ML20236T361

Forwards Tank Waste Remediation Systems Section'S Comments on Bnfl Revised Safety Requirements Document & Integrated Safety Mgt Plan.Comments Range from Difference in Regulatory Requirements to Editorial in Nature

ML20236T361
Person / Time
Site:	07003091
Issue date:	07/23/1998
From:	Pierson R NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:	Gibbs C ENERGY, DEPT. OF
References
NUDOCS 9807280132
Download: ML20236T361 (8)

Text

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l July 23, 1998 Dr. Clark Gibbs, Regulatory Official Regulatory Unit Richland Operations Office U.S Department of Energy L

P.O. Box 550 <

Richland, WA 99353 l I l

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SUBJECT:

NRC TWRS SECTION COMMENTS ON BNFL'S REVISED SRD AND ISMP

References:

1. Safety Requirements Document - Volume I & 11, BNFL-5193-SRD-01,

! Rev.1.

'2. Integrated Safety Management Plan, BNFL-5193-ISP-01, Rev. 2.

Dear Dr. Gibbs:

This letter forwards comments on the referenced documents prepared by the U.S. Nuclear

. Regulatory Commission's Tank Waste Remediation Systems (TWRS) section. In preparing these comments, we focused our efforts on whether our original comment set was appropriately addressed in the revised submittals. To facilitate your usage of our comments we have separated them by functional categories. The comments range from differences in regulatory requirements to editorial in nature.

These comments reflect the initial thoughts of the staff and do not necessarily represent the views of the NRC. The comments have not undergone the NRC review and concurrence process that would be required for an official NRC position. The comments are intended to provide the Regulatory Unit a cursory view as to whether our previously identified comments had been incorporated into the revised standards approval package documents. We would welcome the opportunity to discuss these further with you, your staff, and your consultants.

if you have any questions, please call me at (301) 415-7192, or Mr. Stephen Koenick of my l staff on (301) 415-5228.

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Sincerely, Od#ftd ##% ,

Robert C. Pierson, Chief g

Special Projects Branch Division of Fuel Cyr'e Safety

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and Safeguards, NMSS I

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Tank Waste Remediation System Section i Special Projects Branch Division of Fuel Cycle Safety and Safeguards Office of Nuclear Material Safety and Safeguards j

! US Nuclear Regulatory Commission Comments on TWRS-P Project, Safety Requirements Document - Volume 1 & 61, BNFL-5193-SRD-01, Rev.1 Integrated Safety Management Plan, BNFL-5193-ISP-01, Rev. 2  ;

ENVIRONMENTAL PROTECTION (EP) l EP-1 Environmental Radiation Protection Proaram (ERPP)

In SRD Volume 11, BNFL has committed to addressing the elements listed in safety  ;

criteria, (SC) 5.3-2 in the ERPP. In some cases, BNFL adds explicit requirements for l ERPP elements through additional SC. For example, the ERPP element effluent I monitoring is covered under SC 5.4. However, other ERPP elements are not further  !

addressed, e.g., environmental surveillance. Without supporting SC or other more

detailed descriptions of the ERPP elements, there is not sufficient information at this  ;

time to determine if all elements of the ERPP will satisfy the intended scope upon further  !

development. l l

EP-2 Organizational Responsibility for the ERPP Unlike the Radiation Protection Program, organizational responsibility for the ERPP is not defined in either Section 11.0, " Organization Roles, Responsibilities, and i Authorities," of the ISMP, or in the SRD. Because the ERPP will be under development during the next phase of the project, the roles, responsibilities, and authorities for the ERPP should be clarified. 1 EP-3 C.onsistency in the Titles of Environmental Organizations BNFL has redesignated the Environmental Safety and Health (ES&H) organization as the Licensing Permitting and Safety (LPS) Organization in the ISMP. This change was not reflected in the SRD, Volume ll (see, for example, SC 7 7-8). BNFL should keep organizational designations consistent across all documentation. l EP-4 Environmental Consequence Critmia in SC 1.0-8 of SRD Volume 11, BNFL designated Safety Design Class structures, systems, and components (SSCs) as those that, by performing their specified safety function, prevent workers or the maximally exposed member of the public from receiving

}. a radiological exposure that exceeds the exposure standards defined in the SRD. BNFL defines exposure standards in SC 2.0-1 for the werker, co-located worker, and public.

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In the NRC's comments on BNFL's SRD (Revision 0), the NRC noted that 10 CFR i I .20.1406 requires applicants for licenses other than renewals, after August 20,1997, to l

describe in the application how facility design and procedures for operation will '

- minimize, to the extent practicable, contamination of the facility and the environment. l l Furthermore, revised Part 70, currently under development, is anticipated to contain l environmental consequence criteria for accidents. If transition to NRC regulation was to j l occur, BNFL would need to discuss how it would apply a graded approach to accidents l which lead to an environmental consequence.

l j- NUCLEAR CRITICALITY SAFETY (NCS)

NCS-1 Acolication of Double Contingency '

i in the third paragraph of SC 3.3-3, of the SRD Volume 11, BNFL describes an exception i to the application of double contingency that applies to operations with shielding and confinement as supported by paragraph 5.1 of ANSI /ANS-8.19-1983, R88. NRC has l not adopted this ANSI standard and will, upon any transition to NRC regulatory i authority, expect that double _ contingency will be provided if it has been determined that double contingency is not practicable to implement, the facility will be required to l implement sufficient redundancy and diversity in control parameters for these processes such that at least two unlikely, independent, and concurrent errors, l accidents, or equipment malfunctions, are necessary before a criticality accident is possible.

NCS-2 Criticality Alarm Svstems in SRD Volume 11 SC 3.3-8 states that " coverage of all areas requiring detection may be provided by a single detector." 10 CFR 70.24(a)(1) states that " Coverage of all areas shall be provided by two detectors." NRC would expect to implement this regulation upon any transition to NRC regulatory authority.

I NCS-3 Nuclear Characteristics of Material (Editorial) The fourth sentence of SRD Volume ll SC 3.3-4 states, " Full advantage may be taken of any nuclear characteristics of process materials and equipment." A sentence should be added afterwards to state something equivalent to "Any advantages l provided by nuclear characteristics of process materials or equipment should be verified on a periodic basis to ensure that the characteristics continue to exist." This periodic verification is important because the waste characteristics could unexpectedly change l depending upon the addition of other materials or conditions.

l NCS-4 Moderation l

(Editorial') The fifth sentence of SRD Volume ll SC 3.3-4 should be modified to replace the sentence beginning with "The geometry must be considered..." with " Assumptions for fissile material concentration and distribution, moderating material, reflection, and 2 July 21,1998 l

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uncertainty, will be based on credible conditions that result in maximum nuclear

! , , reactivity." This change better defines the type of analysis expected because the

! current statement is ambiguous in the actual level of moderation to be analyzed, l

PLANT SYSTEMS (PS)

PS-1 SC 4.4-19 Safety Design Class Coolino Water Systems

a. There is still no specification in the SRD Volume !!, SC 4.4-9, of a standard for determining the design basis meteorological conditions for Safety Design Class cooling water systems. {Previously identified as PS-07 in the letter to RU dated October 15,199'7}.

b. Heat exchanger design standards such as TEMA standards are still not listed under implementing standards in the SRD Volume 11, SC 4.4-9 for Safety Design Class cooling water systems. This appears to be an oversight since they are listed for Safety Design Significant (SC 4.4-20). {Previously identified as PS-09 in the letter to RU dated October 15,1997}

PS-2 SC 4.0-2 " Replacement in Kind" In SRD Volume 11, page 4-1, SC 4.0-2 uses the term " replacement in kind." BNFL should define this *u m. Nor is " replacement in kind" discussed or defined in the ISMP Chapter 5.3, Confis cation Management.

{Previously identified as HA-53 in the letter to RU dated October 15,1997 - was to be submitted to BNFL as an editorial comment.)

PS-3 Design Classifications in SRD Volume 11, page 4-13 and page 4-20, SC 4.3-1 and SC 4.4-8 contain criteria for

" engineered safety systems" and " ventilation and offgas systems," respectively.

However, there is no reference to the design classification of the systems. In SRD Volume I, Table 3-1 (Sheet 3), page 3-10, there is a reference that these systems are designated as "important-to-Safety." BNFL should modify Volume ll to be consistent with the table in Volume I.

PS-4 It would be helpfulif the NRC staff could get copies of the following procedures that are referenced in SRD Volume 11, SC 4.2-3, under implementing Codes and Standards.

l a. Document P001/2, " Rules for the Design of Piping Systems," BNFL inc., Richland, Washington,

b. Document V001/2, " Rules for the Design of Vessels," BNFL Inc., Richland, Washington.

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! QUALITY ASSURANCE (QA) l QA-1 'It is no't clear why SRD Volume I, Table 3-2, " Selected Codes and Standards as Referenced by NRC and DOE," does not have a row listing ASME NQA-1, " Quality l l Assurance Requirements for Nuclear Facility Applications," in Column 1; SRP 17.3, {

" Quality Assurance Program Description,"in Column 2; and G-830.120, " Implementation Guide for use with 10 CFR Part 830.120, in Column 3. I 1

QA-2 (Editorial) SC 4.4.3 on page 4-18 of SRD Volume 11 and page 13 of Attachment 4, ,

" Revisions to SRD Safety Criterion Text," of the SRD cover letter both state: "The l

environmental specification for Safety Design Significant equipment must include and be l based on the following as appropriate:* ." The meaning of the asterisk (apparently comparable to the asterisked footnote for Safety Design Class equipment in SC 4.4.2) should be clarified. i QA-3 (Editorial) The Table of Contents of SRD Volume ll does not list the correct page numbers.

QA-4 The first paragraph in Chapter 8.0 of the ISMP refers to BNFL 1997a, Tank Waste System Privatization Project Quality Assurance Program. This document has been superseded by BNFL 1998c, TWRS-P Privatization Project: Quality Assurance Program and / implementation Plan, and the reference should be changed accordingly. ,

QA-5 In Chapter 9.0 and Section 11.1 of the ISMP, BNFL uses an undefined term, " safety-related" activities. BNFL should either define or revise to "important-to-safety" activities i (where "important-to-safety" is defined in DOE /RL-96-0003, DOE /RL-96-0006, and ISMP Chapter 12.0). Also, just as BNFL's graded quality assurance program for W/RS encompasses the impoitant-to-safety structures, systems, and compenents, the quality assurance program should also encompass the important-to-safety activities (using the same general definition of important-to-safety). The ISMP should be revised accordingly,

{ Note that the NRC has consistently recommended elimination of the term " safety-related" since the " Preliminary Comments on Quality Assurance" were transmitted to the Regulatory Unit by letter dated November 20,1996.}

QA-6 The second paragraph in Section 9.1 of the ISMP refers to " key safety-related activities."

If left in the ISMP, this term also needs to be defined.

QA-7 ISMP Section 10.4," Support of the Regulatory Unit's inspection and Corrective Action /

Enforcement Action Programs,"is new in Revision 3. The Regulatory Unit should consider whether BNFL would be expected to include suitable physical facilities to support the subject activities as well as whether the Regulatory Unit may decide to have one or more of its staff located on the BNFL plant site with suitable physical facilities furnished by BNFL. Such facilities do not appear to be addressed in the ISMP.

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a I QA-8 The meaning of the broken (dashed; dotted) lines on ISMP Figures 11-1 and 11-2 (and

,OAP Figure 1-3) should be defined.

! QA-9 The responsibility for quality control during the design and construction phase and during the operations phase should be shown in the ISMP.

OA-10 The text of ISMP Sections 11.1,

• Design and Construction Phase," and 11.2,

" Operations Phase," and the corresponding figures should be clarified in the following

areas

a. Section 11.1 refers to the " manager of the Quality Assurance Organization" while Section 11.2 refers to the " Quality Assurance Manager." Both sections refer to the "TWRS-P Project QA Manager." However, both figures show a "BNFL inc.

Corporate QA Manager" box and a " Quality Assurance" box. It is not clear how these position titles and boxes correlate.

b. The text in Section 11.1 does not assign any safety role or responsibility to five boxes under the Project Manager. These boxes are 1) Operations Support,2)

Procurement Manager, 3) Radiological Nuclear and Process Safety,4)

Environmental Protection, and 5) Area Project Manager - Balance of Plant -

Retreatment - LAW - HLW. Does this indicate that these " boxes" have no safety role or responsibility?

c. Similarly, the text in Section 11.2 does not assign any safety role or responsibility to seven boxes under the BNFL Inc. VP and General Manager. These boxes are 1)

External Corporate Affairs,2) Accounting,3) Production Control,4) Procurement,

5) Human Relations,6) Document Control, and 7) Industrial Safety. Does this indicate that these " boxes" have no safety role or responsibility?

d. Section 11.2 refers to the " Operations and Technical Support Organization," but the figure shows an " Operations Support" box. The text and the table should agree.

QA 11 The third activity of the architect engineer that is overseen by the Architect Engineering Organization is shown in ISMP Section 11.1 as " Preparing specifications for procurement of repurchased equipment." The meaning of" repurchased" should be clarified.

RADIATION PROTECTION (RP)

RP-00 (note on reviewmethod): BNFL's approach to radiation protection appears to be to adopt Part 835 in toto through SC 1.0-10, and include several Part 20 requirements that do not appear in Part 835 as additional safety criteria. Although the Part 20 provisions that have a corresponding Part 835 provision might also be met through compliance with SC $1.0-10, NRC l staff focused on the information presented but did not review Part 835 against Part 20, to determine if there are additional Part 20 excerpts that BNFL should have included as SC .

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RP-01 SC 5.1-2. Resoiratorv Protection

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'In SRD Volume 11, SC 5.1-2 basically transposes requirements of 10 CFR @20.1703, j with the exception of 10 CFR @20.1703(b), which is not addressed, and deals with the i acceptable uses of respirator protection factors and Part 20 Appendix A. BNFL should clarify why it has omitted the requirements for respirator protection factors from the SRD, but included the rest of 10 CFR @20.1703.

RP-02 S_C 5.1-2. Resoiratorv Protection imolementina Standards BNFL should clarify why SC 5.1-2 in SRD Volume 11 does not list ANSI Z88.2 and ANSI Z88.6 as implementing codes and standards, and why SRD Vol. I, Tables 3-2 and 3-3, does not note that these codes were considered. These codes are referenced as regulatory review criteria in NRC's TWRS SRP, NUREG-0800, and in DOE's Radeon (DOE /EH-0256T) manual. Use of ANSI Z88.2 is required by 29 CFR 1910.134. It is also not clear why 29 CFR 1910.134 is not listed in SRD Vol.1, App. E, as an applicable law and contract requirement.

RP-03 SC 5.1-3. Sealed Source Leak Testina in SRD Volume 11, SC 5.1-3 is similar to NRC's Branch Technical Positions, " License Condition for Leak Tes;ing Sealed Byproduct Material Sources," (April 1993); and/or

" License Condition for Leak Testing Sealed Source Which Contains Alpha and/or Beta-Gamma Emitters," (April 1993).

a. The NRC technical positions note a minimum leak testing sensitivity of 0.0005 mmmeuries, but SC 5.1-3 references 0.0005 mci (i.e., mWicuries) as the detection capability. This may be a typographical error that should be corrected.

b. BNFL should clarify / define the meaning of " accountable" as used in the term

" accountable sealed radioactive sources."

c. RU may wish to have BNFL include a requirement for notification to the regulator upon discovery of a leaking source. (The NRC license conditions would require a 5-day notification).

RP-04 SC 5.1-7. Packaae Receiot Surveys and Procedures SC 5.1-7 In SRD Volume 11 appears to be based on 10 CFR 20.1906. In SC 5.1-7, penultimate paragraph, notification of the " Administrator of the NRC Regional office" upon receipt of packages that exceed the radiation / contamination limits, is inappropriate for a non-NRC licensed facility. Please change the reference to DOE's " Regulatory Official."

6 July 21,1998

o RP-05 Section 5 2. Radiological Safety Desian -

'We no"te that SCs 5.2-1,5.2-3, and 5.2-4 in SRD Volume ll, appear to be redundant to f the requirements in 10 CFR 835.1001 and 835.1002, which are already requirements l l incorporated by reference through SRD SC 1.0-10. However, the wording in SCs 5.2-1, 5.2-3, and 5.2-4 has been slightly modified from the wording in Part 835, and cer1ain l provisions (e.g., @835.1002(d)) were not included. NRC staff suggests that the l redundancies be removed or differences in wording be eliminated as they may lead to future confusion or even contradictory requirements.

j GENERAL (GN) l l GN-1 ls the Basis of Design Document referenced on page 3-3 of the SRD Volume I, submitted to the RU, or is it an internal BNFL document? From the description provided in Section 3.2, " Hazards Assessment," the Hazards Analysis Report (part of the SAP) is not going to be revised but subsumed into this document, while a set of potential accidents becomes incorporated into the various SARs. This situation appears to allow for an opportunity of the evolving design to not be incorporated into a document that will be reviewed by the RU or the NRC as part of the construction authorization package.

Clarification is requested to ensure that this does not happen.

GN-2 On page 3-29 of the SRD Volume I, Section 3.6, " Maintenance of the SRD/ there is a brief discussion added with regard to proposed changes to the SRD, including regulatory approval prior to implementing changes that could be considered as decreasing the level of safety; however, it's not until after issuance of the construction approval that the SRD will be coatrolled through the configuration management process.

BNFL should describe some type of configuration management that will occur from now until that point.

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