ML13330A052
| ML13330A052 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 05/22/1980 |
| From: | Baskin K Southern California Edison Co |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578, TAC-44128 NUDOCS 8006030295 | |
| Download: ML13330A052 (13) | |
Text
Southern California Edison Company P.
- 0.
BOX 800 2244 WALNUT GROVE AVENUE
- ROSEMEAD, CALIFORNIA 91770 K. P. BASKIN May 22, 1980 TELEPHONE MANAGER, NUCLEAR ENGINEERING (213) 572-1401 AND LICENSING Director, Office of Nuclear Reactor Regulation Attention: D. M. Crutchfield, Chief Operating Projects Branch No. 5 Division of Project Management U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Gentlemen:
Subject:
Docket No. 50-206 Implementation of Category "A" Lessons Learned Requirements San Onofre Nuclear Generating Station Unit 1 By letter dated May 2, 1980, you forwarded the NRC staff's evaluation of the implementation of Category "A" Lessons Learned Requirements (excluding 2.1.7.a) at San Onofre Unit 1. You indicated that the evaluation was based on our submittals dated January 17, March 25 and April 11, 1980, documenting the actions taken to implement the Category "A" requirements and the discussions held during the NRC staff site visit on March 10 and 11, 1980. Based on the NRC staff's evaluation, your letter concluded that the implementation of the Category "A" requirements is acceptable except for the open items, 2.1.3.b "Instrumentation for Detection of Inadequate Core Cooling," 2.1. 8.a "Post-Accident Sampling," and NRR "Reactor Cooling System Venting."
In order to resolve these open items, you requested that we respond within 14 days from the receipt of your letter with regard to our intention to meet the requirements specified in the NRC staff's evaluation.
The purpose of this letter is to provide our response to each of the open items discussed above as further amplified in the NRC staff's evaluation forwarded by your May 2, 1980 letter. To facilitate your review, we have organized our responses below such that each "open item" from the NRC staff's evaluation is stated followed by our response.
Open Item -
2.1.3.b "Instrumentation for Detection of Inadequate Core Cooling" (Subcooling Meter)
The licensee must, as a minimum, provide automatic selection of the limiting temperature input and commit to upgrade the system to meet criteria 1-5 above by January 1, 1981, prior to restart from the current refueling outage.
O,
Response
As presented in their evaluation, the NRC staff correctly reflected the manner in which we implemented the NRC Category "A" Lessons Learned Requirement to install a subcooling meter. In addition, they accurately stated that they had informed us at the site visit that upgrading of the installed subcooling meter was nee s ry to meet the established NRC criteria for such instrumentation.
D. M. Crutchfield, Chief
-2 However, the NRC staff's evaluation does not discuss the events which led to the commitments made in our April 11, 1980 letter to upgrade the installed subcooling meter to meet the NRC criteria 1, 2, 3, and 5 on a schedule to be proposed about July 1, 1980, and to examine a method to provide redundancy (criterion 4).
The events were as follows:
- 1. The NRC criteria were first discussed with us during the site visit.
- 2. The nature of the NRC staff's request for additional information/
actions resulting from the site visit concerning the need to upgrade the installed subcooling meter to meet the NRC criteria was identified in the enclosure to the NRC's March 13, 1980 letter.
Specifically, the NRC staff requested that we propose a method (with schedule) to meet each of the NRC criteria.
- 3. In response to the NRC's March 13, 1980 letter, we documented in our March 25, 1980 letter, the basis for the acceptability of the design of our subcooling meter indicating that upgrading to meet the established NRC criteria was not warranted.
- 4. During a telephone discussion with members of the NRC staff on April 8, 1980, the NRC staff indicated that the responses provided in our March 25, 1980 letter were unacceptable. The NRC staff stated that we must commit to upgrade the subcooling meter to meet NRC criteria 1, 2, 3, and 5 by January 1, 1981.
The NRC staff further stated that while NRC criterion 4 was not currently an NRC require ment, it was expected to be established as such soon. Accordingly, the NRC staff suggested that consideration should be given to also meeting this criterion.
- 5. During a telephone discussion with members of the NRC staff on April 10, 1980, we indicated our commitments to upgrade our subcool ing meter to meet NRC criteria 1, 2, 3, and 5 and consider a method of meeting criterion 4.
However, we explained that a completion schedule could not be established until sufficient design, engineer ing and procurement information becomes available approximately July 1, 1980.
Our April 11, 1980 letter confirmed these commitments.
Based on the above events, the commitments provided in our April 11, 1980 letter met the NRC staff's criteria as they were best related to us. At no time following submittal of our April 11, 1980 letter were we contacted by the NRC staff regarding the unacceptability of our commitments as discussed in the NRC staff's evaluation. In particular, we were not advised that the NRC staff had (1) decided that criterion 3 must be implemented prior to restart from the current refueling outage, and (2) established criterion 4 as a requirement which we must commit to prior to restart from the current refueling outage.
Not until discussion with the Staff on April 22, 1980, concerning the final form of the Staff's evaluation, were we made aware that the NRC staff
D. M. Crutchfield, Chief
-3 considered the commitments discussed earlier and documented in our April 11, 1980 letter unacceptable. Based on our review of the NRC staff's evaluation, we believe that our commitments as documented in our April 11, 1980 letter could have been revised to completely meet the NRC staff's criteria and resolve this open item.
In accordance with the NRC staff's evaluation, and notwithstanding our initial understanding that our commitments, as revised, were acceptable to the staff, we are proceeding to provide automatic selection of the limiting temperature input for the installed subcooling meter. The automatic selection capability will be completed prior to restart from the current refueling outage. In addition, we will upgrade the installed subcooling meter to meet criteria 1-5 as delineated in the NRC staff's evaluation. This upgrade will be completed by January 1, 1981, or the Unit shutdown on January 1, 1981 for the completion of this upgrade provided that (1) engineering and procurement in support of this upgrade continue to proceed on the present accelerated schedule in support of that deadline, and (2) system and regional reliability considerations at the time of the scheduled shutdown do not dictate otherwise.
Open Item -
2.1.8.a "Post-Accident Sampling" The licensee should provide a more detailed justification for deferral to SEP with the submittal detailing the proposed plant modificatons by June 1, 1980.
Response
As discussed in our March 25, 1980 letter in response to the NRC staff's March 13, 1980 request for additional information/actions concerning this open item, a preliminary, conceptual P&ID and plot plan describing the proposed sampling station and radiological analysis facility will be available by June 1, 1980.
It is expected that this information together with a descrip tion of the capabilities of the facility can be provided by July 1, 1980. At that time, we will also provide a more detailed justification for deferral to SEP. Earlier submittal of this information, as requested by the staff is not possible based on the need for continuing engineering support during the present refueling outage.
Open Item -
"Reactor Cooling System Venting" The licensee must commit to submit design details by a date which allows NRC review and approval prior to installation by January 1, 1981, as well as additional specific detail to support deferral to SEP if this schedule is sought.
Response
As discussed in our March 25, 1980 letter in response to the NRC staff's March 13, 1980 request for additional information/actions concerning this open item, a preliminary, conceptual P&ID describing the proposed reactor coolant system vents will be available by August 1, 1980.
However, complete design details describing the capabilities of the vents such as the method for
D. M. Crutchfield, Chief
-4 actuating the system and any safety analysis (i.e., loss of coolant accident or containment mixing) required to support the design of the vents will not be available until October 1, 1980. Accordingly, it is expected that the design details can be provided by October 1, 1980. At that time, we will also provide a more detailed justification for deferral to SEP.
We trust that the additional information provided above adequately addresses the NRC staff's concerns, and that our implementation of the Category "A" Lessons Learned requirements for 2.1.3.b, 2.1.8.a and NRR, as modified, can be found acceptable. If you have any questions concerning our responses discussed above, please contact me.
Very truly yours,
Southern California Edison Company P.O. BOX 800 2244 WALNUT GROVE AVENUE ROSEMEAD, CALIFORNIA 91770 May 20, 1980 U. S. Nuclear Regulatory Comission IT Region V Suite 202, Walnut Creek Plaza 1990 North California Boulevard Walnut Creek, California 94596 Attention: Mr. R. H. Engelken, Director DOCKET NO. 50-206 SAN ONOFRE - UNIT 1
REFERENCE:
- 1) Letter dated May 7, 1980 from SCE (J. M. Curran) to NRC (R. H. Engelken)
Dear Sir:
The referenced letter provided prompt notification of an indication in excess of the allowable limits of the ASME B&PV Code,Section XI. The indication was found in the seal injection piping to reactor coolant pump B.
This letter constitutes a follow-up report in accordance with the provisions of Section 6.9.2.a of Appendix A to Provisional Operating License No. DPR-13.
While performing dye penetrant testing during inservice inspection on the seal injection system, a socket weld examination revealed indications in weld No. 2011-61.
These indications were cleared during investigation without weld repair; however,Section XI of the ASME Code required examin ation of additional welds.
During these additional examinations, a 1/4 inch rounded indication in the base material next to the toe of socket weld No. 2009-2 was found.
Investigation of the indication revealed a small pin hole extending about 1/8 inch into the base material. There is no leakage from the defect.
A visual inspection of the pin hole indicates that it is a manufactured flaw, not a service induced defect.
This weld had not been examined in previous inspections.
The defect will be removed and repaired under the requirements of Section XI before returning the reactor coolant system to operational pressure.
QAI
Nuclear Regulatory Commission Page 2 In addition as required by Section XI article IWB-2430 all of the remaining socket welds in this seal injection system have been examined to the extent specified in Table IWB-2500 for the inspection interval.
No additional indications in excess of allowable limits were found.
If you should require additional information concerning this occurrence, please contact me.
Sincerely, H. L. Ottoson Manager of Nuclear Operations
Enclosure:
Licensee Event Peport 80-022 cc:
Director, Office of Inspection & Enforcement (30)
Director, Office of Management Information & Program Control (3)
Director, Nuclear Safety Analysis Center (1)
NHC FOHM 366 U. S. NUCLEAR REGULATORY COMMISSION (7-77)
LICENSEE EVENT REPORT CONTROL BLOCK:
1_
(PLEASE PRINT OR TYPE ALL REQUIRED INFORMATION) 6
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8 60 61 DOCKET NUMBER 68 69 EVENT DATE 74 75 REPORT DATE 80 EVENT DESCRIPTION AND PROBABLE CONSEQUENCES During inservice inspection of reactor coolant pu.p Loop B seal injection piping a rounded indication in excess of allowable limits was found. Investigation of the indication revealed a base material flaw extending into the minimum wall.
There was no leakage present and nor an adverse effect on public health and safety.
7 8 9 80 SYSTEM CAUSE CAUSE COMP.
VALVE CODE CODE SUBCODE COMPONENT CODE SUBCODE SUBCODE 0I9 IWIB I@ LBJ@ LBJ@ IP II IP IE IX IX IO L
6 7
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CODE TYPE NO F I 8 0 1 L-1 0 1 2 1 2 1 L***.J 10 11 1 L'J 1i L 21 22 23 24 26 27 28 29 30 31 32 ACTION FUTURE EFFECT SHUTDOWN ATTACHMENT NPRD-4 PRIME COMP.
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[Investigation revealed a small pin hole extending below the minimum wall of the I ipe.
This pin hole is an original fabrication flaw in the base material of the pipe, and is not weld related.
The pin hole will be repaired under a Section E I
repair program before the end of the current refueling outage.
7 8
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% POWER OTHER STATUS DISCOVERY DISCOVERY DESCRIPTION ID LJ@
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10 68 69 80 NAME OF PREPARER M.
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Southern California Edison Company P.O. BOX 800 2244 WALNUT GROVE AVENUE ROSEMEAD. CALIFORNIA 91770 May 20, 1980 U. S. Nuclear Regulatory Commission Region V Office of Inspection and Enforcement Suite 202, Walnut Creek Plaza 1990 North California Boulevard Walnut Creek, California 94596 Attention:
Mr. R. H. Engelken, Director DOCKET No. 50-206 SAN ONOFRE - UNIT 1
Dear Sir:
This letter describes a reportable occurrence involving the failure of portions of the Nuclear Instrumentation System.
Submittal is in accordance with the reporting requirements stipulated in Section 6.9.2(b) of Appendix A to the Provisional Operating License DPR-13.
On April 20, 1980 at 2130 during the preliminary steps of the refueling process after the refueling canal had been flooded, Source Range Nuclear Instrumentation channel 1202 was observed to be indicating erratically.
At 2345 on the same day Source Range Nuclear Instrumentation channel 1201 high voltage failed.
Although an installed spare Source Range Nuclear Instrument ation channel and two intermediate range channels continued to provide visual indication of neutron level, audible indication required by Section 3.8.A.2 of the Technical Specifications was not in service.
In accordance with the requirements of echnical Specification 3.8.A.7 refueling operations were stopped and repairs initiated. At 0925 on April 21, 1980 the neutron detectors from the spare channel were connected to channel 1201 restoring audible indication.
Mr. R. H. Engelken aMty 20, 1980 Page 2 Concurrent with the failure of channel 1201 a Reactor Cavity Hi Level Alarm had been received.
Subsequent investigation revealed that leakage had occurred through the reactor vessel to cavity seal flange allowing water to enter the channel 1201 and 1202 neutron detectors and preanps. The refueling canal was drained and the source range channels 1201 and 1202 repaired on April 23, 1980. After completion of repairs refueling operations were resumed.
Should you have any questions regarding this matter, please contact me.
Sincerely, H. L. Ottoson Manager of Nuclear Operations cc:
Director, Office of Inspection and Enforcement (30)
Director, Office of Management Information & Program Control (3)
Director, Nuclear Safety Analysis Center (1)
NRC FORM 366 U. S. NUCLEAR REGULATORY COMMISSION 77.LICENSEE EVENT REPORT CONTROL BLOCK:
(PLEASE PRINT OR TYPE ALL REQUIRED INFORMATION) 1 6
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8 60 61 DOCKET NUMBER 68 69 EVENT DATE 74 75, REPORT DATE 80 EVENT DESCRIPTION AND PROBABLE CONSEQUENCES During refueling operations water leaked through the reactor vessel to refueling canal flange seal causing failures in source range nuclear instrument channels O 4-11201 and 1202. Audible neutron flux indication required by T.S. 3.8.A.2 was lost. Refueling activities were stopped and repair work initiated.
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33 34 35 36 37(B 40 41 42 43 44 47 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS l Leakage occurred through the reactor vessel to refueling canal flange.
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7 8
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Southern California Edison Company P.O. BOX 800 2244 WALNUT GROVE AVENUE ROSEMEAD. CALIFORNIA 91770 May 20, 1980 U. S. Nuclear Regulatory Camission Region V Suite 202, Walnut Creek Plaza 1990 North California Boulevard Walnut Creek, California 94596 Attention: Mr. R. H. Engelken, Director DOCKET NO. 50-206 SAN ONOFRE -
UNIT 1
REFERENCE:
- 1) Letter dated May 7, 1980 from SCE (J. M. Curran) to NRC (R. H. Engelken)
Dear Sir:
The referenced letter provided prompt notification of an indication in excess of the allowable limits of the ASME B&PV Code,Section XI. The indication was found in the seal injection piping to reactor coolant pump B.
This letter constitutes a follow-up report in accordance with the provisions of Section 6.9.2.a of Appendix A to Provisional Operating License No. DPR-13.
While performing dye penetrant testing during inservice inspection on the seal injection system, a socket weld examination revealed indications in weld No. 2011-61.
These indications were cleared during investigation without weld repair; however,Section XI of the ASME Code required examin ation of additional welds. During these additional examinations, a 1/4 inch rounded indication in the base material next to the toe of socket weld No. 2009-2 was found.
Investigation of the indication revealed a small pin hole extending about 1/8 inch into the base material.
There is no leakage from the defect.
A visual inspection of the pin hole indicates that it is a manufactured flaw, not a service induced defect.
This weld had not been examined in previous inspections.
The defect will be removed and repaired under the requirements of Section XI before returning the reactor coolant system to operational pressure.
Nuclear Regulatory Commission Page 2 In addition as required by Section XI article IWB-2430 all of the remaining socket welds in this seal injection system have been examined to the extent specified in Table IWB-2500 for the inspection interval.
No additional indications in excess of allowable limits were found.
If you should require additional information concerning this occurrence, please contact me.
Sincerely, H. L. Ottoson Manager of Nuclear Operations
Enclosure:
Licensee Event Feport 80-022 cc:
Director, Office of Inspection & Enforcement (30)
Director, Office of Management Information & Program Control (3)
Director, Nuclear Safety Analysis Center (1)
NRC FORM 366 U. S. NUCLEAR REGULATORY COMMISSION (7-77)
LICENSEE EVENT REPORT CINTROL BLOCK:
(PLEASE PRINT OR TYPE ALL REQUIRED INFORMATION)
I lIC IA IS 10i IS16121 I
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8 60 61 DOCKET NUMBER 68 69 EVENT DATE 74 75 REPORT DATE 80 EVENT DESCRIPTION AND PROBABLE CONSEQUENCES O 2 I During inservice inspection of reactor coolant pump Loop B seal injection piping a rounded indication in excess of allowable limits was found. Investigation of the indication revealed a base material flaw extending into the minimum wall.
There was no leakage present and nor an adverse effect on public health and safety.
1E7161 1 I
7 8 9 80 SYSTEM CAUSE CAUSE COMP.
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-47 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS 0 I Investigation revealed a small pin hole extending below the minimum wall of the I
pipe. This Pin hole is an original fabrication flaw in the base material of the I
I pipe, and. is not weld related. The pin hole will be repaired under a Section I I
1 3 repair program before the end of the current refueling outage.
7 89 80 FACILITY METHOD OF STATUS
%POWER OTHER STATUS 0
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068 69 80 NAME OF PREPARER J.
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PHONE:714)42-780