED t
i l
i INITIATIfE EVEiff SItELE FAlll@E SENSITIVITYSTUDIES l
I g jP.
Anticipated Event Made More Severe Bv Singic Failure Reactor Trip / Turbine Trip Main Feedwater Overfeed g ff.
Design Basis Overcooling Double Ended Steam Line Break Main Steam R211e c LOOP at Reactor Trip Valve Stuck Open o LOOP at Low RC Pressure ESFAS Trip LJ7 o Decay IIcat Ps) o llPI Single Failure CE) o Steam Generator Level Control 4>>
Lra o Break on Different OTSGs
_ Tabic 3-4: OVERCOOLI!!G
- /1YSIS I'.Tt:T Asst' PTIO!!S Parancter 205 FA Power Lcyc1 102::
O T,yc, *F 597.5 RCS Operatir.g Pressure (at Pressuri:cr tap),
- 2195, psig Pressurt:cr Lcyc1 (indicated), in.
585 g
PTS Trip Signals IIigh Flux, % FP 105.5 Low Pressure (core outict), psig 1950 ESFAS Trip Sctpoints G
Lou RC Prass., psig 1585 Low SC Press., psig 585 ESl'AS Trip Delay, sec.
2.5 ISIV Closurn Tine, sec.
5 g
1GIV Closure Ti=c (lincat ramped t.rea', sec.
6 Auxiliary Feedwater Design Capacity Turbinc, spa 1620 g
lictor, gp=
810 (onc per generator)
Tenperature. *F 40 Initiation Time After ESTAS, sec.
I?ith Offsite Power 74 4
111th Loss of Offsite Power 40 liain Feedwater Tc=perature, 'F 465 IIPI System Design Capacity per Purp, Sp3 2 pumps G 700 each O
Temperature, 'T 40 Boron Conce.ntration, ppm 2270 Initiation Tine After ESFAS, sec.
I?ith Offsite Power 20 0
I?ith Loss of Offsite Power
.25 MSC Outlet Pressure, psig
'1050 O
1521 044
I i
I f
TABLE 3-5:
EQUIPMENT AND RELATED C'lSTDIS ASSUMED f
TO FUNCTION ESFAS i
MSLIS MSIV RC TURBINE TURBINQ TRIP l
EVEN
'tPS/CRDCS FOGG AFU
i I
l
~
t, Reactor Trip / Turbine Trip with MFW Overfced X
X X
X X
X X
X g
Steam Line Break (Doubic-Ended Rupture)
X X
X X
X X
GL) i i
4 i
X Denotes systc= used when needed in the analysia 3
I Denotes syste:s not used in the analysis l
?.
tri (21)
Loss of offsite power cases assume 4 pump coastdown
~
ro m
C 4
U1 t
1
TABIE 4-1 MAIN FEEDI.'ATER OVEnFEED SEQUENCE OF EVENTS s
EVENT TIME (SEC.)
Reactor Trip Signal 0
Rods Begin to Dron
.4 ESFAS Signal on Low Primary Side Pressure 18.4 Main Feedwater Isolation Valves Ecgin to Close 20.9 Auxiliary Feedwater Begins to Flow to Both Generaters 25.8 MSIV Closes 25.9 Main Fecdwater Isolation Valves Close 26.9 IIPI Flou Begins 38.4 Steam Generator B Tube Region Full of Liquid 200 Steam Generator A Tabe Region Full of Liquid 220 (Refer Figs. 4-1 to 4-8) e 1521 046
Table 4-2 SLB Sensitivity Studies RC pumps LOOP at LOOP at LOOP at ESFAS, Steam line break running reactor trip ESFAS with no decay heat liith stuck open relief valve en un-Case 1(")
Case 2 Case 3 affected generator, 211PI pu=ps Case 4 available Case 8 (d)
With failure of one IIPI pump, no Case 5 Case 6 (b) r:uck open relief valve Case 7 (c!
- Maxi-/ Mini-TRAP comparison presented for this case.
(b)The SLB occurs in the LOOP with the pressurizer.
(C)The SLB occurs in the opposite LOOP from the pressurizer.
(
The Case 4 was re-analyzed with failure of FdGG to maintain the affected steam generator isolated.
L9 FN)
-t=,
%1
~
9 9
9 9
9 9
9 9
9 9
9
i TABLE 4-3 DOUBLE ENDED STEAM LINE EREAK CASE 1 - NO LOOP SEQUENCE OF EVENTS EVENT TUfE, s Double Ended Rupture of 28" Steam Line Between SG and MSIV 0.
Closure of Turbine Stop Valves.
O.
High Flux Trip Sctpoint Reached 2.9 Rods Begin to Drop 3.3 ESFAS Signal on Low Primary System Prescurc 9.3 Main Feedwater Isolation Valves Begin to Close 11.8 Pressurizer Emptics 12.0 Auxiliary Fecdwater Begins to Flow to Intact SG 16.7 MSIV Closes 16.8 Main Feedwater Isolation Valves Close 17.8 HPI Flow Begins 29.3 Unisolated SG Drys Out 50.0 Pressurizer Starts to Refill 270.
(Refer Figs 4-9 to 4-20) 152i 040
. -. - _ ~. _....
O TABLE 4-4 DOUBLE E!:DED STFJ.M LI!'E DREAR
,e CASE 2 - LOOP AT TRIP SEQUENCE OF EVE:iTS EVENT TDIE, s e
Doubic Ended Rupture of 28" Steam Line 0.
Between SG and MSIV Closure of *urbine Stop Valves 0.
g liigh Flur. Trip Setpoint Reached 2.9 Loss of Offsite Power; Main Coolant Punps Begin to Coastdat.n 2.9 3.3 Rods Begin to Drop g
ESFAS Signal on Lew Secondary Systes Pressure 8.1 Main Feedwater Isolation Valves Begin to Close 10.6 Pressurizer Emptics 15.
MSIV Closes 15.6 g
Main Fceduater Isolation Valves Close 16.6 Pressurizer Empties 18.
HP1 Flow Begins 33.1 Auxiliary Teedwater Begins to Flow to Intact SG 48.3 g
Unisolated SG Drys Out 46.
Pressuriser Begins to Fill 270.
O (Refer Figs. 4-21 to Figs 4-29) 1521 049
. -... -. ~.......
.,.... ~.
TABLE 4-S
~
DOUBLE ENDED STElli LINE EREAK CASE 3 - LOOP AT ESFAS SEQUENCE OF EVENTS
- EVENT TIME, s.
Double Ended Rupture of 28" Steam Line Between SG and MSIV D.
Closure of Turbine Stop Vcives 0.
High Flux Trip Setpoint Reached 2.9
~
Rods Begin to Drop 3.3 ESFAS Signal on Low Primary Systes Pressure 9.3 LOOP and Main Coolant Pump Coastdown Begins 9.3 Main Feedwater Isolation Valves Begin to Close 11.8 Pressuriner Emptics 14.0 MSIV Closes 16.8 Main Feedwater Isolation Valves Close 17.8 HPI Begins to Flow 34.3 Auriliary Feedeater Begins to Flow to Intact SG 49.3
- 47. -
Unisolated SG Drys Out Pressurizer Begins to Refill 270.
(Refer Figs. 4-30 to 4-38) e d
IscI UJU l
. - - _ ~ -
~
TABLE 4-6 DOUELE ENDED STEAM LINE EREAK CASE 4 - LOOP AT ESFAS NO DECAY llEAT GD SEQUENCE OF EVENTS EVENT TIf!E, s O
Doubic Ended Rupture of 28" Steam Line Ectween SG and MSIV O '.
Closure of Turbine Stop Valves 0.
Iligh Flux Trip Sctpoint ncached 2.9
?
(p Rods Legin to Drop 3.3 l
ESFAS Signal on Low Primary System Pressure 9.3 LOOP and 1:ain Ccolant Pump Coastdoun Begins 9.3 Main Feeduater Isolation Valves Ecgin to Close 11.8 g
Pressuri:cr Emptics 14.
MSIV Closes 16.8 Main Fecduater 1 solation Valves Close 17.8 IIPI Flow Ecgins 34.3 (p
Auxiliary Tecduater Begins to Flow to Intact SG 49.3 Unisolated SG Drys Out 48.
Pressurizer Ecgins to Till 355.
(Refer Figs. 4-39 to 4-47) 1521 051
I TABLE 4-7 DOUBLE ENDED STEAM LINE BREAK CASE 5 - LOOP AT ESFAS IIPI FAILURE SEQUENCE OF EVENTS EVENT TIME, s Double Ended Rupture of 28" Stcan Line Between SG and MSIV 0.
Iligh Flux Trip Setpoint Reached 2.9 Rods Begin to Drop 3.3 ESFAS Signal on Low Primary System Pressure 9.3 LOOP and Main Ceolant Pu=p Coastdown Begins 9.3 Main Feedwater Isolation Valves Begin to Close 11.8 Pressurizer E=pties 14.0 MSIV Closes 16.8 Main Fecduater Tsolation Vnives Close 17.8 IIPI Flow Begins 34.3 Unisolated SC Drys Out 48.
Auxiliary Feeduater Begins to Flow to Intact SG 49.3 Pressurizer Begins to Fill 490.
(Refer Figs. 4-48 to 4-56) 1aiI UsL
-. ~ - -........ _. _ -
GD TALLE 4-8 D0'JELE ENDED STEM! LI::E LREAK
SEQUENCE OF EVENTS e
EVENT TIME, s Double Ended Rupture of 28" Steam Line Between SG and 1:Slv 0.
gp liigh Flux Trip Sctroint neached 2.9 Rods Begin to Drop 3.3 ESFA6 Signal on Low Primary System Pres;urc
- 9. 3' Main Feeduatcr Isolation Valves Begin to Close 11.8 gg Pressurizer Emptics 14.0 MSIV Closes 16.8 Main Feedwater Isolatica Valves Close 17.8 HPI Flow Begins 34.3 gg Unisolated SG Dryc Out 48.
Auxiliary Feeduater Begins to Flew to Intact SG 49.3
> 6 00'.
Pressurizer Begins to Fill 9
(Refer Figs. 4-57 to 4-65) 1521 053
TABLE 4-9 DOUBLE ENDED STEAM LINE BREAK I
CASE 7 - LOOP AT ESTAS NO DECAY HEAT IIPI FAILURE NO STUCK SAFETY VALVE SEQUENCE OF EVENTS EVENT TIME, s Doubic Ended Rupture of 28" Steam Line Between SG and MSIV 0.
High Flux Trip Setpoint Reached-2.9 Rods Begin to Drop 3.3 ESFAS Signal en Low Frimary System Pressurc 9.3 Main Feedwater Isolation Valves Begin to Close 11.8 Pressuri:cr Emptics 14.0 MSIV Closes 16.8
^
Main Feedwater Isolation Valves Close
.17.8 HPI Flow Begins 34.3 Unisolated SG Drys Out 48.
Auxiliary Feedwater Begins to Flow to Intact SG 49.3 Pressurizer Begins to Fill
>600.
s e
I s & !.
bJu h
,O TiBLE 4-10 g
DOULLE ENDED STEAM LINE EREAK i
CASE S - LOOP AT ESFAS NO DECAY HEAT SV STUCK OPEN Gb SEQUENCE OF EVENTS TIME, s EVENT II Double Ended Rupture of 28" Steam Line tetween SG and MSIV and Closurc cf 0.
Turbine E2op Vcives 2.9 High Flux Trip Setpoint Reached GI 3.3 Rods Ecgin to Drop 9.3 ESFAS Signal on Low Primary System Pressurc LOOP and Main Coolant Pump Coastdown Begins 9.3 Mait. Tecdwater I clation Valves Begin to Close 11.8' II 14.
Pressurizer Emptics 16.8 MSIV Closes Main Feedwater Icolation Valves Close 17.8 54.3 HPI Flov_Lenins II Unisolated Stcan Cencrator Leys Out 48.0 Auxiliary Feeduatcr Flev Legins to Intact Stcam 49.3 Generator Isolated Stcan Ccncrater Pressurc Drops ID 58.
Belou 600 psia Auxiliary Feedwater Flow Ecgins to Unisolated 58.
Steam Generator 450.
Pressuri cr Tills Up O
(Refer Figs. 4-75 to 4-83)
O
} [)2 k O
a 0*
0
- 0'3' wv e
..t TABLE 5-1: Operatine Tr=sient Cycles Design Transient Cycles h' umber Transient De::crittien licatug and Cecidewn C;c=al Conditien)
I 10 50 Ffar heatup and cocidcum with no decay hes:
230 50 F/hr heatup and cocidekn with decay heat 0
240 Total 2
Power change O to 15'6 0;o=al Conditi=)
730 24 0
' and 15 to O'4 3
Pouer Loading 8?. to 100f4 power 0;omal 3000 l
. Condition)
~
' Pouer Loading 15'i to 100f4 power 0 o=r.1 15000 Condition) l 4
Power Unionding 200?. to 8'e power 0:c=al 3000 i
Conditien)
Power Unleading 100?4 to 15?2 power C;;= al 15000 Condition) 40000 Step Lead Increase (h'omal tendi:icn) 5 1074 40000 t:
1074 Step Lead Decrease Cic=al Cendi:icn) 7 Step Load Reduction frca 100f4 to S'; power e
(Upse: Cenditi=)
~
Resulting fr= turbine trip 150 Resulting fr = c1cc:ricc1 load re-150
~
\\
joction 300 Totc1 8
Reactor Trip (Upse: Cendition) 120 T)Te A 140 Type B 120 Type C Trips included in transient rt,::focrs 11, 15, 16, 17, L 21 112 492 Totc1 40 9
Rapid Depressuri=tica (Upset Ccndition) 10 Cidnge of Flew CJpse: Ccndition) 30 11 Rod Withdrawal Accident (Upse: Condition) 4.
20 12
!!yhotests (Test Ccnditien) 13 Steady-State tower Vkrictions 0:om:1)
=
Condition) 40 14 Centrol Red Drop (Upset Cend.itien) g3 lJ6l'h30 S
=
.o e
f of 9
e.
agemme-ew-.
.w we***==
N me.g
- m
-em
^
D O
O b
m O
ge uL
.i.i 42-(Cont'd)
Tui t'
- -- I Transient Design 1;ur.ber Trcnsient Cescrintica Cycles 15 Loss of Statien Power (Upse Condition) 40 16 Steca Line Failure (Faulted Condition) 1 17A Loss of Feedwater to Cnc Stecm Generator 20 (Upse
- Condition) 17B Stuck Open At=csphe-ic Dunp Valvo
./. -
- . 10 (Energency Conditien) 18 Loss of Feedwater Heat (Upset Conditien) 40 19 Fced and Bleed Operations 0 cr=n1..._ 18000 Conditica) 20 Misec11anceus A Cic=al Ccndition) 30000 1!isec11anccus B 4x106 Misec11anecus C 20C00 21 Loss of Coolant- (Faulted Condition) 1 I
22 Test Transion:
High Pressure Injection Systen (';c=al Cenditien) 40 Test Transica:s - Cere F1 ceding Sysica O!o=al Ccnditien) 240 23 Stem Generater Fill, Draining, Flushing
\\
and Cleaning ';;c=al Conditica)
Steam Cenerator seconda y side filling 240 S:can Generator Primary side filling 240 Flushing 40 Chemical CIcaning 20 540*
24 Hot Functional Testing Cicr:a1 Condition) 5 25 Leak Testing (.Tes Conditien) 100 e
1 t
g i.
e e
e 6
vw.m.,,..,..
.w e
TABLE 5-2 O
PSS/ESFAS FREQUENCY Actual Data Allowed Number Frequency Frequency g
No. of Reactor Trips (RPS) 228 6.95/yr 10/yr No. of Autocatic ESFAS Actuations 27
.816/yr 1.0/yr No. of Plants Included 9
32.8 Reactor-g Years e
1521 058
.gm a g o g gg). g Em M e M.Jull!L 5
5
...Q=
Qn T3r 13T e
1
.6
~
t q
e u
g
_ ", _ _3 m
m 3
8 g
m e;.,.,, 2 g
.", ". p g
g g
g 3
g s
e e
K 2
g 2
g.".
g ec '. . s m
g g
g g
g p-e 2
8 g
O; R
l,'
'l $
_gI C
2 Oc MD e
er
- e g
g g
g g
g, g ;;;,
,u.
a o
e i,,
5 n 's e n O
8 uO m
o m
M.,,
S-
_h 7.
~
g e
QD
= 00 1521 059 EAIl-TRAF N00 LNG 3CHERE, 205 FA Figur. 3-1
l 9
1 I
ll e
e x
x 2
30 f
f 27 28 T@T e
X X
g
. _.. g e
5 17 20 ft h
25 s
in g
e
~
21 11 e
is 22 2
[, _'e= a g
n g
e i.
2 O
15 m
O-e e
lLi wu O
e EINI*IRAP N0 DING 3CHENE. 205 FA O
Figure 3 2
D""
9f }N E-ow e
=
FIGURE 4-1.
MFW OVERFEED, 205 FA 850 600 HOT LEG
\\
550
% 's N
N p
500 N
g i
~N
=%
N 3
450 COLO LEG E
400 350 300 250 t
I I
O 100 200 300 400 500 600 Time, Sec 1521 061
O O
mm-o
'q-J o e ju o
. 2.
=
FIGURE 4-2.
NFW DVERFEED 205 FA 60 O
STEAN GENERATOR "B" 50 9
, 40
/
g STEAN GENERATOR "A" 5
" 30 E
20 e
10 8
0 I
i i
i g
0 50 100 150 200 250 Time, Sec 1521 062
FIGURE 4-3.
]D *
- g 'dJu 2._}kfru o
16 T g 9' g'
J.o sJu 15 14 13 12 11 3
9 3
8 2
a j
7 t
6 5
4 3
2 1
0 I
I I
I I
O 100 200 300 400 500 600 Time, Sec 152 i O t,a
e FIGURE 4-4.
STEAM GENERATOR A - NFW OVERFEEO, 205 FA 8
1500 e
1400 e
1300 a
d 5
9 a
[ 1200
=
5 e
-a 3 1100 E
f 5
e
~
1000 e
900 e
800 N
700 0
100 200 300 400 500 600 Time, Sec 1521 004 e
FIGURE 4-5.
STEAN GENERATOR 8 - NFW OVERFEED, 205 FA 1500 1400 1300
.2 E
51200 5
- a
" 1100 g
a 5
w a 1000 m
900 000 700 1
I I
I I
l 0
100 200 300 400 500 600 Time, Sec l52}
OOD
e I
e O
FIGURE 4-6.
NFW OVERFEEO, 205 FA 2400 0
2200 e
2000 E.
O 5
1800 E
b j
1600 e
1400
[
e 1200 0
100 200 300 400 500 60s Time, Sec 1521 006
FIGURE 4-7.
MF1 QVERFEEO, 205 FA 600 580
?
i S
S 8.
560 5
04 S.
540 a
520 500 460 460 -
i I
1 Q
100 200 400 500 600 Time, see n
i NU i
FIGURE 4-8.
NFW OVERFEEO, 205 FA 1.4 1.2 1.0 9-0.8 3
=
0.6
- 0. 4 0.2 t
- 0. 0 I ~
0 100 200 300 400 500 600 Time, Sec
\\S2\\ DM e
FIGURE 4-9.
SL3 gAx3. MINI TRAP COMPARISON, 205 FA 4000 3500 1
3000 a
e a
2500 a.
m a
E T>
.y 2000
\\
1 1500
\\
.,x,.1.,,
.,,,.1.,,
x
,000 x
sa 0
20 40 60 80 100 Time, Sec 1571 009
O FIGURE 4-10.
SLB NAXI-MINI TRAP COMPARSION, 205 FA 620 600 O
550 8
i
\\
E 560 F:
540 Z
3 520 MINI-TRAP 0
NAXI-TRAP 480 460_
t I
t t
I e
0 20 40 60 80 100 120 Time, Sec 1521 07u
FIGURE 4-11. SLB NAXI-MINI TRAP COMPARISON, 205 FA 20.138 4
16.782 k k
l 13.425 3
- 10.079 1
5 I
E 1
[
IL 6.712 MINI-TRAP
\\
i 3.356 MAXI-TRAP
\\
~
~
1 I
I I
I I
O 20 40 60 80 100 120 Time, Sec 7-lgn4 sci pJ / l
O O
FIGURE 4-12. STEAM LINE BREAF. CASE 1, 205 FA 1.4 9
1.2 9
1.0 d
.e
- 0. 8 E
e 0.6 0.4 g
1.-
0.2 g
k I
'~
8 0.0 0
100 200 300 400 500 600 Time, Sec bN?l 0/2
FIGURE 4-13.
STEAll LlHE MEAK CASE 1, 205 FA 3500 3000 3 2500
)
[ 2000 8
1500 l
1000 500 I
i
~
0 100 200 300 400 500 600 Time, Sec 1521 073
FIGURE 4-14.
STEAM LINE BREAK CASE 1, 205 FA 625 l _
600 -
575 I
b P
550 i
1 5
I e
i i
525
-1 E
\\
l I
500
\\
\\
\\
HOT LEG I
475 g
\\-
- ~~ s%
J
'%,s COLO LEG 450 s%
% %~~.
=
7.-
-...7 0
100 200 300 400 500 600 Time, sec.
\\S2\\ Di*
FIGURE 4-15.
STEAM LINE BREAK CASE 1, 205 FA 20.138 16.782 13.425 3
10.079 m
5
- 2 6.712 s
3.356 7. _ -. -.- -
7-t 0
100 200 300 400 500 600 Time, Sec 1rn.
n,-
I.J._. l U/D
O O
a
=.
O
=
w E
=
w a
w e
W oc 9
E
=w 5
==
w
=
a z
o, w
=
=
W 5
6 u
m kC s2
=
w C
=
- = ora E
3 e
.=c w
=
N T,
S w
a:
=
a G
Q O
.-T
=
=
=
=
=
=
o o
G Q
Q c
c, o
a
=
=
=
m e
e[sd 'ajnsssJd Jolejeusa meats 1521 076
ao W
o 3
=
.=-
=
=
w a
=
c w w
=
W M N
o
= 5
=
= M w
w w W g w
w M<a kg u
e3 w
E" M
M m
=
a 5
=<W
>=
M Q
.m
~
,w w
M
=3 o
Q
=
w J
1 1
I
=
c
=
=
E
=
a
=
E g
o
=
u elsd 'aJnsssJd JoleJauss mes;g I-.J V/[
4 e
e FIGURE 4-18 STEAM LINE BREAK CASE 1, 205 FA 60 0
50
/
O gg ugn j
40
/
/
~
O
/
5 l
i
/
3 30
/
5
/
e
=
/
/
N
/
20 7
l\\
/
e iV 10 l
SG "A" 0
100 200 300 400 500 Time, sec g
1q?1 n7n l J L.
I U$U q
FIGURE 4-19.
STEAN LINE BREAK CASE 1, 205 FA HOT LEG f
400
/
I 300 SG UPPER PLENUM h
I
~_
~
I d
3 LOOP "A" 3
200 O
i 100 CANDY CANE h
300 l
0 C
100 200 r9-n Time, sec JcI p) / '/
t 4
e e
e 8
m
+
8
=
/
/
E
/
w E
/
- =
=
u Y
e g
W 8
E
/
=
3 o
3
/
A
=
5
(
5 E
w 3
l
m=
e
(
./
=m O
=
=
=
g E
\\
5 E
\\.
5b
' ~ " '
d kB
=
~
~~
e 4
f,
\\
=
.N-(
=e I ~ --
- ~ r-O e
.o a
o il 'Bulp!aA e
1521 080
FIGURE 4-21.
STEAM LINE BREAK CASE 2, 205 FA 1.2 e
1.0
)
- 0. 8 g 0.8 0.4 LII T'd 0.2 c3 CO 0.0 1
0 100 200 300 40E 500 600 Time, Sec.
O O
O
=
S e
=
O
.n aa 9
m' ww3 a
R w
a
=
e.
I un e
g 5
o
=
=
<w
>=.
u O
A 8
w M8 C
3._
o o
=
=
8 3
E E
G n
~
elsd 'sJnsssJd 1a11110 8J00 0
C} '\\
00b
FIGURE 4-23.
STEAll LINE BREAK CASE 2, 205 FA 650 600 i
550 I 1
l
\\
$ 500 -\\
HOT LEG e
E O
N u\\
\\
n <50
\\ /
J
\\
~
N\\
N 400 N
COLO LEG N
350 300 I
I 1
250 0
100 200 300 400 500 600 Time, sec
\\S?)
00D
e FI6URE 4-24.
STEAll LINE BREAK CASE 2, 205 FA e
t 20.138 -
O 16.782 e
3
,; 13.425 -
e
~
E N
e n.
10.079 e
6.712 9
I 3.356 e
I ji i
i i
i i
0 103 200 300 400 500 600 Time, see 1521 084
FIGURE 4-25.
STEAM LINE BREAK CASE 2, 205 FA STEAM GENERATOR A PRESSURE 1600 1403 1200
=
I
=
1 g 1000 5
I 3
800 E-21
'l v,
600 400 t
200 0
100 200 300 400 500 600 Time. see
]gg) gqc
6 9
FIGURE 4-26.
STEAM LINE BREAK CASE 2, 205 FA STEAM GENERATOR B PRESSURE 8
1400 G
1200 E
l 9
J 1000 '-
a 3
5 m
a 3
800 - -
G 5
.a 9
600 400 -
e I
I I
200 g
0 100 200 300 400 500 6C0 Time, sec 1521 086
FIGURE 4 27. STEAM LINE BREAK CASE 2, 205 FA 60
/
50
/
/
i
/
/
/
40
/
/
STEAM
/
GENERATOR 8/p 30
/
3
/
/
a
/
5 20 f
/
/
/
s/
10 STEAM GENERATOR A 0
I O
100 200 300 400 500 600 Time, sec 1521 0217
e e
FIGURE 4-28. STEAll LINE BREAK CASE 2, 205 FA 8
300 4
i il 250
- li LOOP "A" lI lI e
200
-Il CAN0Y CANE g
-=
II e
.y 150
_a
=
Il
_l I
30, I
HOT LEG
'l l
50
-l I
I I
er
/
\\
I I
o i
I i
i i
t li i rs i
0 50 100 150 200 250 300 Time, Sec 1521 080 e
FIGURE 4-29. STEAM LINE BREAK CASE 2, 205 FA 600 6
LOOP "B" 500 (g
/
i v
\\
/
\\
I
\\VN 400 i
N.
I
\\
l
\\
CANDY CANE g
~
300 I
y i
1
'm E
/ SG 3
UPPER PLENUM
\\\\
200
\\
\\
f.
\\
NOT LEG iOO 1
\\
-.(
\\
CV
\\
I I N
0 50 100 150 200 250 300 Time. Sec 1521 089
O FIGURE 4-30.
STEAll LINE BREAK C.8.SE 3, 205 FA 8
1.6 1.4 0
1.2 1.0 I-E a.
.8 S
O
.6 9
.2 k
0.0 0
10C 200 300 400 500 600 Time, sec 4
nn g-m
}JZl U '/ U e
FIGURE 4-31.
STEAN LINE BREAK CASE 3, 205 FA 4500 4000 3500
.5 E. 3000 5
E 3
2500 5
u 2000 1500 1000 I
I I
I 500 O
100 200 300 400 500 600 Time, sec j1}']j Q C) l
O O
FIGURE 4-32.
STEAM LINE BREAK CASE 3, 205 FA 650 9
600 O
550 L
\\
\\
e 1
500 I
HOT LEG t
\\
i
\\
G e
2
\\
f
\\
2
\\/
K 450 9
\\N s
vs E
N O
s COLD LEG 400
,~
350 300 250
-t t
- - - r -- ~~ ~ ~
- ~ ~ T --~~
0 100 200 300 400 S00 600 Time, Sec 1521 092
O FIGURE 4-33.
STEAM LINE BREAK CASE 3, 205 FA 20 138 16.782 13.425 -
E N
310.079 5
=
6.712 3.356 -
1 J
i I
I i
i O
100 200 300 400 500 600 Time, sec jS2j ygg
O FIGURE 4-34 STEAM LINE BREAK CASE 3, 205 FA STEAN GENERATOR A.'RESSURE 1600 9
1400 1200 E
a i
G 3
1000 E
3 Ea 600 G
v, 600 9
e 400 G
200 O
g i
0 100 200 300 400 500 600 Time, sec 1521 094
FIGURE 4-35. STEAM LINE BREAK CASE 3, 205 FA STEAM GENERATOR 8 PRESSURE 1800 1600 1400 1200 3
m.
5 i
1 g 1000 a
=
800 e
2G 600 400 200 I
I I
1 I
Q 100 200 300 400 500 600 Time, sec lb2l b h' D
O O
FIGURE 4-36.
STEAM LINE BREAK CASE 3, 205 FA O
8
/
50
/
/
STEAM 7
GENERATOR "B"
/s 40
-J
/
,r I
/
i30 a
?
/v
/
/
20 7
E
//
10 STEAM GENERATOR "A" i
0 O
100 200 300 400 500 600 Time, Sec 1521 09u
FIGURE 4-37.
STEAM LINE 3REAK CASE 3, 205 FA LOOP "A" h
CAN0Y CANE I I 300 Ii l
^
iI I.*
II
\\
I
\\
li
\\
11 g
II I
\\
I s
200
-l i
g i
I
\\
g 1
\\
i i
i I
\\
1 i
\\
I 3
I i
\\
l i
I
\\
g i
l I
s 100 1
i HDT LEG SG s
l l
i-UPPER PLENUM
\\
l I
\\
l t
'/
j,
\\,
\\'1 0
0 100 200 300 Time, Sec 1521 097
O O
FIGURE 4-38.
STEAN LINE BREAK CASE 3, 205 FA 600 9
/
s/
)
LOOP "E" l
\\
I
\\
CAN0Y I
I
)
400 CANE 1
4 l
\\
I
\\-
~
g zoo l
\\
I
\\
\\
l
~
+luPPER PLENUM I
O
\\
k j
,_II\\
\\
\\
100
}
HOT LEG 7 g
/
\\
J g
I 1
1 I;i
\\
0
- 1. [
l
\\
0 100 200 8
Time, sec 9
1521 Gio e
FIGURE 4-39.
STEAM LINE BREAK CASE 4, 205 FA 1.6 l.4 P
1.2 1.0 E
.8 3
.6
.4
.2 1
f I
i i
0.0 0
100 200 300 400 500 600 Time, see I n'4 099 r
I
O FIGUP.E 4-40.
STEAM LINE BREAK CASE 4, 205 FA O
4500 9
4000 O
3500 0
3 3000 a
J B
\\
O E
2500 2
E I
2000 0
1500 O
1000 500 I
t 0
100 700 300 400 500 600 Time, sec 1521 100 e
FIGURE 4-41.
STEAM LINE BREAK CASE 4, 205 FA 650 -
600 c l
550 li
\\
=T 500 k
3
\\
HOT LEG
=
E
\\
7 450
\\
\\
\\
\\
400 N
I COLD LEG
% %*% N N
350 200 150 0
100 200 300 400 500 600 Time, sec U2l l0
O FIGURE 4 42. STEAN LINE BREAK CASE 4, 205 FA O
9 O
20.138 o
16.782 C
S
- 13.425 2
E O
10.079 6.712 3.356 0
100 200 300 400 500 600 Time, sec 1521 102
FIGURE 4-43.
STEAM LINE BREAK CASE 4, 205 FA STEAM GENERATOR A PRESSURE 1600 1400 -
1200 -
.2 E.
1000 a
a S
m 800 -
o E
3 600 u,
400 200 0
1 O
100 200 300 400 500 600 Time, see 15 1
103
FIGURE 4-44.
STEAM LINE BREAK CASE 4. 205 FA STEAM GENERATOR B PRESSURE e
1600 1400 1200
\\
ng 1000 E
E
. e 800
=
=8 3
600 400 200 I
I I
I 0
O 100 200 300 400 500 600 Time, sec 1521 104
FIGURE 4-45 STEAM LINE BREAK CASE 4, 205 FA 60 50 j
/
/
/
/
40
/
STEAM GENERATOR 'B" p
5
/
/
l
/
g 30
/
/
E l
/
/
/
3 20
/
=
i
/
u, N
h g
/
10 s /.
((i STEAM GEhERATOR '%"
0 100 200 300 400 500 600 Time, Sec
e e
FIGURE 4-46.
STEAM LINE BREAK CASE 4, 205 FA e
e i
I e
h 300 11 Il e
ll CANDY CANE i
i1 e
5 200
-1 l l!
LOOP "A" Il II Il ll e
100 bI HOT LEG g
ll Sa 8
l l
l UPPER PLENUM I
il I
\\
I.
g!
N
(
j gr O
1 i-li i
\\
f O
100 200 300 400 500 600 Time, Sec 1521 100 e
FIGURE 4-47.
STEAM LINE BREAK CASE 4, 205 FA 600
^
500 a
f g
j
/
/
\\
LOOP "B" g
\\
I
\\
l
\\
\\
400 f
g/
\\
CANDY CANE J
\\~-
s I
\\
l
\\
~
i I
\\
l
\\
3 300 s
i
\\
SG UPPER PLENUM
\\I
\\
200
\\
/
s
[G I
HOT LEG -
N I
100 J
j l
i
\\
l-Ts 0
0 50 100 150 200 250 300 Tims, Sec 1521 107
O O
O FIGURE 4-48.
STEAM LINE BREAK CASE 5,205 FA 1.2 9
1.0 i
O
.8 O
5 3
.6 E
e
.4 O
.2 O
L 0.0 1
0 100 200 300 400 500 600 0
Time, see 1521 10B O
FIGURE 4-40.
STEAM l.INE BREAK CASE 5, 205 FA 3000 2500
.2 8.
al f
2000 n.
E
%a 1500 U
u 1000 w
. g - -
- -'~
?
'I O
ggg 200 300 400 500 600 Time, see j J' L. l lU/
O FIGURE 4-50.
STEAM LtHE BREAK CASE 5, 205 FA 9
E25 -
9 600.-
O 575
?
\\
O i
\\
A 550 t il i
\\'
e g
525]1 a:
HOT LEG
\\
\\
e
\\
500 -
g
\\
,-~
l
/
\\
415 - \\
l
\\
\\
l
's g
g vs
\\J
%^
COLD LES
\\
450 -
\\'
~__.%,
S 425 1
~ --' i~
g 100 20"'
300 400 500 600 Time, sec
,^
S
t FIGURE 4 51. STEAM LINE BREAK CASE 5, 205 FA 20.138
)
18.782 13.425 7.
a 33 2
y 10.079 a.
8.712 3.356 0
~ r --
7..
I 200 300 400 500 800 Time, sec Ib2I llj
O O
FIGURE 4 52.
STEAM LINE BREAK CASE 5, 205 FA e
STEAN GENERI, TOR A PRESSURE 12 9
10 8.
E 8
9 a
g 6
o e
e.
4 L
9
\\
2 0
0 100 200 300 400 500 600 Time, sec 1521 112
FIGURE 4-53. STEAM LINE BREAK CASE 5, 205 FA STEAM GENERATOR S PRESSURE 1200 i
1000
.2 a
a 3
800 3
600 400
/
~
--7---.-....-.7....
.-. 7
.7..--._
0 100 200 300 400 500 600 Time, sec 1521 113
O FIGURE 4-54.
STEAM LINE BREAK CASE 5, 205 FA g
9 60 A f
9 50 C
STEAX GENERATOR B f
40
=
E.
30 G
0 20 e
10
\\
\\
\\
STEAN GENERATOR A
~ -~ ' T- - f -
~f
- r-I r
r--
- r-7-- 1 9
t s
100 200 300 Time, sec 1521 114
.e
FIGURE 4-55.
STEAN LINE BREAX CASE 5, 205 FA I
CANDY CANE 300 I
LOOP "A"
!ll f
\\\\
j\\
I, I
e "r
ll
/
'I
/
I '
i 200
-l i fg gj g g
,i/
vi,;A ii IJ f I Il f
ir CANDY CANE II I
tl I
\\
Y (f I
ll l
\\
il l
1 11 l
l 100 i
II i
i I I
HOT LEG i1 l l
g i
ii l,l*4 3,
UPPER PLENUM l
l Il 100 200 300 400 500 600 Time, see E?\\
\\\\0
9 i
FIGURE 4-56.
STEAN LINE BREAK CASE 5, 205 FA e
e f
1 I
SG UPPER I
300 l
fg g
PLENUM lg g
I fs I li fi llI I
\\
I It
\\j
- I
\\\\
e I
\\
ll I\\ i\\
N l
I I I I \\
~
y g
CAN0Y CANE i2aa i
\\
5 I
I
\\
\\
'N.
I e
i 1
I 1
p!!
I g
100 lg1 HOT Li!G I
11 '
~
i \\y, f
1 I
i r
l I
1 I
g
\\
,1
_.3
_... \\,
~
100 200 300 400 500 600 Time, sec 1521 116
FIGURE 4-57.
STEAM LINE BREAK CASE 6, 205 FA 1.2 1.0
.8 a
a.
.6
.4
/
.2
~g O.
100 200 300 400 500 600 Time, sec E
Isa?l e
O FIGilRE 4-58.
STEAll LINE BREAK CASE 6, 205 FA 3000 2500 p
3' 2000 3
1500 o
ea e
1000 500 I~
~~
~
~' - ~ '
~~ ' ~ ~
0 100 200 300 400 500' 600 Time, sec 1521 118
FIGURE 4-59.
STEAll LINE BREAK CASE 6, 205 FA 625 600 575 L
o'
\\
\\
550 q E
I 3
1 525 l
\\
\\
500 -1
\\
\\
\\
HOT LEG 415
\\
f %s~
\\
l N
\\
/
.s
\\
/
N 450 gf s
Il COLD LEG N
~
0 100 200 10 400 500 M
Time, sec 1521 119
e e
FIGURE 4-60.
STEAM LINE BREAK CASE 6, 205 FA 20.138 16.782 e
13.425 i
3 10.079
~
E e
6.712 e
3.356 e
0 100 200 300 400 500 600 Time, sec e
+
1521 120
FIGURE 4-61.
STEAM LINE BREAK CASE 6, 205 FA STEAN GENERATOR A PRESSURE 1200 1000 3
800 --
cm.
J e
3
[
600 Eo E
400 i
G 200 0
0 100 200 300 400 500 600 Time, sec 1521 121 1
~
FIGURE 4-62.
STEAM LINE BREAK CASE 6, 205 FA 1200
\\
1000 800 s
e a-600 a
=
8-l 400 200 O
I I
00 200 300 400 M
Time, sec 1521 122
FIGURE 4-63.
STEAM LINE BREAK CASE 6, 205 FA 60 50 O
-- 40 STEAM GENERATOR 8 3
a E 30 o
m 20 l-
\\
\\
10
-\\
g STEAM GENERATOR A N
\\
0 50 100 150 200 250 300 Time, sec 1521 123
FIGURE 4-64.
STEAM LINE BREAK CASE 6, 205 FA G
LOOP A 400 A
l \\\\
l N
s l
N
\\
l CANDY CANE l
N e
\\
300 f
I SG UPPER PLENUM CANDY e,:
s e
, l} CANE l
g g
i
-l t 1
s a
'lg I
s
~
Ig I
s '
I 00 I;
I I
I
'I e
I I
t 1-I HOT LEG j
i J
[
100 g
g
\\
l
\\:}(
e I
i 1 I
'N,N I
il 3, _ _ _
.Ny 100 200 300 400 500 600 Time, see 1521 124
FIGURE 4-65.
STEAM LlHE 6REAK CASE 6,'205 FA 200 LOOP B i
i i
CAN0Y CANE f
1
.g n
/
a I
.i 1
!I
=
,i 5
\\
\\
100 g,t I
\\q 1-1' s
'.' i 1
i
's s
I 1l s
s I
il s
s, il N
l s
r il N
\\
s,
\\
- g s
1, N
y s
J' g
\\
s 8
ss UPPER PLLMUM
\\
s
~'
100 200 300 400 W
M Time, see 1521 125
e e
FIGURE 4-66.
STEAM LINE BREAK CASE 7, 205 FA e
1.2 e
1.0 )
e
.8 E
E e
.6 O
e
.4 g
.2 e
- 0. 0
~-~~
f
- ~ ~-
t-----
r---
100 200 300 400 500 600 Time, sec 1521 120 e
FIGURE 4-67.
STEAll LINE BREAK CASE 7, 205 FA 3000 2500
=
5 k
2000
~
a.
5 1500
=
3 1000 500
~I I
~I ~
0 100 200 300 400 500 600 Time, sec 1521 127
O e
FIGURE 4-68.
STEAM LINE BREAK CASE 7, 205 FA e
\\
600 e
(
550 1
g 1
\\
I HOT LEG
- 500
\\
g
\\
~
\\
/
m E
\\
'~~__
j
*s 450
\\/
s e
\\
COLD LEG
\\
i
\\
N 400 9
s%
0 100 200 300 400 500 600 Time, see O
1521
]28
FIGURE 4-69.
STEAM LINE BREAK CASE 7, 205 FA 20.138 16.782 1
13.425 10.079 6.712 3.356 J
0.0
- ' ~
?
f I
O 100 200 300 400 500 600 Time, see 1521 129
O O
FIGURE 4-70.
STEAN LINE BREAK CASE 7, 205 FA STEAM GENERATOR A PRESSURE 1200 O
i 1000
=
800 g
8.
2 600 a
E e 400 200 0
I 1
O 100 200 300 400 500 600 Time, sec O
1521 130 0
FIGURE 4-71.
STEAM LINE BREAK CASE 7, 205 FA STEAM GENERATOR B PRESSURE 1200 1
1000 3
800 a
e E
600 E
a 400 G
200
.-w
.se.--
0 100 200 300 400 500 600 Time, sec 1521 131
O O
FIGURE 4-72.
STEAM LINE BREAK CASE 7, 205 FA 9
6 60 9
i 50 e
Y I
40 STEAM GENERATOR 8 O
Eo E
30 G
e 20 I
e 10 STEAM GENERATOR A N/
0
--t-r --
i 0
100 200 300 Time, see S
1521 132 e
FIGURE 4-73.
STEAM LINE BREAK, CASE 7, 205 FA - STEAM GENERATOR A 400 HOT LEG LOO, "A" 300 b
1 I
==
200 i
\\
CANDY CANE h,....E.
e
- mme, I
x~\\-
1
/
/
i m
400 500 600 Time, Sec
FIGURE 4-74.
STEAM LINE BREAK CASE 7, 205 FA - STEAM GENERATOR B 500 9
450 HOT LEG 400 g
LOOP "B" 350 300 5
250 E
STEAM GENERATOR 200 8
150 uf.
i CANDY CANE 100
[A\\
5
-l
\\
w
/
50
/\\
.o f
b i
g,N ~\\b I
/
h li i
i i
0 O
100 200 300 400 500 600 Time, Sec 1521 iM
FIGURE 4-75.
STEAM LINE BREAK CASE 8, 205 FA 1.4_.
1.2 1.0 q
s g-
- 0. 8
%_a 0.6 0.4 0.2 152i135 i
e
- 0. 0 _
0 100 200 300 400 500 600 Time, Cec
O FIGURE 4-76.
STEAM LINE BREAK CASE 8, 205 FA O
3500 3000 2500 h
5 2000 9
5 E
g 1500 8
I 1000 500 i
i i
i i
0 100 200 300 400 500,
600 Time, Sec 1521
,136
FIGURE 4-77. STEAll LINE BREAK CASE 8, 205 FA 600 l -
I i
L i
l I
500 t
\\
\\
k
\\/\\
a
\\
E
\\
400
\\
HOT LEG E
\\
\\v
\\
\\/^\\
coto 300 LEG
\\
\\
v N
\\j
/
\\/
200 I
I I
I I
'l 0
100 200 300 400 500
'600 Time, Sec 152) 137
e e
FIGURE 4-78.
STEAM LINE BREAK OASE 8, 205 FA 20.138 e
16.782 e
i 13.425
~
e a
3 10.079 e
6.712 e
3.356 e
k A
I I
I
_O 0
100 200 300 400 500-600 Time, Sec 1521 138 i
lGURE 4-79.
STEAM LINE BREAK CASE 8, 205 FA STEAM GENERATOR A PRESSURE 1400 9
1200 t
r 3
1000 - -
=-
W E
E 800 a
3 5
600
=
u, 400 e
200 V
I 8
^
0' O
100 200 300 400 500 600 Time, Sec 1521 139
9 9
FIGURE 4-80.- SIEAM LINE BREAK CASE 8, 205 FA STEAM GENERATOR B PRESSURE e
1200 e
1 1000 2
800
=
n.
600 E
8 m
400 200 Y
__ o _
i i
i i
i 0
100 200 300 400 500 600 g
Time, Sec 1521 140 e
FIGURE 4.81.
STEAM LINE BREAK CASE 8.,205 FA 30 STEAM GENERATOR B 25
/
=
}
20
//
3
~
,i
/
is j
Stu M OEn u 10R A
/
10 l
1
/
\\
!~
5
\\
i
\\
/
1521 141
\\
A i
i i
i 0
0 100 200 300 400 500 600 Time, Sec
FIGURE 4-82.
STEAM LINE BREAK CASE 8, 205 FA, STEAM GENERATOR A 400 -
i 350 NO UPPER SG PLENUN V010S 300
- 25r, HOT LEG
"=
i 5 200 3
150 e
. 100 CAN0Y CANE 50
, /T' 7N 0
y i
0 5
10 15 20 25 30 Time, Sec 1521 142
FIGURE 4-83.
31EAM LINE BREAK CASE 8, 205 FA - STEAN GENERATOR B 600 gyptg PLEMD y
(
\\.
r 0\\
I in
/
\\
s
'\\
(l
\\
1s es i
1 il 1
7
),
senm *'
\\
(
/ \\
(r
\\
(
sI
\\'\\
u
[(^b [;
'l t
- m. $l i
I
. IIj
' \\
v) -
k l K '%.\\J-1
.]
\\
1 rsv \\*3
\\
/j
\\
o_
O 100 200 300 400 Time, Sec
ATTACHMENT c P,tovide a schedule of completion of instattation of the identifled sijstems and components.
1521 144
ATTACHMENT C Page 1 of 2 WPPSS NUCLEAR PROJECT N0. 1 System Design **
Procurement
- Fabrication Start Installation Complete Component Complete %
Complete %
Complete %
WNP-1 Complete %
WNP-1 High Pressure Injection Sys.
100 100 95 Started 10 7/81 Auxiliary Feedwater Sys.
100 100 95 Started 5
7/81 Decay Heat Removal Sys.
100 100 95 Started 5
11/81 Core Flood Tank Sys.
100 100 95 2/80 0
12/81 Reactor Coolant Pressure Control System (ICS) 100 100 100 9/81 0
10/82 Make-Up/ Letdown Sys.
100 100 95 Started 0
7/81 Steam Generator Pressure Control System 100 100 95 Started 10 3/82 Steam Generator 100 100 95 Started 0
12/79 Pressurizer 100 100 100 Started 0
12/79 Quench Tank 100 100 100 Installed 100 Installed Control Room Layout 100 100 0
Started 20 9/81
[3'.deactorCoolantSys. Piping 100 100 100 Started 0
10/81 r4
-- Main Feedwater System 100 100 Started Started Started 3/82 s-Contracts Awarded
- The design completion status is based on complete functional design, system description, and piping design.
Minor electrical work remains to complete " punch list" items.
Instrumentation and control design is approximately 85% complete.
ATTACHMENT C Page 2 of 2 WPPSS NUCLEAR PROJECT NO. 4 System Design **
Procurement
- Fabrication Start Installation Complete Component Complete %
Complete %
Complete %
WNP-4 Complete %
WNP-4 High Pressure Injection Sys.
100 100 95 6/80 0
5/83 Auxiliary Feedwater Sys.
100 100 95 6/80 0
10/83 Decay Heat Removal Sys.
100 100 95 6/80 Started 7/83 Core Flood Tank Sys.
100 100 95 6/80 0
7/83 Reactor Coolant Pressure Control System (ICS) 100 100 100 8/81 0
4/83 Hake-Up/ Letdown Sys.
100 100 95 6/80 0
10/83 Steam Generator Pressure Control System 100 100 95 11/81 0
10/83 Steam Generator 100 100 95 3/81 0
5/81 Pressurizer 100 100 100 5/81 0
6/81 Quench Tank 100 100 100 11/79 0
12/79 Control Room Layout 100 100 0
Started 0
9/83 Reactor Coolant Sys. Piping 100 100 100 8/81 0
4/83 Main Feedwater System 100 100 Released for 11/81 0
10/83 Fabrication
\\
- Contracts Awarded The design completion status is based on complete functional design, system description and piping design.
% ** Minor electrical work remains to complete " punch list" items.
Instrumentation and control design is h
approximately 85% complete.
ATTACHME!Pr d Ide>ttify Bte feasibility of Ita.Ltbtg installation of titese systems ans. componeutts as compa.ted to tite feasibiLLty of completing tJicit installation and titen effecting sigutificant citanges to titem.
Potential Areas Requirina Systems / Components Chances The following systems and components have been identified n attachment f as having a potential for modification:
Systems Components Main Feedwater and Condensate System Instrument Systems Cabinets RCS Pressure Control Auxiliary Feedwater (control portion only)
Integrated Control System Reactor Protection System Engineered Safety Features Actuation System Areas Not Requiring Systems / Components Changes There are no changes anticipated for the following systems / components:
Systems Components High Pressure Injection Steam Generators Core Flood Reactor Coolant Piping Decay Heat Removal Pressurizer (except for the PORV, its block valve and its heater controls)
Present Status At the present time (12/3/79), at WNP-1, the following equipment is installed:
o Reactor coolant pump casings (PSAR Figures 1.2 - 11 & 62) o Reactor drain tank (PSAR Figures 1.2 - 10 & 62) o Auxiliary F/W pumps (PSAR Figures 1.2 - 65 & 66) o DHR pumps (PSAR Figure 1.2 - 35) o DHR heat exchangers (PSAR Figures 1.2 - 33 & 36) o Make-up pumps (PSAR Figures 1.2 - 52 & 53) o Main F/W pumps (PSAR Figures 1.2 - 16 & 20) d-l 1521 i FA
O This equipment is located in all three major structures; the Containment, the General Services Building (GSB) and the Turbine Generator Building (TGB). The status of the structural construction of each of these buildings is as follows:
Containment (PSAR Figures 1.2 - 10 through 15 and 59 through 62) o Polar crane wall concrete and all other interior concrete is complete.
o Exterior (Containment shell) concrete is approximately fifty gg percent complete.
Containment liner is approximately fifty-two percent complete.
o General Services Building (GSB) (PSAR Figures 1.2 - 33 through 57)
All concrete affecting the systems listed in attachment c in the o
GSB has been placed; that is, the individual rooms and spaces containing the systems are complete with walls and ceilings.
One exception is the construction access opening in the ceiling of the Auxiliary Feedwater pump compartment (PSAR Figures 1.2 - 65 and 66) from column lines 4.5 to 5.5 and II from column lines T to U.
This construction opening is expected to renain until mid-1981 and provides access to the 395 foot elevation (Auxiliary Feedwater pump location) from the top of the structure. Some of the pipe, valves, and other construction materials are moved into the GSB via this opening.
This construction opening is not, however, the only construction II 8
access into the GSB.
Turbine Generator Building (TGB) (PSAR Figures 1.2 - 16 through 24) o The building structural steel is complete.
Installation of metal siding is approximately ten percent complete and the roof is in place.
GD Construction Plan The construction plan for WNP-1 is as follows:
Containment: The Containment is a reinforced concrete structure with a (D
3/8" thick steel liner. The plan for completing construction is to assemble the remaining vertical liner rings and the liner dome on the ground in an arca adjacent to Containment. With this plan, the top of Containment will be available for construction access and installation of major equipment (steam generators, pressurizer, polar crane, etc.)
utilizing a currently on-site, large mobile crane.
gp The polar crane will be installed in September 1980, followed by installa-tion of the assembled liner rings and liner dome and placement of the remaining reir forcing steel and concrete for the Containment shell structure. Aftu; September 1980 the equipment hatch and the polar crane, both designed to handle all reactor coolant system components, gg including a steam generator, can be used.
any future required major equipment / system modification.
h d-2 O
General Services Building: The concretc for the GSB is expected to be essentially complete by May 1980. All structural steel and concrete have been completed in and around the systems listed in attachment c, except for the construction opening in the ceiling above the auxiliary feedwater pump area.
Turbine Generator Building: The turbine generator building steel structure is essentially complete and installation of metal siding is in progress.
If necessary, future access to the building can be pro-vided by removal of the siding.
The plan for equipment installation, piping, and other mechanical work is as follows:
Containment: The reactor pressure vessel, che let-down coolers, and the reactor coolant drain tank are installed. Installation of the steam generators and the pressurizer is scheduled this month.
Installation of the nuclear steam supply piping has started but is less than ten percent complete.
Installation of process piping for various systems in the lower levels of Containment is in progress and is approximately ten percent complete.
General Services Euilding: The make-up, auxiliary feedwater, and decay heat removal pumps and the decay heat removal heat exchangers are installed and piping installation is in progress.
Turbine Generator Building: Piping installation and mechanical work for the main feedwater pumps is just starting.
WNP-4 Construction Plan The general discussion presented for WNP-1 is applicable to WNP-4.
The construction plan for WNP-4 is similar to the plan for WNP-1 with WNP-4 activities scheduled approximately eighteen months later than WNP-1.
Potential modifications that may be identified by future studies will probably be applicable to both plants but the construction impact of future changes on WNP-4 will be less than the impact on WNP-1 because of the later WNP-4 construction schedule.
Feasibility of Haltina Construction It is possible co suspend construction on all of the potentially affected systems and components; however, the suspension would be expensive in terms of cost, loss of key personnel, and schedule delay. The primary concern is schedule extension. A delay in the installation of the NSSS equipment and piping will cause a direct delay of project completion because the critical path schedule for the project is piping installa-tion in the GSB and Containment. Although the NSS system installation in Containment is not the current critical path schedule, a slight delay of approximately 30 days will make it critical.
In addition, several key items of construction equipment are available under contact for specific time durations and it is likely that contractors would move this equipment elsewhere if construction were suspended.
d-3 1521 149
O Suspension of civil / structure work will have no effect c.
the access to the components and systems listed in attachment c since the concrete is already placed or is not planned to be placed until late 1980.
Feasibility of Completing Installation & Then Effecting Significant Changes II It is feasible to make significant changes to any or all of the listed systems and components during or ifter the construction phase.
In Con-tainment, the polar crane and th'. equipment hatch are designed to handle steam generators and any sma'.ler components. This design feature will permit removal and replacement if required, of a steam generator after ID construction is complete. Tha concrete on the sides of the steam gen-erator compartments has been } laced and until the top of the Containment liner is installed, no concre.e would have to be removed to replace a stean generator.
In the GSB, pumps, valves, heat exchangers, and other components will GD be as accescible when construction is complete as they are now.
Since the concrete th't affects access to all the components listed has been placed, no advantage to suspension of building construction would be achieved.
The primary advantage to continuation of construction and effecting gp changes later is that the total long range effect on the project will be minimized and the impact on the project schedule will be lessened.
It is true that modification of systems or components will increase costs; however, the failure to proceed with initial system installation will delay other work not directly affected by the Once Through Steam Generator concerns. The cost for delay on WNP-1 Project is estimated gg to be $15,000,000 to S20,000,000 per month, not inclurling lost revenue for lost power production. As mentioned previously, access to the large components in Containment exists in two ways: "over the top" until Sepemmar 1980 and via the polar crane and equipment hatch after September :480. Little benefit can be gained by suspension of con-structio:1.
g Conclusion Continued installation of major reactor coolant system components will not adversely affect future systems modification because:
8 A.
Plant design of the polar crane and the equipment hatch permit replacement of a steam generator if required.
B.
Physical systems installation will be in progress until late 1981, thereby allowing ample time for any required design and modification.
O Remaining civil / structural work in Containment, General Services Building, or the Turbine Generator Building will not restrict systems modification because allowances have been made for future access.
O d-4 I521 150
Economic studies show that intentional construction delays would not be beneficial in this case because:
1) the access provisions previously identified are available to permit future modification and 2) the large costs associated with delays. The cost for delay on WNP-1 is estimated to be $15,000,000 to $20,000,000 per month, not including lost revenue for lost power production.
The reccmmended modifications discussed in attachment f involve almost entirely changes to the WNP-1/4 electrical and instrumentation and control systems. Such changes can be practically accormodated during construction even if it becomes necessary to backfit these modifications.
The only primary _ystem component or piping changes involve the pres-surizer power operated relief valve and its block valve. There is the potential for some very minor piping changes to the main feedwater and condensate systems.
It is concluded chat it is much more feasible and prudent to continue construction and integrate the required changes into the normal design and contruction process. WPPSS will continue construction of both WMP-1 and WNP-4 until investigative results require changes and modifications that exceed present cost tradeoff's, or NRC directs otherwise.
d-5 i52i i51
ATTACHMENT e Comment on the OTSG acn44tLvity to feedstet tiansients.
Introduction.
The Babcock & Wilcox Nuclear Steam Supply System employs Once Through Steam Generators (OTSG's) for heat transfer from the primary to the secondary system. The nuclear OTSG is a vertical straight shell and tube boiler in which the primary coolant (heat source) is on the tube side and the secondary coolant is on the shell side. Normal main feed-water enters the generator shell near the bottom of the tube bundle and flows upward. As it gathers heat, steam is generated and superheated before exiting to the steam piping system. The overall primary to seconder heat transfer is controlled by the rate of feedwater intro-ductiu ' o the generator which in turn controls the area of the total tube bu..dle length which is exposed to liquid secondary coolant for a given input of primary power. Increasing feedwater flow increases the heat transfer and decreasing feedwater ficw decreases heat transfer.
The design of the once Through Steam Generator has yielded supericr performance both in safety and efficiency in pressurized water reactors.
The once through design, with its superheated steam, exhibits a higher thermal efficiency than a recirculating steam generator, resulting in less waste heat rejected to the environment, better utilization of the uranium fuel, and a lower cost for electric pcwer generation. The once Through Steam Generator has exhibited an exceptional tube integrity record over its operating experience; this not cnly maximizes steam generator availability but also minimizes the risk of radioactive release via a tube rupture. One inherent feature of this design is the responsiveness to feedwater control mentioned above. This responsive-ness makes possible an accuracy of control which has both operational and safety udvantages. Safety analysis of limiting feedwater and secon-dary system pressure disturbances has demonstrated the ability to main-tain safe core cooling without radioactive release under the applicable licensing assumptions. However, the frequency of feedwater transients leading to disturbances of pressure and/or pressurizer level in the primary system of B&W plants has been higher than desired. This has been somewhat exacerbated by changes to plant operation which have occurred since the TMI-2 accident.
B&W and the Supply System support the concept of defense-in-depth; and existing plant features accomplish this defense-in-depth as indicated on figure e-1, using an overcooling event as an example. We have com-pleted an evaluation of possible modifications to enhance the existing defense in depth. The results of this evaluation are presented in attachment f.
Existing plant features, plus these changes, are sufficient to resolve many of the concerns expressed in enclosure 1 of the NRC October 25, 1979 10 CFR 50.54 letter.
Included below is a point-by-point discussion of enclosure 1.
B&W and the Supply System have concluded that it is neither necessary nor desirable to modify the fundamental operating characteristics of the Once Through Steam Generator in view of its excellent performance record, e-1 1521 152
O Discussion of enclosure 1 Paracraph addressed Discussion (1) Page 2, para. 1 In all B&W plants under construction, including gg WNP-l/4, the auxiliary feedwater system is a safety grade system. Improved reliability and control of AFW in these plants will minbnize system fluctuations following the initiation of auxiliary feedwater.
(2) Page 2c para. 2 Anticipatory reactor trip on loss of main feed-water has, in fact, yielded very smooth system response. This has been confirmed by recent field data. Use of snticipatory trips should be eliminated, however, for those disturbances (such as turbine trip) which can be handled gg by plant control system action without cnal-lenging the plant safety systems. This will reduce the number of plant trips. See iten 4, belew.
(3) Page 2, para. 3 The addition of an automatic reactor coolant pump trip, based upon the coincidence of signals indicating both low coolant system pressure and significant voids in the primary system, would eliminate the necessity for the operator to manually trip the reactor coolant pumps and raise OTSG water level. With the II addition of this automatic function, reactor coolant pump trip should occur only for actual small breaks in the primary system; it should not occur for overcooling events initiated by feedwater transients. Such an automatic coin-cidence system is recommended for WNP-1/4.
9 (4) Page 2, para. 4 The raising of the PORV setpoint and lowering of the high pressure reactor trip appear to have increased the number of reactor trips on the B&W operating plants. For WNP-l/4, modifications are recommended which will restore the controlled GD relief capability of the PORV while maintaining a high level of protection against PORV malfunction.
A study will also be conducted to determine ways to reduce the vulne ability of the NNI/ICS systems to a power supply failure to these systems. A resetting of the reactor protection gp system high pressure setpoint and the PORV setpoint will restore the capability of the B&W nuclear steam system to sustain a wide range of operational transients without a high pressure reactor trip.
O e-2 4lr2!
jg O
c w,
(5) Page 3, para. 1 The B&W Once Through Steam Generator and Nuclear Steam System are designed to be responsive to utility customer's needs ta avoid reactor trip during minor secondary system transients.
s
=_
This responsiveness is an inherent feature of the design, and operators who have operated both B&W plants and plants employing recircu-lating steam generators nave commented that th3 B&W plants are more stable and easier to control. Some B&W cperating plants do place reliance upon the operator to limit feedwater excursions which may resalt from control system failure. However, the current generation of backlog plants under construction, including WNP-1/4, already employ a number of automatic i
measures to reduce this reliance upon the 3
operator. Several additional modifications are being engineered which will further reduce the re tuirements for the operator to act in respoi.se to a control system failure. The proposed changes will improve our defense-in-depth against primary system parameter excur-sions resulting from moderate secondary system upsets.
(6) Page 4, para. 1 Overcooling transients in all PWR systems pro-ceed initially like a small break LOCA, and this is not a unique problem of the Once Through Steam Generator. For example, on a recent reactor trip at the North Anna Power Station, a stuck-open turbine bypass valve with approxi-mately 5% capacity caused an excessive over-cooling which resulted in a prompt loss of reactor system pressure to the setpoint of the automatic safeguards injection cystem, and contraction of primary coolsnt sufficient to take pressurizer level below the range of indication. B&W and the Supply System believe that proper design action requires a reduction in the frequency of such events to the greatest 3
degree possible, combined with training to help plant operators deal confidently and safely when these abnormal eveats occur. Again, our approach to a defense-in-depth concept will improve the tolerance of the feedwater control system for this type of transient and contri-(o
=;
bute to minimizing events of this nature.
1521 154
=
e-3 3
c 68
n-i--i
O (7) Page 4, para. 2 The loss of pressurizer level indication fol-lowing a reactor trip is an operntional concern and should be minimized for expected abnormal occurrences. However, it shculd 30 noted that the loss of indicated pressurizer le',el on B&W q) operating plants is not synonymous with a loss of liquid in the pressurizer. Certain B&W plants, such as Davis Besse Unit 1, have pres-surizer level indicators which do not cover the full span of the pressurizer volume. In the case of Davis Besse, more than 40 inches of gg pressurizer capacity remains be2cw the zero point of the level indication system. Thus, at these plants, a momentary loss of indicated level should not be confused with an emptying of the pressurizer and potential for loss of natural circulation. For B&W plants currently gg under construction, including WNP-1/4, the indicated pressurizer level range more closely relates to the full fluid volume of the pres-surizer. The operational nuisance caused by a loss of indicated prersurizer level will be minimized for these pl.ats.
With this expanded g
indication range, pressw;izer level is expected to remain on s: ale for feedwater upset tran-sients such as those that have occurred at Davis Besse.
(8) Page 5, para. 8 Operation of the pressurizer heaters when not II covered by liquid should be eliminated as a potential occurrence. B&W operating plants include a control grade circuit to remove power from the pressurizer heaters when liquid level is low, and in no instance on an operating plant have the pressurizer heaters been energized while uncovered. For WNP-1/4 it is recommended II that this circuit be upgraded to incorporate a safety grade feature to assure that the pres-surizer heaters will not be energized when liquid level is low.
(9) Page 6, pa13 1 For B&W plants under construction, automatic GD (item 4) equipment is being evaluated which would elimi-nate the reactor coolant pump trip associated with low reactor coolant pressure only. In addition, main feedwater overfeed limiting equipment, independent of che integrated control system, will be investigated as a means to ter-Ep minate main feedwater flow before excessive overcooling occurs.
e-4 1521 155
(10) Page 6, para. 2 It is recommended that the Aux 1]iary Feedwater (item 5)
System and control circuitry be upgraded for WNP-1/4 to minimize the excessive addition of cold water which could lead to emptying of the pressurizer.
(11) Page 6, para. 3 B&W calculations do not predict an interruption (item 6) of core cooling or heat transfer to the once Through Steam Generator as a result of the events sequence outlined. Delivery of cold water by the high pressure injection system will refill the reactor coolant system and quench any voids to provide additional assur-ance of adequate core cooling.
(12) Page 6, para. 4 Criteria for restart of a reactor coolant pump (item 7) are already provided in the current Small Break Operating Guidelines to permit forced flow to be reestablished promptly following repressuriza-tion of the reactor coolant systam.
(13) Page 6, para. 5 The B&W integrated control system is designed to provide smooth and s*able operation of the complete power plant during power operation.
One of its functions is to maintain the reactor plant on line for minor secondary system dis-turbances and eliminate unnecessary challenges to the reactor trip system. Following reactor trip, the ICS has a function in mainta,ining plant conditions stable and within design limits.
Measures are planned in response to the recently completed ICS failure modes and effects analysis to improve the reliability of all control func-tions related to the ICS by improving its input signals and control of the turbine bypass system. Additional control functions independent of the ICS are recommended as discussed in attachment f to limit the effects of failures of the prhnary control system. The auxiliary feedwater control is already performed by a safety-grade system independent of the ICS on B&W plants under construction, including V'aP-1/4.
As noted above, a system separate from the ICS is being investigated to limit main feedwater introduction which might occur as a result of primary control system failure. The combina-tion of these improvements should provide sub-stantial defense-in-depth againse sequences of the sort discussed in this section.
e-5 1521 156
(14) Page 7, last Limiting safety analysis has shown that adequate para.
core cooling will be maintained and radioactive release will be avoided even for the most severe secondary system accidents within the plant's ID licensing basis. The generation of B&W plants currently holding construction permits already incorporata a number of design features which
_ddress the issues raised by improving system reliability and reducing the consequences of secondary system upsets. In addition to this, a carefully censidered group of modifications Gb has bean proposed to reduce primary system response to feedwater disturbances and to reduce the magnitude and frequency of secondary system feetwater upsets. These mudifications will improva plant performance and enhance safety through the defense-in-depth concept by (p
terminating or mitigating transients early in their course before they result in seriously off-norral conditions.
e-6 1521 157
Dofonse-in-depth approach to maintain adequate core cooling PRESENT DESIGN IEAftRES IuPDCyr0 CEFENSES SECAIENCE OF EVENTS 13f A0V1 Ovisait Ptahi Rit talitilf th' in!!CRatto CONTADt stills CChfAOL flaftait 10 ulkisilt
$149t1 plaCicA P0ftR OPERaf10N PR08a9l tiff 0F PlaCICR 1 RIP R:11081 ma8Cl4 tilattu Aceval Plant
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7 C0%3tlichi af Posts the statica PR0fitt t0% St',f te 5tf F01%f 5 Riac 21 RIP 0'JE TO A8h0krat staC10e PR0ttCg On sisite ir tWPPLVI P9t!5tRittA tivtt RESPONil PRESSustilt tittt ON SCatt h0Rua! Ptahl C0h1R0t f 0ttosthG ICittelhG klaCICR TRIPS 001 10 sh0 RCS PRESSURE GRiaf(R TMa4 staC102 TRIP 1600 PSI A8hChuat CCCURAthCt1 1r PatssuRiffs tivtt Ort statt tot 145tatt PL ANI N00ificaI10%s to Bahimilt PkG8aBillit CF Sittpt 0.teCCCt ehG tanksiinis EstRGINCr saltGuaRos aC10al 04 RCS PRE $$URt tt$$ THa4 1600 PSI 5 tills l'
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- This improved defense is proposed for WilP-1/4 Figure e-1 m
co'
ATTACHMENT f Provide recommendatwns on hardaure and ptocedutat changes related to the need fo.1 and methods for damping primary system sensitivity to pcr.twtbations in the OTSG.
Include detuts on any design adequacy studies you have done o.1 have in ptogress.
a.
Design and Analytical 3tudies B&W has completed or has in progress several design and analytical efforts which deal with the primary system responses to perturba-tions which occur in the OTSG. These efforts are listed below, together with identification of the vehicle by which details of these efforts have been provided to the NRC.
Analysis of the most severe overcooling events provided as attachment a to this letter in the report " Overcooling Event Consequence Review."
Review of the typical sequence of events for overcooling events to identify design features which provide defense-in-depth described in section e of this attachment.
Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant; May 7, 1977, previously submitted to the NRC.
Integrated Control System Reliability Analysis, BAW-1564, August 1979, previously submitted to the NRC.
The Abnormal Transient Operating Guidelines (ATOG) develop-ment program, presented to the NRC at meetings in Bethesda on August 8, 1979 and Sep' ember 13, 1979, and at Lynchburg on c
October 15, 1979.
The review of recent feedwoter transients, discussed with the NRC at the meeting in Bethesda on August 23, 1979.
The results of these various efforts suggest certain modifications which could ba made to further improve the performance of the B&W NSS.
These changes are listed in section b.
In addition to these efforts by B&W, several balance of plant design adequacy studies are in progress or will be started. These studies include:
f-l 1521 159
9 Secondary System Reliability Study. During normal operations, secondary system disturoances can cause perturbations on the primary system parameters because of the close coupling of the primary and secondarl systems through the OTSG. The pur-pose of this study is to identify discrete changes to secon-dary system controls and equipment to improve the stability ID of the secondary systems and to reduce the challenges to the safety systems.
Ccntrol Air Supply System. Control valves performing safety functions under emergency conditions already receive air from the Nuclear Instrument Air Supply System. Air for the remain-GD ing control valves is supplied from the non-Nuclear Instrument Air Supply System. The purpose of this study will be to eval-uate the need for more reliability of the secondary systems and also the auxiliary systems. Examples of valves to be considered are main feedwater control valves and makeup control valves.
dp Minimum Final Feedwater Response Study. A study will be conducted to evaluate the control system for maintaining minimum final feedwater temperature. The response of the feedwater temperature to toe transition from extraction steam to main steam heating of the sixth and seventh stage heaters gg on decreasing load will be icvestigated.
Auxiliary Feedwater Turbine Reliability Study. A study is being conducted to review the plant operating experience and its applicability with respect to:
- 1) water accumulation and governor problems resulting in failure of the turbine and 2) gg air binding resulting in failure of the pump.
NNI/ICS Power Supply Study. A study will be conducted to reduce the vulnerability of the NNI/ICS systems to a supply power failure to these systems.
8 Heater Drain Pump Reliability Study. A study will be conducted to evaluate increasing the NPSH to the heater drain pumps by y y continuous coldwater injection.
b.
Recommended Changes O
Much of the concern expressed about the " sensitivity" of the B&W OTSG PWR design is based on the operational experiences with the currently operating 177 FA plants and particularly that experience accumulated since the accident at TMI-2.
It is important to recog-nize that the normal evolution of design that has occurred on the generation of B&W plants now under construction, such as bce-l/4, gg as a result of new regulatory requirements, improvements in the state-of-the-art in hardware, and the feedback of operating experi-ence has resulted in the incorporation of several new features.
These features serve to improve the reliability of the systems and f-2 l$2T 160
equipment and thereby reduce the probability of the challenges to the safety systems, to improve the response of the MSSS to those events that do occur, and to provide better capability to mitigate the events which occur. These changes include:
Addition of a two channel, Class lE Essential Control and Instrumentation (ECI) system to control auxiliary feedwater and provide post accident monitoring instrumentation for the operator.
Initiation of auxiliary feedwater by the IEEE-279 Engineered Safety Features Actuation System (ESFAS).
Addition of Feed Only Good Generator (FOGG) logic to the ESFAS to help insure that auxiliary feedwater is delivered to the intact steam generator following secondary system breaks.
Adoption of newer control system hardware which uses dual, auctioneered power supplies for the logic modules rather than individual power supplies for each logic module as in the earlier designs.
Moving the pressuriser level sensing taps to the top and bottom heads of the pressurizer to expand the range of level indication.
Raising the level of the OTSG *aith respect to the level of the reactor.
As stated in the introduction to this attachment, several design studies of various aspects of the B&W NSS have been conducted since the TMI-2 accident. These studies have shown that some changes are recommended to (1) retain the basic design operating charac-teristics of the OTSG PWR, (2) improve the reliability of the systems whose failure can lead to overcooling transients (thereby enhancing plant availability and reducing the frequency of chal-lenges to the safety systems), (3) further improve the response of the ::SS to the transients which do occur, and (4) improve the capability to mitigate these transients. These recommended changes include:
1.
Changes to retain original OTSG PWR design operating characteristics Qualify the pressurizer PORV and provide reliable power supplies and controls for the PORV.
o Provide a Class lE PORV isolation valve actuated by the ESFAS on low reactor coolant system pressure.
2.
Chances to improve the reliability of systems Increase demineralized water makeup capacity to the condenser hotwell during a runback af ter a turbine trip.
1521 161 f-3
O Prevent a single failure in the ICS from opening more than 15% steam dump capacity.
Improve the existing bypass capability around the con-densate polishers by providing a fast responding bypass Ib valve to open automatically on high SP across the demin-eralizers and/or low main feedwater pump suction pressure.
Improve the control response of the ICS following the failures of important input sensors.
S 3.
Changes to improve the response of the NSS Add a control function to the ICS to provide for positive and rapid reduction of main feedwater flow following reactor trip.
9 Add a Class lE loss of all feedwater. reactor trip.
Add a main feedwater overfill protection system.
Reduce the potential for auxiliary feedwater overfill due to single failures in the auxiliary feedwater controls.
gg Improve the algorithm used for auxiliary feedwater flow control.
4.
Chances to improve the capability to mitigate transients Provide Class lE low level cutoff of pressurizer heaters.
Improve the control of auxiliary feedwater flow following ESFAS actuations.
Automatically trip the reactor coolant pumps on low gg reactor system pressure concurrent with the detection of voids in the reactor coolant.
Provide improved FOGG logic to provide additional assur-ance that auxiliary feedwater is directed to the proper steam generator throughout the course of transients.
O The above discussed changes have been recommended by B&W and United Engineers & Constructors for implementation on WNP-1/4.
The Supply System concurs with these recommendations and will begin immediately the necessary activities to effect these changes. While the feasi-bility and overall safety benefit of some of the recommendations remain to be demonstrated, we believe the above listing sufficiently II characterizes the types of changes likely to be implemented on WNP-1/4 to show that the changes required to deal with the sensitivity concern can be accommodated during the normal construction progress.
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c.
Event Tree Review One of the studies undertaken by B&W as a result of the NRC 10 CFR 50.54 requests of October 25, 1979 was to review the typical sequence of events for overcooling transients, identify the existing plant design features which provide defense-in-depth against the occurrence of inadequate core cooling, and see if additional changes are suggested to improve these defenses. Figure e-1 illustrates the results of this review.
This review reconfirms that an important feature of the B&W OTSG PWR design is the use of the PORV to reduce the magnitude of reactor coolant pressure increases during transients and thereby reduce the likelihood of tripping the teactor on high pressure.
The improvements to the PORV, to its controls, and the provision of reliable, automatic closure of the PORV isolation valve are changes to help assure this feature is maintained with a setpoint below the reactor trip setpoint rather than changing the setpoint as was done on the operating B&W plants. The review also suggested several improvements to the reliability of certain systems and equipment which would both improve overall plant availability and reduce the occurrence of transients leading to reactor trips.
These changes are identified in section b of this attachment.
The review showed that while the existing NSS post trip control functions appear to be adequate, the response of pressurizer level following the transient could be further improved and the reliance on operator action could be educed by making some chan3as. These changes include adding a loss c' all feedwater reactor trip r.id improving the manner in which auxiliary feedwater is controlled.
The review also showed that changes could be made to further minimize the probability of severe overcooling transients. These changes include a more positive and rapid reduction of main feedwater flow following a reactor trip, new features to prevent failures in the ICS frem opening more than 15% steam dump capacity, and the addi-tion of a new system to prevent main feedwater from overfilling the steam generators.
In addition, WNP-1/4 will consider adopting the Abnormal Transient Operating Guideline program currently being pursued by the EsW operating plants in response to NUREG-0578. This would provide the operators with better procedural guidance and thereby reduce the probability that they might take actions which could further complicate overcooling events, f-5 k
d.
Post Trip Responsiveness Analysis B&W has completed an investigation of the post trip responsiveness of WNP-1/4 for the major factors which influence the magnitude of II pressurizer level and reactor coolant system pressure. The results of this investigation are under review by the Supply System. Our initial review of this investigation confirms that implementation of some of the recommended changes discussed in section b of this attachment are sufficient to assure that loss of pressurizer level and HPI initiation will not result for a simple reactor trip.
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