ML20203F556

Notice of Violation from Insp on 971021-24,1208-12 & 980105-09.Violations Noted:Implementation of Validation Program Did Not Adequately Demonstrate Usability of Emergency Procedures

ML20203F556
Person / Time
Site:	Crystal River
Issue date:	02/23/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20203F553	List: ML20203F548 ML20203F556 ML20203F563
References
50-302-97-12, NUDOCS 9803020002
Download: ML20203F556 (6)

Text

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NOTICE OF VIOLATION Florida Power Corporation Docket No. 50-302 Crystal River Unit 3 License No OPR-72 During an NRC inspection conducted on October 20 through 24, 1997. December 8 through 12. 1997, and January 5 through 9. 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. The February 21, 1984. Order modifying the Operating License confirms the Florida Power Corporation's imalementation of NUREG 0737.

" Clarification of TMl Action Plan Requirements." and Supplement 1 to NUREG-0737 " Requirements for Emergency Response Capability." criterion 1.C.1. . " Guidance for the Evaluation and Development of Procedures for Transients and Accidents."

NUREG-0737 criterion I.C.1 provides clarification regardina the requirements for reanalysis of transients and accidents. Item 7 of Supplement 1 to NUREG-0737. " Upgrade Emergency Operating Procedures (EOPs)" requires that licensees develop a procedures generation package (PGP) which included a description of the validation program for the E0Ps.

By Letter dated March 25, 1983. Florida Power Corporation submitted a (PGP) in response to the I.C.1 requirement of NUREG-0737. Supplement 1.

The response contains a discussion of the upgraded E0P validation program which states, in part that the purpose of the validation program 1s to be demonstrate the usability of emergency procedures. The instructions to operators must be complete, understandable and.

compatible with conditions.

Administrative procedure AI-402C AP and E0P Verification and Validation Plan, enclosure 5. Evaluation Criteria for Procedure Validation, requires an assessment to ensure in-plant actions are not hampered by inaccessibility or environmental conditions. Enclosure 3. " Verification of Technical Accuracy," requires differences between the procedure and the Technical Bases Document be documented and justified.

Contrary to the above. the implementation of the validation r.cogram did not adequately demonstrate the usability of emergency procedures in that:

1. As of December 8. 1997, actions designated in E0Ps for chemistry and maintenance personnel to perform were not always possible due to the lack of personnel staffing requirements.

2. As of January 9,1998. instructions for performing E0P actions referred to the Technical Support Center did not always exist.

9803020002 900223 PDR ADOCK 05000302 G PDR

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3. As of December 8, 1997, numerous differences between the E0Ps and the Technical Bases Document had not been adequately technically justified.

4. As of January 5,1998, ste) 3.1 of AP-770, Failed EDG Recovery, contained an incorrect alpla numeric designator and location for the 86 relay causing operators to be unable to recover the diesel generator in accordance with the procedure during a simulated transient in January 1998,

5. As of December 8, 1997, the fitting used in E0P-14. Enclosure 6.

OTSG Blowdown Lineup, step 6,3 was not readily available causing an operator to be unable to vent the blowdown line in accordance with the procedure during a simulated transient in December 1997,

6. As of December 8, 1997, chemistry instructions were not complete and compatible with conditions of differing electrical bus availabilities when E0Ps directed chemistry sampling,

7. As of December 8, 1997, the E0Ps or Technical Support Center i procedures did not direct operators to implement OP-417.

Containment Operating Procedure, for controlling the hydrogen concentration of the post-accident containment atmosphere.

This is a Severity Level IV Violation (Supplement 1).

B. 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, requires that measures be established to assure conditions adverse to quality be 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.

The February 21, 1984, Order modifying the Operating License confirms the Florida Power Corporation's implementation of NUREG 0737,

" Clarification of TMI Action Plan Requirements," and Supplement 1 to NUREG-0737, " Requirements for Emergency Response Capability," Criterion I.C.1., " Guidance for the Evaluation and Development of Procedures for Transients and Accidents "

NUREG 0737, Criterion I.C.1 provides clarification regarding the requirements for reanalysis of transients and accidents. Item 7 of i Supplement 1 to NUREG-0737, " Upgrade Emergency Operating Procedures (EOPs)" recuires that licensees develop a procedures generation package (PGP) which included a description of the validation program for the

.E0Ps.

By Letter dated March 25, 1983. Florida Power Corporation submitted a PGP in response to the I.C.1 requirement of NUREG-0737. Supplement 1.

Enclosure 1

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The response contains a discussion of the upgraded E0P validation program which states, in part that the purpose of the validation program is to be demonstrate the usability of emergency procedures. The instructions to operators must be complete, understandable and, compatible with conditions.

The March 14. 1983. Order modifying the Operating License confirms '.n ,

part the Florida Power Corporation s implementation of NUREG-0737,

" Clarification of TMI Action Plan Requirements." and Supplement 1 to NUREG 0737, " Requirements for Emergency Response Capability," Criterion .

11.B.2, " Design Peview of Plant Shielding and Environmental Qualification of Equipment for Spaces / Systems.Which May be Used in Postaccident Operations." .

NUREG 0737, criterion II.B.2 " Design Review of Plant Shielding and

, Environmental Qualification of Equipment for Spaces / Systems Which May be

, used in Postaccident Operations." requires that licensees provide adequate access to vital areas to increase the capability of operators to control and mitigate the consequences of an accident. Per Criterion II.B.2, a vital area is defined as, 'Any area which will or may require occupancy to permit an operator to aid in the mitigation of or recovery from an accident is designated as a vital area."

Contrary to the above.

1. As of January 9,1998, a condition adverse to quality identified in licensee precursor card 3-C97-1533 dated March 3, 1997 associated with in-]lant operator accessability was not adequately corrected in that t1e corrective actions did not evaluate the compatibility with conditions associated with the projected radiological doses to individuals for necessary occupanc times in vital areas, such as the dose incurred when aligning hig pressure aub .iary spray or equalizing pressure across the MSIVs e.g.. any area which will or- may require occupancy to permit an operator to aid in the mitigation or recovery from an accident), and the corrective action to a condition adverse to quality identified in licensee precursor card 3-C97-7125 dealing with operator radiological doses incurred during steam generator blowdown actions in response to a postulated steam generator tube rupture was not prom)t in that completion of the radiological dose calculation lad yet to be performed and was not scheduled to be performed until after reactor startup.

2. As of October 20.1997, the extent of the licensee's corrective actions to a significant condition adverse to quality. Violation 50 302/97-01-07. " Instrument Loop Uncertainty Set point Calculation Assumptions Not Translated Into Procedures,' was inadequate in that the calibration temperatures were not specified and the procedures for calibration of instruments located in the Enclosure 1

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< Auxiliary Building did not assure that the Auxiliary Building temperatures were mainta1ned within the temperature ranges assumed in the instrument loop uncertainty set point calculations '

supporting E0P related set points, such as calculations 191 0028 and 190 0022. ahich was the same condition adverse to quality identified in Violation 50 302/97 01-07.

This is a Severity Level IV Violation (Supplement 1).

C. Updateo Final Safety Analysis Report. St.; tion 14.2.2.5.4. ECCS Qualification. states that. "In o der to qualify the ECCS. the NRC placed requirements on the ECCS t > 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 comsliance with 10 CFR 50.46 will be assembled and run. Also.

l the "The L3LOCA application report for the 177 FA lowered loop plants is BAW-10103A." Chapter 10 of BAW 10103A and BAW 10104 states in part.

"The duration of long-term cooling is the period between the onset of long-term cooling and the end of core cooling requirements. . . . The exact duration of long-term cooling will vary.... A realistic assessment of the duration for the worst case is approximately one month."

Chapter 10. Long-Term Cooling. of Topical Report BAW-10103A. Rev. 3.

"ECCS Analysis of B&W 177-Fuel Assem]ly Lowered-Loop NSSS." and Topical Report BAW-10104. Rev. 3. "ECCS Analysis Of B&W s 177-FA Lowered Loop

. NSS." states in part that one of the three long-term cooling methods is "one LPI pum) o]erating with injection through its associated injection

! line and wit 1 t1e crossover to the associated HPI string open: the associated HPI pump would be pumping through its HPI lines."

10 CFR 50. Appendix B Criterion IV. Procurement Document Control, requires that measures be established to assure applicable regulatory requirement and design bases are suitably included in the documents for

procurement of equipment.

Contrary to the above, an applicable regulatory requirement was not suitably included in the documents for procurement of equipment in that one day of post-accident o)eration was specified in the original purchase order for the hig, pressure injection pumps.

This is a Severity Level IV Violation (Suppleinent I) l Enclosure 1 i

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5-D, 10 CFR 50. Appendix B. Criterion II. Quality Assurance Program, requires that a quality assurance program be established. This program shall be docu 4nted by written policies procedures or instructions and carried out in accordance with those documents.

TheQualityAssuranceProgramasdescribedintheUpdatedFinalSafety 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. subsection 3.2. states in part ."The design input shall include but is not limited to ... Environmental conditions anticipated during ... operation such as ... nuclear radiation." and ,..

" Operational requirements under various conditions. such as . . . plant emergency operation ..."

ANSI 45.2.11. subsection 4.2. states " Analysis shall be sufficie'ntly detailed-as to purpose, method, assumptions, design input, references and units such that a person technically qualified in the subject can review and understand the analyses and verify the adequacy of the results without recourse to the originator."

Contrary to the above, as of December 8. 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 calculation. M93 0006. used to determine the post accident radiation doses to personnel for purging the reactor building for hydrogen control used an incorrect design input of 25 days after the accident to establish the radiological source term instead of the 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> after the accident as stated in UFSAR Section 148.3.3. and, the time assumptions for operating valves were not validated.

2. The assumption of the corrosion rate for carbon steel piping exposed to a boric acid containment spray in a post accident environment in Calculation E-90 0023. Evaluation for Containment Spray between pH 4.0 and 12.5 could not be understood by a technically qualified person without recourse to the originator.

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

Pursuant to the provisions of 10 CFR 2.201. Florida Power Corporation is hereby required to submit a written statement or explanation to the U.S.

Nuclear Regulatory Commission. ATTN: Document Control Desk. Washington. 0.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 the date of the letter transmitting this Notice of Violation (Notice). This Enclosure 1

6 reply should be clearly marked as a " Reply to a Notice of Violation" and

,. should include for each violation: (1) the reason for tne 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

[ , docteted correspondence if the correspondence adequately addresses the s 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 shown. consideration will be given to extending the. response time.

Because your response will be placed in the NRC Public Document Room 'PDR), to the extent possiale, it should not include any personal privacy. )roprietary, or safeguards information so that it can be placed in the PDR witlout redaction. If personal privacy or proprietary information is necessary to 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 withleid 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 protection described in 10 CFR 73.21.

Dated at Atlanta. Georgia this 23th day of February 1998 Enclosure 1

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