LIC-12-0027, License Amendment Request (LAR) 12-02, Proposed Change to Relocate Operating and Surveillance Requirements for Technical Specifications 217 and 3.13 Concerning Miscellaneous Radioactive Material Sources

License Amendment Request (LAR) 12-02, Proposed Change to Relocate Operating and Surveillance Requirements for Technical Specifications 217 and 3.13 Concerning Miscellaneous Radioactive Material Sources

ML120720151
Person / Time
Site:	Fort Calhoun
Issue date:	03/09/2012
From:	Bannister D Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR 12-02, LIC-12-0027
Download: ML120720151 (47)

Text

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Omaha Public Power District 444 South 16 Street Ma/i Omaha, NE 68102-2247 LIC-1 2-0027 March 9, 2012 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Reference:

Docket No. 50-285.

SUBJECT:

License Amendment Request (LAR) 12-02, Proposed Change to Relocate Operating and Surveillance Requirements for Technical Specifications 217 and 3.13 Concerning Miscellaneous Radioactive Material Sources Pursuant to 10 CFR 50.90, the Omaha Public Power District (OPPD) hereby requests an amendment to the Renewed Facility Operating License No. DPR-40 for Fort Calhoun Station (FCS), Unit No. 1. The proposed amendment would relocate Technical Specifications (TS)

Limiting Condition of Operation (LCO) 2.17, Miscellaneous Radioactive Material Sources, and the associated surveillance requirement (SR) 3.13, Radioactive Material Sources Surveillance, from the FCS TS. NUREG-1432, Standard Technical Specifications, Combustion Engineering Plants, Revision 3, does not contain a TS or SR for radioactive source surveillance. The operability and surveillance requirements for leak checking of miscellaneous radioactive material sources will be incorporated into the FCS Updated Safety Analysis Report (USAR) and associated plant procedures.

The proposed changes conform to NRC regulation 10 CFR 50.36 for the contents of the TS.

OPPD has determined that this LAR does not involve a significant hazard consideration as determined per 10 CFR 50.92. Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment.

The enclosure contains a description of the proposed changes, the supporting technical analyses, and the significant hazards consideration determination. Attachment 1 provides the existing TS pages marked-up to show the proposed changes. Attachment 2 provides the retyped (clean) TS pages.

Eniplovinei,t with Equal Oppoi-tunitv

U. S. Nuclear Regulatory Commission LIC-1 2-0027 Page 2 OPPD requests approval of the proposed amendment by March 1, 2013. Once approved, the amendment shall be implemented within 120 days.

There are no regulatory commitments associated with this proposed change.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State of Nebraska official.

If you should have any questions regarding this submittal or require additional information, please contact Mr. Bill R. Hansher at (402) 533-6894.

I declare under penalty of perjury that the foregoing is true and correct. Executed on March 9, 2012.

D. J. Bannister Site Vice President and CNO DJB/SDS/mle

Enclosure:

OPPDs Evaluation of the Proposed Change c: Director of Consumer Health Services, Department of Regulation and Licensure, Nebraska Health and Human Services, State of Nebraska

LIC-1 2-0027 Enclosure Page 1 OPPDs Evaluation of the Proposed Change License Amendment Request (LAR) 12-02, Proposed Change to Relocate Operating and Surveillance Requirements for Technical Specifications 2.17 and 3.13 Concerning Miscellaneous Radioactive Material Sources 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions 5.0 ENVIRONMENTAL-CONSIDERATION

6.0 REFERENCES

ATTACHMENTS:

1. Technical Specification Page Markups

2. Retyped (Clean) Technical Specifications Pages

3. Updated Safety Analysis Report Markup (For Information Only)

[IC-i 2-0027 Enclosure Page 2 1.0

SUMMARY

DESCRIPTION License amendment request (LAR) 12-02 proposes a change to the Renewed Facility Operating License No. DPR-40 for Fort Calhoun Station (FCS), Unit No. 1. The Omaha Public Power District (OPPD) proposes to relocate the Technical Specification (TS) limiting condition for operation (LCO) 2.17, Miscellaneous Radioactive Material Sources, and the associated surveillance requirement (SR) 3.13, Radioactive Material Sources Surveillance, in their entirety from the FCS TS to the FCS Updated Safety Analysis Report (USAR) and associated plant procedures. Specifically, the proposed change would revise the Renewed Operating License by deleting LCO 2.17 and SR 3.13 in their entirety. The table of contents (TOO) has been revised to reflect the deletion of LCO 2.17 and SR 3.13.

2.0 DETAILED DESCRIPTION The proposed TS changes for LAR 12-02 are as follows:

LCO 2.17 Deleted in its entirety. No clean pages are provided; this section will be removed from the FCS TS.

SR 3.13 Deleted in its entirety. No clean pages are provided; this section will be removed from the FCS TS.

Table of Contents (TOC)

Page 2 of the TOO is revised to remove the titles of [CO 2.17 and SR 3.13 and show both as DELETED. This is an administrative change to make the TOO reflect the TS changes above.

3.0 TECHNICAL EVALUATION

Radioactive material sealed sources are used at FCS for instrument response verification, calibration, startup sources, and fission detectors, and may contain byproduct, source, or special nuclear material. FOS is bound by the Renewed Operating License No. DPR-40 and associated license conditions for receipt, use, and possession of byproduct, source, and special nuclear material. The FCS Renewed Operating License specifically requires compliance with the Commissions regulations found in 10 CFR 30, Rules of General Applicability to Domestic Licensing of Byproduct Material; 10 OFR 40, Domestic Licensing of Source Material; and 10 CFR 70, Domestic Licensing of Special Nuclear Material.

The standard license language for receipt, use, and possession of byproduct, source, and special nuclear material was developed by the NRC in 1974 as part of a continuing effort to simplify the licensing process. Licensees were requested to submit license amendment requests to adopt the proposed generic wording for receipt, use, and possession of byproduct, source, and special nuclear material, as well as a proposed TS and bases for

LIC-1 2-0027 Enclosure Page 3 source leak testing and surveillance (Reference 6.1). On March 31, 1975, OPPD submitted the requested operating license amendment via the Law Offices of LeBoeuf, Lamb, Leiby & MacRae, Washington, DC (Reference 6.2). That submittal was the basis for Amendment 11 which was issued on February 4, 1976. The text of the TS and SR for radioactive material sources was changed once in the ensuing period. Amendment 122, issued on June 2, 1989, made a minor editorial change to SR 3.13 Item 1 to modify the phrase ...intervals not to exceed six months to ...intervals of six months. (Amendments 157 and 176 made administrative changes related to the numbering of SR 3.13 (Amendment 157) and the sequencing of pages following TS 2.17 and SR 3.13 (Amendment 176) but did not change the text of the TS or SR.)

This LAR will relocate TS 2.17 and SR 3.13 in their entirety to the FCS USAR. No reductions to these requirements are proposed. The only changes will be of an editorial nature to format the requirements to the standards of the FCS USAR. The license conditions for byproduct, source, and special nuclear material in the FCS Renewed Operating License will remain as is [paragraphs 1.1.; 2.B.(2); 2.B.(3); 2.B.(4); and 2.B.(5)].

10 CFR 50.36(c)(2)(ii) states that A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

(A) Criterion 1 - Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

(B) Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C) Criterion 3 A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(D) Criterion 4 A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

Reference 6.3, Final Commission Policy Statement on Technical Specifications for Nuclear Power Reactors, dated July 22, 1993, concluded that those existing TS requirements, which do not satisfy the screening criteria specified in regulation 10 CFR 50.36 above, may be deleted from the TS, and the requirements established in licensee-controlled documents, subject to the controls of 10 CFR 50.59.

Consequently, the miscellaneous radioactive material sources TS does not meet any of the screening criteria of 10 CFR 50.36(c)(2)(ii). This is supported by the absence of operability and surveillance requirements for the miscellaneous radioactive material sources in the Standard Technical Specifications for Combustion Engineering Plants presented in

LIC-1 2-0027 Enclosure Page 4 NUREG-1432, Revision 3 (Reference 6.4). Accordingly, the proposed changes are aligned with the NUREG-1432, Revision 3 and the miscellaneous radioactive material sources requirements can be established in licensee-controlled documents. Future changes to miscellaneous radioactive material sources requirements will be subject to the controls of 10 CFR 50.59.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory ReguirementslCriteria 4.1.1 Regulations Code of Federal Regulations Part 20:

Sealed source leakage limitations were established to ensure that leakage from sources made of byproduct, source, and special nuclear material will not lead to intakes of radioactive materials in excess of regulatory limits. While the terminology associated with internal and external dose has changed, the basis for the leakage limit remains the same. The limitation of removable contamination for sources requiring leak testing, including alpha emitters, is based on the 10 CFR 70.39(c) limits for plutonium. This standard has been carried into the leak check requirements for byproduct, source, and special nuclear material sources of various configurations regulated by NRC through the Office of Nuclear Material Safety and Safeguards.

The general survey requirements set forth in 10 CFR 20, Subpart F, § 20.1501, establish the regulatory requirements for radiological surveys. These general requirements have been translated into procedures to evaluate the potential radiological hazards for the workplace to ensure control of radioactive material to prevent unnecessary ingestion or exposure. The potential hazard from a radioactive source is leakage which could result in internal or external exposure.

Performance of surveys on a periodic basis will identify any radioactive sources that could pose an internal or external exposure hazard and promptly remove them from service.

4.1.2 Design Basis Not applicable; there are no USAR accident analyses which are impacted by miscellaneous radioactive material sources and the associated leak check surveillance.

4.1.3 Approved Methodologies

• Regulatory Guide 8.21 Health Physics Surveys for Byproduct Material at NRC-Licensed Processing and Manufacturing Plants, Revision 1, October 1979.

• NUREG-1556, Volume 11, Consolidated Guidance About Materials Licenses:

Program-Specific Guidance About Licenses of Broad Scope, Final Report, April 1999.

LIC-1 2-0027 Enclosure Page 5 4.1.4 Analysis No analyses were conducted in support of the proposed amendment. Leak checking of miscellaneous radioactive material sources is not credited in Probabilistic Risk Assessment (PRA) for operator actions to mitigate the consequences of an event.

4.2 Precedent None 4.3 Significant Hazards Consideration The proposed change would delete Technical Specification (TS) 2.17 and surveillance requirement (SR) 3.13 to allow the relocation of the operability and surveillance requirements for the miscellaneous radioactive sources from the TS to a licensee-controlled document, specifically the Fort Calhoun Station (FCS), Unit No. 1, Updated Safety Analysis Report (USAR). Future changes to miscellaneous radioactive sources requirements will be subject to the controls of 10 CFR 50.59. The proposed change aligns with NUREG-1432, Standard Technical Specifications, Combustion Engineering Plants, Revision 3 (Reference 6.4) as the requirements for leak checking radioactive sources were removed from TS and relocated into licensee controlled documents.

The Omaha Public Power District (OPPD) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

Miscellaneous radioactive sources are not part of any transient or accident analysis.

The proposed changes conform to the Nuclear Regulatory Commissions (NRCs) regulatory guidance regarding the content of plant TS as identified in 10 CFR 50.36 and NRC publication NUREG-1432.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

LIC-1 2-0027 Enclosure Page 6

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change relocates the requirements for leak checking miscellaneous radioactive material sources to a licensee controlled document subject to the controls of 10 CFR 5059. This change does not alter the physical design, safety limits, or safety analysis assumptions associated with the operation of the plant.

Hence, the proposed change does not introduce any new accident initiators, nor does it reduce or adversely affect the capabilities of any plant structure or system in the performance of their safety function.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change relocates the requirements for leak checking miscellaneous radioactive material sources to a licensee controlled document subject to the controls of 10 CFR 50.59. This change does not alter any safety margins.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, OPPD concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.

However, the proposed amendment does not involve (I) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or

LIC-1 2-0027 Enclosure Page 7 cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

6.1 Letter from NRC (G. Lear) to OPPD (J. L. Wilkins), Information Regarding Application for License Amendment Regarding Special Nuclear, Byproduct, and Source Material, dated December 16, 1974 (NRC-74-0089).

6.2 Letter from the Law Offices of LeBoeuf, Lamb, Leiby & MacRae on behalf of Omaha Public Power District, Fort Calhoun Station to NRC (E. G. Case), Application for Amendment of Operating License, dated March 31, 1975.

6.3 NRC Final Policy Statement on Technical Specifications Improvement for Nuclear Power Reactors (58 FR 39132, dated July 22, 1993).

6.4 NUREG-1432, Revision 3, Standard Technical Specifications, Combustion Engineering Plants, dated June 2004.

LIC-1 2-0027 Enclosure, Attachment 1 Page 1 Technical Specification Page Markups

/Word-processor mark-ups using red/The/strike out feature for new text/deleted text respectively.]

TECHNICAL SPECIFICATION TABLE OF CONTENTS (Continued) 2.13 Limiting Safety System Settings, Reactor Protective System 2.14 Engineered Safety Features System Initiation Instrumentation Settings 2.15 Instrumentation and Control Systems 216 River Level 2.17 Miscellaneous Radioactive Material Sources DELETED 2.18 DELETED 2.19 DELETED 2.20 Steam Generator Coolant Radioactivity 2.21 Post-Accident Monitoring Instrumentation 2.22 DELETED 2.23 Steam Generator (SG) Tube Integrity 3.0 SURVEILLANCE REQUIREMENTS 3.1 Instrumentation and Control 3.2 Equipment and Sampling Tests 3.3 Reactor Coolant System and Other Components Subject to ASME XI Boiler and Pressure Vessel Code Inspection and Testing Surveillance 3.4 DELETED 3.5 Containment Test 3.6 Safety Injection and Containment Cooling Systems Tests 3.7 Emergency Power System Periodic Tests 3.8 Main Steam Isolation Valves 3.9 Auxiliary Feedwater System 3.10 Reactor Core Parameters 3.11 DELETED 3.12 Radioactive Waste Disposal System 3.13 Radioactive Material Sources Surveillance DELETED 3.14 DELETED 3.15 DELETED 3.16 Residual Heat Removal System Integrity Testing 3.17 Steam Generator (SG) Tube Integrity 4.0 DESIGN FEATURES 4.1 Site 4.2 Reactor Core 4.3 Fuel Storage TOC - Page 2 Amendment No. 11,27,32,38,13,1851,60,8486-,

93,97,101122,136,152, 160176,183, 211,230, 236, 246,248, 252

TPCHNIiAI PFflIFIATIflNS 2.0 LIMITING CONDITIONS FOR OPERATION 2.17 MISCELLANEOUS RADIOACTIVE MATERIAL SOURCES Applicability Applios to byproduct, source, and special nuclear radloactivo matorial sourcos.

Objective To assure that leakage from byproduct, source, and special nuclear radioactive material sourcos does not oxcoed allowable limits.

Specifications Radioactive :-.iJul t;r-_-. shall be leak testod for contamination. The leakage test shall be capable of detecting the presenco of 0.005 microcurie of radioactive material on the test sample. If the test roveals the presence of 0.005 microcurie or more of removable contamination, it shall immediately be withdrawn from use, decontaminated, and repaired, or be disposed of in accordanco with Commission regulations. Those quantities of byproduct material that exceed the quantitios listed in 10 CFR Part 30, Section 30.71, Schedule B are to be leak tested in accordanco with the schedule shown in Surveillance Requirements. All other sources (including alpha emitters) containing greater than 0.1 microcurie are also to be leak tested in accordance with the Surveillance Requirements.

Basis Ingestion or inhalation of source material may give rise to total body or organ irradiation.

This specification assures that leakage from radioactive material sources does not exceed allowable limits. In the unlikely event that those quantities of radioactive byproduct materials of interest to this specification which are exempt from leakage tosting are ingested or inhaled, thoy represent less then one maximum permissible body burden for total body irradiation. The limits for all other sources (including alpha emitters) are basod upon 10 CFR Part 70, Section 70. 39(c) limits for plutonium.

217 Page 1 Amendment No. 11,176 Dated: February 4, 1976

IL LIC-1 2-0027 Enclosure, Attachment 2 Page 1 Retyped (Clean) Technical Specification Pages

TECHNICAL SPECIFICATION TABLE OF CONTENTS (Continued) 213 Limiting Safety System Settings, Reactor Protective System 2.14 Engineered Safety Features System Initiation Instrumentation Settings 2.15 Instrumentation and Control Systems 2.16 River Level 2.17 DELETED 2.18 DELETED 2.19 DELETED 2.20 Steam Generator Coolant Radioactivity 2.21 Post-Accident Monitoring Instrumentation 2.22 DELETED 2.23 Steam Generator (SG) Tube Integrity 3.0 SURVEILLANCE REQUIREMENTS 3.1 Instrumentation and Control 3.2 Equipment and Sampling Tests 3.3 Reactor Coolant System and Other Components Subject to ASME Xl Boiler and Pressure Vessel Code Inspection and Testing Surveillance 3.4 DELETED 3.5 Containment Test 3.6 Safety Injection and Containment Cooling Systems Tests 3.7 Emergency Power System Periodic Tests 3.8 Main Steam Isolation Valves 3.9 Auxiliary Feedwater System 3.10 Reactor Core Parameters 3.11 DELETED 3.12 Radioactive Waste Disposal System 3.13 DELETED 3.14 DELETED 3.15 DELETED 3.16 Residual Heat Removal System Integrity Testing 3.17 Steam Generator (SG) Tube Integrity 4.0 DESIGN FEATURES 4.1 Site 4.2 Reactor Core 4.3 Fuel Storage TOC - Page 2 Amendment No. 11 ,2732,38,1316,51,60,8I86, 93,97,101,122,136,152, 160,176,183, 211,230, 236, 246,28, 22

LIC-1 2-0027 Enclosure, Attachment 3 Page 1 Updated Safety Analysis Report Markup (For Information Only)

Page 1 of 31 1

USAR 11.2 Radioactive Waste and Radiation Protection and Monitoring Radiation Protection and Monitoring Rev 1-6 Safety Classification: Usage Level:

Safety Information I I I Change No.: EC Reason for Change: Add Section 11.2.5, Miscellaneous Radioactive Materials Sources. The requirements of TS 2.17 and SR 3.13 have been relocated to the USAR per License Amendment XXX.

Preparer:

Issued:

Fort Calhoun Station

USAR-1 1 .2 Information Use Page 2 of 31 Radiation Protection and Monitoring Rev. 1-Table of Contents 11 .2 Radiation Protection and Monitoring 5 11.2.1 General 5 11.2.2 Shielding 5 11.2.2.1 Purpose 5 11.2.2.2 PlantAccessibility 5 11.2.2.3 Design Dose Rates 5 11 .2.2.4 Shielding Arrangements 6 11.2.2.5 Shielding Materials 8 11 .2.2.6 Reactor Radiation Sources 8 11 .2.2.7 Pressurizer Radiation Sources 11 11 .2.2.8 Reactor Coolant System Radiation Sources 13 11 .2.2.9 Waste Treatment Radiation Sources 13 11.2.2.10 Refueling Radiation Sources 17 11.2.3 Radiation Monitoring 20 11.2.3.1 General 20 11 .2.3.2 Containment Building Radiation Monitors 22 11 .2.3.3 Auxiliary Building Exhaust Ventilation Stack Gaseous Effluent Monitors 23 11.2.3.4 Condenser Off-Gas Monitor 25 11 .2.3.5 Monitoring of Liquid or Gaseous Effluents from Steam Generator Blowdown or Steam Leakage in the Turbine Building 26 11.2.3.6 Plant Liquid Effluent Monitor 26 11 .2.3.7 Other Process Monitors 27 11 .2.3.8 Area Radiation Monitors 27 11.2.3.9 Personnel Contamination Monitors 29 11.2.3.10 PowerSupply 29 11 .2.3.11 Post-Accident Main Steam Line Monitor 29 11.2.3.12 Post-Accident Control Room Iodine Monitor (RM-065) 30 11.2.4 Radioactive Material Storage 30 11.2.4.1 Radioactive Waste Storage 30 11.2.4.2 Radioactive Waste Storage 30 11 .2.5 Miscellaneous Radioactive Material Sources 30

USAR-1 1 .2 Information Use Page 3 of 31 Radiation Protection and Monitoring Rev. 3-Listof Tables Table 11 .2-1 - Neutron Spectra Outside Vessel 9 Table 11 .2-2 - Intermediate and Thermal Neutron Fluxes Outside Vessel 9 Table 11 .2-3 - Gamma Spectra Outside Vessel 10 Table 11.2-4 - N-16 Activity Around the Reactor Coolant Loop 10 Table 11 .2-5 - Shutdown Gamma Spectra at Vessel OD 11 Table 11.2-6 - Maximum Deposited Activities in Pressurizer Due to Crud 12 Table 11 .2-7 - Maximum Activity in Pressurizer Steam Section 12 Table 11 .2-8 - Maximum Deposited Activities in Reactor Coolant System Due to Crud 13 Table 11.2-9 - Maximum Activity in CVCS Demineralizer 14 Table 11 .2-1 1 - Specific Activity of Major Isotopes Following Irradiation 18 Table 11.2-12 - Inconel 625 Composition 18 Table 11.2-13 - Total Gamma Attenuation Coefficients 19 Table 11.2-14 - Area Radiation Monitors 28 Table 11.2-15 Leak Check Surveillance Requirements 31

USAR-1 1 .2 Information Use Page 4 of 31 Radiation Protection and Monitoring Rev. .1-s List of Figures The following figures are controlled drawings and can be viewed and printed from the listed aperture card.

Figure No. Title Aperture Card 11 .2-2 Fuel Assembly Fission Product Decay Gammas 36562

USAR-1 1 .2 Information Use Page 5 of 31 Radiation Protection and Monitoring Rev. 4-6 11 .2 Radiation Protection and Monitoring 11.2.1 General Radiation protection measures comprise the shielding of the predictable sources of significant ionizing radiation and operational and administrative controls and procedures. The latter, technically supervised by the radiation protection staff, include controlled access to hazardous and potentially hazardous spaces, permanently installed radiation monitoring systems, and control of contamination. The radiation protection staff and functions are discussed in Section 12.

11 .2.2 Shielding 11.2.2.1 Purpose Radiation shielding is designed to provide radiation protection for operating personnel inside and outside the plant, and for the general public. For this purpose continuous operation with one percent failed fuel is assumed, with end of cycle fission product inventory.

Shielding provides adequate protection against direct radiation exposure during reactor operation at 1500 MWt, refueling, and the design basis accident.

11 .2.2.2 Plant Accessibility Current information concerning plant accessibility is contained within Radiation Protection Manual.

11.2.2.3 Design Dose Rates Design dose rates, for normal operation at the maximum power level of 1500 MWt and during refueling are kept below the limitations set forth in 10 CFR Part 20.

USAR-1 1 .2 Information Use Page 6 of 31 Radiation Protection and Monitoring Rev. 4-11 .2.2.4 Shielding Arrangements Containment Building The containment has two major shielding functions:

a. During normal operation it shields adjacent auxiliary plant and yard areas from radiation originating in the reactor, in the reactor coolant loop, and in other radioactive equipment and piping located inside the containment. It reduces radiation exposure below 0.5 mrem/hr in those areas outside the containment which are occupied by personnel on a routine basis.

b. Under the postulated LOCA conditions the containment would reduce plant, site, and off-site radiation exposure by direct and by air scattered radiation to levels which would allow egress of personnel from the entire plant consistent with the guidelines set forth in 10 CFR Part 100.

Primary Shield The primary shield surrounds and supports the reactor vessel. It is designed to reduce radiation exposure from primary gamma, neutron, and secondary gamma sources sufficiently to prevent appreciable activation of nearby apparatus. The primary shield is adequate to permit routine work during refueling. A neutron shield in the annular gap around the reactor vessel attenuates neutron streaming from the reactor cavity during operation.

Secondary Shield The secondary shield comprises all walls and floors inside the containment which are built around the reactor loop and around other equipment which contains radioactivity either permanently or during refueling. The secondary shield reduces exposure to levels permitting access for limited periods on the ground and operating floors during operation.

The secondary shield provides adequate protection for refueling operations, inspection, repair and maintenance under the general supervision of the health physics staff during refueling and shutdown periods.

USAR-1 1.2 Information Use Page 7 of 31 Radiation Protection and Monitoring Rev. 1-Concrete shielding provides protection in the spaces close to the fuel transfer route and the reactor internals temporary storage area during spent fuel assembly handling. The fuel transfer route is flooded with water which protects the air space above the fuel transfer canal.

Auxiliary Building Eguigment Shielding Adequate shielding is provided around equipment that carries radioactive liquid, slurry or gas. The controlled access area is reached through a single service corridor which is entered at the access control station.

Highly radioactive equipment is isolated in shielded compartments individually or in groups. The entrances of such containments are provided with labyrinths for full protection of adjacent areas. At concentrations of radioactive equipment and piping, the manually operated valves are grouped in adjacent valve rooms. Pumps and compressors serving highly contaminated tanks are in separate shielded compartments. Certain piping systems are shielded by lead blankets.

Two separately shielded rooms are provided for the engineered safeguards pumps.

Control Room The control room is shielded with concrete on all sides. There are no wall openings which are in direct sight of the containment. The radiological consequences of a LOCA on control room personnel are discussed in Section 14.15.

Radioactive Waste Processing Building and CARP Building Shielding Adequate shielding is provided for each room in the Radioactive Waste Processing and CARP buildings to permit continuous occupancy in general access areas. The controlled access areas are entered through either the access point at the southeast entry to the Radioactive Waste Processing Building or the controlled access point at the entry to the auxiliary building from the CARP Building.

USAR-1 1 .2 Information Use Page 8 of 31 Radiation Protection and Monitoring Rev. 1-6 11.2.2.5 Shielding Materials The bulk of shielding material is either ordinary concrete with a minimum density of 2.33 g/cc or concrete block with a bulk density of 1 .92 g/cc. Where space requirements prohibit the use of ordinary concrete, high density concrete of 3.6 g/cc density with magnetite aggregate is used. Neutron shielding is also used in the reactor vessel annulus consisting of borated concrete with a density of 1 .68 g/cm

. Observation windows at the drumming area 3

are of high density glass with the same attenuation properties as the surrounding shielding wall. Holes above the purification filters are plugged with slabs of high density concrete or steel. The mass of these slabs is 340 lb/ft2 horizontal surface. Lead blankets are also permanently installed on certain piping systems and were attached per engineering procedures to ensure that the shielding does not fall off the pipe during a seismic event. The lead shielding is identified on isometrics. Pipe stress analyses were conducted on those piping systems which required permanent lead shielding.

11 .2.2.6 Reactor Radiation Sources Full Power Sources A reactor power of 1500 MWt and an axial peak to average power factor of 1 .5 were used for the calculations. The side gamma and neutron spectra are based on a 140 in. reactor vessel inside diameter with material thicknesses of 0.75, 1 .5, 3 and 7.25 in. for the shroud, barrel, thermal shield and vessel, respectively. A vessel inside length of 451 .75 in. and vessel wall thickness of 3.5 in. and 6 in. for the bottom and top respectively were assumed for the axial gamma and neutron spectra. For purposes of radiation source calculations a core active length of 132 in. was used for the axial spectra. The gamma spectra are not isotopic as the high energy gammas are very directional. For shielding calculations it was assumed that the gamma spectra are currents at the outer pressure vessel surfaces. The fast neutron spectra at the top of the vessel are small and can be neglected. The radiation sources were considered as follows:

USAR-1 1 .2 Information Use Page 9 of 31 Radiation Protection and Monitoring Rev. 1-

a. Groups of neutron fluxes through the reactor vessel: The fast neutron spectra for a point on the top and bottom of the pressure vessel along the vertical centerline and for a point on the interface of the concrete biological shield (assumed two feet from the vessel) are shown in Table 11 .2-1. The fast neutron spectra are given for 10 source groups. The intermediate and thermal neutron fluxes at the outer surfaces are shown in Table 11 .2-2. The intermediate and thermal neutron fluxes at the top and bottom of the vessel are small and can be neglected.

Table 11 .2-1 - Neutron Spectra Outside Vessel Neutron Energy Neutron Flux (n/cm2-sec-MeV)

(MeV) Bottom Skie 0.33 insignificant 9.96 (+1)* 6.10 (+7) 1 insignificant 3.54 (+1) 5.02 (+7) 2 insignificant 2.68 (+1) 4.46 (+7) 3 insignificant 2.31 (+1) 4.23 (+7) 4 insignificant 1.77 (+1) 6.03 (+7) 6 insignificant 1.26 (+1) 6.34 (+7) 8 insignificant 1.00 (+1) 3.40 (+7) 10 insignificant 7.72 (+0) 1.53 (+7) 14 insignificant 3.80 (-1) 5.98 (+5) 18 insignificant 4.54 (-3) 1.02 (+4)

() Denotes power of ten Table 11 .2-2 - Intermediate and Thermal Neutron Fluxes Outside Vessel Neutron Flux (n/cm2-sec)

Energy Group Bottom Side Epithermal 0 5.6 (4)* 7.57 (+9)

Thermal 0 5 (-4) 1.04 (+10)

() Denotes power of ten

a. Core and capture gamma sources at external surfaces of the reactor vessel: The gamma spectra for a point on the top and bottom of the reactor vessel along the vertical centerline and for a point on the inner face of the concrete biological shield adjacent to the position of maximum axial power are shown in Table 11.2-3. The gamma spectra are given for 14 source groups.

USAR-11.2 Information Use Page lOof3l Radiation Protection and Monitoring Rev. 1-Table 11 .2-3 Gamma Spectra Outside Vessel Gamma Energy Gamma Flux Whotons/cm2-sec-MeV)

(MeV) Bottom Side 10 --- --- 9.34 (+6) 9 2.60 (+0)* 4.22 (+3) 8.97 (+7) 8 4.00 (+3) 7.90 (+6) 1.70 (+8) 7 2.30 (+3) 2.81 (+6) 2.05 (+8) 6 6.74 (+3) 1 .71 (+7) 2.23 (+8) 5 6.74 (+3) 1.17 (+7) 2.56 (+8) 4 9.12 (+3) 2.46 (+7) 3.28 (+8) 3 1.17 (+4) 3.42 (+7) 4.84 (+8) 2 1.72 (+4) 5.18 (+7) 9.02 (+8) 1 .375 2.70 (+4) 8.08 (+7) 5.43 (+8) 1 3.74 (+4) 1.13 (+8) 5.00 (+8) 0.75 5.24 (+4) 1.61 (+8) 6.98 (+8) 0.5 8.72 (+4) 2.74 (+8) 1 .20 (+9) 0.25 2.21 (+5) 7.18 (+8) 1.41 (+9)

() Denotes power of ten

a. Reactor coolant nitrogen-16 activity: The N-16 activity for the minimum coolant loop transmit time is shown in Table 11 .2-4.

Table 11.2-4 - N-16 Activity Around the Reactor Coolant Loop Transit Time Disintegrations Location (sec) (cm3-sec)-1 Core Outlet 0 3.31x106 Vessel Outlet 1.07 2.99x106 Steam Generator Inlet 1.347 2.92x106 Steam Generator Outlet 4.797 2.1 1x106 Vessel Inlet 6.352 1.82x106 Shutdown Sources The shutdown sources consist of gamma fluxes outside the vessel. The core midplane sources are as shown in Table 11 .2-5.

The shutdown sources are given for two days and ten days after shutdown. Continuous operation to end of core life at a power of 1500 MWI was assumed for the calculation. The gamma spectrum from the internals and vessel wall is the same at two days and ten days because the source nuclides are long-lived.

The values shown in Table 11.2-5 are maxima since they include the axial peaking factor of 1.5.

USAR-1 1.2 Information Use Page 12 of 31 Radiation Protection and Monitoring Rev. 4-6 Table 11 .2-6 Maximum Deposited Activities in Pressurizer Due to Crud Activity (d/cm2-sec)

Isotorje (based on 1 .5 mg/cm2)

Co 60 8.45 (5)*

Co 58 6.62 (6)

Mn54 1.59(4)

Fe 59 6.00 (3)

Cr 51 2.63 (5)

Zr 95 1 .48 (4)

( ) Denotes power of ten Steam Section The activity in the steam section is due to the build-up of the gaseous fission products Xe and Kr from the assumed 1 percent failed fuel. It was assumed for the calculation that all the Xe and Kr isotopes could build up for 1 year and were supplied to the pressurizer at a continuous reactor coolant spray rate of 1 .5 gpm.

The activity in the steam section of the pressurizer is as shown in Table 11.2-7.

Table 11 .2-7 - Maximum Activity in Pressurizer Steam Section NUCLIDE ANNUAL BASIS ANNUAL BASIS Pressurizer Steam Pressurizer Steam Space Activity Space Concentration (Xe, Kr Only) (Xe, Kr Only)

Ci iiCi/cc Kr-83m 1.98E+OO 1.7iE-O1 Kr-85 5.60E÷02 4.81 E÷O1 Kr-85m 1.03E+O1 8.84-O1 Kr-87 5.86E+OO 5.04E-O1 Kr-88 183E÷O1 1.57E+OO Kr-89 427E-Oi 3.67E-02 Kr-90 7.80E-02 6.7iE-03 Xe-131m 230E+02 1.97E+O1 Xe-133 6.96E+03 5.99E+02 Xe-133m 2.83E+Oi 2.43E+OO Xe-135 5.84E÷O1 5.03E÷OO Xe-135m 9.1OE-O1 7.83E-02 Xe-137 9.93E-Oi 8.54E-02 Xe-i 38 3.49E+OO 3.OOE-O1 Total 788E+O3 6.78E+02

USAR-1 1 .2 Information Use Page 11 of 31 Radiation Protection and Monitoring Rev. .4-s Table 11.2-5 Shutdown Gamma Spectra at Vessel OD Gamma Energy Gamma Flux (photons/cm2-sec)

(MeV) Core Shroud Barrel Thermal Sh. Vessel 2 days after shutdown 4.00 1.77(3)*

3.00 2.86 (4) 2.00 6.40 (5) 1.38 4.82 (5) 1.80 (5) 4.52 (5) 9.01 (5) 1.00 4.67 (5) 1.11 (4) 2.91 (4) 5.81 (4) 2.22 (6) 0.75 7.02 (5) 0.50 1 .29 (6) 0.25 1.54(6) 10 days after shutdown 4.00 1.18(3) 3.00 2.00 (4) 2.00 4.00 (5) 1.38 3.02(5) 1.80(5) 4.52(5) 9.01 (5) 1.00 2.91 (5) 1.11 (4) 2.91 (4) 5.81 (4) 2.22 (6) 0.75 4.40 (5) 0.50 8.12(5) 0.25 9.57 (5)

() Denotes power of ten 11 .2.2.7 Pressurizer Radiation Sources Liquid Section Sources in the liquid section of the pressurizer are plated-out radioactive crud and dispersed fission products in the water. The maximum deposited activity for the liquid section assuming a crud film of 1 .5 mg/cm2 on the pressurizer internals (liquid section only) is shown in Table 11 .2-6. As an upper limit, the liquid section would have dispersed fission product activity equal to the primary coo lant activity with one percent failed fuel.

USAR-1 1.2 Information Use Page 13 of 31 Radiation Protection and Monitoring Rev. .1-s 11 .2.2.8 Reactor Coolant System Radiation Sources The deposited crud activity in the reactor coolant system is shown in Table 11 .2-8. The fission product activity in the reactor coolant loop at 70°F is shown in Table 11.1-7.

Table 11 .2-8 - Maximum Deposited Activities in Reactor Coolant System Due to Crud Activity (d/cm2-sec)

Steam Generator Reactor Coolant Isotope Tubing Piping Co 60 9.00 (4)* 8.45 (5)

Co 58 7.06 (5) 6.62 (6)

Mn54 1.70(3) 1.59(4)

Fe 59 6.40 (2) 6.00 (3)

Cr 51 2.80 (5) 2.63 (6)

Zr 95 1.58 (3) 1.48 (4)

() Denotes power of ten 11 .2.2.9 Waste Treatment Radiation Sources Maximum Activity in the CVCS Demineralizers The maximum activity accumulated in a CVCS mixed bed demineralizer that has been in service for 1 year with 1 percent failed fuel is shown in Table 11 .2-9. The activity shown is for 40 gpm continuous flow rate. The reactor coolant activity is assumed to be at an equilibrium level when the demineralizer is put into operation. The demineralizers are assumed to not remove noble gases, Molybdenum, Rubidium, and Yttrium. The RCS concentration at STP conditions was used for this calculation. For conservatism the demineralizers are assumed to remove 100% of the nuclides that are in the RCS (1% failed fuel).

Also conservatively it was assumed that the activity resulting from an 18 month cycle would be discharged to the CVCS.

USAR-11.2 Information Use Page 14 of 31 Radiation Protection and Monitoring Rev. 4-Table 11 .2-9 Maximum Activity in CVCS Demineralizer KI WASTE UCJDE 1

TREATMENT CVCS One Year Operation Activity Ci Br-82 3,47E-O1 Br-83 9.93E-0l Br-85 4.18E-02 1-129 7.29E-04 1-130 8.94E-01

-131 3.OOE+03

-132 1.06E+01 1-133 1.37E+02

-134 6.45E÷00 1-135 4.11E+01

-136 6.79E-02 Cs-132 i.89E-02 Cs-134 2.82E+03 Cs-134m 3.61E-01 Cs-135m 1.13E-01 Cs-136 5.39E+01 Cs-i 37 2.47E+03 Cs-138 3.30E+00 Cs-i 39 8,90E-0i Cs-140 9.i7E-02 Ag-lb 3.64E-03 Ag-ibm 5.09E+0i Ag-ill 3.50E+0l Ag-112 2.81E-Oi As-76 1 .71 E-03 Cd-115 l.42E÷00 Cd-115m l.31E+00 Ga-72 7.48E-04 Ge-77 2.55E-02 In-115m l.19E-0l Sb-122 2.04E-0l Sb-124 3.47E÷00 Sb-125 l.51E÷02 Sb-127 2.58E÷0l Sb-129 4,53+O0 Sb-130 2.25E-0l Sb-130m l.51E-0l Sb-131 9.68E-0l Sb-132 l.05E-0l Sb-i 32m 6.80E-02

USAR-11.2 Information Use Page 15of31 Radiation Protection and Monitoring Rev. 1-6 WASTE NUCLIDE TREATMENT CVCS One Year Operation Activity Ci Sb-i 33 8.90E-02 Se-83 7.16E-02 Sn-i21 7.05E-Oi Sn-123 5.46E+OO Sn-i 25 3.70E+OO Sn-i27 2.36E-Oi Te-127 2.57E+OO Te-i27m 108E+02 Te-129 i14E+OO Te-129m i.60E+02 Te-13i i.12E÷OO Te-131m i.91E+O1 Te-132 3.58E+02 Te-133 7.52E-O1 Te-133m 2.74E+OO Te-134 4.18E+OO Ba-i37m i.66E-02 Ba-139 8.30E+OO Ba-140 i.80E+03 Ba-142 9.05E-O1 Sr-89 4.15E÷03 Sr-90 1 .93E+03 Sr-9i 4.07E÷Oi Sr-92 i.2iE÷O1 Sr-93 6.15E-Oi Sr-94 1.03E-O1 Pd-109 i.68E+Oi Rh-i 03m 4.75E+OO Rh-i05 i.i5E+02 Rh-105m i.25E-02 Rh-i06 i.67E-02 Ru-i03 4.79E÷03 Ru-i06 8.1OE÷03 Tc-99m 3.i9E+O1 Tc-iOi 1.28E÷OO Tc-i04 1.28E÷OO Tc-105 4.40E-Oi Ce-i4i 4.24E+03 Ce-i43 i.69E+02 Ce-i44 1,69E÷04 Np-239 3.81 E÷03 Pu-238 6.60E+Oi Pu-239 8.52E+OO

USAR-1 1.2 Information Use Page 16 of 31 Radiation Protection and Monitoring Rev. 4-6 WASTE NUCLIDE TREATMENT CVCS One Year Operation Activity Ci Pu-240 1.11E+O1 Pu-241 2.80E+03 Pu-242 4.72E-02 Th-228 6.23E-05 Am-241 3.44E+OO Cm-242 4.95E+02 cm-244 9.24E+O1 Eu-154 1.39E+02 Eu-i 55 5.85E+O1 Eu-i 56 2.73E÷02 Eu-157 1.2iE+OO Eu-158 2.13E-02 Eu-i59 4.26E-03 Gd-159 3.54E-Oi Ho-166 7.iOE-03 La-i40 2.43E+02 La-141 2.iiE+Oi La-142 8.02E+OO La-143 i.20E+OO Nb-95 4.76E+03 Nb-95m 5.60E÷OO Nb-97 6.36E+OO Nb-97m 8.31 E-02 Nd-147 5.71 E÷02 Pm-147 3.63E+03 Pm-148 6.98E÷Oi Pm-148m 105E+02 Pm-149 9.72E+Oi Pm-151 i.82E+Oi Pr-142 3.70E+OO Pr-143 1.63E÷03 Pr-144 1.21E+OO Sm-153 6.64E÷Oi Tb-160 5.40E+OO Zr-95 8.52E+03 Zr-97 8.88E+Oi Rn-220 1.90E-1O Total 7.94E+04

USAR-1 1.2 Information Use Page 17 of 31 Radiation Protection and Monitoring Rev. .1-s Gas Decay Tank Activity The maximum activity in a gas decay tank would exist after degassing the reactor coolant prior to a cold shutdown (e.g., for refueling). Assuming reactor coolant activities as in Table 11.1-7, the total activities in this gas decay tank immediately following degassing and after 30 days holdup are given in Table 11.1-21.

Table 11 .2-10 Gas Decay Tank Activities Activity (Ci) Activity (Ci)

Nuclide Immediately After After 30 Day Holdup Kr-83m 6.40E+2 0 Kr-85 5.62E+1 5.59E+1 Kr-85m 1.35E+3 0 Kr-87 2.72E+3 0 Kr-88 3.80E+3 0 Kr-89 4.77E+3 0 Kr-90 5.13E+3 0 Xe-131m 4.75E+2 8.28E+1 Xe-i 33 3.27E+4 6.20E+2 Xe-133m 3.i8E+2 2.40E-2 Xe-i 35 3.78E+3 0 Xe-135m 2iiE+3 0 Xe-i 37 9.22E+3 0 Xe-i 38 8.79E+3 0 Total Activity (Ci) in Decay Tank after Degassing = 7.58E+4 Total Activity (Ci) after 30 Day Holdup = 7.59E+2 ii.2.2.iO Refueling Radiation Sources CEA Activity The maximum specific activity of the major radioactive isotopes following irradiation for 15 years with two days decay is shown in Table 11.2-il.

USAR-1 1 .2 Information Use Page 18 of 31 Radiation Protection and Monitoring Rev. 1-Table 11 .2-1 1 - Specific Activity of Major Isotopes Following Irradiation Isotope Specific Activity Isotope Specific Activity

[Curies/g] [Curieslg]

51 Cr 1.527E+OO 3.833E-05 Mn 54 2.784E-03 mNb 93 1.595E-05 Fe 55 1.143E-O1 94 Nb 3.585E-04 Fe 59 1.679E-02 Nb 95 4.813E-O1 Co 58 2.091E-O1 Nb 96 1.464E-04 6000 1.089E+OO Mo 93 7.144E-05 59 Ni 1.197E-03 Mo 99 1.638E-O1 Ni 63 2.723E-O1 mTc 99 1.585E-O1 Cu 64 1 .282E-03 Ru 103 6.796E-04 Zn 65 4.267E-05 mRh 3

lO 6.784E-04 The specific activities are given as Curies per Gram in the Inconel 625 CEA cladding. The B4C absorber material does not contribute to the gamma activity.

The maximum radiation source of the CEA is based on residence in the reactor core for 15 years with a core power level of 1775 MWt. A decay time of two days was assumed prior to removal from the reactor. The assumed elemental composition of Inconel 625 is as shown in Table 11.2-12.

Table 11.2-12 - Inconel 625 Composition Element Weight (%) Element Weight (%)

Ni 61.0 Mn 0.15 Cr 22.0 Si 0.3 Fe 3.0 Nb 4.0 Mo 9.0 Co Q*3*

• Based on an assumed Co content in Ni of 0.5 w/o.

USAR-1 1.2 Information Use Page 19 of 31 Radiation Protection and Monitoring Rev. .1-s Spent Fuel Assemblies The following assumptions were made for the calculations:

a. Long term reactor operation at a power level of 1500 MWt.

b. The length of the active section of the fuel assembly is 128 in.

and the overall length is 145-7/8 in.

c. The assembly has decayed at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to handling and transfer to the spent fuel storage pool.

d. The axial peak to average power factor seen by the assembly was 1.5.

The decay gamma sources in spent fuel assemblies are shown in Figure 11.2-2.

The total gamma attenuation coefficients are given in Table 11 .2-13 for the hot operating condition and for cold shutdown. The cross sections are based on volume fractions of 0.341, 0.101 and 0.548 for U02, zircaloy and water respectively.

The average densities of the fuel assemblies are 4.614 and 4.752 g/cc for the hot condition and cold shutdown, respectively.

Table 11 .2-13 Total Gamma Attenuation Coefficients Gamma Energy Coefficient (cm2/g)

(Mev) (Hot) (Cold) 4.00 4.05(2)* 4.04(-2) 3.00 4.21(-2) 4.21(-2) 2.00 4.71(-2) 4.72(-2) 1 .38 5.66(-2) 5.67(-2) 1.00 7.16(-2) 7.15(-2) 0.75 9.84(-2) 9.80(-2) 0.50 1.48(-1) 1.46(-1) 0.25 4.96(-1) 4.87(-1)

() denotes power of ten

USAR-1 1.2 Information Use Page 20 of 31 Radiation Protection and Monitoring Rev. .1-s 11 .2.3 Radiation Monitoring 11.2.3.1 General Permanently installed radiation monitors are provided for surveillance of plant effluents and critical process streams (process monitors), and personnel exposure levels in hazardous and potentially hazardous plant areas (area monitors). Monitoring and recording is required for liquid and gaseous releases. The monitoring program meets the requirements of 10 CFR Part 50, Appendix I and the ODCM.

The following general design provisions and objectives apply to each monitor in containment, the auxiliary building and Radioactive Waste Processing Building regardless of category of functions:

a. Remotely operable check source at detector,

b. Two independently adjustable, high radiation setpoints; both setpoints are annunciated,

c. In addition to alarms, control room readouts include continuous indication, and recording on multi-point recorders (area monitors repeat indication and alarms at remote detector locations),

d. Independent channel power supplies,

e. Sufficient sensitivity, in relation to detector location, to ensure capability for compliance with 10 CFR Part 20.

Additional information concerning radiation monitoring in the vicinity of the plant, both on-and off-site is presented in Section 2.10.

The alarm setpoints are provided in the Operating Manual Technical Data Book (TDB) and will allow the operators to monitor activities between times grab samples are taken.

Two independently adjustable setpoints are provided for each monitor. The lower setpoint alarm, designated ALERT/WARN, warns that the dose rate has reached an abnormal value. The upper setpoint alarm, designated HIGH ALARM, warns that the dose rate has reached or passed the permissible limit. The local indicator, as well as the control room indicator and recorder, indicates the actual dose rate at the detector location.

USAR-1 1.2 Information Use Page 21 of 31 Radiation Protection and Monitoring Rev. 4 Calibration of the radiation monitors is checked on a frequent, regular interval basis by exposing the detector to a known source and verifying that the reading is within the recommended tolerance of a previous calibration setting. If the reading is not within tolerance, the entire monitor components will be calibrated as necessary, e.g., the detector high voltage setting, alarm setpoints, discriminator setting or panel meter indication.

The principal functions of process radiation monitors are:

a. Continuous check of plant effluent activity concentrations and accumulation of data useful in accounting for released activity,

b. Early warning of equipment failures, malfunctions, and deteriorating performance,

c. Final check on proper execution of potentially hazardous periodic operations such as containment purging and release of treated waste batches,

d. To determine release rates of radionuclides from the plant during Post-Accident operation.

e. To provide the primary means for detecting RCS leakage of 1 gpm in four hours (containment building radiation monitors only) (Ref. 11.4.18).

Process radiation monitors were selected to provide adequate sensitivity to protect the public and the plant personnel based on 10 CFR Part 20 limits, and to provide reliable and stable long-term service. The design is modular to facilitate maintenance, updating, or changes to adapt to new requirements.

The process radiation monitor readings are compared with laboratory analysis of process samples. This procedure allows for a continuing evaluation of the process radiation monitor calibration and verifies the internal check source. If any process radiation monitor reading is questionable, corrective action such as testing and calibration will be taken as necessary to assure an accurate reading.

Administrative programs are maintained to ensure the capability to accurately determine the airborne iodine concentrations in vital areas in post accident situations and to obtain and analyze radioactive lodines and particulates in plant gaseous effluents.

USAR-1 1 .2 Information Use Page 22 of 31 Radiation Protection and Monitoring Rev. 1-11 .2.3.2 Containment Building Radiation Monitors The containment building is usually closed during operation at power; ambient temperature is controlled by recirculation of air through internal cooling units.

The principal source of radionuclides in the containment atmosphere is fission products from coolant leaks not confined by leakoffs. The nuclear detector well cooling system is so arranged that argon-41 produced by the interaction of leakage neutrons and ventilation air immediately around the reactor vessel and out-of-core instrument thimbles is confined to the system (see Section 9.10.2.3). Thus significant concentration of Ar-41 is kept from the containment general atmosphere.

The containment air monitoring system comprises a moving filter-paper particulate monitor (channel RM-050) and a sample chamber noble gas monitor (RM-051) installed in a common housing, with pump and flow controls, outside the containment at a location where background is minimal. A continuous air sample is drawn from the containment, passed through the particulate and noble gas monitor in series, and returned to the containment (see P&ID 11405-M-1). The particulate monitor, upstream of the noble gas monitor, serves as a filter to reduce particulate contamination of the noble gas monitor. Sampling and return lines are provided with containment isolation valves. Sample valves allow grab samples of the containment atmosphere for analysis of the gaseous activity, or alternatively to measure the airborne iodine concentration by passing the sample flow through a replaceable, activated charcoal filter cartridge.

Particulate monitor RM-050 is equipped with a BETA scintillator detector.

Noble gas monitor RM-051 is a sample chamber type with a beta scintillator detector.

The RM-050 alarm setpoints alarm only. The RM-051 lower alarm setpoint alarms only, while the upper setpoint alarms and initiates a CRHS signal.

Grab samples are the primary means of analyzing for the presence of various gaseous nuclides and for airborne halogens.

Grab samples will always be taken and analyzed prior to initiating a containment purge. The gaseous effluent monitors described in the next section provide further protection against exceeding release limits.

USAR-1 1.2 Information Use Page 23 of 31 Radiation Protection and Monitoring Rev. 1-Optionally, gas monitor RM-052 can be aligned to sample the containment. RM-052 is located adjacent to RM-0501051 and utilizes the same containment sample lines. RM-052 containment sampling lines have similar design features as RM-050/051 (see P&ID I I405-M-1).

RM-052 is a beta scintillator detector. When aligned to the containment, RM-052 monitors containment atmosphere gas and alarms and initiates a CRHS signal similar to RM-051. RM-052 is intended as a backup to RM-051 to allow normal plant operations when RM-051 is out of service for maintenance or calibration.

11 .2.3.3 Auxiliary Building Exhaust Ventilation Stack Gaseous Effluent Monitors The following radioactive and potentially-radioactive gaseous effluents are released to atmosphere via the Auxiliary Building Exhaust Ventilation Stack, which is a monitored release point:

a. Auxiliary building ventilation exhaust;

b. Containment atmosphere purge exhaust;

c. Containment hydrogen purge exhaust;

d. Waste gas released from RWDS.

The listed effluents are sampled, monitored, and released through the Ventilation Discharge Duct as shown by P&ID 1 1405-M-l.

Auxiliary Building Exhaust Ventilation Stack Monitors:

The following monitors are provided for full-time surveillance of waste gas releases at the release point:

a. RM-062 continuously monitors the ventilation stack gas.

RM-062 is an off-line system with a particulate filter and iodine cartridge, sample pump, flow controls, control room indication alarm and CRHS actuation.

Particulates are collected on a filter. Iodine is collected on an impregnated charcoal filter canister mounted behind the particulates filter.

Following use, contents of the filter and of the cartridge are laboratory counted for accurate assessment of particulate and iodine releases. Integrated sample flowrate is accumulated on the ratemeter.

USAR-1 1.2 Information Use Page 24 of 31 Radiation Protection and Monitoring Rev. 1-6 RM-062 gas monitor is a beta scintillator detector. The lower setpoint alarms only. The higher setpoint alarms and initiates a CRHS signal.

RM-062 (low range), along with RM-063 (high range), sample and monitor the stack to satisfy post accident monitoring requirements.

b. RM-063 is an off-line system with a particulate filter and iodine cartridge, sample pump, flow controls, and control room indication. RM-063 works in conjunction with RM-062.

RM-062 maintains sample flow through the entire post-accident range. As sample activity increases to the upper range limit of RM-062, the RM-062 monitor automatically valves in RM-063 sample lines and starts the high range monitor RM-063 sample pump. The RM-063 sample pump diverts a portion of the RM-062 sample stream to the RM-063 sampler for sampling and monitoring.

RM-063 is a flow through ion chamber detector. Two parallel particulate filters and iodine cartridges, housed in lead shielded samplers, are located upstream of the RM-063 sampler.

c. RM-052 is a dual function gas monitor. This off-line gas monitor can sample from either the auxiliary building vent stack or from the containment. When RM-052 is aligned to sample the stack, it utilizes independent sample lines (see P&ID 11405-M-1).

Optionally, particulates and iodine can be collected for analysis.

Particulates are collected on a filter. Iodine is collected on an impregnated charcoal filter canister mounted behind the particulate filter.

Following use, contents of the filter and of the cartridge are laboratory counted for accurate assessment of particulate and iodine releases. Integrated sample flowrate is accumulated on the ratemeter.

USAR-1 I .2 Information Use Page 25 of 31 Radiation Protection and Monitoring Rev. 1-11.2.3.3.1 Laboratory and Radioactive Waste Processing Building Exhaust Stack Monitor:

The potentially radioactive noble gas, particulate, and iodine effluents from the CARP and Radioactive Waste Processing Building are released to the atmosphere via the Laboratory and Radioactive Waste Processing Buildings Exhaust Stack which is a monitored release point. These effluents are collected, combined, filtered, monitored and released through this exhaust stack as shown on drawing 7753-03-M-1.

RM-043 continuously monitors the Laboratory and Radioactive Waste Processing Building (LRWPB) ventilation stack gas (see P&ID 7753-03-M-1).

RM-043 is an off-line system with a particulate filter and iodine cartridge, sample pump, flow controls, and control room indication alarms.

Particulates are collected on a filter. Iodine is collected on an impregnated filter canister mounted behind the particulate filter. Following use, contents of the filter and of the cartridge are laboratory counted for accurate assessment of particulate and iodine releases. Integrated sample flowrate is accumulated on the ratemeter.

RM-043 gas monitor is a beta scintillator detector.

RM-043 alarm setpoints alarm only.

11 .2.3.4 Condenser Off-Gas Monitor The condenser off-gas line is monitored with an in-line noble gas monitor, RM-057. RM-057 serves as a steam-generator leak detector, and can monitor off-gas before discharge. The lower detection limit of RM-057 is sufficient to detect leak rates across the steam generator tubes on the order of 0.0022 lb/sec per the requirements of Regulatory Guide 1 .97 and 30 gpd (gallons per day) per Reference 11 .5.7. The capability exists to realign the condenser exhaust to the auxiliary building stack as necessary to monitor effluent releases (Reference 11 .4.15). It is also possible to draw a grab sample of the condenser off-gas line for analysis.

Sample piping and valves are installed, in the off-gas header, to allow grab samples of the condenser off-gas.

USAR-1 1.2 Information Use Page 26 of 31 Radiation Protection and Monitoring Rev. 1-6 RM-057 is a Sodium Iodide Gamma Scintillation Detector tube.

RM-057 has control room indication and alarms. The lower setpoint alarms only. The higher setpoint alarms and isolates the sixth stage turbine extraction to the auxiliary steam system.

11 .2.3.5 Monitoring of Liquid or Gaseous Effluents from Steam Generator Blowdown or Steam Leakage in the Turbine Building A monitor is installed in each of the two steam generator blowdown sample lines (RM-054A and B) as discussed in Section 9.13. Like the condenser off-gas monitor, the blowdown monitors would respond only on occurrence of a primary-secondary leak. The functions of the monitors are to give early alarm of significant tube leakage, and at the upper setpoint, isolate all blowdown to prevent release of radionuclides to atmosphere via the blowdown flash tank (see P&lD 11405-M-12),

or to the river via the circulating water.

The possibility of radioactive gases being in the Turbine Building would be indicated by the condenser off-gas monitor RM-057, and the blowdown monitors RM-054A and B.

The existence of significant count rates on either of these monitors, together with observed steam leakage in the Turbine Building, will be cause to use portable monitors to measure the concentrations of radioactive gases in the Turbine Building. From this measurement the release rate can be inferred and will be considered in demonstrating compliance with Technical Specification release limits.

RM-054A and RM-054B are on-line monitors that monitor the steam generator blowdown. (See P&ID 11405-M-12).

RM-054A and RM-054B monitors are gamma scintillator detectors. RM-054A and RM-054B have control room indication and alarms and either monitor will initiate isolation of steam generator blowdown.

11 .2.3.6 Plant Liquid Effluent Monitor Radioactive or potentially-radioactive liquid effluents reach the plant circulating water system, hence the river, via the RWDS overboard header and two raw-water overboard headers.

USAR-1 1 .2 Information Use Page 27 of 31 Radiation Protection and Monitoring Rev. 1-RWDS Overboard Header RM-055 is an in-line liquid monitor located in the Radioactive Waste Discharge System (RWDS) effluent header. RM-055 monitors the waste tank discharge line to the circulating water discharge tunnel (see P&ID I 1405-M-9).

RM-055 liquid monitor is a gamma scintillator detector. RM-055 has control room indication and alarms. The lower setpoint alarms only. The higher setpoint alarms and initiates closure of the discharge valves to the circulating water overboard header, and stops the waste monitor tank pumps.

11 .2.3.7 Other Process Monitors Component Cooling Water Monitor RM-053 is an in-line monitor that continuously monitors the component-cooling water system downstream of the component-cooling/raw-water heat exchangers (see P&lD 11405-M-1O). RM-053 monitor is a gamma scintillator detector with control room indication alarms. RM-053 alarm setpoints alarm only.

The component-cooling water system is a closed system cooled and buffered from the river by the raw-water system.

11.2.3.8 Area Radiation Monitors A twenty-three channel area monitoring system is provided to protect plant personnel in the containment, auxiliary building, and radioactive waste processing building. Detectors, located where potential hazard exists and routine access is required, are air filled ionization chambers.

Readout modules, independent channel power supplies, and a multipoint recorder are installed on radiation monitor panel Al-33 (see Section 7.6.2). Alarms and indication are, however, repeated at the detector locations.

The area monitors respond monotonically over an eight decade range extending from 0.1 mr/hr to 107 mr/hr. The indicator displays all eight decades. The full eight decades are recorded in the control room for each channel.

USAR-1 1 .2 Information Use Page 28 of 31 Radiation Protection and Monitoring Rev. .1-s Two independently adjustable setpoints are provided for each monitor. The lower setpoint alarm, designated warn, warns that the dose rate has reached an abnormal, but still safe value. The upper setpoint alarm, high radiation, warns that the dose rate has reached or passed the permissible limit for continued occupancy. The local indicator, as well as the control room indicator and recorder, indicates the actual dose rate at the detector location.

Area monitor detector locations are shown in Table 11.2-14.

Table 11 .2-14 Area Radiation Monitors Channel Building Number Containment Basement RM-070 Ground Floor RM-071/2 Operating Floor RM-073/415 Auxiliary Building Basement RM-07617/8/9 Ground Floor RM-080/1/2/4 Operating Floor RM-085/617!8 Control Room Mechanical Room RM-089 Radioactive Waste Processing Bldg.

Room 506 RM-095 Room 504 RM-096 Room 502 RM-097 Office Hallway RM-098 Radiation monitor, RM-080, is located in the New Fuel Storage Room 25a to detect excessive radiation levels when fuel is present.

Radiation monitors, RM-091A and RM-091 B, provide high level radiation measurements which would be required during accident conditions. The monitors are located at an elevation of 1 045-0, but in different areas within the containment. The detector range is from 1 to 107 R!h. Readouts for the monitors are on control panel AI-33C in the control room. Both readouts have high radiation, alert and failure alarms and indicating lights. A system test of each monitor can be performed from the control room.

USAR-1 1.2 Information Use Page 29 of 31 Radiation Protection and Monitoring Rev. .4-s This test checks the electrode configuration and electrical operation of the detector/cable/readout system. These monitors extend the range of the containment area radiation monitoring system.

11 .2.3.9 Personnel Contamination Monitors Several personnel contamination monitors are provided at points beyond which persons should not proceed with contaminated clothing, supplies, equipment, or tools. High sensitivity beta monitors are provided at key locations to detect personnel and equipment contamination as well as Station Exit Monitors of the walk-thru portal type in the Security Building.

11.2.3.10 Power Supply Each process and area monitor channel has an individual power supply for development of the required internal operating voltages.

Channel power supplies are distributed among the four plant a-c instrument buses (see Section 8.3).

11.2.3.11 Post-Accident Main Steam Line Monitor RM-064, Post-Accident Main Steam Line Monitor is an off-line monitor designed to measure the steam activity by sampling steam from the two steam headers via two isolation valves HCV-921 and HCV-922. The monitor will be placed on-line in the event of a steam generator tube rupture. The monitor when on auto position can sample steam from both steam headers by alternating between the headers every 7 minutes. The recorded data from this monitor can then be utilized to quantify effluents released through the main atmospheric dump valve (HCV-1040),

the main steam safety valves, and the auxiliary feedwater pump turbine. RM-064 is located in the Turbine Building next to Room 81.

Optionally, the sampling point can be switched from one steam header to another. Sampling point selection valves (HCV-921 and HCV-922) are controlled from panel Al-33C in the Control Room.

The four position selection switch will be utilized for manual and automatic control of the isolation valves.

RM-064 is a Geiger-Mueller (GM) tube. RM-064 has control room indication and alarms.

USAR-1 1.2 Information Use Page 30 of 31 Radiation Protection and Monitoring Rev. .1-s 11.2.3.12 Post-Accident Control Room Iodine Monitor (RM-065)

RM-065 is an off-line monitor used for monitoring iodine in the control room during design basis accident conditions. An isokinetic nozzle probe is installed in the ventilation duct in the control room. The air sample is drawn through the probe by a vacuum pump. The exhaust air is then released back to the ventilation duct downstream of the inlet nozzle.

11 .2.4 Radioactive Material Storage 11.2.4.1 Radioactive Waste Storage Radioactive waste awaiting disposal is stored in the Radioactive Waste Building, the Independent Spent Fuel Storage Installation (ISFSI) located inside the protected area, and the Original Steam Generator Storage Facility located on the West side of the plant site, North of the main access road.

11 .2.4.2 Radioactive Waste Storage Radioactive material is stored in the Auxiliary Building and the Radioactive Waste Building located inside the protected area, and in the Old Warehouse located just West of the protected area.

11.2.5 Miscellaneous Radoacte MateHa ources To assure that leakage from byproduct, source, and special nuclear material radioactive sources does not exceed allowable limits, radioactive sources shall be leak tested for contamination.

The leakage test shall be capable of detecting the presence of 0.005 microcurie of radioactive material on the test sample. If the test reveals the presence of 0.005 microcurie or more of removable contamination, it shall immediately be withdrawn from use, decontaminated, and repaired, or be disposed of in accordance with Commission regulations.

Ingestion or inhalation of source material may give rise to total body or organ irradiation. This USAR Section assures that leakage from radioactive material sources does not exceed allowable limits. In the unlikely event that those quantities of radioactive byproduct materials of interest to this Section, which are exempt from leakage testing are ingested or inhaled, they represent less then one maximum permissible body burden for total body irradiation. The limits for all other sources (including alpha emitters) are based upon 10 CFR Part 70, Section 70.39(c) limits for plutonium.

USAR-1 1.2 Information Use Page 31 of 31 Radiation Protection and Monitoring Rev. .4 Those quantities of byproduct material that exceed the quantities listed in 10 CFR 30.71, Schedule B are to be leak tested in accordance with the schedule shown in Table 11.2-15 Leak Check Surveillance Requirements.

All other sources, including alpha emitters, containing greater than 0.1 microcurie are also to be leak tested in accordance with Table 11.2-15.

Table 11.2-15 Leak Check Surveillance Requirements Tests for leakage and/or contamination shall be performed by the licensee or by other persons specifically authorized by the Commission or an agreement State, as follows:

1. Each sealed source, except startup sources subject to core flux, containing radioactive material, other than Hydrogen 3, with a half-life greater than thirty days and in any form other than gas shall be tested for leakage and/or contamination at intervals of six months.

2. The periodic leak test required does not apply to sealed sources that are stored and not being used. The sources excepted from this test shall be tested for leakage prior to any use or transfer to another user unless they have been leak tested within six months prior to the date of use or transfer. In the absence of a certificate from a transferor indicating that a test has been made within six months prior to the transfer, sealed sources shall not be put into use until tested.

3. Startup sources shall be leak tested prior to and following any repair or maintenance and before being subjected to core flux.