IR 05000275/1990003
| ML16341F582 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 02/12/1990 |
| From: | Tenbrook W, Wenslawski F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML16341F581 | List: |
| References | |
| 50-275-90-03, 50-275-90-3, 50-323-90-03, 50-323-90-3, NUDOCS 9003080163 | |
| Download: ML16341F582 (22) | |
Text
U.
S.
NUCLEAR REGULATORY COMMISSION
REGION V
Report Nos.
50-275/90-03 and 50-323/90-03 License Nos.
Pacific Gas and Electric Company 77 Beale Street Room 1451 San Francisco, California 94106 Facility Name:
Diablo Canyon Power Plant, Units 1 and
Inspection at: Diablo Canyon Site, Seven miles north of Avila Beach, California Inspection conducted:
January 22-26, 1990 Inspected by: ~
AC~
en roo
,
a 1a son pec>a
>s Approved by:
ens aws T,
se Facilities Radiological Protection Section
~Summer:
g-v- 9e a e.
>gne a
e igne Areas Ins ected:
Routine unannounced inspection covering follow-up of open
>tems, o
ow-up of corrective actions for reported events, plant chemistry and radiochemistry, radiological confirmatory measurements and facility tours.
Inspection procedures 30703, 92700, 92701 and 84750 were addressed.
Results:
Program strengths included good water quality in the reactor coolant anon secondary systems (section 4.C),
and insta1lation of improved on-line chemistry, monitoring instrumentation (section 4.B).
Confirmatory measurements indicated that the licensee had employed an improper half-life for Rb-88 analysis.
Also, the licensee had not consistently documented quality control checks on control charts.
These items will be addressed by licensee commitments (section 4.D).
9OO3OSO 1 63 5OPO27',J 9OO214 PDR ADQCK 0 p
A
DETAILS Persons Contacted Licensee Personnel
~D.
Miklush, Assistant Plant Manager
"M. Kelly, Regulatory Compliance Engineer
"J. Hoots, Chemistry Manager
"J. Gardner, Senior Chemical Engineer
~D. Cosgrove, equality Control Specialist
"K. Dinnel, equality Assurance Auditor
- A. Taylor, Chemical Engineer
- F. Guerra, Chemistry Foreman
"B. Giffin, Acting Plant Manager
- R. Johnson, Chemistry General Foreman NRC
- P. Narbut, Senior Resident Inspector
" Denotes those individuals present at the exit interview conducted on January 16, 1990.
In addition, discussions were held with other members of the licensee's staff and contractor personnel.
Follow-u of 0 en Items (92701)
0 en Item 50-275/88-33-01 (OPEN)
This item concerned an NRC/licensee in ercomparison o
r-
,
e-55 and tritium in water.
A sample of liquid waste was obtained during the inspection for analysis by the licensee and the NRC contract laboratory.
This item will remain open pending satisfactory completion of the intercomparison.
0 en Item 50-275/89-25-01 (CLOSED)
This item concerned unidentified no e gas ea age in t e uxi sary building and pressure transients in the Gaseous Radwaste System.
The licensee stated they would conduct air monitoring 'and survey valve contamination on and around the Residual Meat Removal system during its alignment upon Unit 2 shutdown to attempt to determine the source of the leakage.
The licensee also stated that pressure, temperature and activity levels in gas decay tanks would be trended during offgas of the Unit 2 reactor coolant system (RCS)
upon shutdown.
The licensee's planned actions to investigate these concerns were satisfactory.
This item is closed.
Onsite Fo'liow-U of Mritten Re or ts of Non-Routine Events (92700)
Licensee Event Re ort 50-323/2-89-12-LO (OPEN)
This Report concerned an ac ua ion o on ainmen en i a )on so a ion (CVI) caused by RCS gas sample purging to the plant vent.
Procedure CAP E-l, "Sampling of Primary Systems,"
had been amended to include provisions for controlling
sample purge volume and aligning purge flow to the gaseous radwaste system.
Design changes to restore operability of on-line gas analyzers were in progress.
The licensee's continuing corrective actions were adequate.
This item will remain open pending further review of the licensee's evaluation of this event.
4.
Water Chemistr Confirmator Measurements (84750)
A.
Audits B.
The inspector reviewed two Audit Reports issued since the last inspection.
Audit 89806T concerned offsite dose calculations and plant radiological effluent monitoring.
No Audit Finding Reports (AFR) or Nonconformance Reports were issued.
Audit 89813T addressed plant chemistry and radiochemistry.
Continuing issues of Chemistry Department compliance with equality Assurance Manual requirements were identified.
The Audit Report emphasized the need for the Chemistry Department to correct conditions identified in prior audits by the dates agreed upon.
Unresolved AFR issues specifically identified were quality assurance of process chemicals and disposal of radioactive samples.
The chemistry department had submitted multiple procedure changes in response to the AFR, immediately prior to the inspection.
The licensee's quality assurance oversight of chemistry activities was satisfactory.
Chanchee The inspector observed the installation and testing of on-line ion chromatographs during the inspection.
The licensee installed these instruments to significantly improve the sensitivity and accuracy of on-line chemistry monitoring of the condensate polisher effluent.
These improvements were intended to optimize demineralizer bed regeneration and alert operators and chemistry personnel to polisher breakthrough.
No unreviewed changes to the facility were identified.
C.
Primar and Secondar Water The inspector reviewed daily chemistry data collected during the inspection pursuant to Administrative Procedure C-252,
"DCPP Chemical Data,"
and verified the selected data in laboratory logs.
The following conditions were observed for Unit 1 at power:
~Sam le RCS Fluoride RCS Chloride RCS Dissolved Oxygen RCS Activity RCS I-131 Equivalent Steam Generator (SG) Cl SG Cation Conductivity SG Sodium SG Sul fate Results
< 0.02 ppm 5 - 10 ppb 1 - 5 ppm 0.5 uCi/gm 0.003 uCi/gm 1 -
ppb 0.3 uS/cm 1 - 3 ppb 1 - 4 ppb Limit or Action Level 0.15 ppm 150 ppb 100 ppm 320 uCi/gm 1 uCi/gm 20 ppb 0.8 uS/cm 20 ppb 20 ppb The following conditions were observed for Unit 2 at power:
~Sam 1e Results Limit or Action Level RCS Fluoride RCS Chloride RCS Oissolved Oxygen RCS Activity RCS I-131 Equivalent Steam Generator (SG) Cl SG Cation Conductivity SG Sodium SG Sulfate
< 0.02 ppm
<1 - 5 ppb
<1 -
3 ppb 0. 7-0. 8 uCi/gm 0.008-0.009 uCi/gm 1 -
ppb 0.3 uS/cm 1 - 2 ppb 1 -
ppb 0.15 ppm 150 ppb 100 ppb 250 uCi/gm 1 uCi/gm 20 ppb 0.8 uS/cm 20 ppb 20 ppb The quality of primary and secondary water was well within applicable action levels and limiting conditions for operation.
RCS specific activity indicated satisfactory fuel integrity.
Cation conductivity levels were typically elevated due to presence of borate anions from secondary water boric acid addition.
Confirmator Measurements and Radiochemical Anal sis The NRC Regional mobile laboratory was brought onsite for gamma isotopic intercomparisons with the licensee s counting laboratory.
Sample types commonly analyzed for compliance with regulatory requirements were analyzed by the licensee and the inspector, and the results were compared using the NRC verification test criteria (see enclosure).
The first sample obtained was a single 47 mm filter of suspended solids from one liter of reactor coolant.
The radionuclides on the filter had decayed several days prior to the analyses.
The results of the filter comparison are presented in Table I
Table
Reactor Coolant Suspended Solids Analyte Licensee Result (uCi/ml)
NRC Result (uCi/ml)
NRC Random Ratio:
Agreement Uncertainty Licensee/NRC Range (uCi/ml)
Dl Cr-51 Dl Mn-54 Dl Co-58 Dl Co-60 Dl Sn-117m D1 Hf-181 D1 Nb-95 Dl Zr-95 D1 Zr-97 Dl Ba-140 D1 La-140 D2 Cr-51 D2 Mn-54 D2 Co-58 D2 Co-60 D2 Sn-117m D2 Hf-181 D2 Nb-95 D2 Zr-95 D2 Zr-97 D2 Ba-140 D2 La-140 D5 Cr-51 D5 Mn-54 D5 Co-58 D5 Co-60 D5 Sn-117m D5 Hf"181 D5 Nb-95 D5 Zr-95 D5 Zr-97 D5 Ba-140 D5 La-140 D6 Cr-51 DG Mn-54 DG Co-58 D6'o-60 D6 Sn-117m D6 Hf-181 D6 Nb-95 DG Zr-95 1.45E-05 1. 78E-05 2. 04E-06 1. 61E-06 1. 44E-06 1. 82E-07 5. 07E-05 4. 91E-05 5. 30E-07 3. 93E-06 4. 29E-06 3. 45E-07 5. 34E-07 6. 88E-07 1. 06E-07 7. 98E-07 1. 27E-06 2. 29E-07 l. 71E-05 1. GBE-05 3. 50E-07 1. 58E-05 1. 49E-05 4. 50E-07 2. 18E" 05 3. 81E-05 7. 49E-06 7. 71E-05 5. 93E-05 1. 54E-06 2. 02E-04 2. 03E-04 3. 30E-06 1. 57E-05 1. 78E-05 2. 04E-06 1.44E-06 1.44E-06 1.82E-07 4. 78E-05 4. 91E-05 5. 30E-07 3.52E-OG 4.29E-06 3.45E-07 5,74E-07 6.88E-07 1.06E-07 6.73E-07 1.27E-OG 2.29E-07 1.62E-05 1.68E-05 3.50E-07 1.50E-05 1.49E-05 4.50E-07 2.32E-05 3.81E-05 7.49E-06 7. 15E-05 5. 93E-05 l. 54E-06 1. 80E-04 2. 03E-04 3. 30E-06 1. 73E-05 1. 78E-05 2. 04E-06 1.45E-06 1.44E-OG 1.82E-07 4. 22E-05 4. 91E-05 5. 30E-07 3.66E-OG 4.29E-06 3.45E-07 6.62E-07 6.88E-07 1.06E-07 6. 95E-07 1. 27E-06 2. 29E-07 1.67E"05 1.68E"05 3.50E-07 1. 96E-05 1. 49E-05 4. 50E-07 2. 61E-05 3. 81E-05 7. 49E-06 7. 67E-05 5. 93E-05 1. 54E-06 1. 93E" 04 2. 03E-04 3. 30E-06 1. 50E-05 1. 78E-05 2. 04E-06 1.48E-06 1.44E-06 1.82E-07 5. 19E-05 4. 91E-05 5. 30E-07 3.64E-OG 4.29E-06 3.45E-07 6.59E-07 6.88E-07 1.06E-07 8.67E-07 1.27E-06 2.29E-07 1. 73E-05 1. 68E-05 3. 50E-07 1. 61E-05 1. 49E-05 4. 50E-07 0. 81 1. 12 1. 03 0. 92 0. 78 0. 63 1. 02 1. 06 0. 57 1.3
0. 88
0. 97 0. 82 0. 83 0. 53 0. 96 1. 01 0. 61 1. 21 0. 89 0. 97 1. 01 0. 86 0. 85 0. 96 0. 55 0. 99 1. 32 0. 69 1. 29 0. 95 0. 84 1. 03 1. 06 0. 85 0. 96 0. 68 1. 03 1. 08 0. 6-1. 66 0. 5-2. 00 0. 80-1. 25 0. 6-1. 66 0.5-2.00 0.5-2.00 0.75-1.33 0.75-1.33 0.5-2.00 0. 75-1. 33 0. 80-1. 25 0. 6-1. 66 0.5-2.00 0. 80-1. 25 0. 6-1. 66 0. 5-2. 00 0. 5"2. 00 0. 75-1. 33 0. 75-1. 33 0. 5-2. 00 0. 75-1. 33 0. 80-1. 25 0 ~ 6-1. 66 0.5-2.00 0. 80"1. 25 0. 6-1. 66 0.5-2.00 0. 5-2. 00 0. 75" 1. 33 0, 75-1. 33 0. 5-2. 00 0. 75-1. 33 0. 80-1. 25 0. 6-1. 66 0.5-2.00 0.80-1.25 0.6-1.66 0.5-2.00 0.5-2.00 0.75-1.33 0. 75-1. 33
D6 7r-97 1.89E-05 3.81E-05 7.49E-06 D6 Ba-140 8. 10E-05 5.93E-05 1.54E-06 D6 La-140 2.02E-04 2.03E-04 3.30E-06 0. 5 0. 5-2. 00 1. 37 0. 75-1. 33
0. 80-1. 25 The licensee and NRC measurements of the reactor coolant filter were in general agreement, with one Ba-140 measurement on licensee detector 6 marginally outside the agreement range.
The next sample obtained was 10 ml of degassed reactor coolant from Unit 1.
The inspector observed the sampling and degassing of the coolant and verified the method was consistent with Procedure CAP E-l, "Sampling of Primary Systems."
The coolant sample was split into separate aliquots for NRC and licensee analysis.
The initial intercomparison of reactor coolant is given in Table 2.
Table
10 ml of Reactor Coolant, First Sample NRC Licensee NRC Random Ratio:
Agreement
'esult Result Uncertainty Licensee/NRC Range Analyte (uCi/ml) (uCi/ml) (uCi/ml)
Dl Co-58 Dl Cs-134 D1 Rb-88 D1 I-132 D1 I-133 D1 I-134 D1 I-135 D1 Cs-137 Dl Cs-138 5.18E-04 9.88E-04 9.45E-05 1. 77E-03 2. 10E-03 1. 19E-04 3. 16E-03 3. 96E-02 8. 18E-03 7. 72E-03 9.41E-03 l. 69E-04 4.98E-03 5.45E-03 1.11E-04 1.47E-02 1.71E-02 3.90E-04 7. 42E-03 9. 67E-03 4. 31E-04 1.60E-03 1,66E-03 1.34E-04 l. 65E-02 l. 75E-02 6. 90E-04 0. 52 0. 85 0. 08 0. 82 0. 91 0. 86 0. 77 0. 97 0. 95 0.6-1.66 0. 75-1. 33 0. 5-2. 00 0. 80-1. 25 0. 75-1. 33 0. 75-1.'33 0.?5-1.33 0. 6-1. 66 0.75"1.33 Co-58 and Rb-88 measurements of the initial sample of reactor coolant did not agree, Measurements obtained with other licensee detectors were similar to those obtained with detector 1,
as illustrated in Table 1 above.
The inspector determined that the NRC measurement of Co-58 was biased high due to an unresolved contribution to the 810.8 keV Co-58 emission from the 809.8 keV and 812.2 keV emissions of I-132.
The disagreement between the measurements of Rb-88 was caused by differing half-lives employed by the NRC and licensee.
The NRC employed the 17.8 minute half-life of Rb-88.
The licensee used the 170 minute half-life of Kr-88, the radioactive parent of Rb-88 in the reactor coolant.
However, no Kr-88 was identified in the analyses of the degassed sample, suggesting that the Rb-88 in the sample was not in equilibrium with Kr-88.
Therefore, the use of the 170 minute equilibrium half-life to determine Rb-88 in degassed samples significantly underestimated RCS Rb-88 activity.
The licensee agreed, and committed to properly account for Rb-88 deca The inspector will review the licensee's corrective actions in a subsequent inspection (50-275/90-03-01).
A second sample of RCS liquid was obtained in response to the disagreements observed in the first comparison.
The second sample, obtained from Unit 2, exhibited dominant short-lived radioiodine activity.
The results of the comparison are given in Table 3.
Table
10 ml of Reactor Coolant, Second Sample NRC Licensee NRC Random Ratio:
Agr cement Result Result Uncertainty Licensee/NRC Range Analyte (uCi/ml ) (uCi/ml ) (uCi/ml )
D1 I-131 Dl I-132 D1 I-133 D1 I"134 D1 I"135 Dl Cs-138 D2 I-131 D2 I-132 D2 I-133 D2 I-134 D2 I-135 D2 Cs-138 D5 I"131 D5 I-132 D5 I-133 D5 I-134 D5 I-135 D5 Cs-138 D6 I-131 D6 I-132 D6 I-133 D6 I"134 D6 I-135 D6 Cs-138 1. 20E-03 1. 24E-03 1. 37E-04 2.65E-02 3.08E-02 3.80E-04 1.46E-02 1.58E-02 2.00E-04 4.85E-02 5.20E-02 1.06E-03 2.28E-02 2.98E-02 9.00E-04 4.79E-02 4.54E-02 2.45E-03 1. 22E-03 1. 24E-03 1. 37E-04 2.63E-02 3.08E-02 3.80E-04 1.44E-02 1. 58E-02 2. OOE-04 4.86E-02 5.20E-02 1.06E-03 2. 51E-02 2. 98E-02 9. OOE-04 4.63E-02 4.54E-02 2.45E-03 1. 62E-03 1. 24E-03 1. 37E-04 2.84E-02 3.08E-02 3.80E-04 1. 52E" 02 1. 58E"02 2. OOE-04 4. 81E-02 5. 20E-02 1. 06E-03 2. 76E-02 2. 98E-02 9. OOE-04 3.90E-02 4.54E-02 2.45E-03 1. 24E-03 1. 24E-03 1. 37E-04 2.43E-02 3.08E-02 3.80E-04 1. 25E"02 1. 58E-02 2. OOE-04 3. 92E-02 5. 20E-02 1. 06E-03 2. 62E-02 2. 98E-02 9. OOE-04 3.63E-02 4.54E-02 2.45E-03 0. 97 0. 86 0. 93 0. 93 0. 77 1. 05 0. 99 0. 85 0. 91 0. 93 0. 84 1. 02 1.3 0. 92 0. 96 0. 93 0. 93 0. 86
0. 79 0. 79 0. 75 0. 88 0.8 0.6-1.66 0. 80-1. 25 0. 80-1. 25 0. 75-1. 33 0. 75-1. 33 0. 75-1. 33 0. 6-1. 66 0. 80-1. 25 0. 80-1. 25 0. 75-1. 33 0. 75-1. 33 0. 75-1. 33 0. 6-1. 66 0. 80-1. 25 0. 80" 1. 25 0. 75-1. 33 0. 75-1. 33 0. 75-1. 33 0. 6-1. 66 0. 80-1, 25 0. 80-1. 25 0. 75-1. 33 0. 75-1. 33 0. 75-1. 33 The measurements of the second coolant sample generally agreed, indicating that the licensee's surveillance of dose equivalent I-131 was accurate.
Heasurements on detector 6 were marginally out of the agreement range.
The inspector observed that the detector 6 dead time was approximately 8.5X, compared to 3.8X or less on the other detectors, and the sample was counted on a lower shelf, making the analysis vulnerable to geometry error from sample placement.
The
I
Radiochemical Engineer stated he would reevaluate efficiency calibration with a traceable standard.
the detector The next sample was obtained from a liquid primary waste receiver tank.
The measurement results are given in Table 4.
Analyte Licensee Result (uCi/ml)
NRC Result (uCi/ml)
Table
Liquid Maste NRC Random Ratio:
Agreement Uncertainty Licensee/NRC Range (uCi/ml)
Dl Mn-54 Dl Co-57 Dl Co-58 Dl Fe-59 Dl Co"60 Dl Ag-110 Dl Nb-95 D1 Sb-125 D2 Mn-54 D2 Co-57 D2 Co-58 D2 Fe-59 D2 Co-60 D2 Ag-110 D2 Nb-95 D2 Sb-125 D5 Mn-54 D5 Co-57 D5 Co"58 D5 Fe-59 D5 Co-60 D5 Ag-110 DS Nb-95 D5 Sb-125 D6 Mn-54 DG Co-57 DG Co-58 D6 Fe-59 D6 Co-60 DG Ag-110 D6 Nb"95 D6 Sb-125 1. 86E-05 1. 79E-05 6. 10E-07 1.41E-OG 1.26E-OG 2.15E-07 6. 37E" 05 6. 09E-05 8. 60E" 07 2. 75E-06 4. 32E-06 9. 59E-07 2. 01E-04 1. 86E-04 1. 40E-06 9.32E-OG 9.10E-06 6.38E-07 2. 79E-06 5. 14E-06 5. 96E-07 4. 17E-05 4. 64E-05 1. 43E-06 1.87E-05 1.79E-05 6.10E-07 1. 37E-06 1. 26E-06 2. 15E-07 6. 27E-05 6. 09E-05 8. 60E-07 3.30E-06 4.32E-06 9.59E-07 1.97E-04 1.86E-04 1.40E-OG 8.98E-06 9.10E-06 6.38E-07 1. 91E-06 2. 97E-06 5. 96E-07 4.01E"05 4.64E-05 1.43E-OG 1. 81E" 05 1. 79E-05 6. 10E-07 1.34E-06 1.26E-06 2.15E"07 6.24E-05 6.09E-05 8.60E-07 3.43E-OG 4.32E-06 9'9E-07 2.00E-04 1.86E-04 1.40E-OG 9. 21E-06 9. 10E-06 6. 38E-07 5. 76E-06 5. 14E-06 5. 96E-07 4.61E-05 4.64E-05 1.43E-06 1.88E-05 1.79E-05 6.10E-07 1. 24E-06 1. 26E-06 2. 15E-07 6.40E-05 6.09E-05 8.60E-07 2. 98E-06 4. 32E-06 9. 59E-07 2. 01E-04 1. 86E-04 1. 40E-06 8.96E-06 9.10E-06 6.38E-07 5.84E-06 5.14E-OG 5.96E-07 4.62E-05 4.64E-05 1.43E-06 1. 04 1. 13 1. 05 0. 64 1. 08 1. 02 0. 54 0.9 1. 04 1.1 1. 03 0. 76 1. 06 0. 99 0. 64 0. 86 1. 01 1. 07 1. 02 0. 79 1. 08 1. 01 1. 12 0. 99 1. 05 0. 99 1. 05 0. 69 1. 08 0. 98 1. 14
0. 75-1. 33 0. 5-2. 00 0. 80-1. 25 0. 5-2. 00 0. 80-1. 25 0. 6-1. 66 0. 6-1. 66 0. 75-1. 33 0. 75-1. 33 0. 5-2. 00 0. 80-1. 25 0. 5-2. 00 0. 80-1. 25 0. 6"1. 66 0. 5-2, 00 0. 75-1. 33 0. 75-1. 33 0. 5-2. 00 0. 80"1. 25 0. 5-2. 00 0. 80-1. 25 0. 6-1. 66 0. 6-1. 66 0. 75" 1. 33 0. 75" 1. 33 0. 5"2. 00 0. 80-1. 25 0. 5-2. 00 0. 80-1. 25 0. 6-1. 66 0. 6-1. 66 0. 75-1. 33
The licensee and NRC measurements with small uncertainties agreed well.
Measurements where the licensee counting uncertainty exceeded 35K were not included in Table 4.
The inspector reviewed a
riori determinations of liquid effluent lower limits of detection pursuant to technical specification 4. 11. 1. 1.
The data were obtained using blank samples, and the resulting LLDs were satisfactory.
In addition to the onsite gamma isotopic measurements, the inspector and the licensee retained portions of the liquid waste sample for intercomparison of Sr-89/90, Fe-55 and tritium.
These data will be used to address Open Item 50-275/88-33-01.
The final sample was containment atmosphere obtained from the Unit 1 containment purge system.
The results are presented in Table 5.
Table
Containment Atmosphere NRC.
Licensee NRC Random Ratio:
Agreement Result Result Uncertainty Licensee/NRC Range Analyte (uCi/ml) (uCi/ml) (uCi/ml)
Dl Ar-41 4.42E-07 6.48E-07 4.83E-08 0. 68 0. 6-1. 66 Dl Xe-133 5. 77E-07 7. 33E-07 4. 1SE-08 0. 79 0. 75-1. 33 D2 Ar-41 3. 75E-07 6. 48E-07 4. 83E-08 0. 58 0. 6"1. 66 D2 Xe-133 5.88E-07 7.33E-07 4. 15E-08 0. 8 0. 75-1. 33 D5 Ar-41 4.72E-07 6.48E-07 4.83E-08 '.73 0.6-1.66 D5 Xe-133 5.98E-07 7.33E-07 4.15E-08 0.82 0.75-1.33 D6 Ar-41 4.75E-07 6.48E-07 4.83E-08 0. 73 0. 6-1. 66 D6 Xe-133 6.27E-07 7.33E-07 4. 15E-08 0.86 0.75-1.33 The measurements of containment atmosphere noble gas were generally in agreement.
The inspector observed a decrease in the intercomparison ratios versus time, possibly due to leakage, from the licensee's sample container.
The inspector reviewed quality control data for radioanalytical instruments.
The frequency of checks and the parameters monitored were consistent with the guidance of regulatory guide 4. 15, "equality Assurance for Radiological Monitoring Programs..."
and ANSI/IEEE N42. 14
, "Measurement of Gamma-Ray Emission Rates Using Germanium Detectors."
The data were indicative of good instrument performance.
However, the licensee was not employing quality control charts to trend performance checks of the alpha/beta counting system.
Regulatory guide 4. 15, states
"The results of these measurements should be recorded in a log and plotted on a
control chart."
The licensee agreed that trend charts should be kept for the alpha/beta counter, consistent with other instruments, and committed to establish appropriate charts.
The inspector will review the licensee's corrective action in a subsequent inspection (50-275/90-03-02).
The inspector examined reports of the licensee's intracompany radiochemical cross-check program for 1989.
Mixed gamma isotopes, tritium and gross alpha/beta were analyzed on filters and in water.
The agreement criteria were similar to those employed by NRC.
The results were uniformly acceptable.
The licensee's radiochemistry program met regulatory requirements.
Areas which merited improvement included the licensee's determination of sample activity involving radiological parent-daughter relationships, and consistent quality control documentation.
5.
~Filet T
The inspector toured radiologically controlled areas of Units j. and
during the inspection.
Independent radiation measurements were made using NRC ion chamber survey instrument Model R0-2, Serial 008985, due for calibration on March ll, 1990.
The inspector observed the following:
a.
Radiation monitoring equipment was in current calibration.
b.
All personnel observed on tour were wearing proper dosimetry.
c.
Posting and labeling practices were consistent with 10 CFR 19. 11 and 20.203.
d.
A metal cap was removed from a survey port on a Unit 1 Chemical and Volume Control System demineralizer vault, with an electrical cord dropped through the open port and left unattended.
The inspector and the accompanying Radiation Protection Technician verified the dose rate from the port, determined that the area was properly posted, and reported the observation to Radiation Protection supervision.
The radiation protection practices and housekeeping observed during plant tours were satisfactory.
6.
Exit Interview 30703)
The inspector met with licensee management on January 26, 1990, to discuss the scope and findings of the inspection.
The inspector confirmed the commitments described in section 4.0 with licensee managemen Enclosure Criteria for Acce tin the Licensee's Measurements Resolution Ratio
<4
"
8
-
16
"
51
-
200 200 No comparison 0 5
-
2 0 0.6
-
1.66 0.75 -
1.33 0.80 -
1.25 0.85 -
1.18
~Com ari son l.
Divide each NRC result by its associated uncertainty to obtain the resolution.
(Note:
For purposes of this procedure, the uncertainty is defined as the relative standard deviation, one sigma, of the NRC result as calculated from counting statistics.)
2.
3.
Divide each licensee result by the corresponding NRC result to obtain the ratio (licensee result/NRC).
m The licensee's measurement is in agreement if the value of the ratio falls within the limits shown in the preceding table for the corresponding resolution.