ML20153H189
| ML20153H189 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom, 05000000 |
| Issue date: | 07/14/1981 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| Shared Package | |
| ML20151G218 | List: |
| References | |
| FOIA-88-353 NUDOCS 8809090113 | |
| Download: ML20153H189 (2) | |
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MEMORANDUM FOR:
Boyce H. Grier, Of rector, Region I THRU:
(1)
- 2. Keimig, Chief, Projects Branch 2, ORPI, RI (2)
E. J. Brunner, Acting Director, ORPI, RI FROM:
E. C. McCabe, Chief, Reactor Projects Section #28, ORPI SUM ECT:
VIOLATION SEVERITY FOR ORWELL PRESSURE SENSOR ISOLATION AT PEACHBOTTOM UNIT 2 Backaround 4
On March 31, 1981, the licensee found th.ree drywell pressure sensors valved out of service at Peach Bottom Unit 2.
The valves were about 1/8 turn open and functioning to equalize pr6ssure slowly enough to follow ncrmai changes in containment pressure but not enough to provide the specified accident response time.
Two of the valves were for sensors in che same ECCS #ctuation channel.
The third sensor was in a reactor protectica system channel.
Extensive licensee checks for other sensor valving errors found no further problems.
The valves involved were last known to be in the epen position about August 8.1980.
Licensee evaluation and residant inspector confimation showed that the valving error did not prevent scram or ECCS actuation because the rest of the logic remained operable. The licensee detected, promptly corrected, and properly reported this occurrence.
Safety Siontfic. nee ne valving errors 1sfs ECCS and RPS actuation on high drywell pressure suscepti-ble to single failure for up to about 200 days.
That single failure did not occur.
Automatic core protective capability was not lost.
Even if additional drywell pressure sensor failure had o: curred, there would have been protection against a large break LOCA (lo level scram, le-lo level HPCI and RCIC initiation, lo-lo-lo level plus low reactor pressure initiation of LPCI and Core Spray).
For a small break LOCA without drywell pressure input, there is no ADS actuation, ht operator action can compensate, and that sequence is one in which manual operator response time is adequate. This situation represents a reduction in the i
margin of safety, with other safety features and operator training capable of assuring core protection even if the additional failure were experienced.
(This assessment appears consistent with the NSSS supplier's evaluations in NEDO 10189 and NEDO 24708, obtained from the Hatch licensee by P.egion II.
The NSSS supplier analysis states that about 10 minutes is available for the operator to initiate depressurization.)
Violation Severity Section III of the Interin Enforcement Criteria states:
Severity III Violations are of significant regulatory concaten and, in general, involve utual or high i
i 8809090113 000017 ER O-353 PDR
Memo, for Mr. B; H. Gris.
Z potential impact on the public; Severity IV Violations include degradation of engineered systems designed to detect, prevent, or mitigate an event; and Severity IV Viclations in themselves are not cause for significant concern but could lead to matters of significant concern if uncorrected.
The actual event which occurred meets the Severity Level IV definition of a degraded engineered system designed to detect, prevent, or mitigate an eve'nt.
Literal reading of Supplement I to *,he vnforcement criteric could result in fulfillment of the definitions for both Severity III and Severity IV Violations Since the supplements should not be construed te contradict the basic criteria, a Severity IV classification was assigned and is hereby submitted for concurrens.
This position is supported by Enforcement Guidance Memorandum 81-12 dated Februa ry 25, 1981.
9h E. C. McCabe, Jr.
Chief, Reactor Projects Section 2B 4
cc:
- 0. Thompson i
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