ML19343C606

Safety Evaluation Supporting Amend 37 to License DPR-61

ML19343C606
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	02/26/1981
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML19343C599	List: ML19343C596 ML19343C604 ML19343C606 ML19343C612 ML19343C613
References
NUDOCS 8103240585
Download: ML19343C606 (51)

Text

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\\j UNITED STATES y ',,q ( i NUCLEAR REGULATORY COMMlliSION

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E WASHINGTON, D. C. 20S55

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMEN 0 MENT NO.37 TO FACILITY OPERATING LICENSE NO. DPR-61 CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NEC.V PLANT DOCKET NO. 50-213 1.0 Introduction On April 23, 1976, the Commission sent a generic letter to Connecticut Yankee Atomic Power Company (the licensee, CYAPCO) advising them that the inservice inspection and testing requirements for ASME Co e : lass 1, 2 and 3 components for nuclear power plants delineated in 10 CFR Part 50.55a were changed by a revision to the regulations published on February 27, 1976. The revised regulations require inservice inspection and testing to be performed in accert ance with the examination and testing requirements set forth in Section XI of the ASME Boiler and Pressure Vessel Code and Addenda thereto.

To avoid poten-tial conflicts between the ASME Code requirements and the Technical Specifica-1 tions presently in effect for the Haddam Neck Plant, we also advised the licensee that he should apply to the Commission for amendment of the Technical Specifications.

Sample language for such Technical Specification changes was provided as an enclosure.

Our let'er of April' 23,1976 also advised the licensee that if he determines that conformance with certain ASME Section XI inservice inspection and testing requirements is impractical, he should submit information to the Commission to support his determination in accordance with 50.55a(g)f5)(iii) and (iv).

By letter dated November 30, 1976, we provided additional guidance in preparing inservice inspection and testing program descriptions and associated relief requests.

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By letter dated June 29, 1977, the licensee requested a change to the Technical Specifications (Appendix A) appended to Facility Operating License No. DPR-61 for the Haddam Neck Plant.

The proposed amendment and revised Technical Specifications would delete the present inspection and testing requirements in Section 4.10 of the Technical Spacifications and substitute the sample language enclosed with our letter of April 23, 1976. The proposed Technical Specifica-tions would require all inspection and testing to be performed in accordance with the ASME Code except where specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).

The licensee's letter of June 29, 1977 also submitted a proposed Inservice Inspection and Testing Program which includes requests for relief from examin-ing certain components where the licensee determined that it was impossible or impractical to examine or test the specific component because of design, geometry, materials of constr.ction, or operating limitations.

Subsequent letters dated May 26, 1978, May 30, 1978 and July 14, 1978 revised the Inservice Inspection and Testing Programs and the requests for relief, and as a result of a meeting with the staff on January 17-19, 1979, the licensee submitted your addi 'onal revisions on April 27, 1979, June 29, 1979, March 25,1980, ar.d h vember 4,1980.

EVALUATION 2.0 Technical Soecifications The changes proposed by the licensee to the Technical Specifications are identical to the sample Technical Specifications enclosed with our letter of April 23, 1976.

The revised Technical'. Specifications require all inspections and testing to be performed in accordance with the ASME Boiler and Pressure Vessel Code and are, therefore, acceptable.

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3.0 Requests for Relief As required by 10 CFR 50.55a(g), the licensee has updated the Inservice Inspec-tion Program for the Haddam Neck Plant to the requirements of the 1974 Edition through Summer 1975 Addenda of Section XI of the ASME Boiler and Pressure Vessel Code (B&PV Code).

Based on information contained in the revised sub-mittals dated April 27, 1979 and June 29, 1979, it has been determined that certain requirements of the Code cannot be implemented at the facility because of component or system design, geometry, materials of construction, or opera-tional limitations.

Requests for relief from those requirements have been reviewed and evaluated by the staff and our determinations to grant or deny the requests, pursuant to 10 CFR 50.55a(g)(6)(i), are documented below.

3.1 Inservice Inscection 3.1.1 Class 1 Comoonents 3.1.1.1 The licensee requests relief from volumetric examination of the Closure Head Peel Segment-to-Disc Circumferential Weld.

(Item Bl.2, Examina-tion Category B-B)

Code Requirement: Volumetric examination is required. The areas shall include the circumferential welds in the vessel heads.

This i,cludes weld metal and base metal for one plate thickness beyond the edge of the weld.

The examinations performed during each inspection interval shall cover at least 5 percent of the length of each circumferential head weld.

For welds on the reactor vessel, examination may be performed at or near the end of each inspection interval.

Licensee Basis-for Relief Request: The closure head peel segment is completely enclosed within the pattern of CROM penetrations inside the shroud and is not accessible for examination as required by IWB-2601.

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During the February, 1979 refueling outage, an investigation was made by CYAPCO and Westinghouse personr.el to determine inspectibility of the RPV hest peel segment welds. As can be seen from the sketch, Reference Bl.2,* access for the examination of these welds is prohibited by their location relative to the.CRDM penetrations, the permanent insulation and the head ventilation shroud sections. We have also attached photographs of the actual installation which shows the congestion in this area.

Inspection from the inside (clad) surface is not possible because of surface irregularities, and high radiation.

Integrity of these welds will be verified during the periodic pressure tests.

Evaluation: The design of the closure head and control roc drive penetration locations prevent volumetric examination of the Closure Head Peel Segment-to-Disc Circumferential~ Weld. As an alternate and continuing inspection of the weld, the licensee has preocsed to verify the integrity of these welds duri periodic pressure tests.

This is interpreted to be a visual inspection of these welds during each Code required (IWB-5200) system leakage test prior to startup following each reactor refueling outage.

The test will be performed at not less than the system nominal operating pressure at 100% rated react:r power. Other welos on the closure head are examined to code requirements and are subject to additional examinations if unacceptable indications are revealed.

The visual inspection of the Closure Head Peel Segment-to-Disc Circumferential Weld during the system leakage test at each refueling outage and acceptable results from volumetric examination of other closure head welds will provine assurance of the continued structural integrity of the Closure Head.

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Conclusion:==

The licensee has demonstrated that implementation of the Code requirement at the facility is impractical because of the design of the Closure Head.- The staff finds that the proposed alternative examination of the reunar vessel Closure Head Peel Segment-to-Disc Circumferential Weld will provide an acceptable level of safety and therefore concludes that relief from the volumetric examination requirement may'.be granted.

• Letter from licensee to NRC dated April 27, 1979 4

3.1.1.2 The licensee requests relief from 100% volumetric examination of the Steam Generators (Primary Side) Nozzles-to-Safe End Welds (8).

(Item B3.3, Examinati an Category B-F).

Code Reouirement:

Volumetric and surface enninations performed during each inspectiori interval shall cover the circumference of 100 percent of the welds.

The areas shall include dissimilar metal welds (e.g., safe-end welds) between combinations of carbon, low alloy, or high tensile steels and stainless steels, nickel-chromium-iron alloys, nickel-copper alloys.

This shall include the base material for, at least, one wall thickness bemnd the edge of weld.

The configuration of the (8) steam generator nozz co-safe-end welds prevents a complete volumetric examination from being performed. The carbon steel nozzle casting on one side and the geometry of the safe-end forging on the side prevent full ultrasonic coversge from being possible.

To supplement the limited 42 refracted L wave examinations of the weld, transducers of other characteristics will be utilized to the extent practical, as will surface examination techniques.

It will be possible to perform a Code examination of about 60% of this weld design.

Evaluation:

Because of the geometry of the eight (8) steam generator nozzle-to-safe-end welds, only approximately 60% of each nozzle-to-safe-end weld can be volumetrically examined.

The carbon steel nozzle casting on one side and the safe-end forging on the other prevent full ultrasonic coverage from being performed. The licensee proposes to supplement the limited ultrasonic exam-inations with transducers of other-characteristics in order to improve the flaw detection capability.

Surface examination of the welds will be performed'as required by Code. The volumetric examinations which will be performed on the safe-end welds will cover sufficient percentage of the required voltme to provide assurance that the nozzles will not contain unacceptable inservice flaw. The required surface examination will provide further assurance that inservice flaws penetrating the surface of the weld or heat affected zone will be detected.

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Conclusion:==

The staff finds that the proposed examinations will provide sufficient flaw detection capability to maintain the safe end welds in a structurally acceptable condition.

The staff therefore concludes that relief from the 100% volumetric requirement may be granted.

.1.1.3 The licensee requests relief from 100% volumetric examination of the P.egenerative Heat Exchanger integrally-welded supports.

(Item B3.7, Examination Category B-H).

Code Recuirement:

In the case of vessel support skirt, the volumetric examina-tion performed during each inspection interval shall cover at least 10 percent of the circumference of the weld to the vessel.

In the case of support lug attachments, 100 percent of the welding to the vessel shall be examined.

The areas shill include the integrally welded support attachment (e.g., support skirts).

Thi3 includes the welds-to the vessel and the base metal beneath the weld zone and along the support attachment member for a distance of two support thicknesses.

Integral support pads _an nozzles are excluded.

~ icensee Basis for Relief Reouest: The integrally welded supports are attached L

by fillet welds.

The c' ' figurations of.such welds is such that examinations cannot be performed to one extent required by IWB-2600 and only the base material of the component wall can be examined by ultrasonic techniques.

Surface examination will be performed on integrally welded attachments to supplement the volumetric examination.

Evaluation: The design of the support-to-heat exchanger attachment weld prevents performance of a meaningful ultrasonic examination as required by the Code. The licensee-has proposed to volumetrically. examine the base material and to surface examine the weld and portion of the base material which cannot be volumetrically examined.

Because of the design and conditions to which these welds are subjected, surface flaws are most likely to be gen-erated. The examination techniques proposed by the licensee are acceptable for detection of such flaws.

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Conclusion:==

The staff finds that the proposed surface examination on areas inaccessible to perform volumetric examination will provide assurance that the supports' structural acceptance is maintained.

The staff therefore concludes that relief from the Code requirement may be granted as requested.

3.1.1.4 The licensee requests relief from 100% volumetric examination of the circumferential butt weld Number 1 in Loop 1 of the Safety Injection System.

(Item 84.5, Examination Category B-J).

Code Requirement: The volumetric examinations performed during each inspection interval shall cover all of the area of 25% of the circumferential joints.

The areas shall include circumferential welds and the base metai for one wall thickness beyond the edge of the weld.

Licensee Basis for Recuestina Relief:

Limitations may occur for tne examination of piping system circumferential butt welds (Category B-J) when the welds occur at geometric discontinuities such as pipe-to-fitting welds and fitting-to-fitting welds.

For pipe-to-fitting welds, examinations can generally be per-formed to the extent required by T-532 of Section V from the weld and pipe surfaces, providing the weld surface is suitably ground.

Examination from the fitting side would be de,,erJent upon the geometric configuration. Where elbows or tees are concerned,

e. amination can be performed from the fitting side except where the intrados'of the fitting prevents adequate ultrasonic coupling.

No examinations can be performed from the fitting side when it is a valve or a flange.

In all cases,100 percent of the weld material can be examined.

In instances where welds occur at fitting-to-fitting, access restrictions as out-lined above occur on both sides of the weld, in these instances, surface examination would be performed to supplement the limited volumetric examinations.

Evaluation:

Full compliance with the Code required volumetric examination cannot be achieved for the weld listed because of the limited access or geometry of the' fitting-to-fitting configuration.

The licensee proposes to volumetri-cally examine the weld to the extent practical and to supplement the volumetric examination with a surface examination.

The weld and base metal will be examined 100%. ~The limited volumetric examination in combination with surface 7

examination will provide the capability and confidence of detection of inservice flaws.

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Conclusion:==

The staff finds the combination of volumetric and surface

-examination of the welds listed acceptable in providing assurance of the struc-tural integrity of the welds and concludes that relief from 100% volumetric examination may be granted.

3.1.1.5 The licensee requests relief from 100% volumetric examination of integrally welded supports attached by fillet welds to piping.

(Item B4.9, Examination Category B-K-1).

Code Reouirement:

The volumetric examinations performed during each inspection interval shall cover 25 percent of tne integrally welded supports.

The areas-shall~ include the integrally welded external support attachments.

This includes the welds to the pressure-retaining boundary and the base metal beneath the weld zone and along the support attachment member for a distance of two support thicknesses.

Licensee Basis for Reuest:

The integrally welded supports are attached by fillet welds. The configurations of such welds are such that examinations cannot be performed to the extent required by IWB-2600 and only the base mater-ial of the component' wall can be examined by_ ultrasonic techniques.

Surface examination will be performed on integrally welded attachments to supplement

-the volumetric examination.

Evaluation: Because of the attachment weld design, ultrasonic or radiography is an impractical technique to use and, if applied, would result in little added assurance of safety. The licensee has committed to subject these welds to surface examination and to volumetrically examine the base metal.

Based on environmental and loading conditions of these types' of welds, flaws would most likely generate at the weld surface and thus be detectable by surface examination. Ultrasonic examination of the base metal would provide assurance that flaws do not exist in the heat affected zone.

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Conclusion:==

The staff finds that the examination techniques to be used by the licensee will provide assurance that the integrally-welded supports continue to be structurally acceptable and therefore concludes that relief from the 100%

-volumetric examination requirement may be granted.

3.1.1. 6 The licensee requests to defer performing volumetric examination of a reactor coolant pump casing welds and visual examination of a reactor coulant pump internal pressure boundary surface until the 1981 refueling outage.

(Items B5.6 and 5.7, Examination Categories B-L-1 and B-L-2, respectively).

Ccde Recuirement: Category B-L-1:

The volumetric examinations performed during each inspection interval (10 years) shall include 100% of the pressure-retaining welds in at least one pung in each group of pumps performing similar functions in system.

The examinations may be performed at or near the end of the inspection interval.

Category B-L-2:

Tne internal pressure boundary surfaces of one pump in each group of pumps performing similar functions in the system shall be visually examined during each inspection interval.

This examination may be performed on the same' pump selected for the' Category B-L-1 examinations.

The visual examinations may be performed at or near the end of the inspection interval.

" Licensee Bases for Reouesting Relief: ASME Section XI, Table IWB-2500, examination Categories B-L-1 and B-L-2, requires that one of the four RCP's be disassembled and inspected at or near the end of each inspection interval.

Accordingly, these examinations were scheduled for. the 1980 refueling period.

~ Westinghouse (W), the pump manufacturer, has only recently presented informa-tion to CYAPCO which has resulted in our determination that-it would be prudent to delay pump disassembly.

o Pump design did not provide for disassembly of fixed internals, as there were no inservice inspection requirements for any reactor system components until 1970,_ over two years af ter the Hadcam Neck Plant went into service.

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4 These pumps were designed to provide reliable service for the plant life-time without internal maintenance or inspection.

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.The reactor coolant pump casing consists of four Type 316 stainless steel cast rings with minimum thickness of approximately eight inches.

Con-sicering the stainless steel Type 316 material characteristics and the water. chemistry limitations for the reactor coolant system, the potential for any stress corrosion mechanism is considered to be negligible.

There-fore, service induced flaws would most likely be generated from a cyclical

-loading mechanism such as fatigue; however, the pump bowl geometry as well as the endurance limit of the given material make this, also, a negligible concern.

- o Proper planning of pump disassembly-requires construction of a mockup, development of special tools, and detailed planr,ing of every step of the procedure to minimize man-rem exposures.

Radiation levels will be very high - estimated at 60-80 rem at the pump impeller prior ta decontamination.

Even after decontamination, totu exposures of 400-600 man-rem are considered probable.

o Category B-L-1 requires a volumetric examination of the three pump casing pressure-retaining welds.

It-is extremely doubtful that source radio-graphy would be successful because of the high background radiation, even

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if the welds can be located accurately from inside the pump. By delaying the inspections, it is highly probable that the portable Linac unit being

! developed by EPRI will be available for use.

It is our judgment that more acceptable. radiography would be'possible with the Linac.

o There are also risks involved with the mechanics of disassembly.

CYAPC0 has' no spare parts for fixed internals and W has very few in stock.

If a part is damaged during disas;embly or reassembly, a replacement may not be available.

o Complete disassembly of an.in-service RCP of the Haddam Neck Plant's design. has never been accomplished anywhere in the industry. W informed 10 e

CYAPC0 u.at removal of the casing adapter, will be difficult if there has been any significant amount, cf warpage, and reassembly will be even more aifficult, if not impossible.

o After disassembly, the interior surfaces of the pump casing must be cleaned by hydrolaser or an equivalent method, as there is most probably an oxide buildup which'will mask the location of the seam welds.

CYAPC0 is unsure how effective this will be.

High radiation levels will require that a television camera be used to locate the welds and position the source tube for radiography.

If the casing radiation levels cannot be reduced to lower than 700 mrem / hour, source radiography will not be effective.

As an alternate to the B-L-1 and B-L-2 examinations for the 1980 refueling, CYAPC0 intends to visually inspect the pump casing in accordance with IWA-5000 and IWB-5000 during the performance of the system hydrostatic pressure test.

It is, therefore, _our position that the B-L-1 and B-L-2 reactor coolant pump examinations, original.ly scheduled for this year, will be deferred until the 1981' reactor refueling outage-to allow CYAPC0 to procure the necessary tools and equipment required for a meaningful inspection.

Evaluation:

The problems which would be encountered in attempting to perform

.the required examinations during the-1980 refueling outage make the require-ments impractical.

The licensee has requested to defer the requirements until the next refueling outage at which time use can be made of recently developed portable radiography equipment. As an interim measure, the licensee will visually ~ examine the pump casing for signs of leakage during the pressure and leakage -tests required following a refueling outage.

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Conclusion:==

The staff has determined that the reactor coolant pumps' design, materials, and-operation would limit significantly the parameters known to be

' causative factors of pressure boundary failure.

However, the licensee is monitoring.the pumps. structural integrity by performing visual-examinations at pressure after refueling.

These examinations will provide adequate assurance-11 e

of the pump casing structural integrity.

Therefore, the staff concludes that relief from the volumetric examination of the casing welds and visual examin-ation of the internal pressure boundary surfaces may be granted as requested.

3.1.1. 7 The licensee requests to examine 100% of one head-to-shell and 100%

of one shell-to-tubesheet weld in lieu of a smaller percentage of each well (head-to-shell and shell-to-tubesheet) in the Regenerative Heat Exchanger.

(Item B3.1, Examination Category B-B).

Code Reouirement:

The volumetric examirations performed during each inspection interval shall cover at'least 10% of the length of each longitudinal shell weld and meridional head well and 5% of the length of each circumferential shell weld and head weld.

Licensee Basis for Reouestino Relief:

Examination Category B-B requires that for non exempt Class 1 vessels, at leatt 10% of the 'ength of each longitudinal weld and 5% of the length of each ci cumferential weld be examined volumetri-cally each interval.

The regenerative heat exchanger is a three pass vess 1, having a total of six head-to-shell welds.and-six shell-to-tubesheet welds.

In view of the high radiation levels (up to 10 rem / hour) associated with this vessel, it is proposed that examinations be revised to require that 100% of the total of one head-to-shell weld and 100% of one shell-to-tubesheet weld be inspected during each 40-month period rather than a much smaller percentage of each of twelve welds, each interval.

This revised extent of examination would enable CYAPCO to reduce personnel radiation exposure while inspecting many more inches of weld than required by Code.

The requirements of IWB-2430 con-cerning additional examinations would be adhered to if indications are revealed which exceed the allowable standards.

Evaluation: Because of the relatively high radiation levels in the areas required to be examined and the small percentage of the length of each head-to-shell and shell-to-tubesheet weld to be examined, the staff finds that the requirement is impractical.

The licensee has' proposed to examine 100% of the total length' of' one head-to-shell weld and 100% of the length of one shell-to-12

s tubesheet weld.

The selection of the welds and the percentage of the weld length examined will provide a more meaningful and representative sample for detection of service-induced' flaws in the welds than that required by Code.

Deviation from the Code required weld selection and percentage of weld length

. to be examined will not decrease the capability of flaw detection since the welds selected in the proposed examination plan will be subjected to the same inservice conditions.

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Conclusion:==

The staff concludes that the examination proposed by the licensee will provide adequate assurance that service-induced flaws in the Regenerative Heat Exchanger will be detected and the structural integrity of the component maintained, that the proposed examination will not significantly decrease the plant's margin of safety, and that, therefore, relief from the Code requirement as requested may be granted.

3.1.1. 8 Request to defer the visual examination of clad patches in two steam generators and the pressurizer to the 1981 reactor refueling outage.

(Items B3.8 and 2.9, Examination Category B-I-2).

Code Reouirement:

The visual examinations performed during each inspection interval shall cover 100% of the patch areas.

The areas shall include at least one patch-(36 square inches) near each manway in the primary side of the vessel.

- The examination of.the patches may be performed at or near the end of the inspection interval.

Licensee Basis for Requesting Relief:

Examination Category B-I-2, ASME Section XI, Summer, 1975 Adden'dum, requires that clad patches in the four Steam Generator primary waterboxes and pressurizer be visually inspected at or near the end of.each inspection interval.

In that CYAPC0fintends to open only two of the Steam Generators for tubing eddy current examinations, it is intended to defer the inspections on the pressurize'r and the two Steam Generators not hdl

.sc e u ed for ECT to the'1981 reactor refueling outage.

Experience at the Haddam Neck Plant has shown these areas to De free of problems. The primary waterboxes of all four' Steam Generator 3 were entered in 13

the 1979 reactor refueling outage, but a formal inspection of cladding was not made. The pressurizer cladding was inspected furing the 1973 refueling but CYAPC0 does not have a valid inspection report for this examination.

In that there are no plans to enter this component during the forthcoming shutdown, and the man-way is seal welded, it is our intention to eliminate the radiation

-exposure associated (estimated at 2-5 man-rem) and delay the examinations until 1981.

-Evaluation:

To remove seal welds and subject examination personnel to the relatively high radiation levels in the Steam Generator primary waterboxes and

-the pressurizer solely for the visual examination of a thirty-six square inch cladding patch is an impractical requirement which, if imposed, would place a heavy burden on the licensee without a compensating increase in the safety of the facility. The visual examinations were performed on the two steam generators scheduled for tubing eddy current testing during the 1980 refueling outage will provide information relative to the condition of the other steam generators in the manway areas.

Volumetric examinations performed on-the pressurizer will cover sufficient cladding-base metal area to detect significant flaws if any exist.

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Conclusion:==

The staff finds that examinations being performed will provide adequate assurance that the cladding in the steam generators and pressurizer is not cracked and that since the cladding is not a structural part of the components, undetected clad cracks will not compromise the structural integrity of the components or allow significant degradation of.the base material during the period for which relief is requested. The staff therefore concludes that the relief requested may be granted.

3.1.1.9 The licensee requests relief from volumetric examination of steam generator safe end to piping welds (Item B4.5, Examination Category B-J).

' Code Recuirement: Volumetric examination of all of the area of 25% of the

-joints.each inspection interval.

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Licensee Basis for Requesting Relief: The geometric configuration of the safe end to' piping welds and their irregular surface condition prevents a complete volumetric examination from being performed.

Approximately 60% of this weld can be examined volumetrically.

Surface examinations will be used to supplement the limited volumetric examinations.

Evaluation: The licensee provided a drawing illustrating the geometric configuration of the safe end forging to pipe weld and the volume of coverage possible with ultrasonic examination.

A total of 60% of the examination volume can be insnected. All of the inside surface of the weld adjacent base metal will be examined.

This is the location at which most flaws would be expected

-to develop inservice. The 1977 Edition of Section XI requires only the inner one third of the wall be examined since flaws usually initiate from the I.D.

The licensee has committed to conducting a surface examination of the areas

-not scanned by UT.

Flaws originating on the 0.D. will be detected by this examination.

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Conclusion:==

The limited volumetric examination of the safe end to piping welds, supplemented by surface examinations where coverage by ultrasonic test-ing is not possible will provide adequate assurance of the-structural integrity of the piping pressure boundary. We conclude that relief may be granted.from 100% volumetric examination required by the Code.

3.1.1.10 The licensee requests relief from volumetric examination of-Regenerative Heat Exchanger nozzle to vessel welds (Item B3.2, Examination Category B-D).

_ Code Requirement: Volumetric examination of 100% of the volume shown in

. Figure IWB-25000 of Section XI of the Code.

Licensee Basis for Requesting Relief:

The geometric configuration of the weld surface prevents ultrasonic examinations'from-being performed as required by IWB-2600.

Surface examinations will be performed on this weld in lieu of volumetric examination.

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Evaluation:

The licensee has not provided sufficient information for relief to'be granted.

A drawing illustrating the geometric configuration of the nozzle to vessel weld is required.

In addition, the licensee should indicate the feasibility of conducting a partial volumetric examination from i.he vessel side of the weld and the percentage of the code required volume which can be inspected by this' technique.

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Conclusion:==

Based on'cne information supplied, we cannot grant relief at this time.

The licensee must furnish additional information, ac discussed

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~above, before our evaluation can be ccepleted.

3.1.2 Class 2 Comconents 3.1.2.1 The licensee requests relief from volumetric examination of the nozzle-to vessel weld on the Resicual Heat Exchanger.

(Item C1.2, Examination Category C-B).

Code Recuirement:

Volumetric examination shall cover 100% ci tne nozzle-to-vessel attachment weld. This involves weld cetal and base metal for one plate thickness beyond the edge of the weld joint.

Licens'ee Basis for Recuestine Relief: The reinforcing collar on the nozzle-

-to-channel weld precludes volumetric examination.

A surface exacination of these welds will be conducted.

Evaluation: The reinforcement-required by'ASME Section III in the design of shell. openings prevents the access necessary to perform ultrasonic examination of the nozzle-to-channel weld and' base metal.

Based on detailed drawings at

these. welds,cthe reinforcing collar-to-nozzle weld and nozzle-to-channel weld

. form a single full penetration weld which would most likely fail initially at the outer surface, at the weld-nozzie interface.

The licensee has proposed

'to examine the welds by surface (dye penetrant)-method, a method capable of detecting flaws which would most likely appear. initially.

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Conclusion:==

The staff finds the proposed surface examination to detect inservice. flaws a suitable alternative to the solumetric examination required by the Code. The surface examination will proside assurance of tne structural integrity of the nozzle-to-cnannel weid and base metal arid the staff therefore concludes that relief from the Code requirement may be granted.

3.1.2.2 The licensee requests relief from volumetric examination of certain class 2 welds (Items C2.1, C2.2, C2.3, Category C-F; Items C2.1, C2.2, C2.3, Category C-G) in the feedwater, main steam, CVCS, and residual neat removal

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piping.

Code Recuirement:

100% volumetric examination of welds defined in Table IWC-2520 of Section XI.

Licensee Basis for Relief:

The arrangement and details of the Class 2 piping system and components were designed and fabricated before the examination requirements of Section XI of tne Coce were formalized anc scme examinations as required by IWC-2600'are limited or not practical cue to geometric config-uration or accessibility.

Generally these limitations exist at all fitting-to-fitting welds such as elbow to tee, elbow to valve, reducer to valve, etc.,

where geometry and sometimes surface Conditions preclude ultrasonic ccupling or access for the required scan length. The limitations exist to a lesser degree at pipe to fitting welds, where examination can only be fully performed from the pipe side, the fitting geometry limiting or even precluding examina-tion from the opposite side. Welds having such restrictions will be examined to the extent practical.

5 In instances where the location of pipe supports or hangers restrict the access available for-the examination of pipe welds as required by IWC-2600, examina-tions will be' performed to.the extent practical unless removal or re'.ocation of

- the support is permissible without undu'ly stressing the system.

Surface 2.

examination will be utilized to sLpplement volumetric examinations to the extent permitted by access.

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t 4

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The first weld innediately inside the containment penetration or, the feedwater and main steam lines is inaccessible due to the containment liner plate blocking the' weld.

Evaluation: The licensee has not provided sufficient information for relief to be granted. For each weld for which the request applies, the weld identifi-cation number and estimated percentage of material which can be examined volumetrically should be furnished. Where practical, the licensee should conduct surface examinations on all areas not subjected to volumetric examination.

-Conclusion: We will evaluate this request for relief after the licensee supplies the information requested.

Relief is not to be granted at this time.

3.1. 3 System Pressure Tests 3.1.3.1 Request to test the Chemical and Volume Control Charging (Charging pump discharge), Seal Injection, and Letdown systems at lower pressures than Code required as shown below:

Class 2 System Test Pressure Code Test Pressurg Che-ical and Volume Control Charging (Charging Puep-Discharge) 2400 psig 3419 psig Seal Injection 2400 psig~

3419 psig Letdown" 100 psig-625 psig-Code-Recuirement:

(a) The system hydrostatic test pressure shall be at least 1.25 times the

~

system design pressure (P ) and conducted at test temperature not less D

1E

~

than 100 F except as may be required to meet the test temperature require-ments of IWA-5230.

(b) The test pressure may be reduced in accordance with the following table when system hydrostatic testing is required to be conducted at tempera-tures above 100 F in' order to meet the fracture toughness criteria appli-cable to ferritic materials of which the system components are constructed.

Test Temperature Test Pressure 100 F 1.25 PD 200 F 1.20 PD 300 F 1.15 P D 400 F 1.10 PD 500 F 1.05 PD Licensee Basis for Recuesting Relief:

The chemical and volume control, charging, seal injection and letdown systems are in operation during normal plant opera-tion and are continuously monitored to ensure integrity and performance.

In addition,' the potential for inadvertent overpressurization of the reactor coolant system causes. additional concerns on the advisability of pressurizing.

Class 2 system to considerably higher pressure than the adjacent Class 1 system.

In lieu of the Code requirements'and for the reasons stated above, a

' visual examination at operating pressure will be' conducted.

Evaluation:

Because of the design of the systems, the possibility of over-pressurizing the Class 1 portion of the systems makes the Code requirement impractical.

Portions of these systems will be subjected-to other nondestruc-tive examinations as well as visual examination at operating pressure. These examinations will provide assurance of the structural integrity of tia systems.

19 3

==

Conclusion:==

The staff finds that assurance of the structural integrity of the systems will be provided by the alternate test proposed by the licensee.

The staff therefore concludes that relief from the requirement may be granted as requested.

3.1.3.2 Request to test the Component Cooling System at normal operating pressure (85 psig) in lieu of the Code required test pressure (165 psig).

Code Requirement:

(a) The system hydrostatic test pressure shall be at least 1.25 times the system design pressure (P ) and conducted at a test temperature not less D

than 100 F except as may be required to meet the test temperature requirements of IWA-5230.

(b) The test pressure may be reduced in accordance with the following table when system hydrostatic testing is required to be conducted in order to meet the fracture toughness' criteria applicable to ferritic materials of which the system components are constructed.

Test Temperature-Test Pressure 100 F 1.25 PD 200 F 1.20 PD 300 F 1.15 PD 400 F 1.10 PD 500 F 1.05 PD Licensee Basis For Requesting Relief: The component cooling system is needed to cool the RHR heat excharigers..Therefore, since the RHR system must be 20 J'

sn, mwe

+'

+

mm

operable during c;eration as well as shutdown, the component cooling hydro-static test can only be performed when all fuel is removad from the core.

In lieu of a hydrostatic test a visual examination will lie conducted on this system at normal operating pressure.

Evaluation:

Because of the :coling requirements to prevant damage to plant components, the componer,c cooling system cannot be stoppe for the period of time required to test the system in accordance with the Code.

To remove all fuel from the reactor vessel in order to comply with the Code is impractical and places a burden on the licensee far in excess of the compensating increase in safety of the facility.

The visual examination which the licensee proposes will provide adequate assurance of the system's structural integrity.

==

Conclusion:==

The staff. finds that the proposed examinction of the Component Cooling System is an adequate examination to determine the structural accept-ability of the system in lieu of the Code.equirement and concludes that relief'from the requireLent as requested may be granted.

3.1.4 General 3.1.4.1 The licensee requests to use Appendix III of Section XI in the Winter, 1975 Addenda to 1974 Edition of the Code.

Code Reouirement:

The 1974-Edition of Section XI requires the use of Article 5.of Section V for ultrasonic examination of piping welds.

Licensee Basis for Reouesting Relief: Appendix III is the first guideline that has been published in the ASME Codes for the ultrasonic examination of pipe welds and, as such, its use fs essential.

It is recognized that Appan-dix III of Section XI, as issued in the Winter, 1975 Addenda, is not officially recognized by the NRC by reference in 10CFR50.

Evaluation: Appendix III in the 1977 Edition through Summer 1978 Addenda of Section XI was. incorporated into 10CFR50 after the-licensee's submittal.

The requirements in the later: edition' and.adcenda are essentially the same as those 21'

in the Winter, 1975 Addenda.

The licensee can update portions of their program to later NRC approved editions and addenda, as permitted by 50.55a(g)(4)(iv).

We require the following modifications to Appendix III if it is to be used in the licensee's ISI program:

(a) Indications of 50% of DAC or greater shall be recorded.

(b) An indication of 100% of DAC or greater shall be evaluated by a Level II or Level III examiner to the extent necessary to determine the size, shape, identity and location of the reflector.

(c) Any non geometric indication, 20% of DAC or greater, discovered during the ultrasonic examination shall be recorded and investigated by a Level II or Level III examiner to the extent necessary to determine the shape, identity, and location of the reflector.

(d) The cwner shall evaluate and take corrective action for the disposition of any indication investigated and found to be other than gecmetric in nature.

==

Conclusion:==

We conclude that the licensee may update ultrasonic examination procedures to Appendix III of the 1977 Edition through Summer 1978 Addenda of Section XI.

To maintain the sensitivity for detecting flows in Article 5 of Section V,~ additional recording requirements have.been specified.

3.1.4.2 >The licensee requests that calibration blocks be made to the requirements of Article T-434.1 in the Winter 1976 Addenda of Section V in lieu of I-3121 of Section XI.

Code Requirement: Material from which the block is fabricated shall be from one of the following:

(1) the component nozzle dropout; (2) the component prolongation;'or (3) when it is not possible to fabricate the block from material taken from the component,-_it may be fabricated.from a material of a

' specification included in the acplicable examination volumes of the comconent.

22

The acoustic velocity of such a block shall be demonstrated to fall within the range of straight beam longitudinal wave velocity and attenuation found in an unciad component.

Licensee Basis for Reauesting Relief:

The reason this alternative is requested is that the Code requires that calibration blocks for the examination of welds in Ferritic vessels 2-1/2 inches thick and greater be fabricated from material taken from the component nozzle drop out or material from the component pro-lonoation. 'As a third alternative, when it is not possible to fabricate the block from material taken from the component, the block may be fabricated from a material of a specification included in the applicable examination voluces of the component.

It is required that the acoustic velocity and attenuation of such a block be demonstrated to fal within the range of straight beam longituainal wave velocity and attenuation found in the unclad components.

For the c'mponents in Haddam Neck, particularly the pressurizer and steam generators, it will be impossible to meet the requirements of alternatives 1 or 2.

Materials ~of the specification are readily available, but because all the components involved are clad on the inner surface, it would be impossible to obtain a comparison of sound beam velocities and attenuations in the unclad component.

Evaluation:

The licensee has requested relief from specific provisions of Appendix I '! Ultrasonic Examination" concerning the material selection of the calibration blocks. Appendix I was first published in the Summer 1973 Addenda and is limited in scope to Class 1 and 2 ferritic vessels 2-1/2 inches and over in well thickness.

In the 1977 Edition of Section XI, Appendix I was superseded by Article 4 of Section V.

Article T-434.1.1 of Section V, Winter 1976 Addenda, is the current-requirement for calibration block material for these components in accordance with 10 CFR 50.55a(b). Therefore, we conclude that the substitution of Article T-434.1.1 of Section V is an acceptable alternative provision that may be substituted in lieu of Article I-3121 of Section XI.

23

I

==

Conclusion:==

The staff finds that the alternative code requirements for calibration proposed by the licensee are currently approved by 10 CFR 50.55a(b).

Updates to these later requirements is permitted by 50.55a(g)(4)(iv).

3.2 Valve Testing Procram 3.2.1 Testing of Valves Which Perform a Pressure Isolation Function There are several safety systems connected to the reactor coolant pressure boundary that have design pressures that are below the reactor coolant system operating pressure. There are redundant isolation valves forming the inter-face between these high and low pressure systems to prevent the low pressure systems from being subjected to pressures which exceed their design limits.

In this role, the valves are performing a pressure isolation function.

It is our view that the isolation redundancy provided by these valves regarding their pressure isolation function is important.

We consider it necessary to provide assurance that the condition of each of these valves is adequate to maintain this redundant isolation ~and system integrity.

For this reason we believe that some method, such as pressure monitoring, leak testing, radio-graphy and ultrasonic testing, should be used to assure their condition is sufficient to maintain this pressure-isolation function.

In the event that leak testing is selected as the appropriate method for achieving ~this objective, the' staff believes that the following valves should be categorized as A or AC and leak tested in accordance with IWV-3420 of Section XI of the applicable edition of the ASME Code.

These valves are:

+

-FH-MOV-578, 535, 522, 508 (Loop Fill)

- FH-MOV-562, 507, 521, 534,1544, 310 (Loop Drain)

FH-V-502,-516,'521, 534, 525-(toop Drain) 24

We have discussed this matter and identified the valves listed above to the licensee.

The licensee has agreed to consider testing each of these valves and to categorize these valves with the appropriate designation depending on the testing method selected. Whatever the licensee selects as the testing method to be used to determine each valve's condition, the licensee will pro-vide to the NRC for evaluation on a valve-by-valve basis the details of the method used that clearly demonstrates the condition of each valve.

3.2.2 Strokina Reauirements for Section XI Subsection IW-3410(a) of the Section XI Code (which discusses full stroke and partial stroke requirements) requires that Code Category A and B valves be exercised once every three montas, with exceptions as defined in IW-3410(b)(1), (e) and (f).

IW-3520(a) (which discusses full stroke and partial stroke requirements) requires that Code Category C valves be exercised once every three months, with exceotions as defined in IW-3520(b).

In the above cases of exceptions, the Code permits the valves to be tested at cold shutdown where:

-(a) It is not practical to exercise the valves to the position required to fulfill their function or to the partial position during power coeration.

(b) It is not practical to observe the operation of the valves (with fail-safe actuator) upon loss of actuator power.

The staff stated its position to the licensee that check valves, whose safety functio'n is to open, are expected to be full-stroked.

If only limited operation is possible (and it has been demonstrated by the licensee and agreed to by the staff), the check valve shall be partial stroked.

Because disk position is not always observable, the NRC staff stated that verification of the plant's safety analysis for the design flow rate through the check valve would be an adequate demonstration of the full-stroke requirement.

Any flow rate less than design will be considered part-stroke exercising, unless it can be shown 25

i that the check valve's disk position at the lower-flow rate would be equiva-lent to or greater than that required for the design-flow rate through the valve.

The licensee agreed to conduct his flow tests to satisfy the above position.

The licensee has requested relief from the part-stroke requirement of Section XI for all power operated valves.

The' licensee has stated that none of the Category A or B power operated valves identified in this Safety Evaluation Report can be part-stroked because of the design logic of the operating circuits.

These circuits are such that when an open or close signal is received the valve must complete a full stroke before the relay is released to allow the valve to stroke in the other direction.

We find that the licensee relief request from part-stroking is warranted and we therefore grant relief from this requirement because ite required function 4

of the valves involves only full open or full closed positions.

3.2.3 Cold Shutdown Testing Definition Inservice valve testing at cold shutdown is acceptable when the following conditions are met.

It is understood that the licensee is to commence testing as soon as the cold shutdown condition is achieved, but not later than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shutdown, and to continue until all tests are complete or.the plant is ready to return to power.

Completion of all valve testing is not a prerequisite to return to' power.

Any testing not completed at one cold shutdown should be performed during any subsequent cold shutdowns that may occur before refueling to meet the Code specified testing frequency.

For planned cold shutdowns, where the. licensee will ccaplete.all the valves identified in his Inservice Testing (IST) Program for testing in the cold shut-down mode, exception-to the above 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />'. start time may be taken.

It is noted that the staff ' differentiates for valve testing purposes between the cold shutdown mode and the refueling mode.

That is, for. testing purposes, the refueling mode is not. considered as a planned cold shutdown.

26

3. 2.

• Check Valve Testing Frequency The Code states that, in the case of cold shutdowns, valve testing need nctie performed more of ten than once every three months for Category A and B vales and once every nine months for Category C valves.

It is our position that Category C valves should be tested on the same schedule as Category A and E valves.

This is consistent with the testing requirements of the later 1977 edition of ASME Section XI which requires testing every three months for Category A, B and C valves.

The licensee shall modify any procedures as necessary.on cold shutdown to read, "In the case of frequent cold shutdowns, valve testing shall not be performed more often than once every three (3) months for Category A, B and C valves."

3.2.5 Chances to the Technical Soecification In a November 1976 letter to the CYAPCO, we provided an attachment entitled "NRC Guidelines for Excluded Exercising (Cycling) Tests of Certain Valves During Plant Operation." The attachment stated that when one train of a redundant system such as in the Emergency Core Cooling System (ECCS) is ing-erable, nonredundant valves in the remaining train should not be cycled sirre their failure would cause a loss of total system function.

For example, de ing power operation in some plants, there are stated minimum requirements far

.' systems which make up the ECCS which allow certain limiting conditions for operation to exist at any one. time and if the system is not restored to meet the requirements within the time period specified in a plant's Technical Specifications, the reactor is required to be put in some other mode.

Fur-thermore, prior to initiating repairs all-valves and interlocks in the systum that provide a duplicate function are required to be tested to demonstrate operability immediately and periodically thereaf ter during power operation.

For such plants, this situation could be contrary to the NRC guideline as

~

stated in the document mentioned above.

The licensee has agreed to review the Haddam Neck Plant Technical Specifi'catiums and to consider the need to prcpose Technical Specification changes whic' would have the effect of precluding, such testing.

27

t After making this review, if the licensee determines that the Technical Specifications should be changed because the guidelines are applicable, the licensee will submit, in conjunction with the proposed change, the inoperable condition for each system that is affected which demonstrates that the valves' 4

failure would cause a loss-of-system function or if the licensee determines t

t that the TS should not be changed because the guidelines are not applicable or

- cannot be followed, the licensee will submit to the NRC the reasons that led to their determination for each potentially affected section of the Technical Specifications.

1 3.2.6 Safety Related Valves l

4 This review was limited to safety-related valves.

Safety-related valves are defined as those valves that are reeded to mitigate the consequences of an i

accident and/or to shutdown the reactor and to maintain the reactor in a shutdown condition.

Valves in this category would typically include certain

- Code Class 1, 2 and 3 valves and could include some non-Code class valves.

I It should be noted that the-licensee may have included non-safety related valves in his-IST Program &s a decision on the licensee's part to expand the

- scope of their IST Program.

3.2.7' Application.of Appendix J Testing to the IST Program The Appendix-J review for this plantLis a completely separate review from the IST' program review..However,'the determinations made'by that review are

^

directly applicable to the.IST program.

Our review has determined that the current IST program as submitted by the licensee correctly reflects our inter-

[

- pretation of Section XI vis-a-vis Appendix-J.

The licensee has agreed that, should the Appendix J program be amended, they will amend their IST program accordingly.

4.

i

-28

3.2.8 Licensee Recuest for Relief to Test Valves at Cold Shutdown The Code permits valves to be tested at cold shutdown, and the Code conditions under which this is permitted is noted in Section 5.3.

These valves are specifically identified by the licensee and are full-stroked exercised during cold shutdowns.

Therefore, the licensee is meeting the requirements of the ASME Code.

Since the licensee is meeting the requirements of the ASME Code, it will not be necessary to grant relief.

However, during our review of the licensee's IST' program, we have verified that it was not practical to exercise these valves during power operation and that we agree with the licensee's basis.

3.2.9 Specific Relief Reauests 3.2.9.1 Relief Recuest:

The following Category A, containment isolation valves (CIV's) will meet Appendix J leak testing requirements in lieu of Section XI requirements:

Valve Valve Valve Valve DH-TV-1847 CC-CV-885 SS-V-999 CH-CV-305B BD-V-506 CC-TV-1831 SS-V-999A CH-CV-305C

~BD-V-515 CC-V-884 PW-CV-139 CH-CV-3050 BD-V-522 VS-TV-1848 PW-CV-140 DH-TV-1841 BD-V-529 SOV-12-1 HC-V-212 DH-TV-1844

.BD-TV-1312-1~

VS-CV-1104 PU-V-242 FM-MOV-31 BD-TV-1312-2 CC-CV-731 PU-V-242A FW-CV-192 BD-TV-1312-3 FCV-611-HS-CV-295 FW-CV-194 BD-TV-1312-4 VH-V-507 HS-CV-295A FW-CV-196 VH-V-5078 CC-GCV-608 BV-1-1B FW-CV-198 DH-TV-554 CC-CV-721 HCV-1101 FH-CV-296 WD-HICV-1840

-WG-A0V-558 BV-1-1A WDoTV-1846 WG-TV-1845 P50 DH-TV-1843 SS-V-984A SA-V-411A DH-TV-1844 SS-TV-950 SA-V-413

'LM-TV-1811A SS-TV-955 LD-A0V-202 LM-TV-1811B SS-TV-960

'LD-A0V-203 LM-TV-1812 SS-TV-965 LD-A0V-204 CC-CV-853.

DH-TV-1842A CH-TV-334 CC-TV-1411 DH-TV-1842B CH-CV-305A 29

Code Reouirement:

IWV-3420 Valve Leak Rate Test.

Category A valves shall be leak-tested.

Tests shall be conducted at the same (or greater) frequency aa scheduled refueling outages, but not less than once every two years.

Valve seat leakage ~ tests shall be made with the pressure differential in the same direction as will be applied when the valve is performing its function with the following ex.eptions:

(a) Any' globe type valve may be tested with pressure under seat.

(b) Butterfly valves may be tested in either direction, provided their seat construction is designed for sealing against pressure on either side.

(c) Gate valves with two piece disks may be tested by pressurizing them between the seats.

-(d) All valves (except check valves) may be tested in either direction if the function differential pressure is 15 psi or less.

(e).The use of leakage tests involving pressure differentials lower than function pressure differentials are permitted in those types of valves in which service pressure will tend to diminish the overall leakage-channel opening, as by pressing.tne disk into or onto the seat with greater force.

Gate valves, check valves and globe type valves having function pressure differential applied over the seat, ar2 examples of valve applications satisfying this requirement. When leakage tests are made in such cases using pressure lower than function maximum ~ pressure differential, the observed' leakage shall be adjusted to function maximum pressure dif-ferential value by calculation appropriate to the test media and the ratio between. test and function pressure differential assuming leakage to be directly. proportional to the pressure' differential to'the one-half p1wer.

~

C (f) Any. valves not qualifying for reduced pressure. testing as defined in 3420(c)(5) shall-be leak-tested at full maximum function pressure dif-ferential, with adjustment by calculation if needed to compensate for a difference between service and test media.

30

Valve seat leakage may be determined by:

(a) Draining the line, closing the valve, bringing one side to test pressure and measuring leakage through a down-stream telltale connection, or, (b) By measuring feed rate required to maintain pressure between two valves, or between two seats of a gate valve, provided the total apparent leak rate is charged to the valve or gate valve seat being tested, and that the conditions required by IWV-3420(c) are satisfied.

The test medium shall be specified by the Owner.

Basis for Relief Request:

Appendix J leak testing meets the intent of Section XI Requirements.

-0perability. testing of these valves during normal plant operation could cause a loss of containment integrity and/or system function _if a valve failed in a non-conservative position.

Because the safety function of these valves is to provide containment integrity.

.by their leak tightness, these valves are and have been leak tested under

. Technical Specification requirements based on " Appendix J, Type C" tests.

See. Technical Specification 4.4.

In that leakage tests are conducted to satisfy' containment integrity requirements, waivers from IWV-3420(f) " Analysis of Leakage Rates" and (g) " Corrective Action" are' requested.

CYAPC0 proposed to demonstrate leak tightness by_ following containment integrity rules, which observe an integrated acceptance criterion for these test parameters, in lieu of individual valve performance measurement.

~

CYAPC0 proposed to use existing procedures for exercising and leak testing of valves used as containment isolation. Modifications to procedures as agreed

'31 w

upon by CYAPCO and NRC Staff concerning " Appendix J" tests and exemptions will be reflected in this program.

All " active" containment isolation valves are exercised at cold shutdowns by a procedure to demonstrate the Reactor Containment Atmospheric Control System Functions properly should an over pressure situation exist in the reactor containment, by verifying that all automatic actions required of the system under this condition do occur.

Evaluation:

The Category A valve leak rate test requirements of IW-3420(a-e) have been superceded by Appendix J requirements for CIV's.

The staff has concluded that the applicable leak test procedures and requirements for CIVs are determined by 10 CFR 50 Appendix J.

Relief frcm paragraph IW-3420(a-e) for CIVs presents no safety problem since the intent of IW-3420(a-e) is met by Appendix J requirements.

There is insufficient information to evaluate the relief for paragraphs f and g of IW-3420.

It should be noted that this relief request from IW-3420(f and g), applies only where a Type C Appendix J leak test is' performed.

Conclusion:

Based on the considerations discussed above the staff concludes that the alternate testing proposed above, except for relief from paragraphs IW-3420(f) and (g), will give reasonable assurance of valve operability intended by the Code and that the relief thus granted from paragraphs IW-3410(a-e) will not endanger life or property or the common defense and security of the public.

The request for relief from paragraphs IW-;420(f) and.(g) is denied due to insufficient information.

3.2.9.2 Relief Recuest:

The licensee requests a waiver from the requirements of IW-3410(c)(3).

Code Requirement:

IW-3410(c)(3).

If an increase in stroke time of 25% or more from the previous test for valvee,iith stroke times greater than ten

. seconds or 50% or more for valves with stroke times less than or equal to ten seconds is' observed, test frequency shall be increased to once each month unti?

32

corrective action is taken, at which time the original test frequency shall be resuud.

In any case, any abnormality or erratic action shall be reported.

Basis for Relief Request:

The licensee requests a waiver from the requirements of IW-3410(c)(3) to enable engineering dispositions to be made on stroke time variations without necessarily committing to monthly tests.

The stroke times have been derived basically from plant design information and reviewed by plant operations personnel to verify consistency with time restraints dictated by normal and emergency conditions.

Evaluation:

The licensee has not submitted sufficient justification for supporting his request for relief from the requirements of IW-3410(c)(3).

==

Conclusion:==

The staff concludes that the alternate criteria proposed are inadequate, and the relief request is, therefore, denied.

3.2.9 3 Relief Request:

The licensee requests a waiver from the requirements of IW-3410(g).

Code Requirement:

IW-3410(g) Corrective Action.

If a valve fails to exhibit the required change of valve stem or disk position by this testing, correcti<e action shall be initiated immediately.

If the condition is not, or cannot, be corrected within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the valve shall be declared inoperative.

Whsn

. corrective action'is required as a result of tests made during cold shutdown, the condition shall be corrected before startup.

A retest showing acceptable

. operation shall be run following any required corrective action before the valve is returned to service.

. Basis for Relief Request:

The licensee requests a waiver from the requirements of IW-3410(g) in that the operating status of a valve would be best determined by the plant management after an engineering and operations review of the test data.

Evaluation:

The licensee has not submitted sufficient justification for supporting his request for relief from the. requirements of IW-3410(g).

33

==

Conclusion:==

The staff concludes that the alternate criteria proposed are inadequate, and the relief request is denied.

3.2.9.4 Relief Request:

Valves SI-CV-862A, SI-CV-8628, SI-CV-862C and SI-CV-862D will not be exercised in accordance with Code requirements.

Code Reouirement:

Refer to Section 3.2.2.

Basis for Relief Request:

A full-flow test of these valves would require that a HPSI flow be established into the RCS. A test of this nature would require that the reactor pressure vessel (RPV) head be removed, and would result in water chemistry problems and high radiation exposure to plant personnel caused by dislodging crud in the safety injection system piping.

The licensee proposes to part-stroke test these valves at reactor refueling i

outages and full-stroke test them every interval in conjunction with full-flow tests of the safety injection system.

Evaluation:

The safety injection valves SI-CV-862A, 8628, 862C and 8620 are in the HPSI system.

The valves are closed during nnrmal plant operation and are part of redundancy that isolates the operating RCS pressure from the lower design pressure HPSI. system. The check valves open when the HPSI pumps are activated and RWST water is supplied to the RCS during the emergency condition.

These check valves are located in the HPSI pump discharge connections to the reactor coolant system and part of their function is to prevent backflow from the reactor coolant system to the HPSI system. The licensee intends to use fluid flow to demonstrate movement to fulfill their function.

During normal plant operation flow'cannot be initiated because the combined LPSI and HPSI pumps can only develop a pressure of about 1700 psi and this pressure is not sufficient-to overcome the reactor coolant system pressure of about 2250 psi.

During cold shutdown these pumps are disabled for low temperature overpres-surization protection.

Therefore, the only practical interval that these valves can be exercised is at_ refueling outages when a full-flow testing of the HPSI system may be conducted with the reactor vessel head removed.

How-ever, the consequences of such a test are unknown.

The licensee nas stated 34

- that a full-flow test could result in water chemistry problems and high radiation exposure to plant personnel.

==

Conclusion:==

Based on the considerations discussed above the licensee has been granted relief for the current testing period to part-stroke thes-valves at reactor refueling outages and full stroke at the end of the interval.

The licensee informed the staff that the full-stroke test was scheduled for 1980.

The licensee will reevaluate this relief for these valves for the next interval based on the results of the 1980 full flow test.

3.2.9.5 Relief Reouest:

Valves SI-CV-103, SI-CV-107A, SI-CV-1078, CD-SV-872A, and C0-CV-872B will be part-stroked at refueling outages only in lieu of Code requirements.

Code Reouirement:

Refer to Section 3.2.2.

Basis for Relief Reouest:

For valves SI-CV-103, SI-CV-107A and SI-CV-107B full or partial tealve _exerc ing requires that flow be established into the reactor vessel.

System pres

-e downstream of these valvas does not allow flow during normal. operations iuring cold shutdowns, available volume in the reactor vessel is insuffit.

to accommodate flow required for exercising these valves. -In addition, the-b flow generated by the LPSI pumps would stit up crtd in the Safety injectio-oiping and deteriorate the chemistry of the RCS water.

A gravity 4ficu test will be. performed fi.

the RWST through these' lines to

. partially stroke the valves at reactor re.

ling when the RHR system can be taken out of service. This was done at the 19,2 outage. A full flow test is proposed during the 1980 outage when the reactor core is removed.

If this test is satisfactory, it will be done o.pe per interval.

For valves CD-CV-872A and'CD-CV-872B exercising requires that the RCS be

- depressurized and vented, to allow flow.from the lower pressure RHR/LPSI

systems to be estaolished.

The tests cannot be conducted at cold shutdcwns

' because _of. the water chemistry and venting requirements discussed above.

35-e u

These valves will be part-stroke tested at reactor refuelings and full-stroke tested during the safety injection system tests scheduled once each interval.

Evaluation:

The LPSI pump discharge check valves SI-SV-107A and 1078, the check valve SI-SV-103 located in the LPSI pump discharge to the RHR system and core deluge check valves CD-CV-872A and B, are closed during power operation, and are required to open when the LPSI flow is initiated to the reactor during an emergency.

The function of these valves is to prevent a backflow from ;he RCS and RHR systems to the LPSI pumps.

Quarterly Testing: The subject check-valve designs are such that they can only be exercised by disassembling or by flow.

The LPSI system configuration is such that flow can only be initiated through the check valves by activating the LPSI pumps.

During power operation, the RCS pressure is approximately 2000'psig, and acts to close CD-CV-872A and 8728 (assuming CD-MOV-871A and 871B are open).

The discharge pressure of the LPSI pumps is far lower than the RCS operating pressure, and cannot open check valves 872A and 9728 against the RCS head.

SI-CV-103, 107A, and 107B are in series with 872A and 8728 and are also prevented from exercising.

It should be noted that the mini-flow recirculation lines off the LPSI pumps are upstream of the 107A and 1078 which eliminates any part stroke exercise of these check valves when the LPSi pumps are run monthly.

Based on the above, it is agreed that full or part-stroke exercising the subject valves by disassembly (where possible) or by flow is impractical on a 3-month basis.

Cold Shutdown Tests:

Performing a design-flow test using the LPSI pumps is the only method available (without disassembly) to full-stroke exercise the subject check valves. This requires venting the RCS, and removal of the pressurirer manway cover or several safety valves before a LPSI pump can be used for initiating design flow to the reactor.

To partially stroke the check valves would require that the RCS be vented, and that the pressurizer level be established at a low level and visually monitored as flow is established from the LPSI pumps anc introduced into the RCS (pres-36

i l

i surizer).

In either case the initiation of flow to the reactor coolant system during cold shutdown is impractical due to the lack of volume for flow and, therefore, the high probability of an over pressurization event.

In addition, for either case cited, water pumped by the LPSI pumps would originate from the RWST (borated water).

Introduction of this water to the

?CS would require additional waste liquid processing which is time consuming and delays startup.

Disassembly (where possible) of the check valves is imoractical from a time standpoint, it would increase the length of outages.

F1rthermore, because these valves could not be tested after installation, it educes the assurance that they were installed properly.

The licensee's basis for the requested relief has adequately established t"at full 'or part-stroke exercising the subject check valves by flow or disassembly at cold shutdown is highly impractical from a safety, time and operations standpoint.

Testing at Refuelino:

The licensee does not want to full-flow test with LPSI when.the core internals are in place. They feel the high flow couid damage equipment and stir up crud.

The' licensee proposes to part stroke the valves at refueling outages when the RHR system can be taken out of service.

Part stroking of the valves will be accomplished by using a gravity flow from the refueling water storage tank (RWST) through the LPSI system.

This was done at the 1979 outage.

==

Conclusion:==

Based on the considerations discussed above, the licensee has been granted relief for the current testing period to part-stroke these valves at reactor refueling outages and full stroke at the end of the interval.

The licensee will reevaluate this relief for the next interval based on the result of tests during the 1979 and 1980 outages.

3.2.9.6 Relief Recuest: Valves SI-CV-856A and 856B will not be full stroke exercised in accordance with Section XI requirements.

37

Code Recuirements:

Refer to Section 3.2.2.

Basis for Relief Request:

In order to full stroke these valves, flow must be established into the RCS, because the pump recirculation and test line is 3/4", insufficient to simulate design flow conditions.

Full flow testing of the HPSI system may only be conducted with the reactor vessel head removed.

Because of. the water chemistry and radiation problems associated with pumping large amounts of water in through this system, the licensee proposed to test these valves as follows:

Alternate Testing:.These valves will be part-stroke tested quarterly, and full-stroke tested once each interval in conjunction with full-flow tests of the HPSI system.

Evaluation:

The HPSI Pump Discharge check valves SI-CV-856A and 856B are closed during power. operation nd their function is to open when the HPSI pumps are activated and supply PWST water to the RCS during the emergency condition.

The HPSI pump system's configuration is such that full stroke exercising quarterly is considered to be impractical. The 862A, B, C, O check valves that are downstream and in series with the subject check valves are prevented from opening by the RCS pressure (approximately 2250 psig) during power oper-ations. The pressure developed by the HPSI pumps (approximately 1500 psig) is insufficient to overcome this RCS pressure, thereby preventing flow through the~HPSI system and the subject check-valves.

Part-stroke exercising is accomplished monthly during HPSI pump miniflow t4..irculation tests.

==

Conclusion:==

Based on the considerations-discussed above, we conclude that

' granting of relief for the current testing period to part stroke these valves quarterly and full-stroke test once each interval is. acceptable. The licensee has informed the staff that the full-stroke test was scheduled for -1980.

The

-licensee will reevaluate this relief request for these valves for tha next interval based on the results of the 1980 full-flow test.

38

1 3.2.9.7 Relief Reouest:

Check Valves RH-CV-783 an! RH-CV-808A will not be exercised in accordance with Code requirements.

Code Requirement:

Refer to Section 3.2.2.

Basis for Relief Request:

These valves are located in piping which is normal ~g drained and do not function unless water is present in the containment sump, as would be the case during a LOCA.

It is not practical to flood the contain-ment floor and pump to exercise these valves hydraulically.

Alternate Testing:

Each of these valves will be disassembled and inspected to determine interior condition and operability once each interval.

RH-CV-783 was disassembled and inspected in February 1979 and found to be in satisfactow condition.

Evaluation:

The Containment Sump Suction check valves RH-CV-783 and RH-CV-8G5 are in the recirculation lines to the RHR pumps.

The valves are required to open when the sump recirculation mode is required following the LOCA.

The valves are not designed to be exercised by external actuators, and can only be exercised by flow or by disassembling.

Using flow through the valves is impractical at any tir.e as the system's configuration is such that the containment sump would have to be filled with water and the contaminated water directed back through the RHR pumps to the reactor..Therefore, disassembly appears to be the practical alternative.

==

Conclusion:==

The staff agrees that it is impractical to meet the code required testing.

However, we cannot grant this relief because due to insufficient information provided by the licensee, we are unable to determine that conduct-ing this test at less than the Code specified frequency will not endanger' pubic health and safety.

3.2.9.8 Relief Request: Valves CH-FCV-110 and CH-FCV-110A will be full stroh exercised at reactor _refuelings during scheduled safety inspection system-testing.

3

Code Requirement:

Refer to Section 3.2.2.

Basis for Relief Request:

These valves are arranged in parallel flow paths such that either valve can be utilized to control charging flow to the RCS.

The valves are normally open and modulate to control flow.

They remain open in the accident mode but assume a pre-set SIS position.

Exercising these valves full stroke during nortpal operations would cause flow and pressure transients in the charging system.

Cold shutdown testing is not possible as stroking these valves to the accident position requires an integrated SIS test, performed only at refueling outages when the plant is lined up to conduct such testing.

Evaluation:

These control valves are located in the charging pump discharge connection te the reactor coolant system and their function is to control the charging flow to the RCS.

The valves are normally open and remain open in the accident mode but assume a pre-set SIS position.

The valves cannot be part or full-stroke exercised _during operation as this would cause flow and pressure transients in the charging system and a possible reactor shutdown.

The licensee states that testing the valves during cold shutdown would require an integrated SIS test.that is only performed at refueling outages.

This valve is either passive and/or redundant.

The optimum test interval for operability testing passive and/or redundant valves was determined by the staff, using actual valve failure rate data and standard probabilistic techniques, to be in the range of 3 months to 27 months.

Refueling intervals, which have been proposed as the exercise interval for the valve, occur every 12 to 18 months which is_within the optimum range for operability testing of this valve.

Passive, as used in the above paragraph, means any component whose unavailability upon demand is less than or equal to 10.4/ demand.

Check valves are considered

~

. passive since their unavailability has been found to be 10 / demand.

Redundant, as_used above, means the existence of more th'an one valve for performing a given function.

40'

Furthermore, the ASME Code, which requires testing be done quarterly and which has been adopted by the CFR (10 CFR 50.55a), also allows testing at cold shut-downs if quarterly testing is impractical.

Cold shutdowns can occur at intervals up to refueling outages.

Therefore, changing the test interval from quarterly to refueling will not differ significantly from the Code permitted change from quarterly to cold shutdown testing.

==

Conclusion:==

Based on the considerations discussed above, the staff concludes that the alternate testing frequencies propn ed above will give the reasonable assurance of valve operability intended by the Code and that the relief thus granted will not endanger public health and safety.

3.2.9.9 Relief Recuest: Valves SW-MOV-5 and SW-MOV-6 will not be exercised in accordance with Section XI requirements.

Code Requirements:

Refer to Section 3.2.2.

Basis for Relief Reouest:

These valves open to supply service water to the residual heat exchangers in the event of a failure in the componet cooling system. There is no automatic operation.

Stroking these valves requires that the service water system be cross-connected with the component cooling water system to provide emergency flow to the Resid-ual Heat Removal Heat Exchangers (RHX).

The service water system utilizes water from the Connecticut River, and the component cooling water system is a closed system cw:taining potentially conta minated water.

Cross-connecting could result in.21 easing non-conforming 3ater to the river and require a major cis.anup of system components, which would require that they be taken out of service during the restoration process. These valves were satisfactorily cycled during the February 1979 refuelir.g outage when special arrangements could be made to minimize the cleanup. ~They had not been stroked in the past

. ten years, so the tests confirmed that the valves were still above to fulf.ill their function.

The licensee proposes to exercise these valves each five years.

41

m, Evaluation:

The RHR inlet valves SW-MOV-5 and -6 are closed during normal plant operation when the component cooling system provides cooling water to the RHX's.

In the emergency mode, the component cooling system drops out, SW-MOV-5 and -6 open, and RHX cooling is provided by the service water system.

The staff agrees that full or part stroking these valves could result in excessive cleanup operations.

==

Conclusion:==

The '.taff finas the testing frequency required by the Code to be impractical. We cannot grant this relief, however, because due to insufficient information provided by the licensee we are unable te determine that conducting this test at less than the Code specified frequency will not endanger public health and safety.

3.2.9.10 Relief Recuest:

The licensee has reqJested relief from exercising the following valves per Coce requirements for Category A valves.

The valves are CIV's.

Valve Catecory Valve Service PW-CV-139 AC Primary Water to PRT PW-CV-140 AC Primary Water to PRT HC-V-212 AE Space Heater Containment Return PU-V-242 AE Cavity Purification Line PU-V-242A AE Cavity Purification Line HS-CV-295 AC Containment Space Heating Daply HS-CV-295A

.AC Containment Space Heating Supply BV-1-1B AE Containment Purge Air. Exhaust HCV-1101 AE Containment Purge Air Exhaust Bypass BW-1-1A AE Containment Purge Air Supply P50 AE Fuel Transfer Tube SA-V-411A AE Air Monitor Purge SA-V-413 AE Service Air to Containment BD-V-506 AE Steam Generator No. 1 Blowdown BD-V-515 AE Steam Generator No. 2 Blowdown BD-V-5?2 AE Steam Generator No. 3 Blowdown e

BD-V-529 AE Steam Generator No. 4 Blowdown CC-CV-885 AC-

'CC Water to Neutron Shield Tank Cooler CC-V-884 AE Neutron Shield Tank Fill Line VS-CV-1104-AC Air Monitor Sample to Containment.

CC-CV-731 AC CC Water Supply.to Drain Cooler VH-V-507 AE Primary Vent Header VH-V-507B AE Primary Vent Header CC-CV-853 AC CC Water to RCP Oil Coolers CC-CV-721 AC -

CC Water to RCP Thermal Barrier SS-V-999 AE Neutron Shield Tank Sample 42' i

SS-V-999A AE Neutron Shield Tank Sample FW-CV-192 AC Aux. Feed Water Supply to S.G. No. 4 FW-CV-194 AC Aux. Feed Water Supply to S.G. No. 3 FW-CV-196 AC Aux. Feed Water Supply to S.G. No. 2 FW-CV-198 AC Aux. Feed Water Supply to S.G. No. 1 FH-CV-296 AC Loop Fill Header Check SS-V-984A A

PRT Sample Code Requirement:

Refer to Section 3.2.2.

Basis for Relief Request:

The valves are closed during normal plant operation, and their safety related positions is to remain closed.

Evaluation:

The staff considers these valves listed above as passive, i.e.,

a closed valve whose function are to remain closed during the emergency condi-tion.

The staff has determined that the exercising requirement of Code Section XI provides no meaningful information for these passive valves, anc relieves the licensee from the 3 month stroke and stroke timing requirements.

==

Conclusion:==

-We conclude that the quarterly stroke and stroke time measu.ement are meaningless for passive valves and the relief should be granted.

3.2.9.11 Relief Request: The licensee requests relief from quarterly stroking valves SI-MOV-861 A, SI-MOV-861B,- SI-MOV-861C, SI-MOV-8610,

~ 'CD-MOV-871 A, CD-MOV-8718.

Code Requirement: Refer to Section 3.2.2.

Basis for Relief Request: Each time these valves are cycled the potential

'existsi or an interruption loss-of-coolant accident because the check f

valves which are in series with these valves then become the sole pressure retaining boundary between the reactor coolant system and the 'relatively low pressure ECCS systems. The question then becomes one of whether it is more conservative.to (a) cycle these valves.to verify their operability, ano, thus, the ECCS operability, while at the same time,' temporarily increasing the possibility of an later-system 43

loss-of-coolant accident (LOCA), or (b) to avoid the inter-system LOCA possibility, but lose the verification of value operability.

Operational component failures in the ECCS have been analyzed and provided for with redundancy of components. The loss of one component or even one system will, tnus, not result in complete loss of function.

The potential consequences of a failure of a check valve while cycling an MOV dictate that the cycling of the MOV's should not be the over-riding concern. The licensee intends to stroke test these valves each refueling.

Evaluation and

Conclusion:

By a separate etaluation, attached to this safety evaluation, the staff has determined that the potential conse-quences of a failure of a check valve while cycling an MOV override the potential loss of function which could occur if relief would be granted. Therefore, relief is granted from quarterly stroke testing of these valves.

14.0 Sumary - Inservice Insoection and Testing The licensee has submitted information to support his determinations that certain ASME Section XI Code (1974 Edition through Summar 1975 Addenda) requirements' are impractical to implement at the Haddam Neck Plant. We have evaluated the licensee's bases for his determinations and find that relief may be granted as detailed elsewhere in this Safety Evaluation and as sumarized below.

1.

The licensee requests relief from volumetric examination of the Closure Head Peel Segment-to-Disc Circumferential Weld.

(Item Bl.2, Examination Category B-B).

This request is granted.

2..The licensee requests relief from 100% volumetric examination of the Steam Generators -(Primary Side) Nozzles-to-Safe End Welds (8).

(Item B3.3, Examination Category B-F).

This request is granted.

3.

The licensee requests-relief from 100% volumetric examination of

(. Item the Regenerative Heat Exchanger integrally-welded supports.

~ B3.7, Examination Category B-H).

This request is granted.

44

4.

The licensee requests relief from 100% volumetric examination of the-circumferential butt weld Number 1 in Loop 1 of the Safety Injection System. (Item 84.5, Examination Category B-J)

This request is granted.

5.

The licensee requests relief from 100% volumetric examination of integrally-welded supports attached by fillet welds to piping.

(Item B4.9, Examination Category B-K-1)

This request is granted.

6.

The licensee requests to defer performing volumetric examination of a. reactor coolant pump casing welds and visual examination of a reactor coolant pump internal pressure boundary surface until the 1981 refueling outage.

(Items B5.6 and 5.7, Exami-nation Categories B-L-1 and B-L-2, respectively)

This request is granted.

7.

The licensee requests to examine 100% of one head-tr.-shell and 100% of one shell-to-tubesheet weld in lieu of a smo.

percentage of each well (head-to-shell and shell-to-tubesheet) in the Regenerative Heat Exchanger.

(Item 83.1, Examination Category B-B).

This request is granted.

8.

Request to defer the visual examination of clad patches in two steam generators and the pressruzier in the 1981 reactor refueling outage.

(Itesm B3.8 and 2.9, Examination Category B-I-2).

This-request is granted.

9.

The licensee requests relief from volumetric examination of steam generator safe end to' piping welds.

(Item B4.5, Examination Category B-J).

This request is granted.

10.

The licensee requests relief from volumetric examination of Regenerative Heat. Exchanger nozzle to vessel welds.

(Item B3.2, Examination Category B-D).

This request is granted.

11. The licensee requests relief from volumetric examination of the nozzle-to-vessel weld on.the Residual Heat Exchanger.

(Item C1.2, Examination ' Category C-B).

This request is granted.

45 a

- 12.

The licensee requests re',ief from volumetric examination of certain class 2 welds (Items C2.1, C2.2, C2.3, Category C-f, Itesm C2.1, C2.2 C2.3, Category C-G) in the feedwater, main steam, CVCS, and residual heat renoval piping.

This request is denied.

13.

Request to rest the Cherr.ical and Volume Control Charging (Charging pump discharge), Seal Injection, and Letdown systems at lower presst s than Code required as shown below:

Class ~2 System Test Pressure Code Test Pressure nemical and Volume Control Crarging (Charging Pump Disenarge) 2400 psig 3419 psig Seal' Injection.

2400_psig 3419 psig Lets wn 100 psig 625 psig This request _ is granted.

-14.

Request to-test the Component Cooling System at normal operating pressure (85 psig)-in lieu of the Code-required test pressure

~

~

(165 psig).

This request-is granted.

15. The-licensee : requests - to' use Appendix -III. of Section XI-in the -

Winter,1975 Addenda to 1974 Edition of the Code.

This request is granted.

16. The licensee requests that calibration blocks be made..to the requirements of _ Article T-434.1 in the Winter 1976. Addenda of Section V in lieu.of I-3121 of Section XI;

' This request is granted.

Inservice Testing (Val'ves) - Sumary of Relief Requests 1.

The: licensee has; requested relief _ from the part-stroke requirement of Section XI for' all _ power. operated valuves.

This. request is granted.

46:

. -a -.

2.

Relief Reouest: The following Category A, containment isolation valves (CIV's) will meet Appendix J seak testing requirements in lieu of Section XI requirements:

Valve Valve Valve Valve Dh-TV-1847 CC-CV-885 SS-V-999 CH-CV-305B ED-V-506 CC-TV-1831 SS-V-999A CH-CV-305C BD-V-515 CC-V-884 PW-CV-139 CH-CV-3050 50-V-522 VS-TV-1848 PW-CV-140 DH-TV-1841 BO-V-529 SOV-12-1 HC-V-212 DH-TV-1844 20-TV-1312-1 VS-CV-1104 PU-V-242 FM-MOV-31 SD-TV-1312-2 CC-CV-731 PU-V-242A FW-CV-192 EO-TV-1312-3 FCV-611 HS-CV-295 FW-CV-194 BO-TV-1312-4 VH-V-507 HS-CV-295A FW-CV-196 VH-V-507B CC-GCV-608 BV-1-1B FW-CV-198 DH-TV-554 CC-CV-721 HCV-1101 FH-CV-296 WD-HICV-1840 WG-A0V-558 BV-1-1A WD-TV-1846 WG-TV-1845 P50 OH-TV-1843 SS-V-984A SA-V-411A

' OH-TV-1844 SS-TV-950 SA-V-413 LM-TV-1811A SS-TV-955 LD-A0V-202 LY-TV-18119 SS-TV-960 LO-A0V-203 L"-TV-1812 SS-TV-965.

LD-A0V-204 CC-CV-853 DH-TV-1842A CH-TV-334 CC-TV-1411 DH-TV-1842B CH-CV-305A This request is granted, except that the licensee is not relieved of the requirements of paragraphs IWV-3420f f) and (g).

2.

The licensee requests a waiver from the requirements of IWV-3410(c)(3).

This request is denied.

3.

The licensee requests a waiver from the requirements of IWV-3410(g).

This request is denied.

4.

Valves SI-CV-862A, SI-CV-8628, SI-CV-862C and SI-CV-862D will not be exercised in accordance with Code Requirements.

This request is granted as applicable to the current testing period,' but the request must be resubmitted with additional information before relief can be granted from future testing.

5.

Valves XI-CV-103, SI-CV-107A, SI-CV-1078, CD-CV-872A, and CD-CV-872B will be part-stroked at refueling outages only in lieu of Code requirements.

~This request is granted, as applicable to the current testing period, but the request must be resubmitted with additional information before relief can be granted from future testing.

47

6.

Valves SI-CV-856A and 856B will not be full stroke exercised in accordance with Section XI requirements.

This request is granted, as applicable to the current testing period, but the request must be resubmitted with additional information before relief can be granted from future testing.

7.

Check Valves RH-CV-783 and RH-CV-808A will not be exercised in accordance with Code Requirements.

This request is denied.

8.

Valves CH-FCV-110 and CH-FCV-110A will be full stroke exercised at reactor refuelings during scheduled safety inspection system testing.

This request is granted.

9.

Valves SW-MOV-5 and SW-MOV-6 will not be exercised in accordance with Section XI reouirements.

This request is denied.

10. The licensee has requested relief from exercising the following valves per Code requirements for. Category A valves.

The valves are CIV's.

Valve Category Valve Service DW-CV-139 AC Primary Water to PRT PW-CV-140 AC Primary Water to PRT HC-V-212 AE Spaco Heater Containment Return DU-V-242.

AE Cavity Purification.Line DU-V-242A AE Cavity Purificat'an Line

-HS-CV-295 AC Containment Space Heating Supply MS-CV-295A AC Containment Scace Heating Supply EV-1-1B AE Containment Purge Air Exhaust MCV-1101

.AE Containment Purge Air Exhaust Bypass SW-1-1A AE Containment Purge Air Supply P50 AE Fuel Transfer Tube SA-V-411A AE

. Air Monitor Purge SA-V-413 AE Service Air to Containment BD-V-506 AE Steam Generator No. 1 Blowdown

' 3D-V-515 AE

-Steam Generator No. 2 Blowdown 3D-V-52.

AE Steam Generator No. 3 Blowdown 30-V-529 AE Steam Generator No. 4 Bloydown CC-CV-885 AC CC Water to Neutron Shield Tank Cooler CC-V-884 AE.

Neutron Shield Tank Fill Line VS-CV-1104 AC Air Monitor SampleEto Containment

.CC-CV-731 AC CC Water Supply to Drain Cooler VH-V-507 AE Primary Vent Header VH-V-507B AE Primary Vent Header

.CC-CV-853 AC CC Water to RCP Oil Coolers CC-CV-721 AC.

CC Water'to RCP Thermal Barrier SS-V-999 AE Neutron Shield Tank Sample 48

SS-V-999A AE Neutron Shield Tank Sample FW-CV-192 AC Aux. Feed Water Supply to S.G. No. 4 FW-CV-194 AC Aux. Feed Water Supply to S.G. No. 3 FW-CV-196 AC Aux. Feed Water Supply to S.G. No. 2 FW-CV-198 AC Aux. Feed Water Supply to S.G. No. 1 FH-CV-296 AC Loop Fill Header Check SS-V-984A A

PRT Sample This request is granted.

11. The licensee requests relief from quarterly stroking valves SI-MOV-861 A, SI-MOV-861B, SI-MOV-861 C, SI-MOV-861 D, CD-MOV-871 A, CD-MOV-8718.

This request is granted.

5.0 Reactor Coolant System Pressure Isolation Valves The Reactor Safety Study (RSS), WASH-1400, identified in a PWR an inter-system loss of coolant accident (LOCA) which is a significant contributor to risk of core melt accidents (Event V).

The design examined in the RSS contained in-series check valves isolating the high pressure Primary Coolant System (PCS) from the Low Pressure Injection System (LPIS) piping.

The scenario which leads to the Event V accident is initiated by the failure of these check valves to function as a pressure isolation barrier. This causes an overpressurization and rupture of the LPIS low pressure piping which results in a -LOCA that bypasses containment.

In order to better define the Event V concern, all light water reactor licensees were requested by letter dated February 23, 1980, to provide the following -in accordance with 10 CFR 50.54(f):

i. Describe the valve configurations lat your plant and indicate if -

an Event Y _ isolation valve configuration exists within the Class I boundary of the high pressure piping connecting PCS piping to low pressure system piping; e.g., (1) two check valves in series, or;(2) two check valves in series with a motor operated valve-(MOV);

49

2.

If either of the above Event V configurations exist at your facility, indicate whether continuous surveillance or periodic tests are being performed on such valves to ensure integrity.

Also indicate whether valves have been known, or found, to lack integrity; and 3.

If either of the above Event V configurations exist at your facility, indicate whether plant procedures should be revised or if plant modifications should be made to increase reliability.

In addition to the above, licensees were asked to perform individual check valve leak testing prior to plant startup af ter the next scheduled outage.

By letter dated March 18, 1980, the licensee responded to our February letter. Based upon the NRC review of this response as well as the review of previously docketed infomation for your facility, we have concluded that one or more valve configuration (s) of concern exist at your facility.

The attached Technical Evaluation Report (TER)(Attachment 1) provides, in Section 4.0, a tabulation of the subject valves. We note that contrary to the statement in the TER, the licensee did identify the configurations of concern in his March 18, 1980 submittal.

The staff's concern has been exacerbated due not only to the large number of plants which have an' Event V configuration (s) but also because of recent unsatisfactory operating experience.

Specifically, two plants have leak tested check valves with unsatisfactory results. At Davis-Besse, a pressure isolation-check valve in the Low Pressure Injection System failed-and the ensuing investigation found that, valve internals had become disassembled. At the Sequoyah Nuclear Plant, two Residual Heat Removal (RHR) injection check valves 'and one RHR recirculation check valve failed because valves jacuned open against valve over-travel limiters.

50-

r It is, therefore, apparent that when pressure isolation is provided by two ir-series check valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required.

Since these valves are important to safety, they should be tested periodically to ensure low probability of gross f ailure. As a result, the staff has determined that periodic examination of check valves should be undertaken to verify that each valve is seated properly.and functioning as a pressure isolation device. Such testing' will ' reduce the overall risk of an inter-system LOCA.

By letter dated November 4,1980, the licensee proposed changes.to the Haddam Neck Plant Technical Specifications which would remove the require-ment to stroke the MOV's in.the systems of concern on a conthly basis.

Occrational testing of these valves would be performed only at cold shutdown.

We conclude that the changes proposed by the licensee, when combined with..the additional testing requirements for the check vavles in question, will increase the level of' assurance 'that ' multiple pressure isolation barriers are in place and will. remain intact.

Although not explicitly stated, we 'have interpreted the November 4,1980 letter as' a request for relief from quarterly _ stroke testing-these MOV's as required'by ASME Xi.

This request is addressed in _Section 3.2.9.11 of' this evaluation.

In addition to' Event V valve configurations, we are continuing our efforts to review other configurations located at high pressure / low pressure system boundaries for their potential risk contribution to an intersystem LOCA. Therefore, further activity regarding the broader topic of intersystem LOCA's may be expected in the future.

i 51

.,~

6.0 Environmental Consideration We have determined that this amendment and granting of the relief do not auth,orize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.

Having made this determination, we have further concluded that the amendment and the relief involve actions which are insignificant from the standpoint of environmental impact, and pursuant to 10 CFR 51.5(d)(4) that an environmental impact state-ment, or negative declaration and' environmental impact appraisal need not be prepared in connection with the issuance of these actions.

7.0 Conclusion We nave concluded. based on the considerations discussed above, that:

(1) because the amendment does not involve a significant increase in the prob-ability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of.the public will not be endangered by operation in the

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

Date:

February 26, 1981

Attachment:

Primary Coolant System Pressure Isolation Valve Technical Evaluation Report f

52

-