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NUCLEAR REGULATORY COMMisslON
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OCT 2 g $17 Docket Nos. 50-514/515 MEMORANDUM FOR:
D. B Vassallo, Assistant Director for LWRs, DPM FROM:
D. F. Ross, Jr., Assistant Director for Reactor Safety, DSS
SUBJECT:
OPEfl ISSUES ON PEBBLE SPRINGS The issues contained in the enclosed lists were recently identified and addressed on B-SAR-205 and are also applicable to the Pebble Springs docket. Enclosure 1 lists those areas which must be resolved prior to issuance of a Construction Permit, while Enclosure 2 contains those items I
which we feel can be adequately addressed at the Final Design Stage of review.
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D. F Ross, Jr., Assistant Director for Reactor Safety i
Division of Systems Safety i
Enclosures:
As Stated
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S. Hanauer R. Mattson I'
D. Ross i
S. Varga
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S. Israel 5,flewberryk l
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Contact:
Scott Newberry, NRR 49-27341
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ENCLOSURE 1
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OPEN ITEMS TO BE RESOLVED PRIOR TO A CP l.
Decay Heat Removal System Isolation Valves--This issue is identified in Section 7.4.1 of the Pebble Springs SER. The power supply arrange-ment to the DHR system suction isoiation valves is such that during a pipe break outside containment, one train of the system cannot be isolated from the Reactor Coolant System assuming the single failure of one electrical bus. An acceptable design would be separate Class lE power supplies for each valve as was recently submitted in B-SAR-205.
2.
Overpressure Protection at low Operating Temperatures--The staff is currently developing a position which will provide requirem.ents for i
the design of a protection system for these events. The applicant must commit to the following minimum criteria prior to issuance of a CP:
i (1) Credit for operator action. No credit can be taken for operator action until 10 minutes af ter the operatir is made aware that a transient is in progrese..
(2) Single failure criteria.
The pressure protection system should be designed to protect the reactor vessel, given any event initiating a pressure transient, and followed by a single active component failure. Redundant or diverse pressure protection 2
systems will be considered.as meeting the single failure criteria.
1 (3) Testability.
Provisions for periodic testing of the overpressure protection system (s) and components shall be provided. The program of tests, and frequency or schedule thereof, will be selected to assure functional capability when required.
(4) Seismic design and Standard 2/9-1971 criteria.
Ideally, the pressure protection sysemm(s) should meet both Seismic Category 1 i
and Standard 2/9-1971 criteria.
The basic objective, however, e
is that the system (s) should not be vulnerable to an event which both causes a pressere protection (5) Reliability.
The system (s) provided must not reduce the reliability of the emergency core cooling system or residual heat removal systems.
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3.
ECCS Analyses - The applicant references topical reports BAW-10102, IfAW-10074, andBBAW-10104 for the analyses in accordance with 10 CFR 50.46. Since the approval, by the staff, of the emergency core cooling system model described in BAW-10104 several chances to the model have been submitted by Babcock & Wilcox and approved by us. To have a referenceable worst break which is wholly in conformance with 10 CFR 50.46 Appendix K and, to ensure that the selected postulated breaks adequately define the " worst-case" situations, additional analyses are required.
(These analyses were requested in a Commission letter from S. Varga of NRC,.to James H. Taylor of B&W, dated May 10,1977).
4.
High Pressure Injection Line Break--A break in a high pressure injection line (HPI) between the reactor coolant system piping and the last HPI check valve results in a small LOCA. We require additional information to evaluate the consequences and necessary operator actions to mitigate the consequences of this event.
5.
Provisions for Shutdown--Th'e applicant must demonstrate that the plant can remain for a prolonged period in a hot shutdown condition assuming loss of offsite power and using only safety-grade equipment or show that the plant can be cooled and depressurized using only safety-grade dquipment (assuming loss of offsite power) to the level required for decay heat removal system actuation.
6-Makeup Line Break--The applicant must evaluate the required actions and consequences resulting from a break' in the normally pressurized makeup line considering all potential single active component failures.
7.
Passive Failures--It is our position that detection and alarms be provided to al.ert the opera' tor to passive ECCS failures during long-term cooling following a LOCA which allow sufficient time to identify and isolate the faulted ECCS line. The applicant will be required to commit to the staff's position, (see also the' detailed discussion in NUREG 0138).
8.
Essential Manual Valves in ECCS--Experience has shown that considera-tion must be given to the possibility that, prior to an accident locally manual valves (handwheel) might be left in the wrong position and remain undetected. The staff will require remote position indication in the control room for all such manual ECCS valves, the i
mispositioning of which could compromise ECCS performance.
9.
Excessive Heat Removal Events of Moderate Frequency--The applicant is required to show that no fuel damage occurs for such events (DNBR >l.30).
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3-Therefore, it is not appropriate to reference the main steam line break analysis which shows DNBR < l.32 at 3.1 seconds into the transient.
- 10. Decay Heat Removal System Cooler Bypass Valves--The rate of cooldown is normally controlled with these valves. The concern is that loss of air to these valves, causing them to fail closed, may result in maximum flow being directed through the coolers. This could result in an excessive cooldown rate of the reactor coolant system.
11.
Feedwater Isolation-The main' feed system contains two headers, one for each steam generator.
Each header contains only one safety-grade feedwater isolation valve which received redundant ESFAS signals. The applicant must show that the failure of this valve is considered in those Chapter 15 events requiring feedwater isolation.
412. Chapter 15 Events--The applicant will be required to provide a discussion for each Chapter 15 event describing all of the actions required in the recovery mode following a transient. Our. interest is in evaluating the operator's role in achieving and maintaining stable conditions. An example 'of such a situation would be the necessity of the operator to secure the HPI pumps after a steam line break to prevent repressurization of the reactor coolant system at low temperatures.
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ENCLOSURt 2 j'*..
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Credit for Non-Safety-Grade Systems--The applicant must show that no credit is assumed for non-safety-grade systems for the mitigation of any Chapter 15 event.
For exemple, the turbine trip analysis in i
the PSAR assumes power runback by the ICS.
Table 15.0-3 shows the i
" Equipment Assumed Functioning in the Accident Analysis."
2.
Baron Dilution Events--The applicant must provide additional analyses of the boron dilution events considering the plant conditions other
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than power operation or refueling (as specified in Standard Review plan 15.4.6).
In addition, they must discuss all potential dilution i
sources.
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