IR 05000334/1980001
| ML19312E110 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 04/02/1980 |
| From: | Beckman D, Mccabe E, Rhoads G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19312E106 | List: |
| References | |
| 50-334-80-01, 50-334-80-1, NUDOCS 8006030268 | |
| Download: ML19312E110 (29) | |
Text
,
..
o
-
.
i U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF.TNSPECTION AND ENFORCEMENT
REGION I
Report No.
50-334/80-01 Docket No.
50-334 License No.
DPR-66 Priority Category C
--
Licensee:
Duquesne Light Company 435 Sixth Avenue Pittsburgh, Pennsylvania 15219 Facility Name:
Beaver Valley Power Station, Unit 1 Inspection At:
Shippingport, Pennsylvania _
l Inspection Conducted:
January 14 - February 8, 1980 Inspectors:
[
[
V - 1 - 8 'O D. A. Beckman, Resident Mspector date b
4-2-TO G. G. Rhoads, Reactor Inspector date h
_ f.2- 0 0 P. D. Graham, Reactor Inspettor date Approved by:
d k
4-2.-fo E. C. McCabe, ' Chief, Reactor Projects date Section No. 2, RO&NS Branch Inspection Summary:
Inspections on January 14 - February 8, 1980 (Inspection Report No. 50-334/80-01)
Areas Inspected:
Routine inspections by the resident inspector and region-based inspectors (129 hours0.00149 days <br />0.0358 hours <br />2.132936e-4 weeks <br />4.90845e-5 months <br />) of:
Action on previous inspection findings; plant opera-tions; refueling preparations; refueling operations, IE Bulletin followup; licens-ee event review; review of periodic reports; and review of physical protection /
,
plant security.
l
'
Results:
Four items of noncompliance were identified:
(Deficiency - Inadequate implementation of administrative controls for jumpers / lifted leads, paragraph 3; Infraction - Radiation Monitor setpoints set in excess of allowable TS limits, paragraph 3; Infraction - Failure to follow Chemistry Manual Procedures, para-graph 4; and, Deficiency - Failure to calibrate instruments as required, para-graph 3).
Region I Form 167 (August 1979)
8006030 267
_ __ __ _ _ ________
.
.
-
.
_ DETAILS l
1.
Persons Contacted R. Balcerek, Nuclear Engineering and Refueling Supervisor J. Carey, Director of Nuclear Operations W. Glidden, QA Engineer R. Hansen, Maintenance Supervisor J. Kosmal, Radcon Supervisor E. Kurtz, Senior QA Engineer A. Lonnet, Associate Engineer R. Prokopovich, Reactor Engineer L. Schad, Operations Supervisor E. Schnell, Radcon Foreman J. Sieber, Superintendent, Licensing and Compliance J. Werling, Station Superintendent D. Williams, Results Coordinator H. Williams, Chief Engineer Other licensee personnel were also interviewed.
2.
Licensee Action on Previously Identified Inspection Findings (Closed) Unresolved Item 79-17-02:
Revise applicable section of BVPS Opera-ting Manual (0M) to reflect Design Change Package No.139 for removal of MS Valve Cubicle Rupture Discs.
The inspector reviewed the following revi-sions to the OM and determined that the subject revisions had been made:
Section 1.SO.3, Checklist 8 and E, Revision 10; Section 1.45.3, Revision 4; Section 1.50.3, Station Startup Normal fystem Arrangement, Revision 10.
(Cit 3ed) Unresolved Item 79-17-09:
Revise procedures affected by design change which eliminated cooling water valves on 1A Charging Pump.
Inspector review of revisions to OM Section 1.30.3, Revision 6, and OST 1.7.4, Revision 14 found the licensee's actions acceptable.
(0 pen) Infraction 79-17-01:
Inadequate procedure implementation for locked valve control, ECP Calculations, and OST reviews.
Inspector raview of the currective and preventive actions specified in the DLC leti.er of October 20, 1979 showed that action to prevent recurrences was still in prog ess.
Additionally, the records of corrective action implementation with regard to a letter of direction provided to operations personnel were not available.
This item remains open pending completion of the above preventive actions and further NRC review.
(0 pen) Infraction 79-17-07:
Failure to implement / utilize approved procedures for maintenance on valve 1-SI-29.
Inspector review confirmed the corrective action specified in the DLC letter of October 20, 1979 had been implemented as state _ _ _
.
.
-
.
The licensee's letter stated that the event was discussed with Shift
-
Supervisors and responsible maintenance personnel to preclude future operations of the type discussed by the item of noncompliance.
Inspector interview of personnel confirmed that the discussions had been conducted with Shift Supervisors as stated but did not confirm a similar discussion for affected maintenance personnel.
The licensee has not maintained documentation of the discussions.
This item remains open pending discus-sions with the above maintenance personnel to confirm implementation of preventive actions.
The inspectors also discussed the need to maintain documentation of the type referenced above at the exit meeting held on February 13, IS80, stating that such documentation is required by 10 CFR 50, Appendix B, Criterion XVI and the OLC Operations Quality Assurance Program.
Although the inspectors were able to otherwise confirm portions of the specified preventive actions, the licensee must document their accomplishment ir.
accordance with the requirerents above.
The Superintendent stated that action would be taken to provide such documentation for future station activities. This matter will be reinspected during subsequent inspections.
(80-01-01)
3.
Review of Plant Operations a.
Inspection tours of selected plant areas were conducted on the dates noted during the day shift with respect to housekeeping and cleanli-ness, fire protection, radiation control, and Technical Specification compliance.
,
b.
Areas Toured Control Room (January 16, 22, 24, 28, February 4, 1980);
--
Primary Auxiliary Building, 735', 752' and 768' levels, except
--
locked High Radiation Areas and Loose Surface Contamination Areas (January 24, 1980);
Main Steam Valve Room (January 22, 24, 1980);
--
Vital Switchgear Rooms (January 22, 24, 1980);
--
,
Purge Duct Room (January 22, 24, 1980);
--
Blowdown Tank Room (January 22, 24, 1980);
--
j East Cable Vault (January 22, 24, 1980);
--
Main Plant CO 10 Ton Storage Unit Room (January 24, 1980);
--
Diesel Generator Rooms (January 24, 1980);
--
.
.
-
.
Containment Building, all levels, except locked High Radiation
--
Areas and Loose Surface Contamination Areas (January 24, February 4, 1980); and, Fuel Handling Building (January 24, and February 4, 1980)
--
In addition, the inspectors visited the control roor on a daily basis during the normal work week and made general ebservations of activities in progress and plant status.
c.
Observations (1) Control room monitoring instrumentation was observed to verify that instrumentation and systems required to support Mode 6 operation were in conformance with Technical Specification (TS)
and Limiting Condition for Operation (LCO) requirements.
The following instrumentation / indications were observed with respect to the LC0's indicated:
Boric Acid Storage Tank Level and TS 3.1.2.7a
--
Temperature (January 22, 1980)
Refueling Water Storage Tank Level TS 3.1.2.7.b
--
and Temperature (January 22, 1980)
.
Instrumentation Operability (Channel TS 3.3.11
--
Checks) Source Range Neutron Flux -
(January 22, 24, 1980)
Radiation Monitoring Channels (Channel TS 3.3.3.1
--
Checks) (January 22, 24, February 4, 1980 - RM-VS-104A a B, RM-VS-215A & B)
Residual Heat Removal Flow TS 3.9.8
--
(January 22, 24, February 4, 1980)
Reactor Cooling System Baron TS 3.9.1
--
Concentration (January 22, February 5, 1980)
AC/DC Electrical System Availability and TS 3.8.1.2,
--
Distribution (January 22, 1980)
3.8.2.2 and 3.8.2.4
,
(2) Radiation controls established by the licensee, including posting of radiation areas, the conditions of step off pads, disposal of protective clothing, compliance with radiation work I
'
.
.
.
-
.
permits, personnel monitoring devices being worn, cleanliness of work areas, radiation control job coverage, area monitor operability (portable and permanent), area monitor calibration, and perso el frisking procedures were observed in the following areas.
Locked barrierc,t High Radiation Areas were also confirmed to be secure.
The inspector confirmed the above activities to be in accordance with radiation work permits posted in the respective areas on the dates of inspection.
Primary Auxiliary Building, 735', 752' and 768' levels
--
(January 24, 1980);
Containment Airlock area (January 24, February 4,1980);
--
Fuel Handling Building Access (January 24, February 4,
--
1980); and, Containment Building, all levels, (January 24, and February 6,
--
1980).
(3) Plant housekeeping conditions including general cleanliness
,
!
conditions and control of materials to prevent fire hazards were observed in the area listed in paragraph 6 above.
Maintenance of fire barriers and fire barrier penetrations in these areas was also observed.
(4) Control room manning was observed on the dates noted in para-graph 6 above and during other periodic control room visits.
(5) On January 22, 1980, the inspector witnessed the performance of Maintenance Surveillance Procedure (MSP) 2.09, Nuclear Instrument Source Range N31 Calibration, Revision 1, Step VII.I.21 and Section K, and OM 1.2.4, Reaction Excore Instrumentation, Revision 9, Section 4.J, NIS Source Range Attenuator, Discriminator and High Voltage Adjustment.
The procedures were performed by instrument technicians under the direction of the Reactor Engineer in r'sponse to electrical noise spikes induced by e
operating the reactor vessel stud tensioning hoists (3) in containment., The instrument was properly placed in test and the inspector verified TS compliance for redundant equipment requirements.
The digital voltmeter was verified to be in calibration and the procedure completed satisfactorily.
After the instrument was declared operable, the licensee operated the tensioning hoists and noted that the noise spikes (100 cps indicated on NTE SR-NI) persisted.
The licensee elected to continue the de-tensioning procedure in that the noise spikes had been positively identified.
The licensee established continuous communications between containment and the control room to alert the reactor operator to hoist operation.
t I
.
.
.
'
.
Except as further discussed below, the inspection findings were
%
acceptable.
d.
Findinas (1) On January 23, 1980 the inspector verified the alarm / trip set-points for radiation monitoring instruments required by TS 3.3.3.1 for Mode 6 plant operation.
The following setpoints were noted:
T.S. Limit Actual Setting RM-215B (Containment Gaseous 1 7.3 x 102 cpm *
8 x 10s cpm Activity)
RM-VS-104A
< 1.6 x 103 cpm *
9.25 x 103 cpm
-
(Containment Area Monitor)
RM-VS-1048
$ 1.6 x 103 cpm *
9.34 x 103 cpm
!
(Containment Area l
Monitor)
- Counts above background.
Background for RM-215B was approximately 200 cpm and for RM-VS-104A and B were approximately 100 cpm and approximately 200 cpm respectively.
Failure to set the alarm / trip setpoints of the above radiation monitors to the values required by the referenced Technical Specifi-cations constitutes an item of noncompliance. (80-01-02)
The 'icensee immediately reset the alarm / trip setpoints of the affected radiation monitors to comply with the TS requirements.
The redet values were confirmed to be correct by the inspector on January 23,
,
1980.
Ti? licensee additionally initiated Operating Manual Change Notice (OMCN) No. 80-15 which revised Operating Surveillance Test
(OST) 1.49.3, Refueling Operations Prerequisites, to require checking and resetting of the subject instruments when required by operating mode changes.
The inspector confirmed that OMCN No. 80-15 was issued and in effect on January 24, 1980.
I With respect to the oversight which resulted in failure to reset the instrument setpoints, the licensee stated that the requirement had apparently been overlooked during the preparation, review and approval of OT 1.49.3.
The DLC response to this item of noncompliance should address the causes and corrective action for this oversight.
rm
-.
.
-
.
(2) On January 24, 1980, the inspector toured the Diesel Generator Rooms and the 10 Ton Cardox Storage Unit Room and noted the fol-lowing gages had exceeded their specified calibration period:
PI-63-FPC2 Cardox Pressure - Main Cardox Fire Suppression Unit LI-FP-202 Cardox Level Indicator - Main Cardox Fire Sup-pression Unit PI-EE-205 EDG Fuel System 1 Filter In PI-EE-206 EDG Fuel System 1 Filter In i
l PI-EE-207 EDG Fuel System 2 Filter In PI-EE-208 EDG Fuel System 2 Filter In PI-EE-209 EDG Lube Oil Engine PI-EE-210 EDG Lube Oil Engine PI-EE-211 EDG Lube Oil Filter PI-EE-212 EDG Lube Oil Filter Each of the above instruments was noted to have a required cali-bration due date of November 1979.
The instruments associated with Cardox Unit pressure and level are directly utilized for verification of Cardox Unit operabilitiy required by TS Surveil-lance Requirement 4.7.14.3.
The instruments associated with the No. 1 and 2 EDG's are utilized to confirm proper engine and engine auxiliary system operation during performance of Operating Surveil-lance Tests (OST) 1.36.1 and 1.36.2, Diesel Generator No. 1 and 2, Monthly Tests (respectively), Revision 15.
OST 1.36.1 and 1.36.2 are performed pursuant to TS 4.8.1.7 and 4.8.1.1.2.
Data from the EDG instruments is recorded in the OST and each have acceptable ranges of indicated parameters specified by the OST.
The inspector further noted that all of the above instruments, except Cardox Level Indicator LI-FP-202, are incorporated in the calibration program requirements of the BVPS Maintenance Manual (MM), Chapter 1, Section 0, Calibration Program, Revision 4.
10 CFR 50, Appendix B, Criterion XII, the BVPS FSAR, Section A.2.2.12, Control of Measuring and Test Equipment; Quality Assur-ance Procedure No. OP-12, Control of Measuring and Test Equipment, Revision 3, Section 12.2.3; and, the BVPS MM, Chapter 1, Section 0,
_ _ _
_.__
__
.
.
.
Calibration Program, Revision 4, require that all measuring and test equipment be calibrated on or before its due date.
Similarly, the above documents require that instruments utilized to quantita-tively verify operability of a safety-related system be cali-brated in accordance with the requirements of the 0QA program.
Contrary to the above requirements, the instruments listed above were not calibrated on or before their published due dates.
Additionally, Cardox Level Indicator LI-FP-202 is not identifed by the BVPS MM, Chapter 1, as requiring periodic calibration in accordance with an approved procedure.
Failure
'
to implement administrative controls for the calibration of measuring and test equipment in accordance with the requirements above is an item of nonccmpliance. (80-01-03)
(3) During a control room tour on January 16, 1980, the inspector noted an alligator clip jumper installed on RCS Temperature Recorder T-RC-4488, Terminals 9(+) to 9(-), but found no jumper tag or Jumper and Lifted Lead Log entry which identified or authorized the jumper installation.
Subsequent discussion and review by the inspector with the Operations Supervisor and Instrument Foreman determined that:
Although the recorder nor its equipment were not in service
--
during the inspection, the recorder had been administratively returned to service with all outstanding Maintenance Work Requests and Equipment Clearance Permits applicable to the jumper installation having been closed.
The inspector was unable to determine the existence of any
--
control which would have precluded operation of the equipment with the jumper installed or which would have identified its installation to operations or maintenance personnel.
Failure to properly tag and log or remove a temporary jumper from T-RC-448B is contrary to TS 6.8.1, Regulatory Guide 1.33, and the BVPS OM, Section 1.48.5.0.1 and is an item of noncompliance (80-01-04).
The licensee stated that the subject recorder is not categorized as safety-related and is subject to the OM manual controls only on an optional basis.
The~ inspector acknowledged the licensee's comments and stated that the above recorder includes instrument channels which monitor RC Pump parameters which are used to identify (via alarm) and diagnose loss of Reactor Coolant Pump seal injection flow and/or seal failure and thereby constitute an instrument important to safety which should be appropriately controlled.
(4) During inspection of the Control Room on January 16, 1980, the inspector concluded that the area behind the vertical section of the Main Control Board (MCB) was in need of general cleaning.
.
_ _,
.
_ _
_. _ _
.
_
_ _.
.
.
.
i
.
'x
l
.
x N
N.
,
The inspector noted that the cleanliness conditions were mar-ginally acceptable in that a heavy dust accumulation was present i
on the floor, metallic and nonmetallic trash had accumulated on i
the floor, several box and panel covers had apparently been removed for maintenance and not reinstalled, and dust buildup
,
was accumulating on exposed terminal strips and equipment.
I Although no immediate safety hazard was identified, the inspector discussed the potential for future accumulation causing a safety hazard with the Operations Supervisor and Instrument Foreman on January 16, 1980.
During a subsequent conversation
on February 4, 1980, the Operations Supervisor stated that the area will be cleaned pending the procurement of suitable, nonconducting cleaning equipment.
The completion of MCB cleaning will be followed during future inspections. (80-01-05)
4.
Plant Chemistry
On February 7,1980 the inspector witnessed taking of a primary coolant sample (RHR) and performance of the required analyses.
The following j
documents were also reviewed:
BVPS - Chemistry Manual, Chapter 3, Sampling and Testing Part 7,'
--
Revision 1; and, BVPS - Chemistry Manual, Chapter 4, Analytical Methods - Borox
--
Titrimetric, pH, Conductivity and 15 Minute Degassed Gross Activity,
,
Revision 8.
'
,
l
Except as discussed below, inspection findings were acceptable.
Upon completing the primary sample, the plant chemist secured the operation
i by manipulation of one valve.
Chapter 3 of the Chemistry Manual requires a complete isolation of the sample flow path and provides valve-by-valve alignment instructions.
When the inspector discussed this discrepancy
I with the chemist, it was determined that the sample sink apparently was never secured in accordance with the procedure, but left in a continuous purge line-up between the Reactor Coolant System and the Volume Control Tank, even during reactor operation.
Discussions with Chemistry Department supervision indicated that this arrangement was intended to permit expedi-tious sampling under abnormal conditions by minimizing the setup and recirculation time required prior to actually drawing a coolant sample.
Station supervision, including the Results Coordinator and Station Super-intendent, acknowledged that the above alignment was contrary to the requirements of the Chemistry Manual and that manual revision would be required to permit such operation.
The inspector stated that any revisions
.
permitting continuous recirculation of coolant through the sample sink
'
'
should be reviewed with respect to post-accident coolant activity and general area radiation levels to ensure that such alignment would be
'
desirable for all cases.
The licensee acknowledged the inspector's comments.
.
. -,
-
-
.,.-.-.y
-.r.
-..,
,,.
..
,... - -
- - -, -..,
. _..,...
w,,r-
, -. _. ~., - -..
.
.
Failure to properly implement the return to normal alignment provisions of chemistry procedures is in noncompliance with TS 6.8.1.a, Regulatory Guide 1.33, and the BVPS Chemistry Manual, Chapter 3, Revision 7. (80-01-06)
5.
Refueling Preparation a.
New Fuel Receipt and Inspection The inspector resiewed the licensee's procedure for new fuel assembly receipt and inspection against ANSI N18.7-1972 and ANSI N45.2.2-1972.
Receipt and inspection records for 52 new fuel assemblies were reviewed.
Minor discrepancies noted on the fuel inspection records were discussed with the licensee.
All discrepancies had been satis-factorily dispositioned by the licensee in accordance with QA program requirement.
No unacceptable items were noted.
The documents reviewed are noted in paragraph b. belo..
b.
Documents Reviewed / Refueling Prerequisites and Preparations The inspector reviewed the documents listed below to establish that the licensee had technically acceptable and approved procedures, meeting the requirements of TS and ANSI N18.7-1972, for current and planned refueling activities.
The inspector confirmed the availability and adequacy of procedures for fuel handling, transfers, and core verification; inspection of fuel to be reused; handling and inspection of core internals, and establishment of prerequisite conditions.
NSQC 10.2, Revision 1, Fuel Assembly and Shipping Container
--
Receipt Inspection, dated May 5, 1975.
Temporary Operating Procedure, 80-5, Fuel Shuffle Prerequisites
--
Check List, dated January 18, 1980.
OST 1.49.3, Revision 7, Refueling Operations Prerequisites.
--
OST 1.47.3, Revision 23, Containment Integrity Checklist for
--
Refueling.
OST 1.16.4, Revision 0, Fuel Building Ventilation System Veri-
--
fication - Fuel Movement.
,
'
OST 1.7.8, Revision 10, Boric Acid Storage Tanks and RWST Level
--
and Temperature Verification.
OST 1.7.4, Revision 14, Centrifugal Charging Pump (1CH-P-1A).
--
OST 1.7.3, Revision 12, Boric Acid Transfer Pump Operational
--
Test.
l
_
.
_
....
.
l
.
.
OST 1.20.4, Revision 0, Spent Fuel Pool Level Verification.
--
OST 1.43.4, Revision 7, Radiation GM Area Monitor RM-1RM-207
--
Functional Test.
OST 1.44C.4, Revision 1, Containment Purge and Ventilation
--
Test.
OST 1.11.10, Revision 18, Boron Injection Flow Path Power Ope-
--
rated Valve Exercise.
OST 1.11.3, Revision 18, Boron Injection Flow Path Valve Posi-
--
tion Verification.
FP-DLW-R1, Section 9.2.13, Refueling Equipment Maintenance
--
Checkout, dated December 7, 1979.
FP-DLW-R1, Section 9.2.14, Refueling Equipment Operational
--
Demonstration, dated December 7, 1979.
FP-DLW-R1, Section 9.2.15, Fuel Shuffle, dated December 7,
--
1979.
Fuel Handling Instruction, F-1, Handling of New Fuel Assemblies
--
and RCC Elements, Revision 0, dated December 1,1971.
Fuel Handling Instruction, F-3, Fuel Assembly and RCC Element
--
Replacement, Revision 0, dated December 1, 1971.
Fuel Handling Instruction, F-4, Site Removal of Fuel Assemblies
--
from Shipping Containers and Handling of Shipping Containers, Revision 3, dated November 11, 1974.
Fuel Handling Instruction, F-5, Instruction, Precautions and
--
Limitations for Handling New and Partially Spent Fuel Assemblies, Revision 5, dated September 10, 1979.
E-22, Irradiated Fuel Damage While Refueling, Revision 16.
--
Technical Specification, Section 3/4.
--
Except as noted below, the inspector had no further questions in this area.
The licensee's refueling procedures did not contain any provision for reverifying prerequisites after an interruption in fuel handling as required by paragraph 5.3.4.5 of ANSI N18.7.
On February 5, 6, and 8, discussions were held with the Plant Superintendent, and the
.
.
.
s
'
Nuclear Engineering and Refueling Supervisor and his staff.
During these discussions plant management committed to incorporate the required provisions in the subject procedures.
The inspectors confirmed that, for prior minor interruptions in refueling activities, all applicable prerequisites had been reverified and logged in either the refueling procedures or the Shift Supervisor's S-1 series log. This item is unresolved penaing licensee action and inspector review. (50-334/80-01-02)
6.
Refueling Activities In order to support reactor vessel inservice inspection, the licensee elected to completely defuel the reactor, storing all fuel in the Spent Fuel Pool.
On February 4,1980 the inspector observed the licensee remove fuel from the core to the spent fuel pool.
The following were observed:
Core monitoring was in accordance with TS 3.9.2;
--
Boron concentration was in accordance with TS 3.9.1;
--
The individual directing fuel handling activities held a senior
--
operator license, had no other duties, and was constantly present; Containment integrity was maintained as required by TS 3.9.4.a,
--
3.9.4.b and 3.9.4.c.1.
Radiation control personnel were present;
--
Vessel water level was in accordance with TS 3.9.10;
--
Communication between the control room and the refueling floor was
--
established and maintained in accordance with TS 3.9.5; Fuel movement was conducted in accordance with approved procedures;
--
An audible Source Range Nuclear Instrument count rate could be heard
--
in containment; and, Residual heat removal system operation was in accordance with TS -- 3.9.8.
No items of noncompliance were identified.
l i
. -.
-
..
..
-
-
-
-
-
.
.
.
.
i 7.
IE Bulletin Followuq The inspector reviewed licensee actions taken in response to the following IE Bulletins (IEBs) in order to determine that the written response was submitted within the required time period, that the response included the
information required including adequate corrective action commitments, and that licensee management had forwarded copies of the response to respon-l sible onsite management. The review included discussions with licensee personnel and observations and' review of items discussed below.
I IEB 79-03 - Weld Defects in Stainless Steel Pipe Spools Manufactured by
,
Youngstown Welding and Engineering Company:
!
The licensee's response letter, dated April 11, 1979, stated that piping
'
of the type described by the IEB welded without filler metal) had not been l
used in safety-related systems at BVPS.
The DLC Senior Compliance Engineer l
stated that the basis for this response was review of " Pipe Specification,
Class 153A" which required that all stainless steel piping procured for
.
Category I systems be either " seamless" or " welded with filler material",
I and that all piping had been supplied directly by Southwest Fabricating Co.
Additional inspector review of construction quality control records
,
on January 16, 1980 identified six purchase orders (Nes. 12575, 13151, 13928, 6496, 8414, and 3009) issued directly to Youngstown Welding and Engineering Company in May 1972 for the procurement of piping of the type
described by the IEB.
Subsequent review by the inspector and the licensee i
determined that, although the subject piping was procured on safety-related
purchase orders, none of the piping appears to have been installed in safety-re-
'
'
lated systems.
The documentation of use and installation associated with the above purchase orders was reviewed by the inspector confirming the above.
The inspector informed the licensee that the initial DLC submittal
<
discussed above appeared to remain correct but the basis for that submittal
l appears to have been inadequate.
The licensee acknowledged the inspector's comments and stated that the submittal predated the establishment of the DLC Licensing and Compliance Section which is intended to improve the depth and quality of such submittals.
The licensee further stated that future submittals would receive additional review to preclude similar future occur-rences.
The inspector had no further questions on this matter and will continue to review the adequacy of DLC submittals during future inspections.
,
IEB-79-24 - Frozen Lines:
The licensee's response letter, dated November 1, 1979, stated that a Cold Weather Bill is placed in affect every October, which includes monitoring of heat traced piping, installation of a portable heater in pipe trenches, and other similar freeze protection measures.
The inspector reviewed the
,
'
Cold Weather Bill documentation and noted that, as of January 16, 1980, the Cold Weather Bill was not fully implemented in that the air exhaust louvers in the Main Feed Regulating Valve Room were not covered.
During i
...
.
-
-
.
-.-
-.. - -
_ - _ _ _ _
.
-
.
an inspection of Cold Weather Bill implementation on January 16, 1980, the inspector noted that the Cold Weather provisions noted in the licens-ee's letter had been implemented except as follows:
The inspector noted during this tour that the cap on Fire Hydrant No. 4 was missing and water had collected in the hydrant cavity indicating that the drain valve is either inoperative or clogged.
The inspector confirmed that a maintenance work request had been initiated to repair the hydrant.
Heat tracing on systems was observed to be in operation.
The Operations Supervisors confirmed that the louvers in the Main Feed Regulating Valve Room had not been covered since the systems were shutdown and drained due to the ongoing outage.
No inadequacies or other discrepancies were noted by the inspector.
IEB-79-25 Failure of Westinghouse BFD relays in Safety-Related Systems:
The licensee's response, dated December 17, 1979, stated that BFD relays had been found in control circuits, but that none were of the type described in IEB-79-25.
The inspector reviewed Station Elementary Diagrams and Station Equipment lists associated with the licensees submittal and directly observed the following panels in the field to confirm the licensee's submittal.
Solid State Protection System (SSPS) Train A output cabinet.
--
SSPS Train A Logic cabinet.
--
SSPS Train A Input channel II cabinet.
--
3SPS Train A Auxiliary relay cabinet.
--
3SPS Train A Auxiliary safeguards cabinet.
--
SSPS Train B Output cabinet.
--
SSPS Train B input cabinet III.
--
Relay panels 37F, 35F, 35R, 37R.
--
Emergency diesel generator panels 1 and 2.
--
EE-EG-Diesel Generator #2.
--
No BFD relays were found during this inspection.
EFD relays in the Rod Control Motor Generator power panels were observed by the inspector and verified not to be the type discussed in the bulletin.
The inspector had no further questions on this matte _ _ _ _ _ _ _
.
.
x IEB 79-17 Pipe cracks in Stagnant Borated Water Systems at PWR Plants:
As discussed in IE Inspection Report No. 50-334/79-27, the inspector had previously reviewed Inspection Procedure No. NSD-ISI-90, Manual Ultrasonic Procedures for Investigating for Presence of Intergranular Corrosion and had referred the procedure to NRC:RI for additional technical review.
Based upon that review, on January 30, 1980 the inspector informed the OLC Senior Compliance Engineer that Technique C, Calibration, of the above procedure did not appear to be in compliance with Section XI of the ASME Boiler and Pressure Vessel Code, in that it did not utilize a suitable calibration block for pre-inspection calibration of the ultrasonic test equipment.
DLC consulted with their inspection contractor and informed the inspector on February 6, 1980 that Technique C of NSD-ISI-90 is con-sidered acceptable under ASME Section XI, Paragraph IWA 2240, Alternative Examinations, in that the technique had been qualified on actual samples of piping having known defects.
The inspector requested additional data regarding the above qualification including the correlation between the qualification test data and the Technique C calibration method.
The Senior Compliance Engineer stated that this information would be sought from the inspection contractor.
This matter is unresolved pending receipt and review of the additional information. (80-01-08)
8.
Licensee Event Followup a.
Contamination in the Solid Radwaste Disposal Area.
On January 12 and 13, 1980, loose surface contamination was tracked from the Solid Radwaste Disposal Area into previously uncontaminated
,
areas on the 735' level of the Primary Auxiliary Building (PAB).
Personnel involved in operations within the Solid Radwaste Disposal Area also experienced contamination of their clothing and skin.
In-spector review of this event included its cause, detection, corrective actions taken, and compliance with administrative controls intended to prevent or mitigate such events. The event was reviewed with respect to the following portions of the licensee's Radiological Control Manual (RCM), Chapter 1, Revision 3.
Part II, Radiation Exposure Control and Monitoring.
--
Part III, Contamination Controls.
--
-
Part IV, Radi,ological Surveys.
--
Part VIII, Reports and Record Keeping.
-
Appendix 1, Administrative Guide.
--
l l
'
Appendix 4, Frequency of Airborne, Radiation and Contamination
--
Surveys.
. - _ - -
__
.
.
.
.
(1) Description of Event On January 12, 1980, routine radwaste operations were underway in the Solid Radwaste Area, including decontamination of shipping cask liners. To perform valve operations, an unlicensed operator entered the area's Pump Cubicle wearing full anti-contamination clothing (Anti-Cs), including a respirator and double shoe covers.
Upon departure from the cubicle, the Anti-Cs were removed. The individual passed through the Solid Radwaste Control Room (previously uncontaminated) and exited the PAB via the PAB Health Check Area frisking station.
A short time later the remainder of the crew working in the Solid Radwaste Area exited the area via the Solid Radwaste Area Control Room.
All but one of the five man crew detected contamination on the soles of their shoes at levels of approximately 1000 cpm greater than background, or about 4500 picocuries (pCf), using the same frisking station.
The 735' level of the PAB was isolated by Radcon and surveyed.
Contamination levels in the previously uncontaminated areas between the PAB exit and the Solid gadwaste Area were measured to be between 3500 to 5000 pCi/100 cm via swipes.
An air grab sample from the area of the Solid Radwaste Area door showed less than 1 X 10 migfmum microcuries (pC1)/cc, the minimum detectable activity for Co. The affected areas of the PAB, Solid Radwaste Area and the persognel's shoes were decontaminated to less than 450 pCi/100 cm within the next 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.
Access to the Solid Radwaste Area was restricted pending further evaluation.
In an attempt to locate and quantify the source of contamination, two Radiation Control Technicians (RCTs) entered the Solid Radwaste Area on January 13, 1980.
After completing surveys of the Solid Waste Pump Cubicle, the RCTs detected contamination on their bodies during frisking at levels of about 2400 cpm greater than background on the frisking equipment.
The 735'
level of the PAB was again isolated, surveyed and decontgminated.
Contamination levels ranged from 500 - 50,000 pCi/100 cm in the PAB and from 5000 - 17,000 pCi/100 cm in the Solid Radwaste Area Control Room.
Air grab samples outside the Solid Radwaste Area dooro(735' PAB) indicated airborne activity of approxigately 3.9 X 40 pCi/cc (Maximum Permissible Concentration for Co
=
9 X 10 pCi/cc) apparently resulting from contamination dropping from the RCT's Anti-Cs and shoe covers when the survey as undressed and exited the area.
A portable air monitor sample of the Solid Radwaste Area Control Room indicated airborne activity levels of between 1.0 and 3.3 X 10 pCi/cc which is consistent with grab air samples from the grea.
Isotopic s
analysis of all samples indicated that Co was the source of activity.
_
.
-
_
_
.
.
.
.
,
i The RCTs were decontaminated to less than 100 cpm above background as read on a frisker.
Nasal swabs indicated levels of 0.44 nanocuries (nCi) and 1.87 nCi, respectively.
Initial whole body counts of the RCTs indicated a level of 10 nCi and 3 nCi, respectively at 7:20 p.m. on January 13, 1980.
Subsequent body counts on January 14 (and later) indicated internal depositions to be approximately minimum detectable activity (MDA) or about 1 nCi for each man.
The licensee concluded that the initial higher body count results were due to skin pore contamination rather than internal deposition.
By January 14, 1980, the normally uncontaminated areas of the 735' level PAB had been decontaminated.
Decontamination of the Solid Radwaste Disposal Area Pump Cubicle was in progress.
Con-tamigation in thg pump cubicle was measured to be as high as 6 X 10 pCi/100 cm.
Immediate actions on the part of the licensee included submergence of contaminated floor areas under about 1 inch of water to prevent generation of airborne contamination, construction of a temporary poly-covered wood air lock at the access door of the Solid Radwaste Area, an ALARA (As Low As Reasonably Achievable) review of planned decontamination activities. At the close of this inspection, decontamination was still in progress.
During review and evaluation of this event, the licensee determined that the unlicensed operator who performed the valve manipulations on January 12, 1980 did not comply with the requirements of Techni-cal Specification 6.12.1.a. in that he entered a high radiation area without implementing the monitoring requirements of the above TS.
This matter is the subject of Licensee Event Report No. 80-05/03L.
Corrective action for this aspect of the event is further discussed below.
(2) Apparent Cause The apparent cause of this event was deposition of high levels of contamination on the Solid Radwaste Pump Cubicle Floor due to a valve malfunction on December 28, 1979, misleading sample data obtained from the water spilled by the valve, and decisions based upon the misleading information.
This contamination was initially transferred into previously uncontaminated areas by the unlicensed operator on January 12, 1980.
On December 27, 1979, a spent CVCS mixed bed demineralizer was transferred to the Solid Waste Processing System for solidification and subsequent offsite disposal.
Due to the resultant high radi-ation levels, the area was locked for radiation control purposes.
Subsequent to dewatering of the spent resin, the Spent Resin Dewatering Tank contro's valve, LCV-SW-101, apparently failed open, causing a portion of the tank's contents (which apparently l
-
.
.
..
_
-
...
.
-
.
'
-
'
.
,
included highly radioactive resin fines and other particulates)
to drain to the Solid Radwaste Disposal Area Pump Cubicle.
The sump pump for this area was turned off as a precautionary measure to prevent inadvertent pumping if a resin spill occurred.
This resulted in 2 - 3" of water accumulating on the bottom floor of the area when the level control valve malfunctioned.
On December 28, 1979, water from the floor was sampled and analyzed to contain about 3.5 X 10 pCi/ml of radioactivity and was not considered to present an immediate hazard for the generation of loose surface or airborne contamination.
This sample was subsequently (January 13 - 15, 1980) determined to have been nonrepresentative, in that the solids initially suspended in the water had apparently settled to the floor before the sampling.
The water was routinely pumped from the cubicle on January 29, 1980.
At the time of the level control valve malfunction and acquisition of the associated sample results, a shipping liner containing the spent resin was still being processed in the Solid Radwaste Disposal Area, resulting in radiation levels of 150 roentgens (R)/hr contact and approximately 6 R/hr general area at a distance of approximately 6 feet. The general area radiation levels in the area of the spill were reported to be between
'
30 - 150 millirem (mrem)/hr.
On the basis of the above water sample data and the abnormally high radiation levels in the
-
area, the licensee elected to proceed with processing and removal of the spent resin from the area to conducting any decontamination of the pump cubicle spill area.
Inspector intereview of licensee Radcon Supervision on January 15, 1980 indicated that the magnitude of the radioactivity in the pump cubicle was significantly underestimated on the basis of the above information.
In order to determine whether adequate radiological controls and precautions had been implemented in the Solid Radwaste Area during the time between January 28, 1979 and January 13, 1980, the inspectors reviewed Radiation Work Permit Nos. 6184 and 6227 which were applicable to the activities in the subject area during the time period of interest. The inspectors noted that, with the exception of the unlicensed operator's entry into the high radiation area without adequate personnel monitoring, no other deficiencies or discrepancies were noted.
The Radiation Work Permits were annotated to require protective cloting and equipment appropriate to the actual circumstances and provided control of work activities in the area via assignment of an RCT to each evolution.
Although the licensee had not accurately quantified the levels of contamination in the area, the pre-existing controls imposed appeared to be adequate and effective in
,
,, - - -
-
-
-
.
.
.
minimizing personnel and building contamination detected in previously uncontaminated areas.
The personnel contamination detected appears to have resulted from the removal of Anti-C clothing and equtpment which contained such high levels of surface contamination that normal disrobing techniques resulted in the minor contamination.
All contamination was detected at the appropriate frisker location prior to its potential removal from adiologically controlled areas within the PAB.
The South Sump to which the water from the Pump Cubicle was discharged on December 29, 1979 was sampled and analyzed for radioactivity and the presence of resin beads or finest The South Sump sample of 1 liter read approximately 1 X 10 pCi/ml.
'
The gample filtrate activity was reported to be approximately 1 X 10 pCi/ml and no resin was found in the sample.
The inspectors confirmed the sump to be properly posted and controlled.
Ventilation ducts exhausting from the affected areas are monitored by permanent, fixed filter radiation monitors which showed no increase in level throughout the duration of the event, indicating that no detectable contamination was carried into PAS ventilation systems.
(3) Corrective Action The Radcon Department issued an internal report on the event, recommentations for corrective action, and status of corrective actions in progress on January 21,1980 (Letter No. BVPS:JAK:188).
That report provided the following summary of conclusions which the inspectors confirmed, on a sampling basis, by review of licensee survey, exposure, analysis, and instrument records:
No significant internal deposition of radioactive materials
--
occurred in any of the affected individuals.
Followup body counts indicated MDA.
Administrative or regulatory radiation exposure limits
--
were not exceeded.
All bioassay and air sample results indicate that personnel
--
who had entered affected areas without respiratory equipment were not exposed to airborne radioactivity levels equal to or greater than MPC.
The fixed filter exhaust monitor, RM-VS-101A, indicated no
--
detectable activity throughout the duration of the event.
Transferable radioactive contamination did not spread from
--
the controlled, local area of the 735' level of the PAB.
l
.
.
.
.
.
.
.
All personnel clothing was decontaminated to acceptable
--
levels for unrestricted use.
Response and followup to the radiological problem appeared
--
to be sufficient to control the situation.
The following correctiva actions have been initiated.
An Engineering Memorandum has been submitted to initiate a
--
design change which will relocate the Solid Waste Ventilation Controls to the Solid Waste Control Room to expedite venti-lation isolation should high airborne radioactivity levels be identified.
A frisker station has been placed near the Solid Waste Area
--
exit. No frisker had not been located there previously due to high background radiation levels which had made it im-practical.
Safety meeting training material will be provided and onshift
--
training will be conducted to review high radiation area entry rules, respiratory protection procedures, radiation work permit controls and proper frisking techniques with Radcon and Operations Department personnel.
A Radcon Department Memorandum has been forwarded to Radcon
--
Foremen, directing a review be conducted with all RCTs which addressed familiarization of workers with radiation work permit requirements prior to area entry, obtaining worst case samples for radiological evaluation, remote sampling techniques, and proper presentation of radiological data.
On February 5, 1980, the inspector reviewed documentation for the performance of the above training during the period January 25-21, 1980.
The safety meeting training material discussed above was promul-gated for the month of February 1980.
The inspector reviewed the material on February 5,1980.
During discussions with the Operations Supervisor on February 5, 1980, the inspector was informed that this information will be provided to plant operators via onshift training sessions conducted by Radcon personnel.
These training sessions will be completed as soon as practicable.
The completion of the corrective action items above will be con-firmed during a future inspection.
(80-01-09)
i i
Ga
!
- -
- - -
.
-
.
.
. -.
e
l
.
'
.
'
'
x, N
i On February 6, 1980, inspector direct observation of Radiation Work Permit (RWP No. 6286, Replace Diaphragm of Valve 1CH-44 in the PAB East Valve Trench, and RWP No. 6313, Inspect and Repair Base Plates and Pipe Hangers in PAB and Safeguards Area) processing
,
and use was accomplished to establish that the controls provided
'
were being properly implemented and to confirm that existing work areas and activities are properly administered with respect to radiological control requirements.
Records reviewed included:
i Prework Survey Data;
--
Radiological Control Requirements to establish that they
--
were commensurate with survey data; Respiratory Use Logs;
--
Dosimetry Tracking Logs; and,
--
Blanket RWP Dosimetry Record.
--
The RWP briefing and sign-on process were observed for craft and operating personnel entering the various work areas in order to confirm that personnel were being adequately briefed, and were knowledgeable of the radiological hazards and requirements
,
associated with their respective tasks.
The inspectors confirmed, for the two RWPs above, that conditions in the work area were accurately reflected and controlled by the RWPs.
With regard to the Resin Dewatering Tank level control valve, interviews with plant operations and maintenance personnel deter-mined that corrective maintenance on the instrument control loop for the valve has neither identified nor corrected the cause for valve malfunction.
Pending further investigation, the licensee is administrative 1y controlling the valve control in the manual position to prevent inadvertent automatic opening.
As part'of the circuit troubleshooting, the licensee has confirmed that an actual high ta6k level will not cause the valve to automatically open when its control switch is in the manual mode.
Except as noted above, the inspector had no further questions with regard to this event.
b.
Loss of RHR Flow (LER 80-002/03L)
On January 17, 1980, the facility sustained a total loss of Residual Heat Removal System flow, apparently due to air entrainment of.both RHR Pumps. The inspector reviewed this event on January 17-18, 1980
]
with regard to its apparent cause(s), the licensee's immediate corrective l
.
,
e
-+--m-ra
-,. ---
g-
-, - -
.,- -..,,
, - -.. - - - - -
- - -, - - - - -
.
-
.
.
.
action, and action taken to prevent recurrence.
On January 30, 1980, the licensee issued LER 80-002/03L, which the inspector reviewed.
No discrepancies were noted.
On January 17, 1980, the facility was in Operational Mode 5 with prepara-tions for refueling underway.
P.CS coolant level was being maintained at approximately the horizontal centerline of the RCS hot leg piping.
A reactor vessel vent eductor system had been installed and was drawing a suction on the free volume in the reactor vessel head and discharging the air and gases to the Supphmentary Leak Collection and Release System. During initial operation of this temporary system on January 16, 1980 (in accordance with Temporary Operating Procedure (TOP) No.
80-1, dated January 9, 1980), a rapidly increasing RCS level transient i
occurred. RHR flow was maintained while loop level was restored to the desired centerline level.
The apparent cause of the level transient
'
was the draining of excess coolant trapped in the Steam Generator Inverted U-tubes because of the differential pressure placed across the U-tubes by the air eductor system operation.
About ten hours after this level transient the operating RHR pump displayed indications of no flow and very low operating current (consistent with air binding).
The redundant pump was started and displayed the same characteristics.
t l
The pumps were vented.
Flow was reestablished with one RHR pump within approximately 38 minutes, after experiencing an RHR/RCS temperature increase from 100F to 126F.
Licensee review of the RHR flow transient concluded that the water release from the steam generator tubes apparently included a significant volume of entrained air which later air bound the RHR pumps.
Inspector review of operator logs, RCS loop level recorder traces, TOP-80-1, and the facility's NSS vendor recommendations for the design, installation, and operation of the reactor vessel air eductor system confirmed the licensee's evaluation of the event to be adequate.
No other system evolutions or transients had occurred during the period of interest which would have affected pump operation.
Air binding of RHR pumps has been a recurring problem at BVPS when operating with the RCS loops partially drained.
The licensee had previously addressed this problem by reducing RHR flow to minimize vortexing in the RHR suction line region of the RCS piping.
Inspector review of plant operations during January 16 - 18, 1980 against the BVPS Operating Manual, Section 1.10.4, RHR System Operating Procedures, indicated that the system had been operated in accordance with the procedure requirements.
The corrective action stated by the licensee in LER 80-002/03L includes revision of procedures and investigation of a constant vent system for the RHR pumps to precludu similar air binding during future operations with the reactor vessel vent air ductor in service.
At the close of this inspection, implementatioa c^ these actions was still in progress.
Completion of the licensee's actions :n11 be reviewed during a future in'pection. (80-01-10)
s l
-
.
_
- -. -
.
.
.
.
c.
Unplanned Radioactive Steam Release On January 13, 1980, a cooling water supply valve for the 1A Boron Recovery Evaporator overhead condenser failed shut, causing an over-pressure condition within the evaporator.
A 3 inch evaporator relief valve lifted at approximately 100 psig, releasing steam to the Gaseous Waste Discharge Header which discharges to the atmosphere via an outlet on the cooling tower. The relief valve discharge line connects to the discharge header via a loop seal and enters the piping downstream of the Gaseous Waste Disposal Blowers. The gaseous waste disposal header downstream of the relief valve connection is monitored by a particulate / gaseous radiation monitor, RM-GW-108A and B, which was out of service for corrective maintenance of its sample pump.
The event occurred about 12:25 a.m. on January 19, 1980.
The inspector was informed by the licensee and arrived at the site approximately 3:00 a.m.
The initiating events and followup activities were reviewed against the following references and regulatory requirements:
BVPS Environmental Technical Specifications (ETS);
--
BVPS Technical Specifications (TS);
--
BVPS Operating Manual (OM), Chapter 1.57, Emergency Preparedness
--
Plan (EPP), Issue 5; The event was reviewed for the following elements:
Whether applicable requirements of TS and ETS were satisfied.
--
Whether plant systems had been placed ur were being maintained
--
in stable configurations.
Whether licensee had properly notified i RC, state, and local
--
officials.
l Whether reporting requirements of TS and ETS had been satisfied
--
with respect to written followup reports.
(The inspector deter-
'
mined that no written reports were required).
Whether potential or estimated releases of radioactivity were
--
I within regulatory limits.
Whetner significant radiation levels existed in the plant, onsite,
--
or offsite as a result of the event.
l
'
Whether actions implemented by the licensee for monitoring and
--
evaluating the radiological consequences of the event were in
accordance with the EPP and its implementing procedure.
.
.
s s
\\x Whether sample and survey results substantiated the licensee's
--
conclusions regarding the effect of the event on plant operations and the health and safety of the public.
With regard to initiation of the event, the inspector determined that the malfunctioning cooling water valve, TV-CC-126-1, had a faulty position indicator.
Valve position is indicated by stem actuated limit switches which control main control board indicator lights.
A stem traveler which provides the actuation for the limit switches was found to be loose, resulting in an open indication in the control roem when the valve was actually closed.
This condition was corrected on January 19, 1980.
As a result of the cooling water valve malfunction, an operator noted increasing evaporator pressure at 12:25 a.m. and secured steam to the equipment. About 12:28 a.m. transient flow conditions and alarms on the Gaseous Waste Discharge Header occurred.
That corresponds to the time of relief valve actuation. Within the next 30 minutes, the operators secured discharge of normal Gaseous Waste Discharge header blower exhaust and reestablished cooling water to the evaporator, reducing its pressure.
No further lifting of the associated relief valve was evident from review of evaporator pressure recorder traces.
Within the following thirty minutes the Shift Supervisor toured the Primary Auxiliary Building (PAB) and verified that no measurable water level existed in the relief valve discharge loop seal, evaluated the i
immediate indications of the occurrence and notified plant management of the event.
l l
At approximately 1:30 a.m. the EPP was initiated on a limited basis I
as a precautionary measure to provide augmented survey and monitoring capability and EPP telephone call list notifications were begun.
Based on the preliminary information available, the licensee did not declare an emergency condition.
Between 1:45 and 2:15 a.m. initial samples of the Gaseous Waste Discharge header were obtained.
About 2:15 a.m. the NRC:HQ duty officer and resident inspector were notified and an offsite monitoring team was dispatched.
At approximately 3:00 a.m. DLC management arrived in the Control Room and Emergency Control Center to coordinate the monitoring activities.
Between 3:00 a.m. and 5:00 a.m. sample results from two offsite monitoring teams, onsite Radiological Control Department samples and surveys, and data from offsite proportional ion chamber environmental monitors indicated no evidence of offsite releases.
All radiation survey data indicated background levels.
All radiogas, particulate and liquid samples taken from the environment and from gaseous waste discharge header piping indicated no activities greater than Minimum Detectable Activity (MDA) for the respective analysis techniques.
Samples of residual liquid in the Gaseous Waste Discharge header were taken from the discharge header piping and indicated less than MDA.
The sampling
-_
-
. _ _ _ _
- _ _ _ _ _
- _ _ _ -. -
.
.
-
.
times for all analyses performed spanned from 12:50 a.m. (initial radiogas sample at the sample point for the Gaseous Waste Discharge Header radiation monitor) through 4:55 a.m. (particulate and radio-iodine samples in the downwind NE sector of the offsite environs).
The qvaporator bottom liquid was sampled and analyzed to contain 4.82 X 10 pCi/ml gross activity, isotopically analyzed to be primarily Cesium and Cobalt.
The licensee's conservative estimate of total
,
mass released by the relief valve discharge indicated that a maximum of 0.84 pCi of Cobalt and Cesium isotopes could theoretically have been released on the basis of the following assumptions:
Activity in the evaporator steam space at the point of the relief
--
valve inlet connection was a factor of approximately 100 less than that contained in the evaporator bottoms.
A nominal decon-tamination factor for the steam water interface is approximately 1000; All activity released by the relief valve was released in the
--
form of steam vapor carried through the Gaseous Waste Discharge header 7and was diluted by discharge into a flow of approximately 3 X 10 cc/ min.;
Tritium contribution to the released activity is not included
--
in the above.
An assumed tritium contribution of 0.01 Ci was
,
added to the above on the basis of historical radwaste data, assuming a partition factor of 10 at the fluid / vapor boundary in the avaporator.
The inspector vs. viewed the above calculations and the input data and determined them to be sufficiently conservative on the basis of the above assumptions.
In actuality, the possibility of any release of radioactive evaporator steam or is condensed liquid appears very un-likely, in that:
the discharged liquid sampled from the header piping was all analyzed to be less than MDA; the effluent was calculated to have a maximum liquid mass of 383 lbm, equivalent to approximately 6.1 cu. ft. of water; and the approximate volume of the uninsulated discharge header (10 in. diameter piping approximately 1000 ft. in length) is 450 cu. ft.
Based on the presence of multiple loop seals between the relief valve discharge (all sampled as less than MDA)
and the relative volumes above, it appears that all relief valve ef-fluent was likely retained within the piping.
Late in the day of January 19, 1980, the evaporator was repressurized and a steam space sample was drawn and analyzed.
The sample results indicated that, with the same evaporator influent activities, the vapor space contained radioactivity levels less than MDA, confirming the conservatism of the licensee's assumptions for a decontamination factor of 100 in the above calculations.
_
.
-
.
Inspector review of the radioactive effluent discharge limits and re-porting requirements in the ETS determined that, assuming the worst case release discussed above, no regulatory limits on effluent activity nor reporting limits had been met or exceeded.
During the event, telephone liaison had been established with the NRC and with state and local officials.
At approximately 5:10 a.m. the various offsite agencies were informed of the conclusions discussed above and the Emergency Control Center was secured.
No items of noncompliance were identified.
d.
lA Emergency Diesel Generator Sequencing Timer Malfunction On February 1, 1980, the facility sustained a partial loss of offsite power apparently due to spurious operation of 138 KV switchyard pro-tective relay equipment.
As a result, power was lost to the No. 1 138 KV Bus and the No. lA 4160 V Normal Bus and its associated emergency loads.
The No. 1 Emergency Diesel Generator (EDG) started on a bus blackout signal but failed to complete its timed sequence for energi-
-
zation of emergency loads.
During this event, the plant was in a Refueling configuration (Mode 6) with no core alterations in progress.
The redundant source of offsite power via the B Station Service Transformer was maintained throughout the event and the No. 1A 4160 V Bus was reenergized from offsite power within approximately 15 minutes.
The licensee immediately began an investigation into the cause of the switchyard malfunction and the anomalous operation of the EDG sequencer timer.
On February 1, 1980, after the event, the licensee identified and immediately corrected two incorrect wiring connections in the EDG sequencer relay cabinet.
At the close of this inspection, the cause, approximate time of occurrence, and consequences of the wiring errors l
were still being investigated.
Based on the informatien available, l
the licensee was unable to determine whether the wiring discrepancies l
were incident to authorized maintenance or the result of unauthorized
entry into the cabinets.
The Maintenance Department was in the process l
of reviewing equipment history and maintenance records to establish l
the scope and chronology of prior work on the subject equipment.
The wiring discrepancies found collectively resulted in deenergizing the timer circuit.
The inspector reviewed the preliminary information
available and determined that, for Cold Shutdown (Mode 5) and Refueling (Mode 6) operations, the effect on the plant was negligible and the applicable TS Limiting Conditions for Operation was satisfied.
The l
inspector informed the licensee that, unless the time of occurrence l
of the discrepancies could be shown to be after the most recent plant
'
shutdown, the event must be reported, at least, as operation in a degraded mode per TS 6.9.1.9.b.
The Superintendent was requested to
-
-
-
.
i
.
.
~
.
,
s
\\
keep the inspector informed of the progress and findings of the DLC investigation as it progressed.
This matter will remain unresolved pending further investigation by DLC and NRC review.
(80-01-11)
9.
Followup on NSSS 10 CFR 21 Report - Omission of Safety Circuit Test in Westinghouse PWR Plants The subjact report, issued by the facility's NSSS on November 7, 1979, identifies an undetectable potential failure which could exist in Engineered Safety Features (ESF) actuation circuitry and provided recommended testing techniques which will verify the circuits' operability.
The P-4 permissive circuit provides for manual reset and block of a safety injection signal if the P-4 circuit indicates that the reactor trip breaker is open, i.e.
the reactor is tripped.
During normal power operation, this circuit prevents manual actions which could electrically block automatic initiation of safety injection.
The contacts which provide the P-4 signal are mechanically linked auxiliary contacts associated with the reactor trip breakers.
The NSSS has determined that current testing methods for the ESF circuitry do not provide for checking the operation of the P-4 contacts or the intercon-necting wiring thus permitting a potential failure of the circuit to remain undetected.
The failure modes for these circuits could result in inhibiting the manual resetting and blocking of SIV and alter the sequence of switchcover operations from the injection of the recirculation phase or could result in a block of automatic safety injection remaining in effect when the plant had been returned to power following reset of the reactor trip breakers.
The inspector reviewed the licensee's plans for implementing the testing recommendations of the above report, including discussions with the DLC Operations Supervisor and the NSSS Instrument Service Consultant.
A temporary operating procedure is in preparation which will address the report's recom-mendations. Long term plans include the revision of existing Maintenance Surveillance Procedures and BVPS Operating Manual startup procedures to assure that the recommendations are addressed.
The inspector determined that the licensee's plans include provisions for conducting the recommended tests prior to station restart. The inspector also observed that, based upon the P-4 circuit configuration at BVPS, failure of the contacts providing the P-4 signal would likely be evident by a sustained feedwater isolation, thus preventing inadvertent return to power with a contact failure that would also result in an undiscovered block of automatic safety injection initiation.
The completion of the licensee actions above will be followed during subsequent inspections.
(80-01-12)
10.
Review of Periodic Reports The inspector reviewed the following periodic reports to verify, as appli-cable, that the information required to be reported by NRC requirements had been included, that supporting information discussed in the reports is consistent with design predictions and performance specifications, that
.
.
-
.
planned corrective action is adequate for resolution of identified problems, and whether any information included in the report should be classified as an abnormal occurrence.
The reports reviewed were:
Monthly Operating Reports, July 1979 through December 1979 No discrepant items were identified.
11.
Review of Physical Protection / Security Activities During an inspection tour of the facility guard house and protected area perimeter on February 7, 1980 and during the tours of protected and vital areas discussed in paragraph 3 of this report, the inspector observed the physical / security activities below to confirm compliance with the require-ments of 10 CFR 73.55 and the licensee's security plan and procedures:
Accesses to the protected and vital areas were closed and locked when
--
not attended; Isolation zones were free from visual obstructions;
--
Vehicles authorized access to the protected area were escorted;
--
--
Persons, packages and vehicles were searched in accordance with ap-plicable requirements;
--
Operation checks of metal and explosive detectors were being performed and documented as prescribed by licensee procedures;
--
Identification badges were properly displayed and personnel requiring escort were being escorted; and,
--
Security force response to perimeter detection aids was in accordance with the requirements of 10 CFR 73.55(h) and the licensee's procedures.
No items of noncompliance were identified.
12.
Emergency Preparedness Plan Accountability Provisions for New Office Building The licensee has recently completed and occupied a new administration building within the owner controlled site area but outside the BVPS-1 protected area. The building is currently occupied by DLC clerical, engineering, training, and administrative personnel and the NRC Resident Inspector Office.
During this inspection the licensee installed public address and alarm speakers in the new building, providing the occupants with notification
-
_
.
.
.
of emergency conditions from the control room.
The licensee has additionally established interim personnel accountability methods for building occupants, including designation of personnel assembly area and assignment of accounta-bility responsibility to group supervisors.
The inspector noted that the licensee is in the process of developing a new Emergency Preparedness Plan to meet current regulatory requirements.
The licansee stated that the new EPP will include long term provisions for building occupants which address the above. The inspector had no further questions regarding this matter.
13.
Unresolved Items Unresolved items are matters about which more information is required'to determine whether they are acceptable, items of noncompliance or deviations.
Unresolved items addressed during this inspection are discussed in paragraphs 2, 5, 7, 8 and 9 of this report.
14.
Exit Interview Meetings were held with senior facility management periodically during the course of this inspection to discuss the inspection scope and findings.
'
A summary of inspection findings was also provided to the licensee at the conclusion of the report period.
I
,
e e
.
9