ML20005B898

Forwards Proposed FSAR Changes & Responses to NRC Requests for Addl Info.Proposed Changes Will Be Incorporated in Forthcoming Amend to FSAR

ML20005B898
Person / Time
Site:	Grand Gulf
Issue date:	09/10/1981
From:	Dale L MISSISSIPPI POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
AECM-81-351, NUDOCS 8109160091
Download: ML20005B898 (10)

Text

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MISSISSIPPl POWER & LIGHT COMPANY

] Helping Build Mississippi EdhhibhMidE P. O. B O X 1640. J A C K S O N. MIS SIS SIP PI 3 9 2 05 September 10, 1981 g NUCLEAR PRoOUCTloN DEPARTMENT g >

U.S. Nuclear Regulatory Commission f 'g$ H Office of Nuclear Reactor Regulation hk Washington, D.C. 205!5 3 15196F M' Attention: Mr. flarold R. Denton, Director ui .[#m t

Dear Mr. Denton:

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SUBJECT:

Grand Gulf Nuclear St Units 1 and 2 Docket Nes. 50-416 and 50-417 File 0260/0862/L-340.0 Transmittal of Proposed FSAR Changes and Respor.scs to NRC Q:testions AECM-81/351 In response to NRC formal and informal requests for additional information, Mississippi Power & Light Company is submitting the enclosed information.

This information, as noted in the attachments, represents changes to the Grand Gulf Nuclear Station Final Safety Analysis Report (FSAR).

These proposed FSAR changes will be incorporated into a forthcoming amendment to the FSAR. If you have any questions or require further information, please contact this office. ,

Yours truly, L. F. Dale Manager of Nuclear Services RFP/JGC/JDR:Im Attachments: 1. Contsinment Systems Branch, Surveillance Frequency for the Drywell Bypass Leakage Rate Test

2. NRC Question and Response 260.2

3. NRC Question and Response 21.55

4. Common Reference Level, TMI II.K.3.27

5. Failure in Vessel Level Lines Common to Control and Protective Systems cc: (See Next Page) l AE3F1 l

8109160091 810910 Member Middle South Utilities System PDR ADOCK 05000416 A PDR

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AECM-81/351 MIICISSIPPI POWER Q LIGHT COMPANY Page 2 i

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r r:: Mr. N. L. Stampley i Mr. G. B. Taylor 4

Mr. R. B. McGehee Mr. T. B. Conner

. Mr. Victor Stello, Jr. , Director 4 Office of Inspection & Enforcement

] U.S. Nucles:- Regulatory Commission

Washington, D.C. 20555 i

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2 Attachment 1 to AECM-81/351 CONCERN: Drywell Bypass Leakage Rate Test - Surveillance Frequency RESPONSE: The current GGNS Tech Specs (Proof and Review Copy, dated August 4 j 1981) reflects the following surveillance requirement:

4.6.2.2 The drywell bypass leakage rate test shall be conducted at least once per sch duled refueling outage at an initial dif ferential pressure of 3.0 psid and the A/ k shall be calculated from the measured leakage. One drywell airlock door shall remain open during the drywell leakage test such that each a drywell door is leak tested at least everv other drywell leakage rate test.

a. (If any drywell bypass leakage rate test fails to meet the specified limit, the test scitedule for subsequent tests shall be reviewed and approved by the Commission. If two consecutive tests fail to meet the limit, a test shall be performed at least once every 9 months until two consecutive tests meet the limit, at which time the abcve test schedule may be resumed.)

b. Airlocks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.2.3.

c. The provisions of Specification 4.0.2 are not applicable.

The above requirement will be incorporated into the Grand Gulf Technical Specifications (GGTS).

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Attachment 2 to AECM-81/351 GG FSAR 260.2 The following items, added to Table 3.2-1 of the FSAR by Amendment 49, should be subject to the pertinent requirements of the FSAR operational quality assurance program. Revise Table 3.2-1 to provide a commitment that these items are subject to such requirements or justify not doing so.

XLI.1.b Fuel oil Jrip return line piping and valves XLI.1.g Fuel oil drip return pump XLI.1.h Fuel. oil drip return tank l XLI.4.a Combustion air intake and exhaust suosystem piping and valves XLI.4.b Combustion air intake and exhaust subsystem silencers and filters

RESPONSE

The rbove listed items are nonsafety-related. However, the appropriate portions of the operational QA program will be applied to these items.

Appendix B of the Q-List and FSAR Table 3.2-1 will be revised to reflect this response.

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Attachment 3 to AECM-81/351 GG FSAR 021.55 Table 6.2.46 lists the positions of the containm+- t isolation valves for post-accident conditicas. Many of the valves th+t are required to be open or closed during an accident fail in the "as is" position upon loss of power. It is our position that all containment isolation valves fail in the position of greater safety in the event of power faflure to the valve operator during an accident. Therefore, justification should be provided to demonstrate compliance with the above position or the appropriate plant modifications should be made.

RESPONSE

As denoted an Table 6.2-44 all air-operated valves are decigned to fail in the position of greatest safety, i.e., closed, during a power failure. Electric motor-operated valves will remain in the last position upon loss of power. Provided below is the justification for meeting the position that containment motor operseed isolation valves fail in the state of greater safety.

Fssential systems are designed with radundant trains to accomplish the system's function. Thus, in these systems, the failure of an isolation valve in an "as is" position does not prohibit the proper functioning of l the essential aystem's redundant counterpart. In addition, such systems are powered from independent Class IE power supplies.

Non-essential, beneficial, and essential systems (except for HPCS, LPCS, and RHR LPCI injection and suction) are designed with redundant isolation valves powered from independent Class 1E power supplies. This design ensures that either of the two isolation valves can perform the isolation function regardless of the state of operation or failure of the other. Therefore, a single failure or malfunction of a valve operator will not result in loss of the isolation function. In the design of HPCS, LPCS, and RHR LPCI (single train) injection and suction, motor-operated valve (s) for each system are powered from a single divisional power source. Ecch injection system is provided with a testable check valve located inboard of primary containment. If a single divisional power failure were to occur during an accident, the check valve would prnvide the backup isolation function. Beneficial systems, as discussed in Section 18.1.26 and Table 18.1-1, are not required for mitigating the consequences of an accident and their i failure, due to a loss of power, would not worsen an accident situation.

It should also be noted that in the event of n loss of offsite power, onsite power is 4 vailable irr use in a matter of seconds.

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. . Attachm:nt 4 to AECM-81/351 Branch: Human Factors Engineering Concern: Provide common reference level for vessel level instrumentation (II.K.3.27).

Response: In order to satisfy the requirements of a common reactor vessel level reference point for all reactor vessel level loops, Mississippi Power & Light will make the necessary modifications to reference the Fuel Zone instrument from the bottom of the reactor vessel steam dryer skirt (referenced to instrument zero, 533 vessel inches).

New scales will be obtained for the Fuel Zone instrument indicator and recorder, and both will reflect a range from

-117 to -317 inches. Top of Active Fuel (T0AF) will be marked on the scale at -167 inches. These changes will be implemented prior to fuel loading.

The above information will be Ance:porated into a forthcoming FSAR amendment.

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. Attachment 5 to AECM-81/351

BRANCH:

Instrumentauion and Control Systems i

CONCERN:

Failure in Vessel Level Senaing Lines Common to Control and Protective Systems Operating reactor experience indicates that a number of failures have occurred in BWR reactc' vessel level reference sensing lines and that in most cases the failures have reeulted in erroneously high reactor vessel level indication. For BWR's, common reference sensing lines are used for feedwater control and as the basis for establishing vessel level channel trips for one or more of the protective functions (reactor scram, MSIV closure, RCIC, LPCI, ADS or HPCS initiation). Failures in such sensing lines, may cause reduction in feedwater flow and consequential delay in trip within the related protective channel.

If an additional failure, perhaps of electrical nature, is assumed in a protective channel not dependent on the failed sensing line, protective action may not occur or may be delayed long enough to result in unacceptable consequences. This depends on the logic for combining channel trips to achieve actions.

It is our po ition that those reference lines common to the feed-water control function and to any of the protective functions for loss of feedwater events be identified, and that the consequences of failures in such reference lines concurrent with the worst additional single failure in the protective systems (reactor scram, MSIV closure, ADS, RCIC, HPCS/HPCI, LPCI, etc.) or their initiation circuits be analyzed.

RESPONSE

! A postulated break in an instrument line plus an additional failure is beyond the design basis for this plant; however, an assessment of plant response to this event is provided below.

The instrument refe.sance lines common to feedwater control and to protective system sensors have been identified for this plant. An analysis was performed to determine the consequences of failures in such reference lines concurrent with additional single failures in protective channels not dependent on t he f ailed sensing line. The Sequence of Events are denoted in Table 1.

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I In the highly unlikely scenario, the most severe reference line was asrumed to fail such that all attached level instruments erroneously indicated high 12vels. Then, additional worst-case single failures were postulated for each c' the remaining 3 electrical protective system divisions. A single division power supply loss was considered worst-case for ECCS, but this is independent from other single failures which could effect RPS or MSIV closure, etc. (i.e., a power bus failure in RPS would fall

" safe" causing a trip uf that clannel). The worst postulated failure path, from the various combinations, was found to be failure of Division 1 instrument reference line combined with an RF3 scram circuit failure in Division 3. Worst-case was also assumed for the feedwater controller in tint the manual selection switch is on Division 1 instrument line and the operator does not take the option to switch control to Division 2, as he would normally be expected te do when he sees the level mismatch between the indicators. The feedwater controller would then respond to the high level error signal and reduce the fetdwater flow. Siace the worst single random failure is postulated in RPS Division 3 logic, scram would not occur when the real vessel level drops through the normal scram setpoint at level 3, unless the Division 1 instrument line break is sufficiently large to cause a level 8 scram signal on channel "A", resulting in a less severe event. That is, we do not get level 3 scran because we have taken two complimentary failures in the scram equation (A + C) . (B + D), namely, loss of both "A" and "C" for low level trip.

As the postulated water level drop passes through level 2, RCIC and HPCS will start and both recirculation pumps will trip. The watcr level will continue to drop, but now at a slower rate due to inventory assistance provided by HPCS and RCIC (See Table 1

" Sequence of Events"). Considering no benefit of scrams from level 8, high drywell pressure or manual action (which should normally have occurred by now); snd assuming the operator still has not switched feedwater to the alternate control (which he would be expected to do), the water level will ultimately reach a minimum level above level 1, still well above the Top of the Active roel.

No scram would occur; however, an equilibrium state would be established at about 15% Nuclear Boiler Rated power. No fuel failure would occur. The core remains covered at all times. Low pressure systems are also available, but are not necessary because HPCS and RCIC have more than enough capacity to assure adequate water make-up and inventory control for this postulated line break.

Even if the minimum water level had fallen to level 1, MSIV closure and the associated valve position scram would have been initiated.

Neutron power wou1J be terminated by scram in 1 to 2 4econds. The '

water level would drop below level 1 and start to rise in 1 to 2 seconds. The minimum water level would still remain well above the Top of the Active Fuel. No fuel failure would occur for this event either.

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• O The Sequence of Events as denoted in Table 1 shows that-the reactor system can withstand any reactor-vessel level reference.line break coupled with an additional worst single failure in a protective channel not dependen* on the failed sensing line without compromising safet; This is assured by the following evaluations:

1. No part of the aci ve fuel is uncovered at any time. This 3

assures no fuel damage and no degradation of the critical power ra lCPR), or radioactivity release.

2. Both the vessel and the containment remain structurally

sound throughout the postulated event. This provides

, secondary assurance that no radioactivity can be released to the public.

] 3. The scenario postulated is a highly unlikely event (instru-j ment line breakage with coincident random scram channel failure) and compounds it with worst-case conditions through-I out the event. Though no credit is taken in this scenario, l it is highly probable one or more of the following actions t would occur in a real-life situation:

4 A. The false-high level signal would half-scram at level 8 allowing normal scram at real Level 3, thereby signifi-cantly ensing the accident.

i B. The operator would recover feedwrter level immediately by switching the controller to the alternate instrument line (Division 2 in this case).

C. The operator would manually scram the plant.

D. A scram will eventually result from high drywell pressure ,

provided the reference leg line break is sufficiently large.

It is concluded from this assessment of a break in a vessel level sensing line common to control and protective systems plus an 1 additional worst single failure in a protective channel not.

dependent on the fciled sensing line that the resulting accident is less sev7re and bounded by the DBAs already analyzed in Chapter 15 of the F8AR. ,

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Table 1 SEQUENCE OF EVENTS Tine Events Tsr2) 0 One of the water level reference legs break (assume feedwater control relies on this instrument line).

Feedvat.er starts to decrease due to a false hig*: water level reading produced by the faile'd

, instrument line.

3.0 Actual water level drops to L-4.

5 .' O Feedwater flow decrease to zero.

6.9 Actual level drops to L-3. No low level scram or recirculation pump shift to slow speed due to the failure of the reference leg and an RPS channel.

11.9 Water level drops to L-2, trips the recircu-lation pumps and alse initiates RCIC and HPCS.

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33. HPCS and RCIC flows stort to enter vessel.

71. Water level reaches minimum and begins to rise. The minimum level is above the L-1 setpoint.

200 A new equilibrium state is established at 15% NBR power.

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