NRC-89-3419, Forwards Draft Response to NRC Questions 730.1 - 730.4 Re RESAR-SP/90 Pda,Module 2, Regulatory Conformance

Forwards Draft Response to NRC Questions 730.1 - 730.4 Re RESAR-SP/90 Pda,Module 2, Regulatory Conformance

ML20246P254
Person / Time
Site:	05000601
Issue date:	03/15/1989
From:	Johnson W WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Chris Miller Office of Nuclear Reactor Regulation
References
NS-NRC-89-3419, NUDOCS 8903280168
Download: ML20246P254 (11)

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4= w Westinghouse Energy Systems y,jeagay99,qnjea Electric Corporation Box 355 Pmsburgh Pennsylvania 15230-0355

, March 15, 1989 NS-NRC-89-3419 Docket No. STN 50-601 Document Control Desk U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Charles L. Miller, Director Standardization and Non-Power Reactor Project Directorate

Subject:

Submittal of Westinghouse Draft Response to Staff Questions 730.1 - 730.4 in Review of WAPWR RESAR-SP/90 PDA.

Ref(s) 1. T. J. Kenyon (NRC) letter to W. J. Johnson (Westinghouse),

dated February 2, 1989.

l 2. Letter No. NS-NRC-88-3343_(Addendum 7 to SP/90), Johnson l

(Westinghouse) to Rubenstiin (NRC) dated May 23, 1988.

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Dear Mr. Miller:

Enclosed are ten (10) non-Proprietary copies of the Westinghouse draft response to NRC questions 730.1 - 730.4 (Reference 1) in the Staff's review of the Westinghouse response to " Outstanding" Safety Issues (Reference 2) as applied to MAPWR RESAR-SP/90. The responses include H commitments to modify our original resolution of the subje::t issues, and these modifications will be included in our soon to be submitted update of RESAR-SP/90 PDA Module 2, " Regulatory Confromance" Note that the enclosed package is n9_t to be inserted in the RESAR-SP/90 PDA document, but, as noted above, will be part of our next amendment to Module 2.

Very truly yours, iw W. M J hnson, Manager Nu ha Safety Department l -

l l Enclosure (s)

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cc: Thomas J. Kenyon - NRC (MS 11H3)

Trevor Pratt - Brookhaven National Lab I

'8903280168 890315 PDR ADOCK 05000601 ,

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.. .. i WESTINGHOUSE CLASS 3 {

I DRAFT RESPONSE TO NRC OVESTIONS 730.1 - 730.4 RESAR-SP/90 MODULE 2

" REGULATORY CONFORMANCE"

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ATTACHMENT RE0 VEST FOR ADDITIONAL INFORMATION hh h l

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730.1 For future PWRs, the staff has recently concluded that PORVs'and l (GSI-70) block valves and associated controls for.these components should be  ;

constructed' to safety grade standards. This includes having a I minimum of two sets of PORYs and block valves including redundant and diverse control systems, being designed to Seismic Category I requirements, being environmentally qualified, having increased Technical Specifications surveillance requirements and inservice i testing requirements, and inclusion within the scope of a quality l assurance program that is in compliance with 10 CFR 50, Appendix l'

B. In addition, the requirements should include those improvements.

that were imposed subsequent to the TMI-2 accident, such as requirements to be powered from Class IE buses, and to have valve position indication in the control room.

By letter dated May 23, 1988, Westinghouse provided its resolution of GSI-70 as it applies to the RESAR-SP/B0 design. In addition to this information, the staff requires Westinghouse to provide a commitment to providing safety grade PORVs and block valves, and associated controls which include, 'as a minimum, the following features.

o There must be a minimum of two PORVs and block valves, including associated redundant and diverse control systems, o The equipment must be designed to Seismic Category 1 l requirements.

o The equipment must be environmentally qualified.

l o Increased inservice testing must be provided in accordance with applicable ASME Section XI rules.

o The equipment .must be included within the scope of a quality assurance program that is in compliance with 10 CFR- 50, Appendix B. That is, "PORVs and block valves will be included in the safety related 0 List".

o Increased Technical Specifications surveillance requirements must be provided to insure PORV and block valve operability.

RESPONSE

l l The commitments requested by Staff are, for the most part, currently reflected in the RESAR-SP/90 PDA document. For example, WAPWR-RC 730-1 MARCH, 1989 8751e:1d

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'*%o e D see Subsection 5.4.7 and Question / Response 140.Q, / .. %am 82 m ofq RE'5&R',Sl g90 PDA Module 1, " Primary Side Safeguards , abjub Bon 5.4.13 of 'RESAR-SP/90 PDA Module 4, " Reactor. Coolant System."

However, the Westinghouse response to GSI-70, "PORV and Block Valve Reliability," will be revised for our update of RESAR-SP/90 PDA Module 2 " Regulatory Conformance," to provide the following commitments:

o There will 'eb a minimum of two PORVs and block valves including redundant and diverse control systems.

o The PORVs and block valves will be designed to Safety Grade, Seismic Category I requirements..

o The design will allow .for environmental qualification of the PORVs and block valves.

o The design of the PORVs and block valves will allow inservice testing in accordance with Article IWV-3400 of ASME Section'XI.

o The PORVs and block valves will be included in the final safety related 0-list of Table 3.2-1 of RESAR-SP/90 PDA Module 7,

" Structural / Equipment Design."

4 o Surveillance requirements will be included in the FDA Standard Technical Specifications to ensure PORV and block -valve operability.

1 (730.2 In response to the staff's Question 210.71 relative to periodic

((GSI-105) leak testing of pressure isolation valves, Westinghouse referenced its response *; Question 210.38. In that response, a commitment is.

I made to perform periodic leak testing of all pressure isolation

, valves in accordance with the requirements in the Westinghouse Standard Technical Specifications. The staff concludes that this is an acceptable commitment regarding Generic Issue 105 for the PDA i Phase of RESAR-SP/90. The response to this issue in the May 23, 1988 Letter on USIs/GSIs should be revised to reflect this eommitment. J l WAPWR-RC 730-2 MARCH, 1989 s7si..le l

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.. . ',if In addition. the Westinghouse response to Gener' s s that the ISS interconnections to the RCS are ei er vent irec y back to the in containment EWST or consist of 2000 PSIG design pressure piping. Are there any other low pressure systems which are connected to the RCS that do not have either of these features?

RESPONSE

A commitment to perform periodic leak testing of all pressure isolation valves, in accordance with the requirements in the Westinghouse Standard Technical Specifications, will be added to our response to GSI-105, " Interfacing Systems LOCA at LWRs", in our update of RESAR-SP/90 PDA Module 2, " Regulatory Conformance."

In response to the second part of this question, there are other low pressure systems which are connected to the RCS. These connections are such that leak testing similar to that performed on ECCS check valves is not required. A brief description of these connections is provided below:

o RCS drain and vent connections, and the ISS emergency letdown lines are provided with redundant closed isolation valves.

Since these lines do not penetrate the containment, the failure of both isolation valves would not only be a low probability event, but would not result in an "intersystem LOCA."

o Sampling system lines both penetrate the containment and may be open during normal operation. These piping connections are provided with a 3/8-inch flow restrictor such that failure would not require safety system actuation. These lines contain redundant and automatic containment isolation valves.

o The CVCS letdown line is normally open during plant operation.

This line contains redundant valves (normally open) which will automatically close on low pressurizer level, and redundant and automatic containment isolation valves.

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o The CVCS alternate letdown line conta s closed isolation valves, redundant a fn fu a on 11y ent isolation valves, an operator controlled throttle valve and a temperature sensor, o The CVCS charging line contains multiple chack valves and redundant isolation valves.

All piping connections up to and including their isolation valves are designed for full RCS prersure and temperature conditions.

730.3 By letter dated May 23, 1988, Westinghouse stated that it has (GSI-113) assured high reliability of its snubbers by use of stringent requirements in its equipment specifications to resolve GSI-113.

Westinghouse further stated that for the design of the RESAR-SP/90, it advocates the implementation of -environmental and dynamic qualification of its snubbers as well as inservice testing (to the extent that state of the art testing equipment is adequate) and surveillance. Before the staff can complete its evaluation of the applicant's response on this issue, provide the following additional information:

1. Clarify how the use of stringent requirements in equipment specifications has assured high reliability of large bore hydraulic snubbers. Describe what reliability is assumed for LBHSs in the analyses of applicable systems.

2. Clarify your statement regarding the advocation of the implementation of environmental and dynamic qualification, and inservice testing and surveillance of snubbers. Provide more detailed information explaining how these actions will resolve this 1ssue.

RESPONSE

730.3, ITEM 1 Westinghouse equipment specifications help assure LBHS reliability.

by defining a complete set of fabrication and functional testing requirements. The Westinghouse equipment specifications define the l snubber design requirements, environmental requirements, operability requirements, functional requirements, and establish a process that ensures Westinghouse engineering controls and reviews WAPWR-RC 730-4 MARCH, 1989 87sle.1d

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'q r gy, 49 all vendor procedures and processes us he ring cycle. Each step of the fabrication an shop assembly of the 3 snubber, including the review and approval of all procedures, fabrication sequences and test results, is closely monitored by Westinghouse engineering and quality assurance personnel to insure all requirements and testing criteria are adhered to.

Additionally, each snubber must pass numerous performance qualification tests (i.e., control valve lockup, piston bleed  ;

rates, load rating, spring rates, piston drag, etc.) prior to its acceptance and release for shipment.

The Westinghouse equipment specifications also require the vendor to provide functionality analyses, design reports, fluid / seal compatibility analyses, environmental qualification data for " soft j parts", installation procedures, spare parts lists, and recommended l l 1 l maintenance practices. Westinghouse's attention to detail and j l

complete definition of snubber design, function, operability, and j l maintenance requirements help assure a high reliability in LBHSs.

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The steam generator LBHSs are assumed to perform their intended function for the 40 year opert. ting life of the plant. The snubbers are required to provide minimum resistance against normal reactor coolant loop thermal expansion (drag less than or equal to 5000 l lbs. per snubber) and to " lockup" at velocities greater than or equal to 6 inches per minute. In the locked. condition the snubbers are required to bleed at 0.15 to 0.25 inches per minute at full faulted rated load. In addition, the snubbers are to remain fully functional when exposed to a radiation level of 25 RAD /hr, 70%

hum;dity, and 120 degrees Fahrenheit long term operation and 300  !

degrees Fahrenheit short term operation.

Westinghouse has been defining the requirements for and procuring steam generator LBHSs for over twenty years. The technical and material advances in LBHS technology that have taken place have 4

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l been evaluated and adopted into the Westinghouse requirements for LBHSs. This knowledge will,be implemented in the SP/90.

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The majority of LBHS procured by Westinghouse

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l and manufactured by Paul-Munroe Hydraulics. aul-Munroe steam generator LBHS has implemented design features which help ensure the reliability of the snubber. Features such as a self cleaning bleed orifice end a corrosion resistant chrome lined carbon steel cylinder contribute to the reliability of the snubber. Additional information on the reliability of Paul-Munroe steam generator LBHSs is provided in Attachment-A 730.33 ITEM 2 Currently dynamic testing of production scale steam generator snubbers is not performed. As such, Westinghouse requires a rigorous series of static function,a1 tests (drag, lockup, bleed) to be performed on each snubber. These tests, although not dynamic, verify the operability of the snubbers at their rated load capacity. Until recently, the means of dynamically testing these snubbers was not available. However, Paul-Munroe has performed a limited number of dynamic tests of steam generator snubbers. A drop weight test with an equivalent loading exceeding the rated load has been performed. Small scale dynamic testing has been performed including tests for a cyclic force of 90 kips applied at up to 16 Hz. and a cyclic force of 450 kips applied at up to 11 Hz. These tests represent the approximate limits of current testing facilities. The results of these tests demonstrated that the Paul-Munroe snubbers performed as predicted during dynamic loading conditions.

Due to limits on the capacity of the available testing facilities, dynamic tests of steam generator snubbers for the full rated load and a full range of frequencies (1-33 Hz.) is not possible.

Westinghouse however, does support the future development of such facilities that will permit this type of testing and qualification of the steam generator snubbers.

Currently, inservice inspection requirements of steam generator snubbers is defined by individual plant technical specifications. I WAPWR-RC 730-6 MARCH,1989 87sle.1d

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j Many of these technical specifications r e 1 e a st requirements of ASME's OM4 inservice et n s ng-criteria. Westinghouse believes that a certain amount of inservice surveillance (leakage and fluid level checks) is required to provide added assurance that the snubber will continue to function as intended during plant operation.

730.4 By letter dated May 23, 1988, Westinghouse provided a discussion (GSI-135) on GSI-135 which concluded that its steam generator overfill  !

protection system is designed to divert flow to the Emergency Water Storage Tank if the high steam generator water level setpoint is ,

reached. Although this system feature will reduce the probability  !

of the occurrence of a steam line overfill, the current staff position is that a part of the basis for evaluating plant spe:ific j responses to this issue is a commitment from each applicant or- '

licensee that it has evaluated th'e structural adequacy of the main '

steam lines and associated supports under water-filled conditions  !

as a result of steam generator tube rupture. The RESAR-SP/90 i response to Generic Issue 135 should be revised to provide such a commitment. .

l RESPONSE 1

The Westinghouse response to Generic Issue 135 will be revised in I l our update of RESAR-SP/90 PDA Module 2, " Regulatory Conformance,"

as shown below, to provide a commitment to evaluate the structural adequacy of the main steam lines and associated supports: j i

"Although the features described above will reduce the probability of the occurrence of a steam line overfill, Westinghouse will evaluate the structural adequacy of the main steam lines and associated supports under water filled conditions as a result of steam generator tube rupture. This evaluation will assume that the main steam line is water filled from the steam generator out and the first isolation valves are at main steam system design pressure and temperature. This load condition will be evaluated against Service Level D Allowables".

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Attachment-A (730.3)

Hydraulic Snubber Reliability l fg,'

The Paul-Munroe hydraulic snubbers used for steam generator upper lateral supports are reliable. The reliability of the Paul-Munroe hydraulic snubbers has been demonstrated by their effective performance over 1900 snubber years without impact to plant operation.

Westinghouse has primarily used Paul-Munroe hydraulic snubbers in the Westinghouse reactor coolant loop support system. To date, Westinghouse has accumulated over 1000 snubber years of trouble-free operating experience of Paul-Munroe snubbers in Westinghouse designed reactor coolant loop support systems. Throughout this operating service life, Westinghouse is not knowledgeable of any instance in which a Paul-Munroe steam generator hydraulic snubber was found to be inoperable or cause for a plant shutdown.

Inoperability is defined as the snubber'not being able to perform the minimum plant specific functional requirements of locking-up and bleeding (translating under load). To update and confirm Westinghouse knowledge of Paul-Munroe snubber operating experience, Westinghouse recently conducted a limited survey of domestic operating plants with a Westinghouse reactor coolant loop support design and Paul-Munroe steam generator snubbers. In every case the Paul-Munroe snubbers were reported to have experienced no operational problems.

In addition, a domestic utility independently contacted INPO to perform a sort on the INP0 Licensing Event Report, LER, data base to identify all reported failures of Paul-Munroe hydraulic snubbers. The INPO report identified only three LERs referencing Paul-Munroe hydraulic snubbers, j

1. Diablo Canyon Unit 1 - On 9/13/86 with Unit 1 in mode 5, during inservice inspection one steam generator snubber was found to be unpinned at the blind end. All other snubbers were found to be 1 l functional. The cause of the pin movement is unknown. A design change which provided the pins with locking devices was implemented to prevent repetition.

The SP/90 steam generator snubbers will be provided with pins which feature a positive locking device, cotter pins.

2. Calloway Unit 1 - On 3/8/86, during inservice testing of one steam I generator snubber, one hydraulic control valve failed causing the snubber to lock-up. The cause of the failure was attributed to the in plhee test equipment creating an unnatural high rate of cycling on the control valve and thereby causing the valve stem to shear. The snubber was replaced and sent to Paul-Munroe for refurbishment. .

Paul-Munroe has isolated the problem that caused the control valve failure and have modified their in place test equipment to prevent any

( reoccurrence. The control valve failure at Calloway is unique to the inservice testing performed on that snubber using original Paul-Munroe test equipment and is no indication of any generic service life problem with the control valve.

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3. Arkansas Nuclear One Unit 1 -

On 11/9f%, t shutdown, a 200 kip capacity reactor coolant pu d tul sn I was found to be leaking fluid. Examination of the snubber revealed arc strikes on the piston rod. These appeared to have been the result of the snubber being used as a ground for a welding machine. The cause of the leak were scratches on the TEF2EL piston rod seal  !

apparently caused by the arc strikes. The snubber was disassembled, i cleaned, inspected, damaged parts were rebuilt by Paul-Munroe and the snubber was reassembled with new seals.

1 The SP/90 steam generator snubbers will also use TEFZEL seals.

However, the installation, maintenance, and inspection procedures at i SP/90 will preclude misuse of the snubbers and identify any abnormal l snubber leakage.

As additional confirmation on the reliability of Paul-Munroe hydraulic snubbers, Paul-Munroe Hydraulics was contacted and requested to provide a summary of their knowledge of the operating service life of Paul-Munroe ,

hydraulic snubbers similar in load capacity to the steam generator snubbers.

Paul-Munroe entered into the snubber production of large bore (250 to 2000 kip capacity) snubbers in 1969. Since that time, Paul-Munroe has become the major manufacturer of large bore snubbers worldwide with over 1400 snubbers supplied. In addition, Paul-Munroe has been the main supplier of steam generator snubbers for both Westinghouse and Combustion Engineering reactor coolant support systems.

Paul-Munroe has used advanced technology in its snubber seal, pc'ppet-style control valve design, and in pirce testing capability. The Paul-Munroe TEF2EL seal has been an extremely reliable design and is fully qualified for 40 year life performance. The poppet-style control valve provides consistent  ;-

performance and is not subject to the problems experienced by conventional '

control valves. The poppet-style control valve has the bleed orifice in-line with the poppet thus eliminating silting or clogging conditions by r.ature of this self cleaning design. The control valves are set in the factory and are i not subject to field tampering. The in place testing capability allows the '

snubbers to be tested for operability in place. The SP/90 steam generator hydraulic snubbers incorporate all these advanced design features.

Paul-Munroe large bore snubbers have provided nuclear facilities with a total, I effective performance of over 1,900 snubber years of operating service without impact to plant operation. Paul-Munroe has over 390 snubbers that have been in operating service for over 4 years.

Paul-Munroe's statements are supported by EPRI's " Snubber Reliability Improvement Study", EPRI NP-2297, which states that " ..no generic problems have been reported for their (Paul-Munroe) snubbers due to storage, installation, field adjustment, deterioration in service, inspection, or maintenance.". inservice In conclusion, the 1900 plus snubber years of operating experience with no reports of inoperability demonstrate the reliability of the Paul-Munroe steam generator hydraulic snubbers.

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