3F0398-18, Responds to NRC Re Violations Noted in Insp Rept 50-302/97-12.Corrective Actions:Review of EOPs Was Conducted to Identify Actions to Be Performed by Support Organizations

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Responds to NRC Re Violations Noted in Insp Rept 50-302/97-12.Corrective Actions:Review of EOPs Was Conducted to Identify Actions to Be Performed by Support Organizations
ML20217E276
Person / Time
Site: Crystal River Duke Energy icon.png
Issue date: 03/24/1998
From: Holden J
FLORIDA POWER CORP.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
3F0398-18, 3F398-18, 50-302-97-12, NUDOCS 9803300433
Download: ML20217E276 (20)
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Florida Power

=L, o.a msm March 24,1998 3F0398-18 U.S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington, D.C. 20555-0001

Subject:

Reply to Notice of Violations, NRC Inspection Report No. 50-302/97-12, NRC to FPC letter, 3N0298-16, dated Feh uary 23,1998

Dear Sir:

In the subject letter, Florida Power Corporation (FPC) received Notice of Violations. This correspondence provides our response to the violations.

Sincerely, Ichu J. J. Holden Director Site Nuclear Operations JJH/dwh

' Attachments xc: Regional Administrator, Region 11 1 Senior Resident inspector ,,

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ATTACHMENT 1 FLORIDA POWER CORPORATION NRC INSPECTION REPORT NO. 50-302/97-12 REPLY TO NOTICE OF VIOLATIONS VIOLATION 50-302/97-12 01

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.The Febr'uary 21,1984, Order modifying the Operating License confirms the Florida Power l' Corporation's implementation of NUREG-0737, " Clarification of TMl 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,"

LNUREG-0737 criterion I.C.1 provides clarification regarding 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 EOPs.

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 EOP 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.

W Administrative procedure Al-402C, AP and EOP 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 program did not adequately demonstrate the usability of emergency procedures in that:

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

2.- AsJof. January 9,' 1998, instructions for performing EOP actions referred to the

' Technical Support Center did not always exist.

3. L'A's' of December 8,1997, numerous differences between the ' EOPs and the Technical Bases Document had not been adequately technically justified.

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4. As of January 5,1998, step 3.1 of AP-770, Failed EDG Recovery, contained an l incorrect alpha-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 EOP-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.  ;

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6. As of December 8,1997, chemistry instructions were not complete and compatible with conditions of differing electrical bus availabilities when EOPs directed chemistry sampling.
7. As of December 8,1997, the EOPs or Technical Support Center procedures did not direct operators to implement OP-417, Containment Operating Procedure, for controlling the hydrogen concentration of the post-accident containment atmosphere.

ADMISSION OR DENIAL OF THE ALLEGED VIOLATION Florida Power Corporation accepts the violation.

BACKGROUND The controlling procedures for developing and revising Emergency Operating Procedures (EOPs) and Abnormal Procedures (APs) were revised during the first half of 1997.

References used for the changes included NUREG-1358, " Lessons Learned from the Special - Inspection Program for Emergency Operating Procedures," NUREG-1358, Supplement 1, and NUREG-899, " Guidelines for the Preparation of Emergency Coerating Procedures." A human factors consultant familiar with EOP programs assisted in the development of the procedures. New procedures were also developed for the control of the Cross Step Deviation Document and the Set Point and Set Point Basis Documents.

These documents were used for the development of EOP and AP changes and upgrades in

. conjunction with plant modifications and changes to. Technical Specification to support an extended forced outage.

REASON FOR THE VIOLATION The reason for the subject violation examples was inadequate management oversight of EOP/AP procedure development which resulted in the following:

Improperly interpreting and implementing regulatory requirements pertaining to EOP.

' actions that require support from other organizations, such as the Chemistry

. Department and the Technical Support Center.

Insufficient standards and expectations for personnelimplementing the in-plant field verification and validation of actions directed by the emergency procedures.

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lnadequateistandards and expectations for the' proper level of detail to justify

' deviations from the B&WOG Technical Basis Document (TBD).

CORRECTIVE ~ STEPS THAT HAVE BEEN OR WILL BE TAKEN AND THE- RESULTS ACHIEVED issue 1 A review of the _ EOPs _was: conducted to identify actions to be performed by: support organizations. ~ As a result, the affected support organizations developed plans to address these actions.

Administrative Instruction Al-600, " Conduct of Nuclear Plant Maintenance," was revised to require availability of qualified maintenance personnel to perform-

-. maintenance actions delineated in the EOPs and APs, including repair activities for a

. leaking Main'. Steam Safety Valve (MSSV) during back shifts. Preventive

- Maintenance procedure PM-275 was revised to add instructions for installation of a MSSV gag. Maintenance' personnel who are members of the fire brigade have been trained on this activity. This training included use of a MSSV mock-up.

' AL1500, " Conduct of the Chemistry and Radiation Protection Department," was revised to require sufficient qualified Radiation Protection Technicians and Chemistry Technicians on shift for operating modes. Sufficient personnel are now

.on shift to respond as needed.

Al 1505, " Conduct of Chemistry During Abnormal and Emergency Events," was developed to more clearly define the Chemistry Department's actions during abnormal and emergency events. The procedure includes instructions for Chemistry personnel relating to each abnormal and emergency event to which Chemistry is expected to respond. A table-top validation was performed on this procedure.

Al 1506, " Conduct of Health Physics During Abnormal and Emergency Events,"

was developed to more clearly define the Health Physics Department's actions during abnormal and emergency events. A table-top validation was performed on this procedure.

lasue 2 The EOPs were reviewed for steps that referred the control room procedure reader to the l

Technical Support Center. (TSC). For each of these steps, information has been j
incorporated iinto . Emergency Plan Implementing Procedures (ems) for the Accident -!
Assessment
Team (AAT). Several new ems were issued to include the necessary information and guidance.

W  : An enclosure was .added to EM-225,1" Duties of the Accident Assessment Team," to - j b Lpr' ovide the link between items to be monitored,' activities to be performed,' and the actual l

' guidance in separate ems.  ;

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' Page 5 of 20 The information in these procedures includes logs for.the AAT to record data from the control room, specific criteria to use for decision making, and general information to be utilized for broader evaluation of plant conditions.

Additionally, appropriate personnel have been trained on these procedures.

Issue 3 Following the December 1997 NRC visit, FPC completed the enhanced version of the EOP-

.TBD. cross. step documents along with the supporting documentation. This included .

addition of Appendix A~ for each EOP cross step document which provides' a detailed

. explanation of the reason why steps are sequenced differently than the TBD. The cross step document also ' references other supporting technical justification. documents. In addition to the cross step document enhancements, other actions taken to resolve the identified deficiencies included issuance of the new ems in January .1998. These new procedures provide _ an alternate location for TBD actions EOPs were also ' revised to eliminate sequence deviations, issue 4

'The revision to AP-770 containing the alpha-numeric error had not been issued at the time the error was identified. A review of the procedure, as a follow-up to identification of this error, identified an additional typographical error in Step 4.4 of Enclosure 4. The alpha-numeric designator for the reiriy, the location, and the typographical error were corrected and validated prior to the procedure being issued for plant restart.

Similar reviews were performed for the remaining APs. 'AP-604, "Waterbox Tube Failure,"

contained a' typographical error which was' corrected. Field validation documentation for two steps'in AP 990, " Shutdown From Outside the Control Room," could not be located.

'These steps were. revalidated and found to. be correct. Remaining APs either had field validation documentation or contained no field actions.

'S'ubsequent to this review, FPC identified a weakness in the field validations for APs. The specific issue .was descriptors for some equipment in the APs not exactly matching the descriptor on the equipment in the field. Corrective action for this deficiency is in 7 progress. Discrepancies will be' identified and a plan for resolving them will be developed

.and implemented by April 30,1998.

Issue 5 Subsequent to identification of.this issue, FPC concluded that the apparent missing fitting

..was in the box, although it was not easily found. To assure the operator can readily locate the fitting and other equipment needed, the fitting was attached to' the associated hose

.and elbow.i This assembly was bagged with other items needed to perform the procedure step! The bagged equipment is stored in the EOP box. This allows the operator to obtain

. required items readily.-

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Page 6.of 20 Surveillance Procedure SP-306, " Weekly Surveillance Log," has' been revised to ' add j Enciosures .111through 20 for checking the inventory of the EOP/AP key box and tool boxes.

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lssue6 FPC defines the issue. as performance of EOP/AP requested Chemistry sampling and analyses during a Loss Of Offsite Power (LOOP) with loss of one Engineered Safeguards (ES) bus or loss of a single circuit. Chemistry sampling and analyses suggested by the EOPs/APs have been identified. Actions taken'as a result of this include:

' Al 1505, " Conduct of Chemistry During Abnormal and Emergency Events," has ~

been issued. This procedure provides Chemistry personnel with a checklist of items to be performed subsequent to the occurrence of an abnormal or emergency event.

- The Primary Laboratory, Counting Room and Nuclear Sample Room electrical feeds were researched. FPC found an adequate number of circuits on each ES A, ES B and ES A/B bus available for use in performing the requested analyses, with the

exception of the Post Accident Sampling System (PASS). Reflective labels denoting l circuits and power feeds were placed on switches where drawings and previous l labels could be used to provide identity. Electrical failure scenarios are being compared to the circuitry associated with sampling to support the requested.

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L l ' analyses. Based on the results of'the review, appropriate contingency actions, including guidance and equipment, will be in place by August 5,1998.

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! Lighting contingencies for the Count Room and Nuclear Sample Room will be l l-established by May 1,1998.

i- Training of the Chemistry staff who will implement the procedural changes resulting I- from these actions will be completed by October 5,1998.

A review'was performed to determine whether other plant disciplines required special instructions for implementing EOPs. This review resulted in suggestions which have been provided to affected organizations, issue 7 EM-225A, " Post Accident RB [ Reactor Building] Hydrogen Centrol," was issued and replaces the portion of Operating Procedure OP-417 formerly used for post accident RB hydrogen control. EM-225 was revised to refer the AAT to EM-225A for post accident

, hydrogen purge. Also, EM-225A refers to Maintenance Procedure MP-815, " Installation of Post Accident H2 Purge Flow Instruments."

The EOPs and APs were reviewed to identify any other directed actions that did not have procedural guidance established. None were identified.

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Page 7 of 20:

Programmatic lasues

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' Based 'on' the deficiencies identified in the implementation of the field Verification' and

' Validation (V&V), the approach to the field V&V was changed. Additional experienced resources.were utilized and clear expectations were established. . Documentation of the L V&V was' enhanced by the use of digital photography, when determined necessary, to record specific' items during in-plant walk through validations. The field validations that-had been performed were redone and the remaining validations were performed following ithe 'new expectations. Al-402C, "AP and EOP Verification and Validation Plan," has been revised to incorporate these expectations.'

Sir,ilai to the field validation issue, expectations for the thoroughness and completeness of tasks related to the emergency proce'd ures were communicated to the team working on -

lthe procedures.

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iCORRECTIVE STEPS THAT' HAVE BEEN OR WILL BE TAKEN TO AVOID FURTHER

' VIOLATIONS

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The root cause determination developed for issues related to the EOP inspection has been reviewed by the Corrective Action Review Board (CARB) which is composed of senior level CR 3 management personnel. This provided a forum to discuss the identified management deficiencies.

> A core group of experiencad FPC personnel reporting to the Manager, Nuclear Plant Operations Support, is being maintained to support the EOP Program. This will provide continuity of the EOP Program and assure enhancements to the emergency procedure change process will obtain the maximum benefit from the lessons learned.

Administrative Procedures that control the dew.opment of the EOPs will be reviewed and i revised by. May 29, 1998, to ensure EOP actions directed to. departments other than

Operations are included in apprepriate procedures and to ensure they receive V&V consistent with the EOPs.

FPC.will benchmark the EOP programs at other utilities and evaluate further enhancements

.to the EOP Program. Changes based on the benchmark evaluations will be completed by

September 28,1998.

,., ) Administrative' Procedures that"co'ntrol the development of the EOP-TBD Cross Step

~ Document: will be reviewed and' revised by September 28, 1998, to ensure that 7 expectations 'are ; clearly established regarding the level of detail required 'to -justify -j ideviations from the TBD.~ -)

~ DATE WHEN FULL COMPLIANCE WILL 8E ACHIEVED c -

l FPC will achieve full compliance by October 5,' 1998. 1 3

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' 3F0398-18 Page 8 of 20 VIOLATION 50-302/97-12-02 10 CFR 50, Appendix B, Criterion XVI, Correctivo 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 TMl 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,"

NU REG-0737, Criterion I.C.1 provides clarification regarding 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 EOPs.

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 EOP 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 . . cense confirms in part the Florida Power Corporation's implementation of NUREG-070, " Clarification of TMI Action Plan Requirements," and Supplement 1 to NUREG-0737, " Requirements for Emergency Responso Capability," Criterion ll.B. 2, " Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces / Systems Which May be Used in Postaccident Operations."

NUREG-0737, criterion ll.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 i Criterion ll.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-C971533 dated March 3,1997 associated with in-plant operator j accessability was not adequately corrected in that the corrective actions did not  ;

evaluate the compatibility with conditions associated with the projected radiological doses to individuals for necessary occupancy times in vital areas, such as the dose j l

U. S. Nuclear R~gulatory C:mmissi:n 3F0398-18 i Page 9 of 20 incurred when aligning high pressure auxiliary 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 3 blowdown actions in response to a postulated steam generator tube rupture was i not prompt in that completion of the radiological dose calculation had yet to be I performed and was not scheduled to be performed until after reactor startup.

2. As of October 20, 1997, the extent of the licensee's correctiv'e 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 temperaturea were not specified and the procedures for calibration of instruments located in the Auxiliary Building did not assure that the Auxiliary Building temperatures were maintained within the temperature ranges assumed in the instrument loop uncertainty set point calculations supporting EOP related set points, such as calculations 191-0028 and 190-0022, which was the same condition adverse to quality identified in Violation 50-302/97-01-07.

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ADMISSION OR DENIAL OF THE ALLEGED VIOLATION l

1 Florida Power Corporation accepts the violation. j REASON FOR THE VIOLATION lasue 1 The reason for the issue was personnel error in that the regulatory requirements of NUREG- j 0737, item II.B.2, were too narrowly applied by FPC personnel. Mission dose had been )

analyzed for only those actions required for accident mitigation instead of for any operator 1 action performed outside the Control Complex Habitability Envelope (CCHE) during accident conditions. Additionally, industry experience / benchmarking was not utilized in the l development of EOPs. This resulted in not performing an extent of condition review and l not implementing prompt corrective action when the lack of mission dose analyses was l

identified.  ;

l 11MHL2 The reason for the issue was a management decision early in the recent outage restart process to suspend efforts to revise instrument loop uncertainty and setpoint calculations.

This decision was based on the most safety-significant calculations (Technical Specification related) having already been addressed. Additionally, interim administrative controls were in place to address the concerns identified in Violation 50-302/97-01-07.

(See corrective actions that have been or will be taken and the results achieved for a list of i the in-process corrective actions.] I I

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  • U. S. Nuclerr Regulatory C:mmissi:n 13F0398-18 Page 10 of 20 CORRECTIVE STEPS THAT HAVE BEEN OR' WILL BE TAKEN AND THE RESULTS ACHIEVED lasue 1 Individuals responsible for the disposition of Precursor Cards 97-1533 and 97-7125 have been coached and counseled on the subject violation issue.

Operator actions outside the CCHE directed by EOPs during accident conditions have been reviewed. -FPC concluded that either the operator actions could be performed or that not performing an operator action did .not affect the outcome or inhibit mitigation of the accident.

Dose calculations for the Steam Generator Tube Rupture (SGTR) Accident will be finalized by April 30,1998.

Dose calculations for Hydrogen Purging Operations will be finalized and incorporated into Revision 25 to the FSAR which is scheduled to be submitted by August 2,1998.

The remainder of the EOP actions will be evaluated as part of the re-baselining effort of Environmental Qualification and source terms. The re-baselining and EOP mission dose analysis for operator actions outside the control room will be completed prior to restart following the next refueling outage.

Issue 2 Calculations 190-0022, Revision 0, and 191-0028, Revision 1, have been evaluated to assess the impact of the omission of the uncertainty / error contributors identified during the

' inspection. The evaluations produced results that were larger than those in the existing calculations. The additional errors in calculation 190-0022, Revision 0, resulted in revised loop uncertainties for flow indication and control which were conveyed to the EOP Group.

The additional errors in calculation 191-0028, Revision 1, were evaluated against the EOP setpoint basis documents and found to be bounded by the existing uncertainties.

I Concerning the issue of translating design calculation inputs and assumptions into implementing procedures, specific and programmatic corrective actions had been identified and were being implemented prior to the inspection, immediate and long term corrective j actions associated with the resolution of these issues included: l On May 9,1997, a Maintenance Study Book Entry was developed and training was provided to I&C technicians to ensure design calculation temperature assumptions were maintained during the calibration process. This entry include,i guidance on how to stay within temperature limits and developed an " attachment" to be used by  ;

the I&C technicians to ensure that design temperature requirements were satisfied. i A review had been performed to identify temperature sensitive l&C Shop Calibration Surveillance Procedures. By May 30,1997, Nuclear Procedure Observation and Suggestion Tracking -(NUPOST) entries had been initiated against the impacted i

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' 3F0398-18 Page 11 of 20 procedures to ensure required temperature range limitations were appropriately addressed.

Procedural guidance had been developed to ensure that engineering requirements specified in calculations and used for design inputs receive appropriate review.

Administrative processes had been established to maintain the intermediate, Auxiliary, Reactor, and Control Complex Buildings within specified temperatures.

The initial temperature monitoring program was developed and implemented in Chem / Rad instruction CRI-406, " Building Ambient Temperature Monitoring." The station temperature monitoring program ensures proper system performance and validates I&C calculation specified temperature related design inputs and assumptions.

The above administrative controls would have prevented inadvertent introduction of additional temperature uncertainties (beyond those presently assessed) into the calibration process for calculation 190-0022, Revision 0, related instruments. For calculation 191-0028, Revision 1, engineering personnel did not specify temperature related calibration requirements. This occurred due to differences in the skill / knowledge level of the individuals performing calculations 190-0022 and 191-0028, Revision 1, and due to inadequate design guidance for determining loop uncertainties.

CORRECTIVE STEPS THAT HAVE BEEN OR WILL BE TAKEN TO AVOID FURTHER VIOLATIONS lssue 1 Administrative Instruction Al-402C, Revision 6, dated January 8,1998, incorporated a requirement to contact the Safety Analysis Group to determine if changes to EOPs affect radiological dose to the step performer.

Training will be provided to engmeering personnel by April 30,1998, to ensure cognizance of the potential impact of various activities on CR-3 documentation as related to source terms, Environrnental Qualification, mission dose, and accident reanalysis. j l

Nuclear Engineering Procedure NEP-281, " Nuclear Fuel Design," will be revised by June {

30,1998, to incorporate verification that the source terms used in new fuel cycle reload l reports are consistent with the CR-3 licensing and design bases. Additionally, the j verification will include a review for possible impact on dose calculations performed for {

regulatory required programs to ensure the current fuel cycle analysis bounds previous 1 results.

A program will be developed and implemented by September 30,1998, which will ensure consistency and provide guidance for dose assessment calculations.

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' 3F0398 Page 12 of 20 issue 2 The l&C Design Criteria For Instrument Loop Uncertainty Calculations, Revision 1, was -

issued September 27, 1996, to reflect current industry practice for determining loop uncertainties. This criteria will ensure temperature related calibration requirements are considered in loop uncertainty calculations.

In 'FPC to . NRC letter 3F0298-09, dated February 9, 1998, FPC stated that an Analysis / Calculation Upgrade Program was being developed and that a program plan was being created to include scope, prioritization, schedule, and methodology. The program

- will address inadequacies in existing instrument loop uncertainty and setpoint calculations.

Revision to the two subject calculations (190-0022 and 191-0028) will be accomplished in

- accordance with this program. FPC stated that the plan will be completed by June 19, 1998, and that a schedule for completion of the program will be provided to the NRC at that time. '

DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED FPC will achieve full compliance by September 30,1998.

VIOLATION 50-302/97-12-05 10 CFR 50, Appendix B, Criterion 11, Quality Assurance Program, requires 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.

The Quality Assurance Program as described in the Updated Final 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, subsection 3.2, states in part "The design input shall include but is not limited to ... Environmental conditions anticipated during ... operation such as ... nuclear i radiation," and ... " Operational requirements under various conditions, such as ... plant I emergency operation ..."

ANSI 45.2.11, subsection 4.2, states " Analysis shall be sufficiently 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 I 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 j accordance with those documents in that: i

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 l

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  • 3F0398-18 Page 13 of 20 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 14B.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.

ADMISSION OR DENIAL OF THE ALLEGED VIOLATION Florida Power Corporation accepts the violation.

REASON FOR THE VIOLATIQN lasue 1 The reason for the issue was personnel error. Calculation M93-0006 was performed by an outside agency Upon delivery to FPC, an adequate engineering review and validation was not performed on the subject calculation prior to issuance. Calculation design input and assumptions were not validated.

A contributing cause was CR-3 personnel not thoroughly understanding the requirements for hydrogen purging operation. The FSAR Chapter 148 discussion on hydrogen purging provides three examples and times for purging operations ranging from 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> to 980 hours0.0113 days <br />0.272 hours <br />0.00162 weeks <br />3.7289e-4 months <br /> and 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br />. Tables for purging operations only addressed the 980 hour0.0113 days <br />0.272 hours <br />0.00162 weeks <br />3.7289e-4 months <br /> and  ;

1500 hour0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> scenarios. Since the FSAR Tables did not address the 250 hour0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> scenario, FPC personnel mistakenly concluded that the 250 hour0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> example was an explanation of an unlikely purging start time but was not a specific requirement for establishing purging 4 operations. )

1' issue 2 The reason for the issue was personnel error. Calculation E90-0023 was performed by an outside agency. Upon delivery to FPC, an adequate engineering review and validation was not performed on the subject calculation prior to issuance. The design assumption for ,

carbon steel corrosion rates due to interaction with boron in the Building Spray System I was based on preliminary information supplied to one outside agency by another outside agency. The basis and justification for' use of the carbon steel corrosion' rate was not l questioned by FPC personnel.

l CORRECTIVE STEPS THAT HAVE BEEN OR WILL BE TAKEN AND THE RESULTS -

ACHIEVED lasue .1 i The purging operation start time has been re-evaluated using the guidance contained in 10CFR50.44 and. Regulatory Guide 1.7, Revision 2. That preliminary evaluation concluded that hydrogen purge may be required as early as Day 16. Starting the purge at Day 16

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following a design basis accident meets the requirements of 10CFR50.44(g) and is bounded by the previous FSAR Chapter 14B analysis for a continuous purge starting at 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br />. FPC will finalize review of the evaluations and calculations for hydrogen purge by May 31,1998.

FSAR Chapter 14B will be revised to incorporate the results of this evaluation and will be included in Revision 25 to the Updated FSAR, which is scheduled to be submitted by August 2,1998. ,

J A commitment for re-baselining the EOP mission dose analyses for operator actions outside the Control Room is provided under FPC's response to Violation 50-302/97-12-02, Issue 1.

. Issue 2 The corrosion rate for carbon steel was corrected in Revision 2 to calculation E90-0023, issued January 5,1998.

Calculation E90-0023 was reviewed to determine if additional unconfirmed information may have been used as design input. Other metals in contact with the acidic solution had their corrosion rates properly documented (aluminum, inconel, stainless steel).

A sample of the balance of the affected vendor calculations was reviewed for invalidated assumptions and design inputs. No significant unverified assumptions were identified.

CORRECTIVE STEPS THAT HAVE BEEN TAKEN TO AVOID FURTHER VIOLATIONS Since the subject calculations were issued, changes have been made to Nuclear Engineering . Procedure NEP-213, " Design Analyses / Calculations," to enhance the verification of design inputs and assumptions contained in calculations and analyses. On June 26,1995, NEP-213, Revision 9, became effective. That revision specifically required the Design Engineer to assure calculations prepared by external engineering organizations and vendors have a design inputs / assumptions review performed. That review, the results of the review,^ and an acceptability review are required to be conducted for documents which may affect operations prior to issuance of the calculation.

DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED FPC will achieve full compliance by August 2,1998.

VIOLATION 50-302/97-12-08 Updated Final Safety Analysis Report, Section 14.2.2.5.4, ECCS Qualification, states that, "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 ani 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

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further states that BAW-10104, Rev. 3, is the methods report on how the computer model 4 used to ensure compliance with 10 CFR 50.46 will be assembled and run. Also, the "The 1 LBLOCA appHeation 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 tM onset of long-term cooling and the end of core cooling requirernents.... 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 Assembly 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 pump operating with injection through its associated injection line and with the crossover to the associated HPl string open; the associated HPl pump would be pumping through its HPl linsi."

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 operation was specified in the original purchase order for the high pressure injection pumps.

A_DMISSION OR DENIAL OF THE ALLEGED VIOLATION Florida Power Corporation accepts the violation.

BEASON FOR THE VIOLATION 1

The reason for the violation was personnel error. The original Fimergency Core Coohng i System (ECCS) design analysis for CR-3 focused on Large Break LOCA (LBLOCA) scenarios. Small Break LOCA (SBLOCA) scenarios that require High Pressure Injection (HPI) for longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> were not considered when CR-3 was designed.

Requirements for SBLOCA, design changes resulting from the Three Mile Island (TMI) l incident, and new post-TMI requirements for long term cooling capabilities all served to change the intended use of HPl from a short term, less than 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post accident  ;

function, to a system with a much longer post accident mission time. As these changes were made, a review of the original purchase specifications was not made to verify installed equipment was capable of performing to the new requirements.

CORRECTIVE STEPS THAT HAVE BEEN OR WILL BE TAKEN AND THE RESULTS ACHIEVED i

An evaluation concluded that the HPI pumps are capable of long term operation in the piggyback mode for the expected mission time. An additional concern was identified with the ability of the HPl pumps to operate with entrained solids during long-term post accident conditions. Debris within the pumped medium had not been considered in the original pump specifications. The reactor building sump screen is a 1/4 inch mesh. An evaluation

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' 3F0398-18 Page 16 of 20 concluded that debris in the pumped medium allowed past the reactor building sump screen would not affect the HPl pumps' long-term operation capability in the piggyback mode. Quality documentation for the HPl pumps will be upgraded by August 31,1998, to

. reflect the current mission time.

A review of the Purchase Order for the Decay Heat and Building Spray pumps shows that the expected post accident operation was one year for the Decay Heat pumps, and 30 days for the Building Spray pumps. Considering these durations, there is no concern about mission time for these pumps. However, the purchase order does not address any requirements or specifications for debris within the pumped medium. An evaluation of this condition concluded that the pumps are capable of operating with the expected reactor building sump debris during post accident conditions.

With the increased mission time requirements of the HPi system, other potential effects were also evaluated. These included long term effects of radiation exposure during the extended mission time, pump runout, long term temperature effects, and mechanical wear.

The evaluation of these issues concluded that there were no adverse effects that impeded the ability of the HPl system to fulfill current mission time requirements.

An evaluation of the potential for plugging or damage to HPl control valves MUV-23/24/25/26, Stop Check Valves MUV 2/6/10, and HPl Pump recirculation flow orifices MU-82/83/84-FO was performed. Each of these components was determined to be capable of performing their intended safety function with debris in the process fluid.

A review of the piggyback method of operation showed that there are other ECCS components within the plant that are used in a manner not consistent with the type or degree of service normally intended for such components. These components are DHV-5/6 and MUV-2/6/10. The current uses of DHV-5/6 and MUV-2/6/10 were not specifically addressed in the purchase specifications for the components. The ability of DHV-5/6 and MUV-2/6/10 to operate under all anticipated throttling conditions has been evaluated and determined to pose no credible threat to performance of the Low Pressure injection (LPI) or HPl safety function.

CORRECTIVE STEPS THAT HAVE BEEN TAKEN TO AVOID FURTHER VIOLATIONS NEP-213, Exhibit 1, Section E.2, requires the responsible design engineer to specifically consider the impact on and the need for revising documents such as the Enhanced Design Basis Document, Configuration Management Information System, or Vendor Qualification Package. Information related to original purchase specifications can be specified by or located within these documents, in addition, NEP-213, Section C.1, refers to NEP-210,

" Modification Approval Records," for identifying applicable design inputs. Design inputs that could be invalidated by field activities or operating conditions should be clearly 1 identified in the calculation.

NEP-210, Exhibit 2, item B.4, states that attributes of the actual item / service purchased under a specification may have more detail or include exceptions to the specification whmh FPC has accepted. In this case, design input must be either revised in the specific ion or clarified in the design input. Exhibit 2, item C.d, states the Design input Recoid must j

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. Page 17.of 20 contain justification for addition, upgrade or deletion of essential equipment. Additionally, Exhibit 2, Design Data Sheet, items 1 through 11, address applicable design requirements related to component specifications. In many cases, these considerations can only be answered through review of component purchase specifications.

The above .NEP references provide adequate guidance for ensuring calculations and analyses consider the potential impact on installed plant equipment specificati~ons.

DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED FPC will achieve full compliance by August 31,1998.

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'i ATTACHMENT 2 LIST OF COMMITMENTS The following table identifies those actions committed to by FPC in this document.

Response Commitment? < ,

< Due; Date?

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'Sectisns - ~.

Page 4 Al-600, " Conduct of Nuclear Plant Maintenance," was Complete revised to require availability of qualified maintenance personnel to perform maintenance actions delineated in the EOPs and APs, including repair activities for a leaking Main Steam Safety Valve during back shifts.

Page 4 PM-275 was revised to add instructions for installation Complete of a MSSV gag.

Page 4 Al-1600, " Conduct of the Chemistry and Radiation Complete Protection Department," was revised to require sufficient qualified Radiation Protection and Chemistry Technicians be on shift for operating modes.

Page '4 Al-1505, " Conduct of Chemistry During Abnormal and Complete Emergency Events," was developed to more clearly define the Chemistry Department's actions during abnormal and emergency events. The procedure includes instructions for Chemistry personnel relating to each abnormal and emergency event to which Chemistry is expected to respond.

Page 4 Al-1506, " Conduct of Health Physics During Abnormal Complete and Emergency Events," was developed to more clearly define the Health Physics Department's actions during abnormal and emergency events.

Page 4 An enclosure was added to EM-225, " Duties of the Complete Accident Assessment Te a m," to provide the link between itcms to be monitored, activities to be performed, and the actual guidance in separate ems.

Page 5 Discrepancies will be identified and a plan for resolving April 30,1998 them will be developed and implemented by April 30, 1998.

Page 6 Surveillance Procedure SP-306, " Weekly Surveillance Complete ,

I Log," has been revised to add Enclosures 11 through 20 for checking the inventory of the EOP/AP key box and tool boxes.

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Page 6 - Electrical failure scenarios are being compared to the August 5,1998 circuitry associated with sampling to support the g requested analyses. Based on the results of the Yp review, appropriate contingency actions, including guidance and equipment, will be in place by August 5, 1998.'

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Response Commitment - Due Date)

Section1 Page 6 Lighting contingencies for the Count Room and Nuclear May 1,1998 j Sample Room will be established by May 1,1998.

Page 6 Training of the Chemistry staff who will implement the October 5,1998 procedural changes resulting from these actions will be completed by October 5,1998.

Page 6 EM 225A was issued and replaces the portion of OP- Complete 417 formerly used for post accident RB hydrogen control.

Page 6 EM-225 was revised to refer the AAT to EM-225A for Complete post accident hydrogen purge.

Page 6 EM-225A refers to Maintenance Procedure MP-815, Complete

" Installation of Post Accident H2 Purge Flow Instruments."

Page 7 Al-402C, "AP and EOP Verification and Validation Complete Plan," has been revised to incorporate these expectations.

Page 7 Administrative Procedures that control the May 29,1998 development of the EOPs will be reviewed and revised by May 29,1998, to ensure EOP actions directed to departments other than Operations are included in appropriate procedures and to ensure they receive V&V consistent with the EOPs.

Page 7 FPC will benchmark the EOP programs at other utilities September 28,1998 and evaluate further enhancements to the EOP program. Changes based on the benchmark evaluations will be completed by September 28,1998.

Page 7 Administrative Procedures that control the September 28,1998 development of the EOP-TBD Cross Step Document will be reviewed and revised by September 28,1998, to ensure that expectations are clearly established regarding the level of detail required to justify deviations from the TBD.

Page 10 Dose calculations for the Steam Generator Tube April 30,1998 Rupture (SGTR) Accident will be finalized by April 30, 1998.

Page 10 Dose calculations for Hydrogen Purging Operations will August 2,1998 be finalized and incorporated into Revision 25 to the FSAR which is scheduled to be submitted by August 2, 1998.

Page 10 The remainder of the EOP actions will be evaluated as Prior to restart  !

part of the re-baselining effort of Environmental following the next i Qualification and source terms. The re-baselining and refueling outage.

EOP mission dose analysis for operator actions outside the control room will be completed prior to restart following the next refueling outage.

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Page 20 of 20 Response Commitmenti ,

Due Date?

Section P ~

Page 11 Administrative Instruction Al-402C, Revision 6, dated Complete January 8,'1998, incorporated a requirement to contact the Safety Analysis Group to determine if changes to EOPs could affect radiological dose to the step performer.

Page -11 Training will be provided to engineering personnel by April 30,1998 April 30,1998, to ensure cognizance of the potential impact of various activities on CR 3 documentation as related to source terms, Environmental Qualification, mission dose, and accident reanalysis.

Page 11 Nuclear Engineering Procedure NEP-281, " Nuclear Fuel June 30,1998 Design," will be revised by June 30, 1998, to incorporate ver!fication that the source terms used in new fuel cycle (eload reports are consistent with the CR-3 licensing and design bases. Additionally, the j verification will include a review for possible impact on dose calculations performed for regulatory required programs to ensure the corrent fuel cycle analysis bounds previous results.

Page 11 A program will be developed and implemented by September 30,1998 September 30, 1998, which will ensure consistency and provide guidance for dose assessment calculations.  ;

Page 12 in FPC to NRC letter 3F0298-09, dated February 9, June 19,1998 1998, FPC stated that an Analysis / Calculation Upgrade Program was being developed and that a program plan was being created to include scope, prioritization, schedule, and methodology. The program will address inadequacies in existing instrument loop uncertainty and setpoint calculations. Revision to the two subject calculations (190-0022 and 191-0028) will be accomplished in accordance with this program. FPC stated that the plan would be completed by June 19, 1998, and that a schedule for completion of the program would be provided to the NRC at that time.

Page 14 FPC will finalize review of the evaluations and May 31,1998 calculations for hydrogen purge by May 31,1998.

Page 14 FSAR Chapter 14B will be revised to incorporate the August 2,1998 results of this evaluation and will be included in Revision 25 to the Updated FSAR, which is scheduled to be submitted by August 2,1998.

Page 16 Appropriate Quality Assurance Program documentation August 31,1998 for the HPl pumps will be upgraded by August 31, 1998, to reflect the current mission time.

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