ML20069K262
| ML20069K262 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 04/20/1983 |
| From: | Kammer D TENNESSEE VALLEY AUTHORITY |
| To: | Adensam E Office of Nuclear Reactor Regulation |
| References | |
| TASK-2.K.3.05, TASK-TM GL-83-10D, NUDOCS 8304260268 | |
| Download: ML20069K262 (7) | |
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TENNESSEE VAll'EEY AUTHORITY CH ATTANOOGAi TENNESSEE 374ol 400 Chestnut Street Tower II April 20, 1983 Director of Nuclear Reactor Regulation Attentiont Ms. E. Adensam, Chief Licensing Branch No. 4 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C.-
20555-
Dear Ms. Adensas:
In the Matter of
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Docket No.
50-327 Tennessee Valley Authority
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328 Enclosed for our Sequoyah Nuclear Plant is the response to the
- February 8,1983 letter from D. G. Eisenhut to All Licensees with Westinghouse Designed Nuclear Steam Supply Systems (Generic Letter 83-10d) regarding the resolution of TMI Action Item II.K.3 5, automatic trip of reactor coolant pumps. This information is being provided pursuant to the requirements of 10 CFR 50.54(f) and,. based on the enclosure, the operating licenses (DPR-77 and DPR-79) should not be modified, suspended, or revoked.
'If you have any questions concerning this matter, please get in touch with J. E. Wills at FTS 858-2683 Very truly yours, TENNESSEE VALLEY AUTHORITY h
D. S. Kainner Nuclear Engineer
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day of 1 b d 1983 O&M M s-Notar Public
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My C ssion Expires
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Enclosure cc: U.S.-Nuclear Regulatory Commission (Enclosure)
O Region II Attn: Mr. James P. O'Reilly Administrator 101 Marietta Street, NW,-Suite 2900 Atlanta,. Georgia -30303 i
8304260268 830420 DR ADOCK 05000327 PDR t
' An Equal Opportunity Employer
SEQUOYAH*NUEIME PLANT PLAN FOR RESOLUTION OF TWI ACTION ITEM II.K 3.5 INTRODUCTION The criteria for resolution of TMI Action Plan Item II.K.3.5, ' Automatic Trip of Reactor Coolant Pumps,' was stated in the letter from D. G. Eisenhut to All Licensees with Westinghouse Designed Nuclear Steam Supply Systems (Generic Letter 83-10d) dated February 8,1983. The following represents the plan for demonstrating compliance with those c ri t e ria. The overall philosphy and plan will be provided first, then each l
section of the attachment to NRC Generic Letter 83-10d will be e.ddressed as to how the overall plan responds to each NRC criteria.
OVERALL PLAN In the f our years that have passed since the event at Three Mile Island, Westinghouse Electric Corporation and the Westinghouse Owners' Group have held steadf astly -to several positions relative -to postaccident reactor coolant pump (RCP) operation. First, 'there are small break LOCAs f or which delayed RCP trip can result in higher fuel cladding temperatures and a greater extent of zircalloy-water reaction. Using the conservative evaluation model, analyses for these LOCAs result in a violation of the emergency core cooling system (ECCS) acceptance cri teria as stated in 10 CFR 50.46.
The currently approved Westinghouse evaluation model for small break LOCAs was used to perfona these analyses and found acceptable for. use by NRC in Generic Letter 83-10d.
Therefore, to be consistent with the conservative analyses performed, the RCFs should be tripped if indications of a small break' LOCA exist.
Secondly, Westinghouse and the Westinghouse Owners' Group have always f elt that the RCPs should remain operational for non-LOCA transients and accidents where their operation is beneficial to accident mitigation and
_ recovery..This position was taken even though a design basis for the plant is a loss of off site power. Plant safety is demonstrated in the Final Safety Analysis Reports for all plants for all transients and accidents using the most conservative assumption for reactor coolant pump operation.
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In keeping with these two positions, a low RCS pressure (symptom based) RCP trip citerion was developed that provided an indication to the operator to trip the RCPs for small-break LOCAs but would not indicate a need to trip the RCP for the more likely non-LOCA transients and accidents where continued RCP operation is desirable.
The low RCS pressure RCP trip criterion has already been incorporated into our present Sequoyah Emergency Operating Instructions (E01s). These procedures have been reviewed by your staff before Sequoyah received its operating license. The basis for this criterion is included in the generic Emergency Response Guideline (ERG) Backgound Document (E-0 basic revision, appendix A).
Relevant information regarding the expected results of using this RCP trip criterion 1
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t can be derived from the transients which resulted from the stuck open steam dump valve at North Anna in 1979, the steam generator tube rupture at Prairie Island in 1980, and the steam generator tube rupture at Ginna in 1982. The RCPs were tripped in all three cases; however, a study of the North Anna and Prairie Island transients indicated that RCP trip would not have'been needed based on the application of the ERG trip critorion.
The Ginna event, h ow ev e r, indicated a need to review the basis for the RCP trip criterion to allow continued RCP operation for a steam generator tube rupture for low head safety injection (SI) plant.
Thirdly, it has always been the position of Westinghouse and the Westinghouse Owners' Group that if there is doubt as to what type of transient or accident is in progress, the RCPs should be tripped. Again, the plants are designed to mitigate the ' effects of all transients and accidents even.without RCP operation while maintaining a large margin of safety to the public. The existing esorgency operating procedures reflect this design approach.
Lastly, it remains _the position of Westinghouse and Westinghouse Owners' Group that RCP trip can be achieved safely and reliably by the operator when required.
An adequate snount of time exists for operator action for the small-break LOCAs of interest. The operators have been trained on the need for RCP trip, and the emergency operating procedures give clear instructions on this matter.
In fact, one of the initial operator activities is to check if indications exist that warrant RCP trip.
Westinghouse and the Westinghouse Owners' Group will undertake a two part program to address the requirements of Generic Letter 83-10d based on the aforementioned positions for the purpose of providing more uniform RCP trip criteria and methods of determining those criteria.
In the first part of the progan, revised RCP trip criteria will be developed which provides an indication to the operator to trip the RCPs f or small-break LOCAs requiring such action but will allow continued RCP operation *or steam generator tube ruptures, less than or equal to a double-enden tube rupture.
The revised RCP trip criteria will also be evaluated against other non-LOCA transients and accidents where continued RCP operation is desirable in order to demonstrate that a need to trip the RCPs will not be indicated to the operator f or the more likely case s.
Since this study is to be utilized for emergency response guideline development, better estimate assumptions will be applied.in the consideration of the more likely scenarios. The first part of the program will be completed and incorporated into revision 1 of the emergency response guidelines developed by Westinghouse for the We stinghouse Owners' Group. The scheduled date f or completion of revision 1 is July 31, 1983.
The second part of the program is intended to provide the required justification for manual RCP trip. This part of the program must be done af ter the completion of the first part of the program. The schedule f or completion of the second part of the program is the end of 1983.
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The pref erred and safest method of pump operation following a small-break,
LOCA is to manually trip the RCPs before significant system voiding occurs.
No attempt will be made in this program to demonstrata the acceptability of continued RCP operatior. during a small-break LOCA. Further, no request for an exemption to 10 CFR 50.46 will be made to allow continued RCP operation during a small-break.LOCA.
DETAILED RESPONSE 10 GENERIC LETTER 83-10d Each of tho' requirements stated in 'the attachment to Generic Letter 83-10d will now be discussed indicating how they will be addressed. The organization of this section of the report parallels the attachment to Generic Letter 83-10d.
I.
Pump Operation Criteria Which Can Result in RCP Trip During Transients and Accidents A.
Setnoints for RCP Trin The Westinghouse Owners' Group response to this section of requirements will be contained in revision 1 to' the emergency response guidelines scheduled for July 31, 1983. IVA's plans and schedule for implementing revision 1 WOG ERGS at Sequoyah are provided in an April 15, 1983 letter frce L. M. Mills to you regarding Supplement I to NUREG-0737 - Requirements For Emergency Response Capability (Generic Letter 82-33).
1.
As previously stated, Westinghouse and the Westinghouse Owners' Group are developing revised RCP trip criteria which will ensure that the need to trip the RCPs will be indicated to ths operator f or LOCAs where RCP trip is considered necessary. The criteria will also ensure continued forced RCS flow f or:
a.
Steam generator tube rupture (up to the design bases, double-ended tube rupture),
b.
The other more likely non-LOCA transients where forced circulation is desirable (e.g.,
steam line breaks equal to or smaller than one stuck-open PORV).
2.
The criteria being considered for RCP trip are:
a.
RCS wide-range pressure < constant, b.
RCS subcooling < constant, c.
Wide-range RCS pressure < function of secondary pressure.
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Instrument uncertainties.will be accounted for.
Environmental uncertainty will be included if appropriate.
No partial or staggered RCP trip schemes will be considered. Such schemes are unnecessary and increase the requirements for training, procedures and decision making by the operator during transients and accidents.
3.
The RCP trip criteria selected will be such that the operator will be instructed to trip the RCPs before voiding occurs at the RCP.
4.
The criteria developed in item Al above is not expected to lead to RCP trip for the more likely non-LOCA and SGTR transients.
H ow ever, since continued RCP operation cannot be guaranteed, the emergency response guidelines provide guidance for the use of alternate methods for depressurization.
5.
The new generic revision 1 WOG omergency respon I guidelines will contain specific guidance f or detecting, managing, and removing coolant voids that result from flashing. The symptoms of such a situation are described in these guidelines and in detail in the background document for the guidelines.
Additionally, explicit guidance f or operating the plant with a vaporous void in the reactor vessel head is provided in certain cases where such operation is needed. TVA's plans and schedule f or implementing revision 1 WOG EEGs at Sequoyah are provided in an April 15, 1983 letter from L. M. Mills to you regarding Sappiament I to NUREG-0737, Requirements for Energency Response Capability (Generic Letter 32-33).
The present Sequoyah emergency and abnormal operating instructions contain specific guidance on establishment of natural circulation and natural circulation cooldown upon loss of RCPs. These present procedures are written to at old the establishment of voids in the reactor pressure vessel head.
6.
TVA has previously provided a response to item II.E.4.2 of NUREG-0737 on containment isolation dependability. We view the component cooling water system essential for RCP operation and as a desirable system for accident mitigation. Desirable systems are systems that, while not required for accident mitigation, significantly increase the plant's ability to cope with a small steam line break, small LOCA, or steam generator tube rupture. The systems f alling into this category are essential raw cooling water to the reactor coolant pumps and containment coolers, component cooling water to the RCPs and control air.
The systems are automatically isolated upon the 3
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receipt of a phase B isolation signal. Phase B is either manually or automatically initiated by 2 out of 4 logic on high-high containment pressure. The present asetssacy operating procedures require that the RCPs be, tripped within five minutes af ter containment phase B isolation occurs. We believe the present design and procedures to be adequate in this area since containment phase B is indicative of either a LOCA or staan line break.
The reactor coolant pump seal inj ection is supplied f rom either the. positive displacement (PD) pump or centrifugal charging pumps at Sequoyah.
The contifugal charging pumps are the high head amergency core cooling system (ECCS) pumps and are described in section 6.3 of the FSAR.
RCP seal inj ection is therefore provided for all design basis accidents.
7.
Discussed in Al and A3.
B.
Guidance for Justification of Manual RCP Trio._
The Westinghouse Owners' Group response to this section of requirements will be reported separately at the end of 1983.
TVA will review dhe WOG generic report and provide an additional response after our review.
1.
A significant number of analyses have been performed by Westinghouse for the Westinghouse Owners' Group using the currently approved Westinghouse Appendix K evaluation model for small-break LOCA.
This evaluation model uses the WFLASH code.
These ' analyses demonstrate f or small-break LOCAs of concern that the predicted transient is nearly identical to those presented in the safety analysis reports for all Westinghouse plants if the RCPs are tripped 2 minutes following the onset of reactor conditions corresponding to the RCP trip setpoint.
Thus, the safety analysis reports for all plants demonstrate compliance with this requirement. The analyses performed for the Westinghouse Owners' Group will be used to demonstrate the validity of this approach.
t 2.
Better estimate analyses will be performed for a limiting Westinghouse designed plant using the WFLASH computer code with best estimate assumptions. These analyses will be used to determine the minimum time available for operator action for a' range of break sizes such that the ECCS acceptance criteria of 10 CFR 50.46 are not exceeded.
It is expected that the minimum time available f or manual RCP trip will exceed the guidance contained in the draft ANSI Standard N660. This will justify manual RCP trip for all plants.
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C.
Other Considerations 1.-
The RCP criterion in the present emergency operating instructions utilize the wide-range RCS (systems) pressure instrumentation. Wide-range RCS pressure is a postaccident monitoring (PAM) instrument and is described in section 7.5 of
.the FSAR and in TVA's response to NRC in a letter f ras L. M.
Mills to E. Adensam dated March 15, 1982 on Regulatory Guide 1.97.
In addition, the environmental qualification of this instrumentation will be submitted in accordance with the final
- rule - 10 CFR Part 50.49 Environmental Qualification of Electric - Equx; 'ent important to Saf ety f or Nuclear Power Plants.
2.
The revision 1 WOG em:rgency response guidelines contain guidance for the timely restart of the reactor coolant pumps when conditions which will support safe pump startup and operation are established. TVA's plans and schedule for implementing revision 1 WOG lERGs at Sequoyah 'are provided in an April 15,1983 letter from L. M. Mills to y.on regarding Supplement 1 to NUREG-0737 - Requirements For Emergency Response Capability (Generic Letter 82-33).
3.
Before initial startup of unit 1 at Sequoyah, all licensed personnel-received special amergency operating instruction (E0I) training.
Since the RCP trip was already integrated into our EDIs, the special training also included the RCP trip step. The present E0Is are based upon the Westinghouse emergency operating instructions. All licensed personnel periodically receive requalification training on the current E0Is which includes RCP trip. New licensees receive license certification training on the current E01s which includes RCP trip.
II.
Pump Operation Criteria Which Will Not Result in RCP During Transient
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and Accidents The pref erred and safest method of operation following a small-break LOCA is to manually trip the RCPs; therefore, there is no need to address the criteria contained in this section.
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