Text

IR 05000259-08-009, 05000260-08-009 and 05000296-08-009, on 04/27/08 - 05/02/08, Browns Ferry Nuclear Plant, Units 1, 2 and 3, Event Followup
	ML081510829
Person / Time
Site:	Browns Ferry
Issue date:	05/30/2008
From:	Wert L; Division Reactor Projects II
To:	Campbell W; Tennessee Valley Authority
Shared Package
ML090210101	List: IR 05000259/2008007; IR 05000259/2008009; ML081510829;
References
	IR-08-009
	Download: ML081510829 (28)
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UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION II

SAM NUNN ATLANTA FEDERAL CENTER

61 FORSYTH STREET, SW, SUITE 23T85

ATLANTA, GEORGIA 30303-8931

May 30, 2008

Mr. William R. Campbell, Jr.

Chief Nuclear Officer and Executive Vice President

Tennessee Valley Authority

6A Lookout Place

1101 Market Street

Chattanooga, TN 37402-2801

SUBJECT: BROWNS FERRY NUCLEAR PLANT - NRC SPECIAL INSPECTION REPORT

05000259/2008009, 05000260/2008009 AND 05000296/2008009

Dear Mr. Campbell:

On May 2, 2008, the Nuclear Regulatory Commission (NRC) completed a special inspection at

your Browns Ferry facility. The enclosed report documents the inspection findings which were

discussed on May 2 with Mr. R. West and other members of his staff.

Between March 24 and March 30, 2008, your staff disassembled both Division I 3A and 3C, and

Division II, 3B and 3D, residual heat removal heat exchanger service water (RHR HX SW) outlet

flow control valves (FCVs) and found significant degradation in each valve including stem-to-

disc separation in several. During the last Unit 2 refueling outage, your staff had identified

similar degradation in two of the Unit 2 RHR HX SW outlet FCVs.

On April 18, 2008, NRC Region II management established a Special Inspection Team using

the guidance contained in Management Directive 8.3, NRC Incident Investigation Program. The

Special Inspection Team was chartered to identify the degradation mechanisms that had

affected these valves, review the past and current root-cause evaluations, review the interim

and long-term corrective actions, and verify that the interim corrective actions are adequate to

ensure the operability of the Units 2 and 3 RHR HX SW outlet FCVs, until a long-term resolution

is implemented. The inspection examined activities conducted under your license as they relate

to safety and compliance with the Commissions rules and regulations and with the conditions of

your license. The inspectors reviewed selected procedures and records, conducted field

walkdowns, observed activities, and interviewed personnel.

Based on the results of this inspection, the inspectors identified one issue of very low safety

significance (Green) that was determined to involve a violation of NRC requirements. However,

because of its very low safety significance and because it has been entered into your Corrective

Action Program, the NRC is treating this issue as a non-cited violation, in accordance with

Section VI.A.1 of the NRCs Enforcement Policy. If you deny the non-cited violation, you should

provide a response with the basis for your denial, within 30 days of the date of this inspection

report to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC

20555-0001, with copies to the Regional Administrator, Region II; the Director, Office of

TVA 2

Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and

the NRC Resident Inspector at the Browns Ferry facility.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its

enclosure will be available electronically for public inspection in the NRC Public Document

Room or from the Publicly Available Records (PARS) component of NRC's document system

(ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-

rm/adams.html (The Public Electronic Reading Room).

Sincerely,

/JHM RA for/

Leonard D. Wert, Director

Division of Reactor Projects

Docket Nos.: 50-259, 50-260, 50-296

License Nos.: DPR-33, DPR-52, DPR-68

Enclosure: Inspection Report 05000259/2008009, 05000260/2008009 and 05000296/2008009

w/Attachment: Supplemental Information

cc w/encl.: (See page 3)

_________________________

OFFICE RII:DRP RII:DRP RII:DRP RII:DRP RII:DRP

SIGNATURE JBB /RA/ CRS /via email/ RCH /via email/ RLN /via telecon JHM /RA/

NAME JBaptist CStancil BHagar RNease JMoorman

DATE 05/30/2008 05/30/2008 05/30/2008 05/30/2008 05/30/2008

E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO

TVA 3

cc w/encl: Larry E. Nicholson

Ashok S. Bhatnagar General Manager

Senior Vice President Performance Improvement

Nuclear Generation Development and Tennessee Valley Authority

Construction Electronic Mail Distribution

Tennessee Valley Authority

Electronic Mail Distribution Michael A. Purcell

Senior Licensing Manager

William R. Campbell, Jr. Nuclear Power Group

Chief Nuclear Officer and Executive Vice Tennessee Valley Authority

President Electronic Mail Distribution

Tennessee Valley Authority

6A Lookout Place H. Rick Rogers

1101 Market Street Vice President

Chattanooga, TN 37402-2801 Nuclear Engineering and Technical

Services

(Vacant) Tennessee Valley Authority

Vice President Electronic Mail Distribution

Nuclear Support

Tennessee Valley Authority Beth A. Wetzel

3R Lookout Place Manager

1101 Market Street Corporate Nuclear Licensing and Industry

Chattanooga, TN 37402-2801 Affairs

Tennessee Valley Authority

R. G. (Rusty) West Electronic Mail Distribution

Site Vice President

Browns Ferry Nuclear Plant Senior Resident Inspector

Tennessee Valley Authority Tennessee Valley Authority

Electronic Mail Distribution Browns Ferry Nuclear Plant

U.S. NRC

D. Tony Langley 10833 Shaw Road

Manager Athens, AL 35611-6970

Licensing and Industry Affairs

Tennessee Valley Authority Chairman

Electronic Mail Distribution Acting Licensing Manager

Limestone County Commission

John C. Fornicola 310 West Washington Street

General Manager Athens, AL 35611

Nuclear Assurance

Tennessee Valley Authority

Electronic Mail Distribution

General Counsel

Acting Licensing Manager

Tennessee Valley Authority

Electronic Mail Distribution

TVA 4

Letter to William R. Campbell, Jr. from Leonard D. Wert dated May 30, 2008

SUBJECT: BROWNS FERRY NUCLEAR PLANT - NRC SPECIAL INSPECTION REPORT

05000259/2008009, 05000260/2008009 AND 05000296/2008009

Distribution w/encl:

E. Brown, NRR

L. Raghavan, NRR

C. Evans, RII

L. Slack, RII

OE Mail

RIDSNRRDIRS

PUBLIC

U. S. NUCLEAR REGULATORY COMMISSION

REGION II

Docket Nos: 50-259, 50-260, 50-296

License Nos: DPR-33, DPR-52, DPR-68

Report No: 05000259/2008009, 05000260/2008009 and

05000296/2008009

Licensee: Tennessee Valley Authority (TVA)

Facility: Browns Ferry Nuclear Plant, Units 1, 2 & 3

Location: Corner of Shaw and Nuclear Plant Roads

Athens, AL 35611

Dates: April 27 - May 2, 2008

Inspectors: R. Hagar, Senior Resident Inspector - Robinson

C. Stancil, Resident Inspector - Browns Ferry

Approved by: Rebecca L. Nease, Chief

Reactor Project Branch 6

Division of Reactor Projects

Enclosure

SUMMARY OF FINDINGS

IR 05000259/2008009, 05000260/2008009 and 05000296/2008009; 04/27/08 - 05/02/08;

Browns Ferry Nuclear Plant, Units 1, 2 and 3; Event Followup.

The special inspection team (SIT) inspection was conducted by a senior resident inspector and

a resident inspector. One finding of very low safety significance (Green) was identified. The

significance of most findings is indicated by their color (Green, White, Yellow, Red) using

Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). The

NRCs program for overseeing the safe operation of commercial nuclear power reactors is

described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.

A. NRC-Identified and Self-Revealing Findings

Green. The inspectors identified a Green non-cited violation of 10 CFR 50, Appendix B,

Criterion XVI, for the licensees failure in March, 2000, to take action to preclude

repetition of stem-to-disc separation events in residual heat removal heat exchanger

service water outlet valves. The finding is more-than-minor, because if left uncorrected

the condition would become a more significant safety and regulatory concern. In Phase

1 of the SDP, described in IMC 0609, Attachment 4, this finding screened as Green

because it affected the Mitigating Systems cornerstone and was a design deficiency

confirmed not to result in loss of operability or functionality. The finding has a cross-

cutting aspect in the area of problem identification and resolution (P.1(c)) because the

licensee did not thoroughly evaluate a problem such that the resolutions addressed

causes and extents of condition. (Section 4OA3.3)

B. Inspection Results

Degradation that occurred in the residual heat removal heat exchanger service water

outlet flow control valves included broken motor lugs, broken motor leads, separations of

valve discs from their stems, broken stem-to-disc tack welds, stem shear, slipped anti-

rotation collars, separations of valve handwheels from their shafts, a gasket leak, a

broken disc flute, and dislodging of a stem cap.

In all cases, the degradation mechanism was a combination of operating the valves

under conditions that induced high levels of vibration in the valves and the vulnerability

of the valves to vibration-induced damage.

Repairs have been sufficient to support continued operation and the current operability

determination in the near term, because the repairs have reduced the vulnerability of the

valves to vibration-induced damage. However, no evidence suggests that those repairs

have eliminated that vulnerability.

Planned long-term corrective actions include replacing the currently-installed Walworth

and Anchor-Darling valves in Units 2 and 3 with Copes-Vulcan valves identical to those

in Unit 1, to further reduce the valves vulnerability to vibration induced damage.

Enclosure

3

The implemented corrective actions only partially address the root cause, in that those

actions have reduced the vulnerability of the valves to vibration-induced damage, but

have not yet effectively changed the conditions under which the valves are operated.

The planned corrective actions are expected to further reduce the vulnerability of the

valves to vibration-induced damage, but no changes are planned to change the

conditions under which the valves are operated.

The only generic safety issue associated with this inspection (vibration-induced damage

to safety-related components) has been adequately addressed in generic

communications.

Enclosure

REPORT DETAILS

Initial Conditions

During the period of March 24 to March 30, 2008, Unit 3 was shutdown in Operating Mode 5 for

a planned refueling outage which had begun on March 18, 2008, and Units 1 and 2 were

operating at 100% reactor power.

Event Description

On March 24, 2008, the licensee disassembled and inspected the 3A residual heat removal

heat exchanger service water (RHR HX SW) outlet valve and found that it had experienced

stem-to-disc separation with severe erosion of the valve body and internal rib guides. On March

25 and March 28, respectively, the licensee disassembled and inspected the 3C and 3B RHR

HX SW outlet valves and found that they had also experienced significant internal damage, as

well as, stem-to-disc separation. On March 30, the licensee determined that the 3D RHR HX

SW outlet valve had not experienced stem-to-disc separation but had experienced tack weld

breakage and minor valve body erosion.

Special Inspection Team (SIT) Charter

Based on the criteria specified in Management Directive 8.3, NRC Incident Investigation

Program, and Inspection Procedure 71153, Event Follow-up, a special inspection was initiated

in accordance with Inspection Procedure 93812, Special Inspection. The objectives of the

inspection, described in the charter, are listed below and are addressed in the identified

sections:

1. Develop a complete description of the degradation found in the Units 2 and 3 RHR HX

SW outlet flow control valves (FCVs), and the various degradation mechanism(s)

observed over multiple operating cycles. (Section 4OA3.1)

2. Review the licensees repair efforts of all units RHR HX SW outlet FCVs to determine if

repairs are sufficient to support operation and the current operability determination.

(Section 4OA3.2)

3. Review the licensees root cause analysis and extent of condition. Assess the adequacy

of the licensees implemented and/or planned corrective actions to address the root

cause. (Section 4OA3.3)

4. Review industry operating experience and the licensees actions in response to any

related operating experience items including Information Notice 2006-015, Vibration-

Induced Degradation and Failure of Safety Related Valves. (Section 4OA3.4)

5. Identify any potential generic safety issues and make recommendations for appropriate

follow-up actions (e.g., Information Notices, Generic Letters, Bulletins). (Section

4OA3.5)

Enclosure

5

OTHER ACTIVITIES

4OA3 Event Followup

.1 Degradation Found in the RHR HX SW Outlet FCVs

a. Inspection Scope

The inspectors walked down the RHR HX SW outlet FCVs in Units 1, 2, and 3,

interviewed selected plant personnel, and reviewed problem evaluation reports (PERs)

and work orders (WOs) associated with the degradation that occurred in these valves to

develop a complete description of the degradation that had been found in these valves

and to identify the degradation mechanisms.

b. Observations and Findings

The inspectors noted that valves from different manufacturers were installed in the

different units: Copes-Vulcan valves were installed in Unit 1, Walworth valves with fluted

discs were installed in unit 2, and Anchor-Darling valves are installed in unit 3. Licensee

records show that these valves have all been installed in their respective units since

each units restart date, except for the 2C and 2D HX valves in Unit 2. With respect to

those valves:

The 2C RHR HX SW outlet valve (2-FCV-23-0040) operated for a short period with a

non-fluted disc. That period began in March, 2003, after the valve experienced a

broken disc flute and the licensee installed a non-fluted disc to replace the broken

disc. That period ended in October, 2003, when the licensee re-installed a fluted

disc after the valve experienced seat leakage and a stem-to-disc failure.

During the 2007 Unit 2 refueling outage, the licensee identified a stem shear with

severe valve body and rib guide erosion on the 2D RHR HX SW outlet valve (2-FCV-

23-52). The licensee subsequently replaced this valve with a 12-inch Anchor Darling

model like those in Unit 3.

Licensee records also show that the subject valves had experienced the following

damage since each units restart:

Number of damage occurrences since unit restart

Type of Damage in Unit 2 valves in Unit 3 valves

broken motor lugs 2 2

broken motor leads 0 5

stem-to-disc separation 1 4

broken stem-to-disc tack welds 0 1

stem shear 1 0

slipped anti-rotation collar 0 5

handwheel separation 4 1

Enclosure

6

broken disc flute 1 0

stem cap separation 1 1

gasket leakage 1 0

Records reviewed by the inspectors showed no damage to the RHR HX SW outlet FCVs

in Unit 1.

Comprehensive timelines of the damage that occurred to the RHR HX SW outlet FCVs

in Units 2 and 3 are provided in Attachments 1 and 2 to this report. Those timelines

show, in part, that:

In Unit 2, the earliest occurrence of damage to these valves was in December, 1994,

when a handwheel separated from Valve 2-023-040. That date was approximately

42 months after the July, 1991, restart of the unit.

In Unit 3, the earliest damage occurrence was a broken motor lug in March, 1997,

approximately 17 months after the October, 1995, restart of that unit.

Damage events generally occurred earlier in plant life in Unit 3 than they did in

Unit 2.

More damage events affected Unit 3 valves than Unit 2 valves.

The licensees responses to these events included 2 B-priority root-cause analyses,

11 C-priority apparent-cause evaluations, 7 C-priority no-cause evaluations, and 2 D-

priority apparent-cause evaluations.

Licensee evaluations of these damage states consistently indicated that all of the

damage identified above was caused by vibration; several evaluations referred to that

vibration as normal. The inspectors determined that the damage had actually been

caused by a combination of two factors:

During shutdown cooling, the licensee operated these valves in a way that allowed

relatively-severe cavitation to occur immediately downstream of the valve discs.

That cavitation induced vibration that affected both the valve body and the valve

stem-disc assembly.

The valves were vulnerable to vibration-induced damage.

.2 Efforts to Repair the Subject Valves

a. Inspection Scope

The inspectors reviewed licensee records to determine whether repairs are sufficient to

support operation and the operability determinations documented in Functional

Evaluations 42538 (Unit 2) and 42520 (Unit 3) .

Enclosure

7

b. Observations and Findings

The inspectors determined that the licensee had repaired the damage the valves had

experienced, and that the licensee had also made several changes to the valves to

reduce the valves vulnerability to vibration-induced damage. The damage states and

the corresponding changes made by the licensee are summarized in the table below.

Damage state Change(s) made to address

Broken motor lugs & leads Removed all motor terminal blocks (no longer

requiring lugs) and replaced with Raychem splices

Stem-to-disc separation Replaced threaded and tack-welded connections

between the valve stems and the discs with full-

welded connections

Slipped anti-rotation collar Slightly drilled the valve stems at the set-screw

locations and replaced the standard set screw with a

dog-point-style set screw

Handwheel separation Removed set screws, spot-drilled the handwheel

shafts, and then on each shaft installed two set

screws back-to-back

The inspectors considered that the damage states that had the greatest safety

significance were broken motor leads and lugs, because those can disable the valve

actuators, and slippage of the anti-rotation collar, because that event can prevent the

valve from responding to actuator operation. The other damage states would degrade

the valves, but would not render the valves incapable of performing their safety

functions, which are to open to remove reactor decay heat and to close to isolate flow

from a heat exchanger tube rupture.

Using engineering judgment, the inspectors determined that the changes made by the

licensee had reduced the vulnerability of the valves to vibration-induced damage. In

particular, as shown in the table above, the licensee has made changes to address the

damage states that had the greatest significance (broken motor leads and lugs and

slippage of the anti-rotation collar). However, the inspectors noted that the licensee had

no evidence to indicate that these changes had eliminated that vulnerability.

Furthermore, the inspectors noted that the licensees long-term plan to address this

issue included replacing the Walworth and Anchor-Darling valves currently installed in

Units 2 and 3 with Copes-Vulcan valves identical to those currently installed in Unit 1.

Because the Copes-Vulcan valves contain approximately 50% more mass than do the

Walworth and Anchor-Darling valves, the inspectors considered that completing that

replacement should further reduce the vulnerability of the valves to vibration-induced

damage. However, the inspectors also noted that the licensee had no evidence to

indicate that Copes-Vulcan valves are immune to vibration-induced damage.

Enclosure

8

For this charter item, because the licensee has taken action to reduce the vulnerability of

these valves to vibration-induced damage, the inspectors concluded that repairs are

sufficient to support operation and the operability determination. However, continued

monitoring of these valves during shutdown-cooling operation will be required to

determine whether additional actions are necessary.

.3 Root-Cause Analyses and Extent of Condition

a. Inspection Scope

The inspectors reviewed the licensees root-cause analyses and extent-of-condition

determinations to determine whether the licensees implemented and/or planned

corrective actions address the root cause.

b. Observations and Findings

As mentioned above, the inspectors determined and the licensee staff agreed that the

damage the RHR HX SW outlet FCVs had experienced had been caused by the

combination of operating those valves in a way that induced vibration and the

vulnerability of those valves to vibration-induced damage. The root cause was

therefore that combination.

As the timelines in attachments 2 and 3 show, the licensee completed two root-cause

analyses that address the damage these valves have experienced: (1) PER 35419,

which the licensee initiated in April of 2000 after the first stem-to-disc separation event;

and (2) PER 104621, which the licensee initiated in June of 2006, after the RHR SW

system had exceeded the performance criteria established in accordance with 10 CFR

50.65 (Maintenance Rule) and had, therefore, been classified under paragraph (a)(1) of

that rule.

With respect to the combination of causes described above, the inspectors noted the

following in PER 35419:

This PER had been initiated to address a stem-to-disc separation event that

occurred in Valve 3-FCV-023-0046 in Unit 3.

The licensee determined that the root cause of the subject event had been fatigue

placed on the valve disc when the flow rate was low causing a high differential

pressure across the valve.

With respect to the actual root cause described above, this statement mentions

low-flow conditions but does not mention cavitation, vibration resulting from

cavitation, or the valves vulnerability to vibration-induced damage. This PER,

therefore, did not fully identify the root cause of the damage.

The corrective action to address the identified root cause was to change the

operation instruction for the RHR SW system so that when one RHR SW heat

exchanger (and its associated outlet FCV) is operated at low flow rates, operators

Enclosure

9

would establish flow through another RHR HX so that the combined flow through

both HXs would be approximately 4000 gallons/minute. The licensee expected

this change to lower the pressure the valve has to work against thereby

lessening the probability of damage to the valve. The corresponding procedure

change was implemented in Revision 45 of Procedure 0-OI-23, Residual Heat

Removal Service Water System, which was issued on May 30, 2000.

Subsequent experience demonstrated that this procedure change was not effective

at preventing stem-to-disc separation events. Implementation of corrective actions

for this PER, therefore, constituted the performance deficiency documented below

(following the observations regarding PER 104621).

The record for this PER did not include a review of the effectiveness of this

corrective action. This record, therefore, did not indicate that the licensee had

attempted to verify that the procedure change had the intended effect.

The record for this PER indicated that the valve vendor (Anchor Darling) had

specified the maximum differential pressure that the RHR HX SW outlet FCVs

should experience, and had recommended that the valves, which had been

subjected to a differential pressure outside the recommended maximum, should be

disassembled and inspected. In this PER, the licensee asserted without support

that the valve that experienced the failure (3-FCV-023-0046) was the only valve

that had been subjected to the conditions that caused the failure, and therefore

determined that disassembly and inspection of the other valves was not necessary.

However, that assertion was not consistent with operating records, which indicated

that the four RHR HX SW outlet FCVs in Unit 3 and the four RHR HX SW outlet

FCVs in Unit 2 had all been operated for approximately the same amount of time

during shutdown cooling.

The inspectors noted the following in PER 104621:

This PER had been initiated after valve 3-FCV-23-34 had failed to stroke

electrically due to two broken motor leads, and thereby caused the 3A RHR heat

exchanger train to exceed its Maintenance Rule performance criteria.

The identified cause of the damage was hard strain and excessive vibration, and

the identified cause of the cause was excessive vibration at lower flow

conditions. The valves are not big enough physically to minimize this vibration.

With respect to the actual root cause described above, this statement mentions

low-flow conditions and vibration, but does not mention the valves vulnerability to

vibration-induced damage. However, the corrective actions taken (described

below) indicate that the licensee recognized that vulnerability. The inspectors,

therefore, considered that in this PER, the licensee identified the actual root

cause of the damage.

Enclosure

10

Corrective actions taken to prevent recurrence (of broken motor leads) were to

initially replace hard-strained lugs and insulated lugs with uninsulated lugs; retrain

the motor-lead wiring; and shrink Raychem on the barrel of the lugs, the insulation

on the motor leads, and the control wiring for additional support. In this PER, the

description of these corrective actions includes the statement, These WO's are for

the interim fix, long term fix will be the replacement of the valves with valves similar

to those used on U1.

Final valve motor actuator corrective actions, as developed in this PER, resulted in

removal of all terminal blocks and installation of Raychem splices. By February,

2008, these corrective actions were completed on all of the RHR HX SW outlet

FCVs in Units 2 and 3.

Since implementing these corrective actions, no broken motor leads have occurred

on the affected FCVs.

With respect to this charter item, these observations indicate that the corrective actions

implemented by the licensee as a result of their root-cause analyses do not fully address

the root causes, in that:

The only root-cause analyses initiated to address damage to these valves were

PERs 35419 and 104621;

The corrective action described in PER 35419 to change how the valves are

operated during shutdown cooling was not effective at preventing stem-to-disc

separation events; and

The corrective actions described in PER 104621 had addressed only the broken-

motor-lead damage state.

However, as noted above in section 4OA3.2.b, the licensee had taken corrective action

through other PERs and/or work orders to address the other failure states as well.

Considering the corrective actions described in PER 104621 along with the other

corrective actions discussed above, the inspectors consider that the corrective actions

implemented by the licensee has addressed part of the root cause, in that those

corrective actions have reduced the vulnerability of the valves to vibration-induced

damage. However, the licensee has not yet planned or implemented corrective actions

to reduce the vibration that these valves experience by changing how they operate these

valves during shutdown cooling.

Development and implementation of corrective actions for PER 35419 constituted the

performance deficiency described below.

Introduction: The inspectors identified a Green non-cited violation of 10 CFR 50,

Appendix B, Criterion XVI, for the licensees failure in March, 2000, to take action to

preclude repetition of stem-to-disc separation events in RHR HX SW outlet FCVs.

Enclosure

11

Description: In March, 2000, after RHR HX SW outlet FCV 3-FCV-023-0046

experienced a stem-to-disc separation event, the licensee initiated PER 35419 to

address that event, completed a root-cause analysis of that event, and developed a

corrective action that was intended to prevent recurrence of stem-to-disc separation

events. That corrective action, implemented in May 2000, was to revise an operating

procedure to reduce the differential pressure the valves would experience during

shutdown-cooling operations. That corrective action was shown to be not effective when

in March, 2008, the licensee discovered that stem-to-disc separation had occurred in

Valves 3-FCV-023-0034, 3-FCV-023-0040, and 3-FCV-023-0046.

Analysis: This finding was more-than-minor because if left uncorrected the condition

would become a more significant safety and regulatory concern, in that failure to

adequately address the conditions that caused a stem-to-disc separation event in one

valve could allow those conditions to cause not only stem-to-disc separation events in

other valves, but also more-risk-significant damage that could render the valves

incapable of accomplishing their safety functions. In Phase 1 of the Significance

Determination Process described in MC 0609, Attachment 4, this finding affected the

Mitigating Systems cornerstone and was a design deficiency confirmed not to result in

loss of operability or functionality. The finding, therefore, screened as Green.

The inspectors determined that the cause of this finding was that while evaluating the

event that prompted PER 35419, the licensee did not determine that the cause of the

event had been a combination of operating the valves under conditions that induced

vibration and the vulnerability of the valves to vibration-induced damage, and

consequently did not develop effective corrective actions to address both of those

factors. The inspectors also determined that although the performance deficiency

associated with this finding occurred in 2000, this finding is representative of current

licensee performance, because since 2000, the licensee has not changed their

corrective action program root cause determination methodology in a way that clearly

addresses the weaknesses the inspectors noted in the PER 35419 evaluation. The

finding therefore has a cross-cutting aspect in the area of problem identification and

resolution, because the licensee did not thoroughly evaluate a problem such that the

resolutions address causes and extent of conditions, in that the licensee did not

thoroughly evaluate the April, 2000, stem-to-disc separation of valve 3-FCV-023-0046

such that the resolutions addressed the causes of the vibration. (P.1(c))

Enforcement: 10 CFR 50, Appendix B, Criterion XVI requires, in part, that for significant

conditions adverse to quality, licensees assure that corrective action is taken to preclude

repetition.

Contrary to the above, the licensee failed to take corrective action to preclude repetition

of a significant condition adverse to quality, in that:

the stem-to-disc separation event described in PER 35419 was a significant

condition adverse to quality with respect to Criterion XVI, because for that event,

licensee procedures required both determination of the cause and corrective action

to preclude repetition; and

Enclosure

12

the licensees actions taken in April, 2000, in response to the stem-to-disc separation

event described in PER 35419, failed to preclude repetition of stem-to-disc

separation events, in that three such events occurred in 2008.

Because this finding was of very low safety significance and has been entered into the

licensees corrective action program as PER 143502, consistent with Section VI.A of the

NRC Enforcement Policy, this violation is being treated as a non-cited violation, and is

designated as NCV 05000260/2008009-01, Failure to prevent recurrence of stem-to-

disc separation events in residual heat removal service water heat exchanger outlet

valves.

.4 Response to Industry Operating Experience

a. Inspection Scope

The inspectors reviewed industry operating experience and the licensees actions in

response to related operating experience items including:

NRC Information Notice (IN) 2006-015, Vibration-Induced Degradation and Failure

of Safety Related Valves;

NRC IN 83-70 and Supplement 1, Vibration-Induced Valve Failures;

NRC IN 2005-23, Vibration-Induced Degradation of Butterfly Valves;

NRC IN 2002-26, Failures of Steam Dryer Cover Plate After a Recent Power Uprate

at a BWR;

INPO Significant Event Report (SER)02-005, Lessons Learned from Power

Uprates; and

INPO SER83-20 Supplement 1, Improper Seating of Velan Swing Check Valves

Due to Disc/Hangar Arm Binding.

b. Observations and Findings

The inspectors determined that the licensee had reviewed and appropriately responded

to the operating experience items identified above.

.5 Potential Generic Safety Issues

a. Inspection Scope

The inspectors reviewed the circumstances within the scope of this inspection to

determine whether those circumstances included any potential generic safety issues.

b. Observations and Findings

The inspectors determined that no new generic safety issues were associated with

damage to the subject valves, because the inspectors considered that the only related

generic safety issue (vibration-induced damage to safety-related components) had been

adequately addressed in generic communications.

Enclosure

13

4OA6 Meetings, Including Exit

On May 2, 2008, the inspectors presented the inspection results to Mr. R. West and

other members of the plant staff. The inspectors confirmed that proprietary information

was not provided or examined during the inspection.

ATTACHMENT: SUPPLEMENTAL INFORMATION

Enclosure

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

S. Armstrong, Performance Improvement

T. Chan, Corporate Engineering

S. Douglas, General Manager Site Operations

B. Eberly, Corporate Engineering

R. Godwin, Site Support Manager

D. Hughes, Operations Superintendent

K. Harvey, Site Engineering

W. Justice, Engineering Manager

D. Langley, Site Licensing Manager

B. Mitchell, Performance Improvement Manager

B. Trappett, Site Engineering

R. West, Site Vice President

A. Yarbrough, Site Engineering

NRC personnel

T. Ross, Senior Resident Inspector - Browns Ferry

Attachment 1

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

None

Closed

None

Opened & Closed

05000260/2008009-01 NCV Failure to prevent recurrence of stem-to-disc separation

events in residual heat removal service water heat

exchanger outlet valves (4OA3.3)

Closed

None

Previous Items Closed

None

Discussed

None

Attachment 1

LIST OF DOCUMENTS REVIEWED

Problem Evaluation Reports

35419, Valve 3-FCV-023-0046 Disk Found Separated from Stem

38168, Anti-Rotation Plate for BFN-3-FCV-023-0046 Found On Top of Packing Gland

38915, 2-FCV-23-52, 2D Outlet Valve Handwheel Fell Off

39727, 2-FCV-23-40 Would Not Electrically Cycle to the Full Closed Position

41912, 3-FCV-23-46 Anti-Rotation Device Out of Position Several Occasions

44050, 2-FCV-023-0040 Valve Seat Requires Machining for Proper Leakage Prevention

44056, 2-FCV-23-46 Failure to Operate

50732, Valve 2-FCV-23-0040 Repeat Maintenance Due to Gasket Leakage

50734, 2-FCV-023-0040 Would Not Completely Close

52672, Valve 2-FCV-023-0046 Three Motor Leads Broken at Terminal Lug Connections

59437, Handwheel for 2-FCV-23-52 Dislocated From Limitorque Stem

69087, Valve 3-FCV-23-34 Indication Remained Double Lit at Full Open

80790, 3-FCV-23-34 Would Not Open from the Control Room

81108, Broken Motor Lead Found on 3-MVOP-23-34

85910, Unplanned Maintenance Rule Unavailability 3A RHR HX

91267, 2-FCV-23-34 Failed to Open On A2 Pump Start

99498, Broken Motor Lead for 3-MVOP-23-34

101897, 2-FCV-23-52 Stem Cover Fell Off

104621, MR PC exceeded on 3A RHR HX

114173, Handwheel for valve 2-FCV-23-52 Found On the Floor

122218, Stem failure of 2-FCV-23-52

127137, 3-FCV-23-34 Failed to Open or Close

136712, RHR SW 3-FCV-23-40 Valve Electrical Failure

140768, 3-FCV-23-34 Stem Separated from Disc

140824, 3-FCV-23-40 Stem Separated from Disc

141137, 3-FCV-23-46 Stem Separated from Disc

141219, 3-FCV-23-52 Broken Tack Welds

Work Orders

94-020161-000, Reinstall 2-MVOP-23-40 Handwheel

96-004063-000, Repair 2-MVOP-23-34 Motor Lugs

97-002514-000, Troubleshoot and Repair 3-MVOP-23-52 Thermalling Out

98-011029-000, Troubleshoot 3-MVOP-23-46 Not Opening

98-011271-001, Reinstall 3-MVOP-23-46 Stem Anti-Rotation Collar

00-003802-000, Disassemble and Refurbish 3-FCV-23-46

00-008114-000, Take Vibration Data On 3-FCV-23-46

02-003214-000, Reinstall 2-MVOP-23-52 Handwheel

02-004605-000, Reinstall Handwheel With Mod On 2-MVOP-23-52

02-004605-001, Install Handwheel Mod On 2-MVOP-23-46

02-004605-002, Install Handwheel Mod On 2-MVOP-23-40

02-004605-003, Install Handwheel Mod On 2-MVOP-23-34

02-004605-004, Install Handwheel Mod On 3-MVOP-23-52

02-004605-005, Install Handwheel Mod On 3-MVOP-23-46

02-004605-006, Install Handwheel Mod On 3-MVOP-23-40

Attachment 1

4

02-004605-007, Install Handwheel Mod On 3-MVOP-23-34

02-006589-001, Repair Operator Shaft and Reinstall Handwheel On 2-MVOP-23-52

03-003010-000, Disassemble, Inspect, and Repair 2-FCV-23-40

03-006967-001, Install New Valve Disc On 2-FCV-23-40

03-007163-000, Realign 3-MVOP-23-46 Anti-Rotation Collar

03-021389-000, Repair 2-MVOP-23-46 Broken Motor Leads

04-714325-000, Realign 3-MVOP-23-46 Anti-Rotation Collar

04-717160-000, Install Anti-Rotation Collar Mod On 3-MVOP-23-46

04-717160-001, Install Anti-Rotation Collar Mod On 3-MVOP-23-52

04-722062-000, Realign 3-MVOP-23-34 Anti-Rotation Collar

04-722067-000, Realign 3-MVOP-23-40 Anti-Rotation Collar

05-714220-001, Troubleshoot and Repair 3-FCV-23-34 Not Opening

05-722104-000, Machine 2-FCV-23-40 Fluted Disc

06-713538-000, Troubleshoot and Repair 3-FCV-23-34 Not Opening

06-716018-000, Reinstall 2-MVOP-23-52 Stem Cap

06-718716-000, Replace 3-MVOP-23-52 Stem Cap

06-725311-000, Disassemble and Inspect 2-FCV-23-52

07-711409-000, Repair 3-MVOP-23-34 Motor Leads

07-720665-000, Replace 3-FCV-23-34 Packing, Handwheel Key, and MOVATS

07-721848-000, Replace 3-MVOP-23-40 Motor Terminations With Raychem Splices

08-710819-000, Inspect and Repair 3-FCV-23-40 Not Moving

08-711543-000, Disassemble, Inspect, and Refurbish 3-FCV-23-34

08-711544-000, Disassemble, Inspect, and Refurbish 3-FCV-23-40

08-711545-000, Disassemble, Inspect, and Refurbish 3-FCV-23-46

08-711546-000, Disassemble, Inspect, and Refurbish 3-FCV-23-52

Procedures

TVA Corporate Procedure SPP-3.1, Corrective Action Program, Rev. 13

TVA Corporate Procedure SPP-3.1, Corrective Action Program, Rev. 1

TVA Business Procedure BP-250, Corrective Action Handbook, Rev. 12

MCI-0-000-GLV001, Generic Maintenance Instruction For Globe Valves, Rev. 21

Surveillance Instruction 2-SI-3.2.1, First and Augmented In-Service Test Valve Performance,

Rev. 23

Surveillance Instruction 3-SI-3.2.1, First and Augmented In-Service Test Valve Performance,

Rev. 9

Surveillance Instruction 2-SI-4.5.C.1(3), Residual Heat Removal Service Water Pump and

Header Operability and Flow Test, Rev. 98

Surveillance Instruction 3-SI-4.5.C.1(3), Residual Heat Removal Service Water Pump and

Header Operability and Flow Test, Rev. 30

Functional Evaluation 42538 (PER 141380) Unit 2 Heat Exchanger Outlet Valves 2-FCV-23-34,-

40,-46 & 52

Functional Evaluation 42520 (PER 140768) Unit 3 Heat Exchanger Outlet Valves 3-FCV-23-34,-

40,-46 & 52 and 2-FCV-23-52

Operations Instruction 0-OI-23, Residual Heat Removal Service Water System, Rev. 45

Operations Instruction 2-OI-74, Residual Heat Removal System, Rev. 137

Attachment 1

5

ML081510829

Text

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