ML20141L864

FOIA Request for Documents That Refer to Operation &/Or Mgt of Plant NPPs That Were Disseminated,Discussed,Referred to &/Or Reviewed at 970114,15 & 17 NRC Semiannual Meeting.Plant Performance Overview Encl

ML20141L864
Person / Time
Site:	Point Beach
Issue date:	03/17/1997
From:	Dums D CITIZENS' UTILITY BOARD
To:	Racquel Powell NRC OFFICE OF ADMINISTRATION (ADM)
Shared Package
ML20141L862	List: IA-97-110, Final Response to FOIA Response for Documents.Records Subj to Request Encl & Identified in App a ML20141L860 ML20141L864
References
FOIA-97-110 NUDOCS 9706030045
Download: ML20141L864 (35)

Text

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CITIZENS' UTILITY BOARD l

+ 16 N. Carroll St., Suite 300 + Madison, WI 53703 . (608) 251-3322

• FAX (608) 251-7609 - i l

March 17,1997 CaseNtr D//0 -

i gg .S- 42/ 'l 7-Mr. Russell Powell Aden 09: Nd t a Branch Chieff0lA/LPDR pew rue: -

Nuclear Regulatory Commission l Washington, D.C.

20555 Re: Freedom ofInformation Act Request ]

Dear Mr. Powell:

1 Pursuant to the Freedom ofInformation Act,5 U.S.C. 522, as amended and 10 C.F.R.  ;

Part 9.41, the Wisconsin Citizens' Utility Board (CUB) requests the following documents:

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1. Please provide any and all documents that refer to the operation and/or management of the  ;

Point Beach nuclear power plants, Dockets 50-266 and 50-301, that were disseminated, discussed, referred to, and/or reviewed at the NRC's January 14,15, and 17th,1997 semiannual meeting that was held to determine whether the safety performance of various licensees warrants increased NRC attention;

2. Please provide any and all documents created after April 30,1996 that discuss and/or refer to the performance of Wisconsin Electric Power Company's management level employees.

For purposes of this request, please consider " documents" to include reports, studies, test l results, correspondence, memoranda, meeting notes, meeting minutes, working papers, graphs, charts, diagrams, notes and summaries of conversations and interviews, computer records, and any other form of written communications including internal NRC memoranda.

1 Pursuant to and in compliance with 10 C.F.R. 9.41 of the agency's regulations and 5 l U.S.C. 552 governing ren,uests for waiver of fees, CUB puts forth the following information.

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CUB was incorporated pursuant to Chapter 181,1979, of the Wisconsin Statutues. CUB is exempt from Federal income tax under Section 501(c)(4) of the Internal Revenue Code. CUB is also exempt from Wisconsin franchise tax.

The restricted fund (public interest fund) is exempt from Federal income tax under Section  !

501(c)(3) of the Internal Revenue Code. The fund is also exempt from Wisconsin income taxes. l The public interest fund was established on February 1,1981, and reorganized on April 28,1986

! i 9706030045 970529 PDR FOIA DUMS97-110 PDR 0C=-

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to correspond to the new structure of the CUB. The fund engages in charitable, scientific, literary and educational activities on behalf of CUB.

CUB has 20,000 members who are residential ratepayers in Wisconsin. CUB represents the more than two million residential ratepayers in the state. CUB's purpose continues to be the statutory purpose set forth under sec.199.02, Wis. Stats.:

..to promote the health, welfare and prosperity of all the citizens of this state by ensuring effective and democratic representation of individual farmers and other individual residential utility consumers before regulatory agencies, the legislature and other public bodies and by providing for consumer education on utility service costs and on benefits and snethods of energy conservation.

CUB seeks the requested information to become informed of the causes of the extended outages ofPoint Beach Unit I and Unit 2. CUB intends to use the information to assist it in its participation in any proceedings before Wisconsin regulatory agencies, and as part of CUE's ongoing effort to educate the public on nuclear energy issues in Wisconsin.

The information sought, is not, to the best of our knowledge, in the public domain. The general public in Wisconsin has displayed great interest in nuclear energy issues and have a direct interest in NRC actions regarding the operation of the Point Beach nuclear plants.

CUB has demonstrated its ability and commitment to inform the public on all important nuclear energy issues in Wisconsin. CUB provides this information free of charge through newsletters, educational pamphlets, and correspondence to its members, other residential ratepayers, legislators, and policy makers, and has neither a commercial nor a private interest in the agency records sought.

Under the amended fee waiver standard, CUB is clearly entitled to a full waiver of all search, review and duplication fees. This standard calls for such a waiver, "if disclosure of the information is in the public interest because it is likely to contribute significantly to the public understanding of the operation or activities of the government and is not primarily in the commercial interest of the requester." 5. U.S.C. 552(1)(4)(A)(iii).

In light of the foregoing, CUB meets this standard on its face. CUB has no commercial interest in this matter, but rather seeks this information to help the general public and state regulatory agencies better understand the role of the federal government in regulating the nuclear power industry.

For the reasons cited above, CUB's request falls squarely within the Congressional intent h enacting the Freedom ofInformation Act and the fee waiver provision. CUB, therefore, asks that the NRC grant a full fee waiver for this FOIA request.

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Thank you for your anticipated cooperation. If you have any questions in regards to this request, please feel free to contact me at your earliest convenience. Please contact me before acquiring and sending the requested information if the fee waiver is not applied.

Sincerely, i

Dennis Dums Research Director 3

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PRE-DECIS10NAL POINT BEACH NUCLEAR PLANT SALP Period

April 28, 1996 through November 22, 1997 Previous SALP Rating: Operations 2 Maintenance 2 Engineering 2 Plant Support 1 I. PERFORMANCE OVERVIEW Point Beach's performance is substantially worse than previously recognized.

This is due to longstanding problems that have only recently come to the NRC's attention. Serious regulatory concerns have been identified that have resulted in numerous violations and a significant enforcement case. These concerns cover a broad range of activities and involve fundamental weaknesses in the functional areas of Operations, Maintenance, and Engineering. The primary cause of these performance issues appears to be a deep rooted plant culture that has focused on keeping the units operating without questioning or resolving long-standing safety issues.

Although the licensee provided the NRC with a large number of proposed corrective actions to address specific concerns, many of the corrective actions completed to date have been ineffective. In addition, little has been done in the way of self or independent assessments of plant activities to fully bound the magnitude of performance issues. As a result, the NRC 1 continues to find significant new issues that the licensee has not identified or addressed. These include problems with adherence to administrative procedures, knowledge of regulatory requirements, poor material condition of safety-related components, weaknesses in the surveillance program, and weaknesses in the quality of safety evaluations and operability determinations. In addition, the NRC has concerns with the licensee's management of the plant's configuration in accordance with the FSAR and design bases. For example, several significant technical interface agreements (TIAs) have been sent to NRR describing design-related issues that could impact the startup of Unit 2 and the continued operation or restart of Unit I should the unit shutdown in the near future. In addition, a recent OSTI inspection identified a large number of design bases document (DBD) open items that had been open two years or more and were not entered into the licensee's corrective action program and were not formally evaluated for plant operability concerns.

At the screening meeting, it was decided that Point Beach Nuclear Power Plant would be discussed at the SMM because of the significance of the identified issues, and the licensee's lack of a strong safety-focused questioning attitude, and an aggressive pursuit to identify the full extent of performance issues at the plant.

During the September 12, 1996, enforcement conference that discussed 16 apparent violations, the licensee made a number of specific commitments to address NRC concerns. On October 5, 1996, the licensee started its Unit 2 steam generator replacement outage. On December 5, 1996, during a management

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l POINT BEACH PRE-DECISIONAL nonconservative natre :cI thvse 'achs, eg:hwerisq idFatified this as non-conservative and 'the Wmnwt tw4 sm te Mr.Ugate operations actions, and the reastw fur tie .pwra a hu ; degraded ;from a previous 20-30 psi margin. (IR 9R03; P1Y iih &rbed TSf%95)

C. Plans In addition to the core resident i,nspection program, a substantial regional ~

initiative effort will be focused in the area of operations. See Section III for details on RIII's special Point Beach Oversight Team and inspection plan.

Maintenance A. Assessment Long-standing deficiencies in the licensee's surveillance and post maintenance .

testing program have led to a decline in performance in the maintenance area.

B. Basis

• Design basis requirements had not been incorporated into the inservice testing (IST) program for several safety-related pumps. (IR 96006; PIM item dated 8/15/96)

• The licensee did not adequately control or maintain gages used for IST of the safety injection pumps. This resulted in increased out-of-service time to test the pumps. (IR 96002 & IR 96006; PIM item dated 8/15/96)

• On September 13, NRC inspectors identified that, upon the completion of maintenance in April, a maintenance work request sticker had not been removed from the control switch for the outlet control valve for the component cooling water return line from the "B" train reactor coolant pump thermal barrier. This provided a false indication to operators that equipment was not available for use. A similar issue had been identified by the inspectors and documented in Inspection Report 50-266/301-95011. (IR 96008; PIM item dated 8/9/96)

• The inspectors identified that the licensee returned an SW pump to service with IST results in the alert range, contrary to the l requirements of Section XI of the ASME Code. (IR 96003; PIM item dated i 12/22/95)

C. Plans j In addition to the core resident inspection program, a substantial regional initiative effort will be focused in the area of maintenance. See Section III for details on RIII's special Point Beach oversight team and inspection plan.

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POINT BEACH PRE-DECISIONAL Enaineerina A. Assessment Overall engineering performance has declined. The dec ine is mainly due to lack of a questioning attitude and rigor when performing and documenting engineering evaluations.

B. Basis

• The Augmented Inspection Team that was dispatched following the unanticipated dry cask hydrogen burn event identified several concerns ,

with dry cask storage activities. These concerns reflect a weakness in  !

the licensee's ability to evaluate conditions outside the design basis of ttn VSC-24 cask as defined in the associated Safety Analysis Report.

(Ik 90005 & 96006; PIM items dated 6/7/96 and 5/28/96)

• Several instances were identified where lack of a questioning attitude and/or lack of engineering rigor by Engineering Department personnel resulted in performing inadequate safety evaluations and operability determinations or failing to perform a required safety evaluation:

The licensee did not document interim operability evaluations to address whether the IST program acceptance criteria for safety-related pumps adequately bounded design assumptions in the FSAR.

This was a followup issue to the escalated enforcement action involving IST of the safety injection pumps. The licensee documented these operability evaluations only after extensive NRC involvement. (IR 96013; PIM item dated 10/3/96)

The licensee did not take proper corrective actions after identifying that the number of SW pumps required to mitigate a design basis accident was greater than the number assumed in the FSAR and Technical Specification bases. (IR 96006; PIM item dated 7/1/96) ,

On September 16, the licensee identified that the containment fan coolers may not meet design requirements in the Technical Specifications with the current cooler discharge throttle valve position. In addition, the inspectors identified several weaknesses with the associated safety evaluation regarding the ability to take a post-accident containment air sample and bound containment integrity analysis assumptions in the FSAR for offsite dose. (IR 96008, 96013; PIM item dated 9/16/96)

• The licensee failed to identify that the installation of a temporary flange upon removal of an atmospheric steam dump valve constituted a temporary modification and thus required a safety evaluation. (IR 96004; PIM item dated 5/31/96)

• On November 12, during a review of the results of recent testing on the turbine-driven AFW pump, NRC in- tors noted that the test procedure 5

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POINT BEACH PRE-DECISIONAL did not incorporate recenthy ruised acceptance crnwia. The use of incorrect acceptance criteria in testing was one of the subjects

' discussed at the September 12th 'Predecisiona'l Inforcesent Conference.

(IR 96015; PIM item dated 11/12/96)

• On Nov 19-20, NRC inspectors observed the test of the Unit 2 Train A emergency diesel generator (the G02 diesel). The test required that.the air start motors actuate in a sequence specified in the FSAR. The proper sequence was not obtained on the first attempt, but was obtained on the second attempt. After questioning by the residents, the licensee wrote an operability evaluation to address the initial failure.

The evaluation, however, was superficial and did not adequately discuss the basis of the air start motor sequence. Inadequate operability determinations was also discussed at the Conference. (IR 96015; PIM dated 11/20/96)

C. Plans  !

In addition to the core resident inspection program, a substantial regional ,

initiative effort will be focused in the area of engineering. See Section III  !

for details on RIII's special Point Beach oversight team and inspection plan.

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Plant Support A. Assessment i

Plant support performance was good overall, but some problems were' identified during this assessment period. Problems with radiological work practices were identified during dry cask storage operations and during the steam generator replacement project. Problems with high radiation area (HRA) controls continued, resulting in a violation. Chemistry and radiological environmental monitoring programs conti'nued to be excellent. Performance of the security and emergency preparedness programs was good.

B. Basis

• A HRA barrier violation occurred when RP personnel failed to perform an adequate search of an area before it was posted and controlled as an HRA. This violation was another example of continuing problems with control of HRAs. (IR 96009; PIM item dated 9/25/96) )

• Weaknesses were identified with radiation worker practices. Although these practices did not result in the spread of contamination, they are indicative of a continuing problem with poor personal contamination control. (IR 96004; PIM item dated 5/20/96)

C. Plans Continue with the core inspection program as scheduled.

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POINT BEACH PRE-DECISIONAL III. Future Inspection Activities A. Inspections A dedicated 0versight Team has been established to thoroughly assess Point Beach's performance and the effectiveness of its corrective actions to identified issues. This team will include a third resident that is being t.ssigned to the facility and additional onsite support from DRS engineering staff, the DRP project engineer, and various visiting residents from the better performing RIII sites. This enhanced Point Beach Oversight Team, will concentrate on a number of critical focus areas.

To better define current performance issues and to assess the licensee's actions to date, a 10-person OSTI was performed December 2-20, 1996. This effort was led by a Branch Chief and focused on the areas described below.

Preliminary OSTI findings included the identification of a large number of examples of nonconservative operations and engineering decisions and practices that have resulted in several significant violations. In addition, during the team's review of a list of 93 DBD open items that were two or more years old, they identified that most had not been entered into the licensee's corrective action program; therefore, no operability evaluations were performed on conditions that were outside the design bases and FSAR. Because these open items may have safety significance and could affect the continued operation of Unit 1, the OSTI extended its inspection by at least a week (through 12/20 or later) and moved to the corporate engineering office to perform a detailed review of the licensee's actions associated with the open items. Results of I

this expanded inspection were not complete at the time this paper was written.

l l The level of effort and timing of subsequent followup inspections of OSTI l findings and corrective actions to enforcement issues will be dependent on the licensee's efforts to correct identified problems and perform independent assessments.

Inspection focus areas include:

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• Operations Performance, includes an IP 93802 OSTI evaluation of I

operations performance. -December 2-13, 1996 1

• Surveillance / Post Maintenance Testing, includes an assessment of licensee's M&TE program, procedure and work package adequacy, observations of work in progress, restoration of systems, etc. -December 2-13, 1996

• Maintenance, includes a review of FME program, maintenance procedures, engineering support, material control, and material condition of plant.

-December 1996 a Engineering, includes a 50.59 assessment, modification package review, maintenance package review, operability review, and IST performance review. -December 2-20, 1996 7

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POINT BEACH PRE-DECISIONAL

• Quality Assurance, includes a focussed 1P. 40500 irrepection of corrective action program. -January / February 1997

• FSAR 'lalidation Assessment, includes .a hardware and software assessment..

-December 2-20, 1996

• Steam Generator Replacement. -December 1996

• Followup inspection of Dry Cask Storage program. '

I B. Other Activities- 1

• A' management meeting was held December 5, 1996, to discuss Point Beach's process improvements for conducting operability evaluations, status of facility performance improvements, and issues required to be resolved prior to restarting Unit 2. Similar monthly management meetings are planned for the foreseeable future.

• Extensive followup inspections will be required to address currently identified issues and issues identified as a result of the December 2-20 OSTI.

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POINT bet'H PRE-DECISIONAL l l

DATA

SUMMARY

I. PRA A. PRA Insichts i The Point Beach site contains two units, each of which is a two-loop i Westinghouse with a large, dry containment. The Point Beach units share a number of systems such as AC and DC power, service water, instrument air, and motor-driven AFW pumps.

To cope with station blackout (SBO), Point Beach has four EDGs shared between ,

the two units and a 20 MW Gas Turbine Generator (GTG). Two of the four EDGs 1 were added recently. The IPE did not account for the two new EDGs in the CDF estimate since these EDGs were added after the IPE cut-oft ' ate; however, a sensitivity analysis in the IPE indicated that the addition of these two EDGs decreases the CDF by about 10%. The GTG, which is the alternate AC source for SB0, was modelled as having a 65% reliability in the IPE. However, the reliability may be higher as estimated by the licensee in a letter dated May 14, 1993 which indicated that the reliability of the GTG increased to about 91%.

While SB0 risk is decreased by the addition of two EDGs and better GTG reliability, the batteries have only a one hour lifetime after SB0. The one hour battery lifetime requires special operator actions to provide turbine-driven auxiliary feedwater (TDAFW) for four hours during a SB0. The TDAFW system has, in addition to the dependency on batteries during SB0, a dependency on the diesel driven fire pump during SB0. According to the IPE, the diesel driven fire pump is required after one hour to supply cooling to the bearing coolers of the TDAFW pump.

The IPE reported that failure to provide AFW cooling as a means of decay heat i removal accounted for approximately 38% of the total CDF. The most important I aspect of providing long-term AFW cooling is providing a long term source of suction for the AFW pumps. The condensate storage tanks will be depleted in approximately four hours, and operator action is required to manually align the pumps to an alternate suction source. A sensitivity study in the IPE showed that if the reliability of operators to manually align a long-term suction supply to the AFW pumps was 95%, then the CDF would increase by one I order of magnitude, and, if this action was only 45% reliable, then the CDF l would increase by two orders of magnitude.  !

Also, for LLOCA and MLOCA, operator failure to align the emergency core cooling system for containment sump recirculation accounted for about 30% of l the total CDF. This operator action requires many local manual actions, l according to the IPE.

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POINT BEACH PRE-DECISIONAL B. PRA Profile On June 30, 1993, in response to GL 88-20, Point Beach submitted their IPE.

On December 16, 1994, RES completed their review and found that the licensee had met the intent of GL 88-20. The total core damage frequency, including internal flooding, is 1.15E-4/Rx-yr.

The IPEEE was submitted to NRC on June 30, 1995, and reported a CDF from external events of 6.5E-5/yr for each unit. Completion of review for the IPEEE is scheduled for June 30, 1997.

Initiators that contribute the most to the CDF in the IPE and the percentage they contribute are:

Initiator  % CDF Large LOCA 22 SB0 13 Transient w/o PCS 10 Internal Flooding 9

' ' Medium LOCA 9 LOOP (excluding 5B0) 8 Loss of SW 7 SGTR 5 Transient w/ PCS 5 Loss of CCW 4 l l

The RCP seal LOCA contribution to CDF is 8%, with only 0.3% coming from SBO.  !

This is because the IPE allowed recovery of seal LOCA during SBO, but not during other initiators.

l Twenty-two percent (2.55E-5/yr) of the total CDF represents dual unit core I damage. Ninety-eight percent of this contribution stems from SB0 (1.4E-5/yr) and flooding (1.1E-5/yr). Naturally, dual unit core damage is reduced due to the installation of the two additional EDGs.

The IPE states the following contributions to total containment failure given core damage:

l No Containment Failure 77%

Late Containment Failure 17% l Containment Bypasses 6% l Containment Isolation Failure <1%

Early Containment Failure 0%

i The licensee calculated a fission product release frequency (FPRF) of 2.4E-  :

5/Rx-yr. The largest contributors to FPRF were transients and SB0 sequences, I which would cause the containment to overpressurize and fail at 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. l These initiators contribute 74% of the FPRF. The licensee states the calculated FPRF is conservative.

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l POINT BEACH PRE-DECISIONAL l The CDF profile from external events was determined to be as follows in the IPEEE: fires (79.2%); seismic (20.3%); high winds (0.5%); and accidents at nearby industrial facilities, transportation, and external flooding each contribute <1%.

C. Core Damaae Precursor Events On the basis of the precursors identified by ORNL for 1994 (NUREG/CR-4674 Vols. 21 and 22) and the preliminary precursors for 1995 and for 1996, the staff identified the following event that has a conditional core damage probability (CCDP) of IE-5 or greater.

This event was determined to be a "Significant Event" for the Performance Indicator Program:

On February 7,1994, one of the two shared EDGs (the two additional EDGs were not installed at this time) was declared inoperable due to service for maintenance. Both units were operating at 100% power. Required operability testing of the remaining EDG commenced that day, and the electric fuel pump was discovered failed. The diesel ran unloaded while repairs were made to the electric fuel pump. After repairs were made, the EDG was shut down, and then restarted and loaded. Fourteen minutes later, erratic exciter performance which increased in intensity was observed. Ultimately, the EDG was declared inoperable. A stationary ,

brush jumper cabie in the EDG's exciter was found to be contacting a rotating bus bar, shorting out the DC excitation voltage. This condition was repaired, and the EDG declared operable on February 9, 1994. An ASP study was performed by AEOD where it was assumed that the EDGs could not have supplied long-term emergency power for 47 hours5.439815e-4 days <br />0.0131 hours <br />7.771164e-5 weeks <br />1.78835e-5 months <br />, from the moment the first EDG was taken out of service due to maintenance till the time the exciter on the other EDG was repaired.

(However, it is possible that the EDG could have run successfully for an extended time in a loaded configuration.) The estimated CCDP associated with this event at each unit is 1.2E-5.

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POINT BEACH PRE-DECISIONAL II. ENFORCEMENT HISTORY 10/95 -SEVERITY LEVEL III VIOLATION (Sucolement III. Syfeauards: EA 95-158):

The action nos vased on the failure to properly protect or control safeguards information on July 5, 1995.

Because this was the first escalated action within 2 years, the NRC considered whether credit was warranted for corrective action. Credit was warranted for prompt and comprehensive corrective action, thereby fully mitigating the civil penalty.

12/96 - CIVIL Reactor PENALTIES AND EXERCISE OF ENFORCEMENT DISCRETION (Suoplement I_1 Operation: EA 96-215 and EA 96-273): The action was based on four Severity Level III issues identified during two inspections that took place in September 1996. Three Severity Level III problems were cited for the failure to adequately: 1) conduct control room activities; 2) maintain plant configuration control; and 3) conduct independent fuel dry cask storage activities. In addition, a single Severity Level III violation was cited for the licensee failing to take prompt corrective actions following identification that the technical specifications for the safety-related service water system were non-conservative. Multiple civil penalties were

' issued to emphasize the need for full compliance with NRC regulatory requirements. Because this was not the first escalated action within 2 years, the NRC considered whether credit was warranted for identification and corrective action. Credit for identification was not warranted because the NRC identified all of the violations. Credit was not warranted for corrective action associated with the three reactor operations issues, thereby resulting in twice the base civil penalty in these cases. Credit was warranted for the corrective action for the Severity Level III problem associated with the independent spent fuel installation. Although the base civil penalty of 1

$12,500 would normally be issued because credit was warranted for corrective {

j action, the staff exercised discretion and issued double the base civil  !

penalty to emphasize the importance of properly conducting spent fuel cask loading operations. ($325,000) {

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• DRAFTIREGIONAL REVIEW Legend:

5""'d "" < .,,, x. 72 hrs ! St., tun , ,

POINT BEACH 1 Refueling R Operation ==<a industry . Avg. Trend Shutdown N O Y Not Shown Using Op. Cycle E35335 ops. ops.

! it F, I It It F, It i I

' 94 1 94-3 %1' %3' 51' 96 3 6

1. 6

%1 '94 3' %1 %3- %1 W Year - Quarter Year - Quarter Automatic Scrams While Critical ,

Safety System Actuations 4

! a g

! 3- g 3-m

• M o 2- . 2-o

'~ $5 '~

=

h z h h i 941i 94-3' 51' %3 '~ % 1' 53'~ ' W1' 94 *- Wi i %3' %13 96 3 Year - Quarter Year - Quarter Significar.t Events , Safety System Failures 2.0

$ 3-

], 1.5 - ,,

01.0-i) 2 og  ?

b 0

z '5 ~ g l 1~

g' g' M' M' 0.0 0 EE E D' I

94 16 94-3 i

%16 %3'  % 1' 9M

• 94 1' 94 3 i %1' %3' bli 53' Yes - Quarter Year - Quarter Equipment Forced Outagest Forced Outage Rate P/.) 1000 Commercial Critical Hours 3on 7

7 80-16 f ~

60 Th I- E b #~ 0 2

1 g20- - PA

u. "

e -

0 w 0 i 9 4-15 it.4.y Wii W3' %16 %3' ' S4-1' %31 E in %3' Wil W3' Year- Quarter Year- Quarter Cause Codes

a. Admin b Uc Oper c. Other' Per Collective Radiation Exposure a a n 200 ,

m.

y m. m.

E $0- s. s s.

u n-, S . ,, . ..m i g 3,o_

d. Maint e. Design f. Misc E " *

• 50- w -

a w

.m w 0 i 1 S wy g36 b'gy *-

w g, gs.3 i s. s. s.

9436

,"', Year- Quarter ..@ - -

• - N *, " " "

Site Average Radiation Exposure

_- __ _ _ _ _ _ . ._..m._ . _ _ _ . . - . . . _ _ . -_ _ _ _ __. ._._..__._m.,_.-_..__..._ _ . _ _ .

DRAFT **EGIONAL REVIEW .

POINT BEACH 1 '9ad st===$c='S;+ r- w High 1glllgll

, Peer Group:Wesangtouse Small Medium geen 93-4 to 96-3 *' '

Trends and Deviabons i

' Deviations Frem Plant Peer Group

Self-Trend Median

Short Term Long Term Declined improved Worse Better

{. OPEPATIONS (including startup) 1- )

J

Automatic Scrams While Critical - 0 -

0 I l

Safety System Actuations - 0 -

0.90 l 4

j Significant Events - 0 -

0 i

f Safety System Failures - gg. -0.72 -

0 Cause Codes (All LERs)

a. Aammisesave ContrelPreWees - 0 -

O

t. ucensea o,erseer arrers -

]0.11 -

O c.comrPasonnaisevers - 0 -

0-

d. heamennance Pressoms - )0.06 -

E 0.38

e. r - -- -

earisemen Preunam - -0.25 [ -

0.60 l l r.acessemaneus - 0 -

l0 Safety System Actuations - 0

.o.90 Significant Events - 0 0

Safety System Failures - 0 0

Cause Codes (All LERs)

a. Amedeemerseve Cosered Premamens - 0 -

-0.43 E

h. usessed opermeer errors - 0 -

O c.osserParnesselervers - 0 -

-0.37 I I

a. - Pressoas - 0 -

-0.27 e.

Pressens - 0 -

4.07 g, - - - - o - 0 FORCED OUTAGES Forced Outage Rate * - o 0

Equipment Forced Outages /

• _ _

0.26 1000 Commercial Critical Hours _-.

1.0 -0!5 0.0 o'.5 1.0 1.0 -0!5 O!0 O!5 1.0 )

Performance index Performance index s

• eene Casseness ser operassear cyce.

DRAFT /RE2IONAL REVIEW POINT BEACH 1 PI EVENTS FOR 95-4 NONE ,

PI EVENTS FOR 96-1 NONE PI EVENTS FOR 96-2 SSA 04/05/96 LER# 26696001 50.72#: 30237 PWR HIST: REFUELING DESC  : A LPSI PUMP STARTED ON A SPURIOUS SAFETY INJECTION SIGNAL. THE CAUSE WAS LINKED WITH THE INVESTIGATION OF A GROUND ON THE DC SYSTEM. AN INADEGUATE GROUND SEARCH TECHN!QUE WAS USED. THERE WAS No ECCS INJECTION INTO THE RCS.

SSF 06/21/96 LER# 26696003 5 0.72#:

PWR HIST: CONDIT!0er EXISTED FOR AN INDETERMINATE PERIOD OF TIME

~ GROUP LOW TEMPERATURE /0VERPRESSURE PROTECTION GROUP SYSTEM : LOW TEMP /0VERPRESSURE PROTECTION SYSTEM OTH UNIT: THIS EVENT WAS AS$1GNED TO UNITS 1 AND 2.

DESC  : PLANT OPERATING PROCEDURES ALLOWED OPERATION WITH TWO HPS! PUMPS AVAILABLE DURING LTOP CONDITIONS.

THE DISCHARGE OF TWO HPSI PUMPS WOULD HAVE RENDERED LTOP INCAPABLE OF PERFORMING IT'S SAFETY t FUNCTION.

I PI EVENTS FOR 96-3 SSF- 08/14/96 LER# 26696007 50.72#: 30874

'PWR HIST: CONDITION EXISTED IN ALL MODES UP TO 100% POWER SINCE INITIAL OPERATION GROUP AUXILIARY / EMERGENCY FEEDWATER SYSTEMS GROUP l SYSTEM : AUXILIARY / EMERGENCY FEEDWATER SYSTEM DESC  : A SINGLE CONDITION COULD HAVE RENDERED ALL THREE AFW PUMPS INOPERABLE.' PHYSICAL SEPARATION OF REDUNDANT SAFETY-RELATED CIRCUITS INSIDE THE MAIN CONTROL BOARD WAS INADEQUATE TO PROTECT THE CIRCUlfS FROM ELECTRICAL FAULTS.

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Cause Codes (All LERs)

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DRAFT /RESIONAL REVIEW ,

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'PI EVENTS FOR 95-4 NONE PI EVENTS FOR 96-1 l

NONE PI EVENTS FOR 96-2 SCRAM 05/18/96 LERf 30196001 50.72#: 30492 PWR HIST: POWER OPERATIONS AT 90%

DESC  : A RX TRIP OCCURRED ON A TURBINE TRIP. A SOLEN 0!D VALVE ON THE TURBINE ELECTR0 HYDRAULIC FLUID SYSTEM SPURIOUSLY OPENED CAUSING THE TURBINE STOP VALVES TO CLOSE.

SSF 06/21/96 LER* 26fE003 50.72#2 PWR HIST: CONDITION EXISTED FOR AN IEETERMINATE PERIOD OF TIME ,

GROUP : LOW TEMPERATURE /DVERPRESSUPI PROTECTION GROUP l SYSTEM : LOW TEMP /0VERPRESSURE PROTECTION SYSTEM l OTH UNIT: TNIS EVENT WAS AS$1GNED TO UNITS 1 AND 2. ]

DESC  : PLANT OPERATING PROCEDURES ALLOWED OPERATION WITH TWO HPSI PUMPS AVA:LABLE DURING LTOP CONDITIONS.

THE DISCHARGE OF TWO HPS! PUMPS WOULD HAVE RENDERED LTOP INCAPABLE OF PERFORMING 178S SAFETY FUNCTION.

PI EVENTS FOR 96-3 NONE ,

1 v'

PLANT ISSUES LIST 27-Nov-96 Point Beach DATE ID BY SALP DESCRIPTION CAUSE REF 11/20/96 NRC Engineering He proper sequence (from the FSAR) for the air start motors for the Inadequate IR 96015 Unit 2 Train A emergency diesel generator was not obtained on the first Oversight try during a surveillante, but was obtained on the second try. No operability evaluation was written until the residents raised questions.

De subsequent evaluation was superficial and did not adequately discuss the basis of the air start motor sequence.

11/15/96 Licensee Enf ming An auxiliary operator re-read a pressure gauge resulting in a successful Inadequate IR 96015 survei!!ance of a motor-driven AFW pump. In addition, Oversight nonconserva@mly, the pump was not declared inoperable during the ,

~

subsequent investigation of why the pressure margin had degraded.

II/12/96 NRC Engineering Procedure used for a test of a turbine-driven AFW pump had not been inadequate IR 96015 revised to incorporate recently changed acceptance crite.;a. His was an Oversight example ofinadequate follow-through by the licensee to commitments made at the 9/12/96 enforcement conference.

10/3/96 NRC Engineering From a review of the licensee's safety evaluation, " Containment Integrity Enineering/Desi 96013 Evaluation with reduced fan Cooler and Containment Spray gn Deficiency Performance," the inspectors identified 6 concerns related to containment pressure profiles, PASS, CFCs, and the use of RHR/CCW for post-accident heat removal. Dese concerns were relayed to NRR via a TIA.

10/3/96 NRC Engineering He licensee did not document interim operability evaluations of safety- Inadequate IR 96013 related pumps that had been completed to ensure IST program Oversight acceptance criteria for the pumps bound FSAR design assumptions.

9/25/96 Licensee Plant Support kP personnel did not adequately verify that no unauthorized personnel Personnel IR 96009 were present within an area once a high radiation area boundary was Performance erected and prior to allowing a spent resin transfer to occur. His Deficiency resulted in a high radiation area control violation as an unauthorized individual was located within the area while the resin transfer occurred.

9/16/96 Licensee Engineering The containment fan coolers may not meet TS design requirements with Enineering/Desi IR 96008 the current cooler discharge throttle valve position. gn Deficiency Page 1 of 8

PLANT ISSUES LIST 27 uov.96 PJ i~ t Be~ch DATE ID BY SALP DESCRIPTION CAUSE REF 9/1666 NRC Engineering PASS requirements of NUREG-0737: The ability of the PASS to Enineering/Desi IR 96008 function during an accident is questionable given a new containment gn Deficiency pressure analysis using 1.5 fan coolers and I containment spray pump.

8/23/96 Licensee All . An all-day staff standdown was held to discuss declining station Personnel performance and to propose methods for improvement. IR 96003 Performance Deficiency 8/20/96 NRC Operations Barely audible control room annunciators and failure oflicensed Personnel IR 96007 operator to respond to an alarm until prompted by another operator. Performance Deficiency 8/20/96 NRC Engineering Post-maintenance and inservice testing inadequacies were identified for Inadequate IR 96010 the containment fan coolers, spent fuel heat exchangers, and service Procedure /

water pump P-32E.

Instruction 8/16/96 Licensee Engineering Licensee reported that indication and alarm circuit cables from the Ul Enineering/Desi IR 96010 TDAFW pump recirculation valve to the main control board (MCB) ran gn Deficiency in the same wireway as the control power circuits for the MDAFw pumps. The MDAFW pumps are shared beween both units. A potential fault in the TDAFW pumps could afTect control power to the MDAFPs, disabling all three UI AFW pumps. Afler operaiblity determinations and circuit mods, the licensee declared a!I three pumps operable.

8/15/96 NRC Maintenance Deficiencies were identified regarding the incorporation of design basis Inadequate IR 96006 requirements into the inservice testing program for the Si pumps, and Procedure / Instr with the proper calibration of measuring and test equipment for these uction pumps.

8/15/96 Licensee Operations Duty Technical Advisor alloived to leave the site while remaining within inadequate IR 96007 10 minutes of the control room. TS requires the DTA to be located on- Procedure / Instr site within 10 minutes of the control room. uction 8/13/96 Licensee Operations Licensee removed the UI SW pump from service before returning U2 Personnel IR 96007 "B" train EDO to service. Although this configuration was only in place Performance for approximately 30 seconds, it was not allowed by TS. Deficiency Page 2 of 8

- PLANT ISSUES LIST 27-Nov Point Beach DATE ID BY SALP _.

_ DESCRIPTION CAUSE

-REF 8/12/96 NRC Operations Operations swc.id well to unanticipated reduction in condenser Personnel IR 96007 vacuum due to fish infestation on traveling water screens.

Performance Deficiency *

, 8/9/96 Licensee Engineering Dry cask reflood system may be inadequate: Discrepancies were Enineering/Desi IR 96008 identified between the licensee's calculations on the number and size of gn Deficiency system heat exchangers and a vendor's calculations.

8/9/96 NRC Operations Violation for inadequate Procedures: OPS procedures allowed reactor inadequate IR 96008 power to reach 3.5% with only one reactor coolant pump and there were Procedure / Instr ..'

no procedures for assessing leakage from a patch on a service water pipe. uction 8/9/96 NRC Maint/Eng Violation for Failure to Follow Procedures: A work request sticker had Personnel - IR 96008 not been removed from a control switch in the main control room after Performance ,

completion of maintenance on a component cooling water valve; a Deficiency Condition Report was not written for the improper installation of a washer on an auxiliary feedwater pump; and tools were tied oft ,

improperly during work near the SFP.

! 8/9/96 NRC Engineering Possibly inadequate Appendix J testing ofcontainment isolation valves: .f Enineering/Desi

( One liquid waste and several heating / ventilation inboard containment IR 96008 I gn Deficiency i l' isolation valves were tested in the reverse-flow direction, which the -

licensee has not demonstrated is equivalent to testing in the normal-flow direction. j

! 8/8/96 NRC Maintenance Safety injection pumps discharge gage were within tolerance only once inadequate IR 96006 -

in twenty calibrations over the last 3.5 years. '

Procedure!!nstr uction i 8/8/96 NRC Operations On April 22, UI was taken critical with the TDAFW pump discharge inadequate IR 96006 ~!

flow path isolated, contrary to TSs. Also, the licensee did not adequately Procedure / Instr I control the operability status of the pump and associated discharge flow uction path after replacement of the governor valve stem. The licensee considered the pump operable after the replacement without performing any required PMT.

• Page 3 of 8 '

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r PLANT ISSUES LIST 27-Nov-96 P-i-t Be ch DATE ID BY SALF DESCRIPTION _ _ ____.._ _ _ ___

CAUSE REF 8/6/96 Licensee Operations Inner door of the UI containment upper hatch failed an airlock door seal Inadequate LER 96008 test. ne calculated leak rate was greater than the TS limit. TS required Procedure / Instr that the upper containment hatch should have been full pressure tested. uction Instead the inner door seal was repaired and another partial pressure test was performed with satisfactory results.

8/6/96 Licensee Operations Missed full pressure test of containment airlock.

Inadequate LER 96008 Procedure / Instr uction 8/3/96 NRC Maintenance In December 1992, the licensee installed four new Ashcron 1000-1500 Inadequate IR 96006 psig gages to measure SI pump discharge pressure during the quarterly ' Procedure / Instr inservice test of the high head SI pumps. The gages were listed as having uction an accuracy of 0.5% of span (+/- 2.5 psig). The gages have not performed well in this application. During calibrations over the last 3.5 years, the gages have been within the 2.5 psig tolerance only once in twenty calibrations or Sh !the time.

7/31/96 NRC Operations During routine tour, the inspector noted that the control room operator Personnel IR 96006 left his normal watchstation to,make cofTee with no designated relief, Performance contrary to plant procedure.

Deficiency 7/28/96 NRC Operations inadequate log entries regarding operability of the Unit I containment Personnel IR 96006 upper air-lock door led to confusion with the actual status of the door. Performance Licensee failed to declare the outer door out-of-service in the official Deficiency station log and failed to indicate that Unit I was in an LCO requiring .

verification that the inner door was locked closed once per 31 days.

7/15/96 NRC Operations On shift crew watching video tape in the control room with their backs to Inadequate IR 96006 the control board.

Procedurc/ Instr uClion 7/1/96 NRC Engineering Licensee did not properly conclude that an unreviewed safety question Personnel IR 96006 existed after identifying that the number of SW pumps required to Performance mitigate a design basis accident was greater than the number assumed in Deficiency ,

the FSAR and TS bases.

Page 4of 8

PLANT ISSUES LIST 27-nor-96 .

Point Beach DATE ID BY SALP DESCRIPTION CAUSE REF 6/21/96 Licensee Operations Plant operation outside of design basis of the LTOP system. Inadequate t

LER %003 Procedure / Instr uction 6/20/96 NRC Engineering Lack of timely corrective action for potential inability of SW system to Engineering /De IR 96006 meet accident design basis. First suspected on 1/17/96, but final sign Deficiercy resolution still not achieved.

i 6/7!96 NRC Engineering AIT identified deficiencies during dry cask hydrogen burn. Design Enineering/Desi

, IR %005 i information not included in procedures / instructions; inadequate safety gn Deficiency evaluation (10 CFR 72.48); cask unloading procedures not updated; and deficiency reporting system not used.

6/4/96 NRC Operations Licensee failed to perform a safety evaluation when Unit 2 atmospheric Enineering/Desi IR 96004 steam dump system was in a configuration difTerent than the FSAR.

gn Deficiency 5/31/96 NRC Operations Failure to follow iwosdw. on temporary modifications for the Personnel IR %004 installation / removal of a Unit 2 atmospheric steam dump blank flange. Performance Deficiency 5/28/96 Self-Revealed Engineering flydrogen gas ignition while welding shield lid on VSC-24 dry cask Enineering/Desi IR 96005 gn Deficiency 5/20/96 NRC Plant Support

' Weaknesses in FME and contaimination controls during dry cask loading Personnel IR 96004 Performance Deficiency l 5/18/96 NRC Operations Operators ioyssa well to turbine trip / reactor trip Teamwork / Skill IR %004 Level 5/18/96 Self-Revealed Operations Turbine trip / reactor trip due to relay failure Equipment IR 96004 Malfunction 4/29/96 Licensee Operations Safety injection pump flow meter isolated for unknown reason. No Personnel IR 96003 control function but did have EOP function. Performance Deficiency i

Page 5 of 8 l

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PLANT ISSUES LIST 27-Nov-96 ' '

Point B-~ch - '

DATE \

ID BY ~ SALP . DESCRIPTION CAUSE REF 4/22/96 Licensee Operations Reactor taken critical with turbine-driven AFW pump inoperable due to Personnel . IR%003 IR96006 both discharge valves being closed. ,

Performance  ;

Deficiency i 4/18/96 Self-Revealed Engineering "A" motor-driven AFW pump tripped on overcurrent during testing: Engineering /De

. lR 96003 .

EDG speed droop set too high for low load conditions caused elevated sign Deficiency frequency and higher than normal pump speed and flow. e

. 4/12/96 Self-Revealed Operations  !

Unanticipated radioactive gas release due to improper line-up of reactor inadequate - IR 96003 [

drain tank pump: no procedure for pumping down refuelirg cavity with Procedure / Instr -t this pump. Left open normally closed and locked reactor coolant pump discharge valve allowing path from holdup tanks to primary aux building uction , j sump to exhaust vent stack. i 4/5/96 Self-Revealed - Maintenance ESF actuation due to improper ground faalt search techniques on 125 Personnel IR NO3 -

VDC supply to instrument inverter (started EDGs and train B SW Performance pumps, low head SI pump, motor- drive i AFW pump, and containment Deficiency L isolation).

4/4/96 Licensee Plant Support Individual entered posted high radiation area without required dosimetry Personnel IR'5003 and RWP: Followed authorized individuals under roped boundary. Performance  !

Dird example in 13 months.

Deficiency e

3/31/96 Self-Revealed Operations Momentary lift oflow temperature over-pressure (LTOP) power Personnel IR 96003 operated relief valve (PORV): Operator inattention durirg simultaneous Performance i securmg of teactor coolant pump and lowering cooldown rate resulted in Deficiency  ;

reduced letdown and increased RCS pressure. E 3/21/96 Licensee Engineering 4 kV bus uif.q m;y to reactor coolant pump trip relays had never f inadequate IR %003 received functional testing as required by technical specificaitons. Procedure / Instr uction .

3/1/96 NRC Engineering Failure to update FSAR regarding known discrepancies with plant Personnel IR 96002 [

design or operations. Examples included minimum service water flow to . Performance containment accident fan coolers, operating average coolant temperature, Deficiency -  !

and service water temperature. i Page 6 of 8  !

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PLANT ISSUES LIST 27-Nov-96 ~

Point Beach DATE ID BY SALP DESCRIPTION CAUSE REF 2/12/96 NRC Operation, > allure to recognize previous safety evaluation required diesel generator Personnel IR 96002 ~

40 be declared inoperable when component cooling water pump Performance emergency power supply was unavailable.

Deficiency 2/9/96 NRC Engineering Failure to correct inadequate methodology for perfonnance testing of Personnel IR 96002 spent fuel pool heat exchanger since 1989 or take other actions to ensure Performance design performance per Generic Letter 89-13.

Deficiency 1/25/96 NRC Maintenance Unsecured scaffolding and other equipment in safe-shutdown area: Personnel IR 96002 Problems controlling portable equipment was long term weakness and Performance noted in 1994 and 1995 inspection repe-ts. Deficiency 1/22/96 NRC Maintenance Operators failed to note seal or pump leakage during RilP. pump Personnel IR96003 IR96002 surveillance although significant long tenn boric acid buildup impeded Performance observation. Noted again during testing on 3/23/96. Deficiency 12/22/95 NRC Maintenance Untimely or incomplete engmeering evaluation ofIST data. Allowed Personnel IR 96002 IR 96003 service water pump to operate longer than necessary in alert range Performance without correction and safety injection pump to be prematurely removed Deficiency from increased testing frequency.

12/12/95 Licensee Engineering Material transfer cask (MTCj load testing by vendor did not meet ANSI Inadequate IR 96002 N 14.6 as required by Certificate of Compliance. Oversight 12/1/95 NRC Maintenance FME controls and inspection signoff not in reactor head installation inadequate IR 95015 procedure as required by FME procedure. Procedure / Instr uction i1/25/95 Licensee Maintenance PORV post-maintenance testing not performed prior to establishing Personnel LER 95-006 LTOP. Performance Deficiency 11/7/95 NRC Operations Simultaneous ESF logic testing on both units. Personnel IR 95013 Performance Deficiency f 11/3/95 Self-Revealed Plant Support Two contract workers inadvertently entered high radiation area in Personnel IR 95013

} containment. Performance Deficiency Page 7of 8

PLANT IS3UES LIST - 37-Nov-96 i

Point Beach l

1 DATE ID BY SALP . _ _ ,

DESCRIPTION CAUSE REF

___ _ ._ _ . __ _ .. I 11/1/95 Licensee Es.hig Significant steam generator tube degradation results in extensive tube l Equipment plugging and re-roll repairs. _d 95015\LER 95-005 Malfunction i1/1/95 NRC - Maintenance Weak implementation of FME contro!s.

Personnel IR 95013 Performe.te Deficiency =

10/2695 Self-Revealed Maintenance Following calibration, techniciaa mis-installed time delay relay and - Personnel IR 95013\LER 95-004 caused auto-start of motor driven AFW pump. Performance Deficiency 10/12/95 Self-Revealed Operations i 40 gallon reactor coolant spill due to RCP seal vent valve not properly Personnel '

isolated. IR 95013 Perfonnance Deficiency 10/9/95 Self-Revealed Maintenance Unit 2 reactor trip signal while shutdown due to IAC tech leaving one Personnel . IR 95013 channel in trip and taking another to test during surveillance. Performance '

Deficiency 9/12/95 Self-Revealed Plant Support Security x-ray operator fa2ed to detect a non-operating pistol during an Personnel IR 95011 unannounced NRC-requested. drill.

Performance Deficier:cy

-)

9/11/95 Self-Revealed Operations STRENGTH:

Teamwork / Skill IR 9501i Good operator response to power transient whea a turbine governor Level valve limiter failed.

r b

i Page 8 of 8 '

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ATTACHMENT 1 Summary of the January 1997 NRC Senior Management Meeting i

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

NRC Senior Management Meeting Summary January 14, 15 and 17, 1997 Region IV Following the June 1985 loss of feedwater event at Davis-Besse, '

one resulting NRC action was that senior NRC managers periodically meet to discuss the plants of greatest concern to the agency and to plan a coordinated course of action. The NRC senior managers held their twenty-second such meeting in Region IV.on January 14-15 and continued the meeting in NRC Headquarters on January 17, 1997. The last meeting was held in Region III in June 1996. The meeting in Region IV was structured to review the.

status of the Watch List plants identified at the last meeting and to review the performance of other plants to determine if any changes should be made to the list of facilities which require

) close monitoring by NRC.

In preparation for the meeting, NRR and NMSS, in conjunction with  !

~the four regional offices, AEOD, OE, and RES, prepared background k documents on the plants and materials licensees to be discussed.

Inputs for each operating reactor plant included a summary of the l most recent SALP and SALP history, a discussion of current operating experience, current NRC and licensee activities, and performance indicator data. Data pertaining to safety significant hardware issues at the plants were also provided.

This information was distributed to meeting attendees prior to the meeting. It provided the basis for review and discussion of l

each plant's performance and for eenior management identification of those plants warranting increased NRC attention.

In reviewing the reactor plants that have experienced significant j performance problems, the NRC managers have set the following I

categories of performance based upon plant actions to date to correct the problems and to achieve improved operations.

j 1. Plants removed from the list of problem facilities.

Plants in this category have taken effective action to l correct identified problems and to implement programs for improved performance. No further NRC special attention is f necessary beyond the regional office's current level of i

monitoring to ensure improvement continues.

l 2. Plants authorized to operate that the NRC will monitor closely.

Plants in this category have been identified as having weaknesses that warrant increased NRC attention from both headquarters and the regional office. A plant will remain in this category until the licensee demonstrates a period of  !

improved performance. I l s i

i i

l

3.  !

l Shutdown plants requiring NRC authorization to operate and Which the NRC will monitor closely.

l Plants in this category have been identified as having significant weaknesses that warrant maintaining the plant in a shutdown condition until the licensee can demonstrate to the NRC that adequate progran.s have been established and implemented to ensure substantial improvement.

The following charts list conclusions reached by the senior managers e.t this meeting and from the previous meeting for nuclear power plants and for materials licensees:

Meetina Dates Cateaorv 3 Cateaorv 2 l Catecorv i i JAN 14-15, 1996 Millstone Crystal River 3 1,2&3 Dresden 2&3 i Indian Point 3 Lasalle 1&2 Maine Yankee Salem 1&2 Zion 1&2 JUN 4-5, 1996 (1) Dresden 2&3 Browns Ferry 3 Indian Point 3 Millstone 1,2&3 Meetina Dates Facilities for Priority Attention JAN 14-15, 1997 None JUN 4-5, 1996 None NRC senior management will continue to hold meetings to review the status of all reactor and materials licensees on an approximate six-month frequency. Recommendations will be made during those meetings to add or delete licensees from the list of facilities requiring increased NRC attention based on demonstrated performance. This program represents a concerted effort by the NRC senior management to focus NRC resources on those plants and issues that need to be addressed to assure adequate protection of public health and safety.

(1) Because a decision regarding the restart of Browns Ferry Unit i has been indefinitely deferred, the senior managers concluded, in June 1996, that it should no longer be identified as a Category 3 plant. However, if TVA were to decide to vae"--

operation and restart activities at Browns Ferry Unit 1, this plant will be reinstated as a Category 3 plant requiring Commission authorization prior to resumption of operation.

17 The senior managers discussed the merits for increasing agency attention at Clinton. The senior managers reviewed the licensee's operational safety perfornance since the June 1996 SMM, in light of the above discussion. The senior managers noted i that the safety performanca indicators for Clinton exhibited good ,/

perfor mance, consistent with ..le agency's pre"'-" assessments of

clinton. They also discussed the fact that Clinton was currently shutdown and that confirmatory action letters were issued to the l licensee in September 1996 and January 1997 to record the staff's l understanding of the actions that the licensee would take prior i to restart in response to the April 1996 reactor scram / pump seal s event, and in response to the September 1996 recirculation pump f seal failure event, respectively. The senior managers observed '

that the recent decline in Clinton's overall performance was due largely to a lack of a conservative operational focus within the l organization.

The senior managers acknowledged the recent management and I operating crew changes at Clinton, the licensee initiatives aimed ,

at instilling conservative dec!.sion making, and the actions to j j resolve the numerous poor material condition issues. The senior '

managers also recognized the licensee's efforts in the engineering area, including its intention to perform an independent engineering assessment, and its commitments to develop and implement long term engineering improvements and to { ,

conduct special training and programs aimed at enhancing engineering support to operations. However, because of concern about the licensee's reduced emphasis on safe plant operation, which was apparent in the recent plant events, the willingness to accept degraded material condition, and the weaknesses in i procedural adequacy and adherence, the senior managers concluded that additional agency attention was warranted. The senior managers recommended that the Executive Director for Operation send a trending letter to Illinois Power Company informing the Chief Executive Officer of the agency's concern regarding the decline in operational safety performance at the Clinton Power ,

Station.

POINT BEACH Point Beach Nuclear Plant is being discussed for the first time at the Senior Management Meeting because of the significance of recently identified performance issues, management's failure to aggressively pursue identification of the full extent of these issues, and the licensee's Jack of'a strong safety-focused questioning attitude. The plant's performance has declined substantially since the Systematic Assessment of Performance (SALP) was issued in the April 1996, and longstanding problems have only recently been identified. Weaknesses in operations, engineering, and maintenance led to numerous violations and a significant enforcement actions within the last six months. The primary cause of these performance issues appears to be a pervading focus on keeping the units operating without guestioning or resolving long-standing safety issues.

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During the last six months, the station was plagued with inattentiveness to duty by control room operators, the failure to maintain proper equipment configuration control, and ineffective surveillance testing. At times, operations staff and plant management exhibited poor understanding of administrative procedures, technical specifications, and regulatory requirements. For example, the Unit 1 control room operator left his watch-station with no designated relief present, and Shift i Technical Advisors routinely left the site in violation of l technical specification requirements. Also, Unit I was taken 1 critical with the turbine-driven AFW pump discharge isolation .

valves closed in violation of technical specification  !

requirements. Finally, when an auxiliary operator was notified I that he reported a discharge pressure on the motor driven auxiliary feedwater pump that was below surveillance requirements, he "re-read" the gauge and modified his report such  ;

that the pressure was within the requirements. Control room l operators failed to recognize the non-conservative nature of these actions and declared the pump operable.

The material condition of safety-related components continued to be good. However, some long-standing deficiencies in the surveillance and post-maintenance testing program contributed to I

degraded equipment conditions and increased out-of-service times '

for safety-related equipment. For example, a service water pump was returned to service with IST results in the alert range, contrary to the requirements of Section XI of the ASME Code.

There was declining performance in the area of engineering support to the station. Engineering performance was characterized by inadequate operability determinations and engineering evaluations, poor documentation of engineering evaluations, and the failure to manage plant configuration in accordance with the FSAR and design basis. Several instances were identified where lack of a questioning attitude, untimely corrective actions, and lack of engineering rigor resulted in degraded plant safety. For example, proper corrective actions were not taken after engineers identified that the number of service water pumps required to mitigate a design basis accident was greater than the number assumed in the FSAR and Technical Specification bases.

The senior managers discussed the merits for increasing agency attention at Point Beach, in light of the above discussion. The senior managers noted that the safety performance indicators for Point Beach exhibited good performance, which was consistent with the agency's most recent SALP assessment. The senior managers recognized that the staff's current view of the licensee's pe::formance conflicts with the recent SALP, but noted that this dichotomy reflects early intervention by the NRC through its inspection program.

The senior managers reviewed the licensee's operational safety performance since the June 1996 SMM and noted several areas of improving performance and a number of areas where the licensee has committed to improvement initiatives. The operator response to off-normal conditions remains good and control room formality has improved. An improvement in self-identification of issues

, 19 was evidenced in the increase in the number of condition reports.

The senior managers noted that the licensee is in the process of upgrading operations, surveillance-and maintenance procedures, is relocating corporate engineering to the site and developing a system engineering program to improve ownership, and has committed to a design reconstitution effort and a review of the FSAR.

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The senior managers discussed the fact that confirmatory action letters were issued to the licensee in September 1996 and January 1997, respectively, to record the staff's understanding of the actions that the licensee would take regarding spent fuel loading / unloading as a result of the Palisades hydrogen ignition event, and to confirm licensee actions to address significant weaknesses with. procedures, work and test activities, licensin and design basis adherence, and the corrective action program.g The senior managers also discussed, with concern, the multiple instances of poor performance in all areas of plant operations.

In the operations area, the senior managers noted the examples of inattentive control room operators, and discussed the implications of marginal control room staffing and the inconsistent concervative operating philosophy. The senior managers recognized the weaknesses in maintenance and testing, as well as the poor engineering support for operations which resulted in weak operability evaluations, inadequate testing criteria and un-evaluated engin_aring deficiencies. The senior managers also discussed the fact that the NRC rather than the licensee continues to identify performance issues and conditions adverse to quality and that the licensee has demonstrated little self or independent performance assessment.

In light of the overall declining safety performance at Point Beach, and the fact that the licensee's response has, thus far, been ineffective in reversing the trend, the senior managers concluded that additional agency attention was warranted. The senior managers recommended that the Executive Director for Operation send a trending letter to Wisconsin Electric Power Company informing the Chief Executive Officer of the agency's concern regarding the decline in operational safety performance at the Point Beach Nuclear Power Plant.

'~' Additional Topics Discussed F-- l

1. EDO's Opening Remarks The Acting EDO welcomed the attendees to the Senior Management  ;

Meeting (SMM) and began his remarks by briefly highlighting the recent NRC organizational changes and emphasizing the importance j of good communication to ensure a smooth transition. He '

indicated that because the SMM process is receiving close scrutiny, both from within and from outside the agency, it is imperative that the recommendations from the SMM be clear, understandable and defensible. He cited the significant amount ,

I

of effort involved in preparing for the SMM and encouraged j management to use the information appropriately and to

, participate actively in the discussions. The Acting EDO l reiterated the Commission's interest in the results of the SMM, '

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POINT BEACH IRGUMENTS FORINCREASING AGENCY ATTENTION o Conduct of Operation Inattentive Control Room Operators Marginal Control Room Staffing Inconsistent Operating Philosophy o Conduct of Maintenance and Testing Maintenance and Testing Documents Sometimes Lack' Restoration and/or Post Maintenance Testing Requirements, Resulting in Configuration Control and Operability Verification Problems o Conduct of Engineering Examples of Weak Operability Evaluations, Questionable Testing Acceptance Criteria, and Un-evaluated Engineering Deficiencies were-Identified.

o Licensing Bases Plant, Tech Specs, FSAR and Design Bases Not Always in Agreement or Conservative o Corrective Actions and SelfIdentification ofIssues Little Self or Independent Assessment of Performance Corrective Actions Tended to Focus on NRC Identified Issues and Have Not Always Been Effective NRC Residents Continue to Find Significant Performance Issues OSTI Inspection Identified Many Significant Performance Issues Particularly in the Area ofIdentifying and Addressing Conditions Adverse to Quality and Additional Examples of Nonconservative Technical Specification Surveillance Testing i

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. l POINT BEACH '

ARGUMENTS FOR MAINTAINING CURRENT AGENCY ATTENTION ,

o SES Oversight Manager; Dedicated Branch Chief; Recently Assigned i

Resident Staff; Augmented RegionalInspection l o Conduct of Operation 1 Operator Response to Off-Normal Conditions Remains Good and Some Improvements Noted in Control Room Formality o Conduct of Maintenance and Testing Reviewing / Upgrading OPS, Surveillance and Maintenance Procedures to Address Issues; Also Reviewing past Activities -

o Conduct of Engineering Moving Corporate Engineering to Site and Formulating a System Engineering Program to Improve Ownership o Licensing Bases Performing Design Reconstitution; Committed to a Full Review of FSAR -

l Performed Interim Review of One Chapter Substantial Commitment of Resources for SG Replacement and for Additional EDGs to Improve Safety and Reliability o Corrective Actions and SelfIdentification ofIssues  !

Enhanced Performance Improvement and Restart Plan Confirmed Through '

a Confirmatory Action Letter Number of Condition Reports Has Increased Substantially Management and Staffing Changes Have Been Implemented to Better Address Performance Issues Committed to Further Improve Condition Reporting, Operability Evaluation, and 50.59 Evaluation Programs

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ii e) 6 p arco y $ UNITED STATES j

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 y,, # January 27, 1997 Mr. Richard A. Abdoo, Chairman of the Board and l'

Chief Executive Officer Wisconsin Electric Power Company 231 W. Michigan P. O. Box 2046 Milwaukee, WI 53201 l

Dear Mr. Abdoo:

I l On January 14, 15, and 17, 1997, NRC senior managers met to evaluate the nuclear safety performance of operating reactors, fuel facilities, and other  ;

materials licensees. This meeting is conducted semiannually to determine .

whether the safety performance of various licensees warrants increased NRC  !

attention.

The Commission has directed that, during these meetings, NRC senior managers I also identify those plants whose performance is trending adversely and that steps be taken to communicate concerns to the utility's corporate president or board of directors. This early notification to the highest levels within the 1

utility's organization is intended to allow appro by the licensee to address the areas of concern. priate measures to be taken This letter is to advise you that during the January Senior Management Heeting, recent trends in the performance at Point Beach Nuclear Plant raised sufficient concerns that we believe a meeting with you would be appropriate. )

l The NRC began identifying safety concerns at Point Beach in mid-1996 that indicated performance at the plant had declined. NRC inspection findings included examples of inattentive control room operators, inadequate control room staffing, loss of control of equipment configuration, and inadequate surveillance testing. Those concerns covered a broad range of activities indicating weaknesses existed in the functional areas' of Operations, Naintenance, and Engineering. As a result, a number of violations were identified and an NRC enforcement action involving a $325,000 civil penalty was issued. Your staff developed a performance improvement plan. After the initiation of that plan, NRC continued to find new safety issues that your staff had not identified or addressed. These issues indicate that the facility's corrective actions to address specific concerns had not been fully effective.

An obstacle to improving plant performance had been an ineffective problem 4 identification and corrective action program where your plant and corporate staff had not consistently been raising issues to appropriate management levels to assure effective resolution. Further, some engineering decisions had been non-conservative in that they focussed on continued plant operations

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L (. -

" l R. A. Abdoo over assuring that required safety margins were maintained. Recent NRC l observations indicate that your corporate and site management are beginning to understand the significance of tne alsnt's shortcomings in. the ability to ,

identify and resolve deficiencies. NRC has observed improvements in the '

plant's identification and resolution of specific safety issues, i

The NRC acknowledges your recent management and staff changes ar.d the '

i commitments identified in Mr. Grigg's December 12, 1996, letter to address Point Beach's performance problems. NRC confirmed those commitments in a

. Confirmatory Action Letter (CAL No. RIII-96-017) signed by Mr. A. Bill Beach and issued on January 3, 1997. The scope and depth of the performance weaknesses at Point Beach will be revealed through successful implementation i of those commitments and continued improvement in your staff's ability to j identify and resolve problems. Success in resolving these performance concerns will require strong management involvement and commitment to safety.

Mr. A. Bill Beach, the NRC Region III Administrator in Lisle. Illinois, will be contacting you to arrange for a mutually agreeable time and location for a meeting covering Point Beach performance.

If you have any questions regarding this matter, please do not hesitate to call me.

l Sincerely, 4

uK . hompso Jr A ti g Executiv D' rec r r Operatio s Docket Nos. 50-266, 50-301 cc: See next page l

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