ML20209E541
| ML20209E541 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 05/30/1985 |
| From: | Bemis P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Jordan E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| Shared Package | |
| ML20209C203 | List: |
| References | |
| FOIA-86-729 NUDOCS 8507010307 | |
| Download: ML20209E541 (2) | |
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[AS,at:
UNITEJ STATES NUCLEAR REGULATORY COMMISSION o
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REGloN il 3
Ij 101 MARIETTA STREET.N.W.
e ATLANTA. GEORGI A 30323 ME 3 01985 MEMORANDUM FOR:
Edward L. Jordan, Director Division of Emergency Preparedness and Engineering Response, IE FROM:
Paul R. Bemis Director Division of Reactor Safety
SUBJECT:
INTERPRETATIONS OF 10 CFR 50.59 AND 10 CFR 50.72 Recently, two separate equipment problems at the Robinson facility have raised questions concerning the NRC approval portion of 10 CFR 50.59 and the reporting requirements of 10 CFR 50.72. Although both problems were in separate unrelated b,eing outside their " design basis". systems, they share a commonality in that In the first equipment problem, the logic controlling Main Steam Isolation Valve (MSIV) closure,the Itce should an isolation signal be generated from either one of the two redundant Engineering Safety Features (ESF) trains.
(Detatis discussed in inspection report 261/84-53 and LER 85-002.) The updated Final Safety Analysis Report (FSAR) provides a description and analysis for the Steam System Piping Failure Accident.
As stated in the FSAR, the MSIV valve arrangement is expected "to prevent the blowdown of more than one steam generator for any break location, even if one valve fails to close."
As originally constructed, the MSIV logic did not provide this protection for all steam line breaks in the event of a single failure in the ESF "B"
logic train.
A break in the piping downstream of the main steam check valves and the failure of a single active component in the ESF "B" train would result in the blowdown of both the "B" and "C" steam generators.
In the second case, a licensee review of safety-related loads on two 480 VAC Motor Control Centers (MCCs) revealed that several of the loads had oversized overcurrent trip devices, in that an instantaneous trip of one of there devices would also cause the MCC feeder breaker to trip, resulting in loss of power to all the loads on the MCC. (Details discussed in inspection report 261/85-08 and LER 85-008.)
A further review revealed that a design drawing documented the installed devices to be the correct size.
10 CFR 50.2 defines " Design Basis" to be information which identifies the specific functions to be performed by a structure, system, or component of a factitty, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design.
derived from generally accepted " state of the art"These values may be (1) restraints practices for achieving functional goals, or
- 2) requirements derived from analysis (based on calculations and/or expe(riments) of the effects of a postulated accident for which a structure, system, or component must meet its functional goals.
F edrickson me
H Edward L. Jordan 2
MAY 3 01985 10CFR50.72(b 1)(it)(B) requires that the licensee notify the NRC within one hour of any eve)n(t or condition during operation that re plant being in a condition that is outside the design basis of the plant.
Clearly, the MSIV problem was outside the equipment design basis and the basis accident and thus, resulted in "the nuclear power plant being in a
. condition that is outside the design basis of the plant."
In comparison to the basis and thus by extrapolation, was also outside the "
plant."
The difference is that while the MSIV problem was beyond the design basis accident, the trip device was not, in that loss.cf one MCC is analyzed in i
the FSAR.
The trip device difference raises an interpretation concerning the wording of 10 CFR 50.72.
In that the 10 CFR 50.72(b) question (1)(11 requirements state "outside the design basis of the plant" and not "outside the design basis accident," the most conservative position is that any equipment condition that places that item cutside its design basis is reportable, irrespective of the effect on the design basis accident. The Region II position is that " design basis of the plant" refers to the design basis accident analysis and that individual items being beyond their design basis 3
10 CFR 50.2 does not necessarily require reporting under 10 CFR 50.72.as defined in i
L Similar to the 10 CFR 50.72 question, the wording of 10 CFR 50.59 also raises a question of applicability criteria.
Experiments," enables the holder of a10 CFR 50.59, titled " Changes, Tests and t
license authorizing operation of a utilization facility to make changes in the facility "as described" in the FSAR without prior NRC approval, provided several specific conditions are met.
Although the MSIV and MCC systems are " described" in the FSAR, the as-built and thus each system was not "as described" in the FSAR. condi With changes to systems " described" in the FSAR, but not "as described", the literal interpre-tation is that licensees must receive NRC approval prior to making these mod fact, with equipment outside its design basis, a potential unreviewed safety In question exists with the equipment as-built.
The Region 11 position is that if a modification will correct the discrepancy so that the system becomes ' "as i
described" in the FSAR, th'en 10 CFR 50.59 applies and evaluation of the modif t-cation in accordance with 10 CFR 50.59 should be conducted by the licensee.
Region II requests that you review these generic questions as they relate to the intent of 10 CFR 50.5g and 10 CFR 50.72 and inform us if the NRC position differs from that proposed by Region II.
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Paul R. Bemis M
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