3F0498-15, Forwards Response to Violations Noted in Insp Rept 50-302/98-02.Corrective Actions:Listing of Plant Equipment Utilized for in-plant Actions,By Tag Number Equipment Noun Name & Plant Location,Compiled from Review of Each EOP

Forwards Response to Violations Noted in Insp Rept 50-302/98-02.Corrective Actions:Listing of Plant Equipment Utilized for in-plant Actions,By Tag Number Equipment Noun Name & Plant Location,Compiled from Review of Each EOP

ML20216G824
Person / Time
Site:	Crystal River
Issue date:	04/15/1998
From:	Holden J FLORIDA POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
3F0498-15, 3F498-15, 50-302-98-02, 50-302-98-2, NUDOCS 9804200498
Download: ML20216G824 (13)

Text

1 4

.. t Florida Power 22 RAJgoN 7M'e'n",*d7.*s' w.. oen.72 1

April 15,1998 l 3F0498-15 f

l U.S. Nuclear Regulatory Commission Attention: Document Control Desk l i

Washington, D.C. 20555 0001

Subject:

Reply to Notice of Violations, NRC Inspection Report No. 50-302/98-02, NRC to FPC letter, 3NO398-11, dated March 16,1998

Dear Sir:

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

Sincerely, f

J. . Holden Director [ <

Site Nuclear Operations j

i JJH/dwh Attachments l xc: Regional Administrator, Region 11 Senior Resident inspector NRR Project Manager i l

9804200498 980415 PDR ADOCK 05000302 o PDR CRYSTAL RIVER ENERGY COMPLEX: 16760 W. Power Line Street

• Crystal River, Florida 34428 6708 a (352)795-6486 A Florida Progress Company

. l

( ',' U.S. Nucinr R::guictory Commission l 3F0498-15 Page 2 of 13 I

ATTACHMENT 1 l

[

FLORIDA POWER CORPORATION NRC INSPECTION REPORT NO. 50-302/98-02 REPLY TO NOTICE OF VIOLATIONS VIOLATION 50-302/98 02-01 -

l 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, 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.

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 EOP actions were not adequate to preclude repetitive errors in procedures used for in-plant EOP 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 l 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.

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

ADMISSION OR DENIAL QF THE ALLEGED VIOLATIQN l l

Florida Power Corporaiion accepts the violation.

REASON FOR THE VIOLATION lis_qq_1 The reason for the issue was personnel crror in that the extent of condition review performed for a previously documented condition adverse to quality was not broad enough to identify a larger issue with the Emergency Operating Procedure / Abnormal Procedure l

l

(* U.S. Nucinar Rrgulatory Commission 3F0498-15 Pags 3 of 13 (EOP/AP)* field validation process. The previous issue associated with AP-770 had been attributed to an unvalidated change to the draft procedure after the field validation had been completed. The corrective action plan for that issue included a comparison of field validations against final versions of APs to identify steps that had changed without validation. The accuracy of the field validations performed on the AP draft procedure change packages was not questioned. As a result, similar problems in AP-470 and AP-770 were not detected.

Issue 2 The reason for Issues 2a and 2b was personnel error in that requirements for the scenario were misinterpreted. An assumption was made that during a design basis accident, a loss of offsite power would occur and the Auxiliary Build >3 filtration system would not be running. Therefore, dose from Reactor Building leakage or Emergency Core Cooling System (ECCS) leakage would not be accumulating on the filters. This assumption was not stated in the calculation and is a non-conservative application of design basis accident inputs and assumptions. The inputs and assumptions to this calculation were not rigorously reviewed or questioned. A contributing factor was lack of FPC expertise in the area of dose assessments.

The reason for Issue 2c was personnel error in that FPC personnel did not question the standard 30 feet per minute for times of travel on stairs. That assumption appeared reasonable to personnel involved with the calculation and the ability of mission personnel to negotiate the equipment and stairs was not questioned. A contributing cause was lack of FPC expertise in the area of dose assessments.

CORRECTIVE STEPS THAT HAVE BEEN OR WILL BE TAKEN AND THE RESULTS ACHIEVED issue 1 I

)

A listing of plant equipment utilized for in-plant actions, by tag number, equipment noun j name, and plant location, has been compiled from a review of each EOP and AP. A field j validation of this information has been performed.

The validated listing has been compared to EOP and AP steps. Discrepancies have been identified. Corrections to AP-470 and AP-770 have been completed. Other EOPs/APs will be revised by June 30,1998, to resolve the remaining discrepancies.

Issue 2 Operability Concerns Report (OCR)98-004, Revision 1, documents a review performed on the doses received by l&C technicians to install hydrogen purge flow indicators and by operators to operate the purge valves. FPC concluded that these evolutions can be performed without any one person receiving more than 5 Rem and that adequate margin exists to compensate for possible corrections in the final dose calculation. In addition, the OCR documented the review of operator actions outside the Control Complex Habitability Envelope directed by EOPs during accident conditions. FPC concluded that either the l

l 1

. U.S. Nucl:ar Regulatory Commission 3F0498-15 Page 4 of 13 operator ' actions' could be performed or that not performing an operator action did not affect the outcome or inhibit mitigation of the accident.

A field validation of Maintenance Procedure MP-815, " Installation of Post Accident Hydrogen Purge Flow Instruments," including the travel times for personnel carrying equipment upstairs to support hydrogen purging operations, will be completed by April 30, 1998.

Dose calculations for hydrogen purging operations will be finalized by June 30,1998.

l 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.

CORRECTIVE STEPS THAT HAVE BEEN OR WILL BE TAKEN TO AVOID FURTHER VIOLATIONS I lssue 1 The individual responsible for the previously documented condition adverse to quality has been counseled on the inadequacy of the extent of condition review performed.

Expectations for the validation of emergency procedures were clarified to assure APs are j validated to the same level of detail as the EOPs. Administrative instruction Al-402C, "AP

~

and EOP Verification and Validation Plan," has been revised to incorporate expectations for the validation of emergency operating and abncrmal procedures.

Issue 2 A program will be developed and implemented by September 30,1998, which will ensure consistency and provide guidance for dose assessment calculations.

DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED FPC's currently issued emergency operating and abnormal procedures are in full compliance.

VIOLATION 50-302/98-02-02 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-225 A, 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

.. U.S. Nuclear Regulatory Commission 1 3F0498-15  !

Page 5 of 13 I informati6n from the vendor manual or direction to plug the flow instrument's power cord into the receptacle.

ADMISSION OR DENIAL OF THE ALLEGED VIOLATION ,

Florida Power Corporation accepts the violation.

REASON FOR THE VIOLATION The reaaan for the violation was personnel error. Personnel performinr. the validation of emergency plan implementing procedure EM-225A, " Post Acciden Reactor Building Hydrogen Control," assumed the actions to be " skill of the craft" anc did not recognize that the directions were not adequate to t.isure proper installation of the flow I instrumentation. 1 1

l

_QORRECTIVE STEPS THAT HAVE BEEN TAKEN AND THE RESULTS ACHIEVED MP-815, " Installation of Post Accident Hydrogen Purge Flow Instruments," was developed and issued to provide direction for installing the post-accident hydrogen pu:ge flow instruments.

A review of the emergency procedures was conducted to assure that actions to be performed by support organizations had adequate procedural guidance. The review  !

identified no additional examples of inadequate procedural guidance for actions directed to support organizations.

CORRECTIVE STEPS THAT WILL BE TAKEN TO AVOID FURTHER VIOLATIONS J i

i Administrative procedures that control the development of the EOPs will be reviewed and revised by May 29,1998, to ensure that EOP actions taken by departments other than Operations have appropriate procedures. These procedures will also receive verification and validation consistent with the EOPs.

pATE WHEN FULL COMPLIANCE WILL BE ACHIEVED FPC's currently issued emergency operating and abnormal procedures are in full

compliance.

l VIOLATION 50-302/98-02-06 10 CFR 50, Appendix B, Criterion 11, Quality Assurance Program, requires that a quality assurance program be established. This prograrn 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 Safety Analysis Report lists ANSI 45.2.11,1974, " Quality Assurance Requirements for the Design of Nuclear Power Plants," under the committed standards.

l

d U.S. Nuctur Regulatory Commission 3F0498-15 Page 6 of 13 ANSI 45'.2.11,'Section 3, Design input Requirements, Subsection 3.2, Requirements, states that the design input shall include but shall not be limited to instrumentation and control requirements including type of instrument, range of measurement, and location of indications .

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 applicable design inputs were correctly translated into procedures. Subsection 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 i 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, Stroke Time for DHV-42/43 for Boron Precipitation, Rev.1, i dated November 1,1997. j ADMISSION OR DENIAL OF THE ALLEGED VIOLATION Florida Power Corporation accepts the violation.

REASON FOR THE VIOLATION lssue 1 The reason for this violation issue was personnel error in that the individual responsible for preparing Revision 1 to calculation 1-90-0023 did not adequately specify the intent of the NOTE describing when 3.5 percent hydrogen concentration can occur. The purpose of calculation I-90-0023 is to determine the instrument loop accuracy of the hydrogen concer'tration loops that provide post-accident monitoring indication / recording, control room annunciation, and input for post-accident operation. The NOTE contained in calculation I-90-0023 was added as operational information only and states that 3.5 percent hydrogen concentration can occur at 28.6 days post-accident. The intent of the i

_ . - . . - - - _- ___-___~

! .". U.S. Nuclear Regulatory Commission

! 3F0498-15 l Page 7 of 13 NOTE is t'o establish the need to commence purge prior to reaching a 3.5 percent hydrogen concentration to protect the flammability limit of 4.1 percent. A 3.5 percent indicated

! hydrogen concentration can occur over a range of days following an accident. However, initiation of hydrogen purge is based on indicated hydrogen concentration, not on a projected number of days post-accident to reach a 3.5 percent hydrogen concentration.

Issue 2 1

The reason for the issue was a process weakness in that the review of a new vendor l technical manual did not result in a comparison of the existing installed instrument j configuration against the vendor recommended installation instructions. I Modification Approval Record (MAR) 91-05-03 01 modified existing hydrogen purge piping d for installation of redundant flow meters (MAR 91-05-03 02). The flow meters were located in the longest straight runs of pipe, with the majority of the straight pipe positioned j upstream of the flow meter. This was standard design practice for best available flow I meter accuracy. Calculation l-90-0013, Revision 0, was revised to incorporats the new l instrumentation and was based on vendor furnished instrument accuracy data. The application data sheet provided by the vendor did not contain flow meter positioning statements with respect to required upstream or downstream arrangements. 1 Subsequent to the modification and Revision 1 to the calculation, the vendor technical l manual for the flow meters was received, reviewed, approved, and incorporated into plant l records. During reviews of the vendor technical manual, the deviation from the vendor i recommended installation instructions for upstream and downstream piping arrangements was not identified.

1 lasue 3 The reason for the issue was personnel error. Due to a miscommunication between the I Inservice Testing Program Manager and the Design Engineer, an open item identified on the calculation review sheet that identified the need to revise the Inservice Testing Program was incorrectly believed to be complete.

CORRECTIVE STEPS THAT HAVE BEEN OR WILL BE TAKEN AND THE RESUI.TS ACHIEVED issue 1 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 16 days post-accident.

l Calculation I-90-0023, Revision 1, will be revised by June 15,1998, to either correct the l identified value and reference or delete the associated NOTE.

l

2. U.S. Nuctstr Rngulatory Commission 3F0498-15 Page 8 of 13 lasue 2 An evaluation was performed to determine the affect on instrument uncertainty for the deviations from the vendor recommended installation instructions regarding upstream and downstream piping arrangements for the hydrogen purge flow meters. The evaluation a concluded that the existing installation configuration has acceptable instrument accuracy, Calculation I-90-0013 will be revised by September 30, 1998, to address instrument uncertainty as it relates to deviating from the vendor recommended installation instructions concerning upstream and downstream piping arrangements.

A review was performed of installed flow elements utilized to monitor improved Technical Specification (ITS) allowable values and to monitor safety-significant parameters used for accident mitigation. Instances were identified where vendor recommended installation requirements relative to lengths of straight pipe upstream and downstream of the device were not met. For those cases, FPC concluded that either adequate compensating l correction factors had been incorporated into the associated loop uncertainty calculation or that omented uncertainty assessments had been performed to ensure conservative operator actions are taken.

)

Issue 3 Recently performed stroke time testing for valves DHV-42/43 was reviewed. The stroke times were within the new acceptance criteria required by the assumptions in calculation M-97-0120.

The Engineering Data Sheets for Decay Heat System valves DHV-42/43 have been revised to set the open stroke time limits to those values specified in calculation M-97-0120.

Inservice Testing Program Self Assessment STSANEP97-26, issued October 15, 1997, verified information for components in the Inservice Testing Program. Other valves tested  ;

by the Inservice Testing Program have stroke time acceptance criteria within the limits  !

established by the ASME Code. Those cases had been successfully incorporated into the l inservice Testing Program acceptance criteria. This assessment occurred prior to issuance 1 of calculation M-97-0120. l CORRECTIVE STEPS THAT HAVE BEEN TAKEN TO AVOID FURTHER VIOLATIONS Issue 1 The individual responsible for preparing calculation I-90-0023, Revision 1, is no longer at C R-3. An engineering required reading interoffice Correspondence (IOC) discussing this violation issue has been sent to appropriate design engineering personnel. This IOC stressed the importance of ensuring information provided in calculations is conciselv stated so that a person reading the calculation does not misinterpret what is stated.

I

' l I

. n

.. U.S. Nuctsar Regulatory Commission 3F0498-15

'Page 9 of 13 l

! l#RtL2 '

1

Vendor technical information is reviewed in accordance with Al 404A, " Review of l Technical Information." Al-404A, Revision 5, became effective on June 30,1997. The l

primary reason for this revision was to add requirements to ensure that design basis issues are identified and resolved An example of these changes is an enhancement made to the responsibilities of Nuclear Plant Technical Services (NPTS). That enhancement states that l new or different design criteria that go beyond the existing plant design basis must be identified and resolved. A new vendor recommendation for the physicallocation of a flow l instrument in a line would now represent a condition requiring resolution.

l lasue 3 l An engineering required reading IOC discussing this violation issue has been sent to the l Inservice Testing Program Supervisor, Design Engineering Manager, and appropriate engineering personnel. This IOC stressed the importance of concise communications l between the affected organizations and included a copy of the closed Precursor Card (PC).

l The PC included the apparent cause evaluation for the issue, corrective actions and actions to prevent recurrence.

l l DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED l FPC has achieved full compliance.

VIOLATION 50 302/98-02-09 10 CFR 50, Appendix B, Criterion XI, Test Control, requires in part that a test program be l established to assure that all testing required to demonstrate that structures, systems, and l

components will perform satisfactorily in service which incorporate the requirements and acceptance limits contained in applicable design documents. The test program shall '

include operational tests.

(

l Updated Final Safety Analysis Report, Section 14.2.2.5.4, Emergency Core Cooling l 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.

l The criteria contained in Part 50.46 are applicable to all sizes of LOCAs and are necessary in order to verify adherenca. 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 t'1e 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 part that several alternate modes of operation of the ECC systems can be used during long-term cooling, if necessary, while maintenance is being performed on normal equipment and one

U.S. Nuclear R::gulatory Commission 3F049815 3 Page 10 of 13 )

of these inodes'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 )

I would be pumping through its HPllines."

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.

ADMISSION OR DENIAL OF THE ALLEGED VIOL.ATIO.N Florida Power Corporation accepts the violation, figASON FOR THE VIOLATION The reason for the violation was a process weakness. The original ECCS design analysis for CR-3 focused on Large Break Loss Of Coolant Accident (LBLOCA) scenarios. Small Break LOCA (SBLOCA) scenarios that require High Pressure injection (HP!) 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) incident, and new post-TMl requirements for long term cooling capabilities changed the intended use of HPI 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 longer post accident mission time. As these changes were made, a documented review of the original system operational requirements 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 A review of HPl in the piggyback method of operation showed that credit had been taken for some ECCS components within the plant operating in a manner not consistent with the type or degree of service normally intended for such components. These components are J the HPI pumps, Decay Heat System valves DHV 5/6, and Makeup System valves MUV- )

2/6/10. i Although not procured to operate for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a post-accident environment, )

an evaluation concluded that the HPI pumps are capable of long term operation in the l piggyback mode for the expected mission time. An additionalissue was identified with the l ability of the HPl pumps to operate with entrained solids during long-term post accident l 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 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 j mode.

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 threat to performance of the Low Pressure injection (LPI) or HPI safety function.

l l

l l

l l

l

.I U.S. Nuclear Regulatory Commission 3F0498-15 Page 11 of 13 An evaldation '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 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 issue 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.

While the above reviews justify operability of ECCS components, no specific testing has been performed to verify the capability to support long-term piggyback operation. By October 16,1998, a test plan will be developed to demonstrate the ability of valves MUV-2/6/10, MUV-23/24/25/26, and DHV 5/6 to support long-term piggyback operation. That lesi pian will be implemented prior to restart from the next scheduled refueling outage.  ;

CORRECTIVE STEPS THAT WILL BE TAKEN TO AVOID FURTHER VIOLATIONS The subject violation occurred in the late 1970s and early 1980s. Since that time, changes have been made to Nuclear Engineering Procedures (NEPs) and Administrative Instructions to ensure calculations / analyses or procedure changes consider the potential impact of operating conditions on equipment specifications. For example:

NEP-213, " Design Analyses / Calculations," 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 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 identified in the calculation.

NEP-210 states that attributes of the actual item / service purchased under a i specification may have more detail or include exceptions to the specification which l FPC has accepted. in this case, design input must be either revised in the l specification or clarified in the design input. The Design input Record must contain j justification for addition, upgrade or deletion of essential equipment. Additionally, N EP-210 addresses applicable design requirements related to component l specifications.

Al-400C, "New Procedures and Procedure Change Processes," documents the technical review performed by Nuclear Operations Engineering. The purpose of the review is to determine if the plant design basis could be affected by the procedure revision.

I I

l

.'- U.S. Nuclaar Regulatory Commission 3F049815 Page 12 of 13 l

1 Al-400F, "New Procedures and Procedure Change Processes For EOPs, APs, and  !

Supporting Documents," requires documented technical reviews to be performed by l Nuclear Operations Engineering and Nuclear Plant Technical Services. One consideration is the determination of whether the procedure revision makes any l permanent changes to the plant configuration which are beyond normal and accepted use of the equipment.

DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED FPC will achieve full compliance prior to restart from the next scheduled refueling outage.

l l

l

IJ l i

i

.e

.- U.S. Nuclear R*gulatory Commission 3F049815 Page 13 of 13 l ATTACHMENT 2 l LIST OF COMMITMENTS I l

l The following table identifies those actions committed to by FPC in this document.  ;

{Section? Commitment) '

s LDue Date; Page 3 EOPs/APs will be revised to resolve the remaining June 30,1998 discrepancies.

Page 4 A field validation of MP-815, including the travel times for April 30,1998 personnel carrying equipment upstairs to support hydrogen '

purging operations, will be completed.

Page 4 Dose calculations for hydrogen purging operations will be June 30,1998 3 finalized.

Page 4 The remainder of the EOP actions will be evaluated as part of Prior to restart from the the re-baselining effort of Environmental Qualification and next refueling outage, 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.

Page 4 Al-402C has been revised to incorporate expectations for the Complete -

validation of emergency procedures.

Page 4 A program will be developed and implemented which will September 30,1998 ensure consistency and provide guidance for dose .

assessment calculations.

Page 5 Maintenance Procedure MP-815 was developed and issued to Complete provide direction for installing the post-accident hydrogen purge flow instruments.

Page 5 Administrative procedures that control the development of May 29,1998 the EOPs will be reviewed and revised by May 29,1998, to ensure that EOP actions taken by departments other than Operations have appropriate procedures. These procedures will also receive verification and validation consistent with the EOPs.

Page 7 Calculation I 90-0023, Revision 1, will be revised to either June 15,1998 correct the identified value and reference or delete the associated NOTE.

Page 8 Calculation 1-90 0013 will be revised to address instrument September 30,1998 uncertainty as it relates to deviating from the vendor recommended installation instructions concerning upstream and downstream piping arrangements.

Page 11 A test plan will be developed to demonstrate the ability of October 16,1998 valves MUV 2/6/10, MUV-23/24/25/26, and DHV-5/6 to support long term piggyback operation.

Page 11 That test plan [ demonstrate the ability of MUV 2/6/10, MUV- Prior to restart from the 23/24/25/26, and DHV 5/6 to support long term pigtyback next scheduled operation) will be implemented prior to restart from tha next refueling outage I scheduled refueling outage.

I