ML20133G311

Safety Evaluation Supporting Amends 170 & 174 to Licenses DPR-24 & DPR-27,respectively

ML20133G311
Person / Time
Site:	Point Beach
Issue date:	01/08/1997
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20133G303	List: ML20133G300 ML20133G311
References
NUDOCS 9701150297
Download: ML20133G311 (6)

Text

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k UNITED STATES s

g NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20806 0001

\\...../

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 170 AND 174 TO FACILITY OPERATING LICENSE NOS. DPR-24 AND DPR-27 WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR, PLANT. UNIT NOS. 1 AND 2 DOCKET NOS. 50-266 AND 50-301

1.0 INTRODUCTION

By letter dated April 29, 1996, as supplemented October 21, December 2, and

-December 16, 1996, Wisconsin Electric Power Company (WEPCO), the licensee, submitted a request for revision to the Point Beach Technical Specifications (TSs). The proposed changes would relocate the fire protection requirements from the Point Beach TS to the Point Beach Fire Protection Evaluation Report (F.PER).

Specifically, the licensee proposed to revise the operating licenses to include the U.S. Nuclear Regulatory Commission's (NRC's) standard fire protection license condition and to relocate the fire protection requirements of TS Sections 15.3.14 and 15.4.15, both titled " Fire Protection System," to the Point Beach FPER, Section 7.2.

Administrative TSs 15.6.2.2.f (fire brigade staffing), 15.6.4.2_(fire brigade training), and 15.6.9.F (fire j

protection system degradation reporting requirements) would also be relocated to the FPER. The requested changes are consistent with the guidance of NRC 1

Generic Letter (GL) 86-10, " Implementation of Fire Protection Requirements,"

l and GL 88-12, " Removal of Fire Protection Requirements from Technical Specifications." The October 21, December 2, and December 16, 1996, supplements provided revised license and TS pages and a 90-day implementation schedule to reflect clarification of the safety evaluations to be included in i

the license conditions and typographical corrections for the license page for each unit. These supplements were within the scope of the original application and did not change the NRC staff's initial proposed no significant hazards considerations determination.

2.0 BACKGROUND

Section 50.48, " Fire protection," of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50) requires that each operating nuclear i

power plant have a fire protection plan that satisfies General Design Criterion 3 (GDC 3), " Fire protection," of Appendix A to 10 CFR Part 50. The fire protection plan must describe the overall fire protection program for the facility, outline the plans for fire protection, fire detection, and fire suppression capability, and limitations of fire damage. The program must also describe specific features necessary to implement the program, such as administrative controls and personnel requirements for fire prevention and manual fire suppression activities, automatic and manually operated fire detection and suppression systems, and the means to limit fire damage to 9701150297 970108 PDR ADOCK 05000266 F

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structures, systems, or components important to safety so that the capability i

to safely shut down the plant is ensured. The NRC staff approved the Point i

Beach Nuclear Plant, Units 1 and 2, fire protection program in a Safety 1

Evaluation Report (SER) dated August 2,1979 (and supplemental SERs dated i

j October 21, 1980, January 22, 1981, and July 27,1988).

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3 GL 86-10 and GL 88-12 referred to removing fire protection requirements from l

TSs.

License amendments that relocate the fire protection requirements to the

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Final Safety Analysis Report (FSAR) in accordance with GL 86-10 and GL 88-12 i

do not revise the requirements for fire protection operability, testing, or l

inspections. Such amendments simply replace the fire protection TS sections i

with the standard fire protection license condition. The license condition i

implements and maintains the NRC-approved fire protection program, including j

the fire protection requirements previously specified in the TSs, in j

accordance with 10 CFR 50.48. Therefore, such amendments, including the one i

proposed by the licensee, are administrative in nature and have no effect on l

public health and safety.

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Section 182a of the Atomic Energy Act (the "Act") requires applicants for nuclear power plant operating licenses to state TSs to be included as part of l

the license. The Commission's regulatory requirements related to the content j

of TSs are set forth in 10 CFR 50.36. That regulation requires that the TSs j

include items in five specific categories, including (1) safety limits, limiting safety system settings and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; i

and (5) administrative controls. However, the regulation does not specify the particular requirements to be included in a plant's TSs.

The Commission has provided guidance for the contents of TSs in its " Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" (" Final Policy Statement"), 58 FR 39132 (July 22,1993), in which the Commission indicated that compliance with the Final Policy Statement satisfies Section 182a of the Act.

In particular, the Commission indicated that certain items could be relocated from the TSs to licensee-controlled documents, consistent with the standard enunciated in Portland General Electric Co. (Trojan Nuclear Plant), ALAB-531, 9 NRC 263, 273 (1979).

In that case, the Atomic Safety and Licensing Appeal Board indicated that " technical specifications are to be reserved for those matters as to which the imposition of rigid conditions or limitations upon reactor operation is deemed necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety." The criteria set forth in the policy statement have been incorporated into 10 CFR 50.36 (60 FR 36953).

Following the fire at the Browns Ferry Nuclear Power Plant on March 22, 1975, the Commission undertook a number of actions to ensure that improvements were implemented in the fire protection programs for all power reactor facilities.

Because of the extensive modification of fire protection programs and the number of open issues resulting from staff evaluations, a number of revisions and alterations occurred in these programs over the years. Consequently, licensees were requested by GL 86-10 to incorporate the final NRC-approved fire protection program in their FSARs.

In this manner, the fire protection program, including the systems, certain administrative and technical controls,

i the organization, and other plant features associated with fire protection, would have a status consistent with that of other plant features described in the FSAR.

In addition, the Commission concluded that a standard license condition, requiring compliance with the provisions of the fire protection program as described in the FSAR, should be used to ensure uniform enforcement of the fire protection requirements.

Finally, the Commission stated that with the required actions, licensees may request an amendment to delete the fire protection TSs that would now be unnece:;sary. Subsequently, the NRC issued GL 88-12 to give guidance for the preparation of the license amendment request to implement GL 86-10.

3.0 PROPOSED CHANGE

S The specific changes proposed by the licensee are as follows:

1.

Revise License Condition 3.H for Units 1 and 2 to read as follows-Wisconsin Electric shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility and as approved in the SER dated August 2, 1979 (and Supplements dated October 21, 1980, January 22, 1981, and July 27, 1988), and the SER issued

, 1996, for Technical Specification Amendment No.

subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

2.

Delete TS 15.3.14, " Fire Protection System," including the basis and the table, and incorporate them, by reference to the FPER, into the Point Beach FSAR.

3.

Delete TS 15.4.15, " Fire Protection System," including the basis, and incorporate them, by reference to the FPER, into the Point Beach FSAR.

4.

Delete TS 15.6.2.2.f (fire brigade staffing), and incorporate it, by reference to the FPER, into the Point Beach FSAR.

5.

Delete TS 15.6.4.2 (fire brigade training), and incorporate it, by reference to the FPER, into the Point Beach FSAR.

6.

Delete TS 15.6.9.2.F (fire protection system degradation reporting requirements), and incorporate it, by reference to the FPER, into the Point Beach FSAR.

7.

Correct spelling of " actuation" and " unit" in section 3.G., and remove quotation marks in section 3.I.

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' i 4.0 EVALUATION The NRC staff reviewed the license amendment requests for Point Beach against the guidance provided in GLs 86-10 and 88-12. GL 86-10 requested that the licensee incorporate the NRC-approved fire protection program in its FSAR for-the facility and specified a standard fire protection license condition.

GL 88-12 addressed the elements a licensee should include in a license amendment request to remove the fire protection requirements from the plant TSs. These elements are (1) the NRC-approved fire protection program must be incorporated into the FSAR; (2) the Limiting Conditions for Operation (LCOs) and Surveillance Requirements associated with fire detection systems, fire suppression systems, fire barriers, and the administrative controls that address fire brigade staffing would be relocated from the TSs (the existing administrative controls associated with fire protection audits and specifications related to the capability for safe shutdown following a fire would be retained); (3) all operational conditions, remedial actions, and test requirements presently included in the TSs for these systems, as well as the fire brigade staffing requirements, shall be incorporated into the fire protection program; (4) the standard fire protection license condition specified in GL 86-10 must be included in the facility operating license;

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(5) the Manager's Supervisory Staff (Onsite Review Group) shall be given responsibility for the review of the fire protection program and implementing procedures and the submittal of recommended changes to the Off Site Review Committee (Offsite or Corporate Review Group); and (6) fire protection program i

implementation shall be added to the list of elements for which written procedures shall be established, implemented, and maintained. The licensee i

incorporated the NRC-approved fire protection program by reference into the Point Beach FSAR in June 1987. The licensee has, therefore, satisfied Element 1 of GL 88-12.

The licensee stated in its submittal of April 29, 1996, that it will incorporate the current TS LCOs and surveillance requirements for the fire protection system in their entirety, together with the fire brigade staffing and training requirements, into the FPER. Therefore, the licensee will have satisfied Elements 2 and 3 of GL 88-12.

The licensee proposed to modify the current licenses for Point Beach to include the standard fire protection license condition specified in GL 86-10.

Therefore, with the addition of a reference to this safety evaluation as described in the supplement of October 21, 1996, the licensee will have satisfied Element 4 of GL 88-12.

Current TS 15.6.5.1.8.h requires the Manager's Supervisory Staff (onsite review coautittee) to " review the Facility Fire Protection Prosram at least once per 24 months to ensure the program meets established commitments and requirements." Current TS 15.6.5.2.7 requires the Offsite Review Committee to review all changes completed under 10 CFR 50.59 (which would include changes to the FPER). Therefore, the licensee currently meets Element 5 of GL 88-12.

Element 6 of GL 88-12 specified that the licensee add fire protection program implementation to the list of elements for which written procedures shall be established, implemented, and maintained. Current TS 15.6.8.1.8 satisfies

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i 4' this requirement. Therefore, the licensee currently meets Element 6 of GL 88-12.

The licensee's proposed TS amendments for Point 8each are in accordance with NRC staff guidance provided in GLs 86-10 and 88-12.

In summary, the licensee has proposed to incorporate the existing TS fire i

protection requirements as stated above into the fire protection program which i

is, by reference, incorporated into the FSAR. This conforms to staff guidance in GL 86-10, " Implementation of Fire Protection Requirements," and GL 88-12,

" Removal of Fire Protection Requirements from Technical Specifications," for j

removing unnecessary fire protection TSs in four major areas:

fire detection i

systems, fire suppression systems, fire barriers, and fire brigade staffing requirements.

In addition, incorporating these requirements into the FSAR is consistent with NUREG-1431, " Standard Technical Specifications, Westinghouse Plants," and 10 CFR 50.36, as amended, because these TSs do not impact reactor operations, do not identify a parameter which is an initial condition i

assumption for a design-basis accident or transient, do not identify a j

significant abnormal degradation of the reactor coolant pressure boundary, and j

do not provide any mitigation of a design-basis event.

j The fire protection plan required by 10 CFR 50.48, as implemented and l

maintained by the fire protection license condition, provides reasonable assurance that fires will not give rise to an immediate threat to public health and safety. Although there are aspects of the fire detection and i

mitigation functions that have been determined to be risk significant, such that criterion 4 of 10 CFR 50.36 would otherwise seem to apply, the minimum i

i requirements for those functions were established in GDC 3 and 10 CFR 50.48, and further controls are not necessary since the licensee must comply with these minimum requirements regardless of whether they are restated in the TSs or not.

The licensee's fire protection program is required by 10 CFR 50.48, and any changes to that program are governed by 10 CFR 50.48 and license conditions 3.H. (Units 1 and 2), set forth above. Therefore, the requirements relocated to the FSAR may be controlled in accordance with 10 CFR 50.59.

These relocated requirements relating to fire protection features are not required to be in the TSs under 10 CFR 50.36 or other regulations, or by Section 182a of the Atomic Energy Act, and are not required to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety.

Ir. addition, the staff finds that sufficient regulatory controls exist under 10 CFR 50.48 and 10 CFR 50.59 to address future changes to these requirements. Accordingly, the staff has concluded that these requirements may be relocated from the TSs to the licensee's FSAR.

Additionally, the staff has reviewed the typographical changes included in the supplement of October 21, 1996, and has determined that these changes are administrative in nature, have no effect on safety, and are therefore acceptable.

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5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Wisconsin State official l

was notified of the proposed issuance of the amendments. The State official had r.o comments.

6.0 ENVIR0fMENTAL CONSIDERATION l

The amendments change requirements with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR j

Part 20 and change surveillance requirements. The amendments also change administrative procedures or requirements. The staff has determined that the j

amendments involve no significant increase in the amounts, and no significant j

change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational i

radiation exposure. The Commission has previously issued a proposed finding i

that the amendments involve no significant hazards consideration and there has l

been no public comment on such finding (61 FR 28621). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth j

in 10 CFR 51.22(c)(9) and (c)(10).

Pursuant to 10 CFR 51.22(b), no j

environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

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7.0 CONCLUSION

The staff has concluded, based on the considerations discussed above, that:

l (1) there is reasonable assurance that the health and safety of the public j

will not be endangered by operation in the proposed manner, (2) such i

activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

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Principal Contributors:

C. Thomas (by precedent) i K. West (by precedent)

A. Hansen Date: January 8,1997

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i 111. Enaineerina E1 Conduct of Engineering f

E1.1 Unit 2 Division 4 Safeauards Batterv

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a.

Scoce (37551)

J On November 12, operators identified that cell No. 7 on the Unit 2 Division 4 safeguards battery had a low specific gravity. This battery had been replaced in its entirety in October i

1996, due to cell voltage problems (see NRC Integrated Inspection Report 50-352, 353/96-1 07). The battery was declared inoperable, and the cell was replaced. The inspector discussed the cell condition with the system manager.

1 b.

Observations and Findinas The system manager indicated that the cell No. 7 specific gravity was lower than the other cells after the battery replacement in October, but was above the minimum required. A conscious decision was made to not replace the cell, as it was expected that the specific gravity would increase over time and equalize with the other cells. During subsequent i

testing approximately one month later, the specific gravity was below the minimum value.

2 Additionally, it was noted that the electrolyte level for this cell, although in the acceptable range, was lower than the other cells. Engineering personnel indicated that the other parameters for the cell were normal, and that the battery would be expected to perform adequately under design conditions.

i C.

Conclusions i

The inspectors concluded that engineering personnel should have been more proactive in monitoring the condition of the cell, after it was identified as being weaker than the rest.

Although the battery would have adequately performed under design conditions with the low cell, the cell condition was not tracked and trended for a one month period, resulting in the entry into an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> TS LCO. Additionally, the cell electrolyte level should have been l

brought up to the same level as the rest of the cells. The weak cell was replaced expeditiously, and the battery was returned to a fully operable status.

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E1.2 Enaineerino Review of EHC Pressure Switch Failure a.

Inspection Scope (37551)

The inspector reviewed and discussed the engineering review and assessment of the component failures that contributed to the Unit 2 scram on December 6.

4 b.

Observations and Findinas System managers and technicians determined that a fatigue failure of the bracket supporting the No. 2 TCV pressure switch was the cause of the event. The bracket failure

9 b.

Observations and Findinas On November 15, after a thorough walkdown of the Unit 1 and 2 EDGs, the inspectors brought several concerns to the attention oi EDG branch and plant management. Four of the eight EDGs had governor oillevels lows than the rest, and although they were above the minimum level mark, after an EDG start they would probably drop below the minimum i

level. This was based on feedback from an experienced EO, and on direct observation of level change during and after an EDG start and warmup. Two of the EDG's governor oil filter fittings were leaking. These leaks were promptly fixed by the LFIN team. Three of the EDGs had generator oil levels below the standstill level; this was noted on the D24 EDG just prior to the overspeed test start, was mentioned to the attendant system manager, and was promptly corrected (after verification), prior to starting the engine. The new EDG branch had been active for approximately one month, and there was a plant improvement day on November 1, where two of the EDGs had significant painting done.

However, the following material was found on the engines (these items were physically shown to the EDG branch manager, a maintenance foreman, and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> later to the plant manager and the maintenance manager): wire, nuts, bolts, insulation, and rags in the oil trough under the air blower; the plant manager removed some of the rags immediately.

Additionally, the leaking governor oil filter fitting was shown to plant management as were the low generator oil levels.

Essentially all of the EDGs showed evidence of poor housekeeping practices.

On November 17, the inspector performed another walkdown of the EDGs. Governor oil levels appeared to be adequate. Generator oillevels for all EDGs were on the standstill markings as required. While in the D24 EDG room, the inspector noted that personnel were being briefed on where to clean on the engines.

On November 19, plant management had operations personnel, with help from other work groups, start cleaning the EDGs thoroughly, one at a time, one per day. After the thorough cleaning, the inspectors noted that the material condition of the EDGs was much improved.

Additionally, the inspectors noted that the D24 EDG, which completed the 18 month overhaul / inspection in late November, was more thoroughly cleaned than the rest, and was painted; the overall condition of the D24 EDG was noticeably better than the other EDGs.

Finally, the inspectors noted that, in early December, management set up new plans for maintaining the EDGs in an improved state of cleanliness. After each EDG is run, operators will wipe down the EDG to eliminate the need for all the oil pads that have been used in the past, and to make it easier to identify leaks needing repair.

c.

Conclusions Several concerns raised to plant management by the inspectors necessitated corrective action to improve the material condition of the EDGs. The inspectors concluded that the material condition of the EDGs has improved, that management expectations for maintaining the current condition of the EDGs is clear, and that the plans for maintaining the condition of the EDGs are being properly implemented.

i 11 was due to poor design aggravated by main steam piping vibrations, which induced a fatigue crack in the thin brass U-shaped bracket.

The engineering staff had been monitoring the pressure switch bracket associated with the No. 3 TCV in November. The bracket was degrading and engineering was planning to replace it during the upcoming refueling outage. As a result of the failure of the No. 2 TCV bracket, the engineering staff addressed the adequacy of design of all TCV pressure switches during the maintenance outage. The engineering staff modified the design for the pressure switch supports to be like those on Unit 1, which have been installed for all TCVs.

Engineers also noticed fatigue indications on the stainless steel tubing that was connected to some other RETS pressure switches. The engineers had determined that the bracket failure on the No. 2 TCV had accelerated the fatigue failure of the stainless steel tubing.

Engineers recommended that the stainless steel tubing associated with all TCV RETS pressure switches be replaced as a corrective action. This was accomplished during the maintenance outage.

Engineers walked down main steam piping, inside and outside of the drywell and in the _

main condenser area, to identify and evaluate attached components for susceptibility to failure from piping vibration. Also, a review was performed of main steam piping isometrics and branch lines to evaluate characteristics that would produce acoustic vibrations. The engineers found three broken pipe supports on non safety-related main steam and feedwater piping that were repaired during the outage. They assessed that the failure was due to high tensile stress aggravated by combined cycle fatigue stress from vibration.

The engineering staff is continuing to monitor main steam piping vibration. As corrective action, the TCVs and other main steam components have had extensive vibration and acoustics monitoring instruments attached. Engineers will collect data during the remainder of the operating cycle prior to the refueling outage to further assess any vibration problems, c.

Conclusion The engineering staff performed a high quality review and assessment into the failure of the EHC components that caused the Unit 2 scram. The continued effort to assess the main steam vibration problem is a positive initiative to correct the problem in the future.

E1.3 Safe Shutdown Fire issues a.

Scoce (37551)

On December 6, plant personnel notified the NRC of a condition where provisions of the Fire Protection Program were not properly maintained, constituting a non-compliance with a license condition. The NRC reviewed the conditions surrounding the apparent non-compliance, the compensatory actions taken, and the corrective actions taken.

Discussions were held between NRC management and plant management, and corrective actions put in place were reviewed by the inspectors.