3F0686-16, Revises Response to Violations Noted in Insp Rept 50-302/86-12.Corrective Actions:Cap Studs for Check Valve RWV-117 Lubricated & Torqued & Administrative Instruction AI-401 Revised to Redefine Housekeeping Changes

Revises Response to Violations Noted in Insp Rept 50-302/86-12.Corrective Actions:Cap Studs for Check Valve RWV-117 Lubricated & Torqued & Administrative Instruction AI-401 Revised to Redefine Housekeeping Changes

ML20206R421
Person / Time
Site:	Crystal River
Issue date:	06/19/1986
From:	Widell R FLORIDA POWER CORP.
To:	Grace J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
References
3F0686-16, 3F686-16, NUDOCS 8607070140
Download: ML20206R421 (8)

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t. .- s. n: as: o Power C O R P O R AT 6 O N June 19, 1986 3F0686-16 Dr. J. Nelson Grace Regional Administrator, Region II U.S. Nuclear Regulatory Commission 101 Marietta Street N.W., Suite 2900 Atlanta, GA 30323

Subject:

Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 IE Inspection Report No. 86-12 Revised Response

Dear Sir:

Florida Power Corporation provides the attached as our revised response to the subject inspection report.

Sincerely, Rolf Manag)C.Widell er, Nuclear Operations Licensing and Fuel Management AEF/feb Attachment 8607070140 860619 I l ADOCK 05000302 l PDR G PDR , (,

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t GEN ERAL OFFICE 3201 Thirty fourth Street South.P.O. Box 14o42, St. Petersburg, Florida 33733 813-866-5151 ff,!

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. FLORIDA POWER CORPORATION

RESPONSE

INSPECTION REPORT 86-12 VIOLATION 86-12-02 Technical Specification (TS) 6.8.1 requires that written procedures be imple-mented to cover those activities reconinended in Appendix A of Regulatory Guide 1.33, November 1972, and to implement 'the Offsite Dose Calculation Manual (0DCM).

Regulatory Guide 1.33, Appendix A, recommends maintenance procedures and/or work instructions .for the perfomance _of maintenance.

Step 7.2.9 of Maintenance Procedure MP-149, Check Val ve Cap Removal and Reinstallation, requires that check valve cap studs be lubricated prior to initial torquing.

Step 5.2 of Work Request No. 075920 requires that Emergency Feedwater Pump testing be stopped if vibration readings exceed the specified alert limits.

The ODCM Representative Sampling Method No. 3.1-5 requires the reactor building (RB) purge make-up (supply) fans to be shut down whenever both the RB personnel and equipment hatches are open.

Contrary to the above:

a. On March 14, 1986, procedure MP-149, Check Valve Cap Removal and Reinstallation, was not implemented in that check valve cap studs for check valve RWV-117 were not lubricated prior to initial installation of the check valve cap and torquing of the stud nuts,

b. During the period of March 10-11, 1986, Work Request No. 075920, which directed post-maintenance testing of the turbine driven emergency feedwater pump (EFP), was not implemented in that the testing was not stopped when EFP vibration levels reached and exceeded the alert ranges specified in the instructions.

c. On March 17, 1986, during the period from approximately 11:00 a.m. to 11:45 a.m., the RB purge make-up fans were not shut down while both the RB personnel and equipment hatches were open.

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

RESPONSE

1. Florida Power Corporation's Position Florida Power Corporation (FPC) concurs with the above stated vmiation,

2. Appardnt'Cause of Violaticn

a. The check valve cap studs for check valve RWV-117 were not lubricated prior to initial installation of the check valve cap and torquing of the stud nuts due to lack of understanding of the procedural requirements by the craftsmen.

b. The post-maintenance testing of the turbine driven emergency feedwater pump _ (EFP) was not stopped when EFP . vibration levels reached and exceeded the alert ranges specified in the instructions because it was

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not clearly understood by the s test engineer in charge that high vibration readings were, being recorded. The-test engineer was concen-trating on temperature readings, and failed to notice the vibration readings. The alert range was contained in the Limits and Precautions section of the work instructions, and not on the data sheets. This 1 added to the high vibration reading not being a consideration for terminating the test.

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c. The RB purge make-up fans were not shut down while both the RB person-i nel and equipment hatches were open because Maintenance Procedure MP-114 Removal and Reinstallation of Equipment Hatch, did not instruct maintenance personnel to notify the Operations Shift Super-visor that the reactor building equipment hatch was to be temporarily removed.

3. Corrective Actions ,

a. Personnel involved were counselled.

The cap studs for check valve RWV-117 were lubricated and torqued to ensure adequate gasket compression.

b. Following the pump run, the vibration data was reviewed by the engineering staff. The vibration readings were not considered significant enough to have caused pump damage.

c. Operations personnel secured the RB purge make-up fans upon discovery of the condition.

< 4. Action Taken to Prevent Recurrence i

a. A formal training lesson pl an was developed on " Check Valve Cap Removal and Reinstallation (Gasket Style)." This has been included in

, the contractor training nrogram and was initially given to appropriate supervisory and craft personnel,

b. A Management Review Board meeting was held to investigate the event and establish corrective actions to prevent recurrence. Result:: of this meeting will be reviewed by all. plant engineering personnel - by August 1, 1986. Work instruction testing requirements will be reviewed more carefully prior to the test to eliminate over-restric-tive acceptance criteria. In addition, the acceptance criteria will be included on the data sheets as well as the Limits and Precautions section of the test.

c. MP-114 has bien revisid to r: quire that the Nuclear Shift Supsrvisor and Outage Shift Manag:r be nntified prior to removal of tha equipm:nt hatch. A sign has also been installed on the RB equipment hatch to notify the Outage Shift Manager and Shift Supervisor prior to removal and installation.

5. Date of Full Compliance

a. Full compliance was achieved on March 14, 1986,

b. Full compliance was achieved on March 11, 1986,

c. Full compliance was achieved on March 17, 1986.

VIOLATION 86-12-06

, Technical Specification 6.8.2.b requires that certain procedural changes receive intradepartmental and interdisciplinary reviews by qualified reviewers prior to approval by the responsible superintendent or manager. The Plant Review Committee (PRC) shall then review the 10 CFR 50.59 evaluation within 14 days of approval.

Technical Specification 6.8.1.a requires that -written procedures be estah-lished to cover those activities recommended in Regulatory Guide 1.33, November 1972.

Regulatory Guide 1.33, Section A, recommends that administrative procedures be established to provide the method for changing procedures.

Contrary to the above, procedure AI-401, Origination of and Revisions to P0QAM Procedures, which implements TS 6.8.2.b., was found to be inadequate in that it allows Housekeeping Changes (HKCs) to be made to procedures and implemented following approval by the responsible superintendent or manager. These HKCs do not receive the required review by appropriate intradepartmental and inter-disciplinary qualified reviewers nor are the associated safety evaluations reviewed within 14 days by the PRC.

As a result, several curves in the OP-103 series, Plant Curve Book, have been changed using HKCs. These include curves for radiation monitors for the Reactor Building air and Auxiliary Building liquid release paths and the j reactor plant heatup/cooldown curves.

Additionally, on March 20, 1986, a HKC was used to add a quality control (QC) hold point to a maintenance procedure (MP-174) without receiving the reviews required by TS 6.8.2.b.

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

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RESPONSE

1. Florida Power Corporation's Position Florida Power Corporation agrees that housekeeping changes (HKCs) should not have been used to include curves for radiation monitors for the reactor building air and auxiliary building liquid release paths and the reactor plant heatup/cooldown curves. FPC takes exception to the example that a HKC was incorrectly used to add a QC _ hold point to MP-174.

Relative to the observation during the performance of MP-174 on March 20, 1986, no new QC inspection / verification requirements were added. The signature blanks added to steps 7.2.19 and 7.2.21 were redundant to information already included in Enclosure 6, which is ' referenced by steps 7.2.19 and 7.2.21. These signature blanks were added for clarity and consistency with other procedures to remind the person performing the procedure to notify QC prior to proceeding. The addition of the signature blank and words " Hold Point" to step 7.2.31 was a poor choice of words, but was not an unapproved change to MP-174 since that step referred the user ta Enclosure 3 which already contained QC inspector verification signature bl ank . Again, the signatures were added to the applicable procedure step to ensure the requirement to notify QC was not missed.

These housekeeping changes were made in accordance with the instructions in AI-401, and in FPC's opinion did not circumvent the procedure review and approval process required by TS 6.8.2.

2. Apparent Cause of Violation The cause of the curves in the Plant Curve Book being changed through HKCs was a difference of interpretation of T.S. 6.8.2, as it relates to the use of housekeeping changes.

The housekeeping changes referenced in the NRC Inspection Report 86-12-06 item 2, were made in accordance with the instructions in AI-401. It was

the intent of AI-401 to allow corrections to typographical errors, spelling, etc., and updates to the plant curve book (0P-103) without the l need for the review described in Technical Specification 6.8.2. The justification was that updating a curve in the plant curve book did not

constitute a procedure change since these curves are referenced from a procedure that describes how to perform quality aTtivities. The

, procedures that reference the curves in OP-103 have been reviewed and approved in accordance with Technical Specification 6.8.2. AI-401, was insufficiently restrictive to meet the intent of T.S. 6.8.1.a.

3. _

Corrective Actions Records Management was immediately notified to hold all housekeeping changes until a change to AI-401 was issued.

Administrative Instruction AI-401 was revised to redefine housekeeping changes as follows: i s

- I Houseke ping Change Chages to format of pages, paragraphs, or enclosures which do not change t:w information, meaning, or intent of the original document. Corrections of typographical errors, spelling, numbering of pages or paragraphs, change of administrative assignment or responsibility for P00AM procedures as specified in Section 4.6 of AI-400.

NOTE: A housekeeping procedure change cannot be submitted where a change of intent is necessary.

The revision of AI-401 was issued on 4/11/86.

4. Action Taken To Prevent Recurrence The above corrective actions are sufficient to prevent recurrence.

5. Date of Full Compliance Full compliance was achieved on March 21, 1986, when Records Management was notified to hold all housekeeping changes.

VIOLATION 86-12-04 10 CFR 50.73 (a)(2)(v)(B) requires the submittal of a Licensee Event Report (LER) within 30 days of the event for any condition that could result in a loss of the plant's ability to remove residual reactor heat.

Contrary to the above, on December 19, 1985, it was determined that a design error could cause a loss of residual reactor heat removal ability and no LER was issued.

This is a Severity Level V violation (Supplement I).

RESPONSE

FPC disagrees with this violation. This design problem was identified and evaluated by FPC personnel according to established procedures and determined not to be reportable under 10 CFR 50.73(a)(2)(v)(B). That part of the rule requires licensee's to report, "(v)-(a)ny event or condition that alone could have prevented the fulfillment of the safety function of structures or systems

, that are needed to: .. . (B) (r)emove residual heat;". The design problem (failure to protect the bearing flush water supply to the Nuclear Services seawater pumps and the decay heat seawater pumps from fire damage) would not alone have prevented the function of systems required to remove residual heat. It should be noted that upon discovery of the misunderstanding of the vendor information FPC performed an analysis, as suggested by Generic Letter 85-01 (superceded by Generic Letter 86-10) and voluntarily submitted this information to the appropriate NRC office for review and concurrence. Thus, FPC notified the NRC even though we chose not to report it via a particular means (an LER). Our normal practice to carbon copy the NRC Regional Office was not followed due to administrative oversight. However, the NRC Senior Resident Inspector was, in accordance with normal practices, copied via FPC's internal distribution.

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If fire damaged the systems providing tha b:aring flush water supply to the plant during normal power op: ration, tha turbina driven em:rg:ncy feedwater pump would have remained available to remove residual heat. To continue to cool down, or for fire damage occurring at cold shutdown, one train of decay heat removal equipment including one of the decay heat seawater pumps will remain operable. If fire damages all sources of bearing flush water (redun-dant A and B domestic water pumps plus Nuclear Service closed cycle cooling water supplied through redundant motor operated valves) only one operating decay heat seawater pump would be oamaged. The non-operating decay heat sea-water pump would not be affected. The non-operating pump would only auto-start on an ES signal and therefore would remain unaffected, since other plant accidents need not be considered concurrent with a fire. (See Branch Techni-cal Position APCSB 9.5-1 Appendix A, Item A.4).

The loss of bearing flush water supplies could disable both Nuclear Service seawater systems and eliminate the Nuclear Services closed cycle cooling water heat sink. Eowever, the Nuclear Services closed cycle cooling pumps would remain operable for the purpose of supplying bearing flush water for the decay heat seawater pumps through either of the two redundant motor operated val ves. Manual action may be required to open one of the motor operated flush water supply valves depending on specific fire damage. Sufficient time for manual action would be available before the decay heat system was required, and access to the area following a fire would not be a problem since the dura-tion of an unsuppressed fire from the combustion of all ' materials in the area has been conservatively evaluated to be approximately 12 minutes. This fire duration estimate takes no credit for sprinklers or fire barriers around cable trays that now exist in this area.

The Inspection Report fails to note that this entire area would not meet literal Appendix R requirements and thus an Technical Exemption for this fire area was sought and granted. The primary technical basis for this exemption is detection, suppression, amount of combustible materials, and room geometry (high ceiling with hot gas traps inherently present). These continued to exist and supported a reduced level of concern with regard to the identified error. FPC continues to maintain that this design error did not create a reportable condition under 10 CFR 50.73(a)(2)(v)(B).

DEVIATION 86-12-05 In a letter to the Nuclear Regul atory Commission from Florida Power Corporation dated December 19, 1985, the licensee identified a design error that could result in a loss of the Nuclear Services Seawater and the Decay Heat Seawater pumps in the event of a fire. In the corrective actions for this design error, the licensee commhted to continue 20-minute roving fire watches until such time that plant modifications to prevent this loss of pumps was completed.

Contrary to the above, at 2:00 p.m. , on March 14, 1986, the roving fire watch was secured even though the plant modification for these pumps was not completed.

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RESPONSE

FPC disagrees with this deviation. FPC maintained a level of protection for the functions at risk that was appropriate to the hazard that existed. This protection included periodic surveillance by roving security officers and plant operators. Their normal duties include observation of abnormal conditions, including fire, and response to those conditions. The FPC letter of December 19, 1985, referenced in the Notice of Deviation, was a fire protection evaluation forwarded under the guidance of draft Generic Letter 85-01 as reaffirmed by Generic Letter 86-10. FPC chose this approach in a proactive effort to keep the NRC staff aware of our fire protection activities. The alternate approach offered by draft Generic Letter 85-01 was simply retaining this evaluation in our files available for review. Since the second alternative existed FPC felt justified in adjusting our compensation to the hazard at the time.

FPC's letter noted the 20-minute roving fire watch was in place at that time.

In addition, it stated " Roving fire watches will continue in the areas containing components necessary for maintenance of the current bearing flush water supply until modifications are complete." This commitment does not state the 20-minute frequency would be maintained without regard to plant status.

At the time the evaluation was written, Crystal River Unit-3 (CR-3) was in power operations and 20-minute roving fire watches were being used as compen-sation in the auxiliary building for uncompleted fire barrier construction around cable trays and conduits. At the time of the alleged deviation CR-3 had been shutdown for approximately two and one half months and all cable and conduit fire barrier construction had been completed. FPC believes the fire protection provided in the area, which included operable automatic sprinklers, operable automatic fire detection, and roving patrols by security and operations personnel, was adequate for the hazard.

Prior to discontinuing the 20-minute roving fire watches which had been estab-lished as part of Appendix R Schedular Exemption, FPC Licensing contacted both the NRR Fire Protection Reviewer and the NRC Senior Resident Inspector for CR-3. FPC contacted the NRR Fire Protection Reviewer regarding the Appendix R requi.rements when the plant is in cold shutdown. Based on discussion regarding the fact that Appendix R requirements are not clearly defined when the plant is in cold shutdown, FPC was justified in modifying the protection to more closely match the hazard. On March 13, 1986, one day prior to discontinuing the fire watches, FPC Licensing informed the Senior Resident Inspector for CR-3 that the 20-minute roving fire watches were being discontinued. FPC informed the Senior Resident this was because the schedular exemption modifications were complete and because we were not aware of any specific Appendix R guidelines applicable to cold shutdown. The Senior Resident was also informed the seawater pumps bearing flush water modification was not complete. There was no discussion with the Senior Resident Inspector specifically addressing the commitments in FPC's December 19 letter, but there was no objection voiced to the discontinuance of the 20-minute roving fire l watches.

It is FPC's position that sound fire protection was maintained for the seawater pumps, that our actions were taken after careful consideration of the guidance that exists, and after consultation with NRC staff. Based on the foregoing, FPC requests this Deviation be withdrawn.

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