ML20214Q901
| ML20214Q901 | |
| Person / Time | |
|---|---|
| Issue date: | 05/28/1987 |
| From: | Hodges M Office of Nuclear Reactor Regulation |
| To: | Merschoff E Office of Nuclear Reactor Regulation |
| References | |
| REF-QA-99900403 TAC-65230, NUDOCS 8706050251 | |
| Download: ML20214Q901 (4) | |
Text
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My 2 81987 Docket No. 99900403/86-03 MEMORANDUM FOR:
Ellis W. Merschoff, Acting Chief Vendor Inspection Branch Division of Reactor Inspection and Safeguards FROM:
M. Wayne Hodges, Chief Reactor Systems Branch Division of Engineering & Systems Technology
SUBJECT:
SLOW CONTROL ROD MOTION INITIATION EXPERIENCED AT THE VERMONT YANKEE NUCLEAR POWER STATION DUE TO FAILURE OF SCRAM S0LEN0ID PILOT VALVES (TAC # 65230)
References:
1.
Memorandum from E. W. Merschoff to M. W. Hodges, dated April 10, 1987.
2.
Memorandum from M. W. Hodges to R. F. Heishman, dated December 16, 1987.
3.
Memorandum from R. F. Heishman to M. W. Hodges, dated December 3, 1986.
4.
Information Notice No. 85-13: " Consequences of Using Soluble Dams", February 21, 1985.
In your memorandum of April 10, 1987 (Ref. 1) on the same subject, you requested that the Reactor Systems Branch inform you of the acceptability of a GE statement that a 12 to 16 second slow scram in 119 of 121 control rod drives in a BWR could not result in a substantial safety hazard. We had previously provided ycu with a response (Ref. 2) on this matter and concluded that the GE position was not acceptable. Our position was based on the information we reviewed at the time (Ref. 3) from which we concluded that, for design basis transients, the slow scram of 119 of 121 control rods would result in failure of the fuel cladding due to local overheating. This potential fuel cladding failure would come within the 10 CFR Part 21.3(k) definition of substantial safety hazard as
"...a major reduction in the degree of protection provided to the public health and safety...".
We have reviewed the new information that you have compiled and transmitted by Reference 1.
This information includes proprietary GE documents on two (in GE's terminology) Potential Reportable Conditions (PRC):
(a) PRC No. 86-09 on Coritact:
D. Fieno, SRXB 27141 8706050251 B70528
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99900403 PDR
Ellis W. Merschoff Control Drive Scram Anomaly and (b) PRC No. 84-62 on Monticello CRD Inner Filter. PRC No. 86-09 is an evaluation of six out of eighty-nine control rod drives at Vermont Yankee that failed the single rod scram test during planned testing performed at vessel hydrostatic pressure. The failures have been attributed to the installation in the control rod drive hydraulic control unit of scram solenoid pilot valve rebuild kits that were defective. PRC No. 84-62 is an evaluation of incorrect control rod drive movable inner filters that were shipped by GE to the Monticello nuclear plant. At Monticello, 119 out of 121 control rod drives experienced degraded scram insertion times caused by movable inner filter plugging which was exacerbated by the installation of the incorrect filters on some control rod drives. However, both mesh size filters, the incorrect 2 mil and the correct 10 mil filters, can become plugged. Evidently, a significant quantity of foreign matter remained in the reactor after major portions of the recirculation piping had been replaced (see Information Notice No. 85-13 (Ref. 4)). Both of these PRCs are concerned with degraded control rod drive scram insertion performance.
It is noteworthy that both of these control rod drive scram time insertion anomalies were discovered by required Technical Specification scram time testing and subsequently rectified prior to power operation.
We have reviewed, in particular, the analysis used by GE to conclude that there is no substantial safety hazard when 119 out of 121 control rod drives have degraded scram insertion times.
In this analysis, GE assumed that all of the control rods (121) had degraded scram insertion times. The degraded scram insertion times were based on test data such that the test data were bounded at a 95% probability at a 95% confidence level.
For the limiting load rejection without bypass transient, the minimum critical power ratio (CPR) was calculated to be significantly below the plant's safety limit CPR. GE stated that a few percent of the fuel rods would be subject to boiling transition. Such rods are assumed to fail and would result in a radiological release to the primary coolant corresponding to a margin factor greater than 10 to 10 CFR Part 100 guideline levels. Based on this potential radiological consequence for the limiting transient, GE concluded that the slow scram of 119 out of 121 control rods would not constitute a substantial safety hazard.
The slow scram insertion times of 119 out of 121 control rods could, by GE's calculations, result in the violation of the Technical Specification Safety Limit CPR in the event that the limiting load rejection without bypass transient occurred. The violation of the safety limit CPR would result in violation of GDC 10 which states:
"The reactor core and associated coolant, control and protection system shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences."
The violation of a Technical Specification Safety Limit would require, per 10 CFR 50.36(c)(1), the following action:
Ellis W. Merschoff "...If any safety limit is exceeded, the reactor shall be shut down. The licensee shall notify the Commission, review the matter and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence. Operation shall not be resumed until authorized by the Connission."
The purpose of the Technical Specification Safety Limit CPR is to protect the fuel cladding from overheating and possibly failing. Additionally, the Safety Limit CPR meets, in part, the requirements of GDC 10. The intent of the Technical Specification Safety Limit CPR and GDC 10 is to protect the first barrier to fission product release. On this basis, we believe that a substantial safety hazard, as defined in 10 CFR Part 21.3(k), would exist in that there would have been a loss of safety function to the extent that a major reduction in the degree of protection provided to the health and safety of the public could occur. The seriousness of a breach in this barrier is reflected in the 10 CFR Part 50.36(c)(1) requirement that a plant be imediately shutdown under these conditions. The GE position that there would be no substantial safety hazard is based solely on dose calculations which indicated that doses would be well within 10 CFR Part 100 guideline limits.
Our position is, however, that an operating configuration which would result in a violation of a safety limit during an anticipated operational occurrence is of itself a substantial safety hazard.
In sumary, we disagree with the GE assertion that a BWR, with 119 out of 121 control rod drives having slow scram insertion times, would not constitute a substantial safety hazard. On thescontrary, we conclude that such a degraded scram insertion time would invalidats the safety analysis and could result in -
the Technical Specification safety limit CPR being violated, as shown by the GE calculations, thereby, resulting in a substantial safety hazard.
In addition to the specific question concerning a substantial safety hazard, we note that some other items that need to be considered in this appraisal of slow scram insertion times. These are:
(1) Anticipated transients other than a load rejection without bypass transient could also lead to violation of the Technical Specification Safety Limit CPR in the event that they occurred.
(2) Parts of the safety analysis for the cycle in question would not be valiJ and, therefore, an unreviewed safety question would exist.
Ellis W. Merschoff 4
(3) The possibility exists that the defects that cause slow scram insertion times are in themselves reportable per 10 CFR Part 21 (see Part 21.3(d)(4)). These defects include defective scram solenoid pilot valve rebuild kits and incorrect movable filters installed in the control rod drives.
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