ML051020434

Radionuclides for Consideration During Rancho Seco Site Characterization or Final Status Surveys

ML051020434
Person / Time
Site:	Rancho Seco
Issue date:	04/07/2005
From:	Redeker S Sacramento Municipal Utility District (SMUD)
To:	John Hickman Document Control Desk, NRC/FSME
References
+sispmjr200508, MPC&D 05-039 DTBD-04-001
Download: ML051020434 (228)

Text

SMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT 0 6201 S Street, RO. Box 15830, Sacramento CA 95852-1830, 1916) 452-3211 AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA MPC&D 05-039 April 7, 2005 U.S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington, DC 20555 Docket No. 50-312 Rancho Seco Nuclear Generating Station License No. DPR-54 RADIONUCLIDES FOR CONSIDERATION DURING RANCHO SECO SITE CHARACTERIZATION OR FINAL STATUS SURVEYS Attention: John Hickman As discussed in our February 16, 2005, meeting, we are providing you a copy of Decommissioning Technical Basis Document DTBD-04-001 "Radionuclides for Consideration during Rancho Seco Nuclear Generating Station Characterization or Final Status Surveys," Revision 2. Included as Attachment 1, this Decommissioning Technical Basis Document (DTBD) identifies the site-specific suite of radionuclides that could potentially be present in the Rancho Seco environs or as contamination on structural surfaces at the time of site characterization or during the performance of final status surveys.

In addition, Attachment 2 provides a copy of the Rancho Seco Historical Site Assessment (HSA), Revision 0, dated March 2004. Based on a review of various historical site records and correspondence, and written questionnaires and oral interviews with current and past employees, the HSA is a preliminary component of the Radiological Site Survey Investigation process for the Rancho Seco site.

We will submit additional DTBDs as they become available, in anticipation that the review of these documents will facilitate NRC review and approval of the Rancho Seco License Termination Plan to be submitted later this year.

xrynSSbL

J. Hickman MPC&D 05-039 Members of your staff with questions requiring additional information or clarification may contact Bob Jones at (916) 732-4843.

Sincerely, Steve Redeker Manager, Plant Closure and Decommissioning Attachments (2)

Cc Nv/ attachments: B.S. Mallett, NRC, Region IV

Attachment I Decommissioning Technical Basis Document DTBD-04-001 "Radionuclides for Consideration during Rancho Seco Nuclear Generating Station Characterization or Final Status Surveys,"

Revision 2

DTBD-04-001 Rancho Seco Nuclear Generating Station Revision No. 2 .

Decommissioning Technical Basis Document DPT 04-125

. RIC 2A.900 Radionuclides for Consideration During Rancho Seco Nuclear Generating Station Characterization or Final Status Surveys PREPARED BY: LEBrown Date-Author Date REVIEWED BY: . I- i-oS 0 Date REVIEWED BY: le d jH7 3 05 0ev er / hj .j Date 3/.23/S -

APPROVED BY: . 3 1-oSf Principal Decorni'M'66g Engineer Date

Rancho Seco Nuclear Generating Station Review and Approval of Decommissioning Technical Basis Documents

.DT1BD Number 0- 0°° \ en DTBD t ItU- C e \,Title RQ)J- Ch4-cLc4Ca C(' t r Technically Reviewed By 5C. k *e , Date I l1os Yes No N/A

1. Is the document format correct?

2. Are the assumptions made for the DTBD correct? 0 0

3. Has the technical development of material in the DTBD been appropriate? 0 . 0

• 4. Is there a clear boncise explanation and development of the position? 1 0 0

5. Are there clearly stated applicability and limitations of the document? 0l 10

6. Are the reference list and tables adequate? 0 E

7. Are the data and data collection methods.adequate? 0l .

S. Is this DTBD consistent with existing DTBDs? ' 0 0

9. Was an applicable and validated computer program used? 0l. 0

10. Was the choice of software appropriate?  % 0 0

11. Was the selection of input parameters proper? . . y

12. Was the configuration management of the software adequate? A' 0 0

13. Was the interpretation of the results proper? . O

14. Has an annual applicability review has been initiated by completion.

of Form ADM-256 (for-tracking only)? 0O For any item checked "No", provide an explanation below (reference attachments as needed).

DEC-014, Rev. 0 Page 1 of 1

DECOMMISSIONING TECHNICAL BASIS DOCUMENT NUMBER: DTBD-04-001 REVISION: 2 TITLE: Radionuclides for Consideration During Rancho Seco Nuclear PAGE 1 OF 6 Generating Station Characterization or Final Status Surveys 1.0 PURPOSE The purpose of this Decommissioning Technical Basis Document (DTBD) is to identify a site-specific suite of radionuclides that could potentially still be present in the Rancho Seco Nuclear Generating Station (RSNGS) environs or as contamination on structural surfaces at the time of characterization and final status survey (FSS) performance. This DTBD is considered to be a living document until site characterization is complete and will be periodically revised as new information becomes available. This revision narrows the theoretical suite of radionuclides generated in Revision 0 by evaluation of published abundance values of radionuclides as well as historical 10 CFR Part 61 radioactive waste characterizations.

2.0 DISCUSSION NUREG-1757, Volume 2, Consolidated NMSS Decommissioning Guidance, Characterization, Survey, and Determination of Radiological Criteria [Ref. 7.1] provides guidance to identify a suite of radionuclides that could be present at a power reactor.

Appendix 0 to NUREG-1757, Volume 2; Lessons Learned and Questions and Answers to Clarify License Termination Guidance and Plans, states:

'A unique radionuclide profile must be developed for each of the major types of materials expected to remain onsite .after remediation. A commercial light-waterpower reactor facility will likely require profiles for contaminated soil or sediments, surface contaniinated materials, and activated materials. The licensee must consider that activation products in steels and concretes vary with the constituents and operational history.

Concrete will also differ between facilities because of different trace elements. While one generic list cannot be developed that would be applicable to all power reactor licensees and types of contaminated materials, once radioactive decay has been considered to the time when final status surveys (FSSes) will be conducted, a set of radionuclides may be developed for surface contamination and for activated materials.... The licensee should confirm, by using characterization surveys and historical assessments, that the radionuclide lists developed are applicable to the facility arid appropriate for each medium. Technical considerations and limitations are discussed in: NUREG/CR-3474, "Long-Lived Activation Products in Reactor Materials" [Ref. 7.2]; NUREG-0130, "Technology,

'Safety and Cost of Decommissioning" [Ref. 7.3]; and NUREG/CR-4289, "Residual Radionuclide Contamination Within and Around Commercial Nuclear Power Plants" [Ref. 7.4]".

3.0 DEFINITIONS None January5, 2005*

DTBD-04-OO1 R2.doc MDTB4 001R2.doc January 5, 2005 -

DECOMMISSIONING TECHNICAL BASIS DOCUMENT NUMBER: DTBD-04-001 REVISION: 2 TITLE: Radionuclides for Cofisideration During Rancho Seco Nuclear PAGE 2 OF 6 Generating Station Characterization or Final Status Surveys 4.0 TECHNICAL POSITION The theoretical suite of radionuclides that could potentially still be present at RSNGS (based upon the guidance contained in NUREGICR-3474, NUREG-0130 and NUREG/CR-4289)'is provided as Attachment 8.1 along with their half-lives and mode of decay. All gamma spectrometry analyses that .are performed onsite for characterization or FSS surveys should include the detectable gamma emitters listed in Attachment 8.1 in the gamma spectrometry libraries for analysis. At least initially and periodically thereafter, characterization or FSS samples sent to an offsite laboratory for analysis .

should be analyzed for the narrowed suite of radionuclides listed in Attachment 8.6.

5.0 LIMITATIONS The suite of radionuclides listed in Attachment 8.1 is a theoretical list based on NUREG/CR-3474, NUREG-0130 and NUREG/CR-4289 and should not be used as a site-specific suite for developing derived concentration guideline levels (DCGLs). The suite of radionuclides listed in Attachment 8.6 is a site-specific suite of radionuclides for developing site-specific DCGLs.

6.0 TECHNICAL BASES Development of a Theoretical Suite of Radionuclides Development of the suite of radionuclides listed in Attachment 8.1 began with NUREGICR-3474. This NUREG assessed the problems posed to reactor decommissioning by long-lived activation products in reactor construction materials.

Samples of stainless steel, vessel steel, concrete and concrete ingredients were analyzed for up to 52 elements in order to develop a database of activatable major, minor and trace elements. The list of radionuclides was developed by combining those radionuclides listed in Table 5.6, "Activation of PWR Bioshield (Ci/gm) Average Rebar 30 EFPY at Core Axial Midplane," Table 5.13, uActivity Inventory of PWR Internals at Shutdown (Total Ci)," and Table 5.15, "Inventories of PWR and BWR Vessel Walls at Shutdown (Total Ci)." Only radionuclides with half-lives of two or more years were included on the list.. Radionuclides with half-lives less than two years would not be expected to still be observed since two years or less represents seven or more half-lives since final shutdown of the RSNGS reactor.

Second, radionuclides with half-lives of two or more years identified in NUREG/CR-4289 as being present in PWRs were .compared with the list generated above.

NUREG/CR-4289 investigated residual radionuclide concentrations, distributions and inventories at seven nuclear power plants (four shutdown and three operating, including RSNGS) to provide a database for use in formulating policies, strategies and guidelines for the eventual decommissioning of retired nuclear power plants. This study addressed radionuclides (both activation and fission products) transported from the reactor pressure vessel and deposited in all other contaminated systems of each nuclear plant.

Emphasis was placed on measuring the long-lived radionuclides that are of special DTBD-04-001 R2.doc January 5,2005

DECOMMISSIONING TECHNICAL BASIS DOCUMENT NUMBER: DTBD-04-001 REVISION: 2 TITLE: Radionuclides for Consideration During Rancho Seco Nuclear PAGE 3 OF 6 Generating Station Characterization or Final Status Surveys concern from a low-level waste management standpoint. The study resulting in NUREGICR-4289 was a companion study to the study that resulted in NUREG/CR-3474. Any radionuclides identified in NUREG/CR-4289 but not in NUREG/CR-3474, were added to the above list.

Third, radionuclides with half-lives of two or more years identified in Volume I of NUREG/CR-0130 as being present in PWRs were compared with the list generated above. These radionuclides were identified in Table 7.3-9, "Reactor Coolant Radionuclide Concentrations (12) in an Operating PWR," Table 7.3-10, "Radioactive Surface Contamination in the Reference PWR Resulting from Accumulated Coolant Leakage in an Ion Exchanger Vault (Fractional Activity Normalized at Reactor Shutdown)," and Table'7.3-1 1, "Isotopic Composition of Accumulated Radioactive Surface Contamination in the Reference PWR (Renormalized for Each Decay Time)."

Any radionuclides identified in NUREG/CR-0130 but not in either NUREG/CR-3474 or NUREG/CR-4289, were added to the above list.

Finally, an ORIGEN computer code run was used to determine if there were additional radionuclides that should be added to the above list. The ORIGEN code run was based on Cycle 4 through 7 irradiation of selected batch 6 fuel assemblies with a decay period of 13.64 years from shutdown. This resulted in the addition of Pm-147, Pu-241, Am-243 and Cm-243 to the list, which is provided as Attachment 8.1.

Although not identified in the above regulatory guidance, U-234, U-235, U-236 and U-238 were added because they were identified in NCRP Report No. 58 [Ref. 7.5],

Table 16 as being present in power reactor fuel.

Discounting Insignificant Radionuclides Activation Product Considerations Since Attachment 8.1 includes trace-elements that would not likely be found at RSNGS due to their low abundance, an evaluation of activation product radionuclides that may be discounted as being of potential importance was performed. The total inventory for each radionuclide was determined from activity inventories provided in Table 5.13 and

• Table 5.15 of NUREG/CR-3474. From this information, the percentage of total inventory for each radionuclide was calculated. The results of this evaluation are provided in Attachment 8.2 of this.DTBD:

Spent Fuel Radionuclide Considerations The ORIGEN computer code, run also contains -trace radionuclides that would not likely be found at RSNGS due to their low abundance. The total radionuclide inventory was determined from.the run as well as relative contribution from each radionuclide. The results of this evaluation are included in Attachment 8.3.

January 5, 2005 DTBD-04-QO1 R2.doc DTBD-04-001 R2.doc January 5, 2005

DECOMMISSIONING TECHNICAL BASIS DOCUMENT NUMBER: DTBD-04-001 REVISION: 2 TITLE: Radionuclides for Consideration During Rancho Seco Nuclear PAGE 4 OF 6 Generating Station Characterization or Final Status Surveys PotentialDiscountedDose Considerations Based on the above evaluation, it was determined that individual radionuclides which contributed less than 0.1 percent of the total activity in both Attachments 8.2 and 8.3 could be discounted from the list of Attachment 8.1 identified radionuclides providing that potential.dose contributed by the sum of the radionuclides discounted does not exceed one percent of the total calculated dose.

The radionuclides that meet the criteria of contributing less than 0.1 percent-of the total activity include:

CI-36 Ar-39 Ca-41 Mn-53 Se-79 Kr-81 Kr-85 Zr-93 Mo-93 Sn-121m 1-129 Ba-133 Cs-135 Pm-145 Sm-146 Sm-151 Tb-158 Ho-166m Hf-178m Pb-205 U-233 Am-243 Cm-243 Several additional radionuclides meet the criteria of contributing less than 0.1 percent of the total activity but cannot be discounted because they have. other methods of production in addition to activation of reactor components and have been observed in 10 CFR Part 61 waste stream analyses or in site characterization samples. These radionuclides include H-3, C-14, Nb-94, Ag-108m, Eu-152, and Pu-239.

In order to evaluate compliance with the dose criteria fordiscounted radionuclides, the NRC developed computer code DandD, Version 2.1.0 was used to calculate doses for both residential and occupancy scenarios. The DandD code was used with the NRC determined default parameters to represent a conservative screening tool. Input concentrations for each radionuclide used in the residential scenario were their percent of total activity input as concentration in pCilg. Input concentrations-for each radionuclide used in the occupancy scenario were 1,000 times their percent of total activity input as surface contamination in dpm/100 cm 2 . DandD does not support the following radionuclides and could not calculate their dose contribution:

Ar-39 Mn-53 Kr-81 Kr-85 Ag-108m Ba-133 Pm-145 Sm-146 Tb-158 Hf-178m Pb-205 Therefore doses could.be calculated for only the following discounted radionuclides:

Cl-36 Ca-41 Se-79 Zr-93 Mo-93 Sn-121m 1-129 Cs-135 Sm-151 . Ho-166m U-233 Am-243 Cm-243 The calculated total dose from discounted NUREG radionuclides represents only 3.73E-02 percent and dose from discounted ORIGEN radionuclides represents only 4.27E-02 DTBD04-01R2.oc anuay 5,200 MD-04-001R2.doc January 5, 2005

DECOMMISSIONING TECHNICAL BASIS DOCUMENT NUMBER: DTBD-04-001 REVISION: 2 TITLE: Radionuclides for Consideration During Rancho Seco Nuclear PAGE 5 OF 6 Generating Station Characterization or Final Status Surveys percent of the total calculated dose for the residential scenario. The calculated total dose from discounted NUREG radionuclides represents only 1.99E-03 percent and dose from discounted ORIGEN radionuclides represents only 5.53E-01 percent for the occupancy scenario. Therefore, it is appropriate to discount these radionuclides.

Summary reports for the DandD calculations are included in Attachment 8.4.

The activity represented by the radionuclides not supported by the DandD code is calculated to be only 4.23E-03 percent of the total activity presented in NUREG/CR-3474. Of these radionuclides, Ar-39, Kr-81 and Kr-85 are noble gases and it is highly unlikely that theywould still be present in soil and on structural surfaces.

Therefore, it is appropriate to discount Ar-39, Kr-81 and Kr-85.

Potential dose contribution from the remaining radionuclides not supported by the DandD code was evaluated .by comparison of the inhalation and ingestion exposure-to-dose conversion factors (DCFs) contained in Federal Guidance Report No.11, Limiting Values of Radionuclide Intake and Air Concentration and.Dose Conversion Factors for Inhalation, Submersion, and Ingestion [Ref. 7.6]. Weighted DCFs were calculated for each discounted radionuclide and summed for both inhalation and ingestion DCFs.

These totals were then compared to the sum of the weighted DCFs for the two most abundant radionuclides, Co-60 and Ni-63. This resulted in a total of 5.36E-03 percent for inhalation DCFs and 1.25E-03 percent for ingestion DCFs. The calculations to demonstrate these results are provided in Attachment 8.5. Therefore, it is appropriate to discount all of the radionuclides not supported by the DandD code.

Although originally included in the list of theoretical radionuclides, the naturally occurring radionuclides K-40, U-234, U-235, U-236 and U-238 have not been detected in

.characterization survey samples at concentrations distinguishable from naturally occurring concentrations. Therefore, these radionuclides have been discounted from any further consideration.

Waste Stream Evaluation Considerations Radioactive waste streams are periodically sampled and analyzed at RSNGS.

Analyses are performed for radionuclides listed in 10 CFR 61.55 Tables 1 and 2 as well as other supplementary radionuclides on a select basis. The potential radionuclides identified for discounting as described above were compared with the Rancho Seco 2003 Waste Stream Evaluation [Ref. 7.7]. None of these radionuclides were identified as being present at RSNGS. However, an additional radionuclide, Pu-242, had been' identified by waste stream analysis and was added to the site-specific suite of radionuclides.

The radionuclides remaining on the list in Attachment 8.1 constitute the Site-Specific Suite of Radionuclides for Use at RSNGS, which is provided as Attachment 8.6.

January 5, 2005 DTBD-04-001R2.doc DTBD-04-001R.2.doc January 5, 2005

DECOMMISSIONING TECHNICAL BASIS DOCUMENT NUMBER: DTBD-04-bo1 REVISION: 2 TITLE: Radionuclides for Consideration During Rancho Seco Nuclear PAGE 6 OF 6 Generating Station Characterization or Final Status Surveys

7.0 REFERENCES

7.1 NUREG-1757, Volume 2, Consolidated NMSS Decommissioning Guidance, Characterization, Survey, and Determination of Radiological Criteria, September 2003 7.2 NUREGICR-3474, Long-Lived Activation Products in ReactorMaterials, August 1984 7.3 NUREG-0130, Technology, Safety and Cost of Decommissioning, June 1978 7;4 NUREGICR-4289, Residual Radionuclide Contamination Within and Around Commercial Nuclear Power Plants, February 1986 7.5 NCRP Report No. 58, A Handbook of Radioactivity Measurements Procedures, February 1, 1985 7.6 EPA-520/1-88-020, Federal Guidance Report No.11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors For Inhalation, Submersion, and Ingestion, September 1988 7.7 Rancho Seco 2003 Waste Stream Evaluation 8.0 ATTACHMENTS 8.1 Radionuclides Potentially Present at RSNGS 8.2 Evaluation of NUREG/CR-3474 Total Activity Fractions 8.3 Evaluation of ORIGEN Computer Code Run Total Activity Fractions 8.4 DandD Calculation Summary Reports 8.5 Dose Considerations for Radionuclides Not Supported By DandD 8.6 Site-Specific Suite of Radionuclides for Use at RSNGS 9.0 RESPONSIBLE INDIVIDUAL Leon E. Brown January 5, 2005 DTBD-04-001R2.doc MTD-04-001R2.doc January 5, 2005

'Attachment 8.1 Radionuclides Potentially Present at RSNGS March 8. 2004 DTBD-04-OO1 R2.doc DTbD-04-001 R2.doc March B. 2004 -

Radionuclides Potentially Present at.RSNGS .i Half Life Decay Half Life Decay Radionuclide (Years) Mode Radionuclide (Years) Mode

• H-3 1.23E+01 L *Cs135 2.30E+06

• C-14 5.73E+03. Cs-137 3.02E+01 p Na-22 2.60E+00 , Pm-145 1.77E+01 Y

• CI-36 3.01 E+05 P"Y *Sm-146 1.OOE+08 a

• Ar-39 2.69E+02. *Pm-147 2.62E+00 .Lr K-40 1.28E+09 O" *Sm-151 9.30E+01 ,

• Ca-41 1.03E+05 ,r Eu-152 1.36E+01 ,

• Mn-53 3.70E+06 Y Eu-154 8.80E+00 O,

• Fe-55 2.70E+00 Y Eu-1 55 4.96E+00 Pr

• Ni-59 7.50E+04 7 Tb-158 *1.50E+02 L Co-60 5.27E+O0 Y Ho-166m 1.20E+03 j

• Nii-63 1.OOE+02 *Hf-178m 3.OOE+01 IT, y

. *Se-79 6.50E+04 L *Pb-205 1.51 E+07 r

-Kr-8-1-- 2.1OE+05 - *U-233 1.59E+05. a,,y Kr-85 1.07E+01 VY *U-234 2.45E+05 a, y

• Sr-90 2.86E+01 L U-235 7.04E+08 a, y

• Zr-93 1.53E+06 L *U-236 2.34E+07 a, Y

• Mo-93 3.50E+00 *U-238 4.47E+09 a, r

• Nb-93m 1.46E+01 Y *Np-237 2.14E+6 a, y Nb-94 2.03E+04 r *Pu-238 8.78E+01 a, Y

• Tc-99 2.13E+05 , *Pu-239 2.41E+04 a, y Ag-108m 1.27E+02 r *Pu-240 6.60E+03 a, y

• Sn-121 m 5.00E+00 *Pu-241 1.44E+01 I Sb-1 25 2.77E+OO V Am-241 4.32E+02 a,

. *1-129 1.57E+07 . Am-243 7.38E+03 a, Ba-133 1.05E+01 r Cm-243 2.85E+01 a, Cs-134 2.06E+00 , *Cm-244 1.81 E+01 a, r a-Alpha decay p-Beta decay p+- Positron decay y -Gamma decay IT - Isomeric transition

• Hard to Detect Nuclides (HTDN - radionuclides not readily detected by gamma spectroscopy, e.g., Ni-63 or Cm-244, and requiring offsite, contract laboratory analysis)

DTBD-04-001 R2.doc Page 8.1 - I DTBD-04-001, Attachment 8.1

Attachment 8.2 Evaluation of NUREGICR-3474 Total Activity Fractions March 8, 2004 DTBD-04-OO1 R2.doc DTB D-04-001lR2.doc March 8, 2004

Evaluation of NUREG/CR-3474 Total Activity Fractions Activit- Ci Thermal Vessel Radionuclide Shroud Core Barrel Pads Cladding Vessel Walls Total Activity Percent Total H-3 3.48E+01 6.14E+01 6.71 E+00 1.65E-01 6.37E+00 1.1OE+02 2.88E-02 C-14 2.32E+02 5.09E+01 3.69E+00 8.68E-02 2.79E-01 2.86E+02 7.54E-02 CI-36 4.70E+00 1.10E+00 8.10E-02 1.85E-03 1.70E-02 5.90E+00 1.55E.03 Ar-39 1.34E+00 1.34E-01 2.87E-03 3.27E-04 3.54E-02 1.51 E+00 3.98E-04 Ca-41 4.30E-02 9.55E-03 6.97E-04 1.60E-05 2.32E-01 2.85E-01 7.51 E-05 Mn-53 3.00E-02 3.60E-03 8.1OE-05 9.00E-06 4.06E-04 3.41 E-02 .8.97E-06 Fe-55 2.48E+04 5.21E+03 3.82E+02 9.08E+00 2.38E+02 3.07E+04 8.07E+00 Ni-59 1.02E+03 3.24E+02 2.43E+01 5.70E-01 .6.55E-01 1.37E+03 3.60E-01 Co-60 1.31 E+05 2.59E+04 1.75E+03 4.72E+01 1:27E+02 1.59E+05 4.18E+01 Ni-63 1.49E+05 3.64E+04 2.74E+03 6.34E+01 7.08E+01 1.89E+05 4.96E+01 Se-79 5.70E-03 7.85E-04 3.80E-05 2.00E-06 7.90E-07 6.52E-03. 1.72E-06 Kr-81 7.1 OE-03 5.80E-05 2.84E-07 2.77E-09 .2.80E-08 7.16E-03 1.88E-06 Kr-85 2.71 E+00 1.11 E-01 2.64E-03 5.89E-05 1.03E-02 2.83E+00 7.45E-04 Sr-90 1.24E+01 5.66E-02 1.40E-02 3.17E-04 5.53E-04 1.25E+01 3.28E-03 Zr-93 1.OOE-03 6.70E-05 2:OOE-05 1.13E-07 3.00E-06 1.09E-03 2.87E-07 Mo-93 3.00E-01 . 2.30E-02 5.87E-04 5.18E-05 7.94E-03 3.32E-01 8.74E-05 Nb-94 3.70E+00 4.95E-01 2.40E-02 9.89E-04 1.1 OE-02 4.23E+00 1.11 E-03 Tc-99 1.20E+00 1.40E-01 3.40E-03 3.17E-04 5.50E-02 1.40E+00 3.68E-04 Ag-108m 8.49E-01 . 1.37E-01 8.13E-03 2.66E-04 1.05E-02 1.01E+00 2.65E-04 Sn-121m 4.17E-03 5.50E-04 1.31 E-05 1.422-06 3.70E-05 4.77E-03 1.26E-06 1-129 6.00E-06 2.39E-07 6.40E-09 1.45E-10 2.50E-10 6.25E-06 1.64E-09 Ba-133 8.31 E+00 1.57E+00 1.11E-01 2.89E-04 3.08E-02 1.OOE+01 2.64E-03 Cs-134 . 2.38E-01 6.22E-02 5.02E-03 1. IOE-04 9.88E-04 3.06E-01 8.06E-05 Cs-135 4.00E-04 1.50E-05 4.1 OE-07 1.03E-08 3.60E-08 4.15E-04 1.09E-07 Cs-137 1.28E+01 5.74E-01 1.49E-02 3.75E-04 2.03E-03 *1.34E+01 3.53E-03 Pm-145 4.16E-03 1.24E-03 9.25E-05 2.08E-06 6.24E-06 5.50E-03 1.45E-06 Sm-146 . 9.30E-10 2.21 E-10 5.50E-12 5.70E-13 3.00E-12 1.16E-09 3.05E-13 Sm-151 *3.79E-02 6.78E-02 2.03E-02 6.87E-04 3.26E-03 1.30E-01 3.42E-05 Eu-152 O.OOE+00 6.26E-01 5.58E-01 4.04E-02 1.26E+00 2.48E+00 6.54E-04 Eu-154 1.32E+00 2.59E+00 1.27E-01 .4.07E-03 2.57E-01 4.30E+00 1.13E-03 Eu-155 3.26E-01 2.05E-01 3.48E-03 5.93E-05 5.23E-04 5.36E-01 1.41 E-04 .

Tb-158 1.66E-02 2.68E-03 6.38E-04 6.47E-06 .2.03E-04 2.02E-02 5.31 E-06 Ho-166m 1.49E+00 2.19E-01 8.81 E-03 3.96E-04 1.98E-02 1.73E+00 .4.56E-04 Hf-178m 2.70E-01 3.22E-01 1.69E-02 1.08E-03 7.77E-03 6.18E-01 1.63E-04 Pb-205 1.70E-05 2.00E-06 1.05E-07 4.50E-09 2.00E-06 2.11 E-05 5.56E-09 U-233 3.30E-03 1.70E-03 8.90E-03 4.00E-06 5.002-05 1.40E-02 3.67E-06 Pu-239 6.50E-02 3.92E-02 1.OOE-03 . 1.27E-04 1.1OE-03 1.06E-01 2.80E-05 Totals 3.80E+05 1.00E+02 Total percent of activity discounted 6.52E-03 Radionuclides to be discounted appear in bold font.

DTD0-O R2dcPg . TD0-0,Atcmn .

DTBD-04-ObIR2.doc Page 8.2 - 1 DTBD-04-001, Attachment 8.2

Attachment 8.3 Evaluation of ORIGEN Computer Code Run Total Activity Fractions December 20, 2004 DTBD-04-001R2.doc DTBD-04-001 R2.doc December 20, 2004

Evaluation of ORIGEN Computer Code Total Activity Fractions Radionuclide Activity (Ci) Percent Total Radionuclide Activity (Ci) Percent Total H-3 9.37E+01 4.30E-02 Eu-154 3.31 E+03 1.52E+OO Fe-55 1.14E+03 5.25E-01 Eu-155 1.48E+03 6.81 E-01 Co-60 2.20E+03 1.01E+OO Pu-238 1.75E+03 8.01E-01 Ni-59 3.20E+00 1.47E-03 Pu-239 1.46E+02 6.68E-02 Ni-63 4.12E+02 1.89E-01 Pu-240 2.51 E+02 1.15E-01 Sr-90 5.62E+04 1.29E+01 Pu-241 4.59E+04 2. 1E+01 Sb-125 1.33E+03 6.11E-01 Am-241 5.51 E+02 2.53E-01 Cs-134 1.03E+04 4.71 E+OO Am-243 8.78E+00 4.03E-03 Cs-137 7.96E+04 3.66E+01 Cm-243 1.97E+01 9.04E-03 Pm-147 1.22E+04 6.60E+OO Cm-244 8.40E+02 .3.86E-01 Sm-151 1.55E+02 7.1 OE-02 .

Totals 2..18E+05 1.OOE+OO Total percent of activity discounted 8.41 E-02 Radionuclides to be discounted appear In bold font.

Page 8.3-1 DTBD-04-OO1, Attachment 8.3 DTBD-04-QOI R2.doc DTBD-04-OOlR2.doc Page 8.3 - I DTBD-04-001, Attachment 8.3

Attachment 8.4 DandD Calculation Summary Reports January 5, 2005 DTBD-04-OO1 R2.doc DTBD-04-001lR2.doc January 5, 2005

DandD Residential Scenario S.-

5- 0 DandD Version: 2.1.0 Run Date/Time: 1/5/2005 10:33:43 AM Site Name: RSNGS

==

Description:==

Analysis of potential dose from radionuclides discounted from those identified in NUREG/CR-3474.

FileName:C:\Documents and Settings\lbrown\My Documents\Projects\Rancho Seco\DTBDs\04-001R2\04-OOlSoil.mcd Options:

Implicit progeny doses NOT included with explicit parent doses Nuclide concentrations are distributed among all progeny Number of simulations: 229 Seed for Random Generation: 8718721 Averages used for behavioral type parameters External Pathway is ON Inhalation Pathway is ON Secondary Ingestion Pathway is ON Agricultural Pathway is ON Drinking Water Pathway is ON Irrigation Pathway is ON Surface Water Pathway is ON Initial Activities:

2) Distribution NuclidContamnation (

.36CI UNLIMITED CONSTANT(pCi/g)

Justification for concentration: Percent total Value 1.55E-03 41Ca UNLIMITED llCONSTANT(pCilg)

Justification for concentration: Percent total l Value 7.51E-05 l79Se . UNLIMITED llCONSTANT(pCi/g)

Justification for concentration: Percent total IF)Value 1.72E-06 193Zr . UNLIMITED . CONSTANT(pCi/g). 1 Justification for concentration: Percent total l Value 2.87E-07 I121mSn llUNLIMTED llCONSTANT(pCiug)*

DTBD-04-001 R2.doc Page 8.4 - 1 DTBD-04-001; Attachment 8.4

Justification for concentration: Percent total Value 1.26E-06 129i ITED CONSTANTpCig)

Justification for concentration: Percent total Value 1.64E-09 1135CS  ::MNL ITED lCONSTANT(pCg)

Justification for concentration: Percent total Value 1.09E-07 l151Sm jjUNLMITED .

CCNSTANTpCig)

[FcNTNl IF I

Justification for concentration: Percent total [Value 3.42E-05 I 166mHo llUNLMITED ICONSTANT Cig) lJustification for concentration: Percent total Value 4.56E-04 1233UUNL TED lCONSTANT(pCilg) lJustification for concentration: Percent total l Value 3.67E-06 l93MO -1UNLp MTED CONSTANT(pCg)

.Justification for concentration: Percent total Value 8.74E-05 Site Specific Parameters:

General Parameters:

None Element Dependant Parameters None Correlation Coefficients:

None Summary Results:

90.00% of the 229 calculated TEDE values are < 1.03E-01 mrem/year.

The 95 % Confidence Interval for the .0.9 quantile value of TEDE is 9.01E-02 to - .17E-01 mrem/year Page 8.4-2 DTBD-04-OO1, Attachment 8.4 DTBD-04001 R2.doc DTBD-04-001lR2.doc Page 8.4 - 2 DTBD-04-001, Attachment 8.4

DandD Residential Scenario DandD Version: 2.1.0 Run Date/Time: 1/5/2005 10:49:40 AM Site Name: RSNGS

==

Description:==

Analysis of potential dose from radionuclides not discounted from those identified in NUREG/CR-3474.

FileName:C:\Documents and Settingslbrown\My Documents\Projects\Rancho Seco\DTBDs\04-001R2\04-OOlSoil2.mcd Options:

Implicit progeny doses NOT included with explicit parent doses Nuclide concentrations are distributed among all progeny Number of simulations: 189 Seed for Random Generation: 8718721 Averages used for behavioral type parameters External Pathway is ON Inhalation Pathway is ON Secondary Ingestion Pathway is ON Agricultural Pathway is ON Drinking Water Pathway is ON Irrigation Pathway is ON Surface Water Pathway is ON Initial Activities:

Nuclide A of Distribution lContaminatn(i)

[3H XIUNLIMITED CONSTANT(pCi/g) l

[Justification for concentration: Percent total ll Value 2.88E-02 ]

114C . EIMITED CQNSTANT(pCi/g) 1 IJustification for concentration: Percent total Value 7.54E-02 I NLIMEITED U55Fe CONSTANT(PCi/g)

.Justification for concentration: Percent total Value 8.07E+OO j59Ni llUNLIMITED llCONSTANT(pCig)1 Justification for concentration: Percent total Value 3.60E-01 160CO IMITED FCONSTANT(pCiug) l DTBD-04-00IR2.doc Page 8.4 - 3 DTBD-04-001, Attachment 8.4

• stification for concentration: Percent total Value 4.18E+O1.

..3N .i I I I UNLIMITED FCONSTANT(pCi'g)

`mstification for concentration: Percent total Value 4.96E+01 E ZIr UNLIMITED ICONSTANT(pCilg) u.stification for concentration: Percent total Value 3 .28E-03

-9Tc 1UEMTED llCONSTANT(pCilg) I astification for concentration: Percent total Vlalue 3.68E-04

.134CS ZI IMITED CONSTANT(PCi/g)

.'S:stification for concentration: Percent total Value* 8.06E-05 iJpecific Parameters:

l Parameters

-t Dependant Parameters

.Iation Coefficients:

niary Results:

of the 189 calculated TEDE values are.< 2.76E+02 mrem/year.

% Confidence Interval for the 0.9 quantile value of TEDE is 2.74E+02 to

-02 mrem/year Page 8.4-4 DTBD-04-OO1, Attachment 8.4 AR2.doc

.R2.d 1 oc Page 8.4 - 4 DTBD-04-001, Attachment 8.4

DandD Residential Scenario DandD Version: 2.1.0 Run Date/Time: 3/4/2004 9:27:51 AM Site Name: RSNGS

Description:

Analysis of potential dose from radionuclides not discounted from those identified in NUREG/CR-3474.

FileName: C:\Documents and Settings\lbrown\My D6cuments\Projects\Rancho Seco\DTBDs\04-001R2\04-00lSoil3.mcd Options:

Implicit progeny doses NOT included with explicit parent doses

'Nuclide concentrations are distributed among all progeny Number of simulations: 156 Seed for Random Generation: 8718721 Averages used for behavioral type parameters External Pathway is ON Inhalation Pathway is ON Secondary Ingestion Pathway -isON Agricultural Pathway is ON Drinking Water Pathway is ON Irrigation Pathway is ON Surface Water Pathway is ON Initial Activities:

Nuclide Contamination (m2) Distribution 1137CS UNLIMITED lCONSTANT(pCiug)

J)ustification for concentration: Percent total. Value 3.53E-03 152Eu.I lUNLIMITED lCONSTANT(pCilg)

[justification for concentration: Percent total l Value 6.54E-04

[154Eu UNLMTED CONSTANT(pCi/g) .

Justification for concentration: Percent total l Value 1.13E-03

.155Eu UNLIMITED llCONSTANT(pCiug)

Justification for concentration- Percent total llValue 1.41E-04 239Pu UNLIMITED CONSTANT(pCilg)

DTBD-04001R2.doc~~Pae8_5DB-40,Atcmn

~ ~ ~ ~ .

. MD-04-001R2.doc Page 8.4 - 5 .DTBD-04-001, Attachment 8.4

Justification for concentration: Percent total IValue 2.80E-05 I 94Nb UNZIIMITED [CONSTA T(pCi/g)

)Justification for concentration: Percent total Il Value 1.1 E-03 Site Specific Parameters:

General Parameters:

None Element Dependant Parameters None Correlation Coefficients:

None Summary Results:

90.00% of the 156 calculated TEDE values are

• 1.82E-02 mrem/year.

The 95 % Confidence Interval for the 0.9 quafitile value of TEDE is 1.80E-02 to 1.86E-

.02 mrem/year Page 8.4-6 DTBD-04-OO1, Attachment 8.4 DTBD-04-001 R2.doc DTBD-04-OO1 R2.doc Page 8.4 - 6 DTBD-04-001, Attachment 8.4

DandD Residential Scenario I.

DandD.Version: 2.1;0 Run Date/Time: 12/16/2004 4:08:43 PM Site Name: RSNGS

==

Description:==

Analysis of potential dose from radionuclides discounted from those identified in the ORIGEN computer code run for spent fuel.

FileName:C:\Documents and Settings\lbrown\My Documents\Projects\Rancho Seco\DTBDs\04-001R2\04-00lSoil4.mcd Options:

Implicit progeny doses NOT included with explicit parent doses Nuclide concentrations are distributed among all progeny Number of simulations: 142 Seed for Random Generation: 8718721 Averages used for behavioral type parameters External Pathway is ON Inhalation Pathway is ON Secondary Ingestion Pathway is ON' Agricultural Pathway is ON Drinking Water Pathway is ON Irrigation Pathway is ON Surface Water Pathway is ON Initial Activities:.

F.

Nuclide

>L IM Area of Contamination (m )

M151Sm TED Distribution CONSTANT(pCi/g)

IF:Value 7.10E-02 Justification for concentration: Percent total l243Am . U IMITED [CONSTANT(pCilg) I Justification for concentration:-Percent total ll[Value 4.03E-03 1243Cm ll]UNLIMITED llCONSTANT(pCi/g) ]

Justification for concentration: Percent total Value 9.04E-03 Site Specific Parameters:

8.4-7 Page 8.4 - 7 DTBD-04-OO1, Attachment 8.4 R2.doc MD-04-001R2.doc DTBD-04-OO1 Page DTBD-04-001,,,kttachment 8.4

General Parameters:

None Element Dependant Parameters None Correlation Coefficients:

None Summary Results:

90.00% of the 142 calculated TEDE values are < 1.24E-01 mrem/year.

The 95 % Confidence Interval for the 0.9 quantile value of TEDE is 1.17E-01 to 1.35E-01 mrem/year ag .48 TD-4-O, ttcmet .

DTD-400R2dc DTBD-04-00IR2.doc Page 8.4 - 8 DTBD-04-001, Attachment 8.4

DandD Residential Scenario I..

• • AA DandD Version: 2.1.0 Run Date/Time: 12/16/2004 4:40:07 PM Site Name: RSNGS

Description:

Analysis of potential dose from radionuclides not discounted from those identified in the ORIGEN computer code run for spent fuel.

FileName:CA\Documents and Settings\lbrown\My Documents\Proj ects\Rancho Seco\DTBDs\04-001R2\04-O0lSoil5.mcd Options:

Implicit progeny doses NOT included with explicit parent doses Nuclide concentrations are distributed among all progeny Number of simulations: 116 Seed for Random Generation: 8718721 Averages used for behavioral type parameters External Pathway is ON Inhalation Pathway is ON Secondary Ingestion Pathway is ON Agricultural Pathway is ON Drinking Water Pathway is ON Irrigation Pathway. is ON Surface Water Pathwa,' is ON Initial Activities:

Nuclide of () 2Area Distribution Contamination (m )

13H lIUNLIMTED CONSTANT(pCi/g)

Justification for concentration: Percent total Value 4.36E-02 155Fe = ]UNLFLMTED CQNSTANT&jCi/g)

Justification for concentration: Percent total Value 5.25E-01 160CO lUNLAMTED PCONSTANT(paCg) 1 Jutfcto or concentration: Percent total Value 1.01E+00 159Ni lUL TE CONSTANT@pC;/g) .1

[Justification for concentration: Percent total Value 1.47E-03 163Ni llUNLEMTED ilCONSTANT(pCilg)

• MDTB04-001R2.doc; *Page 8.4 - 9 DTBD-04-001, Attachment 8.4

Tustification for concentration: Percent total Value 1.89E-01 19OSr l-NlrMMTED [CONSTANT(pCitg),

Justification for concentration: Percent total Value 1.29E+01 125Sb UNLIMITED FCONSTANT(pCi/g) I Justification for-concentration: Percent total Value 6.1lE-01 1134Cs I y ~MTED lCONSTANT(pCi/g) 1 IJustification for concentration: Percent total Value . 4.71E+00 1137CsUNLIMITED llCONSTNTpCi/g)

Justification for concentration: Percent total Value 3.66E+01 1147Pm I UNLIMITED lCONSTANT(pCi/g)  ;

Justification for concentration: Percent total . Value 5,60E+00 Site Specific Parameters:

General Parameters:

• None Element Dependant Parameters None Correlation Coefficients:

None

.Summary Results:'

90.00% of the 116 calculated TEDE values are < 2.65E+02 mrem/year.

The 95 % Confidence Interval for the 0.9 quantile value of TEDE is 2.31E+02 to 3.19E+02 mrem/year DTD04 _1R.o Pe841.DB ,Atcmn .

DTBD-04-OOlR2.doc Page 8.4 DTBD-04-001, Attachment 8.4

toP¢ Racer DandD Residential Scenario DandD Version: 2.1.0 Run Date/Time: 12/16/2004 4:58:56 PM Site Name: RSNGS

==

Description:==

Analysis of potential dose from radionuclides not discounted from those identified in the ORIGEN computer code run for spent fuel.

FileName:C:\Documents and Settings\lbrown\My Documents\Projects\Rancho Seco\DTBDs\04-001R2\04-001 Soil6;mcd Options:

Implicit progeny doses NOT included with explicit parent doses Nuclide concentrations are distributed among all progeny Number of simulations: 142 Seed for Random Generation: 8718721 Averages used for behavioral type parameters*

External Pathway is ON Inhalation-Pathway is ON Secondary Ingestion Pathway is ON Agricultural Pathway is ON Drinking Water Pathway is ON Irrigation Pathway is ON Surface Water Pathway is ON Initial Activities:

Nuclide Area of 2) Distribution Contamination (mn

[154Eu INLIMITED IFCONSTANT(pCi/g) .

Justification for concentiation: Percent total Value 1.52E+OO 155Eu I UNLIMITED CONSTANT(pCiug)

Justification for concentration: Percent total Value 6.81E-01 238Pu II UNLIMITED CONSTANT(pCifg) I Justification for concentration: Percent total lValue 8.01E-01 239Pu lUNLIMITED llCONSTANT(pCi/g) I Justification forconcentration: Percent total Value 6.68E-02 24 0P LUNLIMITED = CONSTANT(pCifg)

DTD0-O 2dcPg .4-1DB-0 Atahmn 8.4 DTBD-04-00IR2.doc Page 8.4 - 11 DTBD-04-001, Attachment 8.4

'Justification for concentration: Percent total Value 1.15E-01

. 241Pu . l IMITED FCONSTAN1T(pCig)

'Justification for concentration: Percent total Value 2.1 lE+01 41AmMTED ICONSTANT(pCi/g)

.Justification for concentration: Percent total Value 2.53E-01 244Cm UNLIMITED lCONSTANTpCi/g)

IJustification for concentration: Percent total Value 3.86E-01 Site Specific Parameters:

General Parameters:

None Element Dependant Parameters None Correlation Coefficients:

• None Summary Results:

.90.00% of the 142 calculated TEDE values are < 2.54E+01 mrem/year.

The 95 % Confidence Interval for the 0.9 quantile value of TEDE is 2.41E+01 to

• 2.73E+01 mrem/year 8.4-12 Page 8.4 - 12 DTBD-04-OO1, Attachment 8.4 DTBD-04-001R2.doc DTBD-04-O0lR2.doc .Page DTBD-04-001, Attachment 8.4

+tN4* DandD Building Occupancy Scenario DandD Version: 2.1.0 Run Date/Time: 1/5/2005 11:07:43 AM Site Name: RSNGS

==

Description:==

Analysis of potential dose from radionuclides discounted from those identified in NUREG/CR-3474.

FileName:C:\Documents and Settings\lbrown\My Documents\Proj ects\Rancho Seco\DTBDs\04-0011R2\04-001 Building.mcd Options:

Implicit progeny doses NOT included with explicit parent.doses Nuclide concentrations are distributed among all progeny Number of simulations: 100 Seed for Random Generation: 8718721 Averages used for behavioral type parameters External Pathway is ON Inhalation Pathway is ON Secondary Ingestion Pathway is. ON Initial Activities NuclideArea of 2 Contamination (mi ) Distribution F36-C I lUNLIMITED iFCONSTANT(dprn/100 cm**2)

Justification for concentration: Percent total ]Value 1.55E+00 U41Ca NLIMITED ]C0NSTANT(dpn1l00 cm**2)

Justification for concentration: Percent total lValue 7.51E-62 79SeUNLIMITED CONSTANT(dpm/100 crn**2)

Justification for concentration: Percent total Value 1.72E-03 93Zr llUNLIMITED IlCONSTANT(dpm/100 cm**2)

Justification for concentration: Percent total Value 2.87E-04 193MO IMITED llCONSTANT(dpmnl00 cm**2)

Justification for concentration: Percent total lValue 8.74E-02 121mSn [UNLIMITED CONSTANT(dpm/100 cm**2)

[Justification for concentration: Percent total Value 1.26E-03 ITD0-O Page 8.413 I2do TD0 ,Atcmn .

DTBD-04-001R2-doc Page 8.4 - 13 DTBD-04-001, Attachment 8.4'

11291 i UNLIMITED CONSTANT(dpm/100 cm**2)

IJustification for concentration: Percent total llValue 1.64E-06 135Cs .Ii lUNLIMITED. FCONSTANT(dpm/100 cm**2)

.Justification for concentration: Percent total Value 1.09E-04 151Sm IMITED CONSTANT(dpm/lO0 cm**2)

.Justification for concenfration: Percent total Value- 3.42E-02 1166mHo IMITED [CONSTANT(dpm1o00 cm* *2)

[Justification for concentration: Percent total Value 4.5 6E-O 1 1233U UNLIMITED lCONSTANT(dpm!100 cm**2) 1 IJustification for concentration: Percent total Value 3.67E-03 Site Specific Parameters:

General Parameters:

None Correlation Coefficients:

None Summary Results:

90.00% of the 100 calculated TEDE values are < 2.96E-03 mrem/year.

The 95 % Confidence-Interval for the 0.9 quantile value of TEDE is 2.75E-03 to 3.21E-03 mrem/year Page 8.4-14 DTBD-04-OO1, Attachment 8.4 DTBD-04-OO1 R2.doc DTBD-04-001R2.doc Page 8.4 - 14 DTBD-04-001, Attachment 8.4

I.NIF Resow DandD Building Occupancy Scenario.

DandD Version: 2.1.0 Run Date/Time: 1/5/2005 11:16:17 AM Site Name: RSNGS

==

Description:==

Analysis of potential dose from radionuclides not discounted from those identified in NUREG/CR-3474 FileName: C:\Documents and Settings\lbrown\My Documents\Projects\Rancho Seco\DTBDs\04-001R2\04-OOlBuilding2.mcd Options:

Implicit progeny doses NOT included with explicit parent doses Nuclide concentrations are distributed among all progeny Number of simulations: 100 Seed for Random Generation: 8718721 Averages used for behavioral type parameters External Pathway is ON Inhalation Pathway is ON Secondary Ingestion Pathway is ON Initial Activities:

Nuclide l Area A

offDistribution uContamination (m2 )I 3H UNLIMITED CONSTANT(dpm/lO0 cm**2)

Justification for concentration: Percent total Value 2.88E+01 .

14C UNLIMITED 1CONSTANT(dpml00cm**2)

Justification for concentration: Percent total llValue . 7.54E+01 l55Fe . UNLIMITED 1CONSTANT(dprn/00 cm**2)

Justification for concentration: Percent total ll Value . 8.07E+03 IUNLIMITED U59Ni FCONSTANT(dpm/100 cm**2) 1

[ustification for concentration: Percent total ll Value 3.60E+02 l

[6ZCo I UNUMITED [CONSTANT(dpm/100 cm**2)

Justification for concentration: Percent total Value 4.18E+04 63Ni I l IMITED TCONSTANT(dprn/100 cm* *2)

'Justification for concentration- Percent total llValue 4.96E+04

_TD0-0R.o Pae_41 TD0-OAtcmn .

DTBD-04-001R2.doc Page 8.4 - 15 DTBD-04-001, Attachment 8.4

190Sr lENIITD . lCONSTANT(dPrn/100 cm**2) lJustification for concentration- Percent total IValue 3.28E+00 199TC l UNLIMITED IFCONSTANT(dpm/100 cm**2) for concentration: Percent total DJustification Value 3.68E-01 137Cs UNLMITED CONSTANT(dpm/100 cm**2)

Justification for concentration: Percent total Value 3.53E+00 1152Eu I i iIIMITED . ICoNSTANT(dpm/100 cm**2)

IJustification for concentration: Percent total lValue 6.54E-01 1154Eu l LIMITED I[CONSTANT(dpxn/100 cm**2)

IJustification for concentration: Percent total. Value 1.13E+00 1155Eu lUNLIMITED CONSTANT(dpm/lO00 cm**2)

Justification for concentration: Percent total Value 1.41E-01 iII I1MITED l239PU STANT(dpm/100 cm**2) lJustification for concentration: Percent total . lalue 2.80E-02 U94Nb I I URITmTED llCONSTANT(dpm/100 cm**2) lJustification for concentration: Percent total llValue 1.1 lE+00 l134CS -UNLIMITED CONSTANT(dpm/100 cm**2) lJustification for concentration: Percent total Value 8.06E-02 Site Specific Parameters:

General Parameters:

None Correlation Coefficients:

None Summary Results:

90.00% of the 100 calculated TEDE values are < 1.49E+02 mrem/year.

The 95 % Confidence Interval for the 0.9 quantile value of TEDE is 1.47E+02 to 1.52E+02 mrem/year Page 8.4-16 DTBD-04-OO1, Attachment 8.4 DTSD-04-001R2.doc DTBD-04-001R2.doc Page 8.4 - 16 DTBD-04-001, Attachment 8.4

4.4 00 DandD Building Occupancy Scenario I-9 eI-DandD Version: 2.1.0 Run Date/Time: 12/20/2004 1:53:48 PM Site Name: RSNGS

==

Description:==

Analysis of potential dose from radionuclides discounted from those identified in the ORIGEN computer code run for spent fuel.

FileName:C:\Documents and Settings\lbrown\My Documents\Projects\Rancho Seco\DTBDs\04-001 R2\04-00 1Building3 .mcd Options:

Implicit progeny doses NOT included with explicit parent doses Nuclide concentrations are distributed among all progeny Number of simulations: 100 Seed for Random Generation: 8718721 Averages used for behavioral type parameters External Pathway is ON Inhalation Pathway is ON Secondary Ingestion Pathway is ON Initial Activities:

I Nuclide Area of ContAmiaton (2 Distribution IIContaminatio(m)I 115ISm l - CONSTANT(dpm/100 cm**2)

'Justification for concentration: Percent total Value 7.10E+O1 1243Am I UNLIMITED ImCONSTM10T(dPx'00 cm**2) l justification for concentration- Percent total l[Value 4.03E+OO 1243Cm UNLIMITED lCONSTANT(dprn/100 cm**2)

Justification for concentration: Percent total llValue 9.04E+OO Site Specific Parameters:

General Parameters:

None 8.4-17 Page 8.4 - 17 DTBD-04-OQ1, Attachment 8.4 R2.doc DTBD-04-OO1 R2.doe DTBO-04-001 Page DTBD-04-001, Attachment 8.4

Correlation Coefficients:

None Summary Results:

90.00% of the 100 calculated TEDE values are < 9.48E+00.mrem/year.

The 95 % Confidence Interval for the 0.9 quantile value of TEDE is 8.46E+00 to 1.08E+01 mrem/year Page 8.4- 18 DTBD-04-OO1, Attachment 8.4 DTBD-04-OO1 R2.doc DTBD-04-001 R2.doc Page 8.4 -18 DTBD-04-001, Attachment 8.4

+1 A DandD Building Occupancy Scenario a-0 a0 DandD Version: 2.1.0 Run Date/Time: 12/20/2004 2:58:12 PM Site Name: RSNGS

==

Description:==

Analysis of potential dose from radionuclides not discounted from those identified in the ORIGEN computer code run for spent fuel.

FileName:C:\Documents and Settings\lbrown\My Documents\Projects\Rancho Seco\DTBDs\04-001R2\04-O0lBuilding4.mcd Options:

Implicit progeny doses NOT included with explicit parent doses Nuclide concentrations are distributed among all progeny Number of simulations: 100 Seed for Random Generation: 8718721 Averages used for behavioral type parameters External Pathway is ON Inhalation Pathway is ON Secondary Ingestion Pathway is ON Initial Activities:

NuclideArea of Distribution Contamination (m2) 11

[3H UNLIMITED -!CoNSTANT(dpm/100 cm**2)

IJustification for concentration- Percent total Value 4.30E+01 155Fe llUNLIMITED ]FCONSTANT(dpmi/100 cm**2)

[Justification for concentration- Percent total Value . S;25E+02 160Co - UNLIMITED CONSTANT(dpmi/100 cm**2)

IJustification for concentration: Percent total Value 1.01E+03 159Ni 5EUNLIMITED G0NSTANT(dpxn/l0 cm**2)

)justification for concentration: Percent total lValue* 1.47E+00 190Sr

• lUNLIM ED llCONSTANT(dpmnI100 cm**2)

Justification for concentration: Percent total

• Value 1.29E+04 i125Sb IMITED . CONSTANT(dpm/100 cm**2) .

'Justification for concentration: Percent total Value 6.1 1E+02 MTD-04-00IR2.doc Page 8.4 - 19 DTBD-04-001, Attachment 8.4

3dCS -- IMITED [CONSTANT(dpm/lOO a**2)

TustifiCationfor concentration Percent total Value 4.71E+03 l137CS lNLIMITED llCONSTANT(dpm/100 cm**2) 1 Justification for concentration: Percent total lValue 3.66E+04 1147P llUNLIMITEDI CONSTANT(dpm/100 cm**2)

.Justification for concentration: Percent total' Value 5.60E+03 l63N llUNLIMITED mCONSTANT(dpx/100 cm**2) 1 Justifcation for concentration: Percent total Value 1.89E+02

• SiteSpecific Parameters:

General Parameters:

None Correlation Coefficients:

None Summary Results:

90.00% of the 100 calculated TEDE values are < 8.34E+01 mrem/year.

The 95 % Confidence Interval for the 0.9 quantile value of TEDE is 7.92E+01 to 8.88E+01 mrem/year Page 8.4 -20 DTBD-04-0O1, Attachment 8.4 DTBD-04-OO1 R2.doc DTBD-04-001lR2.doc Page 8.4 - 20 DTBD-04-001, Attachment 8.4

8-P

. 11~1 .A. DandD Building Occupancy Scenario I-DandD Version: 2.1.0 Run Date/Time: 12/20/2004 3:08:30 PM Site Name: RSNGS

==

Description:==

Analysis of potential dose from radionuclides not discounted from those identified in the ORIGEN computer code run for spent fuel.

FileName:C:\Documents and Settings\lbrown\My Documents\Projects\Rancho Seco\DTBDs\04-001R2\04-OOlBuilding5.mcd Options:

Implicit progeny doses NOT included with explicit parent doses Nuclide concentrations are distributed among all progeny Number of simulations: 100 Seed for Random Generation: 8718721 Averages used for behavioral type parameters External Pathway is ON Inhalation Pathway is ON Secondary Ingestion Pathway is ON Initial Activities:

Nuclide Distribution 154Eu lUNLIMITED llCONSTANT(dpm/100 cm**2) . l Justification for concentration: Percent total VLalue 1.52E+03 15Eu UNTIMITED ICONSTANT(dpm/100 cm**2) 1 Justification forconcentration: Percent total ll Value 6.8 lE+02 NLIMITED . [CONSTANT(dpmn/l00 cm**2)

Justification forconcentration: Percent total Value 8.0E+02 i239PuIi UNLIMITED llCONSTANT(dpm/100 cm**2) l justification for concentration. Percent total Value .6.68E+01 ]

1240Pu [U IMNTED 0CNSTANT(dpm/100cm**2) I Justification for concentration: Percent total l Value 1.lSE+02 1241Pu lUNLIMITED [CONSTANT(dpm/l00 cm**2)

Justification for concentration: Percent total lValue 2.1 1E+04 8.421 TBD-4-01, ttacmen 8.

DTBD04-01R2docPag DTBD-04-001R2.doo Page 8.4 - 21 DTBD-04-001, Attachment 8.4

241AmZI UNLTED FCONSTANT(dpn/1lO cm**2)

.Iustificationforconcentration ercentotal FValue 2.53E4O2

. l2I4Cm lIUNLIMITED FCONSTANT(dpm/1O00 cm**2)

Justification for concentration: Percent total_ ll Value 3.86E+02

• Site Specific Parameters:

General Parameters:

None Correlation Coefficients:

None Summary Results:

90.00% of.the 100 calculated TEDE values are < 1.63E+03 mrem/year.

The 95 % Confidence Interval for the 0.9 quantile value of TEDE is 1.45E+03 to 1.85E+03 mrem/year Page 8.4-22 DTBD-04-OO1, Attachment 8.4 R2.doc DTBD-04*OO1 R2.doc DTBD-04-001 Page 8.4 - 22 DTBD-04-001, Attachment 8.4

Attachment 8.5 Dose Considerations for Radionuclides Not Supported By DandD Page 8.5-1 DTBD-04-OO1, Attachment 8.5 DTBD-04-001R2.doc DTBD 0400IR2.doc Page 8.5 - I DTBD-04-001, Attachment 8.5

Dose Considerations for Radionuclides Not Supported By DandD Inhalation Inagestion Radionuclide Percent Total DCF* Weighted DCF % Total WDCF DCF* Weihted DCF l% Total WDCF Ag-108rn 2.65E-04 7.66E-08 2.03E-11 8.08E-04 2.06E-09 5.46E-13 1.75E-04 Ba-133 2.64E-03 2.11 E-09 5.57E-12 2.22E-04 9.19E-10 2.43E-12 7.78E-04 Pm-145 1.45E-06 8.23E-09 1.19E-14 4.75E-07 1.28E-1O 1.86E-16 5.95E-08 Sm-146 3.05E-13 8.26E-05 2.52E-17 1.OOE-09 5.51 E-08 1.68E-20 5.39E-12 Tb-158 5.31 E-06 6.91 E-08 3.67E-13. 1.46E-05 1.19E-09 6.32E-15 2.03E-06 Hf-178m 1.63E-04 6.65E-07 1.08E-10 4.32E-03 5.68E-09 9.26E-13 2.97E-04 Pb-205 5.56E-09 1.06E-09 5.89E-18 2.35E-10 4.41 E-10 2.45E-18 7.86E-10

. Total 5.36E-03 Total 1.25E-03 Co-60 -4.18E+01 15.91 E-081 2.47E-06 I7.28E-09l 3.04E-07 I Ni-63 4.96E+01 8.39E-10 4.16E-08 1.56E-10 7.74E-09 Totall 2.51 E-06 Total 3.12E-07 I

• Effective Committed Dose Equivalent per Unit Intake (Sv/Bq)

DTBD-04-00IR2.doc Page 8.5 - 1 DTBD-04-001, Attachment 8.5

Attachment 8.6 Site-Specific Suite of Radionuclides for Use at RSNGS DTBD-04-001R2.doc Page 8.6-I1 DTBD- 04-001, Attachment 8.6

Site-Specific Suite of Radionuclides for Use at RSNGS Half Life Decay Half Life Decay Radionuclide (Years) Mode Radionuclide (Years) Mode

• H-3 1.23E+01 .Cs-1 37 3.02E+01

• C-14 5.73E+03 *Pm-147 2.62E+00 L Na-22 2.60E+00 , Eu-I 52 1.36E+01 r

• Fe-55 2.70E+00 L Eu-1 54 8.80E+00 ,

• Ni-59 7.50E+04 y Eu-1 55 4.96E+00 ,Y Co-60 5.27E+00 , *Np-237 2.14E+6 a_,ry

• Ni-63 1.OOE+02 L *Pu-238 8.78E+01 a,__y

• Sr-90 2.86E+01. L *Pu-239 2.41E+04 a,r_

Nb-94 2.03E+04 , *Pu-240 6.60E+03 a, y

• Tc-99 2.13E+05 , *Pu-241 1.44E+01

• Ag-108m 1.27E+02 y Am-241 4.32E+02 a,y Sb-125 2.77E+00 , *Pu-242 3.76E+05 -a, I Cs-134 2.06E+00 J *Cm-244 1.81E+01 a, y ac-Alpha decay p - Beta decay p+- Positron decay I- Gamma decay

• Hard to Detect Nuclides (HTDN - radionuclides not readily detected by gamma spectroscopy, e.g., Ni-63 or Cm-244, and requiring offsite, contract laboratory analysis)

Page 8.6-1 DTBD- 04-001. Attachment 8.6 DTBD-04.001R2.doc DTBD-04-001 R2.doc Page 8.6 - I DTBD- 04-001, Attachment 8.6

Attachment 2 Rancho Seco Historical Site Assessment, Revision 0

': -0" I.-, :I.- 1-1 ",-

Sacramento Municipal Utility District Rancho Seco Nuclear Generating Station Historical Site Assessment March 2004

C Historical Site Assessment Document Approval Document Preparation oversight by:

Einar T. Ro lgen Principle Decommissioning Radiological Engineer Document Approved by:

Dennis L( Gardiner Project Manager, Decommissioning

Table of Contents 1.0 Introduction ................................ 3 2.0 Objectives of Historical Site Assessment ................................ 7 3.0 Terms, Acronyms, and Abbreviations ................................ 9 4.0 Property Identification.............................................................................................13 4.1 Facility Characteristics ............... 13 4.1.1 Licensee Identification (DPR-54) .13 4.1.2 Location ............  ; 13 4.1.3 Topography .........................  ; 15 4.1.4 Stratigraphy .15 4.2 Environmental Characteristics .. 16 4..1 e l g ........... .. ...................... ..... 16 4.2.1 Geology.16 ...............-. '

4.2.2 Seismology ...............  ; 16 4.2.3 Hydrology .18 4.2.4 Hydrogeology .18 4.2.5 Meteorology ...............  ; 21 5.0 HSA Methodology .25 5.1 'Approach and Rationale . . . 25 5.2 'Documents reviewed '.. . . 25 5.3 Site Reconnaissance . . . 26 5.4 Personnel Interviews . . .26 5.5 Historical Construction PhotographReview . . . 28 6.0 Operational History .. 29 6.1 Introduction ....................... 29 6.2 Decommissioning Plan Chronology . . .32 6.3 Regulatory Overview . . .33 6.3.1 Permits and Licenses.33 ....................  ;. . . . 33 6.4 Waste Handling Procedures . . .36 6.4.1 ProcessConol Program (PCP) .................. 36

.4. . Rancho Seco Administrative Procedures (RSAP)

. . Pr6.4.2 .36 6.4.3 Radiation Control Manual (RCM) .37 6.4.4 Radwaste Control Manual (RWCM) ................................... 37 6.4.5 Chemistry Department Procedures Manual .37 6.4.6 Surveillance Procedures (SP) ................................... 37 6.5 Current Site Usage .37 Revision 0 RSNGS Historical Site Assessment i March 2004

6.5.1 Description of Operations ........................................... 37 6.5.2 Preliminary Site Characterization ........................................... 37 6.6 Site Dismantlement ............................................ 38 6.6.1 Dismantlement activities within the Power Block ........................................... 38 6.6.2 Dismantlement activities outside the Power Block ........................................... 38 6.7 Radiological Sources ............................................ 38 6.7.1 Spent Fuel ........................................... 38 6.7.2 Irradiated Hardware ........................................... 39 6.7.3 Reactor Vessel and Internals ........................................... 39 6.7.4 Plant Systems ........................................... 39 6.7.5 Industrial Area Contamination ........................................... 39 6.7.6 Non-Industrial Area Contamination ........................................... 40 6.8 Waste Stream Description ........................................... 42 6.8.1 Hazardous Materials/Wastes ........................................... 42 6.8.2 Low Level Radioactive Waste (LLRW) ........................................... 44 6.8.3 Spent Fuel ........................................... 46 6.9 Incident Descriptions ............................................ 46 6.9.1 Radiological Spills ........................................... 46 6.9.2 Chemical Spills ........................................... 47 6.9.3 Loss of Material Control ........................................... 47 6.9.4 System Cross-Contamination ........................................... 47 6.10 Survey Unit Identification and Classification ........................................... 48 6.10.1 Site Classification .48 6.10.2 Assessment Perfonnance ....................... 48 6.10.3 Areas .48 6.10.4 Survey Units.48 6.10.5 Initial Designation of Areas .48 6.10.6 Survey Unit Designation Program .49 6.11 Radiological Impact Summaries ............................ 54 6.11.1 Area I (SA01) - Plant Effluent Area .54 6.11.2 Area 2 (SA02) - South Plant Outfall Area .60 6.11.3 Areas 3 - 7 .60 6.11.4 Area 8 (SA08) .61 6.11.5 Area 9 (SA09) .61 7.0 Findings...................................................................................................................... 63 8.0 Conclusions .67 9.0 References .69 10.0 Appendices and Addendums .71 Revision 0 RSNGS Historical Site Assessment ii March 2004

List of Tables Page 4.1 Magnitude / Intensity Comparison ...................... 17 4.2 Expected Extreme Wind Speeds ..................... 22 4.3 Precipitation Climatology Averages (inches) ............................... 23 4.4 Precipitation Intensity ............................... 24 5.1 Personnel Observations Summary ............................... 27 6.1 Operational History - RSNGS ............................... 30 6.2 Licenses and Permits ............................... 34 6.3 MARSSIM Survey Unit Classification Matrix ............................... 52 7.1 Area Designations ........... 63 Revision 0 RSNG S Historical Site Assessment iii March 2004

-- A-List of Figures Page 4.1 Rancho Seco Property Map ................... 14 4.2 Ground Water Contour Map ................... 20 6.1 Impacted Area Designations ................... 50 6.2 Area Designations ................... 51 Revision 0 RSNG' S Historical Site Assessment iv March 2004

Executive Summary The Sacramento Municipal Utility District (herein referred to as the District) has conducted the Historical Site Assessment (HSA) of its Rancho Seco Nuclear Generating Station (RSNGS) in support of the ultimate decommissioning and license termination of the facility.

The HSA, as a preliminary component in the Radiological Site Survey Investigation (RSSI),

will provide guidance for subsequent activities, culminating with the final status survey (FSS) and license termination. The HSA will designate the initial segregation of the site into various Areas and Survey Units and provide guidance in the development of the procedures and maps required to document the characterization, remediation, and ultimate 10 CFR Part 50 license termination of the site when compared to the release criteria referenced in the MARSSIM standard.

The HSA was developed consistent with the methodology described in NUREG-1575, "Multi-Agency Radiation Survey and Site Investigation Manual"; including personnel interviews, and detailed record reviews.

RSNGS operated from 1974 through 1989 within a highly regulated environment, which continues today. This highly regulated and documented environment provided a significant foundation for the development of the HSA.

The evaluations performed to date identify locations outside of the historic power block-where radioactive material c6ntamination'may have occurred due to various causes such as spills or loss of material control. These locations have been designated as Impacted Areas requiring additional investigation prior to the FSS.

The District-owned and controlled property, comprised of some 2480 acres, includes the 87 acre Industrial Area site. Of the nearly 2400 acres outside of the Industrial Area, approximately 80 acres have been impacted by licensed operations. These include the Plant Liquid Effluent Discharge water course way to the Southwest of the site and the Storm Drain outfalls located to the South of the facility. These two areas have been designated as Impacted. The remaining 2300 acres outside of the Industrial Area have been designated as Non-Impacted areas. The Industrial Area is designated as Impacted based on the diversity of operations conducted in this area.

The District believes that the information contained within the HSA and the resulting conclusions for site area classifications accurately describe the radiological conditions that currently exist at the site.

Revision 0 RSNGS Historical Site Assessment 1 March 2004

This page intentionally left blank Revision 0 RSNGS Historical Site Assessment 2 March 2004

1.0 INTRODUCTION

The Sacramento Municipal Utility District (herein referred to as the District) has conducted the Historical Site Assessment of its Rancho Seco Nuclear Generating Station in accordance with the guidance of the Multi-Agency Radiation Survey and Site' Investigation Manual, NUREG-1575 (MARSSIM) [Ref. 9.1] in support of the ultimate decommissioning and license termination of the facility.

The HSA formally began in July 2001, after several preliminary radiological assessments of the facility operations and their impact on remediation necessary prior to the performance of the Final 'Status' Surveys (FSSs). These preliminary'surveys, collectively referred to as the "Radiological Characterization Plan' for the Raiicho Seco Nuclear Power Generating' Station" (RCPRSNPGS) [Ref. 9.2], were conducted 'shortly after the shut down and termination of commercial operation of the RSNGS in June 1989.

This characterization effort was undertaken prior to the implementation of the MARSSIM guidelines and therefore, relied primarily on the' guidance of NUREG/CR-2082 "Monitoring for Compliance With Decommissioning Termination Survey' Criteria" [Ref. 9.3]'and Nuclear Regulatory Commission (NRC) Draft Regulatory Guide DG-1005 For Nuclear Reactor Facilities. "Standard Format and Content for Decommissioning Plans for Nuclear Reactors"

[Ref. 9.4]

Additional surveys had been anticipated for the RCPRSNPGS. These included Phase III surveys and the FSS, to be performed in the 2007 and 2011 time frame respectively (phases I

& II having been completed prior to 1997). With the issuance of MARSSIM, these surveys will be incorporated into the MARSSIM directed site characterization, FSS design, and the District's License Termination Plan (LTP) for the facility, the schedule of which will be determined by the District's senior management.

The HSA consisted of a review of historical:

• Plant incident records;

• Plant maintenance records;

• Plant modification records;

• Plant radiological survey records; and'

• Regulatory reports submitted by the District to various governmental agencies.

The HSA also included written questionnaires and oral interviews with current and past facility employees regarding historical incidents that posed potential impacts to the facility.

A review of historic site aerial photographs and physical inspections of the facility were performed to verify and validate the results of the historical rec6rd reviews.

These efforts were designed to document the District's detailed knowledge of those events with a potential to impact the decommissioning of the site and final termination of its license.

Concurrent with the performance of the HSA was the initial segregation of the facility into individual areasand specific, uniquely identified, survey units. This provides the basis for Revision 0 RSNGS Historical Site Assessment 3 March 2004

iLt development of area/unit specific site drawings and survey maps required to document the characterization, remediation, and final release survey process. A major output from the HSA process was the information used as the basis for the preliminary MARSSIM classifications of the initial survey units.

The initial classification of the site areas is based on the historical information and site characterization data. Data from operational surveys, surveys performed in support of decommissioning, routine surveillance, or any other applicable data may be used to change the original classification of an area up to the time of the FSS as long as the classification reflects the level of residual activity existing prior to any remediation in the area.

To prevent the spread of radioactive material, Rancho Seco was designed with multiple boundaries to contain the plants' radioactive materials within its many components, systems, and structures. During the operation of the plant from 1974 through 1989, many of these systems and structures, (and many that were not designed to become contaminated) have been impacted due to the routine operations, non-routine events and maintenance activities associated with the operational and post operational history of the plant.

The most significant of these systems and structures include:

• Reactor Containment Building;

• Auxiliary Building;

• Spent Fuel Storage Building;

• Interim Onsite (radwaste) Storage Building (IOSB);

• Turbine Building;

• Solidification Building;

• Contractor Fab Shop;

• Tank Farm;

• Regenerate Holdup Tanks;

• Auxiliary Boilers;

• Main and Auxiliary Steam Systems;

• Main and Auxiliary Feed Water Systems;

• Clean Drain System;

• Component/Turbine Cooling Water Systems;

• Nitrogen Gas System;

• Service Air System;

• Discharge piping from RHUTS

• Main circulating water basins Revision 0 RSNGS Historical Site Assessment 4 March 2004

• Retention basins

• Control Rod Drive Cooling System; and

• Nuclear Service Cooling Water System.

The initial MARSSIM classification of these areas, the majority of which lie within that area comprising the historic power-block, was based on the design function of the area of concern or the areas' operational history.

These areas were given Impacted Area designations. Should subsequent investigations over the course of the decommissioning project support reclassification, the circumstances and rationale will be appropriately documented.

During the operational history of the facility, radioactive liquid spills, waste processing and storage activities, and maintenance activities on contaminated equipment and components occurred outside of the Radiologically Controlled Area (RCA) but within the Industrial Area fence line. These events have resulted in the assignment of an Impacted Area designation for the Industrial Area. Should data acquired during the course of the decommissioning project support reclassification; the circumstances and rationale will be appropriately documented.

The District-controlled property areas outside of the Industrial Area have been initially classified as Non-Impacted with the exception of Area 1 because of the land area north of the plant effluent watercourse and the plant effluent watercourse and Area 2 because of the storm drain outfalls within the area. These two locations will be initially managed as Impacted Areas.

The classification assignments of the areas outside of the Industrial Area have been substantiated by the non-Industrial Area surveys performed by Shonka Research Associates, Inc. (Rancho Seco Non-Industrial Area Survey Project, Final Report, June 26, 2001 [Ref.

9.17]). This project provided direct scanning of over 300,000 square meters of surface area and documented over 80,000 gamma spectral samples without detection of radioactive material of plant origin above background (spiked fields and known contamination along the effluent canal the exceptions).

The program to terminate the RSNGS license (DPR-54), including the HSA, followed by scoping, characterization, and remediation surveys will provide the information required to fully characterize the facility.

The MARSSIM process for preparing for FSS provides multiple opportunities to re-evaluate decisions reached during any phase of the program such that areas with preliminary classifications may be reclassified should subsequent information show these preliminary classifications as not justified.

The District believes the investigation associated with the development of this HSA has resulted in a knowledge base that accurately describes the areas of potential impact and provides a conservative basis for the initial classification of the site's survey units.

Based on the recommendations contained in MARSSIM, these HSA results are presented as follows:

• Section 1.0 - Introduction; Revision 0 RSNGS Historical Site Assessment 5 March 2004

AL

• Section 2.0 - Objectives of the Historical Site Assessment: General purpose of the Historical Site Assessment;

• Section 3.0 -. Terms, Acronyms, and Abbreviations: Presents a listing of the abbreviations and acronyms used in the HSA;

• Section 4.0 - Property Identification: Physical and environmental characteristics of the facility;

• Section 5.0 - HSA Methodology: Methodology used in the development of the HSA;

• Section 6.0 - Operational History: Operational history of site including summaries of the documents providing significant information contributing to the characterization of the facility,

• Section 7.0 - Findings: HSA findings on the potential contaminates and Impacted Areas, including descriptions of the major Areas and significant Survey Units;

• Section 8.0 -

Conclusions:

Conclusions of the HSA;

• Section 9.0 -

References:

List of references applicable to the HSA; and

• Section 10.0 - Appendices: List of HSA appendices and addendums.

Revision 0 RSNGS Historical Site Assessment 6 March 2004

2.0 OBJECTIVES OF HISTORICAL SITE ASSESSMENT The Sacramento Municipal Utility District conducted the Historical Site Assessment of the Rancho Seco Nuclear Generating Station to:

• Identify known and potential sources of radioactive material and radioactively contaminated areas including systems, structures and environmental media based on the investigation and evaluation of existing information;

• Identify areas of the site with no conceivable or likely potential for radioactive or hazardous materials contamination and assign a preliminary classification of Non-Impacted while assigning a preliminary classification of Impacted to all remaining portions of the site;

• Evaluate the potential for migration of radiological and hazardous substances beyond the boundaries of the Industrial Area or District property;

• Develop the records to be utilized during the design of subsequent scoping, characterization, remediation, and the FSS; and

• Provide preliminary information necessary to identify and segregate the site into survey units evaluated against the criteria specified in the MARSSIM guidelines for classification. This classification will designate the need for and level of remedial action required within a particular survey unit as well as the level of intensity required during the FSS.

Revision 0 RSNGS Historical Site Assessment 7 March 2004

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. RSNGS Historical Site Assessment 8 March 2004

3.0 TERMS. ACRONYMS. AND ABBREVIATIONS ACM: Asbestos Containing (or Contaminated) Material AEC: Atomic Energy Commission (also USAEC) a - Alpha: contamination with alpha emitting radionuclides when used in the context of radiological surveys. When used in the context of the statistical analysis of survey data, a will denote a Type I error (rejecting the null hypothesis when it is in fact true).

AOC: Area of Concern Beta: contamination with beta emitting radionuclides when used in the context of radiological surveys. When used in the context of the statistical analysis of survey data, 13will denote a Type II error (accepting the null hypothesis when it is in fact false).

BOP: Balance Of Plant BWST: Borated Water Storage Tank CCPM: Corrected Counts per minute. (CPM minus background CPM)

CCR: California Code of Regulations CCW: Component Cooling Water CFR: Code of Federal Regulations CEOA: California Environmental Quality Act.

cm 2 : Square centimeters Co-60: Cobalt-60 (radioactive isotope of cobalt metal) cpm: Counts per minute Cs-137: Cesium-137 (radioactive isotope of cesium)

CSCA: Controlled Surface Contamination Area CST: Condensate Storage Tank DCGL: Derived Concentration Guideline Level dpm: Disintegrations per minute (DPM, Dpm, or dpm) dpm/1 0c0n 2 : Disintegrations per minute per 100 square centimeter surface area (or dpm/l00 cm2)

DSAR: Defueled Safety Analysis Report DTSC: Department of Toxic Substance Control. California agency regulating hazardous materials and waste EPA: Environmental Protection Agency (also USEPA)

FONSI: Finding of No Significant Impact FSAR: Final Safety Analysis Report Revision 0 RSNGS Historical Site Assessment 9 March 2004

AL Terms, Acronyms, and Abbreviations (Continued)

FSS: Final Status Survey Ft: feet Y- Gamma: contamination with gamma emitting radionuclides when used in the context of radiological surveys.

GEIS: Generic Environmental Impact Statement HP: Health Physics HSA: Historic Site Assessment IDAP: Incremental Decommissioning Action Plan IOSB: Interim On-site (Radwaste) Storage Building ISFSI: Independent Spent Fuel Storage Installation LER: License (e) Event Report LLD: Lower Limit of Detection LTP: License Termination Plan IMTARSSIM: Multi-Agency Radiation Survey and Site Investigation Manual MDA: Minimum Detectable Activity mph: miles per hour (or MPH) mR/hr: millirem per hour MSL: mean sea level MWe: megawatt - electrical (output)

MWt: megawatt - thermnal jCi: microcurie (also ttCi, gCi/g (per gram), pCi/ml (per milliliter), pCi/cc (per cubic centimeter)

NOAA: National Oceanographic and Atmospheric Administration NPDES: National Pollutant Discharge Elimination System NRC: Nuclear Regulatory Commission (Also USNRC)

ODCM: Offsite Dose Calculation Manual ODR: Occurrence Description Report OTSG: Once Thru Steam Generator(s)

PAP: Personnel Access Point Revision 0 RSNGS Historical Site Assessment 10 March 2004

Terms, Acronyms, and Abbreviations (Continued)

PASS: Post Accident Sampling System PCP: Process Control Program PDP: Proposed Decommissioning Plan PDO: Potential Deviation from Quality PE: Plant Effluent pCi: picocurie (pCi/l (per liter), pCi/g (per gram))

PUDF: Plan for Ultimate Disposition of the Facility QA: Quality Assurance QC: Quality Control RB: Reactor Building (or RCB Reactor Containment Building)

RCA: Radiologically Controlled Area RCPRSNPGS: Radiological Characterization Plan for the Rancho Seco Nuclear Power Generating Station RCRA: Resource Conservation and Recovery Act RCS: Reactor Coolant System REMP: Radiological Effluent Monitoring Program RP: Radiation Protection RSNGS: Rancho Seco Nuclear Generating Station consisting of an 87 acre Industrial Area containing the nuclear facility and a total site area of 2,480 acres RSSI: Radiological Site Survey Investigation RHUT: Regenerant Holdup Tank(s) (A, B, & C)

RWP: Radiation Work Permit SAR: Safety Analysis Report SER: Safety Evaluation Report SMUD: Sacramento Municipal Utility District SUID: Survey Unit Identification Number UFSAR: Updated Final Safety Analysis Report USAEC: United States Atomic Energy Commission (also AEC)

Revision 0 RSNGS Historical Site Assessment 11 March 2004

II This page intentionally left blank Revision 0 RSNGS Historical Site Assessment 12 March 2004

4.0 PROPERTY IDENTIFICATION 4.1 Facility Characteristics 4.1.1 Licensee Identification (DPR-54)

Sacramento Municipal Utility District Rancho Seco Nuclear Generating Station Physical address 6201 S Street, Sacramento, California 95817-1899 Mailing address PO Box 15830, Sacramento, California 95852-1830 4.1.2 Location Rancho Seco Nuclear Generating Station 14440 Twin Cities Rd Herald, California 95638 The property, herein called the site, is located in the southeast part of Sacramento County, state of California and lies either wholly or partly within Sections 27, 28, 29, 32, 33, and 34 of township 6 North, Range 8E. The nuclear reactor unit lies entirely within section 29.

The site is approximately 25 miles southeast of Sacramento and 26 miles northeast of Stockton in the central valley of California between the foothills of the Sierra Nevada Mountains to the east and the Pacific Coast range bordering the Pacific Ocean to the west. A map ofthe facility and location is included as Figure 4.1.

The RSNGS site consists of an approximately 87-acre fence-enclosed Industrial Area containing the nuclear facility surrounded by District-owned and District-controlled property totaling 2,480 acres.

The District is in the process of constructing a 30-acre natural gas-fired power plant on the RSNGS site, approximately l/2 mile south of the Industrial Area boundary. Also within the 2,480 acre site are the 560 acre Rancho Seco Reservoir and Recreation Area; a 50 acre solar power (photo-voltaic) electrical generating station; and the 0.9 acre, 10 CFR Part 72 licensed Independent Spent Fuel Storage Installation (ISFSI)..

'The 10 CFR Part 72 licensed ISFSI is independent of the 10 CFR Part 50 licensed facility.

Revision 0 RSNGS Historical Site Assessment 13 March 2004

Figure 4.1 Col Revision 0 RSNGS Historical Site Assessment 14 March 2004

4.1.3 Topograph-The plant site's rolling terrain is not directly intersected by any streams; however, drainage from higher levels is well defined and intercepts with runoff streams at lower levels. The plant's grade level of approximately 165 feet above MSL allows excellent drainage without danger of flooding. The elevation of the site' acreage varies from 130 feet to 280 feet above MSL and drainage along natural gullies varies from two to six percent. Runoff from the site drains into a seasonal "No - name" creek that is a tributary to Clay creek. Clay creek empties into Hadselville creek. Hadselville creek then empties in turn into: Laguna creek south, Consumnes River, Mokelomne River, Sacramento River, into the Pacific Ocean via the Sacramento River Delta.2 4.1.4 Stratigraphy Information regarding the stratigraphy of the site is taken, in part, from the FSAR [Ref. 9.5].

The stratigraphy below the site consists of a basement of Mesozoic and Paleozoic metaphoric rock, overlain with several tertiary and quaternary period formations including:

• Recent Alluvium (Qal) consisting of stream deposited gravel, sand, and silt. This material is confined to present drainage courses and ranges in depth from 0 to 5 feet.

• Older Alluvium (Qalo) consists of old stream and terrace deposits of gravel, sand, and silt. This material covers the flood plains in the southwest portion of the site and ranges in depth from 0 to 10 feet.

• Arroyo Seco formation (Qas) consists of deposits of well-rounded cobbles, pebbles, and sand derived chiefly from pre-Cretaceous sediments on pediment surfaces. This formation caps uplands in the eastern portion of the site and ranges in depth from 0 to 15 feet.,

• Laguna formation (Tl) consists of sand, silt, and some gravel; may or may not contain'clay. Fluviatile deposits are poorly bedded, poorly exposed, and non-'

andesitic in composition. This is the predominant formation within the site and ranges'in depth from 0 to approximately 130 feet.,

. Mehrten I consists of fluviatile sandstone, siltstone, and conglomerate dominantly of andesitic detritus. Locally contains horizons of coarse andesitic agglomerate of mudflow origin. This formation has no preconstruction surface exposure and there is little possibility that any'construction excavation entered this formation, which has an approximate thickness of 225 feet.

2 site information contained in Sections 4.1.3 through 4.2.5.8 is based on the current FSAR and has not been updated to the current date of this document.

Revision 0 RSNGS Historical Site Assessment 15 March 2004

it

• Valley Springs formation (Tv) consists of pumice and fine siliceous ash with much greenish-gray clay and some vitreous tuff, glassy quartz sand, conglomerate; commonly well bedded; derived largely from rhyolithic ejectamenta thrown out from the high Sierra Nevada. This formation also has no site surface exposures and an estimated average thickness of 250 feet.

4.2 Environmental Characteristics 4.2.1 Geology Information regarding the geology of the site is taken, in part, from the FSAR.

Rancho Seco is located about 25 miles southeast of Sacramento California in the low foothills of the Sierra Nevada Mountains. The site is founded on the Pliocene Laguna Formation and is underlain by an estimated 1500 to 2,000 feet of Tertiary or older sediments deposited on a basement complex of granite to metamorphic rocks. Field exploration included;

• 1,552 feet of bucket auger holes logged in detail;

• A 602 foot core hole visually and geophysically logged;

• 2,016 feet of small-bore hole borings that were logged and from which, soil samples were taken for laboratory analysis; and

• Approximately 11,500 feet of geophysical refraction profiles.

The resulting data from this exploration strongly indicate a lack of faulting below the Rancho Seco site.

4.2.2 Seismology Information regarding the seismology of the site is taken, in part, from the FSAR.

There are no indications of faulting below the site. The nearest fault, located approximately 10 miles to the east of the site, is the Foothill Fault System. This system has been inactive since the Jurassic Period, some 135 million years ago. The nearest active faults, located over 70 miles to the west, are the Hayward and San Andreas.

In response to questions by the NRC, prompted by a magnitude 5.7 earthquake on a fault previously believed to be inactive near Oroville, Ca. the District commissioned the reinvestigation of potential seismic activity in the vicinity of the site (Response to NRC questions on Geologic and Seismologic Conditions, 1987 [Ref. 9.14]). The results confirmed the lack of any credible faults closer than the Foothill Fault considered in the original licensing documents.

A search of the USGS database for earthquakes with intensities greater than IV on the modified Mercalli scale (Richter scale 4.0 or larger - Table 4.1) within a 200-mile radius of the plant resulted in 846 such events.

Revision 0 RSNGS Historical Site Assessment 16 March 2004

Table 4.1 Magnitude I Intensity Comparison Magnitude and Intensity measure different characteristics of earthquakes. Magnitude measures the energy released at the source of the earthquake. Magnitude is determined from measurements on seismographs.

Intensity measures the strength of shaking produced by.- the earthquake at a certain location. Intensity is determined from effects on people, human structures, and the natural environment.

The following table gives intensities that are typically observed at locations near the epicenter of earthquakes of different magnitudes.

Magnitude Intensity Description '

l1J°-30 l [i I3. Not felt exceptjbyp-!vr~few under especially favorable conditions. -

3.0 - 3.9 I1 Ill IIC.Felt only by a few persons at rest, especially on upper floors of buildings.

. .Ill. Felt quite noticeably by persons indoors, especially on upper floors of

... __ ..buildings. MaW y people do niot recognize it a'san earthquake. Standing motor cars may rock islightly. Vibrations similar to the passing of a truck. Duration estimated. -

4.0 - 4.9 - V-V IVtFelt o rs, u doors*by few 'during tie day 1 Atight,-.some

- wakedeDs e , dosdoors disturbed; walls make;crackjg sound.

Sehsationilik, drnkg building. Standing motor cars rockhed noticeably,,

V. Felt by'ned ayawakened. Someidfshes, windows broken.

- Unstabfe hduldm clocks may stop.

'5.0 - 5.9 VI - VII VI. Felt by all, many frightened. Some heavy furniture moved; a few instances of falleri plaster. Damageslight.

--- 1- !lDarage negigible in buildings of good design and construction; slight to moderate'in well-built ordinary structures; considerable damage in poorly built or badly designed structures; some chimneys broken. -

6.0 - 6.9 Vi-ll - IX ll. DamageaF.eslgned s rcures csitder dam<age In

-ordnnAla partial icollapse(DarFng agjtpoorby built-s stacs, columnsk sonumnsfWn lf%s Heavy fumltureo0 ' "

r 7.0 and

" ."Xii~>- !$ l ar

.structures

._'__-_. partial col ViIl or era Ii :f ecl llydesgne ,srucurs,,wellideslgned frame great in substantial buildings, with nseafoudinedatrtueweteions. frame.

X. Some well-built wooden structures destroyed; most masonry and frame higher-' .' higher - structures destroyed with foundations. Rails bent.

Xl.' Few, if any (masonry) structures remain standing. Bridges destroyed. Rails

'- -- !, . bent greatly. '  : - - '-,  : - - '

XII: Damage total.' Lines of sight and level are distorted.'Objects thrown into the air.

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NEIC Web Team Last modification: Friday, 18-Oct-2002 14:34 Revision 0 RSNGS Historical Site Assessment 17 March 2004

The largest event was the 1989 Loma Prieta earthquake (Magnitude 7.1 (7.2 in other literature), Modified Mercalli IX, 160 km distant) and the nearest was a magnitude 4.3, Modified Mercalli V, quake, 45.36 miles (73 km) from the site.

Restricting the search criteria to a 50-mile (80.5 km) radius results in only three monitored events.

These results, along with the geographical positioning of the site, aerial photos, and mapping of the facility are included in Appendix D.

4.2.3 Hvdrologv Information regarding the hydrology of the site is taken, in part, from the FSAR and the USAR [Ref. 9.6].

As described in section 4.1.3 Topography, above, the plant site's rolling terrain is not directly intersected by any streams; however, drainage from higher levels is well defined and intercepts with runoff streams at lower levels. Runoff from the site drains into an un-named "No-Name" creek, which in-turn empties into Clay creek. Clay creek empties into Hadselville creek. Hadselville creek then empties in turn into: Laguna creek south, Consumnes River, Mokelomne River, Sacramento River, into the Pacific Ocean via the Delta.

Within recent historical times no flooding or inundation from storms or runoff has occurred within the site boundaries. It is highly unlikely that the site could be flooded, even with abnormal rainfall intensities.

Since the commencement of operations in 1974, the only significant change in regional land use had been the conversion of several sections of land near the facility from grazing to grape production. An additional change of some note would.be the population expansion that has occurred in the communities of Galt and lone, Ca. According to the City of Galt Housing Needs Assessment, Administrative Draft, October 2001, the population of this historically-agricultural community, located between 10 and 15 miles from the site, doubled from 1990 to 2000 and the number of residential properties nearly doubled to almost 6000 units. While notable, the lone expansion has not been as dramatic.

Surveys conducted by the County of Sacramento indicate that the land adjoining the site, within at least a 15-mile radius, will remain primarily for agricultural and grazing use; therefore, the rainfall runoff factors will remain constant and not cause any difference in the hydrological properties of the region.

Within this 15-mile radius, seven reservoirs or lakes of note exist. These include small, private impoundments for agricultural use (i.e., Arroyo Seco and Wallace - under 3,000 acre feet) and moderate, municipal reservoirs for recreation and domestic, municipal usage (Comanche and Pardee reservoirs and Lake Amador - up to 435,000 acre feet).

4.2.4 Hydrogeology Ground water in the area is found at depths generally greater than 100 feet in the sediments of the Laguna and Mehrten Formations. The sand and gravel zones of these formations yield water readily to wells predominately west of the facility in the Central Valley. At the site Revision 0 RSNGS Historical Site Assessment 18 March 2004

however, the formations are less permeable, and the Laguna Formation is above the water table; depth to water in the vicinity of the site is approximately 150 feet.

Ground water flow is generally to the west. West of the site the flow is affected by a conical depression resulting from rthe ground water pumping center to the Southwest near the town of Galt, Ca. (Figure 4.2)

Water from the Laguna and Mehrten formations is of generally good quality in the vicinity of RSNGS. It is a sodium bicarbonate-type with low total dissolved solids, generally less than 200 ppm. Potable water for RSNGS site 6omes from four wells producing from the Mehrten formation at a depth interval of 200-350 'feet. Two wells are located within the Industrial Area, one well serving the Rancho Seco Reservoir and Recreation Area and one well serving a residence located at the northeastern corner of the site.

Studies performed during the initial sighting evaluation and documented in the FSAR, as well as several conducted since the commencement of operations (Geotechnical Investigation for Proposed Evaporation Ponds, ERPT-C0104, Rev.1, [Ref. 9.12] and the Final Engineering Report Assessment of Spent Fuel Liner Leakage, ERPT-M0221, Rev.0, 1990, [Ref. 9.13]),

indicate that the'permeability of the site soils result in infiltration rates (from several hundred to several thousand years) that effectively preclude any radiological impact on the aquifer or the closest well to the site by the facility.

Revision 0 RSNGS Historical Site Assessment 19 March 2004

Source: Based on measured spring 2000 water level data from Sacramento County Departmont of Water Resources Contour numbers indicate feet from mean sea level (msl)

Credit: Sacramento County 2002 Zone 40 Water Supply Master Plan EIR J

Ehttp: llwww.saccodwr.orgiFilesIWater/EIRIZ40%20Sect%204.7%2OWater.pdf Figure 4.2 Ground Water Contour Map Revision 0 RSNGS Historical Site Assessment 20 March 2004

4.2.5 Meteorologv 4.2.5.1 General Climatology The climate of the RSNGS site is generally that of the Great Central Valley of California.

Summers are hot and cloudless and winters are mild. The rainy season occurs between October and May with more than two-thirds of the annual rainfall occurring in December through March. Heavy fog occurs in mid-winter, primarily in December and January, and may last for several days.

Incidents of severe weather, such as Tornados and thunderstorms are infrequent.

The most controlling geographical influence on climate results from the mountains, which surround the valley to the west, north, and east. During the winter, storms that pass through the area are moderated by the mountains, which collect much of the precipitation. The precipitation that does occur in the valley is usually accompanied by south to southeast winds. The cold north and northwest winds pass over the mountains to the north where the air is warmed dynamically by the descent into the valley resulting in comparatively warm, dry winds. A similar condition occurs infrequently in the summer when a steep pressure gradient develops, producing a pronounced heat wave.

The Central Valley warms greatly during the day resulting in a marked thermal contrast between the valley and the air over the Pacific Ocean. The Coast Range separates the marine air from the valley air except for a gap through the range formed by the Sacramento and San Joaquin Rivers. The heavy marine air flows through this gap and splits into a northerly flow into the San Joaquin Valley and a southerly flow into the Sacramento Valley.

The divergence zone between the two flows usually lies between Stockton and Sacramento near the site. The divergence zone is typically north of the site during the day, resulting in north to northwest winds. As the air in the valley cools, the flow decreases and calm may set in. If the drainage from the Sierra Nevada is sufficient, the winds may shift to southeasterly and increase in speed.

During the hottest mid-summer months, light westerly winds may persist all night. During the winter, the synoptic gradients prevail much of the time and the wind trajectories over the Sacramento-Stockton-RSNGS region are reasonably uniform.

4.2.5.2 Extreme Winds Wind data from Sacramento Executive Airport from 1951 to 1971 were used to conduct an extreme wind probability distribution approximate to the RSNGS site. Table 4.2 presents the highest expected wind speed that will be expected for the indicated recurrence interval.

Revision 0 RSNGS Historical Site Assessment 21 March 2004

1 Table 4.2 Expected Extreme Wind Speeds Return Period (years) Wind speed (mph) 50 90 100 101 1000 149 10000 169 The highest recorded average wind speed for Sacramento during the period of July 1877 through December 1989 was 70 mph (recorded in both December 1952 and November 1953).

4.2.5.3 Tornados Tornados have been recorded in California but with a frequency of only two per year (National Climatic Summary, 1969). They are generally not severe, and in many cases amount to little more than a whirlwind that may cause damage to trees and light structures.

An examination of newspaper accounts of nine tornados in California indicates that only one may have been accompanied by wind speeds higher than 100 mph.

The location of a possible tornado strike can be approximated by a geometrical point. The probability of a tornado occurring at a specific point can be estimated by the principle of geometric probability. If two tornados per year are used, the return period for RSNGS is approximately 27,855 years. Because the intensity of California tornados is much less than the "classical mid-western types", winds in only one of five of these tornados would be expected to exceed 100 mph.

This information is reasonably confirmed by searches conducted of the National Oceanic and Atmospheric Administration's (NOAA) database which result in the following information From 1950 through 1995, California, as a whole, averaged 5 tornados per year. This relates to an average of 0.3 tornados per year per 10,000 square miles.

The annual average number of strong-violent (F2-F5) tornados in California for the same period is zero (0).

4.2.5.4 Tropical Storms and Hurricanes The possibility of severe storms in the area can be limited to thunder storms and tornados. A discussion of tropical storms and hurricanes is not applicable to RSNGS.

Revision 0 RSNGS Historical Site Assessment 22 March 2004

4.2.5.5 Precinitation Extremes The precipitation Climatology of the Great Central Valley is characterized by a dry season from June through September and a rainy season from October to May. No precipitation records were taken from RSNGS, but because precipitation is associated with large-scale synoptic systems, the data in Table 4.3 below, taken from the ISFSI FSAR [Ref. 9.19], are believed to be representative of the site.

The annual rainfall occurs almost exclusively in the winter months.

Table 4.3 Precipitation Climatology Averages (inches)

Month Sacramento Stockton January 3.18 2.55 February 2.99 2.46 March 2.36 2.05 April 1.40 1.14 May 0.59 0.44 June 0.1 0.07 July 0.01 0.01 August 0.02- 0.01 September 0.19 0.19 October 0.77 0.63 November 1.45 1.17 December 3.24 2.66 Total 16.29 13.37 A frequency of occurrence of a given precipitation intensity for Sacramento is presented in Table 4.4 (from the ISFSI FSAR). As stated above, this data is believed representative of the conditions that exist at the site and shows that virtually all of the precipitation falls at a rate of under a quarter inch per hour.

Revision 0 RSNGS Historical Site Assessment 23 March 2004

I1L Table 4.4 Precipitation Intensity Year Intensity (inches/hour) 0.01-0.09 0.10-0.24 0.25-0.49 0.50-0.99 1961 79.5% 17.7% 2.3% 0.5%

1962 81.8% 17.0% 0.8% 0.4%

1963 80.0% 17.8% 2.2% 0.0%

1964 86.2% 11.3% 2.2% 0.3%

1965 89.0% 10.0% 1.0% 0.0%

Average 83.5% 14.6% 1.7% 0.2%

4.2.5.6 Snow and Ice Storms The possibility of severe storms in the area can be limited to thunderstorms and tornados.

Snow in the Sacramento area is extremely rare. Most snow that has been observed in the Sacramento Valley occurs in January. Given the lack of significant snowfall in the region, a detailed discussion of snow and ice is not applicable to the RSNGS site.

4.2.5.7 Thunderstorms Thunderstorms, and associated lighting strike, occur infrequently in the area, with the mean number of days per year with thunderstorm activity ranging between 5 in the Sacramento area to 3 in the Stockton area.

4.2.5.8 Restrictive Dilution Conditions (Inversions)

Inversions occur in the Great Valley as a result of cold air advection near the ground or cooling of the earth causing a cooling of the air near the ground. Radiational cooling occurs at night when there are no low clouds. Both types occur at RSNGS with the advection type usually associated with the westerly wind bringing in cool air from the Pacific Ocean.

Temperature inversions at the ground can be expected to occur every night during the summer upwards to several hundred feet. These temperature inversions are the result of the flow of cool maritime air in to the area during the late afternoon and evening hours. During the winter, shallow (a few hundred feet) but intense surface inversions can be expected occasionally during the nighttime hours under light wind conditions.

Revision 0 RSNGS Historical Site Assessment 24 March 2004

5.0 HSA METHODOLOGY The methodology used for the RSNGS Historical Site Assessment is that found in NUREG-1575, MARSSIM. As described in MARSSIM, RSNGS, being a NRC licensee has much of the HSA related information within the records management system used to maintain its records throughout its operational history.

5.1 Approach and Rationale The primary objective of the HSA records search process was the identification of those events posing a significant probability of impacting'the hazardous or radiological characterization of the site. These included system, structure, or area contamination from system failures resulting in airborne releases, liquid spills or releases, or the loss of control over solid material management.

Each incident identified that posed a'realistic potential to impact the characterization of the site was further investigated. This investigation focused on the scope of contaminant sampling and analysis, remedial actions taken to mitigate the situation, and any post-remedial action sampling, survey, and analysis in an attempt to identify the "as left" condition of the incident location. The records management system provided the source of a vast majority of the documents inspected.

Also included in the research associated with the development of the HSA were:

• Relevant excerpts from written reports and correspondences;

• Personnel interviews, including the use of questionnaires, of current, former and retired plant personnel to confirm documented incidents and identify undocumented incidents; and

• Site inspection, utilizing historic site drawings, photographs, prints, and diagrams to identify, locate, confirm, and document areas of concern.

Information from this research was used in the HSA development, including the compilation of data, evaluation of results, documentation of findings, and the characterization and identification of Areas and Survey Units.

Relevant information that becomes available following the publication of the HSA during the characterization and remediation phases of the License Termination Program will be evaluated and documented. - -

5.2 Documents reviewed In researching the HSA, the records reviewed include:

License and Technical Specification reports;

• Annual operational "and environmental reports;

• Environmental investigations performed by independent entities;

• Regulatory actions against the site; -

Revision 0 RSNGS Historical Site Assessment 25 March 2004

II

• Documentation from interviews conducted with currently employed and retired/separated site personnel;

• Radiological control surveys associated with identified events;

• Site inspection and surveillance documents associated with identified events;

• Federal, State and local regulations;

• Regulatory and Industry guidance documents;

• Annual Environmental and Operational documents;

• License Event Reports (LERs);

• Occurrence Description Reports (ODRs);

• Quality departure documents, including Potential Deviations from Quality (PDQ) and Deviation from Quality (DQ);

• Radiological and environmental survey documents;

• Routine radioactive release reports;

• Non-routine reports provided to the NRC under the provisions of the facility's technical specifications, 10 CFR Part 20, and 10 CFR Part 50;

• Plant incident or condition reports;

• Radiological assessments; and

• Quality control/Quality Assurance finding documents.

Records maintained to satisfy the requirements of 10 CFR Part 50.75(g)(1) provided a major source of documentation for the HSA records review process.

Appendix A contains a summary of the details found in the documents reviewed. A brief incident description, radiological implications, and finding synopsis, are included.

5.3 Site Reconnaissance As provided for in MARSSIM Section 3.5, a formal site reconnaissance was not performed, based on the continuous occupancy of the site by the licensee, the detailed information available through the records, and the personnel interviews performed. Appropriate site reconnaissance has been performed to verify locations and current conditions of items or issues discovered during these investigations.

5.4 Personnel Interviews Between August 2001 and December 2002, approximately 150 observations were noted from the individuals contacted in the HSA questionnaire program. These individuals represented a combination of current and past employees, primarily from the operations and radiation protection staffs. These two groups were chosen due to their knowledge of and association with the systems and source terms being investigated for this assessment.

Revision 0 RSNGS Historical Site Assessment 26 March 2004

The personnel surveys included a combination of questionnaires completed by a majority of the participants as well as individual and group interviews with several of the participants.

With few exceptions, the personnel observations were corroborated by either the observations of other interviewees or documentation discovered during the records search.

Table 5.1 contains a brief summary of the survey results showing the number of observations recorded for the various general areas identified. Appendix B contains copies of the questionnaire and a response summary sheet.

Table 5.1 Personnel Observations Summary General Area of Observation Number of Observations Auxiliary Boiler, pad, drains sump 10 Auxiliary Building 7 "B" Warehouse 4 Barrel Farm (Waste Storage Area) 4 Balance of Plant (BOP) 14 Building Maintenance/Machine 4 Shop "C" Warehouse 2 Circulating Water Basins 16 surrounding area 16 Fabrication and Weld Shops 2 Building Contractor Fab Shop 8 "GRS" Warehouse 1 Interim Onsite Storage Building 5 (IOSB) 5 Non-Radiological Observations 4 Plant Effluent 4 Quonset Hut 11 Retention Basin 2 RHUT's 13 Storm Drains 5 Training and Records Laboratory 1 Tank Farm 13 Tool Room I Turbine Building 9 Tritium Evaporator 3 Training Simulator Building 1 (offsite)

"Upper/Outer" Storage Yard 1 Sewer Plant 1 Revision 0 RSNGS Historical Site Assessment 27 March 2004

it-5.5 Historical Construction Photograph Review Collections of historical construction photographs were reviewed to assess their contribution to this HSA. A selection of construction photographs is included as Appendix E. Also.

additional original construction photographs are contained in the Construction Report issued by Bechtel Corporation [Ref. 9.20]

Revision 0 RSNGS Historical Site Assessment 28 March 2004

6.0 OPERATIONAL HISTORY The following summary of the facility's'history was determined through a review of site records, documents and personnel interviews.

6.1 Introduction The RSNGS was issued its 10 CFR Part 50 operating license (DPR-54) on August 16, 1974 and attained initial criticality one month later, on September 16, 1974. The facility became commercial on April 18, 1975.

The facility is described in multiple licensing documents including:

• "Rancho Seco Nuclear Generating Station Unit 1 Defueled Safety Analysis Report" [Ref. 9.7]; and

• US Nuclear Regulatory Comnission (formerly the US Atomic Energy Commission), Safety Evaluation by the Directorate of Licensing, US Atomic Energy Commission, in the matter of Sacramento Municipal Utility District Rancho Seco Nuclear Generating Station, Unit 1, "Docket 50-312 (SER) [Ref.

9.8].

RSNGS had a pressurized water reactor (PWR) designed and constructed by Bechtel Power Corporation with its nuclear steam supply system (NSSS), rated at 2,770-MWt, 913 MWe, provided by Babcock and Wilcox. Condenser cooling and make-up water was provided via the Folsom-South canal, constructed by.the Bureau of Reclamation.

The RSNGS site is located in southern Sacramento ,County, California, approximately 25 miles southeast of Sacramento and 26 miles n6rtheast of Stockton. The site is located on 2,480 acres entirely owned by the District. The facility is located between the Sierra Nevada Mountains to the east, and the Pacific Coast range bordering the Pacific Ocean to the west.

The rural area is used almost entirely for agricultural purposes including row and silage crops, cattle graze land, and in recent years, grape production. Within the five-mile radius of the site, there are no significant tourist attractions or variations in population. The nearest population area is approximately 6.5 miles from the site while the closest substantial populations (>20,000) are Galt, and Lodi, CA. at 10 and 17 miles from the site, respectively.

The main access to the site is State Highway 104 (Twin Cities Road), which runs from highway 99 (just north of Galt, CA) in the west, to State Highway 88 (just east of lone, CA) to the east.

After approximately 15 years of operation, RSNGS was'shut down for the last time on June 7, 1989, after passage of a non-binding referendum by the voters of Sacramento County recommending the District discontinue operation of RSNGS..

The reactor was completely defueled on December 8, 1989 Unable to attract a buyer for the facility, the District formally notified the U.S. Nuclear Regulatory Commission (NRC) of its intent to permanently shut down the facility, requesting a possession-only license on April 26, 1990.

Revision 0 RSNGS Historical Site Assessment 29 March 2004

As noted in the "Rancho Seco Nuclear Generating Station Proposed Decommissioning Plan" (PDP) [Ref. 9.9], RSNGS operated for approximately 2,149 effective full power days (seven fuel cycles), over the course of its operating lifetime.

A summary of the operational history is provided in Table 6.1 below.

TABLE 6.1 Operational History - RSNGS Date Event Oct. 1968 Received construction permit Mar. 1969 Commenced site preparation/construction Aug. 1974 Operating License (OL) issued Aug. 1974 Completed initial fuel loading Sept. 1974 Achieved initial criticality Apr. 1975 Commenced commercial operations Jun. 1975 - Oct Two unplanned outages to repair material deficiencies. Full power 1976 achieved in Mar. 1976. Full power regained in Oct. after 7-month stator coil outage.

1977 8 months of full power operations (75% capacity factor Jul-Dec.)

Nov. 1978 Completed cycle three refueling in 35 days Aug. 1980 Turbine rotor failure resolved Jun. 1982 Frequent electrical inverter trip resolution achieved Apr. 1983 Turbine oil system associated trip issues resolved Aug. 1984 Steam Generator repairs and Aux. Feed water modification outage Dec 1985 Extended plant shutdown resulting from overcooling unusual event Extended plant shutdown for post TMI-mod installation, emergency Mar. 1986-88 feed water system modifications, detailed system analysis and test program implemented, and installation of two additional backup diesel generators.

Resolved feed water transient issue, completed restart testing. Public Jun 1989 referendum voted to have SMUD discontinue operation of RSNGS.

Plant shuts down for last time on June 7, 1989.

Aug. 1989 SMUD notifies NRC of its intent to seek a decommissioning amendment to its license.

Revision 0 RSNGS Historical Site Assessment 30 March 2004

TABLE 6.1 Operational History - RSNGS (Continued)

Date Event District fails in its attempts to sell RSNGS or convert to non-nuclear Sept. 1989 oeain operation. .-.-

Dec. 1989 Reactor defuelinrg completed on December 8, 1989.

Jul. 1990 SMUD submits the Plan for Ultimate Disposition of the Facility in response to NRC request.

MAY - SMUD submits RSNGS Proposed Decommissioning Plan (PDP) 1991 October - Board approves California Environmental Quality Act "Negative Declaration" for PDP (State clearinghouse number

- (SCH#) 91062072)

Mar. 1992 RSNGS OL amended to Possession Only Mar. 1995 NRC approves PDP January - SMUD Board approves Incremental Decommissioning Action Plan (IDAP for 1997 through 1999) and California Environmental Quality Act "Subsequent Negative Declaration":

1997 (SCH# 96112047) for IDAP Post Shutdown Decommissioning Activities Report (PSDAR)

Submitted IAW 10 CFR Part 50.82 (PSDAR supercedes the' PDP)

January 1999 - SMUD Board approves IDAP - Rev. #1 (continue decommissioning through license termination) and CEQA 1999-2000 "Subsequent Negative Declaration" for IDAP Rev. #1

-_ _ (SCH#99042092) ,

Jun. 2000 June 30, 2000 - NRC issued SMUD a 10 CFR Part 72 license to store n.* ,. RSNGS's spent nuclearfuel at the ISFSI

.Spent fuel transfer'to ISFSI complete 'TS amendimjents. 129 and 130 take affect -'preclude's'SF posses'sion on'the 10 CFR Part 50 licensed Aug. 2002 facility and eliminates the need for an Operations Shift Supervisor or Certified Fuel Handlers TS axhend. 131 'taks effect eliminating 'security plan requirements from the 10 CFR Part 50 licensed facility Revision 0 RSNGS Historical Site Assessment 31  ! March 2004

The fact that the plant was shut down years before the expiration of its operating license resulted is several significant impacts, two of which include:

• The District's inability to comply with the requirements of 10 CFR Part 50.75 regarding the submission of a preliminary decommissioning plan five years prior to the cessation of operations; and

• A significant shortage of funds within the decommissioning trust fund.

6.2 Decommissioning Plan Chronology Prompted by a NRC staff request, the "Plan for Ultimate Disposition of the Facility" (PUDF), was submitted in July 1990 [Ref. 9.10]. The original intent of the licensee, as outlined in this document, was to decommission RSNGS using the SAFSTOR - Deferred DECON alternative. This alternative was to include Custodial, as well as Hardened, -

SAFSTOR applications as generally defined in the "Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," NUREG-0586, August 1988 (GEIS)

[Ref. 9.11]. Dismantlement following the SAFSTOR period was estimated to occur in the 2008 to 2012 time frame.

On May 20, 1991, the District submitted the PDP for the Rancho Seco facility, dated April 15, 1991, for NRC approval. The District subsequently submitted supplements to the PDP for review dated April 15, August 6, & August 31, 1992; January 7, April 7, & April 19, 1993; and March 23, April 28, July 26, & October 26, 1994. After an extensive NRC staff review, the PDP was approved on March 20, 1995.

Simultaneous with this review was the amendment of the District's Operating license (DPR-54), to reflect a possession-only authorization on March 17, 1992 and the NRC staff's review of the associated safety evaluation and environmental assessment of the impacts associated with the decommissioning of RSNGS resulted an initial Finding Of No Significant Impact (FONSI), issued on June 16, 1993.

Also occurring during this period was the decision by the District to commence, in an incremental manner, the dismantlement of the site during the Custodial SAFSTOR period.

In 1991, the District Board of Directors approved the negative declaration prepared for the original PDP (Resolution No. 91-10-18) on October 17, 1991. (State Clearinghouse No.

91062072)

In January 1997, the District Board of Directors approved (Resolution 97-01-07) a significant revision to the Decommissioning Plan titled "Incremental Decommissioning Action Plan" (IDAP) and a subsequent negative declaration regarding the potential environmental impacts.

(State Clearinghouse No. 96112047)

In April of 1999, the District Board of Directors approved revisions to the IDAP (IDAP -

R1) accelerating the schedule of the decommissioning effort. (State Clearinghouse No.

99042092)

In accordance with the applicable provisions of the California Environmental Quality Act (CEQA), the District prepared and circulated the studies and evaluations necessary to support Revision 0 RSNGS Historical Site Assessment 32 March 2004

the subsequent negative declarations associated with the PDP, IDAP, and IDAP -RI. This included multiple public meetings convened by the District and the NRC.

6.3 Regulatory Overview The RSNGS has been, and continues to be, closely monitored in a highly regulated environment. Regulatory oversight is provided by an extensive collection of Federal, State, Local, and licensee personnel in addition to non-regulatory industrial peer groups and local stalceholders.

This hierarchy of oversight has carried out its various responsibilities during the sighting, licensing, construction, operations, and decommissioning phases of the plant's life and includes;

• United States Atomic Energy Commission (AEC);

• United States Nuclear Regulatory Commission;

• United States Environmental Protection Agency (EPA);

• US Army Corps. Of Engineers - Bureau of Reclamation;

• California Department of Health Services - Radiological Health Branch;

• California Department of Toxic Substance Control (DTSC);

• California Regional Water Quality Control Board;

• State Water Resources Control Board;-

• California Department of Fish'and Game;

• Sacramento Metropolitan Air Quality Management District;

• Local/County Governments; and

• District Regulatory affairs/licensing.

6.3.1 Pernits and Licenses Permits, issued to the District in association with the construction, operation, and dismantlement of RSNGS, are'summarized in Table 6.2 Revision 0 RSNGS Historical Site Assessment 33 March 2004

It-TABLE 6.2 Licenses and Permits Issuing Agency Permit/License Number USAEC CPPR-56 Construction Permit - Pressurized Water Nuclear Plant USAEC SNM-1333 SNM License USAEC 04-14866-01 Byproduct Materials License California Department of Health - Radiation Health Section 2239-34 Radioactive Materials License USAEC #DPR-54 Possession Only License (POL)

USAEC #DPR-54 Facility Operating License (FOL issued 8/16/74)

California Regional Water Quality Control Board # CA0004758 NPDES permit California Regional Water Quality Control Board Annual order renewal Waste Discharge Order California Department of Health - ELAP 1681 (Environmental Lab analysis permit)

County of Sacramento - Air Pollution Control District #292 Permit to operate Steam Boiler #3677 (E-365)

County of Sacramento - Air Pollution Control District #293 Permit to operate Steam Boiler #3680 (E-360)

County of Sacramento - Air Pollution Control District #294 Permit to operate Diesel Generator (G866A)

County of Sacramento - Air Pollution Control District #295 Permit to operate Diesel Generator (G866B)

County of Sacramento - Air Pollution Control District #7731 Permit to operate Diesel Generator (G1OQA)

County of Sacramento - Air Pollution Control District #7732 Permit to operate Diesel Generator (GIOOB)

County of Sacramento - Air Pollution Control District #4175 Permit to operate Gasoline Dispensing Facility County of Sacramento - Air Pollution Control District None Permit to operate Steam Boiler V-200 Reactor Pressure Vessel Sacramento Regional County Sanitation District #LWH-5/98 Class II discharge permit Revision 0 RSNGS Historical Site Assessment 34 March 2004

TABLE 6.2 (continued)

Licenses and Permits Issuing Agency Permit/license number State of California Department of Public Health Laboratory Services None Sanitation and Radiation Laboratory.

Non-Commercial Water Laboratory.

Sacramento County #232 Health Department - Environmental Health Branch Non-Community Water System Permit Sacramento County Health Department - Environmental Health Branch #302 Non-Community Water System Permit Sacramento Metropolitan Air Quality Management District (Formerly - County of Sacramento, Air Pollution Control District)

Operation of abrasive blasting booth and bag house exhaust vent Sacramento Metropolitan Air Quality Management District 11345 Operation of stand-by diesel driven fire pump Sacramento Metropolitan Air Quality Management District 11344 Operation of 80 horse power Diesel back-up electrical generator Sacramento Metropolitan Air Quality Management District 13392 Operation of gasoline storage and 'dispensing station (One 4000 gallon tank with one nozzle)

Sacramento Metropolitan Air Quality Management District 13833 & 13834 Operation of unconfined abrasive blaster and bag house exhaust vent Sacramento Metropolitan Air Quality Management District 13769 & 13770 Operation of abrasive blasting booth and bag house exhaust vent Sacramento Metropolitan Air Quality Management District 11343 Operation of stand-by air compressor (Gasoline driven)

State of California- 2.6.921.84 Agricultural & Services Agency, Depa'rtment of Industrial Relation, Division of Industrial Safety Operate bridge crane (TDI diesel cranes) -'Load down rated State of California Agricultural & Services Agency, DI6partment of Industrial Relation, 2.6.922.84 Division of Industrial Safety.

Operate bridge crane (TDI diesel cranes) - Load down rated -

State of California 6.25 Agricultural & Services Agency, Department of Industrial Relation, .2.1249.85 Division of Industrial Safety Operate bridge crane (IOSB crane No. Y-1 12)

Revision 0 RSNGS Historical Site Assessment 35 March 2004

II TABLE 6.2 (continued)

Licenses and Permits State of California 7955 Department of Industrial Relations Division of Occupational Safety and Health Operation of Auxiliary Building grade level monorail crane No. A-i State of California 8159 Department of Industrial Relations Division of Occupational Safety and Health Operation of Reactor Building polar crane No. Y-204A State of California 8013 Department of Industrial Relations Division of Occupational Safety and Health Operation of Turbine-Building gantry crane No. Y-304 State of Washington 6170 Department of Social and Health Services Disposal site use permit Cooling tower transite removal /Bechtel building asbestos removal Provided by permits contractor State of California Various Department of Industrial relations Pressure vessel Permits 6.4 Waste Handling Procedures Waste materials generated at RSNGS are generally described as radioactive, hazardous, mixed (radioactive/hazardous), universal, or non-regulated.

To ensure the conformance with prescribed regulatory requirements, waste handling evolutions are controlled through various administrative and operational procedures.

6.4.1 Process Control Program (PCP)

The PCP established a program to provide the District and regulators with a reasonable assurance that the radioactive wastes generated at the facility are properly classified, characterized, processed, packaged, manifested, marked, labeled and transported in accordance with the wide spectrum of regulations governing these activities.

6.4.2 Rancho Seco Administrative Procedures (RSAP)

RSAP's provide general departmental guidance in the control of various activities within the facility including those associated with radioactive and hazardous material/waste handling Revision 0 RSNGS Historical Site Assessment 36 March 2004

6.4.3 Radiation Control Manual (RCM)

The RCM provides implementing procedures for-the control of radioactive material including training requirements, material receipt procedures, and controls for the release of personnel and materials from the controlled area. -

6.4.4 Radwaste Control Manual (RWCM)

The RWCM provides implementing procedures for the management of radioactive waste generated at the RSNGS including material receipts, waste classification, container selection, waste-stream specific processing procedures and characterization verifications.

6.4.5 Chemistry Department Procedures Manual These procedures include chemical controls, off-site dose calculation, and radioactive effluent control implementing procedures.

6.4.6 Surveillance Procedures (SP)

SP's are used to document the performance'of tests to demonstrate the effectiveness and efficiency of the various implementing procedures and performance of equipment and personnel activities associated with the waste management program.

6.5 Current Site Usage RSNGS was shut down on June 7, 1989.

6.5.1 Description of Operations -

Current operations, post 'fuel transfer to the ISFSI, center on the administrative, technical and physical tasks associated with the dismantlement of the RSNGS.

In August 2002, the transfer of spent fuel from the' spent fuel pool to the ISFSI was completed and RSNGS transitioned into the Defueled Technical Specifications, greatly reducing the procedural and operational'controls required at the facility.

6.5.2 Preliminary Site Characterization The initial characterization of the RSNGS site resulted from the review and evaluation of surveys and evaluations previously conducted to' determine the extent and nature of residual contamination. In accordance -with the'guidance of MARSSIM, this initial site characterization (as to the.Impacted orNojn-Impacted nature of the site) beganf in 2001 and wa's completed in 2002.'Th'e HSA including the initial site characterization is the product of the evaluations 'and investigation necessar' to define the current condition at the site and assign preliminary Area classifications. This effort also addressed the hazardous material and "state-only" regulated material at the site that may impact future remediatiorn/dismantlement.

Revision 0 RSNGS Historical Site Assessment 37 March 2004

II 6.6 Site Dismantlement 6.6.1 Dismantlement activities within the Power Block As of January 2003, the decommissioning project has removed virtually all (with the exception of imbedded or buried piping) of the secondary plant systems including:

• Main Steam;

• Auxiliary Steam;

• Main Feed Water;

• Main Condensate and Make-up;

• Main Circulating Water Pumps; Main turbine and Condenser; and

• The vast majority of the support systems located in the Turbine Building.

Within the Auxiliary Building, a majority of the systems have been removed. Spent Fuel Building dismantlement began in October 2002.

Within the Reactor Containment Building, significant progress has been made including removal of all four reactor coolant pumps and motors, a substantial portion of the reactor coolant system, reactor building ventilation system, and support/electrical/mechanical systems.

6.6.2 Dismantlement activities outside the Power Block Dismantlement activities outside of the facility power block are directed at the removal of temporary buildings and structures and are being carried out in accordance with standard site procedures for the release of potentially contaminated materials and equipment. Final Status Survey's will be conducted of the "footprint" left from these structures' dismantlement to verify that no residual contamination above the established derived concentration guideline level (DCGL) will remain following license termination.

6.7 Radiological Sources The majority of regulated waste resulting from the decommissioning of the RSNGS will result from the radiological contamination of plant structures and equipment. The primary source of this contamination was the operation of the facility nuclear reactor and its associated support systems. Based on information developed for the IDAP (section 3.1.3), the radiological inventory of the facility is described in the following sections.

6.7.1 Spent Fuel The largest single contributor to the radioactive inventory at the facility was spent fuel.

Based on estimates performed in 1989, 140,800,000 curies, consisting of primarily (-70%)

Cs-137, Pr-144, Ce-144, Ba-137m, Sr-90, Y-90, Pm-147, and Pu-241 remain. This will decay to approximately 39,630,000 curies by 2009 with Cs-137, Ba-137m, Sr-90, Y-90, and Pu-241 representing over 97% of the remaining activity.

Revision 0 RSNGS Historical Site Assessment 38 March 2004

6.7.2 Irradiated Hardware Non-Fuel contributors to the radiological inventory, estimated at approximately 95,000 curies of primarily (>99%) Co-60, Fe-55, and Ni-63, include:

• Orifice rod assemblies (ORA's);

• Burnable poison rod assemblies (BPRA's);

• Retainer assemblies (RA's); and

• Incore instruments.

The ORA's, BPRA's, and RA's were transferred to the ISFSI, along with the spent fuel. The incore instruments were sectioned for shielded storage pending transfer to the ISFSI. The activity associated with these irradiated componenits will decay to under 9,000 Ci of Co-60 and Ni-63 by 2009.

6.7.3 Reactor Vessel and Internals As of May 1,2003, approximately 99,500 curies of primarily (>61 %) Co-60; comprised of reactor pressure vessel internals and the reactor pressure vessel, are contained within the primary shield wall [Ref. 9.21].

6.7.4 Plant Systems Systems internally contaminated by the operation of the RSNGS have been characterized repeatedly during plant operations. The most substantial of these characterizations was performed in 1984 by Pacific Northwest Laboratory (PNL-1546, 1984) [Ref. 9.22] showing an estimated 4,500 curies resulting primarily of (>88%) Fe-55, Co-58, Ni-63, and Co-60.

6.7.5 Industrial Area Contamination Several areas within the Industrial Area have been identified as having been radiologically impacted by the operation of the facility including:

. Retention Basins; Tank Farm; Barrel Farm;

• Regenerant Hold Up Tanks (RHUT's);

• Storm Drains;

• Oily Water Separator;

• Cooling Tower Basins; and  ;:

• Turbine Building drains and sumps.

Revision 0 RSNGS Historical Site Assessment 39 March 2004

II 6.7.6 Non-Industrial Area Contamination Four locations outside of the Industrial Area have historically had radionuclide concentrations detected above background.

6.7.6.1 Discharge Canal Sediment The plant discharge canal sediment has shown detectable concentrations of licensed radioactive material resulting from 10 CFR Part 20.2001(a)(3) authorized radioactive liquid releases. This release path has been the subject of numerous studies by the facility staff as well as the Lawrence Livermore National Laboratory (LLNL) and is routinely monitored via the Radiological Environmental Monitoring Program. As discussed in the PDP, the most recent of the LLNL studies (UCRL-ID-1061 11, November 1990) reported maximum radioactive sediment concentrations of 1.47 pCi/g Co-60 (April 1989), 1.20 pCi/g Cs-134 (January 1989), and 11.00 pCi/g Cs-137 (January 1989) at points within 1,640 feet (0.5 kilometer) of the plant effluent discharge point (0.3 km for January 1989 sampling, 0.5 km for April 1989 sampling).

Current (March 1, 2004) concentrations can be estimated to be < 0.207 pCi/g Co-60, 0.007 pCi/g Cs-134, and 7.74 pCi/g Cs-137 (based on the radioactive decay of the 1989 results).

Washout and other transport mechanisms will have also affected the concentrations of radioactive material in the effluent discharge path. RSNGS will update the status of this source term as additional studies are completed.

Oak Ridge National Laboratory also evaluated the environmental impact of the authorized radioactive liquid releases for the NRC. This evaluation was applied to both onsite and offsite locations. The results of this evaluation are documented in NUREG/CR-4286, Evaluation of Radioactive Liquid Effluent Releases From the Rancho Seco Nuclear Power Plant [Ref. 9.15].

As part of the Radiological Environmental Monitoring Program and reported to the NRC in the 2002 Annual Radiological Environmental Operating Report [Ref. 9.16], 24 samples of sediment were collected from the discharge canal and the Clay/Hadselville/Laguna Creeks during 2002. Gamma spectrometry analysis of these samples indicated the presence of Cs-137 in the range of 0.017 to 0.604 pCi/g with a mean of 0.111 pCi/g and Co-60 in the range of 0.008 to 0.035 pCi/g with a mean of 0.021 pCi/g.

6.7.6.2 Discharge Canal Soil During plant operation and during the period of authorized radioactive liquid releases, discharge canal sediment was dredged from the canal and deposited as a band adjacent to the canal. Because the discharge canal sediment was known to contain radioactive materials of plant origin, sampling of the soil adjacent to the discharge canal was added to the Radiological Environmental Monitoring Program. As reported to the NRC in the 2002 Annual Radiological Environmental Operating Report, eight soil samples were collected from this area. Cs-137 was identified in seven out of eight of these samples at a concentration range of 0.042 to 0.266 pCi/g.

Revision 0 RSNGS Historical Site Assessment 40 March 2004

6.7.6.3 Depression Area Soil The depression area is an onsite location adjacent to "No Name" Creek. The discharge canal, discussed above, -flowsinto "No Name" Creek.` On occasion and during periods of authorized radioactive liquid releases, "No Name" Creek overflowed and collected in the depression area. Because of this, sampling of the soil in the depression area was added to the Radiological Environmental Monitoring Program. As'reported to the NRC in the 2002 Annual Radiological Environmental Operating Report, 14 soil samples were collected from this area. Cs-137 was identified in 12 of these 14 samples'at a concentration range of 0.070 to 48.15 pCi/g. Cs-134 was identified in two samples at a concentration range of 0.060 to 0.177 pCi/g. Co-60 was identified in six samples at a concentration range of 0.086 to 1.10 pCi/g.

6.7.6.4 Storm Drain Outfall The Radiological Environmental Monitoring Program has routinely identified low levels of radioactive materials of potential plant origin in soil samples taken at storm drain locations.

As reported to the NRC in the 2002 Annual Radiological Environmental Operating Report, 30 soil samples were collected from 15 storm drain outfall locations during 2002. Gamma spectrometry analysis of these samples indicated the presence of Cs-137 in the range of 0.013 to 0.102 pCi/g with a mean of 0.043 pCi/g and Mn-54 in one sample at a concentration of 0.007 pCi/g.

During the fourth quarter of 2000, Shonka Research Associates, Inc. (SRA) conducted detailed surveys of selected areas outside of the Industrial Area. These surveys were conducted to support consideration of an area south of the Industrial Area proposed for the Cosumnes Power Plant (CPP) to be constructed on the RSNGS site. The surveys also determined the boundary of any Impacted Areas and determined background survey values for comparison to Impacted Area values. These surveys included scan surveys conducted using the Subsurface Multi-Spectral Contamination Monitor (SMCM) system developed by SRA, fixed point in situ NaI(TI) spectroscopy measurements and soil sampling for laboratory analysis. To manage the surveys, the site was divided into twelve survey areas. Non-Impacted Areas'required 10% areal scan surveys and Non-Impacted Areas bounding Impacted Areas required 50% areal scan suiveys.-

The final report on these surveys noted that due to several factors, including the marshy conditions of the fields to the south of the plant, several in situ sample points had to be relocated. According to the study's authors, this relocated configuration represented the best combination of complete west-east coverage along the storm drain outfall area to the south of the plant.

The SMCM scan and the in situ measurement survey results for the outfall area immediately south of the Industrial Area and for the proposed CPP location showed no evidence of plant-derived contaminants in these areas. Cs-137 MDCs for'the SMCM scans of these areas ranged from 0.26 to 0.77 pCi/g and for in situ measurements from 0.31 to 0.40 pCi/g. Two out of five soil samples from these areas tested positive for Cs-137 at a range of 0.03 to 0.30 pCi/g with an analysis MDA of 0.03 pCi/g.

NUREG/CR-4286 established Cs-137 background concentrations in the vicinity of RSNGS.

Four locations, at distances of 4 to 10 miles from RSNGS and lying approximately north, Revision 0 RSNGS Historical Site Assessment 41 March'2004

It-south, east and west of the site. The average concentration of Cs-137 in these locations was 0.41 pCi/g. Decaying this average value from December 1984 (the approximate sampling date for NUREG/CR-4286) to December 2002 gives a background concentration of 0.27 pCi/g.

6.7.6.5 Comparison of Soil Concentrations with NRC Screening DCGLs As discussed above, Mn-54, Co-60, Cs-134 and Cs-137 have been identified at concentrations above background in four locations outside of the Industrial Area. On December 7, 1999 the NRC published screening DCGL values in the Federal Register [Ref.

9.18] for various common radionuclides. These screening DCGL values may be used to evaluate the significance of the soil contamination found outside of the Industrial Area.

The published NRC generic screening DCGL values are as follows:

Mn-54 15 pCi/g Co-60 3.8 pCi/g Cs-134 5.7 pCi/g Cs-137 11 pCi/g Based on these values, the discharge canal sediment, the discharge canal soil and the storm drain outfall soil will likely not exceed site-specific DCGL values. The depression area soil does exceed the NRC screening DCGL values. However, it must be compared with site-specific DCGL values developed for surface soils before determination if remediation is necessary.

6.8 Waste Stream Description 6.8.1 Hazardous Materials/Wastes The RSNGS site contains a variety of hazardous materials. The use, storage, handling, and disposal of these materials are controlled through the same procedures and programs used.

during the operation of the facility. In addition to the material management programs in place, the District complies with the OSHA Hazard Communication Standard (29 CFR Part 1910.120) that requires all employers to provide information to its employees about the hazardous substances that they may come into contact with. This is accomplished through the District's Hazard Communication Program that includes training, labeling, other forms of warning, and the availability of Material Safety Data Sheets (MSDS). The District is not abandoning the site, nor do they intend to discontinue the possession or use of hazardous materials or any permits associated with their use. Therefore, the review of hazardous material events for this HSA has not been performed in the same detail as would be done for a site that is to have an unrestricted release from the aspect of hazardous materials.

6.8.1.1 Universal Waste Universal waste means any of the following hazardous wastes that are managed under the universal waste requirements of 40 CFR Part 273:

Revision 0 RSNGS Historical Site Assessment 42 March 2004

• Batteries as described in 40 CFR Part 273.2

• Pesticides as described in 40 CFR Part 273.3

• Thermostats as described in 40 CFR Part 273.4.

Additionally, in accordance with California regulatory requirements, "the hazardous wastes listed in this section are exempt from the management requirements of chapter 6.5 of division 20 of the California Health and Safety Code and its implementing regulations except as specified in chapter 23 and, therefore, are not fully regulated as hazardous waste. The wastes listed in this section are subject to regulation under chapter 23 and shall be known as "universal waste."

• Batteries as described in section CCR 66273.2;

• Thermostats as described in section' CCR 66273.4;

• Lamps as described in section CCR 66273.5;

• Cathode ray tube material as described in CCR 66273.6;

• Aerosol cans as specified in Health and Safety Code section 25201.16;

• Mercury-containing motor vehicle light switches as specified in Health and Safety Code section 25214.5 (MQO1 Wastes) and motor vehicles that contain such switches, as described in'section 66273.7.1);

• Non-automotive mercury switche~s and products that contain such switches (including, but not limited to, M002 Wastes), as described in section 66273.7.2;

• Mercury-containing pressure or vacuum gauges, as described in section 66273.7.4;

• Mercury counterweights and dam'pers, as described in section 66273.7.6;

• Mercury thermometers, as described in section 66273.7.7;

• Mercury dilators and weighted tubing, as described in section 66273.7.8;

• Mercury-containing rubber flooring, as described in section 66273.7.9; and

• Mercury gas flow regulators, as described in section 66273.7.10.

6.8.1.2 RCRA Waste -

The California DTSC and EPA regulate the packaging, storage, processing and disposal of listed or'characteristic waste materials. RSNGS must demonstrate compliance with both the federal EPA and State program requirements. Material at RSNGS in this category include:

• Polychlorinated Biphenyls (PCBs);.;

• Asbestos-Containing Material (ACM);

• Laboratory solvents and reagents; '

• Chrome containing waste materials; Revision 0 RSNGS Historical Site Assessment 43 March 2004

__________iL

• Lead waste;

• Mercury waste from instrumentation;

• Cadmium waste;

• Corrosive waste from laboratory and cleaning processes;

• Spent aerosol cans;

• Ignitable waste from Laboratory and maintenance activities; and

• Paint related waste solvents from maintenance activities.

6.8.1.3 Mixed Waste Mixed wastes are those wastes regulated by the EPA or equivalent state agency, that are also contaminated with radioactive material. At RSNGS, these wastes include,

• Mercury from radioactive system sampling or monitoring instrumentation;

• Chromium containing radioactive air filters generated during decommissioning;

• Radiologically contaminated solvent from the laboratory and painting and decontamination processes; and

• Radiologically contaminated lead from shielding and paint.

6.8.2 Low Level Radioactive Waste (LLRW)

Low level radioactive waste means those waste materials contaminated with radioactive material. LLRW is collected, characterized, classified, packaged and shipped for either processing or disposal at appropriately licensed facilities.

Between 1974 and 1990, RSNGS made 459 LLRW shipments in support of facility operations.

309 dry solid shipments - 147,000 ft3 150 bulk liquid shipments - 458,000 gals. (Prior to the discontinuation of liquid waste shipments in 1980).

After the station's shut down in 1989, the site made only 8 waste shipments totaling less than 4,500 ft3 between 1990 and 1992 and did not ship radwaste again until 1997 when dismantlement activities in the Turbine building commenced. The original decommissioning waste volume estimate, based on the 1991 PDP, was estimated to be approximately 200,000 ft3 .

This material can include:

• Dry Active Waste (DAW) - paper, plastic, glass, wood, used PPE, scrap metal, floor sweeping, etc.;

• Contaminated asbestos insulation material Revision 0 RSNGS Historical Site Assessment 44 March 2004

• Soil and soil like debris including rubblized concrete and asphalt from various site yard areas;

• Equipment, tanks, pumps, motors, generator, and other metal components;

• Sludge's - organic and inorganic solids from tanks, pipes, and pumps; and

• Charcoal - contaminated filter media used to filter liquid process system and ventilation systems.

Since the commenc&ment of incremental decommissioning in 1997, several changes to the volume estimates and inventories have been made as well as significant 'progress in the dismantlement and disposal of wastes. Spanning the period 1997 to 1999, the successful incremental phase 'of the decommissioning project demonstrated that decommissioning could be effectively undertaken by completing the dismantlement and disposal of the secondary system at a cost avoidance of approximately 42 million dollars over what had been estimated.

This success provided the basis for the Board's approval of full scale decommissioning, which commenced in 2000.

The progress'of the decommissioning program to date (December 31, 2003) and the revised projection of remaining waste volumes are summarized below.

Class A

• Shipped to date -190,000 ft3 o Processors_ 59,000 ft3 o Disposal - 131,000 ft3

• Remaining - 123,274 ft3 Class B

• Shipped to Date - none

• Remaining - 949 ft3 Class C

• Shipped to date - none':

• Remaining - 424 ft3 GTCC-The estimated volume of Greater than Class C (GTCC) waste, resulting from the dismantlement of the reactor internals, is approximately 48.4 cubic feet. Current plans call for the amendment of the ISFSI 10 CFR Part 72 license to accommodate the storage of the GTCC material within the .ISFSI until its final disposition, anticipated to be in the Spent Fuel Repository.

Revision 0 RSNGS Historical Site Assessment 45 March 2004

II 6.8.3 Spent Fuel The spent fuel transfer to the ISFSI was completed in August 2002. Spent fuel will remain in the ISFSI until the Federal Spent Fuel Repository becomes operational, sometime after 2010 based on current estimates attained from the Department of Energy's Yucca Mountain Project website.

6.9 Incident Descriptions Based on the review of existing plant records (annual and semi-annual reports, licensee notifications, Occurrence description reports, and PDQ's) approximately 260 incidents with radiological or hazardous material implications occurred between commencement of plant operation in 1974 and approval to continue decommissioning through license termination in 1999. A number of these took place within the power block and, while contributing to the radiological contamination of the power block structures, were generally contained within the RCA. Those occurring outside of the power block have contributed to the Impacted classification of substantial portions of the industrial area. These include:

• Airborne releases with structural or geological contamination potential;

• Spills outside of the power block or incidents involving potential contamination based on system leakage from systems that had been historically contaminated by primary to secondary leaks;

• Loss of control of radioactive materials resulting in the potential for contamination outside of the power block;

• Plant liquid radioactive effluents resulting in soil contamination;

• Hazardous material spills or losses of control; and

• Contamination of systems not originally designed as radioactive systems outside of the historic power block.

A summary index of these incidents is included as Appendix A.

6.9.1 Radiological Spills The records search showed that between 1974 and 1999, 158 documented spills occurred at the facility. Less than forty of these documented spills occurred within the power block and, while contributing to the radiological contamination of the power block structures, were generally contained within the radiologically controlled drains and waste systems. These spills and releases can be grouped into three basic categories as described below.

• Spills that were ultimately contained within the site's controlled process drain system (including the oily water separator, RHUT's, and retention basins),

contaminating the surfaces between the spill site and drain; Spills ultimately entering the site's uncontrolled storm drain system contaminating the drain system as well as the surfaces between the spill site, the drain and the outfall; and Revision 0 RSNGS Historical Site Assessment 46 March 2004

• Spills resulting in the saturation'iid contamination of the media in the immediate area surrounding the spill (i.e., c6ncrete, soil, asphalt, gravel' etc.).

These spills generally resulted in the affected areas being designated as Impacted Areas for FSS design purposes.

6.9.2 Chemical Spills The records search revealed that betwee'n 1974 and 1999, twenty-eight documented cases involving the mishandling or loss of control over hazardous chemical materials exist. These range from spills of acids and caustics used in the plant's various systems to anti-freeze and transmission fluid from District vehicles. There were a minimal number of chemical spills occurring outside of the building comprising the historic power block. A majority of these occurred within one of the facility's structures.

These spills were controlled and remediated in accordance with the policies and procedures associated with these occurrences, including:

• Ranch Seco Hazardous Materials Business Plan;

• RSAP - 0229, Hazardous Waste Management;

• RSAP - 0223, Oil Spill Prevention, Control, and Countermeasures,

• OP-C-32, Onsite Oil Spill;.

• OP-C-46A, Hazardous Material Spill/Release; and

• Rancho Seco Em'ergency'Plain. '

6.9.3 Loss of Material Control The records search showed that between 1974 and 1998, there are 12 documented cases regarding the loss of control of radioactive material or material contaminated with radioactive material resulting in the potential for contamination spread in the immediate vicinity. Areas affected by these incidents will be initially classified as Impacted Areas.

6.9.4 System Cross-Contamination  :

Starting in 1975, with indications of cross 'contamination of the CCW system from the RCS and expanding dramatically in 1981 'with th'e'first indications of primary to secondary, leakage through the OTSGs, systems not originally expected to contain radioactivity became contaminated. The level 'of contamination varied from system to system and in general, was minimal.

In accordance with plant chemistry and surveillance procedures, open cycle and closed cycle cooling systems; auxiliary systems, tankage, and standing water were routinely monitored. In accordance with the guidance of NRC IE Notice 80-10, non-contaminated systems were routinely monitored for radioactivity, and those systems with measurable activity were evaluated (typically through an Engineering or 10 CFR Part 50.59 review process) for potential impacts against the 19 CFR Part 50 Appendix I criteria.

Revision 0 RSNGS Historical Site Assessment 47 March 2004

II The potential exists for leaks from these systems to have resulted in the contamination of additional site systems and locations not originally expected to be contaminated.

Based on the records search performed for the HSA investigation, 165 documented cases involving events of this nature occurred during the operation of RSNGS. These areas, primarily within the Turbine Building and Tank Farm, are classified as Impacted Areas.

6.10 Survey Unit Identification and Classification 6.10.1 Site Classification The identification, designation, and classification of individual survey units are an ongoing process that will be completed prior to submittal of the Final Status Survey Plan contained within the License Termination Plan.

6.10.2 Assessment Performance The Site Characterization working group of the Decommissioning Planning Team will perform the assessments required to assign preliminary Area and Survey Unit classifications, and Survey Unit identification codes to the site.

6.10.3 Areas The entire 2,480 acre site is divided into Areas. Areas are typically larger physical sections of the site that may contain one or more survey units depending on their classification. Areas that have no reasonable potential for residual contamination are classified as Non-Impacted Areas. These Areas have no radiological impact from site operations and are typically identified early in decommissioning. Areas with reasonable potential for residual contamination are classified as Impacted. Impacted Areas of the site are depicted in Figure 6.1, Impacted Area Designations. Areas of the 2,480 acre site not depicted in Figure 6.1 as Impacted are classified as Non-Impacted.

6.10.4 Survey Units A Survey Unit is a physical area consisting of buildings, structures, or land areas of specifically defined shapes and sizes, for which a unique decision will be made regarding if the presence of any residual radioactive material meets or exceeds predetermined release criteria. A Survey Unit is a single contiguous area, whose size is dependent upon its physical characteristics (open land vs. structural building, dry hillside vs. wetland marsh), radiological conditions (Impacted vs. Non-Impacted, remote material storage area vs. a CSCA), and whose operational conditions are reasonably consistent with the exposure modeling used to determine the classification. (A Survey Unit will carry a single classification as described in table 6.3. An area whose physical or radiological conditions mandate multiple classifications will be divided such that each Survey Unit will have a single, consistent classification.)

6.10.5 Initial Designation of Areas Using reasonable and available physical and documented references, nine Areas were identified and assigned Area identification numbers. Except as noted below, Areas one (1)

Revision 0 RSNGS Historical Site Assessment 48 March 2004

through seven (7) are located outside of the Industrial Area while Area eight (8) is comprised of the entire Industrial Area. Area nine (9) contains all portions of the 2,480 acre site not included in Areas one through eight.

Current Area designations (coordinates as referenced on SK-RP-0001, Radiological Characterization Plot Map) are:

• Area 1, Plant Effluent Water Course bounded by AA2, AA16, AF16, AIlS and AY2 (back to AA2);

• Area 2, South Plant Outfall bounded by A119, AI39, A039, and A013 (back to Al 19);

• Area 3, Southern region bounded by AY2, AP13, AP39, and AZ39 (back to AY2);

• Area 4, South Eastern region bounded by Y40, Y66, AY66, and AY40 (back to Y40);

• Area 5, North Eastern region bounded by AE37, T37, U35, V35, V38, X38, X66, and AE66 (back to AE37) - Note: Area 5 contains two Impacted Survey Units; one that is bounded by Q40, R40, R37, U37, U35, V35, V38, X38, X44, and Q44 (back to Q40) plus one consisting of those cells through which the access road to highway 104 passes;

• Area 6, Northern region bounded by AE2, L2, L36, and AE36 (back to AE2);

• Area 7, Western region (excluding ISFSI and that portion transversed by the railroad spur) bounded by M2, M20, N20, Q19, U19, W16, Z16, and Z2 (back to M2); and

• Area 8 - Those portions of the District-controlled Rancho Seco property not included in and surrounded by SA01-SA07. Area 8 (SA08) is also commonly referred to as the Industrial Area and lies primarily within the industrial area fence with the notable exception of parking areas located to the east of the site.

(bounded by M21, N21, P19, U19, X17, AF17, AHl 9, AH39, Y39, Y37, W37, W34, T34, T36, and M36 (back to M21)).

• Area 9 - Those portions of the District-controlled Rancho Seco property not included in Areas 1 through 9.

Areas of the site are depicted in Figure 6.2, Area Designations.

6.10.6 Survey Unit Designation Program The Impacted Areas are being further subdivided into survey units and assigned a unique Survey Unit Identification Number (SUID) along with a preliminary classification.

Descriptions of the initial Survey Units identified are provided in Appendix C. These initial Survey Units may be either further divided or combined and will be classified during design of the FSS.

Revision 0 RSNGS Historical Site Assessment 49 March 2004

II The guidance provided by the MARSSIM Classification matrix is provided in Table 6.3 below.

Revision 0 RSNGS Historical Site Assessment 50 March 2004

Figure 6.1 Impacted Area Designations Revision 0 RSNGS Historical Site Assessment 51 March 2004

Figure 6.2 Area Designations Revision 0 RSNGS Historical Site Assessment 52 March 2004

TABLE 6.3 MARSSIM Survey Unit Classification Matrix Area Survey requirements Area Sampling /direct Isrmn Classification Definition Structures Land. Scan masueingt EMC evaluation Instrument MDC Clsiiain.measurement Areas, as determined during Non- the HSA, having no reasonable No limit No limit None required None required Not required N/A Impacted potential for residual radioactive contamination.

Impacted areas not expected to Impacted contain residual contamination C I a ft _C No limit No limit Judgmental Random : Required 0.1 DCGLW Max Class III above fracion of DCGxLw .:.

[10%]

m C plas 10,000 in2 10-100% Systematic Required 0.5 DCGLW Max Class II exceed DCGLw Impacted areas with potential to exceed DCGLW, isolated Impacted areas to exceed DCGLErC, or m100 2 2,000 m2 100% Systematic Required 0.5 DCGLW Max Class I where remediation has been performed to meet DCGL criteria Revision 0 RSNGS Historical Site Assessment 53 March 2004

it-6.11 Radiological Impact Summaries 6.11.1 Area 1 (SAOI') - Plant Effluent Area Available documentation of the radiological impacts associated with the evaluations from specific incidents during the operational and post operational period include:

6.11.1.1 Licensee identified events Document Equipment/System/location Remarks ODR 75-46 RHUT overflow -1,765 gal overflowed to PE before divert (H-3 only)

ODR 76-79 PE diversion for road construction Flow rate calculation re-altered PE flow measurements verified with minimal impact noted.

ODR 81-192 RHUT sample line discharges Cumulative impact unknown directly to PE (Ž 500 jiLCi Co-60)

[Also ODR81-193, 209]

ODR 84-223 CST (T-358) overflow -900 Release within 10 CFR Part gallons 20 limits 96 pCi H-3, 0.21 !LCi Cs-137 to PE ODR 87-764 System drained in contaminated Cs-137 at 1.75E-7 jICi/ml area removed without sample 55 (no mention of any diversion gal. Dumped down uncontrolled of PE) 1987 semi-annual storm drain between Aux and RB report -1000 gal. Max dose 3.33E-4 mrem.

1988 annual report Cs-137 detected during routine 57 ACi Cs-137 in 3.10 E+06 monitoring gal. release - Est. dose 0.0125 mrem 1988 annual report MSR valve leakage between April Turbine Building floor drains and September to PE - -88 gal. / - 3 ALCi H-3, Cs-134, and Cs-137 released.

PDQ 89-512 Radiological survey results (up to No limits were exceeded 58 uR/hr contact) along creek raise (activity resulted from concerns associated with EPA permitted releases) criteria I Revision 0 RSNGS Historical Site Assessment 54 March 2004

6.11.1.2 Independent Evaluations Conducted Document' EquipmentlSystem/location Remarks UCID-20267 'Rancho Seco Liquid Effluent -Studyto establish and define the Pathway Aquatic and potential exposure pathways Terrestrial Dietary Survey associated with the liquid

- Report - November 30, 1984. effluent releases from RSNGS.

UCID-20295 Concentration of Established basic correlations Radionuclides in Fresh Water between species, diet, size, and Fish Downstream of Rancho radiological concentration of Seco Nuclear Generating Plant common game fish in

- December 27, 1984. downstream waterways. Using consumption data from UCID-20267, calculated maximum intakes of Cs-137 in the 70,000-

_ _pCi/year range.

UCID-20298 Radionuclides in Sediments Estimated that only 20% of the Collected Downstream from Cs-134/137 discharged between Rancho Seco Nuclear Power '1981 and 1984 are associated Generating Station. 'with the bottom sediments (to a.

depth of 12 cm.) in Clay, Hadselville, and Laguna Creeks

- 'to a distance of 16.2 miles (26 km) from the plant UCID-20367 Environmental Radiological Primarily summarizes UCID -

Studies Downstream' from 20267,20295, & 20298 and Rancho Seco Nuclear'Power recommends further Generating Station. March 22, investigation of aquatic and 1985' terrestrial food source pathways UCID-20641 Environmental Radiological Part I documents follow-up Studies Downstream from the investigation of radioactivity Rancho Seco Nuclear Power concentrations in fish and Generating Station - 1985. sediment samples. Part II February 6, 1986. contains appendices with

' sample data Revision 0 RSNGS Historical Site Assessment 55 March 2004

Il Independent Evaluations Conducted (Continued)

Document Equipment/System/location Remarks NUREG/CR-4286 Evaluation of Radioactive Based on the analysis of the data (ORNL-6183) Liquid Effluent Releases From gathered, the potential for the Rancho Seco Nuclear exposures above 25 mrem/yr Power Plant. March 1986. appear highly unlikely, stating that in its summary "... it seems reasonable to assume that unless some individual is eating 14 to 18 kg of fish per year caught in the sump, Clay Creek, or Hadselville Creek at Clay Station Road, a 25 mrem/year dose... is not reached by any individual around Rancho Seco."

CID - 20963 Environmental Radiological Documents the continuation of Studies Conducted During the environmental monitoring 1986 in the Vicinity of the research being performed. Cs Rancho Seco Nuclear Power concentration in fish has Generating Station. March 22, returned to background at 1987 distances greater than 7.5 km from the plant effluent boundary.

UCRL-1061 11 Environmental Radiological Documents the 1989 follow-up Studies in 1989 Near the to the environmental effluents Rancho Seco Nuclear Power studies performed in 84-87.

Generating Station. November Recommendations include 1990. suspension of the studies unless a normal or above normal precipitation cycle prompts an evaluation of the potential redistribution of the activity inventory.

None Rancho Seco Non- Determined that there Industrial Area is now "no presence of Survey Project. contamination Shonka Research discernable from Associates, Inc. June background" with the 2001. exception of the effluent path itself and the swales associated with it.

Revision 0 RSNGS Historical Site Assessment 56 March 2004

6.11.1.3 District Initiated Evaluations ' '

Document Equipment/System/location Remarks RPDP 90-001 Over reporting of effluent Based on residual activity release activities for Agl lOi, detected in retention basin Co-57, Co-58, Co-60, Cs-134, sludge during clean up activities Cs-137, Mn-54, & Sb-125 by in 1985 & 1989. Ag-1 10mIO-30%,

up to 40% Co-57-2%, Co-58<1%, Co-60-26%, Cs-134-2%, Cs-137-3%, Mn-54-2.5%, & Sb-

._ . 125-42%

RPDP 90-010 A multiple topical study, Estimate that -350 fWtof including "Field #14" Soil dredging wastes will fail to contamination. decay to less than the anticipated 10 mrem/standard utilized in 1990. (See 91-006 for follow-up)

RPDP 91-006 Radiological characterization Summarized investigation Along the Plant Effluent documentation between 1985 Stream. and 1989 in preparation for further studies. Noted the elevated levels detected in the dredge piles and that -1020 ft 3 of these piles had been containerized as radwaste.

RPDP 92-004 Effluent course , Soil contamination depth profile characterization Revision 0 RSNGS Historical Site Assessment 57 March 2004

District Initiated Evaluations (Continued)

Document Equipment/System/location Remarks RPDP 92-005 Offsite Soil Sector survey Provided characterization data from within an approximate 2000-foot radius of Reactor Containment Building (3600) surrounding facility with direct measurement and soil sample correlations.

RPDP 92-006 Effluent wastewater course Provides a summary of studies radiological characterization to date and established soil contamination half-lives and remediation options.

RPDP 92-008 Soil activity vs. Measured Early attempt to correlate the exposure rate wastewater course soil activity to direct gamma area. readings.

RPDP 92-009 Half-life Calculations for Clay Estimates environmental half-Creek Bank life of effluent creek at - 4 years RPDP 92-010 TEDE calculation for soil 186 mrem/year, decaying to sample taken at grid location AL- 9.9 mrem/year in -4 half-lives 16 (17 years).

RPDP 93-002A Evaluation of Soil in Area AH & Additional data attempting to AL-15 correlate soil activity and direct dose measurements RPDP 93-003 Evaluation of Soil in Area AM-5 Additional data attempting to correlate soil activity - direct dose measurements and various depth of soil removal.

RPDP 93-006 Evaluation of Soil in Area AN-2 Activity concentration vs.

depth to 6" RPDP 93-008 Offsite Soil Sector Survey Provided characterization data within 3-mile radius, 360° surrounding facility with direct measurement and soil sample correlations.

Revision 0 RSNGS Historical Site Assessment 58 March 2004

District Initiated Evaluations (Continued)

Document Equipment/System/location Remarks RPDP 94-003 Soil environmental half-life Estimates environmental evaluation. i half-life of effluent creek at -

4 years RPDP 95-004 Radiological Characterization Summarizes the Report characterization effort and the decision not to remediate the effluent canal.

RPDP 95-007 Offsite uR/hr versus Soil Activity Provides two different Correlation models with which to estimate annual exposure from measured dose rates in the effluent canal area.

. , I . I

. I .

.~~" ... I -. ....

Revision 0 RSNGS Historical Site Assessment 59 March 2004

L 6.11.2 Area 2 (SA02) - South Plant Outfall Area Available documentation of the radiological impacts associated with the evaluations from specific incidents during the operational and post operational period include:

6.11.2.1 Licensee Identified Events Document EquipmentlSystem/location Remarks ODR 82-0248 Leakage from (auxiliary) Plant Effluent large boiler ran down storm H-3 4.5E-06 to 6.6E-06 drain (gCi/ml)

ODR 84-0217 Hydro-pump hose burst - Hydro source CST - H-3 water down storm drain 2.00E-05 iCi/ml ODR 84-0317 Drain hose fails releasing 500 2.20E-05 ,ACilml - 2880 gallons from T-993 to storm ACi total release drain ODR 85-0075 Hole in "B" RHUT releases 2.OOE-04 MCi/ml at storm 1000 gallon to storm drain drain - < 4.30E-06 at the outfall PDQ 90-0367 H-3 Evap (RWS-730) leaks H-3 at 3.8E-02 and Cs-500 gallons across Tank Farm 137 at 3.6E-08 tMCi/ml into storm drain south of East cooling tower PDQ 93-0088 A RHUT agitator leaks 450 Release -

gallons down storm drain. 37 gCi H-3, 8.30E-03 ACi Co-60, 3.15E-03 AtCi Cs-134, 8.52E-02 ILCi Cs-137 PDQ 02-0015 B RHUT agitator leaks 450 H-3 at 4.42E-06 and Cs-gallons down storm drain 137 at 2.80E-09 ,uCi/ml resulting in an unmonitored I release 6.11.3 Areas 3 - 7 With the exception of two Impacted Survey Units contained in Area 5 as described in Section 6.10.5, no radiological impacts were identified that impacted these Areas.

One Impacted Survey Unit within Area 5 consists of the employee parking lot, Parking Area

1. 2 and Parking Area #4. One event was identified in this area, ODR 870301 where a pallet with articles tagged "Contact RP prior to disassembly outside RCA" was found in this area.

Also, this area has been used as a staging area for radioactive material shipments, both incoming and outgoing.

Revision 0 RSNGS Historical Site Assessment 60 March 2004

The second Impacted Survey Unit consists of those cells through which the access road to highway 104 passes. Since this access road serves as the point of egress and ingress of radioactive material shipments, it must be classified as Impacted in accordance with MARSSIM classification guidance.

6.11.4 Area 8 (SA08)

Area 8 is comprised of that area of the site known as the Industrial Area. The identified radiological impacts on the Industrial Area are too numerous to summarize here. A brief summary of each radiological occurrence is included in Appendix A, HSA 10 CFR Part 50.75(g) Document Review Summary.

6.11.5 Area 9 (SA09)

Area 9 is comprised of those areas of the entire 2,480 acre site not contained in Areas 1 through 8.

Revision 0 RSNGS Historical Site Assessment 61 March 2004

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7.0 FINDINGS RSNGS, like all commercial nuclear power plants, is designed with multiple boundaries to contain the unit's radioactive contents within its many systems,'components, and structures.

Many of these systems and structures have been impacted due to routine operations and maintenance activities during the operational and post operational history of the plant.

Structures anticipated to be classified as Impacted by the unit's operation include the Reactor Containment Building, Auxiliary Building, Spent Fuel Storage Building, Interim Onsite (radwaste) Storage Building (IOSB), and much of the Tank Farm and the systems contained within it. Other systems, components and structures that were not originally anticipated to be contaminated have been impacted as the result of system cross contamination between the primary coolant system and secondarysteam systems due to the failure-6f tbes within the unit's OTSGs. Areas and systems impacted as 'a result of these prirnary'to secondary leaks include:

• Turbine Building;

• Emergency Feed pumps;

• CST in the tank farm;

• Regenerate Holdup Tanks;

• Auxiliary Boilers;

• Main and Auxiliary Steam Systems;

• Main Feed Water System;

• Retention basins;

• Condensate System; and

• Auxiliary Feed Water System.

Other major non-nuclear systems became contaminated by leakage directly from the primary system or by materials that had been in contact with primary coolant including;

• NitrogenGas System;

- Control Rod Drive Cooling System;'

• Service Air System;

• Nuclear Service Cooling Water System; and

• Turbine/Component Cooling Water System.

System leakage from these systems in turn contaminated the Clean Drain system.

As referenced earlier in the report, Area 8, comprising the Industrial Area, as well as Areas 1 and 2 and the Impacted portion of Area 5 will be divided into unique survey units consistent with the guidance contained in MARSSIM. The initial MARSSIM classification of these areas will be based on the design function of the area of concern (AOC) or its operational Revision 0 RSNGS Historical Site Assessment 63 March 2004

history. Of particular significance are those areas historically referred to as the power block.

These include:

• Reactor Containment Structure;

• Auxiliary Building;

• Spent Fuel Building;

• Turbine Building; and

• Tank Farm.

In general, these areas are being assigned Impacted Area classifications. Should information be developed during the course of the project supporting reclassification of these areas, the circumstances and rationale will be documented appropriately.

During the operational history of the facility, radioactive liquid spills, radioactive waste processing, storage, and certain maintenance activities on contaminated equipment and components occurred outside of the historic power block. These occurrences have resulted in the preliminary assignment of Impacted Area classifications to the areas affected. These include, in part:

• North and South Turbine Building lay down areas;

• Radioactive Waste Barrel Farm;

• Radioactive Waste Solidification Pad (East of Auxiliary Building grade);

• Machine Shop;

• Auxiliary Building yard area;

• Construction and Pipe Fabrication Shops; and

• "C" warehouse.

Should future survey data support reclassification of these areas, the circumstances and rationale will be documented appropriately.

Several occurrences involving radioactive materials have potentially impacted other areas outside the RCA. These include the storage of radioactive materials in the following locations:

• Turbine Rotor Storage Shed;

• Paved access surrounding the East and West Spray Ponds;

• Quonset Hut;

• Switchyard;

• Main site tool room; and

• North and South Storage Yards.

Revision 0 RSNGS Historical Site Assessment 64 March 2004

Incident specific survey and post remediation survey results have been used in the assignment of a preliminary Survey Unit classification of Impacted.

The District-controlled property outside of the Industrial Area has been initially classified as Non-Impacted with the exception of the storm drain outfalls (Area 02) and the plant effluent water course way (Area 01).

These preliminary classification assignments have been substantiated by the non-Industrial Area survey work performed by Shonka Research Associates, Inc. This project provided direct scanning of over 300,000 square meters accompanied by over 80,000 gamma spectral samples without the detection of any radioactive material of site origin above background.

Table 7.1 Area Designations Area 01 (SA01) Impacted Area 02 (SA02) Impacted Area 03 (SA03) Non-Impacted Area 04 (SA04) Non-Impacted Area 05 (SA05) Non-Impacted*

Area 06 (SA06) Non-Impacted Area 07 (SA07) Non-Impacted Area 08 (SA08) Impacted Area 09 (SA09) Non-Impacted

• Area 05 contains an impacted area within it as described in Section 6.10.5.

Revision 0 RSNGS Historical Site Assessment 65 March 2004

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8.0 CONCLUSION

S The RSNGS HSA provides sufficient evidence to support Impacted Area classification for SA01, SA02, and SA08 only. Area's SA03 through SA07 and SA09 shall be classified as Non-Impacted Areas and excluded from further investigation and survey actions with the exception of two Impacted Areas within Area SA05 as described in Section 6.10.5.

Based on current and historic sample results from the licensees Radiological Environmental Monitoring Program (REMP), there is no indication that surface waters on or near the facility or the ground water off of the site has been affected by the licensed operation of the facility.

However, further evaluations of the groundwater directly below the licensed facility are also planned prior to the LTP submittal. The plant effluent watercourse contains deposits with measurable amounts of radioactive material resulting from liquid releases conducted in accordance with the regulatory and permit requirements imposed on the facility.

There were periods of liquid effluent releases during operation of the plant where it was determined that calculated dose to a maximally exposed individual via the liquid effluent pathway exceeded the design objective level of 10 CFR Part 50, Appendix I. However, it was also determined that these liquid effluent releases did not exceed the concentration limits of 10 CFR Part 20 or the fuel cycle dose limit of 40 CFR Part 190. The need for remediation of this material, the dose from which has already been accounted for in accordance with the regulation governing radioactive effluent from power plants will be determined prior to submittal of the LTP.

Revision 0 RSNGS Historical Site Assessment 67 March 2004

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9.0 REFERENCES

9.1 NUREG-1575 - Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) 9.2 Radiological Characterization Plan for the Rancho Seco Nuclear Power Generating Station (RCPRSNPGS including Quick Look (April 4 and 14, 1990), Phase I (April 14 through May 9, 1990), and Phase 11 (1991 through 1997) 9.3 NUREG/CR-2082 - "Monitoring for Compliance with Decommissioning Termination Survey Criteria".

9.4 NRC draft Regulatory Guide DG-1005 "Standard Format and Content for Decommissioning Plans for Nuclear Reactors 9.5 Sacramento Municipal Utility District, Rancho Seco Nuclear Generating Station, Unit No. 1, Final Safety Analysis Report 9.6 Sacramento Municipal Utility District, Rancho Seco Nuclear Generating Station, Unit No. 1, Updated Final Safety Analysis Report 9.7 Sacramento Municipal Utility District, Rancho Seco Facility, Defueled Safety Analysis Report 9.8 Safety Evaluation by the Directorate of Licensing, US Atomic Energy Commission, in the matter of Sacramento Municipal Utility District, Rancho Seco Nuclear Generating Station, Unit 1,Docket 50-312 9.9 Rancho Seco Nuclear Generating Station Proposed Decommissioning Plan" (PDP) 9.10 Plan for Ultimate Disposition of the Facility" (PUDF), July 1990 9.11 Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities, NUREG-0586, August 1988 (GEIS) 9.12 Geotechnical Investigation for Proposed Evaporation Ponds, ERPT-C0104, Rev.1, 1989 9.13 Final Engineering Report Assessment of Spent Fuel Liner Leakage, ERPT-M0221, Rev.0, 1990 9.14 Response to NRC questions on Geologic and Seismologic Conditions 1987 9.15 NUIREG/CR-4286, Evaluation of Radioactive Liquid Effluent Releases From the Rancho Seco Nuclear Power Plant, March 1986 9.16 Annual Radiological Environmental Operating Report, January- December 2002, Rancho Seco Nuclear Station, Herald, California 9.17 Rancho Seco Non-Industrial Area Survey Project Rev. 2 - Shonka Research Associates, Inc June 26, 2001 .Appendices and Addendums Revision 0 RSNGS Historical Site Assessment 69 March 2004

II 9.18 Federal Register, Vol. 64, No. 234, December 7, 1999, FR Doc. 99-31508 9.19 Rancho Seco Independent Spent Fuel Storage Installation Final Safety Analysis Report Volumes I, II, and III 9.20 Bechtel Corporation, Construction Report for Rancho Seco Nuclear Generating Station Unit No. 1, January 15, 1976 9.21 Report 2041-RE-009, Rev. 1, Rancho Seco Activation Analysis and Component Characterization, WMG, Inc., July 2003 9.22 PNL (Pacific Northwest Laboratory), 1984 "Residual Radionuclide Distribution and Inventory at Rancho Seco Nuclear Generating Station." PNL-5146, June 1984.

Revision 0 RSNGS Historical Site Assessment 70 March 2004

10.0 APPENDICES AND ADDENDUMS Appendix A: HSA 10 CFR Part 50.75(g) Document Review Summary Appendix B: Personnel Interview Program Appendix C: Area Summary and Preliminary Survey Unit Identification Appendix D: Miscellaneous Location and Earthquake Data and Figures Appendix E: Miscellaneous Historical Construction Photographs Revision 0 RSNGS Historical Site Assessment 71 March 2004

IL-This page intentionally left blank Revision 0 RSNGS Historical Site Assessment 72 March 2004

Appendix A HSA 10 CFR Part 50.75(g) Document Review Summary

Date Doe. # I Doc Location Index description Record remarks ,r Survey data 1/2/74 COR 740121 2802-1802 Loss rad monitor RI5021 GSE monitor (pre Primary-sec leak) 6 Loss rad monitor Ri802l bure u land steam exhaust monitor -plant In . i -

1 ~O IJR -750004 28102-0206 motor . .CSD :No release Emergency sump Isolation valve RB 9/7/74 LER 7402 0204-019 filure alve failed to cycle during SP

- -7 - .

- - ... Boron detected at south outfall, area I 9~f7 7001 -ischarges from A spent regen tank and diked and diluted, no reference to 912 OCR 7407 2802-0069 OSdischarge.; > misc waste tank into storm drain . radiological contaminants.

10/3/74 LER 7405 0204-0031 SFP coolant demin pump BWST BST < min. TS level for power ops Failure to follow RWP resulted in Find survey RWP 74-'72 for extet of

[11I/4f74 ODR '740038 2802-0138 conitaminationofworksite IPSV-29 RB Spray system cotmination Flush water down storm drain between H3 .c2.9E2-6 tCilml, SE outflow <3.4E-6, 12/18/74 ODR 740052 2802-0190 Unmonitored radioactive release Aux and Admin bldg PE <5.GE-7

[1/1n5 CR 280-024 7000 ntmintionfCC systmo'

-~- Increasing activity f~rom 102/ 1874 1 2 leakA Letdown cooler RCP seal 1

1122175 LER 7502 0204-0066 Valve demin water - BWST Valve line up Issue

- ~ISt occurrence ofactivity discoveredin' LR 7505 02407 CCW radioactivity CW1/13/75 . H3 6.7E-5puCiml- .

2/1 9/75 LER 7505 0204-0077 H3 In RCP seal water cooler Follow-up - A RCP seal water cooler leak Au.RB. Air ejector and gland seal OCRexhaust monitors - no flow recorders or.

22r5-OR 750011 '2802-0230 Unmonitored radioactive release moitors to estimate release volumes. ministrative Issue 415f5 DR282-25 7501 Ceicl eaag r8g/90 separation Issue - Napretrelease involved ABparticulate pump tnip unmonioe DR 7503 28027-0329 releas MSD 7/95 ODR, 750044 2802.0331 Unuhrzdrdocierelease Overflow RHUTvia overflow line dietRHUT overfow to Basin- -

OR 750046 28020335 UHIToverflow nyH 3/18/76 LER 7603 0204-0125 Bypass line weld leakage M/U pump suction bypass.- No significant volume noted 00R -760014, '2802-04368 Weld - Mek.Upump room,  : No significan volume noted.

Unmonitored release 413f76 ODR 760169 2802-0440 chrcoal/parficulate Isolated Estimate provided for annual report LER, 7603 19134 Leak MU pump suction byps Minor leak InSIM-00l'bypass - No significant volume noted Lek vent valve B Decay Heat Removal 8/5/76 LER 7611 0204-0149 Cooer No significant volume noted ij6ui.R 705 28020531 Weld leak . -

• 1-5. .Nosignificant volume noted 8/22/76 LER 7613 0204-0158 B DHRP leakage Packing leak > 0.63 gph OCR 760061 '2802-054 DH pum la, , , Pump casing leak during testing, osgiiatvlm noted 8/27/76 COR 760063 2802-0545 Plant effluent valve In manual FV95103, 201, 301 In manual Potential release related event ODR -60069 2802.0559 offsite/retentioni basin hdMAzinFEEPUPealakg.709 6.25ppm hydrazine at PE .

A-i

Date Doc. # Doc Location Index description Record remarks Survey data Drain pipes installed in effluent canal during road construction, effluent backed up Into weir affecting discharge rate Unknown affect on dilution, therefore 10112178 ODR 780079 2802-0580 Plant effluent Diverted monitor concentration of effluent 12f8178 ODR.

flT 7~

780090 ifT11Failure

• 20-87 Ji edings to record tilt and imbialace 7 Nteeerae 4/1/77 LER 7704 0204-0188 S radiation monitor Missed sample while monitor 005 - No abnormal release noted 8/2077 LER 7707 0204-0197 . ZSAMPLE ISOVAL.VELEAK avfailed to cycle during Sp 11/17/77 LR 7719 0204-0238 PR vent sample valve leak rate 1 gpm leak from valve < 24 hrs 1//7DER004001 11 additional itemstoivsgae 8/122178 ODR 780032 2802-0700 Hightiie leak rate cluain. Z code safety - F o release noted

- - -77 7 02L0700 ighlek _at ____laio cod saet 8/28178 LER 7807 0204-0282 BDHR pump seal leak Otoard seal 2 gpm Nosignificant volume noted L7j578 LER.

FTT' ~

F~

7808 0204-0279

. -~ i ExesRSleakage (dresser valve)

V-1 506,507.&505. Majority of leakage to Prssrzer relief tank, Nsinfcant volume noted GDT V665A sample Ri 5006 valve ossible unmonitored release from hot lab 3/21/79 ODR 790015 2802-0934 WGSO80 hot lab vave gallery (gas not particulate) No long term impact V~ ODR' 7903I20-~

71Y~ 90480 io ta Generator shell Thermocouples TC-thermocouples-N release notle'd 7/9179 ODR 790044 2802-10 13 Failure weekly sample aux building vent SD - No release E7/17919 LER 7908 0204-0348 BDHR~sala: . Packing leak >0.63 gph I 1.OE2+8 dpm/100 cm2 no volume 7/20179 ODR 790048 2802-1 023 Ovrlwmisc. waste water tank The Aux Building -20 level flooded - noted 11/25/79 *LER 72 02408 edlek -i-..Pin holeIn weld soce Nosignificant volume noted 1217179 LER 7920 0204-038.4 Wed leak Pin hole in weld socket No significant volume noted 127179-. LER 790 0204-0438- lea~kj 7920 postpone ea sj 1/1 210 LER 8004 0204-0419 RCS leak rate PRspray valve PV21509 Up to 2 gpm 1180ODR 8009 -RC unientified leakage *- - 7 consecutive leak rate tetsfailed- No rate Identified-1/17/80 LER 8005 0204-0422 DHRP seal leak 3gpm outboard seal 1/17/80 OC0R 8011 .DHR pumpise lieak Otorsel3p 45inches on -27 floor. sump appears 1/17/80 ODR 8D0010 Water on RX building floor plgged 1180 OR 800011 P2101aaeERO4PZspray valve leakg f.

2/28/80 LER 7920 1899-1428 MU pump discharge header leak No additional information

.3/18/80 OCR 810059 - . Rdiation monitor plugged RBatmosphere sample. j ~ Ntrelease sa 3/24/80 LER 8015 0204-0458 Filed fuel assembly During in SFP Inspection No release impact tI §. ': j. Hecuitground cover front of R:eui II . onroled substance in dumpster (filed Hthblew loose, dumped in 'site dump '-

3/25/80 OCR 800048 3280 .

• 9 esofplant .

A-2

Date 'Doc. # I Doc Location 'Index description Record remarks I Survey data CSCA Polaroid marked with RAM tag 200 cpm loose. 2000 cpm intemnal 4/2280 ODR 800071 Missing radioactive material missing from hot lab cntamination no record of decon

- DR 800085 - . FSAIR analysis of steam .mn- .No actual release Involved 5210 LR 8026 0204-09 O G tube rupture SAR Issue No release Impact I- Barrel farm storage caused west fence doeto exceed limit no contamination 719180 ODR 800100 Environmental monitoring limrits noted , .No actual release involve 7/25/80 LER 8033 0204-0508 Eniviro'n TLD >10>bkgd 2nd qtr Barrel Farm No release Impact W ater leakage/seepage SFBJTurbine. Potential release via TB floor drains TB 8/4/80 ODR, 800111i Bldg wall oo dancnmitoneident 7 H3 .OE-3 pCi/MI 10/10/80 LER 7922 0204-0398 Purge valves Valve positioning/stroke Issue No release noted 10218 LR 80-42 02D4-3 TLD > omit' -rya n orelease impact Environmental liquid release H3 anal.

12/18/80 ODR 800154 005 Effluent H3 monitor 005 LER 8050 '0204-054 s perimeter RD readin igh , Barrel farm .

1/9/81 LER 8050~ 0204-562 West perimeter TRD reading high Barrel farm 1/1J1 LER 8102 - 0204-0572 Leak search CCW/ contaminatioin' A letdown cooler H3 6.4E-6, GB 9.5E-7, Cs137 2.6E-1/12/81 ODR 810004 Primary to CCW leak Unknown component pnimary to CCW 7Ci/ml 1/38 C 108SRB Release H3 (filed 1/23/81) Effluent H3 monitor OOS . - .

1/25/81 ODR 810009 NRB release H3 (filied 1/25/81) Effluent H3 monitor OCS Leaving site without exiling radiation -

[ T C 001 -Ad_ __ __ __ -

064 8 4monitor -m No actual release involved Potential - Unmonitored radioactive POTENTIAL Issue regarding fan 538 1127/81 ODR 810013 reese uring LOCA LER 812 17-1542 C H3 > MDA~ ~f 30 day follow up report~ no new~data' -

2/5/81 LIER 8102 3086-1093 CCW tritium 30 day follow up report, no new data OOR 810039 Contai'nmetnt integnity -Both personnel hatches open at same time No actual release involved RB purge Isolation valves - excessive 3/6/81 OCR1 810045 RB leakage SFV-53504 leakage LR 8119 1870-1584 RM sample line'lugged CD -<24 hour duration . , -No releasea.. .,m 0 3117/81 LER 8111 3086-1133 Blank flange RB sump Admin Issue Re: SF movement No release noted Both personnel hatches open during 7fuel-.

3281 LER 8113 1870-1571 Breeched containment integrity .,vget rlaentc 3/28/81 LER 8120 1870-1581 PE chart excee d TS (parameter?) pH No radiological Impact 811 3088-1185, BWST valv failure -SP valve stroke su.N ees oe 73 gallons at 2E-1 pCi/mi -50O mCi 5/12/81 ODR 810098 Cntaminated liquid spil Sample valve DRCST left open. leached into ground L1B/81 LER 8126 0204-0854 'TGtb ek-OS A-3

II Date Doc. # Doc Location Index description Record remarks Survey data B OTSG tube leak creating rad areas in 5116181 ODR 810097 OTSG leak turbine building I1st OTSG tube leak 5118/81 'LER 8124 3086.1198' - Crnfetemergency sump_ 30Oday follow up reportL oedt 5/18181 LR 8124 17159 RCS transfer to Emergency sump da olwu.no new data -

RCS to emergency sump during valve 4.000 gas ROStoEegnys p u 5/118/8'1 LER 8124 1870-1 599 teng g . . x-connect of A&B DHR systems Lek steam generator radiation 511 8/81 LER 8128 3086-1202 scndary system Initial notification OTSG tube leak Cnenser. Air eject radiation monitor -.

5281LER 8126 3086-1208 alarmi Flwi,1gpm tube leak BOTSG 8/4/81 COR 810119 RHUT LWR dmin - Noactual release involved 1 8/12/81 ODR 810112, OG Lak cool down - ie.-'Noactual release involv~ed, 6/12181 ODR 810120 RHUT discharge min - No actual release involved 6/ 286DR 810123 RH T vleln u .m -- o acua reeleas Involved

~ . _ _ _ _ _ _ _ _ _ _'A L ~

R15002 A&B secured during WGDT Release could be under estimated due to 6/26/81 ODR 810126 release unknown sample volume Gs release - no contamination impact

- II,,A 'moinitor alarm, sampling of AMB steam, 8/4/1. OR 8014 ,. CPD.'CRT indicated primary to secondary r meoj nite c rt alr ar m Q . i~ po te t e nia! > j "i Acivity in AE condenser, polisher demins.

8/20/8 1 COR 810155 Indication of primary to secondary leak and OTSG liquid samples Rdoctivity in secondary system, PZR 30day repot regard PZR contamination of 8/58,LER '8144 1870-i1631 N i ., . . N2 system which contaminated the OTSG Noadditional MInomtn NGS contamination from the PZR to 8/25181 LER 8144 0204-0699 Primary to N2 system leak OSG 8/27/81 LER 8'145' 0204-0701 DHRP leakage! .< . 1GH o signifcicnt vlume note d>>d Cntainment building N2 system 9/1/81 ODR 810157 Prmary to nitrogen system leak cntamninated (from PZR)

I9/15181 ODR 810167 .- paya dtank cotamnation C.Mn. Cs.I. Xe 1.OE-5 to 1.OE.-3pCi/mI - "i BDHRP seal leakage @0.87 GPH - not a Noindication of release or significant 9/1 5181 ODR 810161 OHsystem leakage sgnificant volume in a CSCA. cntamnination 9'/18/81 LER 8144 3086-1273' Raioactivityin secondary system 30day follow up report. no new data § 1 111 K;'

9/23181 LER 8145 3088-1275 OHpump leak Shaft seal leak < 1 gpm No significant volume noted 10/27/81 OR 81019 7 ..

CR 71samle.Consetvative o-monitored release aric late filter torn in half fo release

• ABSL assumption made Plant effluent activity due to RHUT sample line effluent directly to PE (no basin divert) 10/27/81 ODR 810192 Anual report actvity'MDA during neutralization I

110/27/81 11/20/81

_-I OCR ODR I-

• i, 810193 I,

810210

- 1 .1' _-..--- 1, 1.

krinuallreport activity>MDk:

I- 1. -

7contarnination of NGS & ACS

. . I .

_. I Plant effluent activity due to RHUT sample line effluent directly to PE (no basin divert) during neutralization-Fresh fission gases indicate leakage from M/U tank into NGS then into ASCT i : 'I '

' ' -I' I . r I

A-4

Date Doc. # Doc Location .Index description r . Record remarks . Surveydata

• . -. .Plant effluent actiy d u to RHUT iample 811O R

[~~ 209. ual eporactivty>MAline effluent directly to PE (no basin divert) -

1218 8

• civt~D wiAnalrpr uing neutralization.-

CR not notified of RB purge start/stop.- Conservative assumption used for 1218181 ODR 810212 Incorrect radioactivity calculation samples not changed lAW procedure.- release calc.

[I~i 0009 DR8 . adiactie lquidrelaseLeak from DRCST. PLS-089, 0.5 to 25 2exempt Cs137 and 1 exempt Tc99 214/82 ODR 820010 Missing radioactive sources source missing 8202DG cnaminto Noble gases E-3 to -8uCV~c 2/1882 ODR 820031 Missed aux stack sample No long term Impact H3 5.07 Ci. Xe 28.8 Ci 3/17

• Mssedauxetacksampee-assumption made for release ..

3/ 1282 LER 8206 3086-1340 Plant effluent level pH

[LERi~ 8208 - 3086-1342 Plant effluent level . pH 418182 ODR 820038' Overflow MWNHUT T 993 Diverted to basin. Increased dilution flow OR 820042 Basin release without saplse . . .nual An report explanation 4/20/82 LER 8210 3088-1598 OTSG deformation FW header deformation No release noted LR 8212 020.4-0794 DHRP leakage i k' o-ii ia n 5/5/82 LER 8212 0204.0798 DHRP leakage OIl leak No radiological impact

[Ws12 ODR '820050 Leak InDH pump. .il O.. leak ,.* 7 . ~

• Noreleaseim'pact-10 gal water drained during weekly 5126/82 ODR 820058 Contamination of NGS (7/12182) sample H-32E-5. Cs 8E-7 pCi/ml

[~~2 024-019 LER822 ontmintio N2 fro OTG tmintio ofNGSfro PZ <MA aterflu7in 7/2682 ~LER 822 0204-0822 Contarin ation N2 from TGContaminatio'n of NGS from PZR < MDA after flushing

[O~DR 820118' ' Contaminationo OnoLGLS S . .t 1.6122 pd/cc SIM-549 (lest valve cap) leak based on 8/15/82 ODR 820107 RCS liquid spill (1/17/82) 'fast dearease In M/U tank lever Large spill Penetration #21 (WDHCR)

Cs137 3.OE2-7. Cs134 5.6E7 3IE 818/82 ODR, 820088. AB TSG contaminated PZR to OTSG via NGS systern~ 5 9/9/82 ODR 820101 Contamination of NGS Probable source, surge tank hydro Xel3368.28E-2 pCi/ml

• ....... , .Acid .transfer line from storage tank to TB.

10/8/82 ER * . .,- Acid collected in trough drained directly . .

1682 LR 8228' 0204-0847 Lkacid line stormget ankTB oft. . pH 6.2 for- ISminute-10/8/82 LER 8223 3086-1458 e133 In NGS 30day follow up. no new data OCR 820114 IrantaminationofNGS Xe133 atlI.5E-IlpCi/ml .I1 .

10/12(8 LER 8227 0204-0849 jN2 contamination WUJ tank back leakage < MDA after flushing 10/152182 LER' 8227 0204-08l N2 contamination I~lou MDAafterflushing LER 8226 3086-1460 I

IAd- leak storage tank turbine building Drainage into storm-drains caused low effluent pH A-5

- . IL-Date Doc. # I Doc Location Index description Record remarks Survey data

- *el133' peak conicentration calcuiated at 101/2OCR 820123 .High rad on aux stick (unusual event) 7.2E-4 uCi/cc(25C release)-: Nobel gasIrelease 10/22/82 ODR 820125 Gross Beta on NRB (11/82?) Basin release w/o rad release form (GB >5E-8 pCVmi) 1182ODR 820106 Annalreprtmissed ABS~sampl~iIse E :i:/ ME1L _____ - L_

11/1/82 OCR 820132 Culvert clogged, vault flooded Flooded electrical vault Non-Rad issue 11/1/82~ ODR. 820135 .;Storage tank flooed- -;OOsoaetn al o-'Rad issue' Unidentified RCS leakage more than No record of leak location or final

]

11/1/82 ODR 820122 0 1 GPM disposition No record of excessive volume 11/16182 LER, 8227 3086-1471 ei~ system, .~..day follow up repr.nnedaa .iE7 11/18/82 ODR 820111 . ABS sample missed Admin. Issue Before/alter sample OK 11/20/82' LER 8231 02064-0863 OTSG leak -

_J.~--

11/20/82 LER 8231 0204-0866 OTG leak 11/28 LER 8231 0204-0862 OTGleak mR/hr on 3 oisherde mins IpnOTG 'j 11/22/82 OCR 820139 OTG tube leak and Steam Line monitors in alert Ri5004 @ 500,000 cpm Efuent channel course companison

~112/82 LER 23

-0204-0869w- baseline photogra raph.nitr isesue . 1 Effluent channel course companison 11/26/82 LER 8232 020470871 bseline photograph Aministrative issue OCR 820141REffuent watercourse co mparison P..-u Nan-Rad issue i 12/3/82 LER 8231 3086-1487 OTG leak 14 day follow up, no new data 1 */8 5E234 - 0204-083 CH ytmleakage' . . ea estimate 5 mVminute [- Nosignificanvoueotd -

Supplemental report, Make up tank 12/29/82 LER 8227 3086-1494 ein N2 system apparent leak, system cleaned to < MDA Noble gases 1.0E2-3 to 1.OE-2 levels 1/1/83' LER j . 1032-1783 IDX8301-8341 - nfo included in78-83 index I' 1 PE9.57E-4, Site Boundary 9.01 E-5

-2000 gallons T993 overflow to storm CLmi H3 (32%/ and 3% MPC

'1/20/83 LER 8304 0204-0889 Hlup tank overflow drains respectively) (ODR83008)

Ovr wurngicwsecndna-. .200gallons H3'4.61-uimcl 1/583OR 830008 . . PAHUT overflow (LER 8304 tank transfer - .. assume -o85%~ontandiNb 1/28/83 LER 8304 3086-1513 ater holdup tank overflow Transfer from MWCHUT H3 from 3 to 32% MPC

'211/3 OD &VO6 sa in.Besuore/after sample OK 2/1/83 OCR 830021 R15019 OOS Indication works, alarm COS IiPrmig tank float stuck., tank overflowed, Cs&C and H 1E7Cm 2/10/83 ODR 8300 23 Rtntion basiin overflow -100 gal. ... C/i.-

_ _ _ _I 2/25/3 OCR 830028 CCW monitor alert RI15008 Cs I E-7, GB and H3 <MDA 1 71 r_..

'>ntaMination N2 system blind flange'-

3~8* LER' 8311 0204-0913* MU tank,, ' , olgses,'cMA i2 a sI 3/11/83 OCR 830035 Contamination of NGS (4/8/83) e133 i~2E-3puCLcc

~

8t29/83 LER 333 '103 833  : 03.182 ~-,

.:em .. I I spsi o/Close i out,"letter

• " -a-'

  -..

-'; - - . I A-6

Date . Doe. # Doc Location Index description . Record remarks Survey data 9/8183 ODR 830175 RB purge system RDM testing Issue for Spent Fuel handling LER 8332 1032-1818 OTSG tube leak - . A OTSG -1 gpm' 9/17/83 LER 8332 1032-1 821 OTSG tube leak AOTSG - 1 gpm V"3 R 8335 3 MAown B02letdW cooler 9/20/83 ODR 830098 CCW contamination B letdown cooler leakage H3 at 5E-5. GB @ 1.7E-4 pCi/mt

- Contminationi of steam. Condenser -

L 13 ODR 830199 - trg *Bldg Steam Condenser Tank V373 H31.2E-5 pCi/ml 10/1/83 ODR 830211 Outboard seat leak (see odr200) ADHR pump WR# 80318

-7 - 7- -47.-:- - -

83R0221 Lost particulate fllter ' A-femate sampling'in place -I 10/4/83 ODR 830200 Mehnical seat leak DHR pump WR# 80317 3-1E-3pCi/mlin'A&B~ste~am~lires 1483 OR 830197 Leak OTSG 1.25gpm leak rate .- and hotweil Resin catch bag failure, - 300 gal RHUT rlease out SW access gate Into pasture H3 2E-4 pCi/mt. 1131/133 - 3E-7 11/14/83 ODR 830238 RUT release (12/83) area. Ci/MI

- ~- - ~~ - - ~ - -

• - Trouble with resin catch bag cause and 83024 - * . unknown quantity of BRHUT to spray onto 111583 3041RHTrelease OR (12/83)" rurid surrounding the area.

.; .< §K OCR83239:: Truble with resin catch bag unknown 12/1/83 ODR 830235 SFV-26006 leaking -. Minor leakage during cycle testing

- "7 -7. . - 7 830 M228 *rejector monitor alarm AEmonitor and sampling indicate 0.8 gpm H3 fluctuates during feed and bleed 12/1/83 ODR 830233 Air ejector monitor alarm AE monitor and sampling Indicate 0.8 gpm H3 fluctuates during feed and bleed 1/4 LE 1032.1836 Index 84014425 onew itemsidentified .

1/3/84 ODR 840002 354.0759 AB leak waler to drain E-360 mud drum leak down storm drain. Insignificant LAW NRC-RI (H Canter) 26K fixed, 4K loose dpm/lOcr no Contmnte 6blanketsn th . loose Contamination on structure 2 or, 2/1Z84 ODR 8400 .3910 arehouse -..- Blankets removed from e0 CSCAalrnae dY;aalt RHUtageforr <0<0 dpp/0 Oc )

relese ut S acess ateinto pasture 1/24/84 ODR 840048 Hole In resin catch bag (3/84?) area. H3 IE-3 ICVml OD 840023 Olspill washed to drain,,: -

Agae sample > 3x bkgd, resin found in 1/31/84 -ODR 840127 354-1085 Rein catch bag failure sample CoMO Cs1341137. 1131 LE2R *102189 Mssed BAS stack sampler -. 10 gas MPC @ site bounidary. oe asiseyo ogtr mpact

<90% MPC no contamination or release 3/4/84 ODR 840084 2399-1340 No abs sample>1 0% MPC impact E;/ ODR' 840085 239971 342 C ciiymcesrg CM137@C1 E4 pCV/c 5 pallets discovered with measurable 2-55K d~pr/100cmr2 loose. 0.2-8 mR/hr.

3/15/84 ODR 840071 Contaminated pallet - contamnination outside CSCA 1.0-240 mRad/hr A-7

IL-Date I Doc. # Doc Location Index description Record remarks Survey data Date1 8400 Doe. oc Loctio ondexfw descrnptaon fa bewenmBardkSF crd asvl Sureekdata

- ,-. l-,,During surface blow, - 100 gallon on 4/20184 OCR 840111 iF

- Blow line spraying asmiio fudpilnto circ water-pvement and down storm drain Local and offsite samples taken 5/2284. OR 640137. area Calculated release - 24 mCi Noble gas.

7/3/84 LER 8420 1032.1857 B OTSG tube leak 1.37 gpm max calc leak rate 1.8 mCi Iodine 7/5/84. ODR 840165 Unmonitored release AShoggers e and I during hoggl oeaon,7 Noble gas rei1sei

-1200 gal evaporated via V627 (misc.

7/17/84 COR 840176 atr release (7or 8/84) ater evaporator) 1-131 - 0.33 uCi LT311 2OR AW1I3 - eease (7or 8/84) Leakage from AFW pump; - iF7 Gab6oxcoletl -VOO D E 8/4/ OCDR 840189 -100 gal from T674 - T993 transfer No off site release

.OOR- Drin~ginvesigtion into missing sources, 8-20/8 8/20/84-0199d.scovered four sources purcaed outside Missing199radioactive sources (mid 1984) normal procurement system. - ~

- L~

8/21/84 OOR 840200 Missing radioactive sources Tc99, Cs137 (each exempt) spread Aux bldg roof contaminatedI when hogger blew contaminated water H38E-5. Cs137 9.6E-5. Cs134 3.85-94/4 ODR' 840242 Cotminated water (mid 1984) -A during startup. ,. V. - ~.- ~r' 11 31 6.2E-5 (paimi)

Water source from CST, down storm drain 9/10/84 ODR 840217 Burst hose-hydro pump N. TF H3 at 2E-5 pCi/mi

-91/4 OCR 840218 T,-35 (CST) overflow'-- - Down storm drain -diverted to basin Hat 3 2E-5at.5E2-7 - -I

-900 gallons. 96,pCi primarily H3 release 9/15/84 ODR 840223 CST overflow (T-358) *ithin10CFR20 limits . 96CI H3. 0.21,uCi Cs137 to PE

- ,. . I omeanstocalctulate~releasefio~msteam H2.1 E-S. Csl3.41137.4.OE-,7.I131 9/21/84. D 237 -',, ~ u~ieE30. ks on A-360 '--..7.E7,i/l~

7 -.

CTflow path checked at site boundary 9/23/84 ODR 840225 354-1353 CST overflow, H3 atsite boundary (vrfy path) H3at 2.4E-6juCilmI. no gamma peaks.

nfsapl 1 CA(sample taken.I

~/8 423Rnf D uring basin release 10/11/184 LER 8422 1032-1 848 AOTSG tube leak 1.2 gpm tube leak 77.

10/11/84 LER 8422 1032-1850, OTSG tube leak. Supplemental letter 10/11/184 ODR 840252 Tube leak 2. gpm tube leak AOTSG 7 fT .8 pCl Cs1 37 s~ource at I .eept 1//4 ODR 840304 oure r.-- inventory (12/184) - -seen at rerack cut-up tent) .

H37.6E-3puCiml@ source. Z2E-5 DMW.053. water from T-993, associated CimI storm drain Total release 12/27/84 OCDR 840317 Drain hose blew 500 gal H3 to drain ith B/D .estimate. 28580,uCi.

111/85 LER ~~ _1032-1928. ndek8423-8525. Nonwiesdntfd Ovrressure system (reducer failure - 19.9pCI Xe133, 8.4puCI Xel135. 459 1/17/85 ODR 850013 2356-0025 rlief lift) PSV36012A&B MIft Ci H3 I II i, - I

- 15 gallons released from loose::~. Post repair, testing fill with site servic j

!1/285 ODR 850036: .2356-0089 Unmonitored offsite -release .';~ expansion joint -A RHUT - -) . -I water .'; *1-11 ~ .:. . I A-8

Date Doc. # I Doc Location IIndex description Record remarks Surveydata H3 2.24E-4 pCi/mI Into storm drain ND atoutfall of storm drains (<4.26E-6 2J7/85 OOR 850075 2356-0213 B RHUT leak (hole) -1000 gallons PCimI)

- - Coolant waste holdup Tank off gas via, OCR1/5 001 850034 - . .26 Ci Xel133 release ABS .Noble gases - no contamination impc 3/5/85 OOR 850226 2356-0588 Failure to report OOS plant effluent Reporting violation- Not release related 13/1385 80088Contaminated dollar fond rin] change' Smoke generated overheated 3/21/85 ODR 850081 2356-0229 component rad monitor R15001A CSD - no release - No release Impact OD .850095 2358-0261 Unceuled release'.:- B--RHUJT -200 gallons Samples < MDA (not provided) PE

-2000 gallon into storm drains (discharge samples (5000 gat. from 1985 annual 4/28/85 OCR 850112 2356-0300 RHUT leak down storm drain valve removed no blank Installed report)'

- - 7 No indication source fo~und, no 850128 236 = Missing sources CoO#DC-i indication of source activity, 5/12/85 ODR 850149 2356-0391 No unmonitored offsite release No unmonitored offsite release No release - no Impact F-7

[E_[524~ ODR 8501'43 '2356-0379 Lek above acceptance limit LRT RSP containment - No release-noIpc 6112(85 ODR 850179 2356-0455 Loud noise steam in containment 600 gallon PZR drop during steam leak 5FV22006 OCR 850190 .2358-0481 RCS leakage - - TS Igpm leak rate (-5 gpm)-

Non- isolatable RCS leak (Reactor -16.000 gallons released within 6/23/85 LER 8510 1'021-2234 Colnt Drain Tank) -B OTSG high point vent containment before CSD

-62/5 LER 850 1032-1993. BOTSG tube leak Duplicate SOCentry of 8510,'

16,000 gals RCS Into containment from B 6/3/85 LER 8510 1035-0074 LEAK RB DRAIN TANK OTSG high point vent valve oa I to tank fam, someondrain 71M -OR:

'256-577W22 Sp120T gal A OTSG drained to ground In TF ner FWS-021

- HP turbine stop valve leakage routed to 7/30/85 ODR 850237 2356-0644 Release from site storm vs. controlled drain Cs134 0.49 MPC, Cs137 0.48 MPC Clogged dischargei strinr maiking rad- - ..

8/1/85 ODR 850270 2358-0712 monitor Inoperable D~ual lineuplsample issue . No release - no Impact RI 5001 being changed out 812/85 ODR 850275 2356-0724 unmonitored release CSD - H3 release only LE57HEPA fitter leakage . - 0.1 lwbpiisa le No radi6foicafl mpact 9/9/85 OOR 850295 2356-0773 Radioactive leak -missing gauge P122014 Local letdown pressure gauge

.. , gal.From ABStailpiece down: H~B ~ ND atpan7 850 B299 :2358-781 atrdown uncontrolled drain ~ -uncontrolled drain effluent-9/28/85 ODR 850329 2358-842 R15044 OOS Not release Issue -'no Impact OD 535 25-84 R15001 power off R101CDEOS(CA D) R15001B inse~rvice i 11/26/85 1 ODR 850427 2356-1 086 jSamples not being taken' >12 hr mini-purge in progress (CSD)

-I - -I- I

'ODR 850431 2358-11098 - Body to bonnet leak ._ :- -*"J SFV122006 . . jWI..-IR N847/NCR55218. -,I

-.:. : , - t 1

12/15/85 ODR 850442 2358-1128 Leak (secured pump/leak stopped) IDuplicate of 437 A-9

II Date Doc. # Doc Location Index description Record remarks Survey data 4j7 7K7~ K -3gpm Iaksecured MU puip.closed II 12115OCR .850437 -2356-1114,. Leak (secured pump/leak stopped). SIM 01. 003. & 080 MUpupro 12122185 LER 8524 1032-2044 Leak PZR sample isolation valve 50 gallons from PZR sample line 122/5OCR 850441 2356-1128 Lek srtoppedbyc'dosing SFV-7001/2 2gpnileak from SFV70001~

Unable to track gaseous/particulate 12/26/85 ODR 850457 2356-1169 mnitor unable to read monitor Monitor display issue - No release impact 12128/85850459 OC Unble to access radioogical releas -. '

122.SOD 6 4 2358-1173 urnunusual event Mntrdisplay issue - Noreleaempc 3-allon spill of H3 south side of 12128/85 protected area. T.93 spill -1930juCI total

[12J30/85 ODR 850468 2356-1208 H3aayis CR0 system .- . T 8.E4puCVml H3-1/2188 C1OR 880018 1569-1655 RBpurge wrong flow rate No alarm/release limit exceeded 1569Suspect. PASS backc leakae~

I gallon puddle from BWST (fill hole 1112186 ODR 860023 6Wmix tank leakage cvred, not bolted)

- Ii 1~ osignificant vo'lume/activity noted-1/1 3/88 ODR 860022 1589-16888 ExesvDHR pump seal leakage -rtnespill 1/14/88 ODR 860028 1589-1681 RS purge wrong flow rate No alarm/release limit exceeded 1886LER 8827 - 1083-2078 Re oRBD filter lost.Etne CSO - 2N radiologicalipc > .

(Work Request's 08 1434, 069009.

PCinlet to RHUT sample pump leakage 069013. 067432. 72437,08013.

I pint. area isolated and sampled 081833, 080247. 083780. 100770.

1/26/88 PDQ 880152 2335-1035 P99leakage (converted ODR860042) (rsults) 109937 1/28/88 ODR 860043 .1569-1753. CWGDT release instead of B`1'. min. -not arelease issue:*

R15001&049 COS. R15050 in service 1/28/86 ODR 860072 1569-1878 RB purge limit violation monitoring release

• olroomr surveyedwl ditoa' OCR 880074 C21/8 1569-1887. 1000cpm too[ box found in too[room, I1000 ccpm flxed/625 dprn oose cnaiaonI .

<1gallon contained within AB drain 211 8/86 LER 8602 1024-0032 PINHOLE LEAK B DHRP CASING sytm 21/6 OOR 860082- 1569-1931- DRcaigdrain failure P-8-B leaked to floor drain - RS777N significant volume notdZ-OSG Csl 341137 1.5-511-6 Several SPC pad flooded overflowing to storm CoOpeaks, suspect faulty GeLl. No 2127/86 ODR 860090 1569-1968 Spill from open vent/drain valves drain IVH-.0I1. indication of off site release.

308 OOR 86065- .1569-1647. Hcontamination CRD Su'spect. PASS backileakage - 7 T T 3/10186 OCR 880106 2857-1511 H3contamination of CRC PASS sample cooler leak H31.75E-5 pCi/mi I* 1 I.-- - I ~ - 1 3/V14/8 OCR 880128'1 285-1811' jUnmo~nit'ore-d`RB purge - .. . CSD, <1 hr, . .- . ~ '.", I... , .' ; . I

_WR-no release-no gamma peaks noted in 4/7/88 ODR 860148 2857-1691 Gamma scan missing 0og 17 -

uii mi

. .. I .--I---q.

4/9/86 ODR '860150 2857-1695-Liquid samples release 88-48/51, -,

missing' - -

K I

omposfte samnpleretention issud:' .~. Ii,.

4/15/86 1 ODR 1 860163 1 2857-1768 Unmonitored ASS release :SD -- 24 minutes A-10

Date -Doc. # Doc Location Index description -, Record remarks . 1, .. , Survey data LER 8W3 22 ~~Unmonitored RB PurgeExeddCDNrailgclIpt 4/22/86 COR 860187 2926-1629 Cs137 source missing during Inventory 8.6 uCi (1 5 mremn If ingested) Removed from Inventory SW6 OR 880199' 2926-1680 Particulate filter torn .nesig. nhoedno data lost. No urinmonitored release.

7/2/86 LER 8602 1046-0143 B DHRP CASING LEAK T.wo pin hole leaks total < 0.63 gpm No significant volume noted

_[j7/86 OCR 860287 2857-1920' Missing Csl137 source.' Csl37 source # 102981 (1.18yCi) 7/7/86 OOR 860288 2857-1 925 H3 contamination in CR0 PAS sample cooler leak - 3 at 8.8E-6 pCi/ml COR 800416 2348-0077 Contamination In CRC Suspect. PASS 'back leakage .

Potential radioactive release. no high 7/8/86 COR 800299 2348-0029 alarmn Potential issue - No release Impact

[I 7/22186 OCDR LER 860298.

8614 2348-0023 1076-0652 azrdous waste drum DHRP drain line leak rupture odumnhydrochloric acid drum ruptured at rduwsetrgarea -

Closure letter 8414 HO V.053 removed w/o rad survey in -HV03removed from CSCA to tab shop Salvage area surveyfonse ra.

[ =8 DR 860377 2348-042 unotoldarea , oure(bntsurveyed, body not) additional incidents HDV-053 removed w/o rad survey in HODV.053 removed from CSCA to fab shop Salvage area survey found several 7/26/86 CDR 860371 uncontrolled area w/o survey (bonnet surveyed, body not) additional Incidents ODR 880416 . ~ CRC cbntamination Suspect. PASS back leakage 9.87E-6 1"ril 9/17/86 COR 860437 2857-1 978 Temp piping attached to DRCST min issue. (no PRC approval) Not release related

-OCR 860472 2349-1848 ~iqu~iqd radwaste tank overflow - Flooding on floor. mRaW.5 mR 10/1/86 COR 860455 2349-1753 DHP casing drain failure -2 drops/mmn - no Impact 10/286 'ODR' *880482 2349-1800 OHP1 casing drainftailure -~- 2 drops/mmn -no impact.

OHsystem trip following ARC sump 10/3/86 LER 8616 1843-1556 level Indicator - B room sump stack Electrical Issue No radiological Impact 6/6 OOR 880477 .239-1867 Blower seized on radiation monitor m`1 pensatory measures taken -CSD. No release impact -..

Aditional surveys performed to determine source of Cs 137 Packing leak in temp valve wet asphalt contamination - results indicate 10/11/86 ODR 860494 2348-0696 Lek temp line RHUIT to DRCST and soil In area environmental levels (0.2 pCitg max.)

.OCR -860495 '2348-0705 ROM H3 sample not taken or analyzed Procedural issue- No release impact 10/23/86 ODR 860514 Temp line leak T-621 to RHUT Coupling upstream transfer pump Volume and concentrations'

- Fromdraining unidentified system to th'e CDRBW22 2157 Co0.ntamination of polisher sump' H3 DS after OTSG tube leakage H1-3 1.06E-5SjCi/ml.

Fine mist contaminated imm~ediate 11/5/88 COR 860530 2348-084 Frhru wall leak FT-26003 - 1 gallon puddle of RCS on floor surroundings OCR 8858 2348-0862 CCVcontaminationCsl37 K- i.17@0.04pCl/cc, GB it0.028 pCi/cc ytmfeanbld '

SFP liner leak Rx CDT Iso'lation valve H3. Co6O, Csl34/137, Agl 10m, Sb125 11/16/86 LER 8625 1092-0223 BWST boric acid clean drain system TF East wall to TFstorm drain ranging from 5.7E-2 to 4.8E-5

~ODR 880542 2348-0887 SFPvaterleak I >-"

. I U

PP leakage int6 'u nortroiled' drain i Fbetween SF and RB 75 1 A-11

it Date Doc. # Doc Location Index description Record remarks Survey data Composite samples - due to no gamma 11/18/88 ODR 860544 23481-0897 RB AS particulate missing peaks and extended CSD - No release impact

-200 gal FPS water drained through OC6R 86555 2379-250 nseto vtv laki TF " CSCA and dovn storm drain Plant eflet dLD (not stated)

-J L .. ; .. . ' * ' _ _ _ _ _ _ __ _ _

1215/6 107-0243. LR 625 Co60, Csl134/1 37, Agi lOinSb125 121/8 LR 82 0702 FP liner leak thru TB wall ast wall to TF storm drain ranging from 5.7E-2 104.8E-5 uCi/mi LER 1067-0ica CSImpact.

Inoperable Rad Monitor RC leakage 11/6/87 COR 870052 2348-1495 stem Procedural set point issue - No release impact nprbeRad Monitor RC leakage' -

OC7R.1870053: 2348-150 stePrcdral set point issue-, N release impact 1/15/87 ODR 870048 2348-1481 R.15044.45,&4600OS CD -no release - No release impact M7 1118/8 OC 70 0 2348-1486

• noperable Rad Monitor RC leakage '.

syte eural set point issue-

' i'E' 7 7 .

, :Po -. N release Impact 1/19/87 ODR 870064 2349-0211 RBparticulate air sample filter data lost Coservative value (Prelpost release) used LER 86-27R1

~1/2/87 ODR '870085 FT'1~ 2349-0215..

i Rnatlclate air sample fifter data lost RWAVno alpha sample (noalpha ctvtpre orpost release- No release impi 1/2/87 CDR 870100 2379-1432 Rdiation monitor does not exist Procedural - RHUT vs. Basin release point Noimpact 2/(7 LER 8913 3089-0053 ekng sealed sourc 'Failure to perform siemi-annual leak test Noreesn otmnation noe Approx. 5 gal RCS spilled via A RB spray 2/2/87 ODR 870119 2348-1728 RC spill in containment sytm OC 7 1 5 '2349-0 4 Bprticulate air sample filter data lost Cosraievalue (pre/post release) used ~ R ~ j j j Unmonitored RB purge due to power 2/9/87 ODR 870168 2348-1904 osto stack sampler CSD1hr - .. No significant release impact 87 8 2 2349-0 5 otbes m lrABS no unn ,no release - , orelease Impact 2/23/87 LER 8614 1075-1873 DHRP drain line leak Small leak Nosignificant volume noted Radiation mnoniltorC during RB - . .  : j 2r&7 OCR 870203 2349-0402 purgek detectioncapabilitysissue 7 Noreleasmeacmpjjj ABS particulate filter tom, loss of 3/5/87 ODR 870273 1569-0298 release data OnInvestigation, data not lost - Norelease impact Z 1 FGaseous ef.uen dose calculation MA7 OCR .870284 1589-0259 derinq~brnt'samrple . aclationi I'day late'-.N elas AS grab sample not taken [in] required 3/10/87 ODR 870290 1569-0382 time limit Admin. Time issue - No release impact IF Gas sa eanalysis did not meet On investigation, before and afteir samples 7 i j3R/11/87 OCR 1589-0392' requir~ LID' *:id (CSD)~ . -870293 .No release Impac Palet with articles Tagged Contact RP From 'upper-outer storage yard'. N/D 3/16/87 ODR 870301 1569-0428 Loss of control RAO MAT package pror to disassembly outside RCA fixed or loose f l:TT 7 P l arboys W i A m arun sin TF, 13/24/7 _ O6R. 870333 1589-0 04 C na iatdtp In dum pse pst

.. strwesft of toolromd I Of 5 b g (450 ccpm Radwste) ,

Non sufficient notice sample RP tech 4128/87 ODR 870489 1569-1331 RB purge Edof purge sample - (CSD) No release impact 5/8/7 O R 870514 2379- 5 9 L wfo r d o itor' 'No nm nitored release -. N o rele leim act: ~ ~ t  :

A -12

Date Doe. # IDoe Location Index description Record remarks Survey data D8te5o82 DocCLocatio condeadsciptionReodrm ks- Svydaa -

5/19/87 ODR 8752NC oOcnaiain- 1.59E-7 pCi/mi Co6O

. o~ssible attempt by contractor to smuggle 5178 ODR 870819 2928-203 P at PAP (contaminated) cottn (contaminated) PCs offsite. No contamination 1narea detected.

5/28/87 COR 870638 2926-0287 Cntaminated instrument R violation, No release impact ODR 870661 2926-0389 AS samper ntunngCsD <1 hr CS1 37 SE-7 pCi/mi (similar 860555

-200 gal down drain below B Main Steam (done rMI) 1.9E-4 mrem per semi-6/3/87 ODR 870664 2379-1616 FPspill in TF -Lne No Divert annual report Procedural violaion - no significant7 OC24901 774 ekIntm eaBA Evaporator feed pump leakage vlume (pin holelek Spill resulted from open valve pnor to 6/11/187 ODR 870701 2926-0530 dosing vent.. Spill/flooding of EDHCR - Up to 300Kdpm/1 00 cm 2 Cs1 37 1.75E-7 jCitmrl (nometioo ydiversion of PE) 1987 semi-annual Systemn drained fn contamilnateid a`rea gal81 Dumped down urontrolled st6rm report -1 000 gal. Max dose 3.33E-4 8/30/8 ODR -870784 2926-0748 removed vwithout saperain between Aux and RB- mrem.

Cal 37 6E-8 uCi/ml Feed and bleed 7/13/87 ODR 870800 2926-0882 Coling water sample Cs137 CWH 2 lube oil cooler rmoved contamination 9/10/87 TS violation Uquid Effluent Environ.

8/17/87 ODR 870863 2926-1113 Monitoring program No unmonitored release -admin issue No release Impact Imosherc seamdums Lw levels' CO137-5.31E-7. Csi134-.

Atmsphricstem dmps(wet lay-- - 1.08E-7 #Ci/ml (mnax Dose WB/O up/pinned) lift during testing. -2000 gal: =3.34 E-3 mfrem per semi-annual 8287 OR 870903 2926-1252' Contaminated spill in tank farm (controlled drain inTF per FWK) rport) alve leak contaminated water Into FWS-020 leaking Into local storm drain Csl137 5.3E-7. 134 1.1 E-7 pCi/ml (max 9/1/87 ODR 870905 2349-0643 strm drain . - plant effluent diverted to basin dose 512-3 mrem - semi-annual report)

- , 7----~- .- Startup new monitor -old monitor still in 918 ODR' 870998 .'2349-0910 Radiation monitor set wrong servce- . -. a release impact 9/14/87 ODR 870942 2349-0722 Gamma contamination detected CCW - Cs`137 0.04 pCi/ml 9187 ODR 870993 LSI 77 2349-0878 edleak AOH pump

-_ue__no - . D - systeminprbeTisu-pIll-.

ot

.77 9/23/87 ODR 870975 2349-0799 Fan test without rad monitoring Startup testing on A546/R 15546 - No release impact Erooitri~ngew effluent release Unmonitored releases duiing S/U testing

[/24/87 ~LER' 87413 1833-0484 -pit' of' new system .. No release, impac RPnot informed prior to fan' start for 9/24/87 ODR 870979 2349-0812 smple .Startup testing on A546/R15546 - No release Impact

- .. . .. iga.

1- Ltxpaint (acrylic.

-. .. -p. rrainerethylene glycol) hosed into storm rain nea r7Xwarehourse (OWOC 10/12/87 OCR '871022- 23907 praypump seal rupture causing spinl in!ormedd)).

Filter paper missing from RB stack 10/20/87 ODR 871053 2857-0149 smple CD-no gamma-no Sr prior sample 8D71073 - 2379-0769 RPoffice floor spot direct fnsk - cpmn oni carpet~ removdnorsua 11/1/87 ODR 1 871076 2349-1205 IDelta-P across HEPA Charcoal zero JCSD - no release - No release Impact A -13

ii-Date Doc. # Doc Location Index description Record remarks Survey data

~ 2. Cs137 1E-7pC~mi (max dose WB/OJ

[1187 O~~

15. 1 '17 12 2 3 o4w9-K-3 1 4~ oly b ttleOTverf G 0. 5/1. 2 emper se rn n nua l 111/518 ODR $7122LPol botle verlow 239-134-. TGfill vent - -5gal.To TBdrain;. report)

-7 min, CSD. procedures issue, no 11116/87 CDR 871149 2379.1953 RB purge started prior to sample release impactCs7007pimo6.04Cmj LER/87 747 11/2 188 -17 2 Filure to continuously sample RB LR 877 16-72 pu217 rge Similar 88-3,19,22,28,29,22.&43 Noradiologicai impact

-OSdrainage into WDHCR and.' - OSG drain via FWS-5;2V flooding'- .

112587 R 7155 O -RectrContainment Building : -- W DHCR flli DHRP sumf~

. . . .1. ~..._

~ -AL . .-. _ _ _ _ _-

CSbackup thru demin pump into 300-400 gal through the SF demin pump 11127/87 ODR 871156 2379-0993 iked area into diked area

[7T I~ - ' -. i *R inwater smple not counted to 11/30/87 OCR 871150. 2857-0201 -Uncontrolled are rainwater sample: eniomental LLD - -- I 1/14/88 ODR 880039 2857-0519 RB particulate air sample filter lost Plnt CSD - no gross alpha count 1/14488 LER 8801 -.

i t) , .]Lost !ar filter sample 1866-1754'

- - - -Com osie sample retention issue, no maton releases -'. .. . -

N radiologicat impact .

Lekage from BRHUT into drain sump, Loa survey taken, no results 1/18/88 COR 880055 2857-0827 Sample point leaking unto ground and into storm drain rferenced O R 880077 2 5 0727

' C na V~ T i ted sam pl -

~

.3 (AF'ux eilding steam condensate tn )(and CST) contaminaed -~ *

- T o tatlb[S CV I(vaphone with O]

1/27/88 ODR 880083 2857-0774 H3in CST 5E-6 H3 OR 880108 2857-0918 - tsryacupo rain trAatini 88-186 designated a rad release based on non-procedural criteria (contaminated srface oil which was removed and 2/25/88 ODR 880144 2857-1135 RUT radioactive release isposed as LLRW) l l~~~~ ode safety lifted on g on0e C~~ ~ .  : . -

311/88 PDQ 880367 2880-0415' COW cooler (0o80) -. Drained to TF below COWV cool60@2.61E-7pu~imI

-800 gals (two events) contained within 3/11/88 LER 8805 2596.0825 Letdown relief lift to RB sump RBsump E3115/88 PD Q I-.

80 6 I-

- - ~ - - -

2335-0707

. T~ [ T TTT 77T St a ~ e k * .

L turbine steam lea kaginto p oe tozone 52 (doghouse) 4 r7 F F ep o e ti n o e2: .-

3/15/88 P00 880369 DRcony, letdown cooler relief lifting in RB -91 gpm --800 gallon to RB sump

- -i ~- - * -I- - -7

- Ii- - issue concemned volume and inability to+ -

5~8.PDQ 880541 340-0191 - olleakage - 80 Gaflon per Day identiy exa..loctiLekage contained wit'hin the sysem (NCR-7875) MCM-1 65 leak at condenser 3/15/881 PDQ 88095 2334-0901 LAkpntain R#1418581859 PD 1 287~ 1:1141

- HVI 51 0 leaking (converted NCR'-

T 8 L - i L~I E-302C steam isolation (Turbine building) 3/18/88 PDQ 880963 2871-1157 C-029 leakage B to B leak PDQI .88078 2336-1221 -' -

Csl 37 in COW : - I

.Crud burst as no H3 detected -~ 0_s137 Q 7.3E-8 pCi/m I .! 1 3/19/88 PDQ 880933 2871-1 067 FWS.020 leaking -W start up to BOTSG A - 14

Date Doc.# I Doc Location Index description I Record remarks -. ' Surveydata

[PID088 9708M70 28175 Packingleap.k' 'Zraylineservicedrain' Cple dro pshninuntue 3/23'88 PD0 880012 2871-0001 MCM-168 leak B FW RECIRC line leakage I880048-PDO 2871.00812 Atmzr malfunction stem -30 atomizing regulator steam leak 3/29/88 PDQ 880051 2871-0085 Body to bonnet leak FW Sample valve TSS-094 leakage WR145861

- - - -. . Main Steam Reheater to turbine bldg drain (OWS to plant effluent) wrater trail -,(-450 Cs137iat 8E-QpCI/mI. mxrelease 4/5/88, PDQ 88008 2860-01319 Water into storm drain gal.Per semi-annual report) 8.IE-11CI.r mxexposurel1.8E-8mire 8.62E-3 pCi max WB dose - 1.88E-6 4/5/88 NA ANN RPT . Main Steam Reheater leakage 423 gallons to OWS to PE - .mrem.

- . ... - 6 2-48pOI Cs134 4.02E-2 ~Ci-.

[4-1/8 alelekae urbine building floor draini to PE -- ga. dose -4.07E-8 mrarn -

ANN RPT in Steamn Reheater 137 3.dEleakage9p.i toag-la 1.36E-2 pCi Cs137 Max WB dose 4/18/88 ANN RPT Main Steam Reheater valve leakage Turbine building floor drains to PE - -1 gal. 2.6512-9 mrem 137 detected dur rig routine57 in.16areas-66 5/10/88 5r PD0 N

II~i P

~~

881020 N 2860-1333 oitringPE BRHUT liner leak ner melted on aux boiler blowdown line.

water leakage between liner and tank mrem per semi-annual report

. .- . .MS2_=(PV_3014A)(M to ASC -:-

P '8804 21122 VWleak'~ supplyvalve) ..-

5/11/88 PD0 881025 2871-1228 FWS-542 leak FS-021 bypass valve (start up bypass)

-7 PDO1 88093 2334-1968,' Waste InA RHUT` durn maintenance 80 gallon spill Cs13~7 at 1.e-/Ci/mi. detected Several RulO8 particle found, (8 InB. 1 In A)from resin dewatering system pump. POS and associated piping to the (osibly contaminated at Trojan) several RHUT were deconned and Pacific dditional particle found during decon Nuclear Services Inc. equipment 5/0/88 PD0 881077 2860-1417 BRHUT hot particle ctivities. repaced as necessary.

.PQ 881120. inM ta ineshotel -ithin 50.59 bounds. 1-3a372-6 T-950-A RHUT improper hose routing.

6/1188 P00 881131 2334-1738 Uncontrolled release during pump down spillage on pavement D 0a 81179 3480.0255 LV"0 ta ek . E.35(AuxboIler)steamlea:.1' Cs137 Inoil phase pumped from OWS Cs137 5.6E-8 uC~lcc: No measurable 6/2/88 PD0 881181 8482-0169 Oily water separator *. (from turbine deck drains) contamination on August 1989

. r . .. -78.8poO - Z52E-Srmnrm per semi-HoggerRPTreleasei (Unmonitored , a nnual repiort.

'C'MAIN STEAM REHEATER Instrument 6/11/88 P00 881185 3480-0277 MSS-623 steam leak roots ensatory'measures taker% P Q0 PDQ 808 '--233408 nl -64 fie orc hc de--

• No reles inrpact, 6/25/88 ANN RPT Main Steam Reheater valve leakage Turbine building floor drains to PE - -8 gal. 269 pCi Max WB dose 2.27E-9 mrem

- R-15546A missinig'1ilter screern an .-. -

PD 88079 2335407768 tde-n monitoris -Nrease nat 7/11/88 j ~P00 881150 2334-1820 L1-3 inPDS R~edundant to 88100211120 83 In routine Concentration range A- 15

II.-

Date Doc. # Doc Location Index description Record remarks Survey data uvs PD0 881158 ~lO~aiuered Are153a monitor - no release impact.I CCW contamination has contaminated Spills at CCW cooler sample points and 8/388 P00 881502 2860-2141 TCW CCW & TCW contaminated, CCW pump drains PD,881437 2860-1958 L3505 steam leak . -35 (Aux boiier) steam ieak j 8/8/88 LER 8811 3017-1545 FWpump packing failure rog packing size Issue No release noted 8//8 P00 881444 . 2860'-1974 oo L ildu'pon pump stand R Cm '.

811 2188 PD0 881489 2880-2122 Packing problem AWpump Cs134/137 contaminated liqudfrom 0 p'irrfdvertentEi 8/12,88 ANN RPT GSl on drain &C shop sink (P0088-1498) 'MxWB dose 2.7E-7 mremy 3 gallon leaked from improperly tied Mx contamination on solidification pad 8/14/88 PDQ 881434 2860-1930 Slidification liner overflow at pad down fill head, 3K dprn/l OOcm 2

- -in..,. -

• . SS-88 (MS052bypass) main steam -

II ~ conrol TOW main steam reheater .

8/151/88 PDQ 881508 3480-0420 Packing leak,- E-302A. WRI53357 location/survey 2.55E5 ml H3 1760 pCi, Cs137 3.34 8/19/88 ANN RPT Main Steam Reheater valve leakage Turbine building floor drains to PE -75 gal. pCi - Max WB dose 7.36E-78 mrem 82/8ANRT.Min

.T.urbine Steam Reheater valve leakage ga building floo rintoPE-19' 8/23/88 PDQ 881549 2871-1870 T1 OA steam leak MS coil drain TANK leakage

- , ~- Supectfilurof installed C1386 'S'K -ii-e 82/8 PDQ 881587, -2871-1888 Wtcnaiteperimeter monitor: . 500dp/Ocm2-Csl37/Xe133, 1131 in A&B OTSG No stated cause PDQ closed lAW RSAP- 5.59 evaluation has good information 8/29/88 PDQ 881002 2860-1135 sem line 1308 R2 onrelease impacts 8/&29/88 P00~ 881579 777

~'2871-1703 Boyt iI T7T T.7 onet leak ' j T- MS-3087D main steam reheater 2nd saesample - 7.

FWseal leakage into uncontrolled -28130 gal (9.89E7 ml) between 8/13/88 H3 - 247 uCi. Cs137 5.98E-7puCi - Max 9/9/88 ANN RPT drain & W9/988 WB dose 1.63E-4 mrem L9L 21/8 PDC1 880808 2871-0666, oBfoleak DHS-003 ~ onrsaa~dtbneJn- No significantvlm oe jj 9/27/88 PD0 880112 2871-0177 B OTSG lower manway leakage <0.48 GPM II -Cs Co, I.Mn present from 32-7 to 4E--5 -

9g/28/~88 P00 881231 .2871-1459 pfJ er otmination RCS leakage'. . *..

Cs, Co. I, Mn present from 3E2-7 to 4E-5 i/~co disposable resin processed as 9)28/88 P00 881234. 2871-1 459 Jpolisher contamination RCS leakage radwaste 10/5/88' PDQ 880754 2871-1018 Bto BSI-00,. .Bron crystals at Body to bonnet joint - Nosgnfcnt lnoenoed 10/25188 LER 8814 3017-1 568 PZ spray block valve leakage <1 gpm No significant volume noted II ;..-..-

1 ,i *,  :- Beore/fte saples indicate no' 10/27/88 LER 8818 3017-1560 Faire tocontinuously monitor ASS Rpurge . .a . reas 10/28/88 LER 8812 3069-0058 Failure to isolate A RHUT Oly LER cover letter no details Released to basin not PE I I1117/88 PD0

• 1'I:

881839

,I

. I

. I i

II rro CCW P /U .,.

TSS (177) secondary plant chilled Nrater Ii~tem contaminated Csl137/1-3

-, - :s137 3.0012-07. H34.OOE-04,uCi/m Max dose 0.0005 mrem .

1.I.: II . I F ~- -

I 2' I 11/20/88 PDQ 881892 13480-0680 IXe contamination inSAS aux building Noble gases Xel133 6.8E-7. Xe135 1.24E-7 pC!ico A- 16

Date . Doc.# I Doc Location I Index description Record remarks Survey data - -

11 LER 3083-23 RCS leak 30 day follow up report no new data' Failure to continuously monitor AB I1I23/88 LER 8816 3083-2225 grade level vent 30day follow up report, no new data PEcenterline in spec after one hour 11218PO0 881907 T-991 overfiow . . Igallon

. hypo chlonite down storm drain Wflush of drain H3 75.8juCi, CsI137 6.22 pCi, gases and Iodine 78 uCi (max WB3 dose -

12/9/88 ANN RPT Hogger release (Unmonitored) Steam/gaseous release - 9.06E7 cc 2.99E-3 mrem)

LJan489 .PO0 890089 BafomtnInretention'ibasi' sin"n

.- beads and fines detected in NRB Jan489 POO 890091 Bead formation In retention basin Approx one cup resin detected In the SRB 2 prCi (Primiarly H3 w/Cs) 1989; 1//9ANN RPT BRHLFT Instrumentation leak . 20 gals. Down stormn drain near B RHUT smi-annual report, 1/17189 PD0 890058 8481-1155 Furmanite nipple steam leak MSS-623 (PO130227 root valve)

MSS508 steam vent "D"rmain steam Il /9PDQ 890064 8481-1200 Laigpereeater packing leak V2-STLT0518- Cold reheat to B Main Steam reheater 1/18/89 POO 890066 BX06995 Man way cover leakage above ladder to mini-mezzanine 11/18/839 PD0 '890067. 881-1219 Funrmanite nipple steami leak FWS027 A Main Feed Pump to LP 1)19/89 POO 890075 .8481-1253 Body to bonnet leak .- Cndenser ISO

- - -I a~

FWS3O4 BMAIN FEED PUMP to 2nd pt 19/9 POO0 890079. 8481-1274 Vatve leak FW heater drain MSS508 steam vent "0" main steam reheater leaks by causing steam leak from 1/2289 POO 890092' 8481-1305 Valve leak pie cap ASC031 blanketing steam to D main L.12 9 O 890095' 8481-1324. Vav leak team reheater,. ..

• V2-STLTO518- NRB sample containing contaminated 1/26/89 POO 890086 BX06995 Bead formation In retention basin rsin and sit Csl34, 137, Co6O P00 89017conami~aion~ ~folowupper PD0881892 Noble 111 3/19/89 POO '890357 3480-0831 Detectable fission gases Xe133/135 in SAS (clear WAl 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />) -

LER 8904 3071-1677, Error LWR BMW28 mn Issue No release noted 3/26/89 POO 890367 34i80-0740 XecontaminationIn SAS aux building Xe133 9.E-7. Xe135 1.99E-7puCicc' Clear of noble Cases W/l48 hours

[~~i V-STLTO518- Condenser pit sump contamninated -" -A .-.-.

89084TSGlek - - increasing from OTSGleakage ipstnn881120 P00oX06995for replacement request - (basin 4/389 POO 881522 3480-0441 R-15017B pump failure iNet monitor) No release Impact

~PD0 8981583 BX06995 V2SLT0518-duin Leaking relief valve Dea1an)6Mio ekaedrigccl et INoo ycetetconrtamnination orreleaselimpact 4/9/89 PO0 890412 - SAS contamination I Noble gases - no conitamination Impact V2-STLT0518- E380 blow down line near forced draft fan

.POO 890425 ~.BXD6995 .ASC nipp-le failure. .' I_- " . steam leak :.-. '

.i

- I A-17

Date Doc. # Doc Location Index description Record remarks Survey data E360(smail boiler) mud drum to F-3508 4110189 PDQ 890063 8481-1196 Pipe leak in -23 at weld failure north of shack In overhead pipe chase Steam header vent to AMainSteam 4/12/89 POO 890062 8481-1175 Pipe cap leak 3Xfrom MSS-504 Rheater

- . ,. .~, ~ Lmited to bermdaaea ~LIQUID 1i..V2-STLT0618- . iuid effluent release treatment system EFFLUENT RELEASE TREATMENT 4189POO 890431 BX06995 Rpuedisc failure liquid effluent RWS skd. - .. YTM ski V2-SLT0518-4/25/89 PDQ 890451 BX06995 Environmental Air Sample OOS Lss than 7 day sample run 5//9 PDQ 890486 PDV-36541 steam leak Small a'ux boiler atomizer. I . Nosignificant volume noted, V2-SLT0518-5/10/89 PO0 890499 06995 SAS Xe-1 33 contamination e133 2.1 E-7puCiml noble gas I i... -I - Detiled survey -700 feet out from - I ' .

51 P00 890512 CaCreek Radiological survey effluent V2-STLT0518- 3rd point heater drain to E360/365 MCM.

5/16/89 PO0 890536 BX06995 alve leak 084 E6/13/89 PDQ 890392 BX06995 Nrhpimeter monitor failure - a oio.ntrlae related -

20gallons into bemfied area of Tank Farm between Aux and Rx bldg near B MSS No contamination noted in post incident 7/1/89 POO 890600 3480-1009 CCW MIU (leak in system) peetration, srveys 7/89 PQ.V2-SLT0518-718 O'882019 *BX06995 T, 7TTT7-7, monitor high bickgro'und> .

Nor-eaeimactsetoin

, MDC four ordr magnitude beow*ale rt 8E-8juCi/mi (4016 gal lAW semi-8/23/89 POO 890634 8481-0370 B RHUT leak Tank leak into drain annual report)

~ 1f7 I- . ' .

TLO3P mionitoring araof d~licsited

.mtral near clay creek. RTLO.4N0 post Bsdon czrcumstances. no unusual 19 P00 890739' 8482-0630 -fsiedose rate above background onIOSB fence vs. site boundary fence. raig I Dec-89 P00 890701 Cotmination detected in OWS Cs137 6.2E-9juCi/mI H3at 435 pCi/i in well at feed lot SW Subsequent sampleswoevad 15/9 P00 890832 . 8482-0857 , Uepcdwllatrsample resultsstio.I ttumaivy 12/5/89 POO 890831 8482-0857 Unexpected runoff water sample results H3 at 520 to 580 pci/I.

LER 812 3067-1197. Fifur..re-establish'sample flow ABS Zminues,>ayi .

12/15/89 POO 890849 V1-3098-8587 Hzrdous waste inspection Numerous finding and observations No radiological Impact 1F N detctedinpsmpl cor~ined 1writhin plant drains.

12/16/89 POO 89085mS overflow . actviy etctd monmleitoredat basn..

12/22/89 POO 890775 3480-1407 Containment not set prior to head lift Procedural issue. No release indicated 149 POO 96Q004 : i~nRB I S alfeiaghtinfse Noreasehnpm j Secondary plant chilled water system From CCW make up. Drain and flush Cs137 1.62E-7. Csl34 1.23E-8, H3 1/19/90 P00 900015 3480-1821 cntaminated Cs1341H3 cmpleted S.E-4 pCi/mI

l. .

-f--

900016 jContam"inarted material discovered at" Contaminated burke tube recovered from Ngrudcontamination & decon PO0 1site portal - north salvage area noted A- 18

Date Doc. # Doc Location Index description Record remarks . Surveydata Leakage contained In leak chase (by 216/90 POO 900038 SF liner leaks Leakage past studs In wall of liner design)

LUquid sample NGS-556 (IN Training nd ecrd BULDNG cntaminated Liuid from NGS drain handle as 3i7= O 900076 382DO CS137 contamanated during dismnantlement 1.82E-OjuCI/ml Cs137 319/90 PO0 900061 3480-1946 SP failures R-15029. 40 Araradiation monitor - no release impact 3/16/90 'PD 9008 Vi -3098-8587 C lde6itiiat- -,-'"364~17.-osraieasmtos 3119/90 POO 890588 3480-0996 R-1 5045 set point Conservative error. No release Impact m,'.Arinistrative reporting issue - not relea'sme PD'890842 3400-1784 R-1 5017A005S> 30dy dcmnt - -

> 90 day hazardous waste drum 3128/90 POO 881940 3480-0711 storage Admin violation - no material release mall spiU below overflowied bucketnear

[3/29/9 POO 900103 842-1245 ARHUT agitator leak agitator.

Two drums hazardous material (89- Possible drum leakage behind paint 41390 PD0 890746 3480-1363 25/296) over 90 day limit Administrative issue.,. shed prior to transfer to poly drums West end storage tank -diarkstain . ***.,

4//0 P0 900081 3480-2026 indicating leakage caustic storage tank T.~,.-741 Lekgcotndwhibrmdaa 4/6/90 P00 .881422 3480-0575 Rad Monitor alarm on ceiling fan cycle Basin inlet monitor - No release Impact

- - . -. ~~Steiam leak PDV-38541 (e-36 M.,:,;,* ~ -

4/690) P00 90406 3480-0904 atomizer) mIf steam leak . - . . .

4/17/90 PD0 881887. 3480-0680 R-1 5106 filter clogged IOSB monitor - No release Impact 4/~0 PO0 9013)~3 3480-2130, R-l5045A flow transmitefil ,Local samples taken to monitor release . No relea!se Impact

- ump transferred to drum and disposed of 4/23/90 POO 900143 842-1429 Poly bottle emptied Into wrong sump down decon drain

[~/0 POO0 881542 3400-0615 Effluent compurter code error. TSviolation No release Contamination of LP (50# header) N2 ORCST backflow past PLS-630 into LP 4/26/90 PD0 900152 8482-1473 and associated systems H3 from liquid drain at 3.43E-2 uCi/cc N2 header

[~79015 882144 3 n STrepressurization.

CSTcontamination from N2 system during DWS (from CST) feed and bleed/survey drain/filled 8/11 190 <MDA nror to returntouretrced seM Bucket used to collect seal leakage dumped in sump, east of Spent fuel 5/3/90 PO0 900161 8482-1 498 FC demin pump leakage . ers which pump to RHUT.-

Turbine parts stored since 85 (wheni riler #20 (stagedonSuhetra release criteria was 1 mR/hr fixed) parts to basins) discovered to contain Cotminated turbine part west of 'west aoecurrent criteria (1 00 ccpmn) cotaminated equipment also.Miedt 519g O 900149 8402-1465 spray pond cnrlled. IOSB. J 5131/90 P00 900188 8482-1642 R-1 5007 failed SP . *.No Indication of release impact POO.'9019 R-15042882-651 aild o Indication of release Ipc 6/4/90 POO 890503 8482-0421 High coolant activity - Monitor reading concemn

[~P OO 900092 8482-1202 R-5017 severe spring .- 5 Procedure/Admin- -No release indicated A-19

It-Date Doc. # Doc Location Index description Record remarks Survey data Water trapped withfin FW piping drained FWS-010 FWN into (grade level) into grade level dehumidifier while opening 6/6/90 PDQ 900207 dehumidifier FWS-010 31<. .1..>. - .~ cotained within TF leakcas 6/8/90 POO 900210 8411-07ll27 drnkinbird leakagepta WDCbrera '

6111/90 POO 900135 8482-1352 RB exhaust supply fan Fn tripped during re-energizing H4SCB - No indication of release impact 6/12O9"POQ- 908 - 8482.1632 Iodine lflo~wcheckfaied .oindication ofrleaeipc RNGS CSD - no 1131 inventory - no 6129/90 POO 890700 8482-0602 1131at Lodi Environ. Air sample reease implications

• .I ormalControls for Environmental .

Polutin esutsapparent lab sample 7/09 O 624'nwl apl RW 1M) cntrol issue- . .

V2-SLT0518- Sample taken early due to schedule 7/12/90 PDQ 900239 BX06995 error Sample taken 9 vs. 7+/- 25% No unmonitored release.

F71169 POO 900247 ~-oredmg 8-428) 7/16/90 PDQ 900247 8481-878 Damrage/leaking source Ba133 #83-428 No release, no contamination i 7/189 POQ 900 241' '882-174 Monilorsour`cecheck.77. aradmonitor - . - .' orleseimipact Atviy on charcoal of Turbine. No release or contamination impact 7/19/90 PDQ 900220 8481-0811 dhumidifier ' C8& Csl134/137 at < 1.5E-1 2 noted 7/19/90 PDQ 900227 8481.0845 Channe falrs 1504&545 '  : ~ No release Indicated.~

7/25/90 PDO 880110 8482-0001 Plant effluent ODCM non-conformance No indication of release impact Valv eakag into uncontrolled drain in

~IAN R:T TF .I.~,

9/4/90 POO 900216 8481-0769 Basin release to support l&C Mntor set point during surveillance issue, Noimpact on release Ty~gon blew off REACTOR COOLANT

'i-Reatorcoolant i drain tank drain via .DRAIN TANK pump, spraying water on ***

9/4/90, PDQ 900215 8481-0760. tgn 6EDHR pumnproom sump flosand walls 9/11/90 POO 900281 8481-0978 R-15044. 045,545 COS w/o samples Aministrative reporting issue - Not release document 79/1T 1/ PDO 900285' 8481-11211, R115028 failed SP,' arm function issue - , .No related release 9/12/90 PDO 900219 8481-0799 Moitor line up error R-1 5548, 045, 044 Procedure/Admin - No release indicated II ~~ank famradioactive spills (various . . .,

96/0 PDO 900299' 848-1843 location) . No indication remediation completed '

10/18/90 PDO 900201 8482-16888 1131 in milk sample PDO cancelled 10210PDQ' 900198 8482-1657 Inossec n0C onts (CSD) i No release imp.;td Noindication of release or 10/29/90 POO 900272 8482-180 1 Hzardous Matenial program defic ency dmin issue - conaminnation.

BWGDT leaked to Waste Gas Storage POO 900328 BWGDT uncontrolled release-rank to ABS (monitored per 1990 semi-'

annual report);- . . I. <i' I'..

RS out of compliance with 11/29/90 POQ 890731 8481-1 395 29CFRI910.120 Programladmin issue A -20

Date Doc. # Doc Location Index description Record remarks . Surveydata

- Overflow of one drum on 1123/89

-' . .. . included radioactive reference (800 11/29890858 8481-1374'- EPA violation during walk down .. Hazardous 'material. Impact mainly Admin. cp) 12/3/9 PDQ 900129 8481-1438 R-1 5044 alert point Incorrect No release Impact PDO 90016e7'im 8481.1880 R-1 5049 count Emil No release impact No contamination measured in post spill 12/23/90 PDO 900363 VI -3098-8587 ARHUT sample'pump leak surveys Iooer - non-radioactive, post spill surveys PD0 900384 cooler leak I139-57PA'SS . bak ond I , - . . I I *I I.

IWS-730 H3 Evaporator -- 500 gals waterlH3 3.8E2-2. Cs137 3.6E-8 pCi/mI hru TF wall down storm drain south-east j(6.98E-2 Ci H3 &6.52E-2 pCi Cs137 12/25/9 PDQ 900367 V1-3098-587 ISpill inTF )f east tower total)

ANN RPT Valv leak on H-3 evaporator -40 gals. Down uncontrolled rain In TV -70 mCl H3, <lpui Cs137 >~.

OTSG backflow thru check valve during Contamination In sluice header and recirculation contaminated header. Cs1 37 61E-9 pCVml @CST, 2E-8 pCi/mI 1/14/91 PD0 910012 blenderldryer system blender/dryer, and aux. boiler blender - drier. I E-7puCVml @E3601365 Formula 85 MW Evaporator anti fr liPDQ hemical spill In OHT C108. gnt-20f 2 Spill originally reported as floorwa A B RHUT'S, Blender/Dryer, acid addition PDQ 900375 -Freeze damage various systems west tower

.V2-SLT0516- .'.

,910048 BX(06995 Chricallstorage' dministrative issue .Material stored In Aux vs. BS warehouse Standing water noted as no floor drains No contamination levels referenced in 10/1/91 PD0 910130 Blender/dryer (B/D) relief lifted to floor exist PDOD W ater from under asphalt -S feet west oNocontamiainlvesrfr ncdn PDQ 920003 V2-528819 Leak in 'A RHUT line during discharge AR"UT PDQ O H3 1.5E-3. Cs137 2.68E-5. Cs1 34 12/8/92 920075 CCW monitor alarm Inadvertent x-connect RCS to CCW 2.12E2-6 puimI

-PDQ

- -,Rain

.... water leaking thru street crack Into-930011 V2-1 333-934 3pill i -20 20No containiniation releae No contamination levels referenced in 4/13/93 PDO 930026 V2-1 333-9344 ARHUT liner leakage Contamination of A RHUT due to liner leak PDO Contamination in eascolntwr PDQ 930036 -sudge .Csl 37 up to 9.61-7, CoSO at 4E-7 pjCIg Tanker overflowed duning transfer from 9/2/93 PDO 930055 60gallon transformer oil spill stationary tank No offsite release

[]PDO 1930063 R-15045 OOSW/O sarnolng . .- No contamination or release impact H3 37.3 uCi. Co6O 8.33E-3 pCi, Cs1 34 12/26/93 PDO I 930088 ARHUT agitator leak 40gals. Down uncontrolled storm drain 3.15E-3pvCi, Cs1 37 8.52E2 2pCi RM-80stak oveflov' Nocontaniination or release Impact 21/4 PDO - 940022

- ~CRD cooler scupper. A HPI pump room cooler coil drain, SF-A-2 fan drain, and CR Drains isolated/rerouted, no activity Cotmination on Nuclear Service HVAC room drains routed to sanitary detect In sump or sewer system 3/294 P00 940025 Electrical Building sanitary sump grate sump. ARIFLO pump

-' . 1.5 gal ethylene glyco co antrmixture* _ _______

PDO 940052 Enie coolant spill downstorm drain... * -

A -21

Date Doc.# I Doc Location Index description Record remarks Survey data Total of - 1800 gallons lost. Main Feed Pump oil coolers, C condensate pump CCW leakage - 20 gallon per day - cooler, or A Main air compressor/after 8/1/94 PD0 940071 unknown location?? coaler potentially contaminated IF7T~ i Q . .. 90 gallons per year into 71/4PDQ 940074 V2-459-9237 DHS-068 leaking thru fuel transfer tube RW/otinment Norelease impact Floing of TB floor. Nuclear Service 8/17/94 PDQ 940083 PDS overflowed Electrical Building hallway H3, Co, & Cs E-9 to E-7 uCi/mI q CRDM components (contaminated)" 1000 ccpm fixed. 200 pm x0 loose No residual contamkiation reored i~n 4 PDQ 101 .940092 ndtsouth I scrao yard-.w'. hin wooden crate , -exnsvpodicery iurves 5 gals leaked onto graveli/ground near Cs137 at 6E-8 pCi/cc - insignificant 11/21/94 PDQ 940100 CCW leak in TF CCW cooler vent impact 91/5 PDQ 950076 177. TT FT T'~ T RI-UIT leak (agitator) 7 00gals. Spilled. -10gl.down.

unotoldsomdananual

-7pCI (primarilyH reportjJj 1-3w spr19 12/11/95 PDO 950102 Detenioration of radwaste drums No release impact 12/20/95 PD 514 "71777 T':71 T'transfer~ line ruptured.~no acid.

V-333-9344 Frk lift fell into acid line trench` rlae rm rnh H3at site boundary and lake per Cotrols for Environmental Pollution 1/22/96 PDO 960015 analysis report Dring maintenance of CDS 6- 71 5/22/6 PDQ 960041

-Ii V2-1333-9344 52.525,557,and/or 564, - 1 0-1 5gallons IndqaeW-Preview. water spill.o contaminated water was spilled on floor C137 at 1.66E-8 pCi/cc. No release.

Conty seemed to indicate that soil Peeling paint blowing around site, similar tsting and remediation would need to 1/21/97 PD0 97001 V2-1333-9344 Pb paint peeling from DFO tank situation on turbine deck noted beevaluated (Ken Hawkeswood) p dum7777.e7 alarmred recycler Cnainated boiler end bellar 122/7P60 970082. montor.oC1374.2E-3 p/Ci(2000 cpm)

EPA-Radiological Quality Assurance 2/12/98 PD0 980009 V2-1333-9344 rsults COS Aministrative issue, not release related No contamination or release impact Ii. ... . . For to six sticks, contamination up to 400 Suvyof associated materials, 7 Cotmntdsafling discovered ccmno contamination of B warehouse bad,21 more sticks; one clamp fo 3/1/8 POQ 980017. . ner Bwaehose . deectdm 4/29/9 PDQ 980025 Unmonitored ABS release 4/30/98 larmed recycler survey found - 13.000 in2 of material with Total actvt 0.1p xncontaminatio 980026/9monior. .max fixed of200 ccpm- or rele a e m a t j

-16 gals. Of acid escaped bermed area of Acdspill cause basin to divert on low acid/caustic storage tanks and entered 6/3/98 P00 980035 pH drain to PE Jul-98.- PDO 98004 L SWline cutin TB: .pilled into"TB .

T50A H3 concentration may exceed 10/14/98 PD0 980068 NPDES limit 12/26/9 PDQ 980081 rTOu evpo ator ~pilliflooded berm overflow intogravel.~ >. .

-200 gals Concentrated Boric Acid Storage Tank dumped into crud filter, M/U pmp rooms, -20 hallway and -27 pump 2/25/99 PDQ 990012 Spill in -20 aley I .4/1/99- PDO 990029

_j L~ -. _-1 I _ A.B 1-1AC expansion tank drained to AB

' Ime.zzihine floor drain' I  :

% Nitrate-Borated water . .I .. I. J A - 22

Date Doc. # Doc Location Index description I . Record remarks Survey data Doc. U Doc Location Index description Record remarks Survey data Date Detected at exit truck monitors, never left shte. Contaminated tape from BWST Miaterial removed, dumpster surveyed 4/1199 PO0 990031 :ontamination detected In dumpster remediation tent 3nd released PDQ 990028, Spill MIU Pump and B HPI rooms BWST drop line-

- 1135E2-5, Cs1 37ICo60 51E-7 pCi/mIiIn Leak from A RHUT discharge pipe 200 gal on ground. 25 gal down storm water. Soil and gravel In area 9/9/99 POO 990074 during B RHUT release drain, no off site release Csl 37/Co6O 9E-7 uCi/gm

.PDO0 .990087 0.5 by 39 inch split inexhaust duct <1% based on release calculation's)

Potential unmonitored release from B Approximately 450 gal of water leaked RHUT through agitator seal following through the agitator seal and went down a H--3 at 4.42E-06 and Cs-i137 at 2.80E.

2J7102 P00 020015 maintenance on the agitator . nearby storm drain 09 PCi/mI

.. 5gallon can of trichlorotriflouroetharie ODR 840051 5olvent down storm drain.

bu4nd draining into storm drain south side -

iux boilers

• . 1:

'121-14~concentration - not a radioactive ODR 840106 Dffsite release ssue

. Dnim rupture In hazardous mat storage ODR 880097 2379-0112 area. -Hoclrtoi waste - .Non radioactive Issue ~

Large document regarding caustic and POO 930050 Leak spent regen pump acid spills OD 840087 Particulate filter broken COR 840308 3rd qtr dose calculation's Error found and corrected

- ~-No disposition~notedl, (text from 900240

-. -. ' 1445 pCi/I, Controls for Environmental stts analytical error) REMP limit not P00 '890889 . 113In Tripley well (RWW2 I MO)' Polution states resufts valid e PD0 990070 ABvent monitor 005>30 days System abandoned -Last release 8/1811999

-. Sewage to effluent terrace flow rate vs.

CDS61lO open controls flow to sewage National Pollutant Discharge Elimination PO 900123 8482-1328 yse (contamination??) S

.ystem permit Issue No contamnination/no release LER 8323 0204-0948 P-261A (DHR pump) Electrical breaker issue No release noted LI 6R 603 2802-48 13cotmnton (no PA) . F mnovement issue, N ees COR 760046 2802-0507 SRB dilution incorrect Conductivity and TDS Issue

. ~ .Tanker return from vendor, not facility 790088006 02-1074' Residual radioactivity In takrI9 icensed material .

ODR 8011Plant effluent exceeding limits ewer release - Non-radioactive 802-6 aia~tvlqirla~. Lack of signature on permit issue .7 No actual release Involved COR 810129 Radioactive liquid release Lack of signature' on permit Issue No release Impact

-OR810148 Plant effluent w/o sample ,z . .No additional informatonK on ODR COR 820002 Plant effluent w/o sample

[L ODR 820011 wmissed ABS samples Noble gas & H-3 only-01DR 820028 Radwaste transport (3123/82) No impact to facility lmproper handling coontiaminae OCR 8360004 material .* oR is & e .dministrative issue -- .~,

A -23

Date Doc. # Doc Location Index description Record remarks Survey data ODR 830106 Effluent release limit exceeded pH - not radioactive Ventilation system fa; beating failure and potential itcould have on release ODR 830151' Radioactive release monitorin' ODR 830183 Plant Effluent Chlorine spec- not radioactive OR 840020- Hoei oln oeFre protectio'n watersytm ODR 840021 Discharge valve leaking Retention basin outlet valve No measurable release indicated mn ssue -no 50.5gdone whenused ODR twaste holdu ak3/4 as Evaporator bottomrs lank L ODR 840121 Lekage into RS spray add tanks Boron concentration issue 840122.. qd fuetdscaulation(s) mm__ Issue'ii~

ODR 840140 Tube leak South HP conde nser Non-rad issue 84083Retnnbsi ees ncorrect dilution flow - not> MPG ODR 840194 Portal monitors BackgmoundlMDA issue 8D.

40224 Plant Effun  : ETS dose calculation issue ODR 840249 MSs 051 Os issue plant S/U

- -7 ODR 8420MS02 7 pissue pl ntS/IJ, OOR 840254 HEPA filter bank Filed SP No release impact ODR 8422Prticulate filtr iternot retained forCS - eno ODR 840293 354-1533 Ovrfl lank ied beyond spec, not capacityno spill OC F78 ees xeddlmtAuto ' make up ntrlaeeae ODR2356-0726

&50276 fallu 10.000 gpm~ flow ratentrlasiteae Unable to access radiological release ODR 850458 2358.1167 from secondary system Timely access to information. Not relevant to actual release

- 1[TiT~~~it toed relseseuig- W OCR 860164 '2857-1772 srelac ..-  : Potential (admin review of S.P.)~

ODR 860301 2348-0041 HVS possible unmonitored release Administration OCR 800351, 2348-0298 Sorcs231-240 mjssing :. Bekm7nsources used to setup LSC - Rtmed to Becka BBC component security issue, not OCR 860550 2348-0925 Shed not secured radiological OC 707 2349-21 9

__ iF open.

dan 7I

mcocrnR:pool level, not actual

?creCe 1 1 1 OCR '870199 1569-0001 Re: OHUT) N/A - in QHUT -notregarding OHUT ODR,7073 2926dfferS M Electrical is'sue'- notradioactive issue.

Emergency Diesel cooling water pump.-

OCR 870819 2926-0952 Coling water pump seal leakage not radioactive ODR, 871016 2379-0692, m~assfrom polisher samnple ine onrdiologicalissue POO I 880115 I 2335-0915 Missing source #204488-5 Source found A -24

Date Doc. # Doc Location Index description . Record remarks . Survey data I POOt880153 # o oaio ne esrpinRcodrmrs uvydt

~ 7 ODD815 conyv.

0R860153 pill while pulling lower manway OTSG No significant volume noted P00 880156 ODR cony. FP boron concentration Issue POO 880159 , ODR conv A> Diesel SP issue PD0 880161 ODR conv. RI15044&045 flow calibration SP Failure No release Impact P0 814

• R conv. Posting ofseutyurissue~

TS Issue Re: pump comrbination/duration PD0 880165 CR cony. at power

[111-PD0 880167 0Rc6onv. Waste Gas surge tank maintenance issue Admin Issue Re: LCO reference in SF PD0 880168 ODR cony. Handing procedures POO 880171ODcov USAR Cable trary issue POCO07 R cony. Procedure reference tech spec Issue PP0 '880173 . OR cony. - >- Fire Protection SP Issue' USAR issue Re: Release Via NSRW P00 880175 D0Rconv, evaluation NSCW temp Increase 87 to 95 of USAR.

807 -b R6conv, issue.-

Turbine IMain Feed Pump trip set point PO0 880177 OR conv. issue . -

PO0 880178 ORcnv. atry chargerrnaintenance Issue.

POO 880180 ODR conv. Test procedure issue P0 808Oocnv. Posting offre watcssue PO0 880187 ODR cony. Emergency lighting duration issue

P00 880188 ODRcov.' Electrical X-tie load'Issue PO0 880202 ODR cony. No fire watch PO 880264 O0DR crw. ' . R not LAW ECN admin issue

- Failure to perform scheduled maintenance POO 880331 ODR cony. on BV turbine component Issue P 880341 OR conv. -Efflun mxforertng issue' Procedure Issue Re: independent POO 880352 OD R cony, verifications TS Issue Re: mn miu nirnuryfeolneti PO837ODIR COn v..--, during operations..,

POO 880360 QOR cony. EFIC power restoration Issue I- - - Y-*--- 7 POO 880364 *OR conv.. EFIC power restoration Issue . -

PO0 880365 IODR cony. Operability Issue No release Impact PO0 880415 -2871.0377, ae npl box H71537 Ra'in water A -25

Date Doc. X Doc Location Index description Record remarks Survey data POO 880587 2871-0595 Leakage NRW. NRW no record of contamination PO M1 2340 18U.Jrid ~ ,rcpNon-Rad issue

-1Lqi udrgvenrcp L__

P00 881104 2334-1555 Moisture in J-box Non-Rad issue I lwont HUT - spray on Nocontamination monioe, aential'I:

7 L jJ

.Drn LPQ 88112, 2334-1741 euiprnent (Main Transformer/others)ise ..

PO0 881340 2871-1544 C RHUT (contaminated??) Chmistry issue (CWQCB)

- - - .!r 77 I V2-STLT0S5I - Spent regen line acid waste sump -I.

POO 881748' .BX06995 reoe,,Amin issue Noa POO occurrence'.

V2-STLT0518-PDQ 890078 BX06995 Inner cooler relief lift Emergency diesel air start compressor on-rad issue 1;8-0122 related to ie wrap ofrazo ri~bo POO 890112 ' .Bdto bonnet leak . - ascrt ec Pin hole leak between MSS-031 & ASC-POO 890302 Socket weld leak 620 No significant volume noted PDQ 17 890340 STT018

.BX06995 7T T T T xSem weld failure/leak temleak E360/36 d ownstream ASC47 ntaT mT o inificnt voluenoe V2-STLT0518- Steam leak E360/365 steam trap line PDQ 890349 BX06995 Aux Boiler steam leak downstream ASC-473 No signifcant volume noted P0 931 .B069 -etrrelief cover missing Scrity TB to TF access-issue I~ ~

V2-STLT0518.

P00 890354 BX06995 LPturbine line shear Codenser air in- leakage issue P78034 eaingFow transmte -. E12 Sci rin 4 tank releater No significant volume note V2-STLT05 18- .. L. .i..... . e .

POO 890382 BX06995 Fed heater relief cover missing Security TB to TF access issue

~- ~ 9QT9 V2-STLT05aiue s1t8) _ _.

7- -

POO 8934. B095 n) alve electrical failure issue ,.

V2-STLT0518- Monitor failed SP due to failure of BD valve PO0 890418 BX06995 Cooling tower blowdown monitor failure to close L0 941 . .. .Mntrflow rte . . .R15546A failed SP ICompensatoiey'ctistaejjj V2-STt.T0518- On reinspection, resin levels venified as POO 890438 BX06995 Rsin missing treatment normal POO ~890615 ABSwfthoutsamPling -.- >.18 hours CSD O mat3O V2-STLT0518-P00 890758 BX06995 SF pool liner separated from concrete Lner deformation No change in leak or leak rate P00 900331 . SCW contamination H3 ol 50.59 no impact PO0 910086 V2-1331-864 Uncontrolled release pathway MSS steam trap cover 0.005% app. I P00 910148 V2-1331-8864' xinspent regen tank BDcondensate tank ~ - -'Nrlae I*;

H2analyzer Csl 37 source (30,uCi) not POO 910158 V2.1331-8864 Radioactive source incident added to inventory No release 4'. -.

_11 920014 jV2-1331-8864 PDO0

. - -I I lBRHUT6leak FV95008 .

. . , .. . -it Water contained in drain sump to basin.~

6--l-Jorelease ,,:  : I A -26

Date Doc. # Doc Location Index description Record remarks Survey data Yellow II labeled package received by security and transferred to Training and No loss of package Integrity or PDQ 920019 Radioactive package receipt Records building W/O RP notification :ontamination noted dAux g uncontrolled area Auxiliary Building Grade Level, small area.

PDO 930049 V2-1331-8864 contamination. - - decon and released RPOR-93-002.

Eli PDQ P~O 930053 Mislabeled radioactive material V2-1331-8864 package aux building Box DR vs. ID sheet issue Gage issue leakage contained in poly -

No contamination or release impact PD0 950062 V2-1 333-9344 Unmonitored SF leakage- - bottle - .  : --- No contamination or release impact POO 960063 V2-1333-9344 Leaking drum found in IOSB No contamination noted No contamination or release Impact Ir I .1 CSCAHEPAv'acuum used outside Location not noted -no contamination POO 970021 V2-1333-9344 CSCA . . *. noted .- - No contamination or release impact 7919 Weld leak Pin hole in weld socket No significant volume noted

[1211 ODR 8033 810051 West

~ perimeter T' TLreadin

. ai..

Rx Coolant spill (filed 4/19181) h Barrel_e farm*

COULD NOT LOCATE

..t-Li-- .-

ODR I

810200 Unmonitored Radioactive release

- ULD NOT LOCATE -

O(12/4/81) .

.- .  ;-----1-;-:-----:-.

Li ODR ODR 820091 860837 2348-1248:

Water leak from respiratory eaning iler . .

PZR to OTSG via NGS system 'A' No ontamination detected In water released or surfaced wetted  ;

According to semi-annual report 1.36E-2pCI Cs137 released in -56 gal.

V2-STLT0518- MSS524 ADV bypass in TF +20 PO0 890097 BX06995 Steam leak at furmanite fixture mezzanine

- . i

-IIPDO 890538  ; Surge tank level drop -CRI sample valve left open  ; .

- 7 gal. Via FWS-020 Closed lAW POO POO 900316 Unmonitored release 910006

- I - . _, . . . I . . . - . I.. ., . .

P.O 910015 SWSlinebreakinTF Q 250 gals down uncontrolled drain -3OpCTCs137/Co60 : '.. .

I A_ Ax 1. - BA - 1. A d - .SI. . I I .. .

POO 980082 Holes in drum 93-0220 Discovered during shipment /preps A-27

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Appendix B Personnel Interview Program

Rancho Seco Historic Site Assessment Questionnaire (RSHSAQ)

Introduction The historic site assessment is part of the license termination program for the Rancho Seco Nuclear Generating Station.

This assessment is intended to identify areas, facilities, and systems that may have been contaminated with radioactive materials during the plants operational and post-operational phases.

Once these radiological impacted areas are identified, additional investigations will determine the scope and magnitude of the contamination and their potential impact on demonstrating compliance with the license termination release criteria.

History Rancho Seco Nuclear Generating Station operated from Septemberl974 through June 1989. During this period several secondary and auxiliary systems were contaminated as the result of primary to secondary leaks. As a result, there are many areas of the plant that have become contaminated through leakage from or maintenance on these 'systems not anticipated for in the original plant design and procedures.

In order to accurately and completely assess the potential radiological impact on post license termination occupants of the site, 'a thorough knowledge of these areas is required.

Currently known conditions Many of these areas, systems and facilities have already been identified based on the investigation to date.

This investigation has included reviews of:

a) Potential deviations from quality (PDQ's),

b) Occurrence description reports (ODR's),

c) Radiation protection decommissioning projects summaries (RPDP's) d) Interviews with facility staff personnel Attached is a map of the restricted area depicting the primary areas of known radiological impact (highlight yellow) including the Reactor building, Auxiliary building, Spent fuel building, Turbine building, Interim on-site (radwaste) storage building, and Barrel farm.

Informational needs The decommissioning staff believes there may be additional areas of impact (areas where radioactive materials were stored (i.e. the turbine rotor shed), spill locations (i.e. RHUT's and Aux boilers), and temporary maintenance locations (machine and fabrication shops).

The decommissioning staff appreciates your willingness to assist in the process of identifying these additional areas.

The decommissioning staff is confident that, with the extensive experience of personnel still remaining at the facility, only a limited distribution of this questionnaire outside current facility employees will be required.

B -l1

U-Instructions Included in this package, you will find the Rancho Seco Historic Site Assessment Questionnaire (RSHSAQ) and the site layout map showing areas of known potentially impacted areas.

1. Please complete section A of the RSHSAQ regarding personal and contact information.

2. Please complete section B regarding your knowledge of any systems, facilities, or areas of potential radiological impact not already identified. In your description, be as complete as possible including where, when, and what occurred that leads you to believe these may require additional investigation. Use the attached site map (figure 1-4) or draw a map of your own to identify the system, facility, or area as nearly as possible. You may also indicate an area on the attached map and provide an explanation as to what may have occurred leading you to believe that a radiological impact may exist there.

3. Should you have any questions regarding the questionnaire or have difficultly describing the area or circumstances surrounding a potential area of impact, feel free to contact Dan Tallman at ext. 4081.

4. Please complete the questionnaire and return to decommissioning (attn: D.

Tallman MS N103) by Monday, September 17, 2001.

Thank you for your cooperation and assistance in this project.

B -2

Rancho Seco Historic Site Assessment Questionnaire Section A:

Name:

Phone: Ext.

Section B:

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Approximate Date:_

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B -3

Section C: (continued)

Approximate Date:_

Location:

==

Description:==

Approximate dates:

Location:

==

Description:==

Approximate dates:.

Location:

==

Description:==

Approximate dates:.

Location:

==

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B -4

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Comment details documented in impacted area summaries Systems Bldg Maint. Fab/ Weld NPS Fab Aux Boilers Aux Building B Whrs Barrel Farm BOP Shop C Whrs Circ Basin Shop Shop I 1 1 1 2 1 1 3 1 4 1 1 1 1 1 5 1 6 1 1 1 7 1 1 8 1 9

10 1 1 1 1 1 11 1 1 12 1 1 13 1 1 14 1 1 15 1 1 16 1 1 17 1 1 1 1 1 1 1 1 18 1 1 1 1 1 1 1 1 19 1 1 1 1 20 1 1 1 21 1 22 23 1 1 1 24 1 1 25 1 1 1 26 1 27 10 7 4 5 13 4 2 16 2 8 B -6

Upper/

Outer GRS Non- Plant Ret. Storm T&R Tank Tool Turbine H3 Simulator Storage Sewer WRHS ISOB Rad Effluent QHUT Basin RHUT's Drains Lab Farm Room Building Evap Building Yard Plant 1 1 2 1 3 1 4 1 1 1 1 1 5 1 6 1 7 1 1 1 1 8 1 9

10 1 1 1 1 1 1 -1 1 11 1, ' 1 12 1 -- 1 13 1 1 14 1 1 15 1 1 16 1 1 17 1 1 1 1 18 1 1 1 1 19 1 1 1 1 20 1 1 1 1 21 1 1 1 1 1 22 1 1 23 I I 1 1 1 24 1 1 1 I 1 25 1 I 1 5 4 4 11 2 14 5 1 13 1 9 3 1 1 1 B-7

This page intentionally left blank B- 8

Appendix C Area Summary and Preliminary Survey Unit Identification

SUID 800 - Industrial Area - Balance of Plant Elevation Relative to Room Name or Approximate SUID # Building/Area Grade. Room # Grid Coordinates Area (M2)

Reservoir filter/pump AH22 - AF23 of RSNGS Plot 800001 station Grade NA Plan 5575 Helicopter Landing AH24 - AF27 of RSNGS Plot 800002 . pad Grade .NA Plan 11150 AH28 - AF30 of RSNGS Plot 800003 "South" scrap yard Grade NA Plan 8365 AE28 - AC29 of RSNGS Plot Area surrounding Plan WME of "C"RHUT and 800004 Bechtel Building site Grade NA GRS FAB SHOP PADs 5575 AG32, AF32-39 of RSNGS Plot Plan W/E of that portion covered by "C" warehouse and the south most portion of 800005 Road "C" south site Grade NA the TDI building. 9300 South area directly AG32-39, AH31-39 of 800006 adjacent to IA fence Grade NA RSNGS Plot Plan 15800 Area bounded by AC17, X17, V19, P19, N21, M21, M24, T24, T21, & AC21 of RSNGS 800007 East Industrial Area Grade NA Plot Plan 79,000 C-I

SUID 801 - Receiving Warehouse Elevation Relative to Room Name or Approximate SUID # Building/Area Grade Room # Grid Coordinates Area (M2) 500001 Receiving Warehouse Grade NA W 39 of RSNGS plot plan 175 Receiving Warehouse 500002 (Upper - Outer) Grade NA R-S 38 of RSNGS plot plan 150 Storage Yard S 37 - R39 of RSNGS Plot 500003 Bordering 801002 Grade NA Plan 3700 Parking lot east of M40 - X44 of RSNGS plot 500004 Receiving Warehouse Grade NA plan 55750 RSNGS access road Extending through cells 801005 hwy 104 to 801004 Grade NA between S45- E52 3500 SUID 803 - Quonset Hut Elevation Relative to Room Name or Approximate SUID # Building/Area Grade Room # Grid Coordinates Area (M2) 803001 Quonset Hut Pad Grade NA M 34 of RSNGS Plot Plan 465 Historic waste yard R 34 - M35 (plus S 35) of 803002 south of Quonset Hut Grade NA RSNGS Plot Plan 13,000 C-2

SUID 804 - Personnel Access Point (PAP)

Elevation Relative to

• Room Name or Approximate SUID # Building/Area Grade Room # Grid Coordinates Area (i 2 )

X 36 - W 37 of RSNGS Plot 804001 PAP Grade NA Plan 465 X 36 - W 37 of RSNGS Plot 804002 PAP Second Floor NA Plan 465 X 36 - W 37 of RSNGS Plot 804003 PAP Roof NA Plan 465 SUID 805 - Administration Building Elevation Relative to . Room Name or Approximate SUID # Building/Area Grade Room # Grid Coordinates Area (M2 )

AF 35 - X 37 of RSNGS Plot 805001 Admin. Building Grade NA Plan 15,000 AF 35 - X 37 of RSNGS Plot 805002 Admin. Building Second Floor NA Plan 15,000 AF 35 - X 37 of RSNGS Plot 805003 Admin. Building Roof NA Plan 15,000 C-3

SUID 806 - Spray Ponds Elevation Relative to Room Name or Approximate SUID # Building/Area Grade Room # Grid Coordinates Area (M2)

Grade and R26 - N28 of RSNGS Plot 806001 West Spray ponds below NA Plan 6500 Grade and R 31 - N 32 of RSNGS Plot 806002 East Spray ponds below NA Plan 6500 Area bordering Grade including R 25 - N33 of RSNGS Plot 806003 806001 & 002 hill side/berms NA Plan not including 001 & 002 6500 SUID 808 - Cooling Tower Basins Elevation Relative to Room Name or Approximate SUID # Building/Area Grade Room # Grid Coordinates Area (M2)

West Cooling Tower W 26 - T 28 of RSNGS Plot 808001 Basin Below Grade NA Plan 13900 East Cooling Tower W 30 - T 33 of RSNGS Plot 808002 Basin Below Grade NA Plan 18500 Area around East and X25 - S 34 of RSNGS Plot 80800' West Cooling towers Grade NA Plan W/E X30 (823001) 23,300 C-4

SUID 810 -Tank Farm Elevation Relative to . Room Name or Approximate SUID # Building/Area Grade Room # Grid Coordinates Area (mi2 )

NA SF Cooler Pad bounded by 810001 Tank Farm Grade columns H, K, 3.4, & 4.3 NA "B" main steam sump Area bounded by columns 3.4, 4.4, 810002 Tank Farm Grade K, & the containment building.

C-5

SUID 811 - Reactor Containment Building Elevation Relative to Room Name or Approximate SUID # Building/Area Grade Room # Grid Coordinates Area (M2 )

811200 Reactor Bldg From -27 to -19 NA Primary Bioshield (360 0) 811201 -19 to -10 NA Primary Bioshield (360 0) 811202 -10 to -2 NA Primary Bioshield (360 0) 811203 -2 to +5 NA Primary Bioshield (360 0) 811204 +5 to +13 NA Primary Bioshield (360 0) 811205 +13 to +21 NA Primary Bioshield (360 0)

NA A - D-Ring Wall [8 foot vertical W/E of primary Bioshield [12 811001 -27 to -19 feet E&W of PRV centerline]) -82 NA - D-Ring Wall [8 foot vertical W/E of primary Bioshield [12 811002 -19 to -10 feet E&W of PRV centerline]) -90 NA A - D-Ring Wall [8 foot vertical W/E of primary Bioshield [12 811003 -10 to -2 feet E&W of PRV centerline]) -90 NA - D-Ring Wall [8 foot vertical W/E of primary Bioshield [12 811004 -2 to +5 feet E&W of PRV centerline]) -90 NA A - D-Ring Wall [8 foot vertical W/E of primary Bioshield [12 811005 +5 to +13 feet E&W of PRV centerline]) -90 NA A - D-Ring Wall [8 foot vertical W/E of primary Bioshield [12 811006 +13 to +21 feet E&W of PRV centerline]) -90 811007 +21 to +29 NA A - D-Ring Wall [8 foot] -108 811008 +29 to +37 NA A - D-Ring Wall [8 foot] -108 811009 +37 to +45 NA A - D-Ring Wall [8 foot] -108 C-6

SUID 811 - Reactor Containment Building Elevation Relative to Room Name or Approximate SUID # Building/Area Grade Room # Grid Coordinates Area (M 2) 811010 Reactor Bldg +45 to +53. NA A - D-Ring Wall [8 foot] -108 811011 +53 to +61 NA A - D-Ring Wall [8 foot] -100 811012 +61 to +67 NA A - D-Ring Wall [8 foot] -46 NA B - D-Ring Wall [8 foot vertical W/E of primary Bioshield [12 811013 -27 to -19 feet E&W of PRV centerline]) -82

- NA B - D-Ring Wall [8 foot vertical W/E of primary Bioshield [12 811014 -19 to -10 feet E&W of PRV centerline]) -90 NA B - D-Ring Wall [8 foot vertical WIE of primary Bioshield [12 811015 -10 to -2 feet E&W of PRV centerline]) -90 NA B - D-Ring Wall [8 foot vertical W/E of primary Bioshield [12 811016 -2 to +5 feet E&W of PRV centerline]) -90

- NA - D-Ring Wall [8 foot vertical W/E of primary Bioshield [12 811017 +5 to +13 . feet E&W of PRV centerline]) -90 NA B - D-Ring Wall [8 foot vertical W/E of primary Bioshield [12 811018 +13 to +21 feet E&W of PRV centerline]) -90 811019 +21to.+29 NA B.--RingWall[8foot] -108 811020 " +29 to +37 , NA B - D-Ring Wall [8 foot] -108 811021 . +37to+45 NA B - D-Ring Wall [8 foot] -108 811022 +45 to +53 NA B - D-Ring Wall [8 foot] -108 811023 .

- +53 to +61 NA B - D-Ring Wall [8 foot] -100 811024 . +61 to +67 NA B - D-Ring Wall [8 foot] -46 C-7

SUID 811 - Reactor Containment Building Elevation Relative to Room Name or Approximate SUID # Building/Area Grade Room # Grid Coordinates Area (M2) 811025 Reactor Bldg -27 NA 0-30 degrees lower - 88 811026 -27 NA 0-30 degrees upper - 88 811027 -27 NA 30-60 degrees lower - 88 811028 "-27 NA 30-60 degrees upper - 88 811029 -27 NA 60-90 degrees lower - 88 811030 -27 NA 60-90 degrees upper - 88 811031 "-27 NA 90-120 degrees lower - 88 811032 "-27 NA 90-120 degrees upper - 88 811033 -27 NA 120-150 degrees lower - 88 811034 .. -27 NA 120-150 degrees upper - 88 811035 -27 NA 150-180 degrees lower - 88 811036 -27 NA 150-180 degrees upper - 88 811037 . -27 NA 180-210 degrees lower - 88 811038 -27 NA 180-210 degrees upper - 88 811039 -27 NA 210-240 degrees lower - 88 811040 -27 NA 210-240 degrees upper - 88 811041 -27 NA 240-270 degrees lower - 88 811042 .. -27 NA 240-270 degrees upper - 88 811043 -27 NA 270-300 degrees lower - 88 811044 "-27 NA 270-300 degrees upper - 88 811045 -27 NA 300-330 degrees lower - 88 811046 .. -27 NA 300-330 degrees upper - 88 C-8

SUID 811 - Reactor Containment Building Elevation Relative to Room Name or Approximate SUID # Building/Area Grade Room # Grid Coordinates Area (i 2) 811047 Reactor Bldg -27 NA 330-360 degrees lower - 88 811048 -27 NA 330-360 degrees upper - 89 811049 Grade NA 0-30 degrees lower - 88 811050 Grade NA 0-30 degrees upper - 88 811051 Grade NA 30-60 degrees lower -88 811052 Grade NA 30-60 degrees upper - 88 811053 Grade NA 60-90 degrees lower - 88 811054 Grade NA 60-90 degrees upper - 88 811055 Grade NA 90-120 degrees'lower -88 811056 Grade ' NA 90-120 degrees upper - 88 811057 Grade NA 120-150 degrees lower - 88 811058 Grade ' NA 120-150 degrees upper - 88 811059 Grade NA 150-180 degreeslower - 88 811060 Grade NA 150-180 degrees upper - 88 811061 . Grade NA 180-210 degreeslower -88 811062 Grade NA 180-210 degrees upper -88 811063 Grade NA 210-240 degrees lower -88 811064 . Grade NA 210-240 degrees upper - 88 811065 Grade NA 240-270 degrees lower -88 811066 Grade NA 240-270 degrees upper -88 811067 Grade NA 270-300 degrees lower -88 C-9

IIL-SUID 811 - Reactor Containment Building Elevation Relative to Room Name or Approximate SUID # Building/Area Grade Room # Grid Coordinates Area (M2) 811068 Reactor Bldg Grade NA 270-300 degrees upper - 88 811069 Grade NA 300-330 degrees lower - 88 811070 Grade NA 300-330 degrees upper - 88 811071 Grade NA 330-360 degrees lower - 88 811072 Grade NA 330-360 degrees upper - 88 811073 20 NA 0-30 degrees lower - 88 811074 20 NA 0-30 degrees upper - 88 811075 20 NA 30-60 degrees lower - 88 811076 20 NA 30-60 degrees upper - 88 811077 20 NA 60-90 degrees lower - 88 811078 20 NA 60-90 degrees upper - 88 811079 20 NA 90-120 degrees lower - 88 811080 20 NA 90-120 degrees upper - 88 811081 20 NA 120-150 degrees lower - 88 811082 20 NA 120-150 degrees upper - 88 811083 20 NA 150-180 degrees lower - 88 811084 20 NA 150-180 degrees upper - 88 811085 - 20 NA 180-210 degrees lower - 88 811086 20 NA 180-210 degrees upper - 88 811087 20 NA 210-240 degrees lower - 88 811088 20 NA 210-240 degrees upper - 88 C-10

SUID 811 - Reactor Containment Building Elevation Relative to Room Name or Approximate SUID # Building/Area Grade' Room # Grid Coordinates Area (i 2) 811089 Reactor Bldg 20 NA 240-270 degrees lower -88 811090 20 NA 240-270 degrees upper -88 811091 20 NA 270-300 degrees lower - 88 811092 l 20 NA 270-300 degrees upper - 88 811093 20 NA 300-330 degrees lower - 88 811094 20 NA 300-330 degrees upper - 88 811095 . 20 NA 330-360 degrees lower -88 811096 20 NA 330-360 degrees upper -88 811098 . 40 NA 0-30 degrees lower -88 811099 . 40 NA 0-30 degrees upper - 88 811100 40 NA 30-60 degrees lower -88 811101 40 NA 30-60 degrees upper -88 811102 40 NA 60-90 degrees lower -88 811103 . 40 NA 60-90 degrees upper -88 811104 6 40 NA 90-120 degrees lower -88 811105 40 NA 90-120 degrees upper -88 811106 . 40 NA 120-150 degrees lower - 88 811107 . 40 NA 120-150 degrees upper - 88 811108 . 40 NA 150-180 degrees lower - 88 81 i 109 . . 40. NA 150-180 degrees upper -88 811110 . 40 NA 180-210 degrees lower 88 811111 40 NA 180-210 degrees upper -88 C-11

SUID 811 - Reactor Containment Building Elevation Relative to Room Name or Approximate SUID # Building/Area Grade Room # Grid Coordinates Area (M2) 811112 Reactor Bldg 40 NA 210-240 degrees lower - 88 811113 40 NA 210-240 degrees upper - 88 811114 40 NA 240-270 degrees lower - 88 811115 40 NA 240-270 degrees upper - 88 811116 40 NA 270-300 degrees lower - 88 811117 40 NA 270-300 degrees upper - 88 811118 40 NA 300-330 degrees lower - 88 811119 40 NA 300-330 degrees upper - 88 811120 40 NA 330-360 degrees lower - 88 811121 40 NA 330-360 degrees Upper - 88 811122 60 NA 0-30 degrees lower - 88 811123 60 NA 0-30 degrees upper - 88 811124 60 NA 30-60 degrees lower - 88 811125 60 NA 30-60 degrees upper - 88 811126 60 NA 60-90 degrees lower - 88 811127 60 NA 60-90 degrees upper - 88 811128 60 NA 90-120 degrees lower - 88 811129 60 NA 90-120 degrees upper - 88 811130 60 NA 120-150 degrees lower - 88 811131 60 NA 120-150 degrees upper - 88 811132 60 NA 150-180 degrees lower - 88 811133 60 NA 150-180 degrees upper - 88 C- 12

SUID 811 - Reactor Containment Building Elevation Relative to Room Name or Approximate SUID # Building/Area Grade Room # Grid Coordinates Area (M2) 811134 Reactor Bldg 60 NA 180-210 degrees lower - 88 811135 60 NA 180-210 degrees upper - 88 811136 60 NA 210-240 degrees lower 88 811137 60 NA 210-240 degrees upper -88 81113816 60 NA 240-270 degrees lower -88 811139 60 NA 240-270 degrees upper -88 811140 60 NA 270-300 degrees lower - 88 811141 60 NA 270-300 degrees upper -88 811142 60 NA 300-330 degrees lower - 88 811143 60 NA 300-330 degrees upper - 88 811144 60 NA 330-360 degrees lower - 88 811145 60 NA 330-360 degrees upper - 88

'North Containment stairwell 811146 -27 NA -27 to -20 -50 NA North Containment stairwell 811147 -20 -20 to grade level -50 NA North Containment stairwell 811148 0 Grade level to +40 -50 NA North Containment stairwell 811149 40 +40 to +60 -50 South Containment stairwell 811150 -27 NA -27 to -20 -50 NA South Containment stairwell 811151 -20 -20 to grade level -50 NA South Containment stairwell 811152 0 Grade level to +40 -50 C- 13

i1L SUID 811 - Reactor Containment Building Elevation Relative to Room Name or Approximate SUID # Building/Area Grade Room # Grid Coordinates Area (M2)

... '-.....,,,a..i.:

South Containment stairwell 811153 Reactor Bldg 40 NA +40 to +60 -50 811154 0 NA Emergency Escape hatch -50 811155 0 to 60 NA Elevator/ shaft/ works -100 NA Containment walls from +60 811156 to Containment dome 811157 NA Containment dome -1100 811158 NA NA Tendon gallery access and 811159 gallery proper C- 14

SUID 812 - Spent Fuel Building Elevation .

Relative to - Room Name or Approximate SUID # Building/Area Grade Room # Grid Coordinates Area (M2 )

NA SFB +40 812001 Spent Fuel Building +40 column 9.3/H to 7.6/J 56 NA SFB +40 812002 +40 column 9.3/J to 7.6/K 72 NA SFB +40 walkways 812003 t40 column 7.6/H to 4.3/K 90 NA SFB +40 812004 +40 -column 4.3/H to 3.4/K 65 NA SFB stairwells and platforms above +40 812005 +40 - column 4.3/H to 3.4/K 40 812006 Grade level NA' SFP - cask pit 15 812007 Grade level. NA SFP - SF rack area 95 NA SFP - cask temporary storage 812008 +15" area. 15 812009 NA SFP - Upender pit 35 NA SFB West wall below grade 812010 -4.5' to Grade level (Cask pit wall) 35 NA SFB east wall column 3.4 to 812011 Grade to +20 5.8 (N.1/2 wall) 83 NA SFB east wall column 5.8 to 812012 Grade to +20 7.5 (S. 1/2 wall) 56 NA SFB east wall column 3.4 to 812013 +20 to +40 5.8 (N. 1/2 wall) 83 NA SFB east wall column 5.8 to 812014 +20 to +40 7.5 (S.1/2 wall) 56 NA SFB South wall Column H to 812015 Grade to +20 K 80 C-15 ,--

I SUID 812 - Spent Fuel Building Elevation Relative to Room Name or Approximate SUID # Building/Area Grade Room # Grid Coordinates Area (m2 )

812016 +20 to +40 NA SFB South wall column H to K 80 NA SFB west wall column 3.4 to 812017 Grade to +20 5.8 (N. 1/2 wall) 83 NA SFB west wall column 5.8 to 812018 Grade to +20 7.5 (S. 1/2 wall) 56 NA SFB west wall column 3.4 to 812019 +20 to +40 5.8 (N. 1/2 wall) 83 NA SFB west wall column 5.8 to 812020 +20 to +40 7.5 (S. 1/2 wall) 56 NA SFB North wall Column H to 812021 Grade to +20 K 83 NA SFB North wall column H to 812022 +2 to +40 K 56 NA SFB North wall column H to 812023 +40 to +60 K 125 NA SFB West wall column 3.4 to 812024 +40 to +60 9.3 315 NA SFB East wall column 3.4 to 812025 +40 to +60 9.3 315 812026 Roof NA SFB roof 470 C - 16

SUID 813 - AUXILIARY BUILDING -47 Elevation Relative to Room Name or Approximate SUID # Building/Area Grade Room # Grid Coordinates Area (M2)

West decay heat removal 813001 Auxiliary Bldg -47 -.001 pump room - lower 90 West decay heat removal 813002 -47 . 001 pump room - upper 90 East decay heat removal 813003 -47 002 - pump room - lower 90 East decay heat removal 813004 47 002 pump room - upper 90 Reactor coolant drain tank 813005 -47 003 room - lower 90 Reactor coolant drain tank 813006 -47 003 room - upper 90 Stair well - landing east decay 813007 -38 to -47 002 :heat removal pump room 38 Stair well - decay heat 813008 -47 to -20 001 removal pump rooms 50 056 & West stairwell -20 to grade 813009 -20 to Grade 127 level 45 C- 17

1L SUID 813 - AUXILIARY BUILDING -29 Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2)

Radwaste pump Area North 813045 Auxiliary Bldg -29 036/057 wall to column 9.2 35 Radwaste pump Area 813046 . -29 036/057 Column 9.2 to column 11.7 55 Radwaste pump Area Column 813047 -29 036/057 11.7 to south wall 45 Conc. BA storage Tk Rm -

813048 -29 037 Lower 95 Conc. BA storage Tk Rm -

813049 . -29 037 Upper 95 "B" Coolant Waste Hold-up Tk 813050 "-29 038 Rm - Lower 95 "B" Coolant Waste Hold-up Tk 813051 -29 038 Rm - Upper 95 "A" Coolant Waste Hold-up Tk 813052 -29 039 Rm - Lower 95 "A" Coolant Waste Hold-up Tk 813053 "-29 039 Rm - Upper 95 813054 -29 040 Spent Regen Tk Rm - Lower 95 813055 -29 040 Spent Regen Tk Rm - Upper 95 "A" Coolant Waste Receiver 813056 -29 041 Tk Rm - Lower 95 "A" Coolant Waste Receiver 813057 -29 041 Tk Rm - Upper 95 "B" Coolant Waste Receiver 813058 -29 042 Tk Rm - Lower 95 "B" Coolant Waste Receiver 813059 -29 042 Tk Rm - Upper 95 C- 18

SUID 813 - AUXILIARY BUILDING -20 Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2) 813010 Auxiliary Bldg -20 010 Corridor -DHRC room(s) 25 813011 -20 011 Spray Add Tank Area 55 813012 -20 012 Flash Tank pump room 25 813013 -20 013 Letdown Filter room 10 813014 -20 014 Flash tank room 35 Corridor - Main east/west 813015 -20 015 Columns P-T - lower 75 Corridor - Main east/west 813016 -20 015 Columns P-T - upper 75 Corridor - Main east/west 813017 -20 015 Columns T-V Upper/lower 50 813018 -20 016 Misc.Wastetankpump.room 25 813019 -20 017 - Misc. waste tank room 35 Waste gas decay tank room -

813020 -20 018 lower 100 Waste gas decay tank room -

813021 -20 018 upper 100 DeBorating IX & Misc Waste Cond. Demin. Rm lower &

813022 -20 - - 019 upper 50 813023 -20 59 Waste gas valve gallery 19 Radwaste control panel area -

813024 -20 020 lower 75 Radwaste control panel area -

813025 -20 020 upper 75 Misc. Waste gas cond. Tank 813026 -20 021 rm. - lower 75 C- 19

I1_

SUID 813 - AUXILIARY BUILDING -20 Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2)

Misc. Waste gas cond. Tank 813027 Auxiliary Bldg -20 021 rm. - upper 75 Waste gas compressor room -

813028 -20 022 lower 100 Waste gas compressor room -

813029 -20 022 upper 100 Misc. Waste Evap. Rm. -

813030 023 Lower 75 Misc. Waste Evap. Rm. -

813031 -20 023 upper 75 Misc. Waste concentrate Tk 813032 -20 024 Rm - upper lower 35 813033 -20 025 BA Evap Rm - Lower 75 813034 -20 025 BA Evap Rm - Upper 75 813035 -20 026 Misc. Waste filter Rm 35 813036 -20 27 IX valve gallery 35 813037 -20 28 IX Vault 5 813038 -20 29 IX Vault 5 813039 -20 30 IX Vault 5 813040 -20 31 IX Vault 5 813041 -20 32 IX Vault 5 813042 -20 33 IX Vault 5 813043 -20 34 IX Vault 5 813044 -20 35 IX Vault 5 See - 29' for SUID 045-059 813060 l Auxiliary Bldg -20 043 "B" HPI Pump Rm - Lower 55 C-20

SUID 813 - AUXILIARY BUILDING -20 Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2) 813061 Auxiliary Bldg -20 043 "B" HPI Pump Rm - Upper 55 Make-up Pump Rm - Lower &

813062 -20 044 Upper 45 North-East -20 Corridor -

813063 -20 045 Lower & Upper 35 Seal Return Cooler Rm -

813064 -20 046 Lower & Upper 50 Spent Resin Tk Rm - Lower &

813065 -20 047 Upper 10 813066 -20 048 Crud Tk Rm - Lower & Upper 5 Crud Tk Pump Rm - Lower &

813067 -20 049 Upper 5 Radwaste Air Supply Fan Rm 813068 -20 050 - Lower & Upper 20 East Decay Heat Removal Cooler Rm Column P-T -

813069 -20 051 South of Column 9.7 - Lower 90 East Decay Heat Removal Cooler Rm Column P-T -

813070 -20 051 South of Column 9.7 - Upper 90 East Decay Heat Removal Cooler Rm Column P-T Between Column 9.7 - 9.1 813071 -20 051 Lower 90 East Decay Heat Removal Cooler Rm Column P-T -

Between Column 9.7 - 9.1 813072 -20 051 Upper 90 East Decay Heat Removal Cooler Rm North of Column 813073 -20 051 9.1 Upper & Lower 50 C-21

i1L SUID 813 - AUXILIARY BUILDING -20 Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M 2)

West Decay Heat Removal Cooler Rm West of Column N 813074 Auxiliary Bldg -20 052 Upper 95 West Decay Heat Removal Cooler Rm West of Column N 813075 -20 052 Lower 95 West Decay Heat Removal Cooler Rm East of Column N 813076 -20 052 Lower 100 West Decay Heat Removal Cooler Rm East of Column N 813077 . -20 052 Upper 100 813078 -20 053 "A" HPI pump Rm 40 West -20 to grade level 813079 -20 to grade 056&127 stairwell 30 Sanitary sump Electric cable 813080 -20 055 spreading room - west 1 Electric cable spreading room 813081 -20 055 - West 65 Electric cable spreading room 813082 -20 054 - East 50 813083 -20 061 NSEB Tunnel - West 55 813084 . -20 062 NSEB Tunnel - East 55 C - 22

SUID 813 - AUXILIARY BUILDING - Grade Level Elevation Relative to Room Name or Approx.

SUID # Building/ Area Grade Room # Grid Coordinates Area (i 2 )

813085 Auxiliary Bldg Grade to +20 101 Stairwell - #1 25 813086 Grade 102 Corner, east corridor grade 15 813087 Grade - 103 Corridor 30 813088 Grade 104 Corridor 35 813089 Grade 105 Corridor Column K-S 112 813090 Grade -105 Corridor Column H-K 45 813091 Grade 106 Chemical Storage Rm - Upper 100 813092 Grade 106 Chemical Storage Rm - Lower 100 813093 -20 to Grade 107 Central Stairwell - (#6) 20 Machine shop (change area) 813094 Grade 109 North of column 11.7 55 Machine Shop (change area) 813095 Grade 109 south of column 11.7 40 IX Access area column 813096 Grade 110 . S-T 55 IX Access area column 813097 Grade 110 T-U 55 IX Access area Column 813098 Grade . 110 u-V 55 112&-

813099 Grade 139 Waste Solidification area 50 813100 Grade Stair #4 East stairwell grade to - 10 10 813101 Grade - 113 Make-up Tank Rm. 45 813102 . Grade 114 Make-up value area 10 813103 Gradeto+20 115 - Elevator#1 10 C-23

SUID 813 - AUXILIARY BUILDING - Grade Level Elevation Relative to Room Name or Approx.

SUID # Building/ Area Grade Room Grid Coordinates Area (M2) 813104 Auxiliary Bldg Grade to +20 116 Stairwell #3 18 813105 Grade 117 Service Area Column L-N 43 813106 Grade 117 Service Area Column N-P 43 813107 Grade 117 Service Area column P-R 40 Contaminated Garment Area 813108 Grade 118 Column R-S 40 813109 Grade 119 Nuclear Service Battery Room 38 813110 Grade 120 Nuclear Service Battery Room 38 813111 Grade 121 Switch Gear Room 265 813112 Grade 122 Cable Shaft 8 813113 Grade 123 Cable Shaft 8 813114 Grade 124 Switch GearRoom 265 813115 Grade 125 Nuclear Service Battery Room 38 813116 Grade 126 Nuclear Service Battery Room 38 813117 Grade 128 Elevator Machine Room 12 813118 Grade to +20 129 Elevator #2 10 813119 Grade 130 Diesel Generator Room 112 813120 Grade to +20 131 Stairwell #2 10 813121 Grade 132 Diesel Generator Room 112 813122 +10 133 Chemical Storage Balcony 67 813123 Grade 134 Decon Room 55 813124 Grade 135 Toilet 8 C-24

SUID 813 - AUXILIARY BUILDING - Grade Level Elevation Relative to Room Name or Approx.

SUID # Building/ Area Grade Room # Grid Coordinates Area (m2 )

813125 Auxiliary Bldg Grade 136 Make Up Filter Vaults 10 813126 Grade 137 Make up filter valve gallery 7 Transition room AB grade to 813127 Grade 138 Tank Farm 12 Enclosure W/l solidification 813128 Grade 139 room 10 Area between AB, SFB, & RB 813129 Grade -NA south of column 5.6 120 SUID 813 - AUXILIARY BUILDING +20 Mezzanine Elevation Relative to Room Name or Appr6x.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2 )

813130 Auxiliary Bldg +20 201 Stairwell #1 +20 to +40 25 Mezzanine roof area Columns 813131 +20. T-V & 8.5-13 450 813132 +20 202 Elevator Machine Room 20 813133 +20 to +40 203 Elevator #1 10 813134 - +20 204 A/C Equipment room 160 813135 +20 206 Corridor 45 813136 +20 207 Corridor 150 813137 +20 208 Duct Space 15-813138 +20 209 Electrical Penetration room 90 C-25

1.

SUID 813 - AUXILIARY BUILDING +20 Mezzanine Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2) 813139 Auxiliary Bldg +20 to +40 210 Stairwell #3 18 813140 +20 211 Ventilation Equipment room 245 813141 +20 212 AC/DC Panel room 27 813142 +20 213 AC/DC Panel room 27 813143 +20 214 Electrical equipment room 255 813144 +20 215 Cable shaft 15 813145 +20 216 Cable Shaft 15 813146 +20 217 Electrical equipment room 175 813147 +20 218 AC/DC Panel room 27 813148 +20 219 AC/DC Panel room 27 813149 +20 220 Station Battery room 85 813150 +20 to +40 221 Elevator #2 10 813151 +20 to +40 222 Stairwell #2 18 Storage room (behind 813152 +20 223 Elevator #2) 7 813153 +20 224 Communications room 150 813154 +20 225 Sample cooler chiller room 35 "Fan platforms (2)" south of 813155 +20 NA sample cooler room 41 Corridor & mezzanine landing (bounded by column K8.1-813156 +20 226 9.3/Ki, 9.3/L8.7-8.1) 813157 +20 227 Control Rod Pit 5 813158 +20 228 Pipe Way 15 C-26

SUID 813 - AUXILIARY BUILDING + 40 Elevation -

Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (m2) 813159 Auxiliary Bldg +40 301 Stairwell #1 +40 to +60 25 813160 . +40 302 Corridor 50 813161 . +40 303 Vestibule 6 813162 +40 304 Janitor locker 10 813163 ' . +40 305 , Clean Toilet 25 813164 ' +40 306 Clean Shower 25 813165 . '4 +40 307 Showerdrying area 10 813166 +40 308 Clean Locker room 75 813167 ' +40 NA Woman's locker room 10 Instrument Issue Room (Anti-813168 ' +40 309 C storage) 8 813169 +40 310 Decon Disrobing area 12 813170 +40 311 Control Point 45 813171 ' +40 ' 312 Decon Shower 10 813172 +40 313 Contaminated Toilet 5 813173 +40 314 813174 - +40 315 Elevator#1 10

. Clean Garment Storage (RP 813175 ' - +40 316 FIELD OFFICE) 30 813176 +40 - 317 Containment lay down area 33 813177 ' +40 318 Stairwell #3 18 813178 +40 319 Primary Sample Station 32 C-27

SUID 813 - AUXILIARY BUILDING + 40 Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2) 813179 Auxiliary Bldg +40 320 Radio Chemistry Room 32 813180 +40 321 Radio Chemistry Count Room 25 813181 +40 322 Corridor 95 Chemical Storage (Instrument 813182 +40 323 Repair Room ) 35 Calibration and Source 813183 +40 324 storage Room 32 813184 +40 325 Storage room 12 813185 +40 326 Chemical Storage Room 10 813186 +40 327 Secondary Sample Station 25 813187 +40 328 Secondary Laboratory 38 813188 +40 329 Corridor 156 813189 +40 330 HP office (Tech drone room) 36 813190 +40 331 HP Supervisor's Office 10 Calculation and Data Storage 813191 +40 332 Room 20 813192 +40 333 Corridor 37 Technical Support Center 813193 +40 334 (TSC) 130 813194 +40 335 Conference Room #1 23 813195 +40 336 Conference Room #2 23 813196 +40 337 813197 +40 338 Computer Room 150 C-28

SUID 813 - AUXILIARY BUILDING + 40 Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2) 813198 Auxiliary Bldg +40 339 Control Room 60 813199 +40 340 . Shift Supervisor Office 15 813200 . +40 341 Alcove 20 813201 +40 342 Control Room Kitchen 25 813202 +40 343 Control Room Rest Room 8 813203 +40 344 Corridor 23 813204 +40 345 Stairwell #2 +40 to +60 10 813205 +40 346 Elevator #2 10 813206 +40 347 813207 +40 348 Conference Room (weight 813208 +40 349 . Room) 52 813209 +40 350 Corridor 4 813210 - +40 351 Equipment access room 6 813211 . +40 352 Closet (weight room) 5 813212 +40 353 Water closet 6 SUID 813 - AUXILIARY BUILDING - Roof Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (i 2) 813213 Auxiliary Bldg +60 Roof Roof - West of column N 980 813214 +60 Roof Roof - East of Column N 975 C-29

I.

SUID 814 - Training and Records Building Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2)

AC 33 - AB35 of RSNGS Plot 814001 T&R BLDG Grade NA Plan 1400 AC 33 - AB35 of RSNGS Plot 814002 Second Floor NA Plan 1400 AC 33 - AB35 of RSNGS Plot 814003 Third Floor NA Plan 1400 AC 33 - AB35 of RSNGS Plot 814004 Fourth Floor NA Plan 1400 AC 33 - AB35 of RSNGS Plot 814005 Fifth Flow NA Plan 1400 AC 33 - AB35 of RSNGS Plot 814006 Roof NA Plan 1400 SUID 815 - Nuclear Services Electrical Building (NSEB)

Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2 )

AC 32 - AB 32 of RSNGS Plot 815001 NSEB Grade NA Plan 425

. AC 32 - AB 32 of RSNGS Plot 815002 +20 NA Plan 425 AC 32 - AB 32 of RSNGS Plot 815003 +40 NA Plan 425

. AC 32 - AB 32 of RSNGS Plot 815004 Roof NA Plan 425 C-30

SUID 816 - Control Alarm Station (CAS)

Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2)

AD 33- AC 33 of RSNGS Plot 816001 CAS Grade NA, Plan 200 AD 33- AC 33 of RSNGS Plot 816002 CAS Roof NA Plan 20(

SUID 817 -TDI Diesel Generator Building / Cask Storage Facility Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2)

AF 33 - AE 34 of RSNGS Plot 817001 TDI BUILDING/CSF Grade NA Plan 700 AF 33 - AE 34 of RSNGS Plot 817002 TDI Building/CSF Roof NA Plan 700 SUID 818 - Electrical Fabrication Shop Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2)

Electrical Fabrication Shop to be demolished and released prior to FSSS under existing 818001 rocedures.

C-31

SUID 820 - L&D Building Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2) 820001 L&D Building to be demolished and released prior to FSSS under existing procedures.

SUID 823 - Intake structure Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2)

Intake Pump 823001 Structure Grade NA X 30 of RSNGS Plot Plan 375 C-32

SUID 826 - TURBINE BUILDING -

Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (in2) 826001 Turbine Bldg -7'to grade NA MFP pit - column A/3 to B/5 52 826002 " -7' to grade NA MFP pit - column B/3 to C/5 92 826003 " -7'to grade NA MFP pit - column C/3 to D/5 92

. . Condensate Pump Pit 826004 -15.5' to grade NA column D6 to E8 82 Condenser Pit 826005 -9.5' to grade NA column D7 to E4.6 (approx.) 55 Condenser Pit 826006 -9.5' to grade NA column D7 to E9.4 (approx.) 55 Condenser Tube Pulling Pit West of column B and North 826007 -10' to grade NA of Column 7 91 Condenser Tube Pulling Pit West of column B and South 826008 -10' to grade NA of Column 7 91 Condenser Pit 826009 -9.5' to grade NA column B5 to C7 91 Condenser Pit 826010 -9.5' to grade NA column C5 to D7 91 Condenser Pit 826011 -9.5' to grade NA column B7 to C9 - 91 Condenser Pit 826012 -9.5' to grade' NA column C7 to D9 91

'Grade to Turbine building 826013 Mezzanine NA Column A/2 to H/6 925 Grade to Turbine building 826014 . Mezzanine NA Column A/6 to H/10 ;930 Grade to Turbine building 826015 Mezzanine NA Column Al10 to H/13 745 C-33

1.

SUID 826 - TURBINE BUILDING Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2)

Mezzanine to Turbine building 826016 Turbine Bldg + 39.5' NA Column A/2 to H/6 925 Mezzanine to Turbine building 826017 . + 39.5' NA Column AN6 to H/10 930 Mezzanine to Turbine building 826018 + 39.5' NA Column A/10 to H/13 745 Turbine Deck 0 - 80 feet south column 2 (Including north 826019 Turbine deck NA gantry catwalks) 940 Turbine Deck 80 -160 feet 826020 Turbine deck NA south column 2 940 Turbine Deck 160 - 222 feet south column 2 (Including 826021 Turbine deck NA south gantry catwalks) 725 North laydown column A2 -

826022 Grade level NA Hi 365 North laydown column Al -

826023 Grade level NA H'X, 725 826024 Grade level NA South laydown area column 500 Transformer pad between AB 826025 Grade level NA and NSEB 175 Station service transformer pad adjacent to AB door 826026 + 40' NA AU346 39 Walkway from AB to turbine 826027 +40 NA deck at AU346 18 Control room emergency exit and vital station service 826028 +40 NA transformer enclosure 46 C-34

SUID 826 -TURBINE BUILDING Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (i 2 )

Aux Building ventilation intake 826029 Turbine Bldg +40 NA structure 41 SUID 827 - Tool Room Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2) 827001 Tool Room to be demolished and released prior to FSSS under existing procedures.

SUID 828 - GRS Warehouse Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2) 828001 GRS Warehouse to be demolished and released prior to FSSS under existing procedures.

C-35

SUID 830 - Interim On-Site (Radwaste) Storage Building (IOSB)

Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2) 830001 lOS Perimeter grade NA U24 of RSNGS plot plan 750 830002 grade NA U23 of RSNGS plot plan 425 830003 grade NA U22 of RSNGS plot plan 800 830004 grade NA V22 of RSNGS plot plan 750 830005 grade NA V23 of RSNGS plot plan 425 830006 grade NA V24 of RSNGS plot plan 750 SUID 833 - Warehouse "B" Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2) 833001 "B" warehouse Grade NA Y 26-28 of RSNGS plot plan 1400 SUID 834 - Turbine Rotor Storage Shed Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2 )

834001 Turbine rotor shed Grade Y25-26 of RSNGS plot plan C-36

SUID 836 - Aux Boiler Pad Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade: -Room # Grid Coordinates Area (i 2 )

Aux Boiler Pad Z29-30 of 836001 Auxiliary Boiler Pad. Grade NA RSNGS Plot Plan 1860 SUID 837 - Regenerant Holdup Tanks Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2) 837001 WARHUT == AA29 of RSNGS Plot Plan 60 837002 "B"RHUT AB29 of RSNGS Plot Plan 60 837003 aC" RHUT AC29/30 of RSNGS Plot Plan 60 SUID 838 - FAB Shop (GRS)

Elevation Relative to Room Name or Approx.

2 SUID Building/Area Grade Room # Grid Coordinates Area (n5) 838001 Fab Shop (GRS) Grade NA AD30 of RSNGS plot plan 1025 C-37

IL_

SUID 839 - Transformer Yard Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2)

Transformer Yard West of Turbine AA30 - AC31 of RSNGS plot 839001 Building Grade NA plan 1860 SUID 840 -Warehouse "A" Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2)

AA29 - AB28 of RSNGS plot 840001 Warehouse "A" Grade NA plan 975 SUID 842 - Warehouse "C" Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (m2) 842001 Warehouse "C" Grade NA AF31 of RSNGS plot plan 235 C-38

SUID 851 - Switchyard Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2)

AA22 to AC27 of the RSNGS Plot Plan w/e control building (extends outside of switchyard 851001 Switchyard Grade N/A fencing) 16725 Z2 to Z27 of the RSNGS Plo Plan W/E of Trackage and Turbine Rotor Shed (extends 851002 Switchyard grade N/A outside of switchyard fencing) 4185 Switchyard Control 851003 Building grade N/A AC26 of the RSNGS Plot Plan 225 AD22 to AE27 of the RSNGS Switchyard storage Plot Plan (extends outside of 851004 yard grade N/A switchyard fencing) 11150 SUID 852 - Machine Shop Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2 )

852001 Machine Shop grade N/A AB28 of RSNGS plot plan 465 SUID 853 - Contractor Fab Shop Elevation Relative to Room Name or Approx.

SUID # Building/Area Grade Room # Grid Coordinates Area (M2) 853001 Fab Shop Grade N/A C-39

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Appendix D Miscellaneous Location and Earthquake Data and Figures

1. USGS National Mapping Information for Rancho Seco Power Plant

2. Industrial Area Overhead Photograph

3. USGS Rancho Seco regional map (400 km radius)

4. USGS Rancho Seco regional map (50 km radius)

5. USGS Earthquake Data Base search results - magnitude 4.0 or greater within an 80.5 km (50 mile) radius

6. USGS Seismicity of California 1990 - 2001 (pictogram)

1. USGS National Mapping Information for Rancho Seco Power Plant MUSGS National Mapping Information F 10.1 Feature Name: 10.2 Rancho Seco Power Plant f Feature Type: Building State: Californial County: I Sacramento lUSGS 7.5' x 7.5' Map:l Goose Creek l attude(nnonn'nn"): l. 382040N

[Longltude(nnnonn'nn"): 1210712W U.S. Department of the nteirior 11U.S. Geological Survev 12201 Sunrise Valley Drive, Reston, VA 20192, USA DMS 121 07 12 DMS 38 20 40 DM.m 121 7.2 DM.m 3820.6666666667 D.d 121.12 D.d 38.3444444444 The formulas are as follows:

Degrees Minutes Seconds to Degrees Minutes.m Degrees- Degrees Minutes.m Minutes - (Seconds /60)

Degrees Minutes.m to Decimal Degrees

.d= .m /60 Decimal Deg rees = Degrees -+.d Directions Magazine: Masthead I Contact Us I Advertising

@1998-2003 Directions Magazine. All Rights Reserved. Privacy Statement Questions and corments to: websiteIdirectionsmap.c om D-1

2. Industrial Area Overhead Photograph D-2 6013

3. USGS Rancho Seco regional map (400 km radius) wweatora

_ reka *6Elko lRedding Chico

• Reno W6 ramento '--.

A Rancho Seco6iPower Plant

.Stuultur - -'

San Jose  ;,t1 St. Geo Salinas Fresno Wisalia ' .a z 6BakersField Santa Maria .

122W 12 2fi 12 2W 1 22 1201W 11AR'W 11i Aw Zoom lnZoom Out LEGEND State Expressway Highway Connector LZJ City 0 ,50 ,100 150 200 mi ScalIe 1:6568499 1- '100 '200 '300 '1400 ko

• average--true scale depends on monitor resolution Credit: U. S. Geological Survey A http:jigeonames.usgs.gov/pis/gnisjMapserverf name-Rancho+Seco+Power+plant&fsate-CA&f-lationg=38204ON1210712W&f_ht-8&server=TIGER D -3 D-3S

4. USGS Rancho Seco regional map (50 kin radius)

.2 - Creek Hi~

122 .ObW 121 R"WLI 121 .*.W 121 4'W 121 .2_W 121 I Zooml I Zoom Out LEGEND

- State _ Military Area

- County go, National Park LakelPondlOcean City Expressway County Highway Connector

_Stream 5 10 15 20 25 30 mi Scale 1:821062 lo l' 20 30 14Q '50 km

• average--true scale depends on monitor resolution Credit: U. S. Geological Survey 4 http:Ilgeonames.usgs.gov/pisignis/Mapserver?f name=Rancho+Seco+Power+PIant&fstate=CA&ffJatlong=38204ON1210712W&f_ht=18server=TIGER D-4 Cap n --

5. USGS Earthquake Data Base Search Results --Magnitude 4.0 or Greater within an 80.5 km (50 mile) Radius NEIC: Earthquake Search -Results U. S. GE.OL O'G I C A L S U R V E Y E A R T HQ U A KXE D ATA B A S E FILE CREATED: Thu Jan 8 14:21:59 2004 Circle Search Earthquakes= 4 Circle Center Point Latitude: 38.350N IL.ongitude: 121.120E Radius: 80.500 km.

Catalog Used: PDE Magnitude Range: 4.0 - 12.0 Data Selection: Historical & Preliminarv I)ata CAT YEAR MO DA ORIG TIME LAT LONG DEP MAGNITUDE IEFM DTSVNWG DIST NFPO km TFS PDE 1985 11 23 004713.67 38.68 121.20 33 4.60 mb GS 37 PDE 1991 03 13 163117.99 37.94 121.09 33 4.30 MLBJI 45 PDE 1997 04 12 070507.48 38.31 120.53 33 4.50 mb GS 52 PDE 1997 09 18 063201.35 38.08 121.20 33 4.80 mTb GS 31 Credit: USGS National Earthquake Information Center 4D N-121. 12WCA0 Y -. MT DAYYM- H-UAY .AG-I2.DEPI -&NDEP2-8JO1-&XO2-&3¶IT-Sdxnt+Sech D-5

6. USGS Seismicity of California 1990 - 2001 (pictogram)

-124- -122' -120' -1 18' -116- -114-42e 42' 40 ~40 38 -38.

36' 34- -34.

32- 32'

-124' -122 -120- -1181 *116 -114' I I

.151 -71 .33 0 DEPTH Earthquake locations are from the USGS/NEIC PDE catalog.

Credit: USGS National Earthquake Information Center Al http:/lneic.usgs. gov/neis/states/californialcalifornia-seismicity.html D-6 COp

Appendix E Miscellaneous Historical Construction Photographs

Ii FI F . - h h f . .@ . , .

h . N f .. .. . . . ... - -

F - '. '. . . ..

r , . . . . . . ... .. . . ...

t ., .,,. . , -, -' . ..

i. . . ..

B . . ' , . ' , . ' . .

l' ,

A.

C. . . , -,

l. ,, . '"  ;'-', ' ,' "',..: ' ' -  !  :  :.

F - .- . ....  : - - i .t  :.-.-  :

l . ' .. l  :* . .,, '-' '; 'S - - . . ' ' -:

,, . .,. ...... ..  : ' ' .................. ...  :' . -.- . i ads.. . . - A: . . - . ., - w .

l, S  ;  ; l

. .. ^ ...... .... ,, .. . . , ., . . . . . _ . ..... ..... ....

December 27, 1968 - Clay access road and Clay East road intersection E-1

.11i, . . .

I i 1.

. . I.

iI ii

. I I

-!i I .ji March 10, 1969 - Site Preparation, looking south at rough grading in turbine and reactor area E-2

A'/V Wrnm j

March 10, 1969 - Site Preparation, looking north at first rough grading cut into the turbine area E-3

April 4, 1969 - Site Preparation, looking north along the east edge of the reactor containment structure E-4

fi-t:_r -

May 1, 1969 - Site Preparation,'Twin Cities access road looking north E-5

it z7 September 9, 1969 - forming for the fill concrete under the tendon access gallery floor slab E-6

October 10, 1969 - concrete placement in reactor building tendon access gallery walls E-7

I 'I November 11, 1969 - a view of the shop and warehouse construction taken from the tower crane E-8

February 9, 1970 - installing floor liner plate on reactor building floor slab E-9

March 20, 1970 - a view looking north of the auxiliary building sub-basement slab E-10

March 20, 1970 - overall view of the containment structure with cooling tower no. 1 in background E- 11

• -- - q', ,jj A

p 27 April 23, 1970 - reinforcing steel and tendon sheathing for the first lift of the containment wall E- 12

April 23, 1970 - overall view looking northeast E - 13

,7-

'AlL

~

April 24, 1970 - 108" diameter circulating water pipe forming and reinforcement for concrete encasements E- 14

May 28, 1970 - looking west at the auxiliary building subbasement wall with the turbine building in the background E - 15

May 28, 1970 - view looking south at the turbine building with the reactor and auxiliary buildings on the extreme left E - 16

a - - I May 28, 1970 - view looking south at the turbine and reactor buildings E- 17

July 17, 1970 - auxiliary building showing reinforcing being placed E- 18