ML19352A431
| ML19352A431 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 12/05/1979 |
| From: | Thadani A Office of Nuclear Reactor Regulation |
| To: | Stahle C Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18025B195 | List:
|
| References | |
| FOIA-80-587, REF-GTECI-A-09, REF-GTECI-SY, TASK-A-09, TASK-A-9, TASK-OR NUDOCS 8104170065 | |
| Download: ML19352A431 (5) | |
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C. Stahle FROM:
Ashok Thadant We have reviewed the ATWS emergency operating procedure submitted by TVA on October 17, 1979. The enclosed cormients on this procedure are written in such a manner as to minimize the TVA effort required to develop accep-table ATWS procedure (s).
I reconcend that you request the Operator Licens-ing Branch to review the proposed TVA procedure and our coments.
If requested, we will be pleased to discuss these coments with TVA.
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Ashok Thadani Reactor Systems Branch Division of Systems Safety cc: ATWS Task Force Sf Hanauer
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REVICW OF ATWS_ PROCEDURES FOR SE0UOYAH PLANT A.
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.1 The protecure lists the parameters which cause the reactor to scrar.,
but does not describe the actual indications available to the operators in the centr 51 room which would make him aware that an ATj.[ event has j
occurred. These ATWS symptoms would depend on initiating event and, therefore, they ought to be evaluated for at least the following three key events:
Loss of Main Feedwater Loss of Offsite Power Stuck Open PORY In making the evaluation it is important to show for each event what symptoms would indicate to the operator that scram action was called for but did not occur.
B.
Automatic Action 1.
This section does not address how the automatic actions relate to ATWS.
Some of the automatic actions (e.g., turbine trip) may not even occur after an ATWS. This should be specified in more detail in the procedura.
2.
Why is automatic actuation of HPSI not included in this section of the j
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C.
Immediate Ooerator Action l
1.
The procedure shoula specify critical indications available to the operator consistent with the initiating event and assumption that the reactor trip has not occurred.
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The imediate actions that the operators have to take af ter ATWS h:s occurred and an attempt to ranually scram the reactor from the control room has failed should follow two parallel paths. While one operator should continue the operation of manually scraming the reactor by trip-ping the breakers powering the control rod drive MG sets, the other operator should initiate the other actions leading to safe shutdown of the plant. The procedure should reflect that the actions described in sections A.2.b and A.2.c and those described in sections B.1 and B.2 are to be performed simultaneously. Section B should require sequential actuation of turbine trip, all auxiliary feedwater pumps, and high pressure safety injection system.
(See Figure 1).
3.
Describe the actions taken by the operator when he discovers, during the verification of reactor coolant system status (section C), that the conditions are not within the prescribed limits. What is the impact of 1 css of offsite power on availability of those signals to the operator.
What is the shutoff head of the HPSI pumps? What provisions are taken to prevent pump damage when HPSI is operating against the RCS pressure which is higher than the shutoff head of the ptznp?
D.
Subsequent Operator Action 1.
What is the time frame for these actions?
2.
What criteria are provided to verify that:
- a. The auxiliary feedwater system is providing the necessary flow to the steam generators.
- c. The containment heat removal is being accomplished, if the containment conditions are outside the nomally specified valves.
. 3.
What additional procedure does the operator have to follow in order to bring the plant to and maintain in a. cold shutdown condition after an ATWS? For example, what borori concentration should be mintained in the RCS.
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SUBSEQUErlT ACTIONS Event & Action Transient Failure to Immediate Opera-Verify RCS, Steam Long Tenn Sequence Initiated Scram tor Actions Generator, Contain-Shutdown Symptoms Symptoms Two Operators ment Parameters Values General time t
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4 Sequence i
Operator #1 Manual Scram If outside specified Describe special Attempts limits, describe the actions to bring operator actions, plant to a cold shutdown condi-Operator #2 tion and main-Assure that tain that condition.
a) turbine tripped b) all AFWS provid-ing flow c) HPSI providing flow (shut off head) in that order.
What, if any, is the impact of stuck open PORV.
i Figure 1.
Generalized Approach to be followed for writing ATWS procedure (s)
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Fn 7 na e,f (gf TMI-2 Generic Tasks:
A-3, 4, 5, 9 MEMORANDUM FOR:
L. E. Phillips, Acting Chief Analysis Branch, DSS FROM:
F. Odar, Analysis Branch, DSS
SUBJECT:
BROOKHAVEN NATIONAL LABORATORY TRIP REPORT -
SEPTE!EER 5-7, 1979 The purpose of the trip was to discuss and review 1) an audit calculation performed by BNL in order to assess the impact of TMI-2 incident on Bab:ock & Wilcox plants, 2) the development of the IRT and RAMONA computer programs and 3) the scope of the new ATWS audit calculations.
1.
Audit Calculations - Impact on TMI-2 Incident on B&W Reactors Brookhaven National Laboratory performed an analysis of overfeeding transient for a typical Babcock 8. Wilcox 177 fuel assembly plant using the IRT code.
In this transient it ;s assumed that both of the turbines trip and the ICS fails so that feedwater is continuously pumped into both steam generators.
When ECCS is activated, the main coolant pumps trip as per the new require-ment by B&O task force.
The steam generators are modeled using the Mark I model. The metal heat in the upperhead was not modeled. The plots indicate some quality in the hot leg. This nuality corresponds to approximately 28% vapor fraction according to homcgeneous equilibrium model. Enclosure 1 presents the sequence of events, and the plots of primary variables during this transient.
BNL will perform another computation considering the metal heat cacacity in the upperhead.
2.
ATWS Audit Calculations - Unresolved Safety Issues A-9 The program and the scope of calculations were discussed with the represen-tatives of BNL and with a representative of Battelle Columbus (Dr. Robert Collier). Battelle Columbus will provide the necessary manpower to Brookhaven National Laboratory for the ATWS Audit Calculations (Reference 1 Technical Assistance Contract, FIN A3318 which is produced in Enclosure 3).
Brookhaven National Laboratory will perform BWR ATWS calculations using BNL-TWIGL and RELAP 3B programs with the manpower available at BNL. However, BNL will need additional manpower from Battelle Columbus to perform PWR ATWS audit calculations. BNL will provide computer card decks with new nodalization to Battelle Columbus and Battelle Columbus will perform necessary calculations under BNL guidance.
For this purpose BNL is planning to sub-contract some of the workgin Reference 1 to Battelle Columbus.
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L. E. Phillips NOV ; 3 61 presents the type of computer runs to be performed by BNL and Battelle Columbus under BNL cognizance. The scope of this work is slightly different from that in Reference 1.
Form 189 which will be sent by BNL in response to Reference 1 will be modified by NRC to reflect these changes.
BNL and NRC representatives agreed that the scope of work in Reference 1 did not expand as a result of these modifications. also presents existing nodalization for typical Westinghouse (4 loop), Combustion and Babcock & Wilcox plants.
3.
Analysis of Steamline Break Accident with Concurrent Steam Generator Tube Rupture - Unresolved Safety Issue A-3, 4, 5 BNL gave the new target dates for completion of the analyses. The following table reflects the new dates:
Report to be Issued W
- issued Feb 15/80**
Phase 2 Feb 1/80 Feb 1/80 Feb 15/80 (inc. DNBR calculations) the report has been received by NRC
- see alsc the next item - Development of the IRT program 4.
Development of the IRT Program Mark I steam generator model implementation in the IRT code will be completed by October 6, 1979. Analyses of steam line break accident with the rupture of steam generator tube (s) will be completed by January 15, 1980 and the final report will be issued by February 15, 1980.
Steady state initialization will not be completed by BNL during this calendar year. This task turned out to be more complex than originally estimated. BNL will make a written proposal to accomplish this task within FY80 program.
Improvements in steam generator modeling are being considered by BNL. presents the list of additional work needed for implementation of CANDU steam generator model in the IRT code.
5.
Development of the RAMONA-III Program a.
Verification of the RAMONA-III Program Verification of the RAMONA-III program using Peach Bottcm turbine trip tests continues. The original program as obtained by BNL did not have capability to model the Peach Bottom plant. BNL obtained some updates (including code redimensioning) in order to model the plant. These updates are implemented and the program performs calculations.
L. E. Phillips 9g presents comparisons between two previous BNL models, Peach Bottom test data and the RAMONA-III code for the Peach Bottom Turbine Trip Test #3. The two previous BNL models have been presented in References 2 and 3 and they agree very well with the test data. However, the agreement between the RAMONA code and the test data is very poor.
In.these calculations BNL used i) the cross section data from SCANDPOWER, ii) measured dome pressure from the test data since the RAMONA-III code does not have modeling of steamline dynamics, iii) a jet pump input head curve from SCANDP0WER since the RAMONA-III code does not model the jet pump or recirculation loop and iv) other input parameters as specified by SCANDPOWER.
BNL is preparing a new input deck including the BNL cross section data, feedback model and BNL interpretation of the Peach Bottom geometry. BNL expects to perform the first Peach Bottom transient calculation using these data by October 1,1979.
b.
Development of new Models BNL is developing new models and is in the process of modifying the RAMONA-III code to include these models. The following is the summary of these new models.
Status 1.
Steam Separator Model Completed and implemented in one version of the code.
ii. Jet Pump and Recirculation Completed and it is being implemented Loop Model in one version of the code. Expected date for completion is November 1,1979 iii.
Fuel Model Completed and it is being programmed.
Expected date for completion of imple-mentation is October 1,1979.
iv. Steamline Dynamics Programming is complete.
It is being implemented. Expected completion date is November 1,1979.
References 1.
Letter from F. Schroeder to R. H. Bauer describing Technical Assistance Contract A3318, dated August 24, 1979.
2.
Analysis of Licensing Basis Transients for a BWR/4, BNL & NUREG-26684.
L. E. Phillips NOV 19 373 3.
Thermal-Hydraulic Analysis of Peach Bottom II Turbine Trip Tests, BNL-NUREG-25526.
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- b Fuat Odar Analysis Branch Division of Systems Safety Enclosures Distribution:
F. Schroeder R. Frahm R. Denise B. Sheron S. Hanauer W. Jensen R. Mattson F. Odar R. DeYoung D. Hunter (Region III)
D. Ross P. Matthews V. Stello M. Rubin L. Tong E. Throm i
D. Eisenhut A. Thadani R. Tedesco C. Graves Z. Rosztoczy S. Newberry T. Novak G. Kelly S. Fabic G. Holahan.
K. Kniel H. Silver P. Check S. Salah B. Liaw S. Weiss B. Grimes D. Fieno P. Norian M. Mentonca S. Israel N. Zuber G. Mazetis R. McDermott J. Heltemes W. Kato (BNL)
W. Minners M. Levine (BNL)
M. Aycock W. Shier (BNL)
F. Almeter W. Wulf (BNL)
J. Strosnider G. Knighton K. Parczewski R. Audette 1
ENCLOSURE 1 Overfeeding Transient for a B&W 177 Fuel Assembly Plant l
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DEC 6197C 5-MEMORANDUM FOR:
A. Thadani, Task Manager, A-9
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Karl Kniel, Chief, Core Performance Branch, DSS FROM:
Ralph 0. Meyer, Leader, Reactor Fuels Section, Core Performance Branch, DSS
SUBJECT:
FUEL FAILURE CRITERIA FOR ATWS RULEMAKING It is our understanding that a proposed rule on anticipated transients without scram (ATWS) is to be submitted to the Commission early in 1980 and that the rule will contain information of the sort presented in Apper. dix of NUREG-0460, Vol. 2.
Section IV.$ of that appendix sets forth some proposed acceptance criteria and assumptions to be used in the cal.culation of radiological consequences.
Because the fuel rod cladding serves as the first barrier to fission product release, the first step in a dose calculation for a postulated ATWS requires an estimation of the number of rods that will fail (i.e.,
that will experience cladding perforation or rupture). We have pro-vided you with guidelines for ATWS fuel failure prediction in memoranda spanning the last 1 1/2 years or so. Those guidelines were also presented in Appendices XIV to XVII of NUREG-0460, Vol. 2.
Earlier this year, we restated our position (memorandum, Meyer to Thadani, January 26,1979) so that you could provide the industry with guidelines for the "early verification" effort.
l Our fundamental requirement has been, and continues to be, that all relevant fuel rod failure criteria, whether of thermal / hydraulic or mechanical origin, should be taken into account in the calculation of radiological consequences.
In most cases, existing failure criteria and models are adeouate for use in ATWS fuel behavior analyses. As you know, however, we have had difficulty in dealing with pellet / cladding interaction (PCI) because we have lacked acceptable criteria and models.
i Consequently, our position regarding the calculation of CPI-initiated failure for ATWS has been as follows:
1.
For DWRs, we have stated that the number of rods that fail due to PCI should be calculated, but we had not specified how this was to be done. We had assumed that the vendors would submit PCI failure estimates and models for us to review, but we have received nothing in this area.
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For BWRs, we have stated that the number of rods calculated to be in boiling transition, which is relatively large (s10 to 17",) for a MSIV-closure A1VS, would be likely to encompass the number that would fail due to PCI (in part, because not all rods in boiling transition are sure to fail).
The above-stated position stemed from the fact that, while we believed that there was a significant probability of PCI failure during power-increasing ATWS events, we did not have a PCI model for use in reactor regulation. With the development of the Battelle Northwest PROFIT model, however, that deficiency has been eliminated. We, therefore recommend that the purposed ATWS rule be phrased to require the calculation of PCI-initiated fuel failure for events involving power increases, and that in lieu of an approved vendor model, calculation should be made with a model to be provided by NRC. Because of the need for judgment and flexibility in using a PCI model, the rule should not specify further details regarding the particular model to be used, but PROFIT will be available in case we need it.
To effect as much consistency as possible regarding the treatment of PWRs and BWRs, the above position should apply to both types of reactors.
Because the BWRs also have a large number of rods that are calculated to fail on the basis of thermal / hydraulic criteria, and because we believe it would be overly-conservative to add those rods to the number calculated to fail by PCI, we recomend that the larger of the two estimates should be used in the dose calculation. (Note that for PWRs this is not an issue since no rods are currently calculated to be in DNB for any power-increasing PWR AT'JS).
Except for the modifications indicated above, the remainder of our ATWS fuel failure recomendations remain unchanged, n
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Ralph 0. Meyer,t.eader Reactor Fuel Section Core Performance Branch Division of Systems Safety cc:
S. Hanauer R. Mattson R. Denise F. Cherny F. Akstulewicz M. Tokar K. Kniel
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R. J. Mattson The newly formed task force on ATWS met for the first time on 1/15 to discuss the work required between now and May 1979 and to develop the questions / statements for generic ATWS analyses. The task force members were requested to provide their input to the generic set of questions /
statements before 1/24/79.
During this useful period of briefing of the task members followed by exchange of viewpoints, the following important questions / comments were raised and discussed.
1.
Since MTC value specified is based on estimates of future operation, what, if anything, can we do if the future operation is different than that assumed in the development of the specified value?
My comment: The applicant should be required to recognize that if the plant design or operation changes appreciably in the future such that the plant does not fall within the generic envelope, he may be required to reconsider earlier ATWS conclusions.
If this is reasonable, would the rule or the regulatory guide provide the necessary mechanism for accomplishing this objective.
2.
If the plant were to be permitted to operate at its " stretch" rating, how would we treat such a large (in some cases) change in an important parameter?
My comment: Same as under 1. above.
3.
Some questioned the use of nominal values of parameters in generic analyses and recomended using bounding values. Also, what if the sensitivity to a parameter is very high?
My coment: An objective is to detemine, as well as we can, the l
realistic course of an ATWS and thus we should use nominal values.
If there are small differences in the nominal parameter values for a class of plants, the sensitivity studies could and should be l
relied on to make judgments.
Additionally, my judgment based on review of earlier ATWS analyses is that there is no threshold phenomenon (i.e., extreme consequence dependence on small variation in a parameter value); however, if there is a very important parameter whose initial value is not well understood, use of a conservative value could be required if the preverification approach is to be successful.
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R. J. Mattson JAN 17 79T9 4.
How should PCI failures be treated?
My connent UseNdREG-0460,Vol.2approachorspecifywhatthe penalty might be.
Eliminate, if possible, vague guidelines.
5.
Jim Norberg indicated he would need assistance from someone familiar with ArdS and developing rules and reg. guides.
Perhaps Roger Mattson could ask for John Huang, an ex-member of Standards and now, I believe, in the Division of Operating Reactors.
6.
Frank Cherny emphasized the need for DOR participation in the review of mechanical engineering aspects of operating reactors.
He recom-mended that we request Keith Wichman of D0R to work with us on ATWS.
7.
A difficult question, because of variety of subjects, was the required format for the task force members to prepare their questions and/or comments.
My comment:
I think the most straightforward approach is to state what we want.
Examples:
Identify approved models.
Identify open areas and recommend a way to resolve open areas.
Specify what kind of penalty may be imposed if the vandor does not provide acceptable response.
(Note:
No Q1s or Q2s.)
Specify transients to be analyzed.
l Specify ICs and sensitivity studies.
Specify assumptions for alt. #3 and alt. #4 analyses.
Require list of plants which fall under each set of analyses.
Require list of systems relied on.
Specify requirements for these systems for different alternatives.
Specify what the analysis must include as a minimum.
Specify the constraints on future design or operational variations.
Specify criteria under which dose calculations need not l
be performed.
Specify limits and operability criteria and require vendors I
to show how each class of plants would meet these limits.
Keep in mind different approaches in PWRs on alt. #3 and alt. #4.
Require vendors to specify the necessary plant modifications to satisfy the criteria of Volume 3, NUREG-0460.
Require vendors to provide sufficient detail to ascertain that the mitigating systems. criteria of Vol. 3 of NUREG-0460 shall be satisfied.
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R. J. Mattson JAN 17 ST9 8.
Allotted time for this ambitious approach is too short.
I would appreciate (a7 your comme its, especially if you have any dis-agreements with the above approach and (b) your requesting DOR to add John Huang and Keith Wichman to the ATWS task force.
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Ashok C. Thadani Reacter Systens Branch cc:
R. Tedesco
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G. Hanauer 3 T. Novak F. Cherny T.M. Su H. Richings D. Thatcher F. Odar S. Salah G. Kelly M. Tokar R. Woods R. Lobel V. Rooney G. Chipman E. Jakel J. Norberg '
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