ML20138B183

Schedular Exemption Requests Re Type C Leak Testing

ML20138B183
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	03/31/1986
From:	CONNECTICUT YANKEE ATOMIC POWER CO.
To:
Shared Package
ML20138B132	List: B11994, Requests Exemption from Requirements of App J to 10CFR50. Util Has Initiated Detailed Evaluation of Scope & Schedule for Mods Required to Achieve Compliance W/App J as Part of Facility Isap ML20138B156 ML20138B183 ML20138B200
References
NUDOCS 8603200296
Download: ML20138B183 (27)

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k Docket No. 50-213 i

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I ATTACHMENT 2 I

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t HADDAM NECK PLANT i

SCHEDULAR EXEMPTION REQUESTS RELATED ,

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.TO TYPE C LEAK TESTING '

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! 2-1 March 1986

1.0 GENERIC DISCUSSION OF REQUESTS FOR SCHEDULAR EXEMPTIONS 1.1 Introduction In response to an NRC request dated January 2,1975(l), Connecticut Yankee ~

Atomic Power Company (CYAPCO), submitted proposed changes to the Haddam he Neck Plant Technical requirements of 10CFR50,Specifications which were on Appendix 3. Subsequently, intended May 28, to implement {2),

1973 s o Appendix 3.

CYAPCO requested These exemption requests exemptions from certain were later supplemented of the requiremen}3-9. On May 7 and modifiedt 1982(10), the NRC Staff issued Amendment 49 to Facility Operating License No.

DPR-61 for the Haddam Neck Plant, which issued revised Technical Specifica-tions reflecting the requirements of Appendix 3.

CYAPCO documented (ll,12) the need to re-evaluate containment penetration designs and proposed modifications required to achieve Appendix 3 compliance in light of the results of SEP Topic VI-4, containment isolation, i.e., modifications required to achieve compliance with 10CFR50, Appendix A. Based upon the favorable experiencg of the Systematic Evaluation Program (SEP), CYAPCO -

expressed the belief tl3,141 that a similar integrated evaluation of all pending regulatory requirements would result in the greatest increase in plant safety.

In a letter dated December 28, 1983(15), CYAPCO specified " Containment Penetration Evaluations" as an issue to be evaluated under the auspices of the Haddam Neck Plant Integrated Safety Assessment Program (ISAP). At that time, CYAPCO informed the NRC Staff that modifications required to achieve compliance with 10CFR50, Appendix 3 would be addressed as part of the scope of ISAP. CYAPCO further stated that periodic leak rate ~ testing at the Haddam Neck Plant, coupled with the low probability of a design basis accident, provided adequate justification for deferral of Appendix 3 modifications.

In an April 5, 1984(16) letter, the NRC Staff noted that not all containment penetrations are tested in accordance with Appendix 3. The Staff concluded that it was acceptable to defer implementation of specific Appendix 3 and Appendix A modifications until an integrated assessment, e.g., ISAP, could be performed.

The basis for the StafI's conclusion was that, although the integrated contain-ment (Type A) leak test is not performed as frequently as local leak rate tests would be, the integrated leak rate test does provide an indication of overall containment leak-tightness, including penetrations.

In a July 31,1985 letter (I7), the NRC Staf f formally established the scope of the Haddam Neck Plant ISAP and designated Appendix 3 issues as ISAP Topic 1.03,

" Containment Penetration Evaluations." In this letter, the Staff recognized that some issues would require exemptions to defer action until such time as the Haddam Neck Plant ISAP is completed.

Recently, CYAPCO has completed a comprehensive review of the status of compliance with Appendix 3 for the Haddam Neck Plant. Accordingly, CYAPCO is requesting schedular exemptions for certain penetrations which do not currently comply with 10 CFR 50, Appendix 3. CYAPCO intends to evaluate these areas of noncompliance within the framework of the Haddam Neck Plant ISAP. The purpose of evaluating these Appendix 3 issues within the ISAP 2-2 2

framework would be to perform an integrated assessment of all pending regulatory requirements and utility-initiated projects, ascertain the necessity or appropriateness of implementing potential modifications, establish priorities for the various items in ISAP, and schedule the implementation of each item on a relative basis.

Inherent in CYAPCO's approach is the fact that adequate justification exists for continued operation to ensure that public health and safety are not compromised.

In addition, CYAPCO has been continually encouraged by the NRC Staff at meetings and in formal correspondence of its concurrence with the appropriate-ness of the approach described above. In fact, the Commission's ISAP Policy Statement (49 Federal Register 45113, November 15, 1984) endorses integration of actions such as Appendix 3 compliance into the ISAP, provided good cause is shown. In light of the above summary and the attached justification for the proposed schedular exemptions, CYAPCO believes that adequate basis exists for granting the requested exemptions.

1.2 Exemption Criteria The Commission's regulations, specifically,10 CFR 50.12(a), provide that exemp-tions may be granted from the regulations in 10 CFR 50 provided that they are

" authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest." On December 12,1985, the Commission published the final revisions to 10 CFR 50.1;(a), regarding standards to be applied in granting exemptions (50 Federal Register 50764, December 12, 1985). The purpose of the final rule was to revise and clarify the criteria for granting exemptions. CYAPCO's requests for schedular exemptions is based on the revised 10 CFR 50.12(a). An evaluation of each of the schedular exemptions requested by CYAPCO is presented in the following sections of this at tachment.

Based on the information provided in subsequent sections of this attachment, CYAPCO concludes that schedular exemptions from the requirements of 10 CFR 50, Appendix 3 are justified pursuant to paragraph 50.12(a)(2)(v) of the final rule. That is, they are:

• Authorized by law;

• Will not present undue risk to public health and safety; and

• Are consistent with the common defense and security.

Additionally, the exemptions would provide only temporary relief from the application of regulation, and CYAPCO has made good faith efforts to comply with the regulation, as documented in previous discussions and correspondence with the NRC Staf f.

In general, the intent of the Commission's regulations and other NRC require-ments is to provide reasonable assurance that operation of nuclear power plants does not pose an undue risk to the health and safety of the public. The intent of specific regulations, such as Appendix 3, is to set standards which will provide reasonable assurance that the individual contributions to risk posed by specific issues or concerns will be low, such that overall plant risk remains acceptably low. While compliance with the regulations will, for the most part, provide 2-3

reasonabie assurance that plant operation does not pose undue risk to the health and safety of ' the public, noncompliance does not necessarily represent an unacceptable risk.

In the specific case of the Commission's requirements regarding containment penetrations, the underlying purpose of these requirements is to assure that any radioactive materials released into the containment during a postulated loss-of-coolant accident will be suitably contained and that release to the outside environment will be small, i.e., that the risk associated with such releases'is low. One means for assessing the contribution to risk posed by individual issues is through the use of probabilistic risk assessment (PRA) techniques. The recently- completed Haddam Neck Probabilistic Safety-Study (PSS) (to be released in March, 1986) provides plant-specific information regarding expected core damage frequency. The results of the PSS coupled with conservative predictions of the'offsite public consequences (for Siting Source .

Term 3) for the Haddam Neck ~ Plant, based on the Sandia Siting Study, NUREG/CR-2239, " Technical Guidance for Siting Criteria Development,"

indicate that several of the proposed modifications would provide a small, if not insignificant, reduction in public risk.

When coupled with the results documented in a draf t report prepared by EG&G Idaho, Inc. for the Department of Energy, eatitled, " Benefits to the Light Water Reactor Industry Resulting from Reduced Radiological Source Terms," the consequences of a small amount of leakage from the containment are not as serious as was believed when Appendix 3 was first promulgated in 1973. The draft report concludes that the acceptance criteria for leakage rate testing required by Appendix 3 may be too restrictive. This conclusion is supported by the following:

o All risk assessments performed to date show that accidents with containment leakage at or near the design leakage rate where containment remains intact do not contribute significantly to overall public risk; o Pub!!c risk is dominated by accidents for which:

a. the containment is bypassed (LOCA outside containment)

b. containment heat removal and fission product removal are not available, thereby leading to relatively early and catastrophic containment failure due to overpressure

c. gross failure of containment isolation such as via unisolated purge and vent lines.

At the Haddam Neck Plant, item c above is not a credible scenario due to administrative controls which require that these valves be locked closed during power operation. Also a somewhat more " leak tight" containment, as might be expected from compliance with Appendix 3, will have no impact on Items a and

b. Therefore, the modifications needed to comply with Appendix 3 would have no significa.it impact on public risk.

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On a more plant-specific level, offsite dose consequence calculations for the Haddam Neck Plant, performed in accordance with 10CFR100 and Regulatory Guide 1.4, do not take' credit for plant design features or analytical conservatisms which would tend to mitigate or lessen potential accident consequences.

These design features and analytical conservatisms include the following:

o The closed nature of most systems inside or outside containment which tends to provide a redundant barrier to containment leakage, o Many systems normally operate at pressures greater than Pa. This would eliminate containment atmosphere leakage as long as system pressure is maintained post-accident.

o Charcoal filters in the primary auxiliary building (PAB) which tend to significantly reduce the iodine contribution to of fsite doses. These filters are tested to confirm a 90% iodine removal efficiency, o The more realistic source terms currently being evaluated by the NRC and industry.

o A radiological analysis of a postulated design basis LOCA was performed for the Haddam Neck Plant as part of the Systematic Evaluation Program (SEP). The doses were calculated using the source term assumptions of Regulatory Guide 1.4 (e.g.,100% core inventory of noble gases and 25%

core inventory of iodines). Recent studies have indicated that a more realistic iodine source term would be ten times lower. A reduction in iodine source terms would make the whole body dose more limiting but would accommodate a containment leak rate approximately three times higher than the currently assumed containment leak rate of 0.13% per day.

All Haddam Neck Plant penetrations are affected by some of the above considerations.

4 1.3 Conclusion l

In summary, CYAPCO has concluded that premature address of Appendix 3 noncompilances is clearly not justified, and that schedular exemptions for those areas of noncompliance discussed in the following sections of this attachment are warranted, pursuant to 10 CFR 50.12(a)(2)(v). CYAPCO intends to address the status of Haddam Neck Plant compliance with Appendix 3 as part of the ISAP, as previously discussed.

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1.4 References (1) E. 3. Brunner letter to D. C. Switzer, dated January 2, .1975, " Inspection Report No. 50-213/74-15."

(2) D. C. Switzer letter to R. A. Purple, dated May 28,1975, " Exemption from Provisions of 10CFR50 Appendix 3."

(3) D. C. Switzer letter to R. A. Purple, dated August 20, 1976. (Special Report re Containment Operation and Testing).

(4) D. C. Switzer letter to A. Schwencer, dated December 27, 1976,

" Containment Leak Rate Testing."

(5) D. C. Switzer letter to A. Schwencer, dated August 8,1977, " Proposed Exemptions to 10CFR50, Appendix 3."

(6) D. C. Switzer letter to A. Schwencer, dated September 19,1977, " Proposed Exemptions to 10CFR50, Appendix 3."

(7) W. G. Counsil letter to D. L. Ziemann, dated June 12,1978, " Proposed Exemptions to 10CFR50, Appendix 3."

(8) W. G. Counsil letter to D. L. Ziemann, dated November 13,1978, " Proposed Exemptions to 10CFR50, Appendix 3."

(9) .W. G. Counsil letter to D. L. Ziemann, dated July 24,1979, " Proposed Exemptions to 10CFR 50, Appendix 3."

(10) D. M. Crutchfield letter to W. G. Counsil, dated May 7,1982, " Appendix 3 Requirements (Containment Leakage Testing)."

(11) W. G. Counsil letter to D. M. Crutchfield, dated January 5, 1983,

" Appendix 3 Requirements."

(12) W. G. Counsil letter to D. M. Crutchfield, dated August 16, 1983, "10CFR50 Appendix 3 Requirements, SEP Topic VI-4, Containment Isolation System."

(13) W. G. Counsil letter to W. Dircks, dated June 13,1983, " Integrated Safety Assessment Program."

(14) W. G. Counsil letter to D. G. Eisenhut, dated September 14, 1983,

" Integrated Safety Assessment Program."

(15) W. G. Counsil letter to D. G. Eisenhut, dated December 28, 1983,

" Integrated Assessment of Regulatory Requirements."

(16) D. G. Eisenhut letter to W. G. Counsil, dated April 5,1984, " Expanded integrated Assessments for Haddam Neck and Millstone Unit No.1."

(17) H. L. Thompson letter to 3. F. Opeka, dated July 31,1985, " Integrated Safety Assessment Program."

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r 2.0 SCHEDULAR EXEMPTIONS REQUESTED 2.1 Exemption From the Requirements of Section II.H of Appendix 3 2.1.1 Schedular Exemption Requested Section II.H of 10 CFR 50, Appendix 3 requires that Type C tests be performed for containment isolation valves that:

"1. Provide a direct connection between the inside and outside atmospheres of the primary reactor containment under normal operation, such as purge and ventilation, vacuum relief, and instrument valves;

2. Are required to close automatically upon receipt of a containment isolation signal in response to controls intended to effect containment isolation;

3. Are required to operate intermittently under postaccident conditions; and

4. Are in main steam and feedwater piping and other systems which penetrate containment of direct-cycle boiling water power reactors."

CYAPCO requests a schedular exemption from the Type C testing requirements of Section II.H. of Appendix 3 for the reactor coolant charging (P-8), and containment sump to residual heat removal (P-73), penetrations.

2.1.2 Justification for Exemption 2.1.2.1 Penetration P-8:

CYAPCO had previously requested an exemption from the Type C testing requirements of Appendix 3 for the reactor coolant charging (P-8) penetration.

This previous request was based on the reismic design of system piping inside containment and the proposed seismic qualification of system piping from the isolation valves of Penetration P-8 to its water source. Subsequent evaluations determined such qualification to be a lengthy and costly effort; therefore, CYAPCO intends to modify Penetration P-8 to permit Type C testing. In the '

interim, although Penetration P-8 is not currently Type C tested, the isolation valves in this penetration are exposed to containment pressure through the vented reactor coolant system during the CILRT.

The valves are exposed to containment pressure in the direction of accident I pressure with the backside of the valves depressurized. Further, the portion of the system outside containment is checked for liquid leakage in accordance with Administrative Technical Specification 3.14; system leakage is accounted for in of fsite dose consequence calculations, per 10CFR100. Such leak testing provides additional assurance that the potential for significant containment atmosphere leakage through this penetration is minimized.

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CYAPCO believes a schedular exemption from the Type C testing requirements of Appendix 3 is justified based on the above information as well as the long lead time for equipment needed to modify this penetration, the level of design and analysis required, and the need to integrate these modifications with the ultimate resolution of the SEP review of containment isolation valve configura-tions per 10CFR50, Appendix A. This approach is consistent with the ALARA philosophy in that it avoids repetitive modifications to the same penetration.

The ultimate resolution of this issue will be accomplished within the framework

.of the Haddam Neck Plant ISAP.

2.1.2.2 Penetration P-73:

The containment sump to RHR pump suction line (P-73) penetration has not previously been Type C tested, it is isolated by three normally closed valves.

The motor operated valve RH-MOV-22 and manual valve RH-V-303A are located outside containment and isolate the RHR system, which is operating during the CILRT; the backside of these valves is, therefore, not ventable during the CILRT. Appendix 3 does not require that these valves be tested since the RHR system will be filled with water and operating in the post-accident condition, considering a single active failure. The RHR system, however, has an Administrative Technical Specification leakage limit of 3 liters / hour (Reference LCO 3.14). This !cakage limit encompasses the makeup, seal injection / return, and fill portions of the charging system as well. . Any leakage through these systems is accounted for in of fsite dose consequence calculations per 10CFR100.

The third isolation valve of Penetration P-73 is a check valve CC-CV-802, which has not been local leak rate tested previously. This valve isolates the component cooling water surge tank re!Ief line. Should this valve f all to isolate, relief valve CC-RV-777 would prevent leakage. There are effectively two leakage barriers in this portion of the piping. The possibility of an unisolated leak of containment atmosphere is, therefore, not credible. During a postulated accident reactor coolant and safety injection water will fill the lower elevation of the containment and containment sump for the duration of the long-term cooling phase. Therefore, a 30-day water seal exists for this portion of the penetration.

This penetration will be water tested in the future once physical modifications have been completed to permit such testing. CYAPCO believes a schedular exemption from the Type C testing requirements of Appendix 3 is justified for valve CC-CV-802 based on the above information as well as the long lead time for equipment needed to modify this penetration, the level of design and analysis required, and the need to integrate these modifications with the ultimate resolution of the SEP review of containment isolation valve configurations per 10CFR50, Appendix A. This approach is consistent with the ALARA philosophy in that it avoids repetitive modifications to the same penetration. The ultimate resolution of this issue will be accomplished within the framework of the i Haddam Neck Plant ISAP.

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2.2 Exemption from the Requirements of Section Ill.C.1 of Appendix 3 2.2.1 Exemption Requested Section Ill.C.1 of Appendix 3 to 10 CFR 50 states:

" Type C tests shall be performed by local pressurization.

The pressure shall be applied in the same direction as that when the valve would be required to perform its safety function, unless it can be determined that the results from the tests for a pressure applied in a different direction will provide equivalent or more conservative results."

CYAPCO requests an exemption from the above requirement so that certain valves in the following penetrations may be tested in the reverse direction:

i o P-33 refueling cavity purification; o P-78 pressurizer relief tank drain; and A schedular exemption is requested to allow the reverse direction testing of valve PU-V-242 in Penetration P-33 and DH-TV-554 in Penetration P-78.

In previous correspondence,with the NRC staff, CYAPCO addrcssed the issue of reverse dire _ tion testing of certain valves. The following discussion reflects the most recent status of this issue. In Reference 1, the NRC Staff found CYAPCO's proposals for reverse direction testing of. certain isolation valves to be acceptable, with the exception of valve VH-V-507 (Penetration P-71), which must be tested in the direction of the post-accident LOCA containment 1 pressure. As indicated in Reference 2, Penetration P-71 for the primary vent i header has been modified to allow testing in the direction of post-LOCA containment pressure. Additionally, for Penetration P-62 (Service Air), manual valve SA-V-413 is torqued closed _ to 133 inch-lb handwheel torque, and proce-dures were revised during the 1984 refueling outage so that valve SA-V-413 is in compliance with reverse direction testing criteria. Currently, valve CC-V-884 in Penetration P-63 (neutron shield fil!) is tested in the reverse direction. Since this valve is a manual Grinnel weir-type diaphragm valve which has symmetric internals with a single seating surface, the valve would tend to unseat equally ,

with pressure applied from either direction. Plant procedures require stem and I 4

body-to-bonnett flange leakage to be added to through-valve leakage. Hence, '

reverse direction testing is acceptable and in compliance with Appendix 3; no  ;

i exemption is required to test this penetration in the reverse direction.

2.2.2 Justification for Exemption The refueling cavity purification (P-33) penetration is isolated by two contain-ment isolation valves. Valve PU-V-242A is Type C tested in the direction of accident pressure; valve PU-V-242 is tested in the reverse direction. The pressurizer relief tank drain (P-78) penetration is isolated by valves DT-TV-1844 and DH-TV-554. Valve DT-TV-1844 is tested in the direction of accident 2-9

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pressure; valve DH-TV-554 is tested in the reverse direction. CYAPCO requests a temporary exemption to permit continued reverse direction testing of valves PU-V-242 and DH-TV-554 until modifications can be made to permit testing in the direction of accident pressure. Valves PU-V-242 and DH-TV-554 are tested at the proper direction during the CILRT. As previously concluded by the NRC Staff (Reference 3), although the CILRT is not performed as frequently as Type C leak rate tests, the CILRT does provide an indication of overall containment leaktightness, including penetrations.

CYAPCO believes a schedular exemption to permit reverse direction testing of valves DH-TV-554 and PU-V-242 is justified based on the above information as well as the level of design and analysis required to accomplish modifications to Penetrations P-33 and P-78 to permit testing in the direction of -accident pressure. The ultimate resolution of this issue will be accomplished within the framework of the Haddam Neck Plant ISAP.

2.2.3 References

1. D.M. Crutchfield Letter to W.G. Counsit, dated May 7,1982, " Appendix 3 Requirements (Containment Leakage Testing)."

2. W.G. Counsil letter to D.M. Crutchfield, dated August 16, 1983, "10 CFR 50, Appendix 3 Requirements, SEP Topic VI-4, Containment Isolation System."

3. D. G. Eisenhut letter to W. G. Counsil, dated April 5,1984, " Expanded Integrated Assessments for Haddam Neck and Millstone Unit No.1."

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f 2.3 Exemption From the Requirements of Section Ill.C.2(a) of Appendix 3 2.3.1 Exemption Requested Section III.C.2.(a) of Appendix J to 10 CFR 50 states:

" Valves, unless pressurized with fluid (e.g., water, nitrogen) from a seal system, shall be pressurized with air -

or nitrogen at a pressure of Pa "

A schedular exemption is requested from the requirement to test the following penetrations with air or nitrogen:

P-3 high pressure safety injection P-7 RCP seal water return P-10 RCS letdown P-11 RCS sampling 1

P-24 safety injection recirculation P-28 CCW to RCP oil coolers P-30 space heating steam supply P-34 CCW from RCP thermal barrier P-38 CCW to RCP thermal barrier P-60 CCW to neutron shield cooler P-61 CCW from neutron shield cooler P-63 neutron shield fill P-66 CCW to drain cooler P-67 CCW from drain cooler P-68 primary water to containment P-69 loop fill P-74 RCP seal water supply P-75 RCP seal water supply P-7$ RCP seal water supply P-77 RCP seal water supply 2-11 i

2.3.2 Justification for Exemption All of the penetrations listed above are tested with water during Type C testing.

While no quantitative correlation 'between water and air leakages, has been

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accepted by the NRC Staff to date, CYAPCO believes that the current testing method at the Haddam Neck Plant provides a means to accurately identify those penetrations which represent significant leakage paths. In all cases, plant procedures require repair and retesting of penetrations where specific individual valve liquid leakage limits are exceeded.

While a thirty day water seal cannot be proven for these penetrations if only engineered safeguards equipment must be relied upon, these lines are normally water-filled and are likely to remain water-filled during the early stages of a postulated LOCA, when containment pressure peaks. ' Approximately ten hours af ter the onset of a postulated accident, the containment pressure is less than 1.5 psig. Hence, during the post-accident period when air leakage is most likely to occur, the driving force behind such leakage is substantially reduced.

Offsite dose calculations for the Haddam Neck Plant demonstrated that half of the site boundary thyroid dose due to containment leakage occurs during the first fifteen minutes following the onset of a postulated accident. Temporary water seals in these penetrations are likely during the period when the dose consequences of a postulated accident are greatest.

In previous submittals to the NRC (Reference 1,2 and 3), CYAPCO addressed elements of the scope of the Integrated Safety Assessment Program (ISAP) for the Haddam Neck Plant. In these letters, CYAPCO communicated to the NRC its position that, based upon the favorable experience of the Systematic Evaluation Program for the Haddam Neck Plant, a similar integrated evaluation of all pending regulatory requirements would result in the greatest increase in plant safety. In Reference 2 and 3, CYAPCO identified the need for an integrated evaluation of modifications necessary to resolve 10 CFR 50, Appendix 3, leak testing issues, and 10 CFR 50, Appendix A, containment isola-tion issues. On April 5,1984 (Reference 4), the~NRC concluded that deferral of modifications to permit testing of certain penetrations with air or nitrogen was acceptable, based on CYAPCO's address of this issue as part of the Haddam Neck Plant IS AP. As outlined in the July 31, 1935 letter to CYAPCO (Reference 5), the NRC has designated this issue as IS AP Topic 1.03, "Contain-ment Penetrations."

Based on the above information, CYAPCO believes a schedular exemption from the requirements of Appendix 3 to permit water testing of the penetrations identified above is justified. The ultimate resolution of this issue will be accomplished with the framework of the Haddam Neck Plant ISAP. Additional penetration-specific justification for this schedular exemption is provided below.

Penetrations P-7, 69, 74, 75, 76, and 77:

Leakage for these penetrations, which are the seal water return, loop fill, and seal injection portions of the charging system, is addressed by Administrative Technical Specification 3.14. The Specification requires that these systems outside containment be checked for liquid leakage from seals, flanges, and valves. The leakage limit of 3 liters / hour, established by Administrative Technical Specification 3.14 for these systems, includes leakage from the RHR 2-12

and charging makeup lines as well an'd is included in the offsite dose consequence calculations per 10CFR100. This provision for additional leakage limits provides further assurance that the potential for significant leakage of containment atmosphere through these penetrations is' minimized.

Penetration P-10:

While not safety-related, the letdown system is closed outside containment.

Penetration P-II:

The RCS sampling lines consist in part, of four small, i.e. 3/8", Masonellan globe valve in parallel which are normally isolated against full RCS pressure. Since these containment isolation valves tend to unseat more as pressure increases, testing at 2000 psig is, in fact, conservatively maximizing leakage through these valves.

Penetrations P-28, 34, 38, 60, 61, 63, 66, and 67:

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While not safety-related, the component cooling water system is closed inside containment and maintained at a normal pressure in excess of accident pressure.

With the exception of the component cooling surge tank vent, the system is also closed outside containment. The surge tank vent automatically isolates on high radioactivity in the vent line.

Penetration P-30:

Leakage through this penetration would be filtered through the primary auxiliary building charcoal filters, which tend to significantly reduce the iodine contribution to offsite doses. In addition, more realistic source terms would tend -

to further reduce the offsite consequences of potential leakage.

Penetration P-68:

Although the primary water system is not safety-related, it is closed inside containment and includes two check valves in series near the penetration.

System pressure is normally maintained at a pressure greater than accident pressure.

Penetrations 3 and 24:

The safety grade safety injection system operates at a pressure greater than the accident pressure Pa during the early stages of a design basis accident when containment pressure peaks. Therefore, the potential for containment atmosphere leakage only exists during the long-term post-LOCA period.

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2.3.3 References

1. W.G. Counsil letter to W. Dircks, dated June 13,1983, " Integrated Assessment of Regulatory Requirements."

2. W.G. Counsil letter to .D.G. Eisenhut, dated September 14, 1983,

" Integrated Safety Assessment Program."

3. W.G. Counsil letter to D.G. Eisenhut, dated December 28,-1983,

" Integrated Safety Assessment Program."

4. D.G. Eisenhut letter to W.G. Counsil, dated April 5,1984, " Expanded Integrated Assessments for Haddam Neck and Millstone Unit 1."

5. H.L. Thompson letter to 3.F. Opeka, dated July 31,1985, " Integrated Safety Assessment Program."

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2.4 Exemption from the Requirements of Section III.C.2 of Appendix 3 2.4.1 Exemption Requested Section III.C.2 of Appendix 3 to 10 CFR states:

(a) Valves, unless pressurized with fluid (e.g., water, nitrogen) from a seal system, shall be pressurized with air or nitrogen at a pressure of Pa-(b) Valves, which are scaled with fluid from a seal system, shall be pressurized with that fluid to a pressure not less than 1.10 Pa-A schedular exemption is requested for the following penetrations to allow testing these valves at pressures greater than Pa -

P-Il RCS Sampling P-63 neutron shield fill P-69 loop fill Currently, certain penetrations are tested using a pressure decay scheme of 47 psig to 42 psig, rather than being tested at Pa . While this pressure is slightly in excess of Pa (40 psig), it is not expected to significantly affect test results.

CYAPCO believes that the existing test method satisfies the intent of Section III.C.2 of Appendix 3 and that penetrations so tested do not require exemptions. Therefore, only the three penetrations listed above, which are tested at pressures significantly in excess of Pa, require further address below.

2.4.2 Justification for Exemption In a preceding section of this Attachment, CYAPCO requested a temporary exemption from the requirements of Appendix 3 to permit continued Type C testing of the above penetrations with water rather than air. Currently, these water-filled, water-tested penetrations are tested at pressures which are roughly equivalent to the normal system pressure. The test pressures are listed below.

P-11: 2000 psig P-63: 70 psig P-69: 2000 psig ,

in previous submittals, (References 1, 2, and 3), CYAPCO discussed plans to modify these penetrations to permit Type C testing at accident pressure Pa -

CYAPCO believes that the requested schedular exemption is justified based on the above information as well as the long lead time for equipment needed to modify these penetrations, the level of design and analysis required, and the need to integrate these modifications with the ultimate resolution of the SEP review of containment isolation valve configurations per 10CFR50, Appendix A. The ultimate resolution of this issue will be accomplished within the framework of 2 -,

T the Haddam Neck Plant ISAP. Additional penetration specific justification are provided below.

Penetration P-ll:

This penetration is tested at reactor coolant system pressure rather than Pa-The RCS sampling lines consist,in part, of four small, i.e., 3/8" Masonellan globe valves in parallel. These containment isolation valves would tend to be unseated more as pressure increases. Therefore, these valves would leak more when tested at RCS pressure (2000 psig) than when tested at accident pressure (40 psig). CYAPCO believes that the current test pressure of 2000. psig conservatively maximizes leakage through these valves.

Penetration P-63:

As previously discussed, valve CC-V-834 in Penetration P-63 is a manual Grinnell weir-type diaphragm valve which has symmetric internals and a single seating surface. The valve is currently tested in the reverse direction and would tend to unseat equally with pressure applied in either direction. Through-valve leakage due to a test pressure greater than accident pressure, i.e., at CCW system pressure, will be conservatively larger than through-valve leakage at accident pressure Pa-Nevertheless, Haddam Neck Plant procedures have been revised to conserva-tively maximize the total penetration leakage to include any leakage from the containment side of the penetration. To accomplish this, leakage will be co!!ected from the valve stem and body-to-bonnet flange with the valve in the open position and the system pressurized. Leakage measured in this fashion will be added to measured through valve leakage. This process will conservatively maximize the total leakage to include any leakage on the containment side of the diaphragm. It is, therefore, appropriate to test this penetration at a pressure greater than Pa.

Penetration P-69:

This penetration is tested at RCS pressure. Any post-accident leakage would be filtered by the PAB charcoal filters. As previously discussed, Administrative Technical Specification 3.14 requires that the combined leakages of the seat water return (P-7), loop fill (P-69), and seal injection charging (P-74, 75, 76, and

77) systems outside containment be limited to 3 liters / hours. This leakage limit is also shared with the RHR and charging makeup lines and is included in ,

10CFR100 dose calculations. This provision for additional leakage limits j provides further assurance that the leakage of coptainment post-accident atmosphere through Penetration P-69 is minimized.

1 2-16

2.4.3 References

1. W. G. Counsil letter to D. L. Zieman, dated June 12,1978, " Proposed Exemptions to 10CFR50, Appendix 3."

2. W. G. Counsil letter to D. M. Crutchfield, dated August 16, 1983, "10CFR50 Appendix 3 Requirements, SEP Topic VI-4, Containment Isolation System."

3. 3. F. Opeka letter to T. E. Murley, dated December 23,1985, "10CFR50, Appendix 3 Compliance."

I 2-17

. _ _ - - = - - _ _ _ _ _ - - _ _ _ . _- __ , _ _ - . , _

Docket No. 50-213 ATTACHMENT 3 HADDAM NECK PLANT STATUS OF COMPLIANCE WITH 10CFR50, APPENDIX 3 March,1986

)

l - - - - - - - - .. ._

Page 1 of 8 CONNECTICUT YANKEE APPENDIX J TESTING STATUS TESTED LEAKAGE BARRIER

• PENEIRATION PROCFDURE NUMBER DESCRIPTION DESIGNATION LOCATION TYPE IEST EXEMPTION SUR P-A Personnel llatch -

I GS OK NO-1 5.1 - 62 0 GS OK N0-1 Equalizing Valve I GL OK NO-1 Equalizing Valve 0 GL OK NO-1 P-B Electrical Penetrations - I GS OK NO 5.7 - 100 0 GS OK N0 P-C Equipment Ifatch -

I GS OK fl0 5,7 - 48 0 GS OK N0 P-D. Dome Vent flange (Top) -

I GS OK NO 5.7-- 50

- 0 GS OK NO P-E Dome Penetration Flange (Side) -

I GS OK NO 5.7 - 49

- 0 GS OK NO P-F-A, B Equipment Hatch Penetrations -

I .GS OK fl0 5.1 - 113 0 GS OK NO P-1 Residual lleat Removal - - -

OK NO-1 -

P-2 Residual lleat Removal - - -

OK NO-1 -

P-3 liigh Pressure Safety SI-CV-862 A to D S Cil LIQ YES-1, 3 5.7 - 65

Injection P-4 Pressurizer Relief Tank WG-TV-1845 I AGA OK NO-1, 4 5.7 - 73 Vent WG-A0V-558 0 AGA OK NO SS-V-981A 0 GA OK NO l

• A key for the codes follows this table 4

Pa ge 2 ,o f 8 CONNECTICUT YANKEE APPENDIX J TESTING STATUS TESTED LEAKAGE BARRIER

• PROCTIMRE PENETRATION DESCRIPTION DESIGNATION LOCATION TYPE TEST EXEMPil0M SIR NUMBER P-5 Spare - -

SW OK i NO-2 -

P-6 flydrogen and Post Accident SS-50V-150 A, D I SGA OK NO 5.7 - 25 Sampling SS-50V-150 B, C 0 SGA OK NO l

P-7 RCP Seal Water Return Cil-TV-334 S AGA LIQ YES-1, 3 5.7 - 60 Cll-RV-332 S RV LIQ YES Cil-CV-262 S Cil LIQ YES NT I YES -

RCS Charging - -

P-8 .-

Ilydrogen and Post Accident SS-50V-151 A D I SGA OK NO 5.7 - 26 P-9 Sampilng SS-SOV-151 B, C 0 SGA OK NO P-10 RCS letdown LD-FCV-202 to 204 S AGL LIQ YES-1, 3 5.7 - 2R P-ilA Loop & Pressurizer Drain SS-TV-950 S AGL LIQ YES-3 5.7 - 29 Sample IIP YES P-llB Pressurizer Steam Sample SS-TV-955 S AGL LIQ YES-3 5.7 - 29 LIP YES Pressurizer liquid Sample SS-TV-960 S AGL l.1Q YES-3 5.7 - 29 P-llc IIP YES Loop 1 flot leg Sample SS-TV-965, AGL LIQ YES-3 5.1 - 29 P-IID S IIP YES l

• A key for the codes follows this table

a CC e e W N CC N & C C C w g M m c a.O e e v Eg i i i i i e i

~ ~ ~ ~ ~ ~ ~

y "h Oc m' m e

m e

m e

W e

. in e

m e

to e

g.n to G

L E

> M M M M M M e I e I t e e E mmmm (A m to m EM IM th M W CC C CC CC wwww wwww wwww x ZZ .. z Zz ZZ >>>> >>>> > > > - >

> C C C C C C m uM w MM MM - CL - CL -C-c -C-A W CC C CC CC J Z -J Z JZWZ ==J Z ==l 4

GC W < <

== W << << << <

44 CC h CC W CC CC C --.! C -J C -J p- CC > << z << << < C < C < C w C &

C 8C m W C

La C 3:

3: ad C

C =

m m

'e- W e- e-C in -C -O to so to to m eJ eEl:

W W

>= Cl3 C W -J

• 3

• th x W.

p-

==

E W E C .-* N M c C <

NN cc e Me N N i

N e

b ** CO Lo C >= cv ac: == v ve === .-= -

C M to M W M N W

W E

C CO CC

-*-a I e Ch Ch Ch

>O O I I C

==

CO CD

===*

I e i W C -

I

==e W

I N

L.D 3d == > > e -> >> > e > I > B K m >.- w > 2> >e- W > > > w >

C W s e a a e e e e e i a e e

> C == to CC ==

CC CD cc C

cc C

cc C

cc C

cc C

cc CC EA 33 M.

D U

e.

>=

W W 0 C C U - Cn - -

C. L #c .c2 E e ' to to .'C CD a 3: to U C th X

.to m X === C .-

>= w C C fC .c L w eg 6-= a i

• • • O & L l E .x E t C C C to i U ec *D 3 80 3 3 3 3 m C - 4A C O O C C W -J W = "C *C *C -

C - a 3 3 3 -

E C C - O o o w o

o m a) no - - -

a E C) CD CD CD to C C 84 to O *

• N M G)

C) L to L  %  %  % *C

> a a C o

- O C CL C C C U

1. c C) C 'O N N N

> Z U > LA m 4M C

.C a

E C s-

• O

>= CC .C CD w CW N N M e LD LC N j E w e= .=e *=* .-e m w h i P= I I I I I I B c A C A C 1 G A X w C E i l

l 1

Page 4 of 8-CONNECTICUT YANKEE APPENDIX J TESTING STATUS ,

t TESTED LEAKAGE BARRIER

• PENETRATION PROCIIMRE NUMBER DESCRIPTION DESIGNATION LOCATION- TYPE TEST EXEMPTION Stm P-18 S/G #4 Blowdown BD-TV-1312-4 S AGA LIQ YES-3 5.7 - 40.

IIP YES BD-V-529 S GL LIQ YES-3 IIP YES P-19 Spare - -

SW OK NO-2 -

P-20 Nitrogen Supply NG-C V-557 i Cil OK N0 5.7'- 75 NG-50V-470 0 SGA OK NO SW OK I NO-2 Spare - -

P-21 -

P-22 Space lleating Condensate IIC-V-220B S GA OK NO 5.7 - 33 I

P-23A Containment Open Bulb Lii-TV-1811A I SGA OK NO 5.7 - 34 System Lil-TV-18118 0 SGA OK NO P-23B Containment Closed Bulb Lii-TV-1812 S SGA OK f NO 5.7 - 35 System P-23C Spare -

S CAP OK NO-2 5.7 -110

> P-23D Air Monitor Purge Vil-V-588 S GA OK N0 5.7 - 37 P-24 Safety injection Recirc. SI-V-863A to D S GL LIO YES-3 5.7 - 66

]

I P-25 Spare - -

SW OK NO-2 -

4

• A key for the codes follows this table

1 t o . .

i E 9 R 1 0 1 R 3 3 6 1 5 t

p H - - - - -

pR - - - - - - -

5 TU 7 7 7 CS 7 7 e O .

5 5

5 5

g R 5 a P P

N O ,

I T 3 1 3 2 - 2 2 - 4 - 2 2 2 P 2 - S - -

S SS - S -

M - -

O EE O O E ,O E O O O E O 0 E N Y3N Y N N N X N N Y N YY N E

T. Q QQ V Q K K K K I I K K E K I K K S I O O O E O O L O LL O O R O L T

R E

- I E A L S R P W W l G l 11 W W G W W W A

l CC S S - A S S S

- U R Y S S lC -

T A T A B T

S E G

G A N N K O I A I 0 5 - -

T E T - - S S I 0 - - I -

S L A E C T D O

- E L J T S

X E I T D

N E N P O P I 1 A 8 A T 3 1 55 A 0 A 5 4 99 2 2 6 4 - - - -

E N - - 8 1 22 - - 4 E G - - - - 2 2 V K I V V VV - - C N S C T CC V V F E - - - - - -

A -

U C Y D C C SS I U C C l I l P P C T

U C

I T

C E n N o r N i e O y t i e C l a r l

- s p c r b r p i a a s e u f B t N r l S i s

O e o r l I l o m u a i T o C a P m r

h t

P o e y e I C 1 t s

R 1 S t h C 1 0 i T w S 1 g v o E 0 P n a P l D C i C C l P R t R o C a g f R m e n m s o l l i o e o r l r e e t f e e e e f e e e d r r c r r u r r r o a a W W a a a f W a a a c p p C C p p p e C p p p S S C C S S S R C S S S e h

t N r O o I

f TR 5 6 7 AE 6 7 8 9 0 1 2 3 4 RB 2 2 2 2 3 3 3 3 3 3 3 3 y T M - - - - - - - - - - - - e EU P P P P P P P P P P P P k NN E A P *

' l i' l ' ,~  !  !

1 Page 6 hf 8 CONNECTICUT YANKEE APPENDIX J TESTING STATUS -

4 TESTED LEAKAGE BARRIER

• PROCFDURE PENEIRAT10N NUMBER DESCRIPTION DESIGNATION LOCATION TYPE TEST EXEMPTION SIR t

- P-38 CCW to RCP Thermal Barrier CC-C V-721 S Cll llQ YES-3 5.7 - 51 P-39 Purge Air Exhaust BV-1-1B S -

OVs N0 5.7 - 42 IIC-V-I l01 S --

OK NO VS-V-497 5 -

OK NO P-40 Purge Air Supply BV-1-1A S -

OK NO 5.7 - 41 P-41 Loop Drain ifcader Dif-TV-1841 i AGL OK NO 5.7 - 46 Dll-TV-1847 0 AGL OK NO Dil-RV-1847 S RV OK N0 N0-1

~

P-42 to 45 Main Steam Lines - - -

OK -

P-46 to 49 Feedwater Lines - - -

OK NO-1 -

P-50 Fuel Transfer Tube Blank Flange S GS OK NO 5.7 - 45 P-51 to 58 Service Water for Cooling - - - OK N0-1 -

Coils 1 P-59 Spare - -

SW OK NO-2 -

P-60 CCW to Neutron Shield Cooler CC-CV-885 S Cil LIQ YES-3 5.7 - 93 P-61 CCW from Neutron Shield CC-TV-lS31 'S AGA LIQ YES-3 5.7 - 21 Cooler d

i

!

• A key for the codes follows this table i

Page 7,of 8 CONNECTICUT YANKEE APPENDIX J TESTING STATUS TESTED LEAKAGE BARRIER

• PENETRATION PROCFDimE NUMBER DESCRIPTION DESIGNATION LOCATION TYPE TEST EXEMPTION Stm P-62 Service Air to Containment SA-V-413 S GA OK NO-1, 4 5.7 - 44 P-63 Neutron Shield Fill CC-V-884 S DI LIQ YES-3 5.7 - 55 IIP YES P-64 Air Monitor Sample from VS-TV-1848 I AGA OK N0 5.7 - 90 Cont.

VS-50V-12-1 0 SGA OK NO P-65 Air ifonitor Sample to VS-CV-Il04 I Cil OK NO 5.7 - 57 Containment VS-CV-1103 0 Cll OK NO P-66 CCW to Drain Cooler CC-CV-731 S Cil LIQ YES-3 5.7 - 57 P367 CCW from Drain Cooler CC-FCV-611 S AGA LIQ YES-3 5.7 - 57 P-68 Primary Water to Containment PW-CV-140 1 CII l.10 YES-3 5.7 - 78 PW-CV-139 0 Cll LIQ YES-3 P-69 Loop Fill Fil-CV-296 S Cil LIQ YES'3 5.7 - 54 IIP YES P-70 Instrument Air Supply Blank Flange S GS OK NO 5.7 - 47 4

• A key for the codes follows this table 1

Page8[of8 CONNECTICUT YANKEE APPENDIX J TESTING STATUS 1ESTED LEAKAGE BARRIER

• PENEIRAIION PROCF. DURE NUNHI R DESCRIPTION DESIGNATION LOCATION TYPE TEST EXEMPTION SUR l

P-71 Primary Vent lleader VII-V-522 I GA OK N0 5.7 - 74 Vil-V-525 0 GL OK NO C

P-72A Containment Pressure - - -

OK NO -

Instruments P-72B, C Spare - -

SW CK NO-2 -

P-73 Containment Sump to RilR CC-CV-802 S CV NT YES -

RII-110V-22 S f1GA OK NO Ril-V-808A S GA OK NO P-74 to 77 RCP Seal Water Supply Cil-CV-305 A to D S Cll LIQ YES-3 SPL-10.7-228 l P-78 Pressurizer Relief Tank Drain Dil-TV-554 I AGL REV YES-3 5.7 - 5'9 Df-TV-1844 0 AGL Or, NO P-79 Spare - -

SW OK NO-2 P-80 Aux. Spray from Fire System Ril-MOV-31 S MGA LIQ YES-3 5.7 - 53 ItP YES REV YES P-81 AuxiIIary fee (Iwater Supply IW-CV-192, 194, S Cil LIQ YES-3 5.7 - 56 196, 198 IIP YES 8

• A key for tiie cott'es follows tiils table

. s .

KEY BARRIER LOCATION I - Inner Barrier 0 - Outer Barrier .S - Single Barrier NOTE: The barrier location code does not specify location inside or out-side containment ; it specifies the relative position of engineered barriers to leakage.

BARRIER TYPE Prefix:

A - Air Operated; M - Motor Operated; S - Solenoid Operated; No Prefix .- Manual Ba rri er :

BU - Butterfly Valve ' GL - Globe Val ve GS - Gasket CH - Check Valve NE - Needle Valve SW - Seal Weld DI - Diaphragm Valve PL - Plug or Ball Valve GA - Gat e Val ve RY - Relief Valve Local Leak Rate Test:

OK -

Tested in full compliance with Appendix J NT -

No local leak rate test applied LIQ -

Tested with liquid REV -

Tested in a non-conservative direction HP -

Tested at system pressure greater than P A

Exemption NO -

No exemption required, complies with Appendix J YES - Exemption or modification required, not yet completed EX -

Exempted by NRC Where applicable, the exemption code will be followed by one of the reference numbers listed below:

1. D. M. Crutchfield letter to W. G. Counsil (Amendment 49) dated May 7,1982

2. A. Schwencer letter to D. C. Switzer dated March 11, 1977  !

3. W. G. Counsil letter to D. M. Crutchfield dated August 16, 1983  !

4. D. C. Switzer letter to A. Schwencer dated September 19, 1977