ML19289G312
| ML19289G312 | |
| Person / Time | |
|---|---|
| Site: | Crane, Midland  |
| Issue date: | 09/28/1977 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19289G307 | List: |
| References | |
| NUDOCS 7908220450 | |
| Download: ML19289G312 (5) | |
Text
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211.0 REACTOR SYSTEMS BRANCH 211.1 Provide plots of DNBR vs time for those events required by (15.0)
R. G. 1.70 Rev. 2.
211.2 The description of the steam pressure regulator malfunction and (15.1.3) the inadvertent opening of a steam generator relief or safety valve indicated that the consequences of these events are bounded by the main steam ifne break. Since the staff criterion to ensure no fuel damage is DNER>l.30, provide the specific analyses for these cooldown transients to show that DNBR remains greater than 1.30 for each event or show that DNBR remains greater than 1.30 for the worst case main steam line break.
211.3 The information provided in section 15.1.5 for main steamline break (15.1.5) is not adequate. Provide analysis to locate the worst case break, considering the most limiting single active ccmponent failure, (F4IV, MSIV etc.) the assumption of offsite power available or not available, and the break location. Provide as a minimum the following plots for the worst brsak:
1.
PCT 5.
pressurizer level 2.
DNBR 5.
steam generator levels 3.
reactivity margin 7.
steam generator pressures J. break flow rate 211.4 Operational analyses or failure made and effects analysis of the (15.0) varicus plant responses to the Chapter 15 events are recuired.
To c:molement the SAR discussions in -his regard, provide a su::Tnary of a systematic functional analysis of c:rconents recuired for each event analyzed in Chaoter 15.0.
ine summary snould be shewn in the form of simple bicek diagrams beginning with the event, branching out to the various possible protecticn secuences for each safety action recuired to mitigate One c:nsecuences of the event (e.g., core cooling, containment isolation, ;ressure relief, scram, operator action, etc.), and endinc with an identification of the specific safety actions being proviced.
'4 hen cerclete, each oratection sequence ciagram snould clearly iden-tify (for each event) the safety systems recuired to function to provide the safety actions necessary to mitigate the c:nsecuences of the transient or accident (curing any clant operating state).
An examole of such a systematic functional analysis is c:ntained in " Transactions of the American Nuclear Society 1973 Winter Meeting",
Novem er 11-15, pages 339-340.
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211.5 Provide complete NPSH calculations for the ECCS pumps in both the (6.3) injection and recirculation modes. Provide all assumptions and appropriate justifications.
211.6 The reference to SAW-10104 and BAW-10103 for the required ECCS (6.3) analyses in accordance with 10 CFR 50.46 is insufficient. Provide appropriate calculations and sensitivity studies (or references) which consider the impact of more recent model or equipment changes (such as vessel U-baffle modifications). Also, provide a discussion with references, of all applicable calculations using the small break model.
211.7 The turbine trip analysis assumes credit for ICS and turbine bypass.
(15.2.3)
Provide or reference an analysis for turbine trip taking no credit for any non-sifety grade equipment.
2.11.8 Submit an analysis of the worst case overpressure transient during -
(5.2.2) startup and shutdown. Provide all assumotions. Plots should in-clude p sssure vs. time, reactor coolant temperature vs. time and safety valve flows versus time. Show that the pressure-temperature limits in Technical Saecifications are not exceeded.
The following position is currently being considered for
. implementation by the NRC staff. Provide a discussion for the Midland design with respect to each of these points:
1.
A system shall be designed and installed which will prevent exceeding the applicable Technical Specifications and App. G limits for the reactor pressure vessel while operating at low temceratures. The system shall be ca:able of relieving pressure during all potential overoressuri:ation events at a rate sufficient to satisfy the Technical Scecification limits, particularly while the Reactor Coolant System is in a water-solid condition.
2.
The system must be able to cerform its function assuming any single active ccmconen: f ailure. Analyses using a:or:oria:e calculational tecnnicues must be proviced wnich demonstrate that the system will provide the required :ressure relief ca:acity assuming the most limiting single active failure.
The cause for initiation of the event, i.e., ocerator error, comoonent malfunction, etc., will not be censidered as the single active failure. The analysis should assume the most limiting allowable acerating conditions (e.g., one RHR rain c:erating or availaole for let:own, c:her ::moonents in normai oceration when the system is water solid sucn as
- ressurizer heaters and charging pumes). All cotantial over-ressurization events must be consicered wnen estaoiishing the wors casa event.
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3.
The system must operate automatically, providing a completely independent backup protective feature for the operator. The design must not require manual actions to enable or " turn on" the system or to mitigate the consequences of a potential overpressure event.
4.
To assure operational readiness, the overpressure protection system must be tasted in the following manner:
a.
A test must be performed to assure operab.ility of the system electronics prior to each shutdown.
b.
A test for valve operability must, as a minimum, be con-ducted as specified in the ASME Code Section XI.
c.
Subsequent to system, valve, or electronics maintenance, a test on that portion (s) of the system must be performed prior to declaring the system operational 5.
The system must meet the design requirements of IEEE-279, Regulatory Guide 1.25, and Section III of the ASME Code.
6.
The protection system does not have to meet Seismic Category I requirements if it can be shown that an earthquake would not initiate an overpressure tran'sient. The postulated earth-cuake should be of a magnitude equivalent to the SSE.
If the earthquake can initiate an overpressure transient, then it should be assumed that loss of offsite power is an excected consequence of the event and the protection system should be designed to Seismic Category I recuirements and not require the availability of offsite power to :erfom its function.
Should the aoplicant show than a postulated earthquake could not cause an overpressure event, the overpressure protection system design must not comoromise the design critaria of any other safety-grade system with wnich it wcula interface. The requirements of Regulatory Guide 1.29 must be satisfied.
7.
The loss of offsita power shall be considered as an anticicatac transient which could occur while in a shutdcwn condition.
If this event can initiate an overaressure transient, the over-cressure protection system must be indecencent of offsite ower, in addition to per# arming its function assuming any single active railure.
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8.
Plant designs which take credit for aa active component (s) to mitigate the consequences of an ou rpressurization event must include additional analyses considering inadvertent initiation / actuation or provide justification to show that existing analyses bound such an event.
211.9 Show how the Midland Plants can be maintained at hot shutdown (none) with only safety grade systems assuming the loss of offsite power. How long can the plant be kept in this condition prior to requiring cooldown?
211.10 Provide the following infomation considering a pipe break in a (6.3) high pressure injection (HPI) line between the reactor coolan system piping and the last HPI check valve:
a.
operator action (s) required b.
indications provided for the operator c.
time operator action required d.
HPI pu=o perfomance and availability during this event e.
flow splits in HPI piping f.
surm:ary table of scenario listing each event and associated times.
211.11 Provide a discussion for each Chacter 15 event describing the (15.0) operator actions rectired in both the short and long tem. Our interest is in evaluating the operator's role in achieving and maintaining, stable conditions. (Stable conditions can be assumed to be achieved when the decay heat removal system is placed in oceration). An example of such a situation would be tne necessity of the acerator to secure the HPI pumos after a steam line creak to prevent r:cressuri:ation of the reac::r c olant system at low taccerature.
211.12 provide an analysis of a break in the nor ally pressuri:ed makeu::
(none) line considering all potential single active c:mponent failures.
As a minimum, submi: the folicwing:
a.
Table decicting the secuence of events b.
Indications and alams available c.
Ocerator action (s) recuired d.
?- vice adciticnal analyses of the boren cilution even: c:nsidering (15.2.5) the plan concitions other than just power a:eration or cefueling (as s:ecified in Standard Review Plan 15.1. 5). Discuss all :otential dilution sour:es for the Micland Plart and accress :he design aspects wnich ;;recluce or minimi:e tne ;otantial for a cilution event.
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211.14 The Decay Heat Removal System incorporates low-flow CH pump trip
'(5.4.7) interlock. Discuss this featurei potential contribution to the prob-ability of a completa loss of low pressure injection during a LOCA.
Balance this risk with tne gain in availability of the DH function.
211.15 Discuss the loss of instrument air for the Midland Plants showing (none) that it meets the appropriate acceptance criteria for a moderate frequency event. Provide a detailed failure modes and effects discussion consistent with question 211.4. Causes and. potential systems interactions should be particularly addressed and the loss of instrument air should be considered during all phases of reactt-operation. Also, present your plans and capability for preope ational or startup tests to substantiate the analyses.
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