AECM-87-0209, Application for Amend to License NPF-29,supporting 870813 & 1023 Applications for Amends.Amend Modifies Standby Liquid Control Sys Discharge Piping,Per 871120 Meeting.Addl Info on ATWS Mods Also Encl
| ML20236U632 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 11/25/1987 |
| From: | Kingsley O SYSTEM ENERGY RESOURCES, INC. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML20236U634 | List: |
| References | |
| AECM-87-0209, AECM-87-209, NUDOCS 8712030222 | |
| Download: ML20236U632 (12) | |
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SVETEM ENERGY RESOLJRCES, INC.
Ouvtp D KtE1H JR vcemon November 25, 1987 nmn cwowns U.S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Document Control Desk Gentlemen:
SUBJECT:
Grand Gulf Nuclear Station Unit 1 Docket No. 50-416 License No. NPF-29 10CFR50.62 ATWS Modifications PCOLs 87/06 and 87/09 AECM-87/0209 In letters dated August 13, 1987 and October 23, 1987, System Energy Resources, Inc. (SERI) submitted proposed amendments to implement the 10CFR50.62 ATWS modifications and additional information supporting the proposed amendments.
On November 20, 1987 SERI met with members of the NRC Staff to discuss the design of proposed modifications to the Standby Liquid Control System (SLCS) discharge piping which are being implemented as part of the ATWS modifications.
Thet meeting resulted in SERI proposing an alternate design. The alternate design and associated changes to the technical specifications are discussed in Attachment II to this letter. The proposed changes to the technical specifications associated with the SLCS discharge piping modification supercede those requested in the August 13, 1987 letter. Attachment II also provides a compilation of the no significant hazards considerations for the previous submittals as well as the alternate design proposal for the SLCS discharge piping.
Attachment III provides additional information concerning the ATWS modifications requested by the Grand Gulf NRC Project Manager. Attachment IV provides marked up pages to the GGNS Unit 1 Technical Specifications showing proposed changes associated with the SLCS modifications and Bases pages revised to provide additional information on the ATWS Reactor Pump Trip instrui_ntation.
In accordance with the provisions of 10 CFR 50.4 and 50.30, the signed original of the requested amendment is enclosed and the appropriate copies will be distributed. The additional information and proposed changes have been reviewed and accepted by both the Plant Safety Review Committee and the Safety Review Committee.
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AECM-87/0209 Page 2 Based on the guidelines presented in 10 CFR 50.92, SERI has concluded that this proposed amendment involves no significant hazards considerations.
I Since the application fee of $150 in accordance with the requirements of 10 CFR 170.21, was submitted in the original submittal and supplement, SERI has determined that no application fee is required for the submittal.
SERI requests a response to this letter by January 6, 1988 to support second refueling outage schedule requirements.
Yoes,huly, fs %
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ODK:rg f Attachments: 1. Affirmation per 10 CFR 50.30
- 2. SLCS Discharge Piping Modifications and Combined No Significant Hazards Consideration
- 3. Response to Request for Additional Information
- 4. Affected Pages of the GGNS Unit 1 Technical Specifications cc: Mr. T. H. Cloninger (w/a)
Mr. R. B. McGehee (w/a)
Mr. N. S. Reynolds (w/a)
Mr. H. L. Thomas (w/o)
Mr. R. C. Butcher (w/a)
Dr. J. Nelson Grace, Regional Administrator (w/a)
U. S. Nuclear Regulatory Commission Region II 101 Marietta St., N. W., Suite 2900 Atlanta, Georgia 30323 Mr. L. L. Kintner, Project Manager (w/a)
Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, Maryland 20814 Dr. Alton B. Cobb (w/a)
State Health Officer State Department of Health Box 1700 Jackson, Mississippi 39205 J12AECM87102901 - 2
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BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION LICENSE NO. NPF-29 DOCKET NO. 50-416 IN THE MATTER OF MISSISSIPPI POWER & LIGHT COMPANY and !
SYSTEM ENERGY RESOURCES, INC.
and SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION AFFIRMATION I, O. D. Kingsley, Jr., being duly sworn, stated that I am Vice ,
President, Nuclear Operations of System Energy Resources, Inc.; that on behalf '
of System Energy Resources, Inc., and South Mississippi Electric Power Association I am authorized by System Energy Resources, Inc. to sign and file with the Nuclear Regulatory Commission, this application for amendment of the t Operating License of the Grand Gulf Nuclear Station; that I signed this ;
application as Vice President, Nuclear Operations of System Energy Resources, !
Inc.; and that the statements made and the matters set fo therein are true and correct to the best of my knowledge, information an ief.
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i "0. D/ Kiritjsl , h STATE OF MISSISSIPPI COUNTY OF HINDS SUBSCRIBED AND SWORN T0 befor me, a Notar Public in and for the County and State above named, this d'I ulay of 32>em3,u , , 1987.
(SEAL)
N j ). Y l A.)
Notary Public' I
My commission expires: l 14 Dmmhsb9 bptc.; /.u,1. 5, D]l J12AECM87102901 - 4
l Attachment TI to AECM-87/0209 i
l 10CFR50.62 ATWS MODIFICATIONS A. PROPOSED MODIFICATIONS
- 1. Previously Proposed Modifications The following changes to GGNS were proposed and justified in letters to i the NRC dated August 13, 1987 and October 23, 1987.
- a. RPT modifications including:
- 1) Adding the redundant trip feature of the "Monticello" RPT l design.
- 2) The trip logic will be 2-out-of-2 for RPV pressure or level and will be energized to trip.
- 3) Revising the RPV pressure high trip setpoint to bound less severe transients initiated at less than rated power.
- b. Installation of the ARI system which includes:
- 1) Three parallel vent paths from the scram pilot air header consisting of two valves per vent path.
- 2) Utilizing the same trip system as RPT with the same trip logic.
- c. SLCS modifications including:
l' .tilizing simultaneous operation of both existing SLCS pumps to achieve a minimum injection rate of 82.4 gpm.
- 2) Increasing the sodium pentaborate concentration to greater than or equal to 13.6% weight.
- 3) Increasing surveillance requirement for demonstrating minimum pu.np flow from 1220 to 1300 psig pump discharge pressure.
- 4) Restricting SLCS storage tank temperature to no greater than 130 F. .
- 5) Increasing SLCS design pressure to 1500 psig for portions of the system.
- 2. Newly Proposed SLCS Discharge Piping Modifications Resulting From the NRC/SERI Meeting of November 20, 1987 l
By letter dated August 13, 1987 SERI submitted proposed changes to the SLCS discharge piping which would reroute the SLCS discharge piping to the HPCS injection piping (AECM-87/0152). In response to NRC staff j l
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Attachment II to AECM-87/0209 concerns regarding the location of the outermost check valve and the appropriate code classification of the proposed piping, SERI met with the NRC Staff on November 20, 1987. As a result of that meeting, SERI is proposing an alternate design for rerouting that I piping. The following discussion and proposed changes to the technical specifications supercede those proposed changes to Technical Specification Table 3.6.4-1 (pages 3/4 6-42 and 3/4 6-45) provided as part of the August 13, 1987 request.
- a. SUBJECT j j
- 1) Proposed Modifications to the SLCS Discharge Piping !
- 2) Affected Technical Specifications:
Containment and Drywell Isolation Valves, Table 3.6.4 page 3/4 6-42
- b. DISCUSSION As part of the proposed SLCS discharge piping modification, the low point drain will be moved to a point on the injection line interior to the drywell between the existing valve C41 F007 and the drywell interior wall. One of the valves isolating this drain line will be designated as an inboard drywell isolation valve for penetration number 328 (see Figure 1). This valve will be added to Table 3.6.4-1, page 3/4 6-42.
- c. JUSTIFICATION The SLCS discharge piping will be rerouted to the High Pressure Core Spray (HPCS) injection piping to provide more effective boron mixing as described in AECM-85/0322 dated October 14, 1985. The modification to the SLCS discharge piping will require the relocation of the low point drain. The drain line will be located interior to the drywell between the existing C41 F007 valve and the drywell interior wall. This line will be isolated from the SLCS discharge line by two normally closed valves. One of the two valves (C41-F218) will serve as an inboard drywell isolation valve for drywell I penetration 328. The two drywell isolation valves (C41-F006 and F007) for the SLCS discharge line will remain in their current position (see Figure 1).
The rerouted piping will conform to the requirements for Class 1 {
components of ASME Boiler and Pressure Vessel Code,Section III '
and Seismic Category I.
I The proposed Class 1 boundary will extend up to and include the j C41 F006 valve. The original Class 1 boundary extended up to and {
included the C41 F004 explosive valves (A and B). The proposed configuration complies with the requirements of 10CFR50.55a =
(c)(2)(ii) in that the Class 2 components upstream of the C41 1 F006 valve are isolated from the reactor coolant system by the l J12AECM87102901 - 6
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c Attachment II to AECM-87/0209 normally closed ASME Class 1 isolation valves (F006 and F007).
Because the valves are normally closed, isolation time is not a consideration. Should a failure of one of the normally closed isolation valves occur, the other valve will serve to isolate the reactor coolant pressure boundary.
B. N0 SIGNIFICANT HAZARDS CONSIDERATION The following represents SERI's. evaluation of the ATWS modifications
. submitted by letters dated August 13, 1987 (AECM-87/0152) and October 23, 1987(AECM-87/0190) and.the proposed SLCS discharge piping modifications resulting from the NRC/SERI meeting.of November 20, 1987.
- 1. The proposed amendment does not involve a significant increase in the l
probability or consequences of an accident previously evaluated.
- a. The probability of an ATWS event occurring does not increase due to these changes since they are of a mitigative nature and do not affect the ATWS event precursors. These changes do not involve a significant increase in the consequences of an accident previously l . evaluated. The ATWS-RPT system provides a fully redundant trip of the recirculation pump motors including the low frequency motor generator set so that the pumps coast down to zero speed. .This trip function reduces core flow creating steam voids in the core, thereby decreasing power generation and limiting any power or pressure excursions.
- b. The ARI system'that will be installed during the second refueling outage is described in letters to the NRC dated October 14, 1985, t April 3, 1987 and August 13, 1987. .The ARI system uses the same setpoints and trip channels (transmitters and trip units) as the i RPT system. Both the ARI and RPT' systems are designed to perform a !
mitigative function during an ATWS event. The probability of an j ATWS event occurring does not increase due to the commonality of :
the ARI and RPT trip channel since these systems perform a !
mitigative function and do not affect the ATWS event precursors.
Both the ATWS topical report NEDE-31096-P and the NRC Staff's safety evaluation associated with the report endorse the use of j existing RPT instrumentation where possible for the ARI system. i The consequences of an ATWS event are not significantly increased J by connecting the ARI valves to the RPT trip circuits. The ARI !
system has been designed to minimize the possibility of an inadvertent trip action by use of series vent valves, energize to trip solenoids and required two out of two logic,
- c. The operation of.the two SLCS pumps in conjunction with the I increased sodium pentaborate weight percent concentration merely I provides a backup to other safety-related systems in accordance i with the requirements of 10CFR50.62. ,
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L. Attachment !!
E to AECM-87/0209 The rerouted SLCS discharge piping inside the drywell-will provide' )
more effective boron mixing in the reactor vessel. The piping will '
~be constructed to ASME Section III Class 1 and Seismic Category I requirements and therefore will not increase the consequences of a LOCA or Seismic event. The relocated low point drain inside the -
drywell will be isolated from the containment by two normally -
closed isolation valves and therefore will not create a new leakage ~ '
path to the containment. Movement of the ASME Class 1 boundary from the F004 (A and B) valves to the F006 will not increase the probability'or consequences of an accident because the reactor ;
coolant' system will still be isolated from other Class 2-components by two normally closed ASME Class 1 valves.
The proposed increase in the SLCS pump relief valve setpoint is within ASME Code allowables and will not increase the probability of piping failure. The~ consequences of previously evaluated accidents ~ remain unchanged since the relief valve will still perform its intended function of preventing the SLCS discharge piping from exceeding its design pressure.
Based'upon the above, the probability or consequences of an accident will not be significant1y' increased.
- 2. The proposed amendment does not create the possibility of a new or different kind of accident from any previously analyzed.
- a. The ATWS changes proposed and the accompanying plant modifications only serve as backups to already existing safety-related systems. '
The proposed changes.will ensure that the ATWS-RPT and SLCS systems are maintained such that they are capable of fulfilling the ~
operability requirements of 10CFR50.62.
- b. The ARI system is being installed to help mitigate the consequences of an ATWS event. As discussed above, the ARI system is designed to minimize the possibility of inadvertent trips. If the ARI ,
valves do open inadvertently, a reactor scram will occur resulting in plant shutdown. This event has been analyzed and is not a new !
or different kind of accident. The ARI system is designed with ;
redundant trip systems in order to help ensura system function when ;
required. The system design features do not create the possibility of a new or different kind of accident from any previously analyzed. I
- c. The rerouting of the SLCS discharge piping and the relocation of the low point drain will not create a new or different path for drywell bypass leakage because adequate isolation is provided at the drywell penetration. Therefore, there is no possibility of a new or different kind of accident from any previously analyzed.
The possibility of a new or different kind of accident is not created by the movement of the Class 1 boundary from the F004 (A & B) valves to the F006 valve because the Class 2 components are. isolated from the reactor coolant system by two normally closed ASME Class 1 valves.
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Attachment II to AECM-87/0209 The proposed increase in the'SLCS pump ~ relief valve setpoint does-not adversely affect the safety function performed by SLCS or_the operability of the SLCS system. .The increase in SLCS. design pressure does not affect any accident precursors and cannot create the possibility of a new or different kind of accident from any previously evaluated.
- 3. The proposed amendment does not involve a significant reduction'in a margin of safety,
- a. The proposed revisions are in accordance with the' requirements of 10CFR50.62 and provide additional assurance that systems exist that are capable of safely shutting down the reactor should an ATWS event occur. The ATWS-RPT and SLCS systems operability-do not decrease. the margin of safety since they serve as backups to other safety-related systems.
-The proposed change in the ARI/RPT reactor pressure-high trip setpoint does not involve a significant reduction in a margin.of safety. As stated in AECM-87/0152 dated August 13, 1987, the basis for the new ATWS reactor pressure high setpoint is to ensure that the relief valve capacity below the Nominal Trip Setpoint (NTSP) is.
less than 15% Nuclear Boiler Rated (NBR) when operating at less than rated power. This steam flow limitation allows 10 minutes for SLCSinitiatfonwithoutexceedingthesuppressionpooltemperature ;
limit of 185 F. '
The design basis event for this setpoint is Main Steam Isolation Valve (MSIV) closure with a subsequent failure to scram. The safety limit is' peak Reactor Pressure Vessel (RPV) bottom pressure of 1375'psig for active components. The current setpoint was established such that this safety limit would not be exceeded.
Lowering the setpoint will increase the safety margin by initiating ARI and RPT earlier in the event, thereby providing additional margin to the safety limit.
The normal Reactor Protection System (RPS) scram path setpoint is 1064.7 psig based on an analytic limit of 1095 psig. Therefore, since the ATWS RPV pressure high NTSP is less than or equal to 1095 psig, it is possible (but highly unlikely) under extremes of ,
instrument accuracy and drift to initiate the ATWS ARI and RPT functions prior to t h normal RPS scram. However, since the ARI and normal scram paths perform the same function (i.e., insert all control rods) this would not be of any safety significance.
Additionally, tripping the recirculation pumps via the ATWS RPT prior to the normal RPS scram would only serve to minimize the pressure rise in the RPV in the first few seconds of the event and to reduce the reactor thermal power which in turn reduces steam flow which might need to be discharged to the suppression pool.
Therefore, initiation of the ARI and RPT functions prior to the normal RPS scram is of no safety significance and does not involve a significant reduction in a margin of safety.
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' Attachment It'
.to AECM-87/0209 l b. .The connection of the ARI valves'to the RPT trip system will not l
adversely impact the RPT function. The ARI system serves as a i backup to the normal reactor scram system by depressurizing the scram pilot air header. 'Therefore, installation of the ARI system will increase the margin of safety during an ATWS event,
- c. .The current amount of suppression pool bypass. leakage.that would occur during a LOCA is not affected by the modified SLCS piping
! design. The modified piping will.be constructed.to the Class'1.
standards of ASME Section III and Seismic Category I requirements and will be provided with adequate isolation capability.
The margin of safety is not decreased by.the movement of the Class 1 boundary from the F004 (A & B)' valves to the'F006 valve because the safety-function of isolating the reactor coolant system from Class 2 portions of the system is still maintained by the-ASME' Class 1 violation valves (F006 and F007).
, The increase in the SLCS pump relief valve setpoint is in accordance with the ASME Code. The proposed change will provide additional assurance that the SLCS will be able to deliver its rated flowrate to the reactor without. the possibility of some flow being diverted through the relief valve. The proposed change will allow an upgrade in SLCS' design pressure to 1500 psig which will increase the present margin to the relief valve setpoint by 100 psi. As such,'the proposed change will increase the margin of safety.
Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
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Attachment II to AECM-87/0209 l
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Attachment 111 to AECM-87/0209 REQUESTS FOR ADDITIONAL INFORMATION j The following requests were received in an October 28, 1987 telephone call between the Grand Gulf NRC Project Manager and the SERI staff.
REQUEST 1 By a [[letter::AECM-87-0190, Application for Amend to License NPF-29,changing Standby Liquid Control Sys (SLCS) Pump Relief Valve Setpoint from Present within 3% of 1,400 Psig to Actuate within 3% of SLCS Design Pressure.Fee Paid|letter dated October 23, 1987]], SERI described simultaneous testing of both SLCS pumps. Will this testing verify the conservatively calculated ;
maximum system pressure drop of 150 psi from the pumps to the reactor vessel?
RESPONSE
As part of the post installation SLCS testing during the second refueling outage, SERI will verify the calculated 150 psi pressure drop from the pumps to the reactor vessel.
REQUEST 2 Provide changes to the Bases for RPT to indicate that the system complies with the requirements of 10CFR50.62 ATWS Rule and that the RPT Actuation Instrumentation also actuates the ARI valves.
RESPONSE
The proposed changes to Bases page B 3/4 3-2 are attached.
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- - Attachment Ill to AECM-87/0209.-
Insert Number 1 to Page B,3/4 3-2 LA L _
Theanticipatedtransientwithoutscramrecirculationpumptrip(ATWS-RPT) y' system provides a means of limiting the consequences of the unlikely occurrence ,
'of a failure to. scram during an anticipated transient. :The response of the 4-plant to this postulated event has been evaluated in General Electric: company report NEDC-32408 dated March,.1987. The results of the analysis show that the Grand Gulf ATWS-RPT design provides adequate protection for these events in:
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.which the normal scram paths fail. 1e 4,.
Tiie ATWS-RPT provides fully redundant trip of the recirculation. pump [
motors so that the pumps coast down to-zero speed. This trip function reduces f core flow creating steam voids in the core, thereby decreasing power generation s-and limiting any power or pressure excursions ~ The Grand Gulf ATWS-RPT design provides compliance with the requirements of the NRC ATWS Rule 10CFR50.62.
The ATWS-RPT and Alternate Rod Insertion (ARI) system use common setpoints and trip channels (transmitters and trip systems). Therefore, the ARI trip function and the RPT.. trip. function will be initiated simultaneously. The instrumentation setpoints for the RPV pressure and water level trip channels are established such-that the normal scram paths for these variables would already be. initiated.
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