ML20050B684
| ML20050B684 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 04/05/1982 |
| From: | Dale L MISSISSIPPI POWER & LIGHT CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-0831, RTR-NUREG-737, RTR-NUREG-831, TASK-2.K.3.17, TASK-2.K.3.18, TASK-2.K.3.21, TASK-TM AECM-82-131, NUDOCS 8204070197 | |
| Download: ML20050B684 (9) | |
Text
MISSISSIPPI POWER & LIGHT COMPANY Helping Build Mississippi adMMMddB P. O. B O X 16 4 0. J A C K S O N, MISSISSIPPI 39205 a
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NUCLE AR PHoOUCTioN DEPARTMENT
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U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation A.V kg*
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g Attention: fir. Harold R. Denton, Director
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Dear ffr. Denton:
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SUBJECT:
Grand Gulf Nuclear Stati'on Units 1 and 2 Docket Nos. 50-416 and 50-417 File 0260/L-334.0/L-860.0 Containment Purge, ADS Logic tiodification AECt!-82/131 Attached are responses or clarifications pertaining te several issues discussed in the Grand Gulf Nuclear Station Safety Evaluation Report (SER), NUREG-0831. The attachments addrets the following:
0utstanding Issues 1.9(8)
Containment Isolation 1.9(9)
Containment Purge Confirmatory Issues 1.10(31) ADS Logic Some revisions to the Grand Gulf Final Safety Analysis Report (FSAR) are discussed in the attachments. The content of the last FSAR amendment prior to the projected fuel load of April 23, 1982, has been finali7ad. Thus, the incorporaLion of any proposed FSAR revisions, as dicuss
.n the attachments, will be made pending the receipt of further guidance requested informally from the NRC in regard to post-operating license FSAR amendments.
If additional information is required, please advise.
Yours
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L. F. Dale
,y Planager of Nuclear Services JGC/JDR:Im
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Attachments 8204070197 820405
{DRADOCK 05000416 PDR Member Middle South Utilities System
AECM-82/131 MISSISSIPPI POWER Q LICHT COMPANY Pete 2 cc:
Mr. N. L. Stampley (w/a)
Mr. G. B. Taylor (w/a)
Mr. R. B. McGehee (w/a)
Mr. T. B. Conner (w/a)
Mr. Richard C. DeYoung, Director (w/a)
Office of Inspection & Enforcement U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Mr. J. P. O'Reilly, Regional Administrator Office of Inspection & Enforcement U.S. Nuclear Regulatory Commission Region II 101 Marietta St., N.W., Suite 3100 Atlanta, Georgia 30303 i
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Attachment I to AECM-82/131 BRANCH:
Equipment Qualification CONCERN:
The MP&L submittal on containment purge operation (MP&L letter AECM-82/78, dated March 15, 1982) discussed an isolation valve operability analysis conducted by the valve supplier to ensure the isolation valves can close against accident differential pressures. Clarify the use of drywell isolation valves during containment purge operations and discuss the impact of that use on the valve operability analysis conducted. Refer also to Grand Gulf SER Outstanding Issues 1.9(8) and 1.9(9).
RESPONSE
As discussed in Section 5.2 of Attachment 1 to the above referenced MP&L report, a peak accident differential pressure of 3 psi was assumed for the containment isolatici. valve operability analysis. This value is obtained from FSAR Figure 6.2-10, depicting the containment, drywell, and wetwell pressure response to the design basis main steam line break. The relatively low response profile for containment pressure is a characteristic of the Mark III containment which benefits from the energy absorption capacity of the drywell-suppression pool configuration and presents a less severe challenge to containment purge isolation valves.
The isolation signal, based on high drywell pressure, is initiated within the first second of the accident.
Allowing five seconds for valve closure, it can be seen from Figure 6.2-10 that containment pressure at time five seconds is slightly less than 3 psi. This value is considered conservative for the valve operability analysis because it ignores lower containment pressures at times prior to five seconds.
The above referenced operability analysis assumes that drywell isolation valves, associated with drywell purging and venting are closed. See FSAR Figure 9.4-13 for drywell purge supply valves, M41-F013 and F015 (locators 1-D and 2-D), penetration No. 345, and drywell purge exhaust valves, M41-F016 and F017 (locators 8-G and 8-H),
penetration No. 347. These valves and associated penetrations are 20" in diameter.
The MP&L position on drywell purge operation, i.e., the supply and exhaust of air to and from the drywell, is currently under re-evaluation per. ding the completion of valve operability analyses for drywell purge isolation valves. Drywell purge operation is restricted to hot shutdown, cold shutdown, refueling, and startup operations as discussed in FSAR 9.4.8.2.2.
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The drywell purge system can also be used to vent the drywell in the above mentioned plant conditions. Venting involves the momentary " cracking" of the drywell purge exhaust valves (M41-F016 and F017).
It is anticipated that occasional venting of the drywell may also be required during power operations to relieve pressure buildup due to instrument air and steam leakage in the d rywell.
To assess the impact of drywell venting on the containment isolation valve operability analysis, an additional calculation was performed. This evaluation assumed that the drywell was being vented to containment, coincidental with the purging of the containment and a-design basis LOCA in the drywell.
The major assumptions and conservatisms incorporated into this calculation are described below:
1.
The containment is being purged through the 6" lines (low volume purge).
Whenever primary containment is required to be set, the containment will not be purged through the 20" lines'(high volume purge) shile venting the drywell' to containment.
2.
The drywell is being vented through one 20" line (through the drywell purge exhaust).
3.
Drywell valves are assumed to close five seconds-after the event begins (includes one second for conservatism).
4.
Drywell air was assumed to be released through the 20" vent line for a total of five seconds. A constant flow rate through the open line was calculated' based on the maximum drywell pressure of 23 psig (
Reference:
FSAR Figure 6.2-10).
This is a conservative approach since lower flow rates due to lower drywell pressures and the partial closing of the drywell isolation valves at times prior to five seconds are ignored.
5.
Worst case pressures, predicted a-a result of a postulated design basis LOCA occurring while the plant is in power operation, were used.
The results of this evaluation indicated that a containment pressure increase of 0.6 psi can be expected.
This increase, combined with the 3 psig predicted from FSAR Figure 6.2-10, presents a maximum differential pressure of 3.6 psid across the containment isolation valves. The isolation valve supplier (Henry Pratt) was contacted to determine the impact of this additional 0.6 psid MP&L was informally advised by the valve supplier q
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that such an increase in differential pressure would not alter the conclusions of the valve operability analysis conducted for 3 psid (as discussed in Section 5.2 of Attachment I to AECM-82/28),_i.e., the valves remain operable under the application of the 3.6 psid and function as required.
On the basis of this evaluation and that information presented in the MP&L letter, AECM-82/28, dated March 15, 1982, it is MP&L's position that adequate justification has been provided to assure containment isolation valve operability while purging the containment. This discussion provided here also demonstrates that the momentary venting of the drywell during power operations presents no adverse impact to'the containment isolation valve operability analysis.
MP&L' has taken steps to obtain from the valve supplier an operability analysis of drywell purge isolation valves under LOCA conditions. The results of this study are expected in July, 1982, and will be reported to the NRC following MP&L review and evaluation. Appropriate FSAR revisions will be provided upon resolution of these issues with the NRC.
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. to AECM-82/131 BRANCH:
Equipment Qualification CONCERN:
-MP&L's recent submittal on the containment purge issue stated that surveillance testing included verification of isolation valve _ closure time.
Clarify the method used to verify this parameter and confirm _the testing criteria and frequency employed.. Refer also to Grand Gulf SER Outstanding Items _1.9(8) and 1.9(9).
RESPONSE
As. stated in Section 5.1 of Attachment I to AECM-82/28, dated March 15, 1982, MP&L's surveillance program includes periodic testing to verify proper valve closure time.
_ Table 3.6.4-1 of the Plant Technical Specifications
-identifies all containment-or drywell isolation valves involved in this issue. This table specifies that the maximum isolation time allowed during-testing for-the subject valves is four seconds.. This is consistent with FSAR Table 6.2-44;(Amendment 54, 3/82).
As noted in Section 4.2 of Attachment I to the above referenced MP&L letter, for analysis purposes the valve isolation times were conservatively assumed to be five seconds.
The periodicity for valve closure. time testing is in accordance with Plant Technical Specification 4.0.5.
This specification is implemented by_ the surveillance procedures of the Grand Gulf Surveillance Program.
M3Il
, to AECM-82/131 i
- BRANCH:
- Reactor Systems CONCERN:
Modifications of Automatic.Depressurization Systems Logic - Feasibility for Increased Diversity for Some Event Sequences. NUREG-0737, Item II.K.3.18.
Refer also to Grand Gulf SER Confirmatory Issue 1.10(31).
RESPONSE
As indicated in FSAR-subsection 18.1.30.6, Mississippi Power & Lig' t, in. participation with the BWR Owners' Group, committed.to providing a plant specific analysis by fuel load and a design modification by the first refuelding outage te address this concern. That modification consisted of the addition:of -a bypass of the drywell pressure trip if the reactor water level remains below the low pressure ECCS initiation setpoint for a sustained period.
a However, because of the relationship of this~ concern with current evaluations underway of-the Anticipated Transient Without Scram (ATWS) event, it is necessary that the:
i above MP&L commitment be withdrawn. The withdrawal of this commitment was discussed in a telephone conservation between Tim Collins of Reactor Systems Branch and Guy Cesare of MP&L,' held March 23,.1982, and found acceptable to the NRC. The overalll intent is to arrive at a position responsive to the' concerns of Item II.K.3.18 and consistent with design and procedural modifications necessary to mitigate the consequences of the ATWS event.
MP&L will continue to participate with the BWR Owners' Group and the NRC to develop a consistent and acceptable approach to these issues.
Proposed changes to FSAR subsection 18.1.30.6 are 1
attached.
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GG FSAR also be included in the feasibility study. Those changes which are shown to reduce relief valve challenges without compromising the performance of the relief valves or other systems should be implemented.
Challenges to the relief valves should be reduced substantially (by an order or magnitude).
RESPONSE
Mississippi Power & Light Company has participated in a BWR Owners' Group evaluation of possible ways to reduce challenges to safety / relief valves. The results of that evaluation were forwarded to the NRC in a letter from D. W. Waters to D. G. Eisenhut dated March 31, 1981.
It is Mississippi Power & Light's position that further modifications to the Grand Gulf Nuclear Station would not significantly reduce the frequency of SRV events.
18.1.30.5 Report on Outages of Emergency Core Cooling Systems Licensee Permit and Proposed Technical Specification Changes (II.K.3.17)
REQUIREMENT Several components of the emergency core cooling (ECC) systems are permitted by Technical Specifications to have substantial outage times (e.g., 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for one diesel generator; 14 days for the HPCI system).
In addition, there are no cumulative outage time limitations for ECC systems.
Licensees should submit a report detailing outage dates'and lengths of outages for all ECC systems for the last 5 years of operation. The report should also include the causes of the outages (i.e., controller failure, spurious isolation).
RESPONSE
Mississippi Power & Light Company commits to reporting a summary of emergency core cooling system outages annually.
18.1.30.6 Modification of Automatic Depressurization System Logic - Feasibility for Increased Diversity for Some Event Sequences (II.K.3.18)
REQUIREMENT The automatic depressurization system (ADS) actuation logic should be modified to eleminate the need for manual actuation to assure adequate core cooling. A feasibility and risk assessment study is required to determine the optimum approach. One possible scheme that should be considered is ADS actuation on low reactor vessel water level provided no high pressure coolant injection (HPCI) or high pressure coolant system (HPCS) flow exists and a low pressure emergency core cooling M3H2
GG FSAR (ECC) system is running.. This logic would complement, not replace,Lthe existing ADS actuation logic.
RESPONE.
MississfrpifPower &' Light Company has participated in a BWR Owndrs' Group study to modify ADS actuation without degrading other functionally >
related ECC systems. This study is. complete and the results have been provided to the NRC. Five options, including retaining the current design, were considered. The results indicated that the addition of a bypass of the high drywell pressure trip if the~ reactor water level remains below the low pr'_ssure ECCS initiation setpoint for.a sustained period or the elimination of the high drywell pressure trip are the preferred methods.
Additional study is underway to determine the appropriate design and procedural changes ' responsive to this concern and, at the same time,-
consistent with those changes required to mitigate the consequences of-an Anticipated Transient Without Scram (ATWS) event. MP&L is contin'uing to participate with the BWR Owners' Group and the NRC to determine a consistent and acceptable approach to these concerns.
18.1.30.7 Restart of Core Spray and' Low Pressure, Coolant Injection Systems (II.K.3.21)
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REQUIREMENT The core spray and_ low pressure, coolant injection (LPCI) system-flow may be stopped by the operator. These systems will-not restart automatically on loss of water level if an initiation signal is'still present. The core spray and LPCI system logic ~ should be modified so that these systems will restart, if required, to ensure adequate core-cooling.
Because this design' modification affects-sevdral core cooling modes under accident conditions, a preliminary designishould be submitted for staff review and approval prior to' making' the actual modification.
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