B17391, Forwards Response to NRC GL 98-04, Potential for Degradation of ECCS & CSS After LOCA Because of Construction & Protective Coating Deficiencies & Foreign Matl in Containment

Forwards Response to NRC GL 98-04, Potential for Degradation of ECCS & CSS After LOCA Because of Construction & Protective Coating Deficiencies & Foreign Matl in Containment

ML20195G253
Person / Time
Site:	Millstone
Issue date:	11/12/1998
From:	Bowling M NORTHEAST NUCLEAR ENERGY CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
B17391, GL-98-04, GL-98-4, NUDOCS 9811200239
Download: ML20195G253 (11)

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, Northeast "C"ry adda ute 1s6), wantum 063n p Nucl::arEnergy um. tone Nuclear Power Station Northeast Nuclear Energy Company P.O. Box 128

. Taterford, CT 06385-0128 (860) 447-1791 Fax (860) 444-4277

%e Northeast Utilities System IOf I 2 M Docket No. 50-423 B17391 Re: 10CFR50.54(f)

GL 98-04 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Millstone Nuclear Power Station, Unit No. 3 Response to Generic Letter 98-04 Pr.>tential For Dearadation Of The Emeraency Core Coolina System And The Containment Sorav System After A Loss-Of-Coolant Accident Because Of Construction And Protective Coatina Deficiencies And Foreian MaterialIn Containment This letter provides our response to the Generic Letter (GL) 98-04' regarding potential 4

. for degradation of the Emergency Core Coo!!ng System (ECCS) and the Containment Spray System (CSS) after a design basis Loss-Of-Coolant Accident (DBLOCA)  !

because of construction and protective coating deficiencies and foreign materialin containment.

Generic Letter (GL) 98-04 was issued to; (1) alert addressees that foreign material continues to be found inside operating nuclear power plant containments. In addition, l construction deficiencies and problems with the material condition of ECCS systems, structures, and components (SSCs) inside the containment continue to be found.

Design deficiencies also have been found which could degrade the ECCS or safety-

,,a related CSS; (2) alert the addressees to the problems associated with the material hWh condition of Service Level 1 protective coatings inside the containment and to request information under 10CFR50.54(f) to evaluate the addressees' programs for ensuring

.gy ' that Service Level 1 protective coatings inside containment do not detach from their ,

g substrate during a DBLOCA and interfere with the operation of the ECCS and the //,

hg safety-related CSS.

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h Northeast Nuclear Energy Company (NNECO) has reviewed the current licensing and 1

. Sat design-basis as it pertains to programs which are used to ensure that Service Level 1 protective coatings inside containment do not detach from their substrate during a o GL 98-04, " Potential For Degradation Of The Emergency Core Cooling System And The m22m t2* Containment Spray System After A Loss-Of-Coolant Accident Because Of Construction And

{ { { ] g g Protective Coating Deficiencies And Foreign Materialin Containment," dated July 14,1998.

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U.S. Nuclerr Regul: tory Commission B17391 \ Page 2 DBLOCA and interfere with the operation of the ECCS and the safety-related CSS.

Information requested under 10CFR50.54(f) for evaluation of these programs is provided in Attachment 1.

There are no commitments contained within this letter. Should you have any questions regarding this submittal, please contact Mr. David W. Dodson at (860) 437-2346.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY Martin L. Bowling, Jr. /

Recovery Officer - Technical Services Sworn to and subscribed before me this da o 1998

& mM'-

Qotary Fublic MY COMMISSION EXPIRES My Commission expires JUNE 30,2002

References:

1. EPRI TR-109937: Guidelines on the Elements of a Nuclear Safety-Related Coatings Program, dated April 1998 Attachment (1):

1. Response To Request For Information Under 10CFR50.54(F) cc: H. J. Miller, Region i Administrator A. C. Cerne, Senior Resident inspector, Millstone Unit No. 3 J. W. Andersen, NRC Project Manager, Millstone Unit No. 3 RWF:rf I

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Docket No. 50-423 B17391 Attachment 1 Millstone Nuclear Power Station, Unit No. 3

- Response To Re_ quest For Information Under 10CFR50.54(F) l:

November 1998 l-I t-a- eo

4 U.S. Nuclsar Regulatory Commission l l

B17391 \ Attachment i \ Page 1 Attachment 1 Response To Reauest For Information Under 10CFR50.54fF)  !

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Introduction:

This report piovides information applicable to the Millstone Nuclear Power Station, Unit No. 3, (MP3), that v!as requested by the Generic Letter (GL) 98-04, " Potential For Degradation Of The Emerger cy Core Cooling System And The Containment Spray System After A Loss-Of-Coolant Accident Because Of Construction And Protective Coating Deficiencies And Foreign Materialin Containment." The request for information made by the GL 98-04 is stated below.

Question (1)

(1) A summary descrip a of the plant-specific program or programs irnplemented to ensure that Service Level 1 protective coatings used inside ,

the containment are procured, applied, and maintained in compliance with I applicable regulatory requirernents and the plant-specific licensing basis for the facility. Include a discussion of how the plant-specific program 1 meets the applicable criteria of 10CFRPart 50, Appendix B, as well as information regarding any applicable standards, plant-specific procedures, or other guidance used for: (a) controlling the procureme.nt of coatings i and paints used at the facility,(b) the qualification testing of protective coatings, and (c) surface preparation, application, surveillence, and j maintenance activities for protective coatings. Maintenance activities )

involve reworking degraded coatings, removing degraded coatings to sound coatings, correctly preparing the surfaces, applying new coatings, I and verifying the quality of the coatings. I RESPONSE to (1):

Northeast Nuclear Energy Company (NNECO) has implemented controls for the  ;

. procurement, application, and maintenance of Service Level 1 protective coatings used inside the containment in a manner that is consister t with the i licensing basis and regulatory requirements applicable to Millstone Nuclear Power Station, Unit No. 3 (MP3). The requirements of 10CFRPart 50 Appendix B are implemented through specification of appropriate technical and quality requirements for the Service Level 1 coatings program which includes ongoing maintenance activities.

For Millstone Unit No. 3, Service Level 1 coatings are subject to the requirements of ANSI N101.2 and ANSI N101.4. Service Level 1 applies to

U.S. Nuclear R gulatory Commission B17391 \ Attachment 1 \ Page 2 I

' coatings used in the primary containment which are procured, applied and maintained by NNECO or their contractor. Tests on coatings are performed in

,_ accordance with Section 4 of ANSI N101.2 to meet or exceed plant Design Basis l Accident (DBA) conditions. Quality assurance program recommendations stated i I in Regulatory Guide 1.54 are followed for all major equipment and structures,  ;

l except for the inspection defined in Section 6.2.4 of ANSI N101.4-1972. l Inspection is in accordance with ANSI N5.12-1974, Section 10, " Inspection for I Shop and Field Work". Compliance is not invoked for equipment of a

, misceilaneous nature and allinsulated surfaces. Due to the impracticability of

imposing Regulatory Guide requirements, it is not invoked for standard shop l

processes used in painting valve bodies, handwheels, electrical cabinetry and L control panels, loudspeakers, emergency light cases and other miscellaneous equipment since the total surface area for such items is relatively small when compared to the total surface area for which the requirements are imposed. For this equipment, the specifications require that high quality coatings be applied i

using good commercial practice. Quantities of unqualified coatings are j discussed below in paragraph (b).

l Adequate assurance tha -he applicable requirements for the procurement, application, inspection, ano 'aintenance are implemented is provided by l procedures and programmat controls, approved under the NNECO Quality l Assurance program. A Service Level 1 specification and procedures for MP3

provide technical, quality and procedural requirements for the application of protective coatings inside containment. NNECO is evaluating the guidance provided in EPRI TR-109937 " Guideline on Nuclear Safety-Related Coatings"

!. and, as appropriate, improvements to our existing programs and procedures for  ;

Service Level 1 coatings will be made.

(a) Procurement of Service Level 1 coatings used for new applications or repair / replacement activities are procured from vendor (s) with a quality  :

assurance program meeting the applicable requirements of 10CFRPart 50 Appendix B. The applicable technical and quality requirements that the i vendor is required to meet are specified by NNECO in procurement .

documents. Acceptance activities are conducted in accordance with I L procedures that are consistent with ANSI N45.2 requirements. This specification of required technical and quality requirements combined with appropriate acceptance activities provides assurance that the coatings received meet the requirements of the procurement documents.

(b) The qualification testing of Service Level 1 coatings used for new applications or repair / replacement activities inside containment meets the applicable requirements contained in the standards and regulatory commitments t referenced above. These coatings have been evaluated to meet the t

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U.S. Nucl ar Regulatory Commission B17391 \ Attachment 1 \ Page 3 applicable standards and regulatory requirements previously referenced. The approximate amount of unqualified coatings inside containment (7000 square feet) and basis for its acceptability has been documented in previous correspondence with the NRC (NRC Question 281.3). Items with unqualified j coatings include control valves, hangers and snubbers, miscellaneous  ;

eqt'.pment items and portions of the polar crane. I j (c) The surface preparation and application of Service Level 1 coatings for both new applications or repair / replacement activities inside containment meet the applicable portions of the standards and regulatory commitments referenced l above. The technical and quality requirements for application of protective I coatings to exposed surfaces of steel, concrete, and equipment within l Primary Containment areas at MP3 are established by an approved Service l Level 1 coatings specihcation and procedures. Documentation is consistent '

with the applicable requirements.

NNECO does not have a formal program for conducting a condition assessment L of Service Level 1 coatings inade containment for MP3. Repair of coatings in l accordance with approved proce dures are performed in conjunction with miscellaneous inspectior), maintenance and modification activities. As a result of l miscellaneous activities inside containment, any other localized areas of degraded coatings can be identified. Those coated areas would then be l evaluated and scheduled for repair or replacement, as necessary. NNECO is evaluating the guidance provided in EPRI TR-109937 " Guideline on Nuclear Safety-Related Coatings" for performing condition assessments of containment ,

coatings.

l Question (2)(1)

(2) Information demonstrating compliance with item (i) or item (ii):

(i) For plants with licensing-basis requirements for tracking the amount i of unqualified coatings inside the containment and for assessing the impact of potential coating debris on the operation of safety-related SSCs during a postulated design basis LOCA, the following information shall be provided to demonstrate compliance:

! (a) The date and findings of the last assessment of coatings, and the planned date of the next assessment of coatings, i

(b) The limit for the amount of unqualified protective coatings allowed in the containment and how this limit is determined, f

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U.S. Nucl:ar Regulatory Commission B17391 \ Attachment 1 \ Page 4 1

Discuss any conservatism in the method used to determine this '

limit.

l (c) If a commercial-grade dedication program is being used at your facility for dedicating commercial grade coatings for Service  !

Level 1 applications inside the containment, discuss how the ,

program adequately qualifies such a coating for Service Level 1 l service. Identify which standards or other guidance are i currently being used to dedicate containment coatings at your j facility; or, RESPONSE to (2)(i):

Not Applicable Question (2)(li)

(2) Information demonstrating compliance with item (i) or item (ii):

(ii) For plants without the above licensing-basis requirements, information shall be provided to demonstrate compliance with the  !

requirements of 10CFR50.46b(5), "Long-term cooling" and the functional capability of the safety-related CSS as set forth in your licensing basis. If a licensee can samonstrate this compliance without quantifying the amount of unqualified coatings, this is acceptable.

RESPONSE to (2)(ii):

The following description and referenced materials describe the licensing basis for MP3 relative to conformance with 10CFR50.46(b)(5), "Long-term cooling," l specifically with regard to MP3's ability to provide extended decay heat removal  !

including related assumptions for debris that could block containment emergency sump screens:

Millstone Unit No. 3 FSAR, Section 6.2.2 " Containment Heat Removal System" describes the licensing basis for systems that provide for containment heat removal. These systems consist of the quench spray system (QSS) and the containment recirculation system (RSS). The Containment Heat Removal System is designed in accordance with design criteria documents provided in FSAR Section 6.2.2.1. Assumptions for insulation debris and paint debris are provided in FSAR Sections 6.2.2.2 " System Design" and 6.2.2.3 " Design i

U.S. Nuclear Regulatory Commission i B17391 \ Attachment 1 \ Page 5 Evaluation." The section on paint debris references the FSAR response to NRC Question Q281.3.

The Emergency Core Cooling System (ECCS) provides borated water to cool the I reactor core following a major Loss of Coolant Accident (LOCA) After the injection mode of emergency core cooling, long term cooling is maintained by recirculating the water from the containment structure sump by the containment recirculation pumps, through the containment recirculation coolers, and into the reactor coolant loops directly and via the charging and safety injection pumps.

The containment heat removal system consists of the quench spray and containment recirculation systems. Following the postulated DBA, the containment pressure is reduced by employing both systems. The quench spray system sprays a mixture of borated water from the refueling water storage tank (RWST). The recirculation spray system draws suction from the containment sump, the content of which consists of the primary or secondary system effluent and quench spray.

Conformance with Regulatory Guide 1.82 Millstone Unit No. 3 sump design has been reviewed against the guidance of Regulatory Guide 1.82, Revision 0, " Sumps for Emergency Core Cooling and Containment Spray Systems." in accordance with this guidance, MP3 has assumed the Emergency Core Cooling Systems (ECCS) and Containment Spmy Systems (CSS) systems which draw from this sump may experience j sump blockage of up to 50% of the effective sump area as a result of debris j generated during a LOCA with no loss of function. The analyses submitted )

as part of the licensing basis for MP3 demonstrated that even with this i blockage, the ECCS and CSS will continue to provide sufficient cooling flow to fulfill the long-term cooling functions required to conform with 10CFR50.46(b)(5).

Background

in the Millstone Unit No. 3 Safety Evaluation Report (SER), the NRC staff reviewed the information in the FSAR and identified sump design aspects which were not strictly in conformance with the recommendations of i Regulatory Guide (RG) 1.82. Specifically, MP3 calculated the sump water j approach velocity at the fine mesh screens at a value which exceeded the valued recommended by RG 1.82. This required MP3 to re-evaluate sump screen blockage using an acceptable methodology and consider the types and quantities of insulation that are to be installed to justify the assumption of 50% blockage. However, the staff concluded that, subject to the resolution of

U.S. Nucl:ar Regulatory Commission B17391 \ Attachment 1 \ Page 6 1

i the matter concerning sump screen water approach velocity, the containment  !

sump design satisfied the requirements of RG 1.82. The issue of sump water j approach velocity was resolved in SER Supplement No. 4 (SSER No. 4), as '

discussed below. It also concluded that the containment heat removal systems satisfy the requirements of 10CFR50, Appendix A criteria 38,39, and 40.

In SSER No. 4, the Staff reviewed and accepted additionalinformation confirming that the containment sump design meets the intent of RG 1.82.

- The staff concluded that there is adequate assurance that insulation debris generated by postulated pipe breaks will not interfere with containment sump operation at Millstone Unit No. 3. The matter was classified as Confirmatory item 27 and this item was satisfactorily resolved.

Millstone Unit No. 3 SSER No 4 iridicates that by letters dated April 16,1985, I and October 15,1985, NNECO provided an assessment of debris formation and associated sump flow head loss for several postulated primary and secondary pipe ruptures. The analysis considered quantity of debris generated to be instantaneously transported to the sump. The sump screen was assumed to be both fully submerged and in a subsequent analysis only partially submerged as is the case during the early part of the transient when the recirculation pumps are started. Based on the estimated insulation i thickness on the sump screen and screen approach velocity, NNECO calculated head loss using empirical relationships provided in NUREG-0897 and showed adequate Net Positive Suction Head (NPSH) margin. l The NRC reviewed the method and assumptions used by NNECO analyses to assess the effects of insulation debris. It was found to be consistent with the information and methodology developed as part of USI A-43,

" Containment Emergency Sump Performance," and described in NUREG/CR-2791, NUREG/CR-3170, and NUREG-0897, Revision 1.

1 In a response to FSAR Question 281.3 dated May 3,1983, NNECO j estimated the quantity of unqualified coating debris and provided qualitative justification that it will not have an adverse affect on the performance of Residual Heat Removal (RHR) or CSS.

The NRC accepted these analyses and these systems as meeting the requirements of 10CFR50.46(b)(5) in SSER No. 4. The NRC concluded that there is adequate assurance that insulation debris generated by postulated pipe breaks will not interfere with containment sump operation and that the containment sump design at MP3 meets the intent of RG 1.82.

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U.S. Nuclear Regulatory Commission B17391 \ Attachment 1 \ Page 7 Sumniary l

The licensing basis for MP3, as accepted within the SER, provides both the j regulatory and safety basis for safety system performance. Coatings are not treated separately in the licensing basis for MP3 because the sump screen blockage assumption does not distinguish among the source terms for the ,

LOCA-generated debris. Therefore, a separate demonstration of the regulatory and safety basis for safety system performance is not required.

References:

1. FSAR Section 6.2.2 Containment Heat Removal Gystem.

2. RAI Docket 50-423, Question No. Q281.3, Q480.3.

3. NUREG - 1031 Safety Evaluation Report related to the operation of Millstone Nuclear Power Station , Unit No. 3, Docket No. 50-423.

4. NUREG - 1031 Supplement No. 4.

5. NNECO letter dated October 15,1985 Docket no. 50-423, B11771 - SER Confirmatory item #27.

6. NNECO letter dated October 25,1985 Docket No. 50-423, B11835 -

Revised Response to Confirmatory item #27.

7. NNECO letter dated January 21,1998 Docket No. 50-423, B16948 -

Assurance of Sufficient NPSH for Emergency Core Cooling and Containment Heat Removal Pumps.

Question (2)(li)(a)

(2) Information demonstrating compliance with item (i) or item (ii):

(ii) For plants without the above licensing-basis requirements, information shall be provided to demonstrate compliance with the requirements of 10CFR50.46b(5), "Long-term cooling" and the functional capability of the safety-related CSS as set forth in your licensing basis. If a licensee can demonstrate this compliance without quantifying the amount of unqualified coatings, this is acceptable. The following information shall be provided:

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U.S. Nucber R:gulatory Commission ,

B17391 \ Anachment 1 \ Page 8 (a) If commercial-grade coatings are being used at your facility for Service Level 1 applications, and such coatings are not dedicated or controlled under your Appendix B Quality Assurance Program, provide the regulatory and safety basis for not controlling these coatings in accordance with such a program. Additionally, explain why the facility's licensing basis does not require such a program.

RESPONSE to (2)(ii)(al:

NNECO does not currently employ commercial grade dedication for Service Level 1 coatings used inside containment at Millstone Unit No. 3.

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