ML19249E525
| ML19249E525 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 05/23/1978 |
| From: | Israel S Office of Nuclear Reactor Regulation |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7910010632 | |
| Download: ML19249E525 (8) | |
Text
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Distribution:
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l'AY 2 3 1378 Kelly chron l
Occket llos. 50-416/417 i
PEliORATIDU!i FOR:
D. B. Vassallo, AssisAant Director for LURs, DPil Ti!RU:
T. !!. i!ovak, Chief, Reactor Systems Ccanch, DSS FRC'4:
S. L. Israel, Section Leader, itcactor Systcas Branch, DSS i
SUBJECT:
ACCEPTA!!CE REVIEli - GRAliD GULF Plant !!ame:
Grand Culf Units 1 & 2 Docket !! umbers:
50-416 & 50-417 Licensing Stage:
OL Mt1estone i?unber :
01-21 Responsible Branch LUR-1 l
and Project lianager:
C. Thomas Systems Safety Branch Involved: Reactor Systeas Granch Description 6f i'eview:
Acceptance Review Requested Cenpletion Date:
liay 22, 1978 Revieu Stz*.us:
Complete Reactor Systeus Cranch has revicued Sections 3.5.1.2, 4.6, 5.2.2, 5.2.5, 5.4.6, 5.4.7, 6.3, and 15.0 of the Grand Gulf, Units 1 and 2, draft FSAR.
These sections are acceptable for docketing provided responses to the enclosed questions are subnitted. The staff notes that Grand Gulf has proposed operating limits based upon a recategorization of the turbine-c:c : ' :i'ri-di' ot '17 ss cu.t.
r-t. >>l o t2 n Ly '.h2 s'.J f h;is n'cutifi;d a set of cmarr.s as yet unresolved on this issue. Accordingly, the proposed operating itCPR rrust be nodified prior to pouer operation, unless staff concerns are resolved. Chapter 15.0 cay also have to !;c re-evaluated.
nt:.frweci tin Sanford L. Israel, Section leader Reactor Systems Branch Division of S. stems Safety
Enclosure:
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S. Ucnauer C. Thocas Questions R. Itattson T. I ovak D. Ross S. Israel 8
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212.0 REACTOR SYSTEMS BRANCH 212.1 Per Reg. Guide 1.70, Revision 2 provide elevation and plan (3. 5.1 )
drawings clearly denoting those barriers protecting structures, systems and components listed in Section 3.5.1.2.
When iden-tifying a system to be protected and possible missiles which might affect it, reference these elevation and plan drawings to show how the system is protected.
212.2 Modify Table 3.2-1 to include a listing of those sections of (3.5.1) the SAR which describe the saftey-related structures, systems and components inside containment required for safe shutdown.
212.3 Provide a discussion of the ability of the structures, systems (3.5.1) and components described in Table 3.2-1 to withstand the effects of selected internally generated missiles.
212.4 Under Section 5.2.2.1 cn overpressure protection, provide a (5.2.2) discussion of and identify the postulated events or transients on which the design requirements are based.
Include in your discussion, the assumptions of plant initial conditions and pa rame ters.
212.5 Acceptance Criterion II.2.b of SRP 5.2.2 states that, "All (5.2.2) system and core parameters are at the values within the normal operating range, including uncertainties and technical specifi-cation limits, which would result in the highest transient pressure."
Insufficient infcrmation is presented in the FSAR to detemine that this acceptance criterion will be met.
The applicant shculd confirm that the overpressure analysis will be based on an initial operatino essure (up to the Technical Specification limitj which will t in the most limiting peak pressure.
The applicant r also confirm that the overpressure analysis will ine
.he effects of the ATWS reactor recirculation pump tri un high reactor pressure.
r Acceptance Criterion II.2.c of SRP 5.2.2 states that, "The reactor scram is initiated either by the high pressure signal or by the second signal from the reactor protection system, which-ever is later."
The applicant has stated that the safety valve sizing analyses can take credit for the first indirect scram, which is the high neutron flux scram.
The neutron flux scram occurs before the high pressure scram and results in a lower calculated peak pressure.
The applicant should confirm that the safety valve sizing analyses will be based on the SRP acceptance criterion for reactor scram initiation.
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212-2 212.6 Provide a reference for your studies on overpressurization (5.2.2) which examine the sensitivity of the performance of the system to variations in system and equipment conditions, parameters, and performance.
212.7 Under Section 5.2.2.4.2 provide a discussion of the number and (5.2.2) type of operating cycles for which each component in the over-pressure protection system is designed.
212.8 Per Reg. Guide 1.70, Revision 2, provide the follcwing discus-(5.2.5) sions in Section 5.2.5:
1.
Discuss the reliance placed on the j
systems employed to detect leakage, proper functioning of k
2.
Describe how signals from various leakage detection systems are correlated to provide information to the plant operators on conditions of quantatitive leakage flow rate, 3.
Clearly identify those systems which are not alarmed and which are backups to the alarmed systems, 4.
provide the sensitivity of each detection system.
Justify the ability of these systems to achieve such sensitivity given their normal operating environments, 5.
Discuss full compliance with Reg. Guide 1.45.
6.
Identify fluid systems connected to the primary coolout system and discuss detection and control of intersystem leakage.
212.9 Under Section 5.4.7.1.3 discuss pressure relief capacity in the (5.4.7)
RHR system with respect to operator errors during plant startup and shutdown when the RHR system is not isolated from the RCS.
212.10 Under Section 5.4.7.c.1 provide a complete description of RHR (5.4.7) system interlocks.
212.11 Under Section 5.4.7.2 state the RHR system relief valve capacity, (5.4.7) settings, and state the method of collection of fluids discharged through the relief valve.
212.12 (6.3.2)
Under Section 6.3.2 provide a description of the significant design parameters (including pressure and temperature with explanation of bases for their selection) and design require-ments for ECC delivery lag times for each system.
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212-3 212.13 Provide a failure modes and effect analysis for the ECCS.
(6.3) l 212.14 Identify all ECCS valves which may be potentially submerged (6.3) following a LOCA.
Discuss the consequences cf their spurrious movement or inability to perform as required if submerged.
212.15 It is not clear whether portions of the recirculation pump (3.2) seal cooling water are or are not seismic Category I (Reg.
Guide 1.29).
The staff requires additional information to show that a complete loss of pump seal cooling water would not lead to unacceptable consequences.
212.16 The description or reference to the Standby Liquid Control (4.6)
System should be presented in Section 4.6.
Address the
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requirements of Standard Review Plan (SRP) 4.6.
212.17 Section 5.2.2.10 which addresses safety / relief valve inspection and testing does not provide an adequate discussion of quality (5.2.2) assurance programs to assure that S/R valves will meet speci-fications. The safety / relief valves must be subject to a Q.A. program which meets Appendix B to 10 CFR 50.
212.18 All P & I Diagrams which have their cross hatches deleted are unacceptable for review.
Either provide the cross hatches or provide a suitable method of referencing inter-connections betweca diagrams such that the staff can distinguish interfaces easily.
212.19 The acceptance criteria of SRP 5.4.6 (page 5.4.6-3) state that, (5.4.6)
"As a system which must respond to certain abnormal events, the RCIC system must be designed to seismic Category I standards, as defined in Regulatory Guide 1.29."
The condensate storage tank which is the nomal suction supply for the RCIC is not seismic Category I.
The suppression pool provides a seismic Category I backup source of water, but the switchover requires operator action.
The applicant should confirm that Grand Gulf will conform to the above acceptance criterion.
Either of the following alter-natives would be acceptable approaches for meeting the acceptance criteria:
(1) seismic Category I supply, or (2) safety-grade switchover to a seismic Category I supply, or (3) manual ~ itch-over to a seismic Category I supply if appropriately just The applicant should discuss the approach to be used for Gru Gulf.
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212-4 212.20 The SRP 5,4.7 states the residual heat removal system (RHRS)
(5.4.7) should meet the requirements of General Design Criterion (.GDC) 34 of Appendix A to 10 CFR Part 50.
The RHR ty itself cannot accomplish the heat removal functions as required by GDC 34 To comply with the single failure criterion the FSAR describes an alternate method of achieving cold shutdown in Section 5.4.7.1.5.
Insufficient information is provided to allow an adequate evalua-tion of this alternate method.
In particular, the staff has recently approved 'levision 2 to SRP 5.4.7 (containing Branch Technical Position RSB 5-1) which delineates acceptable methods for meeting the single failure criterion.
This Branch Technical Position requires testing to demonstrate the expected performance of the alternate method for achieving cold shutdown.
The applicant should describe plans to meet this requirement.
In addition, we i
require that all components of the alternate system be safety grade (seismic Category I and IEEE-279).
As a result of this requirement, l
the air supply to the automatic depressurization system (ADS) valves, including the system upstream of the accumulators, must be safety grade.
This air supply must be sufficient to account for air consumption necessary for valve operation plus air loss due to system leakage over a prolonged period with loss of off-site power.
212.21 The SRP 6.3 does not allow credit for operator action for 20 (6.3) minutes following a loss-of-coolant accident (LOCA).
The FSAR states no operator action is required for at least 10 minutes.
The applicant should confirm that no operator action is required until 20 minutes after the LOCA, or provide technical justification and an associated data base to mpport a time less than 20 minutes.
The applicant should identify the iaanual ections,,hich must be performed to prevent safety criteria from being exceeded following a LOCA over the break spectrum, including single failures.
It should also be shown that adequate alarms, instrumentation, and time will be availaba to the operacor to perform manual actions necessary to prevent safety criteria from being exceeded.
212.22 Review procedure III.20 of SRP 6.3 requires that long-term cooling (6.3) capacity folicwiig a LOCA should be adequate in the event of failure of any single active or passive component of the ECCS, Insufficient information is presented in the FSAR to determine that this requirement will be satisfied with regard to passive failurts.
The ECCS should retain its capability to cool the core in the avent of a passive failure during the long-term recircula-tion cocling phase following an accident. We will require Grand Gulf to address the following:
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212-5 Detection and alarms must be provided to alert the operator to passive ECCS failures during long-term cooling which allow suffi-cient time to identify and isolate the faulted ECCS line.
The leak detection system should meet the following requirements:
1.
Identification and justification of maximum leak rate should be provided.
2.
Itaximum allowable time for operator action should be provided and justi fied.
3.
Demonstration should be provided tnat the leak detection system will be sensitive enough to initiate (by alarm) operator action, permit identification of the faulted line, and isolation of the line prior to the leak creating undesirable consequences such as flooding of redundant I
equipment.
The minimum time following initiation of an alarm before operator action is permitted is 30 minutes.
4.
It should be shown that the leak detection system can identify the faulted ECCS train and that the leak is isolable.
5.
The leak detection system must meet the following standards:
a.
Control Room Alarm, b.
IEEE-279, except single failure requirements.
In addition, Grand Gulf should determine the effects on ECCS of passive failures such as pump seals, valve seals, and measurement devices.
This analysis snould address the potential for ECCS flooding and ECCS inoperability that could result from a depletion of suppression pool water inventory.
The analysis should include consideration of (1) the flow paths of the radioactive fluid through floor drains, sump pump discharge piping, and the auxil-iary building; (2) the operation of the auxiliary systems that would receive this radioactive fluid; (3) the ability of the leakage detection system to detect the passive failure; and (4) the ability of the operator to isolate the ECCS passive failure, including the case of an ECCS suction valve seal failure.
212.23 Review procedure III.5 of SRP 6.3 requires that prior to instal-(6.3) lation, revesentative active components used in the ECCS will be proof-tested under environmentel conditions and for time periods representative of the most severe operating conditions to which they may be subjected.
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212-6 Insufficient information is presented in the FSAR to determine that proof testing has been performed for ECCS pumps which must function during the long tem following a loss-of-coolant accident.
Grand Gulf should demonstrate that the design of the ECCS pumps which must function during the long term following a loss-of-coolant accident have been qualified by representative testing.
212.24 Provide analyses to show that diversion of ECCS to containment (6.3) cooling at or less than 10 minutes after a LOCA will not result in exceeding any safety criteria for the entire break spectrum with consideration of single failure.
212.25 Address the inadvertent closure of the reactor recirculation (6.3) system line suction valve as a single failure in the LOCA analysis, for the break size most affected by this failure.
212.26 Provide an analysis of "The Loss of Instrument Air" transient.
(15.0) g 2.12.27 Provide an evaluation which assesses whether the consequences of (5.5.7) a single valve malfunction or operator error could result in possible damage to a heat exchanger of the RHR system while in the steam condensing mode.
The evaluation should examine the consequences from two aspects:
(1) overpressurization and, (2) hydrodynamic forces.
Show that the systems will respond in an acceptable manner.
212.28 Modify NS0A drawings to include benefits of non-safety grade (15.0) equipment which mitigate transients and accidents.
Such equip-ment includes relief valves, rod block monitors, and vessel level (high) trip.
212.29 In the analyses for the generator load rejection and turbine trip (15.2) transients, credit is taken for inmediate reactor scram and recir-culation pump trip obtained from a valve closure signal (turbine control valve for load rejection and turbine stop valve for turbine trip).
Analyze these transients without taking credit for immediate reactor scram and recirculation pump trip.
Take credit only for safety-grade, seismic Category I equipment and assume loss of offsite power.
What is the effect of the failure of a single safety-grade component?
Present curves similar to those of Figures 15.2-2 and 15.2-4 and give values of maximum vessel pressure and minimum MCPR with the times at which these values occur.
Evaluate the percent of fuel rods which would reach boiling transition.
Since this event is not an anticipated transient, limited fuel failure can be allcwed if dose consequences are acceptable.
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212-7 212.30 Identify the limiting transient for each category in Section 15.0.2.
(15.0)
For MCPR limiting transients, provide the fiCPR versus time plots.
Large scale time plots of these parameters presented in Chapter 15.0 should be presented for the limiting transient in each category.
212.31 The applicant must provide assurance that the pressure-time plots (15.0) in Chapter 15 are consistent with the initiation logic for the safety-relief valves.
For example, modifications may have been made to the safety / relief system to prevent subsequent reopening of these valves during pressure increase transients to meet con-tainment design base loadings.
212.32 Provide assurance that the limiting pump trip is assumed in (15.0) analyzing decrease in reactor coolant system flow rate transients.
i The trip initiated from a loss of power may be different than I
a trip initiated from the recirculation pump trip (RPT) system since the location of the electrical breakers may be different and, thereby, cause different coastdown char acteristics.
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