ML20115H655
| ML20115H655 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 07/18/1996 |
| From: | Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML19311C115 | List: |
| References | |
| AW-96-991, NUDOCS 9607230265 | |
| Download: ML20115H655 (19) | |
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Westinghouse Energy Systems Ba 355 Electric Corporation Pmsb>gh Pennsylvanta 15230-0355 AW-96-991 July 18,1996 pocument Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ATTENTION:
T.R. QUAY APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE y
SUBJECT:
WESTINGHOUSE RESPONSES TO NRC REQUESTS FOR ADDITIONAL INFORMATION ON THE AP600
Dear Mr. Quay:
The application for withholding is submitted by Westinghouse Electric Corporation (" Westinghouse")
pursuant to the provisions of paragraph (b)(1) of Section 2.790 of the Commission's regulations. It contains commercial strategic information proprietary to Westinghouse and customarily held in confidence.
The proprietary material for which withholding is being requested is identified in the proprietary version of the subject report. In conformance with 10CFR Section 2.790, Affidavit AW-96-991 accompanies this application for withholding setting forth the basis on which the identified proprietary information may be withheld from public disclosure.
Accordingly, it is respectfully requested that the subject information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10CFR Section 2.790 of the Commission's regulations.
Correspondence with respect to this application for withholding or the accompanying affidavit should reference AW-96-991 and should be addressed to the undersigned.
Very truly yours, s
I x-Brian A. McInty Manager Advanced Plant Si ty and Licensing
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Kevin Bohrer NRC 12H5
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AW-%-991 i
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J COMMONWEALTH OF PENNSYLVANIA:
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COUNTY OF ALLEGHENY:
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L Before me, the undersigned authority,- personally appeared William R. Rice, who, being by.
I me duly sworn acco' ding to law, deposes and says that he is authorized to execute this Affidavit on i
r behalf of Westinghouse Electric Corporation (" Westinghouse") and that the averments of fact set forth' in this Affidavit are true and correct to the best of his knowledge, information, and belief-1 WA t
William R. Rice, Interim Manager Regulatory and Licensing Initiatives Sworn to and subscribed
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Janct A. Sclwab. No:ary Pubnc i
Monroevil:e Boio, AFogheny Cour'y My Commscion Expires May 22,2000 Member, Pennsylvania Association of Notanes Notary Public l
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AW-96-991 l
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L (1)
I am Interim Manager, Regulatory and Licensing Initiatives, in the Nuclear Services Division, of the Westinghouse Electric Corporation and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rulemaking proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Energy Systems'
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Business Unit.
(2)
I am making this Affidavit in conformance with the provisions of 10CFR Section 2.790 of the
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l Commission's regulations and in conjunction with the Westinghouse application for l
withholding accompanying this Affidavit.
I (3)
I have personal knowledge of the criteria and procedures utilized by the Westinghouse Energy l
Systems Business Unit in designating information as a trade secret, privileged or as l
confidential commercial or financial information.
f (4)
Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Commission's j
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regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.
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The information sought to be withheld from public disclosure is owned and has been i
held in confidence by Westinghouse.
l (ii)
The information is of a type customarily held in confidence by Westinghouse and not j ~
customarily disclosed to the public. Westinghouse has a rational basis for determmmg L
the types of infonnation customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types c"information in confidence. The application of that sysicm and the substance of that system i
constitutes Westinghouse policy and provides the rational basis required.
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Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:
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AW-96-991 4
(a)
The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.
(b)
It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.
(c)
Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.
(d)
It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.
(e)
It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.
(f)
It contains patentable ideas, for which patent protection may be desirable.
There are sound policy reasons behind the Westinghouse system which include the following:
(a)
The use of such information by Westinghouse gives Westinghouse a l
competitive advantage over its competitors. It is, therefore, withheld 'from disclosure to protect the Westinghouse competitive position.
(b)
It is information which is marketable in many ways. The extent to which such
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information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.
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(c)
.Use by our competitor would put Westinghouse at a competitive disadvantage -
- I by reducing his expenditure of resources at our expense.
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(d)
Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive j
advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving l
Westinghouse of a competitive advantage.
l (e)
Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the 1
competition of those countries.
i (f)
The Westinghouse capacity to invest corporate assets in research and j
i development depends upon the success in obtaining and maintaining a competitive advantage.
(iii)
The information is being transmitted to the Commission in confidence and, under the j
provisions of 10CFR Section 2.790, it is to be received in confidence by the
.l Commission.
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(iv)
The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.
(v)
Enclosed is Letter NSD-NRC-%-4773, July 18,1996 being transmitted by Westinghouse Electric Corporation (E) letter and Application for Withholding Proprietary Information from Public Disclosure, Brian A.- McIntyre (W), to Mr. T. R. Quay, Office of NRR. The proprietary information as submitted fonise by Westinghouse Electric Corporation is in response to questions concerning the AP600 plant and the associated design certification application and is expected to be applicable in other licensee submittals in response to certain NRC requirements for justification of licensing advanced nuclear power plant designs.
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j AW-96-991 l
This information is part of that which will enable Westinghouse to:
.(a)
Demonstrate the design and safety of the AP600 Passive Safety Systems.
i (b)
Establish applicable verification testing methods.
j (c)
' Design Advanced Nuclear Power Plants that meet NRC requirements.
-l (d)
Establish technical and licensing approaches for the AP600 that will uhimately result in a certified design.
i (e)
Assist customers in obtaining NRC approval for future plants.
1 Further this information has substantial commercial value as follows:
(a)
Westinghouse plans to sell the use of similar information to its customers for j
purposes of meeting NRC requirements for advanced plant licenses.
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(b)
Westinghouse can sell support and defense of the technology to its customers j
t in the licensing process.
Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance.the ability of competitors to provide similar advanced nuclear power designs and licensing defense
.j services for commercial power reactors withou', commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC 4
requirements for licensing documentation without purchasing the right to use the information.
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The development of the technology described in part by the information is the result of j
applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.
l In order for competitors of Westinghouse to duplicate this information, similar.
technical programs would have to be performed and a significant manpower effort, l
having the requisite talent and experience, would have to be expended for developing analytical methods and receiving NRC approval for those methods.
I Further the deponent sayeth not.
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 NSD-NRC-96-4773 i
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NRC REQUEST FOR ADDITIONALINFORMATION Question 410.280 SSAR Section 9.3.5.2.2 states that the drain tanks are vented to the atmosphere. Since these tanks cou contaminated, what would prevent the release of airborne radioactivity to the af.nosphere?
Response
The " Sumps and Drain Tanks" portion of SSAR subsection 9.3.5.2.2, Revision 7. states that no tanks are vented to atmosphere of the room in which they reside. The reactor coolant drain tank is v gaseous radwaste system. The waste holdup tanks can receive water from floor drains. Th which are serviced by the radiologically controlled area ventilation system. Each sump is vented into where necessary to control airborne radioactivity the sump vents are routed to the ventilation syste for the room.
SSAR Revision: NONE 1
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NRC REQUEST FOR ADQlTIONAL INFORMATION
!!M Question 410.281 Why was the sentence "Each sump is provided with two pumps" removed in Revision 4 to SSAR Section 9.3.57
Response
The equipment and floor drainage systems are nonsafety-related and serve no safety-related function except for the backflow preventers in drain lines from containment cavities to the containment sump. Therefore, there is no i
requirement for each sump to be provided with two pumps. SSAR section 9.3.5.2.3, Revision 7. discusses operational modes for both those sumps with two pumps and those with only one.
SSAR Revision: NONE l
A 410.281-1 TW
4 NRC REQUEST FOR ADDITIONAL INFORMATION I
Question 440.120 l
The staff is concerned with boron dilution events for PWR designs. A slow, inadvertent dilution due to a ma function of the chemical and volume control system (CVCS) or faulty operator actions is a design basis event that must be shown to satisfy stringent acceptance criteria. Recently, the question of whether additional failures or scenarios other than the CVCS malfunction events might lead to inadvertent criticality and fuel damage has received considerable attention in Europe and the United States..t ur example, a preliminary study by the Finnish Center for j
Radiation and Nuclear Safety indicates that an inherent mechanism for boron dilution exists in the cold leg loop seals
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or transients and accidents, e.g., a small break LOCA, involving heat removtl by reflux-or boiler-condensation natural circulation. Under certain conditions and scenarios, such as during the restart of RC pumps, substantial boron j
dilution could result in the core, leading to a reactivity induced accident.
a.
Although the AP600 design does not have a loop seal in the cold leg, has Westinghouse evaluated the possibility of accumulating deborated (a highly dilute slug) water in the reactor coolant loop, especially in the steam generator cold leg channel head, as a result of reflux / boiler condensation natural circulation in j
an accident? Address this concern.
b.
For those transients or accidents that may result in the accumulation of a deborated water slug in the RCS loop, provide an analysis to demonstrate that recriticality will not occur as a result of the deborated water slug entering the core, either through natural circulation or by restarting the pump (s). The analysis should include an evaluation of the degree of mixing between the deborated water slug and the existing borated concentration, the reactivity insertion, and the total reactivity. Describe the methodologies use in the analysis.
If recriticality occurs, provide an analysis of the consequence, such as whether the calculated peak fuel c.
enthalpy (due to insertion of reactivity) has exceeded the limiting value of 280 calories per gram.
d.
What emergency operating procedures are there to prevent the restart of RC pump that could result in criticality during transients and accident events? What are other protective measures?
I
Response
Several different scenarios for all PWR designs have been postulated that could cause the accumulation of unborated water in the RCS loops. The postulated scenarios addressed herein, which include those that are unique to the AP600 design, are:
The "Finnish Center" scenario, which is addressed by item (a.) below, with supplementary analytical a
discussions given in item (b.).
The introduction of relatively unborated water is possible as a result of reverse break flow following a steam s
generator tube rupture (SGTR). This scenario is discussed in items (b.) and (d.) below.
440,120-1
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NRC REQUEST FOR ADDITIONAL INFORMATION It can be postulated that the actuation of the AP600 CMTs could potentially yield pockets of coolant that s
may not receive the higher concentration borated water. Thus, subsequent loop recovery under cold conditions may be a concern if the critical boron concentration for the temperature of interest is higher than the boron concentrations present in the stagnant regions of the loop (s). This postulated situation is discussed in items (b.) and (d.).
During a dilution to criticality, if a loss of power occurs, the subsequent emergency diesel generators startup e
and loading would allow the charging /make-up pumps to continue the dilution without the RCPs in operation, thus providing the means to accumulate unborated water in the RCS loop (s). This situation, which has been referred to as the " French" scenario in other studies / reports, is discussed in item (d.).
Various RCS maintenance procedures have the potential for low, or zero boron concentration water to
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accumulate in the RCS. This situation is also discussed in item (d.).
These five scenarios are addressed with respect to the AP600 design below. This response has been structured as items (a.) through (d.), which correspond to the four parts of this RAI question.
(a.) "Finnish Center" Scenario The AP600 is not subject to the "Finnish Center" scenario during small break LOCA events. The AP600 design possesses no loop seals entering the reactor coolant pumps (RCPs). Only a small amount oflow boron content water could collect in the bottom of the four RCP casings (approximately 21 ft' per RCP casing; this equates to approximately 1.8% of the AP600 reactor vessel inlet plenum volume) and potentially be present in small break LOCA scenarios for sudden transport into the reactor vessel upon RCP restart.
The "Finnish Center" scenario as such is not relevant to the AP600 because the steam generators are not cooling the RCS to potentially generate boron-free condensate for any significant length of time during a LOCA event.
During small break LOCAs in conventional PWRs, decay heat is removed through the steam generator safety valves. In AP600, the PRHR rapidly becomes the dominant RCS heat sink following the generation of an "S" signal during postulated small break LOCA events; the steam generators become heat sources rather than heat sinks. A potential means for generating a volume of unborated water during a small break LOCA is via operation of the PRHR. Steam condensed in the PRHR is delivered to the loop 1 steam generator outlet plenum during small break LOCA events. However, with no RCP loop seals the AP600 layout is such that the PRHR effluent will drain continuously from the steam generator channel head into the loop I cold legs and flow into the reactor vessel. A deborated water slug cannot accumulate in the RCS*Imop cold legs. Within the reactor vessel the cold leg fluid entry point is above the direct vessel injection line elevation, which receives passive safety injection water from the core makeup tanks and/or accumulators with a high boron concentration which provides a significant reactivity margin to recriticality. Since the downcomer annulus remains full to approximately the core top elevation or higher throughout the AP600 small break LOCA events, the dilute l
PRHR water must pass through and mix with more than twelve feet of borated water present in the downcomer l
prior to reaching the lower plenum and then the core. De relatively low flow rate of fluid from the downcomer l
into the core during the post-RCP trip phase of AP600 small break LOCA events enable mixing to occur in the 440.120-2 W Westinghouse l
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r NRC REQUEST FOR ADDITIONAL INFORMATION core and lower plenum. No unmixed " slugs" of highly dilute liquid from the PRHR are present in the downcomer to enter the core during LOCA design basis scenarios.
In the later stages of a small break AP600 LOCA event, once ADS stages 1-3 are active, the PRHR receives very little flow (Reference 440.120-1). Therefore, the downcomer boron concentration will not diminish sigmficantly due to continued PRHR operation. Moreover, once ADS fourth stage valves are actuated, the PRHR receives even less flow and produces even less condensate, so no substantial boron dilution occurs within the core.
(b.) Transients or Accidents Addressed by Analysis The safety-related method for decay heat removal for the AP600 c'onsists of heat transfer to the IRWST by the PRHR, and borated make-up water addition to the RCS froni the CMTs. Operation of the CMTs require that the RCPs are tripped. As the residual heat from the core is removed by the PRHR and CMTs, boric acid is added to the RCS by CMT injection flow. He RCS flow associated with the operation of the PRHR and CMT sy' tems is caused by the thermal driving head established by the convective heat transfer. Analyses have been s
performed (Reference 440.120-2) to investigate the flow behavior throughout the RCS while the PRHR and CMT systems are removing core decay heat, in order to quantify the resulting boron distributions. For this study a loss of normal feedwater transient was chosen.
The Reference 440.120-2 analysis effort utilized the TRAC-PFl/ MOD 2 c-ie to pstform transients that are very similar to the design basis loss of normal feedwater transient (Reference 440.120-3, Section 15.2.6). Conditions corresponding to beginning of life, equilibrium cycle, no xenon were assumed, as this would be the most limiting plant conditions in the event core recriticality were predicted. Benchmarking between the TRAC-PF1 code with the SSAR data, which is based upon output from the Westinghouse LOFIRAN-AP code, indicated good agreement. An acceptable comparison of the neutronic model was obtained with Westinghouse reference core j
data.
The results of the loss of normal feedwater transients indicate that all regions of the RCS become sufficiently borated following RCP tiip and CMT actuation as a result of RCS flow remaining high enough in all regions of the AP600 primary side system for a sufficient duration. The affects of reduced decay heat were also included in the analysis. The low decay heat analysis arbitrarily assumed 1% of the ANS 1979 decay heat curve. Reduced heat generction in the core results in the passive cooling systems to lose their thermal driving head earlier in the transient, thereby providing a shorter duration for the CMTs to inject the higher concentration boron into the RCS. He results demonstrate that boron concentrations throughout the RCS were much greater than the critical boron concentration required for cold (200*F) N-1 rods inserted (most reactive RCCA assumed to be stuck out of the core), no Xenon conditions. Therefore, it can be concluded that subsequent RCS loop recovery, following CMT actuation and RCS cooldown to equilibrium temperatures, will not pose a recriticality potential.
Additional analyses were performed as part of the Reference 440.120-2 study to quantify the volume of unborated water that could collect in the RCP casings and steam generator channel head without resulting in localized core inlet boron concentrations to decrease to the critical boron concentration following the restart of the RCPs. De affects of nominal and reduced decay heat situations were also consider ' ne initial conditions 440.120-3
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NRC REQUEST FOR ADDITIONAL INFORMATION for these investigations were obtained from the pseudo-equilibrium conditions (i.e., transient times >
4000 seconds) for the loss of normal feedwater transients discussed previously. The findings of this unborated water investigation can be directly applied to the SGTR reverse-break flow scenario and also supplement the previously discussed "Finnish Center" scenario, as discussed below.
A high order solute tracker, which is described extensively in Reference 440.120-4 (and is also included as Appendix D of Reference 440.120-2), was employed to significantly reduce numerical diffusion. This high order solute tracking method employed for the unborated slug investigation has been benchraarked against experimental mixing data from a 1/5 scale model of a three loop Westinghouse PWR, also discussed further in Reference 440.120-2. The results of the comparison between the TRAC-PFI high order solute tracker with the experimental data clearly demonstrate that the high order method is conservatively under-predicting the mixing that would occur, as indicated by the experimental mixing data. This is primarily due to the fact that the high order solute tracker calculations do not account for the mixing that results from the iinpinging jet of coolant onto the downcomer walls of the reactor vessel. As such, the application of the high order solute tracker to the mixing transient calculations discussed below have significant conservatism inherent in the results. Furthermore, the mixing that would occur from the highly turbulent flow caused by the RCP impellers has not been credited.
Thus, larger volumes of unborated coolant could be shown to be acceptable if the mixing that would occur from these ignored effects (i.e., inlet coolant jet impingement on the downcomer and RCP impellers), were explicitly modeled.
This high order solute tracking scheme was not employed for the previously discussed loss of normal feedwater transients, as the boron transport was determined to be mainly convective, and numerical diffusion plays a very small role, if any,in driving the solute distribution within the system. As such, the runs not modeling unborated slugs of coolant were not repeated with the high order solute transport methods, since the expected results would be basically the same.
The results of this unborated slug analysis, where the RCPs were started in the loop containing the unborated water, yielded unborated volumes greater than 115 ft' for the situation where nominal decay heat had been assumed, and unborated volumes greater than 66 ft' for the situation where the decay heat had been assumed to be 1% of the ANS 1979 curve. In contrast, one RCP casing can collect less than 21 ft' before being exposed to the cold leg connection to the RCP casing. In the absence of cold leg loop seal piping, volumes of unborated water larger than 21 ft' per RCP casing, would begin to spill irto the cold leg piping to be mixed with the borated coolant in the RCS before reaching the reactor vessel. Thus, the maximum volume of unborated water j
1 that could collect in a steam generator channel head region cannot be greater than 42 ft' (i.e., two RCPs per steam generator outlet channel head; this equates to approximately 3.5% of the AP600 teactor vessel inlet plenum volume). The analysis results presented above indicate that approximately one and one-half times this credible value can be accommodated (i.e., this volume can theoretically accumulate and not result in the core inlet boron concentration dropping below the critical concentration following RCP restart in the affected adjacent loops) under low decay heat condithns, and more than twa and one half times as much under nominal decay heat conditions.
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NRC REQUEST FOR ADDITIONAL INFORMATION
........W Unborated slug analyses were also performed assuming that the unborated slug of coolant existed in one loop, and the RCPs were restarted in the opposite loop. The results from this set of analyses are directly applicable to SGTR recovery, as the recovery procedures regarding RCP restart will identify that the RCPs in the intact RCS loop must be restarted first. This analysis demonstrated that the resulting mixing due to the reverse flow through the faulted steam generator and associated RCS loop can accommodate extremely large volumes of unborated water in the faulted steam generator U-tubes and channel head and localized core inlet boron concentrations remain well above the critical boron concentration.
(c.) Recriticality has not been predicted for the rapid boron dilution mechanisms / scenarios addressed herein.
(d.) Protective Measures: EOPs. Others The Emergency Operating Procedures (EOPs) will be written by the combined license applicant. Westinghouse provides input to the EOPs through the Emergency Response Guidelines (ERGS). He ERGS stipulate that prior to restarting RCPs, there must be indication of subcooling based upon core exit thermocouple readings, and indication of pressurizer level. These two conditions allow for single phase natural circulation. He results of the analyses discussed in part (b.) of this response demonstrate that adequate mixing of the boron injected from the CMTs occurs under natural circulation conditions, even assuming a low level of decay heat.
RCP re-start is specifically addressed for the steam generator tube rupture accident by ERG # AE-3 j
(Reference 440.120-5). Steps include a note of caution regarding the potential of inadvertent criticality following any natural circulation or PRHR cooldown if the first RCP started is in the ruptured loop. This potential is significantly reduced when the first RCPs restarted are those in the intact loop, which is supported analytically as noted previously in section (b.) of tnis response.
Regarding the postulated loss of AC power during a dilution to criticality (also referred to as the " French" scenario), where it is assumed that the emergency diesel generators startup and provide power for the CVCS pumps. The addition of unborated makeup water to the RCS would continue without the RCPs in operation, thereby providing the means to accumulate unborated water in the RCS loop (s). However, it should be noted that the AP600 design includes a Battery Charger Input Voltage low signal which causes the DWS (demineralized water supply) isolation valves to close and aligns the BAT (boron acid tank) to the makeup pumps. Therefore, this postulated scenario is not a concern with respect to the AP600 design, as logic exists to prevent such an occurrence.
Concerns of RCP restart following maintenance that has a potential for the formation of low, or zero boron concentration water to accumulate in the RCS, are recommended to be addressed procedurally, for tny PWR design, as discussed in Reference 440.120-6. The means to prevent such a maintenance initiated scenario is that steps be included as part of the maintenance procedures to remove / mix this low, or unborated water volume.
l Measures include, but are not limited to, verifying that sufficient mixing will be present outside of the vessel, l
using feed and bleed, or drain and fill of the affected area.
j Regarding other protective measures, there will be interlocks in the logic controlling the RCP power supply.
L These interlocks, together with the AP600 ERGS /EOPs, will preclude the inadvertent restart of the RCPs following the actuation of the passive core cooling systems.
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I NRC REQUEST FOR ADDITIONAL. INFORMATION Conclusions It has been demonstrated that after a loss of heat sink event, following CMT actuation and RCS cooldown to equilibrium temperatures, RCS loop recovery will not pose a recriticality potential, as the boron addition from the CMTs becomes sufficiently mixed with the entire RCS. It has also been shown that unborated water accumulated upstream of idle RCPs, as is postulated during the "Finnish Center" scenario, to volumes approximately one and one-half times that physically possible to stagnate in the AP600 RCP casings of one of the steam generators, will not result in recriticality following RCP restart. Furthermore, conservatively calculated unborated volumes greater than the entire primary-side of a steam generator can be accommodated without recriticality concerns under reverse RCS loop flow circumstances (i.e., the RCPs are restarted in the loop opposite from that containing the unbo:/?+d coolant). The exceptional mixing that occurs in the RCS loops under the reverse flow configuration is utihzed procedurally for those instances where it will be apparent to the operator where the low, or unborated water volume may exist (e.g., the steam generator tube rupture ERGS).
Therefore, the AP600 design has shown that substantial boron dilution can occur, however unlikely, without leading to recriticality. Even though analyses indicate recriticality would not occur, additional steps have been prescribed to minimize boron dilution potential, thereby maintaining a " defense in depth."
References 440.120-1 Andreychek, T. S., et al., "AP6001.ow-Pressure Integral Systems Test at Oregon State University Test Analysis Report," WCAP-14292, Revision I, September 1995.
440.120-2 Macian, R., K. Ivanov, and G. E. Robinson, " Analysis of Boron Dilution Transients in the AP600". The Pennsylvania State University, Nuclear Engineering Department, June 1996.
440.120-3. Simplified Passive Advanced Light Water Reactor Plant Program, AP600 Standard Analysis Report, Section 15.2.7, " Loss of Normal Feedwater Flow," Revision 5. February 29,1996.
440.120-4 Macian, R., and John H,Mahaffy, "High Order Numerical Modeling of rd.ote Transport in System Codes" The Pennsylvania State University, Nuclear Engineering Department, September 1995.
440.120-5 " Revision 1 of the AP600 Emergency Response Guidelines," NTD-NRC-95-4525 / DCP/NRC0376, Docket No.: STN-52-003, August 9,1995.
440.120-6 Burnett, Toby, et al., " Risk of PWR Inadvertent Criticality During Shutdown and Refueling," NSAC-183 Westinghouse Electric Corporation, Electric Power Research Institute, December 1992.
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no,no-6 W westinghouse
NRC REQUEST FOR AD.DITIONAL INFORMATION Question 440.568 Re: ADS Test Analysis Report Figure 4.21 in the FDR and Figure 4-170 in the TAR purport to show the same tests (210,211 and 212). However, the actual flow and quality shown for the tests are substantially different in the two figures.
a)
Please explain the apparent discrepancy between the figures.
b)
Please provide a discussion of the relationship of ADS test data to projected AP600 operating conditions, and explain the (i) acceptability of tests nominally outside the range of projected operating conditions (ii) adequacy of the test program to provide data to model AP600 ADS operation, especially when some of the test conditions fall substantially outside the " intended" range. This applies to the above-mentioned tests, plus others in the ADS Phase B1 program, including:
i.
Test 250: a single test is not representative of the entire range of operating conditio.ns.
ii.
Tests 330,331 (multiple runs),320,321,322,350,351: discrepancies between FDR and TAR; acceptability of tests not meeting " intended" conditions; tests lying outside of envelope of operating conditions.
Response
a)
Figures 4.2-1 and Figure 4-170, Figures 1 and 2, attached respectively, do show the same tests. 'Ihe differences are as follows:
The steam quality calculation in the FDR is a simplified relationship:
[N -#,,) 100 (1) y Quality = (#,-#,,]
Where.
Water enthalpy at the bottom of supply tank (BW/lb)
Ha
=
Saturated water enthalpy at ADS inlet (BW/lb)
Hws
=
Saturated steam enthalpy at ADS inlet (BTU /lb)
Hs
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e i
440.568-1
l NRC REQUEST FOR ADDITIONAL.lNFORMATION IBil whereas, the TAR uses a more exact expression:
Qualby=
e,-N-100 (2)
V" q
Where:
h " = Latent heat at the ADS inlet (BTU /lb)
V
= Liquid mass in supply tank (Ib) 4*
= Combined (vapor and liquid) mass flow rate at ADS inlet (lb/sec)
- V"
= Liquid internal energy in supply tank (BTU /lb)
The TAR steam quality is considered to te a more rigorous and accurate expression. In -
general the expression in the TAR provides steam qualities higher than the values reported in the FDR.
The two equations are related by:
M n '0'un j
u s( &
3 Where internal energy is:
(3)
]
ya=H -PY e
g (4) 1 where:
E
= specific volume of fluid at bottom of supply tank (ft'llb)
P
= Pressure at bottom of supply tank (BRl/ft')
Therefore equation I assumes that:
V"
'V" (5) 0=PY+
d== ( &
4 where:
440.568 2 y
1 NRC REQUEST FOR ADDITIONAL INFORMATION i
M Nar,Mrr,pg.,yg (6) 2 4
4 The details of this relationship were reviewed for test A037210. Equation 6 makes
)
significant contributions to the steam quality from the term, #"ar, which is negative
(- -1.). %e P and V terms are of the same magnitude and of opposite sign and therefore cancel out. De PV terms of equation 5 is positive in sign but of a smaller magnitude than the second term (-10 versus - -65). It should be noted that the steam quality in the FDR was based on PT4 and the temperature of TElW. De exit enthalpy used was the
)
smaller of the enthalpy at FT4 or FT4 and TE1W. Use of the latter would have increased j
the FDR's estimate of steam quality by a couple percentage points. The TAR's enthalpy was based on FTlW.
There are some small differences shown in the mass flow rates as a result of the anist's renderings. He equivalent data for the two figures is presented in Table 1.
b)
As in (i) above, the shaded area for these tests defines a range of anticipated ADS operating conditions under various plant conditions. The plots were provided as a reference to put the test performed into perspective and to illustrate that the tests performed bound the range of expected operating conditionr. It was not the intent of the test program to test all possible combinations of j
4 operating conditions but rather to bracket the range of operating conditions, hus data was j
provided to validate the analytical tools and thereby confirm by analysis any other operating i
conditions as needed. In a number of cases the tests did not actually duplicate the intended conditions as identified in the test matrix but still provide valid test data on the performance of the system and were therefore suitable for analysis. Very high intended flow rates were not achieved due to the limited capacity of the pressurizer (steam / water supply tank) and other facility operating j
limitations. However, sufficient data was obtained to validate the application of the Henry Fauske Homogeneous Equilibrium Critical Flow Model for the ADS valves.
SSAR Revision: NONE l
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