ML18022A479

Forwards Addl Responses to ACRS Recommendations Documented in ACRS 840116 Ltr to Chairman Palladino.Control Room Air Ventilation Sys Testing,Westinghouse D-4 Steam Generator Tube Degradation & Documentation of Operations Discussed

ML18022A479
Person / Time
Site:	Harris
Issue date:	12/11/1986
From:	Zimmerman S CAROLINA POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
NLS-86-462, NUDOCS 8612160420
Download: ML18022A479 (13)

Text

REGULATORY INFORMATION DISTRIBUTI N SYSTEN (RIDS>

ACCESSIQN NBR: 8612160 0 DQC. DATE: 86/12/11 TARIZED: NO DOCKET FACIL: 50-400 Shearon Harris Nuclear Power Plant> Unit 1> Carolina 05000400 y'AUTH NAl'1E AUTHOR AFFILIATION ZINNERNAN> S. R. Carolina Power Sr Light Co.

REC IP. NANE 'ECIPIENT AFFILIATION DENTON> H. R. Office of Nuclear Reactor Regulation> Director (post 851125

SUBJECT:

Forwards addi responses to ACRS 'recommendations documented in ACRS 840116 ltr to Chairman Pal ladino. Control room air ventilation sos testing> Westinghouse D-4 steam generator tube degradation 8c documentation of operations discussed.

DISTRIBUTION CODE: A001D COPlES RECEIVED: LTR ENCL SIZE:

TITLE: OR Submittal: General Distribution NOTES: Application for permit'enewal f i led. 05000400 REC I P IENT COPIES RECIPIENT COPIES ID CODE/NAl'lE LTTR ENCL ID CODE/MANE LTTR ENCL PWR-A EB 1 1 PWR-A EICSB 2 2 PWR-A FOB 1 1 PWR-A PD2 'LA 1 0 PWR-A PD2 PD 01 5 5 BUCKLEY> B 1 1 PWR-A PSB 1 PWR-A RSB INTERNAL: ACRS 09 6 6 ADN/LFNB 1 0 ELD/HDS1 1 0 B~HF TSC9 1 1 NRR/GRAS 1 0 R QFI 04 1 R(:N2 1 1 EXTERNAL: E('M BRUSKE> S 1 1 LPDR 03 NRC PDR 02 1 NSIC 05 TOTAL NUBBER OF COPIES REQUIRED: LTTR 29 ENCL 25

'l r

T'

CMH, Carolina Power & LIght Company QEC i~ lsb6 SERIAL: NLS-86-062 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-000/LICENSE NO. NPF-53 ACRS RECOMMENDATIONS

Dear Mr. Denton:

Carolina Power R Light Company (CPRL) hereby submits additional information responding to the recommendations of the NRC Advisory Committee on Reactor Safeguards (ACRS) for the Shearon Harris Nuclear Power Plant. The ACRS recommendations are documented in the ACRS letter to NRC Chairman N. Palladino, dated January 16, 1980. Since the ACRS meeting in January 1980, CPRL has provided additional information to the NRC staff in response to these recommendations. The purpose of this letter is to summarize CPRL's position on each of the ACRS recommendations with appropriate references to previously submitted information.

Carolina Power R Light Company's position on each of the ACRS recommendations is attached.

If you have any questions on the attached information or require additional information, please contact me.

Yours very truly,

. R. Zi merman nager Nuclear Licensing Section JHE/mf (5096 JDK)

Attachment Cct Mr. B. C. Buckley (NRC)

Mr. G. F. Maxwell (NRC-SHNPP)

Dr. J. Nelson Grace (NRC-RII)

Sbi2ib0420 Bbi2ii PDR ADOCK 05000400 0 PDR 411 Fayettevilte Street + P. O. Box 1551 ~ Raleigh, N. C. 27602 pggl

ACRS RECOMMENDATION(1):

During the Committee's consideration of this plant, the control room design was reviewed. The Applicants informed us that they intend to perform an operational test of the control room emergency air recirculation system. As a part of this exercise, control room habitability during the recirculation mode will be evaluated. We wish to be kept informed.

CPRL POSITION:

By letter dated December 2, 1986, CPRL described the progress of testing the Control Room Area Ventilation System (CRAVS) relative to pressurization and cooling capacity while in the recirculation mode. At that time the CRAVS ability to pressurize the Control Room area to design levels, effectively 1/0 inches w.g.

relative to outside atmospheric pressure, had been verified. The system's cooling coil capacity was also checked, but a test which measured the Control Room area air temperature directly had not yet been performed. Since December 2, 1986, CPRL has completed additional work and testing including final flow balancing within the CRAVS and performance of temperature monitoring showing that the Control Room area temperature was maintained at or below the test acceptance criteria of 77'F (75' 2'). The system checks and surveillance tests, both on the component and system level, show. that the CRAVS does perform as designed in the recirculation mode and:that habitability of the Control Room area will be maintained if operation while in the recirculation mode is necessary.

(5096JOK/vaw)

ACRS RECOMMENDATION(2):

The Shearon Harris Nuclear Power Plant uses Westinghouse D-4 steam generators.

Steam generators of this design have experienced tube degradation related to flow-induced vibrations in the preheater region. Internal modifications have been developed by Westinghouse which include expanding some steam generator tubes and directing some of the main feedwater flow through the auxiliary feedwater nozzle. We expect to be kept informed regarding the operating experience of these steam generators.

CPM POSITION:

CPRL was a member of a DO/D5/E Steam Generators Owners'roup that addressed a generic problem concerning the potential for tube degradation caused by flow-induced vibration in the preheater section of the steam generator. The NRC Staff reviewed the modification proposed by'the Owners'roup that consisted of expanding the steam generator tubes in the preheater region and directing some of the main feedwater flow through the auxiliary feedwater nozzle. The results of the NRC Staff's thorough evaluation of this modification was reported in NUREG-1010. This modification has been completed for the DO steam generators in use at SHNPP.

In addition, CPRL has modified the Westinghouse Standard Technical Specification for inservice inspection of the steam generators to specifically address the inspection of expanded tubes in the preheater region. CPRL s preservice inspection of the steam generators established baseline data for all steam generator tubes, including the expanded tubes in the preheater region.

A letter from Westinghouse to CPRL dated November 25, 1986 documenting the operating experience of the DO/D5 steam generators used at SHNPP has been provided to the NRC Staff for information. 0 (5096 JDK/vaw )

ACRS RECOMMENDATION(3):

The Committee believes that written evidence of an improvement in CPRL's nuclear operations, which could, for example, be reported in the two scheduled SALP reviews prior to fuel load should be available prior to full power operation.

We wish to be kept informed.

CPdtL POSITION:

SALP Re orts NRC's most recent Systematic Assessment of Licensee Performance (SALP) for CPRL was released in February 1986 for all three nuclear sites and an additional pre-license SALP report for Harris was released in September 1986. All of CPRL's ratings were "1's" and "2's" with the exception of a "3" in one category, Electrical Equipment and Cables, at Harris in the February report; this category was, however, raised to a "2" in the pre-licensing SALP. Based on the comments provided both verbally and in the written reports, NRC sees our plants as continuing to demonstrate proper concern for nuclear safety. Of the total 31 categories assessed for all three sites in 1986 (for 1985 and 1986 performance), 18 categories were noted as "improving", 13 were "constant", and none were decreasing, and the pre-licensing SALP for Harris sh'owed 0 categories which had improved and none which decreased.

This SALP assessment is positive, an indication of continuing progress and increasing excellence.

Radwaste Volume CPdcL continues to show a reduction in the amount of radwaste generated and shipped. This reduction from 171,000 cu. ft. in 1982 to 05,381 cu. ft. so far in 1986 is indicative of improved maintenance, fewer modifications, and positive operational efforts to minimize radioactive materials.

Contaminated Area Surface area containment levels at both Brunswick and Robinson are being maintained below the established, historically conservative, goals set. Both sites have less area contaminated to date than in the past few years. This reflects a management attitude that is focused on continuing improvements in overall housekeeping, maintenance, and personnel exposure reduction.

Personnel Skin Contamination Events The number of skin contamination events for 1985 and 1986 has decreased substantially in comparison to historical trends. Again, these continued improvements reflect a management attitude that is focused on excellence in housekeeping and personnel exposure reduction.

Licensee Event Re orts (LERs)

LERs represent errors or omissions from Operating License Technical Specification requirements or regulations that have been found by the NRC or plant operating (5096 JDK/va>> )

personnel. While they may represent events of safety significance, in most instances they relate only to some condition that was outside of requirements for a brief interval, but without any major safety significance. They do, however, represent one measure that is used to denote compliance with regulations and presumably to reflect the effectiveness of operations.

Our trend on LERs continues to be encouraging. From a high of 327 LERs in 1982, the total decreased to 105 in 1985 (under a new NRC reporting system) and decreased further to 66 for 1986 through the month of October. The 1986 data also reflects addition of the Harris Unit into the reporting scheme.

Further, personnel errors specifically identified in LERs have been reduced from a total of 20 in 1985 to 19 through October of 1986 and, again, this reflects the addition of Harris to the reporting data base.

In addition, it should be noted that equivalent availability (which CPRL strongly believes can be equated directly with operational safety) for 1985 at Robinson was approximately 8596 compared to the PWR average of 62% and through November of this year, Robinson is already at 76%. Brunswick Unit No. 1 has a 1986 through November equivalent availability of 86.896 and Unit 82 has operated at an equivalent availability of 75.396 since completing (on schedule) an outage in June of this year which started in late 1985. In fact, NRC recently recognized CPRL's Brunswick Plant Manager for his ANS Award for Excellence in Nuclear Operations.

~Trainin CPdcL was the first multiplant utility to achieve membership in the National Academy for Nuclear Training by attaining INPO accreditation of all ten eligible training programs at its operating plants. Harris Plant also has more training programs accredited than other recent plants, with remaining eligible training programs to be accredited during hot operations.

This improving operational history combined with the above discussed indicators demonstrate that CPRL is pursuing policies that. will continue to produce improvements in our nuclear operations and strengthen our already strong commitment to safety.

(5096 JDK/vaw )

ACRS RECOMMENDATION(0):

The ACRS has on several occasions recommended that evaluations be made of the capability of light water nuclear power plants to be shut down safely in the event of an earthquake of greater severity and lower likelihood than the safe shutdown earthquake. In a letter dated January 11, 1983, the ACRS made recommendations concerning a possible broad approach to deal generically with the question of seismic margins. In the meantime, for the Shearon Harris Nuclear Power Plant, we recommend that, in addition to items already considered, specific attention be given to assurance of adequate seismic capability of the emergency AC power supplies, the DC power supplies, and small components such as actuators and instrument lines that are important to the accomplishment of safe shutdown and decay heat removal. We suggest also that specific attention be given to the adequacy of clearances between adjacent buildings.

CPRL POSITION:

The emergency AC power supply, including associated distribution equipment and raceway, and the DC power supply, including batteries and racks necessary for safe shutdown and decay heat removal are designed as Seismic Category I to function in the event of a Design Basis Earthquake. The review process to determine Category I components and equipment includes identification of:

1) the design basis accidents (e.g., loss of coolant accident) and the resulting environmental conditions,

2) the major safety functions required to mitigate the event (e.g., emergency core cooling),

3. the systems necessary to perform the primary mitigation function

'e.g., safety injection),

0. the components of those systems and the devices necessary to sense accident conditions and initiate actuation of the safety system, and

5. the auxiliary supporting features and components'required to support the primary safety function (e.g., component cooling water and electrical power).

Components within the Residual Heat Removal System (RHRS) meet General Design Criterion 30 and are qualified per IEEE-300-1975. Seismic margins for specific components associated with RHRS were presented to the ACRS previously. An example was the RHR heat exchangers which exhibit margins between 2 and 3 above the SSE. Smaller components such as actuators and instrument lines exhibit margins based on conservative test response spectrum used for qualification, in the case of actuators, and generic worst case seismic design for tubing and supports utilizing low damping valves and SRSS for forces.

Instrument tubing was designed and installed per ASME Section III from the root valve to the instrument. These aspects demonstrate margin based on standard seismic criteria. CPRL is actively supporting the Seismic Qualification Utility Group (SQVG) in which additional seismic margins are being demonstrated through experience data with similar equipment undergoing actual earthquakes. CPRL is also closely following the EPRI seismic margins research activity.

(5096JDK/vaw)

r

'L

Specific information concerning the adequacy of clearances between buildings was provided to the staff by CPRL in a letter dated 3uly 19, 1985. This letter contained a tabulation of clearances between various buildings and the maximum displacements at the interfaces during an SSE.

(5096JDK/vaw)

ACRS RECOMMENDATION(5):

During this review, there was discussion of the reliability and the fracture resistance of the chilled water system. The Applicants and the NRC Staff reported orally that the system is satisfactory in these respects. The ACRS would like a detailed discussion of the chilled water system in a supplement to the Safety Evaluation Report.

CPBrL POSITION:

CPRL is confident that the Harris Plant Essential Services Chilled Water System (ESCWS) is of high quality and is highly reliable. We have confirmed that the design and construction of redundant, single failure proof ESCWS systems has been carried out. This'includes the nuclear safety related purchase specification for piping, valves, and system components both mechanical and electrical. The piping and pressure retaining components of the ESCWS are designed ASME Section 3 and were fabricated, installed, and supported under CPRL's NRC-approved Quality Assurance Program. The entire nuclear safety related system is included in CPRL's Seismic and Dynamic Qualification of Mechanical and Electrical Equipment Program. The system has also been reviewed and included in the SHNPP fire protection program and protected against pipe failures.

CPRL, in a letter dated May 15, 1980, provided system information to the NRC in response to questions raised by the ACRS. Additionally in a letter dated January 23, 1980, Westinghouse provided a report on the fracture toughness of carbon steel pipe. CPdcL has also discussed portions of the system with an ACRS Subcommittee on June 27, 1986. In October 1986, CPRL provided through the NRC staff design drawings of the switchgear room air handling units that demonstrated the optimum orientation of the rotating fans and their motors in relation to the air handling units'oils and ESCWS entering and exiting the unit.

The reliability of the ESCWS is expected to be high and the continued operability of the system is assured via the plant Technical Specifications (3/0.7.13). The final detailed design basis document for this system is currently in revision and will be made available for subcommittee review, as desired, upon completion.

CPRL is confident that the ESCWS at the Harris Plant is well designed and constructed and will be maintained to provide reliable service throughout the life of the plant.

(5096 JDK/vaw )

ACRS RECOMMENDATION(6):

One of the confirmatory issues concerning this application is "turbine missiles."

Because of the nonoptimum orientation of the turbine relative to vital components in this plant, we recommend that a structured test program for evaluating overspeed protection of the turbine be prepared and submitted to the NRC Staff for review and approval before full power operation.

CPRL POSITION:

CPRL has provided the staff with information concerning the SHNPP Turbine Overspeed Protection System testing by letters dated November 21, 1986 and December 5, 1986. As discussed in these letters, the Turbine Overspeed Protection System is tested during preoperational testing (1-5015-P-01), power ascension testing (9103-5-27), and per Technical Specifications every 18 months by Operations Surveillance Test (OST-1075) and Maintenance surveillance Test (MST-I-319). In addition to Overspeed Protection System testing, these letters provided information concerning the design of the SHNPP Overspeed Protection System and the basis for specific overspeed setpoints.

(5096JDK/vaw)

t F

/