ML20248L617

Notice of Violations from Insp on 980126-30.Violations Noted:Test Program for ECCS Operating in Piggyback Mode Did Not Demonstrate That Sys Would Perform Satisfactorily in Service

ML20248L617
Person / Time
Site:	Crystal River
Issue date:	03/16/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20248L615	List: IR 05000302/1998002 ML20248L612 ML20248L617
References
50-302-98-02, 50-302-98-2, NUDOCS 9803230283
Download: ML20248L617 (4)

Text

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NOTICE OF VIOLATION Florida Power Corporation Docket No. 50-302 Crystal River Unit 3 License No. DPR-72 During an NRC inspection conducted-on January 26 through 30. 1998. violations .;

of NRC requirements were identified. In accordance with the " General

-Statement of Policy and Procedures for NRC Enforcement Actions'" NUREG-1600, the violations are listed below:

A. 10 CFR 50, Appendix B. Criterion XVI, Corrective Action, requires that .

measures be established to assure conditions adverse to quality be j promptly identified and corrected. In the case of significant conditions adverse to quality the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition.

-Contrary to the above.

1. As of January 11, 1998, corrective actions for a significant condition adverse to quality regarding errors in procedures used for in-plant E0P actions were not adequate to. preclude repetitive errors in procedures used for in-plant E0P' actions such as Abnormal Procedure AP-770. Emergency Diesel Generator Actuation, and Abnormal Procedure AP-470. Loss of Instrument Air.

2. As of January 26, 1998. corrective actions for a condition adverse to quality regarding not performing radiological mission doses for personnel installing flow elements to accomplish reactor.

building purging for post-accident hydrogen concentration control were inadequate in that:

(a) The radiological mission dose evaluation did not account for the radiological dose from the loading of the Auxiliary Building ventilation filter banks. ,

(b) The radiological mission dose evaluation did not account for the radioactive loading from a 50 gpm Residual Heat Removal (RHR) pump seal leak on the loading of the- Auxiliary Buildings filters.

(c) The time validation inputting into the mission dose evaluation'used a non-conservative time for two people carrying a cart with approximately 50 pounds of equipment on it up the stairs to the Auxiliary Building location.

l This is a Severity Level IV Violation (Supplement I).  !

B. 10 CFR 50. Appendix B. Criterion XI. Test Control, requires in part that I a test program be established to assure that all testing required to demonstrate that structures, systems, and com)onents will perform  ;

satisfactorily in service which incorporate t1e requirements and  !

Enclosure 1 9903230283 980316 PDR ADOCK 05000302  :

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2 acceptance limits contained in applicable design documents. The test program shall. include operational tests.

Updated Final Safety Analysis-Report. Section 14.2.2.5.4. Emergency Core Cooling System (ECCS) Qualification, states in part. "In order to qualify the ECCS. the NRC placed requirements on the ECCS to ensure that '

.the health and well being of the public is not impacted. These requirements are specified in 10 CFR 50.46 and 10 CFR 50. Appendix K.

The criteria contained in Part 50.46 are applicable to all sizes of LOCAs and are necessary in order to verify adherence. These criteria are as follows ... A path to long-term cooling must be established."

This section further states that BAW-10104. Rev. 3. is the methods report on how the computer model used to ensure compliance with 10 CFR 50.46 is assembled and run. Section 14.2.2.5.4 further states "The LBLOCA application report for the 177 FA lowered loop plants is BAW-10103A.,

Topical Report BAW-10103A. Rev. 3. "ECCS Analysis of B&W 177-Fuel Assembly Lowered-Loop NSSS." and Topical Report BAW-10104. Rev. 3. "ECCS Analysis Of B&W's 177-FA Lowered-Loop NSS." Chapter 10. Long-Term Cooling. Section 10.2 states in aart that several alternate modes of operation of the ECC systems can )e used during long-term cooling, if necessary, while maintenance is being performed on normal equipment and one of these modes is: "One LPI pump operating with injection through-its associated injection line and with the crossover to the associated HPI string open: the associated HPI pump would be pumping through its HPI lines."

Contrary to the above, as of January 26, 1998, the test program for the emergency core cooling system operating in the piggyback mode did not demonstrate that the system would perform satisfactorily in service.

This is a Severity Level IV Violation (Supplement I).

C. 10 CFR 50. Appendix B. Criterion II. Quality Assurance Program, requires l that a quality assurance program be established. This program shall be documented by written policies, procedures or instructions and carried out in accordance with those documents.

l The Quality Assurance Program as described in the Updated Safety Analysis Report lists ANSI 45.2.11. 1974 " Quality Assurance Requirements for the Design of Nuclear Power Plants." under the committed standards.

ANSI 45.2.11. Section 3. Design Input Requirements. Subsection 3.2.  :

Requirements. states that the design input shall include but shall not i be limited to instrumentation and control requirements including type of instrument, range of measurement, and location of indications.  !

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3 ANSI 45.2.11..Section 4. Design Process, Subsection 4.1. General, states in part that design activities shall be prescribed and accomplished in accordance with procedures of a type sufficient to assure that ap]licable design inputs were correctly translated into procedures Su)section 4.5. Other Design Documents, states, in part, that procedures shall be established for the preparation and control of test procedures.

Contrary to the above, as of January 26, 1997, the quality assurance program as documented by written policies, procedures or instructions was not carried out in accordance with those documents in that:

1. The design input for Calculation I-90-0023. Reactor Building Hydrogen Concentration Loop Accuracy Calculation. Revision 1.

dated November 19, 1997, did not adequately include the instrumentation and control requirements of instrument uncertainty affecting the post-accident time duration before initiating reactor building purge for hydrogen concentration control.

2. The design input for Calculation I-90-0013. Post Accident RB Hydrogen Purge Instrument Accuracy. Revision 2. dated December 29.

1994, did not adequately include the instrumentation and control requirements of instrument location affecting the accuracy of the reactor building purge flow rate indication used for post-accident hydrogen concentration control.

3. The design activity of controlling test procedures was not adequately accomplished such that the stroke time acceptance criteria of the applicable surveillance procedures for valves DHV-42 and 43 was less conservative than that indicated in Calculation M-97-0120. Stoke Time for DHV-42/43 for Boron Precipitation. Rev.

1. dated November 1. 1997.

This is a Severity Level IV Violation (Supplement I).

D. 10 CFR 50. Appendix B. Criterion V. Instructions. Procedures and Drawings, requires, in part, that activities affecting quality be prescribed by documented procedures.

Contrary to the above, as of January 26, 1998 an activity affecting quality was not adequately prescribed by documented procedures in that step 1.6 of Emergency Plan Procedure. EM-225A. Post Accident Reactor Building Hydrogen Control, directing installation of the flow elements to be used for purging the reactor building to maintain post-accident hydrogen concentrations, did not include flanged connection torquing information from the vendor manual or direction to plug the flow instrument's power cord into the receptacle.

This is a Severity Level IV Violation (Supplement I).

1 Pursuant to the provisions of 10 CFR 2.201. Florida Pcwer Corporation is i Enclosure 1

4 l hereby required to' submit a written statement or explanation to the U.S.

Nuclear Regulatory Commission. ATTN: Document Control Desk. Washington. D.C.

20555 with a copy to the Regional Administrator. Region II. and a copy to the NRC Resident Inspector at the Crystal River Unit 3-facility, within 30' days of j the date of the letter transmitting this Notice of Violation (Notice). This  ;

reply should be clearly marked as a " Reply to a Notice of Violation" and l should include for each violation: (1) the reason for the violation, or, if '

contested. the basis for disputing the violation. (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full com)liance will be achieved. Your response may reference or include previous I docceted correspondence.e if the correspondence adequately addresses the required response. If an adequate reply is not received within the time.

specified in this Notice, an order or a Demand for Information may be issued as to why the license should not be modified suspended, or revoked, or why  ;

such other action as may be proper should not be taken. Where good cause is i shown, consideration will be given to extending the response time. '

Because your res)onse will be placed in the NRC Public Document Room (PDR). to 4 the extent possi)le. it should not include any personal privacy. 3roprietary, or safeguards information so that it can be placed in the PDR witlout i redaction. If personal privacy or proprietary information is necessary to l provide an acceptable response, then please provide a bracketed copy of your  ;

response that identifies the information that should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material you must s)ecifically identify the portions of your response that you seek to have with1 eld and provide in detail the bases for your claim of withholding (e.g., explain why the disclosure of'information  ;

will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.790(b) to support a request for withholding confidential commercial or financial information). If safeguards information ,

is necessary to provide an acceptable response, please provide the level of l

. protection described in 10 CFR 73.21.

Dated at Atlanta Georgia this 16th day of March 1998 ,

Enclosure 1