ML20090H603
| ML20090H603 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/14/1984 |
| From: | Houston R Office of Nuclear Reactor Regulation |
| To: | Miraglia F Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML082380335 | List:
|
| References | |
| FOIA-91-106, RTR-NUREG-0916, RTR-NUREG-916 TAC-49346, NUDOCS 8402270070 | |
| Download: ML20090H603 (11) | |
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MEMCo A FORT Frank J. Mira011a, Assistant Director for Safety Assesment Division of 1,' r,ensing
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FROM:
R. Wayne Houston, Assistant Director for Reactor Safety Division of Systens Integration
SUBJECT:
SAFETY EVALUATION FOR R. E. GINNA Nutt. EAR P(NER PLANT --
LONG TERM RESTART REQUIREENTS 6 AND 10 Plant Names R. E. G1Ma Nuclear Power Plant l
U Docket Number 50 244 TAC Numbert 49346 Responsible Branch:
ORB H Project Managet:
G. Dick Review Status:
Complete The Reactor Systans Branch has completed its evaluation of long ters itens 6 and 10, as provided in the licensoe's response of November 22. 1982, to r
the requirements of NUREG-0916 "$ER Related to the Restart of R. E. Ginra Nuclear Power Plant". The safety evaluation is provided in the enclosure.
L An SER for the other itens of TAC 49346 (itans 11,12120) was prepared by PSRB.
Crit;,lnalSigned By H. Wayne Houston -
R. Wayne Houston. Assistant Director for Re6ctor Safety Division of Systems Integration Enclosures
-As stated cc:
R. Mattson
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SAFETY EVALUATION REPORT R. E. GINNA NUCLEAR POWER PLANT GINNA STEAM GENERATOR TUBE RUPTURE (SGTR) PROCEDURES TAC NO. 49346 g
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INTRODUCTION d
,s a result of the SGTR sccident on January 25,14'.2, at the Ginna Plant, an i
NRC task force was formed to report the circumstances surrounding the tube
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failure. The task force documented its findings in Reference (1).
Subsequently, a safety evaluation report was prepared to detemine if the Ginna Diant cocid be returned to full power operatien (Reference 2), This report included relevant information from Reference (1), licensee submittals and significant task force findings.
The report retornended restart of the Ginna plant.
The bases for this recomendation included the staff's review
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of the.oechanism that led to the Ginna SGTR, the staff's findings including the edcuM1 af the Ginna repair and inspection program and the operaters' compliance with aoplicable inna procedures. Additionally, indications are that ue recctor vessel was not subject to pressurized thermal shock during this event and that the of fsite radiological conseque, ices of this event were well within 10 CFP Part 100 dose guideline levels. The licensee was committed to a series of short term and long term requirewnts. The licensee responded to its long term commitments in Reference (7). The purpose of this SER is to present the staff evaluation of the licensee's compliance with long range items 6 and 10 The evaluation of other peccedural items is contained
-in Reference (4).
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2 ITEM 6 Within six months, review the requirement for a safety injection signal to be present for automatic transfer of safety injection pump suction from the boric acid storage tank to the refueling water storage tank.
LICENSEE'S, RESPONSE The requirement for a safety injection signal to be present for the autoniatic transfer to take place has been reviewed.
The results of the review indicate that it it acceptable to remove this dependency, A modification will be made o
to the automatic switchover logic that will cause the switchover to occur on
,,. boric acid storage tank level only. The presence of an SI signal will not be require,d for the automatic switchover to oscur. The modification will be implemented prior to startup from the 1983 refueling outage.
STAFF EVALUATION In the event of a large steam line break (SLB) rapid addition of concentrated boric acid solution is required to maintain the reactivity and consequent power level, within acceptable limits.
Therefore the Ginna safety injection (SI) pumps take suction from the boric acid tanks (BATS), which contain concentrated boric acid (21,000 ppm boron). Howaver the BATS only contain 7200 gallons-and are thus quickly depleted in the event of rapid depressurization of the RCS. The SI pump suction -is therefore automatically switched to the refueling water storage tank (RWST) on low BAT level. The
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plant design previously required that the SI sign'als be present f' this switchover to occur. The licensee therefore revised the emergency procedures to specify that the SI signal be reset only af ter MOV 825A or D (tae SI suction valves from the RWST) *>ere open.
Actuation of the SI signal also automatically isolates the containment resulting in isolation of letdown and interruption of the reactnr coolant pump (RCP) seal water return flow and instrument air supply, Reset ot' the $1 signal would permit reestablishment of normal RCP seal flow, nonnal letdown and charging and allow operation of the pressurizer spray.
The limitations
.. imposed by the transfer logic circuitry and the-emergency procedures could cause de, lay in the e.ilization of equipment which can mitigate the consequence 3 of a SGTR event.
The staf# determined that the licensee should perform a review of the need for a coincident SI signal for automatic transfer lof SI pump suction from the BATS to the RWST on BAT low level, (Reference 2).
l As indicated in the licensee's response the requirement for an SI signal to-be present for the automatic t ransfer has been reviewed, and the results of this review indicate that it is acceptable and desirable to re ove ; is dependency. This is because SI reset and reestablishment of necessary systems could be accomplished quicker without defeating the SI switchover a
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requirement. We conclude that the licensee has adeouately responded to this requirement.
ITEM 10 Within six months, review plant procedures to provide any additional gteidance J
required for operator act'lons to be taken in response to real cr suspected reactor vessel upper head vo,iding.
LICENSEF'S RESPONSE Additional guidance beyond that present in the Ginna procedures on January 25 regarding real or. suspected reactor vessel upper head voiding has been found necessary in two areas, safety injection termination and reactor coolant pump restart,- Additional guidance has been added to the S/G Tube Rupture and Loss of Secondary Coolant procedures to permit SI termination with a upper RV head voio as long as natural circulation and other Si c
termination criteria are met.
Guidance -has. also been added to the "E" series procedures (major accident -
The procedural) concerning upper RV head void collapse during RCP start.
procedures permit RCP start with an upper head-void-as long as adequate pressurizer level and RCS subcooling_are present.
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ys STAFF EVALUATION Ouring the Ginna SGTR event, void formation apparently cccurred first during the initial depressurization following reactor trip, and again after the PORY stuck open.
The latter was the more severe case. However there never was any Indication that the water in the core did not stay subcooled. The licensee agreed to perform detailed thermal-hydraulic analyses for the SGTR event. These analyses included Westingnouse LOFTRAN calculations and auxiliary ceiculdtions employing standard ms:s :nd energy balance techniques tu address the limitations of the LOFTQAN results.
The analyses are evaluated in Enclosure 1 of Reference (3).
The staff concluded that, in
.. - spite of some liraitations in the LOFTRAN' program, the analyses supported the verification of the system phenonena including void formation, as required in Reference (2), and that the information provided by the licensee was therefore acceptable.
The licensee's evaluation of RCP restart requirements following an SGTR event is presented in Attachment D of reference (7). This evaluation assesses the potential-for coolant flashing and loss of pressurizer pressure control during pump startup. Depressurization of the RCS following an SGTR may l
generate a steam bubble in the reactor vessel upper head region if the RCP's i
l are not operating. This bubble could rapidly condense on RCP restart,
. drawing liquid from the pressurizer and reducing RCS subcooling. This could-result in loss-of level indication and pressurizer heater unavailability, l
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. thus losing the ability for pressure control and direct indication of coolant inventory.
In addition, local flashing in the RCS could result in erratic system response. These conditions would make plant control more dif ficult and may confuse the operator.
The licensee has performed calculations to determine RCS pressure rerponse to the collapse.of an upper head void. Based on these, minimum indicated levels war; cciculsted that would assure: (1) no heater uncovery; (2) no loss of level indication. Emergency operating procedures for Ginna establish a mininum level of 80 percent before restarting a RCP. This criterion assures that an indicated level will be maintsined for initial RCS pressure greater than 62Q psia. For large voids, pressurizer heaters may not remain available, but guidance is provided to restore level using the charging pumps, and if necessary, reinitiate tafety injection.
Minimum reactor subcooling requirements, consistent with an_ initial-pressurizer level of 80 percent, were calculated for different RCS pressures.
For RCS pressures less than 1100 psia, the required subcooling is less tnan 49'F.
These results include instrument uncertainties.
SGTR emergency procedures require a minimum of 50'F subcooling.
RCP restart is only per--
g nitted after the primary and secondary pressure are equalized. The maximum l
secondary pressure would be 1100 psia (approximate safety valve set point).
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Therefore the RCS would remain subcooled following RCP restart with the Ginna 3
subcooling criteria.
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The licensee has stated that the sequence of recovery actions follou the i
Westinghouse Owners Group (WOG) Emergency Response Guidelines (ERG's) and should ensure early termination of the break flow (Reference 5).
In Ref erence (6) the staff concluded that actions prescribed in the WOG ERG's for the SGTR accident are generally acceptable.
Areas requiring improvement include Sl termination criteria, guidelines 'or combination SGTR/LOCA, and clarification of the use of non-safety related equipment for accident mitigotion.
The SI termination criteria require that once the primary system and ruotured steam generator pressures are equalized, primary system pressure mus t aga t, be increased by 200 psi by SI flow.
This action would reestablish leak flow from the RCS. The NRC position is that the criteria of pressurizer
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e level and RCS subecoling also prescribed in the ERG ere adequate to protect the core without the additional requirement of RCS repressurization.
These issues will be addressed in future ERG revisions.
We conclude that the licensee has' adequately. responded to this requirement, subject to adequate resolution by the WOG of the ERG areas requiring improvement and additional information, particularly with regard to SI termination criteria.
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REFERENCES, 1.
NUREG-0909 "NRC Report on the January 25, 1982 SGTR at R. E. Ginna Nuclear Power Plant", April 1982.
2.
NUREG-0916 "SER Related to the Restart of the R. E. Ginnc Nuclear Power Plant", May 1982.
I 3.
Kerx; from R. W. Hou3 ton to G. C. Lainas, "SER fcr R. E. Ginna Nuclear Power Plant", July 27, 1983.
4.
Memo from D. L. Ziemann to F. J Miraglia " Evaluation of Rochester Gas 1. Electric Corporation Respons_ to Long Term Items Contained in the Ginna Restart SER, SGTR Incident", August 23, 1983.
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5.
Telephone conversation with RG&E, July 15, 1983.
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Memo from R. J. Mattson and H. L. Thompson to D. G. Eisenhut "SER for q Emergency Response Guidelines", May 19, 1983.
7.
Letter from John E. Maier, Rochester Oas & Electric Co., to Dennis M. Crutchfield, NRC, " Response to Safety Evalu' tion Report NUREG-0916.
SGTR Incident - R. E. Ginna Nuclear Power Phnt" November 22, 1982.
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o SALP Inout for TAC 49346 The purpose 0." this attachment is to document our evaluation of the licen-
-see's performance during D51's review of the subject operating reactor action. The following criteria from NRC Appendix 0516 are the only ones sa I
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relevant to this evaluation:
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Management involvement in Assuring Quality e
This action was handled by personrel at the appropriate level of manage-nent. The utility involved :he necessaiy in-house technical staff, to i
help bring about a solution to the salient issues. The analytical effort was performed by Westinghouse.
Rating: Category 2 2.
Anoroach to Resolution of Technical Issues The licensee approached the resolution of the tecanical issues involved in responding to the long term requirements of reference 1.2) in a competent ::ianner. Their resources were used properly, and the work was I
submitted on time. There was no need to obtain additional clarification s
from the staff after completion of reference (2).
Rating: Category 1
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Response to NRC Initiatives The licensee's was generally responsive to NRC Initiatives.
Rating:
Category 2 4.
Overall Rating:
Category 2 I
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SALP Inout for TAC,49346 The purpose of this att:chment is to document our evaluation of the licen-see's performance durin; DSI's review of the subject operating reactor action. The following criteria from NRC Appendix 0516 are the only ones relevant to this evaluation:
1.
Management involvement in Assuring Quality This action was handled by personnel at the appropriate level of manage-ment. The utility involved the necessary in-house technical staff, to help bring about a solution to the-salient issues. The analytical effort was performed by Westinghouse.
Rating: Category 2 2..
Approach to Resolution of Technical Issues The licensee approached the resolution of the technical issues involved in responding to the long term requirements of reference (2) in a competent manner. Their resources were used properly, and the work was
- submitted on time. T; sre was no need to-obtain additional clarification from the staff af ter completion of reference (2).
i Rating: Category 1 f
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Response to NRC Initiatives The licensee's was generally responsive to NRC Initiatives.
Rating:
Category 2 4.
Overall Rating:
Category 2 O
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