IR 05000528/1985022

Forwards Insp Repts 50-528/85-22,50-529/85-23 & 50-530/85-17 on 850624-0712,23-24 & 25-31 Telcons.Violations Will Be Addressed in Separate Correspondence

ML20134E357
Person / Time
Site:	Palo Verde
Issue date:	08/02/1985
From:	Scarano R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:	Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
Shared Package
ML20134E360	List: IR 05000528/1985022
References
RTR-NUREG-0737, RTR-NUREG-737 NUDOCS 8508200171
Download: ML20134E357 (2)

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Arizona Nuclear Power Project

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Attention:

Mr. E. E. Van Brunt,:Jr..

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Executive Vice President

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Gentlemen:

Subject: ~NRC Inspection of1Palo.-Verde Units 1, 2 & 3

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- This~ refers to the inspection conducted'by Mr. C. I. Sherman of this office on June 24 through July-12, 1985, and by Messrs. G. P. Yuhas, A. D. Johnson ~and

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J..L.-Crews'on July 23-24, 1985 and telephone conversations July 25-31, 1985

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of" activities. authorized by NRC License No. NPF-41 and Construction Permit

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Nos. CPPR-142 and CPPR-143 and to the discussion of our findings held by

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Mr. Crews 4with-Mr. Haynes and other members of your staff at the conclusion of

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the inspection.

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Areas? examined during this inspection are described'in the enclosed inspection

report. Within these areas, the inspection consisted of selective

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. exam nations of procedures and representative records, interviews with i

personnel, and observations by the inspectors.

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NRC enforcement _ action based on potential violations identified within the

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scope of this, inspection will be the subject of separate correspondence.

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'In accordance with 10 CFR 2.790(a), a copy.of this letter and'the enclosure

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Should you h' ave any questions concerning this inspection', we will be glad to-discuss them with'you.

Sincerely,

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Ross A.- Sc arano,- Director

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Enclosure:

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REGION V==

Report Nos. 50-528/85-22, 50-529/85-23, 50-530/85-17 Docket Nos. 50-528, 50-529, 50-530 License Nos. NPF-41, CPPR-142, CPPR-143 Licensee:

Arizona Nuclear Power Project P. O., Box 52034 Phoenix, Arizona 85072-2034 Facility Name:

Palo Verde Nuclear Generating Station - Unit 1, 2 & 3 Inspection at:

Palo Verde Site - Wintersburg, Arizona Inspection Conducted:

June 24 - July 12, 1985 and July 23-24, 1985 and telephone conversations July 25-31, 1985 Inspectors:

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.h, adiaLion Specialist Date Signed

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i Vs, Senior Reactor gineer D6td Signed Y

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/ Facilities Radiological P otection Section Approved By:

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/6 F. A. Wedslawski, Chief Dat.e' Signed Emergency Preparedness and Radiological Protection Branch Summary:

Inspection during the period of June 24 - July 12 and July 23-24,1985 and telephone conversations July 25-31, 1985 (Report No. 50-528/85-22, 50-529/85-23, 50-530/85-17)

Areas Inspected: Routine unannounced inspection of Unit I startup testing program for radiation protection and chemistry; Unit 2 preoperational test program for radiation protection - staffing, and training. This inspection involved 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> onsite by four regional based inspectors. During this inspection the following I&E inspection modules were performed:

83521; 84521; 84522; 84523; 84524; 83524; 83525; 2515/65.

Results: Of the 12 areas inspected, one apparent violation in one area.

Technical Specification, Section 6.8 - failure to implement the post accident sampling program is described in paragraph 2 of this report.

8508200176 G50802 PDR ADOCK 05000528 O

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DETAILS 1.

Persons ~ Contacted

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Arizona'Public Service (APS)

1. • J. Haynes, Vice President, Nuclear-

1. • D. Karner, Assistant' Vice President

1. • J. Bynum, Plant Manager

• W. Ide, Director, Corporate QA/QC

1. • W. Quinn, Manager, Licensing

• J. Hesser, Nuclear Engineer

• F. Hicks, Training Manager

• K. McCandless, Licensing C. Emmett, Compliance

1. L. Souza, Assistant Corporate QA/QC Manager W. Doyle, Radiation Protection Supervisor, Unit 2

• G. Perkins, Manager, Radiological Services

• B. Cederquist, Manager, Chemical Services

1. L. Brown, Manager, Radiation Protection and Chemistry T. Bloom, Licensing Engineer

• C. Russo, Manager, Quality Audits and Monitoring

• K. Oberdorf, Radiation Protection Supervisor, Unit 1

• T. Warren, Unit 1 Chemistry Supervisor R. Johnson, Unit 2 Chemistry Supervisor J. Scott, Unit 3 Chemistry Supervisor D. Fuller, Unit 1 Lead Chemistry Technician R. Cunningham, Chemistry Technician L. Drinovsky, Chemistry Technician C. Gray, Chemistry Technician J. Mann, Corporate Health Physics / Chemistry Supervisor G. Nelson, Senior Radiation Protection Technician G. Hampton, Senior Radiation Protection Technician K. Kunter,-Senior Radiation Protection Technician W. Hood, Radition Protection Technician V. Nativio, Radiation Protection Technician R. Logan, Instrument Calibration Technician R. ' Jenkins, Chemistry Technician G. Burhanan, Radiation Protection Technician K. Akers, Lead Radiation Protection Technician

• R. NeVille, Reactor Engineer

• T. Green, Supervisor, Training Support

• W. Montefour, Quality Assurance Engineer

• S. Pennick, Quality Monitoring Supervisor

• D. Hoppes, Reactor Engineer M. Toole, Radiation Protection Technician M. Wagner, Senior Radiation Protection Technician

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Gavin, Radiation Protection Technician G. Sanks, Radiation Protection Technician

1. 0. Zeringue, Technical Support Manager

1. S. Shepherd, Deputy Group Supervisor, Bechtel M. Moon, Radiological Engineer

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• Denotes-those. individuals participating in the preliminary exit interview on July.12, 1985.

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1. Denotes those individuals participating in the exit interview on July 24, 1985.

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2.

Post Accident Sampling System Inspection in this area have been described in NRC Inspection Report Numbers 50-528/83-12, -28, -30, 50-528/84-34, -49, -64 and 50-528/85-09.

This inspection was a post implementation examination of the system the licensee installed to meet licensee condition 2.C.17 of NPF-34 which stated:

"(17)

Post Accident Sampling System (Section 22.2, SSER 7)

Prior to exceeding five percent of full power, APS shall install and have operable a Post Accident Sampling System which meets the provisions of NUREG-0737 (II.B.3)."

The PVNGS Lessons Learned Implementation Report (LLIR)Section II.B.3 describes the system the licensee designed and installed to meet the criteria of NUREG-0737.

Item II.B.3 " Post-Accident Sampling Capability" of the LLIR describes an automated in-line system with chemistry and

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radiochemistry modules located at the auxiliary building 70' level with piping to a remote grab sample device located in the chemistry laboratory at the 140' level. Drawing 13-N-SSP-002 shows the general sampling layout.

Sample sources described include safety injection, reactor coolant' hot leg, sumps and the containment atmosphere.

FSAR Section 12.1.2.4, " General-Design Considerations to Keep Post Accident Exposures ALARA" states in part,

"The facility layout will assist in keeping occupational exposures

- ALARA even after a design basis accident. While exposures will be significantly higher than during normal operation, required access is provided to vital areas and systems without exceeding 5 rem /hr....

To provide sampling capability with exposures kept ALARA, PVNGS will incorporate a remote, automated, post-accident sampling system that meets the requirements of NUREG 0737 and Regulatory Guide 1.97, Revision 2.

This system can be operated from the radiochemistry counting room, the control room, or the technical support center.

Backup grab sample capability will be provided in the hot lab sample room using microchemistry techniques to keep the volume of source as small as possible."

NRC examination of the licensee submittal was described in the PVNGS Safety Evaluation Report (SER).

Pages 22-13 and 22-14 of the Palo Verde SER, section II.B.3 states in part,

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" Planned modifications are being made...to permit personnel to obtain and analyze samples within three hours after an accident, without incurring radiation exposure to an individual in excess of GDC 19, assuming source terms given in Regulatory Guide 1.4....

Gaseous hydrogen will be sampled by the containment hydrogen control system.....

Based on the above evaluation, the staff has concluded that the provision in the post-accident sampling system, when completed, will meet partially the reqcirements of Item II.B.3 in NUREG-0737."

Section II.B.3 of the PVNGS SER identified five areas to be resolved before the review could be completed. These items involved: backup sample capability for any in-line analysis; a description of analytical capabilities in an aEcident chemistry and radiation environment; a core damage assessment procedure; testing and training; and ventilation exhaust from the sample station. These potential license conditions were resolved in SSER 7 which states,

"On the basis of the above evaluations, the staff now concludes the applicant's post-accident sampling system meets all of the eleven criteria of Item II.B.3 in NUREG-0737.

Prior to utilizing the in-line system in accordance with Item II.B.3, the applicant shall demonstrate the capability of the backup grab sample system."

L By l'etter' dated September27, 1984, ANPP-30593-EEVBJ/WFQ/MAJ, E. Van Brunt-(APS) to G. Knighton.(NRC), the licensee described how the five potential license conditions would be met.

This letter described PASS Unit 1 status and identified the need to perform "certain modifications to improve sampling capability". The licensee further stated that until a 60 day operability test had been' completed, " backup sampling meeting the requirements of NUREG-0737 will be available with sample analyses

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described in the attached Table 1".

Table 1 of ANPP-30593 list capabilities for the PASS Backup Grab Sample under high radiation field conditions. For each analysis, a range, sensitivity, accuracy, procedure and calibration technique are identified.

Clarification (2)(b) of NUREG-0737 identifies " dissolved gases (e.g.,

H )" as one of the analysis capabilities for the reactor coolant sample.

9CIarification (4) states that either total dissolved gas or hydrogen could be measured and the unpressurized reactor coolant samples could be used if the licensee could quantify dissolved gases by sampling in this manner.

The description of total gas analytical capability in ANPP-30593 states only to refer to note 2.

Note 2 on page 5 states, "2.

Grab samples may not be applicable because of 1.' excessive dilution, 2. sample size because of high radiation field effects".

.A summary of recent NRC inspections in the PASS area is as follows:

Inspection Report 50-528/84-34 conducted August 20-24 and August 27-28, 1984, S2, pp. 7-8 described inspection findings with respect to the PAS.

O That report identified the fact that required backup capability to

. perform total gas and dissolved oxygen did not exist.

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Inspection Report 50-528/84-49 $2, pp. 4-5 conducted November 5-9 and November 13-14, 1984, described additional review of PASS status and identified the need to perform radiological reviews to define the shielding envelope for backup sampling analytical procedures.

Inspection Report 84-64 conducted December 13-18, 1984 stated that the PASS area would remain an open item until, "an anticipated license condition requirement is completed", and that additional inspection effort in this area would be deferred until the system is required to be operational.

'During-the November inspection, the inspector reviewed with the licensee's representatives, the description of total gas analytical capability in AKPP-30593 and how it could potentially mislead a reviewer into believing that backup gas capability existed when in fact, it did not.

Based on this review, the inspector reminded licensee representatives of the need for accurate reporting because NRC relies on the veracity of licensee submittals.

The inspector made the observation to licensee representatives of the need for the licensee to clarify the submittal.

The licensee submitted ANPP-31333-EVB/WFQ/MAJ dated December 5,1984, E.

Van Brunt (APS) to G. Knighton (NRC),to. clarify the September 27 letter, to provide additional information and to' request a schedule exemption from a prior commitment to have PASS operable prior to OL.

The letter stated that site testing disclosed that further modifications would be required to provide backup sampling capability and that the in-line PASS modifications and testing were expected to be complete by October, 1987 following the restart after the first refueling outage.

The letter ~further stated that, " Prior.to having the in-line system operable, the primary method for performing Post Accident Sampling activities will be remote grab sampling." Both ANPP 31333 and 30593 were submitted under oath.

Attachment A to ANPP-31333 describes the history of PASS testing, problems and an action plan.

Section IV. ' Subsequent Actions" states in part,

"As NUREG-0737 requires backup sampling and analysis for installed in-line analysis equipment, we are currently reviewing, revising and generating the chemistry procedures necessary for obtaining, transporting and analyzing the samples that will be available.

These procedures incorporate provisions for handling undiluted samples while maintaining dose commitments within the requirements of GDC-19, and are expected to be in place (approved), personnel trained and qualified by April 15, 1985, or prior to exceeding five percent power."

Section V. ' Action Plan' states in part,

"The planned action is to proceed with an integrated method to modify PASS to provide manual capability of sampling and analysis to

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By April 15, 1985, or prior to exceeding 5% power PASS is expected to be available to demonstrate the capability to promptly obtain reactor coolant and containment atmosphere samples in an accident chemistry and radiation environment. The PASS will utilize grab sample methodology of depressurized reactor coolant for the parameters except dissolved gases, and will provide dissolved gases capability with a modified remote grab sampler."

Table B of ANPP-31333 lists gas sample capabilities as hydrogen, oxygen and total gas.

These capabilities are listed as " Complete, System Operable".

The NRC issued operating license NPF-34 on December 31, 1984, for Palo Verde Unit 1.

This license contained the condition 2.C.(17) which was previously referenced.

By letter dated May 22, 1985, ANPP-32695-EEVB/JK0, Van Brunt (APS) to Knighton (NRC), the licensee stated in part, "Therefore, as previously committed, a Post Accident Sampling System as described in Reference (5), will be complete before exceeding 5 percent power." Reference (5) is ANPP-31333.

This license condition was resolved in SSER 7 which states:

" Prior to utilizing the in-line system in accordance with Item II.B.3, the applicant shall demonstrate the capability of the backup grab sample system.

The applicant has advised the staff that because of the need for design modifications, the in-line PASS will not be available until after first refueling.

In the interim the backup grab-sampling system will be used to meet the requirements of NUREG-0737 Item II.B.3.

This capability will be demonstrated prior to exceeding 5% power operation and the license will be conditioned accordingly."

Palo Verde SSER 8 issued with NPF-41, the full power operating license states in part, "Several of the conditions are no longer necessary since the applicant has provided acceptable letter commitments. The commitments along with the dates of the applicant's letters are listed below....(4) Prior to exceeding 5% of full power, the post accident sampling system will be operable (ANPP letter 32695, dated May 22, 1985)."

By letter dated June 13, 1985, ANPP-32820-EEVB/MAJ Van Brunt (APS) to Knightor. (NRC), the licensee stated in part, "We wish to inform you that we have completed modifications, testing and have made operable a PASS which meets the provisions of NUREG-0737 Item II.B.3.

The primary method for performing Post Accident Sampling is the remote grab sampling capability described in Reference (2) (Reference (2) is ANPP-31333)."

FSAR Section 12.1.2.4 describes a remote system with backup grab sample capability in the hot lab. ANPP-31333 did not describe changes to the basic system as described in the FSAR or LLI *

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As the PASS had not been inspected since prior to the OL, the purpose of this inspection was, in part, to verify operability of the PASS and to verify that the PASS program required by technical specification (TS) 6.8.4 was adequately implemented.

The inspection began with a demonstration of the RCS sampling capability, a review of the manual' remote grab sample capability and the procedures and training described in the December 5,1984 letter. The inspection also included review of analytical data to support the capabilities described in submittals and a radiological analysis of the manual sample handling techniques. These items were specifically requested by the inspector during a prior inspection as necessary for evaluation of the modified PASS system. The PASS system in place at the time of the inspection consisted of a new valve module, piping to the 140 foot level, a control panel and a liquid grab sample point under the normal chemistry sample sink.

The dose assessment was based on sample volumes in the 0.05 m1 to 0.3 ml range. The analytical verification and shielding analysis calculations were based on these sample sizes. The shielding calculations for liquid samples showed that clarification 6 of NUREG-0737 (GDC-19) could be easily met.

In review of the shielding calculations, the inspector noted that energies were combined into a single E-bar for each time interval and beta energies were added to the gamma values. The shape of the energy fluence conversion to exposure curve shows that this technique will produce non-conservative results. An independent review performed by the corporate health physics staff also identified the potential for-non-conservative calculations in this regard. The inspector also noted that what appeared to be lead glass and metal syringe shields were available for use in the chemistry lab.

Bremsstrahlung production from the beta flux was not considered in the initial calculation. The inspector requested that the licensee verify that the calculations were acceptable given these considerations. NUREG-0737 (II.B.2) specifies that a Regulatory Guide 1.4, ' Assumption... Radiological Consequences

...LOCA for PWR's'

source term shall be assumed. Regulatory Guide 1.4 states the acceptable assumptions that may be used, these are summarized briefly as follows:

Iodine 25% of core immediately available for leakage from containment

Noble Gas 100% of core immediately available for leakage from containment

Decay should be taken into account The individuals who performed the radiation dose evaluation for PASS analysis stated in their analysis that the source term assumptions were taken from the CE Series 80 Radiation Design Guide. For the containment atmosphere analysis, only 5% of the halogens were considered in the atmosphere.

For the reactor coolant source term 50% of the halogens and 0.5% of the noble gases were considered in the coolant. These assumptions are not consistent with RG 1.4 and were not specifcally justified by the licensee as appropriate alternatives. The review by

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• corporate HP did not identify this concern. The overall calculation package presented to the inspector was not a controlled or approved document. The licensee took steps to address these technical concerns during the inspection.

With"one exception, the analytical capabilities were found acceptable.

The licensee stated in a submittal dated September 27, 1984 that a range of 2 ppm - 6000 ppm could be attained for the boron analysis. The licensee stated during the inspection that this range could not be achieved using the current procedure and the 0.1 m1 sample volume used in the shielding analysis.

The licensee representative agreed to correct this matter. The inspector observed two chemistry technicians take a RCS liquid sample. The inspector requested that a radiation protection technician be assigned to the job to describe what he would do if a sample were necessary under accident conditions. The HP technician stated that he received no special training for coverage of this job but the individual was knowledgeable of proper health physics practices for such a job. A sample was obtained I hours after beginning the

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procedure.

Due to staffing requirements, the demonstration was continued to the next day.

One of the chemistry technicians involved in the second demonstration stated he was trained 4 weeks ago on the PASS system. This individual expressed unfamiliarity with the procedure because it had been changed 5 times in the past 4 weeks. The individual stated he had not read these changes.

Procedure 74CH-92Z98 Rev. 2 section 7.3 sets forth the chemistry section policy to issue a required reading notice for each procedure change or revision.

Individuals in the chemistry section indicated that this practice was stopped as a result of numerous procedure changes. On July 29, the Chemical Services Manager confirmed that this requirement was reinstituted on July 9, 1985.

The inspector noted that the chemistry technicians were not totally familiar with the valve and flow path arrangements, however they appeared capable of establishing the proper flow paths.

Based on discussion with chemistry personnel, the inspector identified the following points for improvement of procedure 740P-SS02, ' Operation of the Backup PASS':

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Warning that improper flow path selection could possibly produce full RCS pressure in undesirable places;

Instructions to restart the valve alignment process if critical valves are mispositioned;

Instruction on how to restart the valve alignment process if the PASS appears to be working improperly.

Samples taken as a result of this process were then analyzed using inicrochemical techniques.

PASS analytical values were compared to recent RCS samples taken by standard methods. Agreement for boron was excellent once the proper analytical technique was selected. The licensee agreed to make minor procedural revisions to insure this would not be a problem agai.

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Agreement for hydrogen ion concentration (ph) was excellent once a faulty micrb ph probe was replaced. lAgreement for chloride was acceptable.

Poor results were obtained for dissolved total gases although dissolved hydrogen could be 4etermined accu'rately by a slightly different

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methodology The licensee was aware of problems with the dissolved gas method and.had initiated corrective steps.

During'a system walkdown, the inspector examined the PASS valve module and' signal conditioning box IJSSNE016 located at the 70' level of the auxiliary building. This box was observed with lead shielding partially stacked around it.

The licensee stated that the lead served to protect unqualified electronic equipment in a radiation environment. The lead was unstacked because of ongoing work on the equipment, when identified by the inspector, the licensee took steps to immediately replace the lead shielding. This item was under the licensee's work control system and would have been replaced when certain repairs were complete. The licensee did not know what integrated dose would cause failure but stated that the shielding could be quickly installed if necessary. The inspector noted it was unlikely that the licensee would be aware to do this if the system were needed. A licensee representative stated that if these electronics failed, a sample could not be obtained from the PASS.

Procedures and training related to the PASS program for liquid sampling were examined and found adequate.

A list of PASS related procedures and training are provided as Table 1.

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Notwithstanding the exceptions noted, the inspector considered the liquid sample portion of the PASS operable and an acceptable interim method until the in-line PASS could be completed.

The inspector also examined PASS surveillance, leakage monitoring and performance test packages for the backup PASS. These items appeared satisfactory. All unsatisfactory items regarding the liquid PASS sample point were discussed with the Chemical Services Manager who stated that they would be promptly corrected. The inspector noted that the PASS system installed does not fully meet the licensee's commitments in the LLIR. This is acceptable because the system in place is an interim system that will be replaced before restart from the first refueling outage.

During the PASS demonstration, the inspector learned that the containment atmosphere samples described in NUREG-0737 and licensee submittals were not available at the 140' elevation remote grab sample point described in the PVNGS LL1R and FSAR. The licensee representative stated that this sample would be taken from RU-1 sample lines. RU-1 is the containment atmosphere radiation monitor required by TS for RCS leak detection.

RU-1 is located as shown on 13-P-ZAL-205 in electrical penetration room A-127 at the 100' elevation. The dose evaluation previously described assumed that this sample would be available in the chemistry hot lab and therefore did not address time to access, sample.and leave the sample point. The inspector decided to depart from the planned inspection to examine this matter in greater detail because this sample point is not described in licensee submittals, the location is not considered by the inspector to be remote as described in the June 13, 1985 letter and the

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inspector was aware of the potential for high dose rates under accident conditions in this area.

Chemistry technicians initially questioned by the inspector stated that radiation protection personnel would bring this sample to the chemistry lab. The Unit 1 RP supervisor when asked about this matter stated that

'there was no procedure for taking a PASS sample from RU-1, there was no shield container (Pig) for transporting this sample and no training to take a PASS sample from RU-1.

This individual stated he was not aware of any dose evaluation having been performed for this sample location. This individual also expressed his belief that a sample could be obtained if necessary using present procedures and sample techniques normally used with RU-1 for routine sampling.

The inspector discussed the matter of sampling containment atmosphere with the Manager of Chemical Services and the Manager of Radiological Services. The inspector was informed that EPIP-27, " Post Accident Sampling and Analysis" and 75-RP-92Z64 "RMS Radioactive Sample Collection" were the relevant procedures and that Health Physics personnel were given training in use of the grab sample system used at RU-1 and were familiar'with techniques of sampling high activity effluent streams from their training on handling particulate and iodine filters from the high range effluent monitors.

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The inspect'or examined 75RP-9ZZ64 and EPIP-27.

Based on this examination and discussion with licensee 'and contractor personnel, the inspector noted the following:

Containment isolation valves for R'U-1 must be opened from the control room following a CIASl signal; The grab' sample of containment atmosphere may be taken from RU-1 directly or'from special grab sample taps upstream of RU-1;

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RU-1 valve V-1 isolates to protect.the monitor on a reactor containment building pressure of about 10 PSI. This isolation cannot be overridden by normal means; Procedures-do not describe the details of proper hookups for sampling when a PASS sample is to be taken; Procedures do not describe the proper valving sequence to be used, flush times to assure representative sampling, flow rates or sample times when a PASS sample is to be taken or state they are different

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than for normal samples; Procedures do not describe if the containment atmosphere sample is-brought to the chemistry lab by the RP technician or if the chemistry technician is to collect the sample for analysis at the sample point; Procedures do not describe the in plant route to be used to obtain this sample if necessary;

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Procedures do not describe specific handling techniques, special shields for samples, potential dose rate from samples or the number of individuals needed to get a sample; Procedure 75RP-92Z64 identifies the PASS sample only as a containment atmosphere hydrogen sample.

Procedure 74RP-9ZZ64 did identify the correct sample vessel to be used.

EPIP-27 was initially written to cover handling of effluent samples under accident conditions and was modified by revision 2 to include the containment as a backup to the inline PASS system. EPIP-27 included the

. caution note that dose rates in the vicinity of this sample point may exceed 100 R per hour. This was included before a decision was made to utilize that sample point to be the primary sample point to meet the PASS requirement.

EPIP-27 does provide the following information: General precautions for handling high activity samples; specific information for taking effluent release sample; personnel dosimetry and instrumentation needs; a statement that the emergency coordinator will direct sampling activities. Valve alignment and sample point location is important because the containment air sampling represents a potential path for leakage outside of containment.

The inspector discussed training and knowledge of the licensee's planned implementation of NUREG-0737 item II.B.3 with several health physics

technicians employed by ANPP in order to determine if their knowledge and

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training on the normal sample system was adequate for use under accident conditions. Of seven individuals interviewed at random, only one appeared fully knowledgeable of the sample system operation for meeting PASS requirements. The individuals expressed a general lack of knowledge of expected dose rates following a LOCA. Only one individual was aware of the proper sample vessel, only one individual stated that RU-l would isolate on high (10 psi) containment pressure. One individual did not know that the grab sample technique was the primary PASS method. The individuals interviewed all stated that training for this specific sample under accident conditions was non existent. The individuals appeared to have general knowledge on handling accident samples from training on EPIP-27 and effluent monitors.

Concerns raised were discussed with the Supervisor, Radiological Services and the Fianager, Radiological Protection / Chemistry.

The inspector requested a meeting with cognizant individuals, plant management, QA and licensing to discuss the licensee's position on dose evaluation for this activity, procedural adequacy, training adequacy and a statement of position on readiness to take a post accident sample.

On July 2, 1985, a meeting was held with Radiation Protection, Chemistry, QA, Licensing, the Plant tfanager and Corporate IIcalth Physics. The NRC Senior Resident Inspector was also present. The licensee stated that the PVNGS LLIR item II.B.2 provided the dose evaluation for the sample area and that exposure rates were greater than 100 R/ hour.

The dose evaluation for sample handling was performed for a small volume sample at the chemistry sample statio.

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The licensee acknowledged that an overall dose evaluation to demonstrate that a sample could be taken within GDC-19 criteria had not been performed prior to July 1, 1985. The licensee stated that using an exposure rate of 100 R/hr, a sample could be taken within GDC-19 using i

two individuals. The licensee committed to performing a complete dose evaluation. Regarding the licensee position on readiness, the licensee representative stated that a sample could be taken under the assumed LOCA

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conditions using two separate entries to maintain individual doses below l

GDC-19 criteria.

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Regarding adequacy of training, the Manager, Radiological Services stated

that the sampling operation required knowledge not available in the

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procedure but that personnel could be directed by himself or by the emergency plan designated alternates, the ALARA Supervisor or the Unit One Radiation Protection Supervisor. The inspector asked what additional knowledge was necessary but not in the procedure. The licensee representative identified several items previously listed in this report.

This licensee representative stated the information he thought needed to go into the EPIP and stated that the EPIP would be promptly corrected, The inspector later learned that the Manager of Radiological Services was i.

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not aware that the RU-1 sample location was in fact the primary sample point until after this inspection began. The general position of ANPP management was that the procedures were adequate as they stood, but agreed that improvement could be made. The licensee acknowledged that no specific training was given on taking the containment atmosphere samples under post accidents conditions, but stated that the normal training and practice on RU-l would be adequate if the system were needed. The licensee did commit to provide more specific training.

i The licensee further stated that containment atmosphere samples for PASS I

were not given the same attention as liquid samples and that hardware concerns caused oversight. The radiation protection manager stated his opinion that the program was not deficient to the point where it would prevent performance. The licensee stated that no walkthrough test under assumed accident conditions had been performed but that a PASS drill was planned for July 24, 1985. At this point the inspector asked the licensee if they were willing to demonstrate their capability. The response was that they were willing but would have to call in personnel due to startup support requirements.

The next morning the inspector offered the licensee the opportunity to demonstrate their ability to perform. The licensee was willing to perform this test. The inspector requested that Unit One Radiation Protection Supervisor direct the test rather than a member of management and that the inspector would select the technicians and provide the scenario. The Supervisor, Radiological Services stipulated that only ANPP technicians, not contractor personnel be used. The inspector also requested that two EP engineers accompany the demonstration. One to observe for the licensee and one to act as a time keeper.

At 0930 on July 3, 1984 the inspector stated that the decision was made to take a containment atmosphere sample to be analyzed for hydrogen and radioisotopes in particulate, iodine and gaseous forms. The inspector stated that radiation 1cvels were per the LLIR and high airborne activity

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existed. The entry team performed in a acceptable manner in preparing for entry in the areas of protective clothing, dosimetry, instrumentation and respiratory protection.

At 1140 hours0.0132 days <br />0.317 hours <br />0.00188 weeks <br />4.3377e-4 months <br />, the team of 4 radiation protection technicians and one chemistry technician entered the radiation control area wearing SCBA.

The chemistry technician was added as a result of the inspector identifying the fact that the gas sample holder would not fit into the lead shield available. The licensee then planned to have the chemistry technician take a small sample via gas syringe directly from the sample vessel while still attached to the sample apparatus.

Due to several problems the first sample did not reach the lab until about 1320 hours0.0153 days <br />0.367 hours <br />0.00218 weeks <br />5.0226e-4 months <br />.

Several uncontrolled delays associated with the logistics of an unplanned emergency exercise were partly responsible for the length of time needed to take this sample.

Rather than describe in detail all the problems identified, the following significant facts are presented:

One technician was not aware that the RU-1 monitor sample taps were not appropriate due to the tendency of RU-1 to isolate at 10 psi or more containment pressure. This individual was aware that on low j

flow, RU-1 would have a local alarm and that the RU-1 sample taps I

might not be the proper place to sample; I

A fitting provided to make the connection at the proper grab sample point had the wrong pitch thread.

The licensee stated that this was

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due to a non-standard evolution; The two valves that must be opened manually took 30 seconds each to open; Based on the assumption that the greater than 100 R/hr dose rates predicted in the LLIR would be no greater than 100 R/hr and timing obtained during the inspection, it appeared to the inspector that a minimum of three radiation protection technicians would be required and that each would receive between 3 and 5 rem whole body dose. One chemistry technician would likely receive 1 to 3 rem.

If problems were encountered, the numbers and doses would easily increase.

Based on observations during this test and previous findings the inspector concluded that training was not adequate for this task, nor did procedures provide enough guidance to ensure that the task could be properly accomplished.

Further the inspector noted that the licensee had not performed a dose evaluation for the tasks, nor was a drill under accident conditions conducted to demonstrate the feasibility of performing this evolution within the CDC-19 dose criteria.

The inspector concluded that based on projected dose rates and the need to possibly overexpose 4 or more people based upon 10 CFR 20, licensee representatives should reconsider their position that the system is actually operable. The inr,pector noted that a flow path does exist and

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that provisions to take a sample do exist however these appear inadequate

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for meeting the II.B.3 criteria and TS requirements.

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f In order to confirm radiation exposure rates at the monitor location, the t

inspector examined Bechtel calculation 13-NC-2A-304 Rev. 2 which was the

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shielding review for the 100' level of the auxiliary building.

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'of this calculation showed that the dose rate in room 127 from sources at I

the 88' elevation from safety injection line SIB-072-FCBA-20 would be l

1077 R per hour under the LOCA conditions assumed for the II.B.2 analysis.

Because such high dose rates would rule out any possibility of

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using the RU-1 sample points, the inspector immediately reported this information to the Chemical Services Manager on July 5, 1985 and requested that he provide the inspector an official position interpreting the Bechtel calculations.

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On July 11, 1985', the licensee representatives met with the inspector to discuss status of the containment atmosphere Post-Accident sample point.

The' licensee described steps taken since identification of the potential high dose rates in Room 127 on July 5, 1985. These steps included:

Request Bechtel to perform' additional shielding calculations; f-

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Perform additional in-house do'se assessments; Installation of. alternate sample points on the A&B trains of the post LOCA' hydro'en' analyzers located.outside of penetration rooms; g

Initiation of procehure modifications and additional training for radiation protection and chemistry technicians.

l The licensee took:th6 steps to relocate the sample point when preliminary indications were that dose rates from a spare penetration would preclude access to that penetration room. Dose rates from the 100% gas source term through a spare penetration at the manual grab sample isolation valve in room 127 were initially estimated at 500 R/hr.

Penetration sources were not considered in the LLIR (II.B.2) analysis as the penetration room was not considered a vital area to aid in the mitigation

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of or recovery from an accident. NUREG-0737 item II.B.2, " Clarification" identifies the sampling station as an area that must be included, the clarification further states that as a minimum, necessary modifications must be sufficient to provide vital system operation and for occupancy of the control room, TSC, sampling station, and sample analysis area.

Section (3) Dose Rate Criteria, '(b) Areas Requiring Infrequent Access:

GDC 19' states that for areas requiring access on an irregular bases such as containment gas sample stations, shielding should be provided to allow access at a frequency and duration estimated by the licensee. The licensee did not reexamine the II.B.2 analysis when the containment gas sample point location was relocated to the hydrogen monitor area.

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The licensee indicated that dose assessment for the revised sample location would be completed by July 15, 1985. On July 31, 1985, the licensee provided a revised position on dose rate at the PASS sample points (see p. 16). The licensee stated that the scope of the Units 2/3 PASS would address problems identified at Unit 1.

The licensee

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representative reaffirmed the ANPP commitment to complete the PASS as described in the LLIR at the meeting on July 11, 1985.

A summary of information regarding the PASS is presented below:

Based on testing and NRC inspection effort, the licensee determined the vendor supplied PASS would not be fully tested prior to the anticipated license issuance and described in their September 27 letter the backup sample capabilities that will be in place at the OL date. Additional problems caused delays which the licensee could j

not resolve before five percent power and a request for schedule exemption was made in the December 5, 1984 submittal;

License NPF-34 was issued December 31, 1984, with a license condition regarding PASS (2.C.17);

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The licensee installs, tests, and makes operable a " mini pass" which provides a grab sample for RCS liquid and gas in the chemistry lab.

The normal grab sample point on the containment atmosphere radiation

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monitor located in Room 127 on the 100' level of the auxiliary l

building is selected as the sample point for this portion of the PASS.

• Licensee submitted letter on May 22, 1985 stating PASS will be

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operable prior to exceeding full power.

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On June 1, 1985, NPF-41 was issued, permitting full power operation.

License condition 2.C.17 was removed.

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Licensee submitted letter June 13, 1985 stating that testing is l

complete, the PASS is operable, and meets NUREG-0737 Item II.B.3.

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A summary of inspection findings is as follows:

The system described in LLIR Item II.B.3 was not implemented in that backup sample capability at the hot lab (chemistry) sample station shown on a P&lD drawing and stated in the FSAR was not provided for

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the containment atmosphere sample.

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The licensee stated in a December 5, 1984 letter that the PASS requirement would be met using backup grab sampling techniques.

This letter and the September 27, 1984 letter did not identify different sample locations.

• At the time of inspection, the only evaluation performed on radiation doses to show that GDC-19 could be met for the containment atmosphere sample was performed assuming the sample could be obtained from the hot lab.

• At the time of inspection, no specific training had been performed for taking the containment sample under accident conditions;

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The ' mini-PASS' providing backup liquid sample capability was considered by the inspector acceptable in the interim until the full system could be installed.

At the time of the inspection, the inspector considered procedures inadequate for performing the containment atmosphere sampling evolution under accident conditions; The procedure (EPIP-27) correctly identified the potential high (greater than 100 R/hr) dose rates in the penetration room where the sample point was located.

• The licensee did not reconsider the penetration room a vital area pursuant to h'UREG-0737 Item II.B.2.

The licensee did not determine times to perform the sample evolution, did not determine the basis for the dose rates listed in LLIR Item II.B.2 and did not determine how much higher the dose rates could be under accident conditions.

The inspector concluded that the licensee did not establish a factual basis that GDC-19 could be met.

A demonstration conducted by the licensee indicated that problems existed with the procedures, equipment, and training. The licensee was able to establish a flowpath and obtain a sample using the equipment available.

• Radiation exposure rates in the penetration room in the vicinity of the monitor could reach 400 R/ hour 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after shutdown. This information was reported by the licensee on July 31, 1985 by telephone. This exposure rate would mean, based on observations by the inspector during the previously described demonstration, that sampling could require possibly in excess of 10 individuals to take the sample under the design basis conditions.

Based on the above preliminary dose calculations, the licensee believed that the source term from a capped penetration in Room 127 would preclude access for sampling. An alternate sample point was established approximately three days later.

Technical Specification requirements for PASS are contained in Section 6.

TS 6.8.1 states that written procedures shall be established, implemented and maintained for Post-Accident Sampling System Implementation.

Technical Specification 6.8.4.e states;

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Post-Accident Sampling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions.

The program shall include the following:

(1) Training of personnel, (2) Procedures for sampling and analysis equipment...."

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The licensee's failure to establish procedures containing information unique to sampling containment atmosphere under accident conditions including details of valve alignment, sampling connection points, sample purge time and other important information is considered a significant inadequacy.

The licensee's failure to perform adequate evaluations of potential radiation doses associated with the entry, sampling and retrieval of post-accident samples represented a failure to ensure the capability to obtain the cont,ainment atmosphere samples under accident conditions.

This is an apparent violation of Technical Specification 6.8.4.e (50-528/85-22-01).

In response to these observations, the Chief, Facilities Radiological Protection Section, the Enforcement Officer and a Senior Reactor Engineer of the Region V staff visited the site on July 23 and 24, 1985 to review the facts, determine the root cause of this failure, and evaluate compliance with GDC 19 for the RU-1 PASS sample location and the new PASS containment atmosphere sample ~ locations.

Based on tours of Unit 1, discussions with licensee and Bechtel representatives, review of documents and telephone converations the inspectors found:

An adequate evaluation to assure compliance with the 5, rem or equivalent dose criteria expressed in GDC-19 for obtaining a containment air sample at RU-1 was not performed in that the post accident radiation levels expected in Room 127.were not' considered with respect to the time technicians would have to be present in order to obtain the sample.

• Calculations performed by the licensee subsequent to the inspection indicated that post LOCA dose rates in room 127 could be as high as 400 R/hr three hours following shutdown.

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In a telephone call on July 31, 1985 the licensee reported

a preliminary calculation summarized below based on a TID-14844 source term release:

Present Core Equilibrium Core Source Irradiation History Irradiation R/hr R/hr Containment 140 280 RU-1 Monitor Piping

70 Piping Below 100'

30 Containment Leakage

20 200 400 The licensee also reported that Combustion Engineering had confirmed their calculation that the maximum cladding temperature following shutdown would not exceed 1000*F from latent and decay heat present in the core as calculated using the radionuclide inventory present when the reactor shutdown on July 24, 198.

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The Chemical Services Manager approved the decision to utilize RU-1 as the primary containment atmosphere sample point and failed to assure that all of the criteria of NUREG 0737, II.B.3 had been satisfied, specifically GDC-19.

All levels of management including the Chemical Services Manager, Radiation Protection Manager,. Plant Manager, Nuclear Engineering Manager, Licensing Manager and the Assistant Vice President were aware of the 0737 requirements, problems with the PASS at PVNGS and industry experience with other PASS systems.

No detailed independent technical review of compliance with the NUREG 0737 criteria, specifically GDC-19 for the containment air sample point was made prior to submission of the June 13, 1985 letter. This is considered a lack of adequate control of technical work and indication of inadequate management involvement to assure accurate presentations in correspondence to NRC.

• Although schedule pressure and preoccupation with development of the PASS liquid sample capability may have been a factor, no facts were revealed which would cause the inspectors to believe the apparent violation was deliberate, involved willfulness, careless disregard or an intentional false statement.

• In response to the initial inspection findings the licensee had relocated the containment ai'r sample point to the containment hydrogen monitor locations based on the LLIR radiation zone designations in the FSAR. On July 24, 1985 the reactor was shut down from 50% power due to a main condenser leak.

Because of questions raised by the NRC with respect to the Bechtel dose assessment of the RU-1 and the Hydrogen Monitor locations, the licensee asked the Bechtel representative to come to the site on July 23, 1985 and explain the preliminary results. On the morning of July 24, 1985 the licensee after discussions with Bechtel informed NRC that due to an apparent error in the radiation zone designation for the hydrogen monitor location, these sample points are not considered accessible following an accident. The licensee stated that the situation would be fully reviewed and Unit I would not be returned to power operation until the NUREG 0737 requirements were fully met in accordance with Technical Specifications. This commitment was confirmed by the licensee in a letter to Region V (E.E. Van Brunt to J. B. Martin) dated July 26, 1985. A Confirmatory Action Letter, acknowledging this commitment, was issued by Region V on July 29, 1985.

One apparent violation (50-528/85-22-01) of TS 6.8.4.c was identified 3.

Startup Testing Review The inspector verified that the Test Results Review Group (TRRG)

performed preliminary review of test results as specified by 72PA-12Z01 for the Biological Shield and Chemistry tests prior to procecding beyond 20 percent power.

No violations or deviations were identified.

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Bioassay Analysis 10 CFR 20.108 requires the use of bioassay as necessary to evaluate internal intake of radionuclides. Regulatory Guide 8.26, ' Application of Bioassay for Fission and Activation Products' identifies methods acceptable to the NRC staff. The inspector examined licensee procedure 75RP-9ZZ13 Bioassay Analysis.

The inspector noted that the procedure did not clearly address the need to determine the annual intake and compare it to the NRC limit. The procedure was generally adequate in other respects. The licensee's review and corrective steps will be examined in a future inspection (0 pen, 50-528/85-22-02).

No violations or deviations were identified.

5.

TLD Data The inspector examined a licensee report dated March 20, 1985 on preoperational radiological environmental monitoring for external radiation. The 45 stations monitored had annual average dose rates ranging from 9.3 to 13.6 microrem per hour.

A comparison of NRC data from 1982 to licensee data in 1983 showed reasonable agreement. The licensee is aware of the elevated background radiation dose rates in-Arizona as compared to other U.S. locations.

No violations or deviations were identfied.

6.

Survey Records

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Based on review of licensee procedures, discussions with personnel and observations of work practices during tours of Unit 1, the inspector

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brought to the licensee's attention an apparent potential to violate the record keeping requirements expressed in 10 CFR 20.401. Specifically, although no loose surface radioactive material was present, tools and equipment leaving the radiation controlled area including containment were being surveyed. There were no records to document the results of these surveys. The licensee representative stated that materials leaving the radiation control area are frisked or scanned for radioactive contamination in accordance with station procedure. The licensee acknowledges that records of these surveys are made only when items are identified as contaminated. The licensee also indicated that vehicles leaving the restricted area are surveyed and records are made. The inspector discussed the matter with the radiation protection manager who stated that he interpreted 10 CFR 20 differently than the inspector and that based upon his understanding frisks or scans were not surveys defined by 10 CFR 20 and records were not required. The RPM stated that when positive contamination was identified, a record of this would be made.

This matter was again discussed with the licensee, by the Chief, Facilities Radiological Protection Section, NRC Region V and the

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inspector by telephone on July 17, 1985. At that time the licensee was

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informed that surveys of tools and equipment having the potential of being contaminated with radioactive material from activities within the restricted area must be surveyed prior to release to the unrestricted area pursuant to 10 CFR 20.201(b) and therefore the records of such survey results are required by 10 CFR 20.401(b). The licensee agreed to review the matter. This matter will be reexamined in a subsequent inspection (0 pen, 50-528/85-22-03).

7.

Bioshield Survey NRC Inspection Report 50-528/85-09 $7.a described the inspector's review of the survey program prior to initial criticality.

At that time the inspecto'r discussed guidance of ANSI /ANS 6.3.1-1980, which was incorporated in the PVNGS Bioshield Survey Plan document provided by Bechtel.

The inspector noted that more detailed guidance on survey technique and identification of individual penetrations needed to be included in~the procedure.

Procedure 75PA-12Z01, " Biological Shield Survey" contains very" limited detail on special considerations for survey technique. During this inspection, the inspector verified action taken to improve the survey program ^

The licensee provided copies of sections

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of the Bioshield survey plan to technicians performing the survey and conducted briefings prior to each survey. The licensee did not change the written procedure.

The inspector did not witness any briefings but accompanied two of three survey teams performing the containment 50 percent power bioshield survey.

Based on this tour on July 9, 1985, the inspector made the following observations:

Some but not all penetrations were surveyed;

Scanning measurements of secondary shield walls were not performed consistent with ANSI /ANS guidance;

Readings were not always taken between survey points;

Neutron measurements were not always taken exactly on the survey marker;

Scanning measurements for gamma radiation were performed at a fast rate.

The inspector did not advise or attempt to correct these perceived inadequacies during the surveys.

The inspector then asked the lead radiation protection technician responsible for the survey how he expected the survey to be performed by demonstrating proper technique. This individual demonstrated scanning a shield wall at a fairly slow pace in a repetitive pattern. This individual stated that each penetration was to be surveyed, that scanning measurements were to be performed where access was reasonably cicar, that neutron dose rate measurements were to be made by placing the instrument exactly on the survey point marker, and that the meters should both be observed between survey point _

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The inspector then advised the lead technician and the unit supervisor, that either direction to technicians was unclear, inadequate or that technicians were not following the guidance given them. The unit radiation protection supervisor verified that the survey was to include scanning measurement of the shield wall to verify shield integrity and that each penetration was to be surveyed.

The following day the licensee stated that additional steps would be taken to ensure surveys were carried out properly including:

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resurvey of some points at the 50 percent plateau; the Unit I supervisor accompanying the team;

additional emphasis on survey technique at the pre-survey briefing.

The observations indicate the effect of failure to include adequate detailed guidance in a radiation protection procedure.

This matter will be examined again after the survey program is completed as part of the routine inspection program (open, 50-528/85-22-04).

No violations or deviations were identified.

8.

Pccting The licensee has established Plant Policy No. 13, Subject: Posting of Legal Notices in order to establish a method of complying with NRC requirements.

10 CFR 19.11 contains the requirement for posting parts 19, 20 and 21 of 10 CFR. The inspector examined this policy, the

" Official NRC Bulletin Board" established at several locations and reviewed 10 CFR 19.11(d) with the licensee.

In order to provide adequate assurance that all personnel observe the posting, additional notices were placed at several locations during the inspection period.

No violations or deviations were identified.

9.

Radiation Monitoring System The inspector examined a memo dated February 1, 1985, J. Ong to File, Subject: Detection Limits for the RMS. The licensee has established that radiation monitors have the detection limits specified in the FSAR and that HU-8 and RU-14 have the capability to detect MPC concentrations in any room of the auxiliary building or radwaste building respectively as stated in FSAR Section 11.5.2.1.3.

This matter is considered closed.

10.

Employee Protection t

10 CFR 50.7(a) identifies protected activities which apply to licensee's and permittee's.

10 CFR 50.7(e) states that NRC Form 3 shall be posted at locations sufficient to permit employees protected by 950.7 to observe

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a copy on the way to or from their place of work. The inspector did not observe posting of NRC-3 at the south entrance to the Units 2/3 construct:on site nor at the crafts entrance on the north side of the Units 2/3 site. The inspector and a licensee QA monitor looked for but could not find NRC-3 posted in the two story startup building located within the Units 2/3 area. The inspector also observed no postings at the Unit 3 fuel building north entrance or at the containment hatch.

The construction site is posted at. other locations including the Bechtel construction building and some tool cribs. The licensee stated that signs would be relocated to insure that 50.7(e) would be met for all employees.

The posting deficiencies were not considered to be violations as adequate signs were on the construction site and the licensee established additioaal locations during the inspection period.

11.

Unit-2 Readiness for Operation The inspector conducted reviews of equipment, procedures, personnel staffing, preoperational testing and held discussion with cognizant individuals in order to ascertain the state of readiness for operation of Unit 2.

Areas examined in this inspection are as follows:

RCS sample line routing for ALARA ESF IIVAC system generic and preoperational testing

llealth physics and chemistry qualification and staffing

Radiological effluent release control (release permit system)

Setpoints for effluent monitors for TS compliance

Radioactive waste processing system status Radiation monitoring system status Normal !!VAC system generic and preoperational testing

Selected radiation protection procedures on dosimetry and respiratory protection No violations or deviations were identified.

12.

Biological Shield Survey 20 Percent Results The inspector examined the official test copy of 75PA-lZZ01 and a memo documenting the 20% results.

The memo identified eight survey points above expected levels, all within containment.

These were in low occupancy areas and were not expected to present operating difficulties.

Of note was survey point 72 at the cast end of the refueling cavity. The measured dose rate at this point was 560 mrem per hour whereas only

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400 mrem per hour was expected. This value would lead to an expected dose rate of 2.8 rem / hour at full power. This value is not unusual for CE reactors.

No violations or deviations were identified.

13.

Respiratory Protection The inspector attended the licensee's basic respiratory protection training class, examined class materials and lesson plan NGR02-C-001-84.

Training for filter respirator use meets the NRC requirements. The inspector also examined procedure 75RP-9ZZ34, ' Respiratory Equipment Maintenance, Inspection and Repair'. This document is consistent with NRC requirements.

No violations or deviations were identified.

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Exit Interview Regarding PASS, at the preliminary exit interview the inspector characterized the inspection findings as a failure to ensure the capability to take samples of the containment atmosphere.

Based on the high dose rates projected for this area and the inadequate procedures and training, the inspector characterized the containment atmosphere sample portion of the PASS as inoperable.

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The plant manager stated that there was no question that the procedures were in need of upgrading but that the procedures were adequate because the licensee demonstrated the ability to take a sample.

Additionally, he confirmed that dose assessments were not performed prior to declaring the system operational. The plant manager also reaf firmed the licensee's commitment to make operable the PASS described in the LLIR.

Corrective measures for PASS taken by the licensee were not verified by the inspector during this inspection period.

At the final exit interview, NRC representatives were informed by the licensee that an adequate evaluation was not performed and the licensee does not now consider RU-l accessible based on preliminary calculations.

This was confirmed in a telephone conversation on July 31, 1985.

The NRC enforcement officer discussed the potential enforcement aspects of the inspection findings.

The NRC Senior Reactor Engineer discussed root causes of the apparent violation as a failure to perform quality technical work in this area.

The ANPP Vice President stated that the plant would not be restarted until the PASS sample problem was corrected and that a review would be conducted on the implications of the radiation zone designation error on other activities.

The NRC commented in closing that during all discussions, the licensee's personnel were cooperative and candid with the staf f.

The ANPP Vice President made a followup phone call on July 25, informing Region V that a written statement concerning action on restarting the unit would be (

submitted, confirming statements made at the exit interview on July 24,

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TABLE I PASS RELATED PROCEDURE INDEX PROCEDURE NUMBER PROCEDURE DESCRIPTION 74CH-9XC13 Analytical Instrument Calibration Verification 74CH-9XC15 Sampling Instructions 74CH-9XC33 Post Accident Radioactive Sample Analysis and Handling 74CH-9ZZ06 Boron (Autotitrator)

74CH-92Z34 Operation and Calibration of the Nuclear Data 682B 74CH-92Z35 pH (Potentiometric)

74CH-9ZZ46 Autotitrator Operation and Calibration 74CH-92247 Core Damage Assessment 74CH-9ZZ48 Operation and Calibration of the Stand-alone MCA 74CH-92263 Gamma Energy Analytical System Operation and Calibration 74CH-9ZZ65 Operation and Calibration of the Gas Chromatograph 74CH-92Z66 Determination of Primary to Secondary Leak Rate 74CH-9ZZ67 Gravimetric Equipment Operation and Calibration 74CH-9ZZ70 pH/ Specific Ion Electrode Operation and Calibration 74CH-9ZZ72 Operation and Calibration of the Ion Chromatograph 74CH-92Z90 Operation and Calibration of the Post Accident Sampling System 74CH-92Z98 Chemistry Section Training 74CH-9ZZ99 Records and Reports 74HF-1SS03 Post Accident Sampling System Performance Test 74HF-1SS05 Post Accident _ Sampling System Performance Test Unit I Interim Backup

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740P-ISS01 Nuclear Sampling Instructions 740P-ISS02 Operation of the Backup PASS 74ST-ISS01 Post Accident Sampling System Surveillance 74ST-ISS02 Post Accident Sampling System Leakage Monitoring 74ST-1SS03 Backup Post Accident Sampling System Surveillance EPIP-27 Post Accident Sampling and Analysis 75AC-9ZZO4 Shipment, Receipt and Storage of Radioactive Materials 75RP-92257 Packaging, Marking and Labeling of Radioactive Materials

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75RP-92259 Shipping of Radioactive Material

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75RP-9ZZ64 RMS Radioactive Sample Collection

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