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{{Adams
{{Adams
| number = ML20136J230
| number = ML20203F328
| issue date = 11/07/1985
| issue date = 04/18/1986
| title = Insp Rept 50-293/85-27 on 850916-20.Deviation noted:20 to 30 Ft of Redundant Drywell Detector Cable Not Installed in Conduit
| title = Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-293/85-27
| author name = Cheung L, Nimitz R, Pasciak W
| author name = Martin T
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
| addressee name =  
| addressee name = Harrington W
| addressee affiliation =  
| addressee affiliation = BOSTON EDISON CO.
| docket = 05000293
| docket = 05000293
| license number =  
| license number =  
| contact person =  
| contact person =  
| case reference number = RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-2.F.1, TASK-3.D.3.3, TASK-TM
| document report number = NUDOCS 8604250123
| document report number = 50-293-85-27, NUDOCS 8511250309
| title reference date = 12-13-1985
| package number = ML20136J205
| document type = CORRESPONDENCE-LETTERS, NRC TO UTILITY, OUTGOING CORRESPONDENCE
| document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| page count = 2
| page count = 37
}}
}}


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t U.S. NUCLEAR REGULATORY COMMISSION
  ::c 1-  APR -181986 p
 
k h ' Docket No. 50-293.
==REGION I==
Report No. 50-293/85-27 Docket N License No. OpR-35 Priority --
Category C Licensee: Boston Edison Company M/C Nuclear 800 B6ylston Street Boston, Massachusetts 02199 Facility Name: Pilgrim Nuclear Power Station Inspection At: Plymouth, Massachusetts Inspection Conducted: September 16-20, 1985    !
Inspectors: SLd d R. L. Nimitz, Sen%r Radiation Specialist 11/ 6 ! 6 5 date 2A.0%
L. '5. Ihueng, Reactor Engin er n//
date t'W A. P. Hull, Brookhaven National Laboratory W. H. Knox, Knox Consultants, (contractor to Brookhaven boratory)
fatin   /
Approved by: . (A.o k / A__
W. J .' /lasciak, Chief, BWR Radiation
_ // (7 f[  *
      ''/ d/ti Prot & tion Section    ( (
InspectionSummary:(
Inspection on September 16-20, 1985 (Report No. 50-293/85-27)
Areas Inspected: Special, announced safety inspection of the licensee's implementation and status of the following task action items identified in NUREG-0737: II.B.3, Post Accident Sampling Capability; II.F.1-1, Noble Gas Effluent Monitors; II.F.1-2, Sampling and Analyses of Plant Effluents; II.F.1-3, Containment High-Range Radiation Monitor; III.D.3.3, Improved Inplant Iodine Monitoring. The inspection involved 140 hours onsite by two region-based inspectors and two contractors from Brookhaven National Laborator Results: No violations were identifie Several areas requiring improvements were identifie PDH ADOCK 050 g3      1 G
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_. -- _. _ - _ _ _ _ _ _ _ _ _ . - _ _ _ .__ _ ._______ _________ ___ _ _ _ _ _ _ _ _
Boston Edison Company M/C Nuclear ATTN: Mr. William D. Harrington Senior Vice President, Nuclear 800 Boylston Street-Boston, Massachusetts .02199
=>
  ' Gentlemen:
r DETAILS 1.0. Persons Contacted f
!  The individuals contacted during this inspection are listed in Attachment I to this inspection report.
 
l 2.0 Purpose
;
!  The purpose of this inspection was to verify and validate the adequacy of i  the licensee's implementation of the following task actions identified in i  NUREG-0737, Clarification of TMI Action Plant Requirements:
l l  Task N Title l  II. Post-Accident Sampling Capability l  II.F.1-1  Noble Gas Effluent Monitors j  II.F.1-2  Sampling and Analysis of Plant Effluents i
II.F.1-3  Containment High-Range Radiation Monitor III.D. Improved Inplant Iodine Instrumentation under Accident Conditions
'  In addition, and as part of the inspection, a review was performed to verify and validate the adequacy of the licensee's design and quality assurance program for the design and installation of the Post-Accident
!
Sampling System (PASS). The findings in this area are presented in Section 9 of this report.
 
j 3.0 TMI Action Generic Criteria and Commitments l The licensee's implementation of the task selection actions specified in Section 2.0 were reviewed against criteria contained in the following
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documents:
l  * NUREG-0578, THI-2 Lessons Learned Task Force Status Report and l  Short-Term Recommendations, dated July 1979, I  *
Letter from Darrell G. Eisenhut, Acting Director, Division of
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Operating Reactors, NRC,'to all Operating Power Plants, dated October 30, 1979,
Subject: Inspection:No. 50-293/85-27 .
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iThis refers to your letter dated December. 13, 1985, in response to our letter
  *
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NUREG-0737, Clarification of TMI Action Plan Requirements, dated l Ovember, 1980, l
dated November 13, 1985.
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:Thank you for informing us of the corrective and-preventive actions documented in your letter. These actions will be examined during a future inspection of your. licensed program.
  *
Generic Letter 82-05, Letter from Darrell G. Eisenhut, Director, l
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Division of Licensing (DOL), NRC, to all Licensees of Operating Power Reactors, dated March 14, 1982,
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Pilgrim Nuclear Power Station, Unit 1, Updated Final Safety Analysis Report, dated July 11, 1982, l
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    .-.  . . - - . -. _ _ _
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Letter From Darrell G. Eisenhut, Director, Division of Licensing, NRR to Regional Administrators, " Proposed Guidelines for Calibration and Surveillance Requirements for Equipment Provided to Meet Item II.F.1, Attachments 1, 2, and 3 NUREG-0737," dated August 16, 1982,
  *
Order confirming Licensee Commitments on Post-TMI Related Issues, dated March 14, 1983,
  * Modification of March 14, 1983 Order, dated June 15, 198 *
Regulatory Guide 1.3 " Assumptions Used for Evaluating Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors,"
  * Regulatory Guide 1.97, Rev. 2, " Instrumentation of Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," and
  *
Regulatory Guide 8.8, Rev. 3, "Information Relevant to Ensuring that Occupational Radiation Exposure at Nuclear Power Station will be As Low As Reasonably Achievable."
 
i In addition specific review criteria and/or commitments relative to each task item numbers are included in Attachments 2-7 of this repor .0- Post-Accident Sampling System, Item II. .1~ Position NUREG-0737, Item II.B.3, specifies that licensees shall have the capabil-ity to promptly collect, handle, and analyze post-accident samples which are representative of conditions existing in the reactor coolant and con-tainment atmosphere. Specific criteria are denoted in commitments to the NRC relative to the specifications contained in NUREG-073 Documents Reviewed The implementation, adequacy and status of the licensee's post-accident sampling, monitoring, and analysis systems were reviewed against the
,
criteria identified in Section 3.0 and in regard to licensee letters,
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memoranda, drawings and station procedures as listed in Attachment 2 of this Inspection Repor The licensee's performance relative to these criteria was determined from
!  interviews with the principal personnel associated with post-accident l  sampling, reviews of associated procedures and documentation, and the l  conduct of a performance test to verify hardware, procedures and personnel capabilities.
 
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4.2 Findings Within the scope of the review, the following items items were identified:
Your cooperation with us is appreciated.
4. System Description and Capability The licensee.has installed a Post Accident Sampling System which is a standard General Electric design. It has the ability to obtain unpressurized undiluted and diluted samples of reactor coolant from the jet pump and the RHR System. Also, samples can be obtained from the drywell, suppression pool and reactor building atmosphere Redundant containment hydrogen analyzers provide a hydrogen analysis back-up capabilit Analysis for chloride, boron, and hydrogen _are conducted in the laboratery, using an ion chromatograph, plasma spectrometer and gas chromatograph, respectively. Back-up analysis capability is being negotiated with other nearby utilitie .
4.2.2 PASS Performance Testing Grab samples of reactor coolant and of the drywell atmosphere were collected during an operational test on September 18, 1985. During the test, licensee personnel verified the intergrated ability to collect and analyze samples within the time constraints of NUREG-0737, II.B.3. (See Attachment 3 of this report.)


4.2. Reactor Coolant (Findings)
Sincerely, h (
The reactor coolant sampling system is designed to obtain samples of liquids and dissolved gasses during all modes of operation. Although samples could be obtained from all sampling points, the following matters requiring licensee attention were identified: Demonstrate the adequacy of system purge times. Data were not presented to demonstrate the adequacy of the purge times speci-fied in procedures. Adequate purge time is needed to ensure representatives sampling. (50-392/85-27-01) Demonstrate the adequacy of the sample dilution method and related equipment. Data were not available to demonstrate the adequacy of the sample dilution method and related equipmen Sampling dilution is a key element in the quantification of sample result (50-293/85-27-02)  , Determine the volume of the coolant collection ball valv During the preoperational testing, the ball valve, wh!,ch collects
      %cw Thomas T. irtin, Director Division of Radiation Safety and Safeguards cc:
 
  . A. V.~ Morisi, Manager, Nuclear Management Services Department C. J.- Mathis, Station Manager Joanne Shotwell, Assistant Attorney General Paul Levy, Chairman, Department of Public Utilities W. J.. Nolan, Chainnan,. Plymouth Board of Selectmen Plymouth Civil Defense birector    ..
  .
Senator Edward P. Kirby Public Document Room (PDR)'
 
-Local'Public Document. Room (LPDR)
5
. Nuclear Safety Information Center (NSIC)
  ~
NRC Resident Inspector Connonwealth of Massachusetts (2)
a measured volume of coolant, was determined to be 0.14 ml, instead of the value of 0.10 ml. The valve of 0.10 ml. was presented in procedures. The valve, which was actually_ tested during preops, has since been replaced, but the new valve's volume has not been determined. (50-293/85-27-03) Repair and/or replace the flow control check valve. After the primary system tests had been successfully conducted, the flow control valve stuck open in a fixed position (0.6 GPM). The design flow rate of 1 GPM could not be attaine (50-293/
8604250123 860418 = #"
85-27-04)
PDR- ADOCK 05000293 G PDR-y 0FFICIAL RECORD COPY o
4.2.3.2 Containment Air (Findings)
Atmosphere samples can be obtained from the Drywell, Reactor Build-ing and Suppression Poo The following matters requiring licensee attention were identified: Demonstrate the adequacy of system purge times. Data were not presented to demonstrate the adequacy of the purge times speci-fied in-procedures. Adequate system purge time is needed to ensure representative sampling. (50-293/85-27-05) The capability to obtain a representative sample of containment atmosphere should be demonstrated. (50-293/85-27-06): There has been no line loss or plate-out study conducte . The sa;npling assembly is not heat traced. This could lead to excessive condensation collecting on the iodine car-tridges. This problem would be particularly troublesome with high humidity in containmen Correct the air sampler rotometer reading for differences in air density created by pump suction. The sampling procedure (5.7.11) uses the rotometer reading in the calculation of the radioiodine concentration. (50-293/85-27-07)
4.2.4 Analytical Capability (Findings)
Attachment A to BEC0's May 30, 1985 letter contains the licensee's commitment relative to the range, sensitivity and type of analytical capability availabl .
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Boston Edison Company M/C Nuclear 2
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4.2.4.'1. Chloride (Findings)
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Chloride analysis of diluted coolant is conducted using an ion chromatograph. The lower range sensitivity commitment is 0.01 pp However, the water used for dilution has a higher chloride 1 concentration (approximately 0.03. ppm).
Region I Docket Room (with concurrences)-
Management Assistant, DRMA (w/o encl)
, Section Chief, DRP.


In order.to detect 0.01 ppm, analysis would have to be conducted using an undiluted sample. The licensee's May 30, 1985 letter in-dicated'that'an undiluted sample would be collected and retained for up to 30 days for subsequent confirmatory analysis. However pro-cedures did not provide for the collection and retention of an un-diluted sample for further chloride analysis. The accuracy of +/-10%
W. Raymond, SRI, Vermont Yankee
at the 0.01 ppm level could not be achieved. (See Attachment 3 for test results).
'T. Shedlosky, SRI, Millstone 1&2 lH. Eichenholz, SRI, Yankee P. Leech,. LPM, NRR
/
RI:DRSS RI:D SS Gb.V RI: RSS Nimitz/mmb Pasciak Bellamy
  't 4 12/3c/85 42/8/85 R/8 /85 f!LA 0FFICIAL RECORD COPY 50-293/85-222 - 0002.0.0 12/23/85


The following matters requiring licensee attention was identified:
* The detection limit and sensitivity of the chloride analysis method should be clearly define Procedures'should be revised to include provisions for the retention of an undiluted sample for up. to 30 days for more detailed chloride analysis. (50-293/
85-27-08)
u4.2.4.2 Boron (Findings)
Boron analysis is conducted using two methods: Plasma Spectrometry and Carminic Acid (Spectrometry). The analysis of spiked samples were conducted using the primary method (Plasma Spectrometry) and
: acceptable results were obtaine Reagents needed to conduct back-up analysis using the Carminic Acid method were not available. Hewever, they were on orde The following matter requiring licensee attention was identified:
* Obtain and maintain supplies necessary to conduct boron
' analysis with carminic acid method. (50-293/85-27-09)
.4.2. pH Analyses (Findings)
Although the licensee has purchased the equipment for performing pH analyses,.his procedures for this purpose were not established. The licensee personnel indicated that this analysis.is unnecessar The following matter requiring licensee attention was identified:
* Clarify commitments and capabilities relative to pH analyse I (50-293/85-27-10)
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L 4.2.4.4 Gross Gamma and Isotopic Analyses (Findings)
An isotopic analysis of the PASS grab sample was compared to the normal sink sample. The comparison was within acceptable limit Provisions had not been made to conduct an isotopic analysis on the depressurized dissolved gasses in core damage procedure The following matter requiring licensee attention was identified:
*
Review and evaluate the need to perform isotopic analyses of dissolved gasses. If a need is identified, provision should be made to conduct isotopic analyses on dissolved gasses. (50-293/
85-27-11)
4.2.4.5 Hydrogen and Dissolved Gas (Findings)
Licensee personnel differed in their views on the type of analysis to be conducted and how the data would be used. The core damage assessment procedure uses the hydrogen content to assess core damag Procedures are available for the analysis of hydrogen, using an ion chromatograp Licensee personnel indicated that the primary analysis method to quantify hydrogen was the use of total dissolved gas. However, the procedure for this purpose had not been completed. Additionally licensee personnel indicated that the total gas, once determined, was to be equated to hydrogen and used as such in core damage assessmen Others indicated that the total gas data would be obtained but would not be used. It was not clear that the total gas obtained was equi-valent to total hydrogen. Equating hydrogen to total gas could lead-to an overly conservative estimate of the hydrogen conten Additionally, tests have not been conducted to demonstrate that the committed range and accuracy of either dissolved ges or hydrogen could be achieve The hydrogen analysis procedure required that 0.25 cc of gas be injected into the gas chromatograph. During the test, approximately 5 cc was collected and transported to the laboratory. The bringing of excessive amounts of gas into the laboratory could unnecessarily increase radiation exposur .
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Although the procedure directed the operator to inject 0.25 cc into the gas chromatograph, a method had not been developed to extract this volume from the large syringes in which the gas is transporte The operator indicated that an undetermined volume, up to 5 cc would be injected directly from it. Since the calibration gas volume is 0.25 cc, the GC will respond differently if a 5 cc volume is use The following matters requiring licensee attention were identified:
(50-293/85-27-12): The use of dissolved gas or hydrogen analysis data in the as-sessment of core damage should be clearly specified. Also, it should be demonstrated that the committed range and accuracy can be achieved. Procedures should be established and implemented where neede The amount of gas transported to the laboratory for analysis should be minimize Personnel should be instructed either to follow procedures or the procedures should be revised to reflect actual practice Incorrect gas volumes were injected into the gas chromatograp .2.5 Additional Findings The following additional findings were identifie These matters require _ licensee attention: Establish a routine maintenance program for the PASS. Not all components of the PASS have been included in a regular surveill-ance/ calibration program. Also, there does not appear to be a formal administrative procedure for assuring that non-safety related equipment, such as the PASS, is incorporated in the routine maintenance (50-293/85-27-13) Establish a spare parts program for the PAS A PASS spare parts program is under development. The licensee is coordinating this program development with the BWR Owner's Group. (50-293/85-27-14).
; Establish PASS sample shipping procedure The procedures for the handling, loading and off-site shipment of samples are in a draft form and arrangements are being made through the PIMS to supply the needed shipping cas (50-293/85-27-15)
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9 Complete arrangement for backup off-site analyses of sample Arrangements for backup off-site analysis are incomplete. A verbal agreement exists with Millstone to provide backup sup-por This agreement is being formalized by the licensee's legal staf (50-293/85-27-16) Establish provisions for detection of leakage across chille Provisions have not been made for the detection of leakage of reactor coolant into the chiller cooling water. The pressure differential between the sampling and the cooling lines could result in the buildup of significant activity in the chiller in the event of such a lea (50-293/85-27-17)
f. Provide provisions to preclude PASS from exceeding its design temperature specificatio The chiller will automatically shutdown on low or high pressure signals. However, mechanisms have not been provided for the interruption of the flow of the uncooled water into the PASS or of operator actions procedures to prevent the PASS from exceed-ing its temperature design specifications. (50-293/85-27-18) Improve the safety of the access to the PAS Under accident conditions personnel may be required to climb a 20' ladder to the location of the PASS controls while wearing protective clothing and SCBA gear. The ladder does not have a cage to prevent a person from falling backward. (50-293/
85-27-19) Establish Alarm Set Points for the PASS Area Radiation Monito Consider using an alarm set point based on maximum dose rates that could be encountered such that 10 CFR 50 GDC 19 dose limits will not be exceeded. (50-293/85-27-20) Evaluate the need to install charcoal filters in the exhaust hood of the chemistry laborator A charcoal filter has not been provided for the laboratory ventilation exhaust system. During an accident radioiodine releases may result from the operation of the gas chromatograph, plasma spectrometer and laboratory hood. (50-293/85-27-21) Calibrate the PASS radiation detector in accordance with manufacturer's specification The three radiation detectors associatsa with the PASS are not calibrated in accordance with manufacturer's specifications and
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recommendations. For example, the manufacturer requires the one electronic channel to be set at 2200 volts and the other two channels at 2500 volts. Based on the calibration procedure, all electronic channels were set at 2500 volts. Also, there was no in-situ check of the instruments response following rein-stallatio (50-293/85-27-22) Reevaluate the adequacy of the " time and motion" study performed for collection of PASS samples. Ensure the re-quirements of 10 CFR 50 GDC 19 can be me The " Time and Motion Sedy" was conducted using generic methods before the actual procedves were developed. It is not clear that the samples can be co'lected and analyzed within GDC limits using the existing procedures. (50-293/85-27-23). Clarify use of pressure in procedure Procedure 5.7.4.1.9 does not include provisions for the speci-fication of the pressure in units of inches of Hg. Procedures use units of psig. Subatmospheric readouts in " inches of Hg" are the most appropriate readouts. (50-293/85-27-24) Establish reliable backup power for the chiller.
l A reliable source of backup power has not been provided for the
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chiller used for cooling the incoming reactor coolant in the
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event of a loss of off-site power. While a backup method of {
l  cooling has been devised, it has not been tested for proper i
hose fitting. Also, the heat removal capability of this method nas not been establishe (50-293/85-27-25) Identify and provide " carrying devices".
l Procedure 5.7.3.1.2 states that a " carrying device" would be used to transport the syringe containing radioactive gasse It is not clear what type of " carrying device" is to be use ;
,  (50-293/85-27-26)    '
4. Item for Improvement The following improvement item was identified:
Consider / identify methods to ensure adherence to PASS procedure One operator was solely responsible for the operation of the system and the collection of the sample. A mechanism for verifying that the correct procedure steps had been properly followed or assuring that the correct information had been recorded has not been provide ,. .
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5.0 Noble Gas Effluent Monitor, Item II.F.1-1 5.1 Position NUREG-0737, Item II.F.1-1 requires the installation of noble gas monitors with an extended range designed to function during normal operating and accident conditions. The criteria, including the de-sign monitors for individual release pathways, power supply, cali-bration and other design considerations are set forth in Table I F.1-1 of NUREG-073 Documents Reviewed The implementation, adequacy, and status of the licensee's monitoring systems were reviewed against the criteria identified in Section and in regard to licensee's letters, memoranda, drawings and station procedures as listed in Attachment 4 of this Inspection Repor The licensee's performance relative to these criteria was determined by interviews with the principal per sons and consultants associated with the design, testing, installation and surveillance of the high range gas monitoring systems, a review of the associated procedures and documentation, an examination of personnel qualifications and direct observation of the system .2 Findings Within the scope of this review, the following was identified: .
5. Description and Capability There are three atmospheric release locations at Pilgrim Nuclear Power Station (PNPS), the free-standing main stack (to which the effluent from the SBGTS is ducted), the reactor building exhaust stack and the turbine building roof exhaus The licensee has i. stalled high re.nge ion chambers to supplement the pre-existing normal range monitors for the main stack and for the reactor building vent. An ion chamber has also been installed to provide high range monitoring for the turbine building roof vent, which is not normally monitore .2.1.1 Normal Range Description The normal range monitors for the main stack and the reactor building vent consist of identical GE designed shielded gamma sensitive Na! detectors which view a volume of gas in a shielded chamber. Each has a seven decade range. From calculations supplied by the licensee's consultant, ENTECH, the sensitivity of the main stack monitor appears to be 4.4 x 10 -7 -4.4 uCi/cm3 and that for the reactor building vent appears to be 6 x 10 ~0 - 0.6 uC1/cm3 .
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12 The turbine building is not normally monitored by installed instru-mentatio .2.1.1 High Range Description General Description General Atomic Model RD-2A fon chambers, with a range of 10 -I -10 4 R/hr are employed for high range monitoring. For the main stack one is installed externally to a 20" horizontal duct at its base, while those for the reactor and turbine buildings are externally mounted at elevated levels on the vertical ducts near the roof vent High Range (Main Stack)    i The licensee's submittal of February 27, 1981 indicated a range for the stack monitor of 10 1
  - 106 uti/cm 3equivalent 133 Xe concentra- I tion. A subsequent licensee submittal of June 4,1984 indicated an upper limit of 2 x 105 uC1/cm 3 133 Xe equivalent. The basis for these l
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submittals could not be established at the time of the inspection.
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From calculations supplied onsite by ENTECH, the apparent range of l the ion chamber monitor for the high range main stack is 60 - 6 x 6 133 10 for Xe, in which case overlap with the normal range monitor
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would not be provided. However, the attenuation of the 3/8" thick duct well is embodied in ENTECH's calculations. When this is taken into account, the effective range of this monitor appears to be
0.7 - 7 x 10 uC1/cm3 on a 133 Xe equivalent basi ENTECH's calculations showed that in the event of a design basis accident, there would be an initial overlap for the immediate post-accident mixture. However, the overlap would narrow with elapsed time post-accident, as the average energy of the mixture decrease For a design basis accident, it would be very narrow before the upper range of the low range monitor was reached some 100 hours late High Range (Reactor Butiding}
In its submittal of February 24, 1981, the licensee indicated a range for the High Range monitor for the reactor building vent of 8 x 10 -2 3 3 to 8 x 10 uC1/cm while that of June 14, 1982 indicated an upper
limit of 2 x 10 uC1/cm ,3 133Xe equivalen From ENTECH's calculations reviewed onsite, the range of this monitor
appears to be 1 - 10 uC1/cm3 of 133 Xe. This monitor also showed a range overlap for the immediate mixture which also narrowed with time post-acciden _ _ - _ _ _ _ _ _
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s High Range (Turbine Vent)          ,.,
The licensee's submittal of February 17,\ 1981 indicated that the
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apparent range of the turbine vent monit'or was 1.5 x 10-2 3,5 x      *
            .
103 uCi/cm 3
  , which agreed with the upper limit indicated in the sub-      - ,'
mittal of June 14, 1982. From ENTECH's calculations, its apparent      >
range was approximately 3 x 10 -2 -3 x 103 uCi/cm3 of    133 Xe. The upper limit meets the criterion of the NURCG-0737, II.F.1 Attachment      -
1. Since NRR has accepted the licensee's contention that low range      -
monitors are not appropriate for the turbine roof vents, as they are        '
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            ,
not normal release points, the question of range overlap is not ap-        ,
plicabl .
5.2.1.3 General Findings (High Range Monitor)
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The possible effect on monitor indications of the deposition of post accident radiotodines on duct walls has not been considered in ENTECH's calculation The initial type calioration of the RD-2A ^
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          -  e-ion chambers by the vendor (General Atomics) remains in open item
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from Inspection 50-293/83-02. However, data were supplied to dem-      -
onstrate that the installed chambers are regularly calibrated against a solid source throughout their range. Although.ENTECH's code has not been verified with gaseous sources, they supplied a report to demonstrate that the one used to establish the radiation levels at the installed location of the ion chambers had been bench-marked against an ANSI approved cod Due to the limited range overlap between the low and high range      ._
monitors for the main stack and the reactor building vent, the ability of the former to function throughout their stipulated ranges      ,
and beyond for extended period of time (up to a few days) and to re-      ,
cover therefrom is an important element of the licensee's ability to      ;
follow a post-accident release. This ability has not been documente Local readout and remote readout and recording in the Control Room of the indications of the high range monitors in units of R/nr is        'b provided. The interpretations of these indications in terms of re-lease rates and/or concentrations can be niade by means of nomograms      '
or a computer progra Initial operator responses for unusual rele'ases are based on in-      ,
dications of the low range monitors, for which appropriate alarm        '
levels have been established. There are no alarms on the high range        s channels, nor are there specific provisions in emergency procedures whereby the operator is directed to consult the *
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The licensees indicated that an ion chamber was normally maintained in its spare parts inventory. However, it had been utilized as a replacement following a recent lightning induced failure at the main stack. Another spare was on orde .2.2 High Range Monitors (Findings)
The installed system meets the guidance for high range noble gas monitoring as contained in NUREG-0727, II.F.1, Attachment 1. The following matters requiring license attention were identified: Clearly specify the range of all High Range Noble Gas Monitor The reasons for the inconsistencies between the range of capa-bility of the high-range monitors as can be derived from the reports by ENTECH and from information supplied by the licensee to NRC should be investigated. Also range overlap should be clearly specifie (50-293/85-27-27) The ability of the low range monitors to function for sustained periods of time at concentrations close to and beyond their upper range for and to recover therefrom during a post-accident sequence should be established. If this cannot be satisfac-torily acccomplished, provisicns for turning off the power to thc,and/or bypassing them during these periods of time should be c.onsidere (50-293/85-27-28) The possible effect of radioiodines deposited within ducts on the response of the high range monitors should be considere If it is appreciable, considerations should be given to relo-cating the detectors to a shield cave within which they view a suitable volume of an off-line aliquot of the stack / vent flo (50-293/85-27-29).
6.0 Sampling and Analyses of Plant Effluents, Item II.F.1-2 6.1 Position ,
NUREG-0737, Item II.F.2, requires the provision of a capability for the collection, transport, and measurement of representative samples of radio-active iodines and particulates that may accompany gaseous effluents following an accident. It must be performable without exceeding specified dose limits to the individuals involved. The criteria including the design basis shielding envelope, sampling media, sampling consideration, and analysis considerations are set forth in Table II.F.1- .
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6.2.1' Description and Capability The licensee has elected to utilize the pre-existing normal sampling arrangements for compliance with NUREG-0737, II.F.1. Attachment As such, they are integral wfth the low-range gas monitors which are routinely employed for the sampling of iodines and particulates in the effluent from the main stack and the reactor building ven .The probe for the main stack is installed at an elevation of about 25', at which point it samples the combined flow from the SBGTS duct of 4,000 cfm and of stack dilution air at a rate of 16,000 cfm. The sampling line is heat traced and has a relatively-short run to the sampling station, which is located at the base of the main stac The probe for the reactor building vent is installed at the 163'
level near the top of this 9' x 9' exhaust duc The licensee's-calculations indicate that this probe is sized to be isokinetic for a stack flow rate of 65,000 cfm. No correction features were available for other flow rates, such as the much lower ones which would occur during accident condition The sample line has over a 100' vertical run, which is followed by a relatively short horizontal run to the sampling and monitoring station, which is located on the turbine deck adjacent to the stack. This sample line is not heat trace For routine sampling, a particulate filter and charcoal base iodine collection canister are installed in'a filter holder which is mounted with quick disconnect fittings on the sample station rack ahead of
'
the low range gas monitor. This filter holder is unshielded. The emergency procedure calls for.the replacement of the in place can-ister with one containing a silver zeolite collection medium at the onset of accident-condition The licensee's procedures envision the use of a nomogram to estimate-the 131 I activity released to the environment, based on gamma dose rates of the cartridg However, this is not specifically called for in the sampling procedures for the main stack and the reactor build-ing vent. Also the nomograms which are contained in the procedures permit only an estimation of the activity on the filter. Further-more, as written, these procedures call for the setting aside of sample holders with contact dose-rates that are above the limit for
" hand-touching". The sample holders are placed in a shielded cave adjacent to the sampling station, prior to any other measurement of the These procedures also call for a purge step prior to the removal of the filter holder from the sample rack. During the inspection, it was found that the purge enters the sample line subsequent to the holder. As a result, the purge only purges the gas monitor chamber The sample racks also contain a number of unlabeled valve .
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CrevoM EDCCM COMPANY eDO GovLStoM STRes?
SOSTON, M ASSACNuSETTS O2199
,
11MLLIAM O. MARRINGTON
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===*aaa December 13, 1985 SECo Ltr. #85-222 Thomas T. Martin, Director Division of Radiation Safety and Safegaurds U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406 License No. DPR-35 Docket No. 50-293 Subject: Response to Deviation as contained in NRC Inspection Report 85-27


The procedure for sampling the effluent from the turbine building fo iodines and particulates under the accident conditions calls for the use of a portable air sample pump and filter holder to collect a 10 minute sample at a pre-established location on the turbine deck lev-e It also calls for the measurement of the gamma dose rate of the filter with a survey instrument and the estimation of the activity on the filter, using the same nomogram that is contained in the stack and reactor building vent emergency sampling procedure The licensee was unable to furnish any calculations that at the shielding design basis, plant personnel could remove samples, replace the sampling media and transport samples to the on-site analysis facility with radiation exposures that would not exceed the GDC-19 criteri A review of the licensee's manual dose assessment procedures indi-cated that they provided only for the use of assumed iodine to noble gas ratios and did not provide for the use of the measured radio-fudine activity on effluent sampling media for the estimation of field iodine dose rate . Sampling Plant Effluents (Findings)
==Dear Mr. Martin:==
The following matters requiring licensee attention were identified:
This letter is in response to the subject deviation contained in NRC Inspection Report 85-27, conducted by Mr. R. Nimitz of your office on September 16-20, 1985 at Pilgrim Nuclear Power Station.
! Establish provisions in procedures for the sampling of the main stack and the reactor building vent (similar to those now con-tained in these for sampling in the turbine building) for the measurement of samples of radiation levels up to and including design basis, so as to accomplish continuous sampling throughout a post-accident sequence. (50-293/85-27-30) Perform a time and motion study to ascertain that the system design will make it possible to remove and transport design basis samples within GDC-19 criteria. All appropriate source terms should be used for this study. (50-293/85-27-31) Provide the necessary nomograms and calculator / computer pro-cedures, whereby measurement of the collected radioiodine activities can readily be translated into release concentra-tions and rate (50-293/85-27-32) Develop correction factors for the non-isokinetic sampling rates for the range of stack flow rates of the unit vent especially for those anticipated under accident conditions. (50-293/
85-27-33)


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The subject deviation and Boston Edison's response is enclosed as an attachment to this letter.
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  . Evaluate the capabilities of the sampling system to collect
If you should have any further questions regarding this matter, please do not hesitate to contact me.
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representative samples under accident condition. (50-293/
85-27-34)
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6.2.3 - Additional Items for Review / Resolution In order to reduce the anticipated dose rate of the collected sample, and/or to facilitate its analysis, the licensee should review and/or resolve (as appropriate) the following matters: (50-293/85-27-35)
f Provide shielding for the effluent sample holder ' The provision of' labels and a flow diagram for the valves and indicators on the sample rack The heat tracing of the sampling line for reactor building vent,
.
  'in view of the possibility that it may contain steam leakage or  ,
  - moisture therefrom under accident condition ' The provision of featJres which will- enable the purging of in-place sample canister and nearby sample' lines with an clean air supply, prior.to their removal for transport and analysis, e The addition to manual dose assessment-procedures of nomograms which would make it possible to estimate field iodine dose rates on-the basis of measured radiciodine activity (or release rates derived therefrom).


7.0 Containment High-Range Radiation Monitor, Item II.F.1-3 7.1 Position .
Very truly yours, WS M W. D. Harrington Attachment l
NUREG-0737, Item II,F.1-3, requires the installation of two in-con-tainment radiation monitors with a maximum range of~1 rad /hr to 10'
n
,-  rad /hr (beta and gamma) or alternatively l' R/hr to 107 R/hr(gamma
      ,
        ,
    -
only). The monitors shall be physically separated to view a large portion of~ containment and developed and qualified to function in an accident environment. :The monitors are also required to-have an energy response as specified in NUREG-0737,~ Table II.F.1- ~
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Review Criteria
    ' )
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L  .The implementation adequacy, and status of the installed in-contain-ment high range' monitors were reviewed against the criteria set
  .forth in Section 3.0 of this report and in regard to interviews with cognizant licensee personnel, licensee letters, station procedures, as-built prints and drawings as listed in Attachment 5 to this
,  Inspection Report.
 
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,w- e r e r~~--.e ,<-m , , - - , ,- ~-wr .,e, ,r~r- -
e ,, --ww, , . - - -w,n-- wr
 
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The licensee's performance relative to these criteria was determined by:
* Interviews with cognizant personnel;
*
Review of applicable operational and emergency plan procedures;
*
Review of applicable lesson plans and training records;
* Review of calibration data; and
* Direct observation of installed equipmen .2 Findings Within the scope of this review, the following was identified:
Installation / Placement The licensee has installed four- (4) gamma sensitive ion chambers for monitoring of gamma radiation emanating from primary containmen Two of the detectors monitor the torus at the North and West areas of the Torus respectively (-17' elevation). The remaining two de-tectors are mounted in close proximity to each other (about 2 feet apart) at approximately the north mid plane area of the Drywel These latter two detectors project into primary containment through two caped penetrations. The location and number of the detectors was verified. The number and location of these detectors was found acceptable to the NR The detectors read out at the Post-Accident Monitoring (PAM) Panel, located in.the Main Control Room. The read outs were operating properl Procedures The licensee has established procedures which included monitor response curves. An estimate of core damage may be made by: obtaining a reading from a detector; determining time after shutdown; and usirig an appropriate curve to estimate core damag Different curves are used for the Torus and Drywell detectors.
 
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.The monitor response curves incorporate corrections for inherent shielding of the steel penetration liner in which the detectors are inserte Procedures are in place for calibration and periodic verification of operability of the detector Environmental / Seismic Qualification See section 9 of this repor Training / Qualification of Personnel The licensee has provided training and qualifications of both oper-ations and radiological controls personnel in use/ interpretation of detector readings. Appropriate personnel are provided periodic re-fresher training in use/ interpretation of the read out Calibration-Range The detectors were found to be calibrated in accordance with NUREG-0737. The range of the detectors is consistent with NUREG-0737 Table II.F.1- .3 Acceptability The installed system can be considered to meet the guidance specified in NUREG-0737, Attachment II.F.1-3. However, the following matters requiring licensee attention were identified: Fully establish and implement the operator Training Modules for operation /use of the PAM Panel (Module CT-SH/IG-S-120). l (50-293/85-27-36). Train and qualify appropriate personnel on Procedure 5.7.5,
  " Estimating Core Damage". Personnel have not been trained in the new procedur (50-293/85-27-37) Include North / East Torus Monitor Response Curve in Procedure 5. The procedure contains only the Drywell monitor response curve. However, the Drywell response curve is being incorrectly used with the Torus monitors (50-293/85-27-38) Review dose / damage response curves for the Dryweli Monitor At less than 100% core damage, no other radiation ';curce ( primary lines in area of detectors) are used as contributors to the detector readings. A review should be performed to ensure that sources in the area of the detectors (other than the gas-eous activity in primary containment) do not adversely effect the core damage estimates. (50-293/85-27-39)
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8.0 Improved In-Plant Iodine Instrumentation Under Accident Conditions, Item III.D. .1 Position NUREG-0737, Item III.D.3.3, requires that each licensee shall-provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during an acciden Review Criteria The implementation, adequacy and status of the licensee's in plant iodine monitoring under accident conditions were reviewed against the criteria in Section 3.0 of this report and in regard to the documents identified in Attachment 6 to this Inspection Report. The licensee's performance relative to these criteria was determined by:
      -. - _ _ _ _ _ _ _ _ _ _ _
  * Interviews with cognizant licensee personnel;
    ' ; .1. i
* Review of applicable operational and emergency plan procedures;
        '
* Review of applicable lesson plans and training records;
,v- '
* Direct observation of performance during a walkthrough; and
  .a y r$43A:.i$.'ss.*.. M J~fi.W .h W s4 :.e i'*'<.. 4 . .w.' '.,.'}.b W, .: \. v. ?' .'d'''
* Verification of equipment availability and storag .2 Findings Within the scope of this review, the following was identified:
Equipment The licensee was found to have appropriate equipment for sampling and quantifying airborne iodine in areas within the facility where plant personnel may be present during an acciden Procedures The licensee has established procedures for calibration / operation of the equipment used to determine concentration of airborne iodin Training The licensee provides initial training and refresher training to personnel responsible for calibration / operation of airborne iodine collection and analysis equipmen .
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8.3 Acceptability    .
The licensee's program to sample and analyze iodine against a back-ground of noble gases can be considered to meet the guidance spec-ified in NUREG-0737, section III.D. However, the following matters requiring licensee attention were identified: The licensee should obtain additional SAM-2s. Currently all SAM-2s (4) have been assigned to specific locations. The licensee does not have any spare units. Spare units should be obtained in the event an assigned SAM-2 becomes defective or needs calibration. (50-293/85-27-40) Certain aspects of the procedures for calibration and use of the SAM-2 are written in such a manner that literal reading of the procedures would cause errors in quantification of airborne iodine: (50-293/85-27-41)
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the method for determination and use of iodine measurement efficiency is not clear / consistent between SAM-2 calibra-tion /use procedures,
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  ' CPM' is not defined (i.e. net or gross) Some of the procedures for use of the SAM-2 do not include the correction factors for determination of total iodine dos The procedures only include correction factors for dose due to I-13 (50-293/85-27-42) No verification of acceptability of air flow calibration devices used to calibrate iodine air samplers has been performed (50-293/85-27-43) Procedures for use of the SAM-2s for analysis of iodine activity collected on sample media allow the analyses to be made in up to a radiation field of 4 mR/h The licensee should perform and document an evaluation that demonstrates that a SAM-2 can detect 1x10 7uci/ml in a 4 mR/hr fiel (50-293/
85-27-44)
9.0 Quality Assurance (QA) and Design Review 9.1 General Design and Installation Records The following design records were reviewed:
The design and installation records reviewed in this area are presented in Attachment 7 (Section A) to this repor ___
 
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The inspector found that there were sufficient sample points to obtain samples after the accident and that the installation documents were readily retrievable. No unacceptable conditions were identifie .2 Alternate Power Supplies to' PASS The Pilgrim SER on PASS, Section 2.0, criterion 1 indicated that during loss of off-site power, alternate power sources were avail-able for both gas and liquid sampling systems that could be energized in sufficient time to meet the three hour sampling and analysis time limi The inspector reviewed pertinent documents concerning the power supplies to the PASS and its associated electrical equipment, (e.g.,
sample cooling water pumps, valves that require operation to draw samples) to ascertain whether all power supplies can be switched to the diesel generator after a loss of off-site powe Items examine included:
  *
  *
BECO Drawing S-E-155 " Station Electrical Single Line Composite Diagram" Rev. 2 dated November 18, 198 *
.. ,   ATTACHMENT
Bechtel Drawing No. E8 " Single Line Meter & Relay Diagram 4160V Breaker A409, 480V load Center 88 480 Volt". Revision E *
BECO Drawing M239 " Post Accident Sampling & H and 0 Analyzer 2 2 System" Revision El dated September 13, 198 The inspector verified that the power supplies to the PASS could be switched to the diesel generator after a loss of off-site power. A self contained cooling unit was used to cool the sample. The unit would be inoperable after loss of off-site power. Should this happen after an accident, water from Fire Main System can be supplied to the sample cooler by manually open two hand valves (one for supply, one for discharge to sump). The water from Fire Main System does have sufficient head to perform this functio No unacceptable conditions were identifie .3 Environmental Qualification of PASS Valves The Pilgrim SER on PASS, section 2.0, Criterion 3 indicated that the PASS valves which were not accessible after an accident were environ-mentally qualified for the conditions in which they need to operat _
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Deviation The licensee's May 18, 1982 letter (BECo Letter #82-145) to the NRC provided a description of his High Range Containment Monitoring System.  \
This system is used to monitor Drywell and Torus Radiation levels during a Post-Accident situation. The radiation levels are used to estimate core damage. The attachment to the licensee's letter (Page 2, Design and Qualification) states in part that, "The coaxial cable used to carry the signal between the detectors and the Post-Accident Monitoring Panel is located in a harsh environment under accident conditions. ...The cable is installed in conduit, ..."
Contrary to the above, a visual inspection of the detector cables by the NRC on September 19, 1985 found that an estimated 20-30 feet of cable for the redundant Drywell detectors, at the location where they entered the Drywell penetrations, were not installed in conduit. The cable was found laying in an unprotected manner on various valves and pipes at that location.


The PASS solenoid valves required to be operable for post accident sample collection are SV-5065-63 through SV-5065-86. (24 valves total)
Pursuant to the provisions of 10 CFR 2.201, Boston Edison Company is hereby required to submit to this office within thirty days of the date of the letter which transmitted this Notice, a written statement or explanation in reply, including: (1) the corrective steps which have been taken and the results achieved; (2) corrective steps which will be taken to avoid further violations; and (3) the date when full compliance will be achieved. Where good cause is shown, consideration will be given to extending this response time.
The inspector randomly selected the EQ files of 4 valves (SV-5065-63 and 64, SV-5065-69 and 70) for review. The documents reviewed in-cluded the Equipment Qualification Evaluation Sheet (EQES) for each of the 4 valves, all dated June 11, 1985, and Wyle's Qualificattun Report 47066-50V-8, Revision B dated June 21, 198 Within the scope of this review, the following matters requiring licensee attention were identified: The model of the installed solenoid valves differed from the test mode The licensee tried to qualify the installed model by similarity. However, evaluation of similarities and differences between the installed model and the test model (e.g. organic materials used and their thermal aging effect, coil temperature rating, physical size and construction etc.) were not performe (50-293/85-27-45) The valve manufacturer recommended the 0-rings be replaced once every five years. This was not addressed in the EQ file, Jus-tification should be provided if this recommendation is not to be implemente (50-293/85-27-46) Information Notice 84-68 identified field cable degradation when connected to high power solenoid valves. This cable degra-dation was caused by substantial temperature increase in the solenoid housing. The effect of this should be addressed in the EQ fil (50-293/85-27-47)
9.4 Physical Observation of System Installation The inspector physically observed the installed PASS and discussed its operation with the licensee's I&C personnel. The licensee stated that the instruments on the PASS panel were calibrated during the preoperational test early this year. Although there were no evidences that the calibrations of the PASS instruments were overdue, there were no calibration schedules set up for these in-strument Within the scope of this review, the following matter requiring licensee attention was identified:
*
Review and evaluate the need to incorporate the periodic cal-ibration of PASS instrumentation into the routine maintenance /
calibration program. Critical instruments should be included in such a progra (50-293/85-27-48)


_
Response BECo Letter #82-145 stated "the cable is installed in conduit" This statement was consistent with plant design change PDC 79-61 providing for cable installation in conduit and flex conduit into the penetration sleeve. The PDC also required that slack cable be coiled up and placed inside the penetration sleeve. We admit that the position of the cables, as observed by the inspector on September 19, 1985, deviated from the statement in that sections of loose cable were coiled up and lef t on the floor in an unprotected manner. Most probable cause was that during a subsequent calibration of the detectors, the cables were not placed back in the originally intended configuration.
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9.5 Containment'High Range Radiation Monitor (CHRRM)
9. EQ of CdRRM Detectors
  .The inspector reviewed the EQ files of the CHRRM detectors to ascertain whether the files contained sufficient evidence that these detectors were qualified for the environmental conditions in which they need to operate after an acciden There were four CHRRM detectors in Pilgrim, two for Drywell (RE1001-606A&B), and two for Suppression Pool (RE1001-607A&B).
 
;  These detectors were manufactured by General Electric Compan RE1001-606A&B were mounted in capped pipe sections extending into the drywell atmosphere from drywell outside wal RE1001-607A&B were located in the secondary containment adjacent to the suppression poo The following documents were reviewed:
  * BECO Equipment Qualification Evaluation Sheets for RE1001-606A&B, RE1001-607A& * Advanced Systems Engineering Memo. Qualification of Gamma Sensitive Ionization Chambers for Cams Post-Accident En-vironmental Conditions, Report No. 943-81-003, April 24, 198 * Environmental Qualification Test Report of Raychem WCSF-N
  ' Nuclear In-Line Cable Splice Assemblies for Raychem Cor-poration Menlo Park, California. Report No. 58442-1, May 15, 198 The inspector physically observed the installation conditions of these four CHRRM detector The following matters requiring licensee attention were identified: The radiation exposure qualification data was not contained in the EQ files. This data should be made available for NRC review (50-293/85-27-49). The coaxial cables for the radiation detectors at the Drywell were found laying on the floor and were subject to being stepped on. The licensee's May 18, 1982 letter in-dicated the cables were-in conduit. This appears to be a
,'
Deviation from the information provided to the NR (50-293/85-27-50)
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9. Power Supply to CHRRM Detectors The inspector reviewed pertinent documents to ascertain that safety related redundant emergency power supplies were used in these four detector The documents reviewed are identified in Attachment 7 (Section B) of this repor The inspector verified that these four radiation detectors were powered by safety related redundant power supplies channels A & Within scope of this review no unacceptable conditions were f identifie j
      )
9.6 Environmental Qualification of Containment Hydrogen Monitors (Analyzers)
The inspector reviewed the EQ files of Containment Hydrogen Monitors (C172 and C173) to ascertain whether the file contained sufficient evidences that these hydrogen analyzers were qualified for the en-vironmental conditions in which they were required to operate after an acciden The following documents were reviewed:
*
BECO Equipment Qualification Evaluation sheets for Hydrogen Analyzer C172 & C173, both dated June 28, 198 *
Test Report 1035-1 " Prototype Qualification Test for Hydrogen Analyzer System K-III & K-IV" Revision 1 dated September 1981 (COMSIP Inc.).
Each of the containment hydrogen monitors was a rather complex system. It consisted of numerous solenoid valves and automatic sequencing circuits, sample pump, pressure regulator, flow meters, etc. and a control cabine Within the scope of this review, the following matter requiring licensee's attention was identified:
*
Page 18 of the Test Report 1035-1 described the yearly, 5 year and 10 year maintenance requirements for the H 0 analyze The yearly maintenance requirement states " carefully inspect for degradation, replace as necessary". This description appears to provide less than acceptable guidance relative to performing an inspection of this safety related system. The qualification l l
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maintenance for this-system was not available for review. The licensee. should clearly identify the inspection acceptance criteri (50-293/85-27-51)
10. Exit Interview The Post-Accident Sampling and Analysis Team met with licensee rep-resentatives at the conclusion of the inspection on September 20, 1985. The Team Leader summarized the purpose, scope, and findings of the inspectio At no time during the inspection was written material provided to the license _
 
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ATTACHMENT 1 TO INSPECTION REPORT 50-293/85-27 PERSONS CONTACTED Licensee Personnel
*C. J. Mathes, Station Manager
*K. P. Roberts, Outage Coordinator
*J. F. Crowder, Senior Compliance Engineer
*A. Shatas, Acting Chief Chemical Engineer
*B. Eldridge, Assistant Chief Radiological Engineer, Operations-
*E. T. Graham, Compliance Management, Group Leader J. Smallwood, Senior Chemical Engineer
*L. Dooley, Training Supervisor, Technical
* Hoey, Senior ALARA Engineer
*T. L. Sowdon, Radiological Section Head
*E. Rochelle, Energy Support Services
*T. Kelley, Bartlett Nuclear
* Eisenmann, CYGNA
* Andrew, I&C Engineering
* Velez, Project Manager, EQ
* .
De Lemos, Project Manager, TMI Modifications
* Sanford, Manager, Training
* J. Moraites R. Fairbank, Deputy Manager of Engineering,-Nuclear J. Pawlak, Power System Group Leader L. Perfetti, Power System Engineer R. Sherry, Assistant System Chief Maintenance Engineer J Burbank, I&C Technician
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M. Akhtar, Senior Modification Engineer NRC
* McBride, Senior Resident Inspector 0ther members of the licensee's staff were contacted and/or participated in an exercise of post-accident and effluent monitoring systems during the inspectio * Denotes attendance at the exit interview on September 20, 198 _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _  _ _ _ _ _ - _
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ATTACHMENT 2 TO INSPECTION REPORT 50-293/85-27 Documentation for NUREG-0737, II. Pilgrim Nuclear Power Station Emergency Procedures      .
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2.2.113 "H2/02 Analyser and C-19 Systems," dated August 21, 198 .7.4. " PASS Small Volume Liquid Sample From Jet Pump Flow Sensing Line," dated September 4, 198 .7.4. " PASS Undfluted Liquid Sample and Dissolved Gas Grab Sample From Jet Pump Flow Sensing Lines," dated September 4, 198 .7.4. " PASS Small Volume Liquid Sample From Residual Heat Removal System," dated September 8, 198 .7.4. " PASS Undiluted Liquid Sample and Dissolved Gas Grab Sample From Residual Heat Removal System," dated September 4, 198 .7.4. " PASS Dissolved Gas Measurement From Jet Pump Flow Sensing Lines," dated September 4, 198 .7.4. " PASS Dissolved Gas Measurement From RHR," dated September 4, 198 .7.4. " PASS Iodine / Particulate Sample From Drywell," dated September 4, 198 .7.4.1.10 " PASS Iodine / Particulate Sample From Torus," dated l  September 4, 1985.
 
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5.7.4.1.11 " PASS Iodine / Particulate Sample From Reactor Building,"
dated September 4, 1985.
 
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5.7.4.1.12 " PASS 14 ml Gas Sample From Drywell," dated September 4 198 .7.4.1.13 " PASS 14 ml Gas Sample From Torus," dated September 4, l  198 . " Estimating Core Damage," dated August 21, 1985.
 
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6.5-305 " PASS Source Calibration of PASS Radiation Monitoring Instruments," dated March 27, 198 .1 " Analysis of Liquid Samples for Baron (By Spectrophoto-metry) Under Accident Conditions," dated June 12, 198 .1 " Analysis of Liquid Samples for Boror (By Plasma Spectro-metry) Under Accident Conditions," dated June 12, 198 .1 " Analysis of Liquid Samples for Chloride Under Accident Conditions," dated July 17, 1985.
 
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7.1 Radioisotopic Analysis of Liquid Samples Under Accident Conditions," dated June 12, 198 .1 " Radioisotopic Analysis of Gas Samples Under Accident Conditions," dated June 17, 198 .1 " Radioisotopic Analysis of Iodine Cartridges Under Accident Conditions," dated June 21, 1985.
 
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7.1 " Radioisotopic Analysis of Particulate FtIters Under Accident Conditions," dated June 12, 198 .1 " Analysis of Dissolved Gas Sample (By Gas Chromatograph)
Under Accident Conditions," dated September 11, 198 '
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TP 84-226 " Pre-operational Test for the Post Accident Sampling System," dated October 26, 198 Other Pilgrim Nuclear Power Station Procedur_es
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" Interim Safety Evaluation by the Office of Nuclear Reactor Regulation Relative to Technical Specifications Requested by Generic Letter 83-36,"
dated July 5, 1985.
 
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" Safety Evaluation by The Office of Nuclear Reactor Regulation Post-
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Accident Sampling System (NUREG-0737, Item II.B.3) " dated July 1,198 ,.
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ATTACNeeEMT 3 TO 198SPECTIOtt REPORT 50-293/85-27 COMPARISOM OF AMALYTICAL RESULTS Chemical Analysis Bo ron The test data were:
Analysis Licensee NUREG-0737 Actual Commitment Standard Results Coenitment Pecuireme m %Erro r Meet 100 pps 90 ppa +/- 10% +/-50 ppa  -10%  Yes 500 ppm 460 ppe +/- 1G% +/-5%  - 3%  Yes 1000 ppa 1000 ppa +/- 10% +/-5%  0%  Yes Chloride 1 ppa 1.13 ppe +/- 10%  10%  +13%  No 2 ppa 1.99 ppa +/- 10%  10%  - 1%  Yes 5 pps 4.97 ppa +/- 10%  10%  -
1%  Yes 10 paa 9.34 ppa +/- 10%  10%  - 7%  Yes Isotopic Anasysis 8toraa I  Licensee NUREC-0737  Comeitaent Isotope Sink PASS Cormitment Requirements %Erro r Meet 5-131 1.29D-4 1.53E-4 -50/+200-% -50/+200% +19%  Yes 8-132 4.62E-3 4.75E-3 -50/+200% -50/+200% + 3%  Yes 8-133 1.67E-3 2.30E-3 -50/+200% -50/+200% +30%  Yes I-134 1.39E-2 1.65E-2 -50/+200% -50/+200% +19%  Yes 8-135 5.01E-3 4.67E-3 -50/+200% -50/+200% - 7%  Yes
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ATTACHMENT 4 TO INSPECTION REPORT 50-293/85-27 Documentation for NUREG-0737. II.F.1- Pilgrim Nuclear Power Station Emergency Procedures
-
2.2.55 " Reactor Building Exhaust Radiation Monitoring System,"
Rev. 6, dated May 1, 198 .7.2.18 "Offsite Dose Projections and Protective Action Guides for the General Public," Rev. 4, dated June 8, 198 .7.2.22 "Use of the Emergency Dose Assessment System (EDAS), Re dated
-
5.7.2.23 "Use of Off-site Dose Rate Nomograms", Rev. 3, dated June 8, 198 .7.2.25 "Use of HP-85A Off-site Dose Calculators", Rev. 3, dated August 7, 198 Pilgrim Nuclear Power Station Drawings
-
P. + I. D. M-282, " Plant Ventilation Diagram," Rev. E-4, dated March 25, 198 Pilgrim Nuclear Power Station Construction Package
-
PDCR 79-62, " Noble Gas Effluent Monitoring System".


Licensee Correspondence
Therefore, as corrective action to correct the condition, the sections of flexible conduit that were in place were properly clamped to the edge of the two respective drywell penetration pipes. Also, all loose coaxial cable was coiled up, brought up from the floor and positioned inside the penetration pipes so as to protect it from undue damage. This corrective action was completed on December 7, 1985 and returned the cable to its original design configuration. Since placing the cable in its proper position, we have also determined that there was no. cable degradation based on our physical inspection and the fact that there have been no indicative downscale alarm indications on the associated panel.
-
W. H. Deacon, Actg. Mgr., Nuc. Opns BECO to D. G. Eisenhut, Dir. DOL, dated January 25, 198 W. H. Deacon, Actg. Mgr., Nr Opns BECO to D. G. Eisenhut, Dir. 00L, dated April 16, 198 D. B. Vassallo, Chief, ORD #2 to A. V. Morisi, Mgr. Nuc1 Opns BECO, dated May 10, 198 A. V. Mortsi, Mgr. Nuc. Opns BECO to 0. B. Vassallo, Chief, ORB #3,' 00L, dated May 18, 198 , . _-
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A. V. Morisi, Mgr. Nuc. Opns, to D. G. Eisenhut, Dir. DOL, dated June 9, 198 D. B. Vassalo, Chief, ORB #2 to A. V. Morisi, Mgr. Nuc. Opns BECO Boston, dated August 9, 198 '
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A. V. Morist, Mgr. Nucl. Opns BECO, to D. B. Vassallo, Chief, ORB #2 00L, dated September 14, 198 W. D. Harrington, Sr. VP, Nuclear BEC0, to D. B. Vassallo, Chief ORB #2 00L, dated August 9, 198 l
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D. B. Vassallo, Chief, Operating ORB #2 DOL to W. D. Harrington, Sr. VP Nuclear, BECO, dated December 17, 198 W. D. Harrington, Sr. VP Nuclear, BECO to D. B. Vassallo, Chief, ORB #2      '
DOL, dated February 27, 198 W. D. Harrington, Sr. VP Nuclear, BECO to D. B. Vassallo, Chief, ORB #2 DOL, dated May 18, 1985
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D.-B. Vassallo, Chief, ORB #2 00L to W. D. Harrington, Sr. VP Nuclear, BECO, dated July 5, 1985.


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W. D. Harrington, Sr. VP Nuclear, BECO to D. B. Vassallo, Chief ORB #2      o
i Corr ctiva cction 'to prc.clud) rccurrcnco is, that. the Health Physics Group will modify the detector calibration procedure to include a precautionary step which will require the technician to visually ensure, upon finishing the I calibration, that the coaxial cable is left in a secured, protected manner inside the penetration pipe. Incidentally, the accepted installation method then (as it is now) is to leave several feet of extra coaxial cable coiled up inside the pipe penetration in the vicinity of the detector. This is to allow for ease in removing and calibrating the detectors when required.
. DOL, dated May 30, 198 ,
NRC Memoranda
,
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W. V. Johnston Asst. Dir. Materials, CEB to G. C. Lainas, Asst. Dir. for
: Oper. Reactors, DOL, dated December 14, 1984, i
) Pilgrim Nuclear Power Station Drawings J
-
. Drawing No. M239, Sh. 1-3, " Post Accident Sampling and H2 and 02 Analyzer System," Rev. E3, dated September 13, 1985.
 
,
Pilgrim Nuclear Power Station References 11.t.1-1 and II.F.1-2 Procedures
-
5.7. Sampling, Transport and Analysis of Effluent Iodines and Particulates from the Main Stack Under Emergency Condi-
,   tions, PIL 52-F1, dated December 28, 1982.
 
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  - _ . . _ _ . _ _ . . . _ . _ _ . . . _ . . - _ , - - _ . - _ _ _ . . _ _ . _ . _ _ . _ _ _ _ _ . _ _ . . _ . _ _ _ _ _ _ _ . _ _ _ _ _
 
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We are confident that the above mentioned corrective actions will provide adequate physical protection of the subject cable. Full compliance was achieved on December 7,1985 the date upon which the cables were returned to a satisfactory configuration.


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We would also like to clarify our environmental qualification position on the coaxial cable. Since the issuance of BECo Letter #82-145 it has been determined that the radiation monitors and the coaxial CXG cable are not required to function for a PBOC (Pipe Break Outside Containment).
5.7. " Sampling, Transport and Analysis of Effluent Iodines and Particulates from the Reactor Building Vent Under Emergency Conditions", Rev. 3, PIL 52-01, dated December 28, 198 ~5.7. " Sampling, Transport and Analysis of Effluent Iodines and Particulates.from the Turbine Building Under Emergency
  . Conditions", PIL 52-115, Rev. 2, dated December 28, 198 .4. " Source Calibration of General Monitor High Range Noble Gas Monitors", Rev. 1, dated November 24, 198 Pilgrim Nuclear Power Station Memoranda
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R. A. Smith BECO to A.R. Trudeau, BECO, "Representativenats of
;  Samples from the Main Stack and the Reactor Building.Ver.t", PNPS i  File HP-83-151, dated March 31, 1981.
 
!
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R. A. Smith BECO to A. R. Trudeau, BECO, "NRC Inspection 83-12/
Follow-up Item 83-02-02, Representativeness of Samples from the Main Stack and the Reactor Building Vents" PNPS File HP-83-283, dated June 30, 198 ENTECH Engineering Co. Report '
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P10368 "A nomogram for Correlating the Effluent Sample-Line Filter Radioactivity to the Amount of I-131 Released to the Atmosphere", dated May 24, 198 P100-R2 " Didos-III, A Three-Dimensional Point - Kernel Shielding Code for Cylindrical Sources", dated December 198 P103EC1 " Pilgrim Station Unit 1 -Main Stack and Reactor building Vent Monitors: Correlation Between Monitor Readings and Off-site Whole Body Gamma Doses", Volumes I-III, dated June 198 P103-EC3 " Pilgrim Station Unit 1, Emergency Plan Engineering Calcula-tions to Correlate Radiation Monitor Readings with Plant Re-leases and Off-site Dose Rates", Vol. I-III, dated June 1980.


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Environmental qualification has been established for the LOCA condition.
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P100-R3 "A Nomogram for the Interpretation of I-131 Field Sample Measurements Without the Need of Numberical Calculations", dated l  January 1981.


!
Since the cable is not required to function under a PBOC, conduit protection outside the drywell penetration sleeve is not essential for environmental qualification. Revision (7) of the PNPS E.Q. Master List will further reflect this determination.
Licensee r +
  - ondence l
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W. ' :iarritt, Mgr. Nuc. Engr. , BECO to D. G. Eisenhut, Dir. DOL, dated February 27, 1981.


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A. V. Morist, Mgr. Nuc. Opns. BECO to T. A. Ippolito, Chief, OR8 #2, DOL, dated October 16, 198 T. A. Ippolito, Chief ORB #2, 00L to A. V. Morisi, Mgr. Nuc. Opns. BECO, dated December 8, 198 D. B. Vassallo, Chief ORB #2, DOL to A. V. Morisi, Mgr. Nuc. Opns. BECO, dated March 1, 198 W. H. Deacon, Accts. Manager, Opns, BECO to D. G. Eisenhut, Dir. 00L, dated April 16,-198 ' D. B. Vassallo, Chief, ORB #2, 00L to A. V. Morisi, Mgr. Nuc. Opns.,
BEC0, dated May 10, 198 A. V. Morisi, Mgr. Nuc. Opns, BECO to D. B. Vassallo, Cheif ORB #2 00L, dated June 4, 198 _ A.1V. Morisi, Mgr. Nuc. Opns, BECO to D. G. Eisenhut, Dir. 00L, dated-June 9, 198 D. B. Vassallo, Chief, ORB #2 00L, to A. V. Morisi, Mgr. Nuc. Opns. BECO, dated August 25, 198 P. H. Leach, Proj. Mgr., ORB #2, to W. D. Harrington, Sr. VP Nuclear BECO, dated November 7, 198 D. Harrington, Sr. VP Nuclear BECO to D. B._Vassallo, Chief ORB #2 DOL, dated May 28, 198 NRC Memoranda
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W. E. Kregar, Asst. Dir. Rad. Prot. DSI, to G. C. Lainas, Asst. Di Safety Assessment D0L, dated October 23, 198 R. W. Houston, Asst. Dir. Rad. Prot. DSI to T. Novak, Asst. Dir. OR. DOL, dated January 29, 198 R. W. Houston, Asst. Dir. Rad. Prot. DSI to T. Novak, Asst. Dir. OR. 00L, dated April 26, 198 T. Eipen, R-1, Inspection Report 50-293/82-13 dated May 15, 198 .
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M. Shambecky R-1, Inspection Report 50-293/83-0 G. C. Lainas, Asst. Dir. OR, 00L to D. M. Crutchfield, Asst. Dir. Safety Assessment, DOL, dated October 17, 198 .
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ATTACHMENT 5 i
TO INSPECTION REPORT 50-293/85-27  l Documentation for NUREG-0737  ,
ITEM, II.F.1.-3, High Range  '
Containment Monitor G. E.- Product System and Sensors Engineering MEMO No. 127-82-004, dated i December 20, 1982.
 
,
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Letter, W. H. Deacon (BECO) to D.13. Eisenhut (NRC), dated March 25, 1982 (
(BECO ltr. 82-91).
 
.
I Letter, H. R. Balfour (BECO) to D. B. Vassallo (NRC), dated May 18, 1982
; (BECO Ltr. 82-145). t Letter, D. B. Vassallo (NRC) to A. V. Morisi (BECO), dated May 13, 198 [
;
Letter, W. D.- Harrington (BECO) to D. 8. Vassallo (NRC), dated June 1,198 NRC Inspection Report 50-293/83-02, dated February 23, 198 !
Procedure No. 5.7.5, Revision 0, estimating Core Damage, dated March 6, 198 ,
      !
; Procedure No. 5.7.2.18, Revision 4, Offsite Dose Projections and Protection Action Guides for the General Public, dated June 8, 198 ! Procedure No. 6.5.-296, Revision 1, Source Calibration of the Containment High Radiation Monitoring System, dated February 10, 1984.
 
s Varicus Emergency Plan Qualification Card !
  * *
1 daedJu$y$b,i9 *
r l Procedure No. 2.3.2.4, Revision 5, Panel 904 Left Control Room, dated i December 14, 1984.
 
t
      '
: Procedure No. 5.7.2.22, Revision 3, Use of the Emergency Dose Assessment System (EDAS),datedAugust7,198 ;
Procedure No._2.2.124 Revision 4, Containment High Rad Monitor System, dated '
May 22, 1985, i
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D'
o ATTACHMENT 6 TO INSPECTION REPORT 50-293/85-27 Documentation for NUREG-0737 Item III.D.3.3. Inplant Iodine Letter, D. 8. Vassallo (NRC) to A. V. Morist (BECO), dated April 8,198 Letter, A. V. Morisi (BECO) to D. B. Vassallo (NRC), dated May 5,1982 (BECO)
Ltr82-118).
 
Procedure No. 5.7.2.19, Revision 0, In-Plant I-131 Air Sampling and Analysis, dated April 1, 198 Procedure No. SI-RP.5700, Revision 0, Calibration of Eberline RAS-1 Regulation Air Samples, dated August 22, 198 Procedure No. SI-RP.4701, Revision 0, Operation of the RADS Co Model H809V Air Sampler, dated August 22, 198 Procedure No. PNPS SI-RP 4700, Revision 0, Operation of the RAS-1 Am Sampler, dated August 22, 198 Procedure No. SI-RP.5701, Revision 0, Calibration of Radus Model H809V Air Sampler, dated August 22, 198 Procedure No. 6.5-287, Revision 9 Calibration of Eberline SAM-2, dated July 12, 198 .
 
_ _ _ _ _ _ _ _ _ _
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ATTACHMENT 7 PASS QA AND DESIGN / INSTALLATION RECORDS General Design and Installation
  *
Boston Edison P&ID No M239 " Post Accident Sampling and H &0 2 2 Analyzer System". Sheet 1 Revision E4 dated June 1985, Sheet 2 Re-vision E3 dated September 13, 1985, Sheet 3 Revision E4 dated September 13, 198 *
GE Document No. C5474-SP-1 "BWR Generic PASS Design Requirements" Revision 1 dated November 24, 198 * GE Document No. C5475-SP-5 "LOCA Sampler Installation" Revision 2 dated March 3, 198 * Boston Edison Field Revision Notice FRN 80-31-120 dated April 21, 1983 " PASS Modification Sample Supply and Return Piping Installa-tions & Tie-in".
 
* Boston Edison FRN 80-31 Attachment A "Special Procedures for Welding to Valcor Engineering Model V5265-5295 Solenoid Valves - to avoid overheating of valve, rubber seat welding."
 
*
,  Boston Edison FRN 80-31-142 " Work Installation for the Removal and Replacement of Valve Nos. SV-5065-72, SV-5077 and 5078, Liquid Sample Return Line Containment Isolation Valves". 4 pages B. Power Supplies to High Range Containment Monitors
  * BECO Schematic Diagram E550 Sheet 1 " Containment High Radiation PAM System - Channel A" Revision El dated May 7, 198 * BECO Schematic Diagram E550 Sheet 2 " Containment High Radiation PAM System - Channel B" Revision El dated May 7, 198 * BECO Drawing SE 155 " Station Electrical Single Line Composite Diagram, 4.16KV & 480V AC System" Sheet 1, Revision E10 dat.d Feb-ruary 18, 1985 and Sheet 2 Revision Ell dated December 25, 198 * BECO Wiring Diagram M227A3 " Post Accident Monitoring C170" Sheet 1, Revision E * Beco Wiring Diagram M227A5 " Post Accident Monitoring C171" Sheet 1, Revision E2.
}}
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Latest revision as of 12:02, 7 December 2021

Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-293/85-27
ML20203F328
Person / Time
Site: Pilgrim
Issue date: 04/18/1986
From: Martin T
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To: Harrington W
BOSTON EDISON CO.
References
NUDOCS 8604250123
Download: ML20203F328 (2)
v • d • e


Text

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c 1- APR -181986 p

k h ' Docket No. 50-293.

Boston Edison Company M/C Nuclear ATTN: Mr. William D. Harrington Senior Vice President, Nuclear 800 Boylston Street-Boston, Massachusetts .02199

' Gentlemen:

'

Subject: Inspection:No. 50-293/85-27 .

iThis refers to your letter dated December. 13, 1985, in response to our letter

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dated November 13, 1985.

Thank you for informing us of the corrective and-preventive actions documented in your letter. These actions will be examined during a future inspection of your. licensed program.

Your cooperation with us is appreciated.

Sincerely, h (

%cw Thomas T. irtin, Director Division of Radiation Safety and Safeguards cc:

. A. V.~ Morisi, Manager, Nuclear Management Services Department C. J.- Mathis, Station Manager Joanne Shotwell, Assistant Attorney General Paul Levy, Chairman, Department of Public Utilities W. J.. Nolan, Chainnan,. Plymouth Board of Selectmen Plymouth Civil Defense birector ..

Senator Edward P. Kirby Public Document Room (PDR)'

-Local'Public Document. Room (LPDR)

. Nuclear Safety Information Center (NSIC)

NRC Resident Inspector Connonwealth of Massachusetts (2)

8604250123 860418 = #"

PDR- ADOCK 05000293 G PDR-y 0FFICIAL RECORD COPY o

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Boston Edison Company M/C Nuclear 2

,*

,

bcc:

Region I Docket Room (with concurrences)-

Management Assistant, DRMA (w/o encl)

, Section Chief, DRP.

W. Raymond, SRI, Vermont Yankee

'T. Shedlosky, SRI, Millstone 1&2 lH. Eichenholz, SRI, Yankee P. Leech,. LPM, NRR

/

RI:DRSS RI:D SS Gb.V RI: RSS Nimitz/mmb Pasciak Bellamy

't 4 12/3c/85 42/8/85 R/8 /85 f!LA 0FFICIAL RECORD COPY 50-293/85-222 - 0002.0.0 12/23/85

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CrevoM EDCCM COMPANY eDO GovLStoM STRes?

SOSTON, M ASSACNuSETTS O2199

,

11MLLIAM O. MARRINGTON

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===*aaa December 13, 1985 SECo Ltr. #85-222 Thomas T. Martin, Director Division of Radiation Safety and Safegaurds U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406 License No. DPR-35 Docket No. 50-293 Subject: Response to Deviation as contained in NRC Inspection Report 85-27

Dear Mr. Martin:

This letter is in response to the subject deviation contained in NRC Inspection Report 85-27, conducted by Mr. R. Nimitz of your office on September 16-20, 1985 at Pilgrim Nuclear Power Station.

The subject deviation and Boston Edison's response is enclosed as an attachment to this letter.

If you should have any further questions regarding this matter, please do not hesitate to contact me.

Very truly yours, WS M W. D. Harrington Attachment l

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.. , ATTACHMENT

.

Deviation The licensee's May 18, 1982 letter (BECo Letter #82-145) to the NRC provided a description of his High Range Containment Monitoring System. \

This system is used to monitor Drywell and Torus Radiation levels during a Post-Accident situation. The radiation levels are used to estimate core damage. The attachment to the licensee's letter (Page 2, Design and Qualification) states in part that, "The coaxial cable used to carry the signal between the detectors and the Post-Accident Monitoring Panel is located in a harsh environment under accident conditions. ...The cable is installed in conduit, ..."

Contrary to the above, a visual inspection of the detector cables by the NRC on September 19, 1985 found that an estimated 20-30 feet of cable for the redundant Drywell detectors, at the location where they entered the Drywell penetrations, were not installed in conduit. The cable was found laying in an unprotected manner on various valves and pipes at that location.

Pursuant to the provisions of 10 CFR 2.201, Boston Edison Company is hereby required to submit to this office within thirty days of the date of the letter which transmitted this Notice, a written statement or explanation in reply, including: (1) the corrective steps which have been taken and the results achieved; (2) corrective steps which will be taken to avoid further violations; and (3) the date when full compliance will be achieved. Where good cause is shown, consideration will be given to extending this response time.

Response BECo Letter #82-145 stated "the cable is installed in conduit" This statement was consistent with plant design change PDC 79-61 providing for cable installation in conduit and flex conduit into the penetration sleeve. The PDC also required that slack cable be coiled up and placed inside the penetration sleeve. We admit that the position of the cables, as observed by the inspector on September 19, 1985, deviated from the statement in that sections of loose cable were coiled up and lef t on the floor in an unprotected manner. Most probable cause was that during a subsequent calibration of the detectors, the cables were not placed back in the originally intended configuration.

Therefore, as corrective action to correct the condition, the sections of flexible conduit that were in place were properly clamped to the edge of the two respective drywell penetration pipes. Also, all loose coaxial cable was coiled up, brought up from the floor and positioned inside the penetration pipes so as to protect it from undue damage. This corrective action was completed on December 7, 1985 and returned the cable to its original design configuration. Since placing the cable in its proper position, we have also determined that there was no. cable degradation based on our physical inspection and the fact that there have been no indicative downscale alarm indications on the associated panel.

Page 1 of 2

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i Corr ctiva cction 'to prc.clud) rccurrcnco is, that. the Health Physics Group will modify the detector calibration procedure to include a precautionary step which will require the technician to visually ensure, upon finishing the I calibration, that the coaxial cable is left in a secured, protected manner inside the penetration pipe. Incidentally, the accepted installation method then (as it is now) is to leave several feet of extra coaxial cable coiled up inside the pipe penetration in the vicinity of the detector. This is to allow for ease in removing and calibrating the detectors when required.

We are confident that the above mentioned corrective actions will provide adequate physical protection of the subject cable. Full compliance was achieved on December 7,1985 the date upon which the cables were returned to a satisfactory configuration.

We would also like to clarify our environmental qualification position on the coaxial cable. Since the issuance of BECo Letter #82-145 it has been determined that the radiation monitors and the coaxial CXG cable are not required to function for a PBOC (Pipe Break Outside Containment).

Environmental qualification has been established for the LOCA condition.

Since the cable is not required to function under a PBOC, conduit protection outside the drywell penetration sleeve is not essential for environmental qualification. Revision (7) of the PNPS E.Q. Master List will further reflect this determination.

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