ML041040208

Assessment of the Licensee Response to Concerns Raised to the NRC Regarding Activities at FitzPatrick Nuclear Power Plant

ML041040208
Person / Time
Site:	FitzPatrick
Issue date:	03/31/2004
From:	Carrasco J Division of Reactor Safety I
To:
References
FOIA/PA-2004-0107
Download: ML041040208 (15)

Text

JOSEPH E. CARRASCO, ENGINEERING INSPECTOR ASSESSMENT OF THE LICENSEE RESPONSE TO CONCERNS RAISED TO THE NRC REGARDING ACTIVITIES AT FITZPATRICK NUCLEAR POWER PLANT Concerns Resolution/Corrective Action _ NRC proposed response to concerns:

1.-"Management has not pursued concerns with need to upgrade the -The technical evaluation performed by Design The licensee initiated qualification of several vital and protected area doors, specifically the new Engineering on the safety classification issue AQCR Adverse Quality control room door.- No. 92.360 were reviewed and determined to be Condition Report CAQCR) adequate with the exception of a failure to consider 92-0360 to address the one design issue of one cable tunnel door. New action vital and protected area items have been developed to address these issues. door [V&PAD) issues. It appears that the licensee has completed the resolution of AQCR No.92-360 with the following exceptions: a) classifying the Electrical Bay doors and b) omission from the evaluation of one cable tunnel door. The licensee has developed new action items to address these

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/t~/e~'.concluded~ g Thethat NiRCthe Am w , censee actions to Zig r \address this concern ww . . )Cto know about h licensee plans to close AQCR 92-360.

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JOSEPH E. CARRASCO, ENGINEERING INSPECTOR ASSESSMENT OF THE LICENSEE RESPONSE TO CONCERNS RAISED TO THE NRC REGARDING ACTIVITIES AT FITZPATRICK NUCLEAR POWER PLANT I T 2.- 'Management was not responsive to concerns raised with the hot water -Plant drawings were reviewed to establish the The licensee reviewed boiler modification to that NYPA failed to qualify the existing 170,000 minimum separation between the control room and compared the gallons if 2 fuel oil tank to NFPA 30 standards, but used a loophole allowing (including Its air intakes) and the fuel oil storage tank applicability of NFPA 31 qualification to NFPA 31. This is a concern due to the proximity of the tank (AQCR 92-360). These distances are fully In and NFPA 30 standards to the control room and its air intakes." compliance with NFPA 31 and the fuel oil storage tank used to determine the does not poses a threat to the control room or its air minimum separation intakes. between the control room and the fuel storage-tank.

The licensee concluded that standard NFPA 31 is the correct standard for this application. Further, the licensee found no IA r .- 2 6 safety concerns In the regarding the uses of standard NFPA 31 to qualify the existing 170,000 gallon capacity tank. The distances between the control room and the fuel oil storage tank are above the minimum separation distances prescribed In NFPA 31 Standard. The licensee's actIon to address this concern was adequate. Therefore, no NRC follow-up Is needed.

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JOSEPH E. CARRASCO, ENGINEERING INSPECTOR ASSESSMENT OF THE LICENSEE RESPONSE TO CONCERNS RAISED TO THE NRC REGARDING ACTIVITIES AT FITZPATRICK NUCLEAR POWER PLANT f

3.-"ACTS item 8977 (involving a DER written in 1993-1994) regarding a -There was no design or licensing basis for the The licensee self-building/structures configuration has not been approved. There is a letter of evaluation of a System 52' component since system assessed their response commitment to the NRC in this matter (i.e., GL 83-28 response). 52 was not included in the MCM-6A (System Safety to the NRC Generic Letter Function Sheet). -System 52 was added to MCM- 83-28. This self-6A and components presently existing in system 52 assessment showed that were reviewed for appropriate designation. they made no

-.-.- A review of the system 52 Work Requests in a commitments to the maintenance database did not find safety significant NRC. However, the concerns. licensee noted that there was no design or licensing basis to evaluate System 52 and Its associated components. They attributed this omission to the fact that they did not include System 52 in the MCM-6A (System Safety Function Sheet).

Subsequently, the licensee reviewed the components presently existing within the scope of System 52 to ensure their appropriate safety designation. In addition, a review of System 52 Maintenance Work Requests IMWRs) did not find safety significant concerns. The licensee's actions to address these concerns were adequate.

Therefore, no NRC follow-up Is needed.

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JOSEPH E. CARRASCO, ENGINEERING INSPECTOR ASSESSMENT OF THE LICENSEE RESPONSE TO CONCERNS RAISED TO THE NRC REGARDING ACTIVITIES AT FITZPATRICK NUCLEAR POWER PLANT 6-

/2 1Wle Concerns Resolution/ Corrective Action NRC proposed response to concerns:

4.- 'The corrective actions to DERs94-111

• 97-045 & 95-997 have not yet -DER 94-111: The FA and FB, standard Although the licensee started some been completed." stick drawings do not show the present corrective actions prescribed in DERs plant configuration.-licensee's94-111, 97-045, and 95-997, we assessment Is unclear on the status of may review the licensee actions listed DER 94-111. in DERs94-111, 97-045 and 95-997 v X DER 97-045:Type A drawings have not to ensure the proper and timely been updated. - 2 out of 3 corrective closing of these DERs. The licensee's actions of DER 97-045 are closed, but the actions to address these concerns 3th remain open. were adequate. Therefore, no NRC

-DER 95-0997: The single boiler may not follow-up Is needed. f meet the single failure criteria of FSAR

< 1section 8.7.2.3. - Corrective action for

,(t 1the DER proposed installation of a redundant boiler was completed.

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JOSEPH E. CARRASCO, ENGINEERING INSPECTOR ASSESSMENT OF THE LICENSEE RESPONSE TO CONCERNS RAISED TO THE NRC REGARDING ACTIVITIES AT FITZPATRICK NUCLEAR POWER PLANT 5.- 'The response to DERs and/or corrective actions is given back to the -Frequently DERs written within Regarding concerns about DERs that concernee (and others) to handle in addition to regular duties." engineering disciplines will be returned to were given back to the originator, the the writer for evaluation since this NRC assessed the licensee distribution Individual is the most knowledgeable of DERs for disposition, and it appears person available to perform this task. that frequently the licensee supervisors return DERs that are written within engineering disciplines to the DER originator for evaluation.

This Is done because the DER originator Is the most knowledgeable person available to disposition DERs; this approach appears logical to the NRC. The NRC concluded that the distribution Is an internal responsibility of the licensee, and unless safety Is compromised by starting a poor work distribution, then the NRC may Intervene to ensure that the Ncensee work distribution does not create a negative Impact on the overafl safety of the plant. Therefore, this concern Is closed.

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JOSEPH E. CARRASCO, ENGINEERING INSPECTOR ASSESSMENT OF THE LICENSEE RESPONSE TO CONCERNS RAISED TO THE NRC REGARDING ACTIVITIES AT FITZPATRICK NUCLEAR POWER PLANT r

6.-'Concernee's supervisors discouraged the writing of DERs. -Engineering supervisors, and managers Overall in this concern, it appears that have been instructed to not discourage the licensee has instructed the use of DERs to report problems. engineering supervisor, and managers not to discourage the use of DERs to report problems. Therefore, the alleger's concern is not substantiated because there is no specific indication that employees, including the alleger, are systematically discouraged from using the DER process.Therefore, the alleger's concern Is not substantiated because there Is no specific evidence that employees, Including the alleger, are systematically discouraged from using the DER process. Here, no NRC follow-up Is required.

I ________________________ I. ______________________

Concerns Resolution/ Corrective Action NRC proposed response to concerns:

7.- Concerns expressed regarding NYPA's follow-on actions committed to - Through interviews with personnel Through licensee's interviews with their 10CFR 50.54(f) reply.' assigned to the preparation of the personnel assigned to the preparation 10CFR50.54(f) response. No adverse of the 10CFR50.54(f) response, it data of safety related the noncompliance appears that no adverse data of safety was found. related nature was found. In

, conclusion, In routine Inspections at FitzPatrick the NRC will further assess the licensee ability to maintain the configuration of the plant current.

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JOSEPH E. CARRASCO, ENGINEERING INSPECTOR ASSESSMENT OF THE LICENSEE RESPONSE TO CONCERNS RAISED TO THE NRC REGARDING ACTIVITIES AT FITZPATRICK NUCLEAR POWER PLANT T r 8.- "Concerns expressed with an instrument air system moisture sensor and - The plant Instrument Air System meets The alleger's concern expressed with NYPA's response to GL 88-14 involving instrument air systems." design requirements for moisture, regard to instrument air system Instrurment air dewpoint is measured on a moisture sensors is not substantiated.

quarterly basis per procedure No. RT- Since at FitzPatrick, the plant 01.01. -This meets the plant's Instrument Air System meets design commitment relative to NRC GL 88-14 for requirements for moisture, the moisture in the Instrument Air System. licensee measures instrument air dewpoint quarterly per procedure No.

RT-01.01. These licensee actions meet the plant's commitment about AN S-9 as Itoo cc-Ce NRC GL 88-14 for moisture in the Instrument Air System. These licensee actions meet the plant's commitment about NRC GL 88-14 for moisture In the Instrument Air System.

Therefore, this concern Is closed, and no NRC follow-up Is required.

9.-"NYPA purportedly knew in the 1989 time frame that snubbers were past - An evaluation of snubber elastomeric On November of 1989, the licensee theirs rebuild date and took no immediate action." seals service life was performed. This performed an evaluation of snubber evaluation determined longer life span of elastomeric seals. This evaluation the snubbers. determined a longer life span of the

,§g 7_- fefsv snubbers. Recently, the licensee Quality Assurance (OA) reviewed the te e snubber program as presently carried

'1 out and concluded that the snubber technical documentation was accurate. The licensee's actions to address these concerns were adequate. No NRC follow-up Is needed.

____________________________________ I ____________________ I ___________________

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JOSEPH E. CARRASCO, ENGINEERING INSPECTOR ASSESSMENT OF THE LICENSEE RESPONSE TO CONCERNS RAISED TO THE NRC REGARDING ACTIVITIES AT FITZPATRICK NUCLEAR POWER PLANT Concerns Resolutionl Corrective Action NRC proposed response to concerns:

10.- Concerns were expressed with the Speakout Program, specifically: 1) - During the 1993-1994 period there The Speakout program Is Independent no action was taken with a list of concerns brought to speakout in a 1993- were a Senior Speakout investigator and of the site or Nuclear Generation 1994 time frame by OA inspector, and 2) since the speakout representative two contractor Investigators. According management. The purpose of the communicates directly with senior management, employees are discouraged to information available in Speakout files, program is to provide an outlet for from raising issues." the concernee did not clearly annunciate employees and contractors to express his intentions at the time he came to nuclear safety concerns. An Speakout. evaluation of concern No. 10 showed that during the 1993-1994 period

-The Speakout program Is independent there was a Senior Speakout of the site or Nuclear Generation Investigator and two contractor management. The purpose of the program investigators. According to is to provide an outlet for employees and information available in Speakout files, contractors to express nuclear safety the alleger did not clearly annunciate concerns. his intentions at the time the alleger came to Speakout. The NRC closed this concern due to a lack of specifics.

Therefore, no NRC follow-up Is needed.

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JOSEPH E. CARRASCO, ENGINEERING INSPECTOR ASSESSMENT OF THE LICENSEE RESPONSE TO CONCERNS RAISED TO THE NRC REGARDING ACTIVITIES AT FITZPATRICK NUCLEAR POWER PLANT t 7 11.- "(Related to Concern # 6) Examples provided were DERs were turned (Note: There is no text in this cell) To decide if the alleger was singled back to the writer, purportedly as 'punishment': ACTs 25549, ACTS out for 'punishment' by having DER 22356. ACTS 8997. DER 95-0997 and an issue with the auxiliary boiler responses assigned to him (the room oil water separator being radiologically contaminated." alleger) by his supervisor, the licensee started an Investigation that shows that plant supervision did not treat him unfairly with respect to the assignment of DERs. About the issue of contamination of the boiler rooms oil water separator, the licensee is aware that these rooms and the components inside the rooms have the potential to be radiologically contaminated. Therefore, the licensee has radiation contamination controls for these rooms and the components in these rooms through the use of procedure SP-01.1 1, 'Unmonitored Paths Sampling and Analysis." The NRC concluded that the alleger was not singled out by his supervisor. The NRC found the licensee actions to control radiation contamination controls for these rooms and the components In these rooms adequate.

No NRC follow-up Is required.

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JOSEPH E. CARRASCO, ENGINEERING INSPECTOR ASSESSMENT OF THE LICENSEE RESPONSE TO CONCERNS RAISED TO THE NRC REGARDING ACTIVITIES AT FITZPATRICK NUCLEAR POWER PLANT 12.- '(Related to Concern #4) System 52 buildings not on SSC List; no list - The Issue of designation of System The alleger's concern involving the of components on PEDB for 'structures,' e.g. doors. (References made to 052 for buildings, and addition of issue of designation of System 52 for GL 83-28 response and DER 93-061 1). components assigned to the system to buildings and associated components the PEDB is discussed in response to was extensively discussed in concern 3 (therefore, Response 3 responses to concern No.3. This provides the response for concern 12, as concern Is closed. No NRC follow-up well). Is needed.

13.- 'The reactor Building roof started leaking in 1995 and is near of its -Currently, as leaks are Identified, Regarding the condition of the Reactor useful life. As a result, the steel roof decking may be rusting, potentially corrective action is planned, The reactor Building Roof, it appears that the impacting on the future operation of the Standby Gas Treatment System." building roof has sustained only minor licensee has successfully performed leaks, which have been successfully repairs to this roof in the past. The repaired. current condition of this roof and the existence of minor leaks do not Design Basis for the Secondary Containment ' -The current condition of the Reactor adversely affect the functional and

_ The Reactor Building (RB) is designed to be sufficiently leak tightto allow Building roof and the existence of roofing structural integrity of the secondary the Standby Gas Treatment System ISGTS) to reduce the reactor building Yaks do not adversely affect secondary containment. Therefore, the licensee pressure to a minimum subatmospheric pressure of 0.25 in. of water (under cotA inment integrity. Because of the maintains the secondary containment neutral wing conditions) when the SGTS fans are exhausting RS atmosphere followinlo. , design basis by using Technical.

at a rate of 200% (6,000 cfm) per day of RB free volume. - This safety / Specifications surveillance and basis takes into account expected leakage into a structure of the size of the JAF Technical Sp cifications contain maintenance rule walkdowns. The RB. surveillance requirements to ensure this NRC considers these actions safety design basis (0) is met (last test acceptable and no follow-up Is was performed on Octoberl 6, 1996.). required.

-To support 10CFR 50.56 (Maintenance Rule) performance monitoring, a baseline walkdown of the RB roof was performed in 1996. Further inspection is required once every two l___ years.

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JOSEPH E. CARRASCO, ENGINEERING INSPECTOR ASSESSMENT OF THE LICENSEE RESPONSE TO CONCERNS RAISED TO THE NRC REGARDING ACTIVITIES AT FITZPATRICK NUCLEAR POWER PLANT 7 1 14.- '(Related to Concern i 1i A question regarding the adequacy of -A new Control Room door was installed The licensee has installed a new tornado missile protection for the new control roorii door to the in accordance with plant modification F1- Control Room (CR3 door following administration building."90-013. This mod. installed a new OA plant modification No. F1-90-013.

Cat. I door designed for tornado pressure This modification Installed a new GA loading along with other design attributes. Category "I" door designed for tornado pressure loading along with other design attributes. However, regarding the adequacy of tornado missile protection of the new control room door, the licensee upgraded the access bridge installed between the new support and administration building and the control room. The new access bridge was Installed to provide missile protection to the new CR door. The licensee has Initiated a DER to further assess the adequacy of the missile protection for the CR door.

The NRC needs additional Information on the general layout of the missile protection for the CR door.

4 i.

15. 'Concern expressed with ESW pump room ventilation isolation due to a -This problem was previously described Regarding the alleger's concern possible fire damper isolation.' to the NRC in LER 91-021-00. -. expressed with the emergency service Inadequate ventillation following fire water (ESWM pump room ventilation as damper closure resulted from Inadequate is described in License Event Report analysis of the effects of the closure of (LER) 91-021-00. It appears that the dampers installed to meet NRC licensee has completed the corrective requirements in 1980.-m-* Corrective actions described In LER 91-021-00.'t actions described in LER 91-021-00 have Therefore, the NRC considers this all been completed. concern closed with no further follow-up. g

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JOSEPH E. CARRASCO, ENGINEERING INSPECTOR ASSESSMENT OF THE LICENSEE RESPONSE TO CONCERNS RAISED TO THE NRC REGARDING ACTIVITIES AT FITZPATRICK NUCLEAR POWER PLANT Concerns Resolution/ Corrective Action NRC proposed response to concerns:

16.- 'Concern expressed with the design of a contaminated drain line from - Resulting from a review of modification The licensee evaluated piping the administration building RCA since it is a standard line buried In gravel F1-90-013, the Radiological and configuration of drains for the without a guard pipe." Environmental Services (RES) department personnel and equipment evaluated this configuration against the decontamination located under and Question for the licensee: guidance of Regulatory Guide 1.143. between the administration building

'There was no documentation of the basis for choices made by the design RES determined the design was and the power block. This evaluation engineering organization with respect to the modified piping configuration acceptable since the regulatory guide was perfomed following the guidance Included with ECN-024." We need more Information on this Issue. exempts drain lines from requiring welded of NRC Regulatory Guide 1.143. As a connections, result of this evaluation, the licensee How many branches of this piping was hydro tested? - ECN 024 was issued against Issued ECN 024 to install welded modification Fl 90-013 to Install welded piping in place of bell and spigot piping In place of bell and spigot connections. Further, the licensee connections. - ECN A hydrostatic test performed a hydrostatic test of-the of the drain line (prior to construction drain line (before the construction outlined on ECN-024) demonstrated outlined on ECN-024) that showed satisfactory leak tightness. satisfactory leak tightness, It appears that the licensee has taken the proper steps to address this concem.

However, the NRC has requested additional Information on this Issue.

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JOSEPH E. CARRASCO, ENGINEERING INSPECTOR ASSESSMENT OF THE LICENSEE RESPONSE TO CONCERNS RAISED TO THE NRC REGARDING ACTIVITIES AT FITZPATRICK NUCLEAR POWER PLANT 17.- Concern expressed with the resolution of DER 97-45. Mod F1-92-145 - Three corrective actions were specified About the concern expressed with the involving seismic 11over I piping. in response to DER 97-045: 1) resolution of DER 97-45. the licensee Engineering Change Notice (ECN) was addressed Modification No. FI written against mod. F1-92-145 145 involving seismic II over I piping documenting the current as-left in corrective action No. 2 of DER 97-configuration of the nitrogen purge 045. Specifically, the licensee vaporizer stem and condensate piping and completed a calculation Justifying the to updateapplicable type 'A' drawings. acceptability of abandoning portions This corrective action is completed 2) of the nitrogen purge vaporizer steam complete a calculation justifying the and condensate piping in place. The acceptability of abandoning portions of results of the calculation concluded the nitrogen purge vaporizer steam and that this piping will not have an condensate piping in place. The results of adverse impact on the safe shutdown the calculation concluded that this piping of the plant nor will create seismic 11 will not have impact on the safe over I situation during a seismic event.

shutdown of the plant during a seismic The licensee's analysis to address this event. 3) The 3th corrective action was concern was adequate. No NRC to address the issue of a partially installed follow-up Is needed.

mod. by revising the mod. to delete the remaining and then closing the mod. The action will be completed October 22.

1997.

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JOSEPH E. CARRASCO, ENGINEERING INSPECTOR ASSESSMENT OF THE LICENSEE RESPONSE TO CONCERNS RAISED TO THE NRC REGARDING ACTIVITIES AT FITZPATRICK NUCLEAR POWER PLANT 18.- AQCns92-289, 92-290 and 92-291 were never entered into the - AOCR 92-289 was initiated on 8/28192 The licensee's document control corrective action system and resolved." and closed on 01/15/93. record showed the following:

- AOCR 92-290 was initiated on 10/2/92 and was closed 1/12/93. AQCR 92-289 was initiated on

- ACOR 92-291 was not issued. August 28. 1992 with concurrence from GA management. OA management accepted response to AQCR 92-289 (following independent verification of corrective action) and closed the document on January 15, 1993.

AOCR 92-290 was initiated on October 2, 1992 with concurrence from QA management. QA management accepted response to AQCR 92-290 (following independent verification of corrective action) and closed the document on January 12, 1993.

The licensee did not issue AOCR 92-291. The same Individuals who were Involved with preparation of AQCRs92-289 and 92-290 had also originally

.reserved' AQCR 92-291. There Is no indication that they did not allow items into the system. Rather, It appears that after initiating the first two AOCRs, the individuals found that they did not need to use the third reserved number. Therefore, the alleger has no substantiation for his concerns regarding AQCR 92-291.

Therefore, the alleger has no

, _I substantiation for his concerns 14

f JOSEPH E. CARRASCO, ENGINEERING INSPECTOR ASSESSMENT OF THE LICENSEE RESPONSE TO CONCERNS RAISED TO THE NRC REGARDING ACTIVITIES AT FITZPATRICK NUCLEAR POWER PLANT 19.- 'The resolution of DERs related to the CAD steam line The CAD steam line DER (97-0045) was Issues concerning the Containment Air modification/condensate thermosiphon heat exchanger mod provided an discussed in the response to concern 4 Dilution (CAD) steam line DER (97-example of the overall safety culture at FitzPatrick." and 17. 0045) were discussed in the response to the alleger's concerns No.4 and Regarding the issue of the condensate No.17. Regarding the condensate The issue regarding the condensate thermosiphon heat exchanger thermosiphon heat exchanger thermosiphon heat exchanger modification is as follows: modification as is described in (1) and (2). modification, the licensee Is in the process of carrying out the corrective (1) actions prescribed in DER 94-0471 that include an evaluation to decide if PID 55112, written December 21, 1993 and subsequently converted Into Corrective actions for DER 94-0471 the present configuration can maintain WR 93-04347-00 documented the following: 'CST tank temperature Is not included evaluating whether it was the CST temperature to above 40° F.

maintained by the reboiler or the auxiliary boiler system (the steam supply to possibly for CST temperature to lower to The corrective actions of DER 97-the thermosiphon heat exchangers was Isolated in the late 1970's) as 40° F. revising FAF-SE-91-095 is 0471 are Incomplete. About the Issue described In the FSAR Section 10.9.3. Instead, the licensee use condensate incomplete. of nuclear safety culture, an transfer pump minimum flow to maintain CST teperature above the FSAR independent study conducted late in prescribed 400 F required by the FSAR. In 1991, the licensee prepared 1996 concluded that the licensee Safety Evaluation No. JAF-SE-91-095 to address the substitution of made significant progress at Auxiliary Boiler Steam Supply. FtzPatrick In establishing a strong nuclear safety culture. The NRC will In 1994, the licensee prepared DER 94-0471 to revise SE No. JAF-SE91- need additional Information to ensure 095, or to prepare a new SE. The contents of DER 94-0471 is described as that the corrective actions prescribed follows: by DER 97-0471 are addressed In reasonable time.

(2)

The thermosiphon heat exchangers, 33E-24A/B, are steam supplied heat exchangers that were designed to maintain the water in condensate storage tank [CST) above 40° F. 40° Fis mentioned in the FSAR Section 4.2 as being nil-ductility transition temperature (NDTT for carbon steel. The NDTT is the temperature below which ferritic steel breakage is brittle rather than ductile.

_________________________________________ _______________________ I _____________________

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