ML20040A702
| ML20040A702 | |
| Person / Time | |
|---|---|
| Site: | 05000561 |
| Issue date: | 09/22/1977 |
| From: | Mazetis G Office of Nuclear Reactor Regulation |
| To: | Cox T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201210436 | |
| Download: ML20040A702 (2) | |
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UNITED STATES
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, I+4 NUCLEAR REGULATORY COMMISSION
.D W ASHINGTON, D. C. 20555 e
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SEP 2 21977 Docket No. 50-561 MEMORANDUM FOR:
T. Cox, LPM, LWR Branch No. 3. DPM THRU:
T. M. Novak, Chief, Reactor Systems Branch, 055 FROM:
G. R. Mazetis, Section Leader, RSB, DSS
SUBJECT:
B-SAR-205 OPEN ISSUE ON PROVISIONS FOR SHUTDOWN
REFERENCES:
1.
Report to the Advisory Committee on Reactor Safeguards by the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, in the Matter of Babcock & Wilcox Company Reference Safety Analysis Report, B-SAR-205, July 8, 1977.
2.
Ltr., J. H. Taylor to R. S. Boyd, "B-SAR-205 - Outstanding Issues," July 21, 1977.
Open issue number 7, as described in Reference 1, requires that B&W provide additional interface specifications for a referencing applicant to assure that such applicant can provide a B0P capability to reach and maintain hot shutdown conditions from normal operating conditions, using only safety-grade equipment and assuming the loss of offsite, power.
The int'erface requirement in Reference 2 was acceptable with respect to the above requirement; however, the interface requirement; contained in Amend-ment 17 to B-SAR-205 does not specify the same requirement as described in the July 21 letter and should be revised.
To clarify the staff's position, the following in,terfac'e requirement would be acceptable:
A sufficient quantity of auxiliary feedwater must be provided to reach and maintain hot shutdown with only safety-grade equipment and assuming the loss of offsite pcuer for a.suf-ficient period of time to allow restoration of main feedwater.
,1 erald R. F is, Sec' tion Leader Reactor Syst Branch Division of Systems Safety
Contact:
Scott Newberry, NRR 49-27911 8201210436 810403 I
MADDEN 80-515 PDR
.a 2-T. Cox SEP 2 2 577 cc:
D. Ross
- 0. Parr S. Burwell T. Novak S. Israel G. Mazetis RSB IGInbers e
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UNITED STATES 3%i, NUCLEAR REGULATORY COMMISSION
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j WASWNGTON, D. C. 20555 l
S Docket Nos. 50-329 i
& 330 SEP 2 81977 i
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MEMORANDUM FOR:
D. B. Vassallo, Assistant Director for LWR's, DPM THRU:
Thomas M. Novak, Chief, Reactor Systems Branch, DSS;f' i
FROM:
G. R. Mazetis, Section Leader, RSB, DSS
SUBJECT:
ACCEPTANCE REVIEW 0F THE CONSUMERS POWER COMPANY MIDLAND PLANTS 1 AND 2 FINAL SAFETY ANALYSIS REPORT Plant Name: Midland Plant Units 1 and 2 Docket Nos.:
50-329 and 50-330 Licensing Stage: OL Milestone No.: 01-21 Responsible Branch & Project Manager: LWR-4, D. Hood Systems Safety Branch Involved: Reactor Systems Branch f.s Description of Review: Acceptance Review Requested Completion Date: September 27,1977 ~
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Review Status: Complete The Reactor Systems Branch has' reviewed sections 1.5, 3.5.1.2, 4.6, 5.2.2, 5.2.5, 5.4.1, 5.4.2, 5.4.3, 5.4.7, 5.4.10, 6.3 and 15 of the Midland Plants 1 and 2 FSAR for completeness, using the guidelines specified in the Regulatory Guide 1.70, Revision 2, " Standard Fonnat and Content of Safety Analysis Reports for Nuclear Power Plants". 'This license applica-tion does not presently provide adequate information for us to complete an in-depth review. provides a list of that information which must be submitted in time for us to complete our Round One questions.
We_ suggest submittal of aoplicant resconses at least four weeks prior to.
the Round One due date would be sufficientiv timely. Our acceptance of 56cEeting is contingent on this coninitment ~ by the applicant.
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Twelve hours were devoted to this review.
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G. R. Mazeti etion Leader Reactor Systems Branch Division of Systems Safety
Enclosure:
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List of Info.y Os n S AA M (seenextpaN
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Contact:
S. Newberry, NRR 27911 x
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- D. B. Vassallo '
SEP 2 81977 cc:
S. Hanauer R. Mattson
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S. Israel G. Mazetis t
S. Newberry 1
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j 211.0 REACTOR SYSTEMS BRANCH
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.211.1 Provide plots of DNBR vs time for those events required by 1
(15.0)
R. G. 1.70 Rev. 2.
211.2 The description of the steam pressure regulator malfunction and l
(15.1.3) the inadvertent opening of a sMam generator relief or safety 4
valve indicated that the consequences of these events are i
bounded by the main steam line break. Since the staff criterion to ensure no fuel damage is DNBR>1.30, provide the specific analyses for these cooldown transients to show that DNBR remains j
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 infomation provided in section 15.1.5 for main steamline break 1
(15.1.5) is not adequate. Provide analysis to locate the worst case break, considering the most limiting single active component failure, (FWIV, 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 break:
1.
PCT 5.
pressurizer level 2.
DNBR,
6.
steam generator levels 4
3.
reactivity. margin 7.
steam generator pressures O
4.
break flow rate N~] 211.4 Operational analyses or failure mode and effects analysis of the (15.0) various plant responses to the Chapter 15 events are required.
To complement the SAR discussions in this regard, provide a sumary of a systematic functional analysis of components required for each event analyzed in Chapter 15.0.
The summary should be shown in the form of simple block diagrams beginning with the event, branching out to the various possible protection sequences for i
each safety action required to mitigate the consequences of the event (e.g., core cooling, containment isolatio'n, pressure relief, scram, operator action, etc.), and ending with an identification of the specific safety actions being provided.
When complete, each protection sequence diagram should clearly iden -
tify (for each event) the safety systems required to function to provide the safety actions necessary to mitigate the consequences of the transient or accident (during any plant operating state).
An example of such a systematic functional analysis is contained in " Transactions of the American Nuclear Society 1973 Winter Meeting",
l November 11-15, pages 339-340.
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211.5 Provide complete NPSH calculations for the ECCS pumps in both the
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(6.3) injection and recirculation modes. Provide all assumptions and q
appropriate justifications.
211.6 The reference to BAW-10104 and BAW-10103 for the required'ECCS l
(6.3) analyses in accordance with 10 CFR 50.46 is insufficient. Provide appropriate calculations and sensitivity studies (or ieferences) which consider the impact of more recent model or equipment changes I
(such as vessel U-baffle modifications). Also, provide a discussion I
with references, of all applicable calculations using the small i
break model.
i 211.7 The turbine trip analysis assumes credit for ICS'and turbine bypass.
(15.2.3)
Provide or reference an analysis for turbina trip taking no credit for any non-safety grade equipment.
211.8 Submit an analyris of the worst case overpressure transient during (5.2.2) startup and shut, w. Provide all assumptions. Plots should in-clude pressure vs. time, reactor coolant temperature vs. time and safety valve flows versus time. Show that the pressure-temperature limits in Technical Specifications are not exceeded.
The following pcsition is currently being considered for implementation by the NRC staff.
Provide a discussion for the Midland design with respect to each of these points:
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1.
A system shall be designed and installed which will prevent V'
exceeding the applicable Technical Specifications and App. G
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limits for the reactor pressure vessel while operating at low temperatures. The system shall be capable of relieving pressure during all potential overpressurization events at a rate sufficient to satisfy the Technical Specification limits, particularly while the Reactor Coolant System is in a water-solid condition.
2.
The system must be able to perform its fungtion assuming any single active component failure. Analyses using appropriate calculational techniques must be provided which demonstrate that the system will provide the required pressure relief capacity assuming the most limiting single active failure.
The cause for initiation of the event, i.e., operator error, component malfunction, etc., will not be considered as the single active failure. The analysis should assume the most limiting allowable operating conditions (e.g., one RHR train operating or available for letdown, other components in nonnal operation when the system is water solid such as pressurizer heaters and charging pumps).
All potential over-pressurization events must be considered when establishing the worst case event.
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The system must operate automatically, providing a completely l
independent backup protective feature for the operator. The j
design must not require manual actions to enable or " turn on" the system or to mitigate the consequences of a potential i
overpressure event.
4.
To assure operational readiness, the overpressure protection system must be tested in the following manner:
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a.
A test must be performed to assure operability of the system electronics prior to each shutdown.
b.
A test for valve operability must, as a minimum, be con-j ducted as specified in the ASME Code Section XI.
c.
Subsequent to system, valve, or electronics maintenance,
-i a test on that portion (s) of the system must be performed prior to declaring the system cperational 5.
The system must meet the c'.esign requirements of IEEE-279, Regulatory Guide 1.26, and Section III of the ASME Code.
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6.
The protection system does not have to meet Seismic Category I requirements if it can be shown that an earthquake would not m
initiate an overpressure transient.
The postulated earth-
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quake should be of a magnitude equivalent to the SSE.
If the earthquake can initiate an overpressure transient, then i
it should be assumed that loss of offsite power is an expected consequence of the event and the protection system should be designed to Seismic Category I requirements and not require the availability of offKite power to perform its function.
Should the applicant show tnan a postulated earthquake
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could not cause an overpressure event, the overpressure 4
protection system design must not compromis.e the design criteria of any other safety-grade system with which it would l
interface. The requirements of Regulatory Guide 1.29 must be satisfied.
7.
The loss of offsite power shall be considered as an anticipated ^
transient which could occur while in a shutdown condition.
If this event can initiate an overpressure transient, the over-pressure protection system must be independent of offsite i
power, in addition to performing its function assuming any l
single active failu e.
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Plant designs which take credit for an active component (s)
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to mitigate the consequences of an overpressurization event 1
must include additional analyses considering inadvertent i
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?
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i 211.10 Provide the following information considering a pipe break in a
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(6.3) high pressure injection (HPI) line between the reactor coolant 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 pump performance and availability during this event e.
flow splits in HPI piping i
f.
sumary table of scenario listing each event and associated times, i
2 Provide a discussion for each Chapter 15 event describing the F (11.11\\ 15.0) operator actions required in both the short and long term. Our
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interest is in evaluating the operator's role in achieving and i
maintaining stable conditions. (Stable conditions can be assumed to be achieved when the decay heat removal system is placed in operation). An example of such a situation would be the necessity of the operator to secure the HPI pumps after a steam line break i
to prevent repressurization of the reactor coolant system at i
low temperature.
l 211.12 Provide an analysis of a break in the,~ normally pressurized makeup (none) line considering all potential single active component failures.
As a minimum, submit the following:
i a.
Table depicting the sequence of events
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b.
Indications and alanns available c.
Operator action (s) required l
d.
Plots of RCS pressure, RCS water level l
211.13 Provide additional analyses of the boron dilution event considering (15.4.6) the plant conditions other than just power operation or refueling (as specified in Standard Review Plan 15.4.6). Discuss all potential dilution sources for the Midland Plant and address the design aspects which preclude or minimize the potential for a dilution event.
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211.14 The Decay Heat Removal System incorporates low-flow DH pump trip
'(5.4.7) interlock. Discuss this features potential contribution to the prob-ability of a complete loss of low pressure injection during a LOCA.
Balance this risk with the 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 reactor operation. Also, present your plans and capability for preoperational or startup tests to substantiate the analyses.
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